university of missouri · 2014-02-07 · university of missouri research reactor center january 27,...

UNIVERSITY of MISSOURIRESEARCH REACTOR CENTER

January 27, 2014

U.S. Nuclear Regulatory CommissionAttention: Document Control DeskMail Station P1-37Washington, DC 20555-0001

Reference: Docket 50-186University of Missouri-Columbia Research ReactorAmended Facility License R-103

Enclosed you will find the University of Missouri-Columbia Research Reactor's (MURR) revisedTechnical 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, assupplemented.

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

Sincerely,

Ralph A. Butler, P.E.Director

RAB/jlb

Enclosures

-.. , u,, MARGEE P. STOUTNOTAR)-.V My Comrrrission Eores

":* W March24,2016Mwontgomery Coury

Commission #12511436

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

UNIVERSITY of MISSOURIRESEARCH REACTOR CENTER

January 27, 2014

U.S. Nuclear Regulatory CommissionAttention: Document Control DeskMail Station P1-37Washington, DC 20555-0001

REFERENCE: Docket 50-186University of Missouri - Columbia Research ReactorAmended 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: LicenseRenewal, Safety Analysis Report, Complex Questions (TAC No. MD3034)," datedMay 6, 2010, and the "University of Missouri at Columbia - Request for AdditionalInformation 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 arequest to the U.S. Nuclear Regulatory Commission (NRC) to renew Amended Facility OperatingLicense R-103.

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

On June 1, 2010, the NRC requested additional information and clarification regarding the renewalrequest in the form of one hundred and sixty-seven (167) 45-Day Response Questions. By letter datedJuly 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 theremaining 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 the45-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 theremaining 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 ResponseQuestions and two (2) of the remaining Complex Questions.

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

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

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

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

On May 27, 2011, via email, MIURR requested additional time to respond to the remaining the remaining45-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 andComplex Questions.

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

By letter dated January 6, 2012, MURR responded to four (4) of the remaining 45-Day Response andComplex 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 theremaining 45-Day Response and Complex Questions. By letter dated January 26, 2012, the NRC grantedthe request.

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

By letter dated June 28, 2012, MURR responded to the remaining six (6) 45-Day Response and ComplexQuestions. 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

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

If there are questions regarding this submittal, please contact me at (573) 882-5319 [email protected]. I declare under penalty of perjury that the foregoing is true and corTect.

Sincerely,

John L. FruitsReactor Manager

ENDORSEMENT:Reviewed and Approved,

Ralph A. Butler, P.E.Director

MARGEE P. STOUTMY Commision Em*es

MArch 24, 2018MonWgome y Cut

Commi~ssio #12511436

Enclosed:

Attachment 1:

Attachment 2:

Summary of Changes Made to the MURR Technical Specifications (From versionsubmitted 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 CommitteeReactor Safety SubcommitteeDr. Robert Hall, Interim Vice Chancellor for ResearchMr. Geoffrey Wertz, U.S. NRCMr. Alexander Adams, U.S. NRCMr. Johnny Eads, U.S. NRC

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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 Calibration2. 1.5, Channel Check3. .15, Operating4. 1.17, Reactivity Worth of an Experiment5. 1.26, Reference Core Condition6. 1.29, Research Reactor7. 1.30, Research Reactor Facility8. 1.31, Rod Run-In System9. 1.36, Should, Shall, and May10. 1.38, Surveillance IntervalsII. 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, InstrumentChannel 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 ofANSI/ANS-I15.1-2007.

d. Definition 1. 17, Reactor Containment Integrity was incorporated into LCO 3.4, ReactorContainment Building as Specification 3.4.a because it more closely follows the guidance ofANSI/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.b3. 3.1.c was split- it is now 5.3.c and 3.2.d

4. 3.1.d is now 3.2.e5. 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.p8. 3.1.h is now 3.8.q9. 3.1.i is now 3.8.r

10. 3.1 .j is now 3.8.s11. 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.b3. 3.2.c remains 3.2.c

m. Section 3.3, Reactor Safety System:

1. 3.3.a is now 3.2.gn. Section 3.4, Reactor Instrumentation:

1. First two Specifications of 3.4.a are now 3.5.a2. 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.f5. 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.b2. 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

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2. 3.8.b is now 3.l.e3. 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.b3. 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.bI. Section 4.3, Control Blades:

I. 4.3.a is now 4.2.a

2. 4.3.b is now 4.2.bm. 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.a3. 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

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2. 4.6.b is now 4.6.a3. 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 changesg. Section 5.2, Reactor Containment Building:

I. 5.2.a is now 5.5.a2. 5.2.b is now 5.5.b3. 5.2.c is now 5.5.c4. 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.d2. 5.4.b is now 5.3.e3. 5.4.c is now 5.3.f4. 5.4.d is now 5.3.g5. 5.4.e is now 5.3.hi6. 5.4.fis now 5.3.i7. 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 Controlsa. 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.a2. 6.l.b remains 6.1.b3. 6.1.c remains 6.1.c4. 6.1.d changed to 6.1.e

g. Section 6.2, Review and Audit:1. 6.2.a remains 6.2.a

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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.c4. 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.c4. 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.d5. 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.bb. 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

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6. 3.2.f was old 3.4.c7. 3.2.g was old 3.3.a8. 3.2.h - New Specificationc. Section 3.3, Reactor Coolant Systems:1. 3.3.a was old 3.9.a

2, 3.3.b was old 3.9.b3. 3.3.c was old 3.9,c4, 3.3.d - New Specification5. 3.3.e - New Specification6, 3.3.f-New Specification7. 3.3.g- New Specification8, 3.3.h - New Specification9. 3.3.i was old 3.9.dd. Section 3.4, Reactor Containment Building:1. 3.4.a was previous Definition 1. 172. 3.4.b was old 3.5.a

3. 3.4.c was old 3.5.be. Section 3.5, Reactor Instrumentation:1. 3.5.a was old 3.4.a2. 3.5.b was old 3.4.b3. 3.5.c was old 3.4.d4. 3.5.d was old 3.4.ef. Section 3.6, Emergency Electrical Power System:1. 3.6.awasold3.10.a

g. Section 3.7, Radiation Monitoring Systems Airborne Effluents:1. 3.7.a was old 3.4.a2. 3.7.b was old 3.7.ah. Section 3.8, Experiments:I. 3.8.a was old 3.6.a

2. 3.8.b was old 3,6.b3. 3.8.c was old 3.6.c4. 3.8.d was old 3.6.d5. 3.8.e was old 3.6.e6. 3.8.fwasold3.6.f7. 3.8.g was old 3.6.g8. 3.8.h was old 3.6.h9. 3.8.i was old 3.6.i10. 3.8.j was old 3.6.j11. 3.8.k was old 3.6.k12. 3.8.1 was old 3.6.1

13, 3.8.m was old 3.6.m14. 3.8.n was old 3.6.n15. 3.8.o was old 3,6.o16. 3.8.p was old 3. l.g17. 3.8.q was old 3.l.h

18. 3.8.r was old 3. L.i

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

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

1. 3.9.awasold3.10.b2. 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 Specification2. 4.1 .b- New Specification

3. 4.1 .c was old 4.5.ab. Section 4.2, Reactor Control and Reactor Safety Systems:

1. 4.2.a was old 4.3.a2. 4.2.b was old 4.3.b

3. 4.2.c - New Specification4. 4.2.d - New Specification5. 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 Specification10. 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.c5. 4.3.c - New Specification6. 4.3.d - New Specification

7. 4.3.e - New Specification8. 4.3.f- New Specification

9. 4.3.g was old 4.2.ad. Section 4.4, Reactor Containment Building:

1. 4.4.awasold4.1.a2. 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.b2. 4.7.b - New Specification (previously covered by old 4.4.a)

h. Section 4.8, Experiments:

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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.a2. 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.e6. 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.i10. 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.a5. 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.d8. 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:

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1. 6.1.a was old 6. L.a2. 6. L.b was old 6. .b3. 6.1.c was old 6.1.c4. 6. l.d - New Specification5. 6.1.e was old 6.1.d

b. Section 6.2, Review and Audit:1. 6.2.a was old 6.2.a2. 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 thesecond paragraph "once during each calendar quarter" was replaced by "quarterly."

3. 6.2.c was old 6.2.c4. 6.2.d was old 6.2.d5. 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.a2. 6.4.b was old 6.3.b3. 6.4.c was old 6.3.c4. 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 Specificationf. Section 6.6, Reportable Events and Required Actions:

I. 6.6.a was old 6.5.a2. 6.6.b was old 6.5.b3. 6.6.c was old 6.5.c4. 6.6.d was old 6.5.d5. 6.6.e was old 6.5.e

g. Section 6.7, Records:1. 6.7.a was old 6.4.a2: 6.7.b was old 6.4.b

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APPENDIX A

TECHNICAL SPECIFICATIONS

FOR

THE UNIVERSITY OF MISSOURI RESEARCH REACTOR

FACILITY OPERATING LICENSE R- 103DOCKET 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 inorder to protect the environment and the health and safety of the facility staff and the generalpublic in accordance with 10 CFR 50.36.

This document is divided into the following six sections:

Section 1 - DefinitionsSection 2 - Safety Limits (SL) and Limiting Safety System Settings (LSSS)Section 3 - Limiting Conditions for Operation (LCO)Section 4 - Surveillance RequirementsSection 5 - Design FeaturesSection 6 - Administrative Controls

Specific limitations and equipment requirements for safe reactor operation and for dealing withabnormal situations are called specifications. These specifications, typically derived from thefacility descriptions and safety considerations contained in the Safety Analysis Report (SAR),represent a comprehensive envelope of safe operation. Only those operational parameters andequipment requirements directly related to preserving that safe envelope are listed in theTechnical Specifications. Procedures or actions employed to meet the requirements of theseTechnical Specifications are not included in the Technical Specifications. Normal operation ofthe reactor within the limits of the Technical Specifications will not result in off-site radiationexposure in excess of 10 CFR 20 guidelines.

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

" 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 asystem or operation. This is the most important statement in the TechnicalSpecifications; and

" Bases - This provides the background or reasoning for the choice of specification(s), orreferences 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.

i

TABLE OF CONTENTS

1.0 D E F IN IT IO N S .................................................................................... A -11.1 A bnorm al O ccurrences .................................................................. A -i1.2 C enter T est H ole .......................................................................... A -11.3 C hannel ............................................................................. . . . . A -11.4 C hannel C alibration ................................................................... A -11.5 C hannel C heck ........................................................................ A -11.6 C hannel T est .............................................................................. A -21.7 C ontrol B lade (R od) ..................................................................... A -21.8 E xcess R eactivity ......................................................................... A -21.9 E xperim ent ................................................................................. A -21.10 F lu x T rap .................................................................................. A -21 .11 Irrad iated F uel ............................................................................ A -21.12 Limiting Safety System Settings ....................................................... A-21.13 M ovable Experim ent ..................................................................... A -21.14 O p erab le .................................................................................... A -31.15 O p eratin g .................................................................................. A -31 .16 O perational M odes ........................................................................ A -31.17 Reactivity Worth of an Experiment .................................................... A-31.18 Reactor Containment Building ......................................................... A-31 .19 R eactor C ore .............................................................................. A -31.20 R eactor O perator .......................................................................... A -31.2 1 R eactor in O peration ..................................................................... A -31.22 R eactor Safety System ................................................................... A -31.23 R eactor Scram ............................................................................. A -41.24 R eactor Secured ........................................................................... A -41.25 R eactor Shutdow n ........................................................................ A -41.26 Reference C ore C ondition ............................................................... A -41.27 R egulating B lade (R od) .................................................................. A -41.28 Removable Experiment .............................................................. A-51.29 R esearch R eactor ......................................................................... A -51.30 Research Reactor Facility ............................................................ A-51.31 R od R un-In System ................................................................... A -51.32 S afety L im its .............................................................................. A -51.33 Secured Experim ent ................................................................... A -51.34 Senior Reactor Operator .............................................................. A-51.35 Shim B lade (R od) ..................................................................... A -51.36 Shall, Should, and M ay .................................................................. A -61.37 Shutdow n M argin ........................................................................ A -61.38 Surveillance Intervals .................................................................... A -61.39 T rue V alue ................................................................................. A -61.40 U nscheduled Shutdow n .................................................................. A -61.4 1 U nsecured Experim ent ................................................................... A -6

ii

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS .................... A-72 .1 S afety L im its .............................................................................. A -72.2 Limiting Safety System Settings ...................................................... A-12

3.0 LIMITING CONDITIONS FOR OPERATION ............................................. A-133.1 Reactor Core Parameters ............................................................ A-133.2 Reactor Control and Reactor Safety Systems .................................... A-153.3 Reactor Coolant Systems ............................................................... A-223.4 Reactor Containment Building ........................................................ A-253.5 R eactor Instrum entation ............................................................... A -273.6 Emergency Electrical Power System ................................................. A-293.7 Radiation Monitoring Systems and Airborne Effluents ........................... A-303 .8 E xperim ents .............................................................................. A -323.9 A uxiliary System s ...................................................................... A -37

4.0 SURVEILLANCE REQUIREMENTS ....................................................... A-384.1 Reactor Core Parameters ............................................................... A-384.2 Reactor Control and Reactor Safety Systems ....................................... A-404.3 Reactor Coolant Systems ............................................................... A-424.4 Reactor Containment Building ........................................................ A-444.5 R eactor Instrum entation ................................................................ A -454.6 Emergency Electrical Power System ................................................. A-464.7 Radiation Monitoring Systems and Airborne Effluents ........................... A-474 .8 E xperim ents .............................................................................. A -4 84 .9 A uxiliary System s ....................................................................... A -49

5.0 DESIGN FEATURES ........................................................................ A-505.1 S ite D escription ......................................................................... A -505.2 Reactor Coolant Systems ............................................................... A-525.3 R eactor C ore and Fuel .................................................................. A -545.4 F uel S torage .............................................................................. A -565.5 Reactor Containment Building ........................................................ A-575.6 Emergency Electrical Power System ................................................. A-59

6.0 ADMINISTRATIVE CONTROLS ............................................................ A-606 .1 O rganization .............................................................................. A -606.2 R eview and A udit ....................................................................... A -6 16.3 R adiation Safety ......................................................................... A -636 .4 P rocedures ............................................................................... A -636.5 Experiment Review and Approval .................................................... A-646.6 Reportable Events and Required Actions ............................................ A-646 .7 R eco rd s ................................................................................... A -6 6

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THIS PAGE INTENTIONALLY LEFT BLANK

iv

UNIVERSITY OF MISSOURI RESEARCH REACTORTECHNICAL SPECIFICATIONS

Docket 50-186, License R-103

1.0 DEFINITIONS

1.1 Abnormal Occurrences - An abnormal occurrence is any of the following whichoccurs during reactor operation:

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

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

c. A reactor safety system component malfunction which renders or couldrender the reactor safety system incapable of performing its intended safetyfunction unless the malfunction or condition is discovered duringmaintenance tests or periods of reactor shutdowns;

d. An unanticipated or uncontrolled change in reactivity in excess of 0.006Ak/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 minorleaks), which could result in exceeding prescribed radiation exposure limitsof personnel or environment, or both; or

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

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

1.3 Channel - A channel is the combination of sensor, line, amplifier, and outputdevices 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 suchthat its output corresponds with acceptable accuracy to known values of theparameter that the channel measures. Calibration shall encompass the entirechannel, including equipment actuation, alarm, or trip, and shall be deemed toinclude a channel test.

1.5 Channel Check - A channel check is a qualitative verification of acceptableperformance by observation of channel behavior. This verification, wherepossible, shall include comparison of the channel with other independent channelsor systems measuring the same variable.

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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 intochannel and the observation of proper channel response. When applicable, thetest shall include verification of proper safety trip operation.

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

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

1.9 Experiment - An experiment is any operation, hardware, or target (excludingdevices such as detectors or foils) which is designed to investigate non-routinereactor characteristics or which is intended for irradiation within an irradiationfacility. Hardware rigidly secured to a core or shield structure so as to be part oftheir 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 thecore bounded by the 4.5-inch inside diameter tube and 15 inches above and belowthe reactor core horizontal center line.

1.11 Irradiated Fuel - Irradiated fuel is any fuel element which has been irradiatedand 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-dayand with a decay time of less than 7 days since last irradiation;

ORc. Less than or equal to 1.0 kilowatt-day and with a decay time of less than 24

hours since last irradiation.

1.12 Limiting Safety System Settings - Limiting Safety System Settings (LSSS) aresettings for automatic protection devices related to those variables havingsignificant safety functions. Where a limiting safety system setting is specifiedfor a variable on which a safety limit has been placed, the setting shall be sochosen that automatic protective action will correct the most severe abnormalsituation anticipated before a safety limit is exceeded.

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

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1.14 Operable - Operable means a component or system is capable of performing itsintended function.

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

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

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

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

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

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

1.18 Reactor Containment Building - The reactor containment building is areinforced 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 thereactor pressure vessels occupied by eight or less fuel elements.

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

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

1.22 Reactor Safety System - The reactor safety system is that combination of sensingdevices, electronic circuits and equipment, signal conditioning equipment, andelectro-mechanical devices that serves to either effect a reactor scram, or activatesthe engineered safety features.

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1 .23 Reactor Scram - A reactor scram is the insertion of all four shim rods bygravitational force as a result of removing the holding current from the shim roddrive 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 withoptimum available conditions of moderation and reflection with all fourshim rods removed,

ORb. 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 keylocked in the key box or in custody of a licensed operator,

ORii. The dummy load test connectors are installed on the shim rod drive

mechanisms and a licensed operator is present in the reactorcontrol room;

(3) No work is in progress involving the transfer of fuel in or out of thereactor core;

(4) No work is in progress involving the shim rods or shim rod drivemechanisms with the exception of installing or removing the dummyload test connectors; and

(5) The reactor pressure vessel cover is secured in position and no work isin progress on the reactor core assembly support structure.

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

1.26 Reference Core Condition - Reference core condition is the condition of the corewhen it is at ambient temperature (cold) and the reactivity worth of xenon isnegligible (< 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

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reactor at the desired power level. The regulating blade (rod) may be controlledby the operator with a manual switch or push button, or by an automaticcontroller.

1.28 Removable Experiment - A removable experiment is any experiment which canreasonably 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 supporta self-sustaining neutron chain reaction for research, development, educational,training, or experimental purposes and that may have provisions for theproduction of radioisotopes.

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

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

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

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

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

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

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1.36 Shall, Should, and May - The word "shall" is used to denote a requirement; theword "should" is used to denote a recommendation; and the word "may" is usedto denote permission, neither a requirement nor a recommendation.

1.37 Shutdown Margin - Shutdown margin is the minimum shutdown reactivitynecessary to provide confidence that the reactor can be made subcritical by meansof the control and reactor safety systems starting from any permissible operatingcondition and with the most reactive shim blade and the regulating blade in thefully withdrawn positions, and that the reactor will remain subcritical withoutfurther operator action.

1.38 Surveillance Intervals - Surveillance intervals are the maximum allowableintervals established to provide operational flexibility and not reduce frequency.Established frequencies shall be maintained over the long term. The surveillanceinterval is the time between a check, test or calibration, whichever is appropriateto the item being subjected to the surveillance, and is measured from the date ofthe last surveillance. Allowable surveillance intervals shall not exceed thefollowing:

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 unplannedshutdown, that occurs after all "Blade Full-In Lights" have cleared, caused byactuation of the reactor safety system, rod run-in system, operator error,equipment malfunction, or a manual shutdown in response to conditions thatcould adversely affect safe operation, not including shutdowns that occur duringtesting or checkout operations.

1.41 Unsecured Experiment - An unsecured experiment is any experiment which isnot secured as defined by Definition 1.33, or the moving parts of securedexperiments when they are in motion.

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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

2.1 Safety Limits

Applicability:This specification applies to the interrelated variables associated with reactor corethermal and hydraulic performance. These measurable operating or process variablesinclude reactor power level, core flow rate, reactor inlet water temperature, andpressurizer pressure.

Objective:The objective of this specification is to define a four-dimensional safety limit envelopesuch that operation within this envelope will assure that the integrity of the fuel elementcladding is maintained.

Specification:Reactor power level, core flow rate, reactor inlet water temperature, and pressurizerpressure 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 Figures2.0, 2.1 and 2.2. The limits are considered exceeded if, for core flow rates greaterthan or equal to 400 gpm, the point defined by reactor power level and core flowrate is at any time above the curve corresponding to the true values of reactor inletwater temperature and pressurizer pressure. To define values of the safety limitsfor 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. Forpressurizer pressures greater than 85 psia, the 85 psia curves (Figure 2.2) shall beused 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 flowrate less than 400 gpm. Reactor operation with a core flow rate below 400 gpmwill occur only after a normal reactor shutdown when the primary coolantcirculation pumps are secured or following a loss of flow transient. Under theabove conditions, the maximum fuel element cladding temperature shall notapproach a temperature that would challenge the integrity of the fuel elementcladding.

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

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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 watertemperature and core flow rate to the reactor power level corresponding to adeparture from nucleate boiling ratio (DNBR) of 1.2. This is based on theburnout heat flux data experimentally verified for Advanced Test Reactor (ATR)type fuel elements. Curves are presented for pressurizer pressures of 60, 75 and85 psia. The safety limits were chosen from the results of this analysis for Mode Iand 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 gpmby the low flow scram settings in the reactor safety system. The region below 400gpm will only be entered following a reactor shutdown when the primary coolantcirculation pumps are secured or during a loss of flow transient where the reactorscrams., the flow coasts to zero, reverses, and natural convective cooling isestablished through the decay heat removal system. The analysis of a loss of flowtransient presented in Section 13.2.4 of the SAR, from the ultraconservativeconditions of 11 MW of power, a core flow rate of 3,800 gpm, and a reactor inletwater temperature of 155 'F, indicated a maximum fuel plate centerlinetemperature 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) ispresented in Section 4.6.1 of the SAR.

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22

20 Figure 2.0MvU-RR Mode I and II

Safety Limit Curve

18 Pressurizer at 60 psiaCore Flow Rate > 400 GPM

1l6 Reactor Inlet WaterTemperature, IF

~14

L 12

S10 -

S8

6

1 20140

, 160 _ "

180 -- -

.200 . ........

4

2

NOTEFor Core Flow Rates below400 GPM the Safet, Limit is

established by Specification 2.1.b.

CD

00

00

H~r1

cj~-e

H0zcj~

tll

0

cj.~

0

~r1

H0

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

Core Flow Rate (GPM)

22

Figure 2.1MURR Mode I and It 1

Safety Limit Cur.vePressurizer at 75 psia ...............

18 Cowe Flow Rate > 400 GPM Reactor Inlet WaterTemperature. 'F

14 - . . -16

121 2 ... ............ .. . .... ....... ....... ....... ..... ..... . ....... .... . .. .... ..... ...... ... ... .. .... .... ... , ........ ..... .. .... ...... . 2 0 0

QL

108 .. . . . . . .. . .. . .... ..... ...... ... .. ! . . . . . . .. .. . . . .. ... . . .. .... .. ..... ... ..... ........ ... ... ....... ... - . .. .. .... . ........ ........ ...... ..- ---

4 NOTEFor Core Flow Rates belowv400 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 Mode I and I1Safety Limit Curv-eSafey Liit CrveReactor Inlet Water'

18 .......... Pressurizer at 85 psia . ea n leture. i1 8r F lo ..... ... .4 0 .G P. .......... .. ....... . I e lm p e ra tt re , •'T-ii• .................. .... ............ ...... ....... . . .. .. ..........

Core Flow Rate > 400 GPM4

16 - - 120 -- - .- ,- -- ---- - - - - -- - _ _

0 2120

10

• . 1 26 - ... . . . . . . . ..... . .......... ... ........... ..... ... .. .... .. . .. . . . . . ... . ... . . .. .... .. .. . ....... . .... . 1 4 ..... . .. ....... .. ... ... ... ... ...

U

1 0 .- - ........ ... ..

S 8 .- .~ .... . . . ..... ... -... . . . ...- - -..--.....

4. NOTE

For Core Flow Rates below

->

400 GPM the Safety Limit is2 established by Specification 2 I >

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

400 800 1200 1600 2000 2400 2800 3200 3600 4000 3

Core Ilow Rate (GPM) n'O



2.2 Limiting Safety System Settings

Applicability:This specification applies to the set points for the reactor safety channels monitoringreactor power level, primary coolant flow, reactor inlet water temperature andpressurizer pressure.

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

Specification:

a. Mode I Operation

b.

C.

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)

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)

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 chosensuch that the true value of any of the four safety-related variables, i.e., reactorpower level, core flow rate, reactor inlet water temperature and pressurizerpressure 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 andII operation meet this criterion.

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

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3.0 LIMITING CONDITIONS FOR OPERATION

General: Limiting Conditions for Operation (LCO) are those administratively establishedconstraints on equipment and operational characteristics that shall be adhered to duringoperation of the facility. The LCOs are the lowest functional capability or performancelevel 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 andshut down at all times and that the fuel elements are operated within acceptable designconsiderations 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 mostreactive 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 onless than eight assemblies for the purposes of reactor calibration or multiplicationmeasurement 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 detectedor in which the dimensional changes of any coolant channel between the fuelplates 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 fullelements to assure the validity of the safety limit curves (Specification 2.1) and

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3.1 Reactor Core Parameters - Continued

other safety analyses. When it may be important to conservatively determine theactual critical core loading, Specification 3.1.c allows operation with less thaneight fuel elements up at a lower level not to exceed 100 watts. This maximumpower limit is low enough to ensure no fuel damage will occur. This provides fora conservative approach to criticality with less than eight new fuel elements.

Typically, the first approach to critical would be with a number of fuel elementsinsufficient to achieve criticality but be able to observe subcritical multiplication.Then one additional fuel element would be added at a time in between approachesto critical. The reactor would be operated in this manner only to performnecessary conservative approaches to criticality.

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

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

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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 thereactor control system, thus avoiding conditions which could jeopardize the integrity ofthe fuel element cladding or endanger personnel health and safety, and to specify theminimum number of reactor safety system instrument channels that must be operablefor 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 distancebetween 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 inless than 0.7 seconds.

d. The maximum rate of reactivity insertion for the regulating blade shall not exceed2.5 x 10-4 Ak/k/sec.

e. The maximum rate of reactivity insertion for the four shim blades operatingsimultaneously 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 areoperable. Each of the rod run-in functions shall have 1/N logic where N is thenumber 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 belowcenterline (Min)

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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 disengagedWith Magnet from any rod

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

7. Truck Entry 1 1 1 Loss of entry doorseal pressure

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

9. Manual Rod Run-In 1 1 1 Push button onControl Console

(I) Not required (a) below 50 kW operation with the natural convection flangeand reactor pressure vessel cover removed or (b) in operation with the reactorsubcritical 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 channelsnecessary to provide the following scrams shall be operable whenever the reactoris in operation. Each of the safety system functions shall have 1/N logic where Nis the number of instrument channels required for the corresponding mode ofoperation.

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

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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 belowLevel centerline (Min)

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

15. Primary Coolant 1 1 I1 • Either valve offIsolation Valves open position507A/B Off OpenPosition

16. Pool Coolant Isolation 1 1 0 Valve 509 off openValve 509 Off Open positionPosition

17. Power Level Interlock I1 1 Scram as a result ofincorrect selectionof operating mode

18. Facility Evacuation 11 1 Scram as a result ofactuating thefacility evacuationsystem

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19. Reactor Isolation I I 1 Scram as a result ofactuating the reactorisolation system

20. Manual Scram 1 1 1 Push button onControl Console

,)(6)(6()21. Center Test Hole 2(6) 2(6) Scram as a result ofremoving the centertest hole removableexperiment test tubesor strainer

(I) Not required (a) below 50 kW operation with the natural convection flangeand pressure vessel cover removed or (b) in operation with the reactorsubcritical by a margin of at least 0.0 15 Ak/k.

(2) Flow orifice or heat exchanger AP (psi) in each operating heat exchanger legcorresponding 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 ofSpecification 3.8.q. This safety function shall only be bypassed with specificauthorization from the Reactor Manager.

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

Minimum NumbersInterlock Function Operable

I. Rod Withdrawal Prevents the control rods fromProhibit being withdrawn unless certain

control system logic functionshave been satisfied

2. Automatic Control Prevents placing the reactor inProhibit automatic control unless certain

control system logic functionshave 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 thatcan occur in the core to ensure the validity of the power peaking factors describedin Section 4.5 of the SAR.

c. Specification 3.2.c assures prompt shutdown of the reactor in the event a scramsignal is received as analyzed in Section 13.2.2 of the SAR. The 20% level isdefined as 20% of the shim blade full travel as measured from the frilly insertedposition. Below the 20% level, the fall of the shim blade is cushioned by adashpot assembly. Approximately 91% of the shim blade total worth is insertedat the 20% level.

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

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

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

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The low pool level rod run-in provides assurance that the radiation level fromdirect core radiation above the pool will not exceed 2.5 mR/h (Ref. Section11.1.5.1 of the SAR).

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

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

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

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

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

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

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

With the reflector plenum natural convection valve V547 in the open position anda pool coolant flow rate at 850 gpm, the pool coolant low flow scram assures theadequate 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 pressurescram provides a backup to the low pool coolant flow scram.

The pressurizer high pressure scram provides assurance that the reactor will beshut down during a high pressure transient before the relief valve set point or the

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pressure limit of the primary coolant system is reached as analyzed in Section13.2.9.4 of the SAR.

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

The pool water low level scram assures that the radiation level above the reactorpool from direct core radiation remains below 2.5 mR/h (Ref. Section 11.1.5.1 ofthe 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 ofprotection for a loss of flow accident (in their respective system) initiated by aninadvertent closure of the isolation valve(s).

The power level interlock (PLI) scram provides assurance that the reactor cannotbe operated with a power level greater than that authorized for the mode ofoperation selected on the Power Level Switch. The PLI scram also provides theinterlocks to assure that the reactor cannot be operated in Mode I with a primaryor pool coolant low flow scram bypassed.

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

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

The center test hole scram provides assurance that the reactor cannot be operatedunless the removable experiment test tubes or the strainer is inserted and latchedin the center test hole. This is required anytime the reactivity worth of the centertest hole removable experiment test tubes and the contained experiments or thestrainer 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 theremovable experiment test tubes and the contained experiments or the strainerdoes not exceed the limit of Specification 3.8.q and is authorized by the ReactorManager.

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

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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 toprevent 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 isoperable,

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 primarycoolant 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 secondarycoolant system exceeds the limits of 10 CFR 20, Appendix B, Table 3, forradioisotopes with half-lives greater than 24 hours.

e. The conductivity of the water in the primary coolant system shall be maintained atless 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 between5.0 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 atless 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 5.0and 7.0 when averaged over a period of one month.

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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 addair to obtain the specified pressure. The system pressure will be verified,recorded, and readjusted as required every 4 hours as part of the facility routinepatrol. Procedures will be established for manual verification of water level in theanti-siphon system for conditions when system pressure has an unexplained riseof 4 psi or more. If water level is 6 inches or more above the anti-siphon isolationvalves or a system leak or other malfunction prevents the maintenance of pressurein the specified range, the reactor will be shut down until the malfunction can becorrected.

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 bythe invert loop and the anti-siphon system on the outlet line. Upon opening, theanti-siphon isolation valves will admit a fixed volume of air to the highest point ofthe invert loop, thus preventing the reactor core from becoming uncovered bybreaking 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 primarycoolant lines as close as practicable to the biological shield. Proper operation ofthese 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 ofthe 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 theformation 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/mlwould contain a total iodine-131 inventory of 0.038 Ci in the system. Based onthe iodine-131 activity in the reactor core provided in Section 13.2.1.2 of theSAR, this iodine-131 concentration would equate to less than 0.000022 % of thetotal core iodine-131 inventory in the primary coolant. Specifications 3.3.b and3.3.c provide for the early detection of a leaking fuel element so that correctiveaction can be taken to prevent the release of fission products.

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

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

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corrosion and limits concentrations of particulate and dissolved containments thatcould be made radioactive by neutron irradiation (NUREG-1537).

Specification 3.3.i ensures that the anti-siphon system will perform its intendedfunction as designed by imposing certain operational limits on the system (Ref.Section 6.3 of the SAR).

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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 maintainedwhen required so that the health and safety of the general public is not endangered as aresult 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 tomaintain 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 inthe 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 wassatisfactory.

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 beinghandled.

c. While reactor containment integrity is required, the reactor containment buildingshall be automatically isolated if the activity in the ventilation exhaust plenum orat the reactor bridge indicates an increase of 10 times above previouslyestablished levels at the same operating condition. Exception: The containmentisolation set point may temporarily be increased to avoid an inadvertent scram andisolation during controlled evolutions such as experiment transfers or minormaintenance in the reactor pool area. The pool area shall be continuously

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3.4 Reactor Containment Building - Continued

monitored, and, if necessary, a manual containment isolation actuated, until theautomatic 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 arelease of radioactivity is negligible.

c. Radiation monitors located at the reactor bridge and in the reactor containmentbuilding ventilation exhaust plenum supply input signals to meters located in thereactor control room. A containment isolation will occur when radiation levels inthese areas exceed a predetermined value. During operations such as the removalof experiments or equipment from the pool, the radiation level at the level of thereactor bridge or in the exhaust plenum can increase significantly for shortperiods. To prevent inadvertent containment isolations, it may be necessary toraise the set point on the reactor bridge or exhaust plenum monitor. Duringperiods in which the set point is raised to more than one decade above the normalreading, the radiation level in the area of the monitor will be continuouslymonitored. Thus, should the radiation level increase from unknown causes orfrom material which could be released to the unrestricted environment,the reactor containment building can be quickly isolated by manually actuatingthe isolation system.

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3.5 Reactor Instrumentation

Applicability:This specification applies to the instruments that provide information which must beavailable to the operator during reactor operation.

Objective:The objective of this specification is to ensure that sufficient reliable information ispresented 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 OperableChannel 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 arenot 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)

(2)Not required for Mode III operation.Reactor Pool Temperature limit is a maximum of 100 'F when in Mode IIIoperation and (a) below 50 kW with the natural convection flange and reactorpressure cover removed or (b) with the reactor subcritical by a margin of atleast 0.015 Ak/k.

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

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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 1count per second and the Wide Range Monitor is indicating a power levelgreater 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 neutrontest source or movement on the Source Range meter demonstrating thatthe 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 lowneutron flux levels present during a startup.

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

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

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

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

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

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

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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 availableto 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 entrydoors, facility ventilation exhaust fans, emergency lighting panel, and reactorinstrumentation and control systems. Therefore, on a loss of normal electricalpower, the emergency electrical power system is not required for protection of theintegrity of the fuel elements. In the extremely unlikely event of a simultaneousloss of normal electrical power and fuel element failure, the operation of theemergency electrical power system would be required to provide for continuouscontainment isolation (Ref. Section 13.2.7 of the SAR).

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3.7 Radiation Monitoring Systems and Airborne Effluents

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

Obiective:The objective of this specification is to assure that sufficient radiation monitoringinformation is available to the reactor operator during reactor operations and exposureto the public resulting from the radioactivity released from the reactor facility to theunrestricted environment will not exceed the limits of 10 CFR 20.

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

Minimum Numbers OperableChannel Mode I Mode II Mode III

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

2. Reactor Containment Building Exhaust 1 1 IPlenum 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 maintenanceand sample handling. During these periods, the radiation monitor indicationwill be closely observed.

(2) The off-gas (stack) radiation monitor may be placed out of service for up to 2hours for calibration and maintenance. During this out-of-service time, noexperimental or maintenance activities will be conducted which could likelyresult in the release of unknown quantities of airborne radioactivity.

b. The maximum discharge rate through the ventilation exhaust stack shall notexceed the following:

Max. Concentration Max. ControlledType 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

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3.7 Radiation Monitoring Systems and Airborne Effluents - Continued

AEC = Air Effluent Concentration as listed in Appendix B, Table 2, Column Iof 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 ofradioactivity to the surroundings and so that there will be sufficient time toevacuate the facility should it be necessary to do so.

b. Dispersion calculations based upon standard reference material and experimentdata obtained at the reactor show that argon-41 concentrations under averageconditions will be 0.008 of the AEC limits in the unrestricted area surrounding thereactor facility. Dilution factors under conservative conditions are in the range of5 x 104 under both average and stable conditions at ground level from the facilitybuilding.

The normal short burst releases at the facility are five to ten seconds in durationand occur on an average of ten times per day five days per week. The short burstsaffect 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 hazardto the health and safety of the public.

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3.8 Experiments

Applicability:This specification applies to all experiments which directly utilize neutrons or otherradiation produced by the reactor. Radioactive sources shall meet the requirements forexperiments.

Objective:The objective of this specification is to prevent an accident which would jeopardize thesafe operation of the reactor or would constitute a hazard to the safety of the facilitystaff 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 themaximum strontium-90 inventory is no greater than 300 millicuries.

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

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

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

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

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

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

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3.8 Experiments - Continued

h. Cooling shall be provided to prevent the surface temperature of a submergedirradiated experiment from exceeding the saturation temperature of the coolingmedium.

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

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

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

1. The normal operating procedures shall include controls on the use or exclusion ofcorrosive, flammable, and toxic materials in experiments or in the reactorcontainment building. These procedural controls shall include a current list ofthose materials which shall not be used and the specific controls and proceduresapplicable to the use of corrosive, flammable, or toxic materials which areauthorized.

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 afactor 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-135greater than 1.5 curies or strontium-90 greater than 5 millicuries shall be inirradiation containers that satisfy the requirements of Specification 3.8.i or bevented to the facility ventilation exhaust stack through high efficiency particulateair (HEPA) and charcoal filters which are continuously monitored for an increasein radiation levels.

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

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

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

s. The absolute value of the reactivity worth of each unsecured experiment shall notexceed 0.0025 Ak/k.

t. The absolute value of the reactivity worth of all unsecured experiments which arein 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 agreater inventory of fission products has not been completed, and therefore theiruse 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 inthe reactor core or water annulus surrounding the center test hole by restrictingmaterials which could generate or accumulate gases or vapors.

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

d. Specification 3.8.d is intended to reduce the likelihood of damage to reactor orpool 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 thecenter test hole while the reactor is operating to those that will not introducereactivity transients more severe than one that can be controlled without initiatingsafety 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 toassure that anticipated transients will not result in radioactivity release and thatexperiments will not jeopardize the safe operation of the reactor.

h. Specification 3.8.h is intended to reduce the likelihood of reactivity transients dueto accidental voiding in the reactor or the failure of an experiment from internal orexternal heat generation.

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

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

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

1. Specification 3.8.1 assures that corrosive materials which are chemicallyincompatible with reactor components, highly flammable materials, and toxicmaterials are adequately controlled and that this information is disseminated to allreactor users.

m. The extremely low temperatures of the cryogenic liquids present structuralproblems that enhance the potential of an experiment failure. Specification 3.8.mprovides for the proper review of proposed experiments containing or usingcryogenic materials.

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

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

p. Specification 3.8.p provides assurance that any inadvertent insertion/removal orcredible malfunction of a secured removable experiment would not introducepositive reactivity whose consequences would lead to radiation exposures inexcess of the 10 CFR 20 limits. The step reactivity insertion is analyzed inSection 13.2.2 of the SAR.

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

r. Specification 3.8.r provides assurance that the movement of movable experimentsor movable parts of any experiment will not introduce reactivity transients moresevere than one that can be controlled without initiating a reactor safety systemaction as analyzed in Section 13.2.2 of the SAR.

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s. Specification 3.8.s prevents the installation of an unsecured experiment whichcould introduce, as a positive step change, sufficient reactivity to place the reactorin a transient that would cause a violation of a safety limit as analyzed in Section13.2.2 of the SAR.

t. Specification 3.8.t assures that the reactivity worth of all unsecured experimentsshall not exceed the maximum value authorized for a single secured removableexperiment.

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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 auxiliarysystems 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 primarygrade water.

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

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 approximately1,000 gpm to the reactor pool. This supply assures that the water level in the poolwill remain above the reflector in case a 6-inch beamport or a 6-inch pool coolantline is sheared (Ref. Sections 13.2.9.1 and 13.2.9.2 of the SAR).

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4.0 SURVEILLANCE REQUIREMENTS

General: Surveillance frequencies denoted herein are based on continuing operation ofthe reactor. Surveillance activities scheduled to occur during an operating cycle whichcannot be performed with the reactor operating may be deferred to the end of that currentreactor operating cycle. If the reactor is not operated for a reasonable time, a reactorsystem or measuring channel surveillance requirement may be waived during theassociated period. Prior to reactor system or measuring channel operation, thesurveillance shall be performed for each reactor system or measuring channel for whichsurveillance was waived. A reactor system or measuring channel shall not be consideredoperable until it is successfully tested. Surveillance intervals shall not exceed thosedefined by Definition 1.38. Discovery of noncompliance with any of the surveillancespecifications listed in this Section shall limit reactor operations to that required toperform the surveillance.

4.1 Reactor Core Parameters

Applicability:This specification applies to the surveillance requirements of the reactor coreparameters.

Obiective:The objective of this specification is to verify reactor core parameters which aredirectly 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 bladechange.

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

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

Bases:a. - b. Annual measurements, coupled with measurements made after changes that can

affect reactivity values, provide adequate assurance that core behavior resultingfrom configuration changes are adequately characterized.

c. The specified fuel element inspections along with the continuous primary coolantsystem fission product monitoring and the weekly radiochemical analysis of theprimary coolant provide for the detection of anomalies resulting from reactoroperation and reduces the possibility of fission product release to the primary

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4.1 Reactor Core Parameters - Continued

coolant system. Inspecting the fuel elements at the end of their life has the addedadvantage of allowing for the decay of the fuel elements and, thus, reduction ofexposure to personnel.

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4.2 Reactor Control and Reactor Safety Systems

Applicability:This specification applies to the surveillance requirements on the reactor control andreactor safety systems.

Obiective:The objective of this specification is to reasonably assure proper operation of thereactor 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 thatevery blade is inspected every two years. The reactor shall not be operated with acontrol blade that exhibits abnormal swelling or abnormalities that affectperformance.

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

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

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

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

g. The reactor safety system shall be channel tested before each reactor startupinvolving a refueling, a shutdown greater than 24 hours or quarterly.

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

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

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

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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. Inover 40 years of operation, to date, the shim blades have never failed to meetSpecification 3.2.c.

b. Periodic inspection of the shim blades provides detection of singular bladeabnormalities and any potential generic blade design deficiencies. Specification4.2.b further assures that the reactor will not be operated using shim blades withsuspected 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 speedmechanical devices and withdrawal and insertion speeds should not vary exceptas a result of mechanical wear. The surveillance is chosen to provide a significantmargin over expected failure or wear rates of these mechanical devices.

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

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

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Applicability:This specification applies to the surveillance requirements on the reactor coolantsystems.

Objective:The objective of this specification is to reasonably assure proper operation of thereactor 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 testedprior 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 operationand a radiochemical analysis performed to determine the concentration of iodine-131.

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

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

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

f. The conductivity and pH of the water in the pool coolant system shall bemeasured 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 monthlytesting is adequate to provide assurance of continued operability.

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

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Docket 50-186., License R-103


release of fission products. Specification 4.2.b establishes the frequency ofverification of compliance with Specification 3.3.c.

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

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

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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 thecontainment 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 tothe test.

b. The containment actuation (reactor isolation) system, including each of itsradiation 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 thedesign limits outlined in Specification 5.5.c. No special maintenance will beperformed prior to the test so that the results demonstrate the historic integrity ofthe containment structure.

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

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4.5 Reactor Instrumentation

Applicability:This specification applies to the surveillance requirements of the reactorinstrumentation systems.

Obiective:The objective of this specification is to reasonably assure proper operation of thereactor 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 Specification3.5.b shall be channel calibrated on a semiannual basis.

c. All nuclear instrumentation channels shall be channel-tested before each reactorstartup. This test shall not be required prior to a restart within two (2) hoursfollowing a normal reactor shutdown or an unplanned scram where the cause ofthe scram is readily determined not to involve an unsafe condition or a failure ofone 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 areoperable.

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Applicability:This specification applies to the surveillance requirements of the emergency electricalpower system.

Obiective:The objective of this specification is to reasonably assure proper operation of theemergency 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 electricalloads 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 providingreasonable assurance that the emergency power generator electrical control anddistribution system will remain operable.

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4.7 Radiation Monitoring Systems and Airborne Effluents

Applicability:This specification applies to the surveillance requirements of the radiation monitoringinstrumentation.

Obiective:The objective of this specification is to reasonably assure proper operation of theradiation 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 bechannel 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 longtime period.

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

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4.8 Experiments

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

Objective:The objective of this specification is to prevent the conduct of experiments which maydamage the reactor or release excessive amounts of radioactive materials as a result ofexperiment 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, asappropriate, 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 ofthe reactor or exceeding the limits specified in the Technical Specifications.

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4.9 Auxiliary Systems

Applicability:This specification applies to the surveillance requirements of the reactor auxiliarysystems.

Objective:The objective of this specification is to reasonably assure proper operation of theauxiliary 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 semiannualbasis.

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 virtuallyunlimited source of raw water for the emergency pool fill system. Water supplyis 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.

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5.0 DESIGN FEATURES

General: Major alterations to safety-related components or equipment shall not be madeprior 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 milesouthwest of the University of Missouri at Columbia's main campus. Thiscampus is located in the southern portion of Columbia, the county seat and largestcity in Boone County, Missouri.

Approximate distances to the University property lines from the reactor facilityare 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 lotsurrounding the MURR facility itself. Within the restricted area the ReactorFacility Director has direct authority and control over all activities, normal andemergency. There are pre-established evacuation routes and procedures known topersonnel 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; andthe MKT Nature and Fitness Trail. The area within these boundaries is ownedand controlled by MU and may be frequented by people unacquainted with theoperation of the reactor. The Reactor Facility Director has authority to initiateemergency actions in this area, if required.

'Providence Road crosses MU property separating the University Research Parkfrom another MU-owned tract of land lying to the east. The road runs north and southwith the closest point of approach being approximately 400 meters east of the reactorfacility. MU has the authority to determine all activities including the exclusion orremoval of personnel and property and to temporarily secure the flow of traffic on thisroad during an emergency.

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5.1 Site Description - Continued

Bases:

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

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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. Thefollowing 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 primarycoolant heat exchangers, two pool coolant circulation pumps, one pool coolantheat exchanger, and one pool water hold-up tank, plus all associated piping andvalves.

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

c. The circulation pumps and heat exchangers of the primary coolant system shallconstitute two parallel systems separately instrumented to permit safe operation atfive megawatts on either system or ten megawatts with both systems operatingsimultaneously.

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

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

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 thereactor 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 IIIoperation except for operation with the reactor subcritical by a margin of at least0.015 Ak/k.

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

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

Exceptions:

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

b. Some materials in off-the-shelf commercial components may be excepted fromSpecification 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 asdescribed.

Specification 5.2.a as excepted, permits reactor operation at 50% of full power inthe event of a major component failure in which repairs cannot be accomplishedin a reasonable period of time. The reactor was designed and has extensive safeoperating 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 systemintegrity.

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

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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 configurationand 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 morenegative 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 nominalactive length of 24 inches and a nominal plate thickness of 0.050 inches. Thenominal distance between the fuel plates shall be 0.080 inches. Plate nominalcladding 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 fuelstorage 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 andberyllium reflector. Four blades shall be for coarse control (shim blades) and onefor 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

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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 damagewill not occur following any credible step reactivity insertion as analyzed inSection 13.2.2 of the SAR.

b. The average core void coefficient of reactivity also limits the step reactivityinsertion 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 suchthat 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 dispersionfuel system) successfully and extensively used for many years in test and researchreactors. Specifications 5.3.d, 5.3.e and 5.3.f require fuel content and dimensionsof the fuel elements to be in accordance with the design and fabricationspecifications (Ref. Section 4.2.1 of the SAR).

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

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

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

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

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5.4 Fuel Storage

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

Objective:The objective of this specification is to assure that fuel which is stored shall not becomecritical 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 ageometrical array where the value of Keff is less than 0.9 under all conditions ofmoderation.

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

Bases:

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

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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 accidentalrelease of radioactivity to the environment.

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

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

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

c. The containment building leakage rate shall not exceed 16.3 cubic feet per minuteat STP with an overpressure of one pound per square inch gauge or 10% of thecontained volume over a 24-hour period from an initial overpressure of twopounds 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 orcombination under control of the Reactor Manager.

Bases:

a. No credible accident scenario has been identified which can result in a significantoverpressure condition in the reactor containment building. However,Specification 5.5.a assures that a sufficient free volume exists to prevent anypressure 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 adequateatmospheric dispersion.

c. Specification 5.5.c assures that the containment building will have sufficientintegrity to limit the leakage of contained potentially radioactive air in the event

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5.5 Reactor Containment Building - Continued

of any reactor accident to ensure exposures are maintained below the limits of 10CFR 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.

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Applicability:This specification applies to the facility emergency electrical power system.

Obiective:The objective of this specification is to assure adequate emergency electrical power inthe 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 providingemergency electrical power to the emergency lighting system, the facilityventilation exhaust system, reactor instrumentation, and the personnel air lockdoors.

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 electricalpower to monitor the reactor systems and assure personnel safety in the event of anormal power failure to the reactor facility.

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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 asshown in Figure 6.0.

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

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

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

i. Compliance with Technical Specifications and licenserequirements regarding reactor operation, maintenance andsurveillance; and

ii. Oversight of the experiment review process.

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

i. Compliance with Technical Specifications and licenserequirements regarding radiation safety, byproduct materialhandling 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 personnelat the facility. One of these individuals shall be a Reactor Operator or a SeniorReactor Operator licensed pursuant to 10 CFR 55. The other individual must beknowledgeable 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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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 thefacility or readily available on call at all times during operation, and shall bepresent at the facility during all startups and approaches to power, recovery froman unplanned or unscheduled shutdown or non-emergency power reduction, andrefueling activities.

6.2 Review and Audit

a. A Reactor Advisory Committee (RAC) shall provide independent oversight inmatters pertaining to the safe operation of the reactor and with regard to plannedresearch activities and use of the facility building and equipment. The RAC shallreview:

(1) Proposed changes to the MURR equipment, systems, or procedures whensuch changes have safety significance, or involve an amendment to thefacility operating license, a change in the Technical Specificationsincorporated in the license, or a question pursuant to 10 CFR 50.59.Changes to procedures that do not change their original intent may bemade without prior RAC review if approved by the TS designatedmanager, either the Reactor Health Physics Manager or Reactor Manager,or a designated alternate who is a member of Health Physics or a SeniorReactor Operator, respectively. All such changes to the procedures shallbe documented and subsequently reviewed by the RAC;

(2) Proposed experiments significantly different from any previouslyreviewed or which involve a question pursuant to 10 CFR 50.59;

(3) The circumstances of reportable occurrences and violations of theTechnical Specifications or license and the measures taken to prevent arecurrence;

(4) Violations of internal procedures or operating abnormalities having safetysignificance; and

(5) Reports from audits required by the Technical Specifications.

b. The RAC may appoint subcommittees consisting of knowledgeable members ofthe public, students, faculty, and staff of MU when it deems it necessary in orderto effectively discharge its primary responsibilities. When subcommittees are

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6.2 Review and Audit - Continued

appointed, these are to consist of no less than three members with no more thanone student appointed to each committee. The subcommittees may be authorizedto act on behalf of the parent committee.

The RAC and its subcommittees are to maintain minutes of meetings in which theitems considered and the committees' recommendations are recorded.Independent actions of the subcommittees are to be reviewed by the parentcommittee at the next regular meeting. A quorum of the committee or thesubcommittees consisting of at least fifty percent of the appointed members mustbe present at any meeting to conduct the business of the committee orsubcommittee. The RAC shall meet at least quarterly.

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

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

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

e. Audits:

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

i. Facility Operations, for conformance to the TechnicalSpecifications and license conditions, at least annually;

ii. Operator Requalification Program, for compliance with theapproved program, at least every two years; and

iii. Corrective Action items associated with reactor safety, at leastannually.

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

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6.3 Radiation Safety

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

6.4 Procedures

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

(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 onreactor safety;

(4) Surveillance checks, calibrations and inspections that may affect reactorsafety;

(5) Administrative controls for operations and maintenance and for theconduct of irradiations and experiments that could affect reactor safety orcore reactivity; and

(6) Implementation of the Emergency Plan.

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

c. The Reactor Manager shall approve and annually review the procedures fornormal operations of the reactor and the Emergency Plan implementingprocedures. The Reactor Health Physics Manager shall approve and annuallyreview the radiological control procedures and the procedures for the preparationfor shipping and the shipping of byproduct material.

d. Deviations from procedures required by this Specification may be enacted by aSenior Reactor Operator or member of Health Physics, as applicable. Suchdeviations shall be documented and reported within 24 hours or the next workingday to the Reactor Manager or Reactor Health Physics Manager or designatedalternate.

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6.5 Experiment Review and Approval

Approved experiments shall be carried out in accordance with established and approvedprocedures. Procedures related to experiment review and approval shall include thefollowing:

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

b. Substantive changes to previously approved experiments shall be made only afterreview 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 followingactions shall be taken:

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

(2) The safety limit violation shall be promptly reported to the NRC. Promptreporting of the violation shall be made by MU, by telephone or email, tothe NRC Operations Center no later than the following working day;

(3) A detailed follow-up report shall be prepared. The report shall include thefollowing:

i. Applicable circumstances leading to the violation including, whenknown, 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; andv. Corrective actions to prevent recurrence.

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

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

(1) Reactor conditions shall be returned to normal or the reactor shall be shutdown;

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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 oremail, to the NRC Operations Center no later than the following workingday;

(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 willbe submitted within fourteen (14) days to the NRC Document ControlDesk.

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

(Note: Where components or systems are provided in addition to those requiredby these Technical Specifications, the failure of the extra components or systemsis not considered reportable provided that the minimum numbers of componentsor 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 oremail, to the NRC Operations Center no later than the following workingday;

(2) A detailed follow-up report shall be prepared. The follow-up report willbe submitted within fourteen (14) days to the NRC Document ControlDesk; and

(3) The reactor shall be shut down or placed in a safe condition and return tonormal reactor operations will not be allowed until authorized by theReactor Manager.

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

(1) Any significant change(s) in the transient or accident analyses as describedin the SAR; and

(2) Permanent changes in the facility organization involving the Office of theProvost or the Director's Office.

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6.6 Reportable Events and Required Actions - Continued

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

(1) A brief narrative summary of (a) operating experience (includingoperations designed to measure reactor characteristics), (b) changes in thereactor facility design, performance characteristics, and operatingprocedures 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, includingthe reasons therefore and corrective action, if any, taken;

(4) Discussion of the major maintenance operations performed during theperiod, 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 theprocedures, tests and experiments carried out under the conditions of 10CFR 50.59;

(6) A summary of the nature and amount of radioactive effluents released ordischarged to the environs beyond the effective control of the licensee asmeasured at or prior to the point of such release or discharge;

(7) A description of any environmental surveys performed outside the reactorfacility; and

(8) A summary of radiation exposures received by facility staff,experimenters, and visitors, including the dates and time of significantexposure, and a brief summary of the results of radiation andcontamination surveys performed within the facility.

6.7 Records

Records of the following activities shall be maintained and retained for the periodsspecified below. The records may be in the form of logs, data sheets, or other suitableforms or documents. The required information may be contained in single or multiplerecords, or a combination thereof.

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6.7 Records - Continued

a. Lifetime Records - The following records are to be retained for the lifetime of thereactor facility: (Note: Applicable annual reports, if they contain all of therequired 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 TechnicalSpecifications;

(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 atleast five years or for the life of the component involved, whichever is shorter:

(1) Normal reactor facility operation (but not including supporting documentssuch as checklists, log sheets, etc. which shall be maintained for a periodof 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 byapplicable 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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Board of Curators ofthe University of Missouri

IIIII

I Reporting Lines

.... Communication Lines

FIGURE 6.0UNIVERSITY OF MISSOURI RESEARCH REACTOR (MURR)

ORGANIZATION

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