ML20056H280

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Requests Addl Info to Continue Review of Application for Amend of OL R-95,per 921222 Submittal
ML20056H280
Person / Time
Site: Rhode Island Atomic Energy Commission
Issue date: 08/24/1993
From: Mendonca M
Office of Nuclear Reactor Regulation
To: Tehan T
RHODE ISLAND, STATE OF
References
NUDOCS 9309090083
Download: ML20056H280 (128)


Text

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Docket No. 50-193 Aus:ust 24, 1993 Hr. Terrence Tehan, Director Nuclear Science Center Rhode Island Atomic Energy Commission South Ferry Road Narrangansett, Rhode Island 02882

Dear Mr. Tehan:

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION We are continuing our review of your application for amendment of Facility Operating License No. R-95 for the Rhode Island Atomic Energy Commission research reactor Technical Specifications which you submitted on December 22, 1992. During our review of your application, questions have arisen for which we require additional information and clarification.

Please provide responses to the enclosed Request for Additional Information within 120 days of the date l

of this letter.

Following receipt of the additional information, we will continue our evaluation of your application.

Also enclosed is a copy of the University of Texas Technical Specifications and a guidance document on Technical Specifications that is out for review.

These documents may provide further guidance although the exact format and wording may not be appropriate for your application.

Your response should be executed in a signed original under oath or affirmation in accordance with 10 CFR 50.30(b).

This request affects nine or fewer respondents and, therefore, is not subject to Office of Management and Budget review under Public Law 96-511.

If you have any questions regarding this review or request, please contact me at (301) 504-1128.

MebbfibafS$bYPMectManager rv n Non-Power Reactors and Decommissioning Project Directorate Division of Operating Reactor Support Office of Nuclear Reactor Regulation

Enclosures:

1.

Request for Additional Information 2.

University of Texas Technical Specifications 3.

Guidance Document on Technical Specifications cc w/ enclosures:

See next page DISTRIBUTION:[RH0DEISL.TS] (Mendonca-lb disk)

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wassiscTon. o.c. 2asss.cooi August 24, 1993 Docket No. 50-193 Mr. Terrance Tehan, Director Nuclear Science Center Rhode Island Atomic Energy Commission South Ferry Road Narrangansett, Rhode Island 02882

Dear Mr. Tehan:

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION We are continuing our review of your application for amendment of Facility Operating License No. R-95 for the Rhode Island Atomic Energy Commission research reactor Technical Specifications which you submitted on December 22, 1992. During our review of your application, questions have arisen for which l

we require additional information and clarification. Please provide responses to the enclosed Request for Additional Information within 120 days of the date of this letter. Followbg receipt of the additional information, we will continue our evaluatidn of your application.

Also enclosed is a copy of the University of Texas Technical Specifications and a guidance document on Technical Specifications that is out for review.

{

These documents may provide further guidance although the exact fe m at and wording may not be appropriate for your application.

Your response should be executed in a signed original under oath or l

affirmation in accordance with 10 CFR 50.30(b).

This request affects nine or fewer respondents and, therefore, is not subject to Office of Management and Budget review under Public Law 96-511.

If you have any questions regarding this review or request, please contact me i

at (301) 504-1128.

Sincerely, l

C & r-i Marvin M. Mendonca, Senior Project Manager i:

Non-Power Reactors and Decommissioning Project Directorate Division of Operating Reactor Support Office of Nuclear Reactor Regulation

Enclosures:

1.

Request for Additional Information a

2.

University of Texas Technical Specifications 3.

Guidance Document on Technical Specifications cc w/ enclosures:

See next page

Rhode Island Atomic Energy Commission Docket No. 50-193 cc:

President, Town Council Town of Narragansett Town Hall Narragansett, Rhode Island 02882 Governor of Rhode Island Providence, Rhode Island 02903 1

1 1

l l

ENCLOSURE 1 i

REQUEST FOR ADDITIONAL INFORMATION RHODE ISLAND ATOMIC ENERGY COMMISSION DOCKET NO. 50-193 4

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1.

Provide a Table of Contents for the Technical Specifications.

j 2.

Provide a definition for " Certified Operators" (ANSI 15.1 may provide l

guidance) including definitions for Senior Reactor Operator and Reactor Operator to assure consistent use in the subsequent staffing Specifications or provide rationale and reference why this definition is not needed (e.g., they are acceptably defined in other specific sections of Technical Specifications).

3.

Provide in Section 1.0 definitions for:

a.

" Confinement" or " Containment" as applicable (see later question to assure consistency of this definition).

b.

" Excess reactivity" 1

I c.

" Experiment" d.

" Shutdown Margin" ANSI 15.1 may provide additional guidance.

4.

In Section 1.0 DEFINITIONS of " Measured Value,"

a.

Either provide the reason for the use of the term " process variable"

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or change it to parameter as in ANSI 15.1 which may provide additional guidance if needed, and b.

Either provide the reason for not using the term " measuring channel 1

output" rather than simply " measuring channel" or change it to the former per ANSI 15.1 which may provide additional guidance.

l 5.

In Section 1.0 DEFINITIONS of " Measuring Channel" and "True Value" either

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provide the reason for the use of the term " process variable" or change i

it to " parameter" per ANSI 15.1 which may provide additional guidance.

6.

In Section 1.0 DEFINITIONS of " Moveable Experiment," either provide the i

reason for the use of this definition with specific applicability to your facility, or change the definition to be consistent with that of the definition from ANSI 15.1.

7.

In Section 1.0 DEFINITIONS of " Operable and Operating," either provide the reason for the use of the phrase "in its normal manner," or delete this phrase to be more consistent with ANSI 15.1.

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REQUEST FOR ADD'TIONAL INFORMATION ENCLOSURE 1 8.

In Section 1.0 DEFINITIONS either provide the reason for not using the definition of " Protective Action" from ANSI 15.1, or add the definition.

9.

In Section 1.0 DEFINITIONS of " Potential Reactivity Worth of an Experiment," either provide the reason for the use of this definition, or change the definition to be consistent with that of the definition for

" Reactivity Worth of an Experiment" from ANSI 15.1.

10.

In Section 1.0 DEFINITIONS of " Reactivity Limits,"

a.

Either provide the reason for the use of this definition, or delete the use of this limit in favor of more specific parameters such as shutdown margin (in addition to excess reactivity) and include this definition as part of the definition of " Reference Core Condition" in accordance with ANSI 15.1.

b.

Also, note the definition, that was proposed by the Rhode Island Nuclear Science Cenwr, is not consistent with the subsequent use in the title for Technical Specification 3.1.1.

Correct this discrepancy or provide the reason it is not inconsistent.

11.

In Section 1.0 DEFINITIONS of " Reactor Operation," either provide the reason for the use of this definition, or change the definition to be I

consistent with that of the definition for " Reactor Operating" from ANSI 15.1.

12.

In Section 1.0 DEFINITIONS of " Reactor Safety System," either provide the reason for the use of this definition, or change the definition to be consistent with that of the definition for " Reactor Safety Systems" from ANSI 15.1.

13.

In Section 1.0 DEFINITIONS of " Reactor Secured," either provide the reason for the use of this definition, or change the definition to be consistent with that of the definition from ANSI 15.1.

14.

In Section 1.0 DEFINITIONS either provide the reason for not using the definition of " Reactor Shutdown" from ANSI 15.1, or add the definition.

15.

For the definition of *Readily Available on Call" in Section 1.0 DEFINITIONS, provide a more objective requirement or definition of what

" reasonable driving time from the reactor building" means (e.g.,

30 minutes or 15 miles per ANSI 15.1) or provide, in the response to this j

request, reference to procedures and controls which establish this i

definition.

4

REQUEST FOR ADDITIONAL INFORMATION ENCLOSURE 1 i

16. Related to surveillances:

a.

The definition of " Time Intervals" in Section 1.0 may be interpreted to require the conduct of the activity only within the specified time frame.

For example Tech Spec 4.4.1 requires "The pH of the primary coolant shall be measure weekly." With the current time interval definitions, this could be interpreted to mean that it is only allowed to be conducted between the seventh and tenth day since the last measurement, rather than any time before the tenth day since the last measurement.

Provide appropriate definitions or modify such Specifications to avoid this possible misinterpretation (ANSI 15.1 may provide additional guidance in this regard).

b.

Additionally, if the extended interval is used for a particular surveillance test, a shorter interval should be used as soon afterwards as possible to adhere to the average.

Provide verification that your procedures specify such policy or reason why it is not needed (e.g., incorporated in Technical Specifications or administratively controlled).

17. Technical Specification 2.1 provides Safety Limits based on preventing Departure from Nucleate Boiling. NUREG-1313, " Safety Evaluation Report Related to the Evaluation of Low-Enriched Uranium Silicide-Aluminum Dispersion Fuel for use in Non-power Reactors," provides data and temperature limits for various fuels. The NRC has found temperature limits of 530 degrees C as acceptable fuel and cladding temperature limits based on NUREG-1313.

In the " Bases" provide verification and references that the proposed Safety Limits are consistent with this acceptance criteria or other analyses which verify that fuel integrity will be maintained.

18. Technical Specification 2.1.1.2 has a maximum value of reactor coolant inlet temperature at 2 MW. Describe how this also applies at different power levels (e.g., up to the 2.4 MW safety limit in the previous Specification 2.1.1.1).

If the reactor coolant inlet temperature value should be decreased for higher power levels and increased for lower power levels, provide the relationship. Should the reference to the 2 MW be deleted since the inlet temperature limit should apply to all power level operations?

19. Technical Specification 2.1.1 " Bases" specifies "...at the licensed power level of 2 MW," but power levels of 2.4 MW are specified as a safety limit. Describe how the safety limit relates to licensed power level.

For example, for any transient condition from the 2 MW licensed power level, DNB will not be reached as long as maximum power is maintained below 2.4 MW.

If appropriate, provide a reference to the analyses that were performed to establish this safety limit. Should the phrase be i

deleted as the prevention of DNB is for all the conditions combined up to 2.4 MW?

l i

REQUEST FOR ADDITIONAL INFORMATION ENCLOSURE 1 l

20. Technical Specification 2.1.2 " Bases" specifies "...at a pool depth of 23.54 feet," but no reference to the power level or pool temperature portion of the Specification.

In the bases, describe how power level and pool temperature affect this safety limit. Provide a brief description of the analyses that were performed to establish this safety limit i

and/or, if appropriate, a reference for this analyses.

21. Technical Specification 2.2.1 " Applicability" should include the fact that this Specification is for forced convective conditions.

j 22.

a.

Technical Specification 2.2.2 " Applicability" should include that it is for natural convective conditions.

b.

Provide in the " Bases" a brief description of the safety analyses which establish and verify the acceptability of these LSSSs and/or, l

if appropriate, references for these associated analyses.

4 23.

a.

Technical Specification 3.1 does not provide a limit for shim safety blade reactivity worth although Technical Specification 4.1.1 seems to appropriately require such a surveillance. Therefore, provide a limit for shim safety blade reactivity worth and appropriate acceptance criteria for 4.1.1, or justification for providing no limit.

b.

Is the only safety analysis limit the reactivity insertion rate limit

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of 3.2.4?

24. Non-power reactors should specify appropriate reactivity coefficients for fuel temperature, moderator temperature, and void volume and a power l

i defect, as used in the safety analyses. The SAR analyses of both routine operation and potential accident scenarios should show that the effect of j

i these core characteristics is considered in the analysis of anticipated event or postulated accident scenarios. The values of the coefficients l

and the power defect are acceptable if they assure that the assumptions and initial conditions of the analyses are enveloped to prevent compromise of the fuel integrity during reictor transients and other applicable accident scenarios. Values for surveillance should be l

specified to provide this assurance. An acceptable schedule for surveillance of reactivity coefficients has been at initial reactor startup and when any change in the reactor core configuration or fuel type requires changes in the Specifications of Section 5.

If needed, these Specifications should be consistent with those of the current Technical Specifications K.3.e(3)(c) and (4)(c). Therefore, provide appropriate reactivity coefficient limits and associated surveillance requirements.

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REQUEST FOR ADDITIONAL INFORMATION ENCLOSURE 1 i

25. Limiting Conditions of Operation (LCOs) and surveillances should be specified for certain fuel parameters or characteristics, such as the following:

l a.

No operation with damaged fuel except to locate such fuel. The i

definition of damaged fuel may specify limits on longitudinal growth, bowing, or bending and limits on detectable amounts of fission products that could escape through the primary barrier. Provide appropriate Specification to preclude operation with damaged fuel consistent with the current Technical Specification K.I.a, or justification why such is not needed.

l b.

Periodic visual inspection of fuel should reference fuel manufacturers' guidance or recommendations for detecting deterioration, if available. The intervals and methods of fuel inspection should be specified in Section 4 of the Technical l

Specifications. The ptrpose of inspection is to detect cladding deterioration that results from erosion, corrosion, or other damage.

As background, inspections for reactors with plate fuels have not been required by Technical Specification except for higher power reactors that frequently refuel. However, for reactors that remove l

plate fuel from service because the fuel has reached its burnup i

limit, there should be a requirement to inspect representative fuel elements (e.g., 1 in every 10) for excessive corrosion / erosion, i

mechanical wear or damage, or plate swelling.

In all cases, the l

t Specification should describe briefly how the inspection will be l

performed. Therefore, provide appropriate Specification to inspect fuel or justification why such is not needed.

26. For Technical Specification 3.1.1 " Bases", provide a brief description of the safety analyses which establish and demonstrate the acceptability of each of the reactivity limits, and/or, if appropriate, a reference for associated safety analyses.

j

27. For Technical Specification 3.2.1, a.

Include the operating mode in which the different safety channels are required to be operable, or provide rationale why it is not needed.

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b.

Are there interlock functions which require Specification (e.g.,

Technical Specification F.2.e Control Element Withdrawal Interlocks and Technical Specification F.2.f Servo System Control-Interlock)?

1.

Provide justification for not'needing such Specifications or provide them in response to this request.

ii. Also provide associated surveillance requirements consistent with the current Technical Specification K.3.f(2).

l c.

Provide an indication of which of the parameters in these Specification require and have safety related instrumentation which must be operable.

REQUEST FOR ADDITIONAL INFORMATION ENCLOSURE 1

28. For Table 3.1, a.

Should the coolant flow rate function be automatic scram at 51600 gpm? Provide change or reason why not needed.

b.

Provide a requirement for manual scram or reason why not needed.

c.

Should backup scrams (e.g, those at experimental stations) be in the Technical Specifications? Provide reason or Specification, as appropriate.

d.

Pool water level scram is at 2" max decrease.

Provide in the Specification from what reference point this setpoint is established to ensure that the LSSS is satisfied.

29. For Table 3.2, would it be appropriate to include a column for applicable operating mode Table 3.2?

Provide reason and reference to safety analyses which establish the need or not for this requirement.

30. For Technical Specification 3.2.2, to assure that this Specification is verified against the correct surveillance provide a phrase at the end of the Specification' that states "in accordance with Technical Specification 4.1.1 & 2" and/or other surveillance which verifies this function, or provide a reason why this is not needed.

31.

For Technical Specification 3.2.3, to assure that this Specification is verified against the correct surveillance provide a phrase at the end of the Specification that states "in accordance with Technical Specification 4.2.5 or 6," and/or other Technical Specification surveillance or test which verifies this function, or provide a reason why it is not needed.

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32. For Technical Specification 3.2.4, a.

Provide in the bases the rational to assure that this Specification will not be exceeded (e.g., the maximum allowed rod worth of each rod j

for given speed of withdrawal). Verify exactly what the specification applies to (e.g., does it apply to all rods or to each individually?).

b.

Describe why surveillance requirements to verify this limiting condition of operation are not needed, and where and how they are accomplished. Alternatively provide a Technical Specification l

requirement to accomplish this surveillance.

l 33.

For Technical Specification 3.2 " Bases," provide a brief description of the analyses which establish and demonstrate the acceptability and/or associated references as appropriate for all components required by this Specification.

REQUEST FOR ADDITIONAL INFORMATION ENCLOSURE 1 34.

For Technical Specification 3.3 " Specification,"

a.

Provide the " Applicability" to include other appropriate modes or conditions (e.g., during fuel movement and handling of other radioactive materials), or reason why it is not needed.

b.

Should other modes requirement also apply to other radiation monitors in Table 3.27

35. Technical Specifications 3.3.2 and 3.4 reference a containment function rather than confinement as referred to in Technical Specification 4.3.

Clarify here and through out Technical Specifications the confinement or containment function. ANSI 15.1 may provide additional guidance in this regard if needed.

36.

For Technical Specification 3.3 " Table 3.2," provide which monitors function to meet the requirements for a " constant air monitoring unit located in the containment building" and "one radiation monitor on the ground floor level of the reactor building."

37.

For Technical Specification 3.4, a.

These sort of Specifications are usually applicable during irradiated fuel and control rod movement and inspection activities, and other conditions where reactivity increases or radioactive material releases are credible (ANSI 15.1 may provide additional guidance in this regard).

Either provide rationale based on safety analyses that the containment and emergency exhaust systems are not required during these sort of activities, or provide changes to the Specification to ensure that the containment and emergency exhaust systems are available for applicable activities in addition to the already specified reactor operations.

b.

Provide the instrumentation and conditions for " initiation system for containment isolation." Should these have associated Technical Specifications?

If not, provide the reason.

If additional Specifications are appropriate, provide them.

In either case, refer to and use the safety analyses which use this system in mitigation of radioactive releases.

c.

The Specification of "All isolation valves" is somewhat open ended.

Provide some identification of the valves that are specifically required to avoid future confusion.

38. Technical Specification 3.5 " Applicability" refers to resistivity rather than conductivity per Technical Specification 3.5.2.

Provide consistent terminology or provide the correlation between resistivity and conductivity.

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h 8-ENCLOSURE 1 j

REQUEST FOR ADDITIONAL INFORMATION L

i j

39. For Technical Specification 3.5, j

a.

Provide discussions in the " Objective" and " Bases" of the function of maintaining radioactive contaminants in the pool water at acceptable levels as a function of the conductivity Specification, as i

i appropriate.

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b.

Further, the " Applicability" section refers to " radioactivity," but i

the Specification does not directly limit radioactivity. The prompt detection of fission products from failed fuel or experiment malfunctions may be determined from LCOs that limit radioactivity in both the primary and secondary coolants. The Technical Specifications should require periodic sampling and appropriate t

l analyses to detect and quantify radioactivity in both the primary and secondary coolant. The coolant should be sampled for gross activity 4

on a short interval, for example, weekly, and sampled for isotope j

identification on a longer interval, for example, quarterly. The i

purpose of this LCO should be to detect deterioration of components in the primary coolant loop, and leakage in a heat exchanger into the secondary coolant loop.

It should also provide assurance that l

the safety analyses assumptions for radioactive content and 10 CFR Part 20 requirements are satisfied. These Specifications j

j should be stated in such a way that significant changes in l

radioactivity, as defined in the SAR, trigger remedial action. The j

method could be a radiation detector placed in the primary coolant flow loop or a strategically located continuous air monitor in the reactor room or a ventilation duct. Temporary substitutions, in case the fission product monitor is inoperable, should follow guidance in l

2 Section 3.7.1 of ANSI /ANS 15.1.

The specified fission product monitor should be able to initiate action, such as a reactor scram or reactor room isolation. The safety analysis should prcvide the bases and describe how fission products are distinguished from other waterborne or airborne radioactivity and be consistent with the current Technical Specification K.3.c(5). Therefore, either provide discussion how the radioactivity is determined and the reason why no i

further Specification on primary and secondary coolant radioactivity

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is needed, or provide proposed Technical Specifications for these purposes.

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40. Technical Specification 3.5.2 specifies conductivity in units of gmhos rather than gmhos/cm. Describe the measured parameter and assure that units are correct for the Specification.

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41. For Technical Specification 3.5 " Bases" provide a description of the analyses for the detailed analyses where the requirements of Technical Specifications 3.5.3 and 4 were used, and/or, if appropriate, references for the analyses. For example, the LOCA analyses that demonstrates that the pool gate need not be in storage for operations at 0.1 MW or below.
42. For Technical Specification 3.6.1 and others that. reference or use 10CFR Part 20 (e.g., 3.6.2, 3.7, and 5.1), note that 10 CFR Part 20 is changing effective January 1, 1994.

For example, changes in values and terminologies have occurred in the new 10 CFR Part 20 (e.g., the Ar-41 m.

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REQUEST FOR ADDITIONAL INFORMATION ENCLOSURE 1 1

effluent level decreased by a factor of 4 and 10 CFR Part 20, Appendix 8, Table II becomes 10 CFR Part 20, Appendix B, Table 2). Provide your plans to assure that the associated procedures, analyses and Specifications will be consistent with both the old and new 10 CFR 20 when they are required, and any changes to the Technical Specifications to ensure that they will continue to apply after the new rule takes effect. Provide needed changes.

43. Technical Specifications 3.6.2.a and b require radiation monitors.

If these requirements are the same as those in Technical Specification 3.3, they should be combined to assure all conditions are properly specified and avoid unnecessary redundancy.

44. Technical Specifications 3.6.2.b limits operation as in 3.6.2.a above.

However in order to avoid confusion, consider repeating the requirement in 3.6.2.b.

45. For Technical Specification 3.6 " Bases," briefly describe the dilution factor calculations or, if appropriate, provide a reference.

Provide this dilution factor calculation including meteorological and wind frequency data and sources, and any associated uncertainty analyses.

46. Technical Specification 3.7 " Objectives" indicates "... liquid effluents will be minimized" when a more specific indication of objective may be

... liquid effluents will be within regulatory limits and consistent with as low as reasonably achievable requirements." Provide rationale for the for the word " minimized" or changes to appropriately indicate objective of this Specification.

47. For Technical Specification 3.8, provide analyses and reference to analyses that verify that the provisions of ANSI 15.1 section 3.8.3 are satisfied. That is, (1) credible experimental failure will not result in radioactive material releases in excess of 10 CFR Part 20 limits and (2) experiments will not be designed such that they would contribute to the failure of other experiments, core components, or barriers to radioactive material release or be caused to fail by a reactor transient.

Provide Specifications for the limits on materials (e.g., fissile or radioactive) contained in experiments that was assumed in the safety analyses, or justification why they are not needed. The provided example of Technical Specifications may provide additional guidance. Additionally, Specifications such as those indicated in Section C.2.a of Regulatory Guide 2.2 have been previously found acceptable.

48. Technical Specification 3.8.1 requires that " Corrosive material shall be doubly encapsulated," but does not include such a requirement for double encapsulation for materials highly reactive with water, or explosive material which are typically required.

Provide such Specification or reason why it is not needed.

49. Technical Specifications 3.8.4 precludes the applicability of 3.8.3.

Can the term "in any quantity" be changed to "in excess of 25 mg?" Provide rationale and assure that safety analyses are consistent with any proposed change in response to this request for additional information.

l

REQUEST FOR ADDITIONAL INFORMATION ENCLOSURE 1 As guidance, for experimental facilities, the upper limit may be 25 mg TNT or its equivalent, as indicated ir. Section C.2.d of Regulatory Guide 2.2.

For the overall reactor facility, the upper limit will be no higher than 100 milligrams TNT or its equivalent unless analyzed in the SAR and approved by NRC.

I

50. A Specification should limit the quantities of unknown materials that could be placed in certain experimental facilities for exploratory

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studies. Conformance with Section C.2.1 of Regulatory Guide 2.2 has been found acceptable for other facilities.

Provide such a specification or the reasons why it is not needed.

51. Specifications that address the failure and malfunction of an experiment and limit the experiment parameters should be included on a case-by-case basis, as discussed in the SAR. The guidance of Section 3.8.3(2) of ANSI /ANS 15.1 should be followed.

For experiments that may off-gas, sublime, volatilize, or produce aerosols, standard assumptions are often specified for calculating the activity that could be released under normal operating conditions, accident conditions in the reactor, and accident conditions in the experiment. Such Specifications ensure conservatism in the safety analysis of the experiment. These Specifica-tions have included such assumptions as (1) if an experiment fails and releases radioactive gases or aerosols to the reactor bay or atmosphere, 100 percent of the radioactive gases or aerosols escape; (2) if an effluent holdup tank isolates on a high radiation signal, at least 10 percent of the radioactive gases or aerosols escape; (3) if the effluent exhausts through a filter with 99 percent efficiency for 0.3 micron particles, at least 10 percent of the vapors escape; and (4) if an experiment fails that contains materials with a boiling point above 130 *F (54 *C), the vapors of at least 10 percent of the materials escape through an undisturbed column of water above the core.

The particular assumptions used, if any, must be derived from the SAR.

Applicable limits for specific experiments are normally not part of the Technical Specifications and should be derived from the experiment safety review.

Provide verification that such review processes are included in the operating procedure requirements of Technical Specification 6.5 or reasons why such review processes are not needed.

52.

For Technical Specification 3.8.6 " Bases," the use of the terms " Safety Limit calculations" and " Safety Limits" may be too limiting.

Should the term be replaced with safety analyses? Provide the reasoning for your response.

53. For Technical Specification 3.9, a.

Provide in the bases a reference that established this fission density limit, b.

Provide an analysis that demonstrates that fissions /cc is equivalent to the NVT from the LEU conversion submittal (letter from T. Tehan, Rhode Island AEC Nuclear Science Center to M. Mendonca, USNRC, dated December 22,1992).

REQUEST FOR ADDITIONAL INFORMATION ENCLOSURE 1

54. Technical Specification 4.0 " Surveillance Requirements" do not allow for deferral (for example when the reactor is in prolonged shutdown conditions or when surveillance can not be completed during operations) or for completion prior to return to operation as recommended in ANSI 15.1.

Provide such requirements or justify why they are not needed.

55. Provide surveillance requirements for shutdown margin and excess reactivity in Technical Specification 4.1.

ANSI 15.1 and the provided example Technical Specifications may provide additional guidance.

56. For Technical Specification 4.1.1 " Bases," provide a brief description and a reference, if appropriate, to the analyses which demonstrate that shim safety blade werth need only be measured when inore than three interior core locations are replaced per 4.1.1.b or whenever a fuel element is replaced in an interior core location with a element fuel mass change of more than 20% per 4.1.1.c.

57.

a.

The first sentence of Technical Specification 4.1.3 infers that reactivity worth of experiments above some unspecified limit will be measured prior to use.

Provide a brief description of the administrative controls for this function, and a description of and reference to the safety analyses which established this limit.

If appropriate, provide a Technical Specification to include specific experiment reactivity worth estimates which require measurement and the bases for that requirement.

b.

Furthe., this appears to be more appropriate to experimental limitations. Should this be included in those Specifications?

58.

For Technical Specification 4.1.3, a.

Provide the specific sections of Technical Specification 3.1 which will be verified against the experimental worths for this Specification.

b.

Provide definition of "those experiments whose safety review indicates a need for..." reactivity worth determination and

" reasonably be expected." For example, would experimental worths of some percentage of the limits of Technical Specification 3.1 provide this assurance? Provide reason for selection criteria which may be l

also contained in facility procedures or Technical Specifications, or the reason such criteria is not needed.

59. Technical Specification 4.2.4 refers to "... operation...shall be verified..." and "...an operation and setpoint verification..." Define

... operation... verification..." or change the verification to assure the appropriate condition (e.g., operating or operable which are already defined) is verified. Alternatively provide the reason why such change is not needed.

60.

For Technical Specification 4.3 verify that surveillance of emergency generator system is required consistent with manufacturer or other appropriate recommendations and include that fact in the " Bases."

Alternatively, provide the reason such is not appropriate.

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REQUEST FOR ADDIT 10NAL INFORMATION ENCLOSURE 1

61. For Technical Specification 4.3.1, a.

The "... operation... shall be tested" is specified but there is no definition of operation. Provide specific definition of operation or use previously defined conditions (e.g., operable) or reason why this l

change is not needed (e.g., operation is defined in facility procedures).

b.

Provide acceptance criteria for this testing. The acceptance criteria should include at least those for the current Technical i

Specifications K.3.b(2)(a) and K.3.b(3).

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62. For Technical Specification 4.3.2 l

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a.

In order to avoid confusion, provide in the " Bases" the conditions to l

l be examined, such as, wear, deterioration, or other physical l

conditions that may impair function, or provide rationale why such is l

not needed.

l b.

In addition, an efficiency test of any filters assumed in the SAR i

should be performed annually or in accordance with manufacturer j

recommendations and acceptance criteria.

63.

If the primary coolant level is not continuously displayed during operation, the primary coolant level in the pool or tank should be verified daily if the reactor is operating or before reactor startup.

1 Provide such Technical Specification requirement or the rationale why it is not needed (e.g., the level is continuously indicated).

64. For Technical Specification 4.4, a.

The surveillance for inservice inspection of the reactor coolant system was one of the considerations in reducing the risk of LOCA and is also recommended by ANSI 15.1.

Provide such a Specification or the reason it is not needed (e.g., controlled by procedure).

b.

If any other inservice inspections of cooling system components are i

identified in the SAR, they should be performed according to manufacturer's recommendations.

If the manufacturer's recommendation i

is not available, the frequency should be as established in the SAR i

from engineering judgement and similar component inservice inspection requirements and experience. Therefore provide proposed Specifications for this function or the reason why it is not needed.

65. For Technical Specification 4.5.3, provide a definition of "The operation i

of all airborne activity monitors shall be checked..." Does this mean a channel check or other surveillance ~l Specify the method to accomplish this check to avoid ambiguity.

j i

66. For Technical Specification 4.8, a.

The surveillance is in the form of a limiting condition of operation.

l Either provide an LC0 in section 3.0 and convert this section to more of a surveillance, or provide cationale why this is not needed.

l J

t i

f

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REQUEST FOR ADDITIONAL INFORMATION ENCLOSURE 1 l

b.

Provide a surveillance frequency for the determination of neutron flux in the Specification, or reason it is not needed.

j

67. Limitations on core configurations are intended to ensure that reactor s

physics and thermal-hydraulic parameters specific to the core are within the limits analyzed in the SAR. Core configuration parameters specified in item 3.1 are already verified, but Technical Specification 3.5.3, 3.5.4 and S.3 are not currently verified by surveillance when there are changes to the core. Therefore, provide a surveillance Specification to verify compliance with all applicable Specifications in those sections when any change occurs in the reactor core configuration.

68. The proposed Technical Specifications (section 3.0 and 4.0) do not contain all the requirements that are currently in the Technical i

Specifications.

Provide the reason why not or propose appropriate Technical Specifications or alternate documents to keep this information.

Include evaluation of the following:

i a.

Technical Specification F.2.a The Reactor Control System includina all safety systems from Tables F.1 and F.2 b.

Technical Specification F.2.b Process Instrumentation c.

Technical Specification 2.c Master Switch d.

Technical Specification F.2.d Power Level Selector Switch e.

Technical Specification F.2.e Control Element Withdrawal Interlocks (Previously requested to be included in section 3) f.

Technical Specification F.2.f Servo System Control Interlock 9

Technical Specification G.2.c h.

Technical Specification K.1.b (particularly the requirement related to neutron source) 1.

Technical Specification K.I.c (particularly the requirement related 1

to communications between the control room and the senior reactor operator) h.

Technical Specifications K.1.e and f i

1.

Technical Specification K.3.d(l)

j. Technical Specification K.3.e.(4)(d) and (e) k.

Technical Specification K.3.f.(1), (2), (3), (5), and (6) 1.

Technical Specification K.4(a) l 69.

For Technical Specification 5.0 " Design Features" a.

The current Technical Specifications provide considerable design detail for the reactor. Ensure that this design detail will be maintained (e.g., in the SAR, Emergency Plan or Security Plan),

specify appropriate design features in Section 5 of the Technical Specifications, provide evidence of appropriate procedural control, or provide the reason it is not needed or an alternate document in which they will be maintained. Topics to consider include:

1.

Technical Specification A.4 Principal Activities 2.

Technical Specification B.1 Reactor Buildina 3.

Technical Specification C Reactor Pool and Primary Coolant System 4.

Technical Specification D Secondary Coolant System 5.

Technical Specification E Reactor Core and Control Elements 6.

Technical Specification G.1 Waste Disposal and Facility

REQUEST FOR ADDITIONAL INF0PJiATION ENCLOSURE 1 7.

Technical Specifications G.2.a b and d 8.

Technical Specification G.3 Other Radiation Monitorina Eouipment 9.

Technical Specification G.4 Hiah Radiation Area

10. Technical Specifications H.1 New Fuel Storace and H.2 Irradiated Fuel Storace (Portions of these are already specified in proposed Technical Specifications)
11. Technical Specification I Experimental Facilities
12. Technical Specification J.5 Site Emercency Plans
13. Technical Specification K.3.a Etta
14. Technical Specifications K.3.c(2) and (3)
15. Technical Specification K.3.e(2)
16. Technical Specification K.3.f(7)
17. Technical Specification K.3.g(4), (5), (6) and (7)
18. Technical Specification K.3.h Fuel Storaae
19. Technical Specification K.4(b) and (c) b.

The proposed Technical Specification 5.0 " Design Featrees" does not provide a description of reactor coolant system per ANSI 15.1 recommendations. Either provide such a description or reference to such a description, or reason why it is not needed (e.g., Technical Specifications in 3.0 and 4.0 provide adequate verification of important reactor coolant system design features). Other important design features, such as syphon protection, should also be described and addressed in response to this request for additional information.

70. For Technical Specification 5.1, provide Figure A.1 which shows the location of the reactor building.
71. Technical Specification 5.3 references the "SAR."

Provide a specific reference (e.g., a letter of transmittal, or an internally maintained document or procedure) 72.

For Technical Specification 5.4, provide the reason for the use of the word containment or change to the confinement or appropriate terminology.

73.

For Technical Specification 5.4.1, the minimum free volume of the containment is usually provided if this value was considered in the safety analyses per ANSI 15.1.

Provide this value in Specification or reason why it is not appropriate.

74. Technical Specifications 5.4.2 usually contain ventilation system parameters that are important to radiological safety and monitoring (e.g., recirculation and exhaust flow rates, and filters) as used in the safety analyses.

Provide such appropriate information or the reason why tt is not needed.

75.

For Technical Specification 5.5, provide the reason for the use of the word " quiescent."

76.

For Technical Specification 6.1:

a.

In the organization chart, the position of Senior Clerk Typist is shown with dotted lines (which usually indicate lines of communication) to the Assistant Director for Reactor Operation. The

REQUEST FOR ADDITIONAL INFORMATION ENCLOSURE 1 usual lines of comunications for this sort of organization are from the Radiation Protection Officer to the Radiation Safety Comittee.

Clarify the lines of comunication in the figure to ensure that the Radiation Protection Officer has a means to raise safety concerns to i

the appropriate level of management.

b.

Define the meaning of the dotted lines in the organization chart.

1 c.

For Technical Specification 6.1.2 define what "at the controls" means, or provide a description of the administrative guidance (procedures) that defines this.

d.

Technical Specifications 6.1.2 and 3, do not specify for another individual at the facility when the senior operator is "readily available on call" per ANSI 15.1 recomendations.

Provide the reason why this is not needed or provide for such coverage in the Specifications.

e.

This section nor any other in the administrative section provides a requirement for a list of reactor facility personnel to be available to the operator in the control room per ANSI 15.1 and the example Technical Specifications from the University of Texas. These lists generally include management personnel, radiation safety personnel, and other operations personnel.

Provide such a requirement, or verify in response to this request for additional information by reference that this list is appropriately required by the emergency plan or other procedure, or provide other rationale why this requirement is not needed.

77. Techna..; Specification 6.2.1 and 2 have non-specific requirements such as " trained in reactor technology," " professionally trained" and

" experience such as may have been gained through employment in a responsible Technical position in the field of health physics." To avoid subjective interpretation of this Specification, provide more specific requirements such as time at related job activities such as those in ANSI 15.4 which is referenced by ANSI 15.1.

Alternativcly, briefly describe such administrative controls and reference procedures which provide this guidance, or provide the reason it is not needed.

78. Technical Specification 6.2.2 should say " senior reactor operators" instead of " senior operators" to be consistent with the rest of Technical Specifications. Also assure this wording is consistently used in the rest of the Technical Specifications.
79. For Technical Specification 6.4.2.a, a.

The Reactor Utilization Comittee will review and approve records.

This is generally a function of facility supervision to review records and is not recommended by ANSI 15.1.

Either provide the reason for the review of records (e.g., this fulfills the Reactor l

Utilization Committee's audit function), or change the proposed Technical Specification.

If this records review is related to audit function per ANSI 15.1, clarify this point in the Specifications.

i

REQUEST FOR ADDITIONAL INFORMATION ENCLOSURE 1 b.

The review of procedures is compared to the current Technical Specification J.l.c which states "(r)eview at least annually the operating and emergency procedures..." Provide a discussion if this requirement, which infers the review of all operating and emergency procedures every year, is to continue.

If so, clarify or provide associated Technical Specifications.

If not, provide reason.

c.

The review of procedures is compared to the current Technical Specification J.l.c which states "(r)eview at least annually... the overall radiation safety aspects of the facility." This infers an audit of radiation safety at the facility which should continue.

Provide Specifications that ensure that this audit function of the overall radiation safety will continue or rationale why it should not.

d.

If there is no audit review function for the Reactor Utilization Committee, provide the reason or such a specification.

80. Technical Specification 6.4.2.d refers to unreviewed safety question requiring a change to the Technical Specifications or facility license.

The regulations (10 CFR 50.59) define unreviewed safety question which may be separate from the need to change Technical Specifications or the license. Provide a change to this Specification to be consistent with the 10 CFR 50.59. This may be accomplished by replacing the word

" requiring" with the words "or require."

4

81. Technical Specification 6.4.2 should ensure that the Reactor Utilization Committee reviews any change to their charter per ANSI 15.1.

Provide such Specification or reason why it is not needed.

82. Technical Specification 6.5.6 requires procedures including those for

" periodic surveillance of reactor instrumentation and safety system..."

System should be plural or provide justification.

83. Technical Specification 6.5.7 requires procedures for "(c)ivil disturbance on or near campus," but does not refer to other security measures as suggested in ANSI 15.1, that is, procedures related to the implementation of the security plan.

Provide the reason why not or include words such as those in the ANSI standard.

84. There are no specific experimental review and approval Technical Specifications proposed.

Provide a discussion of how the responsibilities, review and audit functions, and procedure requirements assure that in addition to guidance of ANSI /ANS 15.1, the review and approval of experiments will be consistent with the guidance provided in Section C.3 of Regulatory Guide 2.2 and Regulatory Guide 2.4, " Review of Experiments for Research Reactors."

If appropriate, provide Technical Specifications which provide this assurance. Such Specifications should make it clear that " established and approved procedures" means written procedures, properly reviewed and approved as required for other operating procedures. Changes to these procedures should follow Section 6.4 of ANSI /ANS 15.1 and be consistent with the requirements sf r

the current Technical Specification K.2.b.

~

REQUEST FOR ADDITIONAL INFORMATION ENCLOSURE 1 b.

The review of procedures is compared to the current Technical Specification J.l.c which states "(r)eview at least annually the operating and emergency procedures..." Provide a discussion if this requirement, which infers the review of all operating and emergency procedures every year, is to continue.

If so, clarify or provide associated Technical Specifications.

If not, provide reason.

The review of procedures is compared to the current Technical c.

Specification J.l.c which states "(r)eview at least annually... the overall radiation safety aspects of the facility." This infers an audit of radiation safety at the facility which should continue.

Provide Specifications that ensure that this audit function of the overall radiation safety will continue or rationale why it should not.

d.

If there is no audit review function for the Reactor Utilization Committee, provide the reason or such a specification.

80. Technical Specification 6.4.2.d refers to unreviewed safety question requiring a change to the Technical Specifications or facility license.

The regulations (10 CFR 50.59) define unreviewed safety question which may be separate from the need to change Technical Specifications or the license. Provide a change to this Specification to be consistent with the 10 CFR 50.59. This may be accomplished by replacing the word

" requiring" with the words 'or require."

81. Technical Specification 6.4.2 should ensure that the Reactor Utilization Committee reviews any change to their charter per ANSI 15.1.

Provide such Specification or reason why it is not needed.

82. Technical Specification 6.5.6 requires procedures including those for

" periodic surveillance of reactor instrumentation and safety system..."

System should be plural or provide justification.

83. Technical Specification 6.5.7 requires procedures for "(c)ivil disturbance on or near campus," but does not refer to other security measures as suggested in ANSI 15.1, that is, procedures related to the implementation of the security plan.

Provide the reason why not or include words such as those in the ANSI standard.

84. There are no specific experimental review and approval Technical Specifications proposed.

Provide a discussion of how the responsibilities, review and audit functions, and precedure requirements assure that in addition to guidance of ANSI /ANS 15.1, the review and approval of experiments will be consistent with the guidance provided in Section C.3 of Regulatory Guide 2.2 and Regulatory Guide 2.4, " Review of Experiments for Research Reactors."

If appropriate, provide Technical Specifications which provide this assurance.

Such Specifications should make it clear that " established and approved procedures" means written procedures, properly reviewed and approved as required for other operating procedures. Changes to these procedures should follow Section 6.4 of ANSI /ANS 15.1 and be consistent with the requirements of the current Technical Specification K.2.b.

REQUEST FOR ADDITIONAL INFORMATION ENCLOSURE 1

85. Technical Specification 6.6 refers to " abnormal occurrence" which was never defined. Should this be " reportable occurrence?" Provide reaso s for the response or change.

}

86. Technical Specification 6.8.1, the current preferred mode of notifying the NRC is through the NRC Operations Center (301) 951-0550 and the NRC Region I by telephone rather than to the regional administrator and telegraph. Change this Specification accordingly or provide reason why it should be different.
87. For Technical Specification 6.8.1.c, provide a definition for abnormal occurrence in Paragraph 1.1 to be consistent with this reporting requirement or should it be " Reportable occurrence"?
88. Technical Specification 6.8.2 should be changed to require a 14-day l

written report for any event specified in 6.8.1.a through d per

[

ANSI 15.1.

Provide the appropriate change or the reason for the proposed l

Specification.

i

89. Technical Specification 6.8.3.b p*ovides for a 30-day report for events that are usually under the provisions of Technical Specifications 6.8.1

(

and 2 per ANSI 15.1.

Provide the reason for this discrepancy or change l

to be consistent with this guidance.

i

90. Technical Specification 6.8.3.c provides for conditions which are generally encompassed in Technical Specification 6.8.3.a requirements.

Either provi e the reason for a separate Specification, combine c into a, or delete c.

l

91. Per ANSI 15.1, 30-day reporting is generally required for permanent changes in the facility organization involving the Director or Assistant Director. Provide this requirement or a reason why it is not needed.
92. Technical Specification 6.8.4.a has a phrase "...and of changes in facility design, performance characteristics and operating procedures related to reactor safety..." which seems to be redundant to the reporting of Technical Specification 6.8.4.e.

Provide the reason for the phrase in 6.8.4.a or delete.

93. Technical Specifications 6.8.4.g and h do not use the usual allowance for a simple statement that the exposures or environmental effluents were within 25 percent of allowable limits in accordance with ANSI 15.1 or some other objective criteria (e.g., 10 percent which is the requirement related to the new 10 CFR Part 20).

Provide the reason for this or provide for this allowed reporting mechanism with an appropriate criteria.

94. Technical Specification 6.9.2 items c, e, f, and h are usually contained in Technical Specification 6.9.1 per ANSI 15.1.

Provide rationale or changes to make consistent.

95. Technical Specification 6.9.2.d, refers to " Abnormal occurrences" which are not defined.

Either define " Abnormal occurrences" and provide reason for this usage, or change it to be " Reportable occurrences.

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i REQUEST FOR ADDITIONAL INFORMATION ENCLOSURE 1

96. Provide your plan for changing procedures and training personnel to i

accommodate this proposed change in Technical Specifications.

l i

97. As a general comment, the " Bases" for each Specification should contain a l

brief description of the analyses and assumptions for which the l

Specification applies, and, if appropriate (i.e., it will not change or l

become out-of-date), a reference to the detailed analysis which established and used the equipment, function or condition of the Specification. Such specific references have proven to be of exceptional value in assuring that the licensing bases is clearly understood and maintained.

98. Another general comment relates to " Applicability" which should include the modes of operation for which the Specification applies.

Ensure that it is clearly understood under what conditions each Specification applies.

99. Typos should be corrected.

For example:

a.

For Technical Specification 3.1.1, correct the term "1.0 5AK/K."

b.

For Technical Specification 3.4, correct a typo where the word

" systems" is separated by column from word " designated."

c.

For Technical Specification 5.4.1 and 2, provide a space between these two items consistent with rest of format.

d.

Figure 6-1 is not marked on the figure.

e.

Technical Specification 6.2.1 states "The Director shall have at an advanced degree..."

Should the "at" be deleted?

f.

A space is required above Technical Specification 6.3.4 title line "4.

Radiation Safety Officer" and similarly for 6.3.3.

g.

Technical Specification 6.5.10 has part of what should b2 applicable to the total Specification included in it. Correct this problem.

h.

Technical Specification 6.9.2.f has an "and" at the end. Should it be after Technical Specification 6.9.2 97 100.

Provide a corrected version of the Technical Specifications after incorporation of responses as necessary.

l i

University of Texas Technical Specification

Appendix A Technical Specifications Revision 1 Docket 50-602 The University of Texas at Austin TRIGA Reactor

\\..

Dece=ber 1990

Rsvisicn 1 Tochnien1 Specificatiens i

Table of Contents i

1.0 DEFINITIONS i

1.1 Certified Operators 5

1.1.1 Senior Reactor Operator 5

1.1.3 Reactor Operator 5

1.2 Channel 5

1.2.1 Channel Test 5

i i

1.2.2 Channel check 5

t 1.2.3 Channel Calibration 5

1.3 Confinement 5

l 1.4 Experiment 5

1.4.1 Experiment, Moveable 6

1.4.2 Experiment, Secured 6

1.4.3 Experimental Facilities 6

1.5 Fuel Element, Standard 6

1.6 Fuel Element, Instrument 6

1.7 Mode; Manual, Auto, Square Wave, Pulse

)

6 1.8 Steady State 1

6 i

1.9 Operable i

6 i

1.10 Operating 7

}

1.11 Protective Action 7

1.11.1 Instrument Channel level 7

g 1.11.2 Instrument System level 7

j 1.11.3 Reactor Safety System Level 7

1.12 Reactivity, Excess 7

1.13 Reactivity Limit 7

)

1.14 Reactor Core, Standard 7

1.15 Reactor Core, Operational 8

1.16 Reactor Operating 8

1.17 Reactor Safety System 8

1.18 Reactor Secured 8

i l

1,19 Reactor Shutdown 8

1.20 Reference Core Condition 9

1.21 Research Reactor 9

1.22 Rod, Control 9

I 1.22.1 Shim Rod 9

1.22.2 Regulating Rod 9

1.22,3 Standard Rod 9

1.22.4 Transient Rod 9

1.23 Safety Limit 9

1.24 Scram Time 10 1.25 Shall, Should and May 10 1.26 Shutdown Margin 10 E

1.27 Shutdown, Unscheduled 10 1.28 Value, Measured 10 1.29 Value, True 10 m

1.30 Survelliance Activities 10 1.31 Surveillance Intervals 11 g

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Revisien 1 Tschnical Sp2cifications 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 12 12 2.1 SATETY LIMIT 12 2.2 LIMITING SATETY SYSTEM SETTINGS 12 2.2.1 Tuel Temperature 12 2.2.2 Power Lavel (Non Po'.re) 12 2.2.3 Reactivity Insert'.on (F:tise) 13 3.0 LIMITING CONDITIONS TCA OPERATION 13 3.1 REACTOR CORE PARAFITERS 13 3.1.1 Excess Reactivity 13 3.1.2 Shutdown Margin 13 3.1.3 Transient Insertions 13 3.1.4 Fuel Elements 14 3.2 REACTOR CONTROL AND SATETY SYSTEM 14 3.2.1 Control Assemblies 14 3.2.2 Reactor Control System 15 3.2.3 Reactor Safety System 15 3.2.4 Reactor Instrument Systes 16 3.3 OPERATIONAL SUPPORT SYSTEMS 16 3.3.1 Vater Coolant Systems 16 3.3.2 Air Confinement Systems 17 3.3.3 Radiation Monitoring Systens 18 3.4 LIMITATIONS ON EXPERIMENTS 18 3.4.1 Reactivity 18 3.4.2 Materials 20

(

4.0 SURVEILIANCE REQUIREMENTS 20 4.1 REACTOR CORE PARAMETERS 20 4.1.1 Excess Reactivity 20 4.1.2 Shutdown Margin 20 4.1.3 Transient Insertion 20 4.1.4 Fuel Elements 20 4.2 REACTOR CONTROL AND SAFETY SYSTEM 20 4.2.1 Control Assemblies 21 4.2.2 Reactor Control System 21 4.2.3 Reactor Safety System 21 4.2.4 Reactor Instrument System 22 4.3 OPERATIONAL SUPPORT SYSTEMS 22 4.3.1 Water Coolant Systems 22 4.3.2 Air Confinement Systems 23 4.3.3 Radiation Monitoring Systems 23 4.4 LIMITATIONS ON EXPERIMENTS 23 4.4.1 Reactivity 23 4.4.2 Materials Page 3 12/90 1

=

Revisien 1 Tochnical Sp3cificctiens 1

5.0 DESIGN FEATURES 24 5.1 SITE AND FACILITY DESCRIPTION 24 5.1.1 Location 24 5.1.2 Confinezent 24 5.1.3 Safety Related Systems 24 5.2 REACTOR COO 1 ANT SYSTEM 25 5.2.1 Natural Convection 25 5.2.2 Siphon Protection 25 i

5.3 REACTOR CORE AND FUEL 25 5.3.1 Fuel Elements 25 5.3.2 Control Rods 25 5.3.3 Configuration' 25 5.4 REACTOR FUEL ELEMENT STORACE 26 5.5 REACTOR POOL CAMMA IRRADIATOR 26 6.0 ADMINISTRATIVE 27 6.1 ORGANIZATION 27 6.1.1 Structure 27 6.1.2 Responsibility 2B

[

6.1.3 Staffing 28 6.1.4 Selection and Training of Personnel 29 6.2 REVIEW AND AUDIT 29 6.2.1 Composition and Qualifications 29 i

6.2.2 Charter and Rules 29 6.2.3 Review Function 29 6.2.4 Audit Function 30 6.3 OPERATING PROCEDURES 30 6.4 EXPERIMENT REVIEW AND APPROVAL 31 6.5 REQUIRED ACTIONS 32 6.5.1 Case of Safety Limit Violatien 32 6.5.2 Event of a Reportable Occurrence 32 6.6 REPORTS 33 i

6.6.1 Operating Reports 33 6.6.2 Special Reports 33 6.7 RECORDS 36 6.7.1 Lifetine of the Facility 36 6.7.2 Five Years or the Life of the Co=ponent 36 6.7.3 One Licensing Cycle 36 APPENDIX A.1 Introduction 37 A.2 Objectives & Bases for Safety Limits 38 A.3 Objectives & Bases for Limiting Conditions for Operations 41 A.4 Objectives & Bases for Surveillance Requirements 51 A.5 Objectives & Bases for Design Features SS 12/90 Page 4

Revision 1 Technical Sp:cificatiens a

1.0 DEFINITIONS 1.1 Certified Operators An individual authorized by the U.S.

Nuclear Regulatory Concission to carry out the responsibilities associated with the position requiring the certification.

1.1.1 Senior Reactor Operator An individual who is licensed to direct the activities of reactor operators.

Such an individual hay be referred to as a class A operator.

1.1.2 Resetor Operator An individual who is licensed to manipulate the controls of a reactor.

Such an individual say be referred to as a class B operator.

1.2 Instrumentation Channel A channel is the combination of sensor, line, amplifier, and output device which are connected for the purpose of sessuring the value of a paraseter.

1.2.1 Channel Test Channel test is the introduction of a signal into the channel for A

verification that it is operable.

1.2.2 Channel Check Channel check is a qualitative verification of acceptable perfornance by observation of channel behavior.

This verification, where

possible, shall include co:parison of the channel with other independent channels or systems nessuring the same variable.

1.2.3 Channel Calibration Channel calibration is an adjustment of the channel such that its output corresponds with acceptable accuracy to known values of the para =eter which the channel nessures. Calibration shall encoepass the entire channel, including equipment actuation, alare, or trip and shall be deemed to include a channel test.

1.3 Confinezent Confine:ent neans an enclosure on the overall facility which controls the nove=ent of air into it and out through a controlled path.

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Revisien 1 Technical Spacificctiens 1.4 Experiment Any operation, component, or target (excluding devices such as detectors, foils, etc.), which is designed to investigate non routine reactor characteristics or which is intended for irradiation within the pool, on or in a beam tube or irradiation facility and which is not rigidly secured to a core or shield structure so as to be part of their design.

1.4.1 Experiment, Moveable A moveable 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.

1.4.2 Experiment, secured A secured expericent is any experiment, experiment facility, or co=ponent of an experiment that is held in a stationary position relative to the reactor by mechanical means. The restraining force must be substantially greater than those to which the experiment night be subjected by hydraulic, pneumatic, buoyant, or other forces which are normal to the operating environment of the experiment, or by forces which can arise as a result of credible conditions.

1.4.3 Experimental Facilities Experimental facilities shall mean rotary specimen rack, pneumatic transfer tube, central thimble, beam tubes and irradiation facilities in the core or in the pool.

1.5 Fuel Ele =ent, Standard A fuel elenent is a single TRICA element of standard type. Tuel is U-ZrH clad in stainless steel clad.

Hydrogen te zirconium ratio is no:inal 1.6.

1.6 Tuel Element, Instrumented special fuel elecent fabricated for An instrumented fuel element is a te=perature seasurement.

The element shall have at least one thercoccuple embedded in the fuel near the axial and radial nidpoints.

1.7 Mode; Manual. Auto, Pulse, Square Wave Each mode of operation shall mean operation of the reactor with the mode selection switches in the manual, auto, pulse or square wave position.

1.8 Steady-state Steady-state rede operation shall mean any operation of the reactor with the code selection switches in the nanual, auto or square wave pcsition.

The pulse code switch will define pulse operation.

PaSe 6 12/90

Revision 1 Technical Sp3cificatisns 1.9 Operable Operable means a component or system is capable of performing its intended function.

1.10 Operating Operating means a component or system is performing its intended function.

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

1.11.1 Instrument Channel level At the protective instrument channel level, protective action is the generation and transmission of a trip signal indicating that a reactor variable has reached the specified limit.

1.11.2 Instrument' System Level At the protective instrument system level, protective action is the Eeneration and transmission of the command signal for the safety shutdown equipment to operate.

1.11.3 Reactor Safety System Level f

At the reactor safety system level, protective action is the operation of sufficient equipnent to immediately shut down the reactor.

I 1.12 Reactivity, Excess Excess reactivity is that amount of reactivity that vould exist if all the control rods were noved to the maxicum reactive condition frc= the point where the reactor is exactly critical.

1.13 Reactivity Limits The reactivity licits are those limits imposed on the reactor core excess reactivity. Quantities a e referenced to a reference core condition.

l Fage 7 12/90 l

Rsvisien 1 Tochnical Sp3cifiesticns i

1.14 Reactor Core, Standard A standard core is an arrangement of standard TRICA fuel in the reactor grid plate and say include installed experiments.

1.15 Reactor Core, Operational An operational core is a standard core for which the core parameters of excess reactivity, shutdown margin, fuel temperature, power calibration, and reactivity worths of control rods and experiments have been de te rmined to satisfy the requirements set forth in the Technical Specifications.

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

1.17 Reactor Safety Systems Reactor safety systems are thqse systems, including their associated input channels, which are designed to initiate automatic reactor protection or to provide information for initiation of manual protective r

action.

i 1.18 Reactor Secure The reactor is secure when:

1.18.1 Suberitical :

There is insufficient fissile material or moderator present in the i

reactar, control rods or adjacent experiments, to attain criticality under opticu= available conditions of moderation and reflection, or 1.18.2 The following conditions exist :

a. The ninieu: number of neutron absorbing control rods are fully inserted in shutdown position, as required by technical specifications.
b. The console key switch is in the off position and the key is l

recoved from the lock,

c. No work is in progress involving core fuel, core structure, l

installed control rods, or control rod drives unless they are physically decoupled from the control rods.

d. No experiments are being noved or serviced that have. on covenent, a reactivity vorth equal to or exceeding one dollar.

f Page S 12/90

Revisien 1 Technical Specificatiens 1.19 Reactor Shutdown The reactor is shutdown if it is subcritical by at least one dollar in the reference core condition with the reactivity of all installed experiments included.

4 1.20 Reference Core Condition The condition of the core when it is at ambient temperature (cold) and the reactivity worth of xenon is negligible (<.30 dollars).

1.21 Research Reactor A research reactor is defined as a device designed to support a self-sustaining neutron chain reaction for research, development, ed$cational, training, or experimental purposes, and which may have provisions for the production of radioisotopes.

1.22 Rod, Control A control rod is a device fabricated from neutron absorbing saterial or fuel which is used to establish neutron flux chan6es and to compensate for routine reactivity loses.

A control rod say be coupled to its drive unit allowing it to perform a safety function when the coupling is disengaged.

1.22.1 Shin Rod A shim rod is a control rod having an electric motor drive and scram capabilities.

1.22.2 Regulating Rod A regulating rod is a control rod used to maintain an intended power level and may be varied manually or by a servo-controller.

The regulating rod shall have scram capability.

1.22.3 Standard Rod The regulating and shin rods are standard control rods.

1.22.4 Transient Rod i

A transient rod is a control rod used to initiate a power pulse that is operated by a motor drive and/or air pressure.

The transient rod l

shall have scra= capability.

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Revision 1 Technical Spseifientiens

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l 1.23 Safety Limits Safety limits are limits on important process variables which are found l

to be necessary to protect reasonably the integrity of the principal barriers which guard against the uncontrolled release of radioactivity.

The principal barrier is the fuel element cladding.

1.24 Scram Time Scram time is the elapsed time between reaching a limiting safety system set point and a specified control rod movement.

7 1.25 Shall, Should and May The word shall is used to denote a requirement. The word should is used to denote a recommendation.

The word may is used to denote permission, i

neither a requirement nor a recommendation.

I 1.26 Shutdown Margin Shutdown margin shall mean the minimum shutdown reactivity necessary to provide confidence that the reactor can be made suberitical by means of the control and -safety systems starting from any permissible operating condition and with the most reactive rod in its most reactive position, and that the reactor will remain suberitical without further operator action.

f 1.27 Shutdown, Unscheduled An unscheduled shutdown is defined as any unplanned shutdown of the reactor caused by actuation of the reactor safety system, operator error, equipeent salfunction, or a manual shutdown in response to conditions which could adversely affect safe operation, not including shutdowns

[

which occur during testing or check-out operations.

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't 1.28 Value, Measured The ceasured value is the value of a parameter as it appears on the

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output of a channel.

i 1.29 Value. True The true value is the actual value of a parameter.

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e*visten 1 Tochnical Sp cifiestiens 1.30 Surveillance Activities Surveillance activities (except those specifically required for safety when the reactor is shutdown), may be deferred during reactor shutdown, however, they sust be completed prior to reactor startup unless reactor o;tration is necessary for performance of the activity.

Surveillance activities scheduled to occur durir.g an operating cycle which cannot be performed with the reactor operatin5 may be deferred to the end of the cycle.

1.31 Surveillance Intervals l

Maximum intervals are to provide operatienal flexibility and not to reduce frequency.

Established frequencies shall be maintained over the long tern.

Allowable surveillance intervals shall not exceed the following:

1.31.1 5 years (interval not to exceed 6 years).

l 1.31.2 2 years (interval not to exceed 2-1/2 years).

1.31.3 Annual (interval not to exceed 15 months).

1.31.4 Semiannual (interv.1 not to exceed 7 1/2 months).

1.31.5 Quarterly (interval not to exceed 4 nonths).

1.31.6 Monthly (interval not to exceed 6 weeks).

1.31.7 k'eekly (interval not to' exceed 10 days).

1.31.8 Daily (sust be done during the calendar day).

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L Revision 1 Technicc1 Sp3cificatiens l

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS f

2.1 Safety Limit i

Specification (s)

The maximum temperature in a standard TRICA fuel element shall not exceed 1150*C for fuel element clad temperatures less than 500'c and shall not i

exceed 950*C for fuel element clad temperatures greater than 500*C.

Temperatures apply to any condition of operation.

2.2 Limitinr Safety Syster Settin's 2.2.1 Fuel Temperature Specification (s)

The limiting safety system setting shall be 550'C as measured in an instrumented fuel element.

One instrumented element shall be located e

[

in the B or C ring of the reactor core configuration.

2.2.2 Power Level (Manual, Auto, Square Wave) f Specification (s) f l

1he maximum operating power level for the operation of the reactor shall be 1100 kilovatts in the manual, auto and square wave modes.

>(-

l 2.2.3 Reactivity Insertion (Pulse) l Specification (s)

The maximum transient reactivity insertion for the pulse operation of the reactor shall be 2.2% ak/k in the pulse mode.

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Revision 1 Technical Spscificatiens 3.0 LIMITING CONDITIONS FOR OPERATION 3.1 Reactor Core Parameters 3.1.1 Excess Reactivity Specification (s)

Maximum excess reactivity shall be 4.9% Ak/k.

l 3.1.2 Shutdown Margin Specification (s)

The reactor shall not be operated unless the shutdown margin provided by control rods is greater than 0.2% Ak/k with:

a. The reactor in the reference core condition,
b. The most reactive control rod fully withdrawn.
c. All noveable experiments in their most reactive state.

3.1.3 Transient Insertions Specification (s) i

(

Total worth of the transient rod shall be limited to 2.8% Ak/k, and the total withdrawal time for the rod shall not exceed 15 seconds.

~

3.1.4 Fuel Elements Specification (s)

The reactor shall not be operated with fuel element damage except for the purpose of locating and removing the elements.

A fuel elenent shall be considered damaged and must be removed from the core if:

l

a. In neasuring the elongation, the length exceeds the original length by 2.54 mm (1/10 inch).
b. In seasuring the transverse bend, the bend exceeds the original bend by 1.5875 mm (1/16 inch).
c. A clad defect exists as indicated by release of fission products cr visual observation 9

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Revision 1 TGehnicOI Specificatiens 3.2 Reactor Control and Safety System 3.2.1 Control Assemblies Specification (s)

The reactor shall not be operated unless the control rods are operable, and to the

a. Control rods shall not be operable if damage is apparent rod or drive assemblies.
b. The scram time sensured from the instant a simulated signal reaches the value of a limiting safety system setting to the instant that the slowest scrammable control rod reaches its l

fully inserted position shall not exceed 1 second.

c. Maximum reactivity insertion rate of a standard control rod shall be less than 0.2% ak/k per second.

3.2.2 Reactor Control System Specification (s)

The reactor shall not be operable unless the minimum safety interlocks The following control system safety interlocks shall be are operable.

operable:

k Effective Mode Interlocks Number Rod Drive Control Overable Function Manual

  • Pulse
a. Startup Withdrawal 4

prevent rod X

Standard control rods withdrawal for Transient control rod less than 2 counts per see i

b. Simultaneous Withdrawal 4 prevent rod X

Standard control rods withdrawal for Transient control rod two or more rods

c. Non pulse condition 1

prevent withdrawal X i

Transient control rod for drive not down except square wave

d. pulse Withdrawal 3

prevent withdrawal X

Standard control rods of non pulse rods X

e. Transient Withdrawal 1

prevent rod Transient control rod withdrawal for more than 1 kilowatt power

  • Manual mode includes Auto and Square Wave modes I

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Revision 1 Tcchnical Sp2cifiestiens 1

3.2.3 Reactor Safety System 2

Specification (s) i The reactor shall not be operable unless the minimum safety channels are operable.

The following control rod scram safety channels shall l

be operable.

Number Effective Mode Safety Channel Doerable Function Manual

  • Pulse
a. Fuel Temperature 2

Scram at 5550*C X

X

b. Power Level 2

Scram at $1.1 Mw X

Pulse Power 1

Scram at 52000Mw X

c. High Voltage 2

Scram on loss X

X

d. Magnet Current 1

Scram on loss X

X

e. Manual Scram 1

Scram on demand X

X Console Button f

f. Watchdtg Trip 2

Scram on loss of Microprocessor scan rate timer reset X

X d

  • Manual mode includes Auto and Square Wave modes

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f 3.2.4 Reactor Instrument System Specification (s)

A minimum configuration of measuring channels shall be operable.

The followin5 minimum reactor parameter measuring channels shall be l

operable:

Number Effective Mode Measurine Channel Ocerable Manua1*

Pulse

a. Fuel Tecperature 2

X X

b. Power Level 2

X

c. Pulse Power 1

X X

d. Pulse Energy 1
  • Manual mode includes Auto and Square Wave modes Page 15 12/90

Revision 1 Technical Specificaticns 1

l 3.3 Operational Suerort Systees 3.3.1 Water Coolant Systems Specification (s)

I corrective action shall be taken or the reactor shut down if any of

[

the following (a.-d.) reactor coolant conditions are observed:

a. The bulk pool water temperature exceeds 48'C.
b. The water depth is less than 6.5 meters measured from the pool

~

bottom to the pool water surface.

c. The water conductivity exceeds 5.0 paho/cm for the average value during seacurement periods of one month.

l

d. The pressure difference during heat exchanger op ration is less than 7 kPa (1 psig) sensured between the chilled water outlet.

pressure and the pool water inlet pressure to the heat exchanger.

e. Fool water data from periodic measurements shall exist for water pH and radioactivity.

Radioactivity measurements shall include total alpha-beta activity and gamma ray spectrum analysis.

3.3.2 Air Confinement Systems Specification (s)

Corrective action shall be taken or the reactor shut down if any of the following air confinement conditions do not exist:

a. Equip =ent shall be operable to isolate the reactor area by closure of room ventilation supply and exhaust dampers, and i

shutdown of system supply and exhaust fans.

b. The reactor room ventilation system shall have an automatic l

signal to isolate the area if air particulate radioactivity i

exceeds preset values.

c. An auxiliary air purge system to exhaust air from experiment l

systems shall have a high efficiency particulate filter.

)

d. Room ventilation shall require two air changes per hour or exhaust of pool areas by the auxiliary air purge system.

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Rsvisicn 1 Tochnical Spacificcticns 1

3.3.3 Radiation Monitoring Systems Specification (s)

Radiation monitoring while the reactor is operating requires the following minimum conditions :

a. A continuous air monitor (particulate) shall be operable with

^

readout and audible alarm.

The monitor shall sample reactor room air within 5 meters of the pool at the pool access level.

Alarm set point shall be equal to or less than a acasurement concentration of 2 x 10*' pCi/cm3 with a two hour particulate accumulation.

l The particulate continuous air monitor shall be operating when the reactor is operating.

A set point of the monitor will initiate the isolation signal for the air ventilation system.

The particulate air monitor may be out of service for a period of 1 week provided the filter is evaluated daily, and a signal from the argon-41 continuous air monitor is available to provide information for manual shutdown of the HVAC.

1

b. A continuous air monitor (argon-41) shall be operable with i

readout and audible alarm.

The monitor shall sample exhaust stack air from the auxiliary air purge system when the system is operating.

Alarm set point shall be equal to or less than a seasurement concentration of 2 x 10*5 pCi/cm3 for a daily release.

The argon 41 continuous air. sonitor shall be operating when the auxiliary air purge system is operating.

The average annual concentration limit for release at the stack shall be 2 x 10**

pCi/en3.

If the argon-41 sonitor is not operable, operating the reactor with the auxiliary air purge system shall be limited to a period of ten days.

l l

c. Area radiation monitors (gamma) shall be operable with readout sed audible alarn. Alarm set point shall be a measurement value equal to or less than 100 nr/hr.

One area radiation monitor shall be operating at the pool level when the reactor is operating.

Two additional area radiation l

nonitors shall be operating at other reactor areas when the

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reactor is operating.

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3.4 Linitatiens on Experiments p

3.4.1 Reactivity Specification (s)

The reactor shall not be operated unless the following conditions governing experiment reactivity exist.

a. A moveable experiment shall have a reactivity worth less than l

1.00 dollar.

i

b. The reactivity worth of any single secured experiment shall be i

i less than 2.50 dollars.

I; i

c. The total of absolute' reactivity worths of reactor core 1

experiments shall not exceed 3.00 do11ers, including the potential reactivity which might result from malfunction, I

flooding, voiding, or removal and insertion of the experiments.

j 4,

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)

3.4.2 Materials i

Specification (s)

l I

The reactor shall not be operated unless the following conditions-governing experiment materials exist-i f

s. Experiments containing materials corrosive to reactor components, compounds highly reactive with. water, potentially explosive materials, and liquid fissionable materials shall be l

doubly encapsulated.

Guidance for classification of materials j

shall use the " Handbook of 1.aboratory Safety" Tables of Chemical j

Information published by CRC Press.

j i

b. If a capsule fails and releases material which could dassge 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 l

inspection and any corrective action taken shall be reviewed by 2

the Director, or his designated alternate, and determined to be j

satisfactory before operation of the reactor is resumed.

c. Explosive saterials in quantities greater than 25 milligrams shall not be irradiated in the reactor or experimental facilities. Explosive materials in quantities less than 25 j

milligrams may be irradiated provided the pressure produced upon 5

detonation of the explosive has been calculated and/or experimentally demonstrated to be less than the design pressure of the container, i

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d. Each fueled experiment shall be controlled such that the total inventory of iodine isotopes 131 through 135 in the experinent i

is no greater than 750 mil 11 curies and the maximum strontine inventory is no greater than 2.5 mil 11 curies.

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e. Experiment materials, except fuel materials, which could off-gas, sublime, volatilize, or produce aerosols under (1) normal operating conditions of the experiment or reactor, (2) credible accident conditions in the reactor, (3) 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 room or the atmosphere, the airborne concentration of radioactivity averaged over a year would not exceed the occupational limits for maximum permissible concentraticn.

f. In calculations pursuant to e. above, the following assumptions shall be used:

(1) If the effluent from an experimental facility exhausts through a

holdup tank which closes automatically on high radiation level, at least 10% of the gaseous activity or aerosols produced will escape.

(2) If the effluent from an experimental facility exhausts through a filter installation designed for greater than 991 efficiency for 0.25 micron particles, at least 10% of these vapors can escape.

(3)

For materials whose boiling point is above 55'c and where vapors l

formed by boiling this material can escape only through an undisturbed column of water above the core, at least 101 of these vapors can escape.

(4) 1.imits for maximum permissible concentrations are specified in the appropriate section of 10CTR20.

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Revision 1 Technic Sp3cifiestiens 6

e i

4.0 SURVEIL 1/SCE REQUIREMENTS j

I 4.1 Reactor Core Paraneters l

t h

4.1.1 Excess Reactivity Specification (s) l 1

Excess reactivity shall be determined annually or after significant I

control rod or reactor core changes.

t 4.1.2 Shutdown Margin Specification (s) l Shutdown margin shall be determined annually or after significant 4

control rod or reactor core changes.

i 4.1.3 Transient Insertion l

J Specification (s) l Transient rod function shall be evaluated annually or after I

significant control rod or reactor core changes.

The transient rod drive and associated air supply shall be inspected stinually, and the 1

drive cylinder shall be cleaned and lubricated annually.

l A comparison of pulse data shall be made with previous measurements at f

A annual intervals or each time the interval to the previous seasurement

~

exceeds the annual interval.

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4.1.4 Fuel Elemerts Specification (s)

The reactor fuel elements shall be examined for physical damage by a i

visual inspection, including a check of the dimensional seasurements, j

cade at biennial intervals.

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Revisien 1 i

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a i

i 4.2 Resetor Control and Saferv System l

4 4.2.1 Control Assemblies j

Specification (s)

Control rod worths shall be determined annually or after significant control rod or reactor core changes, and j

a. Each control rod shall be inspected at biennial intervals by visual observation.

l l

2

b. The scram time of a scrammable control rod shall be seasured l

annually or after maintenance to the control rod or drive.

c. The reactivity insertion rate of a standard control rod shall be measured annually or after maintenance to the control rod or' j

drive.

'l j

4.2.2 Reactor Control System

)

i Specification (s) 4 The minimum safety interlocks shall be tested at semiannual intervals cr after repair or modification.

J i-4.2.3 Reactor Safety System Specification (s)

?

The minimum safety channels shall be calibrated annually or after repair or modifications.

A channel test shall be done prior to each

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days operation, after repair or modifications, or prior to each i

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extended period of operation.

9 4.2.4 Reactor Instrument System i

Specification (s)

The minimum configuration of instrument channels shall be calibrated l

l l

annually or after repair or modification.

Calibration of the power seasuring channelt shall be by the calorimetric method.

A channel l

check and channel test of the fuel temperature instrument channels and j

power level instrument channels shall be made prior to each days operation or prior to each extended period of operation.

i J

Page 21 12/90

Revision 1 Tochnical Sp2cificatiens 4.3 Operational Suenort Systems 4.3.1 Vater Coolant Systems Specification (s)

The following acasurements shall monitor the reactor coolant conditions:

a. The pool temperature channel shall have a channel calibration annually, channel check monthly and will be monitored during reactor operation.

a channel calibration

b. The pool water depth channel shall have annually, channel check monthly and will be monitored during reactor operation.
c. The water conductivity channel shall have a channel calibration annually and pool water condoctivity will be seasured weekly,
d. The pressure difference channel shall have a channel test prior to each days operation, after repair or modifications, or prior to each extended period of operation of the heat exchanger and will be monitored during operation.
e. Measure pool water pH with low ion test paper or equivalent quarterly.

Sample pool water radioactivity quarterly for total alpha-beta activity.

Analyze pool water sample by gamma spectroscopy annually for isotope identification.

4.3.2 Air Confinement Systems Specification (s)

The following actions shall demonstrate the air confinement conditions:

a. Annual examination of door seals and isolation dampers.
b. Monthly functional tests of air confinement isolation.
c. Monthly check of the auxiliary air purge system valve alignments for experimental areas.
d. Daily check of ventilation system alignment for proper exhaust conditions prior to reactor operation.

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j Revisien 1 Tochnical Spacificatiens 1

4.3.3 Radiation Monitoring Systems Specification (s)

The following conditions shall apply to radiation monitoring systems:

i

s. Calibrate particulate air monitor at semiannual intervals and l

check operability weekly.

b. Calibrate argon-41 air monitor at biennial intervals and check i

operability monthly.

l

c. Calibrate area radiation monitors at semiannual intervals and l

check operability weekly prior to reactor operation.

l l

4.4 Limitations on Experinents 4.4.1 Reactivity l

Specification (s)

The reactivity of an experiment shall be sessured before an experiment is considered functional.

l 4.4.2 Materials Specification (s)

Any surveillance conditions or special requirements shall be specified as a part of the experiment approval.

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12/90 Fase 23

Revision 1 Tochnical Spscifications I

5.0 DESIGN FEATURES 5.1 Site and Facility Descrietion 4

f 5.1.1 Location Specification (s)

s. The site location is in the northeast corner of The University of Texas at Austin Balcones Research Center.
b. The TRICA reactor is installed in a designated room of a i

building constructed as a

Nuclear Engineering Teaching Laboratory.

c. The reactor core is assembled in an above ground shield and pool structure with horizontal and vertical access to the core.
d. License areas of the facility for reactor operation shall consist of the room enclosing the reactor shield and pool structure, and the adj acent area for reactor control.

(room 1.104, corridor 3.200; and rooms 3.202, 3.204, and 3.208) t 5.1.2 Confinement Specification (s)

[

a. The reactor room shall be designed to restrict leakage and will f

have a minimum enclosed air volume of 4120 cubic meters,

b. Ventilation system should provide two air changes per hour and

{

shall isolate air in the reactor area upon detection of a limit signal related t o the radiation level.

c. An air purge system should exhaust experiment air cavities and i

shall be filtered by high efficiency particulate absorption

filters,
d. All exhaust air from the reactor area enclosure shall be ejected vertically upward at a point above the facility roof level.

5.1.3 Safety Related Systems Specifications Any modifications to the air confinement or ventilation system, the reactor shield, the pool or its penetrations. the pool coolant system, the core and its associated support structure, the rod drive mechanisms or the reactor safety system shall be made and ' tested in accordance with the specifications to which the systems were originally designed and fabricated.

Alternate specifications may be approved by the Nuclear Reactor Committee.

A system shall not be considered operable until after it is tested successfully.

12/90 Page 24

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Revision 1 Technical Sp2cificcticns 5.2 Resetor Coolant System 5.2.1 Natural Convection Specification (s)

The reactor core shall be cooled by natural convection flow of water.

5.2.2 Siphon Protection Specification (s)

Fool water level shall be protected by holes for siphon breaks in pool water system pipe lines.

5.3 Resetor Core and Tuel 5.3.1 Fuel Ele =ents Specification (s)

The standard, TRIGA fuel element at fabrication shall have the following characteristics:

a. Uranium content: 8.5 Ut1 uranium enriched to a nominal 19.7%

Uranium-235.

(

b. Zirconium hydride atom ratio: nominal 1.6 hydrogen to zirconium, ZrH.

x

c. Cladding: 304 stainless steel, nominal.020 inches thick.

5.3.2 control Rods Specification (s)

The shic, regulating, and transient control rods shall have scra:

capability, and

a. Include stainless steel or aluminum clad and say be followed by air or alu=inum, or for a standard rod may be followed by fuel with stainless steel clad.
b. Contain borated graphite, B C powder, or boron and its compounds 4

in solid form as a poison.

c. The transient rod shall have a cechanical licit. An adjustable limit vill allow a variation of reactivity insertions.

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d. Two shim rods, one regulating rod and the transient rod are the

)

einieu= control rods.

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nevision 1 Technical Sp2cificaricns 5.3.3 Configuration i

Speciffcation(s)

The reactor shall be an arrangement of core single grid positions j

occupied by fuel elements, control rods, and graphite elements.

l Single element positions may be occupied by voids, water or experiment facilities.

Special multielement positions or single element positions may be occupied by approved experiments.

5.4 Erneter Fuel Element Storare Specification (s)

a. All fuel elements shall be stored in a geometrical array where the effective multiplication is less than 0.8 for all conditions of moderation.
b. Irradiated fuel elements,and fueled devices shall be stored in t

an array which will permit sufficient natural convection cooling by water or air such. that the fuel element or fueled device i

temperature vill not exceed design values.

t i

5.5 Egaetor Pool Irradiator Specification (s)

The irradiator assembly shallbe an experiment facility.

a. A 10,000 Curie gamma irradiator may be located in the reactor pool. The irradiator isotope shall be cobalt-60.

l

b. Location of the assembly shall be at a depth of at least 4.5 l

and at a distance of at least 0.5 meters from the reactor meter 7 core structure.

I requirements shall monitor pool fog a pool water activity of 2.5x10'gater c.

Pool vater sample pCi/cm source leakage.

At the gamma irradiator corponents shall be tested to locate and r

remove any leaking source.

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Revision 1 Tochnicc1 Sp2cifienticns 6.0 ADMINISTRATIVE CONTROLS 6.1 Orrauiration 6.1.1 Structure The facility shall be under the control of the Director or a supervisory Senior Reactor Operator.

The management for operation of the facility shall consist of the or5anizational structure established as follows:

President of The University of Texas at Austin l

Executive Vice President and Provost

................. 1evel 1 Radiation Dean College of Nuclear Safety Engineering Reactor Co==ittee Co==ittee Radiation Safety Officer I

I Chairman Department of Mechanical Engineering 1

1 1

Director Nuciaar Engineering Teaching Laboratory

................. 1evel 2 i

Reactor Supervisor

                              • "'"I'"'I Health Physicist

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Reactor Operators, 1

Technicians, Others j

................. 1evel 4 f

Responsibility ---

Co==unication i

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Technical Spscificctiens Revision 1 J

I 6.1.2 Responsibility 1

l The Director shall be responsible to the Dean of the College of j

Engineering and the Chairman of the Department of Mechanical Enginer.ing for safe operation and maintenance of the reactor and its i

associated equipnent.

The Director or a supervisory Senior Reactor l

Operator shall review and approve all experiments and experimental procedures prior to their use in the reactor.

Individuals of the i

management organization shall be responsible for the policies and i

operation of the facility, and shall be responsible for safeguarding the public and facility personnel from undue radiation exposures and l

1 1

for adhering to the' operating license and technical specifications.

t 6.1.3 Staffing l

The minimum staffing when the reactor is not shutdown shall be:

1

a. A reactor operator in the control room.
b. A* second person in the facility area that can perform prescribed written instruct?.ons.

Unexpected absence for two hours shall require ine-diate action to obtain an alternate

^

person.

i

c. A senior reactor operator readily available.

The available operator should be within thirty minutes of the facility and reachable by telephone.

(

Events requiring the direction of a senior reactor operator shall be:

a. All fuel element or control rod relocations within the reactor I

core region.

(

b. Relocation of any experiment with a reactivity worth of greater j

i than one dolla::.

c. Recovery from an unscheduled shutdown or significant power reduction.
d. Initial startup and approach to power.

A list of reactor facility personnel by name and telephone number j

shall be available to the operator in the control room.

The list shall include:

a. Management personnel.
b. Radiation safety personnel.

j

c. Other operations personnel.

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. Technical Spscificaticns Revision 1 l

6.1.4 Selection and Training of Personnel i

The selection, training and requalification of operators shall meet exceed the requirements of American National Standard for or l

Selection and Training of Personnel for Research Reactors ANSI /ANS -

Qualification and requalification of licensed operators shall 15.4.

2 be subject to an approved NRC (Nuclear Regulatory Commission) program.

i 6.2 Feview and Audit i

6.2.1 Composition and Qualifications Nuclear Reactor Committee shall consist of at least three (3)

A members appointed by the Dean of the College of Engineering that are knowledgeable in fields which relate to nuclear safety.

The i

university radiological safety officer shall be a member or an ex-.

officio member.

The committee vill perform the functions of review and audit or designate a knowledgeable person for audit functions.

6.2.2 Charter and Rules The operations of the Nuclear Reactor Committee shall be in accordance with an established charter, including provisions for:

l l

a. Meeting frequency (at least once each six months).
b. Quorums (not less than one-half the membership where the l

l operating staff does not represent a majority).

l

c. Dissemination, review, and approval of minutes.

=

d. Use of subgroups.

t l

6.2.3 Review Function i

The review function shall include facility operations related to i

reactor and radiological safety.

The following items shall be 1

reviewed:

s. Determinations that proposed changes in equipment, systems, l

1 tests, experiments, or procedures do not involve an unreviewed safety question.

j

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

~l significance.

c. All new experiments or classes of experiments that could affect reactivity or result in the release of radioactivity.
d. Changes in technical specifications or license.

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Revisien 1 Tcchnical Sp3cificcticns

e. Violations of technical specifications or license.

i

f. Operating abnormalities or violations of procedures having safety significance.

I

g. Other reportable occurrences.
h. Audit reports.

6.2.4 Audit Punction The audit function shall be a selected exstination of operating l

records, logs, or other documents.

An audit vt11 be by a person not directly responsible for the records and may include discussions with i

co5nizant personnel or observation of operations.

The following items shall be audited and a report made within 3 nonths to the Director and Nuclear Reactor Committee

a. Conformance of facility operations with license and technical i

specifications at least once each calendar year.

1

b. Results of actions to correct deficiencies that may occur in reactor facility equipment, structures, systems, or methods of operation that affect safety at least once per calendar year.
c. Function of the retraining and requalification program for reactor operators at least once every other calendar year.

I k

d. The reactor facility emergency plan and physical security plan, and implementing procedures at least once every other year.

6.3 Operatine Procedures Vritten operating procedures shall be prepared, reviewed and approved by supervisory Senior Reactor Operator and the Nuclear the Director or a Reactor Co=mittee prior to initiation of the following activities:

Startup, operation, and shutdown of the reactor.

a

b. Fuel loading, unloading and movement in the reactor,
c. Routine maintenance of major components of systems that could have an effect on reactor safety.
d. Surveillance calibrations and tests required by the technical specifications or those that could have an effect on reactor safety.
e. Administrative controls for operation, maintenance,' and the conduct of experiments or irradiations that could have an effect on reactor safety.

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Revision 1 Technical Sp3cificatiens

. ith applicable

f. Personnel radiation protection, consistent w

t regulations or guidelines, and shall include a sanagement emmitment and programs to maintain exposures and releases as low as reasonably achievable.

g. Implementation of required plans such as the emergency plan or physical security plan.

Substantive changes to the above procedures shall be made effective after approval by the Director or a supervisory Senior Reactor Operator and the Nuclear Reactor Committee.

Minor modifications to the original procedures which do not change the original intent may be made by a senior reactor operator but the modifications must be approved by the Director or a supervirory Senior Reactor Operator.

Temporary deviations from the procedures may be made by a senior reactor operator in order to deal with special or unusual circumstances or conditions.

Such deviations shall be documented and reported to the Director or a supervisory Senior Reactor Operator.

6.4 Exrerirent Review and Aceroval j

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All new experiments or classes of experiments shall be approved by the Director or a Supervisory Senior Reactor Operator and the Nuclear s

Reactor Operations Committee.

a. Approved experiments shall be carried out in accordance with

[

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established and approved procedures.

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b. Substantive changes to previously approved experiments shall require the same review as a new experiment.
c. Minor changes to an experiment that do not significantly alter supervisory senior reactor the experinent may be made by a operator.

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Revision 1 Technical Sp3cifiestiens 6.5 Recuired Actions l

1 6.5.1 Action to be Taken in Case of a Safety Limit Violation In the event of a safety limit violation, the following action shall be taken:

s. The reactor shall be shut down and reactor operation shall not be resumed until a report of the, violation is prepared and i

authorization to restart by the Nuclear Regulatory Commission (NRC) is issued.

b. The safety limit violation shall be promptly reported to the Director of the facility or a designated alternate.
c. The safety limit violation shall be subsequently reported to the NRC.
d. A safety limit violation report shall be prepared and submitted to the Nuclear Reactor. Committee.

The report shall describe:

}

(1)

Applicable circumstances leading to the violation l

including, when known the cause and contributing factors. (2)

Effect of the violation on reactor facility components, systems, or structures and on the health and safety of the public, (3) Corrective actions taken to prevent i'ecurrence.

6.5.2 Action to be Taken in the Event of an Occurrence that is

(

Reportable, j

t In the event of a reportable occurrence, the following action shall be taken:

i

a. Reactor conditions shall be returned to normal or the reactor j

=

shutdown.

If it is necessary to shut down the reactor to correct the occurrence, operations shall not be resumed unless authorized by the Director or his designated alternate.

b. Occurrence shall be reported to the Director or his designated alternate and to the Nuclear Regulatory Commission as required.
c. Occurrence shall be reviewed by the Nuclear Reactor Committee at the next regularly scheduled meeting.

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Revision 1 Tcchnical Spacifications 6.6 Reports All written reports shall be sent within the prescribed interval to the NRC, Washington D.C. 20555, Attn: Document Control Desk, with a copy to l

the Regional Ad=inistrator, Region IV.

6.6.1 Operating Reports Routine annual reports covering the activities of the reactor facility during the previous calendar year shall be submitted within three months following the end of each prescribed year.

Each annual operating report shall include the following information:

l y a. A narrative summary of reactor operating experience including the energy produced by the reactor or the hours the reactor was critical, or both.

b. The unscheduled shutdowns including, where applicable, corrective action taken to preclude recurrence.
c. Tabulation of maj or preventive and corrective maintenance operations having safety significance.
d. Tabulation of maj or changes in the reactor facility and i

procedures, and tabulation of new tests or experiments, or both, that are significantly different from those perforned previously, including conclusions that no unreviewed safety questions were involved.

e. A summary of the nature and amount of radioactive effluents released or discharged to the environs beyond the effective control of the university as determined at or before the point l

of such release or discharge.

The summary shall include to the extent practicable an estimate of individual radionuclides i

prssent in the effluent.

If the estinated average release after dilution or diffusion is less than 25%

of the concentration allowed or recommended, a statement to this effect is sufficient.

f. A summary of exposures received by facility personnel and i

visitors where such exposures are greater than 25% of that allowed or recommended.

g. A summarized result of environmental surveys perforned outside the facility.

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t 6.6.2 Special Reports i

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3 6.6.2.1 A written report within'30 days to the NRC of:

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s. Permanent changes in the facility organization involving l'

l Director or Supervisor.

)

b. Significant changes in transient or accident analysis as described in the Safety Analysis Report.

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j 6.6.2.2 1

A report to NRC Operation Center and Region IV by telephone not later than the following working day and confirmed in writing by telegraph j

or similar conveyance to be followed by a written report within 14 5

2 days that describes the circumstances of the event of any of the i

following:

i

s. Violation of fuel element temper 4ture safety limit.

}

b. Release of radioactivity above allowable limits.

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c. Other reportable occurrences.

i Other events that will be considered reportable events are listed in j

this section.

(Note:

Where components or systems are provided in q[,

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failure of components or systems is not considered reportable l

addition to those required by the technical specifications, the j

(

provided that the minimum number of components or systems specified or required perform their intended reactor safety function.)

l

a. Operation with actual safety-system settings for required j

systems less conservative than the limiting safety systes settings specified in the technical specifications.

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b. Operation in violation of limiting ceditions for operation established in technical specification; unless prompt remedial i

action is taken.

c. A reactor safety system component malfunction which renders or l

could render the reactor safety system incapable of performing i

its intended safety function unless. the malfunction or

]

condition is discovered during maintenance tests 'or periods of l

reactor shutdowns.

i

d. An unanticipated cr uncontrolled change in reactivity greater

.i than one dollar.

Reactor trips resulting from a known cause i

are excluded.

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e. Abnormal and significant degradation in reactor fuel, or i

cladding, or both, coolant boundary, or confinement boundary (excluding minor leaks) where applicable which could result in exceeding prescribed radiation exposure limita of personnel or environment, or both,

f. 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.

6.6.2.3 A written report within 90 days after the completion of startup tests or 9 months after initial criticality, which ever is earlier, of the l

startup test program, to the NRC of:

i Characteristics of the reactor such as critical mass, excess l

reactivity, power calibration, control rod calibrations, shutdown margin and experiment facility worths, describing the measured values of the operating conditions including:

a. Total control reactivity worth and reactivity of the rod of highest reactivity worth,
b. Minimum shutdown margin of the reactor both at ambient and operating temperatures.
c. An evaluation of facility performance to date in comparison with design conditions and sessured operating characteristics, and a reassessment of the safety analysis when measurements indicate that there may -be substantial variance from prior analysis submitted with the license applicatio.n.

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Revisicn 1 Tcchnical Spacificatiens l

6.7 Reeerds The records may be in the form of logs, data sheets, or other suitable forms.

The required information may be contained in single or multiple records, or a combination thereof.

6.7.1 Records to be Retained for the Lifetime of the Reactor Facility:

(Note:

Applicable annual reports, if they contain all of the required information, say be used as records in this section.)

a. Caseous and liquid radioactive affluents released to the environs.
b. Offsite environmental monitoring surveys required by technical specifications.

~

c. Events that impact or effect decommissioning of the facility
d. Radiation exposure for all personnel monitored.

Updated drawings of the reactor facility.

e.

6.7.2 Records to be Retained for a Period of at least Five Years or for the Life *of the Component Involved Whichever is Shorter:

l

a. Normal reactor facility operation (supporting documents such as checklists, log sheets, etc. shall be maintained for a period l

of at least one year).

(
b. Principal maintenance operations.

- c. Reportable occurrences.

d. Surveillance activities required by technical specifications.

< e. Reactor facility radiation and contamination surveys where required by applicable regulations.

f. Experiments performed with the reactor, i

l

g. Fuel inventories, receipts, and shipments.
h. Approved changes in operating procedures,
i. Records of meeting and audit reports of the review and audit group.

6.7.3 Records to be Retained for at Least One Licensing Cycle:

Retraining and requalifications of licensed operations personnel.

Records of the most recent complete cycle shall be saintained at all times the individual is employed.

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AFFENDIX A

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,A.1.0 DOCKET 50-602 INFOR AM TION The Technical Specifications of this document depend on the analysis and conclusions of the Safety Analysia Report.

Descriptive information

'j important to each specification is presented in the form of the l

l applicability, obj ective and bases.

This information defines the l

conditions effective for each technical specification, except ad:ninistrative conditions, for the Docket 50-602 facility.

4 A.1.1 Arolicability P

The applicability defines the conditions, parameters, or equipment to which the specification applies.

I A.1.2 Obiective The objective defines the goals of the specification in terms-of limits, t

frequency, or other controllable item.

1 A.1.3 Esses The bases presents information important to the specification, including f

such things as justification, logical constraints and development methodology.

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I A.2.0 SAFETY LIMITS & LIMITING SATETY SYSTEM SETTINGS I

[

APPLICABILITY, OBJECTIVES AND BASES 2,

A.2.1 Safety 1.imit 2

Applicability 4

This specification applies to the temperature of the reactor fuel in a l"

standard TRIGA fuel element.

Objective 4

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

j i

Bases 4

4 The important parameter for a TRIGA reactor is the fuel element a single specification temperature.

This parameter is well suited as since it can be measured directly.

A loss in the integrity of the fuel element cladding could arise from a build-up of excessive pressure t

between the fuel-moderator and the cladding if the fuel temperature exceeds the safety limit.

The pressure is caused by the presence of air, fission product gases, and hydrogen from the dissociation of the i

hydrogen and zirconium in the fuel-moderator.

Hydrogen pressure is the most significant component.

The magnitude of this pressure is determined by the fuel-moderator temperature and the ratio of hydrogen

[

(.

to zirconium in the alloy.

The safety limit for the standard TRIGA fuel is based on calculations and experimental evidence.

The results indicate that the stress in the cladding due to hydrogen pressure from the dissociation of zirconium hydride vill remain below the ultimate stress provided that the terperature of the fuel does not exceed 1150*C and the fuel cladding does not exceed 500*C.

For conditions that might cause the clad temperatures to exceed 500'C the safety limit of the fuel should be set at 950*C.

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Revision 1 Technical Sp2cificaticns j

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A.2.2 Linitine Safety System Settina 4

i A.2.2.1 Fuel Temperature d

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l Applicability i

4 This specification applies to the protective action for the reactor i

fuel element temperature.

4 i

j Objective 4

The objective is to prevent the fuel element temperature safety limit from being reached.

j i

Bases

.i For non pulse operation of the reactor, the limiting safety system

]

setting is a temperature which, if exceeded, shall cause a reactor l

scram to be initiated preventing the safety limit from being l

1 exceeded.

A setting of 550*C provides a safety margin at the point of measurement of at least 400*C for standard TRICA fuel elements in any condition of operation.

A part of the safety margin is used to a

account for the difference between the true and measured temperatures resulting from the actual location of the thermocouple.

If the thermocouple element is located in the hottest position in the core, i

the difference between the true and measured temperatures will be i

only a few degrees since the thermocouple junction is near the center j

(

and the mid-plane of the fuel element.

For pulse operation of the

reactor, the same limiting safety system setting will apply.

However, the temperature channel will have no effect on limiting the peak powers generated because of its relatively long time constant (seconds) as compared with the width of the pulse (milliseconds).

In 4

this mode, however, the temperature trip will act to limit the energy l

release after the pulse if the transient rod should not reinsert and l

the fuel temperature continues to increase.

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l A.2.2.2 Power 14 vel (Manual, Auto, Square Wave)

A Applicability This specification applies to the protective action for the reactor during non pulse operation.

Obj ective The objective is to prevent the fuel element temperature safety limit from being reached.

l Bases Thermal and hydraulic calculations indicate that standard TRIGA fuel elements may be safely operated at power levels in excess of 1500 kilovatts with natural convection cooling.

Conservative estimates i

indicate that a

departure from nucleate boilins ratio of approximately two will occur at about 1900 itilowatts.

A limiting l

setting for the power level measurement at 1.1 megawatts assures sufficient margin for safety to allow for calibration errors.

The power calibration goal is a measurement accuracy of 5% although an error of 101 may be representative of some measurements.

l A.2.2.3 Reactivity Insertion (Pulse)

Applicability k

This specification applies to the reactivity insertion for the reactor during pulse operation.

Objective The objective is to prevent the fuel element temperature safety limit from being reached.

Bases Calculations indicate that standard TRICA fuel elements may be safely operated at transient conditions in excess of 2.2% Ak/k with ambient cooling conditions.

Conservative estimates indicate that a

l substantial safety sargin exists for the rise of peak fuel l

temperature with reactivity insertions as large as 2.8% ak/k.

t l

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  1. av' *, 1 Technical Spscificatiens A.3.0 LIMITING CONDITIONS FOR OPERATION APPLICABILITY, OBJECTIVES & BASES A.3.1 Reactor Core Parameters A.3.1.1 Excess Reactivity i

Applicability l

This specification applies to the reactivity condition of the reactor core in teras of the available excess above the cold xenon free, critical condition.

Objective The objective is to prevent the fuel element temperature safety limit from beinE reached by limiting the potential reactivity available in the reactor for any condition of operation.

Bases Maximum excess core reactivity is sufficient to provide the core rated power, xenon compensation and reactivity for shutdown.

Analysis of the reactor core demonstrates that no single component

[

represents sufficient potential reactivity to reach the fuel element temperature safety limit during any condition of operation.

k A.3.1.2 Shutdown Margin Applicability This specification applies to the reactivity margin by which the reactor core will be considered shutdown when the reactor is not operating.

Objective I

The objective is to assure that the reactor can be shut down safely by a margin that is sufficient to compensate for the failure of a i

control rod or the movement of an experiment.

Bases

\\

The value of the shutdown margin assures that the reactor can be shut l

down from any operating condition.

These conditions include the assumption that the highest worth control rod remains fully withdrawn and all noveable experiments are in the most reactive condition.

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Revisien 1 Technical Spscificatiens A.3.1.3 Transient Insertions Applicability This specification applies to the total potential worth of the transient rod and the allowable reactivity insertion for reactor pulse operation.

Objective The objective is to limit the reactivity available for pulse insertion to a value that vill not cause the fuel temperature safety limit to be exceeded.

Reses j

Calculations demonstrate that the total insertion of all' the transient rod worth will not exceed the fuel temperature safety limit.

For a 2.8% ak/k pulse a safety margin would exist between the

[

fuel element safety limit and the rise of peak fuel temperature above an assumed ambient pool temperature of 50*C.

A preset timer insures

~

that the transient rod will not remain in the pulse position for an extended time af ter the pulse.

Experiments with pulsed operation of TRIGA reactors by the manufacturer indicate that insertions up to 3.5% Ak/k have not exceeded the fuel temperature safety limit.

A.3.1.4 Fuel Elements Applicability 1

This specification applies to the measurement parameters for the fuel i

elements.

Obj ective The objective is to verify the physical condition of the fuel element cladding.

Bases The elongation limit has been specified to assure that the cladding i

material will not be subjected to stresses that could cause a loss of integrity in the fuel containment and to assure adequate coolant flow.

The limit of transverse bend has been shown to result in no difficulty in disassembling the reactor core.

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 TRICA reactors has shown that fuel element bowing that could result in touching has occurred without deleterious effects.

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i A.3.2 Resetor Control and Safety System A.3.2.1 Control Assemblies I

t Applicability k

~

This specification applies to the function of the control rods.

i objective 4

The objective is to determine that the control rods are operable by specification of apparent physical conditions, the scram times for scrarmable control rods and the reactivity insertion rates for standard control rods.

I Bases The apparent condition of the control rod assemblies will provide assurance that the rods will continue to perform reliably and as designed.

The specification for rod scram time assures that the-reactor will shut down promptly when a scram signal is initiated.

l The specification for rod reactivity insertion rates assures that the reactor will start up at a controllable rate when rods are withdrawn.

Analysis has indicated that for the range of transients anticipated for a TRICA reactor the specified scram time and insertion rate is adequate to assure the safety of the reactor.

j A.3.2.2 Reactor Control System f

Applicability i

These specifications apply to logic of the reactor control system.

{

Objective The objective is to determine the minimum centrol system interlocks l

operable for operation of the reactor.

j l

i Bases Interlocks are specified to prevent function of the control rod l

drives unless certain specific conditions exist.

Program logic of i

the digital processors implement the interlock functions, j

Two basic interlocks control all rod movements in the manual mode.

I The interlock to prevent startup of the reactor at power levels less j

than 2 neutron eps, which corresponds to approximately_4 mil 11 watts, assures that sufficient neutrons are available for controlled reactor startup.

Simultaneous withdrawal of more than one contr'o1 rod is prevented by an interlock to limit the maximum positive reactivity insertion rate available for steady state operation.

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Two basic interlocks control rod movements for the pulse mode.

The i

interlock to prevent withdrawal of the motor driven rods in the pulse mode is designed to prevent changing the critical state of the i

reactor prior to the pulse. A power level interlock controls j

potential fuel temperature changes by setting a limit of less than 1 i

kilowatt for initiation of any pulse.

Interlocks applicable to the transient rod determine the proper rod i

operation during manual mode and pulse mode operation.

The non pulse condition interlock determines the allowable position of. the rod

[

drive for actuation of the TIRE switch.

Actuation of _ the switch applies the air impulse for removal of the transient rod from the l

reactor core.

Auto mode is a special condition of the manual mode with automatic i

I control of the regulating rod.

Square wave mode is also a special case of the manual mode with automatic control except that pulse logic applies to the initiation of the auto mode condition.

i A.3.2.3 Reactor Safety System.

Applicability These specifications apply to operation of the reactor, safety system.

Objective The objective is to determine the minimum safety system scrans operable for the operation of the reactor.

l Bases l

Safety system scram functions consist of three types. These scram types are the limiting safety system settings, operable system conditions, and the manual or program logic scrams. The scrans cause control rod insertion and reactor shutdown.

Scrams for limiting safety system settings consist of signal trip f

levels that monitor fuel temperature and power level.

The trip levels are conservative by a significant margin relative to the fuel element temperature safety limit.

[

Operation without adequate control and safety system power supplies is prevented by scrams on neutron detector high voltage and control rod magnet current.

i Manual action of the scram switch, key switch, or computer actuation of watchdog timers will initiate a protective action of the reactor safety system.

Either of two watchdog circuits provide' updating timers to terminate operation in the event that key digital processing routines fail, such as a display system.

Each watchdog circuit with four resettable timers contains one trip relay and monitors one microcomputer.

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Revision 1 Technical Sp2cifiestiens A.3.2.4 Reactor Instrument System I

Applicability J

These specifications apply to measurements of reactor operating parameters.

Objective l'

The objective is to determine the minimum instrument system channels to be operable for continued operation of the reactor.

Bases I

The minimum measuring channels are sufficient to provide signals for automatic safety system operation.

Signals from the measuring system i

provide information to the control and safety system for a protective l

action.

Instruments provide redundancy by measurements of the same parameters and diversification by measurements of different i

parameters.

Two redundant temperature thermocouple sensors monitor the fuel temperature limiting safety system setting.

Two redundant l

percent power channels monitor the power level limiting safety system.

A digital vide range channel may also function as a safety channel but only by diversification as a supplemental channel to an analog linear power channel.

Pulse parameters of peak power and energy release are measurements of a single detector chamber.

There are, however, two separate peak and energy monitoring circuits.

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i R2vioicn 1 Tcchnien1 Sp2cificctisns A.3.3 Orerational Support Systen A.3.3.1 Vater Coolant Systens Applicability This specification applies to the operating conditions for the reactor pool and coolant water systens.

Obj ective The objective is to assure that adequate conditions are saintained to provide shielding of the reactor radiation, protection against corrosion of the reactor components, cooling of the reactor fuel, and prevent leakage from the primary coolant.

Bases The sp(cifications for conditions of the pool water coolant systen provide controls that are to control the radiation exposures and radioactive releases associated with the

  • actor fission product inventory.
a. The bulk water temperature constraint assures that sufficient core cooling exists under all anticipated operating conditions and protects the resin of the water purification systen from deterioration.
b. A pool water depth of 6.5 naters is sufficient to provide more than 5.25 neters of water above the reactor core so that radiation levels above the reactor pool are at reasonable levels.
c. Average seasurenents of pool coolant water conductivity of 5.0 p=ho/en assure that water purity is naintained to control the i

effects of corrosion and activation of coolant water inpurities.

d. A pressure difference at the heat exchanger chilled wate r outlet and the pool water inlet of 7 kPa vill be sufficient to prevent loss of pool water from the primary reactor coolant systen to the secondary chilling water systen in the event of a leak in the heat exchanger.
e. Periodic sampling of pool water pH and radioactivity are supplenental seasurenents that assist evaluation of the overall conditions of the reactor pool.

Protection of aluminun components requires a pH range of 5 to 6.5.

Measurenents of

)

radioactivity in the pool water provide information to evaluate working hazards for personnel, leakage indications for radioactive sources in the pool, and monitoring for activation of unknown conponents in the water.

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3 Technical Specificatiens 5

A.3.3.2 Air Confinement Systess 1

Applicability 1

This specification applies to the air ventilation conditions in the reactor area during reactor operation.

Objective i

The objective is to control the release of air in the reactor area or i

j l

experimental facilities.

Bases The specifications for exhaust ventilation and isolation of the reactor bay provide control for radioactive releases for both routine and non routine operating conditions.

4

s. Air confinement of the reactor bay includes a provision for i

isoletion of the air flow of the ventilation system.

Danpers in the room supply air ducts and room return air ducts limit the leakage rate and total release of radioactive airborne materials to a fraction of the available volume.

r

b. A signal from a particulate air monitor in the vicinity of the reactor pool initiates the automatic isolation of the supply air dampers and return air dampers.

The isolation process k

takes less than one minute and includes the shutdown of supply fan and exhaust fan.

An equivalent to one maximum permissible l

concentration is the set point.

c. Air from experiment areas within the neutron flux regions of i

i the coro vill ventilate separately from room air by way of a filter bank that includes a high efficiency particulate filter.

J Space is available to install a charcoal filter for special experi=ent conditions,

d. Control of concentrations of argon-41 in reactor room air depends on ventilation of the room air at a rate of two air changes per hour or operation of the auxiliary purge air system.

Operation and isolation of the purge system is by manual control of damper and fan switches.

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Revision 1 A.3.3.3 Radiation Monitoring Systems Applicability F

in radiation monitoring conditions This specification applies to the the reactor area during reactor operation.

Objective The objective is to monitor the radiation and radioactivity m

conditions in the reactor area to control exposures or releases.

as Bases The radiation monitors provide information to operating personnel of from radiation so that there vill be i=pending or existing hazards to control the exposure sufficient time to take the necessary steps the facility.

of radioactivity or evacuate of personnel and release Alarm setpoints do not include measurement uncertainty.

These setpoints are neasured values and not true values radioactivity accumulates on the filter of a

a. Air particulate An alert continuous monitor that records the radiation levels.

including remote readouts at the reactor and alarm set point control console inform the operator of the monitor status and activity levels.

An alarm limit at two thousand detects particulate activity g

picoeurie/ milliliter concentrations at the occupatior.a1 values of 10CTR20.

The alarm set point exceeds occupational values for any single nuclide in the ranges84-105 and 129 149.

f fission product of the particulate isotopes are e.lso detectable Seventy percent The gaseous reference concentrations within two hours.

argon-41 monitor can provide fission product gas monitoring at the during repair of the particulate monitor,

b. Air gaseous radioactivity of argon-41 concentrations require conitorin5 of the levels for effluent release and occupational exposure.

The alars setpoint detects a release concentration that will not exceed ten times either the occupation value at the stack or the reference concentration at the ground.

3 concentration of 1.2 pCi/c Calculations of a stack release indicate that the equivalent ground level concentration is equivalent to 1x10 s pCi/cm3 A license limit for the average annual concentration is necessary to fix the amount of allowable release.

Periods of inoperable argon-41 monitoring the amount of release without equipment of up to 10 days limit neasurecent to a fraction of the total annual release. "

c. Several area radiation eenitors (six) are part of the permanent in which installation.

Some locations are experinent areas the levels of radiation during shield configurations determine radiation levels reactor operation.

At the pool access area substantial enough to be a high radiation level say occur.

if the 100 nr/hr vill conitor radiation areas Alaru levels at limit of 2 or 5 nr/hr is not reasonable.

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Tavision 1 A.3.4 Lief eatf ons on Experittents A.3.4.1 Reactivity

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f 4

Applicability f

i the reactivity of experiments located This specification applies to in the reactor core.

h l

Objective l

The objective is to control the amount of reactivity associated with experiments to values that will not endanger the reactor safety l

4 B

1 limit.

a

?

Bases i

worth of single moveable experiment is limited so that

a. The sudden removal movement of the experiment will not cause prompt i

l will not criticality.

Vorth of a single unsecured experiment cause a reactivity insertion that would exceed the core l

tesperature safety limit.

i

b. The maximum vorth of a single experimenu is limited so that the i

fuel element temperature safety limit will not be exceeded by t

removal of the experiments.

Since experiments of such vorth l

sust be secured in place, removal from the reactor operating at full power would result in a relatively slow power increase j

such that the reactor protective systems would act to prevent F

excessive power levels from being attained.

i 4

c. The maximum worth of all experiments is limited so that removal of the total worth of all experiments will not exceed the fuel l

element temperature safety limit.

A.3.4.2 Materials Applicability These specifications apply to experiments installed in the reactor and its experimental facilities.

2 Objective The objective is to prevent the release of radioactive saterial in the event of an experiment failure, either by failure of the experinent or subsequent dasage to the reactor components.

a. Double encapsulation requirements lessen the leakage hazards of Bases i

some types of experisent materials.

the reactor with the reactor fuel or structure

b. Operation of da: aged is prohibited to avoid release of fission products.

Page 49 12/90

Tcchnieci spec 1 G n h o Revision 1 materials ast a for explosive for

c. Encapsulation requirements of material allowable from the explosive reaction for the amount reference condition Damage and resultant gas any reactor experiment.

depends on the available energy release of for 25 milligrams conditionsf 25 calories (104 joules)

Approximate creation.

explosive material are the release o If a 1 milliliter volume t rial (density of energy and 25 milliliters of gas.is available for the re instantaneous an will represent lease adds another the energy 1.654 ge/cm8),

thin wall, pressure of 1032 atmospheres and the gas re for a calculations for the wall E

Stress capsule specify the requirementsThe relationship 25 atmospheres.

lation.

of the cylindricalthickness and diameter of the encapsufourth the product determines the stress limit diameter to wall thickness ratio.

as one EHur requires a ratio times the capsule l

An aluminum capsule with a 1 milliliter vo umevolume of 5 milliliters cap pressure At a vall thickness -

that does not exceed 5.2.

dimensions with a diameter of 2.6 cm requires aThese f

for experiment of 1 mm.

conitruction components aluminum tubular iodine and facilities and experiments. inventory limits of 750 millicurie 2.5 millieurie strontium fix the potential accident releas

d. Fission product the radioactive isotopes representproduct nuclides with These two concentrations.

i exposure risk to individuals for fiss gnshort (iodine) a If ~ the isotope f lives.

the total inventory release of 750 e

including iodine-131 represents annual average release, the facility will bc building wake dilution of the total inventory, millieuries, of 2x10 ~18 level concentration h

In the case of strontium-90 the release is les to the reference equivalent s

of 5x10'22 pCi/ca,

pCi/cm3 level concentration by manual or 1/5 the reference ventilation system effective total shutdown of the the automatic operation substantially reducesAny release ithin Proper i

for releasw.

the facility, however, in the form o occupational values within the facil ty As f the radionuclides 90 will exceed the the oral ingestion or air inhalation o to maintain the average case the evacuation times 131 and 1 month for annual concentration are I hour for iodine-an extreme that cause airborne strontium.90.

il

e. Accidental release of radioactive mater a s annual limits.

10CTR20 average that cause must meet occupational valueslevel concentrations concentrations apply to Concentration limits exposure within the _ facility and reference

' Calculations facility.

a release from the i l but also must define that may exist as assume a complete release of the mater a conservative or that are release rates and frequencies reasonable estimates of accident conditions, of for the calculation

f. This specification provides guidance conditions in part (e).

Page 50 12/90 e

Technical Specificationa oevisien 1 A.4.0 SURVEII.1ANCE REQUIREMENTS OBJECTIVES & BASES 2,

f i

A.4.1 Resetor Core Parameters F

a a

A.4.1.1 Excess Reactivity Applicability This specification applies to the measurement of reactor excess E

l reactivity.

Objective a

The objective is to periodically determine the changes in core excess reactivity available for power generation.

l Bases Annual determination of excess reactivity and measurements after t

reactor core or control rod changes are sufficient to monitor significant changes in the core excess reactivity.

A.4.1.2 Shutdown Margin 1

Applicability of reactor shutdown This specification applies to the measurement margin.

Objective l

ebjective is to periodically determine the core shutdown The reactivity available for reactor shutdown.

i Bases f

Annual determination of shutdown margin and measurements after reactor core or control rod changes are sufficient to monitor significant changes in the core shutdown margin.

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A.4.1.3 Transient Insertion Applicability This specification applies to surveillance of the transient rod mechanism and to observation of the reactor transient response.

l Objective l

The objective is to assure the function of the transient rod drive and to compare the reactor pulse insertion parameters.

f Bases i

Annual inspections of the pulse rod drive system should be sufficient to detect and correct changes in the system. that could impair operability.

Comparison of pulse parameter-data should detect characteristic changes of reactor core transients.

l 1

A.4.1.4 Fuel Elements i

Applicability This specification applies to the inspection requirements for the i

fuel elements.

i objective The objective is to inspect the physical condition of the fuel i

element cladding.

l Bases The frequency of inspection and measurement schedule - is based on the I

parameters most likely to affect.

Se fuel cladding of a pulsing

{

reactor operated at moderate pulsing levels and utilizing. fuel elements whose characteristics are well known.

Page 52 12/90

Rrefrien 1 Tochnical Sp:cific tiens A.4.2 Reactor Control and Safety Sv.E_t1E A.4.2.1 Control Assemblies Applicability This specification applies to the surveillance of the control rods.

Objective The objective is to inspect the physical condition of the reactor control rods and establish the operable condition of the rod by periodic measurement of the scram times and insertion rates.

Bases Annual determination of control rod worths or measurements after significant core changes previde information about changes in reactor total reactivity and individual rod worths.

The frequency of inspection for the control rods will provide periodic verification of the condition of the control rod assemblies.

Verification will be by measurement of fueled sections and visual observation of absorber sections plus examination of linkages and drives.

The specification intervals for scram time and insertion rate assure operable performance of the rods.

Deviations that are significant from acceptable standards will be promptly corrected.

i A.4.2.2 Reactor Control System Applicability This specification applies to the tests of the logic of the reactor control system.

Objective The objective is to specify intervals for test, check or calibration of the minimum control system interlocks.

1 Bases

)

The periodic test of the interlock logic at semiannual intervals 1

provides adequate information that the function of the control system interlocks are functional.

Changes to the interlock logic consist of revisions to the microcomputer algorithms (hardware, software or firmware) and repair of input or output circuits including devices that are sensors for the interlocks.

Calibrations or checks of the control system logic are not considered applicable functions.

l l

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12/90 Page 53 l

i 1

Kevia. ion 1 Techniccl Spscifications A.4.2.3 Reacter Safety System l

Applicability This specification applies to tests of the function of the reactor safety system.

Objective j

The objective is to specify intervals for test, check or calibration of the minimum safety system scrass.

l Bases

(

The periodic calibration at annual intervals provides adequate l

information that the serpoints of the safety systas scrans are i

functional.

Tests of the safety systet prior to each planned I

cperation assure that each intended scram function is operable.

l i

A.4.2.4 Reactor Instrument System l

Applicabilitf These specifications apply to calibrations, checks, and tests of reactor measurement channels.

i Objective The objective is to specif y intervals for test, check or calibration of the minimum instrument channels.

l Bases 5

Annual calibration of instrument channels are scheduled to allow adjustments for changes in reactor-and instrumentation parameters.

Checks and tests prior to each system operation verify the function of key c'annels and systems.

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12/90 Page 54 i

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l Revision 1 Tcchnical Specificaticns i

A.4.3 overstional Suovert Systens e

A.4.3.1 Vater Coolant Systems Applicability This specification applies to surveillance conditions for the reactor pool and coolant water systems.

Objective The objective is to maintain the reactor coolant conditions within acceptable specifications.

Bases conditions for the reactor coolant are monitored by visual observation of measurements or automatic action of sensors.

Periodic i

' checks and tests of measurement devices for the reactor coolant system parameters assure that the coolant system will perform its intended function.

Measurement frequencies of pool parameters relate the time periods appropriate to detection of abnormal conditions.

to Pool temperature, depth, and heat exchanger pressure differences have an immediate effect on system operation. Water conductivity, pH as a supplemental indicator, and pool radioactive concentrations are conditions that develop at rates detectable at monthly to annual intervals.

A.4.3.2 Air Confinement Systems 1

Applicability This specification applies to surveillance conditions for the air ventilation in the reactor area.

Obj ective The objective is to demonstrate the function of confinement. and release of air from the reactor bay.

Bases Periodic tests and checks of air confinement conditions verify I

appropriate ventilation functions.

Monitoring frequencies verify performance of the confinement system exhaust daily by an align =ent check that includes observation of negative pressures.

Tests of the isolation feature at monthly intervals assure the acceptable operation of the system.

12/90 Page 55

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P.critien 1 Tochnical Specifications f

A.4.3.3 Radiation Monitoring Systems Applicability This specification applies to the surveillance conditions of the radiation monitoring channels.

Objective The objective is to assure the radiation monitors are functional.

Bases specified to maintain Periodic calibrations and frequent checks are reliable performance of the radiation monitoring instruments.

Calibration and check frequencies follow the general recommendations of guidance documents.

i Page 56 12/90

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Rovision 1 Tochnical Spscificctiens i

s A.4.4 Lieftations on Experiments 1

A.4.4.1 Reactivity j

I Applicability i

This specification applies to surveillance of the reactivity of l

experiments.

Objective The objective is assure the reactivity of a.n experiment does not i

exceed the allowable specification.

1 I

Bases The measured reactivity or determination that the reactivity is not significant will provide data that configuration of the experiment or I

experiments is allowable.

a A.4.4.2 Materials f

Applicability This specification applies to the surveillance requirements for materials inserted into the reactor.

Objective 3

The objective is to prevent the introduction of materials that could damage the reactor or its components.

l e

Bases A careful evaluation of all experiments is performed to classify the experiment as an approved experiment.

l Page 57 12/90

I Revision 1 Tochnical sp3cificatisns A.5.0 DESIGN TEATURES OBJECTIVES & BASES a

A.S.1 Site and Facility Descrietions I

A.$.1.1 location Applicability This specification applies to the TRIGA reactor site location and specific facility design features.

\\

Objective The objective is to specify those features related to the safety i

Analysis evaluation, Bases j

a. The TRIGA facility site is located in an area controlled by The l

University of Texas at Austin.

b. The room enclosing the reactor has been designed with characteristics related to the safe operation of the facility.

I

c. The shield and pool structure have been designed for radiation levels of less than 1 arem/hr at locations that are not access ports to the reactor structure.
d. The restricted access to specific facility areas assure that proper controls are established for the safety of the public and for the security of special nuclear materials.

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Page 58 12/90 i

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t Ravisien 1 Technical Sp;cificati ns A.5.1.2 Confinement Applicability This specification applies to the boundary for control of air in the area of the reactor.

Obj ective The objective is to assure that provisions are made to control or j

restrict the amount of release of radioactivity into the environment.

l i

Bases

a. Calculations of the concentrations of released radionuclides within the reactor area depend on the available enclosed air volume to limit the concentrations to acceptable levels.
b. Control of the reactor area air exchange is by fan motors and isolation dampers for the supply and exhaust air which are controlled by a logic si nal from a radiation sensor to provide 5

automatic air confinement.

c. Emergency air ventilation is filtered to control the release of particulates and a pressure difference relative to the external ambient pressure is intended to prevent leakage of air without

/

filtration.

k

d. Exhaust air during reactor operation is released at an elevated level for dispersion and is designed to provide a relative pressure difference to the' external ambient pressure.

A.5.1.3 Safety Related Systems Applicability This specification applies to the requirenents of any system related to reactor safety.

Objective The objective is to assure the proper function of any system related to reactor safety.

Bases l

This specification relates to changes in reactor systems which could affect the safety of the reactor operation.

Changes or substitutions to these systees that meet or exceed the original design I

specifications are assumed to meet the presently accepted operating criteria.

Questions that say include an unreviewed safety question are referred to the reactor operation committee.

i

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12/90 Page 59

I Pevision 1 Techniec1 Spacificctiens A.S.2 Reactor Coolant system i

Applicability This specification applies to the reactor coolant system composed of deionized water.

Objective The objective is to assure that adequate water is available for cooling and shielding during reactor operation.

Bases a.

This specification is based en thermal and hydraulic calculations which show that a standard 85 element TRIGA core can operate in a safe sanner at power levels up to 1,900 kV with natural convection flow of the coolant water and a departure from nucleate boiling ratio of 2.0.

b.

Siphon breaks set the subsequent pool water level for loss of coolant without an associated water return caused by inadvertent pumping or accidental siphon of water from the pool.

A.5.3 Resetor Core and Puel A.5.3.1 Fuel Elements Applicability This specification applies to the fuel elements used in the reactor core.

Objective The objective is to assure that the fuel elements are of such a design and fabricated in such a manner as to permit their use with a hi h degree of reliability with respect to their physical and nuclear S

characteristics.

f Bases The design basis of the standard TRICA core demonstrates that 1.5 megawatt steady or 36 segawatt see pulse operation presents a conservative limitation with respect to safety limits for the maximum temperature generated in the fuel.

The fuel temperatures are not expected to exceed 550*C during any condition of normal operation.

\\

12/90 Page 60

Revision 1 Tcchnical Spscificaticns A.S.3.2 Control Rods 9

Applicability his specification applies to the control rods used in the reactor core.

Objective The objective is to assure 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.

Bases i

The poison requirements for the control rods are satisfied by using neutron absorbing borated graphite, BaC powder, or boron and its i

compounds.

These materials must be contained in a suitable clad j

material, such as aluminum or stainless steel, to insure mechanical stability during movement and to isolate the poison from the pool water environment.

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.

f A minimum configuration of control rods consist of two shim rods, a

regulating rod and ;he transient rod.

The configuration of rods is necessary for the reactor to be operable.

If the appropriate adjustments to the core reactivity are made the removal of one or more of the control rods will facilitate

)

the necessary inspection and repair activities.

Definitions for j

shutdown and suberitical require the reactor core to meet the i

suberitical constraint if any rod is out of the core and the reactor

)

is to be shutdown.

A.S.3.3 Configuration Applicability This specification applies to the configuration of fuel elements, control rods, experiments and other reactor grid plate components.

Objective l

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

Bases Standard TRIGA cores have been in use for years and their characteristics are well documented.

12/90 Page 61

p.vi.4cn 1 Technicc1 Spacific tions A.5.4 Resetor Fuel Element Storare i

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 assure that fuel storage will not achieve criticality and will not exceed design tamperatures.

Bases The limits imposed by these specifications are considered sufficient to provide conservative fuel storage and assure safe storage.

A.5.5 Carta Fool Irradiator Applicability This specification applies to the gamma irradiator experiment facility in the reactor pool.

Objective The objective is to assure that the use of the irradiator does not

(

cause any threat to the reactar or safety question.

i Bases location of the irradiator is at a distance from the reactor sufficient to avoid interference with reactor operation.

Depth of the pool water for adequate shielding of the irradiator is also a constraint of the location 1

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i Page 62 12/90

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A.6.0 NOTES I

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12/90 Page 63

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SAR Standard Format and Content Chapter 14 - Draft August 2, 1993 14 TECHNICAL SPECIFICATIONS 14.1 Introduction This section contains the guidelines to develop the technical specifications for NRC-licensed non-power reactors.

The NRC requires each operating non-power reactor applicant to develop technic.a1 specifications that set forth the limits, operating conditions, and other requirements imposed on facility operation to protect the health and safety of the public in accordance with 10 CFR 50.36. The technical specifi-cations are typically derived from the facility descriptions and safety con-siderations contained in the SAR and represent a comprehensive envelope of safe operation.

Applications for construction permits / operating licenses and renewals of operating licenses must contain proposed technical specifications that are incorporated in the operating license. During its review of the application, the NRC staff will review the SAR and proposed technical specifications to ensure they are complete, comprehensive, and that the public health and safety will be protected. After final acceptance by the NRC staff, the technical specifications will be included as Appendix A to the operating license.

The format and content of the technical specifications discussed in this document follow that of the 1990 revision to ANSI /ANS 15.1, " Development of Technical Specifications for Research Reactors." Additional guidance on the format and content of technical specifications can be found in previously accepted and approved technical specifications for non-power reactors of similar design, operating characteristics, site and environmental conditions, and use.

This chapter of the SAR normally is very short. The applicant should be able to state conclusively in this chapter that the technical specifications were prepared following an accepted format, that normal operation of the reactor within the limits of the technical specifications will not result in offsite radiation exposure in excess of 10 CFR Part 20 guidelines, and that the tech-nical specifications limit the likelihood and consequences of malfunctions.

The reader is referred to the technical specifications, which are contained in a separate document from the SAR. The technical specifications are neii.her derived or justified in this chapter of the SAR. The actual technical speci-fications are determined by the analysis that appears in the other chapters of the SAR. Each of the technical specifications should be supported by the StA and it is useful to refer to the supporting SAR analysis in the basis of each technical specification.

In the text that follows, all sections of ANSI /ANS 15.1 are addressed.

If modifications or clarifications for ANSI /ANS 15.1 are required to provide acceptable technical specifications, the additional guidance is provided.

Sections that provide acceptable guidance as written also are noted.

24

SAR Standard Format and Content Chapter 14 - Draft

]

August 2, 1993 14.2 Format and Content of Technical Specifications The numbering in this part (Sections 1 through 6.8) corresponds to the section numbering in ANSI /ANS 15.1-1990.

l 1

Introduc: tion 1.1 Scope The NRC accepts the guidance provided in this section of ANSI /ANS 15.1.

This sectic 1 confirms that the technical specifications for non-power reactors should include all the categories in 10 CFR 50.36 for production and utilization facilities.

i 1.2 Application I

l 1.2.1 Purpose The NRC accepts the guidance provided in this section of ANSI /ANS 15.1.

Technical specifications represent a set of operating requirements for a reactor that the licensee and the NRC have agreed on.

The specifications i

become part of the operating license.

1.2.2 Format Sections of the technical specifications must be numbered as indicated in Section 1.2.2 of ANSI /ANS 15.1.

Subsections may be left out if not applicable for a particular reactor or altered if necessary, but the subsections included must be numbered in consecutive order.

l For individual specifications in Sections 2, 3, and 4, applicability, objective, specification, and basis information must be included in the specified format. For Sections 5 and 6 of the technical specifications, ANSI /ANS 15.1 suggests the specifications be stated without providing applicability, objective, or basis. Although this format is preferred, it is acceptable to NRC if these sections include applicability, objective, or basis statements.

Technical specifications that use the SAR as a basis should explicitly reference the SAR section number.

In addition, any other sources used to support the technical specification should be explicitly referenced.

1.3 Definitions The NRC and the non-power reactor community have agreed on most of the definitions given in this section of ANSI /ANS 15.1 Those applicable to a particular facility should be included verbatim.

Facility-specific defini-tions may be added to clarify terms referred to in the technical specifica-tions. Modifications and additional definitions presented below help clarify the meaning of terms used in ANSI /ANS 15.1.

i 1

25 i

SAR Standard Format and Content Chapter 14 - Draft August 2, 1993 The following definitions should be modified as indicated:

Class A reactor operator. The term acceptable to the NRC is senior

=

reactor operator.

Class B reactor operator. The term acceptable to the NRC is reactor operator.

reactor shutdown. The reactor is shut down if it is subtritical by at least I dollar both in the reference core condition and for all allowed ambient conditions with the reactivity worth of all installed experiments included.

reference core condition. The reactivity condition of the core when it is at 20 *C and the reactivity worth of xenon is zero (i.e. cold, clean, and critical).

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.

It should be assumed that the most reactive scrammable rod and all non-scrammable rods are in their most reactive position and that the reactor will remain subcritical without further operator action.

Note: Shutdown margin has a single value mutually acceptable to the NRC and the licensee, to be determined on a case-by-case basis.

The following definitions should be added:

secured shutdown.

Secured shutdown is achieved when the reactor meets

=

the requirements of the definition of " reactor secured" and the facility administrative requirements for leaving the facility with no licensed reactor operators present.

shutdown reactivity.

Shutdown reactivity is the value of the reactivity l

=

of the reactor with all control rods in their least reactive positions (e.g., inserted). The value of shutdown reactivity includes the reactiv-ity value of all installed experiments and is determined with the reactor at ambient conditions.

2 Safety Limits and Limiting Safety System Settings 2.1 Safety Limits All reactor licensees are required by 10 CFR 50.36(c) to specify safety limits in the technical specifications. These safety limits will be placed on impor-tant process variables identified in the SAR as necessary to reasonably pro-tect the integrity of the primary barrier against the uncontrolled release of radioactivity.

For non-power reactors, the radioactivity of concern is generally the fission products in the fuel.

For heterogeneous-core non-power reactors, the primary barrier is the cladding of fuel plates, rods, or pins.

26

SAR Standard Format and Content Chapter 14 - Draft August 2, 1993 Cladding integrity could be lost by softening, melting, blistering, or yielding to excess internal pressure, all of which are dependent on temperature and operating history. For homogeneous-core reactors, this primary barrier may be the fuel matrix, the primary vessel or some other component that contains the fuel and the fission products.

Reactor conditions and safety limits should be developed to avoid failure of the fuel and be supported by SAR analyses. Manufacturers have studied failure modes and failure parameters during fuel development programs. The NRC has issued NUREG reports approving the use of some low-enriched uranium fuel types in non-power reactors. The applicant should make maximum use of the i

appropriate references; some of which are listed at the end of this document.

The applicant should consult NUREG-1313, " Safety Evaluation Report Related to i

the Evaluation of Low-Enriched Uranium Silicide-Aluminum Dispersion Fuel for Use in Nonpower Reactors," for evaluating aluminum-clad, aluminum matrix i

plate-type fuels, using both high-enriched uranium (HEU) and new uranium-silicide low-enriched uranium (LEU).

It discusses temperatures from j

experimental irradiation tests at which plate blistering has been observed, a l

possible forerunner of failure. The NRC finds 530 *C an acceptable fuel and l

cladding temperature limit not to be exceeded under any conditions of operation. NUREG-1313 also references tests with HEU plate fuel that have led to similar conclusions (Beeston et al.,1980; Gibson,1967; Nazare et al.,

1975; Stahl,1982).

There are several reports on training reactor and isotope production, General Atomics (TRIGA)-type fuels (NUREG-1282; Simnad et al.,1976 and 1981; Simnad and West, 1986; West et al., 1986). For stainless-steel-clad UZrH LEU 8.5 uranium weight percent (w/o) TRIGA fuel, stainless-steel-clad UZrH,65 HEU (70%

i U-235 enriched) 8.5 w/o fuel lifetime improvement program (FLIP) TNA fuel, 3

and stainless-steel-clad UZrH LEU 20 w/o and 30 w/o TRIGA fuel, General 3 63 Atomics has shown and NRC has, accepted that integrity is not compromised under the following cases and conditions:

for cladding temperature at or less than 500 *C, peak fuel temperature at or less than 1150 *C for cladding temperature greater than 500 *C, peak fuel temperature at or less than 950 *C For aluminum-clad UZrH,o LEU 8 w/o TRIGA fuel, NRC accepts that the peak fuel 3

temperature should not exceed 500

  • C.

For pulsed training assembled reactor (PULSTAR) types, NRC accepts that the UO fuel temperature should not exceed 2400 *C and the Zircaloy-2 cladding 2

temperature should not exceed 1500 *C.

For Aerojet-General Nucleonics (AGN)-201 reactor types, NRC accepts that the fuel temperature should not exceed 200 *C.

The applicant should base SAR analyses on the applicable fuel developer's reported test results to ensure fuel integrity under all operating conditions.

27 1

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j SAR Standard Format and Content i

Chapter 14 - Draft August 2, 1993 i

i 2.1.1 Important Process Variables ANSI /ANS 15.1 proposes a list of parameters that may be acceptable as process I

variables for non-power reactors and states that safety limits will be measur-l i

able parametors. However, as discussed in the Sections 2.1.2 and 2.1.3 below, not all safety limits for non-power reactors must be monitored and actually measurable. Safety limits could be inferred from limitations on other process variables.

i l

It is convenient in discussing fuel integrity to divide the non-power reactors i

t i

into two groups: those with engineered cooling systems (forced-convection cooling) and those without engineered cooling systems (natural-convection 3

cooling or no active cooling system). Safety limits for these reactors are discussed in the following two sections.

2.1.2 Criteria-Reactors With Engineered Cooling Systems The NRC modifies this section of ANSI /ANS 15.1 as follows. Operation of the cooling system for reactors with forced-convection cooling maintains fuel temperature within acceptable limits to ensure cladding integrity.

Important parameters include fuel temperature, coolant flow rate, coolant inlet temperature, height of coolant above core, and reactor power level. These l

parameters should be controlled and measured. When all values are jointly maintained within the limits determined by the safety analyses, fuel cladding integrity will not be lost. These parameters are important process variables 4

on which safety limits should be established and specified in the technical specifications. Safety limits should preclude flow instabilities in the hottest channel and ensure that the minimum departure from nucleate boiling ratio (DNBR) is at least 2.0 (which has been an acceptable margin to the onset i

of nucleate boiling). The analyses should be over the range of all physical and engineering parameters of the fuel components, the core configurations, and the coolant systems, as well as include consideration for uncertainties.

i for reactors that will operate with both natural-convection cooling and forced-convection cooling, safety limits should be specified for appropriate l

process variables in both modes of operation (see Section 2.1.3 below). All non-power reactors should be designed so that both fission heat and decay heat can be dissipated without fuel damage. The analyses also should show that the safety limits are not exceeded during all anticipated modes of operation.

2.1.3 Criteria-Reactors Without Engineered Cooling Systems For reactors that will be licensed to operate without forced-convection cooling, only criterion 2.1.3(2) of ANSI /ANS 15.1 is acceptable according to 10 CFR 50.36(c). The NRC modifies Section 2.1.3(2) of ANSI /ANS 15.1 as follows.

For reactors that either circulate coolant by natural thermal convection or have no specific coolant or cooling systems, thermal-hydraulic coolant parameters are not separately controllable. The SAR should ensure fuel cladding integrity and identify appropriate parameters chosen for safety limits in the technical specifications.

28

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SAR Standard Format and Content Chapter 14 - Draft August 2, 1993 High fuel temperature is the likely precursor of fuel failure. Therefore, a maximum allowable fuel temperature safety limit should be established below which fuel integrity is ensured. On the basis of this fuel temperature, a power level should be calculated using an appropriate margin that ensures the fuel remains below the fuel temperature safety limit.

If the license will contain a provision to measure fuel temperature, the maximum fuel temperature in the core would be the parameter on which a safety limit is established. The SAR should show the relationship between the mea-sured fuel temperature and the maximum fuel temperature for the proposed reactor conditions.

If there is no provision for measuring fuel temperature directly, the calculated power level should be selected as the safety limit, on the basis of the maximum allowable fuel temperature and appropriate margin.

However, the basis for the safety limit still should be the maximum allowable fuel temperature.

Because most TRIGA-fueled cores have at least one instrumented fuel rod, the NRC has accepted fuel temperature alone as the safety limit for these reat-tors.

Because the point of measured fuel temperature is normally not the point of maximum fuel temperature, the SAR must show the relationship between the measured fuel temperature and the maximum fuel temperature.

For plate-type fuel, fuel temperature measuring capability is generally not available.

Therefore, the licensee should determine a fuel cladding temperature Y.ow which claading damage (softening or blistering) can be i

precluded. The licensee should then establish a corresponding power level, reactor conditions, and uncertainties that limits cladding temperature below the damage limit.

i For reactors without fuel elements, such es the homogeneous AGN-201 reactors, safety limits should be based on consideratims similar to those for plate-type fuel. The power level established in the SAR as a safety limit must provide reasonable assurance that fission products will not be celeased from their confining barrier, which could be either the fuel matrix or the fuel canister.

Safety limits should be based on the SAR. The technical specifications should i

discuss the mechanism and magnitude of the fuel limitation, including a primary reference for the fuel development studies that support the safety i

limit presented (NUREG-1282 and -1313; Simnad et al., 1976).

The analysis i

should address authorized core configurations and limiting thermal power levels and conditions for the reactor.

Safety limits acceptable to NRC for various reactor fuels are discussed in Section 2.1 above.

2.2 Limiting Safety System Settings l

The NRC accepts the guidance of this section of ANSI /ANS 15.1.

The SAR should address normal operating conditions, off-normal operations, and all pertinent postulated accident scenarios.

For each parameter on which a safety limit is established by the SAR, a protective channel should be identified that 29

l SAR Standard Format and Content Chapter 14 - Draft August 9, 1993 prevents the value of that parameter from exceeding the safety limit.

The calculated set point for this protective action, providing the minimum acceptable safety margin considering process uncertainty, overall measurement l

uncertainty, and the transient phenomena of the process instrumentation, is defined as the " limiting safety system setting (LSSS)." Because the LSSSs are analytical limits, the protective channels may be set to actuate at more conservative values. The more conservative values may be established as l

limiting conditions of operation (LCOs).

LCOs also may be determined on the basis of experience, which has shown that safety system channels can be set readily within 20 percent of the normal operating value for a measured parameter, if the LSSS is not exceeded, without undue interference on operations.

In many cases the LCO can even be within 10 percent of the operating value.

If LCOs are set more than 20 percent from the operating value of the parameter, the SAR justification should be referenced.

2.2.1 Criteria-Reactors With Engineered Cooling Systems The NRC accepts this section of ANSI /ANS 15.1 for the forced-convection cooling mode of operation. For reactors licensed to operate with forced-convection cooling, this specification should list the LSSS derived in the SAR for each reactor parameter for which a safety limit was established. The bases part of this specification should indicate the SAR assumptions and limits of uncertainty for each analyzed LSSS.

For reactors licensed to operate in forced-and natural-convection cooling modes, there should be appropriate LSSSs for both modes.

2.2.2 Criteria-Reactors Without Engineered Cooling Systems The NRC substitutes the following guidance for Section 2.2.2 of ANSI /ANS 15.1.

Section 2.1.3 above requires safety limits be established by SAR analysis for all licensed reactors; therefore, channels should be established on the basis of SAR analysis to not violate each of these safety limits.

Calculated LSSSs defined in Section 2.2 of ANSI /ANS 15.1 and this document should be provided as technical specifications.

3 Limiting Conditions for Operations t

LCOs are derived from the safety analyses in the SAR, which provide the bases for the LCOs.

LCOs are implemented administratively or by control and monitoring circuitry to ensure that the reactor is not damaged, that the reactor is capable of performing its intended function, and that no one suffers undue radiological exposures because of reactor operations.

The NRC accepts this section of ANSI /ANS 15.1 as amplified in the following sections. Many of the LCOs have evolved from experience. Many are facility specific, depending on reactor type, operating characteristics, and site loca-tion. The NRC accepts the LCOs discussed in this section provided the appli-cant justifies the LC0 and shows the applicability to the specific facility.

Additional specifications may be appropriate for unique facility designs or experimental features or for additional conservatism in operations required by the applicant or NRC. As noted above, LCOs can in many cases be set within 28

l l

l l

I SAR Standard Format and Content Chapter 14 - Draft August 9, 1993 10 percent of the normal operating level of a parameter. Specifications on surveillance intervals for LCOs and other parameters and facility design i

features are provided in Sections 4 and 5, respectively, of ANSI /ANS 15.1.

LCOs should be provided as outlined in the remainder of this section.

3.1 Reactor Core Parameters l

(1) Excess Reactivity f

l The upper limit for allowed excess reactivity should be specified. The l

referenced SAR analyses should discuss all operations that require excess reactivity and the safety implications for the excess reactivity pro-i posed. The discussions should include operational flexibility, potential i

accidents, and the relationship to shutdown margin. The SAR should contain a discussion of the safety implications-of the excess reactivity, including resultant shutdown reactivity with all control rods inserted effects on the reactor of any credible rapid removal of a control or safety rod potential effects of other maximum credible rapid additions of l

excess reactivity l

possible reactivity changes caused by experiment failure or displacement interrelationship between shutdown margin and excess reactivity If none of the postulated events would lead to loss of fuel integrity or to uncontrolled release of radioactivity, the proposed excess reactivity would be acceptable.

(2) Shutdown Margin A single value for the shutdown margin, as defined in Section 1.3 above, i

l should be specified. The specification should state that compliance with the shutdown margin takes precedence over the excess reactivity specifi-cation.

In addition, other reactor parameters that apply to the shutdown margin should be stated. These should include cold, clean core reactivity conditions (e.g., temperature and poisons), core configuration (e.g., fuel and control rods), and the status of experiments (e.g.,

movable experiments in their most reactive state). The value of the shutdown margin should be large enough to be readily determined experimentally, for example, 20.5% Ak/k or 20.50 dollars.

(3) Pulse Limits Because the TRIGA design is the only reactor design the NRC currently i

licenses to pulse at this time, the specific values given in the discussion below apply only to TRIGA design pulsing reactors. However, 1

l 29

1 l

i SAR Standard Format and Content Chapter 14 - Draft August 2, 1993 l

t i

the general design criteria discussed below may be applicable to other j

potential pulsing non-power reactor designs.

The maximum reactivity addition for a pulse is a license condition similar to maximum thermal power and is determined case by case. The value should be based on the SAR analysis for maintaining fuel integrity, which considers fuel type, limiting core configurations, reactivity feedback coefficients, operating history, heat capacity, and peak fuel temperature limitations. This LC0 on the maximum reactivity addition administratively gives assurance that the maximum pulse reactivity addition license condition and the safety limit on maximum fuel temperature will not be exceeded.

The SAR should show that the maximum reactor pulse for a TRIGA reactor with stainless-steel-clad UZrH,,g fuel would not raise the peak fuel temperature of any element above 1000 *C (Simnad et al., 1976).

(This is a conservative limit, proposed by General Atomics and accepted by the NRC, that is not to be confused with the safety limit temperature value.)

For a TRIGA reactor with aluminum-clad UZrH,o fuel, the analysis should show that the peak fuel temperature will nok exceed 500 *C for the maximum reactor pulse. The analysis should be applicable to the specific reactor considering its core size, operating history, fuel types, feedback coefficients, temperature gradients and the power peaking of all authorized core configurations. The potential effects of pulsing on in-l core experiments or detectors should be included in the analysis. Any required limitations on experiments should be included in Section 3.8 of I

the technical specifications document.

I The report by Simnad et al., (1981) discussing fuel damage at Texas A&M University TRIGA Reactor should be reviewed to determine if reactor operating history and power level would require a lower peak pulse fuel temperature because of damage to the fuel during pulsing operation.

If the analysis of the core shows that the worth of the pulse rod could exceed the maximum reactivity insertion limit and allow an amount of reactivity to be inserted into the core that could damage the fuel, there should be a limit on the total worth of the pulse rod. There should be a l

steady-state power level above which pulses shall not be initiated.

For l

TRIGA reactors, the NRC has accepted a power level of I kW in conjunction l

with an interlock that prevents movement of the steady-state control rods l

when the reactor is in the pulse mode.

For other pulsing reactors, the proposed limiting fuel temperature and reactivity insertion should be justified by reference to appropriate tests and analyses.

(4) Core Configurations l

The applicant should specify special core configurations, experimental facilities internal to the core, special neutron reflectors, burnable poisons, or mixed fuel types assumed in the SAR. The following specifications should be included in the LC0 for core configurations:

32

I t

SAR Standard Format and Content Chapter 14 - Draft August 9, 1993 If analysis shows the reactivity effects of waterholes in the core

_i needs to be limited due to reactivity insertion accidents, the reactor should have a closely packed core with acceptable vacancies j

in the core center and periphery described. This does not prevent l

the use of in-core experimental facilities. Reactors with thermal power levels in excess of I megawatt and cross-section area of core j

experimental facilities greater than 16 square inches will be licensed as testing facilities.

No fuel should be inserted or removed from the core unless the reactor is subcritical by more than the worth of the most reactive fuel element.

If control rods need to be removed from the reactor core for l

inspection, an LC0 should state the negative reactivity necessary in the core before a control rod can be removed.

f Non-power reactors should be designed with reactivity and void coefficients and a power defect sufficiently negative that many reactor transients are inherently counteracted to avoid loss of fuel integrity.

l Although the individual reactivity coefficients and power defect are addressed in the new specification below, this LC0 should be used to develop specifications on allowed core configurations to ensure the j

assumptions used in the development of limits on those parameters are l

met.

l l

The specified conditions of core configuration are acceptable to NRC if the SAR shows that none of the conditions analyzed could lead to loss of fuel integrity, uncontrolled release of radioactivity, or potential exposures exceeding 10 CFR Part 20.

l (5) Reactivity Coefficients (Added by NRC) t Non-power reactors should specify reactivity coefficients for fuel temperature, moderator temperature, and void volume and a power defect.

The net effect of the coefficients and the power defect should be negative over most of the range of reactor operations. The SAR analyses of both routine operation and potential accident scenarios should show that the net negative effect of these core characteristics is sufficient to mitigate any anticipated event or postulated accident scenario.

Reasonable values should be designed into the reactor (e.g., by under-moderation of the neutron spectrum). Values for surveillance should be specified for those negative reactivity coefficients and the negative power defect that can be measured. The values of the coefficients and the power defect are acceptable if they ensure that the assumptions and initial conditions of the analyses are enveloped to prevent compromise of the fuel integrity during reactor transients and other applicable accident scenarios.

31

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SAR Standard Format and Content Chapter 14 - Draft August 9, 1993 j

(6) Fuel Parameters (Added by NRC)

An LCO should be specified for certain fuel parameters or l

characteristics.

Design features of the approved fuel should be included in Section 5 of the technical specifications document.

Fuel-related LCOs include the i

following:

l i

(a) All Fuel Types No operation with damaged fuel except to locate such fuel. The

=

definition of damaged fuel should specify limits on longitudinal growth, bowing, or bending and limits on detectable amounts of fission products that could escape through the primary barrier.

l 3

Periodic visual inspection of fuel.

This specification should i

be clear and explicit and reference fuel manufacturers' l

guidance or recommendations for detecting deterioration.. The l

intervals and methods of fuel inspection should be specified in j

Section 4 of the technical specifications document. The purpose of inspection is to detect cladding deterioration that results from erosion, corrosion, or other damage.

I (b) TRIGA Fuel i

Additional technical specifications limit fuel rod elongation, bowing, and uranium burnup. Limits listed below were proposed by General Atomics and accepted by the Atomic Energy Commission (AEC) during initial licensing of pulsing TRIGA reactors.

If these limits are reached, the fuel element is defined as " damaged fuel."

Acceptable specifications for TRIGA fuel for both' steady-state and pulsed operation include the following:

Bowina - For stainless steel-clad UZrH TRI sagitta shall not exceed 0.125 in. (0.$N cm)GA fuel, the over the length of the cladding in a hexagonal-grid core arrangement or 0.0625 in. (0.159 cm) elongation over the original length of the cladding in a circular-grid core arrangement.

For aluminum-clad UZrH,o fuel, the limit on sagitta should be 0.125 in.

(0.318 cm Elonaation - For stainless steel-clad UZrH TRIGA fuel, the 3 65 total length of the fuel element shall not, exceed its original j

length by more than 0.125 in. (0.318 cm).

For aluminum-clad UZrH f

(1.2fcm)uel, the limit on elongation should not exceed 0.5 in.

32

l SAR Standard Format and Content Chapter 14 - Draft August 9, 1993 Burnuo - The burnup of uranium-235 in the UZrH fuel matrix shall not exceed 50 percent of the initial concentration (NUREG

-1282 and Simnad and West, 1986).

t (c) Materials Testing Reactor (MTR)-Type Fuel To prevent fuel swelling there should be burnup limitations on the fuel. Aluminum-clad aluminum-matrix MTR-type fuel plate non-power reactors should have technical specifications that limit U-235 burnup or fission density. The specifications are acceptable if they are consistent with the SAR, which accounts for all relevant thermal-hydraulic and metallurgical considerations. The NRC is specifically concerned with the maximum burnup limit for plate-type fuels because of the build up of oxide on the fuel cladding. This can be a concern when licensees apply to increase the maximum acceptable burnup. The increased resistance to heat transfer to the coolant may affect consequences considered in the accident analysis chapter of the SAR. NRC has accepted uranium burnup densities in i

fuels based on a uran (um aluminide matrix up to a fission density of d

2.3 x 10 fissions /cm and up to 50 percent of the initial concentration of uranium-235 (Beeston et al.,1980; Gibson,1967; i

Nazare et al.,1975; Stahl,1982).

i (d) PULSTAR Fuel j

Burnup of pin-type PULSTAR fuel should be limited by a specification based on testing and the SAR analysis. The NRC has accepted burnup limits up to 20,000 MW.d/ tonne uranium.

LCOs are acceptable if they are analyzed in the SAR and consistent with the values given above. The analyses should verify for these fuel parameter conditions that the fuel will not exceed safety limits for l

normal and off-normal operations.

3.2 Reactor Control and Safety Systems (1) Operable Control Rods The number and type of operable control and safety rods should be speci-fied. There is no prescribed minimum number of operable control and safety rods for non-power reactors. The specification regarding the number of operable control rods is acceptable if the excess reactivity and shutdown margin specifications required by the SAR analyses can be 1

ensured for all operating conditions. The individual or total reactivity worths need not be specifically listed. A rod of lesser worth might be designated the " regulating rod" and is used as a fine power adjustment mechanism.

In some cases the worth of a control rod (s) connected to an automatic control system (which can add reactivity) may be limited to a maximum amount that was assumed in the SAR in this LCO. This regulating rod need not have scram capability but rods without scram capability should not be used when showing compliance with shutdown margin requirements. Other rods of greater worth, with automatic protective 33 l

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l SAR Standard Format and Content i

Chapter 14 - Draft August 9, 1993 l

(scram) function, should be capable of achieving the specified shutdown margin.

The maximum scram time should be specified for each scrammable rod. The specification should ensure that the drop times are consistent with the SAR analysis of reactivity required as a function of time to terminate a reactivity addition event accounting for measurement and calculational uncertainties.

In most non-power reactors, for rods 2-to 3-feet long, full rod insertion time in the absence of excess mechanical friction or interference is less than 1 second.

If a specification proposes a longer l

scram time, it requires appropriate SAR analysis. The NRC finds it l

acceptable to shut down a non-power reactor by intentionally scramming the control and safety rods.

j (2) Reactivity Insertion Rates The maximum rates of adding positive reactivity should be specified for the control and safety rods. The specification should specifically state that gang or multiple rod withdrawal is allowed. Control rod (s) con-nected to an automatic control system may have maximum rates of reactiv-ity addition that differ from the rest of the control rods. The accept-able rates should be based on the SAR, including inadvertent addition of l

ramp reactivity at the maximum rate for the most conservative power, rod position, and reactor conditions.

(3) Pulsed Operation Limitations on reactivity additions are discussed in Section 3.1(3) above and need not be repeated here.

If any hardware systems require special limitations for pulsing, they should be discussed in this section of the technical specification document.

Examples might include (a) special I

core configuration, (b) specific location of pulse rod, (c) number of l

pulse rods, and (d) removal of in-core fueled experiments. These specifications are acceptable if the assumptions of the SAR are ensured and damage to the reactor by authorized pulses is precluded.

If an experiment containing fissionable material could be damaged by reactor pulsing, limiting specifications must be provided to preclude this event in Section 3.8 of the technical specifications document.

(4) Scram Channels A table should specify all required scram channels and set points, the minimum number of channels, other functions performed by the channel, and reactor operating mode, such as steady-state power or pulsed, and cooling method, such as forced-or natural-convection coolant flow. The safety limits that the scram protects should be discussed in the basis for the table. Table 1 provides an example of how the information could be displayed. Reactor scrams should be based on the SAR. There should be at least two completely independent power level scram channels and they should provide diversity and redundancy.

34

1 SAR Standard Format and Content i

Chapter 14 - Draft August 10, 1993 Historically, NRC has accepted power level scrams as high as 1.2 times the licensed power. This is the only non-power reactor scram set point j

that, if reached, violates the license (maximum power level). Some licensees have incorrectly interpreted this scram set point as allowing limited operation above the license power level. Although this is gen-l erally not a safety concern, the NRC staff recognizes it as a regulatory problem.

For example, if the reactor power measuring channels are out of l

calibration, it is possible that the reactor has been operated at several percent above the maximum licensed power. level for a period of time. To ensure that licensed power levels are not exceeded for non-power reac-i tors, power level scrams should be set below the licensed power level.

l The NRC staff, upon submission of a license amendment request, has approved license amendments for non-power reactors that raise the j

licensed power 10 percent above the power value at which the reactor will be operated. This allows the operating power to remain the same while l

retaining the scram set points within the license and near the previous j

values.

In this example, the scram set points will be less than the I

licensed power level and will scram the reactor before the licensed power j

level is exceeded. Power level scram set points above the licensed power j

level will only be acceptable on a case-by-case basis when justified by analysis in the SAR which includes conservative assumptions with regard to fission product inventory, decay heat, fuel temperature, and other pertinent reactor conditions.

l l

l I

35

l SAR Standard Format and Content Chapter 14 - Draft August 10, 1993 l

Table 1 Typical required scrams and power reverses

  • Mininun tuber Channet set point **

regaired and Fmetion Period safety Scram if period 5 3 see 1

Period reverse Rod rm in if period 510 see 1

Power level safety (linear and safety)

Scram if power > 100%

2 Power level reverse (safety)

Rod run in if power > 97%

1 High power /no coolant flow Scram if flow < 56.8 t/sec (900 ppm) 1 and power > 100 kW High power / flapper open Scram if power > 100 kW 1

and flapper open Flapper closed /no coolant flow Scram if flow < 56.8 t/sec (900 ppm) 1 and flapper closed Software (digital) malfunction Scram upon malfunction 1

Loss of high voltage to detectors Scram if voltage lost 1

Pool water level Scram if level < 4.88 m (16 ft) 1 above core top Bridge not clamped Scram when clanps released 1

I Bridge radiation levet and Scram if radiation t 50 mrenuhr ard 1

Building exhaust air radiation level concentration t 2 x 10-' #Ci/nt 1

Manual scram switch Scram when switch depressed 1

Rod magnet power keyswitch Scram when magnet power turned off 1

Fuel tenperature Scram if tenperature t 550 'C (1022 'F) 2 Reactor coolant exit temperature Scram if tenperature t 55 'C (131 'F) 1 Autonatic control system out of limit Rod run in if out of specification 1

Experiment scram if setpoint is violated 1

Loss of site power Scram if power lost 1

l

  • As illustrative values, the set points and channets listed do not apply to any one reactor.

l

    • Values listed are limiting set points. For operational convenience, set points may be changed to more conservative values.

36 SAR Standard Format and Content Chapter 14 - Draft

~

August 2, 1993 retaining the scram set points within the license and near the previous values.

In this example, the scram set points will be hss than the licensed power level and will scram the reactor before the licensed power level is exceeded.

Power level scram set points above the licensed power level will only be acceptable on a case-by-case basis when justified by analysis in the SAR which includes conservative assumptions with regard to fission product inventory, decay heat, fuel temperature, and other pertinent reactor conditions.

(5) Interlocks Required interlocks that inhibit or prevent control rod withdrawal or reactor startup should be specified by a table (see Table 2 as an example).

Interlocks should be specific to the facility and based on the SAR. These interlocks include operability of area or other radiation monitors a

experiment facilities e

confinement and ventilation systems a

initial conditions for pulsing detected neutrons for startup Operability of measuring channel components, such as ion chamber power supplies and recorders as discussed in the SAR Table 2 Typical required interlocks

  • Minin.rn Ntsnber Channel Required Funetion Recorders not operating 3

Prevent rod withdrawal (startup inhibit)

Neutron count rate (startup) 1 Prevent rod withdrawal (startup inhibit) if countrate s 2 cps sinultaneous rod withdrawal 5

Prevent withdrawal of 2 or more rods Nonpulse condition 1

Prevent movement of pulse rod in steady state mode Pulse withdrawal 4

Prevent movement of standard control rods in pulse mode Transient withdrawal 1

Prevent movement of pulse rod with reactor power above 1 kW

  • Values listed are limiting set points. For operational convenience, set points may be changed to more conservative values.

If the reactor will be licensed to operate in more than one mode, the specification should include the mode for which the interlock is required.

If permanent interlocks are established for special experi-39

SAR Standard Format and Content Chapter 14 - Draft August 10, 1993 (5) Interlocks Required interlocks that inhibit or prevent control rod withdrawal or reactor startup should be specified by a table (see Table 2 as an example).

Interlocks should be specific to the facility and based on the SAR. These interlocks include l

operability of area or other radiation monitors i

experiment facilities a

confinement and ventilation systems l

initial conditions for pulsing l

l detected neutrons for startup Operability of measuring channel components, such as ion chamber power supplies and recorders as discussed in the SAR j

i Table 2 Typical required interlocks

  • Mininun f

Nunber Channet Required Funetion l

Recorders not operating 3

Prevent rod withdrawat (startup inhibit)

Neutron count rate (start @)

1 Prevent rod withdrawal (start e inhibit) l if countrate 5 2 cps Simultaneous rod withdrawal 5

Prevent withdrawat of 2 or more rods i

Nonpulse condition 1

Prevent movement of pulse rod in steady state Pulse withdrawal 4

Prevent.kovenent of standard control rods in pulse mode Transient withdrawal 1

Prevent movement of pulse rod with reactor power above 1 kW

  • values listed are timiting set points. For operational convenience, set points may be changed to more conservative values.

If the reactor will be licensed to operate in more than one mode, the specification should include the mode for which the interlock is required.

If permanent interlocks are established for special experi-ments, shields, or access control, they should be included in the technical specificatior.s, as described in the SAR.

37 l

I SAR Standard Format and Content Chapter 14 - Draft August 10, 1993 i

i (6) Backup Shutdown Mechanisms

!i Most non-power reactors are required to use only control and safety rods for shutdown.

If the SAR identifies a-need for backup mechan ums (for example, moderator dump in a critical facility), they should be specified with appropriate requirements placed on their operability (LCOs).

l (7) Bypassing Channels

)

Any individual channels identified in items 4, 5, or 6 above for which bypassing is allor:ed during reactor operation should be justified in the l

SAR and specified under this item. Only minimal bypassing should be permitted in safety systems and never in a system that could compromise i

scram capability of the other channels.

Bypassing temporary scrams or interlocks associated with experiments need not be included in the tech-nical specific &tions but should be addressed in specific experiment protocol.

(8) Cont ol Systems and Instrumentation Requirement: for Operation (Added by NRC) i Non-power resctor technical specifications should have redundant and accurate power level monitors that cover the range from subcritical J

source multiplicaticn to above the full licensed power _ level. Not all monitors are required to include scram capability (see Table 3 for a typical minimum set). These include a startup channel, linear power monitor, logarithmic power mor.itor, and safety channel (s).

In addition, most non-power reactors have a period channel (meter), including a period scram. One.thtuld be specified as analyzed in reactor transient response l

i section of the SAR.

l l

Table 3 Typical required minim.im measuring channels

  • i Minian I

ru ter Channel required Fmetion Startup 1

Monitor suberitical nuttiplicution for -

startup I

Power level 2

Input for safety power level scram Pulse pcwer 1

Input for pulse power level scram Fuel temperature 2

Input for fuel tenperature scram Log W/ period 1

Wide range power level and input for period meter and period scram Linear power level 1

Display power for control w 16 1

Display power level

  • As illustrative values, these channels do not apply to any one reactor. Minlaun channels for a 3_,;icular facility are determined from the SAR analysis.

38

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i SAR Standard Format and Content I

Chapter 14 - Draft j

August 10, 1993 Some non-power reactors with forced-convection cooling have a channel that displays the radiation level of N-16 in the primary flow. Although the N-16 channel is not required in the SAR for mitigating transients it can be an important channel because (1) it has greater stability than delta temperature across the core during power changes and (2) it is not affected by changes in core flux distribution car cd by fission product i

buildup in the core that can affect the ioniza+

chambers.

If it is

{

necessary for the operators to use the N-16 chan91 in reactor j

operations, it should be on the list of specified channels.

j i

If digital control and safety instrumentation is used, an analog system

)

should be specified to provide diversity and redundancy.

-Specifications in this section should include the entire channel, l

including readout meters and recorders and the protective functions they j

perform, such as to prevent an LSSS from being exceeded.

i l

l Each non-power reactor should have more than one power level channel l

l indication in the control room when operating at full licensed power.

I However, because sensors and channel electronics might not be identical, the channels may indicate slightly different power levels.

Power level is a principal license condition, and each applicant may consider i

designating a primary channel for power level monitoring. That channel should be calibrated for thermal power in the region of maximum licensed power and should be recorded in a way that allows auditing for later proof of authorized operation within the license condition.

Facility procedures should identify this designated channel and allow for alternate designations using analytic comparisons to achieve operational flexibility, if necessary. However, technical specifications or facility procedures that do not include this concept are acceptable to NRC because a dr sgnated channel is not a requirement upon licensees.

3.3 Coolant Systems The basic systems required for cooling the fuel and other components, for limiting corrosion, and for monitoring coolant radioactivity in non-power reactors should be specified in this section. All non-power reactors should have the capability to remove both fission and decay heat to ensure fuel integrity under all potential conditions. All reactors requiring forced-convection cooling systems shall have specifications a.ssuring operability of systems and reactor configurations for fail safe changeover from forced-convection to natural-convection cooling. An adequate heat sink, as described in the SAR, is a necessary component of such a system.

Fev reactors licensed to operate in both forced-and natural-convection cool-4.g modes, the appropriate coolant system configurations and the relevant power levels for both modes should be specified as analyzed in the SAR.

Not all of the following items apply to all types of non-power reactors.

Ilowever, when applicable, they should be limited by technical specifications, on the basis of the analyses and justifications in the SAR.

39 i

SAR Standard Format and Content Chapter 14 - Draft August 10, 1993 (1) Shutdown Cooling or Pump Requirements As a minimum, the requirements for natural-convection cooling and the operability and status of related systems required for shutdown should be specified as LCOs.

If additional requirements are necessary for tem-porary forced-convection cooling following reactor shutdown from extended high-power operation, the technical specifications should state them, using the SAR as the basis.

(2) Isolation Valves The existence, location, operability, and status of any valves required to isolate subsystems or components for operational needs, including removal of decay heat, should be specified as LCOs.

l (3) Coolant level Limits l

Both the coolant pressure (boiling temperature at the fuel) and adequate l

natural-convection flow depend on the level of water above the core.

In addition, vertical and horizontal radiation shielding by the coolant j

might be required. Pool water level should be an LCO for both reasons, i

using the SAR as the basis.

j (4) Detection of Leakage or Loss of Coolant If leakage or other loss of coolant could lead te an uncontrolled release of radioactivity (see items 5 and 8 below) to the environment, an 100 should state the need for operability of detection systems.

If credit is i

taken in the SAR, this includes systems that monitor the pressure differ-ence between the primary and secondary cooling systems at the heat exchanger to detect conditions that would allow loss of primary coolant in the event of a heat exchanger leak. Heavy water systems should always have leak detection capability.

(5) Detection of Fission Product Activity l

l The technical specifications should provide for prompt detection of fission products escaping from the fuel barrier. The method could be a radiation detector placed in the primary coolant flow loop or a strategi-cally located continuous air monitor in the reactor room or a ventilation duct. Temporary substitutions, in case the fission product monitor is inoperable, should follow guidance in Section 3.7.1 of ANSI /ANS 15.1.

This specification may be combined with the specification discussed in Section 3.7.1 (2) on fission product monitors.

The specified fission product monitor should be able to initiate action, i

such as a reactor scram or reactor room isolation. The SAR should pro-vide the bases and describe how fission products are distinguished from other waterborne or airborne radioactivity.

40

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i Chapter 14 - Draft August 10, 1993 l!

i 1

(6) Hydrogen Concentration (Off-Gas) Limits If the SAR has shown any of the isotopes of hydrogen (hydrogen, deuterium, and tritium (H,0,1)) to be a significant risk to personnel or the facility, an LC0 should provide for detection or adequate control,

-I as discussed in the SAR.

l (7) Energency Core Cooling Systems l

If the SAR indicates a need for supplemental core cooling to mitigate a loss-of-primary-coolant event, the technical specifications should con-tain an LCO requiring an operable and adequate system. The system should satisfy the cooling requirements for the SAR scenario and should not depend on continued availability of normal electrical service.

1 (8) Secondary and Primary Coolant Radioactivity Limits In addition to the prompt detection of fission products from failed fuel or experimeht malfunctions, LCOs should limit radioactivity in the coolant. The technical specifications should require periodic sampling and appropriate analyses to detect and quantify radioactivity in both the i

primary and secondary coolant. The coolant should be sampled for gross activity on a short interval, for example, weekly, and sampled for iso-1 i

tope identification on a longer interval, for example, quarterly. The i

purpose of this LCO is to detect deterioration of components in the pri-2 mary coolant loop, such as a control element, and leakage in a heat exchanger into the secondary coolant loop. These specifications should I

be stated in such a way that significant changes in radioactivity, as defined in the SAR, trigger remedial action.

4 (9) Water Chemistry Requirements a

To control corrosion of such components as the reactor fuel, structure, l

and pool, control activation of impurities in the reactor coolant, and maintain visual clarity of the reactor coolant, there should be LCOs on both electrical conductivity and pH of the primary coolant. These specifications also should apply to any water that comes into contact with the fuel, such as in fuel storage tanks and pits. The iimits and ranges of values should be given explicitly and be consiste't with 4

recommended values given by both fuel vendors and water enemistry guidelines. The conductivity should be monitored continuously. There should be a definite schedule for measuring pH, during both operating and shutdown periods. The bases should clearly address the appropriate ranges and give meaningful references. The SAR should justify the values i

for conductivity and pH for the particular reactor. Acceptable ranges for these process variables have traditionally been s5 pmbos/cm for conductivity and between 5.0 and 7.5 for pH. These values can usually be i

achieved by demineralization, filtration, and good housekeeping.

practices, but chemical methods should be described and specified, if j

applicable.

i 41

,, ~. -. _

+

,m,,

,.m.

,,m..,.

m.s e

.,e_.,,y+,,...

SAR Standard Format and Content Chapter 14 - Draft August 10, 1993 3.4 Containment or Confinement Because accidents that result in release of steam and building over pressure are uncommon, most non-power reactors are housed in a confinement system, not a containment. There should be an LC0 requiring that the system specifically described in the SAR exists as stated. The system should be operable during operation and for other applicable times such as before operation and follow-ing shutdown, as noted in Sections 3.4.1 and 3.4.2 of ANSI /ANS 15.1.

If interlocks or administrative controls to ensure operability arc required, there should be appropriate specifications. Whether the facility has confine-ment or containment depends on the reactor design, operating characteristics, and facility location. Specifications should require nominal exhaust rates for air under the operating and accident conditions analyzed in the SAR.

i Specifications should limit building leak rates to those described in the SAR.

i 3.5 Ventilation Systems j

Ventilation and exhaust flow rates and the systems to achieve the controlled release of effluents, as analyzed in the SAR, should be specified as LCOs.

These LCOs should be established to achieve controlled release of effluents.

Automatic fail-safe closure of vents should be specified for confinement sys-tems. Provisions to initiate controlled, filtered, and monitored exhaust and ventilation for radiological accidents should be included.

In some cases, i

depending on the results of the analysis, minimum airflow rates may be LCOs.

l The specified ventilation system should maintain a lower air pressure in the reactor room than in adjacent spaces. Reactor room air should not be distri-buted to other occupied spaces within buildings. The location and height of l

the air exhaust system stack or release point should be specified as an LCO here or as a design feature in Section 5 of the technical specifications document. The dimensions of the stack should be consistent with the assumptions used in the SAR to predict potential radiation doses in the unrestricted environment.

It is acceptable that the concentration of airborne t

l radioactivity at the point of exhaust for normal operation be higher than the J

l regulatory limit for restricted areas, provided that this point is not readily l

accessible to the public, the analyzed doses to the public are well below regulatory limits for unrestricted areas, and the potential doses to the facility staff are within regulatory limits. Requirements for the as-lcy-as-i reasonably-achievable (ALARA) program should be applied in all analyses (see l

Section 3.7 below).

1 3.6 Emergency Power Any requirement for emergency electrical power for non-power reactor facili-l ties should be analyzed in the SAR on a case-by-case basis. Any necessary l

facility functions, such as radiation monitoring, emergency core cooling, or l

isolating containment or confinement, that need to be maintained _ if normal l

electrical power is lost should be described in the SAR.

If emergency power is required, an LCO should ensure operability of the system. The technical i

specification should specify automatic startup of e.mergency electrical power if indicated in the SAR.

42 l

l

i i

SAR Standard Format and Content Chapter 14 - Draft August 10, 1993 3.7 Radiation Monitoring Systems and Effluents Monitoring systems and effluents may be addressed in the technical specifi-cations under separate principal headings. The following discussion is consistent with the corresponding sections of ANSI /ANS 15.1.

3.7.1 Monitoring Systems A separate table in the technical specifications document (see Table 4) should list the required radiation monitors, the function each performs (e.g., scram or containment isolation), the approximate location, the type of radiation detected, and the alarm and/or automatic action setting, as analyzed in the SAR. The set points and calibrations should be listed in terms of radiation exposura rates and concentrations rather than count rates that can change with calibration. Specific count rates for alarms and action settings can be presented in a facility procedure that can be amended in accord with the i

technical specifications document procedures section.

For specified monitors l

that become inoperable, the specification should require that reactor 7

l operations may continue only if the monitor is replaced by a substitute or portable monitor. The replacement monitor must perform essentially the same function until the original monitor is repaired or replaced (generally not to exceed I work week unless justified in the SAR). The specification also should require that if the specified monitor was displayed in the control room, the temporary monitor also should be observable by the operator on duty.

i The applicant should provide a table applicable to the specific facility on the basis of the SAR.

(1) Air Monitors (Gas and Particulate)

I Monitors should be specified for both radioactive gas and those radio-active particulates that might be airborne in the reactor room. There should be at least one continuous air monitor (CAM) with an audible alarm and data recorders. These monitors should be capsble of alerting facil-ity personnel to the presence of radioactivity.

lhey should be cali-brated for anticipated radioactive species.

Potential sources of airborne radioactivity should be analyzed in the SAR.

There should be specifications requiring operability of properly calibrated effluent monitors, preferably with recorded outputs for long-term records that provide documentation of the concentration and total quantity of radioactive effluents, as required in Section 3.7.2 below.

For reactors operated at power levels below a few hundred kilowatts, the concentrations of airborne radionuclides may be too low to measure during normal operation. For these, calculated concentrations of released quantities are acceptable as specifications, using the SAR as the basis.

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August 10, 1993 (2) Fission Product Monitors The specified fission product monitor could be the CAM or the primary coolant monitor, depending on the release scenarios analyzed in the SAR.

Release of fission products both from fuel and fueled experiments should be included. This specification may be combined with the specification discussed in Section 3.3 (5) above.

Table 4 Typical Required Radiation Measuring Channelt,*

Mininun I

runber Set point ecpal Channel recpired F metion or less than 1

Area radiation nonitors 4

Alarm 0.15 mSv/hr 15 mrem /hr i

Hot cell monitor 1

Alarm and door interlock i mSv/hr l

100 mrem /hr I

Reactor bridge 1

Alarm (isolates containment 0.5 mSv/hr l

with building particulate) 50 mrem /hr j

Primary coolant 1

Atarm 0.5 mSv/hr 50 mrem /hr i

building particulate 1

Alarm (isolates containment 2 x 10~' SC1/cm' with reactor bridge) 2-hr particulate i

j Sullding gas (Ar-41) 1 Alarm 2 x 10-' Ati/cm'

)

i Daily release i

l j

Stack particulate 1

Alarm 2 x 10 Sci /cm' 2-hr particulate j

Stack gas (Ar-41) 1 Alarm 2 x 10 sci /ca' Daily release 4 x 10 ' sti/cm' Annual average l

  • As illustrative values, these channels and set points do not apply to any one reactor. Set points for a particular facility must be determined in the SAR analysis.

(3) Area Monitors There should be a specification requiring operable area monitors in and near the reactor room. The type of radiation detected, such as gamma rays or neutrons, should be specified.

Brand name, efficiency, and specific design should be avoided as specifications, but the range of i

1 exposure rates monitored may be specified. These area monitors should provide information on the potential exposure rates from reactor-related radiation. Alarm and automatic action set points should be specified to ensure that personnel exposures and potential doses remain well below limits of 10 CFR Part 20 and are consistent with the facility ALARA j

program.

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SAR Standard Format and Content Chapter 14 - Draft August 10, 1993 1

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(4) Environmental Monitors l

There should be at least one environmental monitoring station in the near l

environment to the facility, preferably at the site boundary or other areas of concern, such as at pcpulation centers or student dormitories.

These monitors should be specified to match the types of radiation anti-cipated and should be either in line-of-sight from the air exhaust point or down-wind in the prevailing wind, as appropriate. The types of moni-tors should be specified (see Section 3.7.1 (4) of ANSI /ANS 15.1). The location and method of determining background readings should be dis-cussed in the SAR and the basis of the specification. The specification should state that environmental monitors are used to verify that the potential maximum dose, annual or other, in the unrestricted environment is within the values analyzed in the SAR. The specification should address both potential accident scenarios and normal operations.

3.7.2 Effluents All radioactive species listed in Section 3.7.2 of ANSI /ANS 15.1 should be i

addressed for normal operations, and the releases should be limited by techni-cal specifications. The NRC accepts the proposed concentration limits, pro-vided the SAR shows that potential doses from these concentrations comply with 10 CFR Part 20 for the maximum exposed member of the public on a facility-specific basis.

Ar-41 is the principal radionuclide released by most non-power reactors.

Even though the doses related to Ar-41 are generally small, a specification should address the average and maximum concentrations in both the restricted and unrestricted areas and the total curies (becperels) released durirq a calen-dar year. The calculated potential doses to the most exposed persons in restricted and unrestricted areas must conform with 10 CFR Part 20 and the l

facility ALARA program.

Because of diffusion and dispersion of the release, if the point of release is inaccessible to the public and generally not accessed by restricted area per-sonnel, the maximum normal concentration at that location may be higher than the concentration allowed in 10 CFR Part 20 for restricted areas.

The SAR must show that the diffused and dispersed release at the point of contact with members of the public is within 10 CFR Part 20 limits. The calculations in the SAR for diffusion and dispersion should be realistic but conservative, and be based on logical models and specified effluent levels.

Because an infinite cloud assumption is extremely conservative for Ar-41 releases, a finite cloud should be considered, as discussed in NUREG-0851 and accepted by the NRC.

l l

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l SAR Sundard Format and Content Chapter 14 - Draft j

August 10, 1993

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l I

3.8 Experiments Experimental facilities should be described in the SAR, and their basic features should be included in Section 5 (Design Features) of the technical l

specifications document. The experiments to be performed in the experimental j

facilities need only be noted briefly.if at all, in the SAR unless they could i

j present a hazard to the reactor facility, the public, or facility staff. Any LCOs for experiments should be performance-based to ensure that no regulations are violated, that experiment safety analysis limits are not exceeded, and i

j that the reactor is not damaged by experiment failure or malfunction.

i Regulatory Guide 2.2, " Development of Technical Specifications for Experiments l

in Research Reactors," provides detailed guidance to applicants'on the scope of the discussions for experiments to be included in the SAR. The regulatory guide also provides guidance on the technical specifications needed to govern the experiments performed. The technical specifications should follow the I

guidance of Section 3.8 of ANSI /ANS 15.1 and Section C of Regulatory Guide i

2.2, as supplemented by the guidance below.

{

3.8.1 Reactivity Limits Limits should be specified on absolute values of reactivity associated with j

each type of experiment:

secured, unsecured, and movable (see ANSI /ANS 15.1 l

and Regulatory Guide 2.2 for definitions).

Generally, the limits on secured i

?

experiments should be approximately twice the limits on unsecured and movable experiments, where the latter should be no more than 1 dollar. The 1-dollar i

limit is such that inadvertent prompt criticality is avoided even if failure of the experiment were to occur. Movable experiments must be clearly defined 2

j to include those to be inserted or removed while the reactor is operating.

Unsecured experiments include those installed before reactor startup that change position or other conditions while the reactor continues to operate.

Reactivity limits of experiments that change position while the reactor is operating must not exceed the ability of the reactor operator or automatic 1

servo system to maintain control of the reactor. The specified reactivity I

limits on movable experiments should not permit the violation of the shutdown margin specification.

]

The specified sum of the absolute values of the reactivity worths of all experiments should not be more than twice the limit on individual secured i

4 experiments. The value should be consistent with the SAR analysis of inad-vertent reactivity insertions, as explained in Section C.I.a. of Regulatory Guide 2.2.

4 There should be a specification requiring that the reactor be shut down during i

the changing or moving of any secured experiment.

1 i

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Chapter 14 - Draft August 10, 1993 l

3.8.2 Materials For fissile materials in evperiments, limits should be specified on the l

allowed thermal power and on the equilibrium or maximum inventory of specific fission products, such as iodines and strontium.

Specifications such as those j

indicated in Section C.2.a of Regulatory Guide 2.2 are acceptable.

j A specification should require double encapsulation of liquid, gas, and potentially corrosive materials. The failure of an encapsulation of material l

that could damage the reactor should require removal and physical inspection of potentially damaged components.

Specifications should limit the quantity of explosive material parmitted in the reactor experimental facilities and in the reactor fscility.

For experi-3 mental facilities, the upper limit should be 25 mg TNT or its equivalent, as indicated in Section C.2.d of Regulatory Guide 2.2.

For the overall reactor i

facility, the upper limit will be no higher than 100 mg TNT or its equivalent 4

unless analyzed in the SAR and approved by NRC. An additional specification j

should require prior testing or analyses of explosive material encapsulations t

to ensure no reactor damage in the event of detonation, regardless of the

' limit.

)

A specification should limit the quantities of unknown materials that could be placed in certain experimental facilities for exploratory studies. Confor-mance with Section C.2.1 of Regulatory Guide 2.2 would be acceptable.

3.8.3 Failure and Malfunctions Specifications that address the failure and malfunction of an experiment and limit the experiment parameters should be included on a case-by-case basis, as discussed in the SAR. The guidance of Section 3.8.3(2) of ANSI /ANS 15.1 i

should be followed, but specifications that require compliance with regu-lations are redundant and are not necessary (see Section 3.8.3(1) of ANSI /ANS 15.1).

2 For experiments that may off-gas, sublime, volatilize, or produce aerosols, standard assumptions are often specified for calculating the activity that could be released under normal operating conditions, accident conditions in the reactor, and accident conditions in the experiment. Such specifications ensure conservatism in the safety analysis of. the experiment. These specifi-cations have included such assumptions as (1) if an experiment fails and releases radioactive gases or aerosols to the reactor bay or atmosphere,100 percent of the radioactive gases or aerosols escape; (2) if an effluent holdup tank isolates on a high radiation signal, at least 10 percent of the radioac-tive gases or aerosols escape; (3) if the effluent exhausts through a filter with 99 percent efficiency for 0.3 micron particles, at least 10 percent of the vapors escape; and (4) if an experiment fails that contains matericis with a boiling point above 130 *F (54 *C), the vapors of at least 10 percent of the i

materials escape through an undisturbed column of water above the core. The particular assumptions used, if any, must be derived from the SAR.

47

SAR Standard Format and Content Chapter 14 - Draft August 10, 1993 Applicable limits for specific experiments are normally not part of the l

technical specifications anc' should be derived from the experiment safety l

review discussed in Section 6.5 below.

3.9 Facility-Specific LCOs The LCOs discussed above apply to most non-power reactors.

Each reactor may also have technical specifications containing facility-unique LCOs. These should be based in the SAR and facility design.

4 Surveillance Requirements Certain LCOs established in Section 3 of the technical specification document should be accompanied by a surveillance requirement in Section 4.

These surveillance-related specifications should clearly identify the parameter or function to be measured or tested, the method, the frequency, and the accept-able deviation or error. Acceptable deviations might be limited by license conditions (such as thermal power level) or by regulations (such as 10 CFR Part 20).

The NRC accepts the surveillance frequencies stated in this section of ANSI /

ANS 15.1 as amplified in th following sections. The actual wording of the

)

specifications should not be ambiguous. k'ording in ANSI /ANS 15.1 has been interpreted incorrectly by some licensees to allow the extended ir.terval (interval not to exceed statement) as the average.

If the extended interval is used for a particular surveillance test, a shorter interval should be used as soon afterwards as possible to adhere to the average. The intervals of ANSI /ANS 15.1 should be explicitly listed in the applicant's technical specifications.

l In addition to surveillance verification of LCOs, other surveillance activi-ties should be specified.

These include specific specifications, such as periodic pulse rod maintenance and cleaning, thermal power level calibration, preventive maintenance and inspection of control / safety rod drive systems, fuel element inspections, preventive maintenance on other important components to provide assurance of operability, and calibration of effluent monitoring systems.

If a surveillance specification is not required for safety while the reactor is shut down, it may be deferred, but must be performed before reactor startup.

If the reactor is not to be operated in a particular mode (e.g.,

pulse mode) for an interval of time that exceeds the surveillance intervals l

for that particular mode, surveillance specifications not required for safety (an example is the requirement for a standard pulse to be performed every year) while the reactor is operated in other modes may be deferred, but must be performed before the reactor is considered operational in the mode in which surveillances were deferred. Scheduled surveillance that cannot be performed while the reactor is operating may be deferred until the next planned reactor shutdown.

Surveillances that may be deferred and the reasons for deferment must be clearly stated in the technical specification, justified in the SAR, and noted in the basis of the specification.

48

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Chapter 14 - Draft j

August 10, 1993 j

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i In general, any time that a reactor system or component is modified or repaired, the surveillance requirements for that system should be performed as part of the operability check of the system or component. This should be done regardless of when the surveillance was last performed or when it is next due.

l This special surveillance may change the due date of the next regularly schedulec' surveillance of thct type.

I 1

I 4.1 Reactor Core Parameters 1

i The excess reacthily and shutdown margin LCOs specified in Section 3 of the technical specification document are applicable for all authorized operating l

conditions. As an example, for a movable experiment, the specifications for

]

excess reactivity ard shutdown margin surveillance measurements should be based on that experiment being in its most reactive location.

In addition, other reactor parameters that affect reactivity during operation should be j

explicitly specified.

For the following specified surveillance requirements, j

the parameters may be determined by an appropriate combination of measurements and calculations.

1 (1) Excess Reactivity Excess reactivity should be determined at least annually and after changes in either the core, in-core experiments, or control rods for which the predicted change in reactivity exceeds the absolute value of the specified shutdown margin.

(2) Shutdown Margin The shutdown margin shoeld be determined at least annually and after changes in either the c. ore, in-core experiments, or control rods.

(3) Pulse Limits (Added by NRC)

The relationship between peak fuel temperature and inserted reactivity for pulses should be determined when changes in the core (see item 1 above) are made.

(4) Core Configuration (Added by NRC)

Limitations on core configurations are intended to ensure that reactor physics and thermal-hydraulic parameters specific to the core are within the limits analyzed in the SAR. Core configuration parameters specified in item (4) of Section 3.1 or in Section 5 of the technical specifica-i tions document must be met during reactor operations. Therefore, an acceptable surveillance specification is to verify compliance with all applicable specifications in those sections when any change occurs in the reactor core configuration.

49 i

SAR Standard Format and Content Chapter 14 - Draft August 10, 1993 (5) Reactivity Coefficients (Added by NRC)

Item 5 of Section 3.1 of the technical specifications document will limit reactivity coefficients, which are largely determined by reactor design and fuel type. Measuring and verifying reactivity coefficients can be a difficult task. An acceptable schedule for surveillance of reactivity coefficients is at initial reactor startup and when any change in the reactor core configuration or fuel type requires changes in the I

specifications of Section 5.

(6) Fuel Parameters (Added by NRC)

All TRIGA fuel should be inspected for damage and all TRIGA non-instrumented fuel measured for length and bend at the following frequencies.

For non-pulsing TRIGA reactors, the fuel should be inspected and measured on at least a 5-year cycle. Approximately 20 percent of the fuel can be inspected and measured annually.

If an annual inspection identifies damaged fuel, then the entire core should be i

inspected and measured.

For pulsing TRIGA reactors, tne fuel should be inspected and measured annually.

If the reactor is pulsed infrequently (less than 10 pulses annually), the annual inspection requirement may be relaxed if analyzed and justified in the SAR.

If the reactor is 3

pulsed to reactivity insertions over 4 dollars, additional inspec-l tion requirements based on the number of pulses may be necessary.

I Facilities in this situation should present and justify inspection frequency requirements determined by the fuel vendor.

Routine inspections of fuel used in AGN-201 and PULSTAR reactors have not i

been required by technical specification.

If the opportunity is pre-sented to conduct an inspection of fuel, such as core disassembly of an l

AGN-201 or disassembly of a PULSTAR fuel element to replace fuel pins, l

the licensee should consider taking advantage of this opportunity. This I

type of inspection need not be a technical specification requirement.

NRC may require fuel inspection as a license condition to increase burnup limits on such fuels. This would be determined on a case-by-case basis.

These inspections are qualitative, to detect evidence of excessive corrosion / erosion and mechanical wear or damage.

Inspections for reactors with plate fuels have not been required by technical specification except for higher power reactors that frequently refuel. However, for reactors that remove plate fuel from service because the fuel has reached its burnup limit, there should be a require-ment to inspect representative fuel elements (e.g., 1 in every 10) for excessive corrosion / erosion, mechanical wear or damage, or plate swel-ling. The surveillance procedures should follow guidance provided by the fuel supplier, if available.

In all cases, the specification should describe briefly how the inspection will be performed, i

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f 3

For reactors with technical specification limits or SAR analyses imposing i

i limits on uranium burnup or fission density, confirmatory estimates i

should be made at intervals during the life of the fuel, such as at 50, 60, and 70 percent of the fuel life or semiannually when the NRC/ DOE Form 742 is submitted. The SAR should justify the surveillance method and intervals that ensure that the limit is not exceeded.

1 i

For reactors with HEU fuel and subject to 10 CFR 73.6(b) exemption (self-l protection), confirmatory radiation measurements or analyses should be l

made at intervals justified in the SAR.

i 4

4.2 Reactor Control and Safety Systems I

(1) Reactivity Worth of Control Rods The integral and differential worths of all control and safety rods will be determined at initial fuel loading.

Integral and differential worths i

will be determined at least annually and after changes of the core or control rods, as noted in item 1 of Section 4.1 above.

i (2) Rod Withdrawal and Insertion Speeds ANSI /ANS 15.1 acceptable.

1

}

(3) Transient Rod and Associated Mechanism i

This system will be inspected, disassembled, cleaned, and if applicable, lubricated annually. The reactor should be pulsed at least annually and after changes to core or control rods, as indicated in item 1 of Sec-tion 4.1 above, with a well-documented reactivity insertion.

If the reactor is not routinely pulsed, this standard pulse (a reactivity addition whose results are well know) may be deferred for more than a j

year, but should be performed before resumption of normal pulsing.

(4) Scram Times of Control and Safety Rods A specific interval should be stated for the surveillance intervals given in item 4 of Section 4.2 of ANSI /ANS 15.1.

q (5) Scram and Measuring Channels Channel tests of all scram and power measuring channels required by technical specifications, including scram actions with safety rod release and interlocks, are required before each reactor startup following a 4

i shutdown of greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or each secured shutdown.

If the reat-tor operating schedule calls for no secured shutdowns, the channel tests should be performed at least quarterly. Many facilities perform these tests before each reactor startup and NRC recommends this practice.

51 1

.-_.m.,

s--,-

e.--.--,_,

--m SAR Standard Format and Content Chapter 14 - Draft August 10, 1993 (6) Operability Tests ANSI /ANS 15.1 acceptable.

(7) Thermal Power Calibration for Forced Convection Cooled Reactors Thermal power calibration should be performed at least annually, with a heat balance verification at least monthly.

(8) Thermal Power Calibration for Reactors not Cooled by Forced Convection Thermal power calibration should be performed at least annually. The basis should indicate the method to be used.

(9) Rod Inspection (Added by NRC)

The rod-drive and scram mechanisms of each control and safety rod should be inspected annually. The poison sections of control and safety rods should be inspected biennially for indications of deterioration or damage. This can be a visual inspection or an inspection that requires the rod to pass through a measuring device that detects swelling.

4.3 Coolant Systems Only a small fraction of the licensea non-power reactors will need to consider all of the following surveillance items. The SAR should discuss the applicability and identify the parameters that should be tested. The applicable parameters should be specified, and the functions should be explicitly stated in the specification.

(1) Starting Function of Emergency Shutdown and Sump Pumps ANSI /ANS 15.1 acceptable.

(2) Test of Energency Coolant Sources and Systems ANSI /ANS 15.1 acceptable.

(3) Inservice Inspections If any inservice inspections of cooling system components are identified and required in the SAR, they shall be performed according to manufac-turer's recommendations.

If the manufacturer's recommendation is not available, the frequency should be as established in the SAR from engineering judgement and similar component inservice inspection requirements and experience.

52

SAR Standard Format and Content Chapter 14 - Draft August 10, 1993 i

(4) Analysis of Coolants for Radioactivity Analyses for isotope identification of primary and, if applicable, secondary coolant, should be performed by sampling quarterly. Sampling weekly for gross analysis should be considered to establish trends to quickly identify fuel or heat exchanger failure.

i (5) Hydrogen Concentration in Off-Gas If applicable, this test should be performed at least annually and after maintenance or repair that could affect the system or measurement l

instrumentation.

(6) Conductivity and pH When the reactor is operating on a routine schedule, conductivity and pH should be measured at least weekly. This requirement could be met by a system that monitors conductivity and pH continuously while the reactor i

is operating.

If the reactor is not operated for long periods, the interval between conductivity and pH measurements may be increased to monthly if reasonable justification is provided in the SAR.

If fuel is stored in water in separate fuel storage from the reactor i

pool, the pH and conductivity of this water should be measured at regular intervals as determined and justified in the SAR.

l (7) Primary Coolant level (Added by NRC) l If the primary coolant level above the core is not continuously displayed during operation, the primary coolant level in the pool or tank should be verified daily if the reactor is operating or before reactor startup.

(8) Primary Coolant Sensors and Channels (Added by NRC)

Channel tests of sensor operability and channels not included elsewhere in the technical specifications document that are identified in the SAR should be performed quarterly and before startup after maintenance.

All channels should be calibrated annually and before startup after major i

modification or component replacement.

l 4.4 Containment or Confinement l

4.4.1 Containment few licensed non-power reactors are required to provide a containment l

system. For those required by the SAR, the surveillance intervals given in ANSI /ANS 15.1 are acceptable.

l 53 l

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SAR Standard Format and Content Chapter 14 - Draft August 10, 1993 i

4.4.2 Confinement I

Confinement is a system that provides a temporary holdup or controlled release of radioactive effluents to the environment. Most non-power reactors are equipped with confinements and should have a functional test of the overall system described in the SAR quarterly.

In addition, an efficiency test of the filters should be performed annually or in accordance with mane;acturer recommendations and acceptance criteria.

4.5 Ventilation Systems Ventilation systems at most licensed non-power reactors are an integral part of the containment or confinement system and surveillance activities may be interrelated. An operability check, including dampers and blowers, should be performed quarterly and following repair or maintenance to declare the system operable. A functional and efficiency test of filters should be performed annually or in accordance with manufacturer recommendations and acceptance criteria and following repair or maintenance to declare the system operable.

4.6 Emergency Electrical Power for all emergency electrical power systems, channel checks or other operabil-ity checks should be performed before reactor startup and after maintenance.

Maintenance should be performed according to the manufacturer's recommendations.

If the manufacturer's recommendation is not available, the frequency should be as determined in the SAR.

4.6.1 Diesels and Other Devices The shorter of the surveillance intervals given in ANSI /ANS 15.1 is acceptable.

l 4.6.2 Emergency Batteries l

l The shorter of the surveillance intervals given in ANSI /ANS 15.1 is acceptable.

4.7 Radiation Monitoring Systems and Effluents 4.7.1 Monitoring Systems l

A channel check should be performed daily before reactor startup. Where physically possible, a channel test using a radiation source should be performed at least monthly. The SAR should describe such capability.

All required radiation monitoring systems, including effluent monitors, should be calibrated at least annually according to the manufacturer's recommenda-tions.

Individual systems should have separate specifications.

l 54 1

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4.7.2 Effluents l

1

]

Quantities of radioactive effluents released to the environment v a LCOs.

It l

l the SAR indicates that it is not feasible to monitor such effluents from low l

power reactors in real time at the point of release, calc: dated releases may be substituted. The SAR should specify surveillance methods and intervals for j

confirming these releases or for verifying upper limits.

For gaseous airborne radioactive effluent, confirmation of annual upper limits by integrating dosimeters such as thermoluminescent dosimeters (TLDs) or film i

is acceptable.

For particulate airborne or waterborne radioactive effluent, confirmation of l

annual upper limits by surveillance of environmental factors given in j

Section 4.7.2 of ANSI /ANS 15.1 is acceptable.

1 4.8 Experiments If any experiment discussed in the SAR is designed to operate with emergency

]

systems or with connections to the reactor protective systems, a channel check should be specified both daily and before reactor startup when the particular experiment is being performed.

Surveillance activities for experiments that l

are included in the experiment protocol and the review and approval process

]

need not be included explicitly in the technical specifications.

4.9 Facility-Specific Surveillance There should be applicable surveillance specifications for any facility-specific LCOs in Section 3.9 of the technical specifications document not explicitly included in Section 4.

These surveillances should be to verify significant safety features from the SAR.

5 Design Features I

The SAR forms the basis for NRC to issue an operating license for a non-power 1

reactor. Essential information includes the type and enrichment of fuel, core i

and fuel configurations, fuel storage facilities, thermal power level, potential accident scenarios and mitigating features, environmental conditions at the site, and other factors. To ensure that the issued license remains valid, design features should not be changed without prior NRC review and approval. These major design features are provided in Section 5 of the 4

i technical specifications document, if they have not already been specified in Sections 2 or 3.

i The NRC accepts the guidance in this section of ANSI /ANS 15.1.

The applicant should provide concise but explicit information on all noted features.

i 4

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SAR Standard Format and Content Chapter 14 - Draft August 10, 1993 i

6 Administrative controls The specified information and controls on staffing and operations of the reactor facility will ensure that the management and staff of the facility are acceptably knowledgeable and aware of the technical requirements to operate a safe facility, to comply with regulations and the license conditions, and to practice a meaningful ALARA program, which will protect the health and safety of the public, the facility users, and the staff.

Not all owners and operators of non-power reactors will have the same management organization or office titles.

Regardless of the details of the management organization, or of the complexity of the facility, the administra-tive functions presented in this section of ANSI /ANS 15.1 must be established and specified. The NRC accepts the ANSI /ANS 15.1 position as modified and amplified in the following sections.

6.1 Organization 6.1.1 Structure The information requested by ANSI /ANS 15.1 should be clearly stated, including i

how and when the radiation safety staff communicates with the facility manager and level 1 management to resolve safety issues.

6.1.2 Responsibility Follow ANSI /ANS 15.1.

I 6.1.3 Staffing Applicants should use the terms reactor operator (RO) and senior reactor operator (SRO) instead of Class B and Class A, respectively (see Figure 1 in ANSI /ANS 15.1).

6.1.4 Selection and Training of Personnel Compliance with 10 CFR Part 55 is required of the licensee and licensed operators, unless NRC has issued an exemption. ANSI /ANS 15.4-1988, " Selection and Training of Personnel for Research Reactors," provides additional guidance for non-power reactors.

6.2 Review and Audit The committee established for the review function may be assigned approval authority by the facility manager or the facility manager may retain it. Sec-tion 6.2 of the technical specifications document should explicitly state who holds the approval authority and should specify the committee's authority and how it communicates and interacts with management levels 1 and 2.

There must be a qualified independent review committee.

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j 6.2.1 Composition and Qualifications l

One or more voting members of the committee should be from organizations other than that operating the reactor.

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6.2.2 Charter and Rules l

Follow ANSI /ANS 15.1.

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6.2.3 Review Function j

The fact that this section of ANSI /ANS 15.1 addresses the review function j

required by 10 CFR 50.59 should be explicitly stated in the specifications.

j 6.2.4 Audit Function l

In addition to the emergency plan, all required plans, such as physical security and operator requalification, should be specified for auditing.

The j

requirement to audit these plans may be part of the plan itself.

If that is the case, the requirement to audit does not need to be repeated in the technical specifications.

6.3 Radiation Safety 3

I The technical specifications should clearly state that 10 CFR Part 20 establishes requirements that the radiation safety program must achieve.

Additional guidance for radiation safety programs at non-power reactors may be found in ANSI /ANS 15.11, " Radiological Protection of Research Reactor Facilities."

l The authority of the safety staff to interdict or terminate safety-related 4

activities should be specifically stated. An explicit statement in the tech-nical specifications should state management's commitment to practice an effective ALARA program. This program should apply to facility staff, facility users, general public, and the environment.

6.4 Procedures j

Procedures in addition to those identified in ANSI /ANS 15.1 should be required at facilities to address operational situations recognized in the SAR.

For i

example, if byproduct material whose possession is authorized under the reactor license is used in facility laboratories that are part of the reactor license and/or transferred to other licensees, procedures for control and transfer of this byproduct material should be part of the set of minimum J

procedures required by the technical specifications. The specifications a

should be written to ensure a minimum necessary set of procedures, but i

allowing for future additions as necessary.

57 SAR Standard Format and Content Chapter 14 - Draft August 10, 1993 l

The minor modifications and temporary deviations allowed by ANSI /ANS 15.1 l

should not be spelled out in the technical specifications.

However, the methodology for establishing and changing procedures should be stated in the specifications.

l 6.5 Experiments Review and Approval In addition to guidance of ANSI /ANS 15.1, the review and approval of experi-ments should be consistent with the guidance provided in Section C.3 of Regu-latory Guide 2.2 and Regulatory Guide 2.4, " Review of Experiments for Research Reactors." The specifications should make clear that " established and l

approved procedures" means written procedures, properly reviewed and approved.

~

Changes to these procedures should follow Section 6.4 of ANSI /ANS 15.1.

6.6 Required Actions 6.6.1 Action to Be Taken in Case of Safety Limit Violation Follow ANSI /ANS 15.1.

6.6.2 Action To Be Taken in the Event of an Occurrence of the Type Identified in Sections 6.7.2(1)(b) and 6.7.2(1)(c) l l

The first sentence of this section of ANSI /ANS 15.1 says, " reactor conditions l

shall be returned to normal or the reactor shall be shut down." The i

specification should be written to provide that the licensee establish, in advance, specific criteria for the two alternate actions:

shutdown or return to normal.

For example, a return-to-normal event is a reactor scram resulting from a known cause, such as an electric transient.

6.7 Reports 6.7.1 Operating Reports The technical specifications should state explicitly that operating reports should be sent to the NRC Document Control Desk and a copy to the appropriate i

regional administrator annually.

l Section 6.7.1(4) of ANSI /ANS 15.1 refers to the reporting required by 10 CFR 50.59. The specification should make reference to the rule.

6.7.2 Special Reports The technical specifications should state that special written reports of events should be sent to the NRC Document Control Desk and a copy to the appropriate regional administrator and special telephone reports should be of events made to the NRC Operations Center and the regional staff.

6.8 Records l

Follow ANSI /ANS 15.1.

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August 10, 1993

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REFERENCES i

American National Standards Institute and American Nuclear Society, i'

ANSI /ANS 15.1-1990, "The Development'of Technical Specifications for Research Reactors," LaGrange, Illinois, j

--, ANSI /ANS 15.4-1988, " Selection and Training of Personnel for Research j

Reactors."

--, ANSI /ANS 15.11-1987, " Radiological Protection at Research Reactor Facilities."

P Beeston, J. M., R. R. Hobbins, G. W. Gibson, and W. C. Francis, " Development and Irradiation Performance of Uranium Aluminide Fuels in Test Reactors,"

Nuclear Technoloov,jli, pp. 136-149, June 1980.

Gibson, G. ~W., The Development of Powdered Uranium-Aluminide Compounds for Use as Nuclear Reactor Fuels, IN-ll33, Idaho Nuclear Corporation, Idaho Falls, Idaho, December 1967.

l Nazare, S., G. Ondracek, and F. Thummler, " Investigations on UAlx-Al l

Dispersion Fuels for High-Flux Reactors," Journal of Nuclear Materials 56, 4

pp. 251-259, 1975.

l Simnad, M.

T., et al., " Fuel Elements for Pulsed TRIGA Research Reactors,"

Nuclear Technoloov. 28, p. 31, 1976.

--, " Interpretation of Damage to the FLIP Fuel During Operation of the Nuclear Science Center Reactor at Texas A&M University," GA-A16613, December 1981.

Simnad, M. T., and G. B. West, "Postirradiation Examination and Evaluation of TRIGA LEU Fuel Irradiated in the Oak Ridge Research Reactor," GA-A18599, GA Technologies, Inc., San Diego, California, May 1986.

Stahl, D., Fuels for Research and Test Reactors. Status Review, ANL-85-5, Argonne National Laboratory, Argonne, Illinois, December 1982.

U.S. Nuclear Regulatory Commission, Regulatory Guide 2.2, " Development of Technical Specifications for Experiments in Research Reactors," November 1973.

--, Regulatory Guide 2.4, " Review of Experiments for Research Reactors,"

Revision 0-R, July 1976.

i

--, NUREG-0851, "Nomocrams for Evaluation of Doses From finite Noble Gas Clouds," January 1983.

--, NUREG-1282, Safety Evaluation Report on Hiah-Uranium Content. Low-Enriched Uranium-Zirconium Hydride Fuels for TRIGA Reactors, August 1987.

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--- NUREG-1313, Safety Evaluation Report Related to the Evaluation of low-Enriched Uranium Silicide-Aluminum Dispersion Fuel for Use in Nonoower Reactors, July 1988.

West, G. G., M. T. Simnad, and G. L. Copeland, " Final Results From TRIGA LEU Fuel Postirradiation Examination and Evaluation Following Long-Term Irradiation Testing in the ORR," presented at the International Meetino on Reduced Enrichment for Research and Test Reactors. Gatlinburo. Tennessee.

November 3-6. 1986.

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