ML15098A648

University of Missouri at Columbia - Request for Additional Information Regarding the Renewal of Facility Operating License No. R-103 for the University of Missouri at Columbia Research Reactor (TAC No. ME1580)

ML15098A648
Person / Time
Site:	University of Missouri-Columbia
Issue date:	04/17/2015
From:	Wertz G A Research and Test Reactors Licensing Branch
To:	Rhonda Butler Univ of Missouri - Columbia
Wertz G A
References
TAC ME1580
Download: ML15098A648 (11)

Text

April 17, 2015

Mr. Ralph Butler, Director

Research Reactor Center

University of Missouri-Columbia

Research Park

Columbia, MO 65211

SUBJECT:

UNIVERSITY OF MISSOURI AT COLUMBIA - REQUEST FOR ADDITIONAL INFORMATION REGARDING THE RENEWAL OF FACILITY OPERATING

LICENSE NO. R-103 FOR THE UNIVERSITY OF MISSOURI AT COLUMBIA

RESEARCH REACTOR (TAC NO. ME1580)

Dear Mr. Butler:

The U.S. Nuclear Regulatory Commission (NRC) is continuing its review of your application for

the renewal of Facility Operating License No. R-103, dated August 31, 2006 (a redacted version

of the application is available on the NRC's public web site at www.nrc.gov under Agencywide Documents Access and Management System Accession Nos. ML062540114 - cover letter; ML092110573 - Safety Analysis Report, Chapters 1-9; ML092110597 - Safety Analysis Report, Chapters 10-18), as supplemented, for the University of Missouri - Columbia Research Reactor.

During our review, questions have arisen for which additional information is needed. The

enclosed request for additional information (RAI) identifies the additional information needed to

complete our review. We request that you provide responses to the enclosed RAI within 45

days from the date of this letter. If you need additional time to respond, please contact me.

In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.30(b), you must execute your response in a signed original docum ent under oath or affirmation. Your response must be submitted in accordance with 10 CFR 50.4, "Written communications." Information

included in your response that is considered sens itive or proprietary, that you seek to have withheld from the public, must be marked in accordance with 10 CFR 2.390, "Public inspections, exemptions, requests for withholding." Any information related to security should be submitted

in accordance with 10 CFR 73.21, "Protection of Safeguards Information: Performance

Requirements." Following receipt of the additional information, we will continue our evaluation

of your renewal request.

R. Butler

If you have any questions regarding this review, please contact me at (301) 415-0893.

Sincerely,

/RA/ Geoffrey A. Wertz, Project Manager Research and Test Reactors Licensing Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation

Docket No. 50-186

Enclosure:

Request for Additional Information

cc: See next page University of Missouri-Columbia Docket No. 50-186

cc:

John Ernst, Associate Director

Health and Safety

Research Reactor Facility

1513 Research Park Drive

Columbia, MO 65211

Homeland Security Coordinator

Missouri Office of Homeland Security

P.O. Box 749

Jefferson City, MO 65102

Planner, Dept. of Health and Senior Services

Section for Environmental Public Health

930 Wildwood Drive, P.O. Box 570

Jefferson City, MO 65102-0570

Deputy Director for Policy

Department of Natural Resources

1101 Riverside Drive

Fourth Floor East

Jefferson City, MO 65101

A-95 Coordinator

Division of Planning

Office of Administration

P.O. Box 809, State Capitol Building

Jefferson City, MO 65101

Test, Research, and Training

Reactor Newsletter

University of Florida

202 Nuclear Sciences Center

Gainesville, FL 32611

ML15098A648; *concurrence via email NRR-088

Enclosure OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR ADDITIONAL INFORMATION FOR THE RENEWED LICENSE FOR THE UNIVERSITY OF MISSOURI-COLUMBIA RESEARCH REACTOR LICENSE NO. R-103; DOCKET NO. 50-186

The U. S. Nuclear Regulatory Commission (NRC) is continuing its review of your application for

the renewal of Facility Operating License No. R-103, dated August 31, 2006 (a redacted version

of the application is available on the NRC's public web site at www.nrc.gov under Agencywide Documents Access and Management System (A DAMS) Accession Nos.: ML062540114 - cover letter; ML092110573 - Safety Analysis Report (SAR), Chapters 1-9; ML092110597 - SAR, Chapters 10-18), as supplemented, for the University of Missouri - Columbia Research Reactor (MURR). During our review, questions have arisen for which additional information is needed.

The enclosed request for additional information (RAI) identifies the additional information

needed to complete our review. We request that you provide responses to the enclosed RAI

within 45 days from the date of this letter.

1. In the MURR SAR, Sections 1.4.2, 4.2.2.4, and 4.5.3, the control blade drop time is expressed as "insertion to 20% of the withdrawn position in less than 0.7 seconds." SAR

Section 3.5.2 describes the control blade drop process including the effect of the

dashpot, but does not describe the method for determining the drop time nor does it

explain the basis for the 80 percent insertion times. The scram times and reactivity

worths used or assumed for the various analyses in the SAR are not clearly described or

provided. NUREG-1537, Section 4.5.3, "Operating Limits," provides guidance that the

analysis for the shutdown reactivity for all operational conditions should be described. a. Explain the MURR process for determining the control blade insertion times and the associated control blade insertion reactivity per blade. Provide typical control

blade full insertion scram times and reactivities, or justify why no additional

information is needed. b. Explain which analyses documented in the SAR utilize the assumptions described in Item a. above regarding control blade insertions, withdrawals, and

scrams (e.g., blade withdrawal from subcritical, control blade run in, insertion of

excess reactivity, etc.). For each such event, provide the control blade motion

speeds and reactivities utilized to provide the SAR analyses, or justify why no

additional information is needed.

2. NUREG-1537, Section 9.2, "Handling and Storage of Reactor Fuel", provides guidance that the licensee provide analyses and methods to demonstrate the secure storage of

new and irradiated fuel with a criticality limit of keff <

0.90. The NRC staff's review of the MURR SAR and Hazards Summary Report could not find a criticality analysis supporting

the use of any fuel storage locations outside of the core. Identify the locations that may

be used for the storage of new or irradiate fuel, and provide supporting criticality

analyses, or justify why no additional information is needed.

3. NUREG-1537, Section 4.5.1, "Normal Operating Conditions," and Section 4.5.2, "Reactor Core Physics Parameters," provide guidance that the licensee should identify

their analytical methods, including calculations of individual control blade worths, core

excess reactivity, and coefficients of reactivity, and compare the results with

experimental measurements. The MURR SAR, Section 4.5 states that analyses have been performed using PDQ, EXTREMINATOR, and BOLD-VENTURE codes using R, RZ, and RZ models. The NRC staff noted other analyses (e.g., the RAI responses supporting the NRC staff review of License Amendment No. 36, ADAMS Accession Nos.

ML11237A088 and ML12150A052) used Estimated Critical Position (ECP) comparisons

with the Monte Carlo Neutron Production code. The design code used to support the

T&H analysis appears to be DIF3D. The NRC staff is not clear as to which analytical

method is the final supporting analysis to be reviewed for the MURR license renewal

application. The final supporting analysis should be the source for information used in

accident and event analysis (e.g., peaking factors, control blade worths). Furthermore, in response to RAI 4-14.c., (ADAMS Accession No. ML103060021), it is not clear how

the stuck control blade was determined, what the relative reactivity worth is for the other

control blades in the shutdown margin (SDM) analysis, and whether they are calculated, measured, or compared. The following information is needed:

a. Identify the neutronics code used as the basis for the MURR License Renewal Application, or justify why this information is not needed. b. Using results from that code provide the results of calculations and comparisons of the corresponding measurements for the ECP (or excess reactivity) for a

known critical control blade configuration at zero power, no xenon condition, or

justify why this information is not needed.

c. Provide calculated and measured control blade worths (Shim-1, Shim-2, Shim-3, Shim-4, and Regulating blades) for a given core configuration at a low power, no

xenon condition, or justify why this information is not needed.

d. Provide a calculated and measured temperature coefficient for a given core configuration at a low power, no xenon condition, or justify why this information is

not needed.

4. NUREG-1537, Section 4.5.3, "Operating Limits," provides guidance that licensees demonstrate that their facility has sufficient control blade worth to achieve the required

shutdown reactivity assuming that all scrammable control blades are released upon

scram, but the most reactive blade remains in its most reactive position. The NRC staff

could not find this information in the MURR SAR, but noted a reference in the 1971 Low

Power Testing Program that indicated that the shutdown margin control blade reactivity

was determined using 66 percent of the 4 shim blade insertion worth. Explain how

MURR ensures adequate SDM, whether and if so, how the 66 percent factor from the

1971 Low Power Testing Program is used, or justify why this information is not needed.

5. NUREG-1537, Section 11.1.1.1, "Airborne Radiation Sources," provides guidance for the licensee to characterize the dose for the maximally exposed individual, at the location of the nearest permanent residence, and at any locations of special interest in the unrestricted area.

a. The MURR SAR, Appendix B, contains summary information regarding the radiological impacts of the MURR generated release of Argon 41 (Ar-41) during

normal operations. The MURR methodology includes an equation on SAR page

B-10 that is used to alter the effective stack height used in the dose calculations

to compensate for elevation changes of the receptor due to the local topography.

Although unreferenced in the SAR, the NRC staff reviewed "Plume Rise" by

Briggs (TID-25075) and it seems that this equation is based on the Davidson

empirical model which has limited supporting data. Describe how the effective

stack height calculations are performed for the unique topography surrounding

MURR, and how the results are sufficiently conservative for the estimation of

dose, or justify why no additional information is needed.

b. SAR page B-11 has an equation for /Q that includes the y and z dispersion factors. The NRC staff was unable to validate some of the dispersion values

used in Tables B-2 and B-3. Explain how these values were determined or justify

why no additional information is needed.

6. NUREG-1537, Section 13, provides guidance that the applicant should demonstrate that the facility design features, safety limits, limiting safety system settings, and limiting conditions for operation have been selected to ensure that no credible accident could

lead to unacceptable radiological consequences to people or the environment. The

NRC staff review examined the analyses provided in the MURR SAR, Chapter 13, including the assumptions regarding the initial conditions (e.g., reactor power, reactivity

insertion, etc.), analytical input (e.g., peaking factors and decay times), and results. The

following information is needed:

a. Regarding Insertion of Excess Reactivity - The initial power is 10 MW rather than the Limiting Safety System Setting setpoint in TS 2.2 (12.5 MW). The

temperature feedback coefficient used is -7.0x10

-5 k/k rather than the TS 5.3.a value of -6x10

-5 k/k. It is unclear what peaking factors are employed. SAR Figure 13.2 seems to indicate that the scram time used is faster than the value in

TS 3.2.c. The acceptability of the results is based upon whether the power for

burnout is achieved rather than the safety limit identified in TS 2.1. Provide

additional information justifying and supporting the analysis and the safety

conclusions or provide a justification for why such information is not required.

b. Regarding Loss of Primary Coolant and Loss of Primary Coolant Flow - The initial power is 11 MW rather than the LSSS setpoint in TS 2.2 (12.5 MW). It is

unclear what peaking factors are employed. The acceptability of the results is

based upon the peak fuel temperature attained rather than the safety limit

identified in TS 2.1. Provide additional information justifying and supporting the

analysis and the safety conclusions or provide a justification for why such

information is not required.

c. Regarding the maximum hypothetical accident (MHA) and Failed Fueled Experiment - these events use a 10 minute and 2 minute evacuation time respectively. Provide additional information identifying the limiting evacuation time and then use that time to justify and support the analysis and the safety

conclusions or provide a justification for why such information is not required.

7. NUREG-1537, Section 13.1.1, "Maximum Hy pothetical Accident," provides guidance for the licensee to postulate a failed fuel element scenario and analyze the consequences.

The MURR SAR, Section 13.2.1.2, provides the analysis and related consequences for

a fuel failure involving the melting of four number 1 fuel plates in a core region where the

power is at a maximum. The fuel fails submerged and it is assumed that all iodine, krypton, and xenon isotopes are released into the primary coolant system (PCS) while in

Modes I or II (PCS closed).

a. The iodine and noble gases core inventories are based on a 1200 MWD burnup consisting of twelve 10-day cycles over a 300-day period. These values were

then adjusted using a peaking factor of 1.6. However, in the response to RAI

A.27 (ADAMS Accession No. ML120050315), a peaking factor of 3.0 has been

used. In the MURR SAR, Section 4.5, the peaking factor is listed as 3.676.

Clarify the discrepancies in the peaking factors used, and provide a revised

calculation of the source using the peaking factors determined from the final

analysis, or justify why no additional information is needed.

b. The release is assumed to occur into the PCS with a volume of 2,000 gallons.

Identify what components comprise this volume and provide information to

confirm the 2,000 gallon volume assumption, or justify why no additional

information is needed.

c. The release is assumed to remain in the PCS except for the amount that will enter the pool cooling system as part of the PCS to pool cooling system leakage.

Therefore, the concentration of iodine that is released first enters the pool cooling

system and is diluted once again. This seems to reduce the consequences of

this accident to a fraction of the consequences of the failed fueled experiment as

provided in your response to RAI 13.9 (ADAMS Accession No.

ML103060018).

As such, this event (four failed fuel plates) may not be the MHA. Provide a

confirmation of the dilution assumptions stated above and clarification as to the

MHA for MURR.

d. The released concentrations in the containment are based on the 10-minute leakage between the PCS and the pool cooling system. However, the NRC staff

questions whether the release into the PCS will collect in the vent tanks and

other places in the PCS and eventually be released to the environment after

decay. Provide an explanation for this leakage path, including assumptions and

calculations of the possibility of the isotopic concentrations being released to the

environment, or justify why no additional information is needed.

e. In determining the offsite doses in the unrestricted areas from the releases, the concentrations of the released isotopes are calculated using a method described

in the MURR SAR, Appendix B, which used a simplified joint frequency

distribution of weather data that was prepared in the 1960s. Given the changes

in weather conditions over the last 50 years, it is not clear to the NRC staff whether the listed probabilities and wind speeds for the stability classes are still applicable. Provide available current weather data, and state whether changes

warrant reconsideration of the cited data, or justify why no additional information

is needed.

f. It is not clear to the NRC staff which dispersion factors were used to arrive at the listed concentrations in the cited unrestricted location, which is also not specified.

The calculation of the ratio of the average concentration in the unrestricted

location to the corresponding concentration in containment results in the

reduction factor for iodine twice as large as the value for the noble gases. For

example, for Krypton-85 the ratio is 7.5x10

-14/3.0x10-8 or a reduction of about 4.0x10 5. For I-131, the ratio is 1.36x10

-14/1.1x10-8 or a reduction of about 8.1x10 4. Provide an explanation of all assumptions relating to the calculation of average isotope concentrations, specify all locations where these concentrations

are determined, and explain how dispersion factors are determined and used, or

justify why no additional information is needed.

g. In determining occupational doses, it appears that the MURR SAR calculations use a combination of dose conversion factors (DCFs). It appears that for

radioiodine, the calculation uses DCFs from Federal Guidance Report (FGR)

No. 11 for inhalation pathway (thyroid) and FGR No. 12 for submersion dose (external-deep-dose), whereas for submersion doses from noble gases, it uses

the derived air concentrations from 10 CFR Part 20, Appendix B, Table 1. FGR

12 revises the dose coefficients for air submersion used in FGR 11. Those DAC

values are based on International Commission on Radiation Protection (ICRP)-2

DCFs, whereas the FGR 11 values are based on ICRP-38. In addition, neither

FGR 11 nor FGR 12 lists DCFs for isotopes with very short-half lives. In

10 CFR Part 20, Appendix B Table 1, the regulation provides a DAC value of

1x10-7 micro-Ci/ml for those isotopes with a half-life of less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Overall, the differences in the calculated DCFs result in high values of calculated

doses from noble gas isotopes with a very short half-life. Provide dose

calculations using uniform data and methodology.

8. NUREG-1537, Section 13.1.3, "Loss of Coolant," provides guidance to the licensee to consider the consequences of a loss of coolant accident (LOCA). MURR SAR Section

13.2.3.2 describes the LOCA event for the loss of the PCS integrity, and states that the

accident of greatest consequence is a rupture in the short section of the PCS piping (either the cold leg or the hot leg) between the reactor pool and either isolation valves

(507B or 507A). The SAR describes the consequences of a cold leg break between the

isolation valve 507B and the reactor pool in significant detail. The hot leg break

discussion is more succinct. The SAR also states that how "the anti-siphon system

ensures that the core remains covered differs depending on the location of the rupture."

The NRC staff reviewed the event as described in the SAR and is considering the hot

leg break sequence. It is our understanding that after isolation occurs the coolant

surrounding the core heats up, and because of natural buoyancy it flows upward and out

of the reactor pressure vessel into the in-pool heat exchanger. After passing through the

heat exchanger, the cooled water may then flow downward through what is normally the

upward flow path of the inverted loop and then into the bottom of the pressure vessel.

As this process continues, the water will fill up the downward inverted loop to the bottom of the core reaching to the inverted loop creating an open condition for releasing the

PCS coolant through the broken hot leg pipe. Explain the credibility of this event, and, if

credible, provide a supporting analysis demonstrating acceptable core cooling and peak

fuel temperatures, or justify why no additional information is needed.

9. NUREG-1537, Section 13.1.5, "Mishandling or Malfunction of Fuel" provides guidance that the licensee analyze the consequences of a mishandled fuel event. MURR SAR

Section 13.2.5.2.1 describes damage to a fuel element due to mishandling. It states that

the mishandling could occur during movement and packaging of the irradiated fuel, damage could only occur to the inner or the outer fuel plate, and could only occur during

fuel element relocation activities. Because this accident occurs while the PCS is open

there is minimal containment of fission products by the PCS. The response to RAI A.27 (ADAMS Accession No.

ML120050315), provides an analysis of such an occurrence assuming that the fuel element has decayed for 60 days as part of the spent fuel

movement from storage to a shipping container. However, the NRC staff questions

whether this event could also occur during the initial stages of refueling which would

invalidate the assumption of 60 days of decay. The NRC staff also performed a

confirmatory calculation based on this inventory using the cited values for the MHA

analysis, and it results in an inventory that is seven percent larger than reported by

MURR. a. Explain the possibility of this event occurring during the initial stages of refueling, and the applicability of using the stated decay time in the dose calculation. Also, describe any radioactivity release alarms that are expected to actuate, and

whether containment isolation is expected, including the time required to verify

containment isolation, or justify why no additional information is needed.

b. Provide the details of how the source term is determined, or justify why no additional information is needed.

10. NUREG-1537, Section 13.1.6, "Experiment Malfunction" provides guidance that the licensees analyze the consequences of a failed fueled experiment. SAR Section

13.2.6.2 describes that limiting fueled experiments to 150 curies of radioiodine will result

in a projected dose well within the limits of 10 CFR Part 20. The response to RAI 13.9.a (ADAMS Accession No.

ML103060018) provides radioiodine and noble gas activities for a 5-gram low-enriched fuel target. The response uses a method similar to that used in

the MHA analysis and lists the gaseous fission products to be released into the pool

cooling system. The occupational dose calculation assumes a 2-minute evacuation

time. The NRC staff notes that the submersion dose calculations were performed using

the DAC values, but the DAC data for isotopes with half-lives of less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> that

are not listed in Table 1 of Appendix B are not consistent with the recommended value of

1x10-7 Ci/ml. The NRC staff notes that the 2-minute evacuation time is not consistent with the 10-minute evacuation time assumed in the MHA analysis, or the SAR Section

13.2.1.2 statement that it takes the operations staff approximately 5 minutes to secure

the PCS and verify containment isolation following a containment isolation signal.

a. Please clarify the sequence of events, state which alarms are expected to provide indication that evacuation is required, justify the evacuation time, and use that time to revise the dose assessment employing consistent DAC values, or justify why no additional information is needed.

b. SAR Section 13.2.6.2 states that "Fueled experiments containing inventories of Iodine-131 through Iodine-135 greater than 1.5 curies or Strontium-90 greater

than 5 millicuries shall be vented to the facility ventilation exhaust stack through

high efficiency particulate air and charcoal filters which are continuously

monitored for an increase in radiation levels." This is inconsistent with TS 3.8.o

which states that a fueled experiment can be encapsulated or vented. Clarify

whether fueled experiments are vented or not and revise the TS if required, or

justify why no additional information is needed.

c. If such venting is permitted then explain why those contributions are not included in the inventory of normally released material (such as Ar-41), or justify why no

additional information is needed.