ML19263A663

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Univ. of Missouri - Written Communication as Specified by 10 CFR 50.4(b)(l) Requesting U.S. Nuclear Regulatory Commission Approval to Revise the Technical Specifications Appended to Renewed Facility Operating License No. R-103 Pursuant to 1
ML19263A663
Person / Time
Site: University of Missouri-Columbia
Issue date: 09/18/2019
From: Meffert B, Robertson J
Univ of Missouri - Columbia
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML19263A663 (31)
v • d • e


Text

~University of Missouri '~.I

., Research Reactor Center 1513 Research Park Drive Columbia, MO 65211 PHONE 573-882-4211 WEB murr.missouri.edu September 18, 2019 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001

REFERENCE:

Docket No. 50-186 University of Missouri-Columbia Research Reactor Renewed Facility Operating License No. R-103

SUBJECT:

Written communication as specified by 10 CPR§ 50.4(b)(l) requesting U.S.

Nuclear Regulatory Commission approval to revise the Technical Specifications appended to Renewed Facility Operating License No. R-103 pursuant to 10 CPR § 50.90 Enclosed is an application to amend Renewed Facility Operating License No. R-103 by revising*

University of Missouri Research Reactor (MURR) Technical Specification (TS) 1.26, "Reactor Secured;" TS 3.4, "Reactor Containment Building;" TS 4.4, "Reactor Containment Bui,lding;" and TS 6.4, "Procedures," pursuant to 10 CPR§ 50.90. The proposed, requested changes should be considered permanent changes to the MURR TSs. Additionally, the proposed, requested changes were favorably reviewed on September 10, 2019, by the Reactor Safety Subcommittee (RSS), a subcommittee of the Reactor Advisory Committee (RAC), in accordance with TS 6.2.a(4).

Enclosure 1 provides the basis for revising TS 1.26, Enclosure 2 provides the basis for revising TS 3.4, Enclosure 3 provides the basis for revising TS 4.4, and Enclosure 4 provides the basis for revising TS 6.4. Enclosure 5 contains the proposed, revised TS pages with track changes.

Enclosure 6 contains the proposed, revised TS pages with changes accepted and revision bars.

MURR requests that the proposed license amendment be reviewed within 24 months from the date of this letter. If there are any questions regarding this license amendment request, please contact me at (573) 882-5118 or MeffertB@missouri.edu. I declare under penalty of perjury that the foregoing is true and correct.

Sincerely, ENDORSEMENT:

Reviewed and Approved,

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Su~cribed a n ~ ~ m ~ i s l15 day of , ft

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J A ~ A ° " s , Notary Public

  • . David Robertson My Commission Expires: March 26, 2023 Reactor Manager Reactor Facility Director JACQUEUNE L IM1YAS My Colllnlsabl &pie$

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cc: Reactor Advisory Committee Reactor Safety Subcommittee Isotope Use Subcommittee Dr. Mark McIntosh, Vice Chancellor for Research, Graduate Studies and Economic Development Mr. Geoffrey Wertz, U.S. Nuclear Regulatory Commission Mr. William Schuster, U.S. Nuclear Regulatory Commission

Enclosures:

1. Basis for the Requested Change to Technical Specification 1.26
2. Basis for the Requested Change to Technical Specification 3.4
3. Basis for the Requested Change to Technical Specification 4.4
4. Basis for the Requested Change to Technical Specification 6.4
5. Proposed, revised Technical Specification pages - A-4, A-24, A-25, A-47 and A-70 (with track changes)
6. Proposed, revised Technical Specification pages - A-4, A-24, A-25, A-47 and A-70 (with accepted changes, revision bars)

Page 2 of2

Enclosure 1 - Basis for the Requested Change to Technical Specification 1.26 1.0 Introduction The University of Missouri Research Reactor (MURR) is requesting a change to Technical Specification (TS) 1.26, "Reactor Secured." This change will revise conditions b.(2) and b.(4) of Specification 1.26 in order to allow the installation of shim rod drive mechanism (SRDM) substitute plugs, which will provide greater flexibility in satisfying the requirements of TS 1.26 (Note: the shim rod drive mechanisms are also referred to as control rod drive mechanisms in the MURR Safety Analysis Report). Allowing the use of SRDM substitute plugs to meet the conditions of Specification 1.26 will reduce the time that reactor containment integrity is required by TS 3.4, "Reactor Containment Building," specifically Specification 3.4.b, while the reactor is shut down for maintenance. Minimizing the time that reactor containment integrity is required during reactor maintenance activities will help reduce the chance that equipment failure or human error would cause MURR to deviate from TS 3.4.b. Additionally, minimizing the time that reactor containment integrity is required during reactor maintenance will also allow more scheduling flexibility for maintenance activities with a consequent efficiency increase in both personnel availability and reactor utilization.

The following license amendment request (LAR) provides justification that installation of SRDM substitute plugs confirms that the reactor is secured by ensuring the shim blades (rods) are physically decoupled from the SRDMs. It also ensures that a SRDM cannot be installed without removing the substitute plug. The addition of SRDM substitute plugs to TS 1.26, "Reactor Secured," does not decrease the safety of reactor operations nor does it increase hazards or risks to the health and safety of the public.

2.0 Background Original Facility Operating License No. R-103, issued on October 11, 1966, contained definition TS 1.1, "Reactor Secured," which stated:

"The reactor shall be considered secured whenever it contains insufficient fuel in the reactor to establish criticality with all control rods removed or whenever all of the following conditions are met:

(a) All shim rods are fully inserted.

(b) The "Master Control" switch is on the "off' position and the key is removed and in the custody of a licensed operator..

(c) No work is in progress involving handling of fuel or involving maintenance of the core structure, including control rods or their drives."

Amendment No. 2 to Amended Facility Operating License No. R-103, issued on July 9, 1974, contained definition TS 1.20, "Reactor Secured," which stated:

"The reactor shall be considered secured whenever it contains insufficient fuel in the reactor core to establish criticality with all control rods removed or whenever all of the following conditions are met:

Page 1 of 8

Enclosure 1 - Basis for the Requested Change to Technical Specification 1.26

a. All shim rods are fully inserted.
b. The "Master Control" switch is in the "off' position.
c. The "Master Control" switch key is removed and locked in the key box or is in the custody of a licensed operator.
d. No work is in progress involving transferring fuel in or out of the core.
e. No work is in progress involving control rods or control rod drives.
f. The reactor pressure vessel cover is secured in position and no work is in progress on the pressure vessel or its supports."

Amendment No. 14 to Amended Facility Operating License No. R-103, issued on April 15, 1981, revised definition TS 1.20, "Reactor Secured," to allow dummy load test connectors to be part of the definition.

With dummy load test connectors installed, it is impossible to move the shim control rods, even inadvertently, from the reactor control console. Amendment No. 14 definition TS 1.20 stated:

"The reactor shall be considered secured whenever it contains insufficient fuel in the reactor core to establish criticality with all control rods removed or whenever all of the following conditions are met:

a. All shim rods are fully inserted.
b. One of the following conditions exists:

(1) The "Master Control" switch is in the "off' position with the key locked in the key box or in the custody of a licensed operator.

(2) A licensed operator is present in the Control Room and the dummy load control rod test connectors are installed.

c. No work is in progress involving transferring fuel in or out of the core.
d. No work is in progress involving the control rods or control rod drives with the exception of installing or removing dummy load control rod test connectors.
e. The reactor pressure vessel cover is secured in position and no work is in progress on the pressure vessel or its supports."

The existing TS definition of "Reactor Secured" became effective with the issuance of Renewed Facility Operating License No. R-103 on January 4, 2017. TS 1.26, "Reactor Secured," currently states:

"The reactor shall be considered secured when:

Page 2 of 8

Enclosure 1 - Basis for the Requested Change to Technical Specification 1.26

a. There is insufficient fuel in the reactor core to attain criticality with optimum available conditions of moderation and reflection with all four (4) shim blades (rods) removed, OR
b. Whenever all of the following conditions are met:

(1) All four shim rods are fully inserted; (2) One of the two following conditions exists:

1. The Master Control Switch is in the "OFF" position with the key locked in the key box or in custody of a licensed operator, OR ii. The dummy load test connectors are installed on the shim rod drive mechanisms and a licensed operator is present in the reactor control room; (3) No work is in progress involving the transfer of fuel in or out of the reactor core; (4) No work is in progress involving the shim blades (rods) or shim rod drive mechanisms with the exception of installing or removing the dummy load test connectors; and (5) The reactor pressure vessel cover is secured in position and no work is in progress on the reactor core assembly support structure."

Prior to removing one (1) of the conditions that is required for reactor containment integrity to exist (TS 3.4.a), two (2) licensed operators independently verify that all of the conditions for relaxing reactor containment integrity, as specified by TS 3.4.b, are met. This verification practice is directed by Step 6.1.9.a of reactor administrative procedure AP-R0-110, "Conduct of Operations," and is reiterated by continuous training. The Lead Senior Reactor Operator (LSRO) and another licensed operator, who is designated by the LSRO, check appropriate indications to determine that the conditions of TS 1.26, "Reactor Secured," are met prior to "relaxing containment integrity" and removing one ( 1) of the conditions of TS 3.4.a. These independent verifications are a very good administrative barrier to ensure MURR adheres to TS 3.4.b. However, the installation of SRDM substitute plugs when the SRDMs are removed will provide an additional physical barrier to ensure MURR adheres to TS 3.4.b even ifby human error or equipment failure one (1) of the conditions of TS 3.4.a was not met.

The installation of SRDM substitute plugs during SRDM maintenance activities will allow greater scheduling flexibility for other maintenance activities like opening the truck entry door (Door 101) to allow large items to be transferred into and out of the reactor containment building. This will provide a consequent efficiency increase in both personnel availability and reactor utilization.

Page 3 of 8

Enclosure 1 - Basis for the Requested Change to Technical Specification 1.26 3.0 Justification for Changing Technical Specification 1.26 When the " reactor secured" definition is met, the reactor core and its support equipment are in a condition where attaining criticality inadvertently is nearly impossible. The definition requires either there is insufficient fuel in the reactor core to attain criticality with optimum available conditions of moderation and reflection with all four (4) shim blades removed, or the four ( 4) shim blades are inserted and the SRDMs are in a condition to ensure all four ( 4) shim blades cannot be inadvertently withdrawn. Any SRDM substitute plug installed in place of a SRDM will ensure that the associated shim blade (rod) could not be withdrawn to any position. Furthermore, the SRDM substitute plug does not have a means to withdraw a shim blade (rod); it is only a plug made from aluminum pipe, which prevents the withdrawal of a shim blade (rod) - see Figures 1 and 2.

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Enclosure 1 -Basis for the Requested Change to Tecbnical Specification 1.26 In order to install a SRDM substitute plug, the SRDM must be completely removed, which eliminates the normal ability to withdraw the associated shim blade (rod). In addition, the substitute plug removes any access to the shim blade (rod) anvil and pull rod from inside the upper housing. If some unknown force were to lift a shim blade (rod) upward, the anvil connected to that blade could only move approximately one (1) inch before the SRDM substitute plug prevented further withdrawal. The shim blade (rod) could not significantly move out of the reactor core because when the shim blades (rods) are fully inserted, the tip of the shim blade (rod) extends approximately one (1) inch below the core. this movement would result in an unmeasurable change in reactor core reactivity- see Figure 3.

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UPPER.HOUSING UPPER HOUSING Fl.ANOE FLANGE ANVIL SHIM BLADE SlUMB1..ADE SWMBLADE PULLROD PULL ROD PUUROD SHIM ROD DRIVE MECHANISM SHIM ROD DRIVE MECHANISM SHIM ROD DRIVE MECHANISM INSTALLED & SUBSTITUTE PLUG REMOVED SUBSTITUTE PLUG INSTALLED FIGURE3 ILLUSTRATION OF SRDM AND SRDM SUBSTITUTE PLUG INSTALLATION American National Standard ANSI/ANS-15.1-2007 (R2013), "The Development of Technical Specifications for Research Reactors," Section 1.3, "Definitions," contains the following condition for work on control rods and control rod drives in its definition of "reactor secured:"

"(c) No work is in progress involving core fuel, core structure, installed control rods, or control rod. drives unless they are physically decoupled from the control rods;"

MURR's proposed use of SRDM substitute plugs provides an additional barrier to ensure the SRDMs are physically decoupled from the shim blades (rods) during work on the SRDMs.

Page 6 of 8

Enclosure 1 - Basis for the Requested Change to Technical Specification 1.26 4.0 Conclusion Allowing the installation of SRDM substitute plugs to satisfy TS 1.26 has no adverse effects to reactor safety. Furthermore, the use of SRDM substitute plugs will enhance reactor safety by having the plug installed when the SRDM is removed from service. Therefore, MURR requests approval for the below revised Specification 1.26.

5.0 Proposed Revision to Technical Specification 1.26 Specification 1.26 currently states:

"The reactor shall be considered secured when:

a. There is insufficient fuel in the reactor core to attain criticality with optimum available conditions of moderation and reflection with all four (4) shim blades (rods) removed, OR
b. Whenever all of the following conditions are met:

(1) All four shim blades (rods) are fully inserted; (2) One of the two following conditions exists:

i. The Master Control Switch is in the "OFF" position with the key locked in the key box or in custody of a licensed operator, OR ii. The dummy load test connectors are installed on the shim rod drive mechanisms and a licensed operator is present in the reactor control room; (3) No work is in progress involving the transfer of fuel in or out of the reactor core; (4) No work is in progress involving the shim blades (rods) or shim rod drive mechanisms with the exception of installing or removing the dummy load test connectors; and (5) The reactor pressure vessel cover is secured in position and no work is in progress on the reactor core assembly supp011 structure."

Specification 1.26 will be revised as follows:

"The reactor shall be considered secured when:

a. There is insufficient fuel in the reactor core to attain criticality with optimum available conditions of moderation and reflection with all four (4) shim blades (rods) removed, Page 7 of 8

Enclosure 1 - Basis for the Requested Change to Technical Specification 1.26 OR

b. Whenever all of the following conditions are met:

(1) All four shim blades (rods) are fully inserted; (2) One of the two following conditions exists:

i. The Master Control Switch is in the "OFF" position with the key locked in the key box or in custody of a licensed operator,

' OR ii. Any combination of dummy load test connectors and shim rod drive mechanism substitute plugs are installed on the shim blade (rod) positions and a licensed operator is present in the reactor control room; (3) No work is in progress involving the transfer of fuel in or out of the reactor core; (4) No work is in progress involving the shim blades (rods) or installed shim rod drive mechanisms with the exception of installing or removing the dummy load test connectors. For each shim rod drive mechanism not installed, its shim rod drive mechanism substitute plug shall be installed; and (5) The reactor pressure vessel cover is secured in position and no work is in progress on the reactor core assembly support structure."

No changes to the bases of TS 1.26 are required or requested.

Page 8 of 8

Enclosure 2 - Basis for the Requested Change to Technical Specification 3.4 1.0 Introduction The University of Missouri Research Reactor (MURR) is requesting a change to Technical Specification (TS) 3 .4, "Reactor Containment Building." This change will revise condition ( 6) of Specification 3 .4.a in order to clarify that containment integrity will still exist even if the negative pressure of the reactor containment building to the surrounding areas is less than 0.25 inches of water when the reactor containment building ventilation system's automatically-closing doors and automatically-closing valves are closed. This clarification will help eliminate any potential confusion about the condition of containment integrity when the doors and valves are closed and negative pressure decreases below 0.25 inches of water.

The following license amendment request (LAR) provides justification that the negative pressure requirement of this Specification is not applicable when the automatically-closing doors and valves are closed. In addition, the negative pressure requirem(:nt of Specification 3.4.a(6) is not needed to ensure the health and safety of the general public during a reactor isolation.

2.0 Background The MURR TSs have contained condition (6) of Specification 3.4.a since the relicensing of MURR on January 4, 2017. Prior to January 4, 2017, there was no negative pressure requirement. The negative pressure requirement for containment integrity was established assuming normal operation of the reactor containment building ventilation system. When MURR and U.S. Nuclear Regulatory Commission (NRC) staff discussed the addition of this requirement during relicensing, no one considered that the verbiage of condition (6) was applicable only during normal ventilation system operation. However, when the quick-closing 16-inch isolation valves (designated 16A and 16B) close as a result of a manual or automatic reactor isolation, the hot exhaust line in the reactor containment building is isolated from the facility exhaust system

[See Figure 9.1 on the next page taken from Chapter 9 of the MURR Safety Analysis Report (SAR)].

Without a constant air flow (approximately 2,500 cfm) out of the reactor containment building after valves 16A and 16B close, the negative pressure inside the containment building will slowly decrease until there is no pressure differential to the surrounding areas. The reactor containment building ventilation system was never designed to maintain a negative pressure during a reactor isolation.

For a detailed description ofMURR's ventilation supply and exhaust systems, refer to Section 9.1 of the SAR. For a detailed description of MURR' s containment system and reactor containment building, and reactor isolation system, refer to SAR Sections 6.2 and 7.8, respectively.

Page 1 of 4

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Enclosure 2 - Basis for the Requested Change to Technical Specification 3.4 3.0 Justification for Changing Technical Specification 3.4.a The objective of TS 3.4, "Reactor Containment Building," currently states:

"The objective of this specification is to assure that containment integrity is maintained when required so that the health and safety of the general public is not endangered as a result of reactor operation."

Furthermore, the basis of Specification 3.4.a currently states:

"Specifications 3.4.a and 3.4.b assure that the reactor containment building can be isolated at all times except when plant conditions are such that the probability of a release of radioactivity is negligible."

The MURR containment system was designed to isolate the reactor containment building atmosphere from the facility exhaust system along with isolating the containment building from all surrounding areas.

Therefore, the original design of the system meets both the objective of TS 3.4 and the basis of Specification 3.4.a. In addition, there are no MURR accident analyses that assume a negative pressure of the reactor containment building with respect to surrounding areas is required to mitigate an accident.

4.0 Conclusion In effect, as Specification 3.4.a(6) is currently written, maintaining containment integrity during a reactor isolation is not possible because maintaining a negative pressure in the reactor containment building with respect to the surrounding areas with valves 16A and 16B closed is not feasible. Therefore, MURR requests approval for the below revised Specification 3.4.a.

5.0 Proposed Revision to Technical Specification 3.4.a Specification 3.4.a currently states:

"a. For reactor containment integrity to exist, the following conditions shall be satisfied:

(1) The truck entry door is closed and sealed; (2) The utility entry seal trench is filled with water to a depth required to maintain a minimum water seal of 4.25 feet; (3) All of the reactor containment building ventilation system's automatically-closing doors and automatically-closing valves are operable or placed in the closed position; (4) The reactor mechanical equipment room ventilation exhaust system, including the .

particulate and halogen filters, is operating; (5) The personnel airlock is operable (one door shut and sealed);

Page 3 of 4

Enclosure 2 - Basis for the Requested Change to Technical Specification 3.4 (6) The reactor containment building is at a negative pressure of at least O.25 inches of water with respect to the surrounding areas; and (7) The most recent reactor containment building leakage rate test was satisfactory."

Specification 3.4.a will be revised as follows:

"a. For reactor containment integrity to exist, the following conditions shall be satisfied:

(1) The truck entry door is closed and sealed; (2) The utility entry seal trench is filled with water to a depth required to maintain a minimum water seal of 4.25 feet; (3) All of the reactor containment building ventilation system's automatically-closing doors and automatically-closing valves are operable or placed in the closed position; (4) The reactor mechanical equipment room ventilation exhaust system, including the particulate and halogen filters, is operating; (5) The personnel airlock is operable (one door shut and sealed);

(6) One of the two following conditions exists:

i. The reactor containment building is at negative pressure of at least 0.25 inches of water with respect to the surrounding areas, OR
11. All of the reactor containment building ventilation system's automatically-closing doors and automatically-closing valves are placed in the closed position; and (7) The most recent reactor containment building leakage rate test was satisfactory."

No changes to the bases of TS 3 .4 are required or requested.

Note: With the proposed revision to TS 3.4.a, TS 3.4.b was moved to page A-25; therefore, it is included as part of this LAR. Nothing has been changed to TS 3.4.b nor the bases of TS 3.4, only a repagination caused by the revision to TS 3.4.a.

Page 4 of 4

Enclosure 3 - Basis for the Requested Change to Technical Specification 4.4 1.0 Introduction The University of Missouri Research Reactor (MURR) is requesting a change to Technical Specification (TS) 4.4, "Reactor Containment Building." This change will revise the required periodicity of the reactor containment building leakage rate surveillance, as specified by TS 4.4.a, from annually to biennially thus affording MURR the potential to reduce adverse effects to reactor instrumentation and equipment, and the reactor containment building due to the high humidity and pressure used in conducting the leakage rate test.

Some of these adverse effects could/have led to unscheduled shutdowns and undue stress to the reactor containment building structure. Unscheduled shutdowns reduce MURR's capacity to meet its mission and vision to improve the quality of life for all through nuclear science and technology. Undue stress to the reactor containment building could lead to increased containment building leakage, which would reduce its effectiveness to protect the public.

The following license amendment request (LAR) provides justification that measuring the reactor containment building leakage rate biennially is sufficient to reasonably assure proper operation of the containment system, thereby assuring that conta~ent integrity is maintained when required so that the health and safety of the general public is assured.

2.0 Background The MURR TSs have contained an annual requirement to measure reactor containment building leakage rate since initial licensing in 1966. From 1966 to 1978, either a pressure decay or reference volume technique was used while applying 2.0 pounds per square inch gauge (psig) pressure to the interior of the containment building.

Amendment No. 10 to Amended Facility Operating License No. R-103, issued on July 13, 1978, authorized MURR to perform the reactor containment building leakage rate test at 1.0 psig using the make-up flow technique to avoid unnecessarily stressing the containment structure or evacuating the overpressure relief protection water seal trench, which would invalidate the test results. Amendment No. 10 Specification 4.2.c stated:

"c. The containment building leakage rate shall not exceed 16.3 ft 3/min. (STP) with an overpressure of one pound per square inch gauge or 10% of contained volume over a 24-hour period from an initial overpressure of two pounds per square inch gauge. The test shall be performed by the makeup flow, pressure decay, or reference volume techniques."

Since 1978, MURR has only used the make-up flow technique at 1.0 psig to measure the reactor containment building leakage rate. During the performance of this technique, steam is introduced into the containment building to raise the relative humidity above 90% in order to reduce the evaporation rate of reactor pool water, which affects the accuracy of the test. Although steam is secured prior to starting the measurements, the relative humidity in the containment building remains greater than 80% for the duration of the test, which totals approximately 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. In addition, the 1.0 psig air pressure is maintained for 6-8 hours during the test. Therefore, all electric motors, relays, switches, meters, radiation detectors, and other electronic instrumentation and control equipment, as well as computers that are left in the containment Page 1 of 4

Enclosure 3 - Basis for the Requested Change to Technical Specification 4.4 building, must withstand the high humidity and pressure condition for several hours. The high humidity level creates a sticky film on surfaces including electrical contact faces . The high pressure condition forces moisture into detector and motor enclosures, and it may take several hours or days for the humidity level inside those enclosures to return to normal levels. There are a few select nuclear instrumentation drawers that are removed prior to conducting the test due to their sensitivity to higher-than-normal humidity; however, it would be impractical and impossible to remove all of the electrical equipment that could be affected during the test from the containment building. MURR suspects that the high humidity and pressure condition may have contributed to equipment failures that occurred after the leakage rate test in years 2006, 2013 , 2018 and 2019. Most of these failures eventually caused unscheduled shutdowns of the reactor.

From 1978 through 2006, the average reactor containment building leakage rate was 8.1 standard cubic feet per minute (SCFM), and there was some variance through the years as MURR used different pressure and humidity instruments, as well as converting from a 1980s VAX computer program to using Microsoft Excel.

Since 2007, the MURR process, software, and instrumentation used for measuring containment building leakage rate have been very consistent. Leakage rate values from 2007 to 2019 range from 9.5 to 12.2 SCFM, well below the Specification 5.5 .c design requirement of 16.3 SCFM - see graph below. The red horizontal line at the top of the graph designates the TS leakage rate limit of 16.3 SCFM at a containment building pressure of 1.0 psig.

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~<1) 9.0

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--W-- Ca lculated Lea k Rate 7.0 Max imu m TS 5.5.c 6.0 Design Requirement

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Calendar Year Containment Building Leak Rate Historical Data (in SCFM)-Calendar Years 2007 to 2019 Page 2 of 4

Enclosure 3 - Basis for the Requested Change to Technical Specification 4.4 In the post-measurement analysis, MURR management always looks for any upward trend in leakage rate compared to the last measurement, and the slope of the increase in rate is evaluated. The data is then extrapolated based on that slope to ensure the subsequent year's leakage rate will fall well below the design limit of Specification 5.5.c. If this LAR is approved, MURR would conduct the same analysis but extrapolate the leakage rate out two (2) years from the curre~t measurement and make the same determination.

In addition to measuring reactor containment building leakage rate, MURR conducts preventive maintenance procedure BCl-Ql, "Containment Doors Gasket Inspection," on a quarterly basis to help identify any potential sealing surface defects prior to the sealing surface degrading to a point of significantly affecting containment building leakage rate.

3.0 Justification for Changing Technical Specification 4.4.a The objective of surveillance TS 4.4, "Reactor Containment Building," currently states:

"The objective of this specification is to reasonably assure proper operation of the containment system."

Additionally, the objective of TS 3.4, "Reactor Containment Building," currently states:

"The objective of this specification is to assure that containment integrity is maintained when required so that the health and safety of the general public is not endangered as a result ofreactor operation."

Furthermore, the basis of Specification 4.4.a currently states:

"Annual measurement of the containment building leakage rate has proven adequate to ensure that the leakage rate of the structure will remain within the design limits outlined in Specification 5.5.c. No repairs or modifications will be performed prior to the test so that the results demonstrate the historic integrity of the containment structure."

Historic reactor containment building leakage rate data indicates that if only biennial measurements were made, there is a reasonable assurance that the leakage rate would remain within the design limits of Specification 5.5.c, specifically the leakage rate would remain below 16.3 SCFM at STP with an overpressure of 1.0 psig.

Moreover, American National Standard ANSI/ANS-15.1-2007 (R2013), "The Development of Technical Specifications for Research Reactors," Section 4.4.1(2), establishes the following surveillance frequency for "Containment:"

"(2) Integrated leak rate test: Annually to biennially;"

Changing Specification 4.4.a to a biennial requirement will reduce the cyclic stress occurring to the reactor containment building which should reduce the potential for cracking of the containment structure. In Page 3 of 4

Enclosure 3 - Basis for the Requested Change to Technical Specification 4.4 addition, changing to a biennial requirement will reduce humidity and pressure cycles on electrical equipment which should reduce electrical component failures and unscheduled shutdowns,in the future.

4.0 Conclusion Measuring the reactor containment building leakage rate biennially is sufficient to reasonably assure proper operation of the containment system, thereby assuring that reactor containment integrity is maintained when required so that the health and safety of the general public is assured. Therefore, MURR requests approval for the below revised Specification 4.4.a.

5.0 Proposed Revision to Technical Specification 4.4.a Specification 4.4.a currently states:

"a. The reactor containment building leakage rate shall be measured annually, plus or minus four (4) months. The test shall be performed by the make-up flow, pressure decay, or reference volume techniques. No repairs or modifications shall be performed just prior to the test."

Specification 4.4.a will be revised as follows:

"a. The reactor containment building leakage rate shall be measured biennially. The test shall be performed by the make-up flow, pressure decay, or reference volume techniques. No repairs or modifications shall be performed just prior to the test."

The basis for Specification 4.4.a currently states:

"a. Annual measurement of the containment building leakage rate has proven adequate to ensure that the leakage rate of the structure will remain within the design limits outlined in Specification 5.5.c.

No repairs or modifications will be performed prior to the test so that the results demonstrate the historic integrity of the containment structure."

The basis for Specification 4.4.a will be revised as follows:

"a. Historic containment building leakage rate data indicates that biennial measurement of the containment building leakage rate has proven adequate to ensure that the leakage rate of the structure will remain within the design limits outlined in Specification 5.5.c. No repairs or modifications will be performed prior to the test so that the results demonstrate the historic integrity of the containment structure."

Page 4 of 4

Enclosure 4 - Basis for the Requested Change to Technical Specification 6.4 1.0 Introduction The University of Missouri Research Reactor (MURR) is requesting a change to Technical Specification (TS) 6.4, "Procedures." This change will a_djust the required periodicity of procedure reviews by the Reactor Manager and the Reactor Health Physics Manager, as specified by Specification 6.4.c, from annually to biennially, thus greatly reducing the administrative burden ofreviewing well-developed, time-tested, and mature MURR TS-required procedures.

The following license amendment request (LAR) provides justification that reviewing procedures for normal operations of the reactor, the Emergency Plan implementing procedures, radiological control procedures, and procedures for the preparation for shipping and the shipping of byproduct material biennially is sufficient to assure the continued effectiveness of MURR TS-required procedures.

2.0 Background Original Facility Operating License No. R-103, issued on October 11, 1966, contained administrative TS 9.1, which stated:

"Written procedures shall be in effect for normal operations of the reactor, emergencies to the reactor or facility which could result in significant radioactive releases, and for radiological control. Copies of these procedures shall be available in the reactor control room and shall be reviewed and approved annually by the Reactor Supervisor."

Since original licensing, Amendment Nos. 3, 4, and 30, as well as the relicensing of MURR on January 4, 2017, have changed the wording and numbering of this Specification; however, the essence of an annual review of TS-required procedures by the responsible manager has remained in the TSs.

MURR staff take these procedure reviews seriously and exert much effort into conducting thorough annual reviews. However, the annual review is rarely the initiator of a substantive change to the procedures.

Changes initiated from annual reviews are normally grammatical, editorial, and/or formatting changes that do not change the technical steps in how a procedure guides the operator or technician to complete a task.

Substantive changes to the procedures are normally initiated by modifications to the facility and equipment, component replacements, or facility corrective actions. MURR administrative procedures AP-RR-015, "Work Control Procedure," and AP-R0-115, "Modification Records," direct the organization to ensure maintenance, operating, and other procedures are updated prior to completing and closing a work package or modification record. Furthermore, the MURR Corrective Action Program (CAP) Review Committee, per administrative procedure AP-RR-001, "Corrective Action Program," ensures that procedures are revised prior to closing a CAP report where the corrective action depends on a procedure revision.

Page 1 of 3

Enclosure 4 - Basis for the Requested Change to Technical Specification 6.4 3.0 Justification for Changing Technical Specification 6.4.c American National Standard ANSI/ANS-15.1-2007 (R2013), "The Development of Technical Specifications for Research Reactors," Section 6.4, "Procedures," does not contain any requirement for periodic reviews of procedures.

ANSI/ANS-15.1-2007 (R2013), Section 6.2.4, "Audit Function," contains the following criteria:

"(4) the reactor facility emergency plan and implementing procedures: at least once every other calendar year (interval between audits not to exceed 30 months)."

Additionally, ANSI/ANS-15.1-2007 (R2013), Section 6.2.3, "Review function," states:

"The following items shall be reviewed:

(1) determinations that proposed changes in equipment, systems, test, experiments, or procedures are allowed without prior authorization by the responsible authority, for example, 10 CFR 50.59 or 10 CFR 830; (2) all new procedures and major revisions thereto having safety significance, proposed changes in reactor facility equipment, or systems having safety significance; ... "

A cursory review of the 23 other NRC-licensed research reactors' TS revealed only one (1) facility where a review of TS-required procedures was required annually. Otherwise, most other NRC-licensed research reactors have wording in Section 6.4 of their TSs which closely resembles Section 6.4 of ANSI/ANS-15.1-2007 (R2013), which has no requirement to periodically review TS-required procedures.

Although this proposed TS change will revise the required periodicity of reviewing the Emergency Plan implementing procedures, the NRC-approved MURR Emergency Plan requires that a more restrictive, annual review of the plan and its implementing procedures be conducted, as does the NRC-approved MURR Physical Security Plan and its implementing procedures. Therefore, MURR will continue to conduct annual reviews of emergency preparedness and security procedures even though the revised Specification 6.4.c indicates only a biennial review requirement for Emergency Plan implementing procedures.

4.0 Conclusion Most MURR TS-required procedures have been in place for many years. Due to the maturity of MURR procedures, an annual review is an unnecessary burden on MURR staff. As stated above, the annual review does not normally reveal any technical weaknesses in the procedures. Reviewing TS-required procedures biennially is sufficient to assure the continued effectiveness of MURR procedures to safely operate the reactor, work with radioactive materials, and prepare radioactive shipments and ship byproduct material.

Therefore, MURR requests approval for the below revised Specification 6.4.c.

Page 2 of 3

Enclosure 4 - Basis for the Requested Change to Technical Specification 6.4 5.0 Proposed Revision to Technical Specification 6.4.c Specification 6.4.c currently states:

"c. The Reactor Manager shall approve and annually review the procedures for normal operations of the reactor and the Emergency Plan implementing procedures. The Reactor Health Physics Manager shall approve and annually review the radiological control procedures and the procedures for the preparation for shipping and the shipping of byproduct material."

Specification 6.4.c will be revised as follows:

"c. The Reactor Manager shall approve and biennially review the procedures for normal operations of the reactor and the Emergency Plan implementing procedures. The Reactor Health Physics Manager shall approve and biennially review the radiological control procedures 'and the procedures for the preparation for shipping and the shipping of byproduct material."

No changes to the bases of TS 6.4 are required or requested.

Page 3 of3

Enclosure 5 - Proposed, revised Technical Specification pages - A-4, A-24, A-25, A-47 and A-70 (with track changes) UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-1 86, License No. R-103 1 DEFINITIONS - Continued 1.23 Reactor in Operation - The reactor shall be considered in operation unless it is either shutdown or secured.

1.24 Reactor Safety System - The reactor safety system is that combination of sensing devices, electronic circuits and equipment, signal conditioning equipment, and electro-mechanical devices that serves to either effect a reactor scram, or activates the engineered safety features .

1.25 Reactor Scram - A reactor scram is the insertion of all four (4) shim blades (rods) by gravitational force as a result of removing the holding current from the shim rod drive mechanism electromagnets.

1.26 Reactor Secured - The reactor shall be considered secured when:

a. There is insufficient fuel in the reactor core to attain criticality with optimum available conditions of moderation and reflection with all four (4) shim blades (rods) removed, OR
b. Whenever all of the following conditions are met:

(1) All four shim blades (rods) are fully inserted; (2) One of the two following conditions exists:

1. The Master Control Switch is in the "OFF" position with the key locked in the key box or in custody of a licensed operator, OR
11. Any combination of +he dummy load test connectors and shim rod dri ve mechanism substitute plugs are installed on the shim blade (rod) positions drive mechanisms and a licensed operator is present in the reactor control room; (3) No work is in progress involving the transfer of fuel in or out of the reactor core; (4) No work is in progress involving the shim blades (rods) or shim rod drive mechanisms with the exception of installing or removing the dummy load test connectors. For each shim rod drive mechanism not installed, its shim rod drive mechanism substitute plug shall be installed; and (5) The reactor pressure vessel cover is secured in position and no work is in progress on the reactor core assembly support structure.

A-4

- Proposed, revised Technical Specification pages -A-4, A-24, A-25, A-47 and A-70 (with track changes) UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3.4 Reactor Containment Building Applicability:

This specification applies to the reactor contaimnent building.

Objective:

The objective of this specification is to assure that contaimnent integrity is maintained when required so that the health and safety of the general public is not endangered as a result of reactor operation.

Specification:

a. For reactor contaimnent integrity to exist, the following conditions shall be satisfied:

(1) The truck entry door is closed and sealed; (2) The utility entry seal trench is filled with water to a depth required to maintain a minimum water seal of 4.25 feet; (3) All of the reactor contaimnent building ventilation system's automatically-closing doors and automatically-closing valves are operable or placed in the closed position; (4) The reactor mechanical equipment room ventilation exhaust system, including the particulate and halogen filters, is operating; (5) The personnel airlock is operable (one door shut and sealed);

(6) One of the two following conditions exists:

1. The reactor contaimnent building is at a negative pressure of at least 0.25 inches of water with respect to the surrounding areas,~

OR

11. All of the reactor contaimnent building ventilation system ' s automatically-closing doors and automatically-closing valves are placed in the closed position; and (7) The most recent reactor contaimnent building leakage rate test was satisfactory.

A-24

- Proposed, revised Technical Specification pages - A-4, A-24, A-25, A-47 and A-70 (with track changes) UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No . 50-186, License No . R-103 3.4 Reactor Containment Building- Continued

b. Reactor contaimnent integrity shall be maintained at all times except when :

(1) The reactor is secured, AND (2) No movement of irradiated fuel with a decay time of less than sixty (60) days or experiments with the potential for a significant release of airborne radioactivity outside of containers, systems, or storage areas, AND (3) No movement of experiments that could cause a change of total worth greater than 0.0074 Lik/k.

c. When reactor contaimnent integrity is required, the reactor contaimnent building shall be automatically isolated if the activity in the ventilation exhaust plenum or at the reactor bridge indicates an increase of 10 times above previously established levels at the same operating condition. Exception: The contaimnent isolation set point may temporarily be increased to avoid an inadvertent scram and isolation during controlled evolutions such as experiment transfers or minor maintenance in the reactor pool area. The pool area shall be continuously monitored, and, if necessary, a manual containment isolation actuated, until the automatic set point is reset to its normal value.

Bases:

a. - b. Specifications 3.4.a and 3.4.b assure that the reactor contaimnent building can be isolated at all times except when plant conditions are such that the probability of a release of radioactivity is negligible.
c. Radiation monitors located at the reactor bridge and in the reactor contaimnent building ventilation exhaust plenum supply input signals to meters located in the reactor control room. A contaimnent isolation will occur when radiation levels in these areas exceed a predetermined value. During operations such as the removal of experiments or equipment from the pool, the radiation level at the level of the reactor bridge or in the exhaust plenum can increase significantly for short periods. To prevent inadvertent contaimnent isolations, it may be necessary to raise the set point on the reactor bridge or exhaust plenum monitor. During periods in which the set point is raised to more than one decade above the normal reading, the radiation level in the area of the monitor will be continuously monitored. Thus, should the radiation level increase from unknown causes or from material which could be released to the unrestricted enviromnent, the reactor contaimnent building can be quickly isolated by manually actuating the isolation system.

A-25

- Proposed, revised Technical Specification pages -A-4, A-24, A-25, A-47 and A-70 (with track changes) UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 4.4 Reactor Containment Building Applicability:

This specification applies to the surveillance requirements on the contairunent system.

Objective:

The objective of this specification is to reasonably assure proper operation of the containment system.

Specification:

a. The reactor contairunent building leakage rate shall be measured biennially annually, plus or minus four (4) months. The test shall be performed by the make-up flow, pressure decay, or reference volume techniques. No repairs or modifications shall be performed just prior to the test.
b. The reactor contairunent building leakage rate shall be measured following any modification or repair that could affect the leak-tightness of the building.
c. The contairunent actuation (reactor isolation) system, including each of its radiation monitors, shall be tested for operability at monthly intervals.
d. When required by Specification 3.4.b, contairunent integrity shall be verified to exist within a shift.

Bases:

a. Historic contairunent building leakage rate data indicates that biennial .1\..n nual measurement of the contairunent building leakage rate has proven adequate to ensure that the leakage rate of the structure will remain within the design limits outlined in Specification 5.5.c. No repairs or modifications will be performed prior to the test so that the results demonstrate the historic integrity of the containment structure.
b. Measurement of the contairunent building leakage rate following any modification or repair that could affect the leak-tightness of the building ensures that the leakage rate of the structure will remain within the design limits outlined in Specification 5.5.c.
c. The reliability of the contairunent actuation (reactor isolation) system has proven that monthly verification of its proper operation is sufficient to assure operability.
d. Specification 4.4.d assures that contairunent integrity is verified to exist to limit the leakage of contained potentially radioactive air in the event of any reactor accident to ensure exposures are maintained below the limits of 10 CFR 20.

A-47

- Proposed, revised Technical Specification pages - A-4, A-24, A-25, A-47 and A-70 (with track changes) UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 6.4 Procedures - Continued

c. The Reactor Manager shall approve and biennially annually review the procedures for normal operations of the reactor and the Emergency Plan implementing procedures. The Reactor Health Physics Manager shall approve and biennially annually review the radiological control procedures and the procedures for the preparation for shipping and the shipping of byproduct material.
d. Deviations from procedures required by this Specification may be enacted by a Senior Reactor Operator or member of Reactor Health Physics, as applicable.

Such deviations shall be documented, reviewed pursuant to 10 CFR 50.59, and reported within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the next working day to the Reactor Manager or Reactor Health Physics Manager or designated alternate.

6.5 Experiment Review and Approval

a. Approved experiments shall be carried out in accordance with established and approved procedures. Procedures related to experiment review and approval shall include the following:

(1) All new experiments or class of experiments shall be reviewed by the RAC and approved in writing by the Reactor Manager.

(2) Substantive changes to previously approved experiments shall be made only after review by the RAC and approved in writing by the Reactor Manager.

6.6 Reportable Events and Required Actions

a. Safety Limit Violation - In the event of a safety limit violation, the following actions shall be taken:

(1) The reactor shall be shut down and reactor operation shall not be resumed until authorized by the NRC pursuant to 10 CFR 50.36( c)(1 );

(2) The safety limit violation shall be promptly reported to the Reactor Manager and Reactor Facility Director, or designated alternates; (3) The safety limit violation shall be promptly reported to the NRC. Prompt reporting of the violation shall be made by MU, by telephone and subsequently confirmed in writing or email, to the NRC Operations Center no later than the following working day; (4) A detailed follow-up report shall be prepared. The report shall include the following:

A-70

Enclosure 6 - Proposed, revised Technical Specification pages -A-4, A-24, A-25, A-47 and A-70 (with accepted changes, revision bars) UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 1 DEFINITIONS - Continued 1.23 Reactor in Operation - The reactor shall be considered in operation unless it is either shutdown or secured.

1.24 Reactor Safety System - The reactor safety system is that combination of sensing devices, electronic circuits and equipment, signal conditioning equipment, and electro-mechanical devices that serves to either effect a reactor scram, or activates the engineered safety features.

1.25 Reactor Scram - A reactor scram is the insertion of all four (4) shim blades (rods) by gravitational force as a result of removing the holding current from the shim rod drive mechanism electromagnets.

1.26 Reactor Secured - The reactor shall be considered secured when:

a. There is insufficient fuel in the reactor core to attain criticality with opti1pum available conditions of moderation and reflection with all four (4) shim blades (rods) removed, OR
b. Whenever all of the following conditions are met:

(1) All four shim blades (rods) are fully inserted; (2) One of the two following conditions exists:

1. The Master Control Switch is in the "OFF" position with the key locked in the key box or in custody of a licensed operator, OR
11. Any combination of dummy load test connectors and shim rod drive mechanism substitute plugs are installed on the shim blade (rod) positions and a licensed operator is present in the reactor control room; (3) No work is in progress involving the transfer of fuel in or out of the reactor core; (4) No work is in progress involving the shim blades (rods) or shim rod drive mechanisms with the exception of installing or removing the dummy load test connectors. For each shim rod drive mechanism not installed, its shim rod drive mechanism substitute plug shall be installed; and (5) The reactor pressure vessel cover is secured in position and no work is in progress on the reactor core assembly support structure.

A-4

- Proposed, revised Technical Specification pages - A-4, A-24, A-25, A-47 and A-70 (with accepted changes, revision bars) UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 3.4 Reactor Containment Building Applicability:

This specification applies to the reactor containment building.

Objective:

The objective of this specification is to assure that containment integrity is maintained when required so that the health and safety of the general public is not endangered as a result of reactor operation.

Specification:

a. For reactor containment integrity to exist, the following conditions shall be satisfied:

(1) The truck entry door is closed and sealed; (2) The utility entry seal trench is filled with water to a depth required to maintain a minimum water seal of 4.25 feet; (3) All of the reactor containment building ventilation system's automatically-closing doors and automatically-closing valves are operable or placed in the closed position; (4) The reactor mechanical equipment room ventilation exhaust system, including the particulate and halogen filters, is operating; (5) The personnel airlock is operable (one door shut and sealed);

(6) One of the two following conditions exists:

1. The reactor containment building is at a negative pressure of at least 0.25 inches of water with respect to the surrounding areas, OR
n. All of the reactor containment building ventilation system's automatically-closing doors and automatically"."closing valves are placed in the closed position; and (7) The most recent reactor containment building leakage rate test was satisfactory.

A-24

- Proposed, revised Technical Specification pages -A-4, A-24, A-25, A-47 and A-70 (with accepted changes, revision bars) UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS ,

Docket No. 50-186, License No. R-103 3 .4 Reactor Containment Building - Continued

b. Reactor containment integrity shall be maintained at all times except when:

(1) The reactor is secured, AND (2) No movement of irradiated fuel with a decay time of less than sixty (60) days or experiments with the potential for a significant release of airborne radioactivity outside of containers, systems, or storage areas, AND (3) No movement of experiments that could cause a change of total worth greater than 0.0074 Lik/k.

c. When reactor containment integrity is required, the reactor containment building shall be automatically isolated if the activity in the ventilation exhaust plenum or at the reactor bridge indicates . an increase of 10 tjmes above previously established levels at the same operating condition. Exception: The containment isolation set point may temporarily be increased to avoid an inadvertent scram and isolation during controlled evolutions such as experiment transfers or minor maintenance in the reactor pool area. The pool area shall be continuously monitored, and, if necessary, a manual containment isolation actuated, until the automatic set point is reset to its normal value.

Bases:

a. - b. Specifications 3 .4.a and 3.4.b assure that the reactor containment building can be isolated at all times except when plant conditions are such that the probability of a release of radioactivity is negligible.
c. Radiation monitors located at the reactor bridge and in the reactor containment building ventilation exhaust plenum supply input signals to meters located in the reactor control room. A containment isolation will occur when radiation levels in these areas exceed a predetermined value. During operations such as the removal of experiments or equipment from the pool, the radiation level at the level of the reactor bridge or in the exhaust plenum can increase significantly for short periods .. To prevent inadvertent containment isolations, it may be necessary to raise the set point on the reactor bridge or exhaust plenum monitor. During periods in which the set point is raised to more than one decade above the normal

,reading, the radiation level in the area of the monitor will be continuously monitored. Thus, should the radiation level increase from unknown causes or from material which could be released to the unrestricted environment, the reactor containment building can be quickly isolated by manually actuating the isolation system.

A-25

-Yroposed, revised Technical Specification pages - A-4, A-24, A-25, A-47 and A-70 (with accepted changes, revision bars) UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket No. 50-186, License No. R-103 4.4 Reactor Containment Building Applicability:

This specification applies to the surveillance requirements on the containment system.

Objective:

The objective of this specification is to reasonably assm;e proper operation of the

- containment system.

Specification:

a. The reactor containment building leakage rate shall be measured biennially. The test shall be perfonned by the make-up flow, pressure decay, or reference volume techniques. No repairs or modifications shall be performed just prior to the test.
b. The reactor containment buildihg leakage rate shall be measured following any modification or repair that could affect the leak-tightness of the building.
c. The containment actuation (reactor isolation) system, including each of its radiation monitors, shall be tested for operability at monthly intervals.
d. When required by Specification 3.4.b, containment integrity shall be verified to exist within a shift.

Bases:

a. Historic containment building leakage rate data indicates that biennial measurement of the containment building leakage rate has proven adequate to ensure that the leakage rate of the structure will remain within the design limits outlined in Specification 5.5.c. No repairs or modifications will be performed prior to the test so that the results demonstrate the historic integrity of the containment structure.
b. Measurement . of the containment building leakage rate following any modification or repair that could affect the leak-tightness of the building ensures that the leakage rate of the structure will remain within the design limits outlined in Specification 5.5.c.
c. The reliability of the containment actuation (reactor isolation) system has proven that monthly verification of its proper operation is sufficient to assure operability.
d. Specification 4.4.d assures that containment ~ntegrity is verified to exist to limit the leakage of contained potentially radioact1ve air in the event of any reactor accident to ensure exposures are maintained below the limits of 10 CPR 20.

A-47

- Proposed, revised Technical Specification pages -A-4, A-24, A-25, A-47 and A-70 (with accepted changes, revision bars) UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS DocketNo. 50-186, License No. R-103 6.4 Procedures - Continued

c. The Reactor Manager shall approve and biennially review the procedures for nonnal operations of the reactor and the Emergency Plan implementing procedures. The Reactor Health Physics Manager shall approve and biennially review the radiological control procedures and the procedures for the preparation for shipping and the shipping of byproduct material.
d. Deviations :from procedures required by this Specification may be enacted by a Senior Reactor Operator or member of Reactor Health Physics, as applicable.

Such deviations shall be documented, reviewed pursuant to 10 CPR 50.59, and reported within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the next working day to the Reactor Manager or Reactor Health Physics Manager or designated alternate.

6.5 Experiment Review and Approval

a. Approved experiments shall be carried out in accordance with established and approved procedures. Procedures related to experiment review and approval shall include the following:

(1) All new experiments or class of experiments shall be reviewed by the RAC and approved in writing by the Reactor Manager.

(2) Substantive changes to previously approved experiments shall be made only after review by the RAC and approved in writing by the Reactor Manager.

6.6 Reportable Events and Required Actions

a. Safety Limit Violation - In the event of a safety limit violation, the following actions shall be taken:

(1)

  • The reactor shall pe shut down and reactor operation shall not be resumed until authorized by the NRC pursuant to 10 CPR 50.36(c)(l);

(2) The safety limit violation shall be promptly reported *to the Reactor Manager and Reactor Facility Director, or designated alternates; (3) The safety limit violation shall be promptly reported to the NRC. Prompt reporting of the violation shall be made by MU, by telephone and subsequently confirmed in writing or email, to the NRC Operations Center no later than the following working day; (4) A detailed follow-up report shall be prepared. The report shall include the following:

A-70

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