ML15275A314

University of Missouri, Columbia - Responses to NRC Request for Additional Information, Dated April 17, 2015, Regarding Renewal Request for Amended Facility Operating License

ML15275A314
Person / Time
Site:	University of Missouri-Columbia
Issue date:	10/01/2015
From:	Rhonda Butler, Fruits J Univ of Missouri - Columbia
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML15275A281	List: ML15275A313 ML15275A314
References
TAC ME1580
Download: ML15275A314 (202)

Text

UNWVERSITY of MISSOURI RESEARCH REACTOR CENTER October 1, 2015 U.S. Nuclear Regulatory Commission Attention:

Document Control Desk Mail Station P 1-37 Washington, DC 20555-000 1

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 Regarding the Renewal of Facility Operating License No. R-l103 for the University of.Missouri at Columbia Research Reactor (TACNo. ME1580)," dated April 17, 2015 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, MUJRR 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), MIURR 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, MVURR 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, MURK requested additional time to respond to the remaining sixty-seven (67) 45-Day Response Questions.

On September 30, 2010, MURR responded to sixteen (16) of the remaining 45-Day Questions.

By letter dated October 13, 2010, the NRC granted the extension request.By letter dated October 29, 2010, MURR responded to sixteen (16) of the remaining 45-Day Response Questions and two (2) of the remaining Complex Questions.

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

On December 1, 2010, via email, MURR requested additional time to respond to the remaining 45-Day Response and Complex Questions.

By letter dated December 13, 2010, the NRC granted the extension request.On January 14, 2011, via email, MURK 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 45-Day Response and Complex Questions.

By letter dated July 5, 2011, the NRC granted the request.By letter dated September 8, 2011, MUIIRR responded to six (6) of the remaining 45-Day Response and Complex Questions.

On September 30, 2011, via email, MURR requested additional time to respond to the remaining the remaining 45-Day Response and Complex Questions.

By letter dated November 10, 2011, the N-RC granted the request.By letter dated January 6, 2012, MURK responded to four (4) of the remaining 45-Day Response and Complex Questions.

Also submitted was an updated version of the MUJRR Technical Specifications.

On January 23, 2012, via email, MUJRR requested additional time to respond to the remaining the remaining 45-Day Response and Complex Questions.

By letter dated January 26, 2012, the NRC granted the request.On April 12, 2012, via email, MURR requested additional time to respond to the remaining the remaining 45-Day Response and Complex Questions.

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

With that set of responses, all 45-Day Response and Complex Questions had been addressed.

2 of 86 On December 20, 2012, the NRC requested a copy of the current Physical Security Plan (PSP) and Operator Requalification Program.By letter dated January 4, 2013, MURR provided the NRC a copy of the current PSP and Operator Requalification Program.On February 11, 2013, the NRC requested updated financial information in the form of four (4) questions because the information provided by the September 14, 2009 response had become outdated.By letter dated March 12, 2013, MUIRR responded to the four (4) questions.

On December 3, 2014, the NRC requested additional information in the form of two (2) questions regarding significant changes to the MIURR facility since submittal of the licensing renewal application in August 2006.By letter dated January 28, 2015, MvUIRR responded to the two (2) questions.

On April 17, 2015, the NRC requested additional information in the form of ten (10) questions.

On May 29, 2015, via email, MUJRR requested additional time to respond to the ten (10) questions.

On June 18, 2015, the NRC requested additional information in the form of two (2) questions.

By letter dated July 31, 2015, MUIRR responded to the two (2) questions from the June 18, 2015 request.On September 14, 2015, via telephone, the NRC requested a copy of the Emergency Plan (EP).By letter dated September 14, 2015, the NRC requested additional information in the form of sixteen (16)questions regarding the PSP.By letter dated September 15, 2015, MURR provided the NRC a copy of the current EP.Attached are responses to the April 17, 2015, request for additional information, which were in the form often (10) questions.

• If there are any 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.3 of 86 ENDORSEMENT:

Sincerely, Reviewed and Approved, John L. Fruits Ralph A. Butler, P.E.Reactor Manager Director xc: Reactor Advisory Committee Reactor Safety Subcommittee Dr. Garnett S. Stokes, Provost Dr. Henry C. Foley, Senior Vice Chancellor for Research Mr. Alexander Adams Jr., U.S. Nuclear Regulatory Commission Mr. Geoffrey Wertz, U.S. Nuclear Regulatory Commission Mr. Johnny Eads, U.S. Nuclear Regulatory Commission Attachments:

1. MURR Drawing No. 1905, Sheet 1 of 1, "Control Blade Drop Timer Circuit" 2. Modification Record 72-7, "'Additional In-Pool Fuel Storage Basket" 3. Modification Record 76-3, "Upper Z Spent Fuel Storage" 4. Modification Record 76-3, Revision, "'Spent Fuel Storage" 5. Modification Record 9 1-3, "Temporary Additional In-Pool Fuel Storage Baskets" 6. Modification Record 91-3, Addendum 1, ""Replacement of the Existing X, Y, MIH-X, and MH-Y Fuel Storage Baskets With New X and Y Baskets" 7. Volume of the Primary Coolant System 8. Meteorological Data (Wind Speed and Class) -1961 to 1969 9. Meteorological Data (Wind Speed and Class) -1970 to 1990 10. Meteorological Data (Wind Speed and Class) -1961 to 1990 11. 10 CFR 835, Appendix C, "Derived Air Concentration (DAC) for Workers from External Exposure during Immersion in a Cloud of Airborne Radioactive Material" 12. Micro Shield 8.02 Dose Calculations for a Fuel Handling, Fuel Failure, and Fueled Experiment Failure Accidents 13. Stack Effluent Releases -Calendar Years 2005 to 2014 JACQUELINE L.BOHM '0 "'.STATE OF MISSOURI-MY Commission

26. 2019Comisson#/

u Much ( 1,- "9 ' /4 of 86

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-15 3 7, 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 justfif' why no additional information is needed.Control blade insertion times are determined by a Control Blade Drop Timer Circuit (see Attachment 1). When a reactor scram signal is initiated, the control current to the electromagnet, which engages the control rod drive mechanism (CRDM) to the anvil of the control blade-lift rod assembly, is removed by an electro-mechanical relay contact which allows the control blade to drop and start a blade drop timer/chronometer count. At the 20% withdrawn position (or 80% inserted), a digital fiber optic sensor, which provides a NPN (Not Pointing In) output to the control unit when triggered, causes the electro-mechanical relay to change state stopping the blade drop timer/chronometer.

The control blade drop time is then displayed on a meter on the reactor control room instrument panel. Table 1 provides the minimum, average and maximum drop times of all four (4) shim control blades for the years 2010 to 2014.Table 1 -Control Blade Drop Times (Years 2010 to 2014)Time Control Blade (In Seconds) 'A' 'B' 'C' 'D'Minimum 0.46 0.49 0.45 0.48 Average 0.50 0.54 0.50 0.52 Maximum 0.59 0.58 0.54 0.54 Current MURR Technical Specification 3 .2.c requires the capability of inserting the shim control blades to their 20% withdrawn position (or 80% inserted) in less than 0.7 seconds. This ensures prompt shutdown of the reactor in the event a reactor scram signal, manual or automatic, is received.

The 20% withdrawn position is defined as 20% of the control blade full travel of 26 inches measured from the fully inserted position.

Below the 20% withdrawn position the control blade fall is cushioned by a dashpot assembly.

Approximately 91% of the control blade total worth is inserted at the 20% position.

This is an original design feature of the reactor and its purpose has not been altered in 49 years of operation.

The same Technical Specification will remain in the relicensing Technical Specifications.

The measured and calculated values for reactor core excess reactivity and shutdown margin are provided below to demonstrate the safe shutdown capability with only three (3) out of the four (4)5 of 86 shim control blades inserted to their 20% withdrawn position (also assumes the regulating blade is fully withdrawn).

Some of this information, calculated using older computer programs, can also be found on Table 4-12 of the SAR.Typical MURR operations involve a core change-out every week with eight (8) xenon-free fuel elements in various stages of burnup (mixed core operation) used at startup. The reactor core excess reactivity and shutdown margin values are verified after the weekly core change-out.

The verification is done during reactor startup, when the cold, clean critical control blade height is measured.

This critical control blade position, along with the known integral control blade worth, is used to estimate reactor core excess reactivity.

Measured Values: Table 2 provides the measured values of shim control blade worth, reactor core excess reactivity and shutdown margin in comparison to the Technical Specification limit of -0.020 Ak/k.Table 2 -Summary of Key Measured Reactor Data Value Paraeter(Ak/k)

Typial ota shi cotro blae wrth0.1364 Typical total shim control blade worth at 80% inserted -0.1127 Typical shim control blade worth at 80% inserted with the highest worth -0.0787 control blade excluded (or fully withdrawn)

Maximum reactor core excess reactivity after weekly core change-out

+0.0400 One-year average of reactor core excess reactivity (over 69 core change-outs)

+0.0290 Typical core sub-criticality with 3 shim control blades at 80% inserted and the -038control blade excluded (or fully withdrawn)-037 Minimum shutdown margin allowed by Technical Specifications

-0.0200 Calculated Values: Reactor core excess reactivity and shutdown margin values were also calculated using the detailed MGNP MIIURR core models. Two separate cases were considered for the MCNP calculations:

(1)using all fresh fuel elements (license possession limit only allows 6 fresh fuel elements onsite) and all fresh shim control blades (most conservative), and (2) with a mixed core loading and mixed burnup control blades (typical MUIRR operation).

Table 3 provides the calculated values.6 of 86 Table 3 -Summary of Key Calculated Reactor Data Value Parameter

-All Fresh Fuel and Fresh Control Blades Case (kk Reactor core excess reactivity 0.0865 Total shim control blade worth 0.1740 Core sub-criticality with 3 shim control blades at 80% inserted and highest -0.0324 worth control blade excluded (or stuck fully withdrawn)

Value Parameter

-Mixed Core / Mixed Control Blades Case (Ak/k)Reactor core excess reactivity 0.0445 Total shim control blade worth 0.1517 Core sub-cniticality with 3 shim control blades at 80% inserted and highest -0.0580 worth control blade excluded (or stuck fully withdrawn)

The measured and calculated values for reactor core shutdown margin show that even with three (3)shim control blades at their 20% withdrawn position (and the regulating blade and highest worth shim control blade fully withdrawn), the minimum reactor core shutdown margin required by the Technical Specifications is easily satisfied.

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.The RELAP code is used to perform the accident analyses of the Loss of Coolant Accident (LOCA)and the Loss of Flow Accident (LOFA). The two (2) LOCA analyses determine what would occur if there were a double-ended shear of the 12-inch primary coolant piping on both sides of either the cold-leg isolation valve V507B or the hot-leg isolation valve V507A. To envelope the LOFA, five (5) different scenarios were analyzed.

The inadvertent loss of pressurizer pressure was found to be the worst-case accident so it is the one described in the SAR.In the RELAP analyses, key reactor coolant parameters that are monitored by reactor safety system instrumentation can have trip values set for them at the appropriate coolant loop locations.

In the RELAP modeling, a 150 millisecond time delay is set between the time a scram signal is received and the modeling of when the "insertion" of the control blades start. The insertion is covered by an input table of fission and gamma reactor power as per set time steps after the reactor scrams. The code calculates linear values between these data points.7 of 86 Table 4 below provides the power assumed by RELAP seconds after shutdown compared to the calculated power after shutdown, assuming 30 days of full power operation, using equation 2.66 from Nuclear Reactor Engineering 3 rd Edition by Samuel Glasstone and Alexander Sesonske'.

The equation is given in the upper right corner of the page along with the values of variables a and b to use depending on which time step after shutdown the decay power applies. During the first ten (10)seconds, the RELAP values are very conservative and more than double the calculated decay power except for the values for 8, 9 and 10 seconds. From 10 to 150 seconds, the RELAP values are conservative by 17%. From 180 seconds to 10,000 seconds, the RELAiP values average being 3.8%more conservative than the equation calculated values. Therefore, the RELAP analyses use conservative calculated values of reactor decay heat after the scram, which would correspond to slower insertion of the control blades.See the response to RAI 6.a for control blade drop times related to Insertion of Excess Reactivity accidents.

References:

1 Glasstone, S. and Sesonske, A., Nuclear Reactor Engineering 3 rd Edition, prepared under Technical Information Center, United States Department of Energy.8 of 86 Table 4 -Comparing RELAP Decay Heat to Calculated Decay Heat (Nuclear Reactor Engineering 3 rd Edition: Equation 2.66)ts 0 0.1 0.3 0.7 1.0 2.0 3.0 5.0 6.0 7.0 8.0 9.0 10 20 30 40 50 60 70 80 90 100 120 150 180 200 240 300 400 420 540 600 800 1,000 2,000 4,000 6,000 8,000 10,000 Power MW 11.0 A 7.9906 1.97 17 1.2597 1.0373 0.9309 0.848 1 0.7345 0.693 1 0.6576 0.6292 0.6044 0.5819 0.4506 0.4100 0.3869 0.3698 0.3563 0.3453 0.3360 0.3280 0.3210 0.3090 A 0.3072 0.3053 0.3015 0.2849 0.2659 0.243 1 0.2395 0.2214 0.2142 0.1957 0.1824 0.1462 0.1167 0. 1020 0.0927 0.0860 Power MW 11.0 0.5099 0.4578 0.4201 0.4048 0.376 1 0.3599 0.3400 0.333 1 0.3273 0.3223 0.3 180 0.3 141 Equation 2.66 P/P 0=5E-3 *a *[t 4-b -(To + t4)-~b]

after shutdown To = 30 days operating period prior to shutdown (s)0.1 to 10 10 to 150 150 to 8E8 0.4970 0.43 17 0.3970 0.3740 0.3569 0.3434 0.3324 0.323 1 0.3 150 0.3080 0.296 1 0.2821 a 12.05 15.31 27.43 b 0.0639 0.1807 0.2962 0.3230 0.3050 0.295 1 0.2786 0.2595 0.2368 0.233 1 0.2150 0.2078 0.1893 0. 1760 0.1398 0.1103 0.0957 0.0863 0.0796 Note A: RELAP does not have a value entered for 0.1 seconds, but the linear value between 0 and 0.3 seconds is 7.9906. Value for 150 seconds is linear between 120 and 180 seconds.9 of 86

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.As stated in SAR Section 9.2.1, there are 88 in-pool storage locations for new or irradiated fuel elements.

These storage locations are situated in three (3) areas within the reactor pool and are designated as the "X," "Y" and "Z" storage baskets. The "Z" storage basket contains 48 fuel element storage locations; consisting of two (2) levels, referred to as "upper" and "lower," of 24 locations per level. The "X" and "Y" storage baskets each contain 20 fuel element storage locations.

There are eight (8) storage locations for new, fresh fuel elements in the fuel vault.The MUIRR facility was originally designed and built with only 28 in-pool fuel element storage locations.

The "X" and "Y" storage baskets each had only six (6) storage locations at the time while the "Z" storage basket consisted of 16 storage locations

-two racks (6 and 10) in the lower level. In 1972, due to an increase in operating schedule and with an uprate in power from 5 to 10 MWs in the near future, an additional rack of eight (8) storage locations was added to the lower level of the "Z" basket, thus providing a total of 36 fuel element storage locations in the pool (24 in the "Z" basket). Modification Record 72-7, "Additional In-Pool Fuel Storage Basket," documents the installation of the eight (8) element rack (Attachment 2). On page 2a of the Modification Record, the following is stated: "To determine the safety of installing an additional fuel rack between the present two, the system was modeled using the Exterminator II multi-group neutron diffusion program. The physical model consisted of three adjacent rows of eight clean 775 gram U 2 3 5 fuel elements.

Each fuel element was surrounded by 0.25" thick boral as is the case in the actual design. For the fully loaded rack, the calculated Keff linlit was 0. 714. "A 1/M criticality plot of the storage basket was also performed to verifyi the Exterminator II code results.In 1976, a 14 element rack was added to the upper level of the "Z" storage basket which increased the overall capacity of the "Z" storage basket from 24 to 38. Modification Record 76-3, "Upper 7 Spent Fuel Storage," documented the installation of the additional 14 fuel element storage locations (Attachment 3). On page 4 of the Modification Record, the following is stated: "The addition of another level of elements was modelled using the Exterminator II neutron diffusion code. The presence of 24 rather than 14 elements on the second level was used for a 'factor of safety." The code predicts a value for Keff of 0.748. Thus, the above criteria is satisfied for fuel storage." A l/M criticality plot of the storage basket was also performed to verify the Exterminator II code results.In 1978, a 10 element rack was added to the upper level of the "Z" storage basket which increased the overall capacity of the "Z" storage basket from 38 to 48. Modification Record 76-3, Revision,"Spent Fuel Storage," documents the installation of the additional 10 fuel element storage locations (Attachment 4). A 1/M plot criticality was also performed to verify the Exterminator II code results 10 of 86 stated in Modification Record 76-3, which conservatively modeled 24 fuel elements instead of just 14 elements.Because the criticality analyses for the "Z" storage basket are somewhat dated and vaguely documented, MUJRR performed an updated criticality analysis of the upper and lower levels of the"Z" storage basket using the general-purpose Monte Carlo N-Particle (MCNP) code. The following describes the methodology and results.The "Z" storage basket stores fuel elements that have burnups of 0 to 150 MWds. The baskets are lined with 26- to 29-inch tall sheets of 0.25- to 0.3125-inch thick BaC (BORAL) as the absorbing material to prevent the stored fuel configuration from reaching criticality.

Figure 1 shows the layout (i.e. a detailed MCNP model) of the lower "Z" storage basket configuration.-Stainless Steel-Fuel Element-Pool Water BORALFigure 1 -Detailed MCNP model of the Lower "Z" Storage Basket Configuration The upper 'Z' storage basket configuration layout shown in Figure 2 is very similar to the lower basket with the exception of lead shields surrounding the basket instead of stainless steel, as in the lower basket.* Lead Shield Figure 2 -Detailed MCNP model of the Upper "Z" Storage Basket Configuration (Lead shields instead of stainless steel)11 of 86 v The active region of the fuel elements in the lower and upper baskets is separated in height by approximately seven (7) inches. Each fuel element in every storage location is modeled in full detail, with all 24 aluminum clad UAlx fuel plates. Figure 3 shows very detailed MCNP modeling of an individual MURR fuel element and the elements in their lower and upper "Z" storage basket configurations.

Upper Level of "Z" Storage Basket Lower Level of "Z" Storage Basket Figure 3 -Panels Showing Detailed MCNP Modeling of the Fuel Elements; the Left Panel Showing the Axial Configuration of the Fuel Elements in the Lower and Upper "Z" Storage Baskets and the Right Panel Showing a Cross-sectional View of a MURR Fuel Element Criticality (i.e. KCODE) calculations using MCNP version 5 with the ENDJF/B-VII.O data libraries were performed for two detailed instances of the "Z" storage basket configuration:

(1) a single basket (lower), and (2) both lower and upper baskets together.

All calculations were performed for 20 million source particles.

For the two instances, the basket(s) were filled to their maximum capacities (24 fuel elements) with fresh, highly-enriched uranium (HiEU) UAlx MURR fuel elements.

These configurations describe the most conservative, worst-case conditions for the "Z" storage baskets. Table 1 provides the computed using the MCNP models of the two configurations of the "Z" storage basket.Table 1 -KIf Values for Worst-Case "Z' Storage Basket Configurations Configuration Fuel Status Storage Capacity Ief Lower Fresh Max -24 Fuel Elements 0.49885 Lower + Upper Fresh Max -48 Fuel Elements 0.55862 12 of 86 On receipt, fresh (i.e., un-irradiated fuel) fuel elements may be stored outside the reactor pool in a dry, vaulted location.

The elements are stored separately in a plywood rack filled with (powered)boric acid to prevent reaching criticality.

Figure 4 shows a detailed MCNP model of the dry storage configuration containing the maximum allowable number of on-site stored fresh MURR fuel elements (i.e., six fuel elements).

Note: Amended Facility License No. R-103, Section 2.B.(2), states, ".. .to receive, posses, and use up to 60 kilograms of contained uranium-235 of any enrichment, providing that no more than 5 kilograms of this amount is unirradiated;...".

Six MURR fuel elements, containing 775 grams of uranium-235 each, equals 4.65 kilograms.

Air--Fuel Element Plywood Boric Acid Figure 4 -Detailed MCNP Model Showing Fresh Fuel Dry Storage Filled to Possession Limit To establish a full-scope criticality safety study, in addition to the configuration described in Figure 4, two other configurations were also defined to capture the worst-case scenarios:

(1) a flooded configuration storing the maximum allowable number of on-site stored fresh fuel, i.e., six fresh fuel elements (see Figure 4 where air is replaced with water), and (2) a flooded configuration with the rack filled to its maximum capacity which equals eight fresh fuel elements (see Figure 5).,...Water Fuel Element Plywood Boric Acid Figure 5 -Detailed MCNP Model Showing Fresh Fuel Dry Storage Filled to Physical Capacity 13 of 86 Again, the criticality (i.e. KCODE) calculations performed using MCNP version 5 with the ENDF/B-VII.0 data libraries.

All calculations were performed for 20 million source particles.

The computed K~f f using the MCNP models are reported in Table 2 for all three instances of most conservative, worst-cases (in terms of attaining criticality) for the fresh fuel storage configurations.

Table 2 -K~ff Values for Worst-Case Fresh Fuel Storage Configurations Configuration Fuel Status Storage Capacity Kf Dry (Air) Fresh License Max -6 Fuel Elements 0.02344 Flooded Fresh License Max -6 Fuel Elements 0.36228 Flooded Fresh Storage Max -8 Fuel Elements 0.36258 In 1991, due to the inability to ship spent fuel from the facility because the cask (GE-700) that was used to ship research reactor fuel was removed from service, two (2) new fuel storage baskets were fabricated to increase the onsite storage capacity.

These baskets, which were attached to the "X" and "Y" storage baskets, each held 12 fuel elements and were designated "MH-X" and "MH-Y." Modification Record 91-3, "Temporary Additional In-Pool Fuel Storage Baskets," documented the installation of the additional 24 fuel element storage locations (Attachment 5). On page 2 of the Modification Record, the following is stated: "The evaluation performed for each MHJA basket will include a criticality analysis (KENO), a boral plate verification, thermal analysis and J/M" determination when it is first loaded." In 2004, the "X," "Y," "MH-X" and "MH-Y" fuels storage baskets were replaced with new"X and "Y"' storage baskets, which increased the total storage capacity in these baskets from 36 to 40 locations.

Modification Record 91-3, Addendum 1, "Replacement of the Existing X, Y, MH-X, and MH-Y Fuel Storage Baskets with New X and Y Baskets," documents the installation of the new "X" and "Y" storage baskets (Attachment 6). This Modification Record contains a detailed description of the criticality analysis performed for these two baskets using the MCNP code. On page 4 of the Modification Record, the following is stated: "The MCNP model was used to calculate a Ke,'7 value of 0. 635 for one fuel basket fully loaded with twenty (20) 'fresh" 775 gram U-235 fuel elements.

This predicted value is well below the Technical Specification limit of 0.9.This value will also be validated by 1/M criticality determination.

" In summary, new MCNP modeling of the upper and lower levels of the "Z" storage basket and fresh fuel storage in the vault, using conservative, worst-case assumptions of all fresh fuel elements, indicate IKfr values much less than the MURR Technical Specification Limit of 0.9 (no value was calculated greater than 0.56). Additionally, the 2004 criticality analysis of the "X" and"Y" storage baskets (see Attachment

6) calculated a K~if value of 0.635 for each basket, once again, using conservative, worst-case assumptions of all fresh fuel elements.14 of 86

3. NUREG-153 7, 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 RO, RZ, and ROZ 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 RA1 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.Historically, neutron physics modeling and analyses at MUJRR have been performed using several multi-group and multi-dimensional neutron diffusion theory codes such as PDQ, EXTERMINATOR-fl and BOLD VENTURE. Since the BOLD VENTURE core model was benchmarked against the destructive analysis of a highly-enriched uranium (HEU) MURR fuel element for the license renewal application submitted to the NRC in August of 2006, MURR used results provided by the above set of neutronics codes.Since then, MURR core physics analyses have switched to using newer, state-of-the-art programs such as MCNP for neutronic analysis.

For a compact core such as MURR, it is preferable to use a transport theory code to capture the rapidly changing spectra across the various regions. Therefore, MCNP (in combination with other activation and depletion programs such as ORIGEN) is now routinely used for all calculations of core Ker, critical control blade height, detailed power distribution, and experimental fluxes/reaction rates.As part of the on-going collaboration, which started in 2006, between MURR staff and Argonne National Laboratory (ANL) analysts for the purpose of determining the feasibility of converting MURR from HIEU to low-enriched uranium (LEU) fuel, ANL has assembled a neutronics analysis code suite utilizing WIMS-ALNL, REBUS-DIF3D and REBUS-MCNP.

Figure 1 below illustrates the linkage of the codes in the analysis suite.The suite of programs, or codes, was used to provide detailed (radial, axial and azimuthal) fuel composition for partially burned fuel elements.

Since MURR routinely operates with a fuel cycle utilizing a mixed burnup core, realistic experimental flux, reaction rates and power peaking values have to be evaluated for the typical core weekly cycles rather than for an all-fresh core. The 15 of 86 detailed fuel composition data obtained is then subsequently used in a MCNP calculation to obtain the worst-case power peaking factors and heat flux values used in the thermal-hydraulic analysis.tMCNP Runs (Outside REBUS)Produt *Detailed power distributions Lumped Fission PrdutExperimental fluxes / reaction rates Cross-Sections in 69 groups MCNPinufi-U REBUS Fuel Management Driver* Cross reference materials

& geometry* Transmute materials* Time dependent power & step size* Update materials in geometry* Fuel shuffling* Update control and/or experiments

-I t I DIF3D Neutronics Solver* Cross-section interpolation

• Flux solver I Region Fluxes Region Reaction Rates I 4 Fi~A N~uCross-Section Library Figure 1 -Linkage of the Codes Used in the Analysis Suite The following is a brief description of each of the programs within the ANL neutronic analysis suite: WIMS-ANL:

WIMS-ANL is a one-dimensional lattice physics code used to generate burnup dependent, multi-group cross sections.

The code utilizes either 69- or 1 72-group libraries of cross-section data for 123 isotopes generated from ENDF-6. A customized 10-group structure was developed by ANE based on the neutron spectrum that exists in the MURR core. This multi-group data can be used in MCNP and REBUS-MCNP analyses of depleted cores.REBUS-DIF3D:

DIF3D is a multi-dimensional, multi-group neutron diffusion code that can model systems in a number of geometries.

REBUS is a depletion code that utilizes neutron fluxes from a neutronics solver and cross-section data to solve isotopic transmutation calculations.

A 16 of 86 detailed O-R-Z diffusion MURR model was developed for DIF3D. The depleted core characteristics (plate-by-plate and axially-segmented atom densities) can be saved and passed on to MCNP for more detailed neutronics analyses.MCNP: MCNP is a continuous energy Monte Carlo neutron transport code. MCNP is capable of modeling the heterogeneous details of the MURR fuel elements, core structures, and experimental facilities while capturing the rapidly changing spectra across these various regions. Using the 69-group lumped fission product library generated by WIMS-ANL, the code can be used to model cores of depleted and fresh elements.ANL had performed extensive work to validate the above set of neutron physics codes and models for application to MURR. The MCNP and DIF3D models were benchmarked against available experimental data [Ref. 1].In order to speed up routine neutronics calculations, where such detailed axial, radial and azimuthal fuel composition is not necessary, MURR utilizes the MONTEBURNS program. MONTEBURNS is a coupled MCNP-ORIGEN code system developed by Los Alamos National Laboratory (LANL). It utilizes the capabilities of ORIGEN 2.2 for isotope generation and depletion calculations and that of MCNP5 for continuous energy, flux and reaction rate as well as criticality calculations.

MONTEBUJRNS by itself is not designed to handle transient calculations such as during the period from reactor startup through critical and then on to steady-state reactor operation since it involves control blade motion due to poison buildup as well as from fuel depletion.

However, with the help of in-house developed routines, a code system including MCNP and MONTEBURNS was developed to perform routine reactor physics calculations that can handle transient cases.The flow diagram for the suite of codes implemented at the MURR for routine core-physics analysis is shown in Figure 2.17 of 86 A/

• MONTEBURNS 2.0 -time dependent stepwise MCNP coupled ORIGEN nuclear burnup code Ssystem (LANL)*Critical Rod search routine -adjustments to control rod height based on repeated detailed MCNP KCODE calculations.

• Returns a critical rod height if the KCODE k~e is 1.0000+/-0.03%

and control rod is less than maximum travel Figure 2 -Code Suite Flow Diagram Implemented at MURR for Routine Core Physics Analysis These "routine" calculations utilize a very detailed MCNP MURR core model that has the following key capabilities:

• It can model MURR's "mixed-core" weekly fuel configuration, i.e., atom densities of various isotopes in the fuel matrix, for a range of fuel element burnups from fresh (0 MWd or no fuel depletion) to spent status. For the Estimated Critical Position (ECP) calculations, fuel element definitions can be individually selected from a fuel burnup database to simulate any combination of eight (8), xenon-free fuel elements.*Similarly, it can include a mixture of four (4) independently depleted BORAL shim control blades -each with a different axial and radial boron depletion profile based on its operational history (or core residence time).*It has the ability to account for poison and gas buildup, and the reduction of beryllium atom density within the beryllium reflector based on its run time (from 0 to 8 years).*The multiple samples that are irradiated in the high worth central flux trap region of MURR, as well as in the various positions within the graphite reflector region, are modeled accurately in order to reduce the error in the FCP calculation.

• With the help of a critical control blade height search routine, starting from an initial estimate of the critical control blade height, a series of MCNP5 criticality (KCODE) calculations can be performed in order to calculate the critical control blade height.18 of 86

• In order to predict control blade travel during startup and subsequent steady-state operation, as well as recovery following an unplanned reactor shutdown, it can track the buildup of xenon-i135 and other poisons in the core during reactor operation as well as the buildup and decay of the poisons during shutdown and restart using the isotope buildup and decay/loss capability of MONTEBURNS.

The system of codes and calculation methodology described previously has been benchmarked extensively using actual weekly core refueling and reactor startup data. The response to Question 2.a, which was included in the responses, dated July 31, 2015, to a Request for Additional Information made by the NRC (by letter dated June 18, 2015), contains the benchmark data.

References:

'Stillman, J., et al., Technical Basis in Support of the Conversion of the University of Missouri Research Reactor (MURR) Core from Highly-Enriched to Low-Enriched Uranium -Core Neutron Physics, AINL/RERTRiTM-1 2-30, Argonne National Laboratory, September 2012.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.The code system that is currently used for reactor physics analysis at MURR has been benchmarked extensively.

One of the methods used for the benchmarking was by comparing the Estimated Critical Position (ECP) calculations from the detailed MCNP MUJRR model against the actual startup critical control blade height data from several weekly reactor startups.

The detailed MCNP MURR core model includes depleted control blade data, beryllium aging effect (i.e., more and more gas molecules taking up the place of beryllium atoms with increasing run time), as well as detailed sample information present in the central flux trap region of the reactor core.In Table 1 below, eight (8) separate cores were selected for comparison to verify' consistency in the model's ability to predict the ECP accurately under various core states (mixed burnup) and flux trap sample conditions.

The comparison was performed over an eight (8) month period. Note that the reactor startups at MURR require an occasional "strainer" startup -where initial critical control blade height data is obtained without any samples or sample holder in the central flux trap region, just pool coolant. Two such "strainer" startups are reported in Table 1.19 of 86 Table 1 -Comparison of Estimated Startup Critical Control Blade Height vs. Measured Data Core Actual Critical Predicated Critical Peiae lxTa Control Blade Control Blade Configuration Height (Inches) Height (Inches) I~f Configuration Week of 1/28/2013 16.79 16.67 0.99993 Strainer Week of 2/04/2013 16.52 16.27 0.99975 Samples Week of 4/29/2013 15.98 15.78 1.00017 Samples Week of 6/10/2013 15.44 15.42 0.99995 Samples Week of 8/05/2013 16.74 16.74 0.99985 Strainer Week of 8/12/2013 15.71 15.61 0.99985 Samples Week of 8/19/2013 15.84 15.84 1.00016 Samples Week of 8/26/2013 15.64 15.69 1.00029 Samples A negative bias of ~1.5% is seen in the predictions for the early benchmarks.

After the additional refinements to the MCNP MURR model were made, the variations in the predictions were within+0.8% of the actual critical control blade heights (last 5 entries of the Table).c. Provide calculated and measured control blade worths (Shim-i, 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.Control blade usage at MURR is similar to the mixed core fuel cycle in that, at any given time, the four BORAL shim control blades (Shim-i, Shim-2, Shim-3 and Shim-4, also referred to as control blades 'A,' 'B,' 'C' and 'D') are in various stages of burnup (core residence time) ranging from fresh (no burnup) to approximately 10 years. Every six (6) months, one of the control blades, and its associated offset mechanism, is removed from its installed location for inspection and replaced with another rebuilt offset mechanism and a different control blade with a different burnup status. This schedule satisfies the Technical Specification surveillance requirement of inspecting one (1) out of four (4) control blades every six (6) months so that every blade is inspected every two (2) years. In this way, a given control blade is cycled in and out of the reactor multiple times from the time it is new until it is no longer usable due to burnup.Detailed control blade burnup studies undertaken at MUIRR have shown that the lower 6 to 8 inches of the control blade tip undergoes significant boron depletion with operation.

Only the control blade tip experiences burnup since during steady-state, full power operation the control blades are almost fully withdrawn, resulting in the active neutron absorbing region of the blades being out of any significant neutron flux. Since accurate control blade worth information is crucial for reactor operation, every six (6) months when a control blade is replaced, a blade worth measurement of the installed control blade is performed.

20 of 86 Using the detailed MCNP core model of MUIRR, the differential and integral worth of the four shim control blades, and that of the stainless steel regulating blade, were calculated and the results are shown in Figures 3 through 6. The calculations were performed for fresh (non-depleted) control blades using a fresh core with no xenon. In order to show the effect of control blade depletion with operational history, the differential and integral worth curves of a single blade with a core residence time of over 9 years are also shown in Figures 7 and 8, respectively.

0.004 y l E--07' -7E-0t,', 8E*05x'+ 2E-05X- 7E-06 V = 9E-08,'- 5E-6'.0x 7E-05,' -6E-0OX + 3E-05 0.035 7E-Og, -4E-O6,,3/4+4E-O~x' ,O002O + 1E-O5% y = 1E-O7,'- 5E-O6x; 7E-05,,'.

1E-O4x- 2E-05 0.0025// / CN CpAA N-~shh 0.0015* MCNP hp/ANt All-fresh ShimA S0.0015 /' M(tlP pAol~l All-fresh Shim B...Po. MCNP hp/Al H All-fresh Shim C )0--1 MCN h /AH All-fresh Shim B)/* Poly. (MCNP /AAl-rhSimA 0.0005 --Poly. (MCNP hp/AN A4ll-fresh Shim O '---PI I(CN h/ll All-fresh Shim O* ~/ .PoIy. (MCNP hp""l-fehSi 0510 15 20 25 30 Control Blade Height Withdrawn (inches)Figure 3 -Calculated Differential Worth Curves for All-Fresh Shim Control Blades in an All-Fresh Fuel Core Configuration Shim control blades 'B' and 'C' are worth slightly less than control blades 'A' and 'D' since blades'B' and 'C' are located near two highly "black" fast flux irradiation reflector elements situated on the west side of the core, adjacent to these control blades.21 of 86 0.060000 MCNP Integral Rod Worth Shim A U MCNP Integral Rod Worth Shim B* MCNP ,Iteral Rd Worth Shim D .. ": "'---.Poly. (MCNP Integral Rod Worth Shim A) :s --.......Poly. (MCNP nntegral Rod Worth Shim D) J'* l 0.030000..

' /'./ ! ...... ....... ... ...a/ / y 2E.06xr ÷ .Os5,÷ 1E.OSe'. ZE-ise ÷ZE-13 0.020000 = 9E-07Xa + 1E-05X3+ 9EtDSx3/4 ZE-13X+r 2E-13 0.000000 -,,"1 -05 10 15 20 25 30 Control Blade Height Withdrawn (inches)Figure 4 -Calculated Integral Worth Curves for Ali-Fresh Shim Control blades in an Ali-Fresh Fuel Core Configuration 0.000250/4' MCNP Ag/AH \a" -.-Poly. (MCNP 0.000150 /~/0o.0o01oo

/\/

-..../ y = -8,7729E-1D0'-

2,9792E-8x 3 + 1,0668E-06xz

+ 7.8270E-O6x

/ Rz = 9.9953E-01 0.0o00000

')0 5 10 15 20 25 30 Regulating Blade Height Withdrawn (inches)Figure 5 -Calculated Differential Worth Curve for a Fresh Regulating Blade in an Ali-Fresh Fuel Core Configuration 22 of 86 O.0000 MOWP Integral Rag Blade Worth .Poly. (MCNP Integ'ral Rag Blade Worth)0.0025o00/

S0.0020000

/: 0.0015000/

I.. /-/ y = -2E-IOX" -41r-07x' 4106X'., 1E-14k. 25.14 05 10 15 20 25 30 Regulating Blade Height Withdrawn (inches)Figure 6 -Calculated Integral Worth Curve for a Fresh Regulating Blade in an All-Fresh Fuel Core Configuration 0.003 0.025', Y -4.2163E-O~x'

+ 3.6065E-06x 3 -1.O055E-O4x 7 + 9.1731E-04i q 0.02 I: 9.9715E-01

• I 0.0015 IQ MCNP A/AH I=/ --- Poly. (MCNP h4p/AH) \s 0.000:15 ""20 25-Control Blade Height withdrawn (inches)-0).0005 Control Blade Height Withdrawn (inches)Figure 7 -Calculated Differential Worth Curve for Shim Control Blade 'B'(with 9.0 years of core residence time)23 of 86 0O000 0.05000 0.04000MCNP control blade Id: 6-.05 worth after 9.0 year currently in position B.= 0.02000io/* y = -8E-Ogx 5 + 9E-O7x 4 -3E-O5x 3 + O.O005x' + tE-t3x + 2E-13 0.01000 A)0 5 10 15 20 25 30 Regulating Blade Height Withdrawn (inches)Figure 8 -Calculated Integral Worth Curve for Shim Control Blade 'B'(with 9.0 years of core residence time)As mentioned before, every time a control blade is changed out after a bi-annual inspection, the reactivity worth of the newly installed control blade is measured and the total bank worth curve (combined worth of all four shim control blades) is recalculated for the purpose of reactor physics calculations (such as ECP predictions, reactor core shutdown margin, estimation of unknown sample reactivity worths, etc.). To serve as a benchmark for the calculated control blade worths, a single blade was selected.

Using a detailed mixed-core, mixed-bumup control blade model of the reactor configuration during the last Shim-4 (control blade 'D') inspection and replacement the blade worth 'D' measurements were simulated.

The measured control blade worth curves for blade'D' are compared against the blade worth curves calculated by the MCNP model. The results are shown in Figures 9 and 10.24 of 86 0.003 u0.0025 S0.002 L.* 0.0015*n 0.001-000050 0.-).0005 Control Blade Height Withdrawn (inches)Figure 9 -Comparison of Measured and Calculated Differential Worth Curves for Shim Control Blade 'D'0.0600 U, Measured Worth (02/09/2015) 0.0500 MCNP control blade Id: 6-16 worth after 0.5 year currently In Position 0---Poly. (Measured Worth (02/09/20151) 0.040000 -PolY. (MCNP control blade Id: 6-16 worth after 0.5 year currenty In Position D )S0.0300000 5 01 2 53 Coto ld Hih ihran (n hs Figure 10 -Comparison of Measured and Calculated Integral Worth Curves for Shim Control Blade 'D'25 of 86

d. Provide a calculated and measured temperature coefficient for a given core configuration at a low power, no xenon condition, orjustift why this information is not needed.The primary and pooi coolant temperature coefficients are provided in Table 4-12 (Page 4-42) of the SAR. But since those coefficients were calculated using the older set of neutronics codes, MURR has recalculated the primary temperature coefficient using the newer sets of computer programs that were described in the response to Question 3.a. The results are provided in Table 2 below.Table 2 -MURR Primary Coolant Temperature Coefficient MURR Technical ANL-MCNP Calculation Coefficient Specification Limit (294 to 400 K)All Fresh Core, BOC:-13.2 x 10-5 -2.3 x 10-7 Ak/k/°F Average Core Temperature MxdCr.BC Primary Coolant Coefficient Shall be More -12.8 x 10.5 + 2.2 x 10.7 Ak/k/°F Temperature Negative Than: (Isothermal)

Mixed Core, Eqi. Xe:-6.0x 1 5 Ak//0 F-11.8 x 10s +/-- 2.2 x 10 7 Ak/k/°F Mixed Core, BOC (ENDF7):-12.5 x 10.5 + 2.2 x 10-7 Ak/k/°F Note: BOC = Beginning of Cycle.26 of 86

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.Table 4-12, on Pages 4-41 and 4-42 of the SAR, contains the value for the reactor core shutdown margin. The Table lists the maximum K~ff with the highest worth shim control blade fully withdrawn, or stuck, as 0.93 8. This maximum K~ff, value translates to a minimum reactor shutdown margin value of -0.066 Ak/k. This compares with the Technical Specification minimum reactor core shutdown margin requirement of -0.02 Ak/k.Referring to the response to Question l .a provided earlier, the reactor core shutdown margin value, calculated using the newer suite of reactor physics programs in use at MUJRR as described in the response to Question 3.a, is -0.0875 Ak/k for an all-fresh fuel core case [Note: license possession limit only allows six (6) fresh fuel elements onsite].27 of 86

5. NUREG-1537, Section 11.1.1.], "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-J O 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.MURR calculates effective stack height, for the purposes of determining dose from radionuclide emissions, as the difference in vertical elevation between the point of emission at the end of the MURR exhaust stack and the receptor height at the point of interest, plus the effective stack height calculated using the Davidson equation.

This equation takes into account the stack diameter and exhaust velocity of the gases leaving MURR to calculate an injection height and thus an effective stack height into the atmosphere.

Wind speed is also an input parameter into this formula as it is a function of the particular Pasquill atmospheric stability class that is being modeled for the general wind direction that is being used to calculate the offsite dose; thus it is included in the equation.While G.A. Briggs notes on page 23 of his book "Plume Rise" that the Davidson formula ". .. often greatly underestimates observed rises,..." this underestimation would cause the offsite dose calculations using the Pasquill-Guifford model to overestimate doses to the individual at the point of dose calculation interest.

In fact, dose estimates generated using this model are not out of line with doses calculated using the COMPLY 2 computer code which is used to determine annual doses (demonstrate compliance) to the nearest resident from MURR as part of the facility's annual National Emission Standards for Hazardous Air Pollutants (NESHAPS) compliance report.Additionally, using Briggs' own equations for calculation of effective stack heights from the same reference book "Plume Rise," confirms that while the Davidson model underestimates effective stack heights, these underestimated effective stack heights lead to an overestimation of dose, thus providing a conservative approach to the offsite dose calculations.

Thus, we feel that no additional information is required.

References:

'Briggs, G.A., Plume Rise, AEC Critical Review Series, U.S. Atomic Energy Commission, Division of Technical Information, 1969.2 COMLY is a computerized screening tool for evaluating radiation exposure from atmospheric releases of radionuclides.

May be used for demonstrating compliance with some EPA and U.S.Nuclear Regulatory Commission regulations, including NESHAPS in 40 CFR 61, Subpart H and Subpart I.28 of 86

b. SAR page B-il has an equation for X/Q that includes the cy and arz 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.Tables B-2 and B-3 in SAR Appendix B contained some incorrect values for both the horizontal (oy) and vertical (gz) dispersion coefficients.

These values have been reviewed and updated and are now included in the corrected Tables B-2 and B-3 below.TABLE B-2 MAXIMUM ANNUAL INDIVIDUAL DOSE AT 150 METERS Location:

150 Meters Directly North Elevation at Man Height: 636 Feet (194 Meters) ______ ____Eff. Height ay X %s Dose with %s Clas in I in (n) se/r 3) (tiCi/ml or Comb. (mremly)Class (m)___ (m)_ (m)_ (sec/m3)_

Gi/m 3) _________A 35 33 23 6.27E-05 3.14E-09 2.40E-04 0.00 B 27 23 15 6.09E-05 3.04E-09 5.10E-03 0.08 C 23 17 11 4.55E-05 2.28E-09 1.70E-02 0.19 D 20 12 7 1.14E-05 5.71E-10 6.30E-02 0.18 E 23 8.5 5 4.76E-08 2.38E-12 3.10E-02 0.00 F 30 6 3.2 5.24E-22 2.62E-26 1.50E-02 0.00 Total 0.46 TABLE B-3 MAXIMUM ANNUAL INDIVIDUAL DOSE AT 760 METERS Location:

760 Meters Directly North Elevation at ManHeight:_700 Feet (213 Meters)____________

Eff. Height ay Oz x/Q X %s Dose with %s Class (in) I(in) (mn) (sec/rn 3) i/mi or Comb. (inrem/y)____I 1 _____ __________

Ci/m 3) _________A 16 170 270 3.30E-06 1.65E-10 2.40E-04 0.00 B 8 120 85 1.04E-05 5.18E-10 5.10E-03 0.01 C 4 85 52 1.71E-05 8.55E- 10 1.70E-02 0.07 D 1 55 26 3.97E-05 1.99E-09 6.30E-02 0.63 E 4 42 18 1.03E-04 5.13E-09 3.10E-02 0.80 F 11 30 12 2.23E-04 1.12E-08 1.50E-02 0.84 Total 2.35 29 of 86 Note: Tbe "%s Comb." column was added to Tables B-2 and B-3 to better aid in understanding the calculation of total dose based on the Pasquill-Guifford stability classes and wind direction.

30 of 86

6. NUREG-15 3 7, 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.S MW). The temperature feedback coefficient used is -7.0 x lO Ak/k rather than the TS S.3.a value of -6xlO-5 Ak/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.For the Insertion of Excess Reactivity accident analysis, the licensed maximum power level of 10 MW was used in the SAR as the starting assumption since MURR does not, nor can it legally, operate above this power level. On Page 13-9 of NUREG-1537, Part 2, Standard Review Plan and Acceptance Criteria, for the Insertion of Excess Reactivity accident, "The accident scenario assumes that the reactor has a maximum load of fuel (consistent with the technical specifications), the reactor is operating at full licensed power, and the control system..." The accident was reanalyzed at a much more conservative starting power level (11.5 MW) than required by NUREG-1537 and the results are provided below. 11.5 MW was chosen, instead of the Limiting Safety System Setting (LSSS) set point of 12.5 MW, since the rod run-in system will initiate a rod run-in at 11.5 MW (Technical Specification 3.2.f.1) and shutdown the reactor prior to reaching the LSSS scram set point of 125%.For the SAR analysis of the Insertion of Excess Reactivity accident, the temperature coefficient used was -6.0 x 10.5 Ak/k and not -7.0 x 10-5 Ak/k as stated above. Third paragraph on Page 13-17 of the SAR lists the various reactivity coefficients assumed for the Insertion of Excess Reactivity accident analysis.Details regarding the power peaking factors used were not provided in that section of the SAR. The power peaking values used were values obtained based on the destructive analysis of a MURR fuel element. For the updated analysis, more up-to-date power peaking values, based on the detailed MCNP MURR core model, were used.For both the SAR analyses, as well as for the updated analysis presented here, the control blade insertion times are based on the current and relicensing Technical Specification 3.2.c requirement of insertion to the 20% withdrawn position in less than 0.7 seconds. So the insertion rate was calculated based on shim control blades travelling from 26 inches (fully withdrawn) to 5.2 inches (20% withdrawn or 80% inserted) in 0.7 seconds. This is a conservative assumption since monthly 31 of 86 control blade drop time verifications performed at MURR have always yielded insertion times of 0.6 seconds or less (see response to RAI 1 .a).Similar to the SAR analysis, the Reactivity Transient Analysis program PARET (V7.5), maintained and distributed by the Nuclear Engineering Division of Argonne National Laboratory (ANL) was used. For the Insertion of Excess Reactivity accident analysis, two channels were modeled in PARET; a hot channel representing worst-case conditions inside the core and an average channel representing the rest of the core experiencing "average" conditions.

The axial power profiles used for this 2-channel PARET reactivity transient analysis are given in Table 1 below.Table 1 -Peaking Factors in the Hot and Average Channels Hot Channel Average Channel 2.046 1.058 1.971 0.920 2.145 1.018 2.335 1.132 2.497 1.219 2.672 1.307 2.835 1.360 2.986 1.411 3.105 1.430 3.164 1.437 3.169 1.420 3.098 1.383 2.953 1.326 2.775 1.243 2.542 1.140 2.290 0.989 2.069 0.828 1.888 0.701 1.703 0.615 1.499 0.530 1.277 0.460 1.080 0.386 0.904 0.329 0.880 0.358 As indicated earlier, the transient was started from an initial power level of 11.5 MW with core coolant flow rate as well as core coolant inlet temperatures set at their LSSS values of 3,200 gpm 32 of 86 and 155 'F, respectively.

Also, pressurizer pressure was at 75 psia (LSSS value). Since the Insertion of Excess Reactivity transient was analyzed from a starting power level of 11.5 MW, the rod run-in that would be initiated by the rod run-in system at 11.5 MW was bypassed and only the high power scram set point of 12.5 MW was modeled. Also, a delay of 150 milliseconds was incorporated into the control blade scram model so that the control blades would only start to insert 0.15 seconds after the power level had exceeded the scram set point of 12.5 MW.The results of a step reactivity insertion of 600 pcm (+0.006 Ak/k) are shown below in Figure 1. As expected, due to the higher starting core power level, much lower core coolant flow rate and much higher than normal core coolant inlet temperature conditions assumed for this updated analysis, the peak power during the transient momentarily reaches approximately 37.4 MW compared to a value of approximately 33.0 MW reported in the SAR analysis for the same 600 pcm step reactivity insertion.

40.00 400 35.0030.001 o 20.00-POWER MW-'l'dad "C-I'f maxtC S350 300 l0 S50* 0 3.00 0.100 0.50 1.00 1.50 2.00 Time (seconds)2.50 Figure 1 -Reactor Power, Fuel and Cladding Temperatures vs. Time for a Positive Reactivity Step Insertion of 0.006 Ak/k Several SPERT tests had shown that the reactor can withstand such short duration (few milliseconds) power burst without sustaining any fuel damage and only sustained operation at such high power levels will lead to fuel damage. The peak fuel temperature reached during the Insertion of Excess Reactivity accident in the worst (Hot) channel is only 227.4 0 C -well below the new Safety Limit of 530 0 C for the aluminide fuel.33 of 86

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.The Loss of Coolant Accident (LOCA) and the Loss of Flow Accident (LOFA) are thermo-hydraulic transient accidents based on a departure from long-term, steady-state, full power operation.

MURR does not, nor can it legally, operate above its licensed power level of 10 MW, but assumed 11 MW to add an additional 10% higher steady-state operating heat flux and total decay heat factor.As stated on page 13-7 of NUREG-1537, Part 1, Format and Content, item 13.2(1), "State the initial conditions of the reactor and equipment.

Discuss relevant conditions depending on fuel burnup, experiments installed, core configurations, or other variables.

Use the most limiting conditions in the analyses." On Pages 13-9 and 13-10 of NUJREG-1537, Part 2, Standard Review Plan and Acceptance Criteria, for both the LOCA and LOFA, "The scenario assumes that the reactor is operating at full licensed power and has been operating long enough for the fuel to contain fission products at equilibrium concentrations." Therefore, we believe, the assumed 11 MW steady-state power level was greater than the required assumed power level as stated in NUREG-1537, Part 2.The assumed peaking factors are graphed on Figure C.6 (Page C-l13) and listed in Table C-I (Page C- 14) of Appendix C of the SAR. The cold-leg break LOCA has the highest peak fuel temperature of 311.7 0 F (155.4 0 C) occurring in fuel plate number-3 within the first second. The hot-leg break LOCA peak fuel temperature is 281.2 0 F (138.4 0 C) also occurring within the first second. These peak temperatures occur within the first second because the events start with a loss of normal forced circulation coolant flow with a slight delay in the reactor scram trip. The peak fuel temperature for a LOFA is 280.3 0 F (137.9 °C), which occurs in fuel plate number-i, 0.3 seconds into the transient.

The LOFA and LOCA analyses were redone with the other three (3) Limiting Safety System Setting (LSSS) variables

-core coolant inlet temperature, core coolant flow rate and pressurizer pressure -at their respective set points of 155 0 F, 3,200 gpm and 75 psia. The peaking factors used in the updated analyses are the ones described in the response to RAI 7.a (By letter dated July 8, 2013, the NRC issued Amendment No. 36 to Facility Operating License No. R-103, which revised the MURR Safety Limits). The peak fuel temperature for the cold-leg break LOCA was 413.9 0 F (212.2 0 C) whereas the peak fuel temperature for the LOFA was 292.3 0 F (144.6 °C).Well below the new Safety Limit peak fuel temperature of 530 0 C.MURR feels that having core coolant inlet temperature, core coolant flow rate and pressurizer pressure are their respective LSSS and reactor power at the full licensed limit meets the guidance of NUREG-1537, Parts 1 and 2, and is sufficiently conservative.

34 of 86

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.The basis for the maximum hypothetical accident (MHA) evacuation time is stated on Page 13-5 of the SAR: "It would take approximately 5 minutes for Operations personnel to secure the primary coolant system (PCS) and verify that the containment building has been evacuated following a containment building isolation.

For the purpose of the MIIA calculations, a conservative assumption of 10 minutes is used." This basis will not change; however, the MiHA has now been renamed the "Fuel Failure during Reactor Operation" accident since the dose consequences of a failed fueled experiment are the most severe of all of the radiological accident scenarios.

However, for a failed fueled experiment (now the MHA) or a fuel handling accident (FHA), the primary coolant system (PCS) does not have to be secured. The only required action for Operations personnel is to verify that the containment building has been evacuated following a containment building isolation, which will occur during both of these accidents.

MUJRR performs an evacuation drill every year and the typical time period for all personal to evacuate the containment building, including verification by Operations personnel, is two (2) to two and a half (2.5) minutes. For the purposes of the failed fueled experiment and FHA calculations, a conservative assumption of five (5) minutes is used for both accident scenarios.

35 of 86

7. NUREG-153 7, Section 13.1.1, "Maximum Hypothetical 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 MJVD burnup consisting of twelve 10O-day cycles over a 300-day period. These values were then adjusted using a peaking factor of 1. 6. However, in the response to RAIJA.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.Table 4-14, "

SUMMARY

OF MUIRR HOT CHANNEL FACTORS," in Section 4.5 of the MUJRR SAR, lists a hot spot power peaking factor of 3.6765 with no engineering factors included.

This value applies the product of the radial, axial and azimuthal peaking factors of a fuel plate to determine the hot spot on the plate. The SAR provides an overall peaking factor of 4.35; the hot spot power peaking factor of 3.6765 multiplied by the engineering factors. These two peaking factor values apply to the potential worst-case maximum power density point in the core for the Safety Limits (SL) when the SAR was submitted in August 2006.From Table 4-14, "

SUMMARY

OF MURR HOT CHANNEL FACTORS," of the MURR SAR: On Heat Flux Power-related Factors Nuclear Peaking Factors Radial 2.220 Non-Uniform Burnup 1.112 Local (Circumferential) 1.040 Axial 1.432 Overall 3.676 Engineering Hot Channels Factors on Flux Fuel Content Variation 1.030 Fuel Thickness

/ Width Variation 1.150 Overall Product 4.35 By letter dated July 8, 2013, the NRC issued Amendment No. 36 to Facility Operating License No.R-l103, which revised the MIURR SLs. The revised SLs reduced the overall nuclear peaking factor to 3.4747; with no engineering factors included.

Including the engineering factors, the overall peaking factor increases to 4.116.36 of 86 From Table F.4, "

SUMMARY

OF MURR HOT CHANNEL FACTORS," of Appendix F of Addendum 4 to the MURR Hazards Summary Report (as revised by Amendment No. 36): On Heat Flux From Plate-i Power-related Factors Nuclear Peaking Factors Fuel Plate (Hot Plate Average) 2.215 Azimuthal Within Plate 1.070 Axial Peak 1.3805 Additional Allowable Factor 1.062 Overall 3.4747 Engineering Hot Channels Factors on Flux Fuel Content Variation 1.030 Fuel Thickness

/ Width Variation 1.150 Overall Product: 4.116 This peak heat flux point is at axial mess interval 14 (13 to 14 inches down the fuel plate meat)where the enthalpy rise at that interval is 52.3%. The SL is based on mess interval 18, which has an overall peaking factor of 3.863 and an enthalpy rise of 74.8%; thus producing the most limiting combination of heat flux and enthalpy rise.This overall peaking factor of 4.116 at mesh interval 14 would apply to the 1-inch square assumed in the analysis in the response to RAT A.27. Therefore, since the ratio is 1.372 (4.116 / 3.0), the calculated dose rates in the response to RAI A.27 increased by approximately the 37.2%. The only other fuel plate exposed during handling is plate number-24, which has a lower overall peaking factor than plate number-i.

The assumed peaking factor in the response to RAT A.27 has been revised from 3.0 to 4.116, which increased the whole body (TEDE) "60-Minute Dose from Radioiodine and Noble Gases in Containment" from 0.79 to 1.09 mrem. This change also required a similar revision to the response to RAI A.6 regarding the revised Technical Specification definition for Irradiated Fuel, Definition 1.11, by about the same percentage.

The current MIURR maximum hypothetical accident (MHA) assumes the melting of fuel plate number-i in four (4) different fuel elements.

An unirradiated fuel plate number-i contains, on average, 19.26 grams of U-235, so four (4) unirradiated number-i fuel plates contain 77.04 grams instead of the 78.58 grams assumed in the MHA. These four (4) number-i fuel plates that melt correspond to 1.41% of the total U-235 in the Week 58 Core that was used to determine the high power peaking factor for the revised SLs. The Week 58 Core has a total power history of 576 MWd. This power history results in a total reduced core mass of 5,474 grams of U-235 due to the previous fuel consumption.

This 1.41% of U-235 melting releases 3.42% of the core fission products due to the highest power density fuel plate number-i overall peaking factor of 2.423, which is conservatively assumed to apply to all four number-1 fuel plates (1.41% x 2.423 = 3.42%).37 of 86 Following the response to RAI 7.g below is the revised MHA, which will now be referred to as"Fuel Failure during Reactor Operation" since the dose consequences for an individual in the containment building are less than that for a failed fueled experiment, which is now considered the MURR MHA.b. The release is assumed to occur into the PCS with a volume of 2,000 gallons. Identif what components comprise this volume and provide information to confirm the 2,000 gallon volume assumption, or justif why no additional information is needed.The 2,000 gallon total volume of the primary coolant system (PCS) is based on the volume of all of its individual, major components, including the piping, reactor core, pressure vessels, primary coolant circulation pumps, heat exchangers, and pressurizer.

Attachment 7 is a breakdown of the measurements and calculated volumes, where design capacities are not available, of the individual components following the RELAP Model component designations listed in Appendix C of the SAR. The total calculated volume of the PCS is 2,007 gallons. However, based on the difficulty of measuring some of the in-pool PCS piping and components, this volume is conservatively underestimated by approximately 5 to 10%, thus radionuclide concentrations in the PCS are conservative.

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 RA] 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.In response to these RA~s, all three (3) radiological accidents

-Maximum Hypothetical Accident (MHA), Fuel Handling Accident (FRA), and Fueled Experiment Failure -have been reanalyzed using consistent methodologies and assumptions.

All three (3) new analyses are included in these responses.

The following are the radiological accident scenarios and the whole body exposures (TEDE) to an individual in the containment building associated with them: Maximum Hypothetical Accident:

42.18 mrem Fuel Handling Accident:

687.00 mrem Fueled Experiment Failure: 1212.44 mrem Based on these analyses, the Fueled Experiment Failure accident has been determined to be the new MURR MHA. The current MHA will be renamed "Fuel Failure during Reactor Operation." d. The released concentrations in the containment are based on the 10O-minute leakage between the PCS and the pool cooling system. However, the NRC staff questions whether the release 38 of 86 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.As stated on Page 13-3 of the SAR, a reactor scram and actuation of the containment building isolation system will occur as a result of the gaseous activity collecting in the vent tank system. At this point the containment building is isolated.

As described in Section 9.13.3 of the SAR, the vent tanks will vent through an absolute and charcoal filter if enough gases collect in the vent tanks to cause the water level in the tanks to recede to a point where level controller 925A will signal valve V552A to open and vent the gases. Note: The vent path for the vent tank system, after it goes through the absolute and charcoal filters, is to the pool sweep system which is connected to the containment building 16-inch hot exhaust line (see SAR Sections 6.2.3.8 and 9.1.2.2).

The containment building 16-inch hot exhaust line contains two (2) quick-closing isolation valves, designated 1 6A and 1 6B. During a containment building isolation, both of these valves will close.The volume of gases that are released from four (4) number-l fuel plates is insignificant and would not cause the system to vent. However, if for some reason the system should vent prior to the PCS being secured as part of the actions of Operations personnel during an MHA (vent valves 552A and 552B will not open when the PCS is secured), the gases will be vented into the isolated containment structure and not to the environment.

Any determination to enter the containment building and un-isolate the structure and vent any potential gases after the accident will be part of long-term recovery actions, which will be very well planned and organized.

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.In reviewing the available meteorological data for the Columbia vicinity, newer meteorological data was found from the Columbia Regional Airport. This facility has more current meteorological wind data available and this data was used to generate wind roses for updated time periods closer to the current time frame. Based on the results of the meteorological data review, we believe that the previous data submitted is representative of current wind rose data, in and around the Columbia area, as there appeared to be no substantial difference in wind speed and direction during the original submittal utilizing nine (9) years of data from 1961 to 1969 and subsequent data which included the above 9-year period and an additional 21 years of meteorological data for a total of 30 years (1961 to 1990). Attachment 8 provides the meteorological data for the years 1961 to 1969.Attachment 9 provides the meteorological data for the years 1970 to 1990, while Attachment 10 provides the meteorological data for the years 1961 to 1990.39 of 86

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 7.5x10-1 4/3.0x10-s or a reduction of about 4.0x10s For 1-131, the ratio is 1.36 x10-1 4/1.1x10-s or a reduction of about 8.1 x104. 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.In the case ofi1-131 as noted above the ratio ofi1-131 is 1.24 x 10-°6. This was determined by taking the initial concentration of I-131 in the containment building of 4.4 x 10-°8 (Page 13.6 of the SAR) and multiplying it by 0.25 (plating reduction factor) and dividing into the final offsite concentration of 1.36 x 10-14 pCi/mi. In the case of Kr-85, the ratio is the same, (7.5 x 10-14 / 6.06 x 10-°8) =1.24 x 10-°6. The initial concentration of Kr-85 was used in this case as there is no plating or other phenomena that would hold up the noble gases.g. hn 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 1 x0-7 micro-Ci/mi 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 difjferences 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.

MURR has revised all applicable dose calculations for both occupational and public doses to use limits from either: 10 CFR 20, Appendix B or 10 CFR 835, Appendix C (Attachment 11). Where available we use Derived Air Concentration (DAC) and Effluent Concentrations from 10 CFR 20 Appendix B. The U.S. Department of Energy (DOE) publishes Appendix C (Air Immersion DAC)specifically for isotopes whose principle exposure pathway is via immersion.

For the four short-lived noble gases (T 1 1 2 < 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) that we analyzed in the included accident analyses, ]VUIRR used the 10 CFR 835 Appendix C default DAC value of 6.0 x 10.06 as noted at the end of Appendix C. From this default DAC we estimate the applicable effluent concentration limit based on the description provided in the Table 2, "Effluent Concentrations," footnotes to Appendix B in 10 CFR 20. Thus, all dose calculations now use limits based on the background and methodology provided in ICRP 26 and 30.40 of 86 Revised "Fuel Failure during Reactor Operation" (Formerly the MIIA)13.2.1 Fuel Failure during Reactor Operation 13.2.1.1 Accident-Jnitiating Events and Scenarios Many types of accidents have been considered in conjunction with the operation of the MUJRR. In all cases, safety systems have been designed such that the likelihood of an accident involving the release of a significant amount of fission products has essentially been eliminated.

The safety systems take the form of automatic reactor shutdown circuits and process systems designed to ensure, through redundancy, that the reactor will shut down upon a significant deviation from normal operating conditions.

In addition, the reactor is housed within a containment building, thus providing further protection against a significant release of radioactive material to the environment.

In the "Fuel Failure during Reactor Operation" accident for the MUJRR, it is assumed that an accident condition has caused the melting of the number-i fuel plate in four (4) separate fuel elements (Ref. 13.11). It is further assumed that the four (4) number- 1 fuel plates are in the peak power region of the core.While one might postulate that this accident could result from a partial flow blockage to the fuel, mitigating features such as the primary coolant system strainer, the fuel element end-fittings, and the pre-operational inspection of the reactor pressure vessels and core region following any fuel handling evolution, all prevent an accident of this type from occurring.

In addition, it has been shown that a 75% blockage of coolant flow to the hot channel is insufficient to cause cladding failure (Ref. 13.2).13.2.1.2 Accident Analysis and Consequences The fuel failure accident postulates partial fuel melting with an associated release of fission products into the primary coolant system. The accident is assumed to occur with the primary coolant system operating, resulting in a quick dispersal of the fission products throughout the system. With the design of the primary coolant system and its associated systems, particulate activity will remain in the coolant, and the gaseous activity that comes out of solution will collect in the reactor loop vent system and be retained there. Therefore, the primary coolant system relief valves and pressurizer are the only paths for a release of significant quantities of fission products to the environment.

The potential energy release from the melting of four (4) number-i fuel plates could occur as a possible metal-water reaction (Ref. 13.3). While hydrogen would be formed, it is highly unlikely that in a water environment a hydrogen deflagration reaction would occur. The amount of material which would be involved in a metal-water reaction under the conditions of four (4) number-i fuel plates melting is not predictable as the amount is dependent upon many conditions.

For purposes of calculation, it is conservatively assumed that all the fuel plate aluminum cladding exposed in the 41 of 86 area is involved in the reaction.

The reactor core contains a total of 33.56 Kg of aluminum.

Of this, 1.3% or 436 grams is assumed to react according to the following equation: A1 + n 2-I* A1On +nI-I 2 +heat.The energy release per Kg of aluminum is 18 MW-sec, for a total energy release of: 7.9 MW-sec = 7.5 x 10 3 BTU.This amount of heat would easily be transferred to the adjacent fuel elements and primary coolant in the core. Additionally, any steam that would form in the vicinity of the molten area would also assist in dissipating the heat. Since the fuel failure would result in a negligible release of energy to the primary coolant system, the introduction of pressure surges, which could lift the primary relief valves, are not considered credible.

The pressurizer is an isolated system, and since no significant pressure surges are anticipated, it will not be subject to mixing with the primary coolant system.Any significant gaseous radioactivity entrapped in the reactor loop vent tank will cause a reactor scram and actuation of the containment building isolation system by action of the pool surface radiation monitor. Additionally, following actuation of the anti-siphon system when the primary coolant system is secured, gases could also collect in the anti-siphon pressure tank. The location of these tanks under the pool surface, and the shielding provided by the water and the biological shield, will significantly reduce any radiation exposure to the reactor staff, visitors, or researchers.

Fission products entrapped in the primary coolant system can be removed by the reactor coolant cleanup system. This cleanup procedure would be undertaken under closely monitored and controlled conditions.

The primary coolant system does experience some coolant leakage into the. reactor pool through the pressure vessel head packing and flange gasket. This leakage is typically less than 40 gallons (151 1) per week; an almost imperceptible leakage rate of approximately 4 x 1 0- gallons of primary coolant per minute into the pool. However, for purposes of calculation, a leakage rate of 80 gallons (303 1) per week is used. Based on this assumed conservative leakage rate, the radiation exposure to personnel in the containment building following the fuel failure is calculated below.For operation at 10 MW for 1,200 MWD in twelve 10-day cycles over a 300-day period with 6.2 Kg of 2 3 5 U (normal operating cycle is 6.5 days with a total of less than 700 MVWD on the core), the following radioiodine, krypton and xenon activities will conservatively be present in the core (Ref.13.39).42 of 86 Radioiodine and Noble Gas Activities in the Core 1311 -- 1.7 x 10+0 Ci 85Kr -4.7 x 10+02 Ci '3 Xe -4.2 x 10+°5 Ci 132I -3.3 x 10+05 Ci 85m~r -1.1 x 10+05 Ci '3 5 Xe -9.6 x 10+04 Ci 133I -- 5.1 X 10o0 Ci 87Kr -2.1 x 10+05 Ci 1 3 5mXe -9.4 x 10+04 Ci 14-- 6.3 x 10+° Ci 88Kr -3.0 x 10+05 Ci 1 3 7 Xe -- 4.9 x 10+0 Ci 13SI -5.2 x 10+05 Ci 89Kr -3.8 x 10+° Ci '3 8 Xe -5.2 x 10+05 Ci 9°Kr -3.8 x 10+05 Ci 1 3 9 Xe -- 4.2 x 10"°5 Ci An unirradiated fuel plate number-i contains, on average, 19.26 grams of U-235, so four (4)unirradiated number-i fuel plates contain 77.04 grams instead of the 78.58 grams assumed in the fuel failure analysis.

These four number-i fuel plates that melt correspond to 1.41% of the total U-235 in the Week 58 Core that was used to determine the high power peaking factor for the revised SLs. The Week 58 Core has a total power history of 576 MWd. This power history results in a total reduced core mass of 5,474 grams of U-235 due to the previous fuel consumption.

This 1.41%of U-235 melting releases 3.42% of the core fission products due to the highest power density fuel plate number-1 overall peaking factor of 2.423, which is conservatively assumed to apply to all four (4) number-i fuel plates (1.41% x 2.423 = 3.42%).A conservative value of a 100% release of the radioiodine and noble gas fission products from the fuel is assumed in calculating the fission product inventory in the primary coolant system. It is also assumed that fission products released into the primary coolant are quickly and uniformly dispersed within the 2,000-gallon (7,571-1) primary coolant system volume and, during a normal week's operation, 80 gallons (7.9 x 1 0- gpm) of coolant leaks from the primary coolant system into the pool water. Therefore, the radioactivity released into the reactor pool in 10 minutes -determined to be the maximum personnel occupancy time in the containment building after the accident for necessary operational personnel

-is as follows: (Note: It would take approximately 5 minutes for Operations personnel to secure the primary coolant system and verify that the containment building has been evacuated following a containment building isolation.

For the purpose of the fuel failure calculations, a conservative assumption of 10 minutes is used.)Example calculation of 131I released into the reactor pool:=1 3 1 I in fuel x 0.0342 x 1/2,000 gal x (7.9 xl0"°3 gpm) x 10 min x 10+06 jtCi/Ci= (1.7 x i0+° Ci) x (1 .3509 x 10+° = 2.30 x 10+05 1 iCi Note: Same calculation is used for the other isotopes listed below.43 of 86 Radio iodine and Noble Gas Activities Released Into the Pool after 10 Minutes 3I-- 2.30 x 10+05 1.tCi 85Kr -6.35 x 10+02 1 iCi 1 3 3 Xe -5.67 x 10+05 13I -4.46 x 10+/-05 85m~r -1.49 x 10+°0 5 .Ci 135~ -1.30 x 10+05 1331 -- 6.89 x 10+05 87Kr -2.84 x 10+05 1 iCi l 3 5rage -1.27 x 10+0 ptCi 1341 -- 8.52 x 10+05 j.tCi 88Kr -4.04 x 10+05 gCi 1 3 7 Xe -6.63 x 10+° 13I- 7.02 x 10+05 jtCi 89Kr -5.13 x 10+05 xiCi 1 3 8 Xe -7.02 x 10+05 9°Kr -5.13 x 10+05 gCi 1 3 9 Xe -5.67 x 10+05 1 iCi Fission products released into the reactor pooi will be detected by the pool surface and ventilation system exhaust plenum radiation monitors.

However, for the purposes of this analysis, it is assumed that a reactor scram and actuation of the containment building isolation system occurs by action of the pooi surface radiation monitor.The radioiodine released into the reactor pool over a 10-minute interval is conservatively assumed to be instantly and uniformly mixed into the 20,000 gallons (75,708 1) of bulk pool water, which then results in the following pool water concentrations for the radioiodine isotopes.

The water solubility of the krypton and xenon noble gases released into the pool over this same time period are ignored and they are assumed to pass immediately through the pool water and evolve directly into the containment building air volume where they instantaneously form a uniform concentration in the isolated structure.

Radioiodine Concentrations in the Pool Water 131 -1.15 x 10+01 gxCi/gal '33I -3.44 X 10+01 gtCi/gal 1351 -3.51 x 10+/-01 1321 -2.23 x 10+01 pCi/gal 13I- 4.26 x 10+01 g1Ci/ga1 When the reactor is at 10 MW and the containment building ventilation system is in operation, the evaporation rate from the reactor pool is approximately 80 gallons (302.8 L) of water per day. For the purposes of this calculation, it is assumed that a total of 40 gallons (151 L) of pool water containing the previously listed radioiodine concentrations evaporates into the containment building over the 10 minute period. Containment air with a temperature of 75 0 F (23.9 °C) and 100%relative humidity contains H 2 0 vapor equal to 40 gallons (151.4 L) of water. Since the air in containment is normally at about 50% relative humidity, thus containing approximately 40 gallons (151 L) of water vapor, the assumed addition of 40 gallons (151 L) of water vapor will not cause the containment air to be supersaturated.

It is also conservatively assumed that all of the radioiodine activity in the 40 gallons (151 L) of pool water instantaneously forms a uniform concentration in the containment building air. When distributed into the containment building, this would result in the following radioiodine concentrations in the 225,000 ft 3 (6,371.3 in 3) air volume: Example calculation of 1311 released into containment air:-131I concentration in pool water x 40 gal x 1/225,000 ft 3 x 35.3 147 ft 3/m 3= 1.15 x 10+01 pCi/gal x (6.28 x 10-03 gal/mn 3)44 of 86

-7.22 x 10-o2 ptCi/mn 3 (7.22 x 10-° /.tCi/m 3) x (1 m 3/10 6 ml) =7.22 x 10.08 pCi/ml Note: Same calculation is used for the other isotopes listed below.The average radioiodine concentrations are the sum of the initial concentrations and the concentrations after 10 minutes decay divided by 2.Average Radioiodine Concentrations in the Containment Building Air during the 10 Minutes 1311 -- 7.22 x 10-°8 pCi/ml 132I -- 1.36 x 10-07 pCi/ml 133I -- 2.16 x 10-07 gCi/ml 134I -- 2.53 x 10-07 pxCi/ml 135I -2.18 x 10-° pxCi/ml As noted previously, the krypton and xenon noble gases released into the reactor pool from the primary coolant system during the assumed 1 0-minute interval following the fuel failure (Note: the primary coolant system is shut down and secured, and the leakage driving force is stopped within 10 minutes), are assumed to pass immediately through the pool water and enter the containment building air volume where they instantaneously form a uniform concentration in the isolated structure.

Based on the 225,000-ft 3 volume of containment building air and the previously listed Curie quantities of these gases released into the reactor pool, the maximum noble gas concentrations in the containment building at the end of 10 minutes would be as follows: Example calculation of 85Kr released into containment air:-85Kr activity x 1/225,000 ft 3 x 35.3 147 ft 3/m 3-6.35 x 10+02 uCi x (1.60 x 10-o4 1/in 3)-9.96 x 10.02 gxCi/m 3 (9.96 x 10.02 3) x (1 m 3/10 6 ml) = 9.96 x i0.0 ptCi/ml Note: Same calculation is used for the other isotopes listed below.The average noble gas concentrations are the sum of the initial concentrations and the concentrations after 10 minutes decay divided by 2.Average Noble Gas Concentrations in the Containment Building Air during the 10 Minutes Kr -9.96 x 10-°8 jiCi/ml 85m~r -2.30 x 10-°5 iCi/ml 87Kr -4.27 x 10°5 pCi/ml 88r- 6.22 x 10°5 ptCi/ml 89r- 4.47 x 10"°5 p.Ci/ml 9°r- 4.03 x 10"°5 pCi/ml 1 3 3Xe -l 3 SlXe _'3 8 Xe -'3 9 Xe -8.90 x 10-° pCi/ml 2.02 x 10-° gtCi/ml 1.63 x 10-°5 gCi/ml 6.05 x 10.05 jgCi/ml 8.88 x 10.05 pCi/ml 4.45 x 10.05 pCi/ml 45 of 86 The objective of this calculation is to present a worst-case dose assessment for an individual who remains in the containment building for 10 minutes following fuel failure. Therefore, as noted previously, the radioactivity in the evaporated pool water is assumed to be instantaneously and uniformly distributed into the building once released into the air.Based on the source term data provided, it is possible to determine the radiation dose to the thyroid from radioiodine and the dose to the whole body resulting from submersion in the airborne noble gases and radioiodine inside the containment building.

As previously noted, the exposure time for this dose assessment is 10 minutes.Because the airborne radioiodine source is composed of five (5) different iodine isotopes, it will be necessary to determine the dose contribution from each individual isotope and to then sum the results. Dose multiplication factors were established using the Derived Air Concentrations (DACs)for the "listed" isotopes in Appendix B of 10 CFR 20 and Appendix C of 10 CFR 835 for the"unlisted" submersion isotopes, and the radionuclide concentrations in the containment building (Attachment 11).Example calculation of thyroid dose due to 1311: The DAC can also be defined as 50,000 mnrem (thyroid target organ limit)/2,000 hrs, or 25 mrem/DAC-hr.

Additionally, 10 minutes of one DAC-br is 1.67 x 10-01 DAC-br.1311 concentration in containment 1311 DAC (10 CFR 20)Dose Multiplication Factor= 7.22 x 10.08 giCi/ml= 2.00 x 10.08 g.Ci/ml= (1311 concentration)

/ ('311 DAC)= (7.22 x 10.o8 pCi/ml) / (2.00 x 10.o8 pCi/ml)= 3.61 Therefore, a 10-minute thyroid exposure from 1311 is:= Dose Multiplication Factor x DAC Dose Rate x 10 minutes= 3.61 x (25 mren/DAC-hr) x (1.67 x 10-°1 DAC-hr)= 1.51x 10+°lmrem Note: Same calculation is used for the other radioiodines listed below.Derived Air Concentration Values and 10-Minute Exposures

-Radioiodine Radionuclide 1311 1321 133I 1341 1351 Derived Air Concentration 2.00 x 10-°8 1 Ci/ml 3.00 x 10-06 gtCi/ml 1.00 x 1 007 ptCi/ml 2.00 x 10-0 jiCi/ml 7.00 x 10.o iiCi/ml 10-Minute Exposure 1.51 x 10401 mrem 1.89 x 10-°' mrem 8.99 x 10+°° mrem 5.27 x 10.02 mnrem 1.30 x 10+° mrem Total =25.58 mrem 46 of 86 Doses from the kryptons and xenons present in the containment building are assessed in much the same manner as the radioiodines, and the dose contribution from each individual radionuclide must be calculated and then added together to arrive at the final noble gas dose. Because the dose from the noble gases is only an external dose due to submersion, and because the DACs for these radionuclides are based on this type of exposure, the individual noble gas doses for 10 minutes in containment were based on their average concentration in the containment air and the corresponding DAC.Example calculation of whole body dose due to Kr The DAC can also be defined as 5,000 mrenm/2,000 hrs, or 2.5 mnremiDAC-hr.

Additionally, 10 minutes of one DAC-hr is 1.67 x 1001 DAC-hr.85Kr concentration in containment

= 9.96 x 1 0.0 8 5 Kr DAC (10 CFR 20) = 1.00 x i0"° pCi/ml Dose Multiplication Factor = (8 5 Kr concentration)

/ (SSKr DAC)= (9.96 x 10.08 pCi/mI) / (1.00 x 10o4~ pCi/mi)= 0.001 Therefore, a 10-minute whole body exposure from 85Kr is:= Dose Multiplication Factor x DAC Dose Rate x 10 minutes= 0.001 x (2.5 mrem/DAC-hr) x (1.67 x 10"°1 DAC-hr)=4.15 xl1 0°4 mrem Note: Same calculation is used for the other noble gases listed below.The DACs and the 10-minute exposure for each radioiodine and noble gas are tabulated below.47 of 86 Derived Air Concentration Values and 10-Minute Exposures

-Noble Gases Radionuclide 8 5 Kr 85m~r 8 7 Kr 88Kr 8 9 Kr 90Kr 1 3 3 Xe 1 3 5 Xe l 3 5 rage 1 3 7 Xe 1 3 8 Xe 1 3 9 Xe Derived Air Concentration 1.00 x 10"° xtCi/ml 2.00 x 1005 iiCi/ml 5.00 x 10-°6 itCi/ml 2.00 x 100o6 p.Ci/ml 6.00 x 100o6 gCi/ml 6.00 x 10.o6 iCi/ml 1.00 x 10"04 pCi/ml 1.00 x 10-05 jCi/ml 9.00 x 10-06 xiCi/ml 6.00 x 10.06 gCi/ml 4.00 X 10-06 pCi/mi 6.00 x 10°6 jiCi/ml 10-Minute Exposure 4.51 x 10-°4 mrem 4.80 x 10-°" mrem 3.56 x 10400 mrem 1.30 x 10+01 mrem 3.11 x 10400 mrem 2.80 x 10+00°mrem 3.71 x 10-°1 mrem 8.43 x 10-°1 mrem 7.54 x l0-°' mrem 4.20 x 10+° mrem 9.25 x 1040° mrem 3.09 x 10+0 0 mrem Total = 41.42 mrenl To finalize the occupational dose in terms of Total Effective Dose Equivalent (TEDE) for a 1 0-minute exposure mn the containment building after target failure, the doses from the radioiodines and noble gases must be added together, and result in the following values: 10-Minute Dose from Radioidines and Noble Gases in the Containment Building Committed Dose Equivalent (Thyroid):

Committed Effective Dose Equivalent (Thyroid):

Committed Effective Dose Equivalent (Noble Gases): Total Effective Dose Equivalent (Whole Body): 25.58 mrem 0.77 mrem 41.42 mrem 42.18 mrem By comparison of the maximum TEDE and Committed Dose Equivalent (CDE) for those occupationally-exposed during fuel failure to applicable NRC dose limits in 10 CFR 20, the final values are shown to be well within the published regulatory limits and, in fact, lower than 1% of any occupational limit.Radiation shine through the containment structure was also evaluated when considering accident conditions and dose consequences to the public and MUJRR staff. Calculation of exposure rate from the target failure was performed using the computer program MicroShield 8.02 with a Rectangular Volume -External Dose Point geometry for the representation of the containment structure (Attachment 12). MicroShield 8.02 is a product of Grove Software and is a comprehensive photon/gamma ray shielding and dose assessment program that is widely used by industry for designing radiation shields.The exposure rate values provided below represents the radiation fields at 1 foot (30.5 cm) from a 12-inch thick ordinary concrete containment wall and at the Emergency Planning Zone (EPZ)48 of 86 boundary of 150 meters (492.1 ft). The airborne concentration source terms used to develop the exposure rate values are identical to those used for determining the dose to a worker within containment from noble gases. For radioiodine, the total iodine activity of the target was used for the dose calculations, not the amount that evaporated in 10 minutes. The source term also assumes a homogenous mixture of nuclides within the containment free volume.Radiation Shine through the Containment Building Expo sure Rate at 1-Foot from Containment Building Wall: 1.074 mrem/hr Exposure Rate at Emergency Planning Zone Boundary (150 meters): 0.007 mrem/hr A confirmatory analysis of the accident condition yielding the largest consequence was validated independently by the use of the MCNIP code. This analysis yielded a result 21% less than the Microshield method and results provided above.As noted earlier in this analysis, the containment building ventilation system will shut down and the building itself will be isolated from the surrounding areas. Fuel failure will not cause an increase in pressure inside the reactor containment structure; therefore, any air leakage from the building will occur as a result of normal changes in atmospheric pressure and pressure equilibrium between the inside of the containment structure and the outside atmosphere.

It is highly probable that there will be no pressure differential between the inside of the containment building and the outside atmosphere, and consequently there will be no air leakage from the building and no radiation dose to members of the public in the unrestricted area. However, to develop what would clearly be a worst-case scenario, this analysis assumes that a barometric pressure change had occurred in conjunction with the target failure. A reasonable assumption would be a pressure change on the order of 0.7 inches of Hig (25.4 mm of Hig at 60°C), which would then create a pressure differential of about 0.33 psig (2.28 kPa above atmosphere) between the inside of the isolated containment building and the inside of the adjacent laboratory building, which surrounds most of the containment structure.

Making the conservative assumption that the containment building will leak at the TS leakage rate limit [10% of the contained volume over a 24-hour period from an initial overpressure of 2.0 psig (13.8 kPa above atmosphere)], the air leakage from the containment structure in standard cubic feet per minute (scfmn) as a function of containment pressure can be expressed by the following equation: LR = 17.85 x (CP-14.7)1 1 2;where: LR = leakage rate from containment (scfmn); and CP = containment pressure (psia).The minimum Technical Specification free volume of the containment building is 225,000-ft 3 at standard temperature and pressure.

At an initial overpressure of 2.0 psig (13.8 kPa above atmosphere), the containment structure would hold approximately 255,612 standard cubic feet (scf)of air. A loss of 10%, from this initial overpres sure condition, would result in a decrease in air 49 of 86 volume to 230,051 scf. The above equation describes the leakage rate that results in this drop of contained air volume over 1,440 minutes (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).When applying the Technical Specification leakage rate equation to the assumed initial overpressure condition of 0.33 psig (2.28 kPa above atmosphere), it would take approximately 16.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> for the leak rate to decrease to zero from an initial leakage rate of approximately 10.3 scfm, which would occur at the start of the event. The average leakage rate over the 16.5-hour period would be about 5.2 scfln.Several factors exist that will mitigate the radiological impact of any air leakage from the containment building following target failure. First of all, most leakage pathways from containment discharge into the reactor laboratory building, which surrounds the containment structure.

Since the laboratory building ventilation system continues to operate during target failure, leakage air captured by the ventilation exhaust system is mixed with other building air, and then discharged from the facility through the exhaust stack at a rate of approximately 30,500 cfm..Mixing of containment air leakage with the laboratory building ventilation flow, followed by discharge out the exhaust stack and subsequent atmospheric dispersion, results in extremely low radionuclide concentrations and very small radiation doses in the unrestricted area. A tabulation of these concentrations and doses is given below. These values were calculated following the same methodology stated in Section 5.3.3 of Addendum 3 to the MiURR Hazards Summary Report [1].A second factor which helps to reduce the potential radiation dose in the unrestricted area relates to the behavior of radioiodine, which has been studied extensively in the containment mockup facility at Oak Ridge National Laboratory (ORNL). From these experiments, it was shown that up to 75%of the iodine released will be deposited in the containment vessel. If, due to this 75% iodine deposition in the containment building, each cubic meter of air released from containment has a radioiodine concentration that is 25% of each cubic meter within containment building air, then the radioiodine concentrations leaking from the containment structure into the laboratory building, in microcuries per milliliter, will be: Example calculation of 1311 released through the exhaust stack:-131I activity / (30,500 ft 3/min x 16.5 hr x 60 min/hr x 28,300 ml/ft 3)= 2.30 x 10+05 /8.55 x 10+'1 1ml-2.69 X 10-07 iiCi/ml (2.69 x 10-07 iiCi/ml) x (0.25) = 6.73 x 10.08 iiCi/ml Note: Same calculation is used for the other radioiodines listed below.Radioiodine Concentrations in Air Leaking from Containment 1311 -- 6.73 x 10-°8 jtCi/ml 133I -- 2.02 x 10-07 gtCi/ml 13I- 2.05 x 10-07 jiCi/ml 1321 -1.30 x 10-07 pCi/ml 34 -2.49 x 10-07 pCi/ml 50 of 86 Example calculation of 85Kr released through the exhaust stack:= 85}~. activity / (30,500 ft 3/min x 16.5 hr x 60 mmn/hr x 28,300 mi/fl 3)= 6.35 x 10+02 / 8.55 x 10+11 ml= 7.4 x1-w° ItCi/ml Note: Same calculation is used for the other noble gases listed below.Noble Gas Concentrations in Air Leaking from Containment and Exiting the Exhaust Stack 85Kr -7.43 x 10-1 pCi/ml 87Kr -3.32 x 10-0 jtCi/ml 89Kr -6.00 x 10-0 giCi/ml 85m~r -1.74 x 10-0 g.Ci/ml 88Kr -4.73 x 10.0 pxCi/ml 9°Kr -6.00 x 10-0 jiCi/ml 33e- 6.64 x 10-07 pCi/mil 3 smXe -- 1.49 x 10-07 ptCi/ml 1 3 8 Xe -8.22 x 10-07 pCi/ml 1 3 5 Xe -1.52 x 10-0 pCi/ml 1 3 7 Xe -7.76 x 10-07 pCi/mI '3 9 Xe -6.64 x 10-0 ptCi/ml Assuming, as stated earlier, that (1) the average leakage rate from the containment building is 5.2 scfrn, (2) the leak continues for about 16.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> in order to equalize the containment building pressure with atmospheric pressure, (3) the flow rate through the facility's ventilation exhaust stack is 30,500 scfrn, (4) the reduction in concentration from the point of discharge at the exhaust stack to the point of maximum concentration in the unrestricted area is a factor of 292 and (5) there is no decay of any radioiodines or noble gases, then the following concentrations of radioiodines and noble gases with their corresponding radiation doses will occur in the unrestricted area. The values listed are for the point of maximum concentration in the unrestricted area assuming a uniform, semi-spherical cloud geometry for noble gas submersion and further assuming that the most conservative (worst-case) meteorological conditions exist for the entire 16.5-hour period of containment leakage following target failure. Radiation doses are calculated for the entire 16.5-hour period. Dose values for the unrestricted area were obtained using the same methodology that was used to determine doses inside the containment building, and it was assumed that an individual was present at the point of maximum concentration for the full 16.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> that the containment building was leaking.A worst-case scenario dilution factor of 292 for effluent dilution using the Pasquill-Guifford Model for atmospheric dilution is used in this analysis.

We assume that all offsite (public) dose occurs under these atmospheric conditions at the site of interest, i.e. 760 meters North of MUIRR. In our case at 760 meters it occurs only during Stability Class F conditions; which normally only occur 11.4% of the time when the wind blows from the south. Thus this calculation is conservative.

10 CFR 20 Appendix B Effluent Concentration Limits are used for the "listed" isotopes.

An Effluent Concentration Limit of 2.0 x 10.08 itCi/ml is used for the "unlisted" isotopes, which equals the DAC/300 when using the DOE Part 835 Default DAC limits of 6.0 x 10-06 p.C/ml. Exposure at 1 DAC gives 5000 mrem per year whereas at the effluent concentration limit it is 50 mrem per year.This is a factor of 100 times less for the effluent concentration limit as compared to the DAC.Exposure at the effluent concentration limit assumes you are in that effluent concentration for 8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br /> per year. Thus, the time assumed to be exposed to the effluent concentration limit is a factor of 4.38 longer than the 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> per year that defines a DAC. The isotopes in question are based 51 of 86 on a default DAC limit of 6.0 x 1 0-06 for short-lived

(< 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> half-lives) submersion DAC 's in Appendix C of Part 835. No credit is taken for transit time from the stack to the receptor point nor is credit taken for decay inside containment until release. In the case of Kr-89 and Xe-i137 the transit time alone would be approximately one (1) half-life while the transit time for Kr-90 and Xe-139 would be at least four (4) half-lives.

Thus we believe that a factor of 300 reduction below the DAC value to establish the effluent concentration limit is warranted.

This reduction factor of 300 is consistent, in fact more conservative, than 10 CFR 20 Appendix B, as the N4RC calculates Effluent Concentration Limits for Submersion isotopes by dividing the DAC value by 219.Example calculation of whole body dose in the unrestricted area due to 1311: Conversion Factor: (Public dose limit of 50 mrerr/yr) x (1 yr/8760 hours) = 5.71 x 10.o mremihr 1311 concentration 131i effluent concentration limit 1311 Conversion Factor= 2.30 x 10-1° pCi/ml=2.00 x 10-10 gCi/ml= 5.71 X 10-° mremihr Therefore, a 16.5-hour whole body exposure from 1311 is:=1311 concentration

/ (131I effluent concentration limit x Conversion Factor x 16.5 hrs)= 2.30 x 10"1° giCi/ml / (2.00 x i0 1 0-1 gCi/ml x 5.71 x 10-03 mremihr x 16.5 brs)= 1.09 x10°0 1 mrem Note: Same calculation is used for the other isotopes (radioiodines and noble gases) listed below.Effluent Concentration Limits. Concentrations at Point of Maximum Concentration and Radiation Doses in the Unrestricted Area -Radioiodine Radionuclide 1311 132I 133I 134I 1351 Effluent Limit 2.00 x 10"1° giCi/ml 6.00 x 10.08 iiCi/ml 6.00 x 10-09 Maximum Concentration 1 2.30 x 10-10 giCi/ml 4.47 x 10.1° pCi/ml 6.90 x 10-10 iiCi/ml 8.54 x 10-l0 jiCi/ml 7.03 x 10-1° jCi/ml Radiation Dose 1.09 x 10-01 mrem 2.11 x 10.03 mrem 6.50 x 10-02 mrem 1.34 x 10-° mrem 1.10 x 10.02 mrem Total = 0.19 mrem Note 1: Maximum Concentrations are radioiodine and noble gas concentrations leaking from the containment building and exiting the exhaust stack reduced by a dilution factor of 292.52 of 86 Effluent Concentration Limits. Concentrations at Point of Maximum Concentration and Radiation Doses in the Unrestricted Area -Noble Gases Radionuclide 85Kr~~87Kr 8 8 Kr 8 9 Kr 90Kr 1 3 3Xe l 3 5 mXe 1 3 5mXe 1 3 7Xe Effluent Limit 7.00 x 10.0 ptCi/ml 1.00 x 10.07 pCi/ml 2.00 x 10-°8 jiCi/ml 9.00 x 10-09 pCi/ml 2.00 x 10.08 jiCi/ml 2.00 x 10.08 pCi/ml 5.00 X 10-07 iiCi/ml 7.00 x 10.08 aCi/ml 4.00 x 10.08 !iCi/ml 2.00 x 10.08 aCi/m1 2.00 x 10.08 iiCi/ml 2.00 x 10.0 iiCi/ml Maximum Concentration 1 2.54 x 10-12 pCi/ml 5.97 x 10-1°/Ci/ml 1.14 x 10.0 pCi/ml 1.62 x 1 0-0 gCi/ml 2.06 x 10.0 ptCi/ml 2.06 x 10-0 2.27 x 10.0 pCi/ml 5.21 x 10"1° pCi/ml 5.09 x 10-40 pCi/ml 2.66 x 10-0 aiCi/ml 2.81 x 10.0 pCi/mi 2.27 x 10.0 pCi/ml Radiation Dose 3.43 x 10-0 mrem 5.63 x 10-°4 mrem 5.36 x i0.0 mrem 1.69 x 10.02 mrem 9.69 x 10°3 mrem 9.69 x 1O 0-3 mrem 4.28 x 10-0 mrem 7.01 x 10-o4 mrem 1.20 x 10.03 mrem 1.25 x 10-02 mrem 1.33 x 1 0°2 mrem 1.07 x 10-02 mrem Total = 0.08 mrem Note 1: Maximum Concentrations are radioiodine and noble gas concentrations leaking from the containment building and exiting the exhaust stack reduced by a dilution factor of 292.To finalize the unrestricted dose in terms of Total Effective Dose Equivalent (TEDE), the doses from the radioiodines and noble gases must be added together, and result in the following values: Dose from Radioidines and Noble Gases in the Unrestricted Area Committed Effective Dose Equivalent (Radioiodine)

Committed Effective Dose Equivalent (Noble Gases)Total Effective Dose Equivalent (Whole Body)0.19 mrem 0.08 mrem 0.27 mrem Summing the doses from the noble gases and the radioiodines simply substantiates earlier statements regarding the very low levels in the unrestricted area should a failure of a fueled experiment occur, and should the containment building leak following such an event. Because the dose values are so low, the dose from the noble gases becomes the dominant value, but the overall TEDE is still only 0.19 mrem, a value far below the applicable 10 CFR 20 regulatory limit for the unrestricted area. Additionally, leakage in mechanical equipment room 114 from such items as valve packing, flange gaskets, pump mechanical seals, etc. was also considered in the fuel failure analysis.

A realistic leakage rate of 60 milliliters within the 10-minute time interval was used -after 10 minutes the primary coolant system would be shutdown, isolated and depressurized as part of the control room operator's actions. The additional contaminated water vapor and associated isotopes added to the facility ventilation exhaust system made a minimal (<1%) contribution to the total dose of an individual located in the facility.

Therefore, the dose contribution to the unrestricted area would be expected to be approaching zero.53 of 86 13.2.1.3 Conclusions Generally, the most severe condition which is analyzed with regard to reactor accidents is either a loss of primary coolant or a loss of primary coolant flow during reactor operation.

Both of these accidents are analyzed in this chapter and the results show no core damage. In addition, there are no other accidents that will result in a release of fission products from the reactor fuel, which is assumed in the fuel failure analysis.

Even if such an event were to occur, the anti-siphon and reactor loop vent systems are designed such that any released radioactivity would be contained in the primary coolant system.System design and operational procedures reduce the likelihood of any foreign material being introduced into the reactor core that could cause a partial flow blockage.

Calculations have been performed which indicate that even partial flow blockage to a fuel element will not result in cladding failure (Ref. 13.2). A considerable margin of safety has been designed into the system in this regard. Also, considering the results of the analyses which show no core damage in the event of a loss of primary coolant or a loss of primary coolant flow accident (See Sections 13.2.3 and 13.2.4), and in view of the design of the anti-siphon and reactor loop vent systems, it is concluded that there is no radiation risk to personnel in the reactor containment building or in the unrestricted area should one of these events occur.

References:

Same as those stated on pages v through vii of Chapter 13 of the SAR.54 of 86

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 507A4). 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, iWcredible, provide a supporting analysis demonstrating acceptable core cooling and peak fuel temperatures, or justify why no additional information is needed.In the second paragraph of the above question, the NRC staff's stated understanding is closer to what occurs during a Loss of Flow Accident (LOFA) but not for the hot-leg break Loss of Coolant Accident (LOCA). The difference is that during the hot-leg break LOCA some of the primary coolant that is lost from the primary coolant system (PCS5) piping is located in the reactor pool, but no primary coolant is lost during a LOFA. So, in the LOFA, the natural convention flow path described above is established and provides more than sufficient cooling for the reactor core after shutdown.

During a hot-leg break LOCA, the anti-siphon system actuates and injects air into the PCS vertical 12-inch diameter piping above the inverted loop to the level of the in-pool heat exchanger outlet. The expanding air quickly voids the upper section of the potential PCS natural convention flow path.Key PCS components for the LOFA and LOCA are described in Table 1 along with their RELAP Model component number. These components are also indicated in the vertical cross-sectional view of the reactor pool and in-pool portion of the PCS (Figure 1).55 of 86 Figure 1 -In-Pool Portion of the Primary Coolant System (with RELALP Model components identified) 56 of 86 Table 1 -RELAP Model of In-Pool Portion of Primary Coolant System RELAP No. Component Description 405-1 hn-Pool Heat Exchanger Upper Header 405-2 hn-Pool Heat Exchanger Vertical Finned Tubes 405-3 hn-Pool Heat Exchanger Lower Header 401-2 Last 4 feet of 6-inch Diameter Inlet Piping to hn-Pool Heat Exchanger Upper Header 407 PCS Vertical 12-inch Diameter Pipe above hn-Pool Heat Exchanger Outlet to Flanged Natural Circulation Piping 406 PCS Vertical 12-inch Diameter Pipe Above hnverted Loop to hn-Pool Heat Exchanger Outlet 139 Horizontal PCS Inlet Piping to Upper Section of Pressure Vessel 501 Pressure Vessel Above the Core to Pressure Vessel Head 100-3 Last 4.917 feet of Vertical Hot-Leg Piping Before Joining Pipe No. 101 101 PCS Horizontal 12-inch Diameter Pipe (Section of inverted Loop)102 PCS Downward Vertical 12-inch Diameter Pipe from the Normal Outlet End of No. 101 Towards the PCS Hot-Leg Outlet Isolation Valve With no pipe break occurring in the PCS during a LOFA, all of the above sections of the PCS stay filled with primary coolant. This results in the development of the natural circulation flow path described in Paragraph 2 of the above question.

However, for the hot-leg break LOCA, a double shear on the inlet and outlet sides of the hot-leg isolation valve is assumed, such that the hot-leg isolation valve is functionally eliminated.

With anti-siphon system air being injected into component 406, voiding starts at the higher elevated connected PCS components as indicated in Table 1 and Figure 1. With the air rising vertically, the following occurs:* Component 407 is void of water in approximately 5 seconds.* Component 401-2 is voided of water in about 8 seconds and the in-pool heat exchanger components (405-1, 405-2, 405-3) start draining.

Also, the following components are draining:* Components 139, 101, 406.* Components 405-1, 405-2 and 139 have less than 1% water 16 seconds after the break.* Component 406 joins them by 18 seconds.* By 60 seconds, components 405-3 and 101 have no water in them.It should be noted that components 40 1-2 and 139, which are on the cold-leg PCS inlet side of the reactor core, drain downward with the primary coolant which is flowing down the pressure vessel through the core and then up through the PCS hot-leg outlet piping until their upper level is in 57 of 86 equilibrium with the water level in component 101 or 100-3. With the in-pool portion of the PCS drained to this level, the natural circulation flow path through the in-pool heat exchanger is eliminated.

However, the hot-leg break LOCA RELAP analysis shows that the highest peak fuel center line temperature of 281.2 0 F (138.4 °C) occurs in fuel plate number-i, 0.2 seconds after the LOCA begins. After this initial peak temperature at the start of the transient, the next highest fuel plate centerline temperature of 231.7 °F (110.9 °C) occurs in plate number-22 at 22 seconds as shown in SARk Figure 13.20. The highest coolant channel temperature 219.0 0 F (103.9 °C) occurs in channel 7 at 123.3 seconds and in channel 6 at 123.4 seconds as shown in SARk Figure 13.21. There is sufficient heat transfer from the PCS to the pooi coolant due to conduction through the PCS piping to avoid any fuel damage.58 of 86

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 RAIJA.2 7 (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.Following the response to RAI 9.b is MURR's "Mishandling or Malfunction of Fuel" accident[referred to as the Fuel Handling Accident (FHA)] analysis using the same assumptions and methodologies as used in the Maximum Hypothetical Accident (MHA) (now referred to as the"Fuel Failure during Reactor Operation" accident) and Fueled Experiment Failure. The only exceptions are the source term, which is explained in the accident analysis, as well as the decay prior to the accident (which is once again explained in the analysis).

As discussed in the response to RAI 10O.a, the primary coolant system does not have to be secured for a failed fueled experiment or for a FHA. The only required action for Operations personnel is to verify that the containment building has been evacuated following a containment building isolation, which will occur during both of these accident scenarios.

MUJIRR performs an evacuation dr-ill every year and the typical time period for all personal to evacuate the containment building, including verification by Operations personnel, is two (2) to two and a half (2.5) minutes. For the purposes of the failed fueled experiment and FRA calculations, a conservative assumption of five (5) minutes is used for both accident scenarios.

Additionally, verifying that the reactor has shut down and containment has isolated only takes a few moments -all control blade positions, reactor power meters, and containment isolation valve and door indications are in clear view of the reactor operator in the control room.b. Provide the details of how the source term is determined, or justify' why no additional information is needed.As described in the FHA analysis above, the two most outer fuel plates of a fuel element, number-l and -24, are the plates most likely to be damaged during fuel handling.

The number-i fuel plate contains 19.26 grams of U-235 before irradiation.

The highest peak power density in the various 59 of 86 MUJRR core configurations occurs in fuel plate number-i of a previously unirradiated fuel element, which has a peaking factor of 4.116 -located between 14.75 to 15.75 inches down from the top of the fuel plate. The number-24 fuel plate has a lower peak power density and contains 45.32 grams of U-235, and has the most surface area to be damaged. To be conservative, the analysis assumes that 0.125 grams of U-235 is exposed from plate number-i during the FHA, which corresponds to removing a section of fuel meat from a plate that is 1 inch square and 5 mils thick. A power peaking factor of 4.116 is also applied.60 of 86 "Fuel Handling Accident (FHA)" All fuel handling is performed in accordance with Special Nuclear Material (SNM) Control and Accounting Procedures as outlined in the Operations Procedures.

Irradiated fuel is handled with a specially designed remote tool. The normal fuel handling tool is designed to provide a positive:indication of latching prior to movement of a fuel element. This feature is tested prior to any fuel handling sequence.

Fuel elements are always handled one at a time so that they are maintained in a criticality-safe configuration.

New or irradiated fuel may be stored in any one of 88 in-pool fuel storage locations (not including the core). These storage locations are designed to ensure a geometry such that the calculated Keff is less than 0.9 under all conditions of moderation, thus allowing sufficient convection cooling and providing sufficient radiation shielding.

So the fuel handling system provides a safe, effective and reliable means of transporting and handling reactor fuel from the time it enters the facility until it leaves. All cask lifting equipment, including the 15-ton capacity crane, is rigorously maintained, including preventive maintenance and magnetic particle testing, as appropriate.

Therefore, no specific accidents regarding the handling of fuel have been identified for the MUIRR. The probability of dropping a fuel element while underwater and damaging it severely enough to breach the fuel cladding was considered.

A conservative potential radionuclide release and calculation of the occupational exposure are included below.The following calculations determining the postulated dose from a potential release of radioactivity from a fuel element during a handling accident closely follow the "Fuel Failure during Reactor Operation" calculations for personal exposure due to a release of fission products.

The objective of these calculations is to present a worst-case dose assessment for a person who remains in the containment building for five (5) minutes following the release from a breached fuel element.M~URR's fuel cycle averages having about 40 fuel elements in the cycle -divided into 20 pairs of elements.

Paired elements are always loaded opposite each other in the core. All eight (8) fuel elements are replaced every refueling.

MURR has averaged refueling the core more than 52 times a year since 1977. This type of accident has never occurred at MIJRR during any of these fuel handlings.

The two outer fuel plates of a fuel element, number-i and -24, are the plates most likely to be damaged during fuel handling.

The number-i fuel plate contains 19.26 grams of U-235 before irradiation.

The highest peak power density in the various MURR core configurations occurs in fuel plate number-i of a previously unirradiated fuel element, which has a power peaking factor of 4.116 -located between 14.75 to 15.75 inches down from the top of the fuel plate. The number-24 fuel plate has the most surface area to be damaged; however, it has a lower peak power density and contains 45.32 grams of U-235. To be conservative, the analysis assumes that 0.125 grams of U-235 is exposed from plate number-1 during the FHA, which corresponds to removing a section of fuel meat from a plate that is 1 inch square and 5 mils thick. A power peaking factor of 4.116 is also applied.61 of 86 The following radioiodine, krypton and xenon activities will be present in the MURR core 30 minutes after shutdown from 10 MW full power operation.

Refuelings typically occur no sooner than an hour after shutdown.

This takes into account the time required to shut down the reactor, to secure the primary coolant system (required to stay in operation a minimum of 15 minutes after the control blades are fully inserted), and to remove the reactor pressure vessel head. For the purpose of the FHA calculations, a conservative assumption of 30 minutes is used.Radioiodine and Noble Gas Activities in the Core after 30-Minute Decay 1 3 1 I -9.93 x 10+04 Ci 85Kr -2.47 x 10+01 Ci 1 3 3 Xe -2.73 x 10+05 Ci 132I -2.68 x 10+o Ci 85mKr -1.29 x 10+05 Ci 1 3 5 Xe -1.13 x i0o Ci 1331 -- 5.65 x i0+0 Ci 87Kr -1.67 x 10+° Ci 1 3 5mXe -- 4.79 x 10+04 Ci 134I -- 5.80 X i0+0 Ci 88Kr -2.73 x 10+05 Ci '3 7 Xe -2.37 x 10+0 Ci 1351 -- 5.07 x 10+07 Ci 89}r -5.5 x 10+02 Ci 138X -1.22 x 10+0 5 Ci 9°Kr -6.66 x 10-12 Ci 1 3 9 Xe -8.33 x 10-09 Ci Fission products released into the reactor pool will be detected by the pool surface and ventilation system exhaust plenum radiation monitors.

However, for the purposes of this analysis, it is assumed that an actuation of the containment building isolation system occurs by action of the pool surface radiation monitor. Actuation of the isolation system will prompt Operations personnel to ensure that a total evacuation of the containment building is accomplished promptly, usually within two (2) to two and a half (2.5) minutes. A conservative 5-minute evacuation time is used as the basis for the stay time in the dose calculations for personnel that are in containment during the FRA.The following radioiodine and noble gas activities from 0.125 grams of U-235 from the peak power position of fuel plate number-i in the peak power density fuel element are assumed to instantaneously and homogenously distribute in the reactor pool.Example calculation of 1311 released into the reactor pool:= (1311 in fuel / 2 3 5 U in core) x 2 3 5 U exposed x Power Peaking Factor x 10+06 /xCi/Ci-= (9.93 x 10+04 Ci / 5,474 grams) x 0.125 grams x 4.116 x 10+06 jtCi/Ci= 9.33 10+06 1 iCi Example calculation of 85Kr released into the reactor pool:= (8 5 Kr in fuel / 2 3 5 U in core) x 2 3 5 U exposed x PPF x 10+06 iCi/Ci= (2.47 x 10+°' Ci / 5,474 grams) x 0.125 grams x 4.116 x 10+06 pCi/Ci= 2.32 x 10+03 Note: Same calculations are used for the other isotopes listed below.62 of 86 Radioiodine and Noble Gas Activities Released into the Pool 131I -- 9.33 x 10+06 kCi 85Kr -2.32 x 10+03 gCi 1 3 3 Xe -- 2.56 x 10+07 giCi 132I -2.52 x 10+0 giCi 85mKr -1.21 x 10+07 gxCi '3 5 Xe -1.06 x 10+07 /iCi 133I -5.31 x 10+07 g.Ci 8 7 Kr -1.57 x 10+0 1 iCi l 3 5mXe -4.50 x 10+06 jiCi 1341 -- 5.45 X 10+0 gtCi 88Kr -2.56 x 100 jiCi 1 3 7 Xe -2.22 x 10+05 gCi 35-- 4.76 x 10+07 ptCi 9K~r -5.25 x 10+0 ptCi '3 8 Xe -1.15 x 10+07 plCi 9°~Kr -6.26 x 10-'° ItCi 1 3 9 Xe -7.83 x 10-07 pCi The radioiodine released into the reactor pooi over a 5-minute interval is conservatively assumed to be instantly and uniformly mixed into the 20,000 gallons (75,708 1) of bulk pooi water, which then results in the following pool water concentrations for the radioiodine isotopes.

The water solubility of the krypton and xenon noble gases released into the pool over this same time period are ignored and they are assumed to pass immediately through the pool water and evolve directly into the containment building air volume where they instantaneously form a uniform concentration in the isolated structure.

Radioiodine Concentrations in the Pool Water 131I -- 4.67 x 10+02 pCilgal 1331 -- 2.66 x 10+03 giCi/gal 135I -2.38 x 10+03 PCi/gal 13I -1.26 x 10+03 pCi/gal 1341 -- 2.73 X 10+03 When the reactor is at 10 MW and the containment building ventilation system is in operation, the evaporation rate from the reactor pool is approximately 80 gallons (302.8 L) of water per day. For the purposes of this calculation, it is assumed that a total of 20 gallons (75.7 L) of pool water containing the previously listed radioiodine concentrations evaporates into the containment building over the 5 minute period. Containment air with a temperature of 75 0 F (23.9 °C) and 100% relative humidity contains H 2 0 vapor equal to 40 gallons (151.4 L) of water. Since the air in containment is normally at about 50% relative humidity, thus containing 20 gallons (75.7 L) of water vapor, the assumed addition of 20 gallons (75.7 L) of water vapor will not cause the containment air to be supersaturated.

It is also conservatively assumed that all of the radioiodine activity in the 20 gallons (75.7 L) of pool water instantaneously forms a uniform concentration in the containment building air. When distributed into the containment building, this would result in the following radioiodine concentrations in the 225,000 ft 3 (6,371.3 in 3) air volume: Example calculation of 131I released into containment air:= 131I concentration in pooi water x 20 gal x 1/225,000 ft 3 x 35.3 147 ft 3/m 3-4.67 x 10+02 ptCi/gal x (3.14 x 10.03 gal/in 3)-1.46 pCi/m 3 (1.46 pCi/in 3) x (1 m 3/10 6 ml) =1.46 x 10.06 gxCi/ml Note: Same calculation is used for the other isotopes listed below.63 of 86 The average radio iodine concentrations are the sum of the initial concentrations and the concentrations after 5 minutes decay divided by 2.Average Radioiodine Concentrations in the Containment Building Air during the 5 Minutes 31-- 1.46 x 10-06 pCi/mi 1331 -- 8.32 x 10°06 gtCi/ml 135I -- 7.44 x 10.06 gCi/ml 132j -3.91 x 10.06 gtCi/ml 14-- 8.28 x i0-°5 gCi/ml As noted previously, the krypton and xenon noble gases released into the reactor pool during the 5-minute interval following the EHA, are assumed to pass immediately through the pool water and enter the containment building air volume where they instantaneously form a uniform concentration in the isolated structure.

This assumption is extremely conservative since it ignores the known solubility of krypton and xenon noble gases in the 100 0 F (37.8 °C) pool water, which would reduce their release into the containment building.

Based on the 225,000-ft 3 volume of containment building air, and the previously listed curie quantities of these gases released into the reactor pool, the maximum noble gas concentrations in the containment structure at the end of 5 minutes would be as follows: Example calculation of 85Kr released into containment air:-85Kr activity x 1/225,000 ft 3 x 35.3 147 ft 3/m 3= 2.32 X 10+03 j.Ci x (1.60 x i0 0 4 1/in 3)= 3.64 x 1001 g.Ci/m 3 (3.64 x 10°1 gxCi/m 3) x (1 m 3/10 6 ml) = 3.64 x i0-0 7 plei/ml Note: Same calculation is used for the other isotopes listed below.The average noble gas concentrations are the sum of the initial concentrations and the concentrations after 5 minutes decay divided by 2.Average Noble Gas Concentrations in the Containment Building Air during the 5 Minutes Kr -3.64 x 10-07 FCi/ml 1 3 3 Xe -4.02 x 10-03 pCi/mi S5m~x -1.89 x 10-03 gCi/ml '3 5 Xe -1.66 x 1003 jiCi/ml 87r- 2.41 x 10.03 gCi/ml 1 3 smXe -6.35 x 10-04 iCi/ml 88r- 3.98 x i0-0 3 jiCi/ml 1 3 7 Xe -2.45 x i0-°5 gCi/mI 89Kr -5.49 x 10-0 gtCi/ml '3 8 Xe -1.61 x 10-°3 pCi/mi 9°Kr -4.92 x 10-20 gCi/ml 1 3 9 Xe -6.18 x10-17 Ci/ml The objective of this calculation is to present a worst-case dose assessment for an individual who remains in the containment building for 5 minutes following the ERA. Therefore, as noted previously, the radioactivity in the evaporated pool water is assumed to be instantaneously and uniformly distributed into the building once released into the air.64 of 86 Based on the source term data provided, it is possible to determine the radiation dose to the thyroid from radioiodine and the dose to the whole body resulting from submersion in the airborne noble gases and radioiodine inside the containment building.

As previously noted, the exposure time for this dose assessment is 5 minutes.Because the airborne radioiodine source is composed of five different iodine isotopes, it will be necessary to determine the dose contribution from each individual isotope and to then sum the results. Dose multiplication factors were established using the Derived Air Concentrations (DACs)for the "listed" isotopes in Appendix B of 10 CFR 20 and Appendix C of 10 CFR 835 for the"unlisted" submersion isotopes, and the radionuclide concentrations in the containment building (Attachment 11).Example calculation of thyroid dose due to 1311: The DAC can also be defined as 50,000 mnrem (thyroid target organ limit)/2,000 hrs, or 25 mrem/DAC-hr.

Additionally, 5 minutes of one DAC-hr is 8.33 x 10-°2 DAC-hr.1311 concentration in containment

= 1.46 x 10-o6 g.Ci/ml 131j DAC (10 CFR 20) = 2.00 x 10.o gCi/ml Dose Multiplication Factor =(1311 concentration)

/ (131I DAC)= (1.46 x 10-°6 gCi/ml) / (2.00 x 10.o8 gCi/ml)= 73 Therefore, a 5-minute thyroid exposure from 131j is:= Dose Multiplication Factor x DAC Dose Rate x 5 minutes= 73 x (25 mrenm/DAC-br) x (8.33 x 10.02 DAC-hr)--1.52 X10+°2 mrem Note: Same calculation is used for the other radioiodines listed below.Doses from the kryptons and xenons present in the containment building are assessed in much the same manner as the radioiodines, and the dose contribution from each individual radionuclide must be calculated and then added together to arrive at the final noble gas dose. Because the dose from the noble gases is only an external dose due to submersion, and because the DACs for these radionuclides are based on this type of exposure, the individual noble gas doses for 5 minutes in contaimnment were based on their average concentration in the containment air and the corresponding DAC.Example calculation of whole body dose due to 85Kr: The DAC can also be defined as 5,000 mrem/2,000 hrs, or 2.5 mrem/'DAC-hr.

Additionally, 5 minutes of one DAC-br is 8.33 x l0°2 DAC-hr.85Kr concentration in containment

= 3.64 X 10-07 gtCi/mnl 65 of 86 85Kr DAC (10 CFR 20)Dose Multiplication Factor= 1.00 x 10"04 pCi/ml= (8 5 Kr concentration)

/ (8 5 Kr DAC)= (3.64 x 10-07 ptCi/ml) / (1.00 x 10-04 jiCi/ml)= 0.00364 Therefore, a 5 minute whole body exposure from 85Kr is:=Dose Multiplication Factor x DAC Dose Rate x 5 minutes= 0.00364 x (2.5 mrem/DAC-hr) x (8.33 x 10-02 DAC-hr)= 7.58X10-0 4 mrem Note: Same calculation is used for the other noble gases listed below.The DACs and the 5-minute exposure for each radioiodine and noble gas are tabulated below.Derived Air Concentration Values and 5-Minute Exposures

-Radioiodine Radionuclide 1311 1321 1331 134j 1 3 5 1 Derived Air Concentration 2.00 x 10-°8 3.00 x 10-06 /xCi/ml 1.00 x 10-07/Ci/ml 2.00 x i0.0 gCi/inl 7.00 x 10-07 iiCi/ml 5-Minute Exposure 1.52 x 10+o 2 mrem 2.71 x 10+00 mrem 1.73 x 10+02 mrem 8.62 x 10-°1 mrem 2.21 x 10+°1 mrem Total = 351.44 mrem Derived Air Concentration Values and 5-Minute Exposures

-Noble Gases Radionuclide 8 5 Kr 85m}r 87Kr S88K 89Kr 9 OKr 1 3 3 Xe I 3 5 Xe 1 3 5mXe 1 3 8 Xe 1 3 9 Xe Derived Air Concentration 1.00 x 10-°4 iCi/ml 2.00 x 10.o jtCi/ml 5.00 x 100o6 gCi/ml 2.00 x 100o6 xtCi/ml 6.00 x 10.o6 JiCi/m1 6.00 x 10-°6 pCi/ml 1.00 x 10-04 pCi/mi 1.00 x 10-°5 pCi/ml 9.00 x 10-06 pCi/ml 6.00 x 10-06 gCi/ml 4.00 x 10-06 jiCi/ml 6.00 X 10-06 .tCi/ml 5-Minute Exposure 7.58 x 10-04 mrem 1.96 x 10+°1 me 1.00 x 10+02 mrem 4.14 x 10+02 mrem 1.91 x 10-°1 mrem 1.71 x 10-1 5 mrem 8.36 x 10+00 mrem 3.45 x 10+°1 mrem 1.47 x 10+°1 mrem 8.49 x 10-°1 mrem 8.37 x 10+°1mre 2.14 x 1012 mrem Total = 676.45 mrem 66 of 86 To finalize the occupational dose in terms of Total Effective Dose Equivalent (TEDE) for a 5-minute exposure in the containment building after a FHA, the doses from the radioiodines and noble gases must be added together, and result in the following values: 5-Minute Dose from Radioidines and Noble Gases in the Containment Building Committed Dose Equivalent (Thyroid) 351.44 mrem Committed Effective Dose Equivalent (Thyroid) 10.54 mrem Committed Effective Dose Equivalent (Noble Gases) 676.45 mrem Total Effective Dose Equivalent (Whole Body) 687.00 mrem By comparison of the maximum TEDE and Committed Dose Equivalent (CDE) for those occupationally-exposed during a FHA to applicable NRC dose limits in 10 CFR 20, the final values are shown to be well within the published regulatory limits and, in fact, lower than 15% of any occupational limit.Radiation shine through the containment structure was also evaluated when considering accident conditions and dose consequences to the public and MUIRR staff. Calculation of exposure rate from a FHA was performed using the computer program MicroShield 8.02 with a Rectangular Volume -External Dose Point geometry for the representation of the containment structure (Attachment 12). MicroShield 8.02 is a product of Grove Software and is a comprehensive photon/gamma ray shielding and dose assessment program that is widely used by industry for designing radiation shields.The exposure rate values provided below represents the radiation fields at 1 foot (30.5 cm) from a 12-inch thick ordinary concrete containment wall and at the Emergency Planning Zone (EPZ)boundary of 150 meters (492.1 ft). The airborne concentration source terms used to develop the exposure rate values are identical to those used for determining the dose to a worker within containment from noble gases. For radioiodine, the total iodine activity from the FHA was used for the dose calculations, not the amount that evaporated in 5 minutes. The source term also assumes a homogenous mixture of nuclides within the containment free volume.Radiation Shine through the Containment Building Expo sure Rate at 1-Foot from Containment Building Wall: 54.79 mrem/hr Exposure Rate at Emergency Planning Zone Boundary (150 meters): 0.371 mremn/hr A confirmatory analysis of the accident condition yielding the largest consequence was validated independently by the use of the MCNP code. This analysis yielded a result 21% less than the Microshield method and results provided above.As noted earlier in this analysis, the containment building ventilation system will shut down and the building itself will be isolated from the surrounding areas. A FHA will not cause an increase in pressure inside the reactor containment structure; therefore, any air leakage from the building will occur as a result of normal changes in atmospheric pressure and pressure equilibrium between the 67 of 86 inside of the contaimnment structure and the outside atmosphere.

It is highly probable that there will be no pressure differential between the inside of the containment building and the outside atmosphere, and consequently there will be no air leakage from the building and no radiation dose to members of the public in the unrestricted area. However, to develop what would clearly be a worst-case scenario, this analysis assumes that a barometric pressure change had occurred in conjunction with a FHA. A reasonable assumption would be a pressure change on the order of 0.7 inches of Hg (25.4 mm of Hg at 60 which would then create a pressure differential of about 0.33 psig (2.28 kPa above atmosphere) between the inside of the isolated containment building and the inside of the adjacent laboratory building, which surrounds most of the containment structure.

Making the conservative assumption that the containment building will leak at the TS leakage rate limit [10% of the contained volume over a 24-hour period from an initial overpressure of 2.0 psig (13.8 kPa above atmosphere)], the air leakage from the contaimnment structure in standard cubic feet per minute (scfm) as a function of containment pressure can be expressed by the following equation: LR = 17.85 x (CP-14.7)l" 2;where: LR = leakage rate from containment (scfmn); and CP = containment pressure (psia).The minimum Technical Specification free volume of the containment building is 225,000-ft 3 at standard temperature and pressure.

At an initial overpressure of 2.0 psig (13.8 kPa above atmosphere), the containment structure would hold approximately 255,612 standard cubic feet (scf)of air. A loss of 10%, from this initial overpres sure condition, would result in a decrease in air volume to 230,051 scf. The above equation describes the leakage rate that results in this drop of contained air volume over 1,440 minutes (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).When applying the Technical Specification leakage rate equation to the assumed initial overpressure condition of 0.33 psig (2.28 kPa above atmosphere), it would take approximately 16.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> for the leak rate to decrease to zero from an initial leakage rate of approximately 10.3 scfm, which would occur at the start of the event. The average leakage rate over the 16.5-hour period would be about 5.2 scfm.Several factors exist that will mitigate the radiological impact of any air leakage from the containment building following a FHA. First of all, most leakage pathways from containment discharge into the reactor laboratory building, which surrounds the containment structure.

Since the laboratory building ventilation system continues to operate during a FHA, leakage air captured by the ventilation exhaust system is mixed with other building air, and then discharged from the facility through the exhaust stack at a rate of approximately 30,500 cflm. Mixing of containment air leakage with the laboratory building ventilation flow, followed by discharge out the exhaust stack and subsequent atmospheric dispersion, results in extremely low radionuclide concentrations and very small radiation doses in the unrestricted area. A tabulation of these concentrations and doses 68 of 86 is given below. These values were calculated following the same methodology stated in Section 5.3.3 of Addendum 3 to the MUIRR Hazards Summary Report [1].A second factor which helps to reduce the potential radiation dose in the unrestricted area relates to the behavior of radioiodine, which has been studied extensively in the containment mockup facility at Oak Ridge National Laboratory (ORNL). From these experiments, it was shown that up to 75%of the iodine released will be deposited in the containment vessel [2]. If, due to this 75% iodine deposition in the containment building, each cubic meter of air released from containment has a radioiodine concentration that is 25% of each cubic meter within containment building air, then the radioiodine concentrations leaking from the containment structure into the laboratory building, in microcuries per milliliter, will be: Example calculation of 1311 released through the exhaust stack:-1311 activity / (30,500 ft 3/min x 16.5 hr x 60 mir/hr x 28,300 ml/ft 3)-9.33 x 10+06 1 iCi / 8.55 x 10+11 ml= 1.09 x 10.05 pCi/mi (1.09 x 10-°5 gtCi/ml) x (0.25) =2.73 x 10.06 pCi/ml Note: Same calculation is used for the other radioiodines listed below.Radioiodine Concentrations in Air Leaking from Containment and Exiting the Exhaust Stack 13I- 2.73 x 10.06 gtCi/ml 133I -1.55 x 10°5~ gCi/ml 3I-- 1.39 x 10°s~ pCi/ml 132I --7.37 x 10.06 giCi/ml 134I -- 1.59 x 10-°s pCi/ml Example calculation of 85Kr released through the exhaust stack:-85Kr activity / (30,500 ft 3/min x 16.5 hr x 60 mmn/hr x 28,300 ml/fl 3)-2.32 x 10+03 ptCi / 8.55 x 10+1n ml-2.71 x 10.09 jCi/ml Note: Same calculation is used for the other noble gases listed below.Noble Gas Concentrations in Air Leaking from Containment and Exiting the Exhaust Stack 8 5 Kr -2.71 x 10.09 jCi/ml 87Kr -1.84 x 10-05 gxCi/ml 89Kr -6.14 x 10.08 jCi/ml 8Smjr -1.42 x 10.05 iCi/ml 88Kr -3.00 x 10-05 jCi/ml 9°Kr -7.33 x 10-22 pCi/ml 33e- 3.00 x 10-0 pCi/ml 1 3 5 mXe -5.27 x 10-°6 gxCi/ml 1 3 8 Xe -1.35 x 10.0 gxCi/ml 3Xe- 1.24 x 10.05 pCi/ml 1 3 7 Xe -2.60 x 10-o7 1 iCi/ml 1 3 9 Xe -9.16 x 10-'9 pCi/ml Assuming, as stated earlier, that (1) the average leakage rate from the containment building is 5.2 scfm, (2) the leak continues for about 16.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> in order to equalize the containment building pressure with atmospheric pressure, (3) the flow rate through the facility's ventilation exhaust stack 69 of 86 is 30,500 scfmn, (4) the reduction in concentration from the point of discharge at the exhaust stack to the point of maximum concentration in the unrestricted area is a factor of 292 and (5) there is no decay of any radioiodines or noble gases, then the following concentrations of radioiodines and noble gases with their corresponding radiation doses will occur in the unrestricted area. The values listed are for the point of maximum concentration in the unrestricted area assuming a uniform, semi-spherical cloud geometry for noble gas submersion and further assuming that the most conservative (worst-case) meteorological conditions exist for the entire 16.5-hour period of containment leakage following a FHA. Radiation doses are calculated for the entire 16.5-hour period. Dose values for the unrestricted area were obtained using the same methodology that was used to deternine doses inside the containment building, and it was assumed that an individual was present at the point of maximum concentration for the full 16.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> that the containment building was leaking.A worst-case scenario dilution factor of 292 for effluent dilution using the Pasquill-Guifford Model for atmospheric dilution is used in this analysis.

We assume that all offsite (public) dose occurs under these atmospheric conditions at the site of interest, i.e. 760 meters North of MURR. In our case at 760 meters it occurs only during Stability Class F conditions; which normally only occur 11.4% of the time when the wind blows from the south. Thus this calculation is conservative.

10 CFR 20 Appendix B Effluent Concentration Limits are used for the "listed" isotopes.

An Effluent Concentration Limit of 2.0 x 10.08 iiCi/ml is used for the "unlisted" isotopes, which equals the DAC/300 when using the DOE Part 835 Default DAC limits of 6.0 x 10.06 giC/ml. Exposure at 1 DAC gives 5000 mrem per year whereas at the effluent concentration limit it is 50 mrem per year.This is a factor of 100 times less for the effluent concentration limit as compared to the DAC.Exposure at the effluent concentration limit assumes you are in that effluent concentration for 8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br /> per year. Thus, the time assumed to be exposed to the effluent concentration limit is a factor of 4.38 longer than the 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> per year that defines a DAC. The isotopes in question are based on a default DAC limit of 6.0 x 1 006 for short-lived

(< 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> half-lives) submersion DAC's in Appendix C of Part 835. No credit is taken for transit time from the stack to the receptor point nor is credit taken for decay inside containment until release. In the case of Kr-89 and Xe-1 37 the transit time alone would be approximately one (1) half-life while the transit time for Kr-90 and Xe-139 would be at least four (4) half-lives.

Thus we believe that a factor of 300 reduction below the DAC value to establish the effluent concentration limit is warranted.

This reduction factor of 300 is consistent, in fact more conservative, than 10 CFR 20 Appendix B, as the NRC calculates Effluent Concentration Limits for Submersion isotopes by dividing the DAC value by 219.Example calculation of whole body dose in the unrestricted area due to 1311: Conversion Factor: (Public dose limit of 50 mrem/yr) x (1 yr/8760 hours) =5.71 x 10-°3 mrem/hr 1 3 1I concentration

= 9.35 x 10.0 pCi/ml 131I effluent concentration limit =2.00 x 10-'° pCi/ml 1311 Conversion Factor = 5.71 x 10.03 mnrem/hr 70 of 86 Therefore, a 16.5-hour whole body exposure from 1311 is:=131I concentration

/ (1311 effluent concentration limit x Conversion Factor x 16.5 brs)= 9.35 x 10.09 tCi/ml / (2.00 x 10-'° x 5.71 x 10-03 mremlhr x 16.5 hrs)=4.40 x 10+°° mrem Note: Same calculation is used for the other isotopes (radioiodines and noble gases) listed below.Effluent Concentration Limits. Concentrations at Point of Maximum Concentration and Radiation Doses in the Unrestricted Area -Radioiodine Radionuclide 1311 I321 1331 1341 1351 Effluent Limit 2.00 x 10-1° p.Ci/ml 2.00 x 10.08 pCi/ml 1.00 X 10-0 jiCi/ml 6.00 x 10-0 p.Ci/ml 6.00 x 10.09 tCi/ml Maximum Concentratior 2.52 x 10.0 pCi/ml 5.32 x 10-° jiCi/ml 5.46 x 10-0 jiCi/ml a' Radiation Dose 4.40 x 10+° mrem 1.19 x 10-°1 mrem 5.01 x 10+00 mrem 8.57 x 10-02 mrem 7.49 x 10-°1 mrem Total = 10.37 mrem Note 1: Maximum Concentrations are radioiodine and noble gas concentrations leaking from the containment building and exiting the exhaust stack reduced by a dilution factor of 292.Effluent Concentration Limits, Concentrations at Point of Maximum Concentration and Radiation Doses in the Unrestricted Area -Noble Gases Radionuclide 8 5 Kr 8 7 Kr 8 t Kr 8 9 Kr 9 0 Kr 1 3 smXe Effluent Limit 7.00 x 10.0 pCi/ml 1.00 x 10.07 pCi/mi 2.00 x 10.08 xCi/ml 9.00 x 10-09 pCi/ml 2.00 x 10.08 pCi/mi 2.00 x 10-° jxCi/ml 5.00 x 10-0 gCi/ml 7.00 x 10.08 xCi/ml 4.00 x i0.0 gCi/ml 2.00 x 10.0 gCi/ml 2.00 x 10.08 gCi/ml 2.00 x 10-08 jxCi/ml Maximum Concentration 1 9.30 x 10-12 igCi/ml 4.85 x 10-°8/.tCi/ml 6.29 x 10-0 8 pCi/ml 1.03 x 10-07 pCi/ml 2.51 x 10-24 pCi/ml 1.03 x 10-0 pCi/mI 4.25 x 10-0 pCi/ml 1.80 x 10-08 itCi/ml 4.61 x 10-0 pCi/ml 3.14 x 10-21 p.Ci/ml Radiation Dose 1.25 x 10-06 mrem 4.57 x 10.02 mrem 2.96 x 10.01 mrem 1.07 x 10+0° mrem 9.91 x 10-0 4 mrem 1.18 x 10-'7 mrem 1.93 x 10-0 2 nmrem 5.72 x 10.02 mrem 4.25 x 10.02 mrem 4.19 x 10-0 mrem 2.17 x 10-°1 mrem 1.48 x 10"1 4 mrem Total = 1.76 mrem Note 1: Maximum Concentrations are radioiodine and noble gas concentrations leaking from the containment building and exiting the exhaust stack reduced by a dilution factor of 292.71 of 86 To finalize the unrestricted dose in terms of Total Effective Dose Equivalent (TEDE), the doses from the radioijodines and noble gases must be added together, and result in the following values: Dose from Radioidines and Noble Gases in the Unrestricted Area Committed Effective Dose Equivalent (Radioiodine) 10.37 mrem Committed Effective Dose Equivalent (Noble Gases) 1.76 mrem Total Effective Dose Equivalent (Whole Body) 12.13 mrem Summing the doses from the noble gases and the radioiodines simply substantiates earlier statements regarding the very low levels in the unrestricted area should a FHA occur, and should the containment building leak following such an event. Because the dose values are so low, the dose from the noble gases becomes the dominant value, but the overall TEDE is still only 12.13 mrem, a value far below the applicable 10 CFR 20 regulatory limit for the unrestricted area.72 of 86

10. NUREG-153 7, 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 O CFR Part 20. The response to RAJ 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] l x0-7 jCi/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.Following the response to RAI 10O.c is the revised fueled experiment failure analysis that replaces the one (RAI 13.9.a) that was submitted as part of the responses, by letter dated October 29, 2010, to a Request for Additional Information made by the NRC (by letter dated May 6, 2010). As previously discussed in the response to Question 6.c, and what is stated on Page 13-5 of the SAR, the evacuation time for the MHA is 10 minutes based on the following: "It would take approximately 5 minutes for Operations personnel to secure the primary coolant system and verify that the containment building has been evacuated following a containment building isolation.

For the purpose of the MHA calculations, a conservative assumption of 10 minutes is used." However, the primary coolant system (PCS5) does not have to be secured for a failed fueled experiment or for a fuel handling accident (FHA). The only required action for Operations personnel is to verify that the containment building has been evacuated following a containment building isolation, which will occur during both of these accidents.

MURR perforns an evacuation drill every year and the typical time for all personal to evacuate the containment building, including verification by Operations personnel, is two (2) to two and a half (2.5) minutes. For the purposes of the failed fueled experiment and FHA calculations, a conservative assumption of five (5)minutes is used for both accident scenarios.

Additionally, verifying that the reactor has shut down and the containment building has isolated only takes a few moments -all control blade positions, and containment isolation valve and door indications are in clear view of the reactor operator in the control room.The Derived Air Concentration (DAC) values used for the dose calculations for each accident scenario -MHA (Now Fuel Failure During Reactor Operation), FHA and fueled experiment failure-are now the same. For the isotopes "listed" in Appendix B of 10 CFR 20, those DACs are used 73 of 86 whereas for the "unlisted" isotopes the DACs of 10 CFR 835 are used (published in the Federal Register, 72 FR 31940, June 8, 2007, as amended) (Attachment 11).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 Strontiunm-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. C'larfify whether fueled experiments are vented or not and revise the TS if required, or justify why no additional information is needed.License Amendment No. 34, issued to MURR on October 10, 2008, by the NRC, revised Technical current Specification (TS) 3 .6.o (relicensing TS 3.8.o) such that fueled experiments containing inventories of iodine-13 1 1-1 31) through 1-135 greater than 1.5 curies or inventories of strontium-90 (Sr-90) greater than 5 millicuries can be encapsulated in irradiation containers designed to meet the internal pressure design requirements specified in TS 3.6.i. TS 3.6.i states that "Irradiation containers to be used in the reactor, in which static pressure will exist or in which a pressure buildup is predicted, shall be designed and tested for a pressure exceeding the maximum expected pressure by at least a factor of two (2)." Until then, fueled experiments containing inventories of I-131 through 1-135 greater than 1.5 curies or inventories of Sr-90 greater than 5 millicuries had to be vented to the facility ventilation exhaust stack through high efficiency particulate air (H7EPA) and charcoal filters which were continuously monitored for radiation levels.Since Amendment No. 34 was issued after the SAR was submitted in August 2006 as a part of relicensing, SAR Section 13.2.6.2 is now outdated.

The third bullet on page 13-67 should now read, "Fueled experiments containing inventories of iodine-13 1 through iodine-i135 greater than 1.5 curies or strontium-90 greater than 5 millicuries shall be in irradiation containers that satisfy the requirements of Specification 3.8 .i or be vented to the facility ventilation exhaust stack through high efficiency particulate air (HEPA) and charcoal filters which are continuously monitored for an increase in radiation levels." 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.As discussed in the responses to Questions 1 .a, 1 .b and 1 .c, which are included in the responses, dated July 31, 2015, to a Request for Additional Information made by the NRC (by letter dated June 18, 2015), all air exiting the facility through the ventilation exhaust system is monitored for airborne radioactivity by the Off-Gas Radiation Monitoring System (also see SAR Section 7.9.5).This includes the exhaust from all hot cells, glove boxes, fume hoods, selected areas within the containment building and any experiment that is directly vented to the ventilation exhaust system.74 of 86 Technical Specification 3.7 provides the Limiting Conditions for Operation (LCO) for the radiation monitoring systems and airborne effluents.

As stated in Section B. 1.2 of SAR Appendix B, Argon-41 (Ar-4 1) accounts for greater than 99 % of the radioactivity released from the facility through the ventilation exhaust system; therefore, Ar-4 1 was used to determine the radiological impact of airborne effluents during normal reactor operation.

In addition to At-4 1, all other isotopes greater than 0.0001% of the limits of TS 3.7 are reported to the NRC annually as required by TS 6.6.e.(6), which states, "A summary of the nature and amount of radioactive effluents released or discharged to the environs beyond the effective control of the licensee as measured at or prior to the point of such release or discharge."~

Attachment 13 (also included in the responses, dated July 31, 2015) provides the last 10 years, and average, of air releases from the facility per isotope in percentage of the Technical Specification limit. As you will note, with the exception of argon-4 1, all other isotopes discharged are less than 0.6% of the release limit.75 of 86 Revised "Fueled Experiment Failure" (MURR' s new Maximum Hypothetical Accident)The release of the radioisotopes of krypton, xenon and iodine from a 5-gram low-enriched uranium (LEU) target is the major source of radiation exposure to an individual and will, therefore, serve as the basis for the source term for these dose calculations.

A 5-gram LEU target irradiated for 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> (normal weekly operating cycle) at a thermal neutron flux of 1.5 x 10+13 nlcm 2-sec will produce the following radioiodine, krypton and xenon activities (additionally, approximately 1.40 x 10+04 jiCi of Strontium-90 will be produced):

Radioiodine and Noble Gas Activities in a 5-Gram LEU Target 131I -- 8.400 Ci 85Kr -0.002 Ci 1 3 3 Xe -18.900 Ci 1321 -18.600 Ci 85m~r -7.580 Ci 1 3 5 Xe -13.600 Ci 133I -- 39.900 Ci 87Kr -15.400 Ci l 3 5 mXe -6.760 Ci 134I -- 45.400 Ci 88Kr -21.700 Ci 1 3 7 Xe -35.800 Ci 13I- 37.700 Ci 8 9 Kr -27.740 Ci 1 3 8 Xe -37.400 Ci 9°Kr -27.400 Ci 1 3 9 Xe -30.700 Ci Total Iodine -150.00 Ci Total Krypton -99.822 Ci Total Xenon -143.160 Ci A complete failure of the target is unrealistic for many reasons. The worst that can be expected is partial melting; however, in order to present a worst-case dose assessment for an individual that remains in thle containment building following target failure, 100% of the total activity of the target is assumed to be released into the reactor pool.Fission products released into the reactor pool will be detected by the pool surface and ventilation system exhaust plenum radiation monitors.

However, for the purposes of this analysis, it is assumed that a reactor scram and actuation of the containment building isolation system occurs by action of the pool surface radiation monitor. Actuation of the isolation system will prompt Operations personnel to ensure that a total evacuation of the containment building is accomplished promptly, usually within two (2) to two and a half (2.5) minutes. A conservative 5-minute evacuation time is used as the basis for the stay time in the dose calculations for personnel that are in containment during target failure.The radioiodine released into the reactor pool over a 5-minute interval is conservatively assumed to be instantly and uniformly mixed into the 20,000 gallons (75,708 1) of bulk pool water, which then results in the following pool water concentrations for the radioiodine isotopes.

The water solubility of the krypton and xenon noble gases released into the pool over this same time period are conservatively ignored and they are assumed to pass immediately through the pool water and evolve directly into the containment building air volume where they instantaneously form a uniform concentration in the isolated structure.

76 of 86 Radioiodine Concentrations in the Pool Water 1311 -- 4.20 x 10+02 gCi/gal 31-- 2.00 x 10+0 pCi/gal 1 3 5 1 -- 1.89 x 10+0 gtCi/gal 1321 -- 9.30 x 10+02 pCi/gal 1341 -- 2.27 x 10+03 pCi/gal When the reactor is at 10 MW and the containment building ventilation system is in operation, the evaporation rate from the reactor pooi is approximately 80 gallons (302.8 L) of water per day. For the purposes of this calculation, it is assumed that a total of 20 gallons (75.7 L) of pool water containing the previously listed radioiodine concentrations evaporates into the containment building over the 5 minute period. Containment air with a temperature of 75 0 F (23.9 °C) and 100% relative humidity contains H 2 0 vapor equal to 40 gallons (151.4 L) of water. Since the air in containment is normally at about 50% relative humidity, thus containing 20 gallons (75.7 L) of water vapor, the assumed addition of 20 gallons (75.7 L) of water vapor will not cause the containment air to be supersaturated.

It is also conservatively assumed that all of the radioiodine activity in the 20 gallons (75.7 L) of pool water instantaneously forms a uniform concentration in the containment building air. When distributed into the containment building, this would result in the following radioiodine concentrations in the 225,000 ft 3 (6,371.3 in 3) air volume: Example calculation of 131I released into containment air:=1311 concentration in pool water x 20 gal x 1/225,000 ft 3 x 35.3 147 ft 3/m 3-4.20 x 10+02 pCilgal x (3.14 x 10.03 gal/in 3)-1.32 jiCi/m 3 (1.32 pCi/in 3) x (1 m 3/10 6 ml) = 1.32 x 10.06 gCi/ml Note: Same calculation is used for the other isotopes listed below.The average radioiodine concentrations are the sum of the initial concentrations and the concentrations after 5 minutes decay divided by 2.Average Radioiodine Concentrations in the Containment Building Air during the 5 Minutes 1311 -- 1.32 x 10-°6 ixtCi/ml 133I -- 6.26 x 10-°6 giCi/ml 1351 -5.89 x 10-°6 giCi/ml 1321 -2.88 x 10-06 pCi/ml 1341 -6.90 x 10-°6 ptCi/ml As noted previously, the krypton and xenon noble gases released into the reactor pool from the 5-gram LEU target during the 5-minute interval following failure, are assumed to pass immediately through the pool water and enter the containment building air volume where they instantaneously form a uniform concentration in the isolated structure.

Based on the 225,000-ft 3 volume of containment building air, and the previously listed curie quantities of these gases released into the reactor pool, the maximum noble gas concentrations in the containment structure at the end of 5 minutes would be as follows: 77 of 86 Example calculation of 85Kr released into containment air:= 85Kr activity x 1/225,000 ft 3 x 35.3 147 ft 3/m 3= 1.71 x 10+0 x (1.60 x 10.0 1/mn 3)-2.69 x 10-°1 jiCi/m 3 (2.69 x 10-°1 iiCi/m 3) x (1 m 3/10 6 ml) = 2.69 x 10-07 pCi/ml Note: Same calculation is used for the other isotopes listed below.The average noble gas concentrations are the sum of the initial concentrations and the concentrations after 5 minutes decay divided by 2.Average Noble Gas Concentrations in the Containment Building Air during the 5 Minutes Kr -2.69 x 10-0 pCi/mI 1 3 3 Xe -2.97 x 10-03 pCi/ml 8mr- 1.18 x 10-03 gCi/ml '3 5 Xe -2.13 x 10.03 pCi/ml 87Kr -2.36 x 10-0 jiCi/ml l 3 5 mXe -9.54 x 10-° ~tCi/ml 88Kr -3.37 x 10.0 pCi/mi 1 3 7 Xe -3.95 x 10-0 iiCi/ml 89r- 2.90 x 10.03 xiCi/ml '3 8 Xe -5.23 x 10-0 jiCi/ml 9°Kr -2.15 x i0-0 kCi/ml '3 9 Xe -- 2.42 x 10-0 jiCi/ml The objective of this calculation is to present a worst-case dose assessment for an individual who remains in the containment building for 5 minutes following target failure. Therefore, as noted previously, the radioactivity in the evaporated pool water is assumed to be instantaneously and uniformly distributed into the building once released into the air.Based on the source term data provided, it is possible to determine the radiation dose to the thyroid from radioiodine and the dose to the whole body resulting from submersion in the airborne noble gases and radioiodine inside the containment building.Because the airborne radioiodine source is composed of five different iodine isotopes, it will be necessary to determine the dose contribution from each individual isotope and to then sum the results. Dose multiplication factors were established using the Derived Air Concentrations (DACs)for the "listed" isotopes in Appendix B of 10 CFR 20 and Appendix C of 10 CFR 835 for the"unlisted" submersion isotopes, and the radionuclide concentrations in the containment building (Attachment 11).Example calculation of thyroid dose due to 131I: The DAC can also be defined as 50,000 mrem (thyroid target organ limit)/2,000 brs, or 25 mrem/DAC-hr.

Additionally, 5 minutes of one DAC-hr is 8.33 x 10.02 DAC-hr.131I concentration in containment

=1.32 x 10-06 pCi/ml'3 1 1DAC (10 CFR 20) =2.00 x 10-°8 pCi/ml 78 of 86 Dose Multiplication Factor= (1311 concentration)

/(1311 DAC)= (1.32 x 10-06 pCi/ml) / (2.00 x 10-08 ptCi/ml)= 66 Therefore, a 5-minute thyroid exposure from 1311 is:= Dose Multiplication Factor x DAC Dose Rate x 5 minutes of a DAC-hr= 66 x (25 mrem/DAC-hr) x (8.33 x 10.02 DAC-hr)= 1.37 x10+0 2 mrem Note: Same calculation is used for the other radioiodines listed below.Derived Air Concentration Values and 5-Minute Exposures

-Radioiodine Radionuclide 1311 1321 1 3 3 I 1 3 4 I 1351 Derived Air Concentration 2.00 x 10.08 3.00 x 10-06 gCi/ml 1.00 X 10-0 giCi/ml 2.00 x 10-° iiCi/mnl 7.00 X 10-7/Ci/ml 5-Minute Exposure 1.37 x 10+02 mrem 2.00 x 10+°° mrem 1.30 x 10+02 mrem 7.18 xlO 0-1 mrem 1.75 x 10+01 mrem Total = 287.80 mrem Doses from the kryptons and xenons present in the containment building are assessed in much the same manner as the radioiodines, and the dose contribution from each individual radionuclide must be calculated and then added together to arrive at the final noble gas dose. Because the dose from the noble gases is only an external dose due to submersion, and because the DACs for these radionuclides are based on this type of exposure, the individual noble gas doses for 5 minutes in containment were based on their average concentration in the containment air and the corresponding DAC.Example calculation of whole body dose due to Kr The DAC can also be defined as 5,000 mrem/2,000 hrs, or 2.5 mremiDAC-hr.

Additionally, 5 minutes of one DAC-hr is 8.33 x 10.02 DAC-hr.85Kr concentration in containment 85Kr DAC (10 CFR 20)Dose Multiplication Factor= 2.69 X 10.0 7 pxCi/ml= 1.00 x 10.04 pxCi/ml= (85Kr concentration)

/ (8 5 Kr DAC)= (2.69 X 10-07 g.Ci/ml) / (1.00 x 10.04 pCi/ml)= 0.00269 Therefore, a 5 minute whole body exposure from 8 5 Kr is:= Dose Multiplication Factor x DAC Dose Rate x 5 minutes of a DAC-hr= 0.00269 x (2.5 mnrem/DAC-hr) x (8.33 x 10-02 DAC-hr)79 of 86

= 5.59 x 10-°4mrem Note: Same calculation is used for the other noble gases listed below.Derived Air Concentration Values and 5-Minute Exposures

-Noble Gases Radionuclide 8 5 Kr 85m~r 8 7 Kr 8 8 Kr 89Kr 9 0 Kr 1 3 3mXe 1 3 5Xe'3 9 Xe Derived Air Concentration 1.00 x 10"0 pCi/ml 2.00 x 10.05 pCi/ml 5.00 x 10.06 pCi/ml 2.00 x 10-06 gCi/ml 6.00 x 10-°6 pCi/ml 6.00 x 10-06 pCi/ml 1.00 X 10-04 pCi/ml 1.00 x i0-05 pCi/ml 9.00 x 10-06 pCi/ml 6.00 x 10-06 pCi/mi 4.00 x 10-°6 jiCi/ml 6.00 X 10-°6 ptCi/ml 5-Minute Exposure 5.59 x 10"0 mrem 1.23 x 10+°a mrem 9.85 x 10+°1 mrem 3.51 x 10+02 mrem 1.01 x 10+0 2 mrem 7.48 x 10+°1 mrem 6.18 x 10+°° mrem 4.43 x 1° mrem 2.21 x 10+°1 mrem 1.37 x 10+02 mrem 2.72 x 10+0 2 mrem 8.41 x 10+°1 mrem Total =1203.80 mrem To finalize the occupational dose in terms of Total Effective Dose Equivalent (TEDE) for a 5-minute exposure in the containment building after target failure, the doses from the radioiodines and noble gases must be added together, and result in the following values: 5-Minute Dose from Radioidines and Noble Gases in the Containment Building Committed Dose Equivalent (Thyroid)Committed Effective Dose Equivalent (Thyroid)Committed Effective Dose Equivalent (Noble Gases)Total Effective Dose Equivalent (Whole Body)287.80 mrem 8.63 mrem 1203.80 mrem 1212.44 mrem Note: The addition of Strontium-90 (9°Sr) will increase the above stated TEDE (whole body) by 9.15 mrem (<1%).By comparison of the maximum TEDE and Committed Dose Equivalent (CDE) for those occupationally-exposed during target failure to applicable NRC dose limits in 10 CFR 20, the final values are shown to be well within the published regulatory limits and, in fact, lower than 25% of any occupational limit.Radiation shine through the containment structure was also evaluated when considering accident conditions and dose consequences to the public and MUIRR staff. Calculation of exposure rate from the target failure was performed using the computer program MicroShield 8.02 with a 80 of 86 Rectangular Volume -External Dose Point geometry for the representation of the containment structure (Attachment 12). MicroShield 8.02 is a product of Grove. Software and is a comprehensive photon/gamma ray shielding and dose assessment program that is widely used by industry for designing radiation shields.The exposure rate values provided below represents the radiation fields at 1 foot (30.5 cm) from a 12-inch thick ordinary concrete containment wall and at the Emergency Planning Zone (EPZ)boundary of 150 meters (492.1 ft). The airborne concentration source terms used to develop the exposure rate values are identical to those used for determining the dose to a worker within containment from noble gases. For radioiodine, the total iodine activity of the target was used for the dose calculations, not the amount that evaporated in 5 minutes. The source term also assumes a homogenous mixture of nuclides within the containment free volume.Radiation Shine through the Containment Building Exposure Rate at 1-Foot from Containment Building Wall: 68.87 mrem/hr Exposure Rate at Emergency Planning Zone Boundary (150 meters): 0.467 mrem/hr A confirmatory analysis of the accident condition yielding the largest consequence was validated independently by the use of the MCNP code. This analysis yielded a result 21% less than the Microshield method and results provided above.As noted earlier in this analysis, the containment building ventilation system will shut down and the building itself will be isolated from the surrounding areas. Target failure will not cause an increase in pressure inside the reactor containment structure; therefore, any air leakage from the building will occur as a result of normal changes in atmospheric pressure and pressure equilibrium between the inside of the containment structure and the outside atmosphere.

It is highly probable that there will be no pressure differential between the inside of the containment building and the outside atmosphere, and consequently there will be no air leakage from the building and no radiation dose to members of the public in the unrestricted area. However, to develop what would clearly be a worst-case scenario, this analysis assumes that a barometric pressure change had occurred in conjunction with the target failure. A reasonable assumption would be a pressure change on the order of 0.7 inches of Hig (25.4 mm of Hg at 60 0 C), which would then create a pressure differential of about 0.33 psig (2.28 kPa above atmosphere) between the inside of the isolated containment building and the inside of the adjacent laboratory building, which surrounds most of the containment structure.

Making the conservative assumption that the containment building will leak at the Technical Specification leakage rate limit [10% of the contained volume over a 24-hour period from an initial overpressure of 2.0 psig (13.8 kPa above atmosphere)], the air leakage from the containment structure in standard cubic feet per minute (scfm) as a function of containment pressure can be expressed by the following equation: LR = 17.85 x (CP-14.7)"/

2;where: 81 of 86 LR = leakage rate from containment (scfm); and CP -= containment pressure (psia).The minimum Technical Specification free volume of the containment building is 225,000-ft 3 at standard temperature and pressure.

At an initial overpressure of 2.0 psig (13.8 kPa above atmosphere), the containment structure would hold approximately 255,612 standard cubic feet (scf)of air. A loss of 10%, from this initial overpres sure condition, would result in a decrease in air volume to 230,051 scf. The above equation describes the leakage rate that results in this drop of contained air volume over 1,440 minutes (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).When applying the Technical Specification leakage rate equation to the assumed initial overpressure condition of 0.33 psig (2.28 kPa above atmosphere), it would take approximately 16.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> for the leak rate to decrease to zero from an initial leakage rate of approximately 10.3 scfmi, which would occur at the start of the event. The average leakage rate over the 16.5-hour period would be about 5.2 scfm.Several factors exist that will mitigate the radiological impact of any air leakage from the containment building following target failure. First of all, most leakage pathways from containment discharge into the reactor laboratory building, which surrounds the containment structure.

Since the laboratory building ventilation system continues to operate during target failure, leakage air captured by the ventilation exhaust system is mixed with other building air, and then discharged from the facility through the exhaust stack at a rate of approximately 30,500 cfm.Mixing of containment air leakage with the laboratory building ventilation flow, followed by discharge out the exhaust stack and subsequent atmospheric dispersion, results in extremely low radionuclide concentrations and very small radiation doses in the unrestricted area. A tabulation of these concentrations and doses is given below. These values were calculated following the same methodology stated in Section 5.3.3 of Addendum 3 to the MIURR Hazards Summary Report [ 1].A second factor which helps to reduce the potential radiation dose in the unrestricted area relates to the behavior of radioiodine, which has been studied extensively in the containment mock'up facility at Oak Ridge National Laboratory (ORNL). From these experiments, it was shown that up to 75%of the iodine released will be deposited in the containment vessel [2]. If, due to this 75% iodine deposition in the containment building, each cubic meter of air released from containment has a radioiodine concentration that is 25% of each cubic meter within containment building air, then the radioiodine concentrations leaking from the containment structure into the laboratory building, in microcuries per milliliter, will be: Example calculation of 1311 released through the exhaust stack:= 13aI activity / (30,500 ft 3/min x 16.5 hr x 60 mmn/hr x 28,300 mi/fl 3)= 8.40 x 10+06 g.Ci / 8.55 x 10+1" ml= 9.83 x 10-06 k.Ci/ml (9.83 x 10.06 jiCi/ml) x (0.25) = 2.46 x 10-°6 jiCi/ml 82 of 86 Note: Same calculation is used for the other radioiodines listed below.Radioiodine Concentrations in Air Leaking from Containment and Exiting the Exhaust Stack 1311 -2.46 x 10-o6 gtCi/ml 133I -- 1.17 x 10.05 ptCi/mi 35-- 1.10 x 10.05 jiCi/ml 1 3 2 I -- 5.44 x 10.0 jiCi/ml 14-- 1.33 x 10-°5 gCi/ml Example calculation of 85Kr released through the exhaust stack:= 85Kr activity / (30,500 ft 3/min x 16.5 hr x 60 mir/hr x 28,300 ml/ft 3)= 1.71 x 10+03 gtCi / 8.55 x 10+11 ml-2.00 x 10.o9 pCi/mi Note: Same calculation is used for the other noble gases listed below.Noble Gas Concentrations in Air Leaking from Containment and Exiting the Exhaust Stack 8SKr -2.00 x 10.09 pCi/ml 87Kr -1.80 x 10-° pCi/ml 89Kr -3.25 x 10-0 5 pCi/ml 85mJr -8.87 x 10.o6 gCi/ml 88Kr -2.54 x 10.o pCi/ml 9°'Kr -3.21 x 10-o5 jiCi/mI 3Xe- 2.21 x 100 gCi/ml l 3 5 mXe -7.91 x 10.06 pCi/ml '3 8 Xe -4.38 x 10.o pCi/ml'3 5 Xe -1.59 x 10°5 gtCi/ml '3 7 xe -4.19 x 10-° pCi/ml 1 3 9 Xe -3.59 x 10°0 5 pCi/mi Assuming, as stated earlier, that (1) the average leakage rate from the containment building is 5.2 scfmn, (2) the leak continues for about 16.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> in order to equalize the containment building pressure with atmospheric pressure, (3) the flow rate through the facility's ventilation exhaust stack is 30,500 scfm, (4) the reduction in concentration from the point of discharge at the exhaust stack to the point of maximum concentration in the unrestricted area is a factor of 292 and (5) there is no decay of any radioiodines or noble gases, then the following concentrations of radioiodines and noble gases with their corresponding radiation doses will occur in the unrestricted area. The values listed are for the point of maximum concentration in the unrestricted area assuming a uniform, semi-spherical cloud geometry for noble gas submersion and further assuming that the most conservative (worst-case) meteorological conditions exist for the entire 16.5-hour period of containment leakage following target failure. Radiation doses are calculated for the entire 16.5-hour period. Dose values for the unrestricted area were obtained using the same methodology that was used to determine doses inside the containment building, and it was assumed that an individual was present at the point of maximum concentration for the full 16.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> that the containment building was leaking.A worst-case scenario dilution factor of 292 for effluent dilution using the Pasquill-Guifford Model for atmospheric dilution is used in this analysis.

We assume that all offsite (public) dose occurs under these atmospheric conditions at the site of interest, i.e. 760 meters North of MUJRR. In our case at 760 meters it occurs only during Stability Class F conditions; which normally only occur 11.4% of the time when the wind blows from the south. Thus this calculation is conservative.

83 of 86 10 CFR 20 Appendix B Effluent Concentration Limits are used for the "listed" isotopes.

An Effluent Concentration Limit of 2.0 x 1 0-° pCi/mi is used for the "unlisted" isotopes, which equals the DAC/300 when using the DOE Part 835 Default DAC limits of 6.0 x 10-°6 Exposure at 1 DAC gives 5000 mrem per year whereas at the effluent concentration limit it is 50 mrem per year.This is a factor of 100 times less for the effluent concentration limit as compared to the DAC.Exposure at the effluent concentration limit assumes you are in that effluent concentration for 8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br /> per year. Thus, the time assumed to be exposed to the effluent concentration limit is a factor of 4.38 longer than the 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> per year that defines a DAC. The isotopes in question are based on a default DAC limit of 6.0 x 10-o6 for short-lived

(< 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> half-lives) submersion DAC's in Appendix C of Part 835. No credit is taken for transit time from the stack to the receptor point nor is credit taken for decay inside containment until release. In the case of Kr-89 and Xe-i137 the transit time alone would be approximately one (1) half-life while the transit time for Kr-90 and Xe-139 would be at least four (4) half-lives.

Thus we believe that a factor of 300 reduction below the DAC value to establish the effluent concentration limit is warranted.

This reduction factor of 300 is consistent, in fact more conservative, than 10 CFR 20 Appendix B, as the NRC calculates Effluent Concentration Limits for Submersion isotopes by dividing the DAC value by 219.Example calculation of whole body dose in the unrestricted area due to 1311: Conversion Factor: (Public dose limit of 50 mrem/yr) x (1 yr/8760 hours) = 5.71 x 1 0.0 mremihr I31 concentration

=8.42 x 10.09 iiCilml 131I effluent concentration limit = 2.00 x 10-'° gCi/ml 1311 Conversion Factor = 5.71 x 10.0 mrenm/hr Therefore, a 16.5-hour whole body exposure from 1311 is:= 1311 concentration

/ (1311 effluent concentration limit x Conversion Factor x 16.5 hrs)= 8.42 x 10.09 / (2.00 x 10-1° gCi/ml x 5.71 x i0-0 mremihr x 16.5 brs)=3.96 x 10+°° mrem Note: Same calculation is used for the other isotopes (radioiodines and noble gases) listed below.84 of 86 Effluent Concentration Limits. Concentrations at Point of Maximum Concentration and Radiation Doses in the Unrestricted Area -Radio iodine Radionuclide 1311 1321 133I 1341 1351 Effluent Limit 2.00 x 10-10 p.Ci/ml 2.00 x 10-° xtCi/ml 1.00 x 10-09 pCi/ml 6.00 x 10-° gCi/ml 6.00 x 10-0 pCi/ml Maximum Concentration 1 8.42 x 10-09 pxCi/ml 1.86 x 10-08 pCi/ml 4.00 x 10-° pCi/ml 4.55 x 10.0 pCi/ml 3.78 x 10.08 pCi/ml Radiation Dose 3.96 x 10+00 mrem 8.78 x 10.02 mrem 3.77 x~ 10+00 mrem 7.14 x 10-02 mrem 5.93 x 10-° mrem Total = 8.48 mrem Note 1: Maximum Concentrations are radioiodine and noble gas concentrations leaking from the containment building and exiting the exhaust stack reduced by a dilution factor of 292.Effluent Concentration Limits. Concentrations at Point of Maximum Concentration and Radiation Doses in the Unrestricted Area -Noble Gases Radionuclide 8 5 Kr 85mIKr 5 7 Kr 8 5 Kr 8 9 Kr 9O0r 1 3 5 Xe l 3 smXe 1 3 9 Xe Effluent Limit 7.00 x 10"07 jiCi/ml 1.00 X 10-0 xtCi/ml 2.00 x 10-0 pCi/ml 9.00 x 10-09 2.00 x 10-08 pCi/ml 2.00 x 10-08 ptCi/ml 5.00 X 10-07 iiCi/ml 7.00 x 10°08 iiCi/ml 4.00 x 10-0 ptCi/ml 2.00 x 10-0 pCi/ml 2.00 x 10.0 iiCi/ml 2.00 x 10-08 pCi/ml Maximum Concentration' 6.85 x 10-12 pCi/ml 3.04 x 10.08 gCi/ml 6.17 x 10-08 gtCi/ml 8.70 x 10-08 pCi/ml 1.11 x 10.07 giCi/ml 1.10 x 10.o pCi/mi 5.45 x 10-08 pCi/ml 2.71 x 10-08 gtCi/ml 1.43 x 1 0"07 pCi/ml 1.50 X 10.07 ptCi/ml 1.23 x 10.07 pCi/ml Radiation Dose 9.22 x 10-07 mrem 2.86 x 10.02 mrem 2.91 x 10.01 mrem 9.10 x 10.01 mrem 5.24 x 10.01 mrem 5.17 x 1 0-1 torero 1.43 x 10.02 mrem 7.34 x 10.0 mrem 6.38 x 10-02 mrem 6.76 x 10.01 mrem 7.06 x 10.01 mrem 5.80 x 10.01 mrem Total = 4.38 mrem Note 1: Maximum Concentrations are radioiodine and noble gas concentrations leaking from the containment building and exiting the exhaust stack reduced by a dilution factor of 292.To finalize the unrestricted dose in terms of Total Effective Dose Equivalent (TEDE), the doses from the radioiodines and noble gases must be added together, and result in the following values: Dose from Radioidines and Noble Gases in the Unrestricted Area Committed Effective Dose Equivalent (Radioiodine)

Committed Effective Dose Equivalent (Noble Gases)Total Effective Dose Equivalent (Whole Body)8.48 mrem 4.38 mrem 12.87 mrem 85 of 86 Sunuming the doses from the noble gases and the radioiodines simply substantiates earlier statements regarding the very low levels in the unrestricted area should a failure of a fueled experiment occur, and should the containment building leak following such an event. Because the dose values are so low, the dose from the noble gases becomes the dominant value, but the overall TEDE is still only 12.87 mremn, a value far below the applicable 10 CFR 20 regulatory limit for the unrestricted area.

References:

1 Hlazards Summary Report, Addendum 3, Section 5.3.3, University of Missouri Research Reactor Facility, August 1972 (as revised by the 1989-1990 Operations Annual Report).2 Hlazards Summary Report, Addendum 4, Appendix C, University of Missouri Research Reactor Facility, October 1973.86 of 86

'ATTACHMENT 2 COPY MODIFICATION RECORD Modification Number Modification Title 72-7 .. ..... ......,Page 1 Storage Basket Page Required Date Number Page Title Yes No Completed By 1 2 3 4 S 6 Modification Record x System Proposal (including a detailed x hazards analysis)Crew Evaluation x Safety Evaluation (OSHA) x Safety Subcommittee Review _Reactor Advisory Committee Review ABC Review Modification Approved 8 10 11 12 13 14 Parts Requirement Installation Record Blueprints, spare parts, tech manuals Pre-op test SOP changes Conmpliance checks and PM's MH cards A ervisor Yes No 7_iL_I/616 79~Li4) Z.2 --'I T~, C /7 73 0 te Date'C) 15 73 I -p~'-~3 9 -('-1~vs DaZef ByZ JAm 9, Reactor S ervisor Modification Completed COPY ATTACHMENT 2 Analysis of an Auxiliary Spent Fuel Storage Rack in the MURR Pool The Missouri University Research Reactor (MURR) has as a part of its pool a deep pit with two storage racks capable of holding sixteen spent fuel elements or two full core loadings.

This is inadequate for the present MURR fuel cycle and with the advent of 10 megawatt operation, the situation will be even worse. It is proposed that an addi-tional eight element rack be installed between the existing two. Figure 1 is a sketch depicting a top view of proposed configuration.

The existing two racks are hung from the pool wall along with their respective gamma shields. So as not to stress these supports further, it is proposed that the auxiliary rack have an integral stand to support it 2 feet off the pool floor and level with the existing racks. The new rack will attach to the present ten element rack by brack~ets that engage underneath as shown in Figure 1. The rack is essentially self-supporting but this attachment will lend extra stability.

The fully loaded rack will weigh approximately 300 pounds and will be supported by the pooi floor.Since the existing racks have 1/4" boral on each side, it will only be necessary to place short boral dividers between the elements in the new rack to insure that each element is separated from every other by boral.t This is well within the maximum Keff limit of 0.8 presented in the MURR license R-103.Thus the fully loaded rack will be far subcritical.

• ATTACHMENT 2' Figure I Proposed Fuel Storage Rack Configuration Auxiliary Fuel Storage Rack S. 5" gamma shield (TYP)Existing Fuel Storage Racks Reactor Pool Outline Scale: t/8"= I" ATTACHMENT 2 Spent fuel elements with long operating history emit intense decay gamma radiation which produces heat in the concrete pool walls when attenuated.

To prevent the con-crete from damage due to thermal stresses and excessive temperatures, gamma shields are placed around the storage racks between the spent fuel and the concrete pool walls.The MURR design data establishes a conservative safety cri-terion of a maximum 30°F temperature rise in the concrete wall from pool water temperature.

Figure 1 indicates that to meet this criterion, the racks as constructed have 3-inch thick gamma shields along the sides and 1.5-inch shields on each end. The new rack will have similar 1.5-inch thick shields on each end and utilize the existing side shields.This modification represents no change from the present situa-tion, in that despite the presence of eight additional spent elements in the storage rack area, the strongest contribution to the gamma radiation field will be from the eight elements with the most recent operating history. For example, a spent fuel element with several days decay after its last operation represents less than 10% of the decay gammas that an adjacent element will emit with only two hours decay.Thus it may be concluded that the proposed auxiliary fuel rack may be safely used to extend the MURR spent fuel storage capabilities by one complete core loading.Caudle Julian Reactor Physicist ATTACHMENT 2 Safety Evaluation (non-nuclear)

Modification Number q'r -'Page 4-_L._-This modification must be approved by the plant safety coordinator.

If not ap-proved, state reasons for disapproval and/or areas of non-compliance with 0SHA 1910.Approved _______ Disapproved Plant Safety Coordinator Date-J ATTACHMENT 2.............

Reactor Safety Subcommittee Minutes of Meeting of February 1, 1972 Members present: Partain, Jacovitch, Kuntz, Marriott, Slivinsky Also present: Alger, Julian The meeting was called to order at 1:35 p.m. by Dr. Partain.The minutes were accepted as read. Mr. Alger reported that the test annunciator circuit approved at the December 15, 1971 meeting had not been installed since the reason for prior reactor scrams had been located in a faulty relay. It will be used to locate sites of future scrams. Also, it was reported that the stainless steel fuel tank has been installed.

Reactor utilization request 191 was discussed by Mr. Alger and Mr. Julian. The committee considered sample cooling, reactivity worth of the sample and thermal effects on the spring in the con-tainer. The following recommendations were made: 1. Calculations of the reactivity worth of the sample be made to assure it is in compliance with license limits.2. Calculations of thermal effects be made on spring in sample container.

3. A temperature monitor be used on outer wall of container in initial irradiations to verify calculations.

Further recommendations will be based on these results.4. A preliminary week-long experiment in a low flux position after which the container is opened and examined for damage.5. A~fission product monitoring system for poo1 water be considered if this type of experiment becomes routine.The request was approved with these modifications.Mr. Julian presented plans for a new fuel storage rack in the reactor poo1 to handle increased spent fuel elements expected when 10 MW operation is in effect. These were approved.Meeting adjourned at 2:45 p.m.Robert R. Kuntz Secretary ATTACHMENT 2 Parts Requirement Sheet Modification Number Page 8- F Date Purchase Order No._________

Ordered from________

_____Purchase Order No.Ordered from Date*Date Purchase Order No. _________Ordered from________

_____Purchase Order No.Ordered from Date Date Date Date ATTACHMENT 2 Installation Record Modification Number -Page 9-ft Date Description of Work Accomplished

.,Percent Completed 4.J. 4 4. I+ 1 4. 4 4 *4 I_______ I 4. I 4 4.4 I 9 9*_______ 1~I 4.I 4 L 4.4- t 4- T 4. T

, ATTACHMENT 2 Blueprints Spare Parts Tech Manuals Modification Number Page 10-P New Rev. of*~~ir1ILn IsO. rr+/-Hn.. iitle vrrniiic virrint rKev. P40. uat~e ty Part Description Part No. Purchase Order No. SP No. Date... .._ _,_ _,,_ _,, ,, ,, , Purchase Order No.Ordered from_JA]Purchase Order No.Ordered from A/8q Date Date Purchase Order NO.Ordered from yd h4 Purchase Order No.Ordered from Date Date Manual Title Ordered fromate Ordered Date Rec'd Manua No.i 1~ I.-4 XA' i 4.,,= "4. f -4 , , i,, I. 4. 4. 4 4. 4. 4. 4 1* I I __

ATTACHMENT 2 PROCEDURE FOR Z BASKET MULTIPLICATION MEASUREMENT

1. Scan new baskets with source and a detector to insure boral plate composition.

2. Install source, two detectors, and thermocouples as directed by reactor physicist.

3. Defuel reactor as per refueling procedure.

Wait for evaluation of l/M before transfer of each element.4. Affter all elements are transferred, remove all detectors.

After the pool level is returned to normal, store the source in the deep pool (tag rope).,s. ) .A Reactor Physicist Reactor Supervisor UNIVERSITY OF" MISSOURI* ATTACHMENT 2 COLUMBIA

• KANSAS CITY

• ROLLA

• ST. LOUIS INTER-DEPARTMENT CORRESPONDENCE it-5 February 27, 1973 TO Don Alger SUBJECT Z-Basket Subcriticality Measurement On February 23, 1973 an experiment the degree of subcriticality of the new twenty-four 775-gram elements was conducted to determine Z-basket configuration of Before any fuel was transferred to the new baskets, the plates were scanned to insure boral composition.

As the elements were trans-ferred, a 1/M plot was drawn. The 1/M data indicates that the new Z-basket configuration is far subcritical.

Gerald Schiapper Reactor Physicist kp ATTACHMENT 2 SOP Changes Modi ification Number V Page 12- [____.For each change cite volume, section, part, and paragraph.

Include a copy of each change..S&L,~ , ~-AL -,; I u i- ."2. I I.I t I 1~a +I I I _ ___ ___ ___ ___ ___ ___ ___ ___ ___ ___ ___ ___ __....__ II l ATTACHMENT 2 MBt Cards Modification Number Page 14-R Complete the following data for the system and each major component.

Manufacturer

__ __ __ __ __ __ _ U. of Mo. No. _ _ _ _ _ _ _ _ _ _ _ _ _Ref. Dwg. and Manual No. 15 , 6<c) 1 Specs A ~(ei -A L~~~-Date Incorporated into System LJ) Card No. f~- 5 i O Item __ _ _ _ _ _ _ _ _ _ _ _ _ _ Serial Number ______________

Manufacturer U. of Mo. No.__ ____________

Ref. Dwg. and Manual No.__________________________

Specs Date Incorporated into System _ ______ Card No. ______________

Item _________________

Serial Number _______________

Manufacturer U. of Mo. No. ______________

Ref. Dwg. and Manual No. __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _Specs Date Incorporated into System _ ______ Card No. ______________

Item ________________

Serial Number _______________

Manufacturer U. of Mo. No.__ ____________

Ref. Dwg. and Manual No. _________________

___________

Specs Date Incorporated into System Cr o Card No.

ATTACHMENT 3~JcoPY MODIFICATION RECORD Modification Number "I Modification

....... ..=.r ....Page 1 II Page1.2.3.4.5.6.7.8.Page Title Modification Record System Proposal Preop Test Procedures Reactor Safety Evaluation Crew Evaluation Safety Subcommittee Review Reactor Advisory Committee Review AEC Review Required Yes No x X x x x K x Date ,Comp)l et~ed',I-I / (eI sl.s~f Modification ApprovedL v/A 1 Date Date of Completion 76 Modification Completed Reactor Form Revised 10-31-75 w ° ATTACHMENT 3 Mod 76-3 Spent Fuel Storage The present spent fuel storage for MUJRR is capable of storing 36 elements, 6 in each of the X and Y baskets and 24 in the Z fuel storage. Operation of the reactor with the present fuel cycle plus the 120 day decay time per element before shipment causes the fuel inventory to exceed this capacity.

The criteria for determining what the capacity should be is based on projections of inventories such that at least 8 spaces will always be available to defuel the core Proposal: Install a 14 element storage basket located in the fuel storage area behind the weir. This will be accomplished by the installation of a permanent support stand located above the existing baskets, which rests on the weir floor. The stand is spaced to permit the same access to the existing fuel storage baskets. The support stand will provide the vertical guides for the side lead shields (MURR Print #1170) and will contain lead shields at the end of each row. The side lead shield will rest on top of the existing stainless steel shields used for the lower storage baskets. The fuel basket cradles in a resting pocket which has guide pins mounted on the stand for positioning.

The basket is then secured to the stand by a threaded bolt at each end.Weight Considerations:

The existing fuel baskets and stainless steel shields hang on brackets mounted on the pool liner wall of the spent fuel storage. Each bracket is design rated to carry a vertical load of 2,000 lbs. The large shield and ten element storage basket has three (3) brackets for a total capacity of 6,000 lbs. The large stainless steel shield and ten element basket loaded with elements have a total weight of 1,826 lbs. The added lead shield for the proposed storage will have a total weight of 1,440 lbs. This totals 3,266 lbs of supported weight resting on three brackets or 1,089 lbs.per bracket. This places a load per bracket of 54%of the rated vertical load design. The smaller stainless steel shield and six (6)element storage basket loaded with elements have a total weight of 1,167 lbs. The added lead shield for the proposed storage will have a total weight of 878 lbs.This totals 2,045 lbs of supported weight resting on two brackets or 1,023 lbs.per bracket. This places a load per bracket of 51% of the rated vertical load design.The weight distribution is well within tolerance of safety margin for the vertical load support.

• : ATTACHMENT 3 Materials:

All materials in contact with pool water are aluminum or stainless steel.The boral inserts of the fuel basket will be cut from a coninon sheet which has with it a letter of certification of conformance from the manufacturer that it contains 35% by weight boron carbide.The lead for the shields conforms to A.S.T.M. designation B29-55 and will be poured into the shields prior to sealing closed.Construction Considerations:

The stand, fuel basket and shields are all welded to insure adequate strength except the two 3/8 inch thick spacers (Part 1) mounted on the back of the support stand. These are attached with machine screws so the thickness may be adjusted to insure a proper fit. Extension hooks (part 5) were added at the ends of the ten element row for the placement of future shields. Mounting holes and guide pins (parts 6 and 7) were also incorporated for future use. Consideration is being given to constructing this ten element basket to facilitate transferring elements to necessary locations.

Initial Operation:

Prior to loading fuel in the basket all boral sections will be scanned with a neutron source and neutron detector.d yws, 1f 1UJR O8wib) 1 THE FOLLOWING CRITERIA OUTLINE SAFETY CONSIDERATIONS:

ATTACHMENT 3 Criteria:

Limit dose rate outside biological shield to levels not exceeding those at present.The second level of element storage has lead shields whose attenuation equals or exceeds the present solid stainless steel shields. In addition, the concrete block construction adjacent to the current [ basket area is less dense than the poured concrete which will lie adjacent to the second level of [ basket storage.Criteria:

Limit additional dose rate through water shielding to acceptable levels.With the pool at refuel level there will be approximately 15 feet of water shielding above the second level of spent fuel storage. Conservatively assuming a decay time of I03 seconds Figure 8 of the MURR design data indicates a dose rate of 0.1 mr/hr per newly stored per element should be expected.

The same figure indicated a per element dose rate of less than 0.1 mr/hr for elements stored for a period of one week. Thus the total added dose rate due to 8 elements just removed from service and 6 elements that had been stored for one week would be less than 1.4 mr/hr at the surface of the pool water with the pool at refuel level.Criteria:

Fuel elements shall be stored in a geometry such that under moderation, the maximum value for K eff shall not exceed 0.8 Criteria:

Sufficient thermal shielding or appreciable water thickness must be S provided so that the temperature rise in the concrete shall not exceed 30° F..._.This criteria was addressed in the original design of the spent fuel storage racks (Design Data, Volume I, TM-RKD-62-9).

Results indicated a requirement of 2.0 inches of lead to shield 8 adjacent elements just removed from the reactor (conservatively assumed iO3 seconds decay time).The shields manufactured contain 2 inches of lead. Thus this criteria is satisfied.

"' "ATTACHMENT 3 Criteria:

Th~e heat contributed to the pool by the added 14 elements awaiti~ng ship-ment shall not cause an appreciable pool temperature increase over periods when the pool system is secured.For this calculation it is conservatively assumed that none of the added heat load of the 14 elements is transferred out of the pool.It is also assumed that two of the 14 elements have just been retired...

v- from service. This second assumption is based on the factthat under'the current MURR fuel cycle program, the elements are depleted in pairs.After a one hour decay, the two recently retired elements will contribute a majority of the decay heat load, initially 20 KW. The 12 remaining elements are assumed to have a decay history of only 30 days. These 1:2 contribute 15 KW. To simplify calculations it will be assumed that the decay heat load of the newly removed elements is constant at the 20 KW value for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at which time it is reduced to the heat load level 8.5 KW for 2 elements with one day decay for the remainder of the weekend period. Thus, the total heat load for the 14 spent fuel elements will be 35 KW for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 23.75 KW for the remaining 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. Using the formula q = MCpAT one 8 an determine that the temperature rise over the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is 19v F, while that o.ver the remaining 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is 260, F.The total temperature increase in the pool water over the weekend period will be less than 450 F since this aT would result if no heat were transferred to the surroundings.

The degree of conservatism in this result is illustrated by the fact that at present the pool tempera-ture increase over the weekend resulting from 24 elements stored in the Z basket and 8 elements in the core is approximately 150 F.Standard Operating Procedures require that following a shutdown the pool system shall remain in operation for a minimum of five minutes.Data from pool temperature charts for various times of the year indicates a maximum temperature of 800 F after system shutdowns.

Thus, even with the extremely conservative assumptions made, the final temperature of the pool water will be 140U.F (800.+ 15OF + 45OF) which is .well below the saturation temperature of 2120 F.

ATTACHMENT 3 Criteria:

Safety in moving fuel.Prior to spent fuel shipping the second level of baskets must be moved to the weir area so that elements to be shipped may be transferred to the upper level of baskets. The fuel movement sequence shall be written so that at any time that the baskets are moved there will be no more than six elements contained in the baskets. The six elements shall be secured in the basket, (see design drawings).

Six elements contain insufficient fuel for criticality.

"' ATTACHMENT 3 RTP-l5 Revised 8-6-76 PROCEDURE FOR INSTALLATION OF SPENT FUEL STORAGE BASKET I. After completion of shop work and prior to installation in pool, scan each S element storage box with Pu-Be neutron source and detector to insure presence of boral, record data.2. Manipulate fuel as per sequence to place the 14 elements awaiting shipment in the east two rows of the present Z basket (Zll, to Z24)3. Install shields and basket, secure to stand. Install SRM detector and take a series of base line counts. R'ecord dose rate at this time.4. Remove 14 fuel elements from vault storage and place in new Z basket as per sequence.

A 1/rn plot will be maintained as each element is loaded into the basket.5. Upon completion of transfers to additional Z storage baskets, return the non-irradiated fuel elements to vault storage. Elements shall be bagged, H.P. monitoring will be required.6. Compile data generated in steps 1, 3, and 4, and give to reactor manager for inclusion in mod package.Caudle ~Julian Reactor Manager Date 6'- ?7

° ATTACHMENT 3 SAFETY SUBCOMMITTEE Minutes of Meeting of April 8, 1976 Members Present: W. Meyer, D. Harris, C. Slivinski, 0. MoKown, R. Marriot, 3. dacovitch, H. Danner, C. Julian, T. Storvick.Guests Present: C. McKibben, C. Edwards, G. SchlaPper, G. David.I. The meeting was called to order at 1445.2. The chairman reported to the subcommittee that the parent committee in its last meeting, expressed desire to see more details of the proceedings in the subcommittee minutes.3. The subcommittee reviewed the circumstances of the March 2, 1976 abnormal occurrence report regarding the failure of vent tank level controller 925 B. C. Julian summarized the situation and answered questions.

The subcommittee unanimously concurred with the action taken.4. The subcommittee reviewed the abnormal occurrence report of March 24, 1976 regarding jumpering of the rod run-in functions on regulating blade position.

C. Julian discussed the cause and corrective action. The subcommittee unanimously approved of the action taken.5. The subcommittee reviewed Reactor Utilization Request Number 243 submitted by M. Janghorbani of the Environmental Trace Substances Research Center. The sub-committee suggested editorial changes and D. McKown noted that the RUR limitations were based on actual in-practice experience at the MURR. After discussion, the subcommittee unanimously recommended approval of the RUR as modified.6. The subcommittee began discussion of proposed modification package 76-3 for the installation of additional spent fuel storage in the MURR pool. C. 1 Julian, C. Edwards, and G. Schlapper discussed the need for additional storage and the pro-.... posed design. During this discussion N. Meyer left the meeting~tur~~ng the chair over to T. Storvick.

After questions and explanation, the subcommittee unanimously reconimended approval of the design concept and recommended that the project proceed, providing that any safety related problems or design changes be reported back to the subcommittee for further review.7. The subcommittee reviewed proposed modification 76-4 for the replacement of pool low level rod run-in switch 910. C. Julian explained that this change is an up-grade of originally installed equipment.

The subcommittee unanimously recommended approval of the modification.

..ATTACHMENT 3 Page two Safety Subconhmittee Minutes April 8, 1976 8. New staff members Charles McKibben, Reactor Operations Engineer and Chester Edwards, Reactor Plant Engineer were introduced to the subcommittee.

The meeting was adjourned at 1610..,. ".."..Prepared By: Caudle Julian Secretary Approved By: Dr. Walter Meyer Chai rman CJ:Id ATTACHMENT 3 REACTOR SAFETY EVALUATION Page 4..Modification Number________________

Does this change involve changes to the Technical Specifications or an unreviewed safety hazard as described in 1.0 C£R, secti~on 50.59 A proposed change, test, orexperiment shall be deemed to involve an unreviewed safety question (i) if the probability of occurrence or the consequences of an accident or malfunction, of equipment important to safety previ~ously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created: or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.Yes Signature !l* #EVALUATION,~d~zA -~ t,~ ~ ~. dA ~a , , , .,.. .. i "?_,, Form Revised 10/31/75 jATTACHMENT 3 REACTOR SAFETY ANALYSIS Pa-ge 4-MODIFICATION NUMBER 7,,7-3 Does this change involve a change to the reactor facility as defined in the Hazards Summary and its addenda? .. .- ..Yes No. X Sig nature____, If yes, make an analysis below, if no, outline the basis for the decision.Form Revised 10/31/75 ATTACHMENT 3 Modification 76-3: Upper 2 spent fuel storage Crew evaluation of this proposal was initiated 4/6/76.No constructive suggestions were forthcoming.

Caudle Julian Reactor Manager

"ATiTACHMENT 3Eva luation Modification Number 5 Page 5-All crew members are asked to comment in some manner on this proposal.Name Remarks .-. .""-,-(I t ~r 7~yw,~-'_______ __ -,?6 /l ;;r " %os6 /0 /o, e 44 .. ...__ _ ___ .W .."Z *Ai~c .-ek, ld/e" t */ 5 ,uo\o _ )Ll-.......

..t~ r I * -I ,t r. r -.--/ __________ .; ., it" * .:_.. ,. .,,4y .. _fa,/ .," i K I T /-foe..u G i <-__- !zja..J --.#6 o w > ,. 6/___._,__,_.____,____________

LA.,..."A"U

• ecs~xI N ~ ~ w~ti~.4~-4 6~ALQ 2 /Q?~,1~<-~y~9a~

"' ' ATTACHMENT 3 (b p Brooks & Perkins, Incorporated Mateiials Handling Division.*

P.O. Box 650. Cadillac, Michigan 40001

• 616 776-9715 March 26, 1976 Ref: Certification of Conformance University of Missouri Purchasing Dept.General Services Building Columbia, Mo. 65201 Attention: .Chester Edwards 1 sheet Boral 1/4" x. 48" x 120", 35% B 4 C We hereby certify that the core section of the composite material contains 35% by weight of boron carbide.

Reference:

Invoice #78,139 Sheet #977 Brooks & Perkins, Inc.Charles S. Timmons "-.. "..Quality Assurance EVERT L HANCOCK ,Nlotary 'Public,.

Wexford County My Commnision expires October I 4 1 197B ,4....:-'

._d". .

ATTACHMENT 3 ae16 elementas nunmb~er must be visually confirmed.

Step From To Position and E~lement Nurni Nu-mber : I, love Elemlert ,Number +:Position

. position_

Tim__e Confirmed by (initial.:

-* I I----,-dr .-77f ' L3U!_ ;IUL l -~--rn ! 7 * (.,i4u u-.!tJc3 7 &' J _ 7 _<_ _. ! oa. ..II" -a I-. .....l _________:J T/_. ____1 Y7 F3 I-L-I,-'_C.I

• I:"I' -- " -" -I, ' -..-3Sd _ _ ._ _ _ _ _ _ _ _ _ 1I" .._7 7 5FE7 ' :l._t~i~___ -* --L k -, .zL 1'- ..iJI i 7? FiZ- I a I S I I--I " o I .. ---I I I I I ,. ,______ , .a 3 _____, __ , __________.______--- a_ a- a ....a a a II a*/_. & / , z ..

6ve$wAk?41'i4I~ to ~jl ~v(y X 4 5 G 3.2 1 775 7" F65 F5 7.'I 4-6 ATTACHMENT 3 Date i Note: EaCh.step involves the transfer of a single element.Dungahst 4?2 e element's number must be. Visually confirme~d.

Step From To Position and Element Numb Nwr.,ber : Move Element Number : Position ,, Position ', Time ', Confirmed by (initials*1 I .i~I I a " " ' I.I I 7 I--"I a I7 ?..' F ." ,_ _ __-I, _,....i 7' " r: , .,L 3L4 .Ii I 7"7EF67 -_'___- ", i%_L! 775 F7,/ uLTL.7' .,<)l.J23PA k, Iq. l ..-7 7~5 F I .I7 " ..16~L : 7v P 7'6 L VAL? 7_3S"1 735 LI _,a i h,,T- .', -d !/-- .* I __ .-

X C~1 3, 2)%-,-V 4 5 4-i 3-E 1 2 3 4 5 7 8: , 9 10"7 Y 75 72 725 775: 77': 7 7:' .77?" 775 7 7 775 IS16 17 18 r 1 * -*I: x )I .I.. I .0* " "i ..... .:' " : .:" ' " -." r ._.-._. "_ _ __._ _- -_ _,. "_-,_._ _., -. ._* _ _ _; :_ '_ _ _ "_ -_ _ _ ---_:-i.... ...: ':: ...... .. ' : .'. .:... "'" ... :".'- :-.r,..:'- -F- -:--: , --,-:..." -: 4 -.......T... .. ..: .-.- -..- .. ..-. -: -..... _-. ...:... .. ....._ __ : ._: -: ......... .......2 _2 .... -.... ~ ......., I... ... .-'. I .. ............'.~ .... ... ... .... ... .-"" " ' " " ".. ....... ...I.-... .....-"--"' ' :. .. .... ..... ..."- * * .. .. "

.

_ _','2.. ;. _.. -.. .. .. ... ..-- ' -" ".."-L £--. ."

...* ; '. ... " .... -F-.. : ., -' : " : ." :- .:,. .."' ::: 1 I :-_ _ _ __" _ _ _ _ _ _ _. _ _" _ __'_' _ _ _ __ _ _'-' _'"r; " -t'"." ' ': ' ":: " " "' ' " " "" :: .T --" '"' :' : --° :' : --: ': :'- ' t:; ' '... .Y:

-: -

.:-:, : :-::::_._:_-

I: " --i :-.:,--: E ATTUACHMENT 3_ _ : I 3;7 3 Ji4'ii.il-i):

.1ATTACHMENT 3 LL~~.dAC.4~~

1, "_ _.. ._ _ ..... _ ----_ _... .. _ _ __-_I .,.?.,.7; .*ii~i .. li ii ... 4 , s I, I _________

__________

___________________________________________________________________

9. 2-c 107. -. .. ..~2-3.. ... ....... 4 tc 5... ...........

.............

o ................

....-&--.4 ...............

... .... ...... .... ..... ...$ ..... .. ... ....... ................ .... ..... .. ... ..... ...-72e4>"7 .Q .JI~, % ...

ATTACHMENT 3..........

I-3* 3.... ... .i. ..............

..7/3 1$ii.. ~4 Q-J .....L... .. ...... 1. ...../4) 13 2.7 3 X!0-~Q U~4~P~Q- 1 LoLz(

V ATTACHMENT 4 MODIFICATION RECORD O RIGINAL Page 1 M.odification Nlumber. J-" .co'1 Modification

.y& 6 -/II/Page No.-1.*2.3.4.5.6.7.8.Page Title Modification Record System Proposal Preop Test Procedures Reactor Safety Evaluation Crew Evaluation Safety Subcommittee Review Reactor Advisory Committee Review AEC Review Required Yes No x x X IL_x Date Compl eted_3Q/f7: I0LLZ1L B6 Modification Aproe Date of Completion 2 # T Modification Completed J :_"/Reactor Manager Date Form Revised 10-31-75 ATTACHMENT 4 REVISION TO MODIFICATION PACKAGE 76-3 INSTALLATION OF *14 ELEMENT SPENT FUEL STORAGE BASKET As required by the Safety Committee Meeting of April 8, 1976, implementation of 10 element storage basket addition shall be reported back to the Subcommittee for review. The following report is submitted to meet this requirement.

10 Element Z Basket Installation The present fuel storage capacity at MURR is 38 elements 4 elements in the X and V baskets. The remaining 8 spaces in are required to defuel the core if the situation arises. An fuel storage capacity is necessitated by: in the Z basket and the X and V baskets increase in the 1. NRC regulation of having less than 5Kg of unirradiated fuel in the fuel vault.2. 120 plus days of decay time required per element before spent fuel shipment.3. Operating schedule and unirradiated fuel inventory has increased the number of fuel elements involved in our fuel cycle.The 10 element basket size is dictated by space available in the Z basket storage area. Construction and material of the 10 element basket is similar to the previous 14 element basket. The support stand and shielding for the new 10 element basket was incorporated in the initial construction for Modification Package 76-3. The weight of the 10 element basket, fully loaded, is approxi-mately 340 lbs. This will result in a total support weight of 3,600 lbs. resting on 3 brackets or 1,200 lbs. per bracket. Each bracket is design rated to carry a vertical load of 2,000 Ibs; therefore, load is 60% of rated vertical load design.Safety Considerations Prior a neutron service, than 0.8 r to loading fuel in the basket, all boral sections wi source and neutron detector to verify boral present.a subcritical measurement will be performed to ensure with 10 elements loaded.II be scanned with When placed in that Keff is less ATTACHMENT 42 Decay Heat Build Up Decay heat build up in pool for 3-day upper Z basket.14 element 2 elements retired 12 elements, 30 day decay period for 24 element versus 14 element 24 element 2 elements retired 22 elements, 30 day decay 2 elements 12 elements 20KW 15KW 2 elements 22 elements 20KW 27. 5KW End of 24 hrs.35KW = 19°F increase per 1 st day End of 24 hrs.47.5KW = 25.8°F increase per 1st day 2 elements 12 elements 8.5KW 1 5KW 23.5KW = 12.8°F increase per 2nd & 3rd day 2 elements 8.5KW 22 elements 27.5KW 36.0KW = 19.5°F increase per 2nd & 3rd day End of 3 days 450F increase versus 66°F increase*Calculations are conservative since they assume no heat is transferred out of the pool. The degree of conservatism is illustrated by the fact that during 5-day week operations; pool temperature increase over the weekend resulting from 24 elements stored in the lower Z basket and 8 elements in the core was approxi-mately 15 F. Maximum pool temperature after shutdow 8 from 8 ast experience was 80 F following shutdown procedures.

Thus, 161 0 F (80 F + 15 F + 66 F) would be maximum pool temperature after 3 days which is well below saturation temperature of 212 F.Surface Dose Rate Dose rate increase should be less than 2.Omr/hr at surface of pool water with pool at refuel level for the 10 element basket addition.

A survey will be con-ducted to verify this data.All other safety criteria were considered in the submittal of Modification Package 76-3 since the design model incorporated a 24 element upper Z basket addition versus the 14 element basket installed.

Submitted by Approved by Dave McGinty /Reactor st h/(farl ie McKibbeh SReactor Manager K ATTACHMENT 4 RTP-l 5B October 3, 1978 PROCEDURE FOR INSTALLATION OF UPPER Z FUEL STORAGE BASKET 1. Prior to installation for fuel element loading, a scan will be performed on each element storage box with Pu-Be-neutron source and detector to insure presence of boral. Scan data will be recorded.2. Manipulate fuel as per sequence to facilitate fuel shipment and fuel cycle.3. Ensure Shields are in place, install basket and secure to stand. Install SRM detector and take a series of base line counts. Record dose rate at this time.4. Remove 2 fuel elements from vault storage and place in new Z basket per sequence.Complete the transfer sequence to fully load the Z bas~ket storage facility.Note: A. l/M p lot will be maintained as each element is loaded into the basket. If Keff from graph reaches 0.8, the procedure will be stopped and Reactor Manager informed.5. After completion of transfer sequence and verification, that Keff is less than 0.8; (EstabliSh does rate with Z basket storage area refuel the core according to applicable sequence.6. Compile data generated in steps, 1, 3, and 4. Complete forms are the Reactor Manager for inclusion in mod package.of assembly completely loaded. )to be given to Submitted by Dave McGi nty Reactor Physici st Approved by Charl ie McKi bben ReaCtor Manager ATTACHMENT 4 0 3 5 7/0'37 I/6/ , 5 05.55I ,3 qf/.7qo/4 7g. '.&c &~A A/A1,s~ecI Wa /~ ~? Stor7ffc A't~ con.IOAICJ ~v;t,~ flie e.~c7tuin of 14c /1[ic~ A/~/ ~*-I(- ~ ,ecJs'/>~i c/i si ~ ca~Ic rflorc (sJ7e'/u'../3.13 C/z .... .....S5,/,' 4 ,A-I.-, .'7 IC m 1.0*OI 9*.87 I I .'4S i iiil ii' -ll~ nlt rT .lll rlrlrlil 1 l ri ii lli!!li!l!i!llil

]liil!l!llli!lll i i !!Ii i++Tii+V!

! il 11 ' ;+WN, NNI.NIH, l+hl.!IIi* i Jill i + II i i i II i i I I I Ill + q il i i il I lllllllllllltl!lll!lilil!lil!, .-+" +...$ .....'+"'++'++'+,, ;++ ,, , l + l Njil l f Hji~ l i..... II,. 1.1! I..I, I.., .*j* 4** lily tr II -TiijlP1 IlI.'I I I j j ii[ j*1l!IJN -I ,. ?Vf~l a.+wrtt:ltl ii IH:I N.14I I~*i !)j Ijl ?E i PU LI jjJjjj +/-I ~ fl~~tT I~j~,:I Ii *~ l*! I Ed i~ Ip~1J~yj 3/4 ~ P ii~. .iv i4 ~ j .t1~~i t  ! I j~ h 41 F I ]~ jhI L ~ *1I I I ~ I'1 P b *~. [ I I I m ,3.1]Iit:ilii!i

+' iL + , I ", I ,+ I + i l ++, , I,+LI III i. I.I.I llI'.Iil IN!lil!lltq!li IMI.ttll_ ! l li!!ll!l!llllititltlllltl!!ill!lil!

11 tllhillt!llli.i!ii+

fil l j-:i;1 ju- w

+ ~ ! .i+" ;++ +/-4 JL 14~.IU44PII

+/-$2j jj 11......il'.4..

--.-..........-............-.-.........- ,Ij4t+ .,7 I ~-'4 5 ~, 7 g r~ /0 ,.*1 5J:kI.4 ~'-ATTACHMENT 4REFUELIN'G SEQUENhCE Note': Ea&h.step involves the transfez of a single elei~ent.

During each step 1 the transfs evi, onfiTried.. ,l*Step-Ib 3 1 a I-lore Eleiaert Euniber 775 F 8o°, a S, a a i Position I I I a Position zgl Time I i I i Cos~i$.od .(imitnia I7A/f7". I --..... -_ _ , _I a 3 Il.... ., , J k. 33./ _ __ _ _________

__* u __ .~l ,_' ... --......5_ -? 7 5 F /... _ F_ ' __ ! -___ _-__....___-____ _ _ _________

....a ;* -..a-'a-t/1 77$ fK1.I " -J[ i.-3ID 61% :* : , _...

• __ _ a .,. .... .. .~: t... .... 77 // _ I v7 L EL _ _ _"LJ 7 ' ~7 /D 7"' I ...J L- ..... .... L i c I" K 7*

1 "_Il _775 F27# , P6 _____5 l l il C __ __ __ __ __-'i 77S ,Fio .F7 ________-I__

~~ i ?Z- ........ L#+/-I t* --iJ A .... .~, i '1lV U'#..--.-______

• o .5I 6--~ 4 F~I C; I/'):, S I II i I. S.... L --_____ I ..ai .S II S..... I .. .. II I I ATTACHMENT 4 , 6" ,,, ;,-sk,;-of Sevae ;~ A~*'/2O 5;U, 3 5 S, 7 SI 15*/6 q's 2-o'7 K['0 I, 13//6 zz-20 3o 2j3 30 34, 6Ac4~&,i vgc~2~~j4-G-%~~

ATTACHMENT 4 REACTOR SAFETY EVALUATION Page 4 __ __Modification Number 76 -3C 'Does this change involve changes to the Technical Specifications or an unreviewed safety hazard as described in 10 CFR, section 50.59 A proposed change, test, or experiment shall be deemed to involve an unreviewed safety question (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created: or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.Yes_____No ASignature_________

EVALUAT ION Form Revised 10/31/75 ATTACHMENT 4 REACTOR SAFETY ANALYSIS Page 4-MODIFICATION NUMBER 74- 3. Does this change involve a change to the reactor facility as defined in the Hazards Summary and its addenda?Yes No. / Signature ,zd ...,&If yes, make an analysis below, if no. outline the basis for the. £1 *Form Revised I0/31/75 ATTACHMENT 4 REVISION TO MODIFICATION PACKAGE 76-3 INSTALLATION OF 14 ELEMENT SPENT FUEL STORAGE BASKET As required by the Safety Committee Meeting of April 8, 1976, implementation of 10 element storage basket addition shall be reported back to the Subcommittee for review. The following report is submitted to meet this requirement.

10 Element Z Basket Installation The present fuel storage capacity at MURR is 38 elements 4 elements in the X and Y baskets. The remaining 8 spaces in are required to defuel the core if the situation arises. An fuel storage capacity is necessitated by: in the Z basket and the X and Y baskets increase in the 1. NRC regulation of having less than 5Kg of unirradiated fuel in the fuel vault.2. 120 plus days of decay time required per element before spent fuel shipment.3. Operating schedule and unirradiated fuel inventory has increased the number of fuel elements involved in our fuel cycle.The 10 element basket size is dictated by space available in the Z basket storage area. Construction and material of the 10 element basket is similar to the previous 14 element basket. The support stand and shielding for the new 10 element basket was incorporated in the initial construction for Modification Package 76-3. The weight of the 10 element basket, fully loaded, is approxi-mately 340 lbs. This will result in a total support weight of 3,600 lbs. resting on 3 brackets or 1,200 lbs. per bracket. Each bracket is design rated to carry a vertical load of 2,000 Ibs; therefore, load is 60% of rated vertical load design.Safety Considerations a ne serv than Prior to loading fuel in the basket, all boral sections will eutron source and neutron detector to verify boral present.vice, a subcritical measurement will be performed to ensure tl n0.8 with 10 elements loaded.be scanned with When placed in hat Keff is less ATTACHMENT4 2 Decay Heat Build Up Decay heat build up in pool for 3-day period for 24 element versus 14 element upper Z basket.14 element 2 elements retired 12 elements, 30 day decay 24 el ement 2 elements retired 22 elements, 30 day decay 2 elements 12 elements End of 24 hrs.20KW 15KW 2 elements 22 elements 20KW 27.5KW 35KW = 19°F increase per 1st day End of 24 hrs. 47.5KW = 25.8°F increase per Ist day 2 el ements 12 elements 8. 5KW 1 5KW 23.5KW = 12.8°F increase per 2nd & 3rd day 2 elements 8.5KW 22 elements 27.5KW 36.0KW = 19.5°F increase per 2nd & 3rd day End of 3 days 45° increase versus 66° increase Calculations are conservative sinc~e they assume no heat is transferred out of the pool. The degree of conservatism is illustrated by the fact that during 5-day week operations; pool temperature increase over the weekend resulting from 24 elements stored in the lower Z basket and 8 elements in the core was approxi-mately 15 F. Maximum pool temperature after shutdow 8 from 8 ast experience was 80 F following shutdown procedures.

Thus, 161 F (80 F + 15 F + 66 F) would be max~mum pool temperature after 3 days which is well below saturation temperature of 212 F.Surface Dose Rate Dose rate increase should be less than 2.0mr/hr at surface of pool water with pool at refuel level for the 10 element basket addition.

A survey will be con-ducted to verify this data.All other safety criteria were considered.

in the submittal of Modification Package 76-3 since the design model incorporated a 24 element upper Z basket addition versus the 14 element basket installed.

Submitted by Dave McGint'y /1 Reactor st Approved by c$narlie ~McKibbefi Reactor Manager 9-29-78 a.ATTACHMENT 4 Crew Evaluation M~odification Number Page 5-All crew members are asked to comment in some manner oni this proposal.Name Remarks 10/18/73 ATTACHMENT 4/....[..................

.. ..... J r~~ .".~,j t5 S"'/~., 17 j2* I , " ." / ?17 / io 5 Y...........-....-...--..-...

-.-. .-..-N --.lI _ ............_ ..... ..-.........

....---. ---.--___ -2 4_ .... .........

...................-

i,,# -......c--.1';.l1-I,* La_ .I, 3~z 3/.... .. ... -.-........

....... .z .../6.. ... ._o. .. .i. ...... .........3..3 / ... .... ... .............. .- .......

,ATTACHMENT 5 Page 1 of 15 OkI1GNAL Revised: 2/18/86 App' dj Reactor Manager MODIFICATION RECORD MODIFICATION NO. ___!-,, _Modification Temporary Additional in-pool fuel storage baskets Page N__o.Required YES NO Date Completed Page Title 1 2 3 4 5 6 7 8 9 Modification Record System Proposal Reactor Safety Analysis Reactor Safety Evaluation Crew Evaluation MURR SOP Review Complete Compliance or P.M. Revision " Parts Requirement Sheet Prints, Technical Manual, Spare Parts Change Requirement cation Approved: X X X x x x ,__ 0 17.-I1ll By (Initials) r3 I Modifi(Datff Date of Completion:

Modification Completed:

i~; (PJ Reactor Manager Item No.___REVIEW AND FOLLOW-UP ACTION Required YESS NO Date Documented Completed By (Initials)

]_____________________

1 Safety Subcommittee Review 2 Reactor Advisory Committee Review 3 U.S. NRC Review I I 4 MURR Drawings Updated ATTACHMENT 5 Page 2of 15 flD1(S NAL eie: /88 MODFIATONNO_____

J\~)" ,"App'dtA~

Reactor Manager SYSTEM PROPOSAL The inability of MURR to establish spent fuel shipping capability since the GE-700 cask was removed from service in September 1989 has created the need for temporary additional in-pool fuel storage. This modification package documents the evaluations performed to show that the use of two shipping baskets designed for use in the MHIA cask as temporary in-pool storage facilities does not present an unreviewed safety question.

Each MHIA shipping basket has twelve fuel element storage positions in a three by four matrix with a boral sheet between each row of four elements(see page 13). These baskets will be attached by brackets to the deep pool"X" and"Y" basket fuel element storage to provide stability and lateral support. These brackets are made of 0.25" aluminium angle(see page 15) and provides a position for the OS basket if additional in-pool storage is needed.The evaluation performed for each MHIA basket will include a criticality analysis(KENO), a boral plate verification, thermal analysis and 1/M determination when it is first loaded. A separate evaluation will be made of the OS basket if used as deep pool storage in conjunction with the two MHIA baskets.

ATTACHMENT 5 Page 3of 156>f tC \LtRvsd2/86 MODIFICATION NO.______ evsd: 2/18/8-Reactor Manager REACTOR SAFETY ANALYSIS Does this change involve a change to the reactor facility as defined in the Hazards Summary and its addenda?Yes ___No _X_ Signature:

', ,-_j If YES, make an analysis below and attach a suggested revision to the HSR. If NO, outline the basis for the decision.The Hazards Summary Report (HSR) describes in-pool fuel storage in three parts of the Original HSR. Section 6.4, Spent Fuel Transfer and Storage (p.6-5) and Section 7.1.8, Fuel Handling System (p.7-8). describe irradiated fuel storage in the context of radiation dose outside the biological shield. Section 13.2.11, Refueling Accident provides accident analysis for irradiated fuel transfers within the pool and states "all storage racks have been designed to be safe with regard to criticality"(p.13-14).

The storage of irradiated fuel in the MHX and MHY baskets do not involve a change to the reactor facility as defined in HSR. These storage positions meet the dose rate criteria of Section 6.4 and 7.1.8 of Original HSR.Section 6.4 (Figure 6.6) shows that for storage of eight fuel elements adjacent to the primary reactor shield (with 40 days continuous operation at 10 MW and fission product decay time of 105 seconds) the dose rate at one foot from the outside of the reactor shield would be approximately lmr/hr. This is well within the criteria of 2.5mr/hr at one foot from shield surface required by HSR.The minimum thickness of magnetite concrete between MHX or MHY baskets to the outside of the biological shield is five feet. A further margin from the dose rate criteria is provided by the location of the MH baskets greater than eleven inches from the pool wall (not adjacent);

the fact that each basket represents a dose configuration less than an eight element array adjacent to the pool wall and the fadt that MURR fuel cycle produces irradiated elements with ATTACHMENT 5 Page 4of 15 (

MO I IC T ON N . ~UA pp'd 2 ~ L Reactor Manager a lower activity than the basis fuel cycle of forty days continuous operation at 10MW.Elements stored in MI-X and MHY will have greater than 106 seconds of decay(11.6 days)A criticality analysis and 1IM determination for initial loading of these baskets verify that these storage positions are safe with regard to criticality to meet the requirements of Original HSR Section 13.2.11.Design data volume I section TM-RKD-62-9 Thermal Shielding Requirements for Spent Fuel Storage Facilities provides thermal shielding requirements for spent fuel storage facilities.

Thermal shielding or appreciable water thickness must be provided around the spent fuel storage racks to protect the biological shield concrete from damage due to thermal stresses and excessive temperatures.

Thermal shielding requirements are based on radiation heating in the concrete and resulting temperature conditions within the concrete.

The design criterion is that the temperature rise in the concrete should not exceed 300 F.With an administrative limit of no fuel elements stored in the MHX or MHY position with a decay time less than 106 seconds(11.6 days), all storage positions except 1,8,9,and 10 in MH-X and 1,8,9 and 2 in the MHY have greater than the minimum water thickness for thermal shielding of Table 4-A of Design Data Volume 1.(See page 13) The thickness requirements presented in the table are based on a configuration of a row of eight elements stored adjacent to the biological shield. "Alternate configurations will require less thermal shield thickness" (p.5 of Thermal Shielding Requirements For Spent Fuel Storage Facilities.)

Storage positions 1,8,9 and 10 in MHX and 1,8,9 and 2 in MHY have less shielding than the minimum thickness for water thermal shielding in Table 4-A and will be administratively limited to fuel elements with greater than one year of decay(3 x 10 seconds).

Elements with this decay represent fission product activity and hence gamma heating source, about one twentieth(I

/20) of the activity(and gamma heating source) of a fuel element with 106 seconds of decay [I- Huang M.S thesis, 300 days on cycle, 120 days of irradiation, 180 days out of core alternating in and out of core(See page 11 and 12)1

.'ATTACHMENT 5 Page 5 of 15 9/-5Revisd:

2/1/8 Reactor Manager REACTOR SAFETY EVALUATION Does this change involve changes to the Technical Specifications or an unreviewed safety hazard as described in 10 CFR, section 50.59?A proposed change, test, or experiment shall be deemed to involve an unreviewed safety question (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.Yes___ No X Signature EVALUATION The safety evaluation of MURR by the Division of Reactor Licensing dated July 27, 1966 identified the safety criteria for fuel storage and handling, as providing assurance of not having a critical fuel configuration, even with the unlikely mishap that might occur during fuel handling.

The safety evaluation by the Directorate of Licensing dated May 24, 1974 supporting the MURR power upgrade to 10 MW did not elaborate further on spent fuel storage. The most recent amendments to MURR reactor license R-103 dated May 8, 1991 states: "There are no specific accidents in this type of research reactor associated with the storage of spent fuel in accordance with Technical Specifications.

The maximum hypothetical accident of complete fission product release of four fuel plates is not affected by increasing the amount of stored fuel. Because the fuel will be stored in accordance with Technical specifications, accidents previously evaluated are not changed and no new or different kind of accident is created. Therefore, staff concludes that the temporary increase in the possession limit of U-235 is acceptable." Technical specification 3.8.d states that all fuel elements stored outside the reactor core will be stored in a geometry such that calculated Kef is less than 0.9 under all conditions.

This will be met by first verifying that the two boral plates are installed in each MHIA basket(see attached results), a computer criticality analysis(KENO) will be performed with the "Y" basket and MHY basket full of new elements, the "X" basket and MHX baskets full of new elements and then an analysis with all baskets full of elements(see attached).

When fuel is loaded in each basket for the first time a 1/in plot will be performed (see attached).

This will assure compliance with tech. specs and demonstrate that use of these in-pool storage baskets does not present an unreviewed safety hazard as defined in 10 CFR 50.59.

ATTACHMENT 5 Page 6 of 15 MODIFICATION NO.ORIGINAL Revised: 2/18/86 Reactor Manager CREW EVALUATION All crew members are asked to comment in some manner on this proposal.N am e"q'. A'O'7 Remarks 04'-.0 &

'ATTACHMENT 5 Page 7 Of 15 MODIFICATION NO.?/-3 OR/ 31i,,IAL Revised: 2/18/86 App'd v-Reactor Manager STANDARD PERATINGPrOCEDrECaNGE For each change cite, section, part, and paragraph.

Include a copy of each change.Section Part Paragraph I ATTACHMENT 5 Page 8 of 15 MODIFICATION NO.~iK7 COPIGIHAL Revised: 2/18/86 App'dj Reactor Manager COMPLIANCE CHECK REVISIONS/PREVENTIVE MAINTENANCE REVISIONS Attach a copy of all new or modified compliance checks to this section.Compliance Check P.M. Description Freq umenc Date Incorporated into System]P.M. Number Date Incoroorated into System Date Incorporated into System

,,, ATTACHMENT 5 Page 9 Of 15 MODIFICATION NO.C?!>?0OP G tN AL Revised: 2/18/86 Reactor Manager PARTS REQUIREMENT SHEET_ Parts Description Part No.Purchase Order No. Date Received Purchase Order No._Ordered From_Date Purchase Order No.__Ordered From_Date Purchase Order No. ___Ordered From_____Date _ _ _ _ _ _ _ _Purchase Order No. _____Ordered From _______Date

... ', ATTACHMENT 5 Page 10 of 15 MODIFICATION NO.Revised: 2/18/86 Reactor Manager 0ORG;IN AkL BLUE PRINTS -- SPARE PARTS -TECHNICAL MANUALS BLUE PRINTS: Print No. __Print Title New Print 2306 MH1A fuel holders X SPARE PARTS: SPart Description Part No. Pur Rev, of Old Print N/A'chase Order No.Rev. No.N/A S.P. No.Date Rev.Date Re'd.Purchase Order No.Ordered From ________Date Purchase Order No.Ordered From ________Date TECHNICAL MANUALS Manual Title Ordered From Date Ordered Dt e'.Mna o Date Rec'd.Manual No.

... ATTACHMENT

...5 .....I ' -! :... ,I FI "- -- I. .. .. ..... ...,.-.. .. .--t .- :, .U .!I __.,,_,.___d: ___ __..._ __ :.J:- -..: I: ... .:! , .-:. :: t...:..t-t.i:.:

i .L ..I.... _--- -- * .__. ..L _ .__ 1 i iL r__ : i '___ ! I it*.4: L R:$ ICVi i C_i ' ___ ___ __' .1 ....1 1 .. ' " , ' I ...~ l :._ _ _ _ _ 1 :: : ::': : " : .i ..I F,t-l .: -: I, ... .. i... ...I ..i .... .. ..K~ h ii i~ ..A- : :' _: .- " --.:i : :-' " 0 U')-4" -- '--,.1 .. .... .. ......i .. ...' -.. ..... ......F l ..... .. .......... ... .... .......... .... 1 ._ ......:-::ff:.::::::l::" .-:::I': i':*:l':'"-:.1:: :l::.=1:::::.---::i:::::1:

"': " "-'_ I_ 'I:, ::!777i-i-:!i __ _ 4 -.___________

.-~ .-.-. -.::~.::___________

____I -~~4~24~§t 2 ifL7~Jcw1 Aif4-4~- ~4I7 i i :Di_::i- ri-lTq-7:71:

b t-~--r----

... ...... .... :.. -I ... -= i =.- .. -~--i ... .. ---LL .. ....... ..l.. ..I.. ..!: " F..-- : :':l=:= ': :-f= :° I : : i.. : :... ...___ -.-:_____ .___.__"_._.___-_-__._.____

V -, ., : : -- 2 -. L= :7: : = ==== == = == = == = = " : .' -! '- i' I- -.._ ... .... ..7-. .** .;.-_ --: : .7". ..: .:L _ : -i.2.7-. 1. i .. .... ........ .... ..' I --:.. .. .::':.. ..-.. .:: .... .fz::. .: ...... ........; ! : " : : -: F:, -:: ,- .:Zt:-F ': 2 ? : ::--: .--,j- --- ---- ------ F -" -:"- .... .. ..iF .. ... ........ ...*_ _ .t '~ -. " ' : ...:-i__.....

... ---' -- ---..F:::: :. :::: ."- " ' I '!i:!::>:! 17-;:i:i --. q -.;:a.:nI: -.-*", i"i* '*. I '-.. ....-. .... ..... ...:. .... ..-4o-- -I--" -: .1-, ." : i':': "i',,-:i-

!-i:1:i

-F -F.----.1 .F i -7:. :-,.. *- ..-- :*..i.,-i --.: --- :I " : -* :':-'I ". ::I:::: ..:":[. .:: .: -i ' ::._!.:..

.77-:i: 5-!=__i.::,'

! !.:i- ::-g:-,"t

f.i:L_-:l : ! i i

.t.

-

-: :::l..4::-.:

"" : i i I : ! .! .. !:-T T - t::-:i: L .m T I I -, , i I .i T I -i * , .__:2--: .... ! --"=.!* l:z.:: i! : " ,,'i :* ..:--! i .... t !

ATTACH E 7 6 --w)l 3--0J-4}i I--....8__.2°1 T-I 9-R V 0 In 0 0o 7--3 1 1I O$1 7 4-ci-J SILLI iii IIL I I I i !Ii i I i i I i i I (0 1.n ~.. ~OOId 0 I0.0 V-f e(~a~) 4?~A ee~v~ 4~I42-

...ATTACHMENT 5 I, PooL'I , I TIs dr2/ 5/'ET e//1 3,s" -ro FR O' :z" if!5 ORIGINAL'p.

ORIGINAL tr~Q FRrlNT VIEIW~ ~B TVT ]3 F- VTF-W

' "' ATTACHMENT 5 September 29, 1993 Attachment to Modification Package 91-3 Storage of Irradiated Fuel in OS Basket (1/30/93)The use of the OS basket for temporary fuel storage adjacent to the "Xv basket was first implemented in 1979. The current arrangement of the OS basket with the Mlix and MHY temporary storage positions not in the pooi does not represent an unreviewed safety question.

If the OS basket is used in conjunction with the iVHIY and MIIX baskets an additional evaluation will be performed.

Elements stored in OS basket all have greater than 177 days decay. The edge of OS basket nearest the pooi wall is approximately 12 inches from the wall. Table 4-A of Design Data, Vol. I, .Thermal Shielding Reauirements for Spent Fuel Storage Facilities, indicates 14"' of water needed for fuel with 106 seconds of decay. These elements had 177 days (min) of decay [1.5 x 107 sec] so adequate thermal shielding is available.

The thermal shielding of the 1/4" thick stainless steel bottom and side of the OS basket are not considered, but would further reduce the water thickness required as thermal shield.

ATTACHMENT 6@©[PY AP-RO-l 115 Revision I MODIFICATION RECORD:: SHORT FORM FOR: 1) Addenda to existing Modification Records (e.g., modifications of same nature as ones previously reviewed and approved).

2) Significant modification~s to the facility or facility systems that are not described in the Hazards Summar~y Report.3) Modifications that require engineering decisions/implementation in a time frame that precludes normal licensed operator review prior to implemrentation.

4) Modifications to non-safety systems; for documentation and review only.NOTE: Licensed operators will review these modifications as part of the Operator Requalification Program.The Reactor Safety Subcommittee will review these modifications.

Modification Number: 9 1-3, Addendum 1 Modification Title: .Replacement of the Existing X. Y. MH-X, and MH-Y Fuel Storaqie Baskets With New X and Y Baskets By Page No.2 3 4 6 page Title Modification Record: Short Form Modification Description (Why. Short Form is appropriate)

Hazards Summary Report Evaluation Reactor Safety Evaluation OP, PM, CP, and Print Evaluation Spare Parts Requirements Required Yes Noo x __ _Date Completed X X X X'C x___-_-2-Byiias C- 50.59 Screen Completed:

,/' N.t o~i.m&66..Z

//d,- /Q./(Asst.

Reactor Manager -E£'gineering)

Reactor Safety Subcommittee Review:___________________(Asst. Reactor. Manager -Engineering)

Modification A~pp~roved:

er Date:________

Date: 2 Date:________

Modification Completed: (Reactor Manager)Attachment 8.1]

ATTACHMENT 6 AP-RO-] 115 Revision I Modification Number: 91-3. Addendum 1 MODIFICATION DESCRIPTION Provide a concise description of the system change. InclUde any proposed PRE-OPERATIONAL TESTS required for this change. (If additional pages are necessary, insert after this page.)MURR fuel, new or irradiated, may be stored in any one of five (5) fuel storage loCations in the reactor pool.These five storage locations are designated as X, Y, Z, MH-X, and MH-Y, The X and Y storage locations can each hold 6 fuel elements.

The MH-X and MH-Y locations can each hold 12 elements while the Z* storage location can store a total of 48 fuel elements.

These fuel storage locations have been designed to the following specifications: (a) A geometr~y such that the calculated Keff is less than 0.9 under all conditions of moderation and irrespective of the number of fuel elements stored or the amount of burnup per element;(b) Sufficient natural convection cooling to prevent a fuel element from exceeding its design temperature;(c) Location within the reactor pool atea sufficient depth to provide adequate radiation shielding;(d) Arrangement in the reactor pool to permit efficient handling during the insertion, removal, or interchange of fuel elements; and (e) Fabrication from materials compatible with the fuel elements.Additionally, thermal shielding requirements for the fuel storage locations are presented in the MURR Design Data, Volume I. Thermal shielding or appreciable water thickness must be provided around the spent fuel storage baskets to protect the magnetite concrete from damage due to thermal stresses and excessive temperatures.

The thermal shielding requirements are based radiation heating in the magnetite concrete and the resulting conditions within the concrete.

The design criterion employed is that the temperature rise in the concrete should not exceed 30 degrees F.This Modification Record proposes to replace the current X, Y, MH-X and MH-Y baskets with two (2) new 20 element fuel baskets that will be designated X and Y. The new X basket will replace the old X and MH-X baskets and the new Y basket will replace the old Y and MH-Y baskets. Each new basketwill have essentially the same footprint as the two baskets that they will be replacing but overall fuel storage capacity...*-will increase from 36 to 40 in the deep pool. Additionally, a support plate will be placed between the new X and Y baskets that will provide a storage location for either the OS basket or a Be reflector ring. A separate evaluation will be needed if the OS basket is used at this location for storage.Why a Short Form is appropriate.(At least one of four reasons listed on Page 1, with justification)

• The short form of the Modification Record is appropriate because this modification is an addendum to an existing, previously reviewed and approved Modification Record (9 1-3), "Temporary Additional In-Pool Fuel.Storage.Baskets." Attachment 8.1]

V 1 ATTACHMENT 6 AP-RO-1 15 Revision I Modification Number: 9 1-3, Addendum 1 MODIFICATION DESCRIPTION (con't)The MH--X and M-H-Y baskets were installed in 1991 as additional temporary fuel storage locations during a period when the facility was unable to ship. fuel .because twvo spent fuel shipping casks that were certified to transport MURR fuel were removed from service. The additionial storage locations.

were needed to ensure that no interruption to MURR's operating schedule would be experienced.

The MI-I-X and MH-Y baskets were designed and built for the MH IA shipping cask and were not intended for the everyday use that they have endured at MURR. Over the years, some of the boral and aluminum plates have swelled or warped making certain storage locations unusable.

To ensure that we maintain m~axinmum fuel storage capability during periods of shipment uncertainties, the new baskets were designed and constructed for everyday use, similar to that of the original X, Y, and Z storage baskets. Additionally, the newly designed baskets will increase storage capacity from 36 to 40 at these locations.

Each design specification listed on page 2 will be addressed individually ini the applicable sections of this Modification Record. Specification (a) will be addressed in the Hazards Summary Report and Reactor Safety Evaluation sections.

Specification (b) will be addr'essed in the Reactor Safety Evaluation section. Specification (c) will be addressed in the Hazards Sununary Report Evaluation section. Specifications (d) and (e) will be addressed in this section of the Modification Record.Analysis of the thermal shielding requirements will be discussed in the Hazards Sunmmary Report Evaluation section.The new X and Y fuel storage baskets will be installed in the same locations the current X, Y, MH-X, and MH-Y baskets. These locations meet the requirements of specification (d), which states,"Arrangement in the reactor pool to permit efficient handling during insertion, removal, or interchange of fuel elements." The closest fuel storage location from the new baskets to the reactor pool wall is about 1 4-inches (f'rom the center of X basket storage location 20 or Y basket storage location 16).This is more than sufficient space to satisfy the requirements of specification (d). Included within this Modification Record is a print that shows the new fuel baskets superimposed over the current X, Y, MH-X, and MH-Y baskets thus indicating their similar- footprints.

The new fuel baskets are designed and constructed comparable to that of the original X, Y, and Z storage baskets; baskets that have proven to be very dependable over time. Materials of construction

... ....are boral and aluminum.

The horal for the new baskets are by percent weight less than that of the .Z .,.......baskets (24 versus 35 w%) but still more than sufficient to satisfy the'Keff requirement of specification (a). The materials of construction meet the requirements of specification (e), which states, "Fabrication from materials compatible with the fuel elements." Included in this Modification Record are the design and construction prints for the new baskets (MURR Drawing No. 2640). Also attached is a summary of the QA documentation for the boral plates provided by AAR Cargo Systems, Livonia, Michigan.

The entire QA Boral Data Package will be maintained in Document Control for future reference.

Neutron radiography of eight randomly selected boral sheets that was performed at the University of California-Davis reactor indicated even dispersion of boron in the plates. These radiographs will also.,.,.

..*be i~naintained in Document Control.2a Attachment 8.1 ATTACHMENT 6 AP-RO-l 115 Revision 1 Modification Number: 91-3, Addendum 1 HAZARDS SUMVMARY REPORT EVALUATION Does this change involve a modification to the reactor facility as defined in the Hazards Summary Report?Yes: No: v/ Signature' 4/"Y7 Date: If YES, make an analysis below and p o.ide the suggested revision(s) to the HSR. If NO, outline the basis for the decision.This modification does not involve a change to the reactor facility as defined in the Hazards Summary Report and its addenda. In-pool fuel storage and transfer is described or dis'cussed in the following sections:

HSR -Section 6.4, "Spent Fuel Transfer and HSR -Section 7.1.8, 'Fuel Handling Systems";

and HSR -Section 13.2.11, "Refueling Accident." All of these sections are correct and will remain the same.Section 6.4 describes the required biological shield thicknesses for spent fuel transfer and storage. Shield requirements for fuel storage in the pool are calculated to meet the dose rate criteria of the bulk shielding listed in Section 6.1. Figure 6.6 shows that for the storage of eight fuel elements (based on 40 days.continuous operation at 10 MW and a fission product decay time of I E5 seconds) adjacent to the primary reactor shield the dose rate at one foot from the outside of the reactor shield would be approximately 1 mr/hr.This is well within the design criterion of 2.5 mr/hr at one foot from the shield surface as required by the HSR.The minimum thickness of the magnetite conCrete between either new X or Y fuel storage basket and the outside Surface of the biological shield is five (5) feet. Additional design features that are more conservative than those assumed in Section 6.4 include: (1) the closest section of the new fuel baskets is located approximately 12-inches from the reactor pool wall (tapered section) and not immediately adjacent, (2) the new baskets are in a configuration less than an eight element array, and (3) the current MURR fuel cycle results irn irradiated elements with a much lower activity than the design basis fuel cycle of forty days of continuous operation at 10 MW. Elements stored in the new X and Y baskets for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> will have greater 1E6 seconds of decay (11.6 days). This storage time requirement will be administratively controlled by procedures RP-RO-100 and OP-RP-250.

The depth in the reactor pool at which the new fuel". .-.-.:.:. baskets will be located easily meet the minimum water shielding

~depth requirements listed in Secti6n 6.4.;therefore, the requirements of specification (c) are meet.Attached are the results of calculations performed by the Assistant Reactor Manager-Physics using the Monte Carlo simulation program MCNP that was used to verify that the new fuel baskets have been"designed to be safe with regard to criticality" as specified in Section 13.2.11 of the HSR. The Keff value estimated by IMCNP for the new configuration is 0.635, with a standard deviation of 0.002 -well below Technical Specification 3.8.d limit of 0.9. Using the most conservative approach and assumptions, the baskets were modeled using twenty (20) "fresh" 775 gram U-235 fuel elements -a far greater number of elements than what we are allowed to possess under our current inventory license limits. Additionally, the value of boral used to model the baskets was 0.0624 grams of B-10 atems/cm2.

None of the boral sheets that'were used in the construction of the baskets had a value less than 0.0709 gms/cm2, and the average value of all sheets was 0.0740 gms/cm2. A I/M criticality determination will also be made upon installation of the baskets to verify the results of the MCNP modeling.

A Keff value of 0.635 easily meets the requirements of specification (a).Attachment 8.1 I ' I ATTACHMENT 6 AP-RO-1 15 Revision 1 Modification Number: 91-3. Addendum 1 HAZARDS

SUMMARY

REPORT EVALUATION (con't)Missouri University Research Reactor Design Data Volume I, Design Memoranda TM-RKD-62-9,"Thermal Shielding Requirements for Spent Fuel Storage Facilities," provides the thermal shieldfing thicknesses .for spent fuel storage. Thermal shielding or appreciablk Water thickness must be provided around the spent fuel storage baskets to protect thle magnetite concrete from damage due to th~enna]stresses and excessive temperatures.

The thenmal shielding requirements are based on radiation heating in the magnetite concrete and the resulting conditions within the concrete.

The design criterion employed is that the temperature rise in the concrete should not exceed 30 degrees F.Table 4-A, "Spent Fuel Storage Thermal Shield Requirements," of TM-RKD-62-9 indicates that fuel elements wiflh a decay time of 1E6 seconds (11.6 days) require a minimum of 14-inches of themtla water shielding.

The thickn~ess requirements presented in this table are based on a configuration of a row of eight elements stored adjacent to the biological shield. Page 5 of TM-RI-ID-62-9 also states that "Alternate configurations will require less thermal shield thickness." All fuel storage locations in th~e new X and Y baskets have a minimum of 1 4-inches of water shielding with the exceptib~n of X basket positions 15 through 20 and Y basket positions 11 and 16 through 20.These storage locations will be administratively controlled such that fuel elements can not be stored unless they have greater than 3E7 seconds (one year) of d~ecay. Fuel elements with this decay time have a fission product activity, and hence gamma heating source, of approximately 1/20 of the activity of a fuel element 1E6 seconds of decay: This number was obtained from J. Huang's Master Thesis, pages 11 and 12, which dealt with fuel elements in a 300 day cycle, 120 days of irradiation, 180 days out o~fthe core, and alternating in and out. Graphs from J. Huang's Master Thesis that depict fuel elenment decay are included in Modification Record 91-3.3a Attachment 8.1 4 4 ATTACHMENT 6 AIP-RO-1 15 Revision 1 Modification Number: 9 1-3, Addendum 1 REACTOR SAFETY EVALUATION Does this change involve a revision(s) to the Technical Specifications or a safety hazard as described in 10 CFR 50:597? .NOTE: A licensee may make changes to the facility as described in the I-SR without obtaining a license amaendment only if: (i) A change to the Technical Specifications incorporated in the license is not required, and (ii) The change does not produce any of the following results: 1 .More than a minimal increase in the frequency of occurrence of an accident previously evaluated in the HSR;2. More than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the HSR;3. More than a minimal increase in the consequences of an accident previously evaluated in the HSR;4. More than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the HSR;5. Create a possibility for an accident of a different type than prev'iously evaluated in the HSR;6. Create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in thle HSR;7. Altering or exceeding a design basis limit for a fission product barrier as described in the HSR;8. A departure fiom a method of evaluation described in the HSR used in establishing the design bases or the safety analyses.Yes: ___No: ___ S igna tur e: _________-, ___________

Date: 2 If YES, the change must be performed igaLong Form Modification Record. If NO, outline the basis forthe decision.This modification does not involve a ch'ange to the Technical Specifications or a safety hazard as described in 10 CFR 50.59. A.50.59 Screen is attached and shows that the proposed activity does not have the potential to adversely affect nuclear safety or sate facility operations.

• .: There are two Limiting Conditions for Operation (LOC) regardi~g MURR fuel: Technical Specifi~caiions.3.8.d and 3.8.e.Technical Specification 3.8.d states that "All fuel elements or fueled devices outside the reactor core shall be stored in a geometry such that the calculated Keff is less than 0.9 under all conditions of moderation." The basis for this Specification states that this limit is conservative and assures safe fuel storage. The MCNP model was used to calculate a Keff value of 0.635 for one fuel basket fully loaded with twenty (20) "fresh" 775 gram U-235 fuel elements.

This predicted value is well below the Technical Specification limit of 0.9. This value will also be validated by a I/M criticality determination.

Technical Specification 3.8.e states that "Irradiated fuel elements, shall be stored in an array which will permit sufficient natural convection cooling such that the fuel element temperature will not exceed design values." The design of the new fuel storage baskets is nearly identical to that of the original X, Y and Z baskets with regard to natural convection cooling. This satisfies the requirements of specification (b) stated in the Modification Description.

• Attachment 8.1 ATTACHMENT 6 AP-RO-1 15 Revision 1 Modification Number: 9 1-3. Addendum I REACTOR SAFETY EVALUATION (con't)Furthermore, the Safety Evaluation (SE) performed by the Test & Power Reactor Safety Branch of the Division of Reactor Licensing, documented by letter dated July 27, 1966, was in response to the request by the University of Missouri to operate the MURR at a power level of 5 MW. The SE identified the safety criteria for fuel storage and handling, as providing assurance of not having a critical fuel con'figuration, even with th~e unlikely mishap that might occur during fuel handling.

The SE performed by the Directorate of Licensing, documented by letter dated May 24, 1974, supported MURR's request to operate at the higher power level of 10 MW. This SE did not elaborate any further on spent fuel storage. Additionally, the most recent facility operating license Amendment, Amendment No. 28 dated March 15, 1995, which involved an increase in the possession limit for U-235, stated that "No specific accidents in this type of research reactor are associated with the storage of spent fuel in accordance with the Technical Specifications." 4a Attachment 8.1'I 1[ I *ATTACHMENT 6 AP-RO- l 15 Revision 1 Modification Number: 91-3. Addendum 1 OPERATING, PREVENTATIVE MAINTENANCE, AND COMPLIANCE PROCEDURE, AND PRINT EVALUATION Does this charnge require a revision(s) to any 0pcrating, Preventative Maintenance, o~r Compliance Procedure, or any Print?Yes: ___ No: ____ S ignature:

-.' Date: If YES, provide the suggested revision(s)(,.

This Modification Record does not require a revision to any Preventative Maintenance or Compliance Procedure.

Two operating procedures and one form will require revisions:

RP-RO-100, "Fuel Movement," OP-RO-250, "In-Pool Fuel Handling," and Form-08, "Fuel Movement Sheet.' Suggested revisions to these procedures and form are listed below. New prints associated with the design and construction of the new fuel storage baskets will be maintained by Drafting.Suggested revisions to RP-RO-100:

1. Revise Step 4.12 to read: "Irradiated fuel elements that have decayed for less than one year, must not be stored in the following deep pool storage positions for longer than 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s: X-15, 16, 17, 18, 19, 20, and Y-11, 16, 17, 18, 19, 20 2. Add a precaution to Section 4.0 that states: "Irradiated fuel elements that have decayed for less than two (2) weeks, must not be stored in the X and Y fuel storage baskets for longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />." 3. Add a precaution to Section 4.0 that states: "This procedure only applies to 775 gram U-235 fuel elements.Movement of 1270 gram U-235 fuel elements is not authorized by this procedure." 4. Delete first note box on page 6 -this note is covered by suggested revision number 2 above.5. Delete second caution box -no defective positions will exist.6. Delete the words "MHX & MHY" from the bottom of Attachment 9.1 (Record 8.1).7. Revise Record 8.2, "Fuel Location Map," to depict the new basket configurations.

Suggested revisions to OP-RO-250:

1. Revise Step 3.12 to read: "Irradiated fdel &er~nents that have decayed for tr-an one year, must not be stored in the following deep pool storage positions for longer than 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s: X-15, 16, 17, 18, 19, 20, and Y-11, 16, 17, 18, 19,20 2. Add a precaution to Section 3.0 that states: "Irradiated fuel elements that have decayed for~less than two (2) weeks, must not be stored in the X and Y fuel storage baskets for longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />." 3. Add a precaution to Section 3.0 that states: "This procedure only applies to 775 gram U-235 fuel elements.Movement of 1270 gram U-235 fuel elements is not authorized by this procedure." 4. Delete the words "MHX & MHY" from the bottom of Attachment 8.1.Suggested revision to FM-08: 1. Delete the words "MHX & MHY" from the bottom of the form.Additionally, a new reactor control room fuel status board has been ordered that depicts the new fuel slorage configurations.

Attachmuent 8.1 11 t ATTACHMENT 6 AP-RO-1 115 Revision I Modification Number: 91-3, Addendum 1 SPARE PARTS REQUIREMENTS EVALUATION Does this change require that any new or additional Spare Parts be maintained in inventory?

Ye: __ o ___ inaue: N /~' ae --If YES, provide a list of the spare parts.None required for this modification.

6 Attachment 8.1 ATTACHMENT 6 50.59 SCREEN AP-RR-003 Revision 1 Page 1 of 2 Activity Screening Number: 0 e/- /Title: REDSIGNED DEEP POOL FUEL STORAGE BASKETS -&rt~ui ;',on~j

'.1-3

/Description of Activity ('what is being change..dand why,):______________________

Replace the four currently installed deep: pOol i~uel stora~e baskets designated X, MHX, Y. and MHY with two redesigned baskets that serve .the same function.Safety Determination:

Does the proposed activity have the potential to adversely affect nuclear safety or safe facilit~y (i.e., __'*MURR) Operations? "YE If this question is answered yes, do not__ continue with this procedure.

Identify and report the concern to the Reactor Manager.50.59 Screening Questions:

I. Does the proposed activity involvze a change to an SSC that adversely affects a design function described in the HSR?2. Does the proposed activity involve a change to aprocedure that adversely affects how HSR described SSC design functions are performed, controlled, or t~sted?3. Does the proposed activity involVe revising or replacing an HSR described evaluation methodology.

that is used in establishing the design bases or used in the safety analyses?YES NO YES NO YES, No 4. Does the proposed activity involve a test or experiment not described in the HSR, where an SSC is utilized or controlled in a manner that is outside the reference bounds of the design for that SSC or YES is inconSistent with analyses or descriptions in the HSR?5. Does the proposed activity require a change to the MURR Technical Specifications?

If all screening questions are answered NO, then implement the actiVity per the applicable approved facility procedure~s).

Amendment or a 50.59 Evaluation is not requlired.

...I NO/NO A License If Screen Question 5 is answered YES, then request and receive a License Amendmnet prior to implementation of-the activity.If Screen Question 5 is answered NO and Question 1, 2, 3, or 4 is answered YES, then complete and attach a 50.59 Evaluation form.[Refer to Attachment

2. 3 NOTE: If the conclusion of the screening questions is that. a 50.59 Evaluation is not required, provide justification for the "No'determination.

In addtition, list the documents (HSR, Technical Specifications, and other Licensing Basis documents) reviewed where relevant information was found. Include section/Ipage numbers. Use page 2 of this fo omrtyouzra ements.SPrint Name. ."______e_-__-__Dat Preparer:

Edward L. Murphy kA bj //Reactor Manager: Les Foyto ) /-/- Attachment 1

A T A T~ ITENI t AP -RR-003 R~evision 1 50.59 SCREEN (Cont.)Activity Screening Number: 0 O Y- Page 2of 2 Title: REDSIGNED DEEP POOL FUEL STORAGE BASKETS If t~he conclusion of the five.(5) Screening QUestions is that a 50.59 EvaIluation is no__t required, provide justification to support this determination:

[fUse and attach ihddftional pages as nzecessary.

]1. Does the proposed activity involve a change to an SSC that adversely affects a des'ign function described in the HSR?ANo The ne~w deep p00i fue~l stornge, hn~lkert dn not nffecrany ftinc'tinrn deeicribedc in the T-TqP The.prnopoeAl deign chano-e allow',s only for hetter fuel storage cabhihility The, new h .klets will perfhrrn in an improved manner, the same traction as those currently installed.

No other systems or components will be by this modification

.. ..2. Does the proposed activity involve a change tO a procedure that adversely affects how HrsR described SSC design functions are performed, controlled, or tested?lgohis mortific~ation is physical ifnrntnre_

While minor ndmipistrastive-changes wvill tave to be made to.r-i.rrent ope-rntina proredi-ro-none of the chanoe.s involved will nclvrer~ely nffer-t the manner in wi-hich any HS de.scribed RRC"desi g funnctions ,re. perfo rmned. cnntrolled or- tested 3. Does the proposed activity involve revising or replacing an described evaluation methodology that is used in establishing the design bases or used in the safety analyses?ino The _modi ca'tion to the deep pool fuel prnoro~ed designed using establishedT4qTR descie~hd e.valuatinn methodology to en.sure that design hases were. met~ and fulfills rill safety analysis~e~quirements currently in force ..4. Does the proposed activity in~volve a test o.7 experiment not where an SSC is used or controlled in a manner that is outside tlie reference bounds of the design for that SSC, or-is inconsistent with analyses or descriptions presented in the HSR?No. The redesigned deep pool fuel storage baskets are functionally and operationally the same as those currently installed, and will be used and controlled only in a manner within design boundaries.

All tests required for the proposed change are covered by, and are consistent with, all analyses and descriptions presented in the tHSR.List the documents (HISR, Technjc?,l

...p e..[i~.c~atiopns, and other Licensing.-.asis..documents) reviewed where relevant information was found. [Ihclude section / page numbers. ]1 HSR Section 6.4 "Spent Fuel Transfer and Storage".

HSR Section7..1 "Fuel Handling System". HSR Section 13.2.1t1 "Refueling Accident".

Technical Specification 3.8.d "Fuel Element Storage Geobmetry", Technical Specification 3.8.e "Cooling Requirements for Fuel Element Storage".

OP-RO-250

'Fuel Handling".

RP-RO-100 "Fuel Movement" 1

ATTACHMENT 6 From: Das Kutikkad ..To: Les Foyto, Acting Reactor manager, MURR Date: June 05, 2003 Re: Results of Calculations Performed to Estimate the Keff of the New Deep-Pool" Fuel Storage Bask~et Calculations were performed to estimate the criticality of the newly designed deep pool fuel storage basket slated to replace the X, MHX, Y and MRY baskets. The model used and the results obtained are surmmarized in this memo.For the purpose of simplicity, only one of the new 20-element basket (on one side of the pool) was modeled. A drawing of the new basket is attached to this report. One such basket is expected to replace the combined X & MHX or the combined Y & MHY storage locations.

Since the two sides of the pool are fairly decoupled neutronically (especially with the amount of boron in the storage baskets), this modeling should be adequate to establish the safe storage requirement specified in the Tech :Specs.Monte Carlo simulation program MCNP was used to model the new fuel storage basket and to estimate the criticality.

Several conservative assumptions were used in the modeling such as using all fresh fuel elements (no burn up credit taken) and using a reduced thickness for boral in the outermost surfaces.

A copy of the MCNP input file is also attached for future reference.

The current Z-basket fuel storage baskets have beral~sanldw.vched between Al walls. The boral used is approximately 35 w% of B4C in boral (rest Al). For the new bask~et, we purchased boral that has less boron content. The boral used has 0.0624 grams of B-10 atoms/cm2.

For a boral sheet of 0.265" thick (approx 0.67 cm), this translates to a B4C value of roughly 24 w%. The boron used is natural and not enriched in B-10. The dimensions of the basket and the wall thickness are shown the attached drawing.The Keff value (hie MCNP for this fuel stora ge coii~figiiration (loaded with fresh 775 g U235 fuel elements) was 0.635 with a standard deviation of 0.002. This result shows th~at it is safe to store fuel in the new basket with th~e predicted Keff well below the Tech Spec limit of 0.9.

I I II U B II 19.365 II II U B 11 H B!I/

I IJ YCF('TI fhKIl A_.A QTY. PART NO. DESCRIPTION, 1 2 Alu sheet 1 /8" 3003-H 14 Alum. 24.39"x 33.25" x 1/8'2 5 largeboral 0.265_ B4C Boral stock 24.]10"x 30" " 3 20 4.5alumtube 4.51" Square 606316 Alum, tube 1/8" wall 33.25"long 4 24 smallboral 0.265" 3oB4C boral 4.4375"x 30" J3L 1-c C 52AL sheet2 1 / :!:3003H 14 Alum. 19.615" x 33.25" x 1/8" 6 48 Aluminum stock 1-3/8" Aluminum stock 7 10 Aluminum stock 1/4" x24.14" x 1-3/8" Aluminum stock 23 1 4,)ir,, ,. .." ",,.i .... >'"" " "* :K.v. ". E',II /irdl ii ii ii ii

• ATTACHMENT 6'rodelling of the new 20-element deep pool fuel storage basket 6c this first run is a case with just the new 20-element basket c modelled (as a replacement for the existing mhy and y baskets).c subsequent runs will add the old beryllium next to this storage c basket (in place where os basket was before) to see its effect.c the core is not modelled in this case, so the storage basket c -is a stand alone basket filled with fresh fuel elements.c- " " °c a single fuel el'ement is defined and the "repeated structure" feature c of mcnp is used to construct the storage positions (bins).c some conservatiSm is used during the initial runs. some of these will c he removed during later runs if the keff is found to be unacceptable c (i.e, >0.9) -some of the conservative assumptions are listed below: c c 1) all fresh fuel considered

-i.e., no burnup ctedit taken c 2) less boral thickness for the outermost layers.c c individual "bins" of the new basket are described in an auxiliary c coordinate system. the origin of this auxiliary coordinate system c is at the center of individual bins. These are then tranformed c into the main system centered at one corner of the basket. all the c bins are filled with the same 'universe" (i.e., one fresh fuel-c element plus water, aluminum and boral surrounding the fuel).c c ** histories tracked = 100,000 for this case ***c 1 1 -1.0 (-l40:-146:150:144:-148:149) 130 -151 132 -153 154 -135 imp:n=l $ water surrounding the new basket (approx 30 cm thick)S the following four cells are created since mcnp doesn't like to c complicate any one cell too much. to avoid that problem, the new c basket is artificailly divided in the x-direction to group 5 "bins" c as one unit. this will avoid the problem of having 20 bins in one c basket (thereby complicating that one cell too much).c 2 2 -2.7 140 -150 146 -141 148 -149 #20 #21 #22 #23 #24 imp:n=l $ basket that contins t "bins" along x-axis 3 2 -2.7 140 -150 141 -142 148 -149 #25 #26 #27 #28 #29 imp:n=l $ basket that contins 5 "bins" along x-axis 4 2 -2.7 140 -150 142 -143 148 -149 #30 #31 #32 #33 #34 imp tn=l ** basket that cont-ins-.5

bins" along x-axis 5 2 -2.7 "--150 143 -144 ...48 ...1349 #35 #36 #37 #38 #39 imp:n=l $ basket that contins 5 "bins" along x-axis c 6 0 -130:-132:151:153:-154:135 imp:n=0 $ outside world c 7 9 -2.64 -204:-206:205:207 u=l imp:n=l $ boral of the bins.8 2 -2.7 204 -208 206 -207 u=1 imp:n=l $ al of the bins 9 2 -2.7 209 -205 206 -207 u=1 imp:n=1 $ al of the bins 10 2 -2.7 208 -209 206 -210 u=1 imp:n=l $ al of the bins 11 2 -2.7 208 -209 211 -207 u=l imp:n=l $ al of the bins c c although i~nfinite in dimension, banal-..thickness will be limited c by the'.diiuensions;.of the cell isrille'd with this "universe i".c 12 0 208 -209 210 -211 u=1 imp:n=l fill=2 (-11.00 0 0)c 20 0 200 -201 202 -203 148 -149 imp:n=l trcl=20 fill=1 i above is the definition of a single "bin" that is repeated 20 times c i ATTACHMENT 6 21 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 39 c like like like like like like like like like like like like like like like like like like like 20 20 20 20 20 20 20 20 20 20 20 20 20 20 20 20 20 20 20 but but but but but but but but but but but but but but but but but but but trcl=21 trcl=22 trcl=2 3 trcl=24 trcl=25 trcl=2 6 trcl=27 trcl=2 8 trcl=30 trcl=3 1 trcl=32 trcl=33 trcl=34 trcl=35 trcl=36 trcl=37 trcl=38 trcl=39 c description-of the fuel plates of a single element starts from here c this is the "universe-2" that fills the new storage basket bins.c 101 102 103 104 105 106 1.07 110 283 284 285 286 287 288 289 290 291 c 292 293 294 295 c c C 296 297 c-2.7 2.7 2.7 2.7 2.7 3.88 2.7 1.00 2.7 2.7 2.7 1.00 2.7 2.7 2.7 2.7 2.7 3.88 2*7. 1:.0 1. 0 -98+0.0803 -98+0.0803 -98 3 4 4 3 3 4 5 6 7 8-93 94 95 96 96 95 95 96 97 3 3 3 3 102 102 102 101 104 1,02 102 104 102 102 102 104 102 102 102 101 104 102 102 103 106 105 108-101 -124 125-101 -124 126-101 -127 125-103 -124 125-102 -124 125-101 -126 127-101 -124 125-103 -124 125-101 -124 125-101 -124 126-101 -124 125-103 -124 125-101 -124 125-101 -124 126-101 -127 125-103 -124 125~-102 -124 125-101 -126 127-l1.:a24...12.5 imp :n=!imp :n1l imp :n=1 imp :n=1 imp:l imp:n=1 imp :nfl1 imnp:n=l imp:n=l imp: n~l amp :nfl1 im~p~n=l imp:n=1 imp:n=1 imp: n=l imp:n=l imp:n=1 imp:n=1 imp:n=l imp:n=1 imp:n=l imp:n=1 imp: n 1 u=2 u= 2 u=2 u=2 U=2 U=2 U=2 u= 2 U= 2 U=2 U=2 U=2 U=2 U=2 U=2 U=2 U=2 U=2 U=2 u= 2 U= 2 U=2$ p1$ p1$ p 1$ p1$ p 1$ p1$ p1$ p 1$ p1$ p 1$ p1$ p1$ pl$ p1$ p1$ p1$ p1$ p1$ p1 1 1 1 1 1 1 1 2 2 2 23 24 24 24 24 24 24 24 24 clad clad clad clad on top clad on hot fuel clad wg clad clad clad Swg Sclad Sclad Sclad Sclad on top Sclad on bot Sfuellclad 4-k-105-104-107-106-124-124.-124-124 125 125 125 125$$$$fuel top water fuel hot water fuel top hanger fuel hot hanger side plates of the element are described next 2 2-2 .7-2.7-98 3 108 -107 -122 124-98 3 108 -107 -125 123 imp:n=l u~=2 imp:n=1 u=2$ side plate noAl$ side plate no.2 c the water surrounding the f-uel described next, this will c become a. fini~e amount of watet 'this "universe 2" (single c fresh fuel element plus the water surrounding it) is filled in cells c representing the "bins" of the new storage basket.c 298 1 -1.0 107:98:-123:-3:122:-108 u=2 imp:n=l $water surrounding fuel c * *

• end of cell definitions

• *

• need the following blank line.

ATTACHMENT 6 C c 3 4 5 95 96 97 98 C c 101.102 103 104 105 106 107 108 c C 120 121 c 122 ,Q3'£24 125 126 127 c c 130 1 31 132 133 134 135 c 140 141 142 1.43 144 145 146 147 148 149 c 150 151 152 153.54 c the following surfaces define the single fuel element (origin of this coordinate system different from the main one and from the auxiliary coordinate system # 1 that defines the bins of *the new basket).cz cz cz cz cz" cz cz 99 i00 pz pz pz pz pz pz pz pz 7.036 $ plate 1 7.074 , 7.125 14.630 $ plate 14.668 14.719 14.757 cz 14.986.pz 0 30.48-30.48 32.385-32.385 38.10-38.10 41,275-41.275 24 p -0.4142 1 0 p 0.4142 1 0 P p p p p p px px PY py pz.pz px py py PY py px py py pz p 7 px px PY PY pz-0.4142 0.4142-0.4142 0.4142-0.4142 0.4142-30.0 68.10-30.0 80.80-80.0 80.0 0.0 13 .00 26.00 39.00 52 .00 38.10 0.0 50. 80-42.25 42 .25 64. 50 94. 50 55.50 82.00-50.00 1" 1 1 1 1 0 0 0 0 0 0 0.0 $0.0 .$-0 .0550 0.0550-0.4674 0.4674-0 .6598 0.6598$$$$$$part part part part part part of of of of of of+2'2 .5-22.5+22 .5-22.5+22.5-22 :5 22.5 degree plane-22.5 degree plane degree degree degree degree degree degree side side side side side side plate plate plate plate plate plate ATTACHMVLENT 6 c the following surfaces define the auxiliary coordinate system that defines the individual bins of the new basket, its origin is at the c" center of each bin, a coordinate transformation then connects c this to the main coordinate system -which is centred at one of the c corners of the new basket.c 2O00 201 202 203 204 205 206 207 208 209 210 211 c px px py py px px py PY px px PY PY-6.38 6.38 z.6 38 6.38-6.07 6 .07-6.07 6 .07-5.75 5.75-5 .75 5.75 Rode c c c kcode c ml rmt1 m2 m3 c, c c c C c c m6 mt6 rot m9 c c tr2O tr2 1 tr2 2 tr2 3 tr2 4 t r2 5 tr26 tr2 7 tr2 8 tr29 tr30 tr31 tr3 2 r3 3 cr34 n imp:n 1 18r 2000 0.8 5 1001.5Cc .6667 lwtr. 01 13027.50c 1 13027.50c

• -.600 92235.50c

-.372 92238.50c

-.028 mn4 4009.50c mt4 be. Ol m5 6012.50c 1001. 50c 8016.5Cc 13027.5Cc mut5 grph. Oi 1001.5Cc 0.4170 lwtr. Ol in8 -.. 26000.5Cc 5010.50c -0.035 6000.50c -0.052 0.0 1 197r 55 3000 0 8016.50c .3333 1-.85-.002-.016-.132 8016 .50c 0.208 13027.50c 0.3750 0.24.,...

,000..50lc 0.2 28000.50c 0.1 5011.56c -0.150 13027.50c

-0.763 $ boral with 24 w% b4c 6.40 19.20 32.00 44.80 57.60 6.40 19.20 32.00 44.80 57.60 6.40 19.20 32.00 44.80 57.60 6.50 0.0 6.50 0.0 6.50 0.0 6.50 0.0 6.50 0.0 19.50 0.0 19.50 0.0 1.9.50 0.0 ..." 19.50 0.0" 19.50 0.0 32.50 0.0 32.50 0.0 32.50 0.0 32.5C 0.0 32.50 0.0 tr35 tr38 tr39 C phys : n print ctine cut :fl1 prdinp ksrc ATTACHMENT 6 6.40 45.50 0.0 19.20 45.50 0.0 32.00 45.50 0.0 44.80 45.50 0.0 57.60 45.50 0.0 S15.0 0.0 $cross Sections above 15.0 mev will be expunged 40 50 60 90 5000.j o .o -0.5 j 20 10.0 4.0 0.0 8.0 16.5 0.0 3.0 30.0 0.0 3.0 3.50 0.0 110 120 126-0.1 18.0 21.0 17 .0 4.00 2.0 20.9 33 .0 2 .00 0.0 0.0 0.'0 0.0 34.0 30.0 32.0-3.50 4.5 18.0 29.0-3 .50 0.0 0.. 0 0.0 0.0 55.0 47 .0 59.0 1.50 8.0 22.0 34.0-1i.00 0.0 0.0 0.0 0.0 S I ATTACHMENT 6 OAAR CAR GO SYSTEMS a division of AAR Manufacturing, Inc..Jeff Moore, Sr. Manager -Nuclear Products 12633 lnkster Road, Livonia, MI 48150-2272 USA Phone: (734) 522.2000 Direct: (734) 466-8110 FAX: (734) 622-2240"email:

jmoore@aarcorp.com I. CUSTOMER: A. NAME: B, REQUEST DATE: II. DATES A. CURRENT DATE: B. QUOTATION VALID FOR: III. CONTACTS A. AAR CONTACT I. NAME: 2. TITLE: 3; PHONE: 4. FACSIMILE:

B. CUSTOMER CONTACT 1. NAMvE: 2 PHONE: 3. FACSIMILE:

IV. sPECIFICATION AND PRICING University of Missouri April 8, 2003 April 30, 2003 90 days Jeff Moore Sr. Manager, Nuclear Products (734) 522-2000 x 8110 (734) 522-2240.Mr. Jeff Attebery 1-573-882-5269 1-573-882-6360

+ Shipping to Univ. of Missouri 4i%V. DELIVERIES TO COMMENCE: 60 days ARO VI. TERMS A.B.DELIVERY POINT: PAYMENT: FOB University of Missouri Net 30 da~is VII. SPECIAL INSTRUCTIONS None ATTACHMENT 6 AAR? CARGO SYSTEMS0 A division of AAR Manufacturing Group, INC.CERTIFICATE OF COMPLIANCE CUSTOMER:

University of Missouri QTY. SHIPPED: 68 pcs.DATE OF SHIPMENT:

July 18, 2003 CUSTOMER P.O. NUMBER: COO000009743 AAR CARGO SYSTEMS SALES ORDER NUMBER: 5053667 This is to certify' that the material supplied hereunder has been finspected and tested in accordance with AAR-1 1002 QAP, Revision 23 dated November 7, 2002, and A.AR-10012 QAP, Revision 18 dated April 9, 2003, and Nuclear Quality Program Manual, Revision 29 and meets the requirements of the. purchase order. The Code of Federal Regulations 10OCFR5O Appendix B and 10OCFR2 1 are applicable to the material on this order.SIGNATURE:

TITLE: DATE: Phill Pusilo Lab Manager July 18. 2003 Appendix C AAR-1 0012 QAP Page 1 of 1...systems, components

& more 12633 Inkster Road Livonia Michigan 48150-2272 USA Telephone 1-734-522-2000 Faxc 1-734-522-2240 ATTACHMENT 6 AA R CA RGO0 S YSTEMS A div'ision of AAR Manufacturing Group, INC.BORAL DATA PACKAGE RECORD CHECKLIST SPECIFICATION:

AAR-10012 OAP. REV. 17 D OCUMENT CHECKED BY DATE" Record Checklist Certificate of Compliance Inspection Data Sheets Material Certifications JP/KE JP/KiE JP/KE JP/KE JP/KE JPIKE JP/KE 7-18-03 7-1 8-03 7-12-03 7-1 8-03 7-18-03 7-18-03 7-18-03-BoraI Summary Report, Boxing List Calibrated Equipment Data Sheet REVIEWED BY: TITLE: DATE: Phill Pusilo Lab Manaaer July 18, 2003 APPENDIX D AAR-10012 QAIP PAGE 1 OF I..systems, components

& more 12633 Inkster Road Livonia Michigan 48150-2272 USA Telephone 1-734-522-2000 Fax 1-734-522-2240 t ATTACHMENT 6 Boral Summary Report (Pass)Job Name: University 5O 5053667 Serial Number WM010013-3A

'WM010014-IB wM010015-3A.

WM010016-IA WM010017-2A Y(MO1001 8-2A YM0 10019-1B YM010020-8B YM010O21 -SB YMI f00022-SB YM010023 ,8A YM01]0024

-8 B Lot Number M-21 S M-2 IS M-2 I8 M-218 14-218 M4-220 M-220 M-220 M-220 M-220 M-220 M-220 10B gmns/em2 0.0740 0.0721 0,0709 0.0766 0.0726 0.0754, 0.0758 0.0731 0.0748 0.0750 0.0733 0.0742 Density, 2.5731 2.5507 2.5474 2.5484 2.5427 2.6582 2.5632 2.5781 2.5728 2.5623 2.5847 2,5873 Reviewed By: Phill Pusilo Title: Lab Manager Date: 8/11/2003 Appendix-A AAR10012QAP Pagc: 1 Ptss ATTACHMENT 6 MATERIAL TRACEABILETY bY B ORAL SERIAL NUMBER S.0. # 50536.67 University of Missouri--..~ k-.- WM010013 through WM010017 M-218 YM010018 through YM010024 M-220 WM010013 through YM010024 "AL03-03 WM010013 through YM0 10024 "3-045-C ATTACHMENT 7 Volume of the Primary Coolant System In-Pool Portion Mechanical Equipment Room 114 Portion Scin Area Length Volume Scin Area Length Volume Scin (ft 2) (ft) (ft 3) Scin (ft 2) (ft) (ft 3)135(5) 0.7773 3.828 2.976 133(2-3) 0.7773 10.194 7.924 135(6) 0.7773 3.708 2.882 135(1) 0.7773 2.000 1.555 135(7) 0.7773 3.708 2.882 0.7773 22.374 17.391 137 0.7773 3.250 2.526 133(7-5) 0.7773 22.374 17.391 139 0.7773 2.500 1.943 133(4-3) 0.7773 14.290 11.108 501 0.6048 5.937 3.591 133(2-1) 0.7773 15.584 12.113 575 0.6048 2.269 1.372 132 0.6948 6.000 4.169 100(2) 0.7773 4.917 3.822 131(3-2) 0.6948 4.969 3.452 100(3) 0.7773 4.917 3.822 115(3-2) 0.6948 14.968 10.400 101 0.7773 1.000 0.777 115(1) 0.4948 2.000 0.990 102(1) 0.7773 1.000 0.777 111(7) 0.6948 4.189 2.911 102(2) 0.7773 3.806 2.958 111(6) 0.6948 6.667 4.632 102(3) 0.7773 3.806 2.958 111(2-5) 0.6945 16.264 11.295 102(4) 0.7773 3.806 2.958 111(1) 0.6948 2.167 1.506 102(5) 0.7773 5.097 3.962 105(9) 0.7773 2.167 1.684 401(1) 0.2006 3.975 0.797 105(7-8) 0.7773 17.312 13.457 401(2) 0.2006 3.975 0.797 105(5-6) 0.7773 15.542 12.081 405(1) 3.2150 0.500 1.608 105(1-4) 0.7773 31.832 24.743 405(2) 0.1389 4.708 0.654 102(7) 0.7773 2.000 1.555 405(3) 0.2006 9.163 1.838 102(5-6) 0.7773 10.194 7.924 460 1.3960 4.242 5.922 Total Piping Volume (ft 3) 223.860 406 0.7773 2.500 1.943 407 0.7773 2.333 1.813 Total Piping Volume (gallons)

[1,674.585 Fuel Region (gallons)7.176 Primary Circulation Pumps (gallons) 25.000 Primary Heat Exchangers (gallons) 150.000 Pressurizer (gallons) 150.000 Total Volume of PCS (gallons) 2,006.761 eCFR -- tode of Federal Regulations" ATTACHMENT 8 http://www.ecfr'.gov/cgi-birltext-idx?SlD=a6ddafde7f67322376d64cb..

ELECTRONIC CODE OF FEDERAL REGULATIONS e-CFR data is current as of September 21, 2015 Title 10 .- Chapter III --, Part 835 --. Subpart N -* Appendix-Title 10: Energy PART 835--OCCUPATIONAL RADIATION PROTECTION Subpart N-Emergency Exposure Situations APPENDIX C TO PART 835-DERIVED AIR CONCENTRATION (DAC) FOR WORKERS FROM EXTERNAL EXPOSURE DURING IMMERSION IN A CLOUD OF AIRBORNE RADIOACTIVE MATERIAL a. The data presented in appendix C are to be used for controlling occupational exposures in accordance with§835.209, identifying the need for air monitoring in accordance with §835.403 and identifying the need for posting of airborne radioactivity areas in accordance with §835.603(d).

b. The air immersion DAC values shown in this appendix are based on a stochastic dose limit of 5 reins (0.05 Sv) per year. Four columns of information are presented:

(1) Radionuclide; (2) half-life in units of seconds (s), minutes (min), hours (h), days (d), or years (yr); (3) air immersion DAC in units of pCi/mL; and (4) air immersion DAC in units of Bq/m3. The data are listed by radionuclide in order of increasing atomic mass. The air immersion DACs were calculated for a continuous, nonshielded exposure via immersion in a semi-infinite cloud of airborne radioactive material.

The DACs listed in this appendix may be modified to allow for submersion in a cloud of finite dimensions.

c. The DAC values are given for individual radionuclides.

For known mixtures of radionuclides, determine the sum of the ratio of the observed concentration of a particular radionuclide and its corresponding DAC for all radionuclides in the mixture. If this sum exceeds unity (1), then the DAC has been exceeded.

For unknown radionuclides, the most restrictive DAC (lowest value) for those isotopes not known to be absent shall be used.AIR IMMERSION DAC Radlonlucllde aflf piL)(.qm)Ar-37 __5._2_d_____00__,_____

At-39 269___yr___E-__3__

E+__ 7 Ar-41 157h3-6IEO K~r-74 1. an3-6lE0 Kr-76 __4____h____-__5____+_

5 Kr-79 __5.____h___E-__5______

5 Kr-81 __.__+05_y______-____E_07 Kr-83m 1.83 h 7E-02 2E+09 Kr-85 10.72 yr 7E-04 2E+07 Kr-85m 4.48 h 2E-05 IE+06 Kr-87 76.3 mni 4E-06 1E+05 Kr-88 2.84 h 1E-06 7E+04.Xe-120 40.0 min 1E-05 4E+05 X(e-121 40.1 mai 2E-06 BE+04 Xe-122 20.1 h 8E-05 3E+06 Xe-123 2.14 h 6E-06 2E+05 Xe-125 16.8 h lE-05 SE+05 Xe-127 36.406 d 1 E-05 8E÷05 Xe-129m 8.86 d 2E-04 7E+06 Xe-1 31m 11.84 d 5E-04 1E+07 Xe-133 5.245 d 1 E-04 5E+08 Xe-133m 2.19 d 1 E-04 5E+06 Xe-135 9.11 hi 1E-O5 6E+05 Xe-1 35m 15.36 rni 1 E-05 3E+05 Xe-138 14.13 min 3E-06 IE+05 For any single radlonuclide not listed above with decay mode other than alpha emission or spontaneous fission and I of 2 I of291231201 PM eCFR -Code of Federal RegulationsATTACHMENT 8 http://www.ecfr.gov/cgi-bin/text-idx?SID=a6ddafde7 f67322376d64cb..

with radioactive half-life less than two hours, the DAC value shall be 6 E-06 pCilmL (2 E+04 Bq/m 3).[72 FR 31940, June 8, 2007, as amended at 76 FR 20489, Apr. 13, 2011]Need assistance?

!of 2 9/23/2015 4:37 PM Case Summary of Containment Shine ATTACHMENT 9 Page 1 of 3~MicroShield 8.02 Nathan Hogue (8.00-0000)

Date I By Chke Filename IRun Date I Run Time I Duration Containl1.msd September 29, 2015 1:21:55 PM 00:00:00 Project Info Case Title Containment Shine Description IFuel Accident Analyses Geometr 13 -Rectangular Volume Source Dimensions Length 1 .8e+3 cm (60 ft 0.1 in)Width 1 .8e+3 cm (60 ft 0.1 in)Hei lht 1.8e+3 cm (60 ft0.1 in)________________

DosePoints

_________AIx V z#1 1.9e+3cm(62ft0.1 in) 914.0cm(29ft11.8 914.0cm(29flin 11.8Y#21 1.5e+4cm(492 ft1.5 1914.0cm(29 ft11.8 914.0cm(29ft 11.8 z Sin) in) in)___________Shields Shield N J ~Dimension Material Density ____________

Source 6.12e+09 cm 3 I Air 0.00122 Shield 1 j 30.5 cm I Concrete 2.35 Air Gap j Air 0.00122 Source Input: Grouping Method -Standard Indices Number of Groups: 25 Lower Energy Cutoff: 0.015 Photons < 0.015: Included_______________________

Library: Grove _______ _______Nuclide Ci Bq 3 B q/cm 3 I- 131 8.9329e+000 3.3052e+01 1 1.4600e-003 5.4020e+001 I- 132 2.4168e+001 8.9421 e+011 3.9500e-003 1.4615e+002 I- 133 5.0905e+001 1.8835e+012 8.3200e-003

-3.0784e+002 I- 134 5.2252e+001 1 .9333e+01 2 8.5400e-003 3.1598e+002 1-135 4.5644e+00 1 1 .6888e+0 12 7.4600e-003 2.7602e+002 IK"-85 2.2271 e-003 8.2403e+007 3.6400e-007 1 .3468e-002 Kr-85m 1.1625e+001 4.3013e+011 1.9000e-003 7.0300e+001 Kr-87 1.5051 e+001 5.5690e+011 2.4600e-003 9.1020e+001 Kr-88 2.4596e+001 9.1 006e+01 1 4.0200e-003 1 .4874e+002 Kr-89 5.0416e-002 1.8654e+009 8.2400e-006 3.0488e-001 Kr-90 6.0083e-01 6 2.2231 e-005 9.8200e-020 3.63 34e-0 15 file:///C:/Program%2OFiles%20(x86)/MicroShield%208/Examples/CaseFiles/HTML/Cont...

9/29/2015 Case Summary of Containment Shine ATTACHMENT 9 Page 2 of 3 Xe-133 2.4596e+001I 9.1006e+01 1 4.0200e-003 1 .4874e+002 Xe- 135 1.0157e+001 I 3.7579e+011 1.6600e-003 6.1420e+001 Xe-135m 4.3196e+000 j 1.5983e+011 j 7.0600e-004 2.6122e+001 Xe-137 2.1292e-001 j 7.8781e+009 3.4800e-005 1.2876e+000 Xe- 138 1.1013 e+001 4.0749e+011

[ 1.8000e-003 6.6600e+001 Buildup: The material reference is Shield 1 Integration Parameters X Direction 10 Y Direction I 20 Z Direction 20____________

Results -Dose Point # 1 -(1890,914,914) cm ______Fluence Rate Fluence Rate Exposure Rate Exposure Rate Energy (MeV) iActivity (Photons/sec)

MeV/cm 1/sec MeV/cm 2/sec mR/hr mR/hr_________No Buildup With Buildup No Buildup With Buildup 0.015 1.997e+11I 8.577e-253 2.641e-24 7.357e-254 2.266e-25 0.03 5.752e+11I 6.299e-35 2.648e-23 6.242e-37 2.624e-25 0.08 3.426e+ 11 1.284e-04 3.213e-03 2.031 e-07 5.084e-06 0.1 1.795e+09 8.335e-06

_3.]31e-04 1.275e-08 4.791e-07 0.15 4.738e+l11 3.531 e-02 1.778e+00 5.814e-05 2.928e-03 0.2 6.841 e+ 11 2.179e-01 1.067e+01 3.845e-04 1.883e-02 0.3 3.207e+11 6.259e-01 2.298e+01 1.187e-03 4.359e-02 0.4 1 .055e+1 2 6.904e+00 1 .867e+02 1 .345e-02 3.638e-01 0.5 2.408e+12 3.898e+01 8.084e+02 7.652e-02 1.587e+00 0.6 1.884e+ 12 6.256e+01 1.036e+03 1.221 e-01 2.021 e+00 0.8 4.714e+12 4.684e+02 5.470e+03 8.909e-01 1.040e+01 1.0 1.860e+12 4.187e+02 3.760e+03 7.717e-01 6.931e+00 1 .5 1.544e+ 12 1.406e+03

_8.149e+03 2.365e+00 1.371 e+01 2.0 1.110e+12 2.482e+03

_1.108e+04 3.838e+00 1.713e+01 3.0 8.507e+10 5.850e+02 1.898e+03 7.936e-01 2.575e+00 4.0 8.353e+07 1.1 55e+00 3 .087e+00 1 .429e-03 3.81 9e-03 Totals 1.726e+13 5.470e+03 3.242e+04 8.875e+00 5.479e+01___________

Results -Dose Point # 2 -(15000,914,914) cm _____Fluence Rate Fiuence Rate Exposure Rate Exposure Rate Energy (MeY) Activity (Photons/sec)

MeV/cmz/sec MeV/cm 2/sec mR/hr mR/hr_______No Buildup With Buildup No Buildup With Buildup 0.015 1.997e+ 11 1.798e-263 1.169e-26 1.543e-264 1.003e-27 0.03 5.752e+11 6.868e-38

_1.172e-25 6.807e-40 1.162e-27 0.08 3.426e+ 11 4.039e-07

_1.139e-05 6.392e- 10 1.802e-08 0.1 1.795e+09 2.705e-08

_1.190e-06 4.139e-ll 1.821e-09 0.15 4.738e+11I 1.274e-04

_7.800e-03 2.098e-07 1.284e-05 0.2 6.841le+lI1 8.648e-04 5.172e-02 1.526e-06 9.128e-05 0.3 3.207e+11 2.857e-03 1.255e-01 5.419e-06 2.380e-04 0.4 1.055e+12 3.451e-02 1.091e+00 6.725e-05 2.125e-03 0.5 2.408e+12 2.079e-01 4.939e+00 4.082e-04 9.695e-03 file:///C 9/29/2015

ShinePge3o3 Page 3 of 3 0.6 1 .884e+1 2 3.501e-01 6.536e+00 6.834e-04 1 .276e-02 0.8 4.714e+12 2.798e+00 3.584e+01 5.322e-03 6.817e-02 1.0 1.860e+ 12 2.610e+00 2.523e+01 4.81 le-03 4.651 e-02 1.5 1.544e+ 12 9.284e+00 5.623e+01 1.562e-02 9.461 e-02 2.0 1.110e+l12 1.684e+01 7.709e+01 2.604e-02 1.192e-01 3.0 8.507e+10 4.061e+00 1.323e+01 5.510e-03 1.795e-02 4.0 8.353e+07 8.082e-03 2.146e-02 9.998e-06 2.655e-05 Totals 1.726e+13 3.620e+01 2.204e+02 5.848e-02 3.714e-01 file

%208/Examples/CaseFiles/HTML/Cont...

9/29/2015 ATTACHMENT 9

Case Summary of Containment Shine ATTACMENT 9 Page 1 of 3 MicroShield 8.02 Nathan Hogue (8.00-0000)

Date ByChecked Filename IRun Date I Run Time I Duration J Containl .msd September 29, 2015 1:23:52 PM 00:00:00 Project Info Case Title Containment Shine Description IFuel Element Failure Accident Analyses Geometr 13 -Rectangular Volume Source Dimensions Length 1 .8e+I3 cm (60 ft 0.1 in)Width 1 I.8e+3 cm (60 ft 0.1 in)__________

Hei ht 1.8e+3 cm (60 ft 0.1 in)________________

Dose Points#11.9e+/-3 cm (62 ft0.1lin) 914.Ocm(29ft11.8 914.0cm(29ftl11.8 Y__ _ _ _ _ _ _ _ _in) in)21 5e+4 cm (492 ft 1.5 [914.0Ocm (29 ft 11.8 914.0Ocm (29 ft 11.8 2 in) j n n Shields Shield N Dimension Material Densit ____________

Source 6.12e+09 cm 3 Air 0.00122 Shield 1 I 30.5 cm I Concrete I 2.35 Air Gap Air 0.00122 Source Input: Grouping Method -Standard Indices Number of Groups: 25 Lower Energy Cutoff: 0.0 15 Photons < 0.015: Included__ _ _ _ _ _ _ _ Library: Grove .........._ __ __ __ __ __ _Nuclide Ci Bq JLCi/cm 3 Bq/cm 3 I-131 1 .3399e-00 1 4.9578e+009 2.1 900e-005 8.1 030e-001 1-132 2.6003e-001 9.6213e+009 4.2500e-005 1.5725e+000 I- 133 4.0198e-001 1.4873e+010 6.5700e-005 2.4309e+000 1-134 4.9682e-001 1 .8382e+0 10 8.1 200e-005 3 .0044e+000 I- 135 4.0932e-001 1.5 145e+010 6.6900e-005 2.4753e+000 Kr-85 6.0940e-004 2.2548e+007 9.9600e-008 3.6852e-003 Kr-85m 1.4256e-001 5.2747e+009 2.3300e-005 8.6210e-001 Kr-87 2.7288e-001 1 .0097e+010 4.4600e-005 1 .6502e+000 Kr-88 3.8852e-001 1.4375e+010 6.3500e-005 2.3495e+000 Kr-89 4.9253e-001 1 .8224e+01 0 8.0500e-005 2.9785e+000 Kr-90 4.9253e-00 1 1 .8224e+0 10 8 .0500e-005 2.9785e+000 file:///C:/Program%2OFiles%20(x86)/MicroShield%208/Examples/CaseFiles/HTML/Cont...

9/29/2015 Case Summary of Containment Shine ATTACHMENT 9 Page 2 of 3 Xe-133 5.4454e-001 2.0148e+010 8.9000e-005 3 .2930e+000 Xe-135 1 .2482e-001 4.6182e+009 2.0400e-005 7.5480e-001 Xe.-135m j 1.2176e-001 J 4.5050e+009 j 1.9900e-005 j 7.3630e-001 Xe- 137 [ 6.3632e-001 J 2.3544e+010 J 1.0400e-004 j 3.8480e+000 Xe- 138 j 6.7303 e-001 Jj 2.4902e+010 J 1.1000e-004 4.0700e+000 Buildup: The material reference is Shield 1 Integration Parameters X Direction 10 Y ireto II20 ZDirection 20____________

Results -Dose Point # 1 -(1890,914,914) cm ______Fluence Rate Fluence Rate Exposure Rate Exposure Rate Energy (MeV) Activity (Photons/sec)

MeV/cmn 2 sec MeV/cm 2/sec mR/hr mR/hr_________No Buildup With Buildup No Buildup With Buildup 0.015 5.847e+09 2.51 2e-254 7.735e-26 2.1 54e-255 6.635e-27 0.03 1.247e+ 10 1.365e-36 5.739e-25 1.353e-38 5.688e-27 0.08 7.524e+09 2.819e-06 7.055e-05 4.460e-09 1.116e-07 0.1 6.447e+09 2.994e-05 1.125e-03 4.580e-08 1.721e-06 0.15 6.836e+09 5.094e-04 2.565e-02 8.388e-07 4.224e-05 0.2 1.617e+10 5.150e-03 2.522e-01 9.089e-06 4.45 le-04 0.3 1.051e+10 2.051e-02 7.532e-01 3.891e-05 1.429e-03 0.4 2.244e+10 1 .468e-01 3.969e+00 2.860e-04 7.733e-03 0.5 3.752e+10 6.074e-01 1.259e+01 1.192e-03 2.472e-02 0.6 2.587e+10 8.590e-0I 1.422e+01 1.677e-03 2.775e-02 0.8 5.049e+ 10 5.017e+00 5.859e+01 9.542e-03 1.1 14e-01 1.0 3.060e+ 10 6.885e+00 6.184e+01 1.269e-02 1.140e-01 1.5 2.473e+10 2.251e+01 1.305e+02 3.787e-02 2.195e-01 2.0 2.762e+ 10 6.174e+01 2.755e+02 9.547e-02 4.260e-01 3.0 3.396e+09 2.335e+01 7.576e+I01 3.168e-02 1.028e-01 4.0 8.371e+08 1.158e+01 3.094e+01 1.432e-02 3.828e-02 Totals 2.893e+1 1 1.327e+02 6.649e+02 2.048e-01 1 .074e+00___________Results -Dose Point # 2 -(15000,914,914) cm-Fluence Rate Fiuence Rate Exposure Rate Exposure Rate Energy (MeV) Activity (Photons/sec)

MeV/cm 2/sec MeV/cm 2/sec mR/hr mR/hr_________No Buildup With Buildup No Buildup With Buildup 0.015 5.847e+09 5.266e-265 3.424e-28 4.517e-266 2.937e-29 0.03 1.247e+ 10 1.489e-39 2.540e-27 1.475e-41 2.518e-29 0.08 7.524e+09 8.870e-09 2.501 e-07 1 .404e- 11 3.957e- 10 0.1 6.447e+09 9.718e-08 4.276e-06 1.487e- 10 6.542e-09 0.15 6.836e+09 1.838e-06 1.125e-04 3.027e-09 1.853e-07 0.2 1.617e+10 2.044e-05 1.223e-03 3.608e-08 2.158e-06 0.3 1.051 e+l 0 9.363e-05 4.113e-03 1.776e-07 7.801 e-06 0.4 2.244e+ 10 7.337e-04 2.319e-02 1.430e-06 4.518e-05 0.5 3.752e+ 10 3.240e-03 7.695e-02 6.359e-06 1.51 le-04 file:///C:/Program%2OFiles%20(x86)/MicroShield%208/Examples/CaseFiles/HTML/Cont...

9/29/2015 Cas~1jp~xp~P~aimnent Shine Page 3 of 3 Page 3 of 3 0.6 2.587e+10 4.807e-03 8.974e-02 9.383e-06 1 .752e-04 0.8 5.049e+10 2.997e-02 3.839e-01 5.700e-05 7.301e-04 1.0 3.060e+10 4.292e-02 4.149e-01 7.912e-05 7.649e-04 1.5 2.473e+10 1.486e-01 9.003e-01 2.501e-04 1.515e-03 2.0 2.762e+ 10 4.1 89e-0 1 1.91 8e+00 6.477e-04 2.965e-03 3.0 3.396e+09 1.621e-01 5.280e-01 2.199e-04 7.163e-04 4.0 8.371e+08 8.099e-02 2.151e-01 1.002e-04 2.660e-04 Totals 2.893e+11 8.924e-01 4.555e+OO 1.371e-03 7.339e-03 9/29/2015 ATTACHMENT 9

Case Summary of Containment Shine ATTACHMENT 9 Page 1 of 3~MicroShield 8.02 Nathan Hogue (8.00-0000)

Date I By IChecked I Filename )Run Date I Run Time I Duration j I Containl .msd j September 29,2015 1:15:41 PM 00:00:01 j_________________Project Info Case Title Containment Shine Description Fuel Experiment Accident Analyses Geometry 13 -Rectangular Volume[Source Dimensions

[Length 1.8e+3 cm (60 fl0.1 in)[ Width I1.8e+3 cm (60 ft 0.1 in) __________

Hei ~ht 1.8e+3 cm (60 ft 0.1 in)________________

DosePoints

_________1.9e+3 cm (62 ft0.1 in) 914.0cm(29ft 11.8 914.0cm(29fi 11.8 Y 2 1.5e+4 cm (492it 1.5 914.0em (29 fi11.8 914.0Ocm (29 fi11.8.2 in) in) in)Shields Shield N Dimension Material Density ___________

Source 6.12e+09 cm 3 Air 0.00122 Shield 1 I 30.5 cm I Concrete I 2.35 Air Gap Air 0.00122 Source Input: Grouping Method -Standard Indices Number of Groups: 25 Lower Energy Cutoff: 0.0 15 Photons < 0.015: Included_______________

Library: Grove ________Nuclide Ci Bg JLCi/cm 3 Bg/cm 3 I- 131 8.0763e+000 2.9882e+011 1.3200e-003 4.8840e+001 1- 132 1.7866e+001 6.6104e+011 2.9200e-003 1.0804e+002 1-133 3.8240e+001 1.4149e+012 6.2500e-003 2.3125e+002 1- 134 4.3563e+001 1.611 8e+012 7.1 200e-003 2.6344e+002 I- 135 3.6160e+001 1.3379e+012 5.9100e-003 2.1867e+002 Kr-85 1 .6459e-003 6.0897e+007__

2.6900e-007 9.9530e-003 Kr-85m 7.2810e+000 2.6940e+011 1.1900e-003 4.4030e+001 Kr-87 1 .4807e+001 5.4785e+011I 2.4200e-003 8.9540e+001 Kr-88 2.0864e+001 7.7196e+011I 3.4100e-003 1 .2617e+002 Kr-89 2.6676e+001 9.8703e+011 4.3600e-003 1.6132e+002 Kr-90 2.6309e+001 9.7344e+011 4.3000e-003 1.5910e+002 file 9/29/2015 Case Summary of Containment Shine ATTACHMENT 9 Page 2 of 3 Xe-133 1.81 72e+001 6.7236e+01 1 2.9700e-003 1 .0989e+002 Xe- 135 1,I3093 e+i001 4.8446e+011 2.1400e-003 7.9180e+001 iXe-I135m 6,4856e+'000

[ 2.3997e+01 1 1 .0600e-003 j 3.9220e+001 Xe- 137 3.4386e+001 1 .2723 e+012 5.6200e-003 j 2.0794e+002 Xe- 138 j 3.5915e+001

[ 1.3289e+012 5.8700e-003 2.1719e+002 Buildup: The material reference is Shield 1 Integration Parameters X Direction 10 Y Direction I 20 Z Direction 20__________Results

-Dose Point # 1 -(1890,914,914) cm Fluence Rate Fluence Rate Exposure Rate Exposure Rate Energy (MeV) Activity (Photons/sec)

MeV/cm 2/sec MeV/cmZ/sec mR/hr mR/hr No Buildup With Buildup No Buildup With Buildup 0.015 2.904e+11I 1.248e-252 3.842e-24 1.070e-253 3.296e-25 0.03 5.010e+ll 5.485e-35 2.306e-23 5.436e-37 2.285e-25 0.08 2.546e+ 11 9.536e-05 2.387e-03 1.509e-07 3.777e-06 0.1 3.444e+ 11 1.599e-03 6.008e-02 2.447e-06 9.192e-05 0.15 3.872e+ 11 2.885e-02 1.453e+00 4.751le-05 2.393e-03 0.2 1.110e+12 3.537e-01 1.732e+01 6.242e-04 3.056e-02 0.3 5.920e+11 1 I.156e+00 4.243 e+01 2.192e-03 8.048e-02 0.4 1 .346e+12 8.808e+00 2.382e+02 1.716e-02 4.640e-01 0.5 2.718e+12 4.399e+01 9.123e+02 8.635e-02 1.791e+00 0.6 1.808e+12 6.003e+01 9.936e+02 1.172e-01 1.939e+00 0.8 4.038e+12 4.012e+02 4.686e+03 7.631e-01 8.912e+00 1.0 2.157e+12 4.855e+02 4.360e+03 8.949e-01 8.037e+00 1.5 1.735e+ 12 1.580e+03 9.156e+03 2.658e+00 1.540e+01 2.0 1.593e+12 3.562e+03 1.589e+04 5.508e+00 2.458e+t01 3.0 1.838e+11 1.264e+03 4.101le+03 1.715e+00 5.563e+00 4.0 4.532e+10 6.268e+02 1.675e+03 7.754e-01 2.072e+00 Totals 1.910e+13 8.033e+03 4.208e+04 1.254e+01 6.887e+01 Results -Dose Point # 2 -(15000,914,914) cm F'luence Rate Fluence Rate Exposure Rate Exposure Rate Energy (MeV) Activity (Photons/sec)

MeV/cm 2/sec MeV/cm 2/see mR/hr mRlhr_____________No Buildup With Buildup No Buildup With Buildup 0.015 2.904e+ 11 2.616e-263 1.701 e-26 2.244e-264 1.459e-27 0.03 5.010e-il 1 5.981e-38 1.021e-25 5.928e-40 1.012e-27 0.08 2.546e+11 3.001e-07 8.461e-06 4.749e-10 1.339e-08 0.1 3.444e+ 11 5.191 e-06 2.284e-04 7.942e-09 3.494e-07 0.15 3.872e+11 1,041e-04 6.374e-03 1.714e-07 1.050e-05 0.2 1.110e+ 12 1.404e-03 8.395e-02 2.478e-06 1.482e-04 0.3 5.920e+11I 5.274e-03 2.3 17e-01 1.000e-05 4.394e-04 0.4 1 .346e+12 4.403e-02 1.392e+00 8.579e-05 2.71 le-03 0.5 2.718e+ 12 2.347e-01 5.574e+00 4.606e-04 1.094e-02 fle :///C :/Program%2OFiles%20(x86)/MicroShield%208/Exampies/CaseFiles/HTML/Cont...

9/29/2015 Case 1;n~ainm~ent ShinePae3o3 Page 3 of 3 0.6 1.808e+12 3 .359e-01I 6.271 e+00 6.5 57e-04 1 .224e-02 0.8 4.038e+12 2.397e+00 3.070e+01 4.559e-03 5.839e-02 1.0 2.157e+12 3.026e+00 2.926e+01 5.579e-03 5.393e-02 1.5 1.735e+12 1.043e+01 6.318e+01 1.755e-02 1.063e-01 2.0 1.593e+12 2.416e+01 1.106e+02 3.737e-02 1.711e-O1 3.0 1.838e+11 8.774e+00 2.858e+01 1.190e-02 3.877e-02 4.0 4.532e+10 4.385e+00 1.164e+01 5.425e-03 1.440e-02 Totals 1.910e+13 5.380e+0O1 2.875e+02 8.360e-02 4.694e-01 fie:///C:/Programi%2OFi~es%20(x86)/MicroShield%208/Examples/CaseFiles/HTML/Cont...

9/29/2015 ATTACHMENT 9

ATTACHMENT 10 wIND ROSE PLOT: Station #03945 -COLUMBIAIREGIONAL ARPT, MO DISPLAY: Wind Speed Direction (blowing from)NORTH 15%12%9%WEST EAST WIND SPEED (mis)U >=11 I U8.8-11.1 1 5.7-.88* 3.6- 5.7 E]2.1 -3.62.1 Calms: 1.15%SOUTHDATA PERIOD: COMPANY NAME: Start Date: 11111961 -00:00 End Date: 1213111969

-21:00'MODELER: CALM WINDS: TOTAL COUNT: 1.15% 73020 hr..AVG. WIND SPEED: DATE: PROJECT NO.: 4.70 mls 912312015 WRPLOT View -Lakes Envlonmanlat Software Wind Class Frequency Distribution

z 40-35-30-%/25 20 15 10 5m A Calms 0-0.5- 2.1 2.1- 3.6 3.6- 5.7 Wind Class (mis)0.8>= 11.1 5.7- 8.8 8.8 -11.1 ATTACHMENT 11 WIND ROSE PLOT: Station #03945 -COLUMBIA/REGIONAL ARPT, MO DISPLAY: Wind Speed Direction (blowing from)NORTH 20%16%12%WEST EAST WIND SPEED (mis)U>= 11.1 1n 8.8-11.1 1 57- 8.8 3.6- 5.7 m2.1 -3.6 LI 0.5- 2.1 Calmns: 1.83%SOUTH COMwMENTS:

DATA PERIOD: COMPANY NAME: Start Date: 1/1/1170 -00:00 End Date: 1213111990

-23:00 ___________

____________

MODELER: CALM WINOS: TOTAL COUNT: 1.83% 154387 hrs.AVG. WIND SPEED: DATE; PROJECT NO.: 4.44 mole 912312015 WRPLOT View -Lakes Envirotnmental Software Wind Class Frequency Distribution A 3-z-r'J 4 4~1- 4* 41 I.41 I..6 25--20-15-10-!0.0 0-0.7>= 11.1 Calms 0.5- 2.1 2.1- 3.6 3.6- 5.7 5.7- 8.8 Wind Class (mis)8.8- 11.1 k I VIUW ! .U.U "

ATTACHMENT 12 WIND ROSE PLOT: Station #03945 -COLUMBIA/REGIONAL ARPT, MO DISPLAY: Wind Speed Direction (blowing from)NORTH 20%16%12%WEST EAST WIND SPEED (mis)U = 11.1 n8.8- 11.1 1 57- 8.8*36- 5.7 D 21 -3.6 EJ 0.5 -2.1 Calms: 1.63%:SOUTH COMMJENTS:

DATA PERIOD: COMP~ANY NAAE: Start Date: 11111961 -00:00 End Date: 12131/1990.-23:00 MODELER: CALM WINDS: TOTAL COUNT: 1.63% 227407 hrs.AVG. WINO SPEED: DATE: PROJECT NO,: 4.52 mls 9123/2015 WRPLOT View.- Lakes Environmental Software Wind Class Frequency Distribution Il ~.. I-T.lr-i 40*35.L.4.30 1-4.27.%/25-20-15-10-5-'0-0.7>= 11.1 0.5- 2.1 2.1 -3.6 3.6 -5.7 5.7 -8.8 Wind Class (m/s)8.8- 11.1view 1-rewar 7A.Uo -LX~e uwltrmnmlenltimrw/19 ATTACHMENT 13 Stack Effluent Releases -Calendar Years 2005 to 2014 2005 2006 2007 2008 2009 2010 2011 2012 2103 2014 Average Isotope (% of Technical Specification Limit)____

____ ____Ar-41 76.6876 72.8113 78.3592 77.37 70.3004 58.0857 45.14 68.00 78.1054 74.2642 69.91238 C-14 0.777 0.74 0.793 0.7867 0.613 0.58 0.477 0.723 0.0083 0.0079 0.55059 Os-191 0.0011 0.0018 0.0066 4.1739 0.0294 0.0008 0.0003 0.0001 0.0002 0.46824 1-131 0.0921 0.0435 0.0401 0.0782 0.6035 0.0415 0.0506 0.0503 0.0169 0.2201 0.12368 Ce-144 0.1165 0.0852 0.10085 Co-60 0.0853 0.0792 0.3372 0.0784 0.0084 0.0049 0.0054 0.08554 H-3 0.0732 0.052 1 0.0485 0.0527 0.0328 0.0353 0.0496 0.0426 0.0633 0.0558 0.05059 Kr-79 0.0482 0.0274 0.0378 Sc-46 0.0263 0.0022 0.01425 K-40 0.0093 0.0164 0.01 0.01 19 Cd-109 0.0112 0.01 12 1-125 0.0215 0.0041 0.0021 0.0073 0.0037 0.00774 Fe-59 0.0038 0.0038 Se-75 0.0005 0.0057 0.003 1 Sb-125 0.0026 0.0026 Zn-65 0.0005 0.001 0.0026 0.0009 0.00 125 Htg-203 0.0002 0.001 0.0002 0.0013 0.0033 0.0012 Cs-137 0.0007 0.0013 0.0006 0.0003 0.0004 0.0012 0.00075 Zr-95 0.0005 0.0005 0.0005 1-133 0.0003 0.0001 0.0001 0.0001 0.0003 0.0001 0.0001 0.0001 0.003 0.00047 Sn-i113 0.0009 0.0003 0.000 1 0.00043 Au-196 0.0005 0.0003 0.0004 0.0003 0.0004 0.00038 Gd-153 0.0003 0.0003 1 of 2 ATTACHMENT 13 Stack Effluent Releases -Calendar Years 2005 to 2014 Cu-67 0.0003 0.0003 Pa-233 0.0002 0.0003 0.00025 S-35 0.000 1 0.000 1 0.0005 0.0002 0.00023 Hf-181 0.0004 0.0001 0.000 1 0.0002 0.0002 Ce-141 0.0003 0.0002 0.000 1 0.0002 Xe-133 0.0002 0.0002 Ba-140 0.0003 0.0002 0.0001 0.0002 0.0002 Nb-95 0.0003 0.000 1 0.0002 Br-82 0.0002 0.000 1 0.00015 Co-58 0.0001 0.0001 0.0002 0.00013 As-77 0.0002 0.0001 0.0001 0.00013 Ce-139 0.000 1 0.000 1 0.000 1 Ru-103 0.0001 0.000 1 0.0001 0.0001 Mn-54 0.0001 0.0001 Be-7 0.000 1 0.000 1 Co-57 0.000 1 0.000 1 Hf-175 0.000 1 0.000 1 0.0001 Xe- 135m 0.000 1 0.0001 2 of 2