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Purdue University - Response to Request for Additional Information Regarding the Reactor License Renewal Application, Responses to Letter Dated July 25, 2016
	ML16267A465
Person / Time
Site:	Purdue University
Issue date:	09/19/2016
From:	Townsend C; Purdue University Research Reactor
To:	Cindy Montgomery; Document Control Desk, NRC/NRR/DPR/PRLB
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ML16267A467	List: ML16267A465; ML16267A466;
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PURDUE UNIVERSITY SCHOOL OF NUCLEAR ENGINEERING Document Control Desk US Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, MD 20852 Attn: Ms. Cindy Montgomery, Research and Test Reactors September 19, 2016 5'0 -

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SUBJECT:

PURDUE UNIVERSITY - REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PURDUE UNIVERSITY REACTOR LlCENSE RENEWAL APPLlCATION, RESPONSES TO LETTER - Dated July 25, 2016

Dear Ms. Montgomery,

Enclosed please find the response to the request for additional information (RAI) regarding the Purdue University license Renewal. In addition to the specific RAI response document, changes to the Technical Specifications are enclosed. I am also re-enclosing the PUR-1 emergency plan for the purpose of making it a publicly available document.

Should you have any questions or require further information, please don't hesitate to call me at 765-494-5764, or email me at clive@purdue.edu.

I hereby certify under penalty of perjury with my signature below that the information contained in this submission is true and correct to the best of my knowledge.

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Clive Townsend PURI, Reactor Supervisor Purdue University Attachments:

Purdue University Responses to RAls Dated January 19, 2016 Technical Specifications for the Purdue University Reactor, Draft Amendment 13 PUR-1 Emergency Plan cc:

Leah Jamieson, Dean, Purdue University College of Engineering Jim Schweitzer, Purdue University Radiation Safety Officer, CORO Chair Klod Kokini, Interim Head, Purdue School of Nuclear Engineering School ofNuclear*Engineering Nuclear Engineering Building* 400 Central Drive* West Lafayette, IN 47907-2017 (765) 494-5739

Fax: (765) 494-9570

https://engineering.purdue.edu/NE

1. The regulation in Title 10 of Code of Federal Regulations (CFR) Section 50.36 (c)(4) requires the inclusion of design features of those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered elsewhere in the specifications.

The proposed TS 5.1, "Site Description," Specification c, defines the area under the jurisdiction of the reactor license Oicensed area), in part, as "The licensed areas include the reactor room... "

and "Both of these areas are restricted to authorized... " This wonling does not clearly indicate if these statements are licensing requirements. NUREG-1537 and ANSI/ ANS-15.1-2007 suggest that TSs consist of "shall," "should," or "may" statements where these terms are defined in your TS 1.40, "Shall, Should, or May." For example, "The licensed areas shall include the reactor room... " and "Both of these areas shall be restricted to authorized... " Modify the TS to clarify that it contains requirements, or justify why no change is necessary.

The ambiguity of this Technical Specification has been resolved to include the word "shall".

2. TSs are fundamental criteria necessary to demonstrate facility safety and are required by 10 CFR 50.36 for each license authorizing operation of a production or utilization facility of a type described in 10 CFR 50.21. TSs are derived from the analyses and evaluation included in the SAR and submitted pursuant to 10 CFR 50.34. Due to the importance to overall facility safety of a uniform intemretation by both the licensee and regulator of terms and phrases used in TSs, defuiitions shall be-ID.duded where necessary and consistently used to ensure that the TSs criteria necessary for compliance with regulatory requirements is uniformly understood by the licensee and regulator.

Guidance pertaining to the format and content of TSs previously found acceptable by the NRC staff is provided in NUREG-1537Part1, "Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors - Format and Content" and ANSI/ANS-15.1-2007, "The Development of Technical Specifications for Research Reactors." ANSI/ ANS-15.1-2007, Section 1.3, "Definitions," includes criteria on the uniform inteipretation of terms used in the TS.

The proposed TS 1.7, "Core Configuration," refers to "... fuel elements, reflector elements... "

&'oposed TS 5.3, "Reactor Core and Fuel," Specification a, -refers to fuel assemblies. Modify TS 1. 7, "Core Configuration," to ensure consistency with TS 5.3, Specification a, or justify why no change is necessary.

Technical Specification 1.6 has been updated to ensure consistency with TS 5.3 by specifying that assembly is a synonym for element and now reads, "The core configuration includes the number, type, or arrangement of fuel assemblies (elements), reflector elements, and regulating/control rods occupying the core grid."

3. The regulation in 10 CFR 50.9 requires that information provided to the Commission by a licensee shall be complete and accurate in all material respects. The proposed TS 1.10, "Excess Reactivity," uses non-facility specific tenriinology:

Excess reactivity is that amount of reactivity that would exist if all reactivity control 1 -

devices were moved to the maximum reactive condition from the point where the reactor is exactly critical (k.ir= 1) at reference core conditions or at a specified set of conditions. Modify TS 1.10, "Excess Reactivity," to be specific to the PUR-1 design, or justify why no change is necessary. For example:

Excess reactivity is that amount of reactivity that would exist if all control rods were fully withdrawn from the point where the reactor is exactly critical (keff = 1) at reference core conditions.

The definition of excess reactivity has been clarified to read, "Excess reactivity is that amount of reactivity that would exist if all control rods were moved to the maximum reactive condition from the point where the reactor is exactly critical (keir = 1) at reference core conditions or at a specified set of conditions."

4. As was fully stated in RAIN o. 2, above, the importance of TSs to overall facility safety requires a uniform intemretrtion by both the licensee and regulator of terms and phrases used in TSs.

To that end, -defuritions shall be illduded Where necessary and consistently used to ensure that the TSs criteria necess3iy for compliance with regulatory requirements is uniformly understood by the licensee and regulator.

The proposed TS 1.41, "Shutdown Margin," is not fully consistent with the guidance in ANSI/ANS-15.1-2007, Section 1.3, "Definitions." Revise the definition of shutdown margin to follow the guidance in ANSI/ANS-15.1-2007 applicable to your reactor design, or justify why no change is necessary.

Technical Specification 1.43 has been updated to read identically to the guidance m ANSI/ANS-15.1-2007, Section 1.3.

5. The regulation in 10 CFR 50.9 requires that information provided to the Commission by a licensee shall be complete and accurate in all material respects.

The proposed TSs _l.43 through 1.46, have typographical errors with each definition followed by a colon(:) instead of a dash(-). Revise TSs 1.43 through 1.46 formatted to be consistent with the other definitions in TS 1, "Definitions," or justify why no changes are necessary.

Typographical errors in Technical Specifications 1.43 through 1.46 have been corrected.

6. Due to the importance to overall facility safety of a uniform interoretrtion by both the licensee and regulator of terms and phrases used in TSs, definitions shall be included where necessary and consistentlf used to eruure that the TS criteria -necessary for compliance with regulatory requirements is-uniformly understood by the licensee and regulator.

The _proposed TS 1.43, "Reference Core Condition," defines the allowable reactivity worth of the xenon at reference core condition as negligible or less than 0.30 dollar. Proposed TS 3.1, "Reactivity limits," use reactivity worth values in terms of.Afc/k. Modify TS 1.43, "Reference Core Condition," defining the xenon reactivity worth value in terms of L\\k/k to be consistent with the reactivity units used in the rest of the TSs.

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Technical Specification 1.35 has been converted to~-

7. The regulation in 10 CFR 50.9 requires that information provided to the Commission by a licensee shall be complete and accurate in all material respects.

The proposed TS 1.44, "Rod, Control," uses non-facility specific terminology, "A control rod is a device fabricated from neutron-absorbing material or fuel~ or both, that is used...," which is not applicable to or correct for PUR-1. Revise TS 1.44 to be specific to the PUR-1 design, or justify why no change is necessary. For example: "A control rod is a.device fabricated from neutron-absorbing material that is used... "

The definition of a control rod has been updated to read, "A control rod is a device fabricated from neutron-absorbing material that is used to establish neutron flux changes and to compensate for routine reactivity losses. A control rod can be coupled to its drive unit allowing it to perform a safety function when the coupling is disengaged."

8. Due to the importance to overall facility safety of a uniform intemretation by both the licensee and regulator of terms and phrases used in TSs, definitions shall be included where necessary and consistently used to ensure that the TS criteria necessary for compliance with regulatmy requirements is uniformly understood by the licensee and regulator.

The proposed TS 3.1, "Reactivity Limits," Specification a, defines the shutdown margin in reference to "... to the cold xenon-free condition... " which is not consistent with proposed TS 1.43, "Reference Core Condition." Modify TS 3.1, Specification a to be consistent with TS 1.43, or justify why no change is necessary. For example: "The shutdown margin, relative to the reference core condition... "

Technical Specification 3.1 has been updated to be consistent with TS 1.43.

9. TSs are derived from the analyses and evaluation included in the SAR and submitted pursuant to 10 CFR 50.34. TSs will include items in the following categories: safety limits, limiting safety system settings, and limiting conditions for operation. TS limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the criteria provided in 10 CFR 50.36(c)(2)(ii).

The proposed TS 3.1, "Reactivity Limits," Specification d, establishes a limit on maximum reactivity during operation as "The reactor shall be shutdown if the maximum positive reactivity of the core... " This statement does not specifically refer to the available excess core reactivity.

Modify TS 3.1, Specification d to refer to excess reactivity, or justify why no change is necessary.

For example: "The reactor shall be shutdown if the maximum positive excess reactivity of the core... "

Technical Specification 3.1 now refers to the excess reactivity.

10. The regulation in 10 CFR 50.36 requires the inclusion of smveillance requirements that prescribe the frequency and scope of the smveillance necessary to demonstrate the required 3-

performance in Section 4 of the TS.

The proposed TS 4.1, "Reactivity Limits," Specification a, states that the shutdown margin and the reactivity worth of the control rods are determined biennially. The guidance in ANSI/ ANS-15.1-2007, Section 4.1, "Reactor Core Parameb;;~s," ikm (2) includes additional criteria to perform the smveillance following significant core configuration and/or control rod changes.

Revise TS 4.1, Specification a to include the additional surveillance inANSI/ANS-15.1-2007, or justify why no change is necessary. If the additional smveillance is added to the TS, state what constitutes a significant change. The NRC has acc_epted changes greater than 0.003 fl. k/k. as significant If proposing a less conservative value, provide a justification.

TS 4.1.a has been updated and is attached to this submission. This specification addition lays out the requirement of rod worth measurement following the replacement of an older control rod. With regards to the question of a significant core co11/iguraticm change, TS 4.1.a requires the worths to be measured "... whenever a core configuration is loaded for which shim-safety rod worths have not been me<l:Sured." A core configuration is defmed in TS 1.6 (as a.mended).

Therefore, any change will require a re-measurement of rod worths which is more conservative than the NRC accepted value of 0.003 tik/k.

11. The regulation in 10 CFR 50.36 (c)(4) requires the inclusion of design features of those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, wouM have a significant. effe,i on safety and are not covered elsewhere in the specifications.

The proposed TS 5.3, "Reactor Core and Fuel," Specification a, specifies the reactor fuel as Material Test Reactor (MTR) type in alunrinum_cladding with fuel meat enriched up to 20 percent in the U-235 isotope. The guidance in ANSI/ANS-15.1-2007, Section 5.3, "Reactor Core and Fuel," contains criteria on material specifications, wliich "includes the chemical form of the fuel and the cladding material as aluminum 6061. Revise TS 5.3, Specification a to include the material specifications for the aluminum cladding and fuel form, or justify why no change is necessary.

Technical Specification 5.3 has been updated to include the material specifications for the aluminum cladding and fuel form.

12. TSs are derived from the analyses and evaluation included in the SAR and submitted pursuant to 10 CFR 50.34. TSs will mclude items in the following categories: safety limits, limiting safety system settings, and lin:iiting conditions for operation. TSs limiting condition for operation of a nudear reactor iiuist be*estibliShed for "each item meeting one or more of the criteria provided in 10 CFR 50.36(c)(2)(ii). The regulation in 10 CFR 50.36 also requires the inclusion of smveillallce requii-emellis that prescnhe the frequency and scope or"the surveillance necessary to demonstrate the required performance in Section 4 of the TS.

The guidance in NUREG-1537, ~art 1, Chapter 14, Api>endix 14.1, "Format and Content of Technical Specifications for Non-Power Reactors," Section 3.1, "Reactor Core Parameters,"

item (6)(a) together with Appendix 14.1, Section 4.1, "Reactor Core Parameters," item (6) include guidance on limiting Conditions for Operations (LCOs) and smveillance requirements 4-

for certain fuel parameters including periodic inspection of the fuel.

The proposed TS 5.3, "Reactor Core and Fuel," contains specifications related to fuel inspection and swveillance:

Specification f - inspection requirements for the fuel to detect gross failure or visual deterioration Specification g - inspection criteria to detect gross failure or visual deterioration of the PUR-1 fuel plates Specification h - states the criteria to ensure that the PUR-1 fuel assemblies are not operated in a damaged condition These specifications do not appear to be design features, rather they appear to be LCOs. Move TS 5.3, Specifications f through h and/or develop similar additional LCO specifications in proposed TS Section 3,."Lllniting Conditions for. Operation," with associated swveillance requirements m:TS Section *4;* "Smveillimce ReqUiiements," based on the guidance included in NUREG-1537, Appendix 14.1, or justify why no changes are necessary..

The Technical Specifications have been edited and are attached to this RAI Submission.

13. TSs required by 10 CFR 50.36 provided limitations and operational criteria intent on protecting the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity. The first of these barriers and the most important is the fuel cladding.

The proposed TS does not have a specification for limiting bumup. NUREG-1537, Part 1, Appendix 14.1, Section 3.1, "Reactor Core Parameters," item (6)(c), "Materials Testing Reactor (MTR)-Type Fuel," provides guidance that the fuel matrix should have a limit on U-23.$ bumup. Provide a bumup ]imit. consistent with the guidance in NUREG-1537 and ANSI/ANS-15.1-2007, Section 5.3, "Reactor Core and Fuel," or justify why no change is necessary.

Bumup of up to 50 percent of the initial concentration of Uranium-235 has been accepted by the NRC as well as a fission density of 2.3 x 1021 fissions/cm3.

The PUR-1has190 plates in the core with 12.5 grams ofU-235 per plate giving 2.375 kg of U-235. With a 50% bumup, there would be 1.1875 kg of U-235 fissioned. This would be equivalent to 1 mol 6.022 x 1023 atoms 1187.5 grams* 235 1

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= 3.043 x 1024 fissions grams mo This number of fissions (at 200 Me V per fission) would yield 5-

200 MeV 4.450 x 10-20 kW hours 3.043 x 1024 fissions*

= 2.71x107 kWh fission 1 MeV If the PUR-1 were to operate at 12 kW plus 50 % uncertainty for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months per day, the reactor would need to run for almost two centuries to bum 50% of the Uranium-235 (excluding fission ofU-238).

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1 day 1 year 2 71 x 10 kWh*

= 172 years 18 kW 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 365 days Additionally, the fuel meat has dimensions of 5.96 X 0.0508 X 60.01 cm and a volume of 18.17 cm3

With 190 plates, the total fuel volume is 3452 cm3

Using the guidance from NUREG-1537, this would allow for fissions 2.3 x 1021 3

3452 cm3 = 7.940 x 1024 fissions cm which is less restrictive that the 50% Uranium bumup.

This time period far outlives the recommended time of operation by the fuel manufacturer and therefore, a limit on Bumup is not needed for this power level.

14. TSs required by 10 CFR 50.36 provide_d limitations and operational criteria intent on protecting the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity. The first of these barriers and the most important is the fuel cladding. The guidance in ANSI/ ANS-15.1-2007, Section 5.3, "Reactor Core and Fuel," includes criteria for reactor operation with damaged or leaking fuel.

Provide a specification for reactor operation for locating damaged or leaking fuel following the guidance provided in ANSI/ANS-15.1-2007, Section 5.3, or justify why no LCO is needed.

A Limiting Condition of Operation has already been provided which covers the operation of the PUR-1 with damaged or leaking fuel and was in Technical Specification 5.3. This has been moved to TS 3.6.b in response to RAI #12. TS 3.6.b prohibits use of fuel plates which are unsound for use. Additionally, TS 3.6.b has been clarified to indicate the conditions creating an unsound plate are outlined in TS 3.6.a. The revised TS 3.6.b is attached.

15. The regulation in 10 CFR 50.36(c)(4) requires the inclusion of design features of those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered elsewhere in the specifications.

The guidance in ANSI/ANS-15.1-2007, Section 5.3, "Reactor Core and Fuel," includes criteria for describing the normal core configuration including control rods. Propose a Section 5 design TS describing the PUR-1 core shim and regulating control rods following the 6-

guidance provided in ANSI/ANS-15.1-2007, Section 5.3, or justify why no changes are necessary.

Technical Specification 5.3.fhas been added to satisfy the guidance of ANSI/ANS-15.7-2007, Section 5.3 and is attached.

16. TS limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the criteria provided in 10 CFR 50.36(c)(2)(ii).

The guidance in ANSI/ANS-15.1-2007, Section 3, "Limiting Conditions for Operations,"

includes constraints that shall be adhered to during the operation of the facility. Proposed TS 3.2, "Reactor Safety System," refers to only a critical reactor and states, "The reactor shall not be made critical... " Explain why these conditions do not need to be met if the reactor is in operation but sub-critical, or revise TS 3.2 to refer to reactor operation.

Technical Specification 3.2 has been revised and is attached. Historic rod worth cUIVes have shown that movement of the shim-safeties in their first 10 cm of range inserts approximately 0.002 !::.k/k of positive reactivity which is significantly less than the shutdown margin. The rods should be permitted to be moved less than 10 cm without meeting the rest of the criteria in 3.2 to allow for maintenance and the prestart checklist during normal startup.

17. TS limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the criteria provided in 10 CFR 50.36(c)(2)(ii).

The proposed TS 3.3, "Primary Coolant Conditions," Specification d, specifies the limit for pool water temperature consistent with the thermal-hydraulic analysis in the PUR-1 SAR. It is not clear what steady-state operating power means as used in proposed TS 3.3. Proposed TS 1.26, "Power Level," states that steady state power shorild be 10 kW or less. Provide an explanation if this TS applies during reactor operation. If it does apply during reactor operation, revise the TS accordingly. If it does not, explain and justify when the temperature limit applies.

TS 3.3.c now reads, " The primary coolant (bulk pool volume) shall be maintained at or below 30 °C which the reactor is operating."

18. The regulation in 10 CFR 50.36 requires the inclusion of smveillance requirements that prescribe the frequency and scope of the swveillance necessary to demonstrate the required performance in Section 4 of the TSs. The guidance in ANSI/ANS-15.1-2007, Section. 4, "Swveillance Requirements," includes swveillance criteria_for specific systems specified in Section 3, "Limiting Conditions for Operations."

The proposed TS 4.3, "Primary Coolant System," Specification c states in part, "The reactor pool water will be at a height of the 13 feet over the top of the core whenever the reactor is operated." This appears to be a LCO~ not a sur\\Teillance requirement, and is not consistent with the guidance in ANSI/ANS-15.1-2007. In addition, a LCO already exists for water level in proposed TS 3.3, "Primary Coolant Conditions," Specification c. Modify TS 4.3, Specification c as appropriate, or justify the TS as proposed.

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Technical Specification 4.3.c has been removed and a new TS 4.3.c has replaced it TS 3.3.b is attached.

Additionally, in an effort to improve Technical Specification 3.3, the primary coolant pH requirement has been removed in accordance with the guidance in ANSl/ANS-15.1-2007, which states, "Conductivity or pH, or both: Weekly to quarterly." [Emphasis added]. This is also in conjunction with ML15114A433, Memorandum to Alexander Adams from Cindy Montgomery dated May 11, 2015. Technical Specification 4.3.a has similarly been removed.

19. The regulation in 10 CFR 50.36 requires the inclusion of smveillance requirements that prescribe the frequency and scope of the smveillance necessaiy to demonstrate the required performance in Section 4 of the TSs.

The proposed TS 4.3, "Primary Coolant System," Specification c specifies the weekly inspection of the reactor pool water level. The guidance inANSI/ANS-15.1-2007, Section 4, "Smveillance Requirements," includes criteria under what conditions a smveillance activity can be deferred.

M~dify TS 4, Specification c-specifying whether or not the water level inspection activity can be deferred during long term reactor shutdown to be consistent with the guidance in ANSI/ ANS-15.1-2007, or justify why no change is necessaiy. H the smveillance can be deferred, provide a justification for deferment Technical Specification 3.3.b has been added to read:

The primary coolant shall be maintained at least 13 feet above the core whenever the reactor is operating. The primary coolant shall be maintained at least 13 feet above the top of the core or at a level sufficient for the pool top radiation monitor to indicate less than 1 mRem/hour during non-operational periods.

This specification allows for the reactor water to drop below 13 feet in the event of an extended shutdown while still maintainnl.g dose levels which-are safe to the public directly above the core as specified i.J.1 10 CFR Part 20. The 1 mR/hour limit is half of the public dose rate limit and as the reactor room is a restricted area, dose to the public will be guaranteed to be less than 2 mR/hour.

20. TSs are fundam<:ntal criteria necessary to demonstrate facility safety and are required by 10 CFR 50.36 for each license authorizing operation of a pro-duction or utilization facility of a type descnbed-fu -10 CFR 50.21. The TSs are derived from the analyses and evaluation included in the SAR and submitted pursuant to 10 CFR 50.34. Due to the importance to overall facility safety of a unifo:nn.i:ri.teipretation by both the licensee and regulator of terms and phrases used in TSs, definitions shall be included where necessary and consistently used to ensure that the TSs criteria necessaiy for compliance with regulatory requirements is uniformly understood by the licensee and regulator.

The proposed TS 1.26, "Power Level," defines two different power levels; steady state at 10 kW, and maxim~ power level at 12 kW, which appear to be used in determining the setpoints for the scram and control system. The definition appears to say that average power is 8-

limited to 10 kW, but transients up to the power scram setpoint of 12 kW or less is allowed.

Based on the definition, it is not clear how the steady state power level is determined over time, since periodic deviations are allowed, and how the various setpoints for the scram and control logics are triggered. Explain how the 4-hour average power level is determined and how the average shows that steady-state power is 10 kW or less. Explain how the setpoints for scram and setback are calculated from the steady-state power level.

The proposed TS 1.40 defines "shall" as identifying a requirement NUREG-1537 and ANSI/ ANS-15.1-2007 suggest that TSs consist of "shall, should, or may" statements where these terms are defined in your TS 1.40, "Shall, Should, or May." Clarify what type of statements "is" statements and "will" statements are in TS 1.26, and modify the TS, as appropriate, to meet definition 1.40. TS 1.26, "Power Level," states, in part, that "the steady state operating power level should be 10 kW or less." Since "should" statements are not requirements, explain what the requirements for steady state power are, and modify the proposed TS, as appropriate.

Ambiguities in the definition of Power level have been resolved and the revised Technical Specification is below:

Power Level - There are three important and separately defined power levels.

a. Instantaneous Power Level shall be the power level of the reactor at any given moment, as indicated by the reactor instrumentation.

b. The Operating Power Level is the power level from which setpoints for scram and setback shall be calculated. The Operating Power Level shall be 10 kW or less.

c. The Maximum Power Level is the maximum instantaneous power level allowed by the PUR-1 License. The Maximum Power Level shall be 12 kW, which shall not be exceeded.

There has also been concern noted in the RAI regarding a formerly proposed time interval over which the steady state power level (now Operating Power Level) was to be determined. While this four hour interval has been omitted in the current proposedTechnical Specification 1.26, it is noted by-the responder tharall safety analyses have been done for 12 kW and this value (12kW) remains the limit on the analyzed power, plus 50% uncertainty due to instrumentation aiid other causes. Reactivity ~dditions wlllch immediately initiate a 1 second period have been done previously from this 12+50%-kW value and have been shown to have no effect on the ability of the reactor to operate and shutdown in a ~afe maimer. A similar reactivity addition starting from 10 kw does not chai1ge or make previous analyses less conservative.

21. Due to the importance to overall facility safety of a uniform intemretation by both the licensee and regulator of terms and phrases used in TSs, definitions shall be included where necessacy and consistently ~sed to ensure that the TSs criteria necessacy for compliance with regulatory requirements is uniformly understood by the licensee and regulator.

The proposed TS 1.33, "Reactor Secured," is not consistent with the guidance provided in ANSI/ANS-15.1-2007, Section 1.3, "Definitions," item "Reactor Secured," Section (2)(a), that states, in part, the reactor is secured with the "minimum number of neutron-absorbing control 9-

devices fully inserted." Revise TS 1.33, "Reactor Secured," to match the guidance in ANSI/ ANS-15.1-2007, as follows, or justify why no change is necessary. For example:

2. Orthe following conditions exist

1. All control rods inserted Technical Specification 1.32 has been updated to match the guidance provided with adjustments to provide for facility specific information. This is in the Technical Specifications as attached.

22. Due. to the importance to overall facility safety of a uniform interpretation by both the licensee and regulator of terms and phrases used in TS, definitions shall be included where necessary and consistently used to en8ilre thatthe TSs criteria necessary for compliance with regulat01y requirements is uniformly understood by the licensee and regulator.

The guidance in ANSI/ANS 15.1-2007~ Section 1.2.2, "Format.," includes criteria on the format of specifications stating _ that individual specifications contain "Applicab.ility,"

"Objective," -"Specification~" and ":Basis" sections. -The proposed TS 2.1, "Safety Limit,"

containS a description and justification of the safety limit on process variables before the applicability statement Move the descnption to the Basis section of proposed TS 2.1, "Safety Limit," to be consistent with the guidance in ANSI/ANS-15.1-2007, or justify why no change IS necessary.

Technical Specification 2.1 has been updated to provide clarity in the layout, format and ordering in accordance with the guidance in ANSI/ANS-15.1-2007.

23. Due to the importance to overall facility safety of a uniform interpretation by both the licensee and-regulator of terms an<fphcises-used in TSs,, definitions shall be included where necessary and consistently used to ensure that the TSs _criteria necessary for compliance with regulatory requirements is uniformly understood by the licensee and regulator.

Proposed TS 1.40 defines "shall" as identifying a requirement Revise proposed TS 5.2, "Reactor Coolant System," to be consistent With TS 1.40, or justify why no changes are necessary. Examples of parts of TS 5.2 not consistent with TS 1.40 are:

Specification a - "The PUR-1 primary cooling system~ a pool containing... " to "The PUR-1 primary cooling system shall be a pool containing... "

Specification b - "The process water system ~ assembled... " to "The process water system shall be assembled... "

"The demineralizer contains a removable cartridge... " to "The demineralizer shall contain a removable cartridge... "

Specification c - "Makeup water for the pool ~ taken bat.cp wise... " to "Makeup water for the pool shall be taken batch wise... "

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"A vacuum breaker excludes any possibility... " to "A vacuum breaker shall exclude any possibility... "

"The pool makeup water system, in addition to the demineralizer, includes a... " to "The pool makeup water system, in addition to the demineralizer, shall include a... "

Specification d - "The chiller ~ designed with three loops. Pool water passes through the primary loop, a Freon refrigerant ~ in the secondary loop, and water fro:rµ the building water supply is used to remove* heat, which is then dlschMged to the building-sewer system. The heat-removal capacity of the heat exchanger~ 10.5 kW" to "The d:iill.er shali contaill three loops. Pool water shall pass through the primary loop; a Freon refrigerant shall be in the secondary loop,

- and water from the building water supply shall be used to remove heat, which shall be discharged to the building sewer system. The heat-removal capacity of the heat exchanger shall be at least 10.5kW" Corrections have been made to Technical Specification 5.2 and can be seen in the revised Technical Specifications as attached.

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24. Due to the importance to overall facility safety of a uniform intemretation by both the licensee and regulator of terms and phrases used in TSs, definitions shall be included where necessary and consistently use £.xPiain why the smveillance intervals for pH and conductivity are sufficient to detect a Freon leak.

The installed heat exchanger at the PUR-1 is a three loop system. Pool water exchanges heat with the Freon in the second loop through a shell and tube sy~tem. The primary coolant flow rate through the heat exchanger is at least 5 gallons per minute during operation. The Freon is then piped to a second heat exchanger which transfers the heat to the city sewer system. The Freon in contact with the Primary Coolant is at a higher pressure when the cooling system is activated which eliminates the possibility of the primary coolant leaking into the Freon. The reactor is very seldom run at periods long eilough to heat the pool water to sufficient temperatures which would require the activation of the coolant system.
A potential Freon leak would be detected in two possible ways. Firstly, it is known that Freon is not soluble in water. The University of Arizona Decommissioning Plan, Revision 0 (ML091490074) notes, "a small leak occurring in the cooling coil, in 1997, as observed by a Freon film on the water surface." While the implementation of the cooling system at the University of Arizona is different, this indicates Freon leaks can be determined at facilities through simple visual inspection.
Finally, while the secondary coolant system is necessary for sustained operation at an elevated power level, it has been demonstrated time and again that the rate of bulk pool temperature increase is quite slow. If Freon were to leak into the pool and go undetected, the ability of the heat exchanger to maintain the pool water at appropriate temperatures would be compromised.
As the pool temperature continued to rise, the reactor would be shut down by the operator as it approached the 30 °C limit
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If the facility staff were to discover a leak, the response would be identical to that of all facility anomalies. The reactor would be shut down and the cause of the leak would be investigated.
26. TSs required by 10 CFR 50.36 provided limitations and operational criteria (Limiting Conditions for Operations and their corresponding Smveillance Requirements) intent on
. protecting the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity. The first of these barriers and the most important is the fuel cladding.
The guidance in ANSI/ANS-15.1-2007 includes criteria on Section 4, "Smveillance 12 -
Requirements," which are companion requirements to Section 3, "Limiting Conditions for Operations," specifications demonstrating the minimum performance requirements established in Section 3. The proposed TS 4.3, "Primazy Coolant System," Specification d requires the measurement of the radionuclide content of the pool water for radioactivity content There is no companion Section 3 LCO specification providing the minimum performance level for the radioactivity content of the coolant ensuring -public health and safety. Revise/add a specification of-maximum allowa.Qle radioactivity content of primazy water in proposed TS 3.3, "Primaiy Coolant Condition," to be consistent with the guidance inANSI/ANS-15.1-2007, or justify why no change is necessary.
Technical Specification 3.3.d has been added to limit the amount of radioactivity which may be present during operation to be that of the levels which can be found in 10 CFR Part 20, Appendix B, Table2.
27. The regulation in 10 CFR 50.9 requires that information provided to the Commission by a licensee shall be complete and accurate in all material respects.
The proposed TS 1.6 "Containment.," definition is not applicable to the PUR-1 facility and is not fully consistent with ANSI/ANS-15.1-2007 and NUREG-1537, Appendix 14.1, Section 1.3 that-provides guidance for including facilify specific definitions. Revise or remove TS 1.6 as applicable or justify why no change is necessary.
Former Technical Specification 1.6 has been removed from the PUR-1 Technical Specifications and subsequent items renumbered.
28. Due to the importance to overall facility safety of a uniform interpretation by both the licensee and regulator of terms and phrases used in TSs, definitions shall be included where necessary and consistently used to ensure that the TSs criteria necessary for compliance with regulatory requirements is uniformly understood by the licensee and regulator.
Proposed TS 1.40 defines "shall" as identifying a requirement Revise proposed TS 5.1, "Site Description," to be consistent With. TS 1.40, oijustifywhyno changes are necessary. Examples of parts of TS 5.1 not consistent with TS 1.40 are:
Specification d - "The reactor room remains locked at all times... " to "The reactor room shall remain locked at all times... "
Specification e - "The PUR-1 reactor room~ a closed room... " to "The PUR-1 reactor room shall be a closed room... "
Specification h - "OpeningS into-the reaetor room consist of the following... " to "Openings into the reactor room shall consist of the following... "
The issues with respect to Technical Specification 1.42 and TS 5.1 have been corrected.
and may be reviewed in the attached revision to the PUR-1 Teclmical Specifications.
29. Due to the importance to overall facility safety of a uniform interpretation by both the licensee 13 -
and regulatpr of tenns and phrases used in TSs, definitions shall be included where necessary and consistently used to ensure that the TSs criteria necessary for compliance with regulatory requirements is uniformly understood by the licensee and regulator.
Proposed TS 1.40 defines "shall" as identifying a requirement The proposed TS 3.4, "Confinement," Specification a, is not fully consistent with NUREG-1537, Part 2, Section 9.1, "Heating, Ventilation, and Air Conditioning Systems," acceptance criteria, which states system design should address all nonnal sources of airborne radioactive material. Confinement integrity must be established when radio_active material with the potential for airborne release is being handled. In addition, TS 3.4 "Confinement," Specification a should also be stated as a "shall" requirement to conform to TS definitions. Propose modifications to TS 3.4, as appropriate, or justify why no changes are necessary. For example:
Specification a -
"During reactor operation the following conditions will be met.. " to "During reactor operation and when radioactive material is being handled with potential for airborne release the following conditions shall be met.. "
Technical Specification 3.4.a has been corrected.
The following applies to RAis No. 30, 31and32, below. The regulation in 10 CFR 50.34(b) states, in part, that eac_h apt>li<'.c¢on fo~ an -opera.ting license shall include information that describes the facility, presents the design bases and the limits on its operation, and presents a safety analysis of the structures, systems, and components and of the facility as a whole, and shall include per 10 CFR 50.34(b)(2), a description and analysis of the structures, systems, and
)
components of the facility, with emphasis upon performance requirements, the bases, with technicaljustification therefore, upon which such requirements have been established, and the evaluations required to show that safety functions will be accomplished.
30. NUREG-1537, Part 1, Chapter 7, "Instrumentation and Control Systems," includes guidance on the detail and operating characteristics of the instrumentation & control (I&C) system. The increase in the maximum licensed power to 12 kWt may require changes and/or adjustment:S m the l&C system, which are not included or otheiwise specified in the PUR-1 SAR. As stated in the SAR, the power increase would require the adjustment of two instruments. Provide additional information on the adequacy of the current l&C design and required adjustment to the I&C system required due to the increase in the maximum licensed power level to 12 kWt The PUR-1 original design was-for powerlevels up to and exceeding 10 k Wt The BFa detector location will be adjusted so that actlial power matches or is conservatively close to the indicated po~er. The ~utpi.it ~~~nt fro~ -th~ BF a ~hamber flows through the input impedance of the Sigma Preamplifier and generates a voltage which is proportional to the flux level. This input impedance is made up chiefly of Resistor #16 (R16), the other resistances being small in comparison. To return to an operating power of 10 kWt (with scram setpoints at 12 kWt), will require adjustment of R16 to its original setting.
A detailed start-up plan will be compiled and presented to the Committee on Reactor Operations prior to changing to 10 kW (12 kW set point) operating power. This will include detector movements, console changes, and gold foil calibration procedures. The power level 14-
will be "stepped into" by changing the required settings, operating at 1 kW and doing a power calibration, followed by repeating this for 5 kW and 10 kW.
31. Chapter 7 of the PUR-1 SAR describes an operating mode of the regulating rod using a seIVo-amplifier. Provide additional information on the operation of the regulating rod in this mode.
Is there an interlock preventing the raising of more than one control rod when in servo-mode operation? If so, propose a TS limiting condition of operation for this interlock and related surveillance requirement, or justify why no change is necessary.
The linear servo channel consists of a BF3 ionization chamber and its power supply, a micro-microampere amplifier, a chart recorder whose output current feeds a proportional controller, with a position-adjusting type control unit, and the regulating rod drive unit. Because this channel is the most precise channel in measuring neutron flux level, it is used as the means to obtain m.llform integrated neutron flux and/or reproducible flux. The linear level channel is kept on scale at all times during reactor operation by means of a range switch. This channel will be on scale before startup and is capable of indication well beyond 12 kilowatts. A linear channel with range changes forms a-very precise measuring device. Unfortunately, the BF3 ionization chamber will also detect gamma radiation and consequently becomes less precise as flux is reduced after high power runs. This channel is also called simply the linear channel, or channel 3.
The BF3 ionization chamber provides a d-c current which is proportional to the neutron flux at the chamber. This current is amplified by the micro-microampere amplifier which provides a 0-10 millivolt input signal to the recording controller. The controller provides a signal to the servo control unit which, through a drive milt, drives the regulating rod the necessary amount in the correct direction to maintain the power level at the control point set on the controller.
Whenever the power level exceeds the control point setting the regulating rod is inserted into the reactor the amom1t necessary to decrease the reactivity and restore the power level to the control point setting. Conversely, whenever the power level drops below the control point setting, the regulating rod is withdrawn the amom1t necessary to increase the reactivity and restore the power level to the control point setting.
In practice, the operator sets the regulating rod to the middle of its range of motion (a.bout 30 cm). This places the rod in its most linear portion of the worth curve and the optimum position for level control. Power is increased manually. The range is changed to follow the increasing power level. Then when the desired power level is reached the reactor is leveled off manually.
The red control pointer is moved to align it directly above the black indicating pointer and the control milt may be placed on automatic operation.
The regulating rod may be controlled by the servo amplifier if the reactor operator pushes the servo-permit indicator. The servo system is turned off by pushing the servo indicator. There are interlocks in the system design which prohibit the use of the two shim-safeties as shown on LNP Drawing # 117-3012. Operation of the Raise/Lower Switch would de-energize the Servo Permit Holding Coil which would in tum de-energize the "Servo-On" condition. Thus, attempting to raise/lower any rod manually stops any Servo Control.
No surveillance is necessary as this is a design f ea.ture and is not variable.
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32. NUREG-1537, Part 1, Chapter 14, Appendix 14.1, Section 3.2, "Reactor Control and Safety Systems," item (4), "Scram Channels" and item (5), "Interlocks," include criteria on the detail and operating characteristics of th(! scram channels, scram signals, and interlocks providing guidance that all scram conditions, signals, and interlocks required for operation should be described in the proposed TSs. Revise proposed TS 3.2, Table I and Table II, to include all scram conditions and interlocks described in the PUR-1 SAR, Chapter 7, or justify why no change is necessilIY.
)
All scram conditions, signals, and interlocks, required for operation have already been described in Tables I & II of the Technical Specifications. An additional 15 second, rod withdrawal interlock has been added to Table I which was always present in the reactor design but not credited in the Technical Specifications.,t\\.dditionally, the 0% range setback on the linear channel is now listed in Table I of the Technical Specifications. These setpoints and functions are the ones credited and required for safe operation.
33. Due to the importance to overall facility safety of a uniform interoretation by both the licensee and regulator of terms and phrases used in TSs, definitions shall be included where necessilIY and consistently used to ensure that the TSs criteria necessaIY for compliance with regulatory requirements is uniformly understood by tpe licensee and regulator.
Proposed TS 1.40 defines "shall" as identifying a requirement Revise proposed TS 3.2, "Reactor Safety System," Specification a to be consistent with TS 1.40, or justify why no changes are necessaIY. For example:
Specification a - "The reactor safety channels and safety-related instrumentation are operable... " to "The reactor safety channels and safety-related instrumentation shall be operable containing... "
Technical Specification 3.2.a has been updated to clarify language for the regulator and provide uniform interpretation.
34. The regulation iii lQ_ GFR §0.34(b)(2), requires the_SAR to include a description and analysis of the structures, systems, and components of the facility, with emphasis upon performance requirements, the bases, with technical justification thereupon which such requirements have been established, and the evaluations required to show that safety functions will be accomplished.
Section 7.2.3, "Channel 3 - Linear Power," of the PUR~ 1 SAR, describes the design and operation of the _fl~ m~asuring channel that has an adjustable range instrument The PUR-1 SAR indicates that this channel has two setpoints that will initiate a reactor set back at either zero or 10096 of the selected range and also a 120% range setpoint of the selected range that will initiate a reactor trip. The proposed TS 3.2, "Reactor Safety System," Table I specifies two setpoints, 11096 and 120% :ra.Dge, which initiate a reactor power -setback and trip respectively.
Provide additional details to ensure that the SAR description and TS 3.2 Table I contains consistent information. Does "11096 range" and "12096 range" refer to the selected power range 16 -
of the linear channel? H so, clarify the wording of the setpoint of the linear channel, as appropriate, or justify why no change is necessacy.
The noting of a 100% range setback in the PUR-1 Safety Analysis Report is a typographical error and is correctly noted in the PUR-1 technical specifications as submitted with a value of 110%.
Additionally, to clarify the setback and trip initiations by Charmel 3, these setpoints are of the currently selected :range. The word "Selected" has been added to Table I of the Technical Specifications to clarify.
35. TS LCOs of a nuclear reactor must be established for each item meeting one or more of the criteria provided in 10 CFR 50.36(c)(2)(ii).
NUREG-1537, Part 1, Appendix 14.1, Section 3.2(4), "Scram Channels," includes criteria for
licensees and applicants on the detail and operating characteristics of the scram channels and associated setpoints to include in the SAR and TS.
a. The proposed TS 3.2, "Reactor Safety System," Table I, specifies setpoints for the safety_ channels, but does not indicate the trip signal based on the direction of change in the process variable. Revise TS 3.2, *Table I setpoint values indicating whether the trip is actuated when the signal is "less than...,,_ or "greater than... " the setpoints.
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b. The proposed TS 3.2, "Reactor Safety System," Table I, specifies setpoints based on the reactor power level as "11096 steady state power level" and "120% steady state power level." Consider adding the explicit power level to the setpoint definitions as "11096 steady state power level (11 kW)," and "12096 steady state power level (12 kW)," or justify why no changes are necessacy.
Issues raised in RA.I 35 have been resolved and are shown on Table I as shown below:
Minimum Number Channel Required Setpoint (c)(d)
Function 2 cps or greater 2 cps rod withdrawal interlock Log count rate l(o) 12 sec. or greater Setback and period 7 sec. or greater Slow Scram 15 sec. or greater Rod withdrawal interlock 12 sec. or greater Setback 7 sec. or grea,ter Slow Scram Log N and period 1 (b) 7 sec or greater Fast Scram 15 sec or greater Rod withdrawal interlock 17 -
12kW, 120%
Slow Scram Operating power level, or less 0% Selected Range, or Setback greater Linear 1
110% Selected Range Setback or less 120% Selected Range or less Slow Scram 11kW,110%
Setback Operating power level, or less Safety l(b) 12kW, 120%
Operating power level, Fa.st Scram or less Manual Scram (console) 1 Slow Scram (hallway) 1 Slow Scram (a) Not required after Log N-Period channel comes on scale.
(b) Required to be operable but not on scale at startup.
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(c) All percentage based setpoints shall be tripped when the measured value is greater than or equal to the specified value. Period and com1ts per second (cps)) setpoints are at values less than or equal to the specified value. Exception: Trip point for 0% shall happen as the value goes from the positive to negative value.
(d) Setbacks shall be set such that they will be initiated prior to a Scram
36. The regulation in 10 CFR 50.36 requires the inclusion of smveillance requirements that prescribe the frequency and scope of the smveillance necessacy to demonstrate the required performance in Section 4 of the TSs.
ANSI/ANS-15.1-2007, Section 4, "Suiveillance Requirements," provides guidance on the frequency of functional testing of the reactor safety system channels. Provide clarification for the following:
a. Proposed TS 4.2, "Reactor Safety System," Specification a states, in part, that "a channel test of the reactor safety system channels listed in Table ill shall be performed prior to each reactor staI1np following a shutdown in excess of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ." Discuss the basis for establishing 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months as the limit Does the 8-hour limit apply if the reactor was secured for being unstaffed and there were no reactor operators in the facility (e.g.,
overnight)? Revise the TS as necessacy based on your answers.
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Specification 4.2.a has been deleted and subsequent sections renumbered.
b. Proposed TS 4.2, "Reactor Safety System," Specification b states, in part, that "a channel check of each of the reactor safety system measuring channels in use or on scale... "
Provide clarification if the reactor safety measllring channels are equivalent to the reactor safety channels listed in TS 3.2, "Reactor Safety System," Table I.
Specification 4.2.b has been deleted and subsequent sections renumbered.
c. Proposed TS 4.2, "Reactor Safety System," Specification b states, in part, that"... shall be performed approximately every four hours when the reactor is in operation." Is a channel check performed if the reactor is operated for a period of time less than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ? Hnot.,
provide ajustification. How oft.en does the reactor operator take reactor logs and is this channel check part of the reactor logs? What is the definition of "approximately" as used in the proposed TS? :MOdify-the TS, as -approp:ri3.te, based on your answers, or justify why no change is necessary.
Specification 4.2.b has been deleted and subsequent sections renumbered.
d. Proposed TS 4.2, "Reactor Safety System," Specification c states, in part, that "a channel calibration of the reactor safety channels shall be performed at the following average intervals... " Explain what average intervals means and how that would impact the requirement of "annually, with no intenral to exceed 15 months." How is this different from stating "the following intelVclls?" Clarify what the reactor safety channels referenced in this TS consist of. Modify the TS, as appropriate, or justify why no change necessary.
Specification 4.2 has been clarified to read in part "A channel calibration of the reactor safety channels shall be performed as follows:"
Technical Specification 3.2 Table I is titled,. "Safety Channels Required for Operation." These are the measurement areas which must be operable and calibrated as s~t-forth in 4.2. Additionally, TS 4.2.a has been clarified to read in part, "A channel calibration of the reactor safety channels as described* in Table I shall be performed as follows:".
The following applies to RAis No. 37, 38 and 39, below. Due to the importance to overall facility safety of a uniform interpretation by both the licensee and regulator of terms and phrases used in TSs, definitions shall be included where necessary and consistently used to ensure that the TSs criteria necessary for compliance with regulatory requirements is uniformly understood by the licensee and regulator.
37. Proposed TS 1.40 defines "shall" as identifying a requirement. Revise proposed TS 5.l, "Site Description," to be consistent with TS 1.40, or justify why no changes are necessary. For -
example:
Specification g - "The ventilation system ~ designed... " to "The 19 -
ventilation system shall be designed... "
Technical Specification 5.1 has been updated and is attached to this submission as amended.
38. Proposed TS 1.40 defines "shall" as identifying a requirement. Revise proposed TS 3.4, "Confinement," to be consistent with TS 1.40, or justify why no changes are necessary. For example:
Specification a - "During reactor operation the following conditions will be met... " to "During reactor operation the following conditions shall be met... "
Specification a
1. "The reactor room will be maintained at... " to "The reactor room shall be maintained at... "
Specification c - "Dampers in the ventilation system inlet and outlet ducts are capable... "
to "Dampers in the ventilation system inlet and outlet ducts shall be capable... "
Specification d - "The air conditioner can be shut off... " to "The air conditioner shall be able to be shut off... "
Technical Specification 3.4 has been updated and is attached to this submission as amended.
39. Proposed TS 1.40 defines "shall" as identifying a requirement. Revise proposed TS 4.4, "Confinement," to be consistent with TS 1.40, or justify why no changes are necessary. For example:
Specification a - "The negative pressure of the reactor room will be recorded weekly... " to "The negative pressure of the reactor room shall be recorded weekly... "
Tedmical Specification 4.4 has been updated and is attached to this submission as amended.
40. The regulation in 10 CFR 50.36 requires the inclusion of surveillance requirements that prescribe the frequency and scope of the SwVeillance necessary to demonstrate the required perforIIJallce in Section 4 of the TS. Additionally, 10 CFR 50.36(c)(4) requires the inclusion of design feafures of those features of the facility such as materials of construction and geometric arrangements, which, if cllte~d or modified, would have a significant effect on safety and are not covered elsewhere in the specifications.
The proposed TS 4.4, "Confinement," Specification d appears to be a duplicate statement in part of TS 5.3, "Reactor Core and Fuel," Specification f. The guidance in NUREG-1537, Part 1, Appendix 14.1, Section 3.1 "Reactor Core Parameters," item (6)(a) together with Appendix 20-
14.1, Section 4.1, "Reactor Core Parameters," item 1 (6) include guidance on LCO and swveillance requirements for certain fuel parameters including periodic inspection of the fuel.
Revise both TS 4.4 and TS 5.3, Specificationfbased on the guidance in NUREG-1537, Part 1, Appendix 14.1, also considering RAI No.12, above, or justify why no change is necessary.
The former Technical Specification 5.3.f has removed to prevent a specification which has already been covered.
41. The regulation in 10 CFR 50.36(c)(4) requires the inclusion of design features of those features of the facility such as materials of construction and geometric arnuigements, which, if altered or modified, would have a s~cant effect on safety and are not covered elsewhere in the specifications.
The proposed TS 5_.4, "Fuel Storage," Specification a is not consistent with the guidance in ANSI/ANS-15.1-2007, Section 5.4, "Fissionable Material Stoi-age," which states that fuel including fuel devices and fueled experiments-are to be stored in controlled, configurations. TS 5.4 states, in part, that "... fuel a88emblies shall be stored... " which may not dearly define all sitrnltions~ for example, may exclude the storage of fuel in the form of individual plates. Revise TS 5.4 to be consistent With the guidance, or jmtify why no change is necessary. For example:
Specification a - "All reactor fuel assemblies shall be stored in a geometric array... " to "All
- reactor fuel and fueled devices shall be stored in a geometric array... "
Technical Specification 5.4 has been updated to cover both fuel and fueled devices.
42. The regulation in 10 CFR 50, Appendix E, Section F establishes training requirement for emergency response personnel, including reactor operators. Subsection iv, includes fire control teams.
The PUR-1 reactor room contains two fire extinguishers, which are inspected annually and maintained by the Purdue University Fire Department In the event of a fire, the PUR-1 staff must be able to operate the fire extinguishers. Provide more comprehensive information on the training of reactor staff in the use of the fire extinguishers.
Training is done_ at initial employment and annually thereafter. Purdue's safety program is designed to OSHA Standards and Requirements. This training is done for all reactor staff.
43. Due to their importance to overall facility safety, TSs require a uniform understanding by both the license~ and regulator of terms and phrases used in TSs. To that end, definitions shall be included where necessary and used consistently to ensure that fue TSs criteria necessary for compliance with regulatmy requirements is unlfomtly understood by the licensee and regulator.
The proposed TS 1.47, "Tried Experiment.," states, in part, "... of approximately the same nuclear characteristic," which is not clearly defined. Explain what nuclear characteristics are considered in determining if an experiment is a tried experiment Revise TS 1.47 to provide a clearer definition of which nuclear parameters and their values used to determine whether an 21-
experiment is considered a "tried experiment" or justify why no changes are necessary.
Technical Specification 1.45 has been revised to read in part, "... An experiment of approximately the same nuclear characteristics as an experiment previously tried. These nuclear characteristics include but are not limited to neutron activation cross-sections, absorption cross-sections, and moderating ability."
44. The regulation in 10 CFR 50.9 requires that information provided to the Commission by a licensee shall be complete and accurate in all material respects.
The proposed TS 1.50, "Unsecured Experiment," appears to have a typographical error referring to an incorrect definition "... secured as defined in part 1.37 of this section... "
Should this be part 1.38 of the definitions? Modify the proposed TS as appropriate.
The numbering scheme has been revised. Please see the Technical Specifications as amended.
The following applies to RAis No. 45, 46, 47, and 48, below. TSs LCOs of a nuclear reactor must be established for each item meeting one or more of the criteria provided in 10 CFR 50.36(c)(2)(ii). The LCOs for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility.
45. The proposed TS 3.1, "Reactivity Limits," Specification d limits the total maximum positive reactivity of the core and any installed experiment to 0.006.Afq'k. The irradiation facility in the reflector contains six tubes that are filled with graphite when not in use. ff the six tubes are transitioned from containing experiments to graphite the reactivity change is 0.01 L\\k/k, which is larger t!ian the TS 3. _1 reactivity limit, De~cribe_ the type of irradiation tube facility activities that potentially result in reactivity changes exceeding the TS 3. Llimit for reactivity and how these activities are controlled. Discuss the need for controlling the use of the tubes by TS requirement The definition of "Core Configuration" has been updated to read, "The core configuration indudes the 'number, -type-, --or-arrangement of fuel assemblies (elements)' reflector elements, reflector element configuratioi1, and regulating/control rods occupying the core grid." Therefore, the changing-of the-installation, or lack thereof, regarding the graphite plugs, will alter the core configuration which would thel1 require a re-measurement of the rod worths and other associated Technical Specifications such as Shutdown Margin.
46. The guidance in ANSI/ANS-15.1-2007, Section 3.8.1, "Reactivity Limits," includes criteria for reactivity limits for experiments and states that the limits should be established using the absolute reactivity worth values. Revise proposed TS 3.1, "Reactivity Limits," to be consistent with the guidance or justify why no changes are needed. For example:
Specification e - "The reactivity worth of each experiment shall be ~ted as follows... " to "The absolute value of the reactivity worth of each eX}leriment shall be limited as follows... "
22-
Specification f - "The total worth of all movable and unsecured experiments... " to "The sum of the absolute value of the total worth of all movable and unsecured experiments... "
Specification g - "The total worth of all secured experiments... " to "The sum of the absolute value of the total worth of all secured experiments... "
The Technical Specifications 3.le-g have been revised and are attached as amended.
47. The proposed TS 3.5, "Limitations on Experiments," Specification c stat.es that no explosive material may be placed in the reactor pool. The PUR-1 SAR and TS 3.5 "Limitations on Experiments," Basis, do not indicate whether explosive material can be placed and/or stored in the PUR-1 facility. Provide information that describes the potential placement, movement, storage, and /or limitation on the maximum quantity of any explosive material and its administrative control.
There is no licensed based limitation of explosive material placed and/or stored in the PUR-1 facility. While all material that enters and exits the facility is subject to search, administrative controls limit these materials from being taken into locations such as the reactor confinement which may compromise reactor safety.
48. The proposed TS 3.5 "Limitations on Experiments," Specification d stat.es that experiments will be cooled, to prevent the experiment surface temperature exceeding 100 degrees Celsius.
The safe temperature for an experiment is determined only through the experiment-specific safety evaluation. Provide information why TS 3.5,-Specification d would ensure the safety of specific experiments. Propose and justify changes to TS 3.5, Specification d as needed.
Formerly, TS 3.5.d was an additional safety specification placed on experin!_ents in the reactor pool. This Technical Specification has been removed and subsequent sections renumbered.
The Technical Specifications in 3.5 ensure safety.
49. Due to the importance to overall facility safety of a unifomi intert>retation bv both the licensee and regulator of terms and phrases used in TSs, definitions shall be included where necessacy and consistently used to ensure that the TSs criteria necessacy for compliance with regulatory,
requirements is uniformly understood by the licensee and regulator.
Proposed TS 1.40 defines "shall" as identifying a requirement Revise proposed TS 3.5, "Limitations on Experiments," to be consistent with TS 1.40, or justify why no changes are necessacy. For example:
"The reactor will not be operated unless... " to "The reactor shall not be operated unless... "
Specification b - "All experiments and experimental procedures must receive approval by... " to 23-
"All experiments and experimental procedures shall be approved by... "
Specifications f and g - Contain "should" statements which are defined by TS 1.40 as a recommendation. Explain why these are "should~ statements as opposed to "shall" statements which indicate a requirement Technical Specification 3.5.b, 3.5.f, and 3.5.g are all corrected and included in the revised Technical Specifications as attached.
The following applies to RAis No. 50 and 51, below. The regulations in 10 CFR Part 20 require that dose to members of the public be limited. To support meeting the public dose limits answer the following.
50. The proposed TS 3.5, "Limitations on Experiments," Specification f states that the radioactive exposure dose for a member of the_ public occupying unrestricted areas for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months will be limited below the dose limits in Title 10 of the Code of Federal Regulations (10 CFR) Part 20, "Standards for Protection AgainSt Radiation," after a release. Provide a justification for the 2-hour time period, since the dose limits in 10 CFR 20.1301 are applicable for the full duration of any potential radioactivity release.
Technical Specification 3.5.f has been removed and replaced with Technical Specification 3.5.e which reads, "Any failure of an experiment shall not have a consequence that could exceed dose limits,as set forth in 10 CFR Part 20, as analyzed and approved by the Reactor Supervisor and the Committee on Reactor Operations."
51. The proposed TS 3.5, "Limitations on Experiments," Specification g states the dose limits for a member of the public and facility staff after an. experiment failure resulting in the release of radioactive products. Since these dose limits are in full compliance of the 10 CFR Part 20 limits, consider simplifying TS 3.5, Specification g stating that the dose limits will satisfy the requirements in 10 CFR Part 20.
Technical Specification 3.5.g has been removed and replaced with Technical Specification 3.5.e which reads, "Any failure of an experiment shall not have a consequence that could exceed dose limits as set forth in 10 CFR Part 20, as analyzed and approved by the Reactor Supervisor and the Committee on Reactor Operations."
52. The regulation in 10 CFR 50.36(c)(5) requires administrative controls necessary to assure operation of the facility in a safe manner.
The proposed TS 4.5, "Experiments," Specification b does not clearly define the review and approval requirements for new experiments. Revise TS 4.5 to provide a clear presentation of those requirements, or justify why no changes are necessary.
Also, due to the importance to overall facility safety of a unifonn interpretation by both the licensee and regulator of terms and phrases used in TSs, definitions shall be included where necessary and consistently used to ensure that the TSs criteria necessary for compliance with 24-
regulatmy requirements is uniformly understood by the licensee and regulator.
\\
Proposed TS 1.40 defines "shall" as identifying a requirement Revise TS 4.5, "Experiments,"
to be consistent with TS 1.40, or justify why no changes are necessaiy. For example:
"No experiments will be performed unless... " to "No experiments shall be performed unless... "
Specification b - "The experiment has been properly reviewed... " to "It is a new experiment that has been properly reviewed... ".
Please reference Section 6 for approval by the Committee on Reactor Operations. Please reference the attached Technical Specifications for the revised Section 4.5
53. The regulation in 10 CFR 50.36(c)(5) requires administrative controls necessaiy to assure operation of the facility in a safe manner.
The proposed TS 3.5, "Limits on Experiments," Specification b governs the approval process for _experiments. Proposed TS 6.2, "Review and Audit.," Specification d states that significant changes to previously approved eXJ>erlments are reviewed throiigh the 10 CFR 50.59 process
("Changes, Tests and Experiments"). Provide a clarification whether minor changes to experiments that do not significantly alter the experiment are also entered to the 10 CFR 50.59 process and describe the approval and review structure for such changes.
Changes that do not alter the experiment or the intent of the procedure are not entered into the 10 CFR 50.59 process but rather are screened to see if they fall under the scope of 10 CFR 50.59. Evaluations shall be performed if the experimental change falls under 10 CFR 50.59.
Eiperimental deviations which are permitted are outlined in the experiment itself and those deviations may only be made within the limits of the written approved framework.
54. The regulations in 10 CFR 20.1101 (d) or 10 CFR 20.1301 establish regulatory requirements for a radiation protection program and dose limits for individual members of the public both of which contain criteria addressing, in part issues related to efiluents.
NUREG-1537, Part 1, states that the format and content of the TS follow that of ANSI/ ANS-15.1-2007. ANSI/ANS-15.1-2007, Section 4.7.2, "Effluents," provides guidance for smveillance requirements for monitoring the facility boundaiy with dosimeters and environmental monitoring, specifically "s~pling of soil, vegetation, or water in the vicinity of the facility." Discuss whether the proposed TSs are consistent with the guidance.
Technical Specification 4.7 "Effiuents" has been added to the PUR-1 TSs and prescribes the surveillance to be conducted to ensure the dose to the members of the public remains below those set forth in 10 CFR 20.llO(d) and 10 CFR 20:1301. The additional TS is attached as revised.
The pathways to the environment have been evaluated and we have concluded that there is no credible pathway to the environment under normal conditions of operation. The reactor 25 -
cooling water is sampled on a regular basis and no activity related to fuel leakage or activation is found. There is also no leakage of the reactor pool. Particulates that may be produced in the reactor room would be measured by the continuous air monitor but no activity related to reactor operations has been identified. Additionally, airborne emissions are filtered by a HEPA filter which would prevent the release of particulate emissions. The only remaining pathway to the environment would be the release of gasses. Previous analyses have discussed the production of Ar-41 which was found to be 2.S6 X 10-4 [mRem/hr] at its absolute maximum. Sampling of the any environmental media would not identify these releases as they will disperse in the atmosphere. The dosimetry as specified in Section 4. 7 measures the dose released to the environment
55. TSs are fundamental criteria necessacy to demonstrate facility safety and are required by 10 CFR 50.36 for each license authorizing operation of a production or utilization facility of a type described in 10 CFR 50.21. The TSs are derived from the analyses and evaluation included in the SAR and submitted pursuant to 10 CFR 50.34. TSs LCOs of a nuclear reactor must be established for each item meeting one or more of the criteria provided in 10 CFR 50.36(c)(2) (ii).
LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility.
For the proposed TS 3.2, "Reactor Safety System," Table II, "Safety-Related Channels (Area Radiation Monitors)," address the following items:
a. The Basis does not indicate how the alarm setpoints are established, in order to protect the workers or public. NUREG-1537, Part 1, Appendix 14.1, Section 3.7.1, "Monitoring Systems," item (3) "Area Monitors," provides guidance that the alarm and automatic setpoints should be specified to ensure that personnel exposures and potential doses remain below the limits of 10 CFR Part 20. Provide the basis for the area monitor setpoints, or justify why no change is necessacy.
If the pool top is approximated as a point source, the dose level as a function of distance from the pool top is s
s =--
r 4nr2 where S is the original source strength at approximately 1 foot from the reactor pool, Sr is the reduced dose at the distance r. The distance at which a source strength of SO mR /hr would be reduced to 2 mR /hr is so 2=--
4nr2
/sO
~r= ~~=1.4feet This distance is then added to the initial detector distance to give a total of approximately 2.4 feet for source reduction. There are no unrestricted areas which are less than 26-
2.4 feet in the PUR-1 Reactor room. Similarly for the 7.5 mR/hr limits of the other two radiation area monitors, a reduction of 7.5 mR/hr to 2 mR/hr would require a distance of 175
~ r = ~~
= 0.55 feet There is no unrestricted area that is less than 1.55 feet. Therefore, these limits are adequate to assure there is no area which exceeds the dose limits to the public as set forth in 10 CFR Part 20.
In addition, the CAM is set according to the following procedure in order to alert reactor persom~el of the release of airborne radioactivity. The intention was to establish the setpoint as the value that would have the CAM indicate 10% of the DAC for the radionuclide of interest within 10 minutes of its release. Criterion for the radionuclide of interest was two-fold: (1) Probable neutron activation product; (2) Most restrictive DAC.
Fission products were not considered because of the small probability of fuel failure due to the limitations on reactor operating power as mandated by the technical specifications.
The radionuclide of interest was determined to be 124Sb. As a worst case scenario, the sample containing this radionuclide had a powdery form that facilitated resuspension.
This approach is both conservative and consistent with ALARA. The following methodology assumes instantaneous release and mixing.
Assuming a constant air concentration of the radionuclide of interest, the rate of change of activity on the filter is given by dA
- = CQE-11.A dt (1)
Where C is the concentration of the radionuclide (!)min), Q is the CAM flow rate
(!)min), E is the filter collection efficiency, and A is the decay constant (m.lli1). Using appropriate boundary conditions this differential equation can be solved. The activity on the filter at any time is given by The parameters for the PUR-1 CAM and radioimclide of interest are as follows:
C 0.370 Bq/L Q 84.95 l)m.iJ.1 E
90%
A.
7.996 x 10.. m.lli' 27 -
(2)
t 10 min The calculated activity on the filter is 282.9 Bq or 16,972 cpm. Since the CAM has a detection efficiency of approximately 28%, the expected count rate would be approximately 4 7 52 cpm. It is important to note that this is the activity on the filter above the radon background which is currently measured to be approximately 8000 cpm. The high alarm setpoint'should be set to about 12,700 cpm.
b. It is not clear whether the "Continuous air sampler" refers to the Geiger-Mueller (GM) detector, the monitor, or the whole channel. Provide clarification, whether it refers to a Continuous Air Momtor (CAM) channel and describe the monitor and/or alarm indicator and location. Discuss how footnote "(c)" would be used to replace a continuous air sampler.
The continuous air monitor refers to the whole channel.
The NMC AM-2B CAM is an air particulate monitor designed for optimum detection of gross beta radioactivity. The CAM consists of a structural steel mobile cart supporting the pump assembly, lead shield sampling chamber (with a particulate collector and beta scintillation detector), electrical power compartment, microprocessor based electronics assembly, front panel controls and displays, with top-mounted visual and side-mounted audible, alarms. (Taken directly from the April 2000 NMC CAM manual). Alarm indication is both audible and visual. The traditional location of the CAM is nearest the main entry door but there is not specification of the required location within the room.
Footnote (c) would replace the CAM by providing similar functionality by monitoring the beta activity. Modem hand held dosimetry give live updates on activity levels and allow for setpoints to be manually programed into the device. Additionally, PUR-1 has multiple CAMs which provide identical functionality.
c. The proposed TS 3.2, Table II, footnote (c) allows the replacement of a radiation monitor for a one week duration (or reactor run). Provide the basis for the one week replacement period.
A one week replacement period has been chosen as the maximum time during which an auxiliary CAM channel may be utilized as this is the value referenced in ANS/ ANSI 15.1-2007.
d. The setpoint for the Pool Top Monitor contains a typographical error "... or2x full power... " Modify the TS as appropriate.
This typographical error has been corrected.
56. The regulation in 10 CFR 50.36(c) states TSs will include Surveillances. The regulation in 10 CFR 50.36(c)(3) states, "surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessacy quality of systems and components is maintained, that facility operation will be within safety limits, and that-the limiting conditions for operation will 28 -
be met" For the proposed TS 4.2, "Reactor Safety System," Specification d, address the following items:
a. NUREG-1537, Part 1, states that the format and content of the TS follow that of ANSI/ANS 15.1. ANSI/ANS-15.1-2007, Section 4.7.2, "Eftluents," provides guidance related to operability testing requirements for radiation monitoring systems. The proposed TS 4.2, "Reactor Safety System," Specification d states, "The operation of the radiation monitoring equipment shall be verified... " It is nof clear wliether "verified" refers to a channel check or test or some other action. Discuss whether TS 4.2 is consistent with the standard guidance and modify the TS as appropriate.
Technical Specification 4.2 has been clarified that this verification didrefer to a "Channel Check". The revised specification can be found in the Technical Specifications, Amendment #13 as attached.
b. NUREG-15iJ7, Part 1, states that the format and content of the TS follow that of ANSI/ANS 15.1. ANSI/ANS-15.1-2007 Section 4.7.2, "Eftluents," provides guidance related to calibration requirements for radiation monitoring systems.
The proposed TS 4.2, "Reactor Safety System," Specification d states that the "Calibration of these monitors shall be performed annually... " Provide additional information Whether "radiation monitoring equipment" includes detecting and measurillg 'i.IlStrumentS used -for performing routine radiation surveys in the reactor room.
A correction of the PUR-1 TS 4.2 has been made and can be found in the Technical Specifications, Amendment #13 as attached.
57. The regulation in 10 CFR Part 20, Subpart K - Waste Disposal provides the regulatory requirements related to the disposal of licensed materials.
NUREG-1537, Part 1, Section 11.2.3, "Release of Radioactive Waste," includes criteria on the release and disposal of radioactive waste. The PUR-1 SAR states that all low-level radioactive waste is transferred for disposal to Purdue University under the By-Product License (Broad.scope) and stored until final disposition is determined. Provide a description of the administrative controls for transfer of material to storage. Provide a brief description of the waste storage facility.
Very little radioactive waste _is generated from operations under the R-87 license. This material is transferred to the Purdue Broad Scope License (13-02812-04) and subsequently handled by procedures approved by the Radiation Safety Committee. Specifically, waste is picked up by trained technicians in tl1e Radiological and Environmental Management group to the licensed storage facilities. These facilities are rooms within research buildings or a storage buildings constructed of corrugated metal with a concrete floor which is secured at all times. The material is then decayed in storage, shipped to a licensed disposal site or incinerated to reduce 29 -
volume with the residue shipped to a licensed disposal site. This information has been submitted as a part of the renewal (2015) of the broad scope license.
58. Due to the importance to overall facility safety of a uniform intemret:ation by both the licensee and regulator of terms and phrases used in TSs, definitions shall be included where necessaiy and consistently used to ensure that the TSs criteria necessary for compliance with regulatocy requirement.s is uniformly understood by the licensee and regulator.
ANSI/ANS-15.1-2007, Section 1.3, "Definitions," includes criteria on the uniform interpretation of the terms used in the TS. For the proposed TS Section 1, "Definitions,"
address the following items:
a. TS 1.16, "license," contains the phrase "... by the responsible authority... " Modify the phrase to reflect that PUR-1 is licensed by the U.S. NRC.
TS 1.15 has been modified as recommended.
b. Section 1, "Definitions," contain items which do not appear to be used within the TSs.
Review all definitions and eliminate those which are not used, or justify why they are needed (for example, TS 1.37, "Responsible Authority.")
The definition of Responsible Authority has been removed.
c. TS 1.35, "Readily Available on Call," revise... "senior operator shall insure that he is within... " to "senior operator shall be within... " to provide TSs that are gender neutral.
The definition of "Readily Available on Call" has been revised.
59. The regulation in 10 CFR 50.36(c)(5) requires administrative controls necessary to assure operation of the facility in a safe manner. Due to the importance to overall facility safety of a uniform intemret:ation by both the licensee and regulator of terms and phrases used in TSs, de~tions shall be included where necessary and consistently used to ensure that the TSs criteria necessary fot compliance with regulatocy requirement.s is uniformly understood by the licensee and regulator. NUREG-1537, Part'!, states that the fonnat and content of the TS follow that of ANSI/ANS-15.1 -2007. For the proposed TS Section 6.1, "Organization," address the following items:
a. The proposed TS 6.1, "Organiz.a.tion," does not follow the guidance in ANSI/ANS-15.1-2007, Section 6.1, "Organization," that provides guidance to include an overall description of the organizational functions, structure, staffing, resporuiibility, and selection and training of personnel. Revise TS 6.1 to follow the guidance inANSI/ANS-15.1-2007, or justify why no change is necessary.
Technical Specification Section 6 has been revised. Please see the attached revision.
b. The proposed TS 6.1, "Organiz.a.tion," has a numbering scheme that is not clear (letter followed by letter). Revise the numbering list.s in TS Section 6.1, Specification a, 30-
"Structure," and TS 6.1, Specification c, "Minimum Qualification of Reactor Personnel,"
, or justify why no change is necessary.
Technical Specification Section 6 has been revised. Please see the attached revision.
c. ANSI/ANS-15.1-2007, Section 6.1.1, "Structure," provides guidance related to organizational structure. Proposed TS 6.1, "Organization," Figure 6.1, does not clearly indicate the reporting and communication lines between the organizational structures.
Clarify the reporting structure for the PUR-1 organization, or justify why no change is necessary.
Technical Specification Section 6 has been revised. Please see the attached revision.
d. The proposed TS 6.1, Specification b, "Staffing," item (l)(c) specifies the conditions when the senior operator "... shall be present or readily available on call... " The specification further provides a definition of "R~y Available on Call" that is also specified in proposed TS 1.35, "Readily Available on Call." However, the two definitions are not consistent In order to reduce unnecessary redundancy and ensure consistency, modify TS 6.1, Specification b, item (l)(c) together with TS 1.35, as appropriate, or justify why no change is necessary.
Technical Specification Section 6 has been revised. Please see the attached revision.
e.
ANSI/ANS~15.1-2007; Section 6.1.3, "Staffing," provides guidance for minimum staffing, when the reactor is not secured. The proposed TS, 6.1, Specification b, "Staffing," item (2) states that no senior reactor operator (SRO) or licensed reactor operator (RO) is required in the facility, when the reactor is in secured position. However, the proposed TS definition of reactor secure does not address items such as status and alignment of systems to leave the facility unstaffed and security requirements. What are the additional steps beyond meeting the definition of reactor secure that must be met for leaving the reactor facility unattended?
Some licensees have captured these requirements in a definition titled "secure shutdown."
Modify the TS as appropriate.
Technical Sp~cification Section 6 has been revised. Please see the attached revision.
f.
ANSI/ANS-15.1-2007, Section 6.1.3, "Staffing," item (2) provides guidance for maintaining a list of reactor facility personnel. Evaluate proposed TS 6.1, Specification b, "Staffing," against the guidance, and modify the TS-as appropriate, or provide a basis for not providing a specification for a list of reactor personnel.
Technical Specification Section 6 has been revised. Please see the attached revision.
\\
g. The regulations in 10 CFR 50.54(m)(l) states that recoveiy from a unplanned significant reduction in power requires the presence of a senior reactor operator. In addition, ANSI/ANS-15.1-2007, Section 6.1.3, "Staffing," item (3)(d) includes guidance on events requiring the presence of an SRO at the facility. Evaluate the proposed TS 6.1, Specification b, "Staffing,"
31 -
item (3)(c) against 10 CFR 50.54(m)(l) and the guidance and modify the TS as appropriate, or provide a basis for not including significant power reduction events in the specification. H recovery from a significant reduction in power is added to the TS, define what constitutes a significant reduction in power. Initial startup appears to be defined as the startup following a core change. The NRC staffs inteIJ>retation of initial startup is that it is the first startup after the facility was placed in a status where it could be unstaffed. Proposed a modification of the TS consistent with this inteIJ>retation, or justify how your inteIJ>retation meets 10 CFR 50.54(m)(l).
Technical Specification Section 6 has been revised. Please see the attached revision.
h. Proposed TS 1.4Q defines "~hall" as id!!ntifying a requirement Revise sections of proposed TS 6.1 as identified below to be consistent with proposed TS 1.40, or justify why no changes are necessacy. Additional typographical errors and other issues are also addressed below.
Revise sections of TS 6.1 as identified below to address the typographical errors and other issues, or justify why no changes are necessary.
TS 6.1, Specification a, "Structure,"
Item a - "The Dean of the College of Engineering (Level 1) will be the individual responsible for the facility's license or charter" to "The Dean of the College of Engineering (Level 1) shall be the individual responsible for the facility's license."
Issues: Not consistent with TS 1.40 with "will" statement Reference to charter.
Technical Specification Section 6 has been revised. Please see the attached revision.
TS 6.1, Specification b, "Staffing,"
Item (l)(a) - A licensed reactor operator in the reactor room" to "A licensed reactor operator shall be in the control room" Issue: Not consistent with TS 1.40.
Technical Specification Section 6 has been revised. Please see the attached revis10n.
Item (l)(c) -
"... operation. "Readily Available on Call means... " to
"... operation. "Readily Available on Call" means... "
Issue: Apparent typographical error.
Technical Specification Section 6 has been revised. Please see the attached revision. -
Item (l)(c)(i) - "Has been specifically designates and... " to "Has 32-
been specifically designated and... "
Issue: Apparent typographical error.
Technical Specification Section 6 has been revised. Please see the attached reV1s1on.
Item (l)(c)(iii) - "... within a reasonable time under normal conditions." To "... within a reasonable time (30 minutes) under normal conditions."
Issue: Define reasonable time.
Technical Specification Section 6 has been revised. Please see the attached reVISlOn.
TS 6.1, Specification c, "Minimum Qualifications of Reactor Personnel,"
NUREG-1537, Section 12.1.4, "Selection and Training of Personnel," includes guidance on the selection and training of personnel for research reactors and states that ANSI/ANS-15.4 may also be used for additional guidance_. The proposed TS-6.1, Specification c refers to the ANSI/ ANS-15.4 standard without identifying the year of issue. Add the year of the standard to the TS (current version of the standard is 2016) or justify why no change is necessaiy. The guidance in ANSI/ANS-15.1, Section 6.1.4, "Selection and Training of Personnel," states that operations personnel shall meet or exceed the requirements of ANSI/ANS~l5.4. However, the TS contain a "should"-statement about the applicability of ANSI/ANS-15.4. Justify the use of a "should" statement or propose wording for the TS consistent with ANSI/ANS-15.1.
Itemb
- The guidance in ANSI/ANS-15.4-2016, Section 4.3, "Level 2," states that a Level 2 person
__ ~
~ve a minimum of six years of nuclear experience. Modify TS 6.1, Specification c, item (b), "... shall have a minimum five years... " to "... shall have a minimum six years... " or justify why no change is Iiecessaiy.
Technical Specification Section 6 has been revised. Please see the attached revision.
- The phrase "... meets the certification requirements of the licensing agency... " is unclear.
Modify the phrase to reflect that the license is issued by the U.S. NRC, or justify why no change is necessaiy.
Item c Reactor Supeivisor Technical Specification Section 6 has been revised. Please see the attached revision.
- To have TSs that are gender neutral modify "He shall have a baccalaureate degree... " to "The reactor supeivisor shall have a baccalaureate degree... "
Item d Licensed Senior Operator (Level 4) 33-
Technical Specification Section 6 has been revised. Please see the attached revision.
- ANSI/ANS-15.1-2007, Section 6.1.1, "Structure," includes criteria on the structure of the organization. The J>UR-1 operating staff (Level 4) is not clearly identified in the proposed TS Section 6.1 and Figure 6.1 as Level 4 organi7.ational structure. Clarify the organizational and reporting structure for the Level 4 personnel.
- To have TSs that are gender neutral, modify "He shall hold a valid NRC Senior Reactor Operator's license" to "The senior operator shall hold a valid NRC Senior Reactor Operator's license."
Technical Specification Section 6 has been revised. Please see the attached revision.
Item e licensed Operator
- To have TSs that are gender neutral, modify "He shall hold a valid NRC Reactor Operator's license" to "The licensed operator shall hold a valid NRC Reactor Operator's license."
Technical Specification Section 6 has been revised. Please see the attached reVISlOn.
TS 6.1, Specification d, "Radiation Safety Officer,"
- "... including some fonnal training in radiation protection." Define and justify the use of the word "some" or remove the word from the proposed TS.
Technical Specification Section 6 has been revised. Please see the attached reVISlOn.
- "The RSO should have at least five years... " Justify the use of a "should" statement or propose a "shall" statement Technical Specification Section 6 has been revised. Please see the attached reVISlOn.
- "At least three years of this professional experience should be... " Justify the use of a "should" statement or propose a "shall" statement Technical Specification Section 6 has been revised. Please see the attached revision.
TS 6.1, Specification e, "Reactor SupeIVisor,"
/
34-
- To be consistent with the requalification program, modify "... shall be responsible for the facility retraining... " to
"... shall be responsible for the facility requalification and...," or justify why the current wording is acceptable.
Technical Specification Section 6 has been revised. Please see the attached revis10n.
60. The regulation in 10 CFR 50.36(c)(5) requires administrative controls necessary to assure operation of the facility in a safe manner. Due to the importance to overall facility safety of a uniform intemretation by both the licensee and regulator of terms and phrases used in TSs, definitions shall be included where necessacy and consistently used to ensure that the TSs criteria necessary for compliance with regulatocy requirements is uniformly understood by the licensee and regulator.
Proposed TS 1.40 defines "shall" as identifying a requirement Revise sections of proposed TS 6.2, "Review and.A,~dit", ~
_id~ntified below to be consistent with TS 1.40, or justify why no changes are necessary. Additional typographical errors and other issues are also addressed below. Revise sections of TS 6.2 as identified below to address the typographical errors and other issues, or justify why no changes are necessary.
a. TS 6.2, Specification a. "CORO will advise the Laboratocy Director ~d/or the Reactor Superyisor... "
Issue: Not consistent with TS 1.40.
Technical Specification Section 6 has been revised. Please see the attached revision.
b. The proposed TS Section 6.2 has a numbering scheme that is not clear aetter followed by letter). Revise the numbering lists in TS Section TS 6.2. Specifically, TS 6.2 Specification b, TS 6.2, Specification e, and TS 6.2, Specification f, or justify why no change is necessary.
Technical Specification Section 6 has been revised. Please see the attached revision.
c. TS 6.2, Specification c. "Sub committees may be formed as needed, which may consist of a minimum of 3 (three) members, only one of which may have line... " Issue: Justify second and third use of "may" statements.
Technical Specification Section 6 has been revised. Please see the attached revision.
d. The proposed TS 6.2,' Specification d refers in Items a through c, -to "unreviewed safety question," which is not consistent with 10 CFR 50.59, "Changes, tests, and experiments," or the guidance in ANSI/ANS-15.1-2007. Modify TS 6.2, Specification d, Items a through c, by updating the reference to "unreviewed safety questions" to comply with 10 CFR 50.59 and the guidance in ANSI/ ANS-15.1-2007, or justify why no changes are necessary.
Technical Specification Section 6 has been revised. Please see the attached revision.
35 -
e. TS 6.2, Specification d, Item d "Proposed changes in Technical Specifications or licenses."
Issue: Appears to have a typographical error, more.than one license is referenced.
Technical Specification Section 6 has been revised. Please see the attached revision.
f.
The proposed TS 6.2, Specification d, Item f provides requirements for reviewing operating abnormalities. The requirements do not follow the guidance in ANSI/ANS-15.1-2007 Section 6.2.3(6), "Review Functions." For example:
"Significant operating abnormalities or deviations from normal and expected performance of facility equipment that might affect nuclear safety" to "Operating abnormalities or deviations from normal and expected performance of facility equipment having safety-significance."
Technical Specification Section 6 has been revised. Please see the attached revision.
g. The proposed TS 6.2, Specification d provides requirements for items to be reviewed by the Committee on Reactor Operations (CORO), which does not include a review of the facility radiation protection program. Provide a description of the review process and the organization responsible for the review of the radiation protection program and how it complies with the guidance inANSI/ANS-15.1-2007, Section 6.3, "Radiation Safety."
Technical Specification Section 6 has been revised. Please see the attached revision.
h. The proposed TS 6.2, Specification e, Item b requires an annual audit of the training and qualification of the licensed facility staff. The guidance in ANSI/ANS-15.1-2007 Section 6.2.4(2), "Audit Function," states that the periodic audit is for the retraining and requalification program for the operating staff. Revise TS 6.2, Specification e, Item b to follow the guidance in ANSI/ ANS-15.1-2007, or justify why no change is necessary.
1.
'fh.e guid~ce _in ANSI/ANS-15.1-2007, Section 6.2.4(3), "Audit Function," states the audit requirements for corrective actions that affect reactor safety. The proposed TS 6.2, Specification e, Item c does not fully consistent with the guidance. Revise TS 6.2, Specification e, Item c, as appropriate, or justify why no change is necessary. For example:
"... method of operation that affect nuclear safety... " to
"... method of operation that affect reactor safety... "
Technical Specification Section 6 has been revised. Please see the attached revision.
J*
The proposed TS 6.2, "Review and Audit," Specification f, Items a and b states, in part, that CORO meeting minutes will be distributed to the Reactor Supervisor, and in Specification f, Item c to CORO Chairman. The guidance in ANSI/ANS-i5.1-2007, Section 6.2, "Audit Function," states that the meeting minutes and reactor safety issues are to be provided to Level 1 management and the CORO group members. Revise TS 36 -
6.2, Specification f, Items a through c to include these additional organiV1tional members or justify why no changes are necessaiy.
Technical Specification Section 6 has been revised. Please see the attached revision.
k. The proposed TS Section 6.2, Specification f, Item b refers to items, "... by section 6.2.4 e, f, and g above," while Specification f, Item c refers to
"... Section 6.2.5," which appear to be incorrectly numbered. Revise the numbering lists in TS Section 6.2 together with the appropriate references, or justify why no changes are necessary.
Technical Specification Section 6 has been revised. Please see the attached revision.
The following applies to RAis No. 61 through 65, below. The regulation in 10 CFR 50.36(c)(5) requires administrative controls necessary to assure operation of the facility in a safe manner.
61. The proposed TS 6.3, "Operating Procedures," includes requirements for written operational procedures. The guidance in ANSI/ANS-15.1-2007, Section 6.4 "Procedures," includes
~ditional technical and administrative activities, which require the development of additional procedures, specifically Section 6.4(5), 6.4(6), and 6.4(8). Revise TS 6.3 to include the additional procedures to be consistent with the guidance inANSI/ANS-15.1-2007, Section 6.4, or justify why no changes are necessary.
Technical Specification Section 6 has been revised. Please see the attached revision.
62. The proposed TS 6.3, "Operating Procedures," states that "The Reactor Supexvisor or laboratory Director may make changes to procedures," but does not include a reference that all changes are made under the 10 CFR 50.59 process. NUREG-1537, Part 1, Section 12.3, "Procedures," provides guidance* that the review should include the determinations that the proposed* changes were allowed without prior NRC approval. Revise TS 6.3 to include the revie'Y' for proposed changes consistent with the guidance, or provide a justification for the proposed TS 6.3.
Teclmical Specification Section 6 has been revised. Please see the attached revision.
63. The proposed TS Section 6, "Administrative Controls," is not fully consistent with the guidance in ANSI/ANS-15.1-2007; Section 6.5, "Experiment Review and Approval." Revise TS 6 to include the guidance provided in ANSI/ANS-15.1-2007, Section 6.5, or justify why no changes are necessaiy.
Technical Specification Section 6 has been revised. Please see the attached revision.
64. The proposed TS 6.4, "Operating Records," Specification b includes the word "certified."
Modify TS 6.4, Specification b to reflect that the PUR-1 operational personnel is licensed by the NRC, or justify why no changes are necessary. For example:
37 -
"Record of retraining and requalification of certified operations personnel shall be maintained at all times the individual is employed or until the certification is renewed" to "Record of retraining and requalification of licensed operations personnel shall be maintained at all times the individual is employed or until the license is knewed."
Technical Specification Section 6 has been revised. Please see the attached revision.
65. The proposed TS 6.5, "Required Actions," Specification a, Items (1) through (4), provide requirements in case of safety limit violations. However, TS 6.5, Specification a, Item (1) through (4) are not fully consistent with the guidance in ANSI/ ANS-15.1-2007, Section 6.6.1, "Action to be taken in case of Safety Limit Violation." For example, Specification a, Item 3 states
"... applicable circumstances preceding the violation," but does not mention any requirement for cause and contributing factors as in the ANSI/ANS-15.1-2007, Section 6.6.1 guidance. Revise TS 6.5, Specification a, Items (1) through (4), to include the guidance included inANSl/ANS-15.1-2007, Section 6.6.1, or justify why no changes are necessary.
Technical Specification Section 6 has been revised. Please see the attached revision.
66. The regulation in 10 CFR 50.36 (c)(5) requires administrative controls necessary to assure operation of the facility in a safe~manner. Additionally, due to the importance to overall facility safety of a uniform interpretation by both the licensee and regulator of terms and phrases used in TSs, de~tions shall be included where necessaiy and consistently used to ensure that the TSs criteria necessary for compliance with regulatory requil-ements is uniformly understood by the licensee and regulator.
The proposed TS 1.40 defines "shall" as identifying a requirement Revise sections of TS 6.5, "Required Actions," as identified below to be consistent with TS 1.40, or justify why no changes are necessary. In addition, the TS 6.5 contains phrases, which are not specific to the facility or do not express requirements. Revise TS 6.5 to be specific to the PUR-1 or justify why no changes are necessary. For example:
TS 6.5, Specification a, Item (1)
"The reactor will be shut down immediately and reactor operation will not be resumed without authorU.ation by the Commission" to "The reactor shall be shut down immediately and reactor operation shall not be resumed without authorU.ation by the U.S. NRC."
Technical Specification Section 6 has been revised. Please see the attached revision.
TS 6.5, Specification b, Item (1)
"... unless authorized by Level 2 or designated alternates;" to
"... unless authorized by the Laboratory Director or designated alternates; Technical Specification Section 6 has been revised. Please see the attached revision.
TS 6.5, Specification b, Item (2) 38 -
"Occurrence shall be reported to Level 2 or designated alternates and to chartering or licensing authorities as required;" to "Occurrence shall be reported to the laboratory Director or designated alternates and to the US NRC;"
Technical Specification Section 6 has been revised. Please see the attached revision.
TS 6.5, Specification b, Item(3)
"Occurrence shall be reviewed by the review group at its next scheduled meeting" to "Occurrence shall be reviewed by the CORO at its next scheduled meeting."
Technical Specification Section 6 has been revised. Please see the attached revision.
67. The regulation in 10 CFR 50.36(c)(5) requires administrative controls necessary to assure operation of the facility in a safe manner.
The proposed TS 6.6, "Reporting Requirements," contains phrases, which are not specific to the facility or do not express requirements. Revise TS 6.6, as appropriate, to follow the guidance in ANSI/ANS-15.1-2007, or justify why no changes are necessary. For example:
TS 6.6, Specification a
"... shall be submitted to the Director of the Office of Nuclear Reactor Regulation with a copy to the NRC Regional Administrator by March 31 of each year" to
"... shall be submitted to the U.S. NRC Document Control Desk by March 31 of each year."
Technical Specification Section 6 has been revised. Please see the attached revision.
TS 6.6, Specification a, Item d
"... corrective maintenance (excluding preventive maintenance) performed... " to
"... corrective maintenance and major preventive maintenance performed... "
Technical Specification Section 6 has been revised. Please see the attached revision.
TS 6.6, Specification b, Item a(l)
"... in writing by facsimile or similar conveyance to licensing authorities, to be followed by a written report to NRC Document Control that describes the circumstances of the event within 14 days of any of the following:" to
"... in writing by. facsimile or similar conveyance to the NRC Headquarters Operations 39-
Center, and followed by a written report that describes the circumstances of the event and sent within 14 days to the U.S. NRC Document Control Desk any of the following:"
Technical Specification Section 6 has been revised. Please see the attached revision.
TS 6.6, Specification b, I~m a(l)(b)
"... established in the technical specifications unless prompt remedial action is taken as permitted Di Sec. 3," to
"... established in the technical specifications,"
Technical Specification Section 6 has been revised. Please see the attached revision.
TS 6.6, Specification b, Item a(l)(d)
"... change in reactivity greater than 0.696L\\k/k." to
"... change in reactivity greater than 0.6%8k/k. Reactor trips resulting from a known cause are excluded" Teclmical Specification Section 6 has been revised. Please see the attached revision.
TS 6.6, Specification b, Item a(l)(e)
"... abnormal and significant degradation in reactor fuel or cladding, or both, coolant boundary," to
"... abnormal and significant degradation in reactor fuel or cladding, or both, coolant boundacy, or containment boundacy,"
Technical Specification Section 6 has been revised. Please see the attached revision.
TS 6.6, Specification b, Item a(2)
"There shall be a written report within 30 days to the chartering or licensing authorities of the following:" to "There shall be a written report within 30 days to the U.S. NRC Document Control Desk, of the following:"
i' Technical Specification Section 6 has been revised. Please see the attached revision.
TS 6.6, Specification b, Item b "A written report shall be forwarded within 30 days to the Director, Office of Nuclear Reactor Regulation with a copy to the in the event of:" to "A written report shall be foIWarded within 30 days to the U.S. NRC Document Control Desk in the event of:"
40-
Technical Specification Section 6 has been revised. Please see the attached revision.
The following applies to RAis No. 68 and 69, below. The regulation in 10 CFR 50.36(c)(5) requires administrative controls necessary to assure operation of the facility in a safe manner.
NUREG-1537, Part 1, states thatthe format and content of the TS follow that of ANSI/ANS-15.1-2007.
68. The proposed TS Section 6.6, "Reporting Requirements," Specification a "Annual Reports,"
is not consistent with the guidance in ANSI/ANS-15.1-2007, Section 6.7.1, "Operating Reports," item (5) that includes reporting criteria for the results of environmental swveys.
Revise TS 6.6, Specification a to include the guidance provided in ANSI/ANS-15.1-2007, Section 6.7.1, or justify why no changes are necessary.
Technical Specification Section 6 has been revised. Please see the attached revision.
69. The proposed TS Section 6.6, "Reporting Requirements," Specification h, Item a "Special Reports," Suh-item (1) is not consistent with the guidance in ANSI/ANS-15.1-2007, Section 6.7.2,"Special Reports," item (l)(b) that"includes reporting criteria for the release of radioactivity from the site. Revise TS 6.6, Specification h to include the guidance provided in ANSI/ ANS-15.1-2007, Section 6. 7.2, or justify why no changes are necessary.
Technical Specification Section 6 has been revised. Please see the attached revision.
The following applies to RAI No. 70 through 7 4, below. The regulation in 10 CFR 50.34(b) requires a final SAR that shall include information that describes the facility, presents the design bases and the limits on its operation, and presents a safety analysis of the structures, systems, and components and of the facility as a whole.
70. NUREG-1537, Part 1, Chapter 13, "AccidentAnalyses," states that non-power reactors should analyze events that could affect their safe operation or shutdown including evolution of scenarios and evaluating the consequences of postulated events. For estimating occupational doses after the postulated maximum hypothetical accident (MHA) the PUR-1 SAR assumed that the facility personnel would evacuate from the reactor pool room within a 1-minute timeframe based on "past experience." In the response to RAI No. 96 (datedJune 15, 2012), it was estimated, using analytical est:Unat.es, tha.tt:he iruiximuni. amoUn.t of time required for a staff member to evacuate the reacto!' room is less than 15 seconds._ Discuss whether there is an evacuation procedure, how the strlf members and/or inemher of the public, if escorted in the reactor room, have training to follow the plan,-and Whether-there is any evacuation exercises practiced periodically that would provide data for the evacuation time.
Knowledge of the Emergency Plan is part of the reactor training and may be tested by the U.S.
NRC examiner. All operators are regularly instructed to review the Emergency Plan which has already been duly submitted to as part of the license renewal and power uprate work. As part of the requirement to conduct an Emergency Exercise, the evacuation is practiced periodically.
There have not been data collected during these exercises but in response to this question, facility staff timed a simulated evacuation which took 7 seconds to leave the reactor room via the main entrance, 15 seconds to leave via the reactor room stairway and 51 seconds to leave the building 41 -
via the hallway, 24 seconds to leave via the stainvay. All of these values are less than the numbers which were previously analyzed.
71. NUREG-1537, Part 1, Chapter 13, "Accident Analysis," provides guidance on analyzing credible postulated accidents. The postulated bounding MHA scenario assumes that fission products escape from one of the fuel plates and instantaneously disperse into the reactor room air. One of the potential consequences is a leak of reactor room air containing fission products around the reactor room door into unrestricted areas. Provide a discussion regarding the consequences of the leak around the reactor room door, list any assumptions for the MHA leakage, and provide dose calculations to the maximally exposed member of the public assuming inhalation and submersion of radioactive products in the leakage as well as the direct gamma-ray radiation shine from the radioactive products dispersed in the reactor room.
The nearest unrestricted location which a member of the public can access is near a doorway which has a tight seal. Air will take the least restrictive pathway to diffuse out of the room which would be through the doorway to the restricted part of the facility. Therefore, the only part of the dose will come from the shine from the lliterior of the room which has been previously analyzed.
72. NUREG-1537, Part 1, Chapter 13, "Accident Analysis," provides guidance on analyzing credible postulated accidents including the potential radiation levels in unrestricted areas after the MHA. In the response to RAI No. 14 (datedJuly 19, 2016), an analysis was provided for the gamma-ray exposure and the consequent maximum dose :rate to the member of the public in the unre-stricted area above the reactor room. One of the assumptions used in the model is that members of the public would eventnally evacuate. Discuss whether there is a general evacuation procedure, how that procedure is consistent with the facility emergency plan, how the members of the public are instructed to follow the plan, how it is insured that all members of the public actnally evacuate the building, and whether there are any evacuation exercises practiced periodically.
Purdue University has evacuation procedures for all buildings and can be found in the building emergency plan. If an evacuation is required, the building fire alarm is activated by reactor or fire department personnel as directed by the facility emergency plan. Emergency procedures require evacuation of the building. In cases where it is feasible, the Purdue University Fire Departrpent will verify that all occupants have evacuated the building. While there are no regularly scheduled evacuation exercises, communication on emergency procedures is sent out to faculty, staff, and students on an annual basis.
73. NUREG-1537, Part 1, Chapter 13, "Accident Analysis," provides guidance on analyzing credible postulated accidents includll!g_the potential radiation levels in unrestricted areas after the MHA.
In the response to RAI No.14(dai:edJuly19, 2016) the PUR-1 MHA calculations provided the maximum dose to the member of the public at the maximally exposed location, 100 m away from the facility. Provide a dose calculation to an exposed member of the public located at the closest continuously occupied location such as a private residence, residence hall, and/or donnitocy.
The dose rate to a member of the public if the reactor room fan is left on is calculated below.
42-
The concentration of radioactivity is defined as Activity Concentration = V l oume The activity of a given fission product is a function of time and is given as where II. is the isotopic decay constant and A0 is the initial activity. This yields a concentration of A -ilt oe C(t) = V l oume Additionally, the concentration being expelled from the reactor room is continually reduced due to the replenishment of fresh air. The differential equation describing the reactor room concentration is dC(t)
~=
Cin -Cout The volumetric concentration of radioactive air coming in to the reactor room is zero (clean air supply) and the concentration going out is R
Cout = C(t)
V where R is the rate of air leaving the reactor room and V is the reactor room volume. This yields dC(t) = -C(t)
R dt v
The solution to this equation is This concentration is reduced by the wind dispersion factor and the exhaust rate.
(
Using the equation above for the initial concentration gives A0 (x/Q)Re-ilte-(~)t C(t) =
V l oume 43-
The dose rate from submersion is fom1d with Dose Rate = k
C(t) h k.
c.
"th f mRem/hr I th d
. _.c:
w ere is a convers1on 1actor wi umts o
.1 3
ntegratmg e ose over 11u11nte tllne gwes
µCi cm the dose rate to a member of the public submerged in the plume for infmite time.
-1 00 AoCx/Q)Re-A.te-(~)t Dose -
k V l dt 0
o ume kA 0(x/Q)R Submersion Dose =
R v(il+v)
The submersion and inhalation dose results from the diluted radioiodine *stream from the exhaust fan. An analysis for the activity concentration release from the building can be performed using the equation below as specified in NUREG/CR-2260. The concentration of activity is given as where Ai is the fractional release of activity of the element, x/Q is the atmospheric dispersion factor with milts seconds/m3, and V (m3 /sec) is the volumetric release rate from the fan. The dispersion factor is calculated by taking the most appropriate value of X / Q from the equations below 1
x/Q=---
u103ncrycrz 1
x/Q=---
u1onLyCTz 44-(1)
(2)
(3)
where n = 3.14159, u10 is the wind speed at ten meters above plant grade, ay is the lateral plume spread (a function of atmospheric stability and distance), az is the vertical plume spread (a function of atmospheric stability and distance), Ly is the lateral plume spread with meander and building wake effects, and A is the smallest vertical-plane cross-sectional area of the reactor building in square meters. To find the best version of the dispersion factor to use, the larger dispersion factor from equations (1) and (2) should be compared with the value from equation (3). The lesser of the first result and equation (3) should be the final dispersion factor.
This analysis will conservatively consider moderately stable atmospheric conditions meaning low dispersion. Specifically, a Class F atmospheric stability is used. A wind speed of 1 m/ s, which is much lower than the average wind speed as referenced in the Safety Analysis Report, will be used. The distance from the smoke stack will be considered to be 70 meters as that is the minimum distance to a continuously occupied public residence. At 70 meters for a moderately stable atmosphere, the following plots a.re extrapolated.
Purdue University:
Forney Hall of...
cl< L.)
lrotlon Engineering Fountain Materials and ElectriC81 Engineering Building lr Purdue Exponent Eta Kappa Nu (HKN) Lounge B School of lectrical and ComP,Uter Engineering Purdue U nive r sity~
School of ~n icol...
z Wl, II Ave W LUIZ Av;>
Q,,
2.
~
Delta Tou Delta Fraternity W Fowle* Ave The nearest permanently occupied residence is highlighted in yellow here. The figures below taken from Regulation Guide 1.145 - "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants" gives an extrapolated value of approximately ay = 2 m.
45 -
I I' I I.
5 I
I I
I
/
"~
~
B I;~
I
/ v r/
I
~
v (I/ I; i.
i
~
I/ E' i... I I
.. F*
~
I
~.....
I I
/ v.
I/ /
I.
/ j l.1
1. 1 *
/
l V/ /
v i...-"' i."
I "1
i
i. ~
~
i/ ~i,.
,)' ~
I/
~
I
.I I
~
11-(!XTREMEL'f UNSTABLE:+/-
/
1.....
B-MODERbTELY UNSTABl.E
/
C: - SLIGHTLY UNSTABI.(
f
/.,,,,,I;
/1/..
0- Nl:UTffA~
V/_,, v,/ }"
E ~ SLIGHTLY STbBLE
~,,
I"* MOOliRATELY STA8LE
~
V'
.... /
I v
I I
i I
I 5
103 2
5 104 2
DISTANCE l'ROM $0URC1i lml Figure 1. Lateral diffusion without meander and building wake effects, oy* vs. down*
wind distance from source fOr Pasquill's turbulence types (atmospheric stability) (Ref, 7),
46-
5
/
v
/ /
~
/.
./
~-,. -
I
,, /',,
/
II'" /I
/
0 10 l!
10.
2
.I I
I v 1...,.11' v
~
~
~
~ I-'
s I
ii II II JI l
J
~
~-
~""'
~
/
i,..'lo' I/ v l_/ /
/
~ ~
~..
~
B v
~
~
~....
_c I~
~
~..-
o'
~
II
/"
V/E I-'....
~~,.v
/,...
~,. /
J
.. ~
A - E:KTRcl'ltELY UNSTABL£ 8-MODERATEt.V UNS1'.AS1,.E
(; - SLIGKil.Y 1,JNSTAS\\.E D-NEUTR-"ll E ~ SLIGHTLY ST.OoBLE f - MOOERATEl.Y STA81.I;;
I 4
103 2
5 lO 2
s QISTANCE F"ROM SQURl;E (m)
Figure 2. Vertical diffusion without meandel" and building wake effects~
a
vs. downwind distance from source for Pasquill's turbulence
~(atmospheric stability) (Ref. 7}.
This shows a value of Uz = 1.5 m. The value for Ly for distances less than 800 meters is given as where M is a correction factor based on atmospheric stability and wind speed. With a wind speed of 1 m/ s and Class F stability, M = 4 and therefore Ly= 2
4 = 8 Finally, the cross sectional area is A = 288 m 2
The values of the atmospheric dispersion factor is the1.1 47 -
1 sec x/Q =
m (
288 m2) = 0.00652 m3 1 s n
2 m
1.5 m +
2 (1) 1 sec x/Q = m
= 0.0354 -
3 1-
3
n
2 m
1.5 m m
s (2) 1 sec x/Q = m
= 0.0265 -
3 1-
n
8 m
1.5 m m
s (3)
The larger of equations (1) and (2) is equation (2)'s value of 0.0354 sec/m3 and the smaller of equation (3) and the previous result is equation (3)'s solution of 0.0265 sec/m3.
Therefore, the value of the dispersion factor is taken to be sec x/Q = 0.0265 -
3 m
Given that the exhaust fan expels approximately 0.2 m 3 /sec of contaminated air (in a scenario where the exhaust fan remains running), the room also brings in 0.2 m 3 /sec of clean air. For an isotope such as Iodine-131, the submersion dose is Submersion Dose r_mR/hr ]
[ sec ]
[sec]
67.41 l/lCi/cm3
3600 hour0.0417 days 1 hours 0.00595 weeks 0.00137 months 
103000[µCi]
0.0265 m3 0.2 424 m3
106 [cm3] (9.98 x 10-7 [_!_] + 0.
2 [#c])
m3 sec 424 [m3]
Submersion Dose= 5.12 x 10-5 [mRem]
Similarly, the inhalation dose factor is given by The dose rate is then mRem k =
µCi Dose Rate = k
C(t)
B 0R
[::]
where B.R is the breathing rate by a member of the public. Again, integrating over time yields the 48 -
total dose, f
00 AoCx/Q)Re-ilte-(~)t.
Dose = k V l BR dt 0
o ume kAoCx/Q)R(B.R) f 00 -ilt -(Ii)t d
=
e e v t
v 0
kA 0(x/Q)R(B'R)
V (1L + ~)
32.89 [mRe!11]
103000[µCi]
0.0265 [sec] 0.2 [m 3
]
333 [cm 3
]
µCi m3 sec sec 424 m3
1Q6 [cm3] (9.98 x 10-7 [..l.] + 0.
2 [#c])
m3 sec 424 [m3]
Inhalation Dose= 29.92 [mRem]
The table below shows the dose for all isotopes.
lnhaJarion Submersion Isotope Dose [mRem]
Dose [µRem]
1-131 29.92 0.05 1-132 0.44 0.05 1-133 12.13 0.19 1-134 0.22 0.68 1-135 2.29 0.46 Kr-85m 0.04 0.02 Kr-87 0.32 0.26 Kr-88 1.30 0.41 Xe-13lm 0.00 0.00 Xe-133m 0.00 0.00 Xe-133 0.04 0.01 Xe-135 0.34 0.19 Xe-135m 0.04 0.01 Total By Dose Type 47.08 2.33 Total Overall Dose 47.08 [mRem]
Summing this for all the released fission products yields an inhalation dose of 47.083 mRem, a submersion dose of 2.33 µRem, and a total dose of 47.08 mRem. This is less than the limits as set forth for accident scenarios.
49-
/
7 4. NUREG-1537, Part 1, Chapter 13, "Accident Analysis," provides guidance on analyzing credible postulated accidents including potential consequences of a loss-of-coolant event (LOCA). In the response to RAI No. 15 (dated July 19,. 2016), an analysis was provided for a LOCA event resulting in a partial loss of~oolant., with a reduced amount of coolant., but still leaving the reactor core submerged. In a hypothetical total loss-of-coolant., the reactor core is fully uncovered resulting in an unshielded gamma-ray exposure in the reactor room and floors above the reactor room. In the response to RAI No. 14 (dated July 24, 2015), an analysis was provided for a total LOCA including the maximum dose rates for the occupational staff and the member of the public located above the reactor room shielded only by the concrete floor. The analysis states that the potential dose rate to a member of the public would be 6.6 rem/hr and "... in such an event the building would be immediately evacuated such that the actual dose received by a member of the public would be significantly reduced." Provide the doses to members of the public assuming realistic evacuation times verified by past evacuation drills. Verify that these actions are consistent with your emergency plan.
The first line of RAI #15 dated July 19, 2016 stated, "The response to this RAI is intended to supersede all previous Loss of Coolant Accident Analyses." That response intended to supersede the response in RAI #14 dated July 24, 2015. A detailed description of a highly unlikely Loss of Coolant Accident was given for both a worker as well as for a member of the public. Reference the letter sentJuly 19, 2016 for the completed postulated LOCA analysis.
7 5. The regulation in 10 CFR 50.36(c)(4) requires the inclusion of design features of those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered elsewhere in the specifications.
The proposed TS 5.3, "Reactor Core and Fuel," Specification f describes the periodic inspection requirements for fuel assemblies. NUREG-1537, Part 1, Appendix 14.1, Section 4.1, "Reactor Core Parameters," item (6), provides guidance for fuel inspection. The anticipated gamma-ray exposure during the inspection of the fuel assemblies may increase due to the requested increase in licensed reactor power" to 12 kWt Provide a description of the protective measures and changes in the inspection procedures required to prevent exceeding the occupational dose limits in 10 CFR 20.1201 for staff performing the fuel inspections.
Limits set forth in procedures at the PUR-1 facility require the dose levels of items leaving the pool be less than 1 R/hr at 1 foot In the event it appears an assembly will surpass those lirriits, the assembly will be replaced in the pool and given adequate time to cool prior to inspection.
76. The regulation in 10 CFR 50.34(b) requires a final SAR that shall include information that describes the facility, presents the design bases and the limits on its operation, and presents a safety analysis of the structures, systems, and components and of the facility as a whole. The PUR-1 SAR, Chapter 4 (pages 4 through 44) describes two accident event scenarios, rapid and slow insertion of reactivity, which may be inconsistent with the analysis presented in Chapter 13, where the maximum reactivity insertion is 0.006 AfVk. The slow insertion scenario assumes that the rate is limited by a maximum allowable rate of 0.0004 W(k*s). It is not clear, how this 50-
maximum rate limit is controlled, since the TSs have no limits on control rod reactivity insertion rate. In addition, it is also unclear, what the reactivity insertion rate is due to the control rod operating rates presented in Chapter 4 of the SAR (regulating rods - 17. 7 in/min, shim rods - 4.4 in/min). Provide further information on how the maximum all0Wc1.ble reactivity insertion rates due to control rod movements are controlled and are consistent with the analysis. Propose a limiting condition for operation and associated swveillance requirement or justify why TSs are not necessary.
The values listed in the SAR for the maximum possible reactivity insertion rate from control rod motion are calculated from the reactivity worth of the rods and the maximum withdrawal speed of the drive motors. The withdrawal speed of each control rod is a fixed value, based upon the control rod drive mechanisms, and the reactor operator cannot alter this value. Accordingly, no limiting condition for operation is required with respect to the control rod withdrawal speed.
The values listed in the SAR for the maximum and average reactivity insertion rate due to control rod motion were inconsistent in their units. Table 4-2 was listed with units of (%Llk/k) or
(%Llk/k*s), where Table 4-13 was listed with units of (Llk/k*s). Also, Table 4-5, while it has the units shown as (Llk/k*s), the values given are correct for (%Llk/k*s). These discrepancies have been corrected and the units are now consistently given as (Llk/k*s) and the magnitude of the values are shown correctly for those units. It can be seen that all calculated and measured values of the maximum reactivity insertion due to control rod motion are less than the value analyzed in the SAR, thus leaving the conclusion unchanged. The maximum possible reactivity insertion rate is therefore completely bounded by the analysis. Accordingly, no limiting condition for operation is required with respect to the maximum reactivity insertion rate due to control rod withdrawal.
77. The regulation in 10 CFR 50.34(b) requires a final safety analysis report that shall include information that describes the facility, presents the design bases and the limits on its operation, and presents a safety analysis of the structures, systems, and components and of the facility as a whole.
The PUR-1 SAR Chapter 4 indicates that the radiation monitors would trigger an alann in the reactor room console, if the reactor water tank level decreases below a certain level due to an unexpected leakage. Provide additional information how reactor staff or university personnel would be alerted if the leak. occurred when the reactor is shut down and no operational staff is present in the reactor room to obsexve the alann, for example, on nights and weekends.
Previous analysis has shown that once the reactor is shutdown and following a potential Loss of Coolant Accident, the maximum dose to a member of the public is far below the limits set forth in 10 CFR Pa.rt 20. These calculations bound those which would occur if there was a LOCA following shutdown, as the reactor would have been in a shutdown condition prior to the LOCA postulated in this RAI. While the radiation area monitors are operable at all times in the PUR-1 Facility, reactor staff or university personnel would be alerted by the audible alarm when the dose levels were to rise above the dose limits set forth in Table II of the Technical Specifications.
78. Subpart F of 10 CFR Part 20 provides the regulatmy requirements for radiation suxveys and monitoring.
51 -
The PUR-1 SAR states that routine radiation smveys are performed by health physics staff. It is not clear if the health physics personnel are considered PUR-1 reactor staff or Punlue University personnel. Provide a description of the organizational arrangement for the health physics staff.
Health physics personnel are Purdue University staff which are members of the Radiological and Environmental Management team. They assist in monitoring regulatory compliance with federal, state, and university regulations involving environmental, health, and safety issues. The Radiation Safety Section is respol1sible for complying with regulations set forth by the NRC as well as the Indiana State Department of Health for the safe use of radioactive materials on campus. The health physics staff are overseen by the University Radiation Safety Officer.
52-
Jim:5chw*1tt*r Radiation.51f.iy Officer Plly1f11 Hi l SKrUiry John fffn.ey carol Shelby S*nior DlrilCtor 4*75<M Jim 5chw*imr Di-rtaor.1nd A.SO
~2 150 St*phlni* SU.ti EHS Mifla,ir, IPl"W K*t* arott fHS
>j!OCllist 11obaoldtn vironme tll Hnltti Manactr.and e.losaftty Officer Jennifer Albl.JfV fnvironme:ntal TKhnician A.1Cti11I O.:!l:udd*
e:rMronnwntal T1tttlnid1n
79. The regulation in 10 CFR 50.36(c) (2) requires that limiting condition for operation of a nuclear reactor shall be established in TSs for each item meeting one or more of the criteria provided in 10 CFR 50.36(c)(2)(ii). Additionally, 10 CFR 50.34(b) requires a final SAR that shall include information that describes the facility, presents the design bases and the limits on its operation, and presents a safety analysis of the structures, systems, and components and of the facility as a whole.
53 -
The PUR-1 SAR Chapter 3.5, "Systems and Components," describes the emergency shutdown of the reactor via manual scram buttons, one located on the reactor console, another one out.side the main personnel access door to the reactor room. The scram functions are also included in TS 3.2, Table I, "Safety Channels Required for Operation." Provide additional information on the design and operation of the manual scram button that is located out.side the personnel access door, specifically: (a) is the location considered controlled or do members of the public have access to the button, (b) who controls the operation of the emergency button, (c) what are the design features preventing inadvertent actuation, (d) is there a procedure that controls the activation sequence?
The design and operation of the manual scram button located out.side the personnel access door is identical to that of the scram button located on the operator console. It.s primary purpose is to serve as a backup in the event of a reactor room evacuation where the operator neglects to scram the reactor during the event. This allows staff to initiate the scram without having to re-enter the room. The area in which this secondary scram button under consideration is located is restricted by the outer facility doors which are generally locked. This restricts the public from accessing the location. Additionally, the button is behind a plexiglass sliding window to prevent inadvertent actuation during normal operation. Additionally, it is at eye level which prevents it from being contacted by a shoulder or other body part. The button serves solely as a backup. It is tested during the prestart checklist but is never used in normal operation.
80. The regulation in 10 CFR 50.34(b) requires a final SAR that shall include information that describes the facility, presents the design bases and the limit.s on it.s operation, and presents a safety analysis of the structures, systems, and components and of the facility as. a whole.
The PUR-1 SAR, Chapter 5.5, describes the "Primary Coolant Makeup Water System," and it.s operation to maintain the reactor pool water level at 13 feet above the reactor core. Provide additional information on the operation of the makeup water system, specifically:
(a) how is the filling process initiated, (b) does the magnetrol level switch in the pool tum the filling process on and/or off, (c) how is water loss trended, if the system operates automatically, (d) is there a chance for over pressurizing the water tank and consequently the university water supply line, (e) is the water level maintained, when operation staff is not present (night.s and week-ends), (f) and what is minimilln waterl~ss rare that can be detected?
(a) The filling process is initiated through the proper use of the water make-up system and proper valve arrangement Reactor staff check the water level generally three times per :week and record the amount of water which is added to the pool during each fill. (b) The magnetrol level switch energizes a solenoid valve which additionally activates an LED indicating water should be added to the pool. In practice, the operator or a facility staff member (c) manually fills the pool in order to track the amom1t of water which has been added. The total water added is calculated weekly to note any trend changes. (d) The holding tank from which the water is added to the pool is open to the atmosphere (i.e. not sealed tight) and includes an overflow pipe which removes the chance for over pressurizing the tank or mliversity water supply line.
54-
(e) Water continually evaporates from the pool (approximately 30-40 gallons per week). This
, represents less than 1% of the bulk pool volume (6400 'gallons). For half of the water to evaporate, it would take 80 weeks or over a year. Therefore, continuous monitoring on nights and weekends is not needed. (f) The loss of water is accurate to approximately 1 gallon/week which is well within the variability noted in facility history due to changes in room temperature, humidity and pressure.
81. The regulation in 10 CFR 50.36(c)(2) requires that limiting condition for operation of a nuclear reactor shall be established in TSs for each item meeting one or more of the criteria provided in 10 CFR 50.36(c)(2)(ii). Additionally, 10 CFR 50.34(b) requires a final SAR that shall include information that describes the facility, presents the design bases and the limits on its operation, and presents a safety analysis of the structures, systems, and components and of the facility as a whole.
NUREG-1537, Chapter 14, Section 3.2 "Reactor Control and Safety Systems," item (8)
"Control Systems and Instrumentation Requirements for Operation (Added by the NRC),"
includes criteria on the detail and operating characteristics of the measuring channels required for operation (not necessarily having scram capability). The guidance suggests to include all required measuring channels including protective functions in the TS. Add a table to TS 3.2 to include all measuring channels required for operation or justify why no changes are necessary.
Table I and Table II included in the Technical Specification already detail the required measuring channels and their protective functions. The title of these tables as shown in the attached Technical Specifications are "Safety Channels Required for Operation" and "Safety-Related Channels Required for Operation."
82. The regulation in 10 CFR 50.36(c) states TSs will include Swveillances. The regulation in 10 CFR 50.36(c)(3) states "swveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met" The guidance in ANSI/ANS-15.1-2007 includes criteria on Section 4, "Swveillance Requirements," which are companion requirements to Section 3, "Limiting Conditions for Operation," specifications demonstrating the minimum performance requirements established in Section 3. The proposed TS 3.3, "Primacy Coolant Conditions," Specification d speCifies that the primary cool3.nt bulk temperature is maintained below 30 degrees Celsius.
There is no companion Section 4 SUIVeillance specification ensuring the minimum performance level for the bulk water temperatnre. Revise/add a swveillance specification of the' maximum allowable bulk water temperature in proposed TS 4.3, "Primacy Coolant System," to be consistent with the guidance in ANSI/ ANS-15.1-2007, or justify why no change IS necessary.
A surveillance has been added to monitor the temperature level while the reactor is operating.
Teclmical Specification 4.3.d now reads, "The primary coolant temperature shall be recorded in the log book at no interval to exceed four hours if any shim-safety or regulating rod is at a 55 -
height greater than 6 cm." Response to RAI #5 in the letter dated July 19, 2016 indicates a maximum temperature rise of 2.56 °C /hr if the power level were 12 kW+ 50% instrument uncertainty.
83. The regulation in 10 CFR70.32 provides the regulatory requirements related to conditions of licenses for the possession of special nuclear material (SNM). The regulation in 10 CFR 70.32(b) states, in part, that the Commission may incmporate in any license such additional conditions and requirements with respect to the licensee's ownership, receipt., possession, use, and transfer of special nuclear material as it deems appropriate or necessary. However, the NRC staff review of the license renewal application does not find a possession limit for SNM.
Provide a propose limit for the possession of SNM for irradiation, including a mass (gram) limit and a description of the material form, as needed for operation of the reactor and its experimental programs.
2.B.(2) Pursuant to the Act and 10 CFR Part 70, "Domestic Licensing of Special Nuclear Material," to receive, possess, and use, but not separate: (1) up to 3.8 kilograms of contained uranium-235 of enrichment of less than 20 percent in the form of materials testing reactor (MTR)-type reactor fuel; (2) up to 80.0 grams of plutonium contained in encapsulated plutonium-beryllium sources; and (3) up to 100 grams of contained uranium-235 of any enrichment in the form of fission chambers, flux foils and fueled experiments, all used in connection with operation of the facility; Additionally, the licensee is not requesting the previous provision listed as 2.B.(4).
56-
TECHNICAL SPECIFICATIONS FOR THE PURDUE UNIVERSITY REACTOR, PUR-1 DOCKET NUMBER 50-182 FACILITY LICENSE NO. R-87 PREPARED BY:
C.H. Townsend West Lafayette, IN 4 7907 September 2016
TABLE OF CONTENTS Page
1.
DEFINITIONS................................................................................................ 4
2.
SAFETY LIMIT AND LIMITING SAFETY SYSTEM SETTING..................... 9 2.1 Safety Limit............................................................................................. 9 2.2 Limiting Safety System Setting............................................................... 9
3.
LIMITING CONDITIONS FOR OPERATION............................................... 10 3.1 Reactivity Limits.................................................................................... 10 3.2 Reactor Safety System......................................................................... 11 3.3 Primary Coolant Conditions.............................................................. 1443 3.4 Confinement......................................................................................... 14
\\
3.5 Limitations on Experiments.................................................................. 15 Fuel Parameters.......................................................................................... 16 3.6................................................................................................................... 16
4.
SURVEILLANCE REQUIREMENTS........................................................... 17 4.1 Reactivity Limits.................................................................................... 17 4.2 Reactor Safety System......................................................................... 17 4.3 Primary Coolant System............................................................ '........... 19 4.4 Confinement................................................................................ *......... 19 4.5 Experiments.................. '.**********************************************************************_* 20 4.6 Fuel Parameters........................................... -..'...................................... 20 4.7 Effluents............................................................................................... 21 I
5.
DESIGN FEATURES................................................................................... 22 5.1 Site Description..................................................................................... 22 PUR-1 Technical Specifications 2
Amendment No. 13
5.2 Reactor Coolant System....................................................................... 23 5.3 Reactor Core and Fuel......................................................................... 24 5.4 Fuel Storage......................................................................................... 25
6.
ADMINISTRATIVE CONTROLS................................................................. 26 PUR-1 Technical Specifications 3
Amendment No. 13
1.
DEFINITIONS The following frequently used terms are to aid in the uniform interpretation of these specifications:
1.1 Channel - A channel is the combination of sensor, line, amplifier, and output devices that are connected for the purpose of measuring the value of a parameter.
1.2 Channel Calibration - A channel calibration is an adjustment of the channel such that its output corresponds, within acceptable range and accuracy, to known values of the parameter which the channel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip, and is deemed to include a channel test.
1.3 Channel Check - A channel check is a qualitative verification of acceptable performance by observation of channel behavior. This verification may include comparison of the channel with other independent channels or methods of measuring the same variable.
1.4 Channel Test - A channel test is the introduction of a simulated signal into a channel to verify that it is operable.
1.5 Confinement - Confinement is an enclosure of the overall facility that is designed to limit the release of effluents between the enclosure and its external environment through controlled or defined pathways.
1.6 Core Configuration -
The core configuration includes the number, type, or arrangement of fuel assemblies (elements), reflector elements, reflector element configuration, and regulating/control rods occupying the core grid.
1. 7 Core Experiment - A core experiment is one placed in the core, in the graphite reflector, or within six inches (measured horizontally) of the reflector. This includes any experiment in the-pool directly above or below the core.
1.8 Direct Supervision - In visual and audible contact.
1.9 Excess reactivity-Excess reactivity is that amount of reactivity that would exist if all_ control rods were moved to the maximum reactive condition from the point where the reactor is exactly critical (kett = 1) at reference core conditions or at a specified set of conditions.
1.10 Experiment - Any operation, hardware, or target (excluding devices such as detectors, foils, etc.) -that is designed -to investigate non-routine reactor characteristics or that is intended for irradiation within the pool, on or in a beam port or irradiation facility. Hardware rigidly secured to a core or shield structure so as to be a part of its design to carry out experiments is not normally considered an experiment.
1.11 Experimental Facility-Experimental facilities are:
PUR-1 Technical Specifications 4
Amendment No. 13
a. those regions specifically designated as locations for experiments or
b. systems designed to permit or enhance the passage of a beam of radiation to another location.,
1.12 Experiment With Movable Parts (Secured or Nonsecured) - An experiment with movable parts is an experiment that contains parts that are intended to be moved while the reactor is operating.
1.13 Explosive Material - Explosive material is any solid or liquid which is categorized as a Severe, Dangerous, or Very Dangerous Explosion Hazard in "Dangerous Properties of Industrial Materials" by N. I. Sax, Tenth Ed. (2000), or is given an Identification of Reactivity (Stability) index of 2, 3 or 4 by the National Fire Protection Association in its publication 704, "Identification System for Fire Hazards of Materials."
1.14 Fueled Experiment -A fueled experiment is any experiment planned for irradiation of uranium 233, uranium 235, plutonium 239,, or plutonium 241.
1.15 License -
The written authorization, by the US NRC, for an individual or organization to carry out the duties and responsibilities associated with a personnel position, material, or facility requiring licensing.
1.16 Licensed - See licensee.
1.17 Licensee - An individual or organization holding a license.
1.18 Measured Value - The measured value is the value of a parameter as it _appears at the output of a channel.
1.19 Movable Experiment - A movable experiment is one where it is intended that all or part of the experiment may be moved in or near the core or into and out of the reactor while the reactor is operating.
1.20 New Experiment - A new experiment is one whose nuclear characteristics have not been experimentally determined.
1.21 Non-secured Experiment-See Unsecured Experiment.
1.22 Operable - A system or component is operable when it is capable of performing its intended function in a normal manner.
1.23 Operating -A system or component is operating when it is performing its intended function.
1.24 Pool Experiment - A pool experiment is one positioned within the pool more than six inches (measured horizontally) from the graphite reflector.
1.25 Power Level - There are three important and separately defined power levels.
PUR-1 Technical Specifications 5
Amendment No. 13
a. Instantaneous Power Level shall be the power level of the reactor at any
_;given moment, as indicated by the reactor instrumentation.
b. The Operating Level shall be the power level from which setpoints for scram and setback shall be calculated. The Operating power level shall be 10 kW or less.
c. The Maximum Power Level shall be the maximum instantaneous power level allowed by the PUR-1 License. The Maximum Power Level shall be 12 kW, which shall not be exceeded.
1.26 Protective action - Protective action is the initiation of a signal or the operation of equipment within _the reactor safety system in response to a parameter or condition of the reactor facility having reached a specified limit.
1.27 Reactivity worth of an experiment - The reactivity worth of an experiment is the value of the reactivity change that results from the experiment being inserted or removed from its intended position.
1.28 Reactor-Facility - The reactor facility is that portion of the ground floor of the Duncan Annex of the Electrical Engineering Building occupied by the School of Nuclear Engineering used for activities associated with the reactor.
1.29 Reactor Operating - The reactor is operating whenever it is not secured or shut down.
1.30 Reactor Operator -An individual who is licensed to manipulate the controls of the reactor.
1.31 Reactor Safety System - The reactor safety system is that combination of measuring channels and associated circuitry which forms the automatic protective system of the reactor, or provides information which requires manual protective action to be initiated.
1.32 Reactor Secured -A reactor is secured when
a. Either there is insufficient moderator available in the reactor to attain criticality or there is insufficient fissile material present in the reactor to attain criticality under optimum available conditions of moderation and reflection
b. Orthe following conditions exist:
1.
Both shim-safeties and the regulating rod shall be fully inserted
2.
Electrical power to the control rod circuits shall be switched off PUR-1 Technical Specifications 6
Amendment No. 13
3.
The reactor key shall be out of the key switch and under control of a licensed operator or locked in an approved location
4.
No work shall be in progress involving core fuel, core structure, installed control rods, or control rod drives unless Jhey are physically decoupled from the control rods
5.
No experiments shall be moved or serviced that have, on movement, a reactivity worth exceeding the maximum value allowed for a single experiment 1.33 Reactor Shutdown - That subcritical condition of the reactor where the negative reactivity, with or without experiments in place, is equal to or greater than the shutdown margin.
1.34 Readily Available on Call - Readily available on call shall mean the licensed senior operator shall be within a reasonable driving time (1/2 hour) or less than 15 miles from the reactor building, and the operator on duty is currently informed, and can rapidly contact the senior reactor operator by phone.
1.35 Reference core condition - The condition of the core when it is at ambient temperature (cold) and the reactivity worth of xenon is negligible {<0.003 !::.k/k).
1.36 Removable Experiment -
A removable experiment is any experiment, experimental facility, or component of an experiment, other than a permanently attached appurtenance to the reactor system, which can reasonably be anticipated to be moved one or more times during the life of the reactor.
1.37 Rod. control -A control rod is a device fabricated from neutron-absorbing material that is used to establish neutron flux changes and to compensate for routine reactivity losses. A control rod can be coupled to its drive unit allowing it to perform a safety function when the coupling is disengaged.
1.38 Rod. regulating - The regulating rod is a low worth control rod used primarily to maintain an intended power level that need not have scram capability. Its position may be varied manually or by a servo-controller.
1.39 Rod. Shim-Safety - The control rods used in PUR-1 as described in the definition for Rod, control.
1.40 Secured Experiment -Any experiment, experimental facility, or component of an experiment is deemed to be secured, or in a secured position, if it is held in a stationary position relative to the reactor by mechanical means. The restraining fortes must be substantially greater than those to which the experiment might be subjected by hydraulic, pneumatic, buoyant, or other forces which are normal to the operating environment of the experiment, or by forces which can arise as a result of credible malfunctions.
PUR-1 Technical Specifications 7
Amendment No. 13
1.41 Senior Reactor Operator - An individual who is licensed to direct the activities of reactor operators. Such an individual is also a reactor operator.
1.42 Shall. should. and may - The word "shall" is used to denote a requirement; the word "should" is used to denote a recommendation; and the word "may" is used to denote permission, neither a requirement nor a recomm-endation.
1.43 Shutdown Margin - The shutdown margin is the minimum shutdown reactivity necessary to provide confidence that the reactor can be made subcritical by means of the control an9 safety systems starting from any permissible operating condition and with the most reactive rod in the most reactive position, and the n-onscramable rods in their most reactive posftions and that the reactor will remain subcritical without further operator action.
1.44 Surveillance and Test Intervals - These are intervals established for periodic surveillance and test actions. Established intervals shall be maintained on the average. Maximum intel"Vals are allowed to provide operational flexibility, not to reduce frequency.
1.45 Tried Experiment - A tried experiment is:
a. An experiment previously performed in this facility, or
b. An experiment of approximately the same nuclear characteristics as an experiment previously tried. These nuclear characteristics include but are not limited to neutron activation cross-sections, absorption cross-sections, and moderating ability.
1.46 True Value -The true value of a parameter is its exact value at any instant.
1.47 Unscheduled Shutdown -An unscheduled shutdown is defined as any unplanned shutdown of the reactor by actuation of the reactor safety system, operator error, equipmenf malfunction, or a manual shutdown in response to conditions that could adversely affect safe operation, not including shutdowns that occur during testing or checkout operations.
1.48, Unsecured Experiment - Any experiment, experimental facility, or component of an experiment is considered to be unsecured when it is not secured as defined in this section.
PUR-1 Technical Specifications 8
Amendment No. 13
2.
SAFETY LIMIT AND LIMITING SAFETY SYSTEM SETTING 2.1 Safety Limit Applicability - This specification applies to the temperature of the reactor fuel and cladding under any condition of operation.
Objective - The objective is to ensure fuel cladding integrity.
Specification - The fuel and cladding temperatures shall not exceed 530°C (986°F).
Basis - Safety limits for nuclear reactors are limits/upon important process variables that are necessary to reasonably protect the integrity of certain physical barriers that guard against the uncontrolled release of radioactivity. The principal physical barrier is the fuel cladding.
In the Purdue University Reactor, the first and principal barrier protecting against release of radioactivity is the cladding of the fuel plates. The 6061 aluminum alloy cladding of the LEU fuel plates has an incipient melting temperature of 582°C.
However, measurements (NUREG-1313) on irradiated fuel plates have shown that fission products are first released near the blister temperature (-550°C) of the cladding. To ensure that the blister temperature is never reached, NUREG 1537 concludes that 530°C is an acceptable fuel and cladding temperature limit not to be exceeded under any condition of operation.
2.2 Limiting Safety System Setting Applicability - This specification applies to the reactor power level safety system setting for operation.
Objective - The objective is to assure that the safety limit is not exceeded.
Specification - The measured value of the power level scram shall be no higher than 12.0 kW.
Basis - The LSSS has been chosen to assure that the automatic reactor protective system will be actuated in such a manner as to prevent the safety limit from being exceeded during the most severe expected abnormal condition.
The function of the LSSS is to prevent the temperature of the reactor fuel and cladding from reaching the safety limit under any condition. of operation. During steady-state operation, a power level of 98.6 kW is required to initiate the onset of nucleate boiling.
This is far higher than the maximum power of 18 kW, which allows for 50% instrument uncertainties in measuring power level.
For the transients that were analyzed, the temperature of the fuel and cladding reach maximum temperatures of 43.20°C, assuming reactor trip at 18 kW after failure of the first trip. This temperature is far below the safety limit of 530°C.
PUR-1 Technical Specifications 9
Amendment No. 13
3.
LIMITING CONDITIONS FOR OPERATION 3.1 Reactivity Limits Applicability - These specifications apply to the reactivity conditions of the reactor, and the reactivity worths of control rods and experiments Objective - The objective is to assure that the reactor can be shut down at all times, that the safety limits will not be exceeded, and that operation is within the limits analyzed in the SAR.
Specification - The reactor shall not be operated unless the following conditions exist:
a. The shutdown margin, relative to the reference core condition with the most reactive shim rod fully withdrawn, and the regulating rod fully withdrawn shall be at least 0.010 ~k/k.
b. The reactor shall be subcritical by more than 0.03 ~k/k during core loading changes.
c. No shim-safety rod shall be removed from the core if the shutdown margin is less than 0.01 ~k/k with the remaining shim-safety rod fully withdrawn.
d. The reactor shall be shutdown if the maximum positive excess reactivity of the core and any installed experiment exceeds 0.006 ~k/k.
e. The absolute value of the reactivity worth of each experiment shall be limited as follows:
Experiment Movable Unsecured Secured Maximum Reactivity Worth 0.003 ~k/k 0.003 ~k/k 0.004 ~k/k
f.
The sum of the absolute value of the total worth of all movable and unsecured experiments shall not exceed 0.003 ~k/k.
g. The sum of the absolute value of the total worth of all secured experiments shall not exceed 0.005 ~k/k.
Bases - The shutdown margin required by Specification 3.1.a assures that the reactor can be shut down from any operating condition and will remain shut down even if the control rod of the highest reactivity worth should be in the fully withdrawn position.
Specifications 3.1.b and 3.1.c provide assurance that the core will remain subcritical during loading changes and shim-safety rod maintenance or inspection.
Specification 3.1.d limits the allowable excess reactivity to the value assumed in the SAR. This limit assures that the consequences of reactivity transients will not be PUR-1 Technical Specifications 10 Amendment No. 13
increased relative to transients previously reviewed, and assures reactor periods of sufficient length so that the reactor may be shutdown without exceeding the safety limit.
Specification 3.1.e limits the reactivity worth of secured experiments to values of reactivity which, if introduced as a positive step change, are calculated not to cause fuel melting. This specification also limits the reactivity worth of unsecured and movable experiments to values of reactivify which, if introduced as a positive step change, would not cause the violation of a safety limit. The manipulation of experiments worth up to 0.003 Ak/k will result in reactor periods longer than 9 seconds. These periods can be readily compensated for by the action ofthe safety system without exceeding any safety limits.
A limitation of 0.003 Ak/k for the total reactivity worth of all movable and unsecured experiments provides assurance that a common failure affecting all such experiments cannot result in an accident of greater consequences than the maximum credible accident analyzed in the HSR.
Specification 3.1.g along with 3.1.a assures that the reactor is capable of being shut down in the event of a positive reactivity insertion caused by the flooding of an experiment.
3.2 Reactor Safety System Applicability - This specification applies to the reactor safety system and other safety-related instrumentation.
Objective - The objective is to specify the lowest acceptable level of performance or the
minimum number of acceptable components for the reactor safety system and other safety related instrumentation.
Specification - The two shim-safeties shall not be moved more than 6 cm from the fully inserted position unless the following conditions are met:
a. The reactor safety channels and safety-related instrumentation shall be operable in accordance with Tables I and II including the minimum number of channels and the indicated maximum or minimum set points.
b. Both shim-safety rods and the regulating rod shall be operable.
c. The time from the initiation of a scram condition in the scram circuit until the shim-safety rod reaches the rod lower limit switch shall not exceed one second.
d. The pool top radiation monitor shall be capable of indicating an alarm to off-site reactor staff when a high limit is reached and the reactor has been secured. The alarm may be out of service up to thirty days. Loss of functionality beyond thirty days shall require c;i visual pool level inspection in intervals of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , not to exceed 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
PUR-1 Technical Specifications 11 Amendment No. 13
TABLE I. SAFETY CHANNELS REQUIRED FOR OPERATION Minimum Number Channel Required Log count rate 1 (a) and period Log N and period Linear Safety Manual Scram (console)
(hallway) 1 (b) 1 1 (b) 1 1
Setpoint (c)(d) 2 cps or greater 12 sec. or greater 7 sec. or greater 15 sec. or greater 12 sec. or greater 7 sec. or greater 7 sec or greater 15 sec or greater 12kW, 120% Operating power level, or less 0% Selected Range, or greater 110% Selected Range or less 120% Selected Range or less 11kW,110%
Operating power level, or less 12 kW, 120%
Operating power level, or less (a) Not required after Log N-Period channel comes on scale.
(b) Required to be operable but not on scale at startup.
Function 2 cps rod withdrawal interlock Setback Slow Scram Rod withdrawal interlock Setback Slow Scram Fast Scram Rod withdrawal interlock Slow Scram Setback Setback Slow Scram Setback Fast Scram Slow Scram Slow Scram (c) All percentage based setpoints shall be tripped when the measured value is greater than or equal to the specified value. Period and counts per second (cps) setpoints are at values less than or equal to the specified value. Exception: Trip point for 0% shall happen as the value goes from the positive to negative value.
(d) Setbacks shall be set such that they will be initiated prior to a Scram PUR-1 Technical Specifications 12 Amendment No. 13
TABLE II. SAFETY-RELATED CHANNELS (AREA RADIATION MONITORS)
Minimum Number Channel Required<e>
Setpoint Function Pool top monitor 1
50 mR/hr or 2x full power Slow Scram background Water process 1
7 % mR/hr Slow Scram Console Monitor 1
7 % mR/hr Slow Scram Continuous air sampler 1
Stated on sampler Air sampling (e)
For periods of one week or for the duration of a reactor run, a radiation monitor may be replaced by a gamma sensitive instrument which has its own alarm and is observable by the reactor operator.
Bases - The neutron flux level scrams provide redundant automatic protective action to prevent exceeding the safety limit on reactor power, and the period scram conservatively limits the rate of rise of the reactor power to periods which are manually controllable without reaching excessive power levels or fuel temperatures.
The rod withdrawal interlock on the Log Count Rate and Period Channel assures that the operator has a measuring channel operating and indicating neutron flux levels during the approach to criticality.
The manual scram button and the "reactor on" key switch provide two methods for the reactor operator to manually shut down the reactor if an unsafe or abnormal condition should occur and the automatic reactor protection does not function.
The use of the area radiation monitors (Table II) will assure that areas of the Purdue University R~actor (PUR-1) facility in which a potential high radiation area exists are monitored. These fixed monitors initiate a scram whenever the preset alarm point is exceeded to* avoid high radiation conditions as well as alert facility personnel when the reactor has been secured and an elevated radiation level exists.
Specifications 3.2.b and 3.2.c assure that the safety system response will be consistent with the assumptions used in evaluating the reactor's capability to withstand the maximum credible accident.
In specification 3.2.c. the rod lower limit switches are positioned to measure, as close as possible, the fully inserted position.
Shielding from radiation is one of the primary reasons for the pool's level. An offsite alarm from the pool top radiation monitor alerts facility staff of a rising radiation level which must be mitigated or otherwise addressed and this is addressed in 3.2.d.
PUR-1 Technical Specifications 13 Amendment No. 13
3.3 Primary Coolant Conditions Applicability - This specification applies to the limiting conditions for reactor operation for the primary coolant.
Objective - The objective is to assure a compatible environment, adequate shielding, and a continuous coolant path for the reactor core.
Specification -
a.
The primary coolant resistivity shall be maintained at a value greater than 330,000 ohm-cm.
b.
The primary coolant shall be maintained at least 13 feet above the core whenever the reactor is operating. The primary coolant shall be maintained at least 13 feet above the top of the core or at a level sufficient for the pool top radiation monitor to indicate less than 1 mRem/hour during non-operational periods.
c.
The primary coolant (bulk pool volume) shall be maintained at or below 30 °C while the reactor is operating.
d.
The primary coolant radiation levels shall not exceed the levels for water in 10 CFR 20 Appendix B, Table 2.
Bases - Experience at the PU R-1 and other facilities has shown that the maintenance of r:frimary coolant system water quality in the ranges specified in specification 3.3.a will minimize the amount and severity of corrosion of the aluminum components of the primary coolant system and the fuel element cladding.
The height of water in specification 3.3.b is enough to furnish adequate shielding as well as to guarantee a continuous coolant path.
Ma_iQtaining the primary_ c90Jant temperature in Specification 3.3.c will ensure the margin to the onset of nucleate boiling is maintained and analyses shown in the Safety Analysis Report remain valid.
Limiting the amount of radioactivity in the primary coolant minimizes the health risk to the public as well as to facility personnel.
3.4 Confinement Applicability - This specification applies to the integrity of the reactor room.
Objective - The objective is to limit and control the release of airborne radioactive material from the reactor room.
Specification -
a. During reactor operation and when radioactive material is being handled with potential for airborne release, the following conditions shall be met:
PUR-1 Technical Specifications 14 Amendment No. 13
1. The reactor room shall be maintained at a negative pressure of at least 0.05 inches of water with the operation of the room exhaust fan.
2. All exterior doors in the reactor room shall remain closed except as required for personnel, equipment, or materials access.
b. All inlet and exhaust air ducts and the sewer vent shall contain a HEPA filter or its equivalent.
c. Dampers in the ventilation system inlet and outlet ducts shall be capable of being closed.
d. Concentration of Ar-41 shall not exceed 2.08 x 10-7 µCi/cm3 at the top of the confinement exhaust stack.
Bases - The PUR-1 does not rely on a containment building to reduce the levels of airborne-radioaetive-rfraforial released t6 the -envirorimenf in Hie event of the maximum hypothetical accident. However, in the even_t of such an accident, a significant fraction of the airborne material will be confined within the reactor room, and the specifications stated above will further reduce the release to the environment.
The limit on the concentration of Argon at the top of the confinement exhaust stack is the maximum theoretical concentration of the isotope and therefore a fan malfunction ventilating the room would be the only way to violate this technical specification. It is validated by the dose readings obtained through the effluent surveillances in section 4.7.
3.5 Limitations on Experiments Applicability-This specification applies to experiments installed in the reactor and its experimental facilities.
Objective - The objective is to prevent damage to the reactor or excessive release of radioactive materials in the event of an experiment failure, and to assure the safe operation of the reactor.
Specification - The reactor shall not be opera~ed unless the following conditions are met:
a. All experiments shall be constructed of material which will be corrosion resistant for the duration of their residence in the pool.
b. All experiments and experimental procedures shall receive approval by the Committee on Reactor Operations.
c. Known explosive materials shall not be placed in the reactor pool.
d. No experiment shall be placed in the reactor or pool that interferes with the safe operation of the reactor.
PUR-1 Technical Specifications 15 Amendment No. 13
e. Any failure of an experiment shall not have a consequence that could exceed dose limits as set forth in 10 CFR Part 20, as analyzed and approved by the Reactor Supervisor and the Committee on Reactor Operations.
f.
A fueled experiment shall not produce more than 0.5 Curies of radio-iodine.
Bases - Specification 3.5.a through 3.5.f are intended to reduce the likelihood of damage to reactor components and/or radioactivity releases resulting from experiment failure and serve as a guide for the review and approval of new experiments by the facility personnel and the Committee on Reactor Operations.
Limiting the amount of radio-iodine levels in a fueled experiment will ensure that the Maximum Hypothetical Accident analyzed in the Safety Analysis Report remains the bounding incident which could occur at the PUR-1.
3.6 Fuel Parameters Applicability-This specification applies to fuel plates installed in the reactor and in use during the previous surveillance period.
Objective - The objective is to prevent damage to the reactor or excessive release of radioactive materials in the event of a fuel cladding failure, and to assure the safe operation of the reactor.
Specification - The reactor shall only be operated when the following specifications have been met:
a.
The inspection of fuel assemblies shall be performed to identify any abnormal or previously undocumented defect present on a fuel plate. These defects may include but are not limited to blistering of the cladding on the fuel plate from elevated temperatures beyond the design of the cladding, deep scratches or gouges on the plate due to debris in the coolant flowing along the face, scratches on the edges of the plate due to insertion and removal from the assembly, discc:>loration from the deposition of particulates within the coolant or corrosion of the plate itself.
b.
The reactor shall not operate with fuel plates that have been determined to be unsound for u~e as outlined above in 3.6.a. These plates shall be removed from service and the manufacturer consulted to determine possible causes.
Bases - The fuel parameter Limiting Condition on Operation is intended to limit the possibility radioactivity releases resulting from fuel failure by identifying issues prior to potential release.
PUR-1 Technical Specifications 16 Amendment No. 13
4.
SURVEILLANCE REQUIREMENTS 4.1 Reactivity Limits Applicability - This specification applies to the surveillance requirements for reactivity limits.
Objective - The objective is to ensure that the reactivity limits of Specification 3.1 are not exceeded.
Specification -
a. The shim-safety rod _reactivity worths shall be measured and the shutdown margin calculated biennially with no interval to exceed 2% years and whenever a core configuration is loaded for which shim-safety rod worths have not been measured. This may be deferred with CORO approval during any extended reactor shutdown. Additionally, if a new rod is used, its worth must be measured on the first start-up following installation. In the case of a deferred measurement, the measurement must be performed prior to resuming routine reactor operations.
b. The shim-safety rods shall be visually inspected biennially with no interval to exceed 2% years, which may be deferred with CORO approval during any reactor shutdown. If. the rod is found to be deteriorated, it shall be replaced with a rod of approximately equivalent or greater worth, meeting the limiting conditions of operation specified in 3.1.
In the case of a deferred measurement, the measurement must be performed prior to resuming routine reactor operations.
c. The reactivity worth of experiments placed in the PUR-1 shall be measured during*the first startup subsequent to the experiment's insertion and shall be verified if core configuration changes cause increases in experiment reactivity worth which may cause the experiment worth to exceed the values specified in Specification 3.1 Bases - Specification 4.1.a will assure that shim-safety rod reactivity worths are not degraded or changed by core manipulations which cause these rods to operate in regions where their effectiveness is reduced.
The boron stainles_s steel shim-safety rods have oeen in use at the PUR-1 since 1962, and over this period of time, no cracks or other evidence of deterioration have been observed. Based on this performance and the experience of other facilities using similar shim-safety rods, the specified inspection times are considered adequate to assure that the control rods will not fail.
4.2 Reactor Safety System Applicability - This specification applies to the surveillance of the reactor safety system.
PUR-1 Technical Specifications 17 Amendment No. 13
Objective - The objective Js to assure that the reactor safety system is operable as required by Specification 3~2 Specification -
a. A channel calibration of the reactor safety channels as described in Table I shall be performed as follows:
1.
2.
An electronic calibration shall be performed annually, with no interval __ to __ ~X91?~_q 15 _month$. __ The el~ctror:iic calibration may be deferred with_ CORO approval during periods of reactor shutdown, but shall be performed prior to startup.
A power calibration by foil activation shall be performed annually, with no interval to exceed 15 months. The power calibration may be deferred with CORO approval during periods of reactor shutdown, but shall be performed prior to startup.
b. A channel check on the radiation monitoring equipment shall be completed daily during periods when the reactor is in operation. Calibration of the Safety-Cha_nnels_ specified in Table II and hand held radiation survey instruments shall be performed annually, with no interval to exceed 15 months. *Calibration ma}i be deferred with CORO approval during periods of reactor shutdown, but shall be performed prior to startup.
c. Shim-safety rod drop times shall be measured annually, with no measurement's interval to exceed 15 months. These drop times shall also be measured prior to operation following maintenance which could affect the drop time or cause movement of the shim-safety rod control assembly.
Drop times may be deferred with CORO approval during periods of reactor shutdown, but shall be performed prior to startup.
d. A channel check of each of the Scram capabilities specified in Table I shall be performed prior to each day's startup.
e. A channel check of the pool top radiation monitoring equipment's off-site alarm capability shall be done biannually, not to exceed 7 % months.
Bases - A test of the safety system channels prior to each startup will assure their operability, and annual calibration will detect any long-term drift that is not detected by normal intercomparison of channels. The channel check of the neutron flux level channel will assure that changes in core-to-detector geometry or operating conditions will not cause undetected changes in the response of the measuring channels.
Area monitors will give a clear indication when they are not operating correctly. In addition, the operator routinely records the readings of these monitors and will be aware of any reading which indicates loss of function.
PUR-1 Technical Specifications 18 Amendment No. 13
The area monitoring system employed at the PUR-1 has exhibited very good stability over its lifetime, and annual calibration is considered adequate to correct long-term drift.
The measured drop times of the shim-safety rods have been consistent since the PUR-1 was built. An annual check of this parameter is considered adequate to detect operation with materially changed drop times. Binding or rubbing caused by rod misalignment could result from maintenance; therefore, drop times will b~ checked after such maintenance.
A daily check of the scram functionality ensures functionality of the system.
4.3
-Primary Coolant System Applicability - This specification applies to the average surveillance schedules of the primary coolant system.
Objective - The objective is to assure high quality pool water, adequate shielding, and to detect the release of fission products from fuel elements.
Specification -
a.
The conductivity of the primary coolant shall be recorded monthly, not to exceed six weeks. This cannot be deferred during reactor shutdown.
b.
The primary coolant shall be sampled monthly, not to exceed six weeks, and analyzed for gross alpha and beta activity. This cannot be deferred during reactor shutdown.
c.
During reactor shutdown, the primary coolant level or radiation level shall be monitored monthly with an interval not to exceed six weeks. Primary coolant height shall be measured prior to reactor operation.
d.
The Primary Coolant temperature shall be recorded in the log book at no interval to exceed four hours if any shim-safety or regulating rod is at a height greater than 6 cm.
1 Bases - Monthly surv~illar_tce of pool water quality provides assurance conductivity changes will be detected before signific(int corrosive damage could occur.
)
When the reactor pool water is at a height of 13 feet above the core, adequate shielding during operations is assured. Experience has shown that approximately 35-40 gallons of water will evaporate weekly an-d weekly water make-up is sufficient to maintain the reactor-pool wate-r he-ight. -Analysis has shown radiation levels to remain sufficiently low with excessive water loss during non-operational periods.
Analysis of the reactor water for gross alpha and beta activity assures against undetected leaking fuel assemblies.
4.4 Confinement Applicability - This specification applies to the surveillance requirements for maintaining the integrity of the reactor room and fuel clad.
PUR-1 Technical Specifications 19 Amendment No. 13
Objective - The objective is to assure that the integrity of the reactor room arid the fuel clad is maintained, by specifying average surveillance intervals.
Specification -
a. The negative pressure of the reactor room shall be recorded weekly.
b. Operation of the inlet and outlet dampers shall be checked semiannually, with no interval to exceed 7 1/2 months.
c. Operation of the air conditioner shall be checked semiannually, with no interval to exceed 7 1/2 months.
Bases - Specification a, b, and c check the integrity of the reactor room, and d the integrity of the fuel clad. Based upon past experience these intervals have been shown to be adequate for ensuring the operation of the systems affecting the integrity of the reactor room and fuel clad.
4.5 Experiments Applicability - This specification applies to the surveillance of limitations on experiments.
Objective -
To assure compliance with the provision of the utilization license, the Technical Specifications, and 10 CFR Parts 20 and 50.
Specification - No experiments shall be performed unless:
a. It is a tried experiment.
b. The experiment has been properly reviewed and approved according to Section 6 of the technical specifications.
1. Proposed experiments shall be approved by the Committee on Reactor Operations
2. Submitted proposed experiments shall provide a comprehensive list of steps to be performed, quantities to be measured, hazards to be considered, limiting initial conditions of the reactor, and required available personnel.
Bases - The basis for this specification is to ensure the safety of the reactor and associated components, personnel, and the public by verification of proper review and approval of experiments as specified in Section 6 of these technical specifications.
4.6 Fuel Parameters Applicability-This specification applies to the surveillance requirements for fuel integrity.
Objective - The objective is to assure that the fuel clad remains unblemished and there has been no release of radioactivity to the reactor coolant or facility.
PUR-1 Technical Specifications 20 Amendment No. 13
Specification - Representative fuel plates shall be inspected annually, with no interval to exceed 15 months. Representative is set forth to mean at least one plate from the assembly expected to have the highest burn as well as a plate from one of the 12 remaining, non-control assemblies.
Bases - Specification 4.6 will ensure reactor fuel integrity is not compromised. The inspection period is set forth to verify the integrity of the fuel cladding thereby ensuring there are no unexpected releases of fission products exposing facility workers or members of the public. Inspection of an assembly from the highest power region (as outlined in the PUR-1 SAR) ensures those __ plates _under the largest thermal stress are considered. Inspection of another assembly ensures that a single plate passing inspection does not provide a single false negative data point representing the entire core. Non-control assemblies are chosen to inhibit undue burden on the facility.
4.7 Effluents Applicability - This specification applies to the surveillance requirements for radioactive effluents which may leave the facility through the confinement system.
Objective - The objective is to assure requirements set forth in 10 CFR 20.11 O(d) and 10 CFR 20.1301 are not exceeded and public safety is maintained.
Specification -
a.
Dosimetry shall be placed at the following locations
1. The location inside the reactor room which represents the hypothetical minimum distance a member of the public could reach to the reactor pool.
2. At the exhaust location of the reactor facility which is representative of effluent release from the reactor facility.
b.
Dosimetry shall be changed out according to the guidance of the Purdue Radiological Management on the same time period as facility personnel or semiannually, not to exceed 7 % months, whichever is lesser.
Bases - Specification 4. 7 will ensure that the dose given to member of the public is measured to be below those set forth in 10 CFR 20.110(d) and 10 CFR 20.1301.
PUR-1 Technical Specifications 21 Amendment No. 13
5.
DESIGN FEATURES 5.1 Site Description Applicability-This specification applies to the general design and areas under which the PUR-1 Technical Specifications shall have jurisdiction.
Objective - This section is to clarify those areas which are involved with the PUR-1 Facility.
Specifications -
a.
The reactor shall be located on the ground floor of the Duncan Annex of the Electrical Engineering Building, Purdue University, West Lafayette, Indiana.
b.
The School of Nuclear Engineering shall control approximately 5000 square feet of the Duncan Annex ground floor, which includes the reactor room.
Access to the Nuclear Engineering controlled area shall be restricted except when classes are held there.
c.
The licensed areas shall include the reactor room, and a fuel storage room.
Both of these areas shall be restricted to authorized personnel, or those escorted by authorized personnel.
d.
The reactor room shall remain locked at all times except for the entry or exit of authorized personnel or those escorted by authorized personnel, equipment, or materials.
e.
The PUR-1 reactor room shall be a closed room designed to restrict leakage.
f.
The minimum free volume of the reactor room shall be approximately 15,000 cubic feet.
g.
The ventilation system shall be designed to exhaust air or other gases from the reactor room through an exhaust vent at a minimum of 50 feet above the ground.
h.
Openings into the reactor room shall consist of no more than the following:
1. Three personnel doors
2. One door to a storage room with no outside access.
3. Air intake
4. Air exhaust
5. Sewer vent Bases The bases for the above specifications are the naming of the buildings, city and state at the time of the enactment of this amendment to the PUR-1 Technical Specifications. The PUR-1 Technical Specifications 22 Amendment No. 13
access to the restricted areas is controlled to inhibit the removal of materials, information, or other import aspects of the facility to maintain confidence in safe operation under which the Safety Analysis was completed.
The volume of the reactor room and its leakage properties are so set forth to further ensure safety to facility workers and the general public is maintained during all operational circumstances.
5.2 Reactor Coolant System Applicability - This specification outlines the make-up and properties of the PUR-1 Reactor Coolant System.
Objective - By outlining the systems which are required to be in place during operations, the validity of the Safety Analysis Report calculations is ensured.
Specifications -
a.
Primary Cooling System - The PUR-1 primary cooling system shall be a pool containing approximately 6,400 gallons of water.
b.
Process Water System - The process water system shall be assembled in one unit and contain a pump, filter, demineralizer, valves, flow meters, and a heat exchanger (see 5.2.d). The demineralizer shall contain a removable cartridge that is monitored continuously for radioactivity buildup.
c.
Primary Coolant Makeup Water System - Makeup water for the pool shall be taken batchwise from the Purdue University water line and passed through the demineralizer enroute to the pool. A vacuum breaker shall exclude any
-possibility otsiphoning pool water into the supply line. The pool makeup water system, in addition to the demineralizer, also shall include a normally closed manual shutoff and throttle valve and a check valve.
d.
Primary Coolant Chiller System - The chiller shall be designed with three
-loops. Pool water shall pass through the primary loop, a Freon refrigerant in the secondary loop, and water from the building water supply shall be used to remove heat, which shall then be discharged to the building sewer system.
The heat-removal capacity of the heat exchanger shall be 10.5 kW or greater.
Bases - The basis of having a reactor pool with the listed volume is to ensure there is an adequate cooling path for the PUR-1 core as well as providing a shield to direct shine from the reactor's standard operation. The make-up water to the pool has a set of processing and monitoring equipment to ensure the long-term operation of the facility and fuel integrity by suppressing corrosion and other effects due to submersion in water.
This system shall limit, -by the u-se of filters and ion-exchange resin, the aluminum corrosion rate, corrosion product buildup, and neutron activation of impurities in the coolant.
The chiller system must maintain the reactor pool temperature to be lower than the specified value while operating to-ensure the margin to the onset of nucleate boiling does PUR-1 Technical Specifications 23 Amendment No. 13
not go beyond the values determined in the PUR-1 Safety Analysis Report. It shall be capable of maintain the reactor pool temperature at or below 30°C during steady state operation at 10 kW.
5.3 Reactor Core and Fuel Applicability-This specification outlines the limits on the design and loading of the PUR-1 Core.
Objective - The standard loading, fuel type, and inspection period is given to ensure the
~shutdown margin, accident analysis, and operational characteristics are maintained and remain valid.
Specifications -
a.
The fuel assemblies shall be MTR type consisting of U3Si2-Al, 6061 Aluminum clad plates enriched up to 20% in the U-235 isotope.
b.
A standard fuel assembly shall consist of up to 14 fuel plates containing a maximum of 180 grams of U-235.
c.
A control fuel assembly shall consist of up to 8 fuel plates containing a maximum of 103 grams of LJ;..235.
d.
Partially loaded fuel assemblies in which some of the fuel plates are replaced by aluminum plates containing no uranium may be used.
e.
The core configuration shall consist of 13 standard fuel assemblies as described in b, and 3 control fuel assemblies as described in c.
f.
The core shall include two shim-safety rods and one regulating rod placed within *a control assembly. The two shim-safeties shall be made of solid.
borated 304 stainless steel. The Regulating Rod shall be stainless steel in composition. Each control blade shall be protected by an aluminum guide plate on each side within the control fuel assemblies.
Bases - The basis of enriching.the MTR fuel up to 20% is to allow a core loading compact enough to fit within* existing structures as well as to continue historic operations with approximately the same reactor characteristics while keeping the strategic significance of the material as low as possible.
Limiting the amount of grams of U-235 in each plate and assembly keeps the expected shutdown margin and accident analyses valid.
Those types of fuel plate defects listed in the specification have been exhibited in other facilities but the inspector should be cognizant of any change in plate appearance.
Changes in cladding appearance may be indicative of larger issues within the core and be precursors to failure of cladding integrity.
PUR-1 Technical Specifications 24 Amendment No. 13
5.4 Fuel Storage Applicability-The specification for fuel storage shall apply to the placement of fuel when it is not in the core configuration.
Objective - Ensuring that fuel outside of the highly analyzed reactor core does not go critical ls desirable to maintain safety to facility workers-and members of the public.
Specifications -
a.
All reactor fuel and fueled devices shall be stored in a geometric array where keff is less than 0.8 for all conditions of moderation and reflection.
b.
Irradiated fuel assemblies and fueled devices shall be stored in an array which will permit sufficient natural convection cooling by water or air such that the fuel integrity is maintained per the Safety Analysis Report.
Bases - The requirement to store fuel in such a way that the keff is less than 0.8 will 'be adequate to provide reasonable certainly that an accidental criticality event is not possible.
Placing fuel in an array which allows for adequate cooling will ensure that those elements which have experienced high burnup and have elevated levels of decay heat do not undergo loss of cladding integrity by blistering or other means due to high temperature.
PUR-1 Technical Specifications 25 Amendment No. 13
6.
ADMINISTRATIVE CONTROLS 6.1 Organization The PUR-1 Facility is managed and run by members of the university's College of Engineering, specifically the School of Nuclear Engineering. The Dean of the College of Engineering shall be the final authority on all PUR-1 matters. The Laboratory Director is responsible to the Dean for the_ __ administration and proper and safe operation of the facility. Figure 6.1 shows the administration chart for the PUR-1. The Committee on Reactor Operations advises the director of the PUR-1 on all matters or policy pertaining to safety. The Radiologicar Safety Officer provides advice concerning personnel and radiological safety and provides technical assistance and review in the area of radiation protection.
- President r - - - - - -
Purdue University I
~
Vice President for Research I
I Provost Purdue University Dean College of Engineering Radiation Safety Committee Level 1 Radiation Safety Officer Laboratory Director Level2 Reactor Supervisor Level3 Reactor Operations
--- Primarily Administration Primarily Safety I
I I
I t
Committee On Reactor Operations Figure 6.1: Organization Chart for Reactor Administration PUR-1 Technical Specifications 26 Amendment No. 13
a. Structure
1. A line management organizational structure provides for personnel who shall administrate and operate the reactor facility.
2. The Dean and the Facility Director shall have line management responsibility for adhering to the PUR-1 license and Technical Specifications and for safeguarding the public and facility personnel from undue radiation exposure.
3. Management Levels:
a) Level 1: Dean of the College of Engineering: Responsible for the PUR-1.
b) Level 2: PUR-1 Facility Director: Responsible for reactor facility C?peration and shall report to Level 1.
c) Level 3: Reactor Supervisor: Responsible for the day-to-day operation of the PUR-1 including shift operation and shall report to Level 2.
d) Level 4: Reactor Operating Staff: Licensed _reactor operators and senior reactor operators and trainees. These individuals shall report to Level 3.
e) The reporting structure of Figure 6.1 is such that those personnel below shall report up and those personnel listed above may communicate down.
4. Committee on Reactor Operations (CORO):
The CORO shall be responsible to the licensee for providing an independent review and audit of the safety aspects of the PUR-1.
b. Responsibility Responsibility for the safe operation of the reactor facility shall be in accordance with the line organization established in Section 6.1.a. In all instances, responsibilities of one level may be assumed by designated alternates or by higher levels, conditional upon appropriate qualifications.
)
The reactor facility shall be under the direct control of the Reactor Supervisor, a Senior Reactor *Operator, or Reactor Operator (RO). The RO shall be responsible for ensuring that all operations are conducted in a safe manner and within the limits prescribed by the facility license, procedures and requirements of the Radiation Safety Officer and the CORO.
c. Staffing PUR-1 Technical Specifications 27 Amendment No. 13
1. The minimum staffing when the reactor is not secured shall be as follows:
a) At least two individuals shall be present at the facility complex and shall consist of at least a licensed reactor operator and a second person capable of calling 911. Unexpected absence for as long as 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to accommodate a personal emergency are acceptable provided immediate action is taken to obtain a replacement. During periods when the reactor is not secured, it shall be under the direct control the of the. reactor operator; b) During periods of reactor maintenance the two individuals who shall be present at the facility complex shall consist of a licensed senior reactor operator and a second individual capable of calling 911.
c) A licensed reactor operator or senior reactor operator shall be in the reactor room; d) A Senior Reactor Operator shall be readily available for emergencies or on call (the individual can be rapidly reached by phone or radio and is within 30 minutes or 15 miles of the reactor facility); and e) A list of reactor facility personnel by name and telephone number shall be readily available for use in the reactor room. The list shall include:
i.
Senior Reactor Operator on Call, ii.
Radiation Safety Officer iii.
Other operations personnel, as deemed by the Facility Director
2. Events requiring the presence at the facility of the senior reactor operator:
a) Initial startup and approach to power, b) A Senior Reactor Operator shall direct any loading or unloading offuel or control rods within the reactor core region, c) A senior reactor operator shall direct the recovery from an unplanned shutdown, unscheduled shutdown, or unplanned power reduction of more than 5%.
d. Selection and Training of Personnel The selection and training of operations personnel shall be in accordance with the following:
1. Responsibility: The Reactor Supervisor is responsible for the selection, training, and requalification of the facility reactor operators and senior reactor operators.
PUR-1 Technical Specifications 28 Amendment No. 13
2. Selection: The selection of operations personnel shall be consistent with the standards related to selection in ANSl/ANS-15.4-2007
3. Training Program: The Training Program shall be consistent with the standards related to training in ANSl/ANS-15.4-2007.
4. Requalification Program: The Requalification Program shall be consistent with the standards related to requalification in ANSl/ANS-15.4-2007.
6Property "ANSI code" (as page type) with input value "ANSl/ANS-15.4-2007.6" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..2 Review and Audit
a. Committee on Reactor Operations (CORO)
The CORO s_h~ll be cqmprised of at least 3 voting members knowledgeable in fields which relate to NUCiear Safety. One of these members, the Radiation Safety Officer, will serve as the Chair. If the Chair is unable to attend one or a number of committee meetings, then the Chair may designate a committee member as Chair pro fem. The members are appoin~ed by the Dean of the College of Engineering to serve three--year terms. It is expected that the members will be reappointed each term as long as they are willing to serve so that their experience and familiarity with the past history of the PUR-1 will not be lost to the committee.
b. CORO Charter and Rules The operations of the CORO shall be in accordance with a written charter, including provisions for:
1. Meeting Frequency: The CORO shall meet annually at intervals not to exceed 15 months. (Note: The facility license requires a meeting at least once per year and as frequently as circumstances warrant consistent with effective monitoring of facility activities);
2. Quorum: A quorum shall be comprised of not less than one-half of the voting membership where the operating staff does not constitute a majority;
3. Voting Rules: On matters requiring a vote, if only a quorum is present a unanimous vote of the quorum shall be required; otherwise a majority vote shall be required;
4. Subcommittees: The Chair may appoint subcommittees comprised of members of the CORO to perform certain tasks. Subcommittees or members of the CORO may be authorized to act for the committee; and
5. Meeting Minutes: The Chair shall designate one individual to act as recording secretary. It shall be the responsibility of the secretary to prepare the minutes which shall be distributed to the CORO, including the Dean of the College of Engineering, within three months. The CORO shall review and approve the PUR-1 Technical Specifications 29 Amendment No. 13
minutes of the previous meetings. A complete file of the meeting minutes shall be maintained by the Chair of the CORO and by the Facility Director.
c. CORO Review Function The review responsibilities of the CORO or a designated subcommittee shall include, but are not limited to the following:
1. Review and evaluation of determinations of whether new tests or experiments and proposed changes to equipment, systems, or procedures can be made under 10 CFR 50.59 or would require a change in Technical Specifications or license conditions;
2. Review of new procedures, major revisions of procedures, and proposed changes in reactor facility equipment or systems which have significant safety impact to reactor operations;
3. Review of new experiments or classes of experiments that could affect reactivity or result in the release of radioactivity;
4. Review of proposed changes to the Technical Specifications and U.S. NRC issued license;
5. Review of the PUR-1 radiation protection program;
6. Review of violations of Technical Specifications, U.S. NRC issued license, and violations of internal procedures or instructions having safety significance;
7. Review of operating abnormalities having safety significance;
8. Review of reportable occurrences listed in Section 6.6.a and 6.6.b of these Technical Specifications; and 9: Review of audit reports.
d. CORO Audit Function The audit function shall include selective (but comprehensive) examination of operating records, logs, and other documents. Discussions with cognizant personnel and observation of operations should be used also as appropriate. In no case shall the individual immediately responsible for an area perform an audit in that area. Audits shall include but are not limited to the following:
1. Facility operations, including radiation protection, for conformance to the Technical Specifications, applicable license conditions, and standard operating procedures: at least every 12 months (interval between audits not PUR-1 Technical Specifications 30 Amendment No. 13
to exceed 15 months);
2. The results of action taken to correct those deficiencies that may occur in the reactor facility equipment systems, structures, or methods of operations that affect reactor safety: at least once every 12 nionths (interval between audits not to exceed 15 months);
3. The retraining and requalification program for the operating staff: at least once every other calendar year (interval between audits not to exceed 30 months);
4. The reactor facility emergency plan and implementing procedures: at least once every other calendar year (interval between audits not to exceed 30 months); and
5. The reactor facility security plan and implementing procedures: at least once every other calendar year (interval between audits not to exceed 30 months).
D*eficiencies uncovered that affect reactor safety shall immediately be reported to the Dean of the College of Engineering (Level 1 Management). A written
-report of the "firidings.of the audit shall be submitted to the Dean of the College of Engineering (Level 1 Management) and the review and audit group members within 3 months after the audit has been completed.
e. Audit of ALARA Program The Chair of the CORO or designated alternate (excluding anyone whose normal job function is within the operating staff) shall conduct an audit of the reactor facility ALARA program annually. The auditor shall transmit the results of the audit to the CORO at the next scheduled meeting for its review and approval.
6.3 Radiation Safety The Radiation Safety Officer shall be responsible for implementing the radiation safety program for the PUR-1. The requirements of the radiation safety program are established in 10 CFR 20. The Program should use the guidelines of the ANSl/ANS-15.11-1993; R2004, "Radiation Protection at Research Reactor Facilities."
6.4 Procedures Written operating procedures shall be prepared, reviewed, and approved before initiating any of the activities listed in this section. The procedures shall be reviewed and approved by the Facility Director, the CORO, and shall be documented in a timely manner. Procedures shall be adequate to ensure the safe operation of the reactor but shall not preclude the use of independent PUR-1 Technical Specifications 31 Amendment No. 13
judgment and action should the situation require such. Operating procedures shall be used for the following items:
a. Startup, operation, and shutdown of the reactor;
b. Fuel loading, unloading, and movement within the reactor;
c. Control rod removal or replacement;
d. Routine maintenance of the control rod, drives and reactor safety and interlock systems or other routine maintenance of major components of systems that could have an effect on reactor safety;
e. Surveillance checks, calibrations, and inspections of reactor instrumentation and controls, control rod drives, area radiation monitors, facility air monitors, the central exhaust system and -other systems as required by the Technical Specifications;
f.
Administrative controls for operations, maintenance, and conduct of irradiations and experiments, that could affect reactor safety or core reactivity;
g. Implementation of required plans such as emergency or security plans;
h. Radiation protection program to maintain exposures and releases as low as reasonably achievable (ALARA);
i.
Use, receipt, and transfer of by-product material, if appropriate; and
j.
Surveillance procedures for shipping radioactive materials.
6.5 Experiment Review and Approval Approved experiments shall be carried out in accordance with established and approved procedures.
a. All new experiments or class of experiments shall be reviewed by the CORO as required by TS 6.2.c and implementation approved in writing by the Facility Director or designated alternate.
b. Substantive changes to previously approved experiments shall be made only after review by the CORO and implementation approved in writing by the Facility Director or designated alternate.
6.6 Required Actions
a. Action to be Taken in the Event of a Safety Limit Violation PUR-1 Technical Specifications 32 Amendment No. 13
1. The reactor shall be shut down and reactor operation shall not be resumed until authorized by the U.S. NRC;
2. An immediate notification of the occurrence shall be made to the CORO Chair and the Facility Director, and reports shall be made to the U.S.
NRC in accordance with Section 6.7.b of these specifications; and
3. A report shall be prepared which shall include:
a) Applicable circumstances leading to the violation including, when known, the cause and contributing factors, b) Effect of the violation upon reactor facility components, systems, or structwes and _on_ the ~ealth and safety of_personnel and the public, c) Corrective action to be taken to prevent recurrence.
T~!_s r.~port shall be su~niit!ed to the CORO for review and then submitted to the U.S. NRC when authorization is sought to resume operation of the reactor.
b. Action to be Taken in the Event of a Reportable Occurrence Other Than A Safety Limit Violation
1. PUR-1 staff shall return the reactor to normal operating via the approved PUR-1 procedure or shut down conditions. If it is necessary to shut down the reactor to correct the occurrence, operations shall not be resumed unless authorized by the Facility Director or a designated alternate;
2. The Facility Director or designated alternate shall be notified and corrective action taken with respect to the operations involved;
3. The Facility Director or designated alternate shall notify the CORO Chair who shall arrange for a review by the CORO;
4. A report shall be made to the,CORO which shall include an analysis of the cause of the occurrence, efficacy of corrective
action, and recommendations for measures to prevent or reduce the probability of recurrence; and
5. A report shall be made to the U.S. NRC in accordance with Section 6.7.b of these specifications.
6.7 Reports PUR-1 Technical Specifications 33 Amendment No. 13
a. Annual Operating Report An annual report covering the operation of the reactor facility during the previous calendar year shall be submitted to the NRC before March 31 of each year providing the following information:
1. A narrative summary of (1) reactor operating experience (including experiments performed), (2) changes in facility design, performance characteristics, and operating procedures related to reactor safety and occurring during the reporting period, and (3) results of surveillance tests and inspections;
2. Tabulation of the energy output of the reactor, hours reactor was critical, and the cumulative total energy output since initial criticality;
3. The number of unscheduled shutdowns and inadvertent scrams, including, where applicable corrective action to preclude recurrence;
4. Discussion of the major maintenance operations performed during the period, incl_uding the effect, if any, on the safety of the operation of the reactor and the reasons for any corrective maintenance required;
5. A brief description, including a summary of the safety evaluations of changes in the facility or in procedures and of tests and experiments carried out pursuant to Section 50.59 of 10 CFR Part 50;
6. A summary of the nature and amount of radioactive effluents released or discharged to the environs beyond the effective control of the licensee as measured at or before the point of such release or discharge. The summary shall include to the extent practicable an estimate of individual radionuclides present in the effluent. If the estimated average release after dilution or diffusion is less than 25% of the concentration allowed or recommended,- a statement to this effect is sufficient:
a) Liquid Waste (summarized on a monthly basis)
i. Radioactivity discharged during the reporting period.
I.
Total radioactivity released (in Curies),
11.
The effluent concentration used and the isotopic composition if greater than 1 x 10-7 µCi/cc for fission and activation products, Ill.
Total radioactivity (in Curies), released by nuclide during the reporting period based on representative isotopic analysis, and IV.
Average concentration at point of release (in µCi/cc) during the PUR-1 Technical Specifications 34 Amendment No. 13
reporting period.
ii. Total volume (in gallons) of effluent water (including dilution) during periods of release.
b) Airborne Waste (summarized on a monthly basis)
i. Radioactivity discharged during the reporting period (in Curies) for:
I.
41Ar, and II.
Particulates with half-lives greater than eight days.
c) Solid Waste
i. The total amount of solid waste transferred (in cubic feet),
ii. The total activity involved (in Curies), and iii. The dates of shipment and disposition (if shipped off site).
7. A summary of radiation exposures received by facility personnel and visitors, including dates and time where such exposures are greater than 25% of that allowed or recommended; and
8. A description and summary of any environmental surveys performed outside th~, facility.
b. Special Reports In addition to the requirements of applicable regulations, reports shall be made to the NRC Document Control Desk and special telephone reports of events should be made to the Operations Center as follows:
1. There shall be a report not later than the following working day by telephone and confirmed. in writing by fax or similar conveyance to the NRC Headquarters Operation Center, and followed by a written report that describes the circumstances of the event and sent within 14 days to the U.S.
Nuclear Regulatory Commission, Attn: Document Control Desk, Washington, DC 20555, of any of the following:
a) Violation of safety limit (see TS 6.6.a);
b) Any release of radioactivity from the site above allowed limits; and c) Any of the following:
PUR-1 Technical Specifications 35 Amendment No. 13
i.
Operation with actual safety system settings for required systems less conservative than the limiting safety system settings specified in the technical specifications.
ii.
Operation in violation of limiting conditions for operation established in the technical specifications.
iii.
A reactor safety system component malfunction that renders or could render the reactor safety system incapable of performing its intended safety function. If the malfunction or condition is caused by maintenance, then no report is required.
Note: Where components or systems are provided in addition to those required_ by the technical specifications, the failure of the extra components or systems is not considered reportable provided that the minimum numbers of components or systems specified or required perform their intended reactor safety function.
iv.
An unanticipated or uncontrolled change in reactivity greater than 0.006 !:J.k/k.
v.
Abnormal and significant degradation in reactor fuel or cladding, or both, coolant boundary, or confinement boundary (excluding minor leaks).
vi.
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.
2. A written report within 30 days to the U.S. Nuclear Regulatory Commission, Attn: Document Control Desk, Washington, DC, 20555, of:
a) Permanent changes in the facility organization involving Level 1 and Level 2;and b) Significant changes in the transient or accident analysis as described in the Safety Analysis Report.
6.8 Records Records of facility operations in the form of logs, data sheets, or other suitable forms shall be retained for the period indicated as follows:
a. Records to be Retained for a Period of at Least Five Years or for the Life of the Component Involved if Less Than Five Years
1. Normal reactor facility operation (but not including supporting documents PUR-1 Technical Specifications 36 Amendment No. 13
such as checklists, log sheets, etc. which shall be maintained for a period of at least one year),
2. Principal maintenance operations,
3. Reportable occurrences,
4. Surveillance activities required by the Technical Specifications,
5. Reactor facility radiation and contamination surveys where required by applicable regulations,
6. Experiments performed with the reactor,
7. Fuel inventories, receipts, and shipments,
8. Approved changes in operating procedures, and
9. Records of meeting and audit reports of the CORO.
b. Records to be Retained for at Least One Certification Cycle Records of retraining and requalification of licensed operations personnel shall be maintained at all times the individual is employed or until the license is renewed.
c. Records to be Retained for the Lifetime of the Reactor Facility
1. Gaseous and liquid radioactive effluents released to the environs,
2. Radiation exposure for all personnel monitored,
3. Drawings of the reactor facility, and
4. Reviews and reports pertaining to a violation of the safety limit, the limiting safety system setting, or a limiting condition of operation.
PUR-1 Technical Specifications 37 Amendment No. 13

v t e Letter Sequence Response to RAI
TAC:ME1594, (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request Acceptance, Acceptance Administration Questions 24 March 2010 6 July 2011 8 July 2011 14 July 2011 28 December 2011 29 May 2012 29 August 2014 19 January 2016 23 December 2015 25 July 2016 Answers 4 June 2010 15 November 2011 4 January 2012 1 June 2012 15 June 2012 29 June 2012 13 July 2012 11 August 2012 31 January 2012 29 January 2016 26 February 2016 31 March 2016 9 May 2016 19 July 2016 19 September 2016 Results Approval Other: ML120900571, ML16265A632, ML16277A165
MONTHYEAR2012-06-29 [Table View]

ML16267A465

Text

SUBJECT:

Dear Ms. Montgomery,

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