Text

Letter from the University of Missouri-Columbia Research Reactor Regarding Amended Facility License R-103
	ML14030A132
Person / Time
Site:	University of Missouri-Columbia
Issue date:	01/27/2014
From:	Rhonda Butler; Univ of Missouri - Columbia
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	Download: ML14030A132 (87)
	v • d • e

UNIVERSITY of MISSOURI RESEARCH REACTOR CENTER January 27, 2014 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 Enclosed you will find the University of Missouri-Columbia Research Reactor's (MURR) revised Technical Specifications in support of our renewal request for Amended Facility Operating License R-103, which was submitted to the U.S. Nuclear Regulatory Commission (NRC) on August 31, 2006, as supplemented.

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

Sincerely, Ralph A. Butler, P.E.

Director RAB/jlb Enclosures

-.. u,,, MARGEE P.STOUT NOTAR)-.V My Comrrrission Eores

":* March24,2016 W

Mwontgomery Coury Commission #12511436 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

UNIVERSITY of MISSOURI RESEARCH REACTOR CENTER January 27, 2014 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 responses to the "University of Missouri at Columbia - Request for Additional Information Re: License Renewal, Safety Analysis Report, Complex Questions (TAC No. MD3034)," dated May 6, 2010, and the "University of Missouri at Columbia - Request for Additional Information Re: License Renewal, Safety Analysis Report, 45-Day Response Questions (TAC No. MD3034)," dated June 1, 2010 On August 31, 2006, the University of Missouri-Columbia Research Reactor (MURR) submitted a request to the U.S. Nuclear Regulatory Commission (NRC) to renew Amended Facility Operating License R-103.

On May 6, 2010, the NRC requested additional information and clarification regarding the renewal request in the form of nineteen (19) Complex Questions. By letter dated September 3, 2010, MURR responded to seven (7) of those Complex Questions.

On June 1, 2010, the NRC requested additional information and clarification regarding the renewal request in the form of one hundred and sixty-seven (167) 45-Day Response Questions. By letter dated July 16, 2010, MURR responded to forty-seven (47) of those 45-Day Response Questions.

On July 14, 2010, via electronic mail (email), MURR requested additional time to respond to the remaining one hundred and twenty (120) 45-Day Response Questions. By letter dated August 4, 2010, the NRC granted the request. By letter dated August 31, 2010, MURR responded to fifty-three (53) of the 45-Day Response Questions.

On September 1, 2010, via email, MURR requested additional time to respond to the remaining twelve (12) Complex Questions. By letter dated September 27, 2010, the NRC granted the request.

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

On September 29, 2010, via email, MURR requested additional time to respond to the remaining sixty-seven (67) 45-Day Response Questions. On September 30, 2010, MURR responded to sixteen (16) of the remaining 45-Day Questions. By letter dated October 13, 2010, the NRC granted the extension request.

By letter dated October 29, 2010, MURR responded to sixteen (16) of the remaining 45-Day Response Questions and two (2) of the remaining Complex Questions.

By letter dated November 30, 2010, MURR responded to twelve (12) of the remaining 45-Day Response Questions.

On December 1, 2010, via email, MURR requested additional time to respond to the remaining 45-Day Response and Complex Questions. By letter dated December 13, 2010, the NRC granted the extension request.

On January 14, 2011, via email, MURR requested additional time to respond to the remaining 45-Day Response and Complex Questions. By letter dated February 1, 2011, the NRC granted the extension request.

By letter dated March 11, 2011, MURR responded to twenty-one (21) of the remaining 45-Day Response Questions.

On May 27, 2011, via email, MIURR requested additional time to respond to the remaining the remaining 45-Day Response and Complex Questions. By letter dated July 5, 2011, the NRC granted the request.

By letter dated September 8, 2011, MURR responded to six (6) of the remaining 45-Day Response and Complex Questions.

On September 30, 2011, via email, MURR requested additional time to respond to the remaining the remaining 45-Day Response and Complex Questions. By letter dated November 10, 2011, the NRC granted the request.

By letter dated January 6, 2012, MURR responded to four (4) of the remaining 45-Day Response and Complex Questions. Also submitted was an updated version of the MURR Technical Specifications.

On January 23, 2012, via email, MURR requested additional time to respond to the remaining the remaining 45-Day Response and Complex Questions. By letter dated January 26, 2012, the NRC granted the request.

On April 12, 2012, via email, MURR requested additional time to respond to the remaining 45-Day Response and Complex Questions.

By letter dated June 28, 2012, MURR responded to the remaining six (6) 45-Day Response and Complex Questions. With that set of responses, all 45-Day Response and Complex Questions had been addressed.

During the week of May 6, 2013, Geoffrey Wertz, NRC Relicensing Project Manager for M-URR, discussed with MURR staff additional changes that the NRC felt should be incorporated into the 2 of 3

Technical Specifications. Attached are a summary detailing the changes and the revised version of the MURR Technical Specifications.

If there are questions regarding this submittal, please contact me at (573) 882-5319 or FruitsJ@missouri.edu. I declare under penalty of perjury that the foregoing is true and corTect.

ENDORSEMENT:

Sincerely, Reviewed and Approved, John L. Fruits Ralph A. Butler, P.E.

Reactor Manager Director MARGEE P.STOUT MY Commision Em*es MArch 24, 2018 MonWgome Cut y Commi~ssio #12511436 Enclosed: : Summary of Changes Made to the MURR Technical Specifications (From version submitted on January 6, 2012, to the enclosed version). : Appendix A, Technical Specifications for the University of Missouri Research Reactor, Facility Operating License R-103, Docket 50-186, as revised.

xc: Reactor Advisory Committee Reactor Safety Subcommittee Dr. Robert Hall, Interim Vice Chancellor for Research Mr. Geoffrey Wertz, U.S. NRC Mr. Alexander Adams, U.S. NRC Mr. Johnny Eads, U.S. NRC 3 of 3

Summary of Changes Made to the MURR Technical Specifications (From version submitted on January 6, 2012, to the enclosed version)

Changes to January 12, 2012 Submittal (old version):

==

Introduction:==

a. The following words were deleted from the first sentence, "an agreement between the licensee and the U.S. Nuclear Regulatory Commission (NRC)."

b. Page ii and ii, Table of Contents revised as described in the following sections.

c. Added page number to page iv.

Section 1, Definitions:

a. The following definitions have been added:

1. 1.4, Channel Calibration

2. 1.5, Channel Check

3. .15, Operating

4. 1.17, Reactivity Worth of an Experiment

5. 1.26, Reference Core Condition

6. 1.29, Research Reactor

7. 1.30, Research Reactor Facility

8. 1.31, Rod Run-In System

9. 1.36, Should, Shall, and May

10. 1.38, Surveillance Intervals II. 1.40, Unscheduled Shutdown

b. The following definitions were deleted:

1. 1.2, Calibration or Testing Interval was deleted because it was replaced by 1.38, Surveillance.

Intervals.

2. 1.4, Cold, Clean, Critical was deleted because it was replaced by 1.26, Reference Core Condition.

c. The following definitions were revised:

1. Deleted the word "Instrument" from 1.9, Instrument Channel (new 1.3) and 1.10, Instrument Channel Test (new 1.6) to follow the guidance of ANSI/ANS-1 5.1-2007.

2. The words "in a normal manner" were deleted from 1.14, Operable to follow the guidance of ANSI/ANS-I15.1-2007.

d. Definition 1.17, Reactor Containment Integrity was incorporated into LCO 3.4, Reactor Containment Building as Specification 3.4.a because it more closely follows the guidance of ANSI/ANS- 15.1-2007 (Section 3.4.2).

e. Other Definitions have been renumbered as required.

Section 2.0, Safety Limits and Limiting Safety System Settings:

a. No changes Section 3.0, Limiting Conditions of Operation:

a. Section 3.1, Reactivity renamed to Reactor Core Parameters (ANSI/ANS-15.1-2007).

b. Section 3.2, Control Blades renamed to Reactor Control and Reactor Safety Systems (AN SI/ANS- 15.1-2007).

c. Section 3.3, Reactor Safety System renamed to Reactor Coolant Systems (ANSI/ANS-15.1-2007).

d. Section 3.4, Reactor Instrumentation renamed to Reactor Containment Building (ANSI/ANS-15.1-2007).

e. Section 3.5, Reactor Containment Building renamed to Reactor Instrumentation.

f. Section 3.6, Experiments renamed to Emergency Electrical Power System (ANS1/ANS-15.1-2007).

g. Section 3.7, Facility Airborne Effluents renamed to Radiation Monitoring Systems and Airborne Effluents (ANSI/ANS- 15.1-2007).

h. Section 3.8, Reactor Fuel renamed to Experiments (ANSI/ANS-15.1-2007).

i. Section 3.9, Reactor Coolant Systems renamed to Auxiliary Systems (ANSI/ANS-15.1-2007).

j. Section 3.10, Auxiliary Systems deleted because it became Section 3.9, Auxiliary Systems.

k. Section 3.1, Reactivity Limitations:

1. 3.1.a is now 5.3.a

2. 3.l.b is now 5.3.b

3. 3.1.c was split- it is now 5.3.c and 3.2.d

4. 3.1.d is now 3.2.e

5. 3.1.e is now 3.L.b

6. 3.1 .f is now 3.1 .a

7. -3.1.g is now 3.8.p

8. 3.1.h is now 3.8.q

9. 3.1.i is now 3.8.r

10. 3.1 .j is now 3.8.s

11. 3.L.k is now 3.8.t I. Section 3.2, Control Blades:

1. 3.2.a remains 3.2.a

2. 3.2.b remains 3.2.b

3. 3.2.c remains 3.2.c

m. Section 3.3, Reactor Safety System:

1. 3.3.a is now 3.2.g

n. Section 3.4, Reactor Instrumentation:

1. First two Specifications of 3.4.a are now 3.5.a

2. Other three Specifications of 3.4.a are now 3.7.a

3. 3.4.b is now 3.5.b

4. 3.4.c is now 3.2.f

5. 3.4.d is now 3.5.c

6. 3.4.e is now 3.5.d

o. Section 3.5, Reactor Containment Building:

1. 3.5.a is now 3.4.b

2. 3.5.b is now 3.4.c

p. Section 3.6, Experiments:

1. 3.6.a through 3.6.o are now 3.8.a through 3.8.o

q. Section 3.7, Facility Airborne Effluents:

1. 3.7.a is now 3.7.b

r. Section 3.8, Reactor Fuel:

1. 3.8.a is now 3.l.d 2 of 9

2. 3.8.b is now 3.l.e

3. 3.8.c is now 3.1.c

4. 3.8.d is now 5.4.a

5. 3.8.e is now 5.4.b

s. Section 3.9, Reactor Coolant Systems:

1. 3.9.a is now 3.3.a

2. 3.9.b is now 3.3.b

3. 3.9.c is now 3.3.c

4. 3.9.d is now 3.3.i

t. Section 3.10, Auxiliary Systems:

1. 3.10.aisnow3.6.a

2. 3.10.bisnow3.9.a

3. 3.10.c is now 3.9.b Section 4.0, Surveillance Requirements:

a. Section 4.1, Containment System renamed to Reactor Core Parameters (ANSI/ANS- 15.1-2007).

b. Section 4.2, Reactor Coolant Systems renamed to Reactor Control and Reactor Safety Systems (ANSI/ANS-15.1-2007).

c. Section 4.3, Control Blades renamed to Reactor Coolant Systems (ANSI/ANS-15.1-2007).

d. Section 4.4, Reactor Instrumentation renamed to Reactor Containment Building (ANSI/ANS-15.1-2007).

e. Section 4.5, Reactor Fuel renamed to Reactor Instrumentation.

f. Section 4.6, Auxiliary Systems renamed to Emergency Electrical Power System (ANSI/ANS-15.1-2007).

g. Section 4.7, Radiation Monitoring Systems and Airborne Effluents added (ANSI/ANS-15.1-2007).

h. Section 4.8, Experiments added (ANSI/ANS-15.1-2007).

i. Section 4.9, Auxiliary Systems added (ANSI/ANS-15.1-2007).

j. Section 4.1, Containment System:

1. 4. .a is now 4.4.a

2. 4.1.b is now 4.4.b

k. Section 4.2, Reactor Coolant Systems:

1. 4.2.a is now 4.3.g

2. 4.2.b is now 4.3.a but it was also reworded to include the in-pool convective cooling loop

3. 4.2.c is now 4.3.b I. Section 4.3, Control Blades:

I. 4.3.a is now 4.2.a

2. 4.3.b is now 4.2.b

m. Section 4.4, Reactor Instrumentation:

1. 4.4.a has been broken down into many individual surveillance requirements

2. 4.4.b is now 4.7.a

3. 4.4.c is now 4.5.c

n. Section 4.5, Reactor Fuel:

1. 4.5.aisnow4.1.c

o. Section 4.6, Auxiliary Systems:

1. 4.6.a is now 4.9.b 3 of 9

2. 4.6.b is now 4.6.a

3. 4.6.c is now 4.6.b Section 5.0, Design Features:

a. Section 5.2, Reactor Containment Building was renamed to Reactor Coolant Systems (ANSI/ANS-15.1-2007).

b. Section 5.3, Reactor Coolant Systems was renamed to Reactor Core and Fuel (ANSL/ANS-15.1-2007).

c. Section 5.4, Reactor Core and Fuel was renamed to Fuel Storage (ANSI/ANS-15.1-2007).

d. Section 5.5, Emergency Electrical Power System was renamed to Reactor Containment Building (ANSI/ANS- 15.1-2007).

e. Section 5.6, Emergency Electrical Power System was added (ANSI/ANS-15.1-2007).

f. Section 5.1, Site

Description:

1. No changes

g. Section 5.2, Reactor Containment Building:

I. 5.2.a is now 5.5.a

2. 5.2.b is now 5.5.b

3. 5.2.c is now 5.5.c

4. 5.2.d is now 5.5.d

h. Section 5.3, Reactor Coolant Systems:

I. 5.3.a through 5.3.k (including Exceptions a and b) are now 5.2.a through 5.2.k

i. Section 5.4, Reactor Core and Fuel:

I. 5.4.a is now 5.3.d

2. 5.4.b is now 5.3.e

3. 5.4.c is now 5.3.f

4. 5.4.d is now 5.3.g

5. 5.4.e is now 5.3.hi

6. 5.4.fis now 5.3.i

7. 5.4.g is now 5.3.j

j. Section 5.5, Emergency Electrical Power System:

I. 5.5.a is now 5.6.a Section 6.0, Administrative Controls

a. Section 6.3, Procedures was renamed to Radiation Safety (ANSI/ANS-15.1-2007).

b. Section 6.4, Records was renamed to Procedures (ANSI/ANS- 15.1-2007).

c. Section 6.5, Reportable Events and Required Actions was renamed to Experiment Review and Approval (ANSI/ANS-I5.1-2007).

d. Section 6.6, Reportable Events and Required Actions was added (ANSI/ANS-15.1-2007).

e. Section 6.7, Records was added (ANSI/ANS-I 5.1-2007).

f. Section 6.1, Organization:

I. 6. l.a remains 6. l.a

2. 6.l.b remains 6.1.b

3. 6.1.c remains 6.1.c

4. 6.1.d changed to 6.1.e

g. Section 6.2, Review and Audit:

1. 6.2.a remains 6.2.a 4 of 9

2. 6.2,b remains 6.2.b with the following change: Wording was added to the first sentence of the first paragraph: "knowledgeable members of the public" and in the last sentence of the second paragraph "once during each calendar quarter" was replaced by "quarterly."

3. 6.2,c remains 6.2.c

4. 6.2,d remains 6.2.d

5. 6.2,e remains 6.2.e

h. Section 6.3, Procedures:

1. 6.3,a is now 6.4.a

2. 6.3,b is now 6.4.b

3. 6.3,c is now 6.4.c

4. 6.3,d is now 6.4.d I. Section 6.4, Records:

1. 6.4,a is now 6.7.a

2. 6.4,b is now 6.7.b

j. Section 6.5, Reportable Events and Required Actions:

I. 6.5,a is now 6.6.a

2. 6.5,b is now 6.6.b

3. 6.5,c is now 6.6.c

4. 6.5,d is now 6.6.d

5. 6.5,e is now 6.6.e January 27, 2014 Submittal (new version):

==

Introduction:==

a, As described above Section 1, Definitions:

a. As described above Section 2.0, Safety Limits and Limiting Safety System Settings:

a. No changes Section 3.0, Limiting Conditions of Operation:

Added "General" wording at the beginning of Section.

a. Section 3.1 Reactor Core Parameters:

I. 3.],a was old 3.1.f

2. 3.l,b was old 3.I.e

3. 3.1,c was old 3.8.c

4. 3.1,d was old 3.8.a

5. 3.l.e was old 3.8.b

b. Section 3.2, Reactor Control and Reactor Safety Systems:

1. 3.2,a was old 3.2.a

2. 3.2,b was old 3.2.b

3. 3.2.c was old 3.2.c

4. 3.2.d was old 3.1.c

5. 3.2.e was old 3.l.d 5 of 9

6. 3.2.f was old 3.4.c

7. 3.2.g was old 3.3.a

8. 3.2.h - New Specification

c. Section 3.3, Reactor Coolant

1. 3.3.a was old Systems:

3.9.a 2, 3.3.b was old 3.9.b

3. 3.3.c was old 3.9,c 4, 3.3.d - New Specification

5. 3.3.e - New Specification 6, 3.3.f-New Specification

7. 3.3.g- New Specification 8, 3.3.h - New Specification

9. 3.3.i was old 3.9.d

d. Section 3.4, Reactor Containment

1. 3.4.a was previous Building:

2. 3.4.b was old Definition 1.17 3.5.a

3. 3.4.c was old 3.5.b

e. Section 3.5, Reactor Instrumentation:

1. 3.5.a was old 3.4.a

2. 3.5.b was old 3.4.b

3. 3.5.c was old 3.4.d

4. 3.5.d was old 3.4.e

f. Section 3.6, Emergency Electrical

1. 3.6.awasold3.10.a Power System:

g. Section 3.7, Radiation Monitoring

1. 3.7.a was old Systems Airborne 3.4.a Effluents:

2. 3.7.b was old

h. Section 3.8, 3.7.a Experiments:

I. 3.8.a was old 3.6.a

2. 3.8.b was old 3,6.b

3. 3.8.c was old 3.6.c

4. 3.8.d was old 3.6.d

5. 3.8.e was old 3.6.e

6. 3.8.fwasold3.6.f

7. 3.8.g was old 3.6.g

8. 3.8.h was old 3.6.h

9. 3.8.i was old 3.6.i

10. 3.8.j was old 3.6.j

11. 3.8.k was old 3.6.k

12. 3.8.1 was old 3.6.1 13, 3.8.m was old 3.6.m

14. 3.8.n was old 3.6.n

15. 3.8.o was old 3,6.o

16. 3.8.p was old

3. l.g

17. 3.8.q was old 3.l.h

18. 3.8.r was old

3. L.i 6 of 9

19. 3.8.s was old 3.1 .j (Also, found error in unsecured experiment reactivity worth that was not originally caught. Should have been 0.0025 not 0.025, has been changed)

20. 3.8.t was old 3.1 .k Section 3.9, Auxiliary Systems:

1. 3.9.awasold3.10.b

2. 3.9.b was old 3.10.c Section 4.0, Surveillance Requirements:

Added "General" wording at the beginning of Section.

a. Section 4.1 Reactor Core Parameters:

1. 4.1.a- New Specification

2. 4.1 .b- New Specification

3. 4.1 .c was old 4.5.a

b. Section 4.2, Reactor Control and Reactor Safety Systems:

1. 4.2.a was old 4.3.a

2. 4.2.b was old 4.3.b

3. 4.2.c - New Specification

4. 4.2.d - New Specification

5. 4.2.e - New Specification

6. 4.2.f- New Specification (previously covered by old 4.4.a)

7. 4.2.g - New Specification (previously covered by old 4.4.a)

8. 4.2.h - New Specification (previously covered by old 4.4.a)

9. 4.2.i - New Specification

10. 4.2.j - New Specification

c. Section 4.3, Reactor Coolant Systems:

I. 4.3.a - New Specification (previously covered by old 4.2.b)

2. 4.3.b was old 4.2.c

5. 4.3.c - New Specification

6. 4.3.d - New Specification

7. 4.3.e - New Specification

8. 4.3.f- New Specification

9. 4.3.g was old 4.2.a

d. Section 4.4, Reactor Containment Building:

1. 4.4.awasold4.1.a

2. 4.4.b was old 4.1.b

e. Section 4.5, Reactor Instrumentation:

1. 4.5.a - New Specification (previously covered by old 4.4.a)

2. 4.5.b - New Specification (previously covered by 4.4.a)

3. 4.5.c was old 4.4.c

f. Section 4.6, Emergency Electrical Power System:

1. 4.6.a was old 4.6.b

2. 4.6.b was old 4.6.c

g. Section 4.7, Radiation Monitoring Systems Airborne Effluents:

1. 4.7.a was old 4.4.b

2. 4.7.b - New Specification (previously covered by old 4.4.a)

h. Section 4.8, Experiments:

7 of 9

1. 4.8.a - New Specification

2. 4.8.b - New Specification

i. Section 4.9, Auxiliary Systems:

1. 4.9.a -New Specification

2. 4.9.b was old 4.6.a Section 5.0, Design Features:

Added "General" wording at the beginning of Section.

a. Section 5.1, Site

Description:

1. 5.].a was old 5.l.a

b. Section 5.2, Reactor Coolant Systems:

1. 5.2.a was old 5.3.a

2. 5.2.b was old 5.3.b

3. 5.2.c was old 5.3.c

4. 5.2.d was old 5.3.d

5. 5.2.e was old 5.3.e

6. 5.2.f was old 5.3.f

7. 5.2.g was old 5.3.g

8. 5.2.h was old 5.3.h

9. 5.2.i was old 5.3.i

10. 5.2.j was old 5.3.j

11. 5.2.k was old 5.3.k

c. Section 5.3, Reactor Core and Fuel:

I. 5.3.a was old 3.L.a

2. 5.3.b was old 3.1.b

3. 5.3.c was old 3.1.c

4. 5.3.d was old 5.4.a

5. 5.3.e was old 5.4.b

6. 5.3.f was old 5.4.c

7. 5.3.g was old 5.4.d

8. 5.3.h was old 5.4.e

9. 5.3.i was old 5.4.f

10. 5.3.j was old 5.4.g

d. Section 5.4, Fuel Storage:

1. 5.4.a was old 3.8.d

2. 5.4.b was old 3.8.e

e. Section 5.5, Reactor Containment Building:

1. 5.5.a was old 5.2.a

2. 5.5.b was old 5.2.b

3. 5.5.c was old 5.2.c

4. 5.5.d was old 5.2.d

f. Section 5.6, Emergency Electrical Power System:

1. 5.6.a was old 5.5.a Section 6.0, Administrative Controls

a. Section 6.1, Organization:

8 of 9

1. 6.1.a was old 6. L.a

2. 6. L.b was old 6. .b

3. 6.1.c was old 6.1.c

4. 6. l.d - New Specification

5. 6.1.e was old 6.1.d

b. Section 6.2, Review and Audit:

1. 6.2.a was old 6.2.a

2. 6.2.b was old 6.2.b with the following change: Wording was added to the first sentence of the first paragraph: "knowledgeable members of the public" and in the last sentence of the second paragraph "once during each calendar quarter" was replaced by "quarterly."

3. 6.2.c was old 6.2.c

4. 6.2.d was old 6.2.d

5. 6.2.e was old 6.2.e

c. Section 6.3, Radiation Safety:

1. 6.3.a - New Specification

d. Section 6.4, Procedures:

1. 6.4.a was old 6.3.a

2. 6.4.b was old 6.3.b

3. 6.4.c was old 6.3.c

4. 6.4.d was old 6.3.d

e. Section 6.5, Experimental Review:

1. 6.5.a - New Specification

2. 6.5.b - New Specification

f. Section 6.6, Reportable Events and Required Actions:

I. 6.6.a was old 6.5.a

2. 6.6.b was old 6.5.b

3. 6.6.c was old 6.5.c

4. 6.6.d was old 6.5.d

5. 6.6.e was old 6.5.e

g. Section 6.7, Records:

1. 6.7.a was old 6.4.a 2: 6.7.b was old 6.4.b 9 of 9

APPENDIX A TECHNICAL SPECIFICATIONS FOR THE UNIVERSITY OF MISSOURI RESEARCH REACTOR FACILITY OPERATING LICENSE R- 103 DOCKET 50-186

THIS PAGE INTENTIONALLY LEFT BLANK Introduction The Technical Specifications represent the 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 the general public in accordance with 10 CFR 50.36.

This document is divided into the following six sections:

Section 1 - Definitions Section 2 - Safety Limits (SL) and Limiting Safety System Settings (LSSS)

Section 3 - Limiting Conditions for Operation (LCO)

Section 4 - Surveillance Requirements Section 5 - Design Features Section 6 - Administrative Controls Specific limitations and equipment requirements for safe reactor operation and for dealing with abnormal situations are called specifications. These specifications, typically derived from the facility descriptions and safety considerations contained in the Safety Analysis Report (SAR),

represent a comprehensive envelope of safe operation. Only those operational parameters and equipment requirements directly related to preserving that safe envelope are listed in the Technical Specifications. Procedures or actions employed to meet the requirements of these Technical Specifications are not included in the Technical Specifications. Normal operation of the reactor within the limits of the Technical Specifications will not result in off-site radiation exposure in excess of 10 CFR 20 guidelines.

Specifications in Sections 2, 3, 4 and 5 provide related information in the following format shown:

" Applicability - This indicates which components are involved;

Objective - This indicates the purpose of the specification(s);

" Specification(s) - This provides specific data, conditions, or limitations that bound a system or operation. This is the most important statement in the Technical Specifications; and

" Bases - This provides the background or reasoning for the choice of specification(s), or references a particular section of the SAR that does.

Section 6, Administrative Controls, simply state the applicable specification(s).

Although the applicability, objective and bases provide important information, only the

'"specification(s)" statement is governing.

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TABLE OF CONTENTS 1.0 D E F IN IT IO N S .................................................................................... A -1 1.1 A bnorm al O ccurrences .................................................................. A -i 1.2 C enter T est H ole .......................................................................... A -1 1.3 C hannel ............................................................................. . . . . A -1 1.4 C hannel C alibration ................................................................... A -1 1.5 C hannel C heck ........................................................................ A -1 1.6 C hannel T est .............................................................................. A -2 1.7 C ontrol B lade (R od) ..................................................................... A -2 1.8 E xcess R eactivity ......................................................................... A -2 1.9 E xperim ent ................................................................................. A -2 1.10 F lu x T rap .................................................................................. A -2 1 .11 Irrad iated F uel ............................................................................ A -2 1.12 Limiting Safety System Settings ....................................................... A-2 1.13 M ovable Experim ent ..................................................................... A -2 1.14 O p erab le .................................................................................... A -3 1.15 O p eratin g .................................................................................. A -3 1 .16 O perational M odes ........................................................................ A -3 1.17 Reactivity Worth of an Experiment .................................................... A-3 1.18 Reactor Containment Building ......................................................... A-3 1.19 R eactor C ore .............................................................................. A -3 1.20 R eactor O perator .......................................................................... A -3 1.2 1 Reactor in O peration ..................................................................... A -3 1.22 Reactor Safety System ................................................................... A -3 1.23 Reactor Scram ............................................................................. A -4 1.24 Reactor Secured ........................................................................... A -4 1.25 Reactor Shutdow n ........................................................................ A -4 1.26 Reference C ore C ondition ............................................................... A -4 1.27 Regulating B lade (R od) .................................................................. A -4 1.28 Removable Experiment .............................................................. A-5 1.29 R esearch R eactor ......................................................................... A -5 1.30 Research Reactor Facility ............................................................ A-5 1.31 R od R un-In System ................................................................... A -5 1.32 S afety L im its .............................................................................. A -5 1.33 Secured Experim ent ................................................................... A -5 1.34 Senior Reactor Operator .............................................................. A-5 1.35 Shim B lade (R od) ..................................................................... A -5 1.36 Shall, Should, and M ay .................................................................. A -6 1.37 Shutdow n M argin ........................................................................ A -6 1.38 Surveillance Intervals .................................................................... A -6 1.39 T rue V alue ................................................................................. A -6 1.40 U nscheduled Shutdow n .................................................................. A -6 1.4 1 U nsecured Experim ent ................................................................... A -6 ii

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS .................... A-7 2.1 Safety L im its .............................................................................. A -7 2.2 Limiting Safety System Settings ...................................................... A-12 3.0 LIMITING CONDITIONS FOR OPERATION ............................................. A-13 3.1 Reactor Core Parameters ............................................................ A-13 3.2 Reactor Control and Reactor Safety Systems .................................... A-15 3.3 Reactor Coolant Systems ............................................................... A-22 3.4 Reactor Containment Building ........................................................ A-25 3.5 R eactor Instrum entation ............................................................... A -27 3.6 Emergency Electrical Power System ................................................. A-29 3.7 Radiation Monitoring Systems and Airborne Effluents ........................... A-30 3.8 E xperim ents .............................................................................. A -32 3.9 A uxiliary System s...................................................................... A -37 4.0 SURVEILLANCE REQUIREMENTS ....................................................... A-38 4.1 Reactor Core Parameters ............................................................... A-38 4.2 Reactor Control and Reactor Safety Systems ....................................... A-40 4.3 Reactor Coolant Systems ............................................................... A-42 4.4 Reactor Containment Building ........................................................ A-44 4.5 R eactor Instrum entation ................................................................ A -45 4.6 Emergency Electrical Power System ................................................. A-46 4.7 Radiation Monitoring Systems and Airborne Effluents ........................... A-47 4 .8 E xperim ents .............................................................................. A -4 8 4 .9 A uxiliary System s....................................................................... A -49 5.0 DESIGN FEATURES ........................................................................ A-50 5.1 Site D escription ......................................................................... A-50 5.2 Reactor Coolant Systems ............................................................... A-52 5.3 R eactor C ore and Fuel .................................................................. A -54 5.4 F uel Storage .............................................................................. A -56 5.5 Reactor Containment Building ........................................................ A-57 5.6 Emergency Electrical Power System ................................................. A-59 6.0 ADMINISTRATIVE CONTROLS ............................................................ A-60 6 .1 O rganization .............................................................................. A -60 6.2 Review and A udit ....................................................................... A -6 1 6.3 R adiation Safety ......................................................................... A -63 6 .4 P rocedures ............................................................................... A -63 6.5 Experiment Review and Approval .................................................... A-64 6.6 Reportable Events and Required Actions ............................................ A-64 6 .7 Reco rd s ................................................................................... A -6 6 iii

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UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 1.0 DEFINITIONS 1.1 Abnormal Occurrences - An abnormal occurrence is any of the following which occurs during reactor operation:

a. Operation with actual safety system settings for required systems less conservative than specified in Section 2.2, Limiting Safety System Settings;

b. Operation in violation of Limiting Conditions for Operation established in Section 3.0;

c. A reactor safety system component malfunction which renders or could render the reactor safety system incapable of performing its intended safety function unless the malfunction or condition is discovered during maintenance tests or periods of reactor shutdowns;

d. An unanticipated or uncontrolled change in reactivity in excess of 0.006 Ak/k. Reactor trips resulting from a known cause are excluded;

e. Abnormal and significant degradation in reactor fuel or cladding, or both; primary coolant boundary, or containment boundary (excluding minor leaks), which could result in exceeding prescribed radiation exposure limits of personnel or environment, or both; or

f. An observed inadequacy in the implementation of administrative or procedural controls such that the inadequacy causes or could have caused the existence or development of an unsafe condition involving operation of the reactor.

1.2 Center Test Hole - The center test hole is that volume in the flux trap occupied by the removable experiment sample canister.

1.3 Channel - A channel is the combination of sensor, line, amplifier, and output devices that are connected for the purpose of measuring the value of a parameter.

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

1.5 Channel Check - A channel check is a qualitative verification of acceptable performance by observation of channel behavior. This verification, where possible, shall include comparison of the channel with other independent channels or systems measuring the same variable.

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UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R- 103 1.0 DEFINITIONS - Continued 1.6 Channel Test - A channel test is the introduction of a simulated input signal into channel and the observation of proper channel response. When applicable, the test shall include verification of proper safety trip operation.

1.7 Control Blade (Rod) - A control blade (rod) is either a shim blade (rod) or the regulating blade (rod). The words blade and rod can be used interchangeably.

1.8 Excess Reactivity - Excess reactivity is that amount of reactivity that would exist if all of the control blades were moved to the fully withdrawn position from the point where the reactor is exactly critical (K~ff= 1).

1.9 Experiment - An experiment is any operation, hardware, or target (excluding devices such as detectors or foils) which is designed to investigate non-routine reactor characteristics or which is intended for irradiation within an irradiation facility. Hardware rigidly secured to a core or shield structure so as to be part of their design to carry out experiments is not normally considered an experiment.

1.10 Flux Trap - The flux trap is that portion of the reactor through the center of the core bounded by the 4.5-inch inside diameter tube and 15 inches above and below the reactor core horizontal center line.

1.11 Irradiated Fuel - Irradiated fuel is any fuel element which has been irradiated and used to an integrated power of:

a. Greater than 0.10 megawatt-day; OR

b. Less than or equal to 0. 10 megawatt-day but greater than 1.0 kilowatt-day and with a decay time of less than 7 days since last irradiation; OR

c. Less than or equal to 1.0 kilowatt-day and with a decay time of less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months since last irradiation.

1.12 Limiting Safety System Settings - Limiting Safety System Settings (LSSS) are settings for automatic protection devices related to those variables having significant safety functions. Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting shall be so chosen that automatic protective action will correct the most severe abnormal situation anticipated before a safety limit is exceeded.

1.13 Movable Experiment - A movable experiment is one which is designed with the intent that it may be moved into, out of, or in the near proximity of the reactor while the reactor is operating.

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UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R- 103 1.0 DEFINITIONS - Continued 1.14 Operable - Operable means a component or system is capable of performing its intended function.

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

1.16 Operational Modes - The reactor may be operated in any of three operating modes, depending upon the configuration of the reactor coolant systems and the protective system set points.

a. Operational Mode I - Reactor can be operated safely at a thermal power level often megawatts or less.

b. Operational Mode II - Reactor can be operated safely at a thermal power level of five megawatts or less.

c. Operational Mode III - Reactor can be operated safely at a thermal power level of fifty kilowatts or less.

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

1.18 Reactor Containment Building - The reactor containment building is a reinforced concrete structure within the facility site which houses the reactor core, pool, and irradiated fuel storage facilities.

1.19 Reactor Core - The reactor core shall be considered to be that volume inside the reactor pressure vessels occupied by eight or less fuel elements.

1.20 Reactor Operator - A reactor operator is an individual who is certified to manipulate the controls of a reactor.

1.21 Reactor in Operation - The reactor shall be considered in operation unless it is either shutdown or secured.

1.22 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.

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UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 1.0 DEFINITIONS - Continued 1.23 Reactor Scram - A reactor scram is the insertion of all four shim rods by gravitational force as a result of removing the holding current from the shim rod drive mechanism electromagnets.

1.24 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 shim 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 exits:

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 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.

1.25 Reactor Shutdown - The reactor shall be considered shutdown when all four of the shim rods are fully inserted and power is unavailable to the shim rod drive mechanism electromagnets.

1.26 Reference Core Condition - Reference core condition is the condition of the core when it is at ambient temperature (cold) and the reactivity worth of xenon is negligible (< 0.002 Ak/k).

1.27 Regulating Blade (Rod) - The regulating blade (rod) is a low worth control blade (rod) used for very fine adjustments in the neutron density in order to maintain the A-4

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 1.0 DEFINITIONS - Continued reactor at the desired power level. The regulating blade (rod) may be controlled by the operator with a manual switch or push button, or by an automatic controller.

1.28 Removable Experiment - A removable experiment is any experiment which can reasonably be anticipated to be moved during the life of the reactor.

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

1.30 Research Reactor Facility - A research reactor facility includes all areas within which the owner or operator directs authorized activities associated with the reactor.

1.31 Rod Run-In System - The rod run-in system is that combination of sensing devices, electronic circuits and equipment, signal conditioning equipment, and electro-mechanical devices that serves to effect a rod run-in. A rod run-in is the automatic insertion of the shim blades at a controlled rate should a monitored parameter exceed a predetermined value. This system is not part of the reactor safety system, as defined by Definition 1.22; however, it does provide a protective function by introducing shim blade insertion to terminate a transient prior to actuating the reactor safety system.

1.32 Safety Limits - Safety Limits (SL) are limits placed upon important process variables which are found to be necessary to reasonably protect the integrity of the principal physical barriers which guard against the uncontrolled release of radioactivity.

1.33 Secured Experiment - A secured experiment is any experiment which is rigidly held in place by mechanical means with sufficient restraint to withstand any anticipated forces to which the experiment might be subjected to.

1.34 Senior Reactor Operator - A senior reactor operator is an individual who is certified to direct the activities of reactor operators and manipulate the controls of a reactor.

1.35 Shim Blade (Rod) - A shim blade (rod) is a high worth control blade (rod) used for coarse adjustments in the neutron density and to compensate for routine reactivity losses. The shim blade (rod) is magnetically coupled to its drive mechanism allowing it to perform its safety function when the electromagnet is de-energized.

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UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R- 103 1.0 DEFINITIONS - Continued 1.36 Shall, Should, and May - The word "shall" is used to denote a requirement; the word "should" is used to denote a recommendation; and the word "may" is used to denote permission, neither a requirement nor a recommendation.

1.37 Shutdown Margin - Shutdown margin is the minimum shutdown reactivity necessary to provide confidence that the reactor can be made subcritical by means of the control and reactor safety systems starting from any permissible operating condition and with the most reactive shim blade and the regulating blade in the fully withdrawn positions, and that the reactor will remain subcritical without further operator action.

1.38 Surveillance Intervals - Surveillance intervals are the maximum allowable intervals established to provide operational flexibility and not reduce frequency.

Established frequencies shall be maintained over the long term. The surveillance interval is the time between a check, test or calibration, whichever is appropriate to the item being subjected to the surveillance, and is measured from the date of the last surveillance. Allowable surveillance intervals shall not exceed the following:

a. Annual - interval not to exceed 15 months.

b. Semiannual - interval not to exceed 7.5 months.

c. Quarterly - interval not to exceed 4 months.

d. Monthly - interval not to exceed 6 weeks.

e. Weekly - interval not to exceed 10 days.

f. Within a shift - must be done during a reactor shift.

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

1.40 Unscheduled Shutdown - An unscheduled shutdown is defined as any unplanned shutdown, that occurs after all "Blade Full-In Lights" have cleared, caused by actuation of the reactor safety system, rod run-in system, operator error, equipment malfunction, or a manual shutdown in response to conditions that could adversely affect safe operation, not including shutdowns that occur during testing or checkout operations.

1.41 Unsecured Experiment - An unsecured experiment is any experiment which is not secured as defined by Definition 1.33, or the moving parts of secured experiments when they are in motion.

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UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limits Applicability:

This specification applies to the interrelated variables associated with reactor core thermal and hydraulic performance. These measurable operating or process variables include reactor power level, core flow rate, reactor inlet water temperature, and pressurizer pressure.

Objective:

The objective of this specification is to define a four-dimensional safety limit envelope such that operation within this envelope will assure that the integrity of the fuel element cladding is maintained.

Specification:

Reactor power level, core flow rate, reactor inlet water temperature, and pressurizer pressure shall not exceed the following limits during reactor operation:

a. Mode I and II Operation (Core Flow Rate > 400 gpm)

The combination of the true values of reactor power level, reactor core flow rate, and reactor inlet water temperature shall not exceed the limits plotted on Figures 2.0, 2.1 and 2.2. The limits are considered exceeded if, for core flow rates greater than or equal to 400 gpm, the point defined by reactor power level and core flow rate is at any time above the curve corresponding to the true values of reactor inlet water temperature and pressurizer pressure. To define values of the safety limits for temperatures and/or pressures not shown in Figures 2.0, 2.1 and 2.2, interpolation or extrapolation of the data on the curves shall be used. For pressurizer pressures greater than 85 psia, the 85 psia curves (Figure 2.2) shall be used and no pressure extrapolation shall be permitted.

b. Mode I and II Operation (Core Flow Rate < 400 gpm)

Steady-state power operation in Modes I and II is not authorized for a core flow rate less than 400 gpm. Reactor operation with a core flow rate below 400 gpm will occur only after a normal reactor shutdown when the primary coolant circulation pumps are secured or following a loss of flow transient. Under the above conditions, the maximum fuel element cladding temperature shall not approach a temperature that would challenge the integrity of the fuel element cladding.

c. Mode III Operation Reactor power is limited to a maximum of 150 kilowatts.

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UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 2.1 Safety Limits - Continued Bases:

a. A complete safety limit analysis for the MURR is presented in Section 4.6.3 of the SAR. A family of curves is presented which relate reactor inlet water temperature and core flow rate to the reactor power level corresponding to a departure from nucleate boiling ratio (DNBR) of 1.2. This is based on the burnout heat flux data experimentally verified for Advanced Test Reactor (ATR) type fuel elements. Curves are presented for pressurizer pressures of 60, 75 and 85 psia. The safety limits were chosen from the results of this analysis for Mode I and II operation, i.e., forced convection operation with greater than 400 gpm flow.

b. Steady-state reactor operation is prohibited for core flow rates less than 400 gpm by the low flow scram settings in the reactor safety system. The region below 400 gpm will only be entered following a reactor shutdown when the primary coolant circulation pumps are secured or during a loss of flow transient where the reactor scrams., the flow coasts to zero, reverses, and natural convective cooling is established through the decay heat removal system. The analysis of a loss of flow transient presented in Section 13.2.4 of the SAR, from the ultraconservative conditions of 11 MW of power, a core flow rate of 3,800 gpm, and a reactor inlet water temperature of 155 'F, indicated a maximum fuel plate centerline temperature of 280.3 'F and a maximum coolant channel temperature of 237.5 'F, which is well below the saturation temperature of 277 'F.

c. Analysis of natural convective cooling of the core (Mode III Operation) is presented in Section 4.6.1 of the SAR.

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22 20 Figure 2.0 MvU-RR Mode I and II Safety Limit Curve 18 Pressurizer at 60 psia Core Flow Rate > 400 GPM 1l6 Reactor Inlet Water Temperature, IF

~14 120 140

, 160 _ "

L 12 180 -- -

tll S10 - .200. ........

S8 0 H cj.~

6 ~r1 0

4 NOTE CD For Core Flow Rates below 400 GPM the Safet, Limit is ~r1 2 established by Specification 2.1.b. 00 cj~

-e 0-400 800 1200 1600 2000 2400 2800 3200 3600 4000 00 H

Core Flow Rate (GPM) 0 H zcj~ 0

22 Figure 2.1 MURR Mode I and It 1 Safety Limit Cur.ve Pressurizer at 75 psia ...............

18 Cowe Flow Rate > 400 GPM Reactor Inlet Water Temperature. 'F 14 - . . -

16 12 12 ... .. . ...........

. 2 00 QL 10 8 .. . . . . . .. . . .. .... ..... ...... ... ..! . . . . . . .. .. . . . . ................

4 NOTE For Core Flow Rates belowv 400 GPM the Safety Limit is 2 establisied b Specification 2-l1b- 0 L-r-400 800 1200 1600 2000 2400 2800 3200 3600 4000 = n Core Flow Ratc (GPM)

SOH

22 20 Figure 2.2 _

MURR Safety Mode I and I1 Limit Curv-e Safey Liit CrveReactor Inlet Water' .. .......... i 18r 18 .......... F lo Pressurizer at 850psia P. .4.G ..........

. ... .. ....... . Ieaelm pne ra ttleture. ....... .

re , *'T-ii*.................. ................

Core Flow Rate > 400 GPM4 16

- - -- - .- -- ---- - - - - -- - _ _ 120 ,-

0 2120

. 126 . . . ...... ........................

.. . . . . . . . . ....... . ... ...... .. . .. ... 1 4 . .. .......

10 U

10 - .-........ ... ..

S8 .-.~ .

4. NOTE For Core Flow Rates below 400 GPM the Safety Limit is 2 established by Specification 2 I >

0 -. . .. . -..........

.. .... .. . . . . . .. .. .. ... . . .... .. .. . . .. . .. .. . ... . " . .. . . . . . . ... .I _ _ _ _ _ _ _ - -

400 800 1200 1600 2000 2400 2800 3200 3600 4000 3 Core Ilow Rate (GPM) n

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UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 2.2 Limiting Safety System Settings Applicability:

This specification applies to the set points for the reactor safety channels monitoring reactor power level, primary coolant flow, reactor inlet water temperature and pressurizer pressure.

Objective:

The objective of this specification is to assure that automatic protective action is initiated to prevent a safety limit from being exceeded.

Specification:

a. Mode I Operation Reactor Power Level (10 MW) 125% of full power (Maximum)

Primary Coolant Flow 1,625 gpm either loop (Minimum)

Reactor Inlet Water Temperature 155 'F (Maximum)

Pressurizer Pressure 75 Psia (Minimum)

b. Mode II Operation Reactor Power Level (5 MW) 125% of full power (Maximum)

Primary Coolant Flow 1,625 gpm (Minimum)

Reactor Inlet Water Temperature 155 'F (Maximum)

Pressurizer Pressure 75 Psia (Minimum)

C. Mode III Operation Reactor Power Level (50 kW) 125% of full power (Maximum)

Bases:

a. - b. The limiting safety system settings (LSSS) are set points which, if exceeded, will cause the reactor safety system to initiate a reactor scram. The LSSS were chosen such that the true value of any of the four safety-related variables, i.e., reactor power level, core flow rate, reactor inlet water temperature and pressurizer pressure will not exceed a safety limit under the most severe anticipated transient.

Section 4.6.4 of the SAR presents analyses to show that the LSSS for Mode I and II operation meet this criterion.

c. For Mode III operation, the high power scram set point of 125% of full power will occur at 62.5 kW, thus, there is a margin of 87.5 kW between the LSSS and the safety limit of 150 kW.

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UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 3.0 LIMITING CONDITIONS FOR OPERATION General: Limiting Conditions for Operation (LCO) are those administratively established constraints on equipment and operational characteristics that shall be adhered to during operation of the facility. The LCOs are the lowest functional capability or performance level required for safe operation.

3.1 Reactor Core Parameters Applicability:

This specification applies to the reactor core and fuel elements used in the reactor core.

Obiective:

The objective of this specification is to assure that the reactor can be controlled and shut down at all times and that the fuel elements are operated within acceptable design considerations thus ensuring fuel element integrity is maintained.

Specification:

a. The reactor core excess reactivity above reference core condition shall not exceed 0.098 Ak/k.

b. The reactor shall be subcritical by a margin of at least 0.02 Ak/k with the most reactive shim blade and the regulating blade in their fully withdrawn positions.

c. The reactor core shall consist of eight fuel assemblies.

Exception: The reactor may be operated to 100 watts above shutdown power on less than eight assemblies for the purposes of reactor calibration or multiplication measurement studies.

d. The peak burnup for UAIx dispersion fuel shall not exceed a calculated 2.3 x 1021 fissions per cubic centimeter.

e. The reactor will not be operated using fuel in which anomalies have been detected or in which the dimensional changes of any coolant channel between the fuel plates exceeds ten (10) mils.

Bases:

a. Specification 3.1.a provides additional assurance that Specification 3.1.b is satisfied.

b. Specification 3.1.b assures that a shutdown margin, as defined by Definition 1.37, is maintained.

c. Operation at a power level greater than 100 watts requires a full core of eight full elements to assure the validity of the safety limit curves (Specification 2.1) and A-13

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 3.1 Reactor Core Parameters - Continued other safety analyses. When it may be important to conservatively determine the actual critical core loading, Specification 3.1.c allows operation with less than eight fuel elements up at a lower level not to exceed 100 watts. This maximum power limit is low enough to ensure no fuel damage will occur. This provides for a conservative approach to criticality with less than eight new fuel elements.

Typically, the first approach to critical would be with a number of fuel elements insufficient to achieve criticality but be able to observe subcritical multiplication.

Then one additional fuel element would be added at a time in between approaches to critical. The reactor would be operated in this manner only to perform necessary conservative approaches to criticality.

d. Specification 3.1.d restricts the peak fissions per cubic centimeter burnup to values that have been correlated to result in less than 10% swelling of the fuel plates. It has been found that fuel plate swelling of less than 10% has no detrimental effect on fuel plate performance (Ref.: Change No. 4 to Facility License R-103, Change No. 6 to Facility License R-103, and Application dated September 12, 1986 with supplements).

e. Specification 3.1 .e assures that fuel elements which have been inspected and found to be defective are no longer used for reactor operation.

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UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 3.2 Reactor Control and Reactor Safety Systems Applicability:

This specification applies to the reactor control and reactor safety systems.

Objective:

The objective of this specification is to reasonably assure proper operation of the reactor control system, thus avoiding conditions which could jeopardize the integrity of the fuel element cladding or endanger personnel health and safety, and to specify the minimum number of reactor safety system instrument channels that must be operable for safe reactor operation.

Specification:

a. All control blades, including the regulating blade, shall be operable during reactor operation.

b. Above 100 kilowatts, the reactor shall be operated so that the maximum distance between the highest and lowest shim blade shall not exceed one inch.

c. The shim blades shall be capable of insertion to the 20% withdrawn position in less than 0.7 seconds.

d. The maximum rate of reactivity insertion for the regulating blade shall not exceed 2.5 x 10-4 Ak/k/sec.

e. The maximum rate of reactivity insertion for the four shim blades operating simultaneously shall not exceed 3.0 x 10-4 Ak/k/sec.

f. The reactor shall not be operated unless the following rod run-in functions are operable. Each of the rod run-in functions shall have 1/N logic where N is the number of instrument channels required for the corresponding mode of operation.

Number Required (N)

Rod Run-In Function Mode I Mode I1 Mode III Trip Set Point

1. High Power Level 3 3 3 115% of full power (Max)

2. Reactor Period 2 2 2 10 Seconds (Min)

3. Pool Low Water Level I 1 0 27 feet (Min)

4. Vent Tank Low Level 1 1 0 1 foot below centerline (Min)

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UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R- 103 3.2 Reactor Control and Reactor Safety Systems - Continued Number Required (N)

Rod Run-In Function Mode I Mode II Mode III Trip Set Point

5. Rod Not-In-Contact 4 4 4 Magnet disengaged With Magnet from any rod

6. Anti-Siphon System 1 1 1il' 6 inches above High Level valves (Max)

7. Truck Entry 1 1 1 Loss of entry door seal pressure

8. Regulating Blade 2 2(2) 2(2) < 10% withdrawn Position and bottomed

9. Manual Rod Run-In 1 1 1 Push button on Control Console (I) Not required (a) below 50 kW operation with the natural convection flange and reactor pressure vessel cover removed or (b) in operation with the reactor subcritical by a margin of at least 0.0 15 Ak/k.

(2) Not required during calibration measurements of the regulating blade.

g. The reactor safety system and the number (N) of associated instrument channels necessary to provide the following scrams shall be operable whenever the reactor is in operation. Each of the safety system functions shall have 1/N logic where N is the number of instrument channels required for the corresponding mode of operation.

Reactor Safety System Number Required (N)

Instrument Channel Mode I Mode II Mode III Trip Set Point

1. High Power Level 3 3 3 125% of full power (Max)

2. Reactor Period 2 2 2 8 Seconds (Min)

3. Primary Coolant Flow 4 2 2(l) 1,625 gpm 21 (Min)

4. Differential Pressure 1 0 0 3,200 gpm13 (Min)

Across the Core

5. Differential Pressure 0 1 1(1) 1,600 gpm( 3 ) (Min)

Across the Core A-16

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 3.2 Reactor Control and Reactor Safety Systems - Continued Reactor Safety System Number Required (N)

Instrument Channel Mode I Mode II Mode III Trip Set Point

6. Primary Coolant Low 4 4 4(1) 75 psia(5) (Min)

Pressure

7. Reactor Inlet Water 2 1 1l() 155 'F (Max)

Temperature

8. Reactor Outlet Water 1 1 1i1) 175 'F (Max)

Temperature

9. Pool Coolant Flow 2 2 0 850 gpm(4) (Min)

10. Differential Pressure 1 0 0 2.52 psi (Min)

Across the Reflector 8.00 psi (Max)

11. Differential Pressure 0 1 0 0.63 psi (Min)

Across the Reflector 2.00 psi (Max)

12. Pressurizer High 1 1 l1') 95 psia (Max)

Pressure

13. Pressurizer Low Water 1 1 1i1' 16 inches below Level centerline (Min)

14. Pool Low Water Level 0 0 1 23 feet (Min)

15. Primary Coolant 1 1 I1* Either valve off Isolation Valves open position 507A/B Off Open Position

16. Pool Coolant Isolation 1 1 0 Valve 509 off open Valve 509 Off Open position Position

17. Power Level Interlock I1 1 Scram as a result of incorrect selection of operating mode

18. Facility Evacuation 11 1 Scram as a result of actuating the facility evacuation system A-17

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 3.2 Reactor Control and Reactor Safety Systems - Continued Reactor Safety System Number Required (N)

Instrument Channel Mode I Mode II Mode III Trip Set Point

19. Reactor Isolation I I 1 Scram as a result of actuating the reactor isolation system

20. Manual Scram 1 1 1 Push button on Control Console

,)(6)(6()

21. Center Test Hole 2(6) 2(6) Scram as a result of removing the center test hole removable experiment test tubes or strainer (I) Not required (a) below 50 kW operation with the natural convection flange and pressure vessel cover removed or (b) in operation with the reactor subcritical by a margin of at least 0.0 15 Ak/k.

(2) Flow orifice or heat exchanger AP (psi) in each operating heat exchanger leg corresponding to the flow value in the table.

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

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

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

(6) Not required if reactivity worth of the center test hole removable experiment test tubes and its contents or the strainer is less than the reactivity limit of Specification 3.8.q. This safety function shall only be bypassed with specific authorization from the Reactor Manager.

A-18

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R- 103 3.2 Reactor Control and Reactor Safety Systems - Continued

h. The following reactor control interlocks shall be operable whenever the reactor is in operation.

Minimum Numbers Interlock Function Operable I. Rod Withdrawal Prevents the control rods from Prohibit being withdrawn unless certain control system logic functions have been satisfied

2. Automatic Control Prevents placing the reactor in Prohibit automatic control unless certain control system logic functions have been satisfied Bases:

a. Specification 3.2.a ensures that the normal method of reactivity control is used during reactor operation.

b. Specification 3.2.b provides a restriction on the maximum neutron flux tilting that can occur in the core to ensure the validity of the power peaking factors described in Section 4.5 of the SAR.

c. Specification 3.2.c assures prompt shutdown of the reactor in the event a scram signal is received as analyzed in Section 13.2.2 of the SAR. The 20% level is defined as 20% of the shim blade full travel as measured from the frilly inserted position. Below the 20% level, the fall of the shim blade is cushioned by a dashpot assembly. Approximately 91% of the shim blade total worth is inserted at the 20% level.

d. Specification 3.2.d limits the rate of reactivity addition by the regulating blade to provide for a reasonable response from operator control.

e. Specification 3.2.e assures that power increases caused by control rod motion will be safely terminated by the reactor safety system. The continuous control rod withdrawal accident is analyzed in Section 13.2.2 of the SAR.

f. The specifications on high power level and short reactor period are provided to introduce shim blade insertion on a reactor transient before the reactor safety system trip is actuated.

A-19

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 3.2 Reactor Control and Reactor Safety Systems - Continued The low pool level rod run-in provides assurance that the radiation level from direct core radiation above the pool will not exceed 2.5 mR/h (Ref. Section 11.1.5.1 of the SAR).

The vent tank low level rod run-in prevents reactor operation with a vent tank level which could result in the introduction of air into the primary coolant system (Ref. Section 9.13 of the SAR).

The anti-siphon system high level rod run-in provides assurance that the introduction of air to the invert loop is sufficiently rapid to prevent a siphoning action following a rupture of the primary coolant piping (Ref. Section 6.3 of the SAR).

The rod not-in-contact with magnet rod run-in assures the reactor cannot be operated in violation of Specification 3.2.b due to a dropped rod.

The specification on the truck entry door prohibits reactor operation without the door's contribution to containment integrity as required by Specification 3.4.a.

The regulating blade rod run-ins ensure termination of a transient which, in automatic control, is causing a rapid insertion of the regulating blade.

g. The specifications on high power level, primary coolant flow, primary coolant pressure and reactor inlet water temperature provide for the limiting safety system settings outlined in Technical Specifications 2.2.a, 2.2.b and 2.2.c. In Mode I and Mode II operation, the core differential temperature is approximately 17 'F and, therefore, the reactor outlet water temperature scram set point at 175 'F provides a backup to the high reactor inlet water temperature scram. The core differential pressure scram provides a backup to the primary coolant low flow scrams.

The reactor period scram assures protection of the fuel elements from a continuous control blade withdrawal accident as analyzed in Section 13.2.2 of the SAR.

With the reflector plenum natural convection valve V547 in the open position and a pool coolant flow rate at 850 gpm, the pool coolant low flow scram assures the adequate cooling of the reactor pool, reflectors, control rods, and the flux trap (Ref. Section 5.3.5 of the SAR). The reflector high and low differential pressure scram provides a backup to the low pool coolant flow scram.

The pressurizer high pressure scram provides assurance that the reactor will be shut down during a high pressure transient before the relief valve set point or the A-20

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 3.2 Reactor Control and Reactor Safety Systems - Continued pressure limit of the primary coolant system is reached as analyzed in Section 13.2.9.4 of the SAR.

The pressurizer low level scram provides assurance that the reactor will be shut down on a loss of coolant accident before the pressurizer level decreases sufficiently to introduce nitrogen gas into the primary coolant system.

The pool water low level scram assures that the radiation level above the reactor pool from direct core radiation remains below 2.5 mR/h (Ref. Section 11.1.5.1 of the SAR).

The reactor scrams caused by the primary and pool coolant isolation valves (V507A/B and V509) leaving their full open position provide the first line of protection for a loss of flow accident (in their respective 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 Power Level Switch. The PLI scram also provides the interlocks to assure that the reactor cannot be operated in Mode I with a primary or pool coolant low flow scram bypassed.

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

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

The center test hole scram provides assurance that the reactor cannot be operated unless the removable experiment test tubes or the strainer is inserted and latched in the center test hole. This is required anytime the reactivity worth of the center test hole removable experiment test tubes and the contained experiments or the strainer exceeds the limit of Specification 3.8.q (Ref. Section 13.2.2 of the SAR).

The center test hole scram may be bypassed if the total reactivity worth of the removable experiment test tubes and the contained experiments or the strainer does not exceed the limit of Specification 3.8.q and is authorized by the Reactor Manager.

h. Specification 3.2.h assures that certain system conditions have been met prior to conducting a reactor startup or placing the reactor in automatic control at power.

A-21

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 3.3 Reactor Coolant Systems Applicability:

This specification applies to the reactor coolant systems.

Objective:

The objective of this specification is to protect the integrity of the reactor fuel and to prevent the release of fission product radioisotopes.

Specification:

a. The reactor shall not be operated in Modes I or II unless the following components or systems are operable:

(1) Anti-siphon system; (2) Primary coolant isolation valves V507A/B; and (3) In-pool convective cooling system.

b. The reactor shall not be operated with forced circulation unless:

(1) A continuous primary coolant system fuel element failure monitor is operable, OR (2) The primary coolant system is sampled and analyzed at least once every four hours for evidence of fuel element failure.

c. The reactor shall not be operated if a radiochemical analysis of the primary coolant system indicates an iodine-131 concentration of greater than 5 x 10-ptCi/ml.

d. The reactor shall not be operated if a radiochemical analysis of the secondary coolant system exceeds the limits of 10 CFR 20, Appendix B, Table 3, for radioisotopes with half-lives greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

e. The conductivity of the water in the primary coolant system shall be maintained at less than 5 ýLmho/cm when averaged over a period of one month.

f. The pH of the water in the primary coolant system shall be maintained between 5.0 and 7.0 when averaged over a period of one month.

g. The conductivity of the water in the pool coolant system shall be maintained at less than 5 ptmho/cm when averaged over a period of one month.

h. The pH of the water in the pool coolant system shall be maintained between 5.0 and 7.0 when averaged over a period of one month.

A-22

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 3.3 Reactor Coolant Systems - Continued

i. The anti-siphon system will be maintained pressurized to a value of 30 to 45 psig.

In the event of a system low pressure alarm, immediate action will be taken to add air to obtain the specified pressure. The system pressure will be verified, recorded, and readjusted as required every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months as part of the facility routine patrol. Procedures will be established for manual verification of water level in the anti-siphon system for conditions when system pressure has an unexplained rise of 4 psi or more. If water level is 6 inches or more above the anti-siphon isolation valves or a system leak or other malfunction prevents the maintenance of pressure in the specified range, the reactor will be shut down until the malfunction can be corrected.

Bases:

a. The first line of protection against a loss of core water resulting from a rupture of the primary coolant system is provided by the check valve on the inlet line and by the invert loop and the anti-siphon system on the outlet line. Upon opening, the anti-siphon isolation valves will admit a fixed volume of air to the highest point of the invert loop, thus preventing the reactor core from becoming uncovered by breaking any potential siphon which may have been created by the pipe rupture (Ref. Section 6.3 of the SAR).

The primary coolant isolation valves are located on the inlet and outlet primary coolant lines as close as practicable to the biological shield. Proper operation of these valves is not required for protection of the integrity of the fuel elements; however, their operation provides a means for isolation of the in-pool portions of the primary coolant from the remainder of the system.

The in-pool convective cooling system is not required for core protection (Ref.

Section 13.2.9.3 of the SAR); however, its operation is desirable to prevent the formation of steam in the loop and to reduce thermal cycling of the reactor fuel.

b. - c. The primary coolant system with an iodine-131 concentration of 5 x 10-3 V Ci/ml would contain a total iodine-131 inventory of 0.038 Ci in the system. Based on the iodine-131 activity in the reactor core provided in Section 13.2.1.2 of the SAR, this iodine-131 concentration would equate to less than 0.000022 % of the total core iodine-131 inventory in the primary coolant. Specifications 3.3.b and 3.3.c provide for the early detection of a leaking fuel element so that corrective action can be taken to prevent the release of fission products.

d. Secondary coolant system activity is limited to ensure dose rates are maintained below the limits of 10 CFR 20.

e. - h. Experience at many research reactor facilities has shown that maintaining the conductivity and pH within the specified limits provides acceptable control of A-23

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 3.3 Reactor Coolant Systems - Continued corrosion and limits concentrations of particulate and dissolved containments that could be made radioactive by neutron irradiation (NUREG-1537).

Specification 3.3.i ensures that the anti-siphon system will perform its intended function as designed by imposing certain operational limits on the system (Ref.

Section 6.3 of the SAR).

A-24

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R- 103 3.4 Reactor Containment Building Applicability:

This specification applies to the reactor containment building.

Obiective:

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 must 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 operable; (5) The personnel airlock is operable (one door shut and sealed); and (6) The most recent reactor containment building leakage rate test was satisfactory.

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

(1) The reactor is secured, AND (2) Irradiated fuel with a decay time of less than sixty (60) days is not being handled.

c. While 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 times 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 A-25

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R- 103 3.4 Reactor Containment Building - Continued 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-26

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 3.5 Reactor Instrumentation Applicability:

This specification applies to the instruments that provide information which must be available to the operator during reactor operation.

Objective:

The objective of this specification is to ensure that sufficient reliable information is presented to the operator to assure safe operation of the reactor.

Specification:

a. The reactor shall not be operated unless the following instrument channels are operable:

Minimum Numbers Operable Channel Mode I Mode II Mode III

1. Source Range Nuclear Instrument Channel I1() 1()

2. Reactor Pool Temperature 1 1 1 (1) Required for reactor startup only.

b. Sufficient instrumentation shall be provided to assure that the following limits are not exceeded during steady-state operation:

Parameter Limit

1. Primary Coolant System Pressure 110 psig (Max)

2. Anti-Siphon System Pressure 27 psig(') (Min)

3. Reactor Pool Temperature 120 OF( 2) (Max)

(1)

Not required for Mode III operation.

(2)

Reactor Pool Temperature limit is a maximum of 100 'F when in Mode III operation and (a) below 50 kW with the natural convection flange and reactor pressure cover removed or (b) with the reactor subcritical by a margin of at least 0.015 Ak/k.

c. A minimum of one decade of overlap shall exist between adjacent ranges of nuclear instrument channels.

A-27

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 3.5 Reactor Instrumentation - Continued

d. The reactor shall not be started up unless:

(1) The Source Range Channel is indicating a neutron count rate of at least 1 count per second and the Wide Range Monitor is indicating a power level greater than I watt, OR (2) The Source Range Channel is indicating a neutron count rate of at least 2 counts per second and is verified just prior to startup by a neutron test source or movement on the Source Range meter demonstrating that the channel is responding to neutrons.

Bases:

a. The Source Range Nuclear Instrument Channel provides a neutron monitor that is very sensitive to neutrons and thus provides improved indication of the low neutron flux levels present during a startup.

The reactor pool temperature instrument is required to ensure that pool temperature does not increase to a level which would jeopardize the ability to cool in-pool components.

b. The maximum primary coolant pressure of 110 psig assures that the system design pressure of 125 psig is not exceeded.

Maintaining the minimum anti-siphon system pressure ensures that the system will adequately perform its intended function (Ref. Section 6.3 of the SAR).

The reactor pool temperature limit provides an operating limit to assure the adequate cooling of the reactor fuel or pool components during all modes of operation.

c. Specification 3.5.c ensures that, during a startup, the reactor power level is continuously monitored over the entire range.

d. Specification 3.5.d provides for adequate neutron flux level monitoring to ensure that subcritical multiplication and criticality can be observed during a startup.

A-28

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 3.6 Emergency Electrical Power System Applicability:

This specification applies to the emergency electrical power system.

Obiective:

The objective of this specification is to assure emergency electrical power is available to vital equipment.

Specification:

a. The reactor shall not be operated unless the emergency electrical power system is operable.

Bases:

a. On a loss of normal electrical power, the emergency electrical power system will supply power to the containment ventilation isolation doors, personnel entry doors, facility ventilation exhaust fans, emergency lighting panel, and reactor instrumentation and control systems. Therefore, on a loss of normal electrical power, the emergency electrical power system is not required for protection of the integrity of the fuel elements. In the extremely unlikely event of a simultaneous loss of normal electrical power and fuel element failure, the operation of the emergency electrical power system would be required to provide for continuous containment isolation (Ref. Section 13.2.7 of the SAR).

A-29

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R- 103 3.7 Radiation Monitoring Systems and Airborne Effluents Applicability:

This specification applies to radiation monitoring information which must be available to the reactor operator during reactor operation and the release of gaseous and particulate activity from the facility ventilation exhaust stack.

Obiective:

The objective of this specification is to assure that sufficient radiation monitoring information is available to the reactor operator during reactor operations and exposure to the public resulting from the radioactivity released from the reactor facility to the unrestricted environment will not exceed the limits of 10 CFR 20.

a. The reactor shall not be operated unless the following radiation monitoring channels are operable:

Minimum Numbers Operable Channel Mode I Mode II Mode III

1. Reactor Bridge Radiation Monitor 1(1) 1(1) 1

2. Reactor Containment Building Exhaust 1 1 I Plenum Radiation Monitor

3. Off-Gas (Stack) Radiation Monitor 1(2) 1(2) 1(2)

(l) The trip setting may be temporarily set upscale during periods of maintenance and sample handling. During these periods, the radiation monitor indication will be closely observed.

(2) The off-gas (stack) radiation monitor may be placed out of service for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for calibration and maintenance. During this out-of-service time, no experimental or maintenance activities will be conducted which could likely result in the release of unknown quantities of airborne radioactivity.

b. The maximum discharge rate through the ventilation exhaust stack shall not exceed the following:

Max. Concentration Max. Controlled Type of Averaged Over Instantaneous Release Radioactivity One Year Concentration Particulates and halogens with AEC AEC half-lives greater than 8 days All other radioactive isotopes 350 AEC 3,500 AEC A-30

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 3.7 Radiation Monitoring Systems and Airborne Effluents - Continued AEC = Air Effluent Concentration as listed in Appendix B, Table 2, Column I of 10 CFR 20, "Standards for Protection Against Radiation."

Bases:

a. The radiation monitors provide information of an impending or existing danger from radiation so that corrective action can be initiated to prevent the spread of radioactivity to the surroundings and so that there will be sufficient time to evacuate the facility should it be necessary to do so.

b. Dispersion calculations based upon standard reference material and experiment data obtained at the reactor show that argon-41 concentrations under average conditions will be 0.008 of the AEC limits in the unrestricted area surrounding the reactor facility. Dilution factors under conservative conditions are in the range of 5 x 104 under both average and stable conditions at ground level from the facility building.

The normal short burst releases at the facility are five to ten seconds in duration and occur on an average of ten times per day five days per week. The short bursts affect the concentration by less than 1% when averaged over a one-day period.

It is concluded that these concentrations as specified will not constitute a hazard to the health and safety of the public.

A-31

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 3.8 Experiments Applicability:

This specification applies to all experiments which directly utilize neutrons or other radiation produced by the reactor. Radioactive sources shall meet the requirements for experiments.

Objective:

The objective of this specification is to prevent an accident which would jeopardize the safe operation of the reactor or would constitute a hazard to the safety of the facility staff and general public.

Specification:

a. Each fueled experiment shall be limited such that the total inventory of iodine-131 through iodine-135 in the experiment is not greater than 150 curies and the maximum strontium-90 inventory is no greater than 300 millicuries.

b. No experiments shall be placed in the reactor pressure vessel or water annulus surrounding the center test hole other than for reactor calibration.

c. Where the possibility exists that the failure of an experiment could release radioactive gases or aerosols into the containment building atmosphere, the experiment shall be limited to that amount of material such that the airborne concentration of radioactivity when averaged over a year will not exceed the limits of 10 CFR 20, Appendix B, Table 1. Exception: Fueled experiments (See Specification 3.8.a).

d. Explosive materials shall not be irradiated nor shall they be allowed to generate in any experiment in quantities over 25 milligrams of TNT-equivalent explosives.

e. Only movable experiments in the center test hole shall be removed or installed with the reactor operating. All other experiments in the center test hole shall be removed or installed only with the reactor shutdown. Secured experiments shall be rigidly held in place during reactor operation.

f. Experiments shall be designed and operated so that identifiable accidents such as a loss of primary coolant flow, loss of experiment cooling, etc., will not result in a release of fission products or radioactive materials from the experiment.

g. Experiments shall be designed such that a failure of an experiment will not lead to a direct failure of another experiment, a failure of a reactor fuel element, or to interfere with the action of the reactor control system or other operating components.

A-32

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 3.8 Experiments - Continued

h. Cooling shall be provided to prevent the surface temperature of a submerged irradiated experiment from exceeding the saturation temperature of the cooling medium.

i. Irradiation containers to be used in the reactor, in which a static pressure will exist or in which a pressure buildup is predicted, shall be designed and tested for a pressure exceeding the maximum expected pressure by at least a factor of two (2).

j. Corrosive materials shall be doubly encapsulated in corrosion-resistant containers to prevent interaction with reactor components or pool water.

k. Fluids utilized in loop experiments placed in the beamports shall be of types which will not chemically react in the event of leakage and shall be maintained at pressure and temperature conditions such that the integrity of the beam tube will not be impaired in the event of loop rupture.

1. The normal operating procedures shall include controls on the use or exclusion of corrosive, flammable, and toxic materials in experiments or in the reactor containment building. These procedural controls shall include a current list of those materials which shall not be used and the specific controls and procedures applicable to the use of corrosive, flammable, or toxic materials which are authorized.

m. Cryogenic liquids shall not be used in any experiment within the reactor pool.

n. The maximum temperature of a fueled experiment shall be restricted to at least a factor of two (2) below the melting temperature of any material in the experiment.

First-of-a-kind fueled experiments shall be instrumented to measure temperature.

o. Fueled experiments containing inventories of iodine-131 through iodine-135 greater than 1.5 curies or strontium-90 greater than 5 millicuries shall be in irradiation containers that satisfy the requirements of Specification 3.8.i or be vented to the facility ventilation exhaust stack through high efficiency particulate air (HEPA) and charcoal filters which are continuously monitored for an increase in radiation levels.

p. The absolute value of the reactivity worth of each secured removable experiment shall be limited to 0.006 Ak/k.

q. The absolute value of the reactivity worth of all experiments in the center test hole shall be limited to 0.006 Ak/k.

A-33

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R- 103 3.8 Experiments - Continued

r. Each movable experiment or the movable parts of any individual experiment shall have a maximum absolute reactivity worth of 0.001 Ak/k.

s. The absolute value of the reactivity worth of each unsecured experiment shall not exceed 0.0025 Ak/k.

t. The absolute value of the reactivity worth of all unsecured experiments which are in the reactor shall not exceed 0.006 Ak/k.

Bases:

a. Specification 3.8.a restricts the generation of hazardous materials to levels that can be handled safely and easily. Analysis of fueled experiments containing a greater inventory of fission products has not been completed, and therefore their use is not permitted. (Ref. Section 13.2.6 of the SAR).

b. Specification 3.8.b is intended to reduce the likelihood of accidental voiding in the reactor core or water annulus surrounding the center test hole by restricting materials which could generate or accumulate gases or vapors.

c. The limitation on experiment materials imposed by Specification 3.8.c assures that the limits of 10 CFR 20, Appendix B, are not exceeded in the event of an experiment failure.

d. Specification 3.8.d is intended to reduce the likelihood of damage to reactor or pool components resulting from the detonation of explosive materials (Ref.

Section 13.2.6 of the SAR).

e. Specification 3.8.e is intended to limit the experiments that can be moved in the center test hole while the reactor is operating to those that will not introduce reactivity transients more severe than one that can be controlled without initiating safety system action (Ref. Section 13.2.2 of the SAR).

f. - g. Specifications 3.8.f and 3.8.g provide guidance for experiment safety analysis to assure that anticipated transients will not result in radioactivity release and that experiments will not jeopardize the safe operation of the reactor.

h. Specification 3.8.h 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.

i. Specification 3.8.i is intended to reduce the likelihood of damage to the reactor and/or radioactivity releases from experiment failure.

A-34

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 3.8 Experiments - Continued

j. Specification 3.8.j provides assurance that no chemical reaction will take place to adversely affect the reactor or its components.

k. Specification 3.8.k provides assurance that the integrity of the beamports will be maintained for all loop-type experiments.

1. Specification 3.8.1 assures that corrosive materials which are chemically incompatible with reactor components, highly flammable materials, and toxic materials are adequately controlled and that this information is disseminated to all reactor users.

m. The extremely low temperatures of the cryogenic liquids present structural problems that enhance the potential of an experiment failure. Specification 3.8.m provides for the proper review of proposed experiments containing or using cryogenic materials.

n. Specification 3.8.n is intended to reduce the likelihood of damage to the reactor and/or radioactivity releases from experiment failure.

o. Specification 3.8.o restricts the generation of hazardous materials to levels that can be handled safely and easily. Analysis of fueled experiments containing a greater inventory of fission products has not been completed, and therefore their use is not permitted. (Ref. Section 13.2.6 of the SAR).

p. Specification 3.8.p provides assurance that any inadvertent insertion/removal or credible malfunction of a secured removable experiment would not introduce positive reactivity whose consequences would lead to radiation exposures in excess of the 10 CFR 20 limits. The step reactivity insertion is analyzed in Section 13.2.2 of the SAR.

q. The reactivity worth of experiments in the center test hole is limited by Specification 3.8.q such that the introduction of the maximum reactivity worth of all experiments would not result in damage to the fuel plates as analyzed in Section 13.2.2 of the SAR.

r. Specification 3.8.r provides assurance that the movement of movable experiments or movable parts of any experiment will not introduce reactivity transients more severe than one that can be controlled without initiating a reactor safety system action as analyzed in Section 13.2.2 of the SAR.

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UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 3.8 Experiments - Continued

s. Specification 3.8.s prevents the installation of an unsecured experiment which could introduce, as a positive step change, sufficient reactivity to place the reactor in a transient that would cause a violation of a safety limit as analyzed in Section 13.2.2 of the SAR.

t. Specification 3.8.t assures that the reactivity worth of all unsecured experiments shall not exceed the maximum value authorized for a single secured removable experiment.

A-36

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 3.9 Auxiliary Systems Applicability:

This specification applies to the reactor auxiliary systems.

Objective:

The objective of this specification is to provide for the operation of certain auxiliary systems and thus further protect the reactor fuel and personnel.

Specification:

a. The reactor shall not be operated unless the primary coolant make-up water system is operable and connected to a source of at least 2,000 gallons of primary grade water.

b. The reactor shall not be operated unless the emergency pool fill system is operable.

Bases:

a. Specification 3.9.a provides for an adequate supply of primary grade water for reactor plant make-up during all modes of operation.

b. The emergency pool fill system is capable of supplying water at approximately 1,000 gpm to the reactor pool. This supply assures that the water level in the pool will remain above the reflector in case a 6-inch beamport or a 6-inch pool coolant line is sheared (Ref. Sections 13.2.9.1 and 13.2.9.2 of the SAR).

A-37

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R- 103 4.0 SURVEILLANCE REQUIREMENTS General: Surveillance frequencies denoted herein are based on continuing operation of the reactor. Surveillance activities scheduled to occur during an operating cycle which cannot be performed with the reactor operating may be deferred to the end of that current reactor operating cycle. If the reactor is not operated for a reasonable time, a reactor system or measuring channel surveillance requirement may be waived during the associated period. Prior to reactor system or measuring channel operation, the surveillance shall be performed for each reactor system or measuring channel for which surveillance was waived. A reactor system or measuring channel shall not be considered operable until it is successfully tested. Surveillance intervals shall not exceed those defined by Definition 1.38. Discovery of noncompliance with any of the surveillance specifications listed in this Section shall limit reactor operations to that required to perform the surveillance.

4.1 Reactor Core Parameters Applicability:

This specification applies to the surveillance requirements of the reactor core parameters.

Obiective:

The objective of this specification is to verify reactor core parameters which are directly related to reactor safety.

Specification:

a. The reactor core excess reactivity above reference core condition shall be verified annually and following any significant core configuration and/or control blade change.

b. The shutdown margin shall be verified annually and following any significant core configuration and/or control blade change.

c. One out of every eight (8) fuel elements that have reached their end-of-life will be inspected for anomalies.

Bases:

a. - b. Annual measurements, coupled with measurements made after changes that can affect reactivity values, provide adequate assurance that core behavior resulting from configuration changes are adequately characterized.

c. The specified fuel element inspections along with the continuous primary coolant system fission product monitoring and the weekly radiochemical analysis of the primary coolant provide for the detection of anomalies resulting from reactor operation and reduces the possibility of fission product release to the primary A-38

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 4.1 Reactor Core Parameters - Continued coolant system. Inspecting the fuel elements at the end of their life has the added advantage of allowing for the decay of the fuel elements and, thus, reduction of exposure to personnel.

A-39

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 4.2 Reactor Control and Reactor Safety Systems Applicability:

This specification applies to the surveillance requirements on the reactor control and reactor safety systems.

Obiective:

The objective of this specification is to reasonably assure proper operation of the reactor control system and the reactor safety system instrument channels.

Specification:

a. The drop time of each of the four shim blades shall be measured at quarterly intervals.

b. A different one of the four shim blades shall be inspected each six months so that every blade is inspected every two years. The reactor shall not be operated with a control blade that exhibits abnormal swelling or abnormalities that affect performance.

c. Above 100 kilowatts, the distance between the highest and lowest shim blade shall be verified within a shift.

d. The withdrawal and insertion speeds of the regulating blade shall be verified on an annual basis.

e. The withdrawal and insertion speeds of each shim blade shall be verified on an annual basis.

f. The rod run-in functions required by Specification 3.2.f shall be channel calibrated on a semiannual basis.

g. The reactor safety system shall be channel tested before each reactor startup involving a refueling, a shutdown greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or quarterly.

h. The reactor safety system instrument channels listed in Specification 3.2.g shall be channel calibrated on a semiannual basis.

i. The reactor control interlocks listed in Specification 3.2.h shall be channel calibrated on a semiannual basis.

j. A thermal power verification of power range indication, using coolant flows and differential temperatures, shall be performed weekly when the reactor is operating above 2 MW.

A-40

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 4.2 Reactor Control and Reactor Safety Systems - Continued Bases:

a. Measurement of the drop time of each of the four shim blades is normally made quarterly to demonstrate that the blades are capable of performing properly. In over 40 years of operation, to date, the shim blades have never failed to meet Specification 3.2.c.

b. Periodic inspection of the shim blades provides detection of singular blade abnormalities and any potential generic blade design deficiencies. Specification 4.2.b further assures that the reactor will not be operated using shim blades with suspected generic design deficiencies.

c. Specification 4.2.c assures that shim blade heights will be verified within a shift.

d. - e. The drive mechanisms for the regulating and shim blades are constant speed mechanical devices and withdrawal and insertion speeds should not vary except as a result of mechanical wear. The surveillance is chosen to provide a significant margin over expected failure or wear rates of these mechanical devices.

f. - i. Experience has shown that the identified frequencies will ensure performance and operability for each of these systems or components (NUREG-1537 and ANSI/ANS-15.1-2007).

j. Thermal power verification will ensure that indicated reactor power level is correct. Because of the small primary coolant differential temperature at 10 MW (about 17 'F), these verifications will not be performed below 2 MW.

A-41

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R- 103 4.3 Reactor Coolant Systems Applicability:

This specification applies to the surveillance requirements on the reactor coolant systems.

Objective:

The objective of this specification is to reasonably assure proper operation of the reactor coolant systems.

Specification:

a. The following components or systems shall be tested for operability at monthly intervals except during extended shutdown periods when the valves shall be tested prior to reactor operation:

(1) Anti-siphon system; (2) Primary coolant isolation valves V507A/B; and (3) In-pool convective cooling system.

b. A primary coolant sample shall be taken during each week of reactor operation and a radiochemical analysis performed to determine the concentration of iodine-131.

c. A pool coolant sample shall be taken monthly and a radiochemical analysis performed to determine gross radioactivity.

d. A secondary coolant sample shall be taken quarterly and a radiochemical analysis performed to determine gross radioactivity.

e. The conductivity and pH of the water in the primary coolant system shall be measured on a monthly basis.

f. The conductivity and pH of the water in the pool coolant system shall be measured on a monthly basis.

g. The primary coolant system relief valves shall be tested for operability at two-year intervals, with at least one of the valves tested on an ammal basis.

Bases:

a. The past 40 years of operation of the anti-siphon system, primary coolant isolation valves and in-pool convective cooling system has shown that monthly testing is adequate to provide assurance of continued operability.

b. The weekly radiochemical analysis will provide assurance that a fuel element leak will be discovered so that corrective action can be taken to prevent the A-42

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186., License R-103 4.3 Reactor Coolant Systems - Continued release of fission products. Specification 4.2.b establishes the frequency of verification of compliance with Specification 3.3.c.

c. - f. Experience has shown that the frequency of measurements on the reactor coolant systems for gross radioactivity, conductivity and pH adequately maintain the water quality at such a level to minimize corrosion and maintain safety.

g. Satisfactory performance of both relief valves during the testing program over the past 40 years has demonstrated the reliability of the valves and the assurance of operability gained by the testing frequency outlined in Specification 4.3.g.

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UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License 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 assure proper operation of the containment system.

Specification:

a. The reactor containment building leakage rate shall be measured annually, plus or minus four (4) months. No special maintenance shall be performed just prior to the test.

b. The containment actuation (reactor isolation) system, including each of its radiation monitors, shall be tested for operability at monthly intervals.

Bases:

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 special maintenance will be performed prior to the test so that the results demonstrate the historic integrity of the containment structure.

b. The reliability of the containment actuation (reactor isolation) system has proven that monthly verification of its proper operation is sufficient to assure operability.

A-44

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 4.5 Reactor Instrumentation Applicability:

This specification applies to the surveillance requirements of the reactor instrumentation systems.

Obiective:

The objective of this specification is to reasonably assure proper operation of the reactor instrumentation systems.

Specification:

a. The instrument channels required by Specification 3.5.a shall be channel calibrated on a semiannual basis.

b. The instrumentation required to monitor the parameters required by Specification 3.5.b shall be channel calibrated on a semiannual basis.

c. All nuclear instrumentation channels shall be channel-tested before each reactor startup. This test shall not be required prior to a restart within two (2) hours following a normal reactor shutdown or an unplanned scram where the cause of the scram is readily determined not to involve an unsafe condition or a failure of one or more nuclear instrumentation channels.

Bases:

a. - b. Semiannual calibration of the instrument channels and instrumentation will assure that long-term drift of the channels and instrumentation will be corrected.

c. The nuclear instrumentation channel test will assure that the channels are operable.

A-45

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 4.6 Emergency Electrical Power System Applicability:

This specification applies to the surveillance requirements of the emergency electrical power system.

Obiective:

The objective of this specification is to reasonably assure proper operation of the emergency electrical power system.

Specification:

a. The operability of the emergency power generator shall be verified on a weekly basis.

b. The ability of the emergency power generator to assume the emergency electrical loads shall be verified on a semiannual basis.

Bases:

a. The emergency power generator tests provide assurance that the generator is operable.

b. The semiannual electrical load test has proven satisfactory in providing reasonable assurance that the emergency power generator electrical control and distribution system will remain operable.

A-46

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R- 103 4.7 Radiation Monitoring Systems and Airborne Effluents Applicability:

This specification applies to the surveillance requirements of the radiation monitoring instrumentation.

Obiective:

The objective of this specification is to reasonably assure proper operation of the radiation monitoring instrumentation.

Specification:

a. Radiation monitoring instrumentation required by Specification 3.7.a shall be verified operable by monthly radiation source checks or channel tests.

b. Radiation monitoring instrumentation required by Specification 3.7.a shall be channel calibrated on a semiannual basis.

Bases:

a. Experience has shown that monthly verification of operability of the radiation monitoring instrumentation is adequate assurance of proper operation over a long time period.

b. Semiannual channel calibration of the radiation monitoring instrumentation will assure that long-term drift of the channels will be corrected.

A-47

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 4.8 Experiments Applicability:

This specification applies to the surveillance requirements of experiments installed in the reactor or its experimental facilities.

Objective:

The objective of this specification is to prevent the conduct of experiments which may damage the reactor or release excessive amounts of radioactive materials as a result of experiment failure.

Specification:

a. The criteria of Specification 3.8 shall be evaluated prior to inserting an experiment in the reactor or its experimental facilities.

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

Bases:

a. - b. Experience has shown that experiments which are reviewed by the staff and the Reactor Advisory Committee can be conducted without endangering the safety of the reactor or exceeding the limits specified in the Technical Specifications.

A-48

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R- 103 4.9 Auxiliary Systems Applicability:

This specification applies to the surveillance requirements of the reactor auxiliary systems.

Objective:

The objective of this specification is to reasonably assure proper operation of the auxiliary systems.

Specification:

a. The operability of the primary coolant make-up water system shall be tested on a semiannual basis.

b. The operability of the emergency pool fill system shall be tested on a semiannual basis.

Bases:

a. Specification 4.9.a assures that an adequate supply of primary grade water is available for make-up during all modes of operation.

b. The University of Missouri-Columbia water supply system provides a virtually unlimited source of raw water for the emergency pool fill system. Water supply is maintained at a high pressure by automatically-controlled pumping stations.

The above test, in light of the reliability of the emergency pool fill system, provides assurance that Specification 3.9.b is satisfied.

A-49

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R- 103 5.0 DESIGN FEATURES General: Major alterations to safety-related components or equipment shall not be made prior to appropriate safety reviews.

5.1 Site Description Applicability:

This specification applies to the site of the MURR facility.

Objective:

The objective of this specification is to identify the location of the MURR facility.

Specification:

a. The MURR facility is situated on a 7.5-acre lot in the central portion of the University Research Park, an 84-acre tract of land approximately one mile southwest of the University of Missouri at Columbia's main campus. This campus is located in the southern portion of Columbia, the county seat and largest city in Boone County, Missouri.

Approximate distances to the University property lines from the reactor facility are 2,400 feet (732 m) to the north, 4,800 feet (1,463 m) to the east, 2,400 feet (732 mn) to the south, and 3,600 feet (1,097 m) to the west.

The restricted, or licensed, area is that area inside the fenced 7.5 acre lot surrounding the MURR facility itself. Within the restricted area the Reactor Facility Director has direct authority and control over all activities, normal and emergency. There are pre-established evacuation routes and procedures known to personnel frequenting this area.

For emergency planning purposes, the site boundaries consist of the following:

Stadium Boulevard; Providence Road (Route K)1; the MU Recreational Trail; and the MKT Nature and Fitness Trail. The area within these boundaries is owned and controlled by MU and may be frequented by people unacquainted with the operation of the reactor. The Reactor Facility Director has authority to initiate emergency actions in this area, if required.

'Providence Road crosses MU property separating the University Research Park from another MU-owned tract of land lying to the east. The road runs north and south with the closest point of approach being approximately 400 meters east of the reactor facility. MU has the authority to determine all activities including the exclusion or removal of personnel and property and to temporarily secure the flow of traffic on this road during an emergency.

A-50

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 5.1 Site Description - Continued Bases:

a. The MURR facility site location and description are strictly defined in Chapter 2 of the SAR. The location of the MURR facility and University Research Park is owned and operated by the University of Missouri. Based on the information provided in Chapter 2, and throughout the SAR, the site is well suited for the location of the facility when considering the relatively benign operating characteristics of the reactor.

A-51

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R- 103 5.2 Reactor Coolant Systems Applicability:

This specification applies to the reactor coolant systems.

Obiective:

The objective of this specification is to assure proper coolant for safe operation.

Specification:

The MURR utilizes three reactor coolant systems: primary, pool, and secondary. The following design features apply to these coolant systems:

a. The reactor coolant systems shall consist of not less than a reactor pressure vessel, a primary pressurizer, two primary coolant circulation pumps, two primary coolant heat exchangers, two pool coolant circulation pumps, one pool coolant heat exchanger, and one pool water hold-up tank, plus all associated piping and valves.

b. The secondary coolant system shall be capable of continuous discharge of heat generated at the operating power of the reactor.

c. The circulation pumps and heat exchangers of the primary coolant system shall constitute two parallel systems separately instrumented to permit safe operation at five megawatts on either system or ten megawatts with both systems operating simultaneously.

d. The pool coolant circulation pumps shall be instrumented and connected so as to permit safe operation at five or ten megawatts on either pump or both pumps operating simultaneously.

e. All major components of the reactor coolant systems in contact with pool or primary water shall be constructed principally of aluminum alloys or stainless steel.

f. The pool and primary coolant systems shall have a water clean-up system.

g. The pool and primary coolant piping shall have isolation valves between the reactor and mechanical equipment room.

h. The primary coolant system shall have two anti-siphon isolation valves.

i. The reactor shall have a natural convection coolant flow path for Mode III operation except for operation with the reactor subcritical by a margin of at least 0.015 Ak/k.

A-52

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 5.2 Reactor Coolant Systems - Continued

j. The reactor shall have a decay heat removal system.

k. The primary coolant system shall contain at least two operable pressure relief valves.

Exceptions:

a. The reactor may be operated in Mode II with any component removed from the shutdown leg of the system for emergency repairs.

b. Some materials in off-the-shelf commercial components may be excepted from Specification 5.2.e.

Bases:

a. - k. The reactor coolant systems are described and analyzed in Section 5 of the SAR.

The reactor can be safely operated at 10 MW with the coolant systems as described.

Specification 5.2.a as excepted, permits reactor operation at 50% of full power in the event of a major component failure in which repairs cannot be accomplished in a reasonable period of time. The reactor was designed and has extensive safe operating history for operation at 50% of 10 MW cooling capacity. In this event, the shutdown system shall be secured in a manner such as to assure system integrity.

Specification 5.2.e assures strength and corrosion resistance of the coolantsystem components and excepts some components in the instrumentation of the system which are not commercially available in the materials specified. The size of these components is such that a failure would not result in a hazard to safe operation of the reactor.

A-53

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 5.3 Reactor Core and Fuel Applicability:

This specification applies to the reactor core and fuel elements.

Objective:

The objective of this specification is to specify the general reactor core configuration and to assure that the fuel elements are of a type designed for use in the reactor.

Specification:

The following design features apply to the reactor core and fuel:

a. The average reactor core temperature coefficient of reactivity shall be more negative than -6.0 x I0-5 Ak/k/!F.

b. The average reactor core void coefficient of reactivity shall be more negative than

-2.0 x 10-3 Ak/k/% void.

c. The regulating blade total reactivity worth shall be a maximum of 6.0 x 10-3 Ak/k.

d. Each reactor fuel element shall contain 24 fuel-bearing plates with a nominal active length of 24 inches and a nominal plate thickness of 0.050 inches. The nominal distance between the fuel plates shall be 0.080 inches. Plate nominal cladding thickness shall be 0.015 inches.

e. The fuel material shall be aluminide dispersion UAlx nominally enriched to 93%

in the isotope uranium-235.

f. Each fuel element shall have a maximum uranium-235 loading of 775 grams.

g. The reactor fuel shall be contained in the aluminum pressure vessel, in-pool fuel storage locations, or the fuel storage vault.

h. The reactor shall have a beryllium and graphite reflector.

i. The reactor shall have five control blades between the pressure vessel and beryllium reflector. Four blades shall be for coarse control (shim blades) and one for fine control (regulating blade) of reactor power.

j. The reactor shall have the following experimental facilities:

I. Six beam tubes which penetrate the graphite reflector;

2. A center test hole located in the flux trap;

3. A portion of the graphite reflector;

4. A bulk pool consisting of the water region above and outside the graphite reflector; and A-54

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R- 103 5.3 Reactor Core and Fuel - Continued

5. A thermal column.

Bases:

a. Specification 5.3.a limits one of the parameters which assures that core damage will not occur following any credible step reactivity insertion as analyzed in Section 13.2.2 of the SAR.

b. The average core void coefficient of reactivity also limits the step reactivity insertion accident as analyzed in Section 13.2.2 of the SAR.

c. The regulating blade total reactivity worth is limited by Specification 5.3.c such that any condition resulting in the step insertion of the maximum worth of 6 x 10-3 Ak/k will not result in fuel plate damage.

d.- f. The MURR fuel elements are one of a configuration (aluminide UAlx dispersion fuel system) successfully and extensively used for many years in test and research reactors. Specifications 5.3.d, 5.3.e and 5.3.f require fuel content and dimensions of the fuel elements to be in accordance with the design and fabrication specifications (Ref. Section 4.2.1 of the SAR).

g. Specification 5.3.g assures that the reactor fuel is properly positioned in the pressure vessel during operation (Ref. Section 4.2.5 of the SAR).

h. Specification 5.3.h assures proper neutron reflection as required by design (Ref Section 4.2.3 of the SAR).

i. Specification 5.3.i assures reactivity of the reactor is properly controlled as required by design (Ref. Section 4.2.2 of the SAR).

j. Specification 5.3.j assures that the reactor consists of the experimental facilities as required by design (Ref. Chapter 10 of the SAR).

A-55

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 5.4 Fuel Storage Applicability:

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

Objective:

The objective of this specification is to assure that fuel which is stored shall not become critical and will not reach an unsafe temperature.

Specification:

The following design features apply to fuel storage:

a. All fuel elements or fueled devices outside the reactor core shall be stored in a geometrical array where the value of Keff is less than 0.9 under all conditions of moderation.

b. Irradiated fuel elements shall be stored in an array which will permit sufficient natural convection cooling such that the temperature of the fuel element or fueled device will not exceed its design values.

Bases:

a. - b. The limits imposed by Specifications 5.4.a and 5.4.b are conservative and assure safe fuel storage.

A-56

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 5.5 Reactor Containment Building Applicability:

This specification applies to the building in which the reactor is located.

Objective:

The objective of this specification is to assure adequate restriction to the accidental release of radioactivity to the environment.

Specification:

The reactor containment building is a five-level, poured-concrete structure with 12-inch thick reinforced exterior walls configured to form the shape of a cube, with each side being approximately 60 feet long. Below grade within the containment structure is a space extending to the north that is 15 feet high by 37 feet deep by 40 feet wide. The following design features apply to the MURR reactor containment building:

a. The reactor and fuel storage facilities shall be enclosed in a containment building with a free volume of at least 225,000 cubic feet.

b. Whenever reactor containment integrity, as defined by Specification 3.4.a, is required, containment building ventilation exhaust shall be discharged at a minimum of 55 feet above containment building grade level.

c. The containment building leakage rate shall not exceed 16.3 cubic feet per minute at STP with an overpressure of one pound per square inch gauge or 10% of the 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 make-up flow, pressure decay, or reference volume techniques.

d. The containment building shall have a secured fuel storage room with the key or combination under control of the Reactor Manager.

Bases:

a. No credible accident scenario has been identified which can result in a significant overpressure condition in the reactor containment building. However, Specification 5.5.a assures that a sufficient free volume exists to prevent any pressure buildup in the containment building (Ref. Section 6.2.2.2 of the SAR).

b. Specification 5.5.b assures a sufficient stack height for more than adequate atmospheric dispersion.

c. Specification 5.5.c assures that the containment building will have sufficient integrity to limit the leakage of contained potentially radioactive air in the event A-57

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R- 103 5.5 Reactor Containment Building - Continued of any reactor accident to ensure exposures are maintained below the limits of 10 CFR 20 (Ref. Sections 6.2.10 and 13.2.1 of the SAR).

d. Specification 5.5.d assures safe and secure storage of fresh fuel.

A-58

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 5.6 Emergency Electrical Power System Applicability:

This specification applies to the facility emergency electrical power system.

Obiective:

The objective of this specification is to assure adequate emergency electrical power in the event of normal electrical power failure.

Specification:

The following design feature applies to the emergency electrical power system:

a. The MURR shall have an emergency power generator capable of providing emergency electrical power to the emergency lighting system, the facility ventilation exhaust system, reactor instrumentation, and the personnel air lock doors.

Bases:

a. The emergency electrical power system is described in Section 8.2 of the SAR.

Specification 5.6.a assures that a system exists to provide the necessary electrical power to monitor the reactor systems and assure personnel safety in the event of a normal power failure to the reactor facility.

A-59

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 6.0 ADMINISTRATIVE CONTROLS 6.1 Organization

a. The organizational structure of the University of Missouri-Columbia (MU) relating to the University of Missouri Research Reactor (MURR) shall be as shown in Figure 6.0.

b. The following positions shall have direct responsibility in implementing the Technical Specifications as designated throughout this document:

(1) Reactor Facility Director: Responsible for establishing the policies that minimize radiation exposure to the public and to radiation workers, and that ensures that the requirements of the license and Technical Specifications are met.

(2) Reactor Manager: To safeguard the public and facility personnel from undue radiation exposure, the Reactor Manager is responsible for:

i. Compliance with Technical Specifications and license requirements regarding reactor operation, maintenance and surveillance; and ii. Oversight of the experiment review process.

(3) Reactor Health Physics Manager: To safeguard the public and facility personnel from undue radiation exposure, the Reactor Health Physics Manager is responsible for:

i. Compliance with Technical Specifications and license requirements regarding radiation safety, byproduct material handling and the shipment of byproduct material; and ii. Implementation of the Radiation Protection Program.

c. At a minimum during reactor operation, there shall be two facility staff personnel at the facility. One of these individuals shall be a Reactor Operator or a Senior Reactor Operator licensed pursuant to 10 CFR 55. The other individual must be knowledgeable of the facility.

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

(1) Management personnel.

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UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 6.1 Organization - Continued (2) Health Physics personnel.

(3) Reactor Operations personnel.

e. A Senior Reactor Operator licensed pursuant to 10 CFR 55 shall be present at the facility or readily available on call at all times during operation, and shall be present at the facility during all startups and approaches to power, recovery from an unplanned or unscheduled shutdown or non-emergency power reduction, and refueling activities.

6.2 Review and Audit

a. A Reactor Advisory Committee (RAC) shall provide independent oversight in matters pertaining to the safe operation of the reactor and with regard to planned research activities and use of the facility building and equipment. The RAC shall review:

(1) Proposed changes to the MURR equipment, systems, or procedures when such changes have safety significance, or involve an amendment to the facility operating license, a change in the Technical Specifications incorporated in the license, or a question pursuant to 10 CFR 50.59.

Changes to procedures that do not change their original intent may be made without prior RAC review if approved by the TS designated manager, either the Reactor Health Physics Manager or Reactor Manager, or a designated alternate who is a member of Health Physics or a Senior Reactor Operator, respectively. All such changes to the procedures shall be documented and subsequently reviewed by the RAC; (2) Proposed experiments significantly different from any previously reviewed or which involve a question pursuant to 10 CFR 50.59; (3) The circumstances of reportable occurrences and violations of the Technical Specifications or license and the measures taken to prevent a recurrence; (4) Violations of internal procedures or operating abnormalities having safety significance; and (5) Reports from audits required by the Technical Specifications.

b. The RAC may appoint subcommittees consisting of knowledgeable members of the public, students, faculty, and staff of MU when it deems it necessary in order to effectively discharge its primary responsibilities. When subcommittees are A-61

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 6.2 Review and Audit - Continued appointed, these are to consist of no less than three members with no more than one student appointed to each committee. The subcommittees may be authorized to act on behalf of the parent committee.

The RAC and its subcommittees are to maintain minutes of meetings in which the items considered and the committees' recommendations are recorded.

Independent actions of the subcommittees are to be reviewed by the parent committee at the next regular meeting. A quorum of the committee or the subcommittees consisting of at least fifty percent of the appointed members must be present at any meeting to conduct the business of the committee or subcommittee. The RAC shall meet at least quarterly.

A meeting of a subcommittee shall not be deemed to satisfy the requirement of the parent committee to meet at least once during each calendar quarter.

c. Any additions, modifications or maintenance to the systems described in these Specifications shall be made and tested in accordance with the specifications to which the system was originally designed and fabricated or to specifications approved by the U.S. Nuclear Regulatory Commission (NRC).

d. Following a favorable review by the NRC, the RAC, or the Reactor Facility Management, as appropriate, and prior to conducting any experiment, the Reactor Manager shall sign an authorizing form which contains the basis for the favorable review.

e. Audits:

(1) Audits of the following functions shall be conducted by an individual or group without immediate responsibility in the area to be audited:

i. Facility Operations, for conformance to the Technical Specifications and license conditions, at least annually; ii. Operator Requalification Program, for compliance with the approved program, at least every two years; and iii. Corrective Action items associated with reactor safety, at least annually.

(2) Audit findings which affect reactor safety shall be immediately reported to the Reactor Facility Director.

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UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 6.3 Radiation Safety

a. The Reactor Health Physics Manager shall be responsible for the implementation of the Radiation Protection Program. The requirements of the Radiation Protection Program are established in 10 CFR 20. The program should use the guidelines of American National Standard "Radiation Protection at Research Reactor Facilities," ANSI/ANS- 15.1 1-1993 (R2004).

6.4 Procedures

a. Written procedures shall be in effect for operation of the reactor, including the following:

(I) Startup, operation, and shutdown of the reactor; (2) Fuel loading, unloading and movement within the reactor; (3) Maintenance of major components of systems that could have an effect on reactor safety; (4) Surveillance checks, calibrations and inspections that may affect reactor safety; (5) Administrative controls for operations and maintenance and for the conduct of irradiations and experiments that could affect reactor safety or core reactivity; and (6) Implementation of the Emergency Plan.

b. Written procedures shall be in effect for radiological control, and the preparation for shipping and the shipping of byproduct material produced under the facility operating license.

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.

d. Deviations from procedures required by this Specification may be enacted by a Senior Reactor Operator or member of Health Physics, as applicable. Such deviations shall be documented and reported within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or the next working day to the Reactor Manager or Reactor Health Physics Manager or designated alternate.

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UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 6.5 Experiment Review and Approval Approved experiments shall be carried out in accordance with established and approved procedures. Procedures related to experiment review and approval shall include the following:

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

b. 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 NRC. Prompt reporting of the violation shall be made by MU, by telephone or email, to the NRC Operations Center no later than the following working day; (3) A detailed follow-up report shall be prepared. The report shall include the following:

i. Applicable circumstances leading to the violation including, when known, the causes and contributing factors; ii. Date and approximate time of the occurrence; iii. Effect of the violation upon the reactor and associated systems; iv. Effect of the violation on the health and safety of the facility staff and general public; and

v. Corrective actions to prevent recurrence.

(4) The follow-up report will be submitted within fourteen (14) days to the NRC Document Control Desk.

b. Release of Radioactivity - Should a release of radioactivity greater than the allowable limits occur from the reactor facility boundary, the following actions shall be taken:

(1) Reactor conditions shall be returned to normal or the reactor shall be shut down; A-64

UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 6.6 Reportable Events and Required Actions - Continued (2) The release of radioactivity shall be promptly reported to the NRC.

Prompt reporting of the violation shall be made by MU, by telephone or email, to the NRC Operations Center no later than the following working day; (3) If it is necessary to shut down the reactor to correct the occurrence, operations shall not be resumed until authorized by the Reactor Manager; and (4) A detailed follow-up report shall be prepared. The follow-up report will be submitted within fourteen (14) days to the NRC Document Control Desk.

c. Other Reportable Occurrences - In the event of an Abnormal Occurrence, as defined by Definition 1.1, the following actions shall be taken:

(Note: Where components or systems are provided in addition to those required by these Technical Specifications, the failure of the extra components or systems is not considered reportable provided that the minimum numbers of components or systems specified or required perform their intended reactor safety function.)

(1) The Abnormal Occurrence shall be promptly reported to the NRC.

Prompt reporting of the violation shall be made by MU, by telephone or email, to the NRC Operations Center no later than the following working day; (2) A detailed follow-up report shall be prepared. The follow-up report will be submitted within fourteen (14) days to the NRC Document Control Desk; and (3) The reactor shall be shut down or placed in a safe condition and return to normal reactor operations will not be allowed until authorized by the Reactor Manager.

d. Other Reports - A written report shall be submitted to the NRC Document Control Desk within thirty (30) days of:

(1) Any significant change(s) in the transient or accident analyses as described in the SAR; and (2) Permanent changes in the facility organization involving the Office of the Provost or the Director's Office.

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UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R-103 6.6 Reportable Events and Required Actions - Continued

e. Annual Report - An annual operating report shall be submitted to the NRC within sixty (60) days following the end of each calendar year. The report shall include the following information for the preceding year:

(1) A brief narrative summary of (a) operating experience (including operations designed to measure reactor characteristics), (b) changes in the reactor facility design, performance characteristics, and operating procedures related to reactor safety occurring during the reporting period, and (c) results of surveillance tests and inspections; (2) A tabulation showing the energy generated by the reactor (in megawatt-days);

(3) The number of emergency shutdowns and inadvertent scrams, including the reasons therefore and corrective action, if any, taken; (4) Discussion of the major maintenance operations performed during the period, including the effects, if any, on the safe operation of the reactor; (5) A summary of each modification to the reactor facility or change to the procedures, tests and experiments carried out under the conditions of 10 CFR 50.59; (6) A summary of the nature and amount of radioactive effluents released or discharged to the environs beyond the effective control of the licensee as measured at or prior to the point of such release or discharge; (7) A description of any environmental surveys performed outside the reactor facility; and (8) A summary of radiation exposures received by facility staff, experimenters, and visitors, including the dates and time of significant exposure, and a brief summary of the results of radiation and contamination surveys performed within the facility.

6.7 Records Records of the following activities shall be maintained and retained for the periods specified below. The records may be in the form of logs, data sheets, or other suitable forms or documents. The required information may be contained in single or multiple records, or a combination thereof.

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UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R- 103 6.7 Records - Continued

a. Lifetime Records - The following records are to be retained for the lifetime of the reactor facility: (Note: Applicable annual reports, if they contain all of the required information, may be used as records in this section.)

(1) Gaseous and liquid radioactive effluents released to the environs; (2) Off-site environmental-monitoring surveys required by the Technical Specifications; (3) Radiation exposure for all monitored personnel; and (4) Updated drawings of the reactor facility.

b. Five .Year Records - The following records are to be maintained for a period of at least five years or for the life of the component involved, whichever is shorter:

(1) Normal reactor facility operation (but not including supporting documents such as checklists, log sheets, etc. which shall be maintained for a period of at least one year);

(2) Principal maintenance operations; (3) Reportable occurrences; (4) Surveillance activities required by the Technical Specifications; (5) Reactor facility radiation and contamination surveys required by applicable regulations; (6) Experiments performed with the reactor; (7) Fuel inventories, receipts and shipments; (8) Approved changes to operating procedures; and (9) Records of meetings and audit reports of the review and audit group.

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UNIVERSITY OF MISSOURI RESEARCH REACTOR TECHNICAL SPECIFICATIONS Docket 50-186, License R- 103 Board of Curators of the University of Missouri I

I I

I Reporting Lines

.... Communication Lines FIGURE 6.0 UNIVERSITY OF MISSOURI RESEARCH REACTOR (MURR)

ORGANIZATION A-68

ML14030A132

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