ML11255A003

Univ of Missouri - Columbia Research Reactor'S Responses to NRC Request for Additional Information Regarding a Renewal Request for Amended Facility Operating License R-103

ML11255A003
Person / Time
Site:	University of Missouri-Columbia
Issue date:	09/08/2011
From:	Rhonda Butler Univ of Missouri
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML11255A003 (27)

Text

UNIVERSITY of MISSOURI RESEARCH REACTOR CENTER September 8, 2011 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Mail Station P1-37 Washington, DC 20555-0001

Reference:

Docket 50-186 University of Missouri-Columbia Research Reactor Amended Facility License R- 103 Enclosed you will find the University of Missouri-Columbia Research Reactor's responses to the U.S. Nuclear Regulatory Commission's (NRC) request for additional information, dated May 6, 2010 (Complex Questions) and June 1, 2010 (45-Day Response Questions) regarding our renewal request for Amended Facility Operating License R-103, which was submitted to the NRC on August 31, 2006, as supplemented.

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

Sincerely,

,rcvt A,-ac,0-f /,3c,,,-<_L-it Ralph A. Butler, P.E.

Director RAB/djr Enclosures CHRISTINE M.ERRANTE Qublc -Notary Seal a of Missour Commissioned for Boone Cou My Commission Expires: April 14,2015 Commission Mumher 11528381 vI 1513 Research Park Drive Columbia, MO 65211 Phone: 573-882-4211 Fax: 573-882-6360 Web: www.murr.missouri.edu 4o0oAL Fighting Cancer with Tomorrow's Technology

UNIVERSITY of MISSOURI RESEARCH REACTOR CENTER September 8, 2011 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Mail Station P1-37 Washington, DC 20555-0001

REFERENCE:

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

SUBJECT:

Written communication as specified by 10 CFR 50.4(b)(1) regarding responses to the "University of Missouri at Columbia - Request for Additional Information Re: License Renewal, Safety Analysis Report, Complex Questions (TAC No. MD3034)," dated May 6, 2010, and the "University of Missouri at Columbia - Request for Additional Information Re: License Renewal, Safety Analysis Report, 45-Day Response Questions (TAC No. MD3034)," dated June 1, 2010 On August 31, 2006, the University of Missouri-Columbia Research Reactor (MURR) submitted a request to the U.S. Nuclear Regulatory Commission (NRC) to renew Amended Facility Operating License R-103.

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

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

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

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

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

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

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

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

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

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

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

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

Attached are MURR's responses to six (6) of the remaining 45-Day Response and Complex Questions.

With these responses, the following 45-Day Response and Complex Questions remain unanswered: 4.7, 6.2, 4.15, 4.16, 13.4.a, 13.4.b, 13.6, 13.7, C.1 and C.3. Additionally, MURR is currently evaluating which RAIs will need to be resubmitted based on the recent Amendment request to revise the MURR Safety Limits (ML 1123A088).

Because of the newly revised Safety Limits, and because additional RELAP work is required to answer the thermal-hydraulic questions, MURR is requesting additional time to answer the remaining 45-Day Response and Complex Questions. The extension request will be discussed with our NRC Senior Project Manager.

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

ENDORSEMENT:

Sincerely, Reviewed and Approved, AC-l~- A C/; C-( 1 John L. Fruits Ralph A. Butler, Reactor Manager Director CHRISTINE M.ERRANTE xc: Reactor Advisory Committee Reactor Safety Subcommittee Commissioned for Boone Coun My Commission Expires: April 14,2015 Dr. Robert Duncan, Vice Chancellor for Research Commission Numter: 11528381 Mr. Craig Basset, U.S. NRC Mr. Alexander Adams, U.S. NRC Page 2 of 26

Chapter 4 4.14 Section 4.5.3, Operating Limits.

d. Page 4-40 and Technical Specification (TS) 3.2.b. Discuss how the one inch limit restricts flux tilting to ensurepower peakingfactors.

A summary of the MURR Hot Channel Factors is provided in Table F.4, Appendix F, of the August 24, 2011 University of Missouri-Columbia submittal to the U.S. Nuclear Regulatory Commission (NRC) requesting revised reactor Safety Limits (SLs) (Ref. 1). The MURR SL Analysis is based on an enthalpy rise overall peaking factor product of 2.4416 - the nuclear peaking and the engineering hot channel factor components are 2.3014 and 1.061, respectively.

The peak heat flux power-related factors have an overall product of 4.116 - the nuclear peaking and the engineering hot channel factor components are 3.475 and 1.184, respectively. The difference in control rod heights could affect the nuclear peaking factor in both the enthalpy rise and the hot spot heat flux.

Therefore, the effect of unbanked control rods on the nuclear peaking factors was checked using the computer code MCNP5. The models used were developed by a combined staff from MURR and Argonne National Laboratory (ANL) for the future low-enriched uranium (LEU) fuel conversion of MURR. The benchmarking models include the MIURR 1971 graphite reflector configuration to the detailed physics analysis on the original 6.2 Kg 235U aluminide fuel core.

The current graphite reflector configuration was in place in 2008 when the current model for benchmarking to our mixed core (elements in various stages of burnup) fuel cycle was performed.

Burnup for the mixed cores was modeled using REBUS/DIF3 to compare the current highly-enriched uranium (HEU) fuel cycle to the proposed LEU fuel cycle.

To answer this question, the highly detailed, equilibrium (MCNP5) models of various core configurations were used to determine the worst case peaking factors. In particular, the geometry of each modeled core configuration included highly segmented fuel meat regions, where each fuel meat region (in every element) was divided into 24 one-inch axial sections. By tallying the energy deposited from the MCNP5 criticality calculations in each of the axial sections of any given fuel meat region, the nuclear peaking factors were derived as a function of axial fuel length.

The largest axial peaking factors were found to be in the inner and outer most fuel meat positions (i.e. in fuel plate-1 and -24, respectively) with fuel plate-I being the highest. The highest peaking factors can be found in the following two extreme core conditions: "Fresh Core" - eight fresh fuel elements or "Week 58 Core" - a mixed core with two fresh fuel elements (elements 1 and 5) located next to two fuel elements (elements 4 and 8) which will reach a burnup of 150 MWD at the end of the run.

The MCNP5 MURR core configurations used to obtain peaking factors were modeled based on the following five factors, each of which includes two extreme cases:

1. Variation in control rod height - all the same heights (banked) vs. rods 'A' and 'D' having a 4-inch height difference when compared to rods 'B' and 'C';

2. Fuel loading - "Fresh Core (0 MWD) vs. worst case "Week 58 Core" (two elements each with 0, 81, 65, and 142 MWD);

3. Graphite reflector configuration - 1971 vs. present (2008);

4. Flux Trap (FT) - all water vs. sample holder loaded with samples inserted into the FT; and

5. Control rod critical position - xenon free vs. equilibrium xenon heights.

Page 3 of 26

The modeled core configurations based on these five factors include the extreme power peaking MURR core configurations. To begin with, the worst case heat flux peaking is found in the "Fresh Core," which has not been used at MJURR in about 30 years because of license possession limits. This core requires 6.2 Kg of 235U fresh fuel, which exceeds the current 5 Kg unirradiated 235U possession limit. Since 1973, in order to obtain the maximum burnup from the fuel elements, MiURR has operated with a mixed core fuel cycle loaded with elements with burnups from low MWD to high MWD. With equilibrium xenon in the core, the control rods are fully withdrawn with a little more than 700 MWD on the core. During the weekly scheduled maintenance day, the MIURR fuel cycle requires reloading the core with eight xenon free fuel elements with the average MWD burnup on a fuel element being around 70 MWD. The typical core has four pairs of fuel elements with a pair of elements having the same MWD burnup and being loaded across from each other, i.e. fuel positions 1 and 5, 2 and 6, etc. After "Fresh Core,"

the next worst case heat flux peaking occurs in the extreme mixed burnup core loading.

In performing our LEU conversion feasibility study, the MURR fuel cycle had to be modeled for both the current HEU and the proposed LEU fuel element. For the two different types of fuel elements, over a year's worth of refueling or different core loadings were modeled. The worst case mixed burnup for the HEU modeled fuel cycle was labeled the "Week 58 Core." It consists of two fuel elements each with the following power history: 0, 81, 65 and 142 MWD. This worst case peaking occurs in the 0 MWD element loaded next to a 142 MWD element, which results in higher azimuthal peaking between fuel elements due to the difference in 235U loading. These peaking factors will be higher than those of the average MURR core. The mixed burnup fuel cycle was modeled using the 2008 reflector configuration so that it could be benchmarked against xenon free mixed core loadings. The fuel element burnup atom densities obtained for the "Week 58 Core" were used in the week 58 cases for both. the 1971 and 2008 graphite reflector configurations.

The graphite reflector configuration has varied over the years. The two reflector configurations selected are the extreme cases regarding their effect on overall reactivity. The original graphite reflector configuration, which provided the highest excess reactivity for a fresh core, was still being used in 1971 when the physics measurements of the first 6.2 Kg 235U aluminide fuel core were performed. The 2008 graphite reflector configuration has the lowest excess reactivity of the various reflector configurations that have been used. This is because there is less graphite in the reflector elements due to the increase in irradiation positions and certain elements containing boron and cadmium. Results from MCNP5 KCODE calculations using a xenon free, mixed core model, with an empty (flooded) flux trap and banked control rods, showed that the difference in reactivity from the 1971 to the current graphite reflector configuration is -0.00638 Ak/k.

For these cases, the control rod heights describe the core in either xenon free or an equilibrium xenon state. The highest power peaking factors are always in fuel plate-i, therefore the xenon free control rod critical height and the flux trap region only containing pool water and not samples results in the higher peaking factors. The lower critical control rod height pushes the power density in and downward in the core increasing the peaking in fuel plate-1. The water filled flux trap region also increases the thermal flux reaching plate-1. Therefore, the results included in this answer contain only the peaking factors and graphs for the xenon free and water filled flux trap cases.

First, these worst case power peaking cases modeled show that the peaking factors for the heat flux power and enthalpy peaks are within the MURR hot channel factors on which the SLs are based. Then the control rods are modeled with a four-inch difference between them to see if the peaking factors are still within MURR hot channel factors given in Table F.4 of the revised SLs.

Page 4 of 26

For the 1971 reflector configuration, control rods 'B' and 'C' are placed four inches above the height of control rods 'A' and 'D.' This results in the highest peaking in fuel element position 5 of the 1971 reflector configuration. For the current reflector configuration, control rods 'A' and

'D' are placed four inches above the height of control rods 'B' and 'C.' This results in the highest, peaking in fuel element position 1 for the current reflector configuration because the two reflector elements containing boron and cadmium have control rods 'B' and 'C' between them and fuel elements 4, 5and 6, so raising them does not cause as much increase in the peaking factor.

Eight MCNP5 MURR core configurations were modeled based on the above listed five factors.

The peaking factors of fuel plate-I and -24 in fuel elements 1, 3 and 5 are displayed in Figures 1 through 8 for the different core configurations. The MCNP5 calculated highest heat flux and enthalpy rise nuclear peaking factors for each configuration are presented in Table 1 for the hot spot and hot stripe on the hottest fuel plate, which is always fuel plate-i of the hottest fuel element.

Figures 1 through 8 are for the following control rod, reactor core and reflector configurations (all configurations are xenon free and the flux trap full of water):

Figure L.a - Control Rods banked, Fresh Core, 1971 Reflector (Plate-1)

Figure 1.b - Control Rods banked, Fresh Core, 1971 Reflector (Plate-24)

Figure 2.a - Control Rods 4-inch diff., Fresh Core, 1971 Reflector (Plate-1)

Figure 2.b - Control Rods 4-inch diff., Fresh Core, 1971 Reflector (Plate-24)

Figure 3.a - Control Rods banked, Week 58 Core, 1971 Reflector (Plate-i)

Figure 3.b - Control Rods banked, Week 58 Core, 1971 Reflector (Plate-24)

Figure 4.a - Control Rods 4-inch diff., Week 58 Core, 1971 Reflector (Plate-i)

Figure 4.b - Control Rods 4-inch diff., Week 58 Core, 1971 Reflector (Plate-24)

Figure 5.a - Control Rods banked, Fresh Core, 2008 Reflector (Plate-i)

Figure 5.b - Control Rods banked, Fresh Core, 2008 Reflector (Plate-24)

Figure 6.a - Control Rods 4-inch diff., Fresh Core, 2008 Reflector (Plate-I)

Figure 6.b - Control Rods 4-inch diff., Fresh Core, 2008 Reflector (Plate-24)

Figure 7.a - Control Rods banked, Week 58 Core, 2008 Reflector (Plate-I)

Figure 7.b - Control Rods banked, Week 58 Core, 2008 Reflector (Plate-24)

Figure 8.a - Control Rods 4-inch diff., Week 58 Core, 2008 Reflector (Plate-I)

Figure 8.b - Control Rods 4-inch diff., Week 58 Core, 2008 Reflector (Plate-24)

Page 5 of 26

Power Peaking Factors In Fuel Plate 1 of ELEMENT 1, 3 and 5 for no Variation Between Control Blade Height: HEU MURR Fresh Core - Day 0, 1971 Graphite Reflector, Water-Filled Flux-Trap 3

E 2.5 U

E (U

CL 2

14

'ka 1.5 CL U-CL 0.5 0

0 5 10 15 20 25 Axial Fuel Length (inches)

Figure 1.a Power Peaking Factors In Fuel Plate 24 of ELEMENT 1, 3 and 5 for no Variation Between Control Blade Height: HEU MURR Fresh Core - Day 0, 1971 Graphite Reflector, Water-Filled Flux-Trap 3

E b.

2.5 U

2 U

m dl LL 1.5 0.b 0

W 1

-UW-ELEMENT 3 Plzte 24 Peaking facto, 0.5 0

0 5 10 15 20 25 Axial Fuel Length (inches)

Figure 1.b Page 6 of 26

Power Peaking. Factors In Fuel Plate 1 of ELEMENT 1, 5 and 3 for 4-inch Variation Between Control Blade Height: HEU MURR Fresh Core - Day 0, 1971Graphite Reflector, Water-Filled Flux-Trap 3

Q E

2.5 U

U C 2 CL m-FL 1.5 UL

i 1 0.

0.5 0

0 5 10 15 20 25 Axial Fuel Length (inches)

Figure 2.a Power Peaking Factors In Fuel Plate 24 of ELEMENT 1, 5 and 3 for 4-inch Variation Between Control Blade Height: HEU MURR Fresh Core - Day 0, 1971 Graphite Reflector, Water-Filled Flux-Trap 3

C ci E

2.5 U

C 2

U N ci C

m

0. 1.5 CL ELENME'T 1 PItc~24 Peak ing factor 1

-UW-ELEME'4T3 PIatC24 Peaking factor ELErvE4T5 Pldtcn21 PejkiiiCfjtur 0.5 0

0 5 10 15 20 25 Axial Fuel Length (inches)

Figure 2.b Page 7 of 26

Power Peaking Factors In Fuel Plate 1 of ELEMENT 1, 3 and 5 for no Variation Between Control Blade Height: HEU MURR Mixed Core - Week 58 Day 0, 1971 Graphite Reflector, Water-Filled Flux-Trap 3

E 2.5 C)

C UL 1.5 0.S 4-9 am IL 0.5 0

0 5 10 15 20 25 Axial Fuel Length (inches)

Figure 3.a Power Peaking Factors In Fuel Plate 24 of ELEMENT 1, 3 and 5 for no Variation Between Control Blade Height: HEU MURR Mixed Core - Week 58 Day 0, 1971 Graphite Reflector, Water-Filled Flux-Trap 3

z a)

E 2.5 U

A

-CL4 W6 1.5 1 C6 ELEMENT1 PHlet24 Pu~kin factor 1 i1

-- HI-M[ FNI !PIP~te)4 I-'EA1Itng fa,.or 0.5 -

0 0 5 10 15 20 25 Axial Fuel Length (inches)

Figure 3.b Page 8 of 26

Power Peaking Factors In Fuel Plate I of ELEMENT 1, 5 and 3 for 4-inch Variation Between Control Blade Height: HEU MURR Mixed Core - Week 58 Day 0, 1971 Graphite Reflector, Water-Filled Flux-Trap 3

E 2.5

.S CL

'-4 U a' CL 1.5 a)

-- #ELEMENT. ýIre ýea~ingfzcto, oeý

-U-WLEMENT3 aIate - eacngf-,cto, 0.5 ELEMENT5:Ijte I ea~inu Lawo 0

0 5 10 i5 20 25 Axial Fuel Length (inches)

Figure 4.a Power Peaking Factors In Fuel Plate 24 of ELEMENT 1, 5 and 3 for 4-inch Variation Between Control Blade Height: HEU MURR Mixed Core - Week 58 Day 0, 1971 Graphite Reflector, Water-Filled Flux-Trap 3

t E

2.5 4 C

0 CL

0. N a'

J. 4.'

fU UL 1.5 U-C ELEMEN-I Plate 24 Pe~ki ig ktor a,

1 0.

-U&- ELEMEN-3 Plate 24 Pezki iq ctur o3 ELEMENS5 PI.t,t24 'eja~it facto,

0. 0.5 0

0 5 10 15 20 25 Axial Fuel Length (inches)

Figure 4.b Page 9 of 26

Power Peaking Factors In Fuel Plate 1 of ELEMENT 1, 3 and 5 for no Variation Between Control Blade Height: HEU MURR Fresh Core - Day 0, 2008 Graphite Reflector, Water-Filled Flux-Trap 3

E P

.S U 2 CL 0._

1.5 LL a-

.J o 0.5 a_

0 0 5 10 is 20 25 Axial Fuel Length (inches)

Figure 5.a Power Peaking Factors In Fuel Plate 24 of ELEMENT 1, 3 and 5 for no Variation Between Control Blade Height: HEU MURR Fresh Core - Day 0, 2008 Graphite Reflector, Water-Filled Flux-Trap 3

.4.

C 0)

E 2.5 CL 2 1.5 (U

0.

1 CL 0.5 0

0 5 10 15 20 25 Axial Fuel Length (inches)

Figure 5.b Page 10 of 26

Power Peaking Factors In Fuel Plate 24 of ELEMENT 1, 5 and 3 for 4-inch Variation Between Control Blade Height: HEU MURR Fresh Core - Day 0, 2008 Graphite Reflector, Water-Filled Flux-Trap 3

E U

2 ko*

04.

u-1.5 0-a.

I-o 0.5 0

0 5 10 15 20 25 Axial Fuel Length (inches)

Figure 6.a Power Peaking Factors In Fuel Plate 1 of ELEMENT 1, 5 and 3 for 4-inch Variation Between Control Blade Height: HEU MURR Fresh Core - Day 0, 2008 Graphite Reflector, Water-Filled Flux-Trap 3

E

1. 2.5 U

2 V-4 4-1.5 CL

-- o-EEMENT I Platlc Peak! g9factor 1

-UW-EEM_=NT3 Plate I Peaki-gfictor I IMENT5 Plate I Peaki-g factor 0.5 0

0 5 10 15 20 25 Axial Fuel Length (inches)

Figure 6.b Page 11 of 26

Power Peaking Factors In Fuel Plate 24 of ELEMENT 1, 3 and 5 for no Variation Between Control Blade Height: HEU MURR Mixed Core Week 58 - Day 0, 2008 Graphite Reflector, Water-Filled Flux-Trap 3

E L. 2.5 U

CL 2

U6 1.5 1

0.5 0

0 5 10 15 20 25 Axial Fuel Length (inches)

Figure 7.a Power Peaking Factors In Fuel Plate 1 of ELEMENT 1, 3 and 5 for no Variation Between Control Blade Height: HEU MURR Mixed Core Week 58- Day 0, 2008 Graphite Reflector, Water-Filled Flux-Trap 3

2.5 0

2 U6 m

-471NfNIIPArtIPpkn i r 1.5 -0M1 PAf IPpkng an 0.

ICU

~ I'7 L l ,IPe iglcn 1

0.

CL 0.5 0

0 5 10 15 20 25 Axial Fuel Length (inches)

Figure 7.b Page 12 of 26

Power Peaking Factors In Fuel Plate 1 of ELEMENT 1, 5 and 3 for 4-inch Variation Between Control Blade Height: HEU MURR Mixed Core week 58 -

Day 0, 2008 Graphite Reflector, Water-Filled Flux-Trap 3

4.

C (U

E 2.5 CL U 2 m -4 U. 03 41

'U K-a.

tu

-. e- ELEVENr I P ate 1 PeakIng factor 0

-- W--l FVFNT' P ate 1 Ptaking lactor CL ELEMENS P ate I Peakingfactor 0.5 0

0 5 10 15 20 25 Axial Fuel Length (inches)

Figure 8.a Power Peaking Factors In Fuel Plate 24 of ELEMENT 1, 5 and 3 for 4-inch Variation Between Control Blade Height: HEU MURR Mixed Core week 58 -

DayO, 2008 Graphite Reflector, Water-Filled Flux-Trap 3

4' E

02 U

2.5 C

0.5 03 1j -ELEMENT I Plia: 21 Pejk rig factor

-U--ELEMENT 5 Plie24 Peakg factor

-mu V/J ELEMENTS Plate 24 Peak ng factor 0

D 5 10 15 20 25 Axial Fuel Length (inches)

Figure 8.b Page 13 of 26

The eight control rod, reactor core and reflector configurations for fuel plate-I nuclear peaking factors listed in Table I (all configurations are xenon free and the flux trap full of water):

1. Control Rods banked, Fresh Core, 1971 Reflector

2. Control Rods 4-inch diff., Fresh Core, 1971 Reflector

3. Control Rods banked, Week 58 Core, 1971 Reflector

4. Control Rods 4-inch diff., Week 58 Core, 1971 Reflector

5. Control Rods banked, Fresh Core, 2008 Reflector

6. Control Rods 4-inch diff., Fresh Core, 2008 Reflector

7. Control Rods banked, Week 58 Core, 2008 Reflector

8. Control Rods 4-inch diff., Week 58 Core, 2008 Reflector Table 1 - Summary of the Nuclear Peaking Factors at the Hotspots of Eight MURR Core Extreme Configurations (i.e., "Fresh Core" and "Week 58 Core")

Core Peaking Factor Peaking Factor Configuration Hot Spot' With Engr. PF Enthalpy Rise' With Engr. PF 1 3.08 3.65 1.97 2.09 2 3.17 3.75 2.03 2.15 3 3.06 3.62 2.01 2.14 4 3.06 3.62 2.05 2.18 5 3.08 3.64 1.95 2.07 6 3.12 3.70 1.99 2.11 7 3.01 3.57 1.99 2.11 8 3.06 3.62 2.03 2.15 Safety Limits' 3.475 4.116 2.301 2.442 Note 1: The results presented in Table I reports the overall nuclear peaking factors for the hot spot heat flux and the hot channel enthalpy rise for the various MURR core configurations. The azimuthal peaking factor across the one-inch sections of fuel plate-i is 1.07. Therefore the hot spot peaking factor is the highest peaking factor in the MCNP run for that model multiplied by 1.07 (circumferential peaking within plate-1). Then the hot spot peaking factor is multiplied by 1.03 and 1.15 (engineering hot channel factors on flux). The enthalpy rise peaking factor is the highest average plate-I and -2, which heat coolant channel 2, peaking factors in the MCNP run for that model multiplied by 1.0921 (circumferential peaking within the channel 2). Then the enthalpy rise factor is multiplied by 1.03 and 1.03 (engineering hot channel factors on enthalpy rise).

Note 2: The SL peaking factors are taken from Table F.4, Appendix F, of the August 24, 2011 University of Missouri-Columbia submittal to the U.S. Nuclear Regulatory Commission requesting revised reactor SLs.

Considering the power peaking factors for these highest peaking factor configurations with a four-inch difference in the control rod heights do not exceed the SAR Table 4-14 peaking factors, the one inch limit between the highest and lowest control blades when the reactor is at a power level greater than 100 kW insures that an excess in flux tilting is not created.

Page 14 of 26

REFERENCES:

'Letter Request to the U.S. Nuclear Regulatory Commission, Written communication as specified by 10 CFR 50.4 regarding an application to amend Amended Facility License R-103 by revising the University of Missouri-Columbia Research Reactor Technical Specification 2.1, "Reactor Core Safety Limit," pursuant to 10 CFR 50.90, University of Missouri-Columbia Research Reactor, August 2011 (ML 1123A088).

CHAPTER 13 13.2 Section 13.2.2.1, Accident initiating events and scenarios, Page 13-17.

d Please discuss the effect of build up of an oxide layer on the fuel cladding.

During reactor operation, a thin film of aluminum oxide (A120 3.H20 or boehmite) forms on the fuel plate surface. Based on fuel element oxide thickness measurements taken in July 1987 and using the Griess Correlation, the average hot spot oxide thickness was calculated to be 0.631 mils on a fully burned up MURR fuel element. The "worst case" hot spot oxide thickness was calculated to be 1.27 mils1 . Previous studies of the oxide formation on aluminum 6061 have shown that the spallation of the oxide layer does not occur until a thickness in the range of 2 mils has developed2 .

To investigate the effect of this oxide layer on reactivity transients, a 2.0 mil thick oxide layer was added to the MURR PARET reactivity transient analyses model. Typically, the oxide layer gradually builds up on fuel plate surface during reactor operation and, by the time the layer reaches a thickness of 2.0 mils, that fuel element is near its end-of-life and hence will have a lower power load. However, to be conservative in modeling, the maximum power load was applied to the fuel as if it is a fresh fuel element. Additionally, conservative values were selected for the oxide thermal properties in the PARET model.

With these conservative assumptions incorporated, the MURR limiting step reactivity insertion transient of 0.006 Ak/k was reanalyzed. The results obtained are discussed below.

The initial peak power burst reached during the transient went from 34.4 MW to 36.8 MW. This change is to be expected since the insulating oxide layer will delay the heat transfer from the fuel to the coolant and hence will delay the inherent feedback mechanisms that help limit the transient.

This effect was also evident since the peak power is reached in 0.057 seconds from the start of the transient compared to 0.047 seconds for the no oxide layer case.

The peak centerline temperature of the fuel increased from 418.5 'F (214.7 'C) for the no-oxide layer case to 581.7 'F (305.4 'C) with the 2.0 mil oxide layer. Again, this is consistent since the heat transfer from the fuel to the coolant is inhibited by the oxide layer thus raising the peak centerline temperature attained during the transient. The maximum temperature reached is still well below the melting point or the blister temperature of the fuel.

REFERENCES:

'University of Missouri Research Reactor Letter to U.S. Nuclear Regulatory Commission, "Response to NRC Request for Additional Information," In answer Number 1, September 11, 2

1987.

j.C. Griess, H.X. Savage, and J.L. English, Effect of Heat Flux on the Corrosion of Aluminum by Water, Part IV, ORNL-3541, Union Carbide Corp. Nuclear Division, Oak Ridge National Laboratory, 1964.

Page 15 of 26

13.3 Section 13.2.2.1.2, Continuous Control Blade Withdrawal.

Justify why the regulating blade is not part of this evaluation. Explain if reactor operation in Mode II or III changes the results of the evaluation.

The regulating blade was not initially included as part of the "Continuous Control Blade Withdrawal" evaluations since manual withdrawal of the four shim blades, either individually or in gang, while simultaneously withdrawing the regulating blade is prohibited by administrative procedure AP-RO-110, "Conduct of Operations," and operating procedures OP-RO-211, "Reactor Startup - Hot," and OP-RO-212, "Reactor Startup - Recovery From Temporary Power Reduction." With separate switches for manipulating the shim blades and the regulating blade, inadvertent simultaneous manipulation of the two control mechanisms is highly unlikely.

In order to evaluate the consequences of withdrawing the regulating blade at the same time as the four shim blades during a reactor startup, the "Continuous Control Blade Withdrawal" accident, as discussed in SAR Section 13.2.2.1.2, was reanalyzed using the computer code PARET.

For this analysis, the maximum positive reactivity insertion rate allowed by the Technical Specification (TS) due to the withdrawal of the regulating blade was superimposed on top of the maximum allowed reactivity insertion resulting from the simultaneous withdrawal of the four shim blades. It should be noted that since the total reactivity worth of the regulating blade is limited to 0.006 Ak/k (TS 3.1.c), and at a maximum reactivity addition rate of 0.00025 Ak/k/sec (also TS 3.1.c), the reactivity addition due to withdrawal of the regulating blade occurs only during the first 24 seconds of the "Continuous Control Blade Withdrawal" accident scenario.

Beyond that, the only reactivity addition will be due to shim blade motion. It should also be noted that the current total reactivity worth of the regulating blade is less than half of the TS limit.

In addition, by applying the total reactivity worth of the regulating blade using the maximum insertion rate allowed by the TS, very conservative results can be expected. In practice, the regulating blade displays typical control blade worth behavior of a cosine shape, i.e. maximum worth is only in the middle section of the worth curve and much less effect towards either end of blade travel.

The results obtained are shown in Figure 1. As expected, reactor power level reached during the transient is higher when compared to the case where only the four shim blades are continuously withdrawn. Even though the reactivity worth of the regulating blade is considerably less than the reactivity worth of the shim blades, the rate of reactivity addition is significant because of its higher rate of travel (40 inches/min for the regulating blade compared to one inch/min for the shim blades). Figure 1 shows reactor power continuously rising and reaching a value of almost 11.0 kW within 140 seconds of the accident initiation. It should be noted that for this analysis no credit is taken for the reactor protective system initiating a rod run-in or reactor scram once the appropriate short period set points are reached. For example, the short period rod run-in set point is 11 seconds and this value is reached approximately 120 seconds after initiation of the accident (compared to 140 seconds for the shim blade only withdrawal case). At this point, reactor power level is only 17.2 watts. If the reactor protective system fails to initiate a rod run-in, the scram circuit will initiate a reactor scram once reactor period decreases to 9 seconds - which happens about 123 seconds after accident initiation (compared to 143 seconds for the shim blades only case). Reactor power level at this point is only 22.8 watts.

Thus, even though the effect of a "Continuous Control Blade Withdrawal" accident appears to be more serious when the regulating blade is withdrawn simultaneously along with all four shim blades, the higher rate of reactivity insertion will cause the reactor protective system to initiate a Page 16 of 26

rod run-in or scram at an earlier time which will terminate the transient before any fuel damage can occur.

80.00 60.00 Withi Regulathig Blade S40.00 Withoiir Regiihaing Blade 20.00 0.003 7...

0.0 20)o 40.0 60.0 80.0 100.0 120.0 140.0 Time (sec)

Figure 1 - Reactor Power Versus Time For Continuous Control Blade Withdrawal Accident (With and Without the Regulating Blade)

CHAPTER 14, TECHNICAL SPECIFICATIONS 14.1 Section 14.2, Formatand Content and Introductionsection of TSs. The SAR states that Section 5 of the TSs, Design Features,only contains specifications. The regulations in 10 CFR 50.36(a)(1) requires bases for design features TSs. Please amend your proposed design feature TSs to include bases.

The Technical Specifications will be revised as follows to include the bases for the design features:

5.1 Site Description Bases The MURR facility site location and description are strictly defined in Chapter 2 of the SAR.

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

Page 17 of 26

5.2 Reactor Containment Building Bases

a. No credible accident scenario has been identified which can result in a significant overpressure condition in the reactor containment building. However, Specification 5.2.a assures that a sufficient free volume exists to prevent any pressure buildup in the containment building (Ref. Section 6.2.2.2 of the SAR).

b. Specification 5.2.b assures a sufficient stack height for more than adequate atmospheric dispersion.

c. Specification 5.2.c assures that the containment building will have sufficient integrity to limit the leakage of contained potentially radioactive air in the event of any reactor accident to ensure exposures are maintained below the limits of 10 CFR 20 (Ref. Sections 6.2.10 and 13.2.1 of the SAR).

d. Specification 5.2.d assures safe and secure storage of fresh fuel.

5.3 Reactor Coolant Systems Bases The reactor coolant systems are described and analyzed in Section 5 of the SAR. The reactor can be safely operated at 10 MW with the coolant systems as described.

Specification 5.3.a as excepted, permits reactor operation at 50% of full power in the event of a major component failure in which repairs cannot be accomplished in a reasonable period of time.

The reactor was designed and has extensive safe operating history for operation at 50% of 10 MW cooling capacity. In this event, the shutdown system shall be secured in a manner such as to assure system integrity.

Specification 5.3.e assures strength and corrosion resistance of the coolant system components and excepts some components in the instrumentation of the system which are not commercially available in the materials specified. The size of these components is such that a failure would not result in a hazard to safe operation of the reactor.

5.4 Reactor Core and Fuel Bases

a. - c. The MURR fuel elements are one of a configuration (aluminide UAlx dispersion fuel system) successfully and extensively used for many years in test and research reactors.

Specifications 5.4.a, 5.4.b and 5.4.c require fuel content and dimensions of the fuel elements to be in accordance with the design and fabrication specifications (Ref. Section 4.2.1 of the SAR).

d. Specification 5.4.d assures that the reactor fuel is properly positioned in the pressure vessel during operation (Ref. Section 4.2.5 of the SAR).

Page 18 of 26

e. Specification 5.4.e assures proper neutron reflection as required by design (Ref. Section 4.2.3 of the SAR).

f. Specification 5.4.f assures reactivity of the reactor is properly controlled as required by design (Ref. Section 4.2.2 of the SAR).

g. Specification 5.4.g assures that the reactor consists of the experimental facilities as required by design (Ref. Chapter 10 of the SAR).

5.5 Emergency Electrical Power System Bases

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

Specification 5.5.a assures that a system exists to provide the necessary electrical power to monitor the reactor systems and assure personnel safety in the event of a normal power failure to the reactor facility.

CHAPTER 16 16.1 Section 16.1.1, Fuel and Fuel Cladding, TS 3.8, Reactor Fuel, and TS 4.5, Reactor Fuel.

a. The basesfor TS 3.8.b states that the TS assures that fuel elements found to be defective are no longer used for'reactor operation. The TS contains a limit on dimensional changes of coolant channel between fuel plates of 10 mils. What is the basis for the 10 mils and what is the impact of this amount of fuel channel dimensional change on thermal-hydraulic and accident analysis?

The allowance for a 10-mil reduction in a coolant channel gap is to account for swelling due to fuel meat expansion caused by fuel burnup and clad thickening from oxidation buildup. Because of the fuel meat void fraction of aluminide fuel, the reduction in channel gap from fuel bumup is delayed. This results in heat fluxes of the reduced channel being substantially less than those that existed when irradiation of the fuel element began. With a reduction in heat flux caused by fuel burnup, the 62-mil (0.062 inches) end-of-life minimum allowed coolant channel gap was calculated to have a Safety Limit (SL) of 19.690 MW for reactor power [7.19 MW greater than the reactor power Limiting Safety System Setting (LSSS) of 12.5 MW]. This is 32% greater than the 14.894 MW reactor power SL for 80-mil nominal and 72-mil worst case wide coolant channels. Thus, the fuel elements with the highest bumups and coolant channels reduced to the minimum width allowed are operating well within the MURR SLs.

The design gap of all coolant channels in the core that are bounded by two fuel plates is 80+/-8 mils, therefore the narrowest channel gap could potentially be 72 mils (0.072 inches) after fabrication. The assumed additional 10-mil reduction in channel gap due to oxidation and/or fuel swelling causes the minimum channel gap to be decreased to 62 mils. Making the channel thinner reduces both its coolant velocity and its flow area. Both of these effects reduce its flow rate. Section 5 of Reference 1 considered an analogous situation where the limiting channel was assumed to have a gap of 72 mils instead of its nominal width of 80 mils (0.080 inches). The same analysis applies here using 62 mils in place of 72 mils. The equations that were used are:

Page 19 of 26

1+OC VH =(DHFC; (1)

VN \DN)

WH VHX AH VH AH (2)

WN VNX AN VN AN where:

V = velocity; D = hydraulic diameter; A = flow area; a = the exponent in the friction factor versus Reynolds number relationship; and Subscripts H and N represent the hot and nominal channels, respectively.

The friction, f, versus Reynolds number, Re, relationship is f 0Cl/Re'. Based on the Blasius formula for turbulent flow friction factor, f= 0.316/Re025, a is 0.25.

In Reference 1, the hot channel gap is 72 mils, the nominal channel gap is 80 mils, DH is 0.13876 inches, and DN is 0.15828 inches. This value of DH was obtained as 4 times the flow area divided by the wetted perimeter. The channel flow area is the product of the channel arc length along the average of the inner and outer radii and the channel gap. The channel arc length is one-eighth the circumference of a circle reduced by both the thickness of two side plates (0.150 inches each) and the clearance between adjacent elements (0.04 inches). The nominal radii of the limiting channel analyzed in Reference 1, coolant channel 2 of fuel element 1, are 2.820 and 2.900 inches, corresponding to a channel gap of 0.080 inches. For the hot channel, the inner radius was increased by 0.004 inches and the outer radius was decreased by 0.004 inches to account for the channel gap tolerance of 0.008 inches, which reduced the overall channel gap to 0.072 inches.

Thus, for the Reference 1 hot channel, the flow area in square inches is:

{2 7r [(2.820 + 2.900)/2] / 8 - (2 x 0.150 + 0.04)} x 0.072 = 0.13725 and the wetted perimeter in inches is:

2 7 (2.820 + 2.900) / 8 - 2 x (2 x 0.150 + 0.04) + 2 x 0.072 = 3.9565 The corresponding hydraulic diameter, De, in inches is 4 x 0.13725 / 3.9565 = 0.13876. For the 0.062-inch channel gap, the calculations for the flow area, wetted perimeter, and hydraulic diameter are the same as above except that 0.072 is replaced by 0.062. Thus, the new flow area, wetted perimeter, and hydraulic diameter are 0.11819 in2, 3.9365 inches, and 0.1201 inches, respectively.

For the current analysis, the hot channel gap is 62 mils instead of 72 mils and the nominal channel gap is unchanged. The hot channel flow area factor, which is the area ratio AH/AN, is 62/80 = 0.7750. For the 62-mil channel gap, equation (1) yields:

VH/VN = (0.1201/0.15 828)(1+025)/(2-0.25) = 0.8210.

Page 20 of 26

VHiVN is the engineering hot channel factor for velocity identified in the Reference I analysis.

Equation (2) yields:

WH/WN = 0.8210 x 0.7750 = 0.6363.

The Reference 1 methodology also requires the value of the channel heated diameter (D1), defined as the channel heated perimeter divided by 7r. Di is the same for the reduced channel gap as for the nominal channel gap since thinning the channel does not change its heated perimeter.

In summary, changing the limiting channel gaps from 72 to 62 mils reduces AH/AN, which is the hot channel flow area factor from 0.90 to 0.775, VH/VN, which is the engineering hot channel factor for velocity, from 0.9108 to 0.8210, and the hydraulic diameter (De) from 0.13876 inches to 0.1201 inches. Substituting, the three new values, 0.775, 0.8210, and 0.1201, which are the hot channel flow area factor (cell B65), the engineering hot channel factor on velocity (cell B68), and De (cell D86), respectively into the Table 8 model of Reference 1 accounts for the heat removal reduction caused by the 62-mil narrow channel. A substantial amount of fuel burnup must occur before there is sufficient fuel swelling and clad surface oxidation to cause a 10-mil reduction in channel gap. This bumup considerably reduces the element power. This is addressed in the next section.

The thermal limiting criteria for the MURR during steady-state operation are based on avoiding both flow instability and critical heat flux (CHF). The August 24, 2011 University of Missouri-Columbia submittal to the U.S. Nuclear Regulatory Commission requesting revised reactor SLs provides a detailed description of the model for defining the safety envelope for steady-state operation of the MURR (Ref. 2). This model, as described in Reference 1, was used to address this question about reduced channel thicknesses. Reference I includes a very detailed sample problem solution to promote clarity of the analytical model. Table 8 of Reference I provides the computer spreadsheet that was used for the sample problem to predict the allowed reactor power SL of 14.894 MW with all three (3) non-power LSSS variables set at their corresponding limits, i.e. primary coolant pressurizer pressure at 75 psia, total core flow rate at 3200 gpm, and reactor inlet water temperature at 155 'F. The coolant channel chosen for analysis, channel 2 of element 1, is the innermost one bounded by two fuel plates and is the most limiting channel in the reactor.

This sample problem with relatively minor changes to its input was used to calculate the limiting power level for the reduced channel thickness.

Page 21 of 26

Table 1 - Limiting Channel Plate Axial Peaking Factors and Coolant Temperature Rise Distribution Plate iFraction of Bulk Coolant Temperature Rise from Axiali Bulk Inlet, Fraction Level Peaking Cooanprtur 0.0 0.2 0.4 0.6 0.8 1.0 FactorTeprte 0 1 0.461 0.0188 2 2 0.466 0.0377 4 3 0.537 0.0597 4 0.632 0.0857 6 5 0.705 0.1147 8 6 0.803 0.1478 7 0.872 0.1839 10 7J 8 0.967 0.2241 12 9 1.016 0.2667 10 1.105 0.3132 14 11 1.152 0.3616 16 12 1.214 0.4130 13 1.257 0.4658 18 14 1.278 0.5195 20 15 1.299 0.5743 16 1.297 0.6287 22 17 1.310* 0.6833 24 18 1.250 0.7355 0.0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 19 1.247 0.7875 Plate Axial Peaking Factor 20 1.162 0.8359 21 1.122 0.8821 22 1.002 0.9235 Figure 1- Limiting Channel Plate Axial Peaking 23 0.920 0.9614 Factors and Coolant Temperature Rise Distribution 24 0.925 1.0000

• Maximum value Table 1 and Figure 1 show the axial heat flux peaking factors for fuel plate-I and coolant channel 2 in fuel element 8, which is the most limiting high burnup element with 142 MWd of burnup.

The bulk coolant temperature rise is based on a weighted average of the heat fluxes of the two fuel plates that bound the channel. The corresponding core configuration is for week 58, xenon free (day 0), with an average control rod height of 17 inches withdrawn. In Table 2, the peaking factors for fuel plate-1 and coolant channel 2 of the high burnup fuel element 8 are compared to plate-I and channel 2 peaking factors of the 0 MWd fuel element 1 in a week 58 core, which is the basis for the MURR SLs. The power peaking factors were obtained from MCNP modeling of the MURR fuel cycle. The MCNP modeling to determine the power peaking factors only included the nominal coolant channel dimensions. If the hot channel 2 had been modeled as narrower, it would have further reduced the power peaking due to the localized reduction in moderation.

Page 22 of 26

The variation of heat flux along the azimuthal direction of each fuel plate is also taken into account. In the neutronics analysis the heated arc length along the azimuthal direction of each fuel plate is divided into a series of nine equal-radian arc length vertical strips. The average heat flux of each strip is calculated. The ratio of the highest of these nine averages to the plate-average heat flux is called the "azimuthal peaking factor." The plate axial peaking factors provided for each level in Table 1 are based on level-averaged heat fluxes rather than the level azimuthal maximum. The azimuthal peaking factor for plate-I in fuel element 8 is 1.04, which is a lower value than for fuel element 1 because burnout of the hot stripe is greater than the plate average burnout.

As shown in Table 2 below, the power factor for enthalpy rise for element I is 2.4416, whereas for element 8 it is 1.8184, a 25% reduction in enthalpy rise, while the channel flow reduction factor was reduced by 22%. The power factor for heat flux in element 1 at interval 18, which is the limiting point, is 3.863, whereas for element 8 at the limiting interval 19 the power factor is 2.504. This corresponds to a 35% reduction in the heat flux at the limiting point. When all of the new factors due to both the reduction in channel size and fuel burnup are in place in the Reference 1 model and the high burnup axial power shape is represented in lines 78 and 79, the 62-mil channel is predicted to reach its maximum SL power level of 19.69 MW. This is 32%

larger than the 14.894 MW reactor power SL of Reference 0, demonstrating that for the high burnup fuel element the additional 10-mil reduction in the coolant channel gap creates no safety problem.

REFERENCES:

1Earl E. Feldman, "Implementation of the Flow Instability Model for the University of Missouri Reactor (MURR) That is Based on the Bernath Critical Heat Flux Correlation," ANL-RERTR/TM-1 2Letter 1-28, July 2011.

Request to the U.S. Nuclear Regulatory Commission, Written communication as specified by 10 CFR 50.4 regarding an application to amend Amended Facility License R-103 by revising the University of Missouri-Columbia Research Reactor Technical Specification 2.1, "Reactor Core Safety Limit," pursuant to 10 CFR 50.90, University of Missouri-Columbia Research Reactor, August 2011 (ML 1123A088).

Page 23 of 26

Table 2 - Peaking Factors for Week 58 Fuel Elements 1 and 8 Safety Limit Factors On Enthalpy Rise In Channel 2 Element 1 Element 8 Power History of Fuel Element ..................................... 0 MWd 142 MWd Coolant Channel 2 Gap ............................................... 72 mils 62 mils Power-related Factors Nuclear Peaking Factors Fuel Plate-i and -2 Average ............................... 1.9843 1.5310 Azimuthal in the Channel .................................... 1.0921 1.0542 Additional Allowable Factor .............................. 1.0620 1.0620 Engineering Hot Channel Factors Fuel Content Variation ..................................... 1.0300 1.0300 Fuel Thickness/Width Variation ........................... 1.0300 1.0300 O verall Product ................................................... 2.4416 1.8184 Flow-related Factors Core/Loop Flow Fraction ....................................... 1.0000 1.0000 Assembly Minimum/Average Flow Fraction ................. 1.0000 1.0000 Channel Minimum/Average Flow Fraction Inlet V ariation ................................................ 1.0000 1.0000 W idth V ariation ............................................. 1.0000 1.0000 Thickness Variation ......................................... 0.72/0.80 0.62/0.80 Within Channel Minimum Channel Thickness affect on:

V elocity Factor .............................................. 0.9108 0.8210 Overall Factor on Flow Reduction ................................. 0.8197 0.6363 On Heat Flux From Fuel Plate-I For mesh interval between Power-related Factors the following inches down the fuel plate Mesh Interval Number 18(17-18") 19(18-19")

Nuclear Peaking Factors Fuel Plate (Hot Plate Average) ............................ 2.215 1.5345 Azimuthal Within Plate .................................... 1.070 1.040 A xial Peak .................................................... 1.2958 1.247 Additional Allowable Factor .............................. 1.062 1.062 Engineering Hot Channel Factors Fuel Content Variation ..................................... 1.030 1.030 Fuel Thickness/Width Variation ........................... 1.150 1.150 O verall Product ........................................................ 3.863 2.504 Percentage Enthalpy Rise at Hot Spot 74.8% 78.8%

Energy Fraction Generated in Fuel Plate 93%

Safety Limit at LSSS 14.894 MW 19.69 MW Page 24 of 26

APPENDIX A, TECHNICAL SPECIFICATIONS A.48 TS 5.1, Site Description. Clearly describe the area under the authority of the reactorlicense.

Technical Specification 5.1 has been revised as follows to help clarify thearea under the authority of the reactor license.

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

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

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

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

1Providence Road crosses MU property separating the University Research Park from another MU-owned tract of land lying to the east. The road runs north and south with the closest point of approach being approximately 400 meters east of the reactor facility. MU has the authority to determine all activities including the exclusion or removal of personnel and property and to temporarily secure the flow of traffic on this road during an emergency.

Note: Based on the above revised Technical Specification (TS) 5.1, MURR is requesting that TS Definition 1.6, Exclusion Area, be deleted from the submitted TSs. This term is a holdover from the currently approved TSs and is not used in any operating or emergency procedure.

Additionally, MURR requests that Section 2.1.2, Operational Boundaries, of Chapter 2 of the SAR, be revised as follows to be consistent with the revised TS 5.1.

2.1.2 Operational Boundaries There are three areas of concern regarding the normal operation, safety, and emergency actions associated with the reactor facility: the restricted area within the operations boundary; the unrestricted area within the site boundary; and the Emergency Planning Zone (EPZ).

The operations boundary consists of the fencing at the border of the 7.5 acre lot surrounding the MURR facility itself. The area within this boundary is the restricted, or licensed, Page 25 of 26

area where the Reactor Facility Director has direct authority and control over all activities, normal and emergency. -The adjacent reactor cooling tower building is also included within the restricted area. A tunnel connects the cooling tower building to the laboratory building basement.

There are pre-established evacuation routes and procedures known to personnel frequenting this area. The operations boundary is within the site boundary.

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

In addition, an Emergency Planning Zone (EPZ) has been established for which emergency plans have been developed to ensure that prompt and effective actions can be taken to protect the public in the event of an accident. MURR's EPZ is the area bounded by a 150-meter radius from the reactor facility ventilation exhaust stack and lies completely within the site boundary (Figure 2.3).

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

Page 26 of 26