Text

University of Missouri-Columbia Research Reactor, Submittal of Request to Amend the Technical Specifications
	ML110320458
Person / Time
Site:	University of Missouri-Columbia
Issue date:	01/31/2011
From:	Rhonda Butler; Univ of Mississippi
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
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	Download: ML110320458 (45)
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UNIVERSITY of MISSOURI RESEARCH REACTOR CENTER January 31, 2011 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Mail Station P1-37 Washington, DC 20555-0001

Reference:

Docket 50-186 University of Missouri-Columbia Research Reactor Amended Facility License R-103 On August 6, 2009 the University of Missouri-Columbia Research Reactor submitted a request to amend the Technical Specifications appended to Facility License R-103. Enclosed is our response to the U.S. Nuclear Regulatory Commission's request for additional information regarding the proposed Amendment, dated December 27, 2010.

If you have any questions, please contact John L. Fruits, the facility Reactor Manager, at (573) 882-5319.

Sincerely, Ralph A. Butler, P.E.

Director RAB/djr Enclosures 1513 Research Park Drive Columbia, MO 65211 Phone: 573-882-4211 Fax: 573-882-6360 Web: http://web.missouri.edu/-murrwww Fighting Cancer with Tomorrow's Technology

UNIVERSITY of MISSOURI RESEARCH REACTOR CENTER January 31, 2011 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Mail Station P1-37 Washington, DC 20555-0001

REFERENCE:

Docket 50-186 University of Missouri-Columbia Research Reactor Amended Facility License R-103

SUBJECT:

Written communication as specified by 10 CFR 50.4(b)(1) regarding the response to the "University of Missouri - Columbia - Request for Additional Information, Re:

License Amendment, Center Test Hole (TAC No. ME1876)," dated December 27, 2010 By letter dated August 6, 2009, the University of Missouri-Columbia Research Reactor (MURR) submitted a request to the U.S. Nuclear Regulatory Commission (NRC) to amend the Technical Specifications (TS), which are appended to Facility License R-103, that would authorize the implementation of an engineered safety device that would prevent operation of the reactor unless the center test hole canister is inserted and latched onto the inner reactor pressure vessel. Approval of this Amendment would allow greater flexibility and capacity in the center test hole for the irradiation of high specific activity radioisotopes that are used for radiopharmaceutical research and cancer treatments.

By letter dated June 1, 2010, as part of the facility license renewal process, the NRC requested additional information and clarification regarding the proposed Amendment in the form of seven (7) questions. By letter dated August 31, 2010, the MURR responded to those questions.

On December 27, 2010, the NRC requested additional information and clarification regarding the proposed Amendment in the form of five (5) questions. Those questions, and MURR's responses to those questions, are attached. If there are questions regarding this response, please contact me at (573) 882-5319. I declare under penalty of perjury that the foregoing is true and correct.

ENDORSEMENT:

Sincerely, Reviewed and Approved, John L. Fruits Ralph A. Butler, P.E. ,Y.P,; MARGEE P.STOUT Reactor Manager Director ,.TN0.TA& March 24, 2012

-,-... SEAL.;: Montgoery County

.. ..... C d- . 8513 1513 Research Park Drive Columbia, MO 65211 Phone: 573-882-4211 Fax: 573-882-6360 Web: www.murr.missouri.edu Fighting Cancer with Tomorrow's Technology

Attachments: 1. Administrative Procedure AP-RO-135, "Reactor Utilization Requests"

2. Compliance Check Procedure CP-36, "FIRST Scrams"

3. Preventative Maintenance Procedure RX-Q2, "Inspect FIRST Support Rig"

4. Revised Technical Specification 3.3, Pages 3 of 5 and 5 of 5 xc: Reactor Advisory Cdmmittee Reactor Safety Subcommittee Dr. Robert Duncan, Vice Chancellor for Research Mr. Craig Basset, U.S. NRC Mr. Alexander Adams, U.S. NRC 2 of 19

Page 8 of amendment application. In analyzing the accidents associated with the Flux-trap IrradiationsReactivity Safety Trip (FIRST) device, you state that the FIRST device does not alter the results of analyzed accidents. You also state that the most likely accident initiator in the center test hole is the failure of a single experiment sample in the center test hole canister. Your discussion focuses on the reactivity aspects of experimentfailures. The use of the FIRST device could allow an increase in the quantity of material allowed in the center test hole position. Please provide additionaldetail regardingpotential changes in the probabilityor consequences offailures from other causes involving increased sample loading in the center test hole associated with the FIRST device as it is proposed to be used.

The mechanism used to verify that each experiment, or single experiment sample, complies with all of the applicable Hazards Summary Report (HSR) and Technical Specifications (TS), and other limitations based on good operating, engineering, and health physics practices, is called the Reactor Utilization Request (RUR). This formalized process, which is detailed in administrative procedure AP-RO-135, "Reactor Utilization Requests" (Attachment 1), specifically requires that a safety analysis be prepared, reviewed and approved by both the Reactor and Reactor Health Managers before an experiment can be conducted. No material shall be irradiated in the center test hole, graphite reflector region, bulk pool or pneumatic tube system unless it is in full compliance with an approved RUR. Additionally, as required by TS 6.1.c (2), proposed experiments significantly different from any previously reviewed or which involve a question pursuant to 10 CFR 50.59 shall be reviewed by the Reactor Advisory Committee (RAC).

Each safety analysis includes, but is not limited to, the following major criteria: criticality and/or reactivity considerations; heat generation considerations; shielding considerations; and off-gassing and/or chemical reactions. It also includes all credible accident and transient scenarios to ensure that the experiment does not jeopardize the safe operation of the reactor or constitute a hazard to the safety of the facility staff and general public.

As discussed in Section 3.10 of Addendum 3 to the HSR, the most likely accident in the center test hole is the possible failure of any single experiment. The worst case might be the sudden bursting of the sample can and a discharge of its contents and possible damage to adjacent sample cans.

Implementation of the FIRST device does not alter this assumption. The introduction of a reactivity change due to a sample can failure is just one of the potential hazards analyzed. The following are additional safety analyses that are required to be included in the RUR to demonstrate that the experiment meets all applicable HSR and TS requirements:

Thermal Analysis TS 3.6.h

Sample Decomposition-Pressure Analysis TS 3.6.i

Experiment Failure Analysis TS 3.6.c

Loss of Coolant Analysis TS 3.6.f

Failure of Other Experiments Analysis TS 3.6.g

Corrosion Analysis TS 3.6.j The Thermal Analysis* is an estimation of the heat generation and heat transfer rates for an experiment, determining if a cooling design change is required to prevent the surface temperature of a submerged irradiated sample from exceeding the saturation temperature of the liquid it is submerged it. The intent of TS .3.6.h is to reduce the likelihood of reactivity transients due to accidental voiding in the reactor or the failure of an experiment from internal or external heat generation. An example of one of the thermal limits is the heat generation rate of a sample in the 3 of 19

center test hole, which is restricted to a heat flux limit of no greater than 100 w/cm2 on the surface of its irradiation container. This value represents the conservative limit to be used if all experiment samples over the entire length of the center test hole were generating this same amount of heat.

The Sample Decomposition-Pressure Analysis describes the form of the sample and component materials of the experiment during irradiation, with reasonable leeway for normal and abnormal conditions. The analysis confirms that a potential pressure buildup due to a complete decomposition of the sample material will not exceed the design pressure of the irradiation container. The intent of TS 3.6.i is to reduce the likelihood of damage to the reactor and/or radioactivity releases from an experiment failure.

The Experiment Failure Analysis determines if products or components from the experiment have the potential to violate the limits of 10 CFR 20, Appendix B, Table I, if released to the atmosphere.

This analysis ensures compliance with TS 3.6.c.

The Loss of Coolant Analysis describes how a loss of coolant (e. g., loss of pool coolant flow, loss of experiment cooling, etc.) to the experiment will not result in a release of radioactivity to the atmosphere or affect the safe operation/control of the reactor. This analysis ensures compliance with TS 3.6.f.

The Failure of Other Experiments Analysis identifies the possible effects upon reactor control and other experiments due to operating an experiment under abnormal conditions (failure). This analysis ensures compliance with TS 3.6.g.

The Corrosion Analysis ensures that the encapsulation provides enough corrosion resistance to endure the worst case scenario of corrosion for the duration of the experiment if corrosive materials are expected to be generated in an appreciable quantity during normal operation or as a result of the experiment failing. This analysis ensures compliance with TS 3.6.j.

The Explosive Analysis ensures that if explosive materials are present or are expected to be formed during the irradiation then the total mass of the explosive will not exceed the TS limitation. This analysis ensures compliance with TS 3.6.d.

In addition to efforts to ensure that a single experiment failure will not cause other experiments to fail, there are guideline limits on other variables to ensure that the entire irradiation facility, such as the center test hole or the graphite reflector region, is protected from failure. These limits can include mass, flux, fluence, byproduct generation, and number and type of encapsulation.

The RUR process, and the safety analysis within that process, provides a continuity of protection to the entire reactor system, including the center test hole. Therefore, this process ensures that while there may be a minimal increase in the probability of a failure due to an increase in sample loading in the center test hole, the magnitude of a potential failure does not increase accordingly.

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2. RAI 10.5. e. In response,to RAI 10. 5. e, a methodology is proposed to measure reactivity worth of the empty center test hole canister and a loaded canister with back-to-back reactor startups.

Although adjustments to the calculationsdue to temperaturechanges are specifically mentioned, no mention is made regarding adjustmentsfor fission product poisons. Please clarify the proposed methodology in determining the reactivity worth of items in the center test hole as related to adjustmentsfor poisons in the core.

The effects of fission product poisons are very minimal during reactivity worth measurements for the following reason: all reactivity measurements are performed during a scheduled reactor startup following a weekly maintenance day shutdown period. As part of the maintenance day activities, all eight (8) fuel elements are replaced with different elements of varying stages of bumup, and hence, typically, no xenon effects are present during the startup. When the reactivity worth of the loaded center test hole canister is being determined, the core will be loaded with only xenon-free fuel elements. Additionally, the reactivity measurements are conducted at very low power levels specifically to avoid fission product poison build up during multiple startups. Typically, these measurements are performed at or below 50 kW, or less than 0.5% of the MURR full core power level of 10 MW, which is sufficient enough to provide reliable nuclear instrumentation indication yet not enough to produce any significant amount of poisons in the core. Also, the amount of time spent at these low power levels are limited to only a few minutes, again for reducing the amount of poison build up in the core. Because of the above precautions, we do not see any effects from poisons, and hence, no poison corrections are required during reactivity worth estimations.

The stated purpose of the FIRST system is to allow the center test hole canister or strainer to be considered a normal part of the reactorfor reactivity consideration purposes instead of being considered an experiment. Use of the FIRST system will allow what constitutes the "normal" reactorto change depending if a three tube canister,a six tube canister,or straineris in the center test hole position. Rod worth curves should be developed for the reactor in each of its normal configurations. Because of this, please propose technical specification (TS) requirements to insure that correct, current rod worth curves exist and are used specific to the configuration being used with the FIRST system.

Unlike power or typical research reactors, the control elements of the MURR are situated entirely outside of the outer reactor pressure vessel, or the "core proper," which contains the fuel. As can be seen in Figure 1, the eight (8) fuel elements, which make up the reactor core, are situated between the control rod water gap and the flux trap region where the sample filled center test hole canister is positioned. This arrangement reduces the coupling between the control rods from the effects of sample or sample holder changes in the center test hole region and hence will not change the rod worth characteristics significantly depending on whether the strainer or the 3- or 6-barrel center test hole canister is installed. This hypothesis is quantified below using both analytical as well as experimental methods.

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Regulating Blade Control Blade Berylium Reflector Plate 1 2 Inner Pressure Vessel Plate 24 Outer Pressure Vessel Figure 1 Overhead View of Reactor Core Region First, a set of detailed MCNP calculations were performed to determine the total rod worth of two different control rods as well as the neutron spectrum at fuel plate number-24 under several conditions within the central flux trap region. Plate number-24 is the fuel plate that is closest to the control rod gap and furthest from the flux trap region. If this plate sees a significantly different spectrum when changes are made within the flux trap, it would indicate a corresponding change to the control rod worth.

As shown in the Table 1 below, the difference in the estimated total rod worth of two different control rods when the central flux trap region is filled entirely with water (strainer condition) or with a fully loaded 3-tube center test hole canister installed is only less than 0.01%.

Table 1 Control Rod Worth Comparisons with Strainer or 3-Tube Center Test Hole Canister Installed Control Rod 'B' Total Worth I 0.0478 1 0.0486 0.0008 1 Control Rod 'C' Total Worth 0.0488 1 0.0493 0.0005 6 of 19

Similarly, fuel plate number-24 neutron flux spectrums were calculated using MCNP and are provided below in Tables 2 through 7 for a single fuel element in 4-inch axial increments. The maximum change in the spectrum is less than 5% for the strainer versus the loaded 3-tube center test hole canister case. Although results for a single fuel element are shown, similar results were obtained for three (3) other fuel elements.

Table 2 Normalized Neutron Flux Spectrum for Fuel Plate Number-24 of Fuel Element X2 (Between 0 and 4 inches)

Neutron Energy Normalized Flux Normalized Flux ýj (MeV) (Strainer) (Loaded 3-Tube Center Test Hole Canister) 1.OOE-06 1.44E-05 1.47E-05 2.59%

1.OOE-05 8.05E-06 8.38E-06 4.00%

1.OOE-04 9.61E-06 9.82E-06 2.19%

1.OOE-03 1.1E-05 1.12E-05 2.17%

1.00E-02 1.1 8E-05 1.21 E-05 2.08%

1.OOE-01 1.43E-05 1.47E-05 2.45%

1OOE+00 3.26E-05 3.29E-05 0.97%

2,OOE+O1 3.49E-05 3.61EO05 3.37%

Table 3 Normalized Neutron Flux Spectrum for Fuel Plate Number-24 of Fuel Element X2 (Between 4 and 8 inches)

Neutron Energy Normalized Flux Normalized Flux (MeV) (Strainer) (Loaded 3-Tube Center Test Hole Canister) A 1.00E-06 2.50E-05 2.56E-05 2.18%

1.OOE-05 1.50E-05 1.51E-05 1.03%

1.OOE-04 1.78E-05 1.79E-05 0.47%

L.OOE-03 2.06E-05 2.08E-05 1.27%

1.OOE-02 2.20E-05 2.23E-05 1.56%

LOOE-01 2.67E-05 2.70E-05 1.02%

1.00E+00 5.99E-05 6.1OE-05 1.76%

2.00E+01 6.39E-05 6.46E-05 1.03%

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Table 4 Normalized Neutron Flux Spectrum for Fuel Plate Number-24 of Fuel Element X2 (Between 8 and 12 inches)

Neutron Energy Normalized Flux Normalized Flux (MeV),, (Strainer) (Loaded 3-Tube Center Test HoleCanister) 1.00E-06 4.15E-05 4.09E-05 -1.53%

1.OOE-05 2.33E-05 2.36E-05 1.34%

L.00E-04 2.78E-05 2.78E-05 -0.20%

1.OOE-03 3.22E-05 3.22E-05 0.12%

1.00E-02 3.42E-05 3.45E-05 0.81%

L.00E-01 4.19E-05 4.20E-05 0.20%

1.00E+00 9.43E-05 9.46E-05 0.25%

2.00E+01 1.02E-04 1.02E-04 -0.37%

Table 5 Normalized Neutron Flux Spectrum for Fuel Plate Number-24 of Fuel Element X2 (Between 12 and 16 inches)

Neutron Energy N6omalized Flux Normalized Flux (MeV) (Strainer) (Loaded 3-Tube Center Test Hole Canister) A L.OOE-06 1.29E-04 1.23E-04 -4.83%

1.OOE-05 3.89E-05 3.88E-05 -0.24%

L.OOE-04 4.07E-05 4.08E-05 0.37%

1.OOE-03 4.49E-05 4.49E-05 0.03%

L.OOE-02 4.76E-05 4.78E-05 0.43%

L.OOE-01 5.93E-05 5.93E-05 -0.11%

1.OOE+00 1.51E-04 1.50E-04 -1.14%

2.OOE+01 1.83E-04 1.78E-04 -2.52%

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Table 6 Normalized Neutron Flux Spectrum for Fuel Plate Number-24 of Fuel Element X2 (Between 16 and 20 inches)

Neutron Energy Normalized Flux Normalized Flux A (MeV) (Strainer) (Loaded 3-Tube Center Test Hole Canister) 1.00E-06 1.58E-04 1.60E-04 1.39%

1.00E-05 4.25E-05 4.28E-05 0.79%

1.00E-04 4.35E-05 4.42E-05 1.53%

1.OOE-03 4.71E-05 4.80E-05 2.01%

1.00E-02 5.06E-05 5.13E-05 1.46%

1.OOE-01 6.46E-05 6.49E-05 0.51%

1.00E+00 1.68E-04 1.70E-04 0.99%

2.OOE+01 2.08E-04 2.09E-04 0.43%

Table 7 Normalized Neutron Flux Spectrum for Fuel Plate Number-24 of Fuel Element X2 (Between 20 and 24 inches)

Neutron Energy Normalized Flux Normalized Flux (MeV) (Strainer) (Loaded 3-Tube Center Test Hole Canister) 1.00E-06 1.28E-04 1.30E-04 1.57%

L.OOE-05 3.06E-05 3.13E-05 2.49%

L.OOE-04 3.14E-05 3.20E-05 1.70%

1.00E-03 3.39E-05 3.43E-05 1.19%

1.OOE-02 3.66E-05 3.73E-05 1.82%

L.00E-01 4.66E-05 4.73E-05 1.56%

1.OOE+00 1.24E-04 1.27E-04 2.14%

2.00E+01 1.57E-04 1.60E-04 1.30%

Next, to experimentally verify the results obtained analytically, the worth of single control rod (Control Rod 'B') was measured three (3) times with the following three (3) different conditions in the center test hole: (1) with the strainer installed, (2) with a loaded 3-tube center test hole canister installed, and (3) with an empty 6-tube center test hole canister installed.

As shown in Table 8, the difference in the measured total rod worth of control rod 'B' for the three (3) different central flux trap region conditions mentioned above is less than 0.00 15.

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Table 8 Control Rod 'B' Total Worth (with Strainer, 3- and 6-Tube Center Test Hole Canisters Installed)

Control Rod 'B' 0.0350 0.0342

)335 Total Worth Figure 2 shows the measured integral rod worth curve for control rod 'B' for the three cases indicated above.

Control Rod B - Integral Worth 0.03600 0.03000 0.02400 I 0.O1gO0 -- With Strainer

With 39-FT

'~0.01200 With6B-FT F

0.00600 0 .00 00 0 1...

0.0 4.0 8.0 12.0 16.0 20.0 24.0 29.0 Rod B Height (inches)

Figure 2 Control Rod 'B' Integral Rod Worth Curve - 3 Different Center Test Hole Conditions The measured control rod worth curves are typically used for verifying the reactivity worth of a sample or experiment by using the change in rod height with and without the sample or experiment installed. As shown in Table 9, positive reactivity inserted while withdrawing control blade 'B' in 2-inch increments in the region of interest (10 to 20 inches) was estimated using the three (3) integral rod worth curves generated for the three (3) different center test hole conditions. It can be seen that the change in reactivity are well within typical experimental error range.

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Table 9 Reactivity with Strainer, 3- or 6-Tube Center Test Hole Canister Installed Control Rod Wit With Loaded 3-Tube With Empty 6-Tube Heigh'*ivWith'Strainer Height Center Test Hole Canister Center Test Hole Canister 10 - 12 inches 0.0043 0.0045 0.0045 12 - 14 inches 0.0041 0.0044 0.0044 14 - 16 inches 0.0036 0.0039 0.0039 16- 18 inches 0.0030 0.0032 0.0032 18 - 20 inches 0.0022 0.0023 0.0024 Both the analytical and experimental results demonstrate that the worth of a control rod is minimally influenced by the samples or the lack of samples (strainer) or the different types of sample holders in the central flux trap region. MURR feels that no Technical Specification requirement is needed for estimating control rod worth for the different possible conditions in the center test hole or flux trap region of the reactor that are requested in the Amendment.

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3. Proposed TS 3.3. Normally it is expected that the use of bypasses in the reactor safety system, where their use is approved, is rare and under significant administrative control. It appears that the use of bypass in the FIRST system will be an operationaland experimental need decision. The proposed TS 3.3 adds the channel operability requirementfor the Center Test Hole Scram under conditions specified in a footnote. The operationof the reactor with the system bypassed should be through a positive control TS with specific requirements ratherthan an indirect requirementthough a footnote. Pleasepropose new TS wordingfor the use of the bypass that controls the Center Test Hole Scram function and specifies the requirements for its use or provide justification for its omission.

MURR feels that the use of a footnote in a Technical Specification (TS) holds equal enforcement as would any other portion of a TS. In this case, the use of the bypass switch is indeed anticipated to be very rare and thus justifies its use as a footnoted direction to the specific instance when it may be authorized. We agree that specific authorization from the Reactor Manager should be included in this TS. Footnote (6) to TS 3.3 has been revised to read:

(6) Not required if reactivity worth of the center test hole removable experiment test tubes and its contents is less than the reactivity limit of specification 3.6.h. This safety function shall only be bypassed with specific authorization from the Reactor Manager.

MURR will also add specific steps in administrative procedure AP-RO-1 10, "Conduct of Operation," which will require permission from the Reactor Manager for operation of the reactor with any scram bypass switch in the "bypass" position. FM-57, "Long Form Startup Checksheet,"

already contains a step to ensure all bypass switch keys are removed.

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4. RAls 10.5.c., d., and g. The statement is made that periodic surveillance and maintenance inspections will reveal potential switch and cable deterioration and failures. You state that TS 4.4. a. (we assume that this TS refers to the proposed license renewal TS, if issued as a stand-alone amendment prior to issuance of the license renewal TS changes will apply to the current TSs) will apply. This TS (and current TS 5.3.a.) refers to channel calibrations. Will the system also be subject to a channel test prior to each start-up? If not, please justify. Because the center test hole canister could be removed before each start-up, the proper operation of the system should be confirmed before start-up. Pleaseprovide a copy of draftprocedures CP-36 and RX-Q2.

That is correct. The response to relicensing RAI 10.5.g was in reference to the Technical Specifications (TS) that were submitted as a part of relicensing (SAR Appendix A). Currently NRC-approved TS 5.4.a, not TS 5.3.a as stated in the above question, applies to channel calibrations. The FIRST system will be tested for operability on a semiannual basis using Compliance Check Procedure CP-36, "FIRST Scrams," to ensure compliance with TS 5.4.a.

Yes, the system will be subjected to a channel test before each reactor startup as part of performing FM-57, "Long Form Startup Checksheet." Just prior to performing step "Control Rod Operation and Scram Test" with the center test hole canister removed, the operator will attempt to reset the reactor safety system Trip Actuator Amplifiers (TAAs). With the center test hole canister removed, the FIRST system should prevent the TAAs from resetting and energizing. If the TAAs were to reset and energize without the center test hole canister installed, it would give positive indication that a closed circuit failure of the FIRST device had occurred and thus requires corrective action be taken prior to reactor startup. If the TAAs do not reset and energize with the center test hole canister removed, it would indicate proper operation of the system and no closed circuit failure. At this point the center test hole canister would be inserted and latched, and the operator would again attempt to reset the TAAs at which time they should successfully reset and energize if the FIRST system is functioning properly.

Draft copies of Compliance Check Procedure CP-36, "FIRST Scrams," and Preventative Maintenance Procedure RX-Q2, "Inspect FIRST Support Rig," are attached for your review.

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5. Page 7 ofyour amendment applicationand RAI 10.5 f Your amendment applicationstates that the design of the system meets all of the applicable criteria of the Institute of Electricaland Electronics Engineers (IEEE)Standard IEEE-279. However, the answer to RAI 10. 5f states that the system effectively meets the intent of IEEE-279 (emphasis added). Please provide a copy of your design evaluation of this issue which shows the applicablecriteriamet, why they were determined to be the applicablecriteriaand why the system meets the intent ofIEEE-2 79.

The following response discusses the design evaluation with regard to IEEE Standard 279-1971, "Criteria for Protection Systems for Nuclear Power Generating Stations," and is intended to clarify previous discussions with regard to applicable criteria or meeting the intent of IEEE-279. Some of the requirements are not applicable to any part of the MURR reactor safety system, whereas others have limited applicability. For each of the design requirements, a short discussion clarifies the applicability of the requirement, and provides the justification for said applicability.

IEEE Standard 279-1971 proposes the following twenty-two (22) requirements:

1. General Functional: Requirement 12. Operating Bypasses

2. Single Failure Criterion 13. Indication of Bypasses

3. Quality of Components 14. Access to Means for Bypassing

4. Equipment Qualification 15. Multiple Set Points

5. Channel Integrity 16. Completion of Protective Action Once Initiated

6. Channel Independence 17. Manual Initiation

7. Control and Protection System Interaction 18. Access to Set Point Adjustments

8. Derivation of System Inputs 19. Identification of Protective Actions

9. Capability for Sensor Checks 20. Information Read-Out

10. Capability for Test and Calibration 21. System Repair

11. Channel Bypass or Removal from 22. Identification Operation

1. General Functional Requirement This requirement is satisfied by the combined physical channel arrangement and periodic testing. The channel arrangement uses precision position sensor switches to detect a small change in the position of the center test hole canister, and using standardized relays, converts that change to one or more logic inputs to the reactor safety system. Each component chosen has been tested in its intended working environment or has a proven reliable track record at the MURR to meet the full range of transient and steady-state conditions anticipated. Credible malfunctions have been considered and are discussed in more detail below. Performance requirements will ensure that the instrument channels provide an accurate indication of this small change in position to ensure automatic protective action is completed prior to exceeding step reactivity insertion limits.

2. Single Failure Criterion This requirement is satisfied primarily by the provision of redundant instrument channels to provide the protective function. Efforts to provide channel independence, and consideration of any credible single failure, provide additional confidence that this requirement is met. The worst-case credible single failure of these instrument channels would be an unintended closed circuit condition that serves to mask the true detected condition. This single failure could occur in many different locations within the physical channel, but would be minimized by efforts to 14 of 19

provide channel independence. A more detailed discussion of those locations are discussed in the channel independence requirement below.

3. Quality of Components This requirement is satisfied by the component specifications with regard to materials compatibility, reliable operation, availability of spare parts, and access to components for inspection, testing and calibration.

4. Equipment Qualification This requirement is satisfied by the test data gained through previous bench testing and the ongoing in-situ testing of the temporary mock instrument channels. With the exception of actuating the reactor safety system proper, the mock instrument channels and their components have performed reliably in their intended working environment, and have continued to produce timely and accurate channel responses within the proposed operating range. The means of actuating the safety system proper uses existing proven components with a high degree of operational reliability.

5. Channel Integrity This requirement is satisfied by the consideration of the extremes of applicable conditions. The proposed instrument channels will be susceptible only to the same extremes of conditions as the remainder of the reactor safety system. These extremes, if produced by the reactor systems, or if produced by external environmental factors would have initiated and completed a safe reactor shutdown by other means (including Manual SCRAM) prior to impacting the operability of these instrument channels.

6. Channel Independence This requirement is effectively satisfied by the provision of two independent instrument channels from detection through reactor safety system actuation. While the two instrument channels will be fully independent electrically, they will provide only an effective independence physically.

Starting at detection, the switches, though separate, will share a common support rig. No credible single failure could be envisioned that would allow both switches to close. Two separate cable bundles will be routed together through the pool from the support rig to the pool bridge mezzanine level. This arrangement will reduce the likelihood of physically interfering with the cable bundle during routine pool operations. Two separate cable bundles will be routed together through a dedicated conduit through the Instrument Cabinet and Reactor Console to the bypass switch. This co-routing of safety system instrument channel cable runs is not a common practice at the MURR, but has been accepted where the conduits are physically protected by a concrete chase or other structural support, as will be the case here.

The bypass switch will provide an effective mechanical separation of the channels by its proven reliable track record. Each contact block will be a separate physical component of the key switch, so electrical independence will remain intact. The common failure point then will be the mechanical operation of the key tumbler. These standardized key switches have a long track record at the MURR, with no record of failure or replacement due to reliability concerns.

Mechanically, they are of sound construction and not prone to unintended movement without the key properly fitted. A high level of administrative control will be utilized in the operation of this bypass switch, such that the key will not be present or accessible unless the Reactor Manager has granted permission [proposed TS 3.3, footnote (6)] for its use and directed the 15 of 19

channels to be placed in bypass. Therefore, no credible single event, physical or administrative, could result in a reduction of channel independence with regard to the bypass switch.

Two separate cable bundles will be routed together from the bypass switch to the two independent relays in the safety system K-relay drawer. Again, while not common practice, this co-routing of cable runs can be accepted in this case due to the protected nature of the routing.

The independent relays will then provide the channel inputs to the reactor safety system via a single series contact in each of the Yellow and Green input legs to the Non-Coincidence Logic Units (NCLU). This system arrangement and the remainder of the safety system will be in accordance with the existing MURR convention for safety system channels.

7. Control and Protection System Interaction This requirement is not applicable to these instrument channels. The proposed instrument channels will provide only protective system inputs and annunciator indication of those inputs.

No control function will be provided by these instrument channels. Therefore, none of the remaining criteria of this requirement (isolation devices, single random failure, or multiple failuresfrom a credible single event) are applicable to these instrument channels.

8. Derivation of System Inputs This requirement is effectively satisfied by directly measuring the position of the center test hole canister. While reactivity is the variable to be controlled, only indirect means are available to measure this variable. The proposed instrument channels will provide a positive protective action in the event that a very small amount of motion is measured. This tight range between normal conditions and actuation conditions will ensure that a timely response is generated in the safety system before any significant reactivity change can occur in the core due to unintended movement of the center test hole canister.

9. Capability for Sensor Checks This requirement is not applicable to these instrument channels. The proposed instrument channels will produce a logical state output, which through the existing 1/N logic system, will either continue to operate or SCRAM the reactor. Thus, only a single state of the instrument channel operability can be verified during reactor operation.

10. Capability for Test and Calibration This requirement is ,satisfied by the performance of the semiannual surveillance Compliance Check Procedure CP-36, quarterly physical inspection Preventive Maintenance Procedure RX-Q2, and weekly operability checks. Each of these tests provide specific detailed instructions for their respective purposes, and collectively ensures that the entire system can perform its intended safety function in the intended manner. No testing interval will be less than the normal time interval between shutdowns, so no provision will be allowed for testing during reactor operation.

11. Channel Bypass or Removal from Operation This requirement is not applicable to these instrument channels. The proposed instrument channels will be either in service during reactor operation, or in bypass during reactor operation. This condition will be determined administratively by the Reactor Manager. No change in the bypass status of these instrument channels will be allowed during reactor operation.

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12. Operating Bypasses This requirement is not applicable to these instrument channels. The proposed instrument channels will be either in service during reactor operation, or in bypass during reactor operation. This condition will be determined administratively by the Reactor Manager. No permissive conditions will be implemented for these instrument channels.

13. Indication of Bypasses This requirement is satisfied by the combination of the key-type bypass switch, its location on the Reactor Console, and the high level of administrative control required to operate the bypass switch. The bypass switch for these instrument channels will be located in the Reactor Console adjacent to the existing bypass switches. The switch will be a standardized key-type. The presence of the key in the switch will provide the operator a direct and unmistakable indication that these instrument channels are bypassed. Due to the high level of administrative control required for use, key presence and not key position is the first indication necessary to positively inform the operator of the bypass condition.

14. Access to Means for Bypassing This requirement is satisfied by the administrative control of the key used to operate the bypass switch. Reactor Manager's permission will be required to operate the key switch. The key will be kept in a locked key box, which is accessible only by licensed operators. In accordance with existing MURR convention, the key will be clearly identified as a key requiring Reactor Manager's permission to remove from the locked key box.

15. Multiple Set Points This requirement is not applicable to these instrument channels. The proposed instrument channels do not offer different modes of operation or sets of operating conditions. There will be no less restrictive set points associated with these instrument channels.

16. Completion of Protective Action Once Initiated This requirement is satisfied at the system level by the existing Non-Coincidence Logic Units (NCLU) and the Trip Actuator Amplifiers (TAA). The proposed instrument channels will provide a logic input to these safety system major components.

17. Manual Initiation This requirement is satisfied at the system level by the existing Manual SCRAM Switch. This switch remains independent of all other safety system actuation signals.

18. Access to Set Point Adjustments Set point adjustment, calibration adjustment and test points are combined in these instrument channels and represented by the sensor switches. Access to these switches is administratively controlled via performance of semiannual surveillance Compliance Check Procedure, quarterly physical inspection Preventive Maintenance Procedure, and weekly operability checks.

19. Identification of Protective Actions This requirement is satisfied by the Annunciator window and the Safety System Monitoring Circuit. Actuation of either of these instrument channels will result in an Annunciator window illumination, and a Safety System Monitoring Circuit indication that provide unmistakable indication and identification of the responsible instrument channel and the resulting SCRAM to the operator.

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20. Information Read-Out Actuation of either of these instrument channels will result in an Annunciator window illumination, and a Safety System Monitoring Circuit indication that provide unmistakable indication without providing anomalous or confusing indication to the operator.

21. System Repair Provisions for identifying malfunctioning components will be provided by the semiannual Surveillance Compliance Check Procedure, quarterly physical inspection Preventive Maintenance Procedure, and weekly operability checks. The mechanical components of this system can be removed from the reactor pressure vessel head and relocated to the pool bridge deck for inspection and repair. The electrical components of this system can be readily accessed for inspection and repair. Replacement parts are maintained in Spare Parts, including a complete support rig with switches installed.

22. Identification The visible channel components including cable runs and conduits will be identified in accordance with the existing MURR convention for reactor safety system channels. Safety System Yellow Leg channel components will be identified with yellow tape or other indelible marking. Safety System Green Leg channel components will be indentified with green tape or other indelible marking.

How would the reactor operator become aware of a closed circuit failure? How long could a closed circuitfailure exist without the reactor operatornot being aware of the failure? If this type of failure could exist for more than a short period of time, -'please justify the time period an undetectedfailure could exist.

In addition to the semiannual surveillance Compliance Check Procedure and the quarterly physical inspection Preventive Maintenance Procedure, an operability check will be performed prior to each reactor startup to ensure that a closed circuit failure has not occurred. The means used to accomplish this are detailed in the answer to Question 4 of this RAI. This condition in a single channel of the proposed instrument could exist for a maximum of one week without the operator becoming aware of the failure. During this time, the other instrument channel would continue to provide proper indication and be capable of performing its intended function. Also during this time, the root variable being detected, namely a change in reactivity, will be indirectly measured and protected against by the Nuclear Instrumentation and its independent input to the reactor safety system.

It appears that a single bypass switch (1S28) is used to bypass both channels of the FIRST system.

Why is a single switch used? Could this switch fail and create a closed circuitfailure in both channels of the FIRST system? If not, please explain.

No credible failure has been identified that would allow the bypass switch to fail and create a closed circuit failure in both channels of the FIRST system. As discussed in the design evaluation with regard to IEEE Standard 279-1971 above, the bypass switch provides full electrical independence and effective mechanical independence due to its construction quality, reliable track record, and high level of administrative control.

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Each contact block will be a separate physical component of the key switch, so electrical independence will remain intact. The common failure point then will be the mechanical operation of the key tumbler. These standardized key switches have a long track record at the MURR, with no record of failure or replacement due to reliability concerns. Mechanically, they are of sound construction and not prone to unintended movement without the key properly fitted. A high level of administrative control will be utilized in the operation of this bypass switch, such that the key will not be present or accessible unless the Reactor Manager has granted permission for its use, as required by proposed TS 3.3 Footnote (6), and directed the channels to be placed in bypass.

Therefore, no credible single event, physical or administrative, could result in a reduction of channel independence with regard to the bypass switch.

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ATTACHMENT 1 AP-RO-135 Revision 0 MURR MASTER COPY ADMINISTRATIVE PROCEDURE ISSUED SEP 2 7 ?i27 AP-RO-135 REACTOR. UTILIZATION REQUESTS RESPONSIBLE GROUP: Reactor Operations PROCEDURE OWNER: Les P. Foyto APPROVED BY: Les P. Foyto Date 9,- 7-z7 This procedure contains the following:

Pages 1 through 10 Attachments 1 through 4 Tables None through Figures None through Appendices None through Check-Off Lists None through

AP-RO-135 Revision 0 TABLE OF CONTENTS Section Page Number IN TRO D U CTION ................................................................................................................................ 3 1.0 PU RPO SE ................................................................................................................................. 3 2.0 SCO PE ...................................................................................................................................... 3 3.0 DEFIN ITION S ......................................................................................................................... 4 4.0 RE SPON SIBILITIES ....................................................................................................... 5 5.0 PRECAU TIO N S A N D LIM ITA TION S ................................................................................ 6 6.0 PRO CED U RE ........................................................................................................................... 7 7.0 REFEREN CES ....................................................................................................................... 10 8.0 RECO RD S .............................................................................................................................. 10 9.0 ATTA CHM EN TS ................................................................................................................... 10 2

AP-RO-135 Revision 0 REACTOR UTILIZATION REQUESTS INTRODUCTION All experiments conducted at the University of Missouri Research Reactor (MURR) must be reviewed and approved by the Reactor and Reactor Health Physics Managers (Reference 7.1).

The mechanism for obtaining such approval is a Reactor UtilizationRequest (RUR). The RUR describes the experiment in considerable detail. It presents the activities and isotopes that are produced and details the methods of handling the radioactive waste. The most important section of the RUR, and the one which is given paramount consideration in its preparation, is the safety analysis. The safety analysis includes all credible accident and transient scenarios to ensure that the experiment does not jeopardize the safe operation of the reactor or constitute a hazard to the safety of the facility staff and general public.

Experiments conducted at the MURR are subdivided into two general classifications: (1) neutron beam, and (2) neutron irradiation and isotope production. The neutron beam experiments are those research projects which utilize one of the beamports or the thermal column. The neutron irradiation and isotope production experimental facilities include the center test hole (flux trap),

the graphite reflector region, the pneumatic tube system, and in-pool locations external to the graphite reflector (bulk pool).

1.0 PURPOSE 1.1 To provide instructions for requesting, preparing, evaluating and approving an RUR. The review and approval process is based on requirements stated in the Hazards Summary Report (Reference 7.1) and the Technical Specifications (Reference 7.2) - both part of Amended Facility License R-103.

2.0 SCOPE 2.1 This procedure establishes the following:

Instructions on how to initiate an RUR,

" Instructions on what information is required to be part of the request to conduct an experiment,

Instructions on the format and content of the safety analysis to assure that the experiment meets the requirements of the Hazards Summary Report (Reference 7.1) and the Technical Specifications (Reference 7.2), and

The review process for an RUR.

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AP-RO- 135 Revision 0 3.0 DEFINITIONS 3.1 50.59 Evaluation - A documented evaluation that is prepared using the eight criterion presented in 10 CFR 50.59(c)(2). This documented evaluation is performed to determine if a proposed change, test or experiment requires prior NRC approval by means of a license amendment pursuant to 10 CFR 50.90 (Reference 7.3).

3.2 50.59 Screen - An assessment performed using established screening criteria to determine if an activity requires the preparation of a 50.59 Evaluation (Reference 7.3).

3.3 Experiment - An experiment is (a) any device or material which is exposed to significant radiation from the reactor and is not a normal part of the reactor or (b) any operation designed to measure or monitor reactor characteristics or parameters (Reference 7.2).

3.4 Hazards Summary Report (HSR) - The complete document consists of the original report and its five (5) addenda. For research and test reactors, the HSR includes information that describes the facility, presents the design bases and the limits on its operation. It also presents a safety analysis of the facility's structures, systems, and components (Reference 7.3).

3.5 PrincipleExperimenter- An individual who is responsible for conducting the RUR safety analysis. This individual may seek technical expertise from other MURR staff in completing the safety analysis.

3.6 Reactor Utilization Request (RUR) - The mechanism for the review and approval of a reactor experiment. The RUR is submitted by the RUR Requester with the assistance of other MURR staff using the "Instructions for Submitting a Reactor Utilization Request" (Attachments 1 and 2).

3.7 RUR Summary Sheet - A single page sheet which summarizes most of the limitations placed on the experiment.

3.8 Safety Analysis - The safety analysis, which will include a 50.59 Screen, demonstrates that the experiment meets all of the Hazards Summary Report (Reference 7.1) and Technical Specification (Reference 7.2) requirements and does not have the potential to adversely affect nuclear safety or safe facility operations.

3.9 Technical Specifications (TS) - The Technical Specifications represent an agreement between the licensee and the NRC on administrative controls, equipment availability, operational conditions and limits, and other requirements imposed on reactor facility operation in order to protect the environment and the health and safety of the facility staff and general public in accordance with 10 CFR 50.36 (Reference 7.2).

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AP-RO- 135 Revision 0 4.0 RESPONSIBILITIES 4.1 Reactor Manager:

4.1.1 Responsible for the RUR process.

4.1.2 Responsible for submitting the RUR to the Reactor Safety Subcommittee for its review if he determines that the proposed experiment represents a new class of experiment or a change to an existing experiment with safety significance.

4.1.3 Responsible for final approval of an RUR.

4.1.4 Responsible for performing an annual review of the RUR Program.

4.1.5 Responsible to possess and continually develop a precedence of past experiments which he has found to be safely conducted (Reference 7.1).

4.2 Reactor Health Physics Manager:

4.2.1 Responsible for assisting the Reactor Manager as required in evaluating the safety of an experiment; specifically the radiological protection section of the RUR safety analysis.

4.2.2 Responsible to possess and continually develop a precedence of past experiments which he has found to be safely conducted (Reference 7.1).

4.3 Assistant Reactor Manager-Physics:

4.3.1 Responsible for assisting the Reactor Manager as required in evaluating the safety of an experiment; specifically verifying the heat generation, sample decomposition and reactivity effect calculations of the RUR safety analysis.

4.4 PrincipleExperimenter

4.4.1 Responsible for preparing the RUR safety analysis in accordance with "Format and Content of the Safety Analysis" (Attachment 9.3 or 9.4).

4.4.2 Responsible for assisting the Reactor Manager with the annual review of the RUR Program.

4.4.3 Responsible for evaluating the need to update the experiment description.

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AP-RO- 135 Revision 0 4.0 RESPONSIBILITIES (CONT.)

4.5 RUR Requester:

4.5.1 Responsible for gaining approval from the appropriate Senior Manager prior to submitting an R UR.

4.5.2 Responsible for submitting an RUR in accordance with "Instructions for Submitting a Reactor Utilization Request" (Attachment 9.1 or 9.2).

4.6 Reactor Safety Subcommittee (RSSC):

4.6.1 Responsible for acting in behalf of the Reactor Advisory Committee in performing reviews of experiments significantly different from any previously reviewed or which involve a question pursuant to 10 CFR 50.59.

4.6.2 Responsible for referring an experiment to the Reactor Advisory Committee for review if the RSSC feels that the experiment involves unusual hazards, special conditions or that a potential safety hazard does or may exist.

4.6.3 Responsible for recommending alternatives for the experiment and report the conclusion to the Chairman of the Reactor Advisory Committee and the Reactdr Manager if the experiment review results in a negative recommendation.

4.7 Reactor Advisory Committee (RAC):

4.7.1 Responsible to review and make recommendations concerning experimental and operational activities at the facility.

4.7.2 Responsible for reviewing and making recommendations concerning proposed experiments significantly different from any previously reviewed or which involve a question pursuant to 10 CFR 50.59 (Reference 7.2).

5.0 PRECAUTIONS AND LIMITATIONS 5.1 The RSSC shall review proposed experiments that are significantly different from any previously reviewed or which involve a question pursuant to 10 CFR 50.59.

5.2 The precedence of past experiments is captured in the RUR safety analysis approved through MURR history. These safety analyses are retained in the Document Control System.

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AP-RO- 135 Revision 0 6.0 PROCEDURE 6.1 INITIATING A REACTOR UTILIZATION REOUEST 6.1.1 The RUR Requester will obtain approval from the appropriate Senior Manager prior to submitting 'an RUR.

6.1.2 Once approval has been obtained from the Senior Manager, Document Control will assign a number to the RUR.

6.1.3 The RUR Requester will then submit a request to conduct the experiment using the appropriate attachment:

a. For neutron irradiation or isotope production experiments, use Attachment 9.1.

b. For neutron beam experiments, use Attachment 9.2.

6.1.4 Once the information has been collected, the principle experimenter, with the assistance of other MURR staff with technical expertise, will complete a safety analysis using the appropriate attachment:

a. For neutron irradiation or isotope production experiments, use Attachment 9.3.

b. For neutron beam experiments, use Attachment 9.4.

Additional guidance for safety evaluation methodology may also be found in "Guideline/Training Manual for Preparing a Reactor Utilization Request" (Reference 7.4).

6.2 REACTOR UTILIZATION REQUEST REVIEWS NOTE: For the purposes of this procedure, a change to an existing experiment, which has safety significance, is also deemed to involve a question pursuant to 10 CFR 50.59.

6.2.1 Each reviewer in the review chain shall either approve the RUR and forward it to the next reviewer or disapprove the RUR and return to the principle experimenter for further analyses. If this occurs, the review process restarts.

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AP-RO-135 Revision 0 6.0 PROCEDURE (CONT.)

6.2.2 The Reactor Health Physics Manager will perform the initial review of the RUR. He will ensure that all necessary radiological calculations and control measures have been incorporated into the RUR. However, his review is not just limited to the above areas.

He may also recommend limitations or additional analyses in other areas. He will also determine if a Project Authorization will be required, based on information provided by the principle experimenter.

6.2.3 The Assistant Reactor Manager-Physics will perform the second review of the RUR. He will ensure that all necessary heat generation, sample decomposition and reactivity effect calculations have been incorporated into the RUR. However, his review is not limited to the above areas. He may also recommend limitations or additional analyses in other areas.

6.2.4 The Reactor Manager performs the third review of the RUR.

6.2.5 The Reactor Manager may approve the RUR at this time if the following two conditions are met. He may also apply additional limitations on the experiment that must be incorporated into the R UR by the principle experimenter.

a. The experiment does not involve a new class of experiment, AND

b. The experiment does not involve a question pursuant to 10 CFR 50.59.

6.2.6 IF the above two conditions are not met, THEN the Reactor Manager will refer the RUR to the RSSC for review.

6.2.7 The RSSC will review the RUR to determine if the experiment inv'olves a question pursuant to 10 CFR 50.59.

6.2.8 IF the RSSC determines that the experiment does not involve a potential safety hazard, THEN the review process is complete and the RSSC will forward its recommendation to the Reactor Manager.

6.2.9 IF the RSSC determines that the experiment may involve unusual hazards, special conditions or that a potential safety hazard does or may exist, THEN the RSSC will forward the RUR to the RAC for review.

6.2.10 The RAC will review the RUR if it has been referred to it by the RSSC. IF the RAC determines that the experiment does not involve a potential safety hazard, THEN the review process is complete and the RAC will forward its recommendation to the Reactor Manager.

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AP-RO- 135 Revision 0 6.0 PROCEDURE (CONT.)

6.2.11 If the RSSC and/or the RAC feel that a proposed experiment does introduce a potential safety hazard, the experiment proposal must be submitted to the NRC for final review.

The MURR staff will prepare the documents necessary for submittal to the NRC. MURR Management at this time may also decide not to proceed with the experiment.

6.3 REACTOR UTILIZATION REQUEST APPROVAL 6.3.1 After all experiment reviews have been completed, the Reactor Manager will make a final determination and approve the RUR as appropriate.

6.3.2 The Reactor Manager will assure copies of the approved RUR are distributed to:

a. Document Control (original),

b. The principle experimenter who initiated the RUR,

c. The RSSC, and

d. The RAC (if the RAC was involved in the review).

6.3.3 The principle experimenter will assure that an RUR Summary Sheet, if applicable, is generated and approved by the Reactor Manager.

6.4 REACTOR UTILIZA TION REQUEST PROGRAM ANNUAL REVIEW 6.4.1 The. Reactor Manager will review the R UR Program on an annual basis with the assistance of the principle experimenters. This review is not an in-depth review of each RUR, but rather a cursory review to determine if an RUR should remain active and if any recent changes to applicable regulations may have an affect on an RUR. The annual review will also include a review of this procedure.

6.5 REACTOR UTILIZATION REQUEST AMENDMENT PROCESS 6.5.1 A principle experimenter may request an amendment to an approved R UR.

6.5.2 The Reactor and Reactor Health Physics Managers may choose to revise an approved RUR or initiate a new RUR if the amendment is significant.

6.5.3 An amended RUR will follow the same review and approval process as a new RUR (e.g.,

a scale-up in mass or activity).

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AP-RO- 135 Revision 0

7.0 REFERENCES

7.1 MURR Hazards Summary Report 7.2 MURR Technical Specifications 7.3 AP-RR-003 "10 CFR 50.59 Evaluations" 7.4 "Guideline/Training for Preparing a Reactor UtilizationRequest" 8.0 RECORDS 8.1 Original RUR and RUR Amendments 9.0 ATTACHMENTS 9.1 Instructions for Submitting a Reactor UtilizationRequest for Neutron Irradiation and Radioisotope Production 1

Experiments 9.2 Instructions for Submitting a Reactor UtilizationRequest for Neutron Beam Experiments 9.3 Format and Content of the Safety Analysis for Neutron Irradiation and Radioisotope Production Experiments 9.4 Format and Content of the Safety Analysis for Neutron Beam Experiments 10

AP-RO- 135 Revision 0 INSTRUCTIONS FOR SUBMITTING A REACTOR UTILIZATION REQUEST FOR NEUTRON IRRADIATION AND RADIOISOTOPE PRODUCTION EXPERIMENTS INTRODUCTION No material will be irradiated in the center test hole, graphite reflector, bulk pool or pneumatic tube system unless it is in compliance with an approved Reactor Utilization Request (RUR). This request for experiment authorization should be limited to those target materials that are intended to be irradiated in a single primary encapsulation. A safety analysis will be performed by the principle experimenter based on the information provided. This analysis normally deals with, but is not limited to, the following criteria: criticality and/or reactivity considerations; heat generation considerations; shielding considerations; and off-gassing and/or chemical reactions.

A. The request should include a discussion of each item listed under C. The RUR Requester is encouraged to seek assistance from other MURR staff in the preparation of the request.

B. The request may be presented in typewritten or email form and must indicate approval by the MURR Senior Manager sponsoring the experiment before a safety analysis is commenced.

C. The following items should be addressed:

1. Provide a complete list of all target materials. All items that will be irradiated in the reactor such as flux monitor, quartz vials, aluminum vials, etc., should be included (include purity, percent enrichment, trace elements or impurities if known, chemical composition and/or alloy if known, estimated values if unknown).

2. Describe the physical form of all target materials. For liquids and powders include the average density.

3. If required, describe the secondary encapsulation for each target material. The normal minimum secondary encapsulation is sealed quartz or aluminum vials for liquid, powder or solid organic samples, and aluminum foil for inorganic solids (unless specifically requested and approved, the primary encapsulation for all samples will be one of the standard approved aluminum cans).

4. List the maximum target weight per secondary encapsulation unit.

5. List the maximum target weight per primary encapsulation unit.

6. What is the maximum neutron flux requested?

7. What is the maximum irradiation time at the above flux?

Page I of 2 Attachment 9.1

AP-RO-135 Revision 0 INSTRUCTIONS FOR SUBMITTING A REACTOR UTILIZATION REQUEST FOR NEUTRON IRRADIATION AND RADIOISOTOPE PRODUCTION EXPERIMENTS (CONT.)

8. Provide a reasonable estimate of byproduct material activities generated for an irradiation at the above maximum time and flux.

9. What is the sample configuration? If the arrangement of your secondary encapsulation units is critical, describe the arrangement and include drawings with enough detail iso that heat transfer flow paths can be calculated. If it is not critical, the reactor staff will determine the configuration, taking into consideration heat transfer, flux gradients, and sample decapsulation.

10. Include any additional information that you feel will aid in preparing the safety analysis.

Page 2 of 2 Attachment 9.1

AP-RO-135 Revision 0 INSTRUCTIONS FOR SUBMITTING A REACTOR UTILIZATION REQUEST FOR NEUTRON BEAM EXPERIMENTS INTRODUCTION No experiment should be performed in a beamport or the thermal column without an approved Reactor Utilization Request (RUR) or Project Authorization. The Reactor and Reactor Health Physics Managers will determine, based on the information provided, which one is applicable. If an RUR is required, a safety analysis will be performed based on the information provided. This analysis normally deals with, but is not limited to, the following hazards:

1. Changes in reactivity due to activities such as draining or flooding a beamport.

2. Exposure of personnel to radiation as a result of movements of shielding, inadequate shielding or reflected beams.

3. Release of radioactive gases such as argon-41 that may be produced in the beamport.

4. Production of explosive or toxic materials in the beamport or attached experiment.

A. The request should include a discussion of each item listed under C. The RUR Requester is encouraged to seek assistance from other MURR staff in the preparation of the request.

B. The request may be presented in typewritten or email form and must indicate approval by the MURR Senior Manager sponsoring the experiment before a safety analysis is commenced.

C. The following items should be addressed:

1. Detailed description of the experiment. Include a list of facility equipment and services needed (hot cell, laboratory hood, air, gas, vacuum, demineralized water, electricity, etc.)

2. Which facility is requested for this experiment? If a beamport, state which one.

3. Will this experiment change the configuration within the biological shielding of the reactor?

4. Anticipated duration of the experiment?

5. Where materials will be exposed to neutrons, provide a list of materials and their weight in grams. For enriched targets, give the percentage of each isotope.

6. Estimated'maximum activity of byproduct material in millicuries.

Page I of 2 Attachment 9.2

AP-RO-135 Revision 0 INSTRUCTIONS FOR SUBMITTING A REACTOR UTILIZATION REQUEST FOR NEUTRON BEAM EXPERIMENTS (CONT.)

7. Desired decay time before receipt of irradiated materials and estimated isotope activities after this decay time.

8. Description of encapsulation of target materials.

9. Any special operating instructions or experiment limitations. Include as appropriate the limiting irradiation conditions, i.e., the maximum neutron flux, maximum weight of target materials, maximum sample exposure time, etc.

10. State the disposition of all radioactive waste material including experiment apparatus.

Where radioactive material will be shipped, indicate the desired arrangements.

11. Is a liquid nitrogen cryostat required for the experiment?

12. Include any additional information that you feel will aid in preparing the safety analysis.

Page 2 of 2 Attachment 9.2

AP-RO- 135 Revision 0 FORMAT AND CONTENT OF THE SAFETYANALYSIS FOR NEUTRON IRRADIATION AND RADIOISOTOPE PRODUCTION EXPERIMENTS A. Prepare a cover sheet which includes the assigned RUR number, material descriptions, name of the principle experimenter, and signature/date blocks for review and approval for the Reactor Health Physics Manager, Assistant Reactor Manager-Physics, and the Reactor Manager.

B. Prepare a detailed description of the experiment or irradiation to include the objective of the experiment, target description, target form, target enrichment(s), mass limits, primary encapsulation, secondary encapsulation, expected irradiation position, estimates of byproduct activity.

C. Prepare a draft RUR Summary Sheet.

D. Prepare a 50.59 Screen for the irradiation or experiment that will demonstrate and document that the safety analysis meets all applicable HSR and TS Requirements.

1. Thermal Analysis - An estimation of the heat generation and heat transfer rates for an experiment, determining if a cooling design change is required to prevent the surface temperature of a submerged irradiated sample from exceeding the saturation temperature of the liquid it is submerged it.

This analysis ensures compliance with TS 3.6.h, which states, "Cooling shall be provided to prevent the surface temperature of a submerged irradiation experiment from exceeding the saturation temperature of the cooling medium."

This specification is intended to reduce the likelihood of reactivity transients due to accidental voiding in the reactor or the failure of an experiment from internal or external heat generation.

2. Sample Decomposition-Pressure Analysis - Describes the form of the sample and component materials of the experiment during irradiation, with reasonable leeway for normal and abnormal conditions. The analysis should confirm that a potential pressure buildup due to a complete decomposition of the sample material will not exceed the design pressure of the irradiation container.

This analysis ensures compliance with TS 3.6.i, which states, "Irradiation containers to be used in the reactor, in which a static pressure will exist or a pressure buildup is expected, shall be designed and tested for a pressure exceeding the maximum expected pressure by at least a factor of two (2)."

This specification is intended to reduce the likelihood of damage to the reactor and/or radioactivity releases from experiment failure.

Page I of 3 Attachment 9.3

AP-RO-135 Revision 0 FORMAT AND CONTENT OF THE SAFETY ANALYSIS FOR NEUTRON IRRADIATION AND RADIOISOTOPE PRODUCTION EXPERIMENTS (CONT.)

3. Experiment Failure Analysis - Used to determine if products or components from the experiment have the potential to violate the limits of 10 CFR 20, Appendix B, Table I, if released to the atmosphere.

This analysis ensures compliance with TS 3.6.c, which states, "Where the possibility exists that the failure of an experiment could release radioactive gases or aerosols to the reactor bay or atmosphere, the experiment shall be limited to that amount of material such that the airborne concentration of radioactivity averaged over a year will not exceed the limits of Appendix B, Table I of 10 CFR Part 20. Exception:

Fueled experiments (See Specification 3.6.a)."

4. Loss of Coolant Analysis - Describes how a loss of coolant (e. g., loss of pool coolant flow, loss of experiment cooling, etc.) to the experiment will not result in a release of radioactivity to the atmosphere or effect the safe operation/control of the reactor.

This analysis ensures compliance with TS 3.6.f, which states, "Experiments shall be designed and operated so that identifiable accidents such as loss of reactor flow, loss of experiment cooling, etc., will not result in a release of fission products or radioactive materials from the experiment."

5. Failure of Other Experiments Analysis - Used to identify the possible effects upon reactor control and other experiments due to operating an experiment under abnormal conditions (failure).

This analysis ensures compliance with TS 3.6.g, which states, "Experiments shall be designed such that a failure of an experiment will not lead to a direct failure of other experiments, a failure of reactor fuel elements, or to interference with the action of the reactor control elements or other operating components.

6. Corrosion Analysis - If corrosive materials are expected to be generated in an appreciable quantity during normal operation or as a result of the experiment failing, this analysis must show that the encapsulation provides enough corrosion resistance to endure the worst scenario of corrosion for the duration of the experiment.

This analysis ensures compliance with TS 3.6.j, which states, "Corrosive materials shall be doubly encapsulated in corrosion-resistant containers to prevent interaction with reactor components or pool water."

Page 2 of 3 Attachment 9.3

AP-RO-135 Revision 0 FORMAT AND CONTENT OF THE SAFETYANALYSIS FOR NEUTRON IRRADIATION AND RADIOISOTOPE PRODUCTION EXPERIMENTS (CONT.)

7. Explosive Analysis - Ensures that if explosive materials are present or are expected to be formed during the irradiation then the total mass of the explosive will not exceed the TS limitation.

This analysis ensures compliance with TS 3.6.d, which states, "Explosive materials shall not be irradiated or allowed to generate in any experiment in quantities over 25 milligrams."

E. Reactor chemistry analysis to indicate the ability to detect water-soluble radioisotopes in pool water in the event of encapsulation failure.

F. Post irradiation procedures will be specified for sample handling, hot cell activities, and distribution to MURR researchers or MURR radiation material shippers.

G. Health Physics analyses that consider radiological safety and protection issues including but not limited to:

a. Experiment failure analysis.

b. Evaluation of post irradiation handling requirements, including decay time before removing from the reactor pool.

c. Evaluation of Project Authorization limits for samples distributed to MURR experimenters to assure procedures, training and equipment are adequate for the byproduct material activity.

d. Evaluation of adequacy of procedures specified for Preparation and Shipping of Radioactive Byproduct Material.

e. Evaluate the residual activity that may become waste and methods of handling and disposition.

H. Reactor Physics analysis that shows the experiment will meet TS. 3.1 requirements for experiment reactivity worth. An estimate of reactivity worth will be performed for initial irradiation and verified by measurement as required.

Page 3 of 3 Attachment 9.3

AP-RO- 13 5 Revision 0 FORMAT AND CONTENT OF THE SAFETYANALYSIS FOR NEUTRON BEAM EXPERIMENTS A. Prepare a cover sheet which includes the assigned R UR number, material descriptions, name of the principle experimenter, and signature/date blocks for review and approval for the Reactor Health Physics Manager, Assistant Reactor Manager-Physics, and the Reactor Manager.

B. Prepare detailed description of the experiment including objective of the experiment. Include a list of facility equipment and services needed (air, gas, vacuum, demineralized water) and how they will be supplied.

C. Provide scale drawings of materials that will placed inside the biological shield (e.g., in the collimator or collimator liner). Include design information including specific materials of construction. Show compatibility of materials with aluminum and estimates of activation.

D. Provide scale drawings of materials and equipment that will be external to the biological shield (e.g., shielding, monochromator, etc.). Include materials of construction and targets.

Provide estimates of material activation.

E. Provide anticipated duration of experiment and provide names of sponsor and principle users, including students, who will need to document their user training and be included in a Project Authorization, if necessary.

F. If applicable, attach a copy of the Project Authorization which should include any special operating instructions or experiment limitations.

G. Indicate the intended disposition of radioactive material including experiment apparatus after experiment is completed.

H. Prepare a draft RUR Summary Sheet, if applicable.

I. Prepare a 50.59 Screen for the irradiation or experiment that will demonstrate and document that the safety analysis meets all applicable HSR and TS Requirements.

1. Thermal Analysis - An estimation of the heat generation and heat transfer rates for an experiment, determining if a cooling design change is required to prevent the surface temperature of a submerged irradiated sample from exceeding the saturation temperature of the liquid it is submerged it.

This analysis ensures compliance with TS 3.6.h, which states, "Cooling shall be provided to prevent the surface temperature of a submerged irradiation experiment from exceeding the saturation temperature of the cooling medium."

Page I of 3 Attachment 9.4

AP-RO- 135 Revision 0 FORMAT AND CONTENT OF THE SAFETYANALYSIS FOR NEUTRON BEAM EXPERIMENTS (CONT.)

This specification is intended to reduce the likelihood of reactivity transients due to accidental voiding in the reactor or the failure of an experiment from internal or external heat generation.

2. Sample Decomposition-Pressure Analysis - Describes the form of the sample and component materials of the experiment during irradiation, with reasonable leeway for normal and abnormal conditions. The analysis should confirm that a potential pressure buildup due to a complete decomposition of the sample material will not exceed the design pressure of the irradiation container.

This analysis ensures compliance with TS 3.6.i, which states, "Irradiation containers to be used in the reactor, in which a static pressure will exist or a pressure buildup is expected, shall be designed and tested for a pressure exceeding the maximum expected pressure by at least a factor of two (2)."

This specification is intended to reduce the likelihood of damage to the reactor and/or radioactivity releases from experiment failure.

3. Exe i - Used to determine if products or components from the

3. Eperiment Failure Analysis -Ue odtriei rdcso opnnsfo h experiment have the potential to violate the limits of 10 CFR 20, Appendix B, Table I, if released to the atmosphere.

This analysis ensures compliance with TS 3.6.c, which states, "Where the possibility exists that the failure of an experiment could release radioactive gases or aerosols to the reactor bay or atmosphere, the experiment shall be limited to that amount of material such that the airborne concentration of radioactivity averaged over a year will not exceed the limits of Appendix B, Table I of 10 CFR Part 20. Exception:

Fueled experiments (See Specification 3.6.a)."

4. Loss of Coolant Analysis - Describes how a loss of coolant (e. g., loss of pool coolant flow, loss of experiment cooling, etc.) to the experiment will not result in a release of radioactivity to the atmosphere or effect the safe operation/control of the reactor.

This analysis ensures compliance with TS 3.6.f, which states, "Experiments shall be designed and operated so that identifiable accidents such as loss of reactor flow, loss of experiment cooling, etc., will not result in a release of fission products or radioactive materials from the experiment."

Page 2 of 3 Attachment 9.4

AP-RO-135 Revision 0 FORMAT AND CONTENT OF THE SAFETYANALYSIS FOR NEUTRON BEAM EXPERIMENTS (CONT.)

5. Failure of Other Experiments Analysis - Used to identify the possible effects upon reactor control and other experiments due to operating an experiment under abnormal conditions (failure).

This analysis ensures compliance with TS 3.6.g, which states; "Experiments shall be designed such that a failure of an experiment will not lead to a direct failure of other experiments, a failure of reactor fuel elements, or to interference with the action of the reactor control elements or other operating components."

6. Corrosion Analysis - If corrosive materials are expected to be generated in an appreciable quantity during normal operation or as a result of the experiment failing, this analysis must show that the encapsulation provides enough corrosion resistance to endure the worst scenario of corrosion for the duration of the experiment.

This analysis ensures compliance with TS 3.6.j, which states, "Corrosive materials shall be doubly encapsulated in corrosion-resistant containers to prevent interaction with reactor components or pool water."

7. Explosive Analysis - Ensures that if explosive materials are present or are expected to be formed during the irradiation then the total mass of the explosive will not exceed the TS limitation.

J. Reactor Health Physics analyses that consider radiological safety and ALARA considerations including, but not limited to:

1. Experiment failure analysis.

2. Evaluate the residual activity that may become waste and methods of handling and disposition.

K. Reactor Physics analysis that estimates possible reactivity effect of materials inside biological shield for beamport experiments. Will beamport status (flooded or drained) present a potential problem with neutron coupling to Nuclear Instrumentation?

L. Evaluate shielding requirements and design.

.Page 3 of 3 Attachment 9.4

ATTACHMENT 2 NUMBER: CP-36 COMPLIANCE CHECK PROCEDURE PAGE: 1 OF2 REVISION: DRAFT Compliance Check Frequency: Semi-annually FIRST Scrams Plant Conditions Number of Men Needed: 2 Reactor Shutdown Pool System Secured Estimated Time: 30 minutes FIRST Rig Installed Strainer Installed Test Equipment, Tools and Materials

1. Jumpers

2. Shorted Relays

3. Bypass Key References Technical Specifications 3.1.g, 3.1.h Print No. 139 Procedure FIRST - Scram (Yellow Lea) 1 . Install Dummy Load Test Connectors.

2. Install jumper G-3 (bypass green leg of safety system).

3. Install jumper Y-2 (reactor loop 'BI' low flow FT-912G).

4. Install jumper Y-3 (Valve 509 off open).

5. Install jumper Y-4 (reactor outlet low press PT-944A).

6. Install jumper Y-5 (power level interlock).

7. Remove relay K-25 (pressurizer low press PS-938) and install shorted relay in K-25 position.

8. Remove relay K-30 (reactor loop 'Al' low flow FT-912A) and install shorted relay in K-30 position.

9. Remove relay K-31 (pool loop low flow FT-912F) and install shorted relay in K-31 position.

10. Place Master Control Switch 1S1 to 'On' position.

11. Place Magnet Current Switch 1S14 to 'On' position.

12. Reset scram TAAs.

13. Remove strainer.

14. VERIFY scram TAA's and magnet current to zero

15. Place FIRST Scram Bypass Switch 1S28 to 'Bypass' position.

16. Reset scram TAAs.

17. Remove Bypass Key 1S28.

18. VERIFY scram TAAs and magnet current to zero

19. Install strainer.

20. Place Magnet Current Switch 1S14 to 'Off' position.

21. Place Master Control Switch 1S1 to 'Off' position.

22. Remove jumper Y-5.

23. Remove jumper Y-4.

24. Remove jumper Y-3.

25. Remove jumper Y-2.

APPROVED:

Reactor Manager

NUMBER: CP-36 COMPLIANCE CHECK PROCEDURE PAGE: 2 OF 2 REVISION: DRAFT

26. Remove jumper G-3.

27. Remove shorted relay from K-31 position and install relay K-31.

28. Remove shorted relay from K-30 position and install relay K-30.

29. Remove shorted relay from K-25 position and install relay K-25.

FIRST - Scram (Green Leg)

30. Install jumper Y-1 (bypass yellow leg of safety system).

31. Install jumper G-5 (power level interlock).

32. Install jumper G-9 (reactor loop low press PT-943).

33. Install jumper G-10 (reactor outlet low press PT-944B).

34. Install jumper G-1 1 (reactor loop '82' low flow FT-912H).

35. Install jumper G-27 (Reflector D/P PT-917).

36. Remove relay K-26 (pressurizer high press PS-939) and install shorted relay in K-26 position.

37. Remove relay K-37 (pool loop low flow FT-912D) and install shorted relay in K-37 position.

38. Remove relay K-38 (reactor loop 'A2' low flow FT-912E) and install shorted relay in K-38 position.

39. Place Master Control Switch lS1 to 'On' position.

40. Place Magnet Current Switch 1S14 to 'On' position.

41. Reset scram TAAs.

42. Remove strainer.

43. VERIFY scram TAAs and magnet current to zero

44. Place FIRST Scram Bypass Switch 1S28 to 'Bypass' position.

45. Reset scram TAAs.

46. Remove bypass key 1S28.

47. VERIFY scram TAAs and magnet current to zero

48. Place Magnet Current Switch 1S14 to 'Off' position.

49. Place Master Control Switch lS1 to 'Off position.

50. Remove jumper G-27.

51. Remove jumper G-1 1.

52. Remove jumper G-10.

53. Remove jumper G-9.

.54. Remove jumper G-5.

55. Remove jumper Y-1.

56. Remove shorted relay from K-38 position and install relay K-38.

57. Remove shorted relay from K-37 position and install relay K-37.

58. Remove shorted relay from K-26 position and install relay K-26.

59. ENSURE all jumpers, shorted relays and bypass keys have been removed.

60. Remove Dummy Load Test Connectors and connect control rod drive mechanism cables.

61. Sign and date datasheet.

62. Log CP completed in the Console Log Book and Maintenance Day Book.

Date Completed:

LSRO Signature:

ATTACHMENT 3 NUMBER: RX-Q2 PREVENTIVE MAINTENANCE PROCEDURE PAGE: 1 OF 1 REVISION: DRAFT Preventive Maintenance: PM System: Reactor Control Inspect FIRST Support Rig PM Components: FIRST Plant Conditions: Frequency: Quarterly Reactor Shutdown Pool System Shutdown Number of Men Needed: 2 Flux Trap Holder and Strainer Removed Estimated Time: 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Test Equipment, Tools and Materials

1. Various hand tools

2. Blotter paper

3. Contamination controls References Modification Record [preliminary long form]; Flux-trap Irradiations Reactivity Safety Trip (FIRST)

Procedure

1. Disconnect FIRST support rig from Pressure Vessel Head.

2. Using proper radiological precautions, raise FIRST support rig to the bridge area.

3. Inspect FIRST support rig for the following:

a. Structural integrity, tightness, and no warping

b. Smooth operation of both switches

c. Smooth operation of both engagement mechanisms

4. Report any abnormality to the Assistant Reactor Manager - Engineering for further guidance.

5. Lower FIRST support rig to the Pressure Vessel Head area.

6. Reconnect FIRST support rig on Pressure Vessel Head OR place in storage.

7. Log PM completed in the Console Log Book and Maintenance Day Book.

Date Completed:

LSRO Signature:

APPROVED:

Assistant Reactor Manager

ATTACHMENT 4 TECHNICAL SPECIFICATION UNIVERSITY OF MISSOURI RESEARCH REACTOR FACILITY Number 3.3 Page 3 of 5 Date Amendment No.

SUBJECT:

Reactor Safety System (continued)

Manual Scram 1 1 1 Push Button at Control Console Center Test Hole 2(6) 2(6) 2(6) Scram as a result of removing the center test hole removable experiment test tubes (I) Flow orifice or heat exchanger AP (psi) in each operating heat exchanger leg corresponding to the flow value in the table.

(2) Not required below 50 KW operation if natural convection flange and pressure vessel cover are removed or in operation with the reactor subcritical by a margin of at least 0.0 15 AK.

(3) Trip pressure is that which corresponds to the pressurizer pressure indicated in the table with normal primary coolant flow.

(4) Flow orifice AP (psi) corresponding to the flow value in the table.

(5) Core AP (psi) corresponding to the core flow value in the table.

(6) Not required if reactivity worth of the center test hole removable experiment test tubes and its contents is less than the reactivity limit of specification 3.6.h. This safety function shall only be bypassed with specific authorization from the Reactor Manager.

Bases

a. The specifications on high power, primary coolant flow, primary coolant pressure, and reactor inlet temperature provide for the safety system settings outlined in

specifications 2.2.a, 2.2.b, and 2.2.c. In Mode I and II operation the core differential temperature is approximately 17°F.

TECHNICAL SPECIFICATION UNIVERSITY OF MISSOURI RESEARCH REACTOR FACILITY Number 3.3 Page 5 of 5 Date Amendment No.

SUBJECT:

Reactor Safety System (continued)

The scrams from the primary and pool coolant isolation valves (507A/B and 509) leaving their full open position provide a first line of protection for a loss of flow accident in that system initiated by an inadvertent closure of the isolation valve/s.

The power level interlock (PLI) scram provides assurance that the reactor cannot be operated with a power level greater than that authorized for the mode of operation selected on the Mode Selector Switch. The PLI scram also provides the interlock to assure that the reactor cannot be operated in Mode I with a pool or primary coolant flow scram by-passed.

The facility evacuation and reactor isolation scrams provide assurance that the reactor is shutdown for any condition which initiates or leads to the initiation of an evacuation or isolation.

The manual scram provides assurance that the reactor can be shutdown by the operator if an automatic function fails to initiate a scram or if the opera-tor detects an impending unsafe condition prior to the automatic scram initiation.

The center test hole scram provides assurance that the reactor can not be operated unless the removable experiment test tubes or strainer is inserted and latched in the center test hole. This is required any time the reactivity worth of the center test hole removable experiment test tubes and the contained experiments exceeds the limit of specification 3.1 .h.

(Ref. Section 3.5 of Add. 3 to HSR). The center test hole scram may be.

bypassed if the total reactivity worth of the removable experiment test tubes and the contained experiments does not exceed the limit of specification 3.1 .h and is authorized by the Reactor Manger.

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