ML15275A314: Difference between revisions

From kanterella
Jump to navigation Jump to search
(Created page by program invented by StriderTol)
(Created page by program invented by StriderTol)
Line 19: Line 19:


=Text=
=Text=
{{#Wiki_filter:UNWVERSITY of MISSOURIRESEARCH REACTOR CENTEROctober 1, 2015U.S. Nuclear Regulatory CommissionAttention: Document Control DeskMail Station P 1-37Washington, DC 20555-000 1REFERENCE: Docket 50-186University of Missouri -Columbia Research ReactorAmended Facility License R- 103SUBJECT: Written communication as specified by 10 CFR 50.4(b)(1) regarding responses to the"University of Missouri at Columbia -Request for Additional Information Regardingthe Renewal of Facility Operating License No. R-l103 for the University of.Missouri atColumbia Research Reactor (TACNo. ME1580)," dated April 17, 2015On August 31, 2006, the University of Missouri-Columbia Research Reactor (MURR) submitted arequest to the U.S. Nuclear Regulatory Commission (NRC) to renew Amended Facility OperatingLicense R-103.On May 6, 2010, the NRC requested additional information and clarification regarding the renewalrequest in the form of nineteen (19) Complex Questions. By letter dated September 3, 2010, MUJRRresponded to seven (7) of those Complex Questions.On June 1, 2010, the NRC requested additional information and clarification regarding the renewalrequest in the form of one hundred and sixty-seven (167) 45-Day Response Questions. By letter datedJuly 16, 2010, MURR responded to forty-seven (47) of those 45-Day Response Questions.On July 14, 2010, via electronic mail (email), MIURR requested additional time to respond to theremaining one hundred and twenty (120) 45-Day Response Questions. By letter dated August 4, 2010,the NRC granted the request. By letter dated August 31, 2010, MURR responded to fifty-three (53) of the45-Day Response Questions.On September 1, 2010, via email, 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.eduFighting 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 theremaining 45-Day Questions. By letter dated October 13, 2010, the NRC granted the extension request.By letter dated October 29, 2010, MURR responded to sixteen (16) of the remaining 45-Day ResponseQuestions and two (2) of the remaining Complex Questions.By letter dated November 30, 2010, MURR responded to twelve (12) of the remaining 45-Day ResponseQuestions.On December 1, 2010, via email, MURR requested additional time to respond to the remaining 45-DayResponse and Complex Questions. By letter dated December 13, 2010, the NRC granted the extensionrequest.On January 14, 2011, via email, MURK requested additional time to respond to the remaining 45-DayResponse and Complex Questions. By letter dated February 1, 2011, the NRC granted the extensionrequest.By letter dated March 11, 2011, MURR responded to twenty-one (21) of the remaining 45-Day ResponseQuestions.On May 27, 2011, via email, MURR requested additional time to respond to the remaining 45-DayResponse 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 andComplex Questions.On September 30, 2011, via email, MURR requested additional time to respond to the remaining theremaining 45-Day Response and Complex Questions. By letter dated November 10, 2011, the N-RCgranted the request.By letter dated January 6, 2012, MURK responded to four (4) of the remaining 45-Day Response andComplex 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 theremaining 45-Day Response and Complex Questions. By letter dated January 26, 2012, the NRC grantedthe request.On April 12, 2012, via email, MURR requested additional time to respond to the remaining the remaining45-Day Response and Complex Questions.By letter dated June 28, 2012, MURR responded to the remaining six (6) 45-Day Response and ComplexQuestions. With that set of responses, all 45-Day Response and Complex Questions had been addressed.2 of 86 On December 20, 2012, the NRC requested a copy of the current Physical Security Plan (PSP) andOperator Requalification Program.By letter dated January 4, 2013, MURR provided the NRC a copy of the current PSP and OperatorRequalification Program.On February 11, 2013, the NRC requested updated financial information in the form of four (4) questionsbecause 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) questionsregarding significant changes to the MIURR facility since submittal of the licensing renewal application inAugust 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 formoften (10) questions.*If there are any questions regarding this response, please contact me at (573) 882-5319 orFruitsJ@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 Directorxc: Reactor Advisory CommitteeReactor Safety SubcommitteeDr. Garnett S. Stokes, ProvostDr. Henry C. Foley, Senior Vice Chancellor for ResearchMr. Alexander Adams Jr., U.S. Nuclear Regulatory CommissionMr. Geoffrey Wertz, U.S. Nuclear Regulatory CommissionMr. Johnny Eads, U.S. Nuclear Regulatory CommissionAttachments: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 System8. Meteorological Data (Wind Speed and Class) -1961 to 19699. Meteorological Data (Wind Speed and Class) -1970 to 199010. Meteorological Data (Wind Speed and Class) -1961 to 199011. 10 CFR 835, Appendix C, "Derived Air Concentration (DAC) for Workers from ExternalExposure during Immersion in a Cloud of Airborne Radioactive Material"12. Micro Shield 8.02 Dose Calculations for a Fuel Handling, Fuel Failure, and Fueled ExperimentFailure Accidents13. Stack Effluent Releases -Calendar Years 2005 to 2014JACQUELINE L.BOHM '0 "'.STATE OF MISSOURI-MY Commission 26. 2019Comisson#/ u Much ( 1,- "9 ' /4 of 86  
{{#Wiki_filter:UNWVERSITY of MISSOURIRESEARCH REACTOR CENTEROctober 1, 2015U.S. Nuclear Regulatory CommissionAttention: Document Control DeskMail Station P 1-37Washington, DC 20555-000 1REFERENCE: Docket 50-186University of Missouri -Columbia Research ReactorAmended Facility License R- 103
: 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 describesthe control blade drop process including the effect of the dashpot, but does not describe the methodfor determining the drop time nor does it explain the basis for the 80 percent insertion times. Thescram times and reactivity worths used or" assumed for the various analyses in the SAR are notclearly described or provided. NUREG-15 3 7, Section 4. 5.3, "Operating Limits, "provides guidancethat 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 theassociated control blade insertion reactivity per blade. Provide typical control blade fullinsertion 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 (seeAttachment 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 rodassembly, is removed by an electro-mechanical relay contact which allows the control blade to dropand 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 whentriggered, causes the electro-mechanical relay to change state stopping the blade droptimer/chronometer. The control blade drop time is then displayed on a meter on the reactor controlroom instrument panel. Table 1 provides the minimum, average and maximum drop times of allfour (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 0.48Average 0.50 0.54 0.50 0.52Maximum 0.59 0.58 0.54 0.54Current MURR Technical Specification 3 .2.c requires the capability of inserting the shim controlblades to their 20% withdrawn position (or 80% inserted) in less than 0.7 seconds. This ensuresprompt shutdown of the reactor in the event a reactor scram signal, manual or automatic, isreceived. The 20% withdrawn position is defined as 20% of the control blade full travel of 26inches measured from the fully inserted position. Below the 20% withdrawn position the controlblade fall is cushioned by a dashpot assembly. Approximately 91% of the control blade total worthis inserted at the 20% position. This is an original design feature of the reactor and its purpose hasnot been altered in 49 years of operation. The same Technical Specification will remain in therelicensing Technical Specifications.The measured and calculated values for reactor core excess reactivity and shutdown margin areprovided 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 isfully withdrawn). Some of this information, calculated using older computer programs, can also befound on Table 4-12 of the SAR.Typical MURR operations involve a core change-out every week with eight (8) xenon-free fuelelements in various stages of burnup (mixed core operation) used at startup. The reactor coreexcess reactivity and shutdown margin values are verified after the weekly core change-out. Theverification is done during reactor startup, when the cold, clean critical control blade height ismeasured. 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 reactivityand shutdown margin in comparison to the Technical Specification limit of -0.020 Ak/k.Table 2 -Summary of Key Measured Reactor DataValueParaeter(Ak/k)Typial ota shi cotro blae wrth0.1364Typical total shim control blade worth at 80% inserted -0.1127Typical shim control blade worth at 80% inserted with the highest worth -0.0787control blade excluded (or fully withdrawn)Maximum reactor core excess reactivity after weekly core change-out +0.0400One-year average of reactor core excess reactivity (over 69 core change-outs) +0.0290Typical core sub-criticality with 3 shim control blades at 80% inserted and the -038control blade excluded (or fully withdrawn)-037Minimum shutdown margin allowed by Technical Specifications -0.0200Calculated Values:Reactor core excess reactivity and shutdown margin values were also calculated using the detailedMGNP 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) andall fresh shim control blades (most conservative), and (2) with a mixed core loading and mixedburnup control blades (typical MUIRR operation). Table 3 provides the calculated values.6 of 86 Table 3 -Summary of Key Calculated Reactor DataValueParameter -All Fresh Fuel and Fresh Control Blades Case (kkReactor core excess reactivity 0.0865Total shim control blade worth 0.1740Core sub-criticality with 3 shim control blades at 80% inserted and highest -0.0324worth control blade excluded (or stuck fully withdrawn)ValueParameter -Mixed Core / Mixed Control Blades Case (Ak/k)Reactor core excess reactivity 0.0445Total shim control blade worth 0.1517Core sub-cniticality with 3 shim control blades at 80% inserted and highest -0.0580worth 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 worthshim control blade fully withdrawn), the minimum reactor core shutdown margin required by theTechnical 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 withdrawalfrom subcritical, control blade run in, insertion of excess reactivity, etc.). For each suchevent, provide the control blade motion speeds and reactivities utilized to provide the SARanalyses, 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 occurif there were a double-ended shear of the 12-inch primary coolant piping on both sides of either thecold-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 bethe 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 systeminstrumentation can have trip values set for them at the appropriate coolant loop locations. In theRELAP modeling, a 150 millisecond time delay is set between the time a scram signal is receivedand the modeling of when the "insertion" of the control blades start. The insertion is covered by aninput table of fission and gamma reactor power as per set time steps after the reactor scrams. Thecode calculates linear values between these data points.7 of 86 Table 4 below provides the power assumed by RELAP seconds after shutdown compared to thecalculated power after shutdown, assuming 30 days of full power operation, using equation 2.66from Nuclear Reactor Engineering 3rd Edition by Samuel Glasstone and Alexander Sesonske'. Theequation is given in the upper right corner of the page along with the values of variables a and b touse 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 powerexcept for the values for 8, 9 and 10 seconds. From 10 to 150 seconds, the RELAP values areconservative 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 useconservative calculated values of reactor decay heat after the scram, which would correspond toslower insertion of the control blades.See the response to RAI 6.a for control blade drop times related to Insertion of Excess Reactivityaccidents.References:1Glasstone, S. and Sesonske, A., Nuclear Reactor Engineering 3rd Edition, prepared underTechnical Information Center, United States Department of Energy.8 of 86 Table 4 -Comparing RELAP Decay Heat to Calculated Decay Heat(Nuclear Reactor Engineering 3rd Edition: Equation 2.66)ts00.10.30.71.02.03.05.06.07.08.09.01020304050607080901001201501802002403004004205406008001,0002,0004,0006,0008,00010,000Power MW11.0A7.99061.97 171.25971.03730.93090.848 10.73450.693 10.65760.62920.60440.58190.45060.41000.38690.36980.35630.34530.33600.32800.32100.3090A 0.30720.30530.30150.28490.26590.243 10.23950.22140.21420.19570.18240.14620.11670. 10200.09270.0860Power MW11.00.50990.45780.42010.40480.376 10.35990.34000.333 10.32730.32230.3 1800.3 141Equation 2.66P/P0=5E-3 *a *[t4-b -(To + t4)-~b]after shutdownTo = 30 days operating period prior toshutdown(s)0.1 to 1010 to 150150 to 8E80.49700.43 170.39700.37400.35690.34340.33240.323 10.3 1500.30800.296 10.2821a12.0515.3127.43b0.06390.18070.29620.32300.30500.295 10.27860.25950.23680.233 10.21500.20780.18930. 17600.13980.11030.09570.08630.0796Note A: RELAP does not have a value entered for 0.1 seconds, but the linear value between 0 and0.3 seconds is 7.9906. Value for 150 seconds is linear between 120 and 180 seconds.9 of 86  
 
==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 Regardingthe Renewal of Facility Operating License No. R-l103 for the University of.Missouri atColumbia Research Reactor (TACNo. ME1580)," dated April 17, 2015On August 31, 2006, the University of Missouri-Columbia Research Reactor (MURR) submitted arequest to the U.S. Nuclear Regulatory Commission (NRC) to renew Amended Facility OperatingLicense R-103.On May 6, 2010, the NRC requested additional information and clarification regarding the renewalrequest in the form of nineteen (19) Complex Questions. By letter dated September 3, 2010, MUJRRresponded to seven (7) of those Complex Questions.On June 1, 2010, the NRC requested additional information and clarification regarding the renewalrequest in the form of one hundred and sixty-seven (167) 45-Day Response Questions. By letter datedJuly 16, 2010, MURR responded to forty-seven (47) of those 45-Day Response Questions.On July 14, 2010, via electronic mail (email), MIURR requested additional time to respond to theremaining one hundred and twenty (120) 45-Day Response Questions. By letter dated August 4, 2010,the NRC granted the request. By letter dated August 31, 2010, MURR responded to fifty-three (53) of the45-Day Response Questions.On September 1, 2010, via email, 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.eduFighting 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 theremaining 45-Day Questions. By letter dated October 13, 2010, the NRC granted the extension request.By letter dated October 29, 2010, MURR responded to sixteen (16) of the remaining 45-Day ResponseQuestions and two (2) of the remaining Complex Questions.By letter dated November 30, 2010, MURR responded to twelve (12) of the remaining 45-Day ResponseQuestions.On December 1, 2010, via email, MURR requested additional time to respond to the remaining 45-DayResponse and Complex Questions. By letter dated December 13, 2010, the NRC granted the extensionrequest.On January 14, 2011, via email, MURK requested additional time to respond to the remaining 45-DayResponse and Complex Questions. By letter dated February 1, 2011, the NRC granted the extensionrequest.By letter dated March 11, 2011, MURR responded to twenty-one (21) of the remaining 45-Day ResponseQuestions.On May 27, 2011, via email, MURR requested additional time to respond to the remaining 45-DayResponse 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 andComplex Questions.On September 30, 2011, via email, MURR requested additional time to respond to the remaining theremaining 45-Day Response and Complex Questions. By letter dated November 10, 2011, the N-RCgranted the request.By letter dated January 6, 2012, MURK responded to four (4) of the remaining 45-Day Response andComplex 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 theremaining 45-Day Response and Complex Questions. By letter dated January 26, 2012, the NRC grantedthe request.On April 12, 2012, via email, MURR requested additional time to respond to the remaining the remaining45-Day Response and Complex Questions.By letter dated June 28, 2012, MURR responded to the remaining six (6) 45-Day Response and ComplexQuestions. With that set of responses, all 45-Day Response and Complex Questions had been addressed.2 of 86 On December 20, 2012, the NRC requested a copy of the current Physical Security Plan (PSP) andOperator Requalification Program.By letter dated January 4, 2013, MURR provided the NRC a copy of the current PSP and OperatorRequalification Program.On February 11, 2013, the NRC requested updated financial information in the form of four (4) questionsbecause 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) questionsregarding significant changes to the MIURR facility since submittal of the licensing renewal application inAugust 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 formoften (10) questions.*If there are any questions regarding this response, please contact me at (573) 882-5319 orFruitsJ@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 Directorxc: Reactor Advisory CommitteeReactor Safety SubcommitteeDr. Garnett S. Stokes, ProvostDr. Henry C. Foley, Senior Vice Chancellor for ResearchMr. Alexander Adams Jr., U.S. Nuclear Regulatory CommissionMr. Geoffrey Wertz, U.S. Nuclear Regulatory CommissionMr. Johnny Eads, U.S. Nuclear Regulatory CommissionAttachments: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 System8. Meteorological Data (Wind Speed and Class) -1961 to 19699. Meteorological Data (Wind Speed and Class) -1970 to 199010. Meteorological Data (Wind Speed and Class) -1961 to 199011. 10 CFR 835, Appendix C, "Derived Air Concentration (DAC) for Workers from ExternalExposure during Immersion in a Cloud of Airborne Radioactive Material"12. Micro Shield 8.02 Dose Calculations for a Fuel Handling, Fuel Failure, and Fueled ExperimentFailure Accidents13. Stack Effluent Releases -Calendar Years 2005 to 2014JACQUELINE 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 describesthe control blade drop process including the effect of the dashpot, but does not describe the methodfor determining the drop time nor does it explain the basis for the 80 percent insertion times. Thescram times and reactivity worths used or" assumed for the various analyses in the SAR are notclearly described or provided. NUREG-15 3 7, Section 4. 5.3, "Operating Limits, "provides guidancethat 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 theassociated control blade insertion reactivity per blade. Provide typical control blade fullinsertion 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 (seeAttachment 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 rodassembly, is removed by an electro-mechanical relay contact which allows the control blade to dropand 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 whentriggered, causes the electro-mechanical relay to change state stopping the blade droptimer/chronometer. The control blade drop time is then displayed on a meter on the reactor controlroom instrument panel. Table 1 provides the minimum, average and maximum drop times of allfour (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 0.48Average 0.50 0.54 0.50 0.52Maximum 0.59 0.58 0.54 0.54Current MURR Technical Specification 3 .2.c requires the capability of inserting the shim controlblades to their 20% withdrawn position (or 80% inserted) in less than 0.7 seconds. This ensuresprompt shutdown of the reactor in the event a reactor scram signal, manual or automatic, isreceived. The 20% withdrawn position is defined as 20% of the control blade full travel of 26inches measured from the fully inserted position. Below the 20% withdrawn position the controlblade fall is cushioned by a dashpot assembly. Approximately 91% of the control blade total worthis inserted at the 20% position. This is an original design feature of the reactor and its purpose hasnot been altered in 49 years of operation. The same Technical Specification will remain in therelicensing Technical Specifications.The measured and calculated values for reactor core excess reactivity and shutdown margin areprovided 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 isfully withdrawn). Some of this information, calculated using older computer programs, can also befound on Table 4-12 of the SAR.Typical MURR operations involve a core change-out every week with eight (8) xenon-free fuelelements in various stages of burnup (mixed core operation) used at startup. The reactor coreexcess reactivity and shutdown margin values are verified after the weekly core change-out. Theverification is done during reactor startup, when the cold, clean critical control blade height ismeasured. 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 reactivityand shutdown margin in comparison to the Technical Specification limit of -0.020 Ak/k.Table 2 -Summary of Key Measured Reactor DataValueParaeter(Ak/k)Typial ota shi cotro blae wrth0.1364Typical total shim control blade worth at 80% inserted -0.1127Typical shim control blade worth at 80% inserted with the highest worth -0.0787control blade excluded (or fully withdrawn)Maximum reactor core excess reactivity after weekly core change-out +0.0400One-year average of reactor core excess reactivity (over 69 core change-outs) +0.0290Typical core sub-criticality with 3 shim control blades at 80% inserted and the -038control blade excluded (or fully withdrawn)-037Minimum shutdown margin allowed by Technical Specifications -0.0200Calculated Values:Reactor core excess reactivity and shutdown margin values were also calculated using the detailedMGNP 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) andall fresh shim control blades (most conservative), and (2) with a mixed core loading and mixedburnup control blades (typical MUIRR operation). Table 3 provides the calculated values.6 of 86 Table 3 -Summary of Key Calculated Reactor DataValueParameter -All Fresh Fuel and Fresh Control Blades Case (kkReactor core excess reactivity 0.0865Total shim control blade worth 0.1740Core sub-criticality with 3 shim control blades at 80% inserted and highest -0.0324worth control blade excluded (or stuck fully withdrawn)ValueParameter -Mixed Core / Mixed Control Blades Case (Ak/k)Reactor core excess reactivity 0.0445Total shim control blade worth 0.1517Core sub-cniticality with 3 shim control blades at 80% inserted and highest -0.0580worth 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 worthshim control blade fully withdrawn), the minimum reactor core shutdown margin required by theTechnical 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 withdrawalfrom subcritical, control blade run in, insertion of excess reactivity, etc.). For each suchevent, provide the control blade motion speeds and reactivities utilized to provide the SARanalyses, 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 occurif there were a double-ended shear of the 12-inch primary coolant piping on both sides of either thecold-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 bethe 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 systeminstrumentation can have trip values set for them at the appropriate coolant loop locations. In theRELAP modeling, a 150 millisecond time delay is set between the time a scram signal is receivedand the modeling of when the "insertion" of the control blades start. The insertion is covered by aninput table of fission and gamma reactor power as per set time steps after the reactor scrams. Thecode calculates linear values between these data points.7 of 86 Table 4 below provides the power assumed by RELAP seconds after shutdown compared to thecalculated power after shutdown, assuming 30 days of full power operation, using equation 2.66from Nuclear Reactor Engineering 3rd Edition by Samuel Glasstone and Alexander Sesonske'. Theequation is given in the upper right corner of the page along with the values of variables a and b touse 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 powerexcept for the values for 8, 9 and 10 seconds. From 10 to 150 seconds, the RELAP values areconservative 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 useconservative calculated values of reactor decay heat after the scram, which would correspond toslower insertion of the control blades.See the response to RAI 6.a for control blade drop times related to Insertion of Excess Reactivityaccidents.
 
==References:==
1Glasstone, S. and Sesonske, A., Nuclear Reactor Engineering 3rd Edition, prepared underTechnical Information Center, United States Department of Energy.8 of 86 Table 4 -Comparing RELAP Decay Heat to Calculated Decay Heat(Nuclear Reactor Engineering 3rd Edition: Equation 2.66)ts00.10.30.71.02.03.05.06.07.08.09.01020304050607080901001201501802002403004004205406008001,0002,0004,0006,0008,00010,000Power MW11.0A7.99061.97 171.25971.03730.93090.848 10.73450.693 10.65760.62920.60440.58190.45060.41000.38690.36980.35630.34530.33600.32800.32100.3090A 0.30720.30530.30150.28490.26590.243 10.23950.22140.21420.19570.18240.14620.11670. 10200.09270.0860Power MW11.00.50990.45780.42010.40480.376 10.35990.34000.333 10.32730.32230.3 1800.3 141Equation 2.66P/P0=5E-3 *a *[t4-b -(To + t4)-~b]after shutdownTo = 30 days operating period prior toshutdown(s)0.1 to 1010 to 150150 to 8E80.49700.43 170.39700.37400.35690.34340.33240.323 10.3 1500.30800.296 10.2821a12.0515.3127.43b0.06390.18070.29620.32300.30500.295 10.27860.25950.23680.233 10.21500.20780.18930. 17600.13980.11030.09570.08630.0796Note A: RELAP does not have a value entered for 0.1 seconds, but the linear value between 0 and0.3 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 thelicensee provide analyses and methods to demonstrate the secure storage of new and irradiated fuelwith a criticality limit of keff < 0.90. The NRC staff's review of the MURR SAR and HazardsSummary Report could not find a criticality analysis supporting the use of any fuel storagelocations outside of the core. Identify the locations that may be used for the storage of new orirradiate fuel, and provide supporting criticality analyses, or justify' why no additional informationis needed.As stated in SAR Section 9.2.1, there are 88 in-pool storage locations for new or irradiated fuelelements. These storage locations are situated in three (3) areas within the reactor pool and aredesignated as the "X," "Y" and "Z" storage baskets. The "Z" storage basket contains 48 fuelelement storage locations; consisting of two (2) levels, referred to as "upper" and "lower," of 24locations per level. The "X" and "Y" storage baskets each contain 20 fuel element storagelocations. 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 storagelocations. The "X" and "Y" storage baskets each had only six (6) storage locations at the timewhile the "Z" storage basket consisted of 16 storage locations -two racks (6 and 10) in the lowerlevel. In 1972, due to an increase in operating schedule and with an uprate in power from 5 to 10MWs in the near future, an additional rack of eight (8) storage locations was added to the lowerlevel of the "Z" basket, thus providing a total of 36 fuel element storage locations in the pool (24 inthe "Z" basket). Modification Record 72-7, "Additional In-Pool Fuel Storage Basket," documentsthe installation of the eight (8) element rack (Attachment 2). On page 2a of the ModificationRecord, the following is stated: "To determine the safety of installing an additional fuel rackbetween the present two, the system was modeled using the Exterminator II multi-group neutrondiffusion program. The physical model consisted of three adjacent rows of eight clean 775 gramU235 fuel elements. Each fuel element was surrounded by 0.25" thick boral as is the case in theactual design. For the fully loaded rack, the calculated Keff linlit was 0. 714. "A 1/M criticality plotof 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 increasedthe overall capacity of the "Z" storage basket from 24 to 38. Modification Record 76-3, "Upper 7Spent 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 ofanother level of elements was modelled using the Exterminator II neutron diffusion code. Thepresence of 24 rather than 14 elements on the second level was used for a 'factor of safety." Thecode predicts a value for Keff of 0.748. Thus, the above criteria is satisfied for fuel storage." Al/M criticality plot of the storage basket was also performed to verify the Exterminator II coderesults.In 1978, a 10 element rack was added to the upper level of the "Z" storage basket which increasedthe 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 results10 of 86 stated in Modification Record 76-3, which conservatively modeled 24 fuel elements instead of just14 elements.Because the criticality analyses for the "Z" storage basket are somewhat dated and vaguelydocumented, 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 followingdescribes the methodology and results.The "Z" storage basket stores fuel elements that have burnups of 0 to 150 MWds. The baskets arelined with 26- to 29-inch tall sheets of 0.25- to 0.3125-inch thick BaC (BORAL) as the absorbingmaterial to prevent the stored fuel configuration from reaching criticality. Figure 1 shows thelayout (i.e. a detailed MCNP model) of the lower "Z" storage basket configuration.-Stainless Steel-Fuel Element-Pool WaterBORALFigure 1 -Detailed MCNP model of the Lower "Z" Storage Basket ConfigurationThe upper 'Z' storage basket configuration layout shown in Figure 2 is very similar to the lowerbasket with the exception of lead shields surrounding the basket instead of stainless steel, as in thelower basket.* Lead ShieldFigure 2 -Detailed MCNP model of the Upper "Z" Storage Basket Configuration(Lead shields instead of stainless steel)11 of 86 vThe active region of the fuel elements in the lower and upper baskets is separated in height byapproximately seven (7) inches. Each fuel element in every storage location is modeled in fulldetail, with all 24 aluminum clad UAlx fuel plates. Figure 3 shows very detailed MCNP modelingof an individual MURR fuel element and the elements in their lower and upper "Z" storage basketconfigurations.Upper Level of "Z" Storage BasketLower Level of "Z" Storage BasketFigure 3 -Panels Showing Detailed MCNP Modeling of the Fuel Elements; the Left PanelShowing the Axial Configuration of the Fuel Elements in the Lower and Upper "Z" StorageBaskets and the Right Panel Showing a Cross-sectional View of a MURR Fuel ElementCriticality (i.e. KCODE) calculations using MCNP version 5 with the ENDJF/B-VII.O data librarieswere performed for two detailed instances of the "Z" storage basket configuration: (1) a singlebasket (lower), and (2) both lower and upper baskets together. All calculations were performed for20 million source particles. For the two instances, the basket(s) were filled to their maximumcapacities (24 fuel elements) with fresh, highly-enriched uranium (HiEU) UAlx MURR fuelelements. 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 twoconfigurations of the "Z" storage basket.Table 1 -KIf Values for Worst-Case "Z' Storage Basket ConfigurationsConfiguration Fuel Status Storage Capacity IefLower Fresh Max -24 Fuel Elements 0.49885Lower + Upper Fresh Max -48 Fuel Elements 0.5586212 of 86 On receipt, fresh (i.e., un-irradiated fuel) fuel elements may be stored outside the reactor pool in adry, 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 drystorage configuration containing the maximum allowable number of on-site stored fresh MURRfuel 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 anyenrichment, providing that no more than 5 kilograms of this amount is unirradiated;...". SixMURR fuel elements, containing 775 grams of uranium-235 each, equals 4.65 kilograms.Air--Fuel ElementPlywoodBoric AcidFigure 4 -Detailed MCNP Model Showing Fresh Fuel Dry Storage Filled to Possession LimitTo establish a full-scope criticality safety study, in addition to the configuration described in Figure4, two other configurations were also defined to capture the worst-case scenarios: (1) a floodedconfiguration storing the maximum allowable number of on-site stored fresh fuel, i.e., six fresh fuelelements (see Figure 4 where air is replaced with water), and (2) a flooded configuration with therack filled to its maximum capacity which equals eight fresh fuel elements (see Figure 5).,...WaterFuel ElementPlywoodBoric AcidFigure 5 -Detailed MCNP Model Showing Fresh Fuel Dry Storage Filled to Physical Capacity13 of 86 Again, the criticality (i.e. KCODE) calculations performed using MCNP version 5 with theENDF/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 mostconservative, worst-cases (in terms of attaining criticality) for the fresh fuel storage configurations.Table 2 -K~ff Values for Worst-Case Fresh Fuel Storage ConfigurationsConfiguration Fuel Status Storage Capacity KfDry (Air) Fresh License Max -6 Fuel Elements 0.02344Flooded Fresh License Max -6 Fuel Elements 0.36228Flooded Fresh Storage Max -8 Fuel Elements 0.36258In 1991, due to the inability to ship spent fuel from the facility because the cask (GE-700) that wasused to ship research reactor fuel was removed from service, two (2) new fuel storage baskets werefabricated 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 theinstallation of the additional 24 fuel element storage locations (Attachment 5). On page 2 of theModification Record, the following is stated: "The evaluation performed for each MHJA basketwill 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"Xand "Y"' storage baskets, which increased the total storage capacity in these baskets from 36 to 40locations. 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 thenew "X" and "Y" storage baskets (Attachment 6). This Modification Record contains a detaileddescription of the criticality analysis performed for these two baskets using the MCNP code. Onpage 4 of the Modification Record, the following is stated: "The MCNP model was used tocalculate a Ke,'7 value of 0. 635 for one fuel basket fully loaded with twenty (20) 'fresh" 775 gramU-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 andfresh fuel storage in the vault, using conservative, worst-case assumptions of all fresh fuelelements, indicate IKfr values much less than the MURR Technical Specification Limit of 0.9 (novalue 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  
: 2. NUREG-1537, Section 9.2, "Handling and Storage of Reactor Fuel ", provides guidance that thelicensee provide analyses and methods to demonstrate the secure storage of new and irradiated fuelwith a criticality limit of keff < 0.90. The NRC staff's review of the MURR SAR and HazardsSummary Report could not find a criticality analysis supporting the use of any fuel storagelocations outside of the core. Identify the locations that may be used for the storage of new orirradiate fuel, and provide supporting criticality analyses, or justify' why no additional informationis needed.As stated in SAR Section 9.2.1, there are 88 in-pool storage locations for new or irradiated fuelelements. These storage locations are situated in three (3) areas within the reactor pool and aredesignated as the "X," "Y" and "Z" storage baskets. The "Z" storage basket contains 48 fuelelement storage locations; consisting of two (2) levels, referred to as "upper" and "lower," of 24locations per level. The "X" and "Y" storage baskets each contain 20 fuel element storagelocations. 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 storagelocations. The "X" and "Y" storage baskets each had only six (6) storage locations at the timewhile the "Z" storage basket consisted of 16 storage locations -two racks (6 and 10) in the lowerlevel. In 1972, due to an increase in operating schedule and with an uprate in power from 5 to 10MWs in the near future, an additional rack of eight (8) storage locations was added to the lowerlevel of the "Z" basket, thus providing a total of 36 fuel element storage locations in the pool (24 inthe "Z" basket). Modification Record 72-7, "Additional In-Pool Fuel Storage Basket," documentsthe installation of the eight (8) element rack (Attachment 2). On page 2a of the ModificationRecord, the following is stated: "To determine the safety of installing an additional fuel rackbetween the present two, the system was modeled using the Exterminator II multi-group neutrondiffusion program. The physical model consisted of three adjacent rows of eight clean 775 gramU235 fuel elements. Each fuel element was surrounded by 0.25" thick boral as is the case in theactual design. For the fully loaded rack, the calculated Keff linlit was 0. 714. "A 1/M criticality plotof 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 increasedthe overall capacity of the "Z" storage basket from 24 to 38. Modification Record 76-3, "Upper 7Spent 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 ofanother level of elements was modelled using the Exterminator II neutron diffusion code. Thepresence of 24 rather than 14 elements on the second level was used for a 'factor of safety." Thecode predicts a value for Keff of 0.748. Thus, the above criteria is satisfied for fuel storage." Al/M criticality plot of the storage basket was also performed to verify the Exterminator II coderesults.In 1978, a 10 element rack was added to the upper level of the "Z" storage basket which increasedthe 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 results10 of 86 stated in Modification Record 76-3, which conservatively modeled 24 fuel elements instead of just14 elements.Because the criticality analyses for the "Z" storage basket are somewhat dated and vaguelydocumented, 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 followingdescribes the methodology and results.The "Z" storage basket stores fuel elements that have burnups of 0 to 150 MWds. The baskets arelined with 26- to 29-inch tall sheets of 0.25- to 0.3125-inch thick BaC (BORAL) as the absorbingmaterial to prevent the stored fuel configuration from reaching criticality. Figure 1 shows thelayout (i.e. a detailed MCNP model) of the lower "Z" storage basket configuration.-Stainless Steel-Fuel Element-Pool WaterBORALFigure 1 -Detailed MCNP model of the Lower "Z" Storage Basket ConfigurationThe upper 'Z' storage basket configuration layout shown in Figure 2 is very similar to the lowerbasket with the exception of lead shields surrounding the basket instead of stainless steel, as in thelower basket.* Lead ShieldFigure 2 -Detailed MCNP model of the Upper "Z" Storage Basket Configuration(Lead shields instead of stainless steel)11 of 86 vThe active region of the fuel elements in the lower and upper baskets is separated in height byapproximately seven (7) inches. Each fuel element in every storage location is modeled in fulldetail, with all 24 aluminum clad UAlx fuel plates. Figure 3 shows very detailed MCNP modelingof an individual MURR fuel element and the elements in their lower and upper "Z" storage basketconfigurations.Upper Level of "Z" Storage BasketLower Level of "Z" Storage BasketFigure 3 -Panels Showing Detailed MCNP Modeling of the Fuel Elements; the Left PanelShowing the Axial Configuration of the Fuel Elements in the Lower and Upper "Z" StorageBaskets and the Right Panel Showing a Cross-sectional View of a MURR Fuel ElementCriticality (i.e. KCODE) calculations using MCNP version 5 with the ENDJF/B-VII.O data librarieswere performed for two detailed instances of the "Z" storage basket configuration: (1) a singlebasket (lower), and (2) both lower and upper baskets together. All calculations were performed for20 million source particles. For the two instances, the basket(s) were filled to their maximumcapacities (24 fuel elements) with fresh, highly-enriched uranium (HiEU) UAlx MURR fuelelements. 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 twoconfigurations of the "Z" storage basket.Table 1 -KIf Values for Worst-Case "Z' Storage Basket ConfigurationsConfiguration Fuel Status Storage Capacity IefLower Fresh Max -24 Fuel Elements 0.49885Lower + Upper Fresh Max -48 Fuel Elements 0.5586212 of 86 On receipt, fresh (i.e., un-irradiated fuel) fuel elements may be stored outside the reactor pool in adry, 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 drystorage configuration containing the maximum allowable number of on-site stored fresh MURRfuel 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 anyenrichment, providing that no more than 5 kilograms of this amount is unirradiated;...". SixMURR fuel elements, containing 775 grams of uranium-235 each, equals 4.65 kilograms.Air--Fuel ElementPlywoodBoric AcidFigure 4 -Detailed MCNP Model Showing Fresh Fuel Dry Storage Filled to Possession LimitTo establish a full-scope criticality safety study, in addition to the configuration described in Figure4, two other configurations were also defined to capture the worst-case scenarios: (1) a floodedconfiguration storing the maximum allowable number of on-site stored fresh fuel, i.e., six fresh fuelelements (see Figure 4 where air is replaced with water), and (2) a flooded configuration with therack filled to its maximum capacity which equals eight fresh fuel elements (see Figure 5).,...WaterFuel ElementPlywoodBoric AcidFigure 5 -Detailed MCNP Model Showing Fresh Fuel Dry Storage Filled to Physical Capacity13 of 86 Again, the criticality (i.e. KCODE) calculations performed using MCNP version 5 with theENDF/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 mostconservative, worst-cases (in terms of attaining criticality) for the fresh fuel storage configurations.Table 2 -K~ff Values for Worst-Case Fresh Fuel Storage ConfigurationsConfiguration Fuel Status Storage Capacity KfDry (Air) Fresh License Max -6 Fuel Elements 0.02344Flooded Fresh License Max -6 Fuel Elements 0.36228Flooded Fresh Storage Max -8 Fuel Elements 0.36258In 1991, due to the inability to ship spent fuel from the facility because the cask (GE-700) that wasused to ship research reactor fuel was removed from service, two (2) new fuel storage baskets werefabricated 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 theinstallation of the additional 24 fuel element storage locations (Attachment 5). On page 2 of theModification Record, the following is stated: "The evaluation performed for each MHJA basketwill 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"Xand "Y"' storage baskets, which increased the total storage capacity in these baskets from 36 to 40locations. 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 thenew "X" and "Y" storage baskets (Attachment 6). This Modification Record contains a detaileddescription of the criticality analysis performed for these two baskets using the MCNP code. Onpage 4 of the Modification Record, the following is stated: "The MCNP model was used tocalculate a Ke,'7 value of 0. 635 for one fuel basket fully loaded with twenty (20) 'fresh" 775 gramU-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 andfresh fuel storage in the vault, using conservative, worst-case assumptions of all fresh fuelelements, indicate IKfr values much less than the MURR Technical Specification Limit of 0.9 (novalue 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 CorePhysics Parameters, "provide guidance that the licensee should identify their analytical methods,including calculations of individual control blade worths, core excess reactivity, and coefficients ofreactivity, and compare the results with experimental measurements. The MURR SAR, Section 4.5states that analyses have been performed using PDQ, EXTREMINATOR, and BOLD-VENTUREcodes using RO, RZ, and ROZ models. The NRC staff noted other analyses (e.g., the RAI responsessupporting the NRC staff review of License Amendment No. 36, ADAMS Accession Nos.ML11237A088 and ML12150A052) used Estimated Critical Position (ECP) comparisons with theMonte Carlo Neutron Production code. The design code used to support the T&H analysis appearsto be DIF3D. The NRC staff is not clear as to which analytical method is the final supportinganalysis to be reviewed for the MURR license renewal application. The final supporting analysisshould 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.ML10306002 1), it is not clear how the stuck control blade was determined, what the relativereactivity worth is for the other control blades in the shutdown margin (SDM) analysis, andwhether 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, orjustify why this information is not needed.Historically, neutron physics modeling and analyses at MUJRR have been performed using severalmulti-group and multi-dimensional neutron diffusion theory codes such as PDQ,EXTERMINATOR-fl and BOLD VENTURE. Since the BOLD VENTURE core model wasbenchmarked against the destructive analysis of a highly-enriched uranium (HEU) MURR fuelelement for the license renewal application submitted to the NRC in August of 2006, MURR usedresults provided by the above set of neutronics codes.Since then, MURR core physics analyses have switched to using newer, state-of-the-art programssuch as MCNP for neutronic analysis. For a compact core such as MURR, it is preferable to use atransport 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 nowroutinely used for all calculations of core Ker, critical control blade height, detailed powerdistribution, and experimental fluxes/reaction rates.As part of the on-going collaboration, which started in 2006, between MURR staff and ArgonneNational Laboratory (ANL) analysts for the purpose of determining the feasibility of convertingMURR from HIEU to low-enriched uranium (LEU) fuel, ANL has assembled a neutronics analysiscode suite utilizing WIMS-ALNL, REBUS-DIF3D and REBUS-MCNP. Figure 1 below illustratesthe linkage of the codes in the analysis suite.The suite of programs, or codes, was used to provide detailed (radial, axial and azimuthal) fuelcomposition for partially burned fuel elements. Since MURR routinely operates with a fuel cycleutilizing a mixed burnup core, realistic experimental flux, reaction rates and power peaking valueshave to be evaluated for the typical core weekly cycles rather than for an all-fresh core. The15 of 86 detailed fuel composition data obtained is then subsequently used in a MCNP calculation to obtainthe worst-case power peaking factors and heat flux values used in the thermal-hydraulic analysis.tMCNP Runs (Outside REBUS)Produt *Detailed power distributionsLumped Fission PrdutExperimental fluxes / reaction ratesCross-Sections in 69 groupsMCNPinufi-UREBUS 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-ItIDIF3D Neutronics Solver* Cross-section interpolation* Flux solverIRegion FluxesRegion Reaction RatesI4Fi~A N~uCross-Section LibraryFigure 1 -Linkage of the Codes Used in the Analysis SuiteThe following is a brief description of each of the programs within the ANL neutronic analysissuite:WIMS-ANL: WIMS-ANL is a one-dimensional lattice physics code used to generate burnupdependent, 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 wasdeveloped by ANE based on the neutron spectrum that exists in the MURR core. This multi-groupdata 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 canmodel systems in a number of geometries. REBUS is a depletion code that utilizes neutron fluxesfrom a neutronics solver and cross-section data to solve isotopic transmutation calculations. A16 of 86 detailed O-R-Z diffusion MURR model was developed for DIF3D. The depleted corecharacteristics (plate-by-plate and axially-segmented atom densities) can be saved and passed on toMCNP for more detailed neutronics analyses.MCNP: MCNP is a continuous energy Monte Carlo neutron transport code. MCNP is capable ofmodeling the heterogeneous details of the MURR fuel elements, core structures, and experimentalfacilities 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 modelcores of depleted and fresh elements.ANL had performed extensive work to validate the above set of neutron physics codes and modelsfor application to MURR. The MCNP and DIF3D models were benchmarked against availableexperimental data [Ref. 1].In order to speed up routine neutronics calculations, where such detailed axial, radial and azimuthalfuel composition is not necessary, MURR utilizes the MONTEBURNS program. MONTEBURNSis 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 depletioncalculations and that of MCNP5 for continuous energy, flux and reaction rate as well as criticalitycalculations.MONTEBUJRNS by itself is not designed to handle transient calculations such as during the periodfrom reactor startup through critical and then on to steady-state reactor operation since it involvescontrol blade motion due to poison buildup as well as from fuel depletion. However, with the helpof in-house developed routines, a code system including MCNP and MONTEBURNS wasdeveloped 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-physicsanalysis is shown in Figure 2.17 of 86 A/
: 3. NUREG-153 7, Section 4.5.1, "Normal Operating Conditions," and Section 4. 5.2, "Reactor CorePhysics Parameters, "provide guidance that the licensee should identify their analytical methods,including calculations of individual control blade worths, core excess reactivity, and coefficients ofreactivity, and compare the results with experimental measurements. The MURR SAR, Section 4.5states that analyses have been performed using PDQ, EXTREMINATOR, and BOLD-VENTUREcodes using RO, RZ, and ROZ models. The NRC staff noted other analyses (e.g., the RAI responsessupporting the NRC staff review of License Amendment No. 36, ADAMS Accession Nos.ML11237A088 and ML12150A052) used Estimated Critical Position (ECP) comparisons with theMonte Carlo Neutron Production code. The design code used to support the T&H analysis appearsto be DIF3D. The NRC staff is not clear as to which analytical method is the final supportinganalysis to be reviewed for the MURR license renewal application. The final supporting analysisshould 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.ML10306002 1), it is not clear how the stuck control blade was determined, what the relativereactivity worth is for the other control blades in the shutdown margin (SDM) analysis, andwhether 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, orjustify why this information is not needed.Historically, neutron physics modeling and analyses at MUJRR have been performed using severalmulti-group and multi-dimensional neutron diffusion theory codes such as PDQ,EXTERMINATOR-fl and BOLD VENTURE. Since the BOLD VENTURE core model wasbenchmarked against the destructive analysis of a highly-enriched uranium (HEU) MURR fuelelement for the license renewal application submitted to the NRC in August of 2006, MURR usedresults provided by the above set of neutronics codes.Since then, MURR core physics analyses have switched to using newer, state-of-the-art programssuch as MCNP for neutronic analysis. For a compact core such as MURR, it is preferable to use atransport 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 nowroutinely used for all calculations of core Ker, critical control blade height, detailed powerdistribution, and experimental fluxes/reaction rates.As part of the on-going collaboration, which started in 2006, between MURR staff and ArgonneNational Laboratory (ANL) analysts for the purpose of determining the feasibility of convertingMURR from HIEU to low-enriched uranium (LEU) fuel, ANL has assembled a neutronics analysiscode suite utilizing WIMS-ALNL, REBUS-DIF3D and REBUS-MCNP. Figure 1 below illustratesthe linkage of the codes in the analysis suite.The suite of programs, or codes, was used to provide detailed (radial, axial and azimuthal) fuelcomposition for partially burned fuel elements. Since MURR routinely operates with a fuel cycleutilizing a mixed burnup core, realistic experimental flux, reaction rates and power peaking valueshave to be evaluated for the typical core weekly cycles rather than for an all-fresh core. The15 of 86 detailed fuel composition data obtained is then subsequently used in a MCNP calculation to obtainthe worst-case power peaking factors and heat flux values used in the thermal-hydraulic analysis.tMCNP Runs (Outside REBUS)Produt *Detailed power distributionsLumped Fission PrdutExperimental fluxes / reaction ratesCross-Sections in 69 groupsMCNPinufi-UREBUS 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-ItIDIF3D Neutronics Solver* Cross-section interpolation* Flux solverIRegion FluxesRegion Reaction RatesI4Fi~A N~uCross-Section LibraryFigure 1 -Linkage of the Codes Used in the Analysis SuiteThe following is a brief description of each of the programs within the ANL neutronic analysissuite:WIMS-ANL: WIMS-ANL is a one-dimensional lattice physics code used to generate burnupdependent, 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 wasdeveloped by ANE based on the neutron spectrum that exists in the MURR core. This multi-groupdata 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 canmodel systems in a number of geometries. REBUS is a depletion code that utilizes neutron fluxesfrom a neutronics solver and cross-section data to solve isotopic transmutation calculations. A16 of 86 detailed O-R-Z diffusion MURR model was developed for DIF3D. The depleted corecharacteristics (plate-by-plate and axially-segmented atom densities) can be saved and passed on toMCNP for more detailed neutronics analyses.MCNP: MCNP is a continuous energy Monte Carlo neutron transport code. MCNP is capable ofmodeling the heterogeneous details of the MURR fuel elements, core structures, and experimentalfacilities 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 modelcores of depleted and fresh elements.ANL had performed extensive work to validate the above set of neutron physics codes and modelsfor application to MURR. The MCNP and DIF3D models were benchmarked against availableexperimental data [Ref. 1].In order to speed up routine neutronics calculations, where such detailed axial, radial and azimuthalfuel composition is not necessary, MURR utilizes the MONTEBURNS program. MONTEBURNSis 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 depletioncalculations and that of MCNP5 for continuous energy, flux and reaction rate as well as criticalitycalculations.MONTEBUJRNS by itself is not designed to handle transient calculations such as during the periodfrom reactor startup through critical and then on to steady-state reactor operation since it involvescontrol blade motion due to poison buildup as well as from fuel depletion. However, with the helpof in-house developed routines, a code system including MCNP and MONTEBURNS wasdeveloped 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-physicsanalysis is shown in Figure 2.17 of 86 A/
* MONTEBURNS 2.0 -time dependent stepwiseMCNP coupled ORIGEN nuclear burnup codeSsystem (LANL)*Critical Rod search routine -adjustments tocontrol rod height based on repeated detailedMCNP KCODE calculations.*Returns a critical rod height if the KCODEk~e is 1.0000+/-0.03% and control rod is lessthan maximum travelFigure 2 -Code Suite Flow Diagram Implemented at MURR for Routine Core Physics AnalysisThese "routine" calculations utilize a very detailed MCNP MURR core model that has thefollowing key capabilities:*It can model MURR's "mixed-core" weekly fuel configuration, i.e., atom densities of variousisotopes in the fuel matrix, for a range of fuel element burnups from fresh (0 MWd or no fueldepletion) to spent status. For the Estimated Critical Position (ECP) calculations, fuelelement definitions can be individually selected from a fuel burnup database to simulate anycombination of eight (8), xenon-free fuel elements.*Similarly, it can include a mixture of four (4) independently depleted BORAL shim controlblades -each with a different axial and radial boron depletion profile based on its operationalhistory (or core residence time).*It has the ability to account for poison and gas buildup, and the reduction of beryllium atomdensity 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 accuratelyin order to reduce the error in the FCP calculation.*With the help of a critical control blade height search routine, starting from an initial estimateof the critical control blade height, a series of MCNP5 criticality (KCODE) calculations canbe performed in order to calculate the critical control blade height.18 of 86  
* MONTEBURNS 2.0 -time dependent stepwiseMCNP coupled ORIGEN nuclear burnup codeSsystem (LANL)*Critical Rod search routine -adjustments tocontrol rod height based on repeated detailedMCNP KCODE calculations.*Returns a critical rod height if the KCODEk~e is 1.0000+/-0.03% and control rod is lessthan maximum travelFigure 2 -Code Suite Flow Diagram Implemented at MURR for Routine Core Physics AnalysisThese "routine" calculations utilize a very detailed MCNP MURR core model that has thefollowing key capabilities:*It can model MURR's "mixed-core" weekly fuel configuration, i.e., atom densities of variousisotopes in the fuel matrix, for a range of fuel element burnups from fresh (0 MWd or no fueldepletion) to spent status. For the Estimated Critical Position (ECP) calculations, fuelelement definitions can be individually selected from a fuel burnup database to simulate anycombination of eight (8), xenon-free fuel elements.*Similarly, it can include a mixture of four (4) independently depleted BORAL shim controlblades -each with a different axial and radial boron depletion profile based on its operationalhistory (or core residence time).*It has the ability to account for poison and gas buildup, and the reduction of beryllium atomdensity 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 accuratelyin order to reduce the error in the FCP calculation.*With the help of a critical control blade height search routine, starting from an initial estimateof the critical control blade height, a series of MCNP5 criticality (KCODE) calculations canbe 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 ofxenon-i135 and other poisons in the core during reactor operation as well as the buildup anddecay of the poisons during shutdown and restart using the isotope buildup and decay/losscapability of MONTEBURNS.The system of codes and calculation methodology described previously has been benchmarkedextensively using actual weekly core refueling and reactor startup data. The response to Question2.a, which was included in the responses, dated July 31, 2015, to a Request for AdditionalInformation 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 MissouriResearch Reactor (MURR) Core from Highly-Enriched to Low-Enriched Uranium -CoreNeutron 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 thecorresponding measurements for the ECP (or excess reactivity) for a known critical controlblade configuration at zero power, no xenon condition, or justify why this information is notneeded.The code system that is currently used for reactor physics analysis at MURR has been benchmarkedextensively. One of the methods used for the benchmarking was by comparing the EstimatedCritical Position (ECP) calculations from the detailed MCNP MUJRR model against the actualstartup critical control blade height data from several weekly reactor startups. The detailed MCNPMURR core model includes depleted control blade data, beryllium aging effect (i.e., more and moregas molecules taking up the place of beryllium atoms with increasing run time), as well as detailedsample 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 themodel's ability to predict the ECP accurately under various core states (mixed burnup) and flux trapsample conditions. The comparison was performed over an eight (8) month period. Note that thereactor startups at MURR require an occasional "strainer" startup -where initial critical controlblade 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 DataCore Actual Critical Predicated Critical Peiae lxTaControl Blade Control BladeConfiguration Height (Inches) Height (Inches) I~f ConfigurationWeek of 1/28/2013 16.79 16.67 0.99993 StrainerWeek of 2/04/2013 16.52 16.27 0.99975 SamplesWeek of 4/29/2013 15.98 15.78 1.00017 SamplesWeek of 6/10/2013 15.44 15.42 0.99995 SamplesWeek of 8/05/2013 16.74 16.74 0.99985 StrainerWeek of 8/12/2013 15.71 15.61 0.99985 SamplesWeek of 8/19/2013 15.84 15.84 1.00016 SamplesWeek of 8/26/2013 15.64 15.69 1.00029 SamplesA negative bias of ~1.5% is seen in the predictions for the early benchmarks. After the additionalrefinements 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, andRegulating blades) for a given core configuration at a low power, no xenon condition, orjustify 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, thefour BORAL shim control blades (Shim-i, Shim-2, Shim-3 and Shim-4, also referred to ascontrol blades 'A,' 'B,' 'C' and 'D') are in various stages of burnup (core residence time) rangingfrom 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 andreplaced with another rebuilt offset mechanism and a different control blade with a different burnupstatus. This schedule satisfies the Technical Specification surveillance requirement of inspectingone (1) out of four (4) control blades every six (6) months so that every blade is inspected everytwo (2) years. In this way, a given control blade is cycled in and out of the reactor multiple timesfrom 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 inchesof the control blade tip undergoes significant boron depletion with operation. Only the controlblade tip experiences burnup since during steady-state, full power operation the control blades arealmost fully withdrawn, resulting in the active neutron absorbing region of the blades being out ofany significant neutron flux. Since accurate control blade worth information is crucial for reactoroperation, every six (6) months when a control blade is replaced, a blade worth measurement of theinstalled control blade is performed.20 of 86 Using the detailed MCNP core model of MUIRR, the differential and integral worth of the four shimcontrol blades, and that of the stainless steel regulating blade, were calculated and the results areshown in Figures 3 through 6. The calculations were performed for fresh (non-depleted) controlblades using a fresh core with no xenon. In order to show the effect of control blade depletion withoperational history, the differential and integral worth curves of a single blade with a core residencetime of over 9 years are also shown in Figures 7 and 8, respectively.0.004y l E--07' -7E-0t,', 8E*05x'+ 2E-05X- 7E-06V = 9E-08,'- 5E-6'.0x 7E-05,' -6E-0OX + 3E-050.035 7E-Og, -4E-O6,,3/4+4E-O~x' ,O002O + 1E-O5% y = 1E-O7,'- 5E-O6x; 7E-05,,'. 1E-O4x- 2E-050.0025// / CN CpAA N-~shh0.0015* MCNP hp/ANt All-fresh ShimAS0.0015 /' 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-rhSimA0.0005 --Poly. (MCNP hp/AN A4ll-fresh Shim O '---PI I(CN h/ll All-fresh Shim O* ~/ .PoIy. (MCNP hp""l-fehSi0510 15 20 25 30Control Blade Height Withdrawn (inches)Figure 3 -Calculated Differential Worth Curves for All-Fresh Shim Control Bladesin an All-Fresh Fuel Core ConfigurationShim 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 onthe west side of the core, adjacent to these control blades.21 of 86 0.060000MCNP Integral Rod Worth Shim AU 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 &#xf7; .Os5,&#xf7; 1E.OSe'. ZE-ise &#xf7;ZE-130.020000 = 9E-07Xa + 1E-05X3+ 9EtDSx3/4 ZE-13X+r 2E-130.000000 -,,"1 -05 10 15 20 25 30Control Blade Height Withdrawn (inches)Figure 4 -Calculated Integral Worth Curves for Ali-Fresh Shim Control bladesin an Ali-Fresh Fuel Core Configuration0.000250/4' MCNP Ag/AH \a" -.-Poly. (MCNP 0.000150 /~/0o.0o01oo /\/-..../ y = -8,7729E-1D0'- 2,9792E-8x3 + 1,0668E-06xz + 7.8270E-O6x/ Rz = 9.9953E-010.0o00000 ')0 5 10 15 20 25 30Regulating Blade Height Withdrawn (inches)Figure 5 -Calculated Differential Worth Curve for a Fresh Regulating Bladein an Ali-Fresh Fuel Core Configuration22 of 86 O.0000MOWP 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. 25.1405 10 15 20 25 30Regulating Blade Height Withdrawn (inches)Figure 6 -Calculated Integral Worth Curve for a Fresh Regulating Bladein an All-Fresh Fuel Core Configuration0.0030.025', Y -4.2163E-O~x' + 3.6065E-06x3 -1.O055E-O4x7 + 9.1731E-04iq 0.02 I: 9.9715E-01* I0.0015 IQ MCNP A/AHI=/ --- Poly. (MCNP h4p/AH) \s0.000:15 ""20 25-Control Blade Height withdrawn (inches)-0).0005Control 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 0O0000.050000.04000MCNP control blade Id: 6-.05 worth after 9.0 yearcurrently in position B.= 0.02000io/* y = -8E-Ogx5 + 9E-O7x4 -3E-O5x3 + O.O005x' + tE-t3x + 2E-130.01000 A)0 5 10 15 20 25 30Regulating 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, thereactivity 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 physicscalculations (such as ECP predictions, reactor core shutdown margin, estimation of unknownsample reactivity worths, etc.). To serve as a benchmark for the calculated control blade worths, asingle blade was selected. Using a detailed mixed-core, mixed-bumup control blade model of thereactor configuration during the last Shim-4 (control blade 'D') inspection and replacement theblade 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 areshown in Figures 9 and 10.24 of 86 0.003u0.0025S0.002L.* 0.0015*n 0.001-0000500.-).0005Control Blade Height Withdrawn (inches)Figure 9 -Comparison of Measured and Calculated Differential Worth Curvesfor Shim Control Blade 'D'0.0600U, 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 )S0.03000005 01 2 53Coto ld Hih ihran (n hsFigure 10 -Comparison of Measured and Calculated Integral Worth Curvesfor Shim Control Blade 'D'25 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 ofxenon-i135 and other poisons in the core during reactor operation as well as the buildup anddecay of the poisons during shutdown and restart using the isotope buildup and decay/losscapability of MONTEBURNS.The system of codes and calculation methodology described previously has been benchmarkedextensively using actual weekly core refueling and reactor startup data. The response to Question2.a, which was included in the responses, dated July 31, 2015, to a Request for AdditionalInformation made by the NRC (by letter dated June 18, 2015), contains the benchmark data.
: d. Provide a calculated and measured temperature coefficient for a given core configuration ata 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) ofthe 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 computerprograms that were described in the response to Question 3.a. The results are provided in Table 2below.Table 2 -MURR Primary Coolant Temperature CoefficientMURR Technical ANL-MCNP CalculationCoefficient Specification Limit (294 to 400 K)All Fresh Core, BOC:-13.2 x 10-5 -2.3 x 10-7Ak/k/&deg;FAverage Core Temperature MxdCr.BCPrimary Coolant Coefficient Shall be More -12.8 x 10.5 + 2.2 x 10.7 Ak/k/&deg;FTemperature Negative Than:(Isothermal) Mixed Core, Eqi. Xe:-6.0x 15Ak//0F-11.8 x 10s +/-- 2.2 x 107 Ak/k/&deg;FMixed Core, BOC (ENDF7):-12.5 x 10.5 + 2.2 x 10-7 Ak/k/&deg;FNote: BOC = Beginning of Cycle.26 of 86  
 
==References:==
'Stillman, J., et al., Technical Basis in Support of the Conversion of the University of MissouriResearch Reactor (MURR) Core from Highly-Enriched to Low-Enriched Uranium -CoreNeutron 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 thecorresponding measurements for the ECP (or excess reactivity) for a known critical controlblade configuration at zero power, no xenon condition, or justify why this information is notneeded.The code system that is currently used for reactor physics analysis at MURR has been benchmarkedextensively. One of the methods used for the benchmarking was by comparing the EstimatedCritical Position (ECP) calculations from the detailed MCNP MUJRR model against the actualstartup critical control blade height data from several weekly reactor startups. The detailed MCNPMURR core model includes depleted control blade data, beryllium aging effect (i.e., more and moregas molecules taking up the place of beryllium atoms with increasing run time), as well as detailedsample 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 themodel's ability to predict the ECP accurately under various core states (mixed burnup) and flux trapsample conditions. The comparison was performed over an eight (8) month period. Note that thereactor startups at MURR require an occasional "strainer" startup -where initial critical controlblade 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 DataCore Actual Critical Predicated Critical Peiae lxTaControl Blade Control BladeConfiguration Height (Inches) Height (Inches) I~f ConfigurationWeek of 1/28/2013 16.79 16.67 0.99993 StrainerWeek of 2/04/2013 16.52 16.27 0.99975 SamplesWeek of 4/29/2013 15.98 15.78 1.00017 SamplesWeek of 6/10/2013 15.44 15.42 0.99995 SamplesWeek of 8/05/2013 16.74 16.74 0.99985 StrainerWeek of 8/12/2013 15.71 15.61 0.99985 SamplesWeek of 8/19/2013 15.84 15.84 1.00016 SamplesWeek of 8/26/2013 15.64 15.69 1.00029 SamplesA negative bias of ~1.5% is seen in the predictions for the early benchmarks. After the additionalrefinements 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, andRegulating blades) for a given core configuration at a low power, no xenon condition, orjustify 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, thefour BORAL shim control blades (Shim-i, Shim-2, Shim-3 and Shim-4, also referred to ascontrol blades 'A,' 'B,' 'C' and 'D') are in various stages of burnup (core residence time) rangingfrom 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 andreplaced with another rebuilt offset mechanism and a different control blade with a different burnupstatus. This schedule satisfies the Technical Specification surveillance requirement of inspectingone (1) out of four (4) control blades every six (6) months so that every blade is inspected everytwo (2) years. In this way, a given control blade is cycled in and out of the reactor multiple timesfrom 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 inchesof the control blade tip undergoes significant boron depletion with operation. Only the controlblade tip experiences burnup since during steady-state, full power operation the control blades arealmost fully withdrawn, resulting in the active neutron absorbing region of the blades being out ofany significant neutron flux. Since accurate control blade worth information is crucial for reactoroperation, every six (6) months when a control blade is replaced, a blade worth measurement of theinstalled control blade is performed.20 of 86 Using the detailed MCNP core model of MUIRR, the differential and integral worth of the four shimcontrol blades, and that of the stainless steel regulating blade, were calculated and the results areshown in Figures 3 through 6. The calculations were performed for fresh (non-depleted) controlblades using a fresh core with no xenon. In order to show the effect of control blade depletion withoperational history, the differential and integral worth curves of a single blade with a core residencetime of over 9 years are also shown in Figures 7 and 8, respectively.0.004y l E--07' -7E-0t,', 8E*05x'+ 2E-05X- 7E-06V = 9E-08,'- 5E-6'.0x 7E-05,' -6E-0OX + 3E-050.035 7E-Og, -4E-O6,,3/4+4E-O~x' ,O002O + 1E-O5% y = 1E-O7,'- 5E-O6x; 7E-05,,'. 1E-O4x- 2E-050.0025// / CN CpAA N-~shh0.0015* MCNP hp/ANt All-fresh ShimAS0.0015 /' 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-rhSimA0.0005 --Poly. (MCNP hp/AN A4ll-fresh Shim O '---PI I(CN h/ll All-fresh Shim O* ~/ .PoIy. (MCNP hp""l-fehSi0510 15 20 25 30Control Blade Height Withdrawn (inches)Figure 3 -Calculated Differential Worth Curves for All-Fresh Shim Control Bladesin an All-Fresh Fuel Core ConfigurationShim 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 onthe west side of the core, adjacent to these control blades.21 of 86 0.060000MCNP Integral Rod Worth Shim AU 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 &#xf7; .Os5,&#xf7; 1E.OSe'. ZE-ise &#xf7;ZE-130.020000 = 9E-07Xa + 1E-05X3+ 9EtDSx3/4 ZE-13X+r 2E-130.000000 -,,"1 -05 10 15 20 25 30Control Blade Height Withdrawn (inches)Figure 4 -Calculated Integral Worth Curves for Ali-Fresh Shim Control bladesin an Ali-Fresh Fuel Core Configuration0.000250/4' MCNP Ag/AH \a" -.-Poly. (MCNP 0.000150 /~/0o.0o01oo /\/-..../ y = -8,7729E-1D0'- 2,9792E-8x3 + 1,0668E-06xz + 7.8270E-O6x/ Rz = 9.9953E-010.0o00000 ')0 5 10 15 20 25 30Regulating Blade Height Withdrawn (inches)Figure 5 -Calculated Differential Worth Curve for a Fresh Regulating Bladein an Ali-Fresh Fuel Core Configuration22 of 86 O.0000MOWP 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. 25.1405 10 15 20 25 30Regulating Blade Height Withdrawn (inches)Figure 6 -Calculated Integral Worth Curve for a Fresh Regulating Bladein an All-Fresh Fuel Core Configuration0.0030.025', Y -4.2163E-O~x' + 3.6065E-06x3 -1.O055E-O4x7 + 9.1731E-04iq 0.02 I: 9.9715E-01* I0.0015 IQ MCNP A/AHI=/ --- Poly. (MCNP h4p/AH) \s0.000:15 ""20 25-Control Blade Height withdrawn (inches)-0).0005Control 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 0O0000.050000.04000MCNP control blade Id: 6-.05 worth after 9.0 yearcurrently in position B.= 0.02000io/* y = -8E-Ogx5 + 9E-O7x4 -3E-O5x3 + O.O005x' + tE-t3x + 2E-130.01000 A)0 5 10 15 20 25 30Regulating 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, thereactivity 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 physicscalculations (such as ECP predictions, reactor core shutdown margin, estimation of unknownsample reactivity worths, etc.). To serve as a benchmark for the calculated control blade worths, asingle blade was selected. Using a detailed mixed-core, mixed-bumup control blade model of thereactor configuration during the last Shim-4 (control blade 'D') inspection and replacement theblade 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 areshown in Figures 9 and 10.24 of 86 0.003u0.0025S0.002L.* 0.0015*n 0.001-0000500.-).0005Control Blade Height Withdrawn (inches)Figure 9 -Comparison of Measured and Calculated Differential Worth Curvesfor Shim Control Blade 'D'0.0600U, 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 )S0.03000005 01 2 53Coto ld Hih ihran (n hsFigure 10 -Comparison of Measured and Calculated Integral Worth Curvesfor Shim Control Blade 'D'25 of 86  
: d. Provide a calculated and measured temperature coefficient for a given core configuration ata 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) ofthe 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 computerprograms that were described in the response to Question 3.a. The results are provided in Table 2below.Table 2 -MURR Primary Coolant Temperature CoefficientMURR Technical ANL-MCNP CalculationCoefficient Specification Limit (294 to 400 K)All Fresh Core, BOC:-13.2 x 10 2.3 x 10-7Ak/k/&deg;FAverage Core Temperature MxdCr.BCPrimary Coolant Coefficient Shall be More -12.8 x 10.5 + 2.2 x 10.7 Ak/k/&deg;FTemperature Negative Than:(Isothermal) Mixed Core, Eqi. Xe:-6.0x 15Ak//0F-11.8 x 10s +/-- 2.2 x 107 Ak/k/&deg;FMixed Core, BOC (ENDF7):-12.5 x 10.5 + 2.2 x 10-7 Ak/k/&deg;FNote: BOC = Beginning of Cycle.26 of 86  
: 4. NUREG-1537, Section 4.5.3, "Operating Limits, "provides guidance that licensees demonstratethat their facility has sufficient control blade worth to achieve the required shutdown reactivityassuming that all scrammable control blades are released upon scram, but the most reactive bladeremains in its most reactive position. The NRC staff could not find this information in the MURRSAR, but noted a reference in the 1971 Low Power Testing Program that indicated that theshutdown margin control blade reactivity was determined using 66 percent of the 4 shim bladeinsertion worth. Explain how MURR ensures adequate SDM, whether and if so, how the 66 percentfactor from the 1971 Low Power Testing Program is used, or justify why this information is notneeded.Table 4-12, on Pages 4-41 and 4-42 of the SAR, contains the value for the reactor core shutdownmargin. The Table lists the maximum K~ff with the highest worth shim control blade fullywithdrawn, or stuck, as 0.93 8. This maximum K~ff, value translates to a minimum reactor shutdownmargin value of -0.066 Ak/k. This compares with the Technical Specification minimum reactorcore 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 theresponse to Question 3.a, is -0.0875 Ak/k for an all-fresh fuel core case [Note: license possessionlimit only allows six (6) fresh fuel elements onsite].27 of 86  
: 4. NUREG-1537, Section 4.5.3, "Operating Limits, "provides guidance that licensees demonstratethat their facility has sufficient control blade worth to achieve the required shutdown reactivityassuming that all scrammable control blades are released upon scram, but the most reactive bladeremains in its most reactive position. The NRC staff could not find this information in the MURRSAR, but noted a reference in the 1971 Low Power Testing Program that indicated that theshutdown margin control blade reactivity was determined using 66 percent of the 4 shim bladeinsertion worth. Explain how MURR ensures adequate SDM, whether and if so, how the 66 percentfactor from the 1971 Low Power Testing Program is used, or justify why this information is notneeded.Table 4-12, on Pages 4-41 and 4-42 of the SAR, contains the value for the reactor core shutdownmargin. The Table lists the maximum K~ff with the highest worth shim control blade fullywithdrawn, or stuck, as 0.93 8. This maximum K~ff, value translates to a minimum reactor shutdownmargin value of -0.066 Ak/k. This compares with the Technical Specification minimum reactorcore 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 theresponse to Question 3.a, is -0.0875 Ak/k for an all-fresh fuel core case [Note: license possessionlimit 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 licenseeto characterize the dose for the maximally exposed individual, at the location of the nearestpermanent residence, and at any locations of special interest in the unrestricted area.a. The MURR SAR, Appendix B, contains summary information regarding the radiologicalimpacts of the MURR generated release of Argon 41 (Ar-41) during normal operations. TheMURR methodology includes an equation on SAR page B-J O that is used to alter the effectivestack height used in the dose calculations to compensate for elevation changes of the receptordue 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 theDavidson empirical model which has limited supporting data. Describe how the effectivestack height calculations are performed for the unique topography surrounding MURR, andhow the results are sufficiently conservative for the estimation of dose, or justify why noadditional information is needed.MURR calculates effective stack height, for the purposes of determining dose from radionuclideemissions, as the difference in vertical elevation between the point of emission at the end of theMURR exhaust stack and the receptor height at the point of interest, plus the effective stack heightcalculated using the Davidson equation. This equation takes into account the stack diameter andexhaust velocity of the gases leaving MURR to calculate an injection height and thus an effectivestack height into the atmosphere. Wind speed is also an input parameter into this formula as it is afunction of the particular Pasquill atmospheric stability class that is being modeled for the generalwind 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 ". .. oftengreatly underestimates observed rises,..." this underestimation would cause the offsite dosecalculations using the Pasquill-Guifford model to overestimate doses to the individual at the pointof dose calculation interest. In fact, dose estimates generated using this model are not out of linewith doses calculated using the COMPLY2 computer code which is used to determine annual doses(demonstrate compliance) to the nearest resident from MURR as part of the facility's annualNational Emission Standards for Hazardous Air Pollutants (NESHAPS) compliance report.Additionally, using Briggs' own equations for calculation of effective stack heights from the samereference book "Plume Rise," confirms that while the Davidson model underestimates effectivestack heights, these underestimated effective stack heights lead to an overestimation of dose, thusproviding a conservative approach to the offsite dose calculations. Thus, we feel that no additionalinformation is required.References:'Briggs, G.A., Plume Rise, AEC Critical Review Series, U.S. Atomic Energy Commission,Division of Technical Information, 1969.2COMLY is a computerized screening tool for evaluating radiation exposure from atmosphericreleases 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 andSubpart I.28 of 86  
: 5. NUREG-1537, Section 11.1.1.], "Airborne Radiation Sources, "provides guidance for the licenseeto characterize the dose for the maximally exposed individual, at the location of the nearestpermanent residence, and at any locations of special interest in the unrestricted area.a. The MURR SAR, Appendix B, contains summary information regarding the radiologicalimpacts of the MURR generated release of Argon 41 (Ar-41) during normal operations. TheMURR methodology includes an equation on SAR page B-J O that is used to alter the effectivestack height used in the dose calculations to compensate for elevation changes of the receptordue 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 theDavidson empirical model which has limited supporting data. Describe how the effectivestack height calculations are performed for the unique topography surrounding MURR, andhow the results are sufficiently conservative for the estimation of dose, or justify why noadditional information is needed.MURR calculates effective stack height, for the purposes of determining dose from radionuclideemissions, as the difference in vertical elevation between the point of emission at the end of theMURR exhaust stack and the receptor height at the point of interest, plus the effective stack heightcalculated using the Davidson equation. This equation takes into account the stack diameter andexhaust velocity of the gases leaving MURR to calculate an injection height and thus an effectivestack height into the atmosphere. Wind speed is also an input parameter into this formula as it is afunction of the particular Pasquill atmospheric stability class that is being modeled for the generalwind 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 ". .. oftengreatly underestimates observed rises,..." this underestimation would cause the offsite dosecalculations using the Pasquill-Guifford model to overestimate doses to the individual at the pointof dose calculation interest. In fact, dose estimates generated using this model are not out of linewith doses calculated using the COMPLY2 computer code which is used to determine annual doses(demonstrate compliance) to the nearest resident from MURR as part of the facility's annualNational Emission Standards for Hazardous Air Pollutants (NESHAPS) compliance report.Additionally, using Briggs' own equations for calculation of effective stack heights from the samereference book "Plume Rise," confirms that while the Davidson model underestimates effectivestack heights, these underestimated effective stack heights lead to an overestimation of dose, thusproviding a conservative approach to the offsite dose calculations. Thus, we feel that no additionalinformation is required.
 
==References:==
'Briggs, G.A., Plume Rise, AEC Critical Review Series, U.S. Atomic Energy Commission,Division of Technical Information, 1969.2COMLY is a computerized screening tool for evaluating radiation exposure from atmosphericreleases 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 andSubpart I.28 of 86  
: b. SAR page B-il has an equation for X/Q that includes the cy and arz dispersion factors. TheNRC 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 andare now included in the corrected Tables B-2 and B-3 below.TABLE B-2MAXIMUM ANNUAL INDIVIDUAL DOSE AT 150 METERSLocation: 150 Meters Directly NorthElevation at Man Height: 636 Feet (194 Meters) ______ ____Eff. Height ay X %s Dose with %sClas in I in (n) se/r3) (tiCi/ml or Comb. (mremly)Class (m)___ (m)_ (m)_ (sec/m3)_ Gi/m3) _________A 35 33 23 6.27E-05 3.14E-09 2.40E-04 0.00B 27 23 15 6.09E-05 3.04E-09 5.10E-03 0.08C 23 17 11 4.55E-05 2.28E-09 1.70E-02 0.19D 20 12 7 1.14E-05 5.71E-10 6.30E-02 0.18E 23 8.5 5 4.76E-08 2.38E-12 3.10E-02 0.00F 30 6 3.2 5.24E-22 2.62E-26 1.50E-02 0.00Total 0.46TABLE B-3MAXIMUM ANNUAL INDIVIDUAL DOSE AT 760 METERSLocation: 760 Meters Directly NorthElevation at ManHeight:_700 Feet (213 Meters)____________Eff. Height ay Oz x/Q X %s Dose with %sClass (in) I(in) (mn) (sec/rn3) i/mi or Comb. (inrem/y)____I 1 _____ __________ Ci/m3) _________A 16 170 270 3.30E-06 1.65E-10 2.40E-04 0.00B 8 120 85 1.04E-05 5.18E-10 5.10E-03 0.01C 4 85 52 1.71E-05 8.55E- 10 1.70E-02 0.07D 1 55 26 3.97E-05 1.99E-09 6.30E-02 0.63E 4 42 18 1.03E-04 5.13E-09 3.10E-02 0.80F 11 30 12 2.23E-04 1.12E-08 1.50E-02 0.84Total 2.3529 of 86 Note: Tbe "%s Comb." column was added to Tables B-2 and B-3 to better aid in understanding thecalculation of total dose based on the Pasquill-Guifford stability classes and wind direction.30 of 86  
: b. SAR page B-il has an equation for X/Q that includes the cy and arz dispersion factors. TheNRC 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 andare now included in the corrected Tables B-2 and B-3 below.TABLE B-2MAXIMUM ANNUAL INDIVIDUAL DOSE AT 150 METERSLocation: 150 Meters Directly NorthElevation at Man Height: 636 Feet (194 Meters) ______ ____Eff. Height ay X %s Dose with %sClas in I in (n) se/r3) (tiCi/ml or Comb. (mremly)Class (m)___ (m)_ (m)_ (sec/m3)_ Gi/m3) _________A 35 33 23 6.27E-05 3.14E-09 2.40E-04 0.00B 27 23 15 6.09E-05 3.04E-09 5.10E-03 0.08C 23 17 11 4.55E-05 2.28E-09 1.70E-02 0.19D 20 12 7 1.14E-05 5.71E-10 6.30E-02 0.18E 23 8.5 5 4.76E-08 2.38E-12 3.10E-02 0.00F 30 6 3.2 5.24E-22 2.62E-26 1.50E-02 0.00Total 0.46TABLE B-3MAXIMUM ANNUAL INDIVIDUAL DOSE AT 760 METERSLocation: 760 Meters Directly NorthElevation at ManHeight:_700 Feet (213 Meters)____________Eff. Height ay Oz x/Q X %s Dose with %sClass (in) I(in) (mn) (sec/rn3) i/mi or Comb. (inrem/y)____I 1 _____ __________ Ci/m3) _________A 16 170 270 3.30E-06 1.65E-10 2.40E-04 0.00B 8 120 85 1.04E-05 5.18E-10 5.10E-03 0.01C 4 85 52 1.71E-05 8.55E- 10 1.70E-02 0.07D 1 55 26 3.97E-05 1.99E-09 6.30E-02 0.63E 4 42 18 1.03E-04 5.13E-09 3.10E-02 0.80F 11 30 12 2.23E-04 1.12E-08 1.50E-02 0.84Total 2.3529 of 86 Note: Tbe "%s Comb." column was added to Tables B-2 and B-3 to better aid in understanding thecalculation 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 facilitydesign features, safety limits, limiting safety system settings, and limiting conditions for operationhave been selected to ensure that no credible accident could lead to unacceptable radiologicalconsequences to people or the environment. The NRC staff review examined the analyses providedin 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 theLimiting Safety System Setting setpoint in TS 2.2 (12.S MW). The temperature feedbackcoefficient used is -7.0 x lO Ak/k rather than the TS S.3.a value of -6xlO-5 Ak/k. It is unclearwhat peaking factors are employed. SAR Figure 13.2 seems to indicate that the scram timeused is faster than the value in TS 3.2.c. The acceptability of the results is based uponwhether 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 safetyconclusions 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 10MW 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 andAcceptance Criteria, for the Insertion of Excess Reactivity accident, "The accident scenarioassumes 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 wasreanalyzed 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 SafetySystem Setting (LSSS) set point of 12.5 MW, since the rod run-in system will initiate a rod run-inat 11.5 MW (Technical Specification 3.2.f.1) and shutdown the reactor prior to reaching the LSSSscram set point of 125%.For the SAR analysis of the Insertion of Excess Reactivity accident, the temperature coefficientused was -6.0 x 10.5 Ak/k and not -7.0 x 10-5 Ak/k as stated above. Third paragraph on Page 13-17of the SAR lists the various reactivity coefficients assumed for the Insertion of Excess Reactivityaccident analysis.Details regarding the power peaking factors used were not provided in that section of the SAR. Thepower peaking values used were values obtained based on the destructive analysis of a MURR fuelelement. For the updated analysis, more up-to-date power peaking values, based on the detailedMCNP MURR core model, were used.For both the SAR analyses, as well as for the updated analysis presented here, the control bladeinsertion times are based on the current and relicensing Technical Specification 3.2.c requirementof insertion to the 20% withdrawn position in less than 0.7 seconds. So the insertion rate wascalculated 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 monthly31 of 86 control blade drop time verifications performed at MURR have always yielded insertion times of0.6 seconds or less (see response to RAI 1 .a).Similar to the SAR analysis, the Reactivity Transient Analysis program PARET (V7.5), maintainedand distributed by the Nuclear Engineering Division of Argonne National Laboratory (ANL) wasused. For the Insertion of Excess Reactivity accident analysis, two channels were modeled inPARET; a hot channel representing worst-case conditions inside the core and an average channelrepresenting the rest of the core experiencing "average" conditions. The axial power profiles usedfor this 2-channel PARET reactivity transient analysis are given in Table 1 below.Table 1 -Peaking Factors in the Hot and Average ChannelsHot Channel Average Channel2.046 1.0581.971 0.9202.145 1.0182.335 1.1322.497 1.2192.672 1.3072.835 1.3602.986 1.4113.105 1.4303.164 1.4373.169 1.4203.098 1.3832.953 1.3262.775 1.2432.542 1.1402.290 0.9892.069 0.8281.888 0.7011.703 0.6151.499 0.5301.277 0.4601.080 0.3860.904 0.3290.880 0.358As indicated earlier, the transient was started from an initial power level of 11.5 MW with corecoolant flow rate as well as core coolant inlet temperatures set at their LSSS values of 3,200 gpm32 of 86 and 155 'F, respectively. Also, pressurizer pressure was at 75 psia (LSSS value). Since theInsertion of Excess Reactivity transient was analyzed from a starting power level of 11.5 MW, therod run-in that would be initiated by the rod run-in system at 11.5 MW was bypassed and only thehigh power scram set point of 12.5 MW was modeled. Also, a delay of 150 milliseconds wasincorporated into the control blade scram model so that the control blades would only start to insert0.15 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. Asexpected, due to the higher starting core power level, much lower core coolant flow rate and muchhigher than normal core coolant inlet temperature conditions assumed for this updated analysis, thepeak power during the transient momentarily reaches approximately 37.4 MW compared to a valueof approximately 33.0 MW reported in the SAR analysis for the same 600 pcm step reactivityinsertion.40.0040035.0030.001o 20.00-POWER MW-'l'dad "C-I'f maxtCS350300l0S50* 03.000.1000.50 1.00 1.50 2.00Time (seconds)2.50Figure 1 -Reactor Power, Fuel and Cladding Temperatures vs. Timefor a Positive Reactivity Step Insertion of 0.006 Ak/kSeveral SPERT tests had shown that the reactor can withstand such short duration (fewmilliseconds) power burst without sustaining any fuel damage and only sustained operation at suchhigh power levels will lead to fuel damage. The peak fuel temperature reached during the Insertionof Excess Reactivity accident in the worst (Hot) channel is only 227.4 0C -well below the newSafety Limit of 530 0C for the aluminide fuel.33 of 86  
: 6. NUREG-15 3 7, Section 13, provides guidance that the applicant should demonstrate that the facilitydesign features, safety limits, limiting safety system settings, and limiting conditions for operationhave been selected to ensure that no credible accident could lead to unacceptable radiologicalconsequences to people or the environment. The NRC staff review examined the analyses providedin 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 theLimiting Safety System Setting setpoint in TS 2.2 (12.S MW). The temperature feedbackcoefficient used is -7.0 x lO Ak/k rather than the TS S.3.a value of -6xlO-5 Ak/k. It is unclearwhat peaking factors are employed. SAR Figure 13.2 seems to indicate that the scram timeused is faster than the value in TS 3.2.c. The acceptability of the results is based uponwhether 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 safetyconclusions 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 10MW 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 andAcceptance Criteria, for the Insertion of Excess Reactivity accident, "The accident scenarioassumes 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 wasreanalyzed 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 SafetySystem Setting (LSSS) set point of 12.5 MW, since the rod run-in system will initiate a rod run-inat 11.5 MW (Technical Specification 3.2.f.1) and shutdown the reactor prior to reaching the LSSSscram set point of 125%.For the SAR analysis of the Insertion of Excess Reactivity accident, the temperature coefficientused was -6.0 x 10.5 Ak/k and not -7.0 x 10-5 Ak/k as stated above. Third paragraph on Page 13-17of the SAR lists the various reactivity coefficients assumed for the Insertion of Excess Reactivityaccident analysis.Details regarding the power peaking factors used were not provided in that section of the SAR. Thepower peaking values used were values obtained based on the destructive analysis of a MURR fuelelement. For the updated analysis, more up-to-date power peaking values, based on the detailedMCNP MURR core model, were used.For both the SAR analyses, as well as for the updated analysis presented here, the control bladeinsertion times are based on the current and relicensing Technical Specification 3.2.c requirementof insertion to the 20% withdrawn position in less than 0.7 seconds. So the insertion rate wascalculated 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 monthly31 of 86 control blade drop time verifications performed at MURR have always yielded insertion times of0.6 seconds or less (see response to RAI 1 .a).Similar to the SAR analysis, the Reactivity Transient Analysis program PARET (V7.5), maintainedand distributed by the Nuclear Engineering Division of Argonne National Laboratory (ANL) wasused. For the Insertion of Excess Reactivity accident analysis, two channels were modeled inPARET; a hot channel representing worst-case conditions inside the core and an average channelrepresenting the rest of the core experiencing "average" conditions. The axial power profiles usedfor this 2-channel PARET reactivity transient analysis are given in Table 1 below.Table 1 -Peaking Factors in the Hot and Average ChannelsHot Channel Average Channel2.046 1.0581.971 0.9202.145 1.0182.335 1.1322.497 1.2192.672 1.3072.835 1.3602.986 1.4113.105 1.4303.164 1.4373.169 1.4203.098 1.3832.953 1.3262.775 1.2432.542 1.1402.290 0.9892.069 0.8281.888 0.7011.703 0.6151.499 0.5301.277 0.4601.080 0.3860.904 0.3290.880 0.358As indicated earlier, the transient was started from an initial power level of 11.5 MW with corecoolant flow rate as well as core coolant inlet temperatures set at their LSSS values of 3,200 gpm32 of 86 and 155 'F, respectively. Also, pressurizer pressure was at 75 psia (LSSS value). Since theInsertion of Excess Reactivity transient was analyzed from a starting power level of 11.5 MW, therod run-in that would be initiated by the rod run-in system at 11.5 MW was bypassed and only thehigh power scram set point of 12.5 MW was modeled. Also, a delay of 150 milliseconds wasincorporated into the control blade scram model so that the control blades would only start to insert0.15 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. Asexpected, due to the higher starting core power level, much lower core coolant flow rate and muchhigher than normal core coolant inlet temperature conditions assumed for this updated analysis, thepeak power during the transient momentarily reaches approximately 37.4 MW compared to a valueof approximately 33.0 MW reported in the SAR analysis for the same 600 pcm step reactivityinsertion.40.0040035.0030.001o 20.00-POWER MW-'l'dad "C-I'f maxtCS350300l0S50* 03.000.1000.50 1.00 1.50 2.00Time (seconds)2.50Figure 1 -Reactor Power, Fuel and Cladding Temperatures vs. Timefor a Positive Reactivity Step Insertion of 0.006 Ak/kSeveral SPERT tests had shown that the reactor can withstand such short duration (fewmilliseconds) power burst without sustaining any fuel damage and only sustained operation at suchhigh power levels will lead to fuel damage. The peak fuel temperature reached during the Insertionof Excess Reactivity accident in the worst (Hot) channel is only 227.4 0C -well below the newSafety Limit of 530 0C for the aluminide fuel.33 of 86  
Line 34: Line 46:
: 7. NUREG-153 7, Section 13.1.1, "Maximum Hypothetical Accident," provides guidance for thelicensee 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 themelting of four number 1 fuel plates in a core region where the power is at a maximum. The fuelfails submerged and it is assumed that all iodine, krypton, and xenon isotopes are released into theprimary 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 oftwelve 10O-day cycles over a 300-day period. These values were then adjusted using a peakingfactor 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 islisted as 3.676. Clarify the discrepancies in the peaking factors used, and provide a revisedcalculation of the source using the peaking factors determined from the final analysis, orjustify why no additional information is needed.Table 4-14, "SUMMARY OF MUIRR HOT CHANNEL FACTORS," in Section 4.5 of the MUJRRSAR, lists a hot spot power peaking factor of 3.6765 with no engineering factors included. Thisvalue applies the product of the radial, axial and azimuthal peaking factors of a fuel plate todetermine the hot spot on the plate. The SAR provides an overall peaking factor of 4.35; the hotspot power peaking factor of 3.6765 multiplied by the engineering factors. These two peakingfactor values apply to the potential worst-case maximum power density point in the core for theSafety 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 FluxPower-related FactorsNuclear Peaking FactorsRadial 2.220Non-Uniform Burnup 1.112Local (Circumferential) 1.040Axial 1.432Overall 3.676Engineering Hot Channels Factors on FluxFuel Content Variation 1.030Fuel Thickness / Width Variation 1.150Overall Product 4.35By 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 factorto 3.4747; with no engineering factors included. Including the engineering factors, the overallpeaking factor increases to 4.116.36 of 86 From Table F.4, "SUMMARY OF MURR HOT CHANNEL FACTORS," of Appendix F ofAddendum 4 to the MURR Hazards Summary Report (as revised by Amendment No. 36):On Heat Flux From Plate-iPower-related FactorsNuclear Peaking FactorsFuel Plate (Hot Plate Average) 2.215Azimuthal Within Plate 1.070Axial Peak 1.3805Additional Allowable Factor 1.062Overall 3.4747Engineering Hot Channels Factors on FluxFuel Content Variation 1.030Fuel Thickness / Width Variation 1.150Overall Product: 4.116This 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 anoverall peaking factor of 3.863 and an enthalpy rise of 74.8%; thus producing the most limitingcombination of heat flux and enthalpy rise.This overall peaking factor of 4.116 at mesh interval 14 would apply to the 1-inch square assumedin the analysis in the response to RAT A.27. Therefore, since the ratio is 1.372 (4.116 / 3.0), thecalculated dose rates in the response to RAI A.27 increased by approximately the 37.2%. The onlyother fuel plate exposed during handling is plate number-24, which has a lower overall peakingfactor than plate number-i. The assumed peaking factor in the response to RAT A.27 has beenrevised from 3.0 to 4.116, which increased the whole body (TEDE) "60-Minute Dose fromRadioiodine and Noble Gases in Containment" from 0.79 to 1.09 mrem. This change also requireda similar revision to the response to RAI A.6 regarding the revised Technical Specificationdefinition for Irradiated Fuel, Definition 1.11, by about the same percentage.The current MIURR maximum hypothetical accident (MHA) assumes the melting of fuel platenumber-i in four (4) different fuel elements. An unirradiated fuel plate number-i contains, onaverage, 19.26 grams of U-235, so four (4) unirradiated number-i fuel plates contain 77.04 gramsinstead of the 78.58 grams assumed in the MHA. These four (4) number-i fuel plates that meltcorrespond to 1.41% of the total U-235 in the Week 58 Core that was used to determine the highpower peaking factor for the revised SLs. The Week 58 Core has a total power history of 576MWd. This power history results in a total reduced core mass of 5,474 grams of U-235 due to theprevious fuel consumption. This 1.41% of U-235 melting releases 3.42% of the core fissionproducts 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 thecontainment building are less than that for a failed fueled experiment, which is now considered theMURR MHA.b. The release is assumed to occur into the PCS with a volume of 2,000 gallons. Identif whatcomponents comprise this volume and provide information to confirm the 2,000 gallonvolume 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 ofits individual, major components, including the piping, reactor core, pressure vessels, primarycoolant circulation pumps, heat exchangers, and pressurizer. Attachment 7 is a breakdown of themeasurements and calculated volumes, where design capacities are not available, of the individualcomponents following the RELAP Model component designations listed in Appendix C of theSAR. The total calculated volume of the PCS is 2,007 gallons. However, based on the difficulty ofmeasuring some of the in-pool PCS piping and components, this volume is conservativelyunderestimated by approximately 5 to 10%, thus radionuclide concentrations in the PCS areconservative.c. The release is assumed to remain in the PCS except for the amount that will enter the poolcooling system as part of the PCS to pool cooling system leakage. Therefore, theconcentration of iodine that is released first enters the pool cooling system and is dilutedonce again. This seems to reduce the consequences of this accident to a fraction of theconsequences 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 notbe the MHA. Provide a confirmation of the dilution assumptions stated above andclarification 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 reanalyzedusing consistent methodologies and assumptions. All three (3) new analyses are included in theseresponses. 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 mremFuel Handling Accident: 687.00 mremFueled Experiment Failure: 1212.44 mremBased on these analyses, the Fueled Experiment Failure accident has been determined to be the newMURR 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 betweenthe PCS and the pool cooling system. However, the NRC staff questions whether the release38 of 86 into the PCS will collect in the vent tanks and other places in the PCS and eventually bereleased to the environment after decay. Provide an explanation for this leakage path,including assumptions and calculations of the possibility of the isotopic concentrations beingreleased 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 buildingisolation system will occur as a result of the gaseous activity collecting in the vent tank system. Atthis point the containment building is isolated. As described in Section 9.13.3 of the SAR, the venttanks will vent through an absolute and charcoal filter if enough gases collect in the vent tanks tocause the water level in the tanks to recede to a point where level controller 925A will signal valveV552A to open and vent the gases. Note: The vent path for the vent tank system, after it goesthrough the absolute and charcoal filters, is to the pool sweep system which is connected to thecontainment building 16-inch hot exhaust line (see SAR Sections 6.2.3.8 and 9.1.2.2). Thecontainment 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 wouldnot cause the system to vent. However, if for some reason the system should vent prior to the PCSbeing secured as part of the actions of Operations personnel during an MHA (vent valves 552A and552B will not open when the PCS is secured), the gases will be vented into the isolated containmentstructure 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-termrecovery actions, which will be very well planned and organized.e. In determining the offsite doses in the unrestricted areas from the releases, theconcentrations of the released isotopes are calculated using a method described in the MURRSAR, Appendix B, which used a simplified joint frequency distribution of weather data thatwas prepared in the 1960s. Given the changes in weather conditions over the last 50 years, itis not clear to the NRC staff whether the listed probabilities and wind speeds for the stabilityclasses are still applicable. Provide available current weather data, and state whetherchanges warrant reconsideration of the cited data, or justify why no additional information isneeded.In reviewing the available meteorological data for the Columbia vicinity, newer meteorological datawas found from the Columbia Regional Airport. This facility has more current meteorologicalwind data available and this data was used to generate wind roses for updated time periods closer tothe current time frame. Based on the results of the meteorological data review, we believe that theprevious data submitted is representative of current wind rose data, in and around the Columbiaarea, as there appeared to be no substantial difference in wind speed and direction during theoriginal submittal utilizing nine (9) years of data from 1961 to 1969 and subsequent data whichincluded the above 9-year period and an additional 21 years of meteorological data for a total of 30years (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 10provides the meteorological data for the years 1961 to 1990.39 of 86  
: 7. NUREG-153 7, Section 13.1.1, "Maximum Hypothetical Accident," provides guidance for thelicensee 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 themelting of four number 1 fuel plates in a core region where the power is at a maximum. The fuelfails submerged and it is assumed that all iodine, krypton, and xenon isotopes are released into theprimary 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 oftwelve 10O-day cycles over a 300-day period. These values were then adjusted using a peakingfactor 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 islisted as 3.676. Clarify the discrepancies in the peaking factors used, and provide a revisedcalculation of the source using the peaking factors determined from the final analysis, orjustify why no additional information is needed.Table 4-14, "SUMMARY OF MUIRR HOT CHANNEL FACTORS," in Section 4.5 of the MUJRRSAR, lists a hot spot power peaking factor of 3.6765 with no engineering factors included. Thisvalue applies the product of the radial, axial and azimuthal peaking factors of a fuel plate todetermine the hot spot on the plate. The SAR provides an overall peaking factor of 4.35; the hotspot power peaking factor of 3.6765 multiplied by the engineering factors. These two peakingfactor values apply to the potential worst-case maximum power density point in the core for theSafety 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 FluxPower-related FactorsNuclear Peaking FactorsRadial 2.220Non-Uniform Burnup 1.112Local (Circumferential) 1.040Axial 1.432Overall 3.676Engineering Hot Channels Factors on FluxFuel Content Variation 1.030Fuel Thickness / Width Variation 1.150Overall Product 4.35By 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 factorto 3.4747; with no engineering factors included. Including the engineering factors, the overallpeaking factor increases to 4.116.36 of 86 From Table F.4, "SUMMARY OF MURR HOT CHANNEL FACTORS," of Appendix F ofAddendum 4 to the MURR Hazards Summary Report (as revised by Amendment No. 36):On Heat Flux From Plate-iPower-related FactorsNuclear Peaking FactorsFuel Plate (Hot Plate Average) 2.215Azimuthal Within Plate 1.070Axial Peak 1.3805Additional Allowable Factor 1.062Overall 3.4747Engineering Hot Channels Factors on FluxFuel Content Variation 1.030Fuel Thickness / Width Variation 1.150Overall Product: 4.116This 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 anoverall peaking factor of 3.863 and an enthalpy rise of 74.8%; thus producing the most limitingcombination of heat flux and enthalpy rise.This overall peaking factor of 4.116 at mesh interval 14 would apply to the 1-inch square assumedin the analysis in the response to RAT A.27. Therefore, since the ratio is 1.372 (4.116 / 3.0), thecalculated dose rates in the response to RAI A.27 increased by approximately the 37.2%. The onlyother fuel plate exposed during handling is plate number-24, which has a lower overall peakingfactor than plate number-i. The assumed peaking factor in the response to RAT A.27 has beenrevised from 3.0 to 4.116, which increased the whole body (TEDE) "60-Minute Dose fromRadioiodine and Noble Gases in Containment" from 0.79 to 1.09 mrem. This change also requireda similar revision to the response to RAI A.6 regarding the revised Technical Specificationdefinition for Irradiated Fuel, Definition 1.11, by about the same percentage.The current MIURR maximum hypothetical accident (MHA) assumes the melting of fuel platenumber-i in four (4) different fuel elements. An unirradiated fuel plate number-i contains, onaverage, 19.26 grams of U-235, so four (4) unirradiated number-i fuel plates contain 77.04 gramsinstead of the 78.58 grams assumed in the MHA. These four (4) number-i fuel plates that meltcorrespond to 1.41% of the total U-235 in the Week 58 Core that was used to determine the highpower peaking factor for the revised SLs. The Week 58 Core has a total power history of 576MWd. This power history results in a total reduced core mass of 5,474 grams of U-235 due to theprevious fuel consumption. This 1.41% of U-235 melting releases 3.42% of the core fissionproducts 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 thecontainment building are less than that for a failed fueled experiment, which is now considered theMURR MHA.b. The release is assumed to occur into the PCS with a volume of 2,000 gallons. Identif whatcomponents comprise this volume and provide information to confirm the 2,000 gallonvolume 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 ofits individual, major components, including the piping, reactor core, pressure vessels, primarycoolant circulation pumps, heat exchangers, and pressurizer. Attachment 7 is a breakdown of themeasurements and calculated volumes, where design capacities are not available, of the individualcomponents following the RELAP Model component designations listed in Appendix C of theSAR. The total calculated volume of the PCS is 2,007 gallons. However, based on the difficulty ofmeasuring some of the in-pool PCS piping and components, this volume is conservativelyunderestimated by approximately 5 to 10%, thus radionuclide concentrations in the PCS areconservative.c. The release is assumed to remain in the PCS except for the amount that will enter the poolcooling system as part of the PCS to pool cooling system leakage. Therefore, theconcentration of iodine that is released first enters the pool cooling system and is dilutedonce again. This seems to reduce the consequences of this accident to a fraction of theconsequences 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 notbe the MHA. Provide a confirmation of the dilution assumptions stated above andclarification 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 reanalyzedusing consistent methodologies and assumptions. All three (3) new analyses are included in theseresponses. 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 mremFuel Handling Accident: 687.00 mremFueled Experiment Failure: 1212.44 mremBased on these analyses, the Fueled Experiment Failure accident has been determined to be the newMURR 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 betweenthe PCS and the pool cooling system. However, the NRC staff questions whether the release38 of 86 into the PCS will collect in the vent tanks and other places in the PCS and eventually bereleased to the environment after decay. Provide an explanation for this leakage path,including assumptions and calculations of the possibility of the isotopic concentrations beingreleased 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 buildingisolation system will occur as a result of the gaseous activity collecting in the vent tank system. Atthis point the containment building is isolated. As described in Section 9.13.3 of the SAR, the venttanks will vent through an absolute and charcoal filter if enough gases collect in the vent tanks tocause the water level in the tanks to recede to a point where level controller 925A will signal valveV552A to open and vent the gases. Note: The vent path for the vent tank system, after it goesthrough the absolute and charcoal filters, is to the pool sweep system which is connected to thecontainment building 16-inch hot exhaust line (see SAR Sections 6.2.3.8 and 9.1.2.2). Thecontainment 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 wouldnot cause the system to vent. However, if for some reason the system should vent prior to the PCSbeing secured as part of the actions of Operations personnel during an MHA (vent valves 552A and552B will not open when the PCS is secured), the gases will be vented into the isolated containmentstructure 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-termrecovery actions, which will be very well planned and organized.e. In determining the offsite doses in the unrestricted areas from the releases, theconcentrations of the released isotopes are calculated using a method described in the MURRSAR, Appendix B, which used a simplified joint frequency distribution of weather data thatwas prepared in the 1960s. Given the changes in weather conditions over the last 50 years, itis not clear to the NRC staff whether the listed probabilities and wind speeds for the stabilityclasses are still applicable. Provide available current weather data, and state whetherchanges warrant reconsideration of the cited data, or justify why no additional information isneeded.In reviewing the available meteorological data for the Columbia vicinity, newer meteorological datawas found from the Columbia Regional Airport. This facility has more current meteorologicalwind data available and this data was used to generate wind roses for updated time periods closer tothe current time frame. Based on the results of the meteorological data review, we believe that theprevious data submitted is representative of current wind rose data, in and around the Columbiaarea, as there appeared to be no substantial difference in wind speed and direction during theoriginal submittal utilizing nine (9) years of data from 1961 to 1969 and subsequent data whichincluded the above 9-year period and an additional 21 years of meteorological data for a total of 30years (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 10provides 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 listedconcentrations in the cited unrestricted location, which is also not specified. The calculationof the ratio of the average concentration in the unrestricted location to the correspondingconcentration in containment results in the reduction factor for iodine twice as large as thevalue for the noble gases. For example, for Krypton-85 the ratio is 7.5x10-14/3.0x10-s or areduction of about 4.0x10s For 1-131, the ratio is 1.36 x10-14/1.1x10-s or a reduction ofabout 8.1 x104. Provide an explanation of all assumptions relating to the calculation ofaverage isotope concentrations, specify~ all locations where these concentrations aredetermined, and explain how dispersion factors are determined and used, or justify why noadditional information is needed.In the case ofi1-131 as noted above the ratio ofi1-131 is 1.24 x 10-&deg;6. This was determined by takingthe initial concentration of I-131 in the containment building of 4.4 x 10-&deg;8 (Page 13.6 of theSAR) and multiplying it by 0.25 (plating reduction factor) and dividing into the final offsiteconcentration 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 x10-&deg;8) =1.24 x 10-&deg;6. The initial concentration of Kr-85 was used in this case as there is no platingor other phenomena that would hold up the noble gases.g. hn determining occupational doses, it appears that the MURR SAR calculations use acombination of dose conversion factors (DCFs). It appears that for radioiodine, thecalculation 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 submersiondoses 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 FGR11. 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 FGR11 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 thoseisotopes with a half-life of less than 2 hours. Overall, the difjferences in the calculated DCFsresult 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 uselimits from either: 10 CFR 20, Appendix B or 10 CFR 835, Appendix C (Attachment 11). Whereavailable we use Derived Air Concentration (DAC) and Effluent Concentrations from 10 CFR 20Appendix 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 (T112 < 2 hours) that we analyzed in the included accident analyses, ]VUIRR usedthe 10 CFR 835 Appendix C default DAC value of 6.0 x 10.06 as noted at the end ofAppendix C. From this default DAC we estimate the applicable effluent concentration limit basedon the description provided in the Table 2, "Effluent Concentrations," footnotes to Appendix B in10 CFR 20. Thus, all dose calculations now use limits based on the background and methodologyprovided in ICRP 26 and 30.40 of 86 Revised "Fuel Failure during Reactor Operation"(Formerly the MIIA)13.2.1 Fuel Failure during Reactor Operation13.2.1.1 Accident-Jnitiating Events and ScenariosMany types of accidents have been considered in conjunction with the operation of the MUJRR. Inall cases, safety systems have been designed such that the likelihood of an accident involving therelease of a significant amount of fission products has essentially been eliminated. The safetysystems take the form of automatic reactor shutdown circuits and process systems designed toensure, through redundancy, that the reactor will shut down upon a significant deviation fromnormal operating conditions. In addition, the reactor is housed within a containment building, thusproviding 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 anaccident condition has caused the melting of the number-i fuel plate in four (4) separate fuelelements (Ref. 13.11). It is further assumed that the four (4) number- 1 fuel plates are in the peakpower 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, andthe pre-operational inspection of the reactor pressure vessels and core region following any fuelhandling evolution, all prevent an accident of this type from occurring. In addition, it has beenshown that a 75% blockage of coolant flow to the hot channel is insufficient to cause claddingfailure (Ref. 13.2).13.2.1.2 Accident Analysis and ConsequencesThe fuel failure accident postulates partial fuel melting with an associated release of fissionproducts into the primary coolant system. The accident is assumed to occur with the primarycoolant system operating, resulting in a quick dispersal of the fission products throughout thesystem. With the design of the primary coolant system and its associated systems, particulateactivity will remain in the coolant, and the gaseous activity that comes out of solution will collect inthe reactor loop vent system and be retained there. Therefore, the primary coolant system reliefvalves and pressurizer are the only paths for a release of significant quantities of fission products tothe environment.The potential energy release from the melting of four (4) number-i fuel plates could occur as apossible metal-water reaction (Ref. 13.3). While hydrogen would be formed, it is highly unlikelythat in a water environment a hydrogen deflagration reaction would occur. The amount of materialwhich would be involved in a metal-water reaction under the conditions of four (4) number-i fuelplates melting is not predictable as the amount is dependent upon many conditions. For purposes ofcalculation, it is conservatively assumed that all the fuel plate aluminum cladding exposed in the41 of 86 area is involved in the reaction. The reactor core contains a total of 33.56 Kg of aluminum. Ofthis, 1.3% or 436 grams is assumed to react according to the following equation:A1 + n2-I* A1On +nI-I2 +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 103 BTU.This amount of heat would easily be transferred to the adjacent fuel elements and primary coolantin the core. Additionally, any steam that would form in the vicinity of the molten area would alsoassist in dissipating the heat. Since the fuel failure would result in a negligible release of energy tothe primary coolant system, the introduction of pressure surges, which could lift the primary reliefvalves, are not considered credible. The pressurizer is an isolated system, and since no significantpressure 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 reactorscram and actuation of the containment building isolation system by action of the pool surfaceradiation monitor. Additionally, following actuation of the anti-siphon system when the primarycoolant system is secured, gases could also collect in the anti-siphon pressure tank. The location ofthese tanks under the pool surface, and the shielding provided by the water and the biologicalshield, 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 coolantcleanup system. This cleanup procedure would be undertaken under closely monitored andcontrolled conditions.The primary coolant system does experience some coolant leakage into the. reactor pool through thepressure vessel head packing and flange gasket. This leakage is typically less than 40 gallons (1511) per week; an almost imperceptible leakage rate of approximately 4 x 1 0- gallons of primarycoolant 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 exposureto 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 6.2Kg of 235U (normal operating cycle is 6.5 days with a total of less than 700 MVWD on the core), thefollowing 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 Core1311 -- 1.7 x 10+0 Ci 85Kr -4.7 x 10+02 Ci '3Xe -4.2 x 10+&deg;5 Ci132I -3.3 x 10+05 Ci 85m~r -1.1 x 10+05 Ci '35Xe -9.6 x 10+04 Ci133I -- 5.1 X 10o0 Ci 87Kr -2.1 x 10+05 Ci 135mXe -9.4 x 10+04 Ci14-- 6.3 x 10+&deg; Ci 88Kr -3.0 x 10+05 Ci 137Xe -- 4.9 x 10+0 Ci13SI -5.2 x 10+05 Ci 89Kr -3.8 x 10+&deg; Ci '38Xe -5.2 x 10+05 Ci9&deg;Kr -3.8 x 10+05 Ci 139Xe -- 4.2 x 10"&deg;5 CiAn 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 thefuel 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 revisedSLs. The Week 58 Core has a total power history of 576 MWd. This power history results in atotal 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 fuelplate 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 thefuel is assumed in calculating the fission product inventory in the primary coolant system. It is alsoassumed that fission products released into the primary coolant are quickly and uniformly dispersedwithin the 2,000-gallon (7,571-1) primary coolant system volume and, during a normal week'soperation, 80 gallons (7.9 x 1 0- gpm) of coolant leaks from the primary coolant system into thepool water. Therefore, the radioactivity released into the reactor pool in 10 minutes -determinedto be the maximum personnel occupancy time in the containment building after the accident fornecessary operational personnel -is as follows:(Note: It would take approximately 5 minutes for Operations personnel to secure the primarycoolant system and verify that the containment building has been evacuated following acontainment building isolation. For the purpose of the fuel failure calculations, a conservativeassumption of 10 minutes is used.)Example calculation of 131I released into the reactor pool:=131I in fuel x 0.0342 x 1/2,000 gal x (7.9 xl0"&deg;3 gpm) x 10 min x 10+06 jtCi/Ci= (1.7 x i0+&deg; Ci) x (1 .3509 x 10+&deg; = 2.30 x 10+05 1iCiNote: 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 Minutes3I-- 2.30 x 10+05 1.tCi 85Kr -6.35 x 10+02 1iCi 133Xe -5.67 x 10+05 13I -4.46 x 10+/-05 85m~r -1.49 x 10+&deg;05 .Ci 135~ -1.30 x 10+05 1331 -- 6.89 x 10+05 87Kr -2.84 x 10+05 1iCi l35rage -1.27 x 10+0 ptCi1341 -- 8.52 x 10+05 j.tCi 88Kr -4.04 x 10+05 gCi 137Xe -6.63 x 10+&deg; 13I- 7.02 x 10+05 jtCi 89Kr -5.13 x 10+05 xiCi 138Xe -7.02 x 10+05 9&deg;Kr -5.13 x 10+05 gCi 139Xe -5.67 x 10+05 1iCiFission products released into the reactor pooi will be detected by the pool surface and ventilationsystem exhaust plenum radiation monitors. However, for the purposes of this analysis, it isassumed that a reactor scram and actuation of the containment building isolation system occurs byaction of the pooi surface radiation monitor.The radioiodine released into the reactor pool over a 10-minute interval is conservatively assumedto be instantly and uniformly mixed into the 20,000 gallons (75,708 1) of bulk pool water, whichthen results in the following pool water concentrations for the radioiodine isotopes. The watersolubility of the krypton and xenon noble gases released into the pool over this same time periodare ignored and they are assumed to pass immediately through the pool water and evolve directlyinto the containment building air volume where they instantaneously form a uniform concentrationin the isolated structure.Radioiodine Concentrations in the Pool Water131 -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/ga1When the reactor is at 10 MW and the containment building ventilation system is in operation, theevaporation rate from the reactor pool is approximately 80 gallons (302.8 L) of water per day. Forthe purposes of this calculation, it is assumed that a total of 40 gallons (151 L) of pool watercontaining the previously listed radioiodine concentrations evaporates into the containment buildingover the 10 minute period. Containment air with a temperature of 75 0F (23.9 &deg;C) and 100%relative humidity contains H20 vapor equal to 40 gallons (151.4 L) of water. Since the air incontainment 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 causethe containment air to be supersaturated. It is also conservatively assumed that all of theradioiodine activity in the 40 gallons (151 L) of pool water instantaneously forms a uniformconcentration in the containment building air. When distributed into the containment building, thiswould result in the following radioiodine concentrations in the 225,000 ft3 (6,371.3 in3) air volume:Example calculation of 1311 released into containment air:-131I concentration in pool water x 40 gal x 1/225,000 ft3 x 35.3 147 ft3/m3= 1.15 x 10+01 pCi/gal x (6.28 x 10-03 gal/mn3)44 of 86  
: f. It is not clear to the NRC staff which dispersion factors were used to arrive at the listedconcentrations in the cited unrestricted location, which is also not specified. The calculationof the ratio of the average concentration in the unrestricted location to the correspondingconcentration in containment results in the reduction factor for iodine twice as large as thevalue for the noble gases. For example, for Krypton-85 the ratio is 7.5x10-14/3.0x10-s or areduction of about 4.0x10s For 1-131, the ratio is 1.36 x10-14/1.1x10-s or a reduction ofabout 8.1 x104. Provide an explanation of all assumptions relating to the calculation ofaverage isotope concentrations, specify~ all locations where these concentrations aredetermined, and explain how dispersion factors are determined and used, or justify why noadditional information is needed.In the case ofi1-131 as noted above the ratio ofi1-131 is 1.24 x 10-&deg;6. This was determined by takingthe initial concentration of I-131 in the containment building of 4.4 x 10-&deg;8 (Page 13.6 of theSAR) and multiplying it by 0.25 (plating reduction factor) and dividing into the final offsiteconcentration 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 x10-&deg;8) =1.24 x 10-&deg;6. The initial concentration of Kr-85 was used in this case as there is no platingor other phenomena that would hold up the noble gases.g. hn determining occupational doses, it appears that the MURR SAR calculations use acombination of dose conversion factors (DCFs). It appears that for radioiodine, thecalculation 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 submersiondoses 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 FGR11. 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 FGR11 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 thoseisotopes with a half-life of less than 2 hours. Overall, the difjferences in the calculated DCFsresult 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 uselimits from either: 10 CFR 20, Appendix B or 10 CFR 835, Appendix C (Attachment 11). Whereavailable we use Derived Air Concentration (DAC) and Effluent Concentrations from 10 CFR 20Appendix 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 (T112 < 2 hours) that we analyzed in the included accident analyses, ]VUIRR usedthe 10 CFR 835 Appendix C default DAC value of 6.0 x 10.06 as noted at the end ofAppendix C. From this default DAC we estimate the applicable effluent concentration limit basedon the description provided in the Table 2, "Effluent Concentrations," footnotes to Appendix B in10 CFR 20. Thus, all dose calculations now use limits based on the background and methodologyprovided in ICRP 26 and 30.40 of 86 Revised "Fuel Failure during Reactor Operation"(Formerly the MIIA)13.2.1 Fuel Failure during Reactor Operation13.2.1.1 Accident-Jnitiating Events and ScenariosMany types of accidents have been considered in conjunction with the operation of the MUJRR. Inall cases, safety systems have been designed such that the likelihood of an accident involving therelease of a significant amount of fission products has essentially been eliminated. The safetysystems take the form of automatic reactor shutdown circuits and process systems designed toensure, through redundancy, that the reactor will shut down upon a significant deviation fromnormal operating conditions. In addition, the reactor is housed within a containment building, thusproviding 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 anaccident condition has caused the melting of the number-i fuel plate in four (4) separate fuelelements (Ref. 13.11). It is further assumed that the four (4) number- 1 fuel plates are in the peakpower 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, andthe pre-operational inspection of the reactor pressure vessels and core region following any fuelhandling evolution, all prevent an accident of this type from occurring. In addition, it has beenshown that a 75% blockage of coolant flow to the hot channel is insufficient to cause claddingfailure (Ref. 13.2).13.2.1.2 Accident Analysis and ConsequencesThe fuel failure accident postulates partial fuel melting with an associated release of fissionproducts into the primary coolant system. The accident is assumed to occur with the primarycoolant system operating, resulting in a quick dispersal of the fission products throughout thesystem. With the design of the primary coolant system and its associated systems, particulateactivity will remain in the coolant, and the gaseous activity that comes out of solution will collect inthe reactor loop vent system and be retained there. Therefore, the primary coolant system reliefvalves and pressurizer are the only paths for a release of significant quantities of fission products tothe environment.The potential energy release from the melting of four (4) number-i fuel plates could occur as apossible metal-water reaction (Ref. 13.3). While hydrogen would be formed, it is highly unlikelythat in a water environment a hydrogen deflagration reaction would occur. The amount of materialwhich would be involved in a metal-water reaction under the conditions of four (4) number-i fuelplates melting is not predictable as the amount is dependent upon many conditions. For purposes ofcalculation, it is conservatively assumed that all the fuel plate aluminum cladding exposed in the41 of 86 area is involved in the reaction. The reactor core contains a total of 33.56 Kg of aluminum. Ofthis, 1.3% or 436 grams is assumed to react according to the following equation:A1 + n2-I* A1On +nI-I2 +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 103 BTU.This amount of heat would easily be transferred to the adjacent fuel elements and primary coolantin the core. Additionally, any steam that would form in the vicinity of the molten area would alsoassist in dissipating the heat. Since the fuel failure would result in a negligible release of energy tothe primary coolant system, the introduction of pressure surges, which could lift the primary reliefvalves, are not considered credible. The pressurizer is an isolated system, and since no significantpressure 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 reactorscram and actuation of the containment building isolation system by action of the pool surfaceradiation monitor. Additionally, following actuation of the anti-siphon system when the primarycoolant system is secured, gases could also collect in the anti-siphon pressure tank. The location ofthese tanks under the pool surface, and the shielding provided by the water and the biologicalshield, 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 coolantcleanup system. This cleanup procedure would be undertaken under closely monitored andcontrolled conditions.The primary coolant system does experience some coolant leakage into the. reactor pool through thepressure vessel head packing and flange gasket. This leakage is typically less than 40 gallons (1511) per week; an almost imperceptible leakage rate of approximately 4 x 1 0- gallons of primarycoolant 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 exposureto 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 6.2Kg of 235U (normal operating cycle is 6.5 days with a total of less than 700 MVWD on the core), thefollowing 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 Core1311 -- 1.7 x 10+0 Ci 85Kr -4.7 x 10+02 Ci '3Xe -4.2 x 10+&deg;5 Ci132I -3.3 x 10+05 Ci 85m~r -1.1 x 10+05 Ci '35Xe -9.6 x 10+04 Ci133I -- 5.1 X 10o0 Ci 87Kr -2.1 x 10+05 Ci 135mXe -9.4 x 10+04 Ci14-- 6.3 x 10+&deg; Ci 88Kr -3.0 x 10+05 Ci 137Xe -- 4.9 x 10+0 Ci13SI -5.2 x 10+05 Ci 89Kr -3.8 x 10+&deg; Ci '38Xe -5.2 x 10+05 Ci9&deg;Kr -3.8 x 10+05 Ci 139Xe -- 4.2 x 10"&deg;5 CiAn 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 thefuel 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 revisedSLs. The Week 58 Core has a total power history of 576 MWd. This power history results in atotal 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 fuelplate 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 thefuel is assumed in calculating the fission product inventory in the primary coolant system. It is alsoassumed that fission products released into the primary coolant are quickly and uniformly dispersedwithin the 2,000-gallon (7,571-1) primary coolant system volume and, during a normal week'soperation, 80 gallons (7.9 x 1 0- gpm) of coolant leaks from the primary coolant system into thepool water. Therefore, the radioactivity released into the reactor pool in 10 minutes -determinedto be the maximum personnel occupancy time in the containment building after the accident fornecessary operational personnel -is as follows:(Note: It would take approximately 5 minutes for Operations personnel to secure the primarycoolant system and verify that the containment building has been evacuated following acontainment building isolation. For the purpose of the fuel failure calculations, a conservativeassumption of 10 minutes is used.)Example calculation of 131I released into the reactor pool:=131I in fuel x 0.0342 x 1/2,000 gal x (7.9 xl0"&deg;3 gpm) x 10 min x 10+06 jtCi/Ci= (1.7 x i0+&deg; Ci) x (1 .3509 x 10+&deg; = 2.30 x 10+05 1iCiNote: 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 Minutes3I-- 2.30 x 10+05 1.tCi 85Kr -6.35 x 10+02 1iCi 133Xe -5.67 x 10+05 13I -4.46 x 10+/-05 85m~r -1.49 x 10+&deg;05 .Ci 135~ -1.30 x 10+05 1331 -- 6.89 x 10+05 87Kr -2.84 x 10+05 1iCi l35rage -1.27 x 10+0 ptCi1341 -- 8.52 x 10+05 j.tCi 88Kr -4.04 x 10+05 gCi 137Xe -6.63 x 10+&deg; 13I- 7.02 x 10+05 jtCi 89Kr -5.13 x 10+05 xiCi 138Xe -7.02 x 10+05 9&deg;Kr -5.13 x 10+05 gCi 139Xe -5.67 x 10+05 1iCiFission products released into the reactor pooi will be detected by the pool surface and ventilationsystem exhaust plenum radiation monitors. However, for the purposes of this analysis, it isassumed that a reactor scram and actuation of the containment building isolation system occurs byaction of the pooi surface radiation monitor.The radioiodine released into the reactor pool over a 10-minute interval is conservatively assumedto be instantly and uniformly mixed into the 20,000 gallons (75,708 1) of bulk pool water, whichthen results in the following pool water concentrations for the radioiodine isotopes. The watersolubility of the krypton and xenon noble gases released into the pool over this same time periodare ignored and they are assumed to pass immediately through the pool water and evolve directlyinto the containment building air volume where they instantaneously form a uniform concentrationin the isolated structure.Radioiodine Concentrations in the Pool Water131 -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/ga1When the reactor is at 10 MW and the containment building ventilation system is in operation, theevaporation rate from the reactor pool is approximately 80 gallons (302.8 L) of water per day. Forthe purposes of this calculation, it is assumed that a total of 40 gallons (151 L) of pool watercontaining the previously listed radioiodine concentrations evaporates into the containment buildingover the 10 minute period. Containment air with a temperature of 75 0F (23.9 &deg;C) and 100%relative humidity contains H20 vapor equal to 40 gallons (151.4 L) of water. Since the air incontainment 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 causethe containment air to be supersaturated. It is also conservatively assumed that all of theradioiodine activity in the 40 gallons (151 L) of pool water instantaneously forms a uniformconcentration in the containment building air. When distributed into the containment building, thiswould result in the following radioiodine concentrations in the 225,000 ft3 (6,371.3 in3) air volume:Example calculation of 1311 released into containment air:-131I concentration in pool water x 40 gal x 1/225,000 ft3 x 35.3 147 ft3/m3= 1.15 x 10+01 pCi/gal x (6.28 x 10-03 gal/mn3)44 of 86  
-7.22 x 10-o2 ptCi/mn3(7.22 x 10-&deg; /.tCi/m3) x (1 m3/106 ml) =7.22 x 10.08 pCi/mlNote: Same calculation is used for the other isotopes listed below.The average radioiodine concentrations are the sum of the initial concentrations and theconcentrations after 10 minutes decay divided by 2.Average Radioiodine Concentrations in the Containment Building Air during the 10 Minutes1311 -- 7.22 x 10-&deg;8 pCi/ml132I -- 1.36 x 10-07 pCi/ml133I -- 2.16 x 10-07 gCi/ml134I -- 2.53 x 10-07 pxCi/ml135I -2.18 x 10-&deg; pxCi/mlAs noted previously, the krypton and xenon noble gases released into the reactor pool from theprimary coolant system during the assumed 1 0-minute interval following the fuel failure (Note: theprimary coolant system is shut down and secured, and the leakage driving force is stopped within10 minutes), are assumed to pass immediately through the pool water and enter the containmentbuilding air volume where they instantaneously form a uniform concentration in the isolatedstructure. Based on the 225,000-ft3 volume of containment building air and the previously listedCurie quantities of these gases released into the reactor pool, the maximum noble gasconcentrations 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 ft3 x 35.3 147 ft3/m3-6.35 x 10+02 uCi x (1.60 x 10-o4 1/in3)-9.96 x 10.02 gxCi/m3(9.96 x 10.02 3) x (1 m3/106 ml) = 9.96 x i0.0 ptCi/mlNote: Same calculation is used for the other isotopes listed below.The average noble gas concentrations are the sum of the initial concentrations and theconcentrations after 10 minutes decay divided by 2.Average Noble Gas Concentrations in the Containment Building Air during the 10 MinutesKr -9.96 x 10-&deg;8 jiCi/ml85m~r -2.30 x 10-&deg;5 iCi/ml87Kr -4.27 x 10&deg;5 pCi/ml88r- 6.22 x 10&deg;5 ptCi/ml89r- 4.47 x 10"&deg;5 p.Ci/ml9&deg;r- 4.03 x 10"&deg;5 pCi/ml133Xe -l3SlXe _'38Xe -'39Xe -8.90 x 10-&deg; pCi/ml2.02 x 10-&deg; gtCi/ml1.63 x 10-&deg;5 gCi/ml6.05 x 10.05 jgCi/ml8.88 x 10.05 pCi/ml4.45 x 10.05 pCi/ml45 of 86 The objective of this calculation is to present a worst-case dose assessment for an individual whoremains in the containment building for 10 minutes following fuel failure. Therefore, as notedpreviously, the radioactivity in the evaporated pool water is assumed to be instantaneously anduniformly 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 thyroidfrom radioiodine and the dose to the whole body resulting from submersion in the airborne noblegases and radioiodine inside the containment building. As previously noted, the exposure time forthis dose assessment is 10 minutes.Because the airborne radioiodine source is composed of five (5) different iodine isotopes, it will benecessary to determine the dose contribution from each individual isotope and to then sum theresults. 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 25mrem/DAC-hr. Additionally, 10 minutes of one DAC-br is 1.67 x 10-01 DAC-br.1311 concentration in containment1311 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)= 3.61Therefore, 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-&deg;1 DAC-hr)= 1.51x 10+&deg;lmremNote: Same calculation is used for the other radioiodines listed below.Derived Air Concentration Values and 10-Minute Exposures -RadioiodineRadionuclide13111321133I13411351Derived Air Concentration2.00 x 10-&deg;8 1Ci/ml3.00 x 10-06 gtCi/ml1.00 x 1 007 ptCi/ml2.00 x 10-0 jiCi/ml7.00 x 10.o iiCi/ml10-Minute Exposure1.51 x 10401 mrem1.89 x 10-&deg;' mrem8.99 x 10+&deg;&deg; mrem5.27 x 10.02 mnrem1.30 x 10+&deg; mremTotal =25.58 mrem46 of 86 Doses from the kryptons and xenons present in the containment building are assessed in much thesame manner as the radioiodines, and the dose contribution from each individual radionuclide mustbe calculated and then added together to arrive at the final noble gas dose. Because the dose fromthe noble gases is only an external dose due to submersion, and because the DACs for theseradionuclides are based on this type of exposure, the individual noble gas doses for 10 minutes incontainment were based on their average concentration in the containment air and thecorresponding DAC.Example calculation of whole body dose due to KrThe DAC can also be defined as 5,000 mrenm/2,000 hrs, or 2.5 mnremiDAC-hr. Additionally, 10minutes of one DAC-hr is 1.67 x 1001 DAC-hr.85Kr concentration in containment = 9.96 x 1 0.0 85Kr DAC (10 CFR 20) = 1.00 x i0"&deg; pCi/mlDose Multiplication Factor = (85Kr concentration) / (SSKr DAC)= (9.96 x 10.08 pCi/mI) / (1.00 x 10o4~ pCi/mi)= 0.001Therefore, 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"&deg;1 DAC-hr)=4.15 xl10&deg;4mremNote: 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 GasesRadionuclide85Kr85m~r87Kr88Kr89Kr90Kr133Xe135Xel35rage137Xe138Xe139XeDerived Air Concentration1.00 x 10"&deg; xtCi/ml2.00 x 1005 iiCi/ml5.00 x 10-&deg;6 itCi/ml2.00 x 100o6 p.Ci/ml6.00 x 100o6 gCi/ml6.00 x 10.o6 iCi/ml1.00 x 10"04 pCi/ml1.00 x 10-05 jCi/ml9.00 x 10-06 xiCi/ml6.00 x 10.06 gCi/ml4.00 X 10-06 pCi/mi6.00 x 10&deg;6 jiCi/ml10-Minute Exposure4.51 x 10-&deg;4 mrem4.80 x 10-&deg;" mrem3.56 x 10400 mrem1.30 x 10+01 mrem3.11 x 10400 mrem2.80 x 10+00&deg;mrem3.71 x 10-&deg;1 mrem8.43 x 10-&deg;1 mrem7.54 x l0-&deg;' mrem4.20 x 10+&deg; mrem9.25 x 1040&deg; mrem3.09 x 10+00 mremTotal = 41.42 mrenlTo finalize the occupational dose in terms of Total Effective Dose Equivalent (TEDE) for a1 0-minute exposure mn the containment building after target failure, the doses from the radioiodinesand noble gases must be added together, and result in the following values:10-Minute Dose from Radioidines and Noble Gases in the Containment BuildingCommitted Dose Equivalent (Thyroid):Committed Effective Dose Equivalent (Thyroid):Committed Effective Dose Equivalent (Noble Gases):Total Effective Dose Equivalent (Whole Body):25.58 mrem0.77 mrem41.42 mrem42.18 mremBy comparison of the maximum TEDE and Committed Dose Equivalent (CDE) for thoseoccupationally-exposed during fuel failure to applicable NRC dose limits in 10 CFR 20, the finalvalues are shown to be well within the published regulatory limits and, in fact, lower than 1% ofany occupational limit.Radiation shine through the containment structure was also evaluated when considering accidentconditions and dose consequences to the public and MUJRR staff. Calculation of exposure ratefrom the target failure was performed using the computer program MicroShield 8.02 with aRectangular Volume -External Dose Point geometry for the representation of the containmentstructure (Attachment 12). MicroShield 8.02 is a product of Grove Software and is acomprehensive photon/gamma ray shielding and dose assessment program that is widely used byindustry for designing radiation shields.The exposure rate values provided below represents the radiation fields at 1 foot (30.5 cm) from a12-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 theexposure rate values are identical to those used for determining the dose to a worker withincontainment from noble gases. For radioiodine, the total iodine activity of the target was used forthe dose calculations, not the amount that evaporated in 10 minutes. The source term also assumesa homogenous mixture of nuclides within the containment free volume.Radiation Shine through the Containment BuildingExpo sure Rate at 1-Foot from Containment Building Wall: 1.074 mrem/hrExposure Rate at Emergency Planning Zone Boundary (150 meters): 0.007 mrem/hrA confirmatory analysis of the accident condition yielding the largest consequence was validatedindependently by the use of the MCNIP code. This analysis yielded a result 21% less than theMicroshield method and results provided above.As noted earlier in this analysis, the containment building ventilation system will shut down and thebuilding itself will be isolated from the surrounding areas. Fuel failure will not cause an increase inpressure inside the reactor containment structure; therefore, any air leakage from the building willoccur as a result of normal changes in atmospheric pressure and pressure equilibrium between theinside of the containment structure and the outside atmosphere. It is highly probable that there willbe no pressure differential between the inside of the containment building and the outsideatmosphere, and consequently there will be no air leakage from the building and no radiation doseto members of the public in the unrestricted area. However, to develop what would clearly be aworst-case scenario, this analysis assumes that a barometric pressure change had occurred inconjunction with the target failure. A reasonable assumption would be a pressure change on theorder of 0.7 inches of Hig (25.4 mm of Hig at 60&deg;C), which would then create a pressure differentialof about 0.33 psig (2.28 kPa above atmosphere) between the inside of the isolated containmentbuilding and the inside of the adjacent laboratory building, which surrounds most of thecontainment structure. Making the conservative assumption that the containment building will leakat the TS leakage rate limit [10% of the contained volume over a 24-hour period from an initialoverpressure of 2.0 psig (13.8 kPa above atmosphere)], the air leakage from the containmentstructure in standard cubic feet per minute (scfmn) as a function of containment pressure can beexpressed by the following equation:LR = 17.85 x (CP-14.7)112;where:LR = leakage rate from containment (scfmn); andCP = containment pressure (psia).The minimum Technical Specification free volume of the containment building is 225,000-ft3 atstandard temperature and pressure. At an initial overpressure of 2.0 psig (13.8 kPa aboveatmosphere), 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 air49 of 86 volume to 230,051 scf. The above equation describes the leakage rate that results in this drop ofcontained air volume over 1,440 minutes (24 hours).When applying the Technical Specification leakage rate equation to the assumed initialoverpressure condition of 0.33 psig (2.28 kPa above atmosphere), it would take approximately 16.5hours 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 periodwould be about 5.2 scfln.Several factors exist that will mitigate the radiological impact of any air leakage from thecontainment building following target failure. First of all, most leakage pathways fromcontainment discharge into the reactor laboratory building, which surrounds the containmentstructure. Since the laboratory building ventilation system continues to operate during targetfailure, leakage air captured by the ventilation exhaust system is mixed with other building air, andthen 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 bydischarge out the exhaust stack and subsequent atmospheric dispersion, results in extremely lowradionuclide concentrations and very small radiation doses in the unrestricted area. A tabulation ofthese concentrations and doses is given below. These values were calculated following the samemethodology 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 tothe behavior of radioiodine, which has been studied extensively in the containment mockup facilityat 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% iodinedeposition in the containment building, each cubic meter of air released from containment has aradioiodine concentration that is 25% of each cubic meter within containment building air, then theradioiodine concentrations leaking from the containment structure into the laboratory building, inmicrocuries per milliliter, will be:Example calculation of 1311 released through the exhaust stack:-131I activity / (30,500 ft3/min x 16.5 hr x 60 min/hr x 28,300 ml/ft3)= 2.30 x 10+05 /8.55 x 10+'11ml-2.69 X 10-07 iiCi/ml(2.69 x 10-07 iiCi/ml) x (0.25) = 6.73 x 10.08 iiCi/mlNote: Same calculation is used for the other radioiodines listed below.Radioiodine Concentrations in Air Leaking from Containment1311 -- 6.73 x 10-&deg;8 jtCi/ml 133I -- 2.02 x 10-07 gtCi/ml 13I- 2.05 x 10-07 jiCi/ml1321 -1.30 x 10-07 pCi/ml 34 -2.49 x 10-07 pCi/ml50 of 86 Example calculation of 85Kr released through the exhaust stack:= 85}~. activity / (30,500 ft3/min x 16.5 hr x 60 mmn/hr x 28,300 mi/fl3)= 6.35 x 10+02 / 8.55 x 10+11 ml= 7.4 x1-w&deg; ItCi/mlNote: Same calculation is used for the other noble gases listed below.Noble Gas Concentrations in Air Leaking from Containment and Exiting the Exhaust Stack85Kr -7.43 x 10-1 pCi/ml 87Kr -3.32 x 10-0 jtCi/ml 89Kr -6.00 x 10-0 giCi/ml85m~r -1.74 x 10-0 g.Ci/ml 88Kr -4.73 x 10.0 pxCi/ml 9&deg;Kr -6.00 x 10-0 jiCi/ml33e- 6.64 x 10-07 pCi/mil 3smXe -- 1.49 x 10-07 ptCi/ml 138Xe -8.22 x 10-07 pCi/ml135Xe -1.52 x 10-0 pCi/ml 137Xe -7.76 x 10-07 pCi/mI '39Xe -6.64 x 10-0 ptCi/mlAssuming, as stated earlier, that (1) the average leakage rate from the containment building is 5.2scfrn, (2) the leak continues for about 16.5 hours in order to equalize the containment buildingpressure with atmospheric pressure, (3) the flow rate through the facility's ventilation exhaust stackis 30,500 scfrn, (4) the reduction in concentration from the point of discharge at the exhauststack to the point of maximum concentration in the unrestricted area is a factor of 292 and (5) thereis no decay of any radioiodines or noble gases, then the following concentrations of radioiodinesand noble gases with their corresponding radiation doses will occur in the unrestricted area. Thevalues listed are for the point of maximum concentration in the unrestricted area assuming auniform, semi-spherical cloud geometry for noble gas submersion and further assuming that themost conservative (worst-case) meteorological conditions exist for the entire 16.5-hour period ofcontainment 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 thatwas used to determine doses inside the containment building, and it was assumed that an individualwas present at the point of maximum concentration for the full 16.5 hours that the containmentbuilding was leaking.A worst-case scenario dilution factor of 292 for effluent dilution using the Pasquill-Guifford Modelfor atmospheric dilution is used in this analysis. We assume that all offsite (public) dose occursunder these atmospheric conditions at the site of interest, i.e. 760 meters North of MUIRR. In ourcase at 760 meters it occurs only during Stability Class F conditions; which normally only occur11.4% 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. AnEffluent Concentration Limit of 2.0 x 10.08 itCi/ml is used for the "unlisted" isotopes, which equalsthe DAC/300 when using the DOE Part 835 Default DAC limits of 6.0 x 10-06 p.C/ml. Exposure at1 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 8760hours per year. Thus, the time assumed to be exposed to the effluent concentration limit is a factorof 4.38 longer than the 2000 hours per year that defines a DAC. The isotopes in question are based51 of 86 on a default DAC limit of 6.0 x 1 0-06 for short-lived (< 2 hour half-lives) submersion DAC 'sin Appendix C of Part 835. No credit is taken for transit time from the stack to the receptor pointnor is credit taken for decay inside containment until release. In the case of Kr-89 and Xe-i137 thetransit 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 theDAC value to establish the effluent concentration limit is warranted. This reduction factor of 300 isconsistent, in fact more conservative, than 10 CFR 20 Appendix B, as the N4RC calculates EffluentConcentration 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 mremihr1311 concentration131i effluent concentration limit1311 Conversion Factor= 2.30 x 10-1&deg; pCi/ml=2.00 x 10-10 gCi/ml= 5.71 X 10-&deg; mremihrTherefore, 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&deg; giCi/ml / (2.00 x i010-1 gCi/ml x 5.71 x 10-03 mremihr x 16.5 brs)= 1.09 x10&deg;01mremNote: Same calculation is used for the other isotopes (radioiodines and noble gases) listed below.Effluent Concentration Limits. Concentrations at Point of Maximum Concentrationand Radiation Doses in the Unrestricted Area -RadioiodineRadionuclide1311132I133I134I1351Effluent Limit2.00 x 10"1&deg; giCi/ml6.00 x 10.08 iiCi/ml6.00 x 10-09 Maximum Concentration12.30 x 10-10 giCi/ml4.47 x 10.1&deg; pCi/ml6.90 x 10-10 iiCi/ml8.54 x 10-l0 jiCi/ml7.03 x 10-1&deg; jCi/mlRadiation Dose1.09 x 10-01 mrem2.11 x 10.03 mrem6.50 x 10-02 mrem1.34 x 10-&deg; mrem1.10 x 10.02 mremTotal = 0.19 mremNote 1: Maximum Concentrations are radioiodine and noble gas concentrations leaking from the containment buildingand exiting the exhaust stack reduced by a dilution factor of 292.52 of 86 Effluent Concentration Limits. Concentrations at Point of Maximum Concentrationand Radiation Doses in the Unrestricted Area -Noble GasesRadionuclide85Kr~~87Kr88Kr89Kr90Kr133Xel35mXe135mXe137XeEffluent Limit7.00 x 10.0 ptCi/ml1.00 x 10.07 pCi/ml2.00 x 10-&deg;8 jiCi/ml9.00 x 10-09 pCi/ml2.00 x 10.08 jiCi/ml2.00 x 10.08 pCi/ml5.00 X 10-07 iiCi/ml7.00 x 10.08 aCi/ml4.00 x 10.08 !iCi/ml2.00 x 10.08 aCi/m12.00 x 10.08 iiCi/ml2.00 x 10.0 iiCi/mlMaximum Concentration12.54 x 10-12 pCi/ml5.97 x 10-1&deg;/Ci/ml1.14 x 10.0 pCi/ml1.62 x 1 0-0 gCi/ml2.06 x 10.0 ptCi/ml2.06 x 10-0 2.27 x 10.0 pCi/ml5.21 x 10"1&deg; pCi/ml5.09 x 10-40 pCi/ml2.66 x 10-0 aiCi/ml2.81 x 10.0 pCi/mi2.27 x 10.0 pCi/mlRadiation Dose3.43 x 10-0 mrem5.63 x 10-&deg;4 mrem5.36 x i0.0 mrem1.69 x 10.02 mrem9.69 x 10&deg;3 mrem9.69 x 1O0-3mrem4.28 x 10-0 mrem7.01 x 10-o4 mrem1.20 x 10.03 mrem1.25 x 10-02 mrem1.33 x 10&deg;2mrem1.07 x 10-02 mremTotal = 0.08 mremNote 1: Maximum Concentrations are radioiodine and noble gas concentrations leaking from the containment buildingand 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 dosesfrom the radioiodines and noble gases must be added together, and result in the following values:Dose from Radioidines and Noble Gases in the Unrestricted AreaCommitted Effective Dose Equivalent (Radioiodine)Committed Effective Dose Equivalent (Noble Gases)Total Effective Dose Equivalent (Whole Body)0.19 mrem0.08 mrem0.27 mremSumming the doses from the noble gases and the radioiodines simply substantiates earlierstatements regarding the very low levels in the unrestricted area should a failure of a fueledexperiment occur, and should the containment building leak following such an event. Because thedose values are so low, the dose from the noble gases becomes the dominant value, but the overallTEDE is still only 0.19 mrem, a value far below the applicable 10 CFR 20 regulatory limit for theunrestricted area. Additionally, leakage in mechanical equipment room 114 from such items asvalve packing, flange gaskets, pump mechanical seals, etc. was also considered in the fuel failureanalysis. 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 partof the control room operator's actions. The additional contaminated water vapor and associatedisotopes added to the facility ventilation exhaust system made a minimal (<1%) contribution to thetotal dose of an individual located in the facility. Therefore, the dose contribution to theunrestricted area would be expected to be approaching zero.53 of 86 13.2.1.3 ConclusionsGenerally, the most severe condition which is analyzed with regard to reactor accidents is either aloss of primary coolant or a loss of primary coolant flow during reactor operation. Both of theseaccidents are analyzed in this chapter and the results show no core damage. In addition, there areno other accidents that will result in a release of fission products from the reactor fuel, which isassumed in the fuel failure analysis. Even if such an event were to occur, the anti-siphon andreactor loop vent systems are designed such that any released radioactivity would be contained inthe primary coolant system.System design and operational procedures reduce the likelihood of any foreign material beingintroduced into the reactor core that could cause a partial flow blockage. Calculations have beenperformed which indicate that even partial flow blockage to a fuel element will not result incladding failure (Ref. 13.2). A considerable margin of safety has been designed into the system inthis regard. Also, considering the results of the analyses which show no core damage in the eventof a loss of primary coolant or a loss of primary coolant flow accident (See Sections 13.2.3 and13.2.4), and in view of the design of the anti-siphon and reactor loop vent systems, it is concludedthat there is no radiation risk to personnel in the reactor containment building or in the unrestrictedarea 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  
-7.22 x 10-o2 ptCi/mn3(7.22 x 10-&deg; /.tCi/m3) x (1 m3/106 ml) =7.22 x 10.08 pCi/mlNote: Same calculation is used for the other isotopes listed below.The average radioiodine concentrations are the sum of the initial concentrations and theconcentrations after 10 minutes decay divided by 2.Average Radioiodine Concentrations in the Containment Building Air during the 10 Minutes1311 -- 7.22 x 10-&deg;8 pCi/ml132I -- 1.36 x 10-07 pCi/ml133I -- 2.16 x 10-07 gCi/ml134I -- 2.53 x 10-07 pxCi/ml135I -2.18 x 10-&deg; pxCi/mlAs noted previously, the krypton and xenon noble gases released into the reactor pool from theprimary coolant system during the assumed 1 0-minute interval following the fuel failure (Note: theprimary coolant system is shut down and secured, and the leakage driving force is stopped within10 minutes), are assumed to pass immediately through the pool water and enter the containmentbuilding air volume where they instantaneously form a uniform concentration in the isolatedstructure. Based on the 225,000-ft3 volume of containment building air and the previously listedCurie quantities of these gases released into the reactor pool, the maximum noble gasconcentrations 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 ft3 x 35.3 147 ft3/m3-6.35 x 10+02 uCi x (1.60 x 10-o4 1/in3)-9.96 x 10.02 gxCi/m3(9.96 x 10.02 3) x (1 m3/106 ml) = 9.96 x i0.0 ptCi/mlNote: Same calculation is used for the other isotopes listed below.The average noble gas concentrations are the sum of the initial concentrations and theconcentrations after 10 minutes decay divided by 2.Average Noble Gas Concentrations in the Containment Building Air during the 10 MinutesKr -9.96 x 10-&deg;8 jiCi/ml85m~r -2.30 x 10-&deg;5 iCi/ml87Kr -4.27 x 10&deg;5 pCi/ml88r- 6.22 x 10&deg;5 ptCi/ml89r- 4.47 x 10"&deg;5 p.Ci/ml9&deg;r- 4.03 x 10"&deg;5 pCi/ml133Xe -l3SlXe _'38Xe -'39Xe -8.90 x 10-&deg; pCi/ml2.02 x 10-&deg; gtCi/ml1.63 x 10-&deg;5 gCi/ml6.05 x 10.05 jgCi/ml8.88 x 10.05 pCi/ml4.45 x 10.05 pCi/ml45 of 86 The objective of this calculation is to present a worst-case dose assessment for an individual whoremains in the containment building for 10 minutes following fuel failure. Therefore, as notedpreviously, the radioactivity in the evaporated pool water is assumed to be instantaneously anduniformly 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 thyroidfrom radioiodine and the dose to the whole body resulting from submersion in the airborne noblegases and radioiodine inside the containment building. As previously noted, the exposure time forthis dose assessment is 10 minutes.Because the airborne radioiodine source is composed of five (5) different iodine isotopes, it will benecessary to determine the dose contribution from each individual isotope and to then sum theresults. 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 25mrem/DAC-hr. Additionally, 10 minutes of one DAC-br is 1.67 x 10-01 DAC-br.1311 concentration in containment1311 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)= 3.61Therefore, 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-&deg;1 DAC-hr)= 1.51x 10+&deg;lmremNote: Same calculation is used for the other radioiodines listed below.Derived Air Concentration Values and 10-Minute Exposures -RadioiodineRadionuclide13111321133I13411351Derived Air Concentration2.00 x 10-&deg;8 1Ci/ml3.00 x 10-06 gtCi/ml1.00 x 1 007 ptCi/ml2.00 x 10-0 jiCi/ml7.00 x 10.o iiCi/ml10-Minute Exposure1.51 x 10401 mrem1.89 x 10-&deg;' mrem8.99 x 10+&deg;&deg; mrem5.27 x 10.02 mnrem1.30 x 10+&deg; mremTotal =25.58 mrem46 of 86 Doses from the kryptons and xenons present in the containment building are assessed in much thesame manner as the radioiodines, and the dose contribution from each individual radionuclide mustbe calculated and then added together to arrive at the final noble gas dose. Because the dose fromthe noble gases is only an external dose due to submersion, and because the DACs for theseradionuclides are based on this type of exposure, the individual noble gas doses for 10 minutes incontainment were based on their average concentration in the containment air and thecorresponding DAC.Example calculation of whole body dose due to KrThe DAC can also be defined as 5,000 mrenm/2,000 hrs, or 2.5 mnremiDAC-hr. Additionally, 10minutes of one DAC-hr is 1.67 x 1001 DAC-hr.85Kr concentration in containment = 9.96 x 1 0.0 85Kr DAC (10 CFR 20) = 1.00 x i0"&deg; pCi/mlDose Multiplication Factor = (85Kr concentration) / (SSKr DAC)= (9.96 x 10.08 pCi/mI) / (1.00 x 10o4~ pCi/mi)= 0.001Therefore, 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"&deg;1 DAC-hr)=4.15 xl10&deg;4mremNote: 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 GasesRadionuclide85Kr85m~r87Kr88Kr89Kr90Kr133Xe135Xel35rage137Xe138Xe139XeDerived Air Concentration1.00 x 10"&deg; xtCi/ml2.00 x 1005 iiCi/ml5.00 x 10-&deg;6 itCi/ml2.00 x 100o6 p.Ci/ml6.00 x 100o6 gCi/ml6.00 x 10.o6 iCi/ml1.00 x 10"04 pCi/ml1.00 x 10-05 jCi/ml9.00 x 10-06 xiCi/ml6.00 x 10.06 gCi/ml4.00 X 10-06 pCi/mi6.00 x 10&deg;6 jiCi/ml10-Minute Exposure4.51 x 10-&deg;4 mrem4.80 x 10-&deg;" mrem3.56 x 10400 mrem1.30 x 10+01 mrem3.11 x 10400 mrem2.80 x 10+00&deg;mrem3.71 x 10-&deg;1 mrem8.43 x 10-&deg;1 mrem7.54 x l0-&deg;' mrem4.20 x 10+&deg; mrem9.25 x 1040&deg; mrem3.09 x 10+00 mremTotal = 41.42 mrenlTo finalize the occupational dose in terms of Total Effective Dose Equivalent (TEDE) for a1 0-minute exposure mn the containment building after target failure, the doses from the radioiodinesand noble gases must be added together, and result in the following values:10-Minute Dose from Radioidines and Noble Gases in the Containment BuildingCommitted Dose Equivalent (Thyroid):Committed Effective Dose Equivalent (Thyroid):Committed Effective Dose Equivalent (Noble Gases):Total Effective Dose Equivalent (Whole Body):25.58 mrem0.77 mrem41.42 mrem42.18 mremBy comparison of the maximum TEDE and Committed Dose Equivalent (CDE) for thoseoccupationally-exposed during fuel failure to applicable NRC dose limits in 10 CFR 20, the finalvalues are shown to be well within the published regulatory limits and, in fact, lower than 1% ofany occupational limit.Radiation shine through the containment structure was also evaluated when considering accidentconditions and dose consequences to the public and MUJRR staff. Calculation of exposure ratefrom the target failure was performed using the computer program MicroShield 8.02 with aRectangular Volume -External Dose Point geometry for the representation of the containmentstructure (Attachment 12). MicroShield 8.02 is a product of Grove Software and is acomprehensive photon/gamma ray shielding and dose assessment program that is widely used byindustry for designing radiation shields.The exposure rate values provided below represents the radiation fields at 1 foot (30.5 cm) from a12-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 theexposure rate values are identical to those used for determining the dose to a worker withincontainment from noble gases. For radioiodine, the total iodine activity of the target was used forthe dose calculations, not the amount that evaporated in 10 minutes. The source term also assumesa homogenous mixture of nuclides within the containment free volume.Radiation Shine through the Containment BuildingExpo sure Rate at 1-Foot from Containment Building Wall: 1.074 mrem/hrExposure Rate at Emergency Planning Zone Boundary (150 meters): 0.007 mrem/hrA confirmatory analysis of the accident condition yielding the largest consequence was validatedindependently by the use of the MCNIP code. This analysis yielded a result 21% less than theMicroshield method and results provided above.As noted earlier in this analysis, the containment building ventilation system will shut down and thebuilding itself will be isolated from the surrounding areas. Fuel failure will not cause an increase inpressure inside the reactor containment structure; therefore, any air leakage from the building willoccur as a result of normal changes in atmospheric pressure and pressure equilibrium between theinside of the containment structure and the outside atmosphere. It is highly probable that there willbe no pressure differential between the inside of the containment building and the outsideatmosphere, and consequently there will be no air leakage from the building and no radiation doseto members of the public in the unrestricted area. However, to develop what would clearly be aworst-case scenario, this analysis assumes that a barometric pressure change had occurred inconjunction with the target failure. A reasonable assumption would be a pressure change on theorder of 0.7 inches of Hig (25.4 mm of Hig at 60&deg;C), which would then create a pressure differentialof about 0.33 psig (2.28 kPa above atmosphere) between the inside of the isolated containmentbuilding and the inside of the adjacent laboratory building, which surrounds most of thecontainment structure. Making the conservative assumption that the containment building will leakat the TS leakage rate limit [10% of the contained volume over a 24-hour period from an initialoverpressure of 2.0 psig (13.8 kPa above atmosphere)], the air leakage from the containmentstructure in standard cubic feet per minute (scfmn) as a function of containment pressure can beexpressed by the following equation:LR = 17.85 x (CP-14.7)112;where:LR = leakage rate from containment (scfmn); andCP = containment pressure (psia).The minimum Technical Specification free volume of the containment building is 225,000-ft3 atstandard temperature and pressure. At an initial overpressure of 2.0 psig (13.8 kPa aboveatmosphere), 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 air49 of 86 volume to 230,051 scf. The above equation describes the leakage rate that results in this drop ofcontained air volume over 1,440 minutes (24 hours).When applying the Technical Specification leakage rate equation to the assumed initialoverpressure condition of 0.33 psig (2.28 kPa above atmosphere), it would take approximately 16.5hours 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 periodwould be about 5.2 scfln.Several factors exist that will mitigate the radiological impact of any air leakage from thecontainment building following target failure. First of all, most leakage pathways fromcontainment discharge into the reactor laboratory building, which surrounds the containmentstructure. Since the laboratory building ventilation system continues to operate during targetfailure, leakage air captured by the ventilation exhaust system is mixed with other building air, andthen 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 bydischarge out the exhaust stack and subsequent atmospheric dispersion, results in extremely lowradionuclide concentrations and very small radiation doses in the unrestricted area. A tabulation ofthese concentrations and doses is given below. These values were calculated following the samemethodology 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 tothe behavior of radioiodine, which has been studied extensively in the containment mockup facilityat 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% iodinedeposition in the containment building, each cubic meter of air released from containment has aradioiodine concentration that is 25% of each cubic meter within containment building air, then theradioiodine concentrations leaking from the containment structure into the laboratory building, inmicrocuries per milliliter, will be:Example calculation of 1311 released through the exhaust stack:-131I activity / (30,500 ft3/min x 16.5 hr x 60 min/hr x 28,300 ml/ft3)= 2.30 x 10+05 /8.55 x 10+'11ml-2.69 X 10-07 iiCi/ml(2.69 x 10-07 iiCi/ml) x (0.25) = 6.73 x 10.08 iiCi/mlNote: Same calculation is used for the other radioiodines listed below.Radioiodine Concentrations in Air Leaking from Containment1311 -- 6.73 x 10-&deg;8 jtCi/ml 133I -- 2.02 x 10-07 gtCi/ml 13I- 2.05 x 10-07 jiCi/ml1321 -1.30 x 10-07 pCi/ml 34 -2.49 x 10-07 pCi/ml50 of 86 Example calculation of 85Kr released through the exhaust stack:= 85}~. activity / (30,500 ft3/min x 16.5 hr x 60 mmn/hr x 28,300 mi/fl3)= 6.35 x 10+02 / 8.55 x 10+11 ml= 7.4 x1-w&deg; ItCi/mlNote: Same calculation is used for the other noble gases listed below.Noble Gas Concentrations in Air Leaking from Containment and Exiting the Exhaust Stack85Kr -7.43 x 10-1 pCi/ml 87Kr -3.32 x 10-0 jtCi/ml 89Kr -6.00 x 10-0 giCi/ml85m~r -1.74 x 10-0 g.Ci/ml 88Kr -4.73 x 10.0 pxCi/ml 9&deg;Kr -6.00 x 10-0 jiCi/ml33e- 6.64 x 10-07 pCi/mil 3smXe -- 1.49 x 10-07 ptCi/ml 138Xe -8.22 x 10-07 pCi/ml135Xe -1.52 x 10-0 pCi/ml 137Xe -7.76 x 10-07 pCi/mI '39Xe -6.64 x 10-0 ptCi/mlAssuming, as stated earlier, that (1) the average leakage rate from the containment building is 5.2scfrn, (2) the leak continues for about 16.5 hours in order to equalize the containment buildingpressure with atmospheric pressure, (3) the flow rate through the facility's ventilation exhaust stackis 30,500 scfrn, (4) the reduction in concentration from the point of discharge at the exhauststack to the point of maximum concentration in the unrestricted area is a factor of 292 and (5) thereis no decay of any radioiodines or noble gases, then the following concentrations of radioiodinesand noble gases with their corresponding radiation doses will occur in the unrestricted area. Thevalues listed are for the point of maximum concentration in the unrestricted area assuming auniform, semi-spherical cloud geometry for noble gas submersion and further assuming that themost conservative (worst-case) meteorological conditions exist for the entire 16.5-hour period ofcontainment 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 thatwas used to determine doses inside the containment building, and it was assumed that an individualwas present at the point of maximum concentration for the full 16.5 hours that the containmentbuilding was leaking.A worst-case scenario dilution factor of 292 for effluent dilution using the Pasquill-Guifford Modelfor atmospheric dilution is used in this analysis. We assume that all offsite (public) dose occursunder these atmospheric conditions at the site of interest, i.e. 760 meters North of MUIRR. In ourcase at 760 meters it occurs only during Stability Class F conditions; which normally only occur11.4% 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. AnEffluent Concentration Limit of 2.0 x 10.08 itCi/ml is used for the "unlisted" isotopes, which equalsthe DAC/300 when using the DOE Part 835 Default DAC limits of 6.0 x 10-06 p.C/ml. Exposure at1 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 8760hours per year. Thus, the time assumed to be exposed to the effluent concentration limit is a factorof 4.38 longer than the 2000 hours per year that defines a DAC. The isotopes in question are based51 of 86 on a default DAC limit of 6.0 x 1 0-06 for short-lived (< 2 hour half-lives) submersion DAC 'sin Appendix C of Part 835. No credit is taken for transit time from the stack to the receptor pointnor is credit taken for decay inside containment until release. In the case of Kr-89 and Xe-i137 thetransit 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 theDAC value to establish the effluent concentration limit is warranted. This reduction factor of 300 isconsistent, in fact more conservative, than 10 CFR 20 Appendix B, as the N4RC calculates EffluentConcentration 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 mremihr1311 concentration131i effluent concentration limit1311 Conversion Factor= 2.30 x 10-1&deg; pCi/ml=2.00 x 10-10 gCi/ml= 5.71 X 10-&deg; mremihrTherefore, 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&deg; giCi/ml / (2.00 x i010-1 gCi/ml x 5.71 x 10-03 mremihr x 16.5 brs)= 1.09 x10&deg;01mremNote: Same calculation is used for the other isotopes (radioiodines and noble gases) listed below.Effluent Concentration Limits. Concentrations at Point of Maximum Concentrationand Radiation Doses in the Unrestricted Area -RadioiodineRadionuclide1311132I133I134I1351Effluent Limit2.00 x 10"1&deg; giCi/ml6.00 x 10.08 iiCi/ml6.00 x 10-09 Maximum Concentration12.30 x 10-10 giCi/ml4.47 x 10.1&deg; pCi/ml6.90 x 10-10 iiCi/ml8.54 x 10-l0 jiCi/ml7.03 x 10-1&deg; jCi/mlRadiation Dose1.09 x 10-01 mrem2.11 x 10.03 mrem6.50 x 10-02 mrem1.34 x 10-&deg; mrem1.10 x 10.02 mremTotal = 0.19 mremNote 1: Maximum Concentrations are radioiodine and noble gas concentrations leaking from the containment buildingand exiting the exhaust stack reduced by a dilution factor of 292.52 of 86 Effluent Concentration Limits. Concentrations at Point of Maximum Concentrationand Radiation Doses in the Unrestricted Area -Noble GasesRadionuclide85Kr~~87Kr88Kr89Kr90Kr133Xel35mXe135mXe137XeEffluent Limit7.00 x 10.0 ptCi/ml1.00 x 10.07 pCi/ml2.00 x 10-&deg;8 jiCi/ml9.00 x 10-09 pCi/ml2.00 x 10.08 jiCi/ml2.00 x 10.08 pCi/ml5.00 X 10-07 iiCi/ml7.00 x 10.08 aCi/ml4.00 x 10.08 !iCi/ml2.00 x 10.08 aCi/m12.00 x 10.08 iiCi/ml2.00 x 10.0 iiCi/mlMaximum Concentration12.54 x 10-12 pCi/ml5.97 x 10-1&deg;/Ci/ml1.14 x 10.0 pCi/ml1.62 x 1 0-0 gCi/ml2.06 x 10.0 ptCi/ml2.06 x 10-0 2.27 x 10.0 pCi/ml5.21 x 10"1&deg; pCi/ml5.09 x 10-40 pCi/ml2.66 x 10-0 aiCi/ml2.81 x 10.0 pCi/mi2.27 x 10.0 pCi/mlRadiation Dose3.43 x 10-0 mrem5.63 x 10-&deg;4 mrem5.36 x i0.0 mrem1.69 x 10.02 mrem9.69 x 10&deg;3 mrem9.69 x 1O0-3mrem4.28 x 10-0 mrem7.01 x 10-o4 mrem1.20 x 10.03 mrem1.25 x 10-02 mrem1.33 x 10&deg;2mrem1.07 x 10-02 mremTotal = 0.08 mremNote 1: Maximum Concentrations are radioiodine and noble gas concentrations leaking from the containment buildingand 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 dosesfrom the radioiodines and noble gases must be added together, and result in the following values:Dose from Radioidines and Noble Gases in the Unrestricted AreaCommitted Effective Dose Equivalent (Radioiodine)Committed Effective Dose Equivalent (Noble Gases)Total Effective Dose Equivalent (Whole Body)0.19 mrem0.08 mrem0.27 mremSumming the doses from the noble gases and the radioiodines simply substantiates earlierstatements regarding the very low levels in the unrestricted area should a failure of a fueledexperiment occur, and should the containment building leak following such an event. Because thedose values are so low, the dose from the noble gases becomes the dominant value, but the overallTEDE is still only 0.19 mrem, a value far below the applicable 10 CFR 20 regulatory limit for theunrestricted area. Additionally, leakage in mechanical equipment room 114 from such items asvalve packing, flange gaskets, pump mechanical seals, etc. was also considered in the fuel failureanalysis. 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 partof the control room operator's actions. The additional contaminated water vapor and associatedisotopes added to the facility ventilation exhaust system made a minimal (<1%) contribution to thetotal dose of an individual located in the facility. Therefore, the dose contribution to theunrestricted area would be expected to be approaching zero.53 of 86 13.2.1.3 ConclusionsGenerally, the most severe condition which is analyzed with regard to reactor accidents is either aloss of primary coolant or a loss of primary coolant flow during reactor operation. Both of theseaccidents are analyzed in this chapter and the results show no core damage. In addition, there areno other accidents that will result in a release of fission products from the reactor fuel, which isassumed in the fuel failure analysis. Even if such an event were to occur, the anti-siphon andreactor loop vent systems are designed such that any released radioactivity would be contained inthe primary coolant system.System design and operational procedures reduce the likelihood of any foreign material beingintroduced into the reactor core that could cause a partial flow blockage. Calculations have beenperformed which indicate that even partial flow blockage to a fuel element will not result incladding failure (Ref. 13.2). A considerable margin of safety has been designed into the system inthis regard. Also, considering the results of the analyses which show no core damage in the eventof a loss of primary coolant or a loss of primary coolant flow accident (See Sections 13.2.3 and13.2.4), and in view of the design of the anti-siphon and reactor loop vent systems, it is concludedthat there is no radiation risk to personnel in the reactor containment building or in the unrestrictedarea 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 theconsequences of a loss of coolant accident (LOCA). MURR SAR Section 13.2.3.2 describes theLOCA event for the loss of the PCS integrity, and states that the accident of greatest consequence isa rupture in the short section of the PCS piping (either the cold leg or the hot leg) between thereactor pool and either isolation valves (507B or 507A4). The SAR describes the consequences of acold leg break between the isolation valve 507B and the reactor pool in significant detail. The hotleg break discussion is more succinct. The SAR also states that how "the anti-siphon systemensures 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 breaksequence. It is our understanding that after isolation occurs the coolant surrounding the core heatsup, and because of natural buoyancy it flows upward and out of the reactor pressure vessel into thein-pool heat exchanger. After passing through the heat exchanger, the cooled water may then flowdownward through what is normally the upward flow path of the inverted loop and then into thebottom of the pressure vessel. As this process continues, the water will fill up the downwardinverted loop to the bottom of the core reaching to the inverted loop creating an open condition forreleasing 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 fueltemperatures, or justify why no additional information is needed.In the second paragraph of the above question, the NRC staff's stated understanding is closer towhat occurs during a Loss of Flow Accident (LOFA) but not for the hot-leg break Loss of CoolantAccident (LOCA). The difference is that during the hot-leg break LOCA some of the primarycoolant that is lost from the primary coolant system (PCS5) piping is located in the reactor pool, butno primary coolant is lost during a LOFA. So, in the LOFA, the natural convention flow pathdescribed above is established and provides more than sufficient cooling for the reactor core aftershutdown. During a hot-leg break LOCA, the anti-siphon system actuates and injects air into thePCS vertical 12-inch diameter piping above the inverted loop to the level of the in-pool heatexchanger outlet. The expanding air quickly voids the upper section of the potential PCS naturalconvention flow path.Key PCS components for the LOFA and LOCA are described in Table 1 along with their RELAPModel component number. These components are also indicated in the vertical cross-sectionalview 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 SystemRELAP No. Component Description405-1 hn-Pool Heat Exchanger Upper Header405-2 hn-Pool Heat Exchanger Vertical Finned Tubes405-3 hn-Pool Heat Exchanger Lower Header401-2 Last 4 feet of 6-inch Diameter Inlet Piping to hn-Pool Heat Exchanger UpperHeader407 PCS Vertical 12-inch Diameter Pipe above hn-Pool Heat Exchanger Outlet toFlanged Natural Circulation Piping406 PCS Vertical 12-inch Diameter Pipe Above hnverted Loop to hn-Pool HeatExchanger Outlet139 Horizontal PCS Inlet Piping to Upper Section of Pressure Vessel501 Pressure Vessel Above the Core to Pressure Vessel Head100-3 Last 4.917 feet of Vertical Hot-Leg Piping Before Joining Pipe No. 101101 PCS Horizontal 12-inch Diameter Pipe (Section of inverted Loop)102 PCS Downward Vertical 12-inch Diameter Pipe from the Normal Outlet End ofNo. 101 Towards the PCS Hot-Leg Outlet Isolation ValveWith no pipe break occurring in the PCS during a LOFA, all of the above sections of the PCS stayfilled with primary coolant. This results in the development of the natural circulation flow pathdescribed in Paragraph 2 of the above question. However, for the hot-leg break LOCA, a doubleshear on the inlet and outlet sides of the hot-leg isolation valve is assumed, such that the hot-legisolation valve is functionally eliminated. With anti-siphon system air being injected intocomponent 406, voiding starts at the higher elevated connected PCS components as indicated inTable 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 exchangercomponents (405-1, 405-2, 405-3) start draining. Also, the following components aredraining:* 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 thereactor core, drain downward with the primary coolant which is flowing down the pressure vesselthrough the core and then up through the PCS hot-leg outlet piping until their upper level is in57 of 86 equilibrium with the water level in component 101 or 100-3. With the in-pool portion of the PCSdrained to this level, the natural circulation flow path through the in-pool heat exchanger iseliminated.However, the hot-leg break LOCA RELAP analysis shows that the highest peak fuel center linetemperature of 281.2 0F (138.4 &deg;C) occurs in fuel plate number-i, 0.2 seconds after the LOCAbegins. After this initial peak temperature at the start of the transient, the next highest fuel platecenterline temperature of 231.7 &deg;F (110.9 &deg;C) occurs in plate number-22 at 22 seconds as shown inSARk Figure 13.20. The highest coolant channel temperature 219.0 0F (103.9 &deg;C) occurs in channel7 at 123.3 seconds and in channel 6 at 123.4 seconds as shown in SARk Figure 13.21. There issufficient heat transfer from the PCS to the pooi coolant due to conduction through the PCS pipingto avoid any fuel damage.58 of 86  
: 8. NUREG-1537, Section 13.1.3, "Loss of Coolant, "provides guidance to the licensee to consider theconsequences of a loss of coolant accident (LOCA). MURR SAR Section 13.2.3.2 describes theLOCA event for the loss of the PCS integrity, and states that the accident of greatest consequence isa rupture in the short section of the PCS piping (either the cold leg or the hot leg) between thereactor pool and either isolation valves (507B or 507A4). The SAR describes the consequences of acold leg break between the isolation valve 507B and the reactor pool in significant detail. The hotleg break discussion is more succinct. The SAR also states that how "the anti-siphon systemensures 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 breaksequence. It is our understanding that after isolation occurs the coolant surrounding the core heatsup, and because of natural buoyancy it flows upward and out of the reactor pressure vessel into thein-pool heat exchanger. After passing through the heat exchanger, the cooled water may then flowdownward through what is normally the upward flow path of the inverted loop and then into thebottom of the pressure vessel. As this process continues, the water will fill up the downwardinverted loop to the bottom of the core reaching to the inverted loop creating an open condition forreleasing 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 fueltemperatures, or justify why no additional information is needed.In the second paragraph of the above question, the NRC staff's stated understanding is closer towhat occurs during a Loss of Flow Accident (LOFA) but not for the hot-leg break Loss of CoolantAccident (LOCA). The difference is that during the hot-leg break LOCA some of the primarycoolant that is lost from the primary coolant system (PCS5) piping is located in the reactor pool, butno primary coolant is lost during a LOFA. So, in the LOFA, the natural convention flow pathdescribed above is established and provides more than sufficient cooling for the reactor core aftershutdown. During a hot-leg break LOCA, the anti-siphon system actuates and injects air into thePCS vertical 12-inch diameter piping above the inverted loop to the level of the in-pool heatexchanger outlet. The expanding air quickly voids the upper section of the potential PCS naturalconvention flow path.Key PCS components for the LOFA and LOCA are described in Table 1 along with their RELAPModel component number. These components are also indicated in the vertical cross-sectionalview 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 SystemRELAP No. Component Description405-1 hn-Pool Heat Exchanger Upper Header405-2 hn-Pool Heat Exchanger Vertical Finned Tubes405-3 hn-Pool Heat Exchanger Lower Header401-2 Last 4 feet of 6-inch Diameter Inlet Piping to hn-Pool Heat Exchanger UpperHeader407 PCS Vertical 12-inch Diameter Pipe above hn-Pool Heat Exchanger Outlet toFlanged Natural Circulation Piping406 PCS Vertical 12-inch Diameter Pipe Above hnverted Loop to hn-Pool HeatExchanger Outlet139 Horizontal PCS Inlet Piping to Upper Section of Pressure Vessel501 Pressure Vessel Above the Core to Pressure Vessel Head100-3 Last 4.917 feet of Vertical Hot-Leg Piping Before Joining Pipe No. 101101 PCS Horizontal 12-inch Diameter Pipe (Section of inverted Loop)102 PCS Downward Vertical 12-inch Diameter Pipe from the Normal Outlet End ofNo. 101 Towards the PCS Hot-Leg Outlet Isolation ValveWith no pipe break occurring in the PCS during a LOFA, all of the above sections of the PCS stayfilled with primary coolant. This results in the development of the natural circulation flow pathdescribed in Paragraph 2 of the above question. However, for the hot-leg break LOCA, a doubleshear on the inlet and outlet sides of the hot-leg isolation valve is assumed, such that the hot-legisolation valve is functionally eliminated. With anti-siphon system air being injected intocomponent 406, voiding starts at the higher elevated connected PCS components as indicated inTable 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 exchangercomponents (405-1, 405-2, 405-3) start draining. Also, the following components aredraining:* 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 thereactor core, drain downward with the primary coolant which is flowing down the pressure vesselthrough the core and then up through the PCS hot-leg outlet piping until their upper level is in57 of 86 equilibrium with the water level in component 101 or 100-3. With the in-pool portion of the PCSdrained to this level, the natural circulation flow path through the in-pool heat exchanger iseliminated.However, the hot-leg break LOCA RELAP analysis shows that the highest peak fuel center linetemperature of 281.2 0F (138.4 &deg;C) occurs in fuel plate number-i, 0.2 seconds after the LOCAbegins. After this initial peak temperature at the start of the transient, the next highest fuel platecenterline temperature of 231.7 &deg;F (110.9 &deg;C) occurs in plate number-22 at 22 seconds as shown inSARk Figure 13.20. The highest coolant channel temperature 219.0 0F (103.9 &deg;C) occurs in channel7 at 123.3 seconds and in channel 6 at 123.4 seconds as shown in SARk Figure 13.21. There issufficient heat transfer from the PCS to the pooi coolant due to conduction through the PCS pipingto avoid any fuel damage.58 of 86  
: 9. NUREG-1537, Section 13.1.5, "Mishandling or Malfunction of Fuel" provides guidance that thelicensee analyze the consequences of a mishandled fuel event. MURR SAR Section 13.2.5.2.1describes damage to a fuel element due to mishandling. It states that the mishandling could occurduring movement and packaging of the irradiated fuel, damage could only occur to the inner or theouter fuel plate, and could only occur during fuel element relocation activities. Because thisaccident 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 anoccurrence assuming that the fuel element has decayed for 60 days as part of the spent fuelmovement from storage to a shipping container. However, the NRC staff questions whether thisevent could also occur during the initial stages of refueling which would invalidate the assumptionof 60 days of decay. The NRC staff also performed a confirmatory calculation based on thisinventory using the cited values for the MHA analysis, and it results in an inventory that is sevenpercent larger than reported by MURR.a. Explain the possibility of this event occurring during the initial stages of refueling, and theapplicability of using the stated decay time in the dose calculation. Also, describe anyradioactivity release alarms that are expected to actuate, and whether containment isolationis expected, including the time required to verify containment isolation, or justify why noadditional 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 andmethodologies as used in the Maximum Hypothetical Accident (MHA) (now referred to as the"Fuel Failure during Reactor Operation" accident) and Fueled Experiment Failure. The onlyexceptions are the source term, which is explained in the accident analysis, as well as the decayprior to the accident (which is once again explained in the analysis). As discussed in the responseto RAI 10O.a, the primary coolant system does not have to be secured for a failed fueled experimentor for a FHA. The only required action for Operations personnel is to verify that the containmentbuilding has been evacuated following a containment building isolation, which will occur duringboth of these accident scenarios. MUJIRR performs an evacuation dr-ill every year and the typicaltime period for all personal to evacuate the containment building, including verification byOperations personnel, is two (2) to two and a half (2.5) minutes. For the purposes of the failedfueled experiment and FRA calculations, a conservative assumption of five (5) minutes is used forboth accident scenarios. Additionally, verifying that the reactor has shut down and containment hasisolated only takes a few moments -all control blade positions, reactor power meters, andcontainment isolation valve and door indications are in clear view of the reactor operator in thecontrol room.b. Provide the details of how the source term is determined, or justify' why no additionalinformation is needed.As described in the FHA analysis above, the two most outer fuel plates of a fuel element, number-land -24, are the plates most likely to be damaged during fuel handling. The number-i fuel platecontains 19.26 grams of U-235 before irradiation. The highest peak power density in the various59 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 ofthe fuel plate. The number-24 fuel plate has a lower peak power density and contains 45.32 gramsof U-235, and has the most surface area to be damaged. To be conservative, the analysis assumesthat 0.125 grams of U-235 is exposed from plate number-i during the FHA, which corresponds toremoving a section of fuel meat from a plate that is 1 inch square and 5 mils thick. A powerpeaking 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 andAccounting Procedures as outlined in the Operations Procedures. Irradiated fuel is handled with aspecially 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 fuelhandling sequence. Fuel elements are always handled one at a time so that they are maintained in acriticality-safe configuration. New or irradiated fuel may be stored in any one of 88 in-pool fuelstorage locations (not including the core). These storage locations are designed to ensure ageometry such that the calculated Keff is less than 0.9 under all conditions of moderation, thusallowing sufficient convection cooling and providing sufficient radiation shielding.So the fuel handling system provides a safe, effective and reliable means of transporting andhandling 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 andmagnetic particle testing, as appropriate. Therefore, no specific accidents regarding the handling offuel have been identified for the MUIRR. The probability of dropping a fuel element whileunderwater and damaging it severely enough to breach the fuel cladding was considered. Aconservative potential radionuclide release and calculation of the occupational exposure areincluded below.The following calculations determining the postulated dose from a potential release of radioactivityfrom a fuel element during a handling accident closely follow the "Fuel Failure during ReactorOperation" calculations for personal exposure due to a release of fission products. The objectiveof these calculations is to present a worst-case dose assessment for a person who remains in thecontainment 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 ofelements. Paired elements are always loaded opposite each other in the core. All eight (8) fuelelements are replaced every refueling. MURR has averaged refueling the core more than 52 timesa year since 1977. This type of accident has never occurred at MIJRR during any of these fuelhandlings.The two outer fuel plates of a fuel element, number-i and -24, are the plates most likely to bedamaged during fuel handling. The number-i fuel plate contains 19.26 grams of U-235 beforeirradiation. The highest peak power density in the various MURR core configurations occurs infuel plate number-i of a previously unirradiated fuel element, which has a power peaking factor of4.116 -located between 14.75 to 15.75 inches down from the top of the fuel plate. The number-24fuel plate has the most surface area to be damaged; however, it has a lower peak power density andcontains 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 offuel meat from a plate that is 1 inch square and 5 mils thick. A power peaking factor of 4.116 isalso applied.61 of 86 The following radioiodine, krypton and xenon activities will be present in the MURR core 30minutes after shutdown from 10 MW full power operation. Refuelings typically occur no soonerthan an hour after shutdown. This takes into account the time required to shut down the reactor, tosecure the primary coolant system (required to stay in operation a minimum of 15 minutes after thecontrol blades are fully inserted), and to remove the reactor pressure vessel head. For the purposeof the FHA calculations, a conservative assumption of 30 minutes is used.Radioiodine and Noble Gas Activities in the Core after 30-Minute Decay131I -9.93 x 10+04 Ci 85Kr -2.47 x 10+01 Ci 133Xe -2.73 x 10+05 Ci132I -2.68 x 10+o Ci 85mKr -1.29 x 10+05 Ci 135Xe -1.13 x i0o Ci1331 -- 5.65 x i0+0 Ci 87Kr -1.67 x 10+&deg; Ci 135mXe -- 4.79 x 10+04 Ci134I -- 5.80 X i0+0 Ci 88Kr -2.73 x 10+05 Ci '37Xe -2.37 x 10+0 Ci1351 -- 5.07 x 10+07 Ci 89}r -5.5 x 10+02 Ci 138X -1.22 x 10+05 Ci9&deg;Kr -6.66 x 10-12 Ci 139Xe -8.33 x 10-09 CiFission products released into the reactor pool will be detected by the pool surface and ventilationsystem exhaust plenum radiation monitors. However, for the purposes of this analysis, it isassumed that an actuation of the containment building isolation system occurs by action of the poolsurface radiation monitor. Actuation of the isolation system will prompt Operations personnel toensure that a total evacuation of the containment building is accomplished promptly, usually withintwo (2) to two and a half (2.5) minutes. A conservative 5-minute evacuation time is used as thebasis for the stay time in the dose calculations for personnel that are in containment during theFRA.The following radioiodine and noble gas activities from 0.125 grams of U-235 from the peak powerposition of fuel plate number-i in the peak power density fuel element are assumed toinstantaneously and homogenously distribute in the reactor pool.Example calculation of 1311 released into the reactor pool:= (1311 in fuel / 235U in core) x 235U 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 1iCiExample calculation of 85Kr released into the reactor pool:= (85Kr in fuel / 235U in core) x 235U exposed x PPF x 10+06 iCi/Ci= (2.47 x 10+&deg;' 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 Pool131I -- 9.33 x 10+06 kCi 85Kr -2.32 x 10+03 gCi 133Xe -- 2.56 x 10+07 giCi132I -2.52 x 10+0 giCi 85mKr -1.21 x 10+07 gxCi '35Xe -1.06 x 10+07 /iCi133I -5.31 x 10+07 g.Ci 87Kr -1.57 x 10+0 1iCi l35mXe -4.50 x 10+06 jiCi1341 -- 5.45 X 10+0 gtCi 88Kr -2.56 x 100 jiCi 137Xe -2.22 x 10+05 gCi35-- 4.76 x 10+07 ptCi 9K~r -5.25 x 10+0 ptCi '38Xe -1.15 x 10+07 plCi9&deg;~Kr -6.26 x 10-'&deg; ItCi 139Xe -7.83 x 10-07 pCiThe radioiodine released into the reactor pooi over a 5-minute interval is conservatively assumed tobe instantly and uniformly mixed into the 20,000 gallons (75,708 1) of bulk pooi water, which thenresults in the following pool water concentrations for the radioiodine isotopes. The water solubilityof the krypton and xenon noble gases released into the pool over this same time period are ignoredand they are assumed to pass immediately through the pool water and evolve directly into thecontainment building air volume where they instantaneously form a uniform concentration in theisolated structure.Radioiodine Concentrations in the Pool Water131I -- 4.67 x 10+02 pCilgal 1331 -- 2.66 x 10+03 giCi/gal 135I -2.38 x 10+03 PCi/gal13I -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, theevaporation rate from the reactor pool is approximately 80 gallons (302.8 L) of water per day. Forthe purposes of this calculation, it is assumed that a total of 20 gallons (75.7 L) of pool watercontaining the previously listed radioiodine concentrations evaporates into the containment buildingover the 5 minute period. Containment air with a temperature of 75 0F (23.9 &deg;C) and 100% relativehumidity contains H20 vapor equal to 40 gallons (151.4 L) of water. Since the air in containment isnormally at about 50% relative humidity, thus containing 20 gallons (75.7 L) of water vapor, theassumed addition of 20 gallons (75.7 L) of water vapor will not cause the containment air to besupersaturated. It is also conservatively assumed that all of the radioiodine activity in the 20gallons (75.7 L) of pool water instantaneously forms a uniform concentration in the containmentbuilding air. When distributed into the containment building, this would result in the followingradioiodine concentrations in the 225,000 ft3 (6,371.3 in3) air volume:Example calculation of 131I released into containment air:= 131I concentration in pooi water x 20 gal x 1/225,000 ft3 x 35.3 147 ft3/m3-4.67 x 10+02 ptCi/gal x (3.14 x 10.03 gal/in3)-1.46 pCi/m3(1.46 pCi/in3) x (1 m3/106 ml) =1.46 x 10.06 gxCi/mlNote: 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 theconcentrations after 5 minutes decay divided by 2.Average Radioiodine Concentrations in the Containment Building Air during the 5 Minutes31-- 1.46 x 10-06 pCi/mi 1331 -- 8.32 x 10&deg;06 gtCi/ml 135I -- 7.44 x 10.06 gCi/ml132j -3.91 x 10.06 gtCi/ml 14-- 8.28 x i0-&deg;5 gCi/mlAs 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 andenter the containment building air volume where they instantaneously form a uniform concentrationin the isolated structure. This assumption is extremely conservative since it ignores the knownsolubility of krypton and xenon noble gases in the 100 0F (37.8 &deg;C) pool water, which would reducetheir release into the containment building. Based on the 225,000-ft3 volume of containmentbuilding 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 wouldbe as follows:Example calculation of 85Kr released into containment air:-85Kr activity x 1/225,000 ft3 x 35.3 147 ft3/m3= 2.32 X 10+03 j.Ci x (1.60 x i004 1/in3)= 3.64 x 1001 g.Ci/m3(3.64 x 10&deg;1 gxCi/m3) x (1 m3/106 ml) = 3.64 x i0-07 plei/mlNote: Same calculation is used for the other isotopes listed below.The average noble gas concentrations are the sum of the initial concentrations and theconcentrations after 5 minutes decay divided by 2.Average Noble Gas Concentrations in the Containment Building Air during the 5 MinutesKr -3.64 x 10-07 FCi/ml 133Xe -4.02 x 10-03 pCi/miS5m~x -1.89 x 10-03 gCi/ml '35Xe -1.66 x 1003 jiCi/ml87r- 2.41 x 10.03 gCi/ml 13smXe -6.35 x 10-04 iCi/ml88r- 3.98 x i0-03 jiCi/ml 137Xe -2.45 x i0-&deg;5 gCi/mI89Kr -5.49 x 10-0 gtCi/ml '38Xe -1.61 x 10-&deg;3 pCi/mi9&deg;Kr -4.92 x 10-20 gCi/ml 139Xe -6.18 x10-17 Ci/mlThe objective of this calculation is to present a worst-case dose assessment for an individual whoremains in the containment building for 5 minutes following the ERA. Therefore, as notedpreviously, the radioactivity in the evaporated pool water is assumed to be instantaneously anduniformly 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 thyroidfrom radioiodine and the dose to the whole body resulting from submersion in the airborne noblegases and radioiodine inside the containment building. As previously noted, the exposure time forthis dose assessment is 5 minutes.Because the airborne radioiodine source is composed of five different iodine isotopes, it will benecessary to determine the dose contribution from each individual isotope and to then sum theresults. 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 25mrem/DAC-hr. Additionally, 5 minutes of one DAC-hr is 8.33 x 10-&deg;2 DAC-hr.1311 concentration in containment = 1.46 x 10-o6 g.Ci/ml131j DAC (10 CFR 20) = 2.00 x 10.o gCi/mlDose Multiplication Factor =(1311 concentration) / (131I DAC)= (1.46 x 10-&deg;6 gCi/ml) / (2.00 x 10.o8 gCi/ml)= 73Therefore, 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+&deg;2 mremNote: 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 thesame manner as the radioiodines, and the dose contribution from each individual radionuclide mustbe calculated and then added together to arrive at the final noble gas dose. Because the dose fromthe noble gases is only an external dose due to submersion, and because the DACs for theseradionuclides are based on this type of exposure, the individual noble gas doses for 5 minutes incontaimnment were based on their average concentration in the containment air and thecorresponding 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, 5minutes of one DAC-br is 8.33 x l0&deg;2 DAC-hr.85Kr concentration in containment = 3.64 X 10-07 gtCi/mnl65 of 86 85Kr DAC (10 CFR 20)Dose Multiplication Factor= 1.00 x 10"04 pCi/ml= (85Kr concentration) / (85Kr DAC)= (3.64 x 10-07 ptCi/ml) / (1.00 x 10-04 jiCi/ml)= 0.00364Therefore, 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)= 7.58X10-04 mremNote: 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 -RadioiodineRadionuclide131113211331134j1351Derived Air Concentration2.00 x 10-&deg;8 3.00 x 10-06 /xCi/ml1.00 x 10-07/Ci/ml2.00 x i0.0 gCi/inl7.00 x 10-07 iiCi/ml5-Minute Exposure1.52 x 10+o2 mrem2.71 x 10+00 mrem1.73 x 10+02 mrem8.62 x 10-&deg;1 mrem2.21 x 10+&deg;1 mremTotal = 351.44 mremDerived Air Concentration Values and 5-Minute Exposures -Noble GasesRadionuclide85Kr85m}r87KrS88K89Kr9OKr133XeI35Xe135mXe138Xe139XeDerived Air Concentration1.00 x 10-&deg;4 iCi/ml2.00 x 10.o jtCi/ml5.00 x 100o6 gCi/ml2.00 x 100o6 xtCi/ml6.00 x 10.o6 JiCi/m16.00 x 10-&deg;6 pCi/ml1.00 x 10-04 pCi/mi1.00 x 10-&deg;5 pCi/ml9.00 x 10-06 pCi/ml6.00 x 10-06 gCi/ml4.00 x 10-06 jiCi/ml6.00 X 10-06 .tCi/ml5-Minute Exposure7.58 x 10-04 mrem1.96 x 10+&deg;1 me1.00 x 10+02 mrem4.14 x 10+02 mrem1.91 x 10-&deg;1 mrem1.71 x 10-15 mrem8.36 x 10+00 mrem3.45 x 10+&deg;1 mrem1.47 x 10+&deg;1 mrem8.49 x 10-&deg;1 mrem8.37 x 10+&deg;1mre2.14 x 1012 mremTotal = 676.45 mrem66 of 86 To finalize the occupational dose in terms of Total Effective Dose Equivalent (TEDE) for a5-minute exposure in the containment building after a FHA, the doses from the radioiodines andnoble gases must be added together, and result in the following values:5-Minute Dose from Radioidines and Noble Gases in the Containment BuildingCommitted Dose Equivalent (Thyroid) 351.44 mremCommitted Effective Dose Equivalent (Thyroid) 10.54 mremCommitted Effective Dose Equivalent (Noble Gases) 676.45 mremTotal Effective Dose Equivalent (Whole Body) 687.00 mremBy comparison of the maximum TEDE and Committed Dose Equivalent (CDE) for thoseoccupationally-exposed during a FHA to applicable NRC dose limits in 10 CFR 20, the final valuesare shown to be well within the published regulatory limits and, in fact, lower than 15% of anyoccupational limit.Radiation shine through the containment structure was also evaluated when considering accidentconditions and dose consequences to the public and MUIRR staff. Calculation of exposure ratefrom a FHA was performed using the computer program MicroShield 8.02 with a RectangularVolume -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 comprehensivephoton/gamma ray shielding and dose assessment program that is widely used by industry fordesigning radiation shields.The exposure rate values provided below represents the radiation fields at 1 foot (30.5 cm) from a12-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 theexposure rate values are identical to those used for determining the dose to a worker withincontainment from noble gases. For radioiodine, the total iodine activity from the FHA was used forthe dose calculations, not the amount that evaporated in 5 minutes. The source term also assumes ahomogenous mixture of nuclides within the containment free volume.Radiation Shine through the Containment BuildingExpo sure Rate at 1-Foot from Containment Building Wall: 54.79 mrem/hrExposure Rate at Emergency Planning Zone Boundary (150 meters): 0.371 mremn/hrA confirmatory analysis of the accident condition yielding the largest consequence was validatedindependently by the use of the MCNP code. This analysis yielded a result 21% less than theMicroshield method and results provided above.As noted earlier in this analysis, the containment building ventilation system will shut down and thebuilding itself will be isolated from the surrounding areas. A FHA will not cause an increase inpressure inside the reactor containment structure; therefore, any air leakage from the building willoccur as a result of normal changes in atmospheric pressure and pressure equilibrium between the67 of 86 inside of the contaimnment structure and the outside atmosphere. It is highly probable that there willbe no pressure differential between the inside of the containment building and the outsideatmosphere, and consequently there will be no air leakage from the building and no radiation doseto members of the public in the unrestricted area. However, to develop what would clearly be aworst-case scenario, this analysis assumes that a barometric pressure change had occurred inconjunction with a FHA. A reasonable assumption would be a pressure change on the order of 0.7inches of Hg (25.4 mm of Hg at 60 which would then create a pressure differential of about0.33 psig (2.28 kPa above atmosphere) between the inside of the isolated containment building andthe 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 ratelimit [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 feetper minute (scfm) as a function of containment pressure can be expressed by the followingequation:LR = 17.85 x (CP-14.7)l"2;where:LR = leakage rate from containment (scfmn); andCP = containment pressure (psia).The minimum Technical Specification free volume of the containment building is 225,000-ft3 atstandard temperature and pressure. At an initial overpressure of 2.0 psig (13.8 kPa aboveatmosphere), 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 airvolume to 230,051 scf. The above equation describes the leakage rate that results in this drop ofcontained air volume over 1,440 minutes (24 hours).When applying the Technical Specification leakage rate equation to the assumed initialoverpressure condition of 0.33 psig (2.28 kPa above atmosphere), it would take approximately 16.5hours 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 periodwould be about 5.2 scfm.Several factors exist that will mitigate the radiological impact of any air leakage from thecontainment building following a FHA. First of all, most leakage pathways from containmentdischarge into the reactor laboratory building, which surrounds the containment structure. Since thelaboratory building ventilation system continues to operate during a FHA, leakage air captured bythe ventilation exhaust system is mixed with other building air, and then discharged from thefacility through the exhaust stack at a rate of approximately 30,500 cflm. Mixing of containment airleakage with the laboratory building ventilation flow, followed by discharge out the exhaust stackand subsequent atmospheric dispersion, results in extremely low radionuclide concentrations andvery small radiation doses in the unrestricted area. A tabulation of these concentrations and doses68 of 86 is given below. These values were calculated following the same methodology stated in Section5.3.3 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 tothe behavior of radioiodine, which has been studied extensively in the containment mockup facilityat 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% iodinedeposition in the containment building, each cubic meter of air released from containment has aradioiodine concentration that is 25% of each cubic meter within containment building air, then theradioiodine concentrations leaking from the containment structure into the laboratory building, inmicrocuries per milliliter, will be:Example calculation of 1311 released through the exhaust stack:-1311 activity / (30,500 ft3/min x 16.5 hr x 60 mir/hr x 28,300 ml/ft3)-9.33 x 10+06 1iCi / 8.55 x 10+11 ml= 1.09 x 10.05 pCi/mi(1.09 x 10-&deg;5 gtCi/ml) x (0.25) =2.73 x 10.06 pCi/mlNote: Same calculation is used for the other radioiodines listed below.Radioiodine Concentrations in Air Leaking from Containment and Exiting the Exhaust Stack13I- 2.73 x 10.06 gtCi/ml 133I -1.55 x 10&deg;5~ gCi/ml 3I-- 1.39 x 10&deg;s~ pCi/ml132I --7.37 x 10.06 giCi/ml 134I -- 1.59 x 10-&deg;s pCi/mlExample calculation of 85Kr released through the exhaust stack:-85Kr activity / (30,500 ft3/min x 16.5 hr x 60 mmn/hr x 28,300 ml/fl3)-2.32 x 10+03 ptCi / 8.55 x 10+1n ml-2.71 x 10.09 jCi/mlNote: Same calculation is used for the other noble gases listed below.Noble Gas Concentrations in Air Leaking from Containment and Exiting the Exhaust Stack85Kr -2.71 x 10.09 jCi/ml 87Kr -1.84 x 10-05 gxCi/ml 89Kr -6.14 x 10.08 jCi/ml8Smjr -1.42 x 10.05 iCi/ml 88Kr -3.00 x 10-05 jCi/ml 9&deg;Kr -7.33 x 10-22 pCi/ml33e- 3.00 x 10-0 pCi/ml 135mXe -5.27 x 10-&deg;6 gxCi/ml 138Xe -1.35 x 10.0 gxCi/ml3Xe- 1.24 x 10.05 pCi/ml 137Xe -2.60 x 10-o7 1iCi/ml 139Xe -9.16 x 10-'9 pCi/mlAssuming, as stated earlier, that (1) the average leakage rate from the containment building is 5.2scfm, (2) the leak continues for about 16.5 hours in order to equalize the containment buildingpressure with atmospheric pressure, (3) the flow rate through the facility's ventilation exhaust stack69 of 86 is 30,500 scfmn, (4) the reduction in concentration from the point of discharge at the exhauststack to the point of maximum concentration in the unrestricted area is a factor of 292 and (5) thereis no decay of any radioiodines or noble gases, then the following concentrations of radioiodinesand noble gases with their corresponding radiation doses will occur in the unrestricted area. Thevalues listed are for the point of maximum concentration in the unrestricted area assuming auniform, semi-spherical cloud geometry for noble gas submersion and further assuming that themost conservative (worst-case) meteorological conditions exist for the entire 16.5-hour period ofcontainment leakage following a FHA. Radiation doses are calculated for the entire 16.5-hourperiod. Dose values for the unrestricted area were obtained using the same methodology that wasused to deternine doses inside the containment building, and it was assumed that an individual waspresent at the point of maximum concentration for the full 16.5 hours that the containment buildingwas leaking.A worst-case scenario dilution factor of 292 for effluent dilution using the Pasquill-Guifford Modelfor atmospheric dilution is used in this analysis. We assume that all offsite (public) dose occursunder these atmospheric conditions at the site of interest, i.e. 760 meters North of MURR. In ourcase at 760 meters it occurs only during Stability Class F conditions; which normally only occur11.4% 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. AnEffluent Concentration Limit of 2.0 x 10.08 iiCi/ml is used for the "unlisted" isotopes, which equalsthe DAC/300 when using the DOE Part 835 Default DAC limits of 6.0 x 10.06 giC/ml. Exposure at1 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 8760hours per year. Thus, the time assumed to be exposed to the effluent concentration limit is a factorof 4.38 longer than the 2000 hours per year that defines a DAC. The isotopes in question are basedon a default DAC limit of 6.0 x 1 006 for short-lived (< 2 hour half-lives) submersion DAC'sin Appendix C of Part 835. No credit is taken for transit time from the stack to the receptor pointnor is credit taken for decay inside containment until release. In the case of Kr-89 and Xe-1 37 thetransit 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 theDAC value to establish the effluent concentration limit is warranted. This reduction factor of 300 isconsistent, in fact more conservative, than 10 CFR 20 Appendix B, as the NRC calculates EffluentConcentration 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-&deg;3 mrem/hr131I concentration = 9.35 x 10.0 pCi/ml131I effluent concentration limit =2.00 x 10-'&deg; pCi/ml1311 Conversion Factor = 5.71 x 10.03 mnrem/hr70 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-'&deg; x 5.71 x 10-03 mremlhr x 16.5 hrs)=4.40 x 10+&deg;&deg; mremNote: Same calculation is used for the other isotopes (radioiodines and noble gases) listed below.Effluent Concentration Limits. Concentrations at Point of Maximum Concentrationand Radiation Doses in the Unrestricted Area -RadioiodineRadionuclide1311I321133113411351Effluent Limit2.00 x 10-1&deg; p.Ci/ml2.00 x 10.08 pCi/ml1.00 X 10-0 jiCi/ml6.00 x 10-0 p.Ci/ml6.00 x 10.09 tCi/mlMaximum Concentratior2.52 x 10.0 pCi/ml5.32 x 10-&deg; jiCi/ml5.46 x 10-0 jiCi/mla' Radiation Dose4.40 x 10+&deg; mrem1.19 x 10-&deg;1 mrem5.01 x 10+00 mrem8.57 x 10-02 mrem7.49 x 10-&deg;1 mremTotal = 10.37 mremNote 1: Maximum Concentrations are radioiodine and noble gas concentrations leaking from the containment buildingand exiting the exhaust stack reduced by a dilution factor of 292.Effluent Concentration Limits, Concentrations at Point of Maximum Concentrationand Radiation Doses in the Unrestricted Area -Noble GasesRadionuclide85Kr87Kr8tKr89Kr90Kr13smXeEffluent Limit7.00 x 10.0 pCi/ml1.00 x 10.07 pCi/mi2.00 x 10.08 xCi/ml9.00 x 10-09 pCi/ml2.00 x 10.08 pCi/mi2.00 x 10-&deg; jxCi/ml5.00 x 10-0 gCi/ml7.00 x 10.08 xCi/ml4.00 x i0.0 gCi/ml2.00 x 10.0 gCi/ml2.00 x 10.08 gCi/ml2.00 x 10-08 jxCi/mlMaximum Concentration19.30 x 10-12 igCi/ml4.85 x 10-&deg;8/.tCi/ml6.29 x 10-08 pCi/ml1.03 x 10-07 pCi/ml2.51 x 10-24 pCi/ml1.03 x 10-0 pCi/mI4.25 x 10-0 pCi/ml1.80 x 10-08 itCi/ml4.61 x 10-0 pCi/ml3.14 x 10-21 p.Ci/mlRadiation Dose1.25 x 10-06 mrem4.57 x 10.02 mrem2.96 x 10.01 mrem1.07 x 10+0&deg; mrem9.91 x 10-04 mrem1.18 x 10-'7mrem1.93 x 10-02nmrem5.72 x 10.02 mrem4.25 x 10.02 mrem4.19 x 10-0 mrem2.17 x 10-&deg;1 mrem1.48 x 10"14 mremTotal = 1.76 mremNote 1: Maximum Concentrations are radioiodine and noble gas concentrations leaking from the containment buildingand 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 dosesfrom the radioijodines and noble gases must be added together, and result in the following values:Dose from Radioidines and Noble Gases in the Unrestricted AreaCommitted Effective Dose Equivalent (Radioiodine) 10.37 mremCommitted Effective Dose Equivalent (Noble Gases) 1.76 mremTotal Effective Dose Equivalent (Whole Body) 12.13 mremSumming the doses from the noble gases and the radioiodines simply substantiates earlierstatements regarding the very low levels in the unrestricted area should a FHA occur, and shouldthe containment building leak following such an event. Because the dose values are so low, thedose from the noble gases becomes the dominant value, but the overall TEDE is still only 12.13mrem, a value far below the applicable 10 CFR 20 regulatory limit for the unrestricted area.72 of 86  
: 9. NUREG-1537, Section 13.1.5, "Mishandling or Malfunction of Fuel" provides guidance that thelicensee analyze the consequences of a mishandled fuel event. MURR SAR Section 13.2.5.2.1describes damage to a fuel element due to mishandling. It states that the mishandling could occurduring movement and packaging of the irradiated fuel, damage could only occur to the inner or theouter fuel plate, and could only occur during fuel element relocation activities. Because thisaccident 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 anoccurrence assuming that the fuel element has decayed for 60 days as part of the spent fuelmovement from storage to a shipping container. However, the NRC staff questions whether thisevent could also occur during the initial stages of refueling which would invalidate the assumptionof 60 days of decay. The NRC staff also performed a confirmatory calculation based on thisinventory using the cited values for the MHA analysis, and it results in an inventory that is sevenpercent larger than reported by MURR.a. Explain the possibility of this event occurring during the initial stages of refueling, and theapplicability of using the stated decay time in the dose calculation. Also, describe anyradioactivity release alarms that are expected to actuate, and whether containment isolationis expected, including the time required to verify containment isolation, or justify why noadditional 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 andmethodologies as used in the Maximum Hypothetical Accident (MHA) (now referred to as the"Fuel Failure during Reactor Operation" accident) and Fueled Experiment Failure. The onlyexceptions are the source term, which is explained in the accident analysis, as well as the decayprior to the accident (which is once again explained in the analysis). As discussed in the responseto RAI 10O.a, the primary coolant system does not have to be secured for a failed fueled experimentor for a FHA. The only required action for Operations personnel is to verify that the containmentbuilding has been evacuated following a containment building isolation, which will occur duringboth of these accident scenarios. MUJIRR performs an evacuation dr-ill every year and the typicaltime period for all personal to evacuate the containment building, including verification byOperations personnel, is two (2) to two and a half (2.5) minutes. For the purposes of the failedfueled experiment and FRA calculations, a conservative assumption of five (5) minutes is used forboth accident scenarios. Additionally, verifying that the reactor has shut down and containment hasisolated only takes a few moments -all control blade positions, reactor power meters, andcontainment isolation valve and door indications are in clear view of the reactor operator in thecontrol room.b. Provide the details of how the source term is determined, or justify' why no additionalinformation is needed.As described in the FHA analysis above, the two most outer fuel plates of a fuel element, number-land -24, are the plates most likely to be damaged during fuel handling. The number-i fuel platecontains 19.26 grams of U-235 before irradiation. The highest peak power density in the various59 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 ofthe fuel plate. The number-24 fuel plate has a lower peak power density and contains 45.32 gramsof U-235, and has the most surface area to be damaged. To be conservative, the analysis assumesthat 0.125 grams of U-235 is exposed from plate number-i during the FHA, which corresponds toremoving a section of fuel meat from a plate that is 1 inch square and 5 mils thick. A powerpeaking 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 andAccounting Procedures as outlined in the Operations Procedures. Irradiated fuel is handled with aspecially 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 fuelhandling sequence. Fuel elements are always handled one at a time so that they are maintained in acriticality-safe configuration. New or irradiated fuel may be stored in any one of 88 in-pool fuelstorage locations (not including the core). These storage locations are designed to ensure ageometry such that the calculated Keff is less than 0.9 under all conditions of moderation, thusallowing sufficient convection cooling and providing sufficient radiation shielding.So the fuel handling system provides a safe, effective and reliable means of transporting andhandling 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 andmagnetic particle testing, as appropriate. Therefore, no specific accidents regarding the handling offuel have been identified for the MUIRR. The probability of dropping a fuel element whileunderwater and damaging it severely enough to breach the fuel cladding was considered. Aconservative potential radionuclide release and calculation of the occupational exposure areincluded below.The following calculations determining the postulated dose from a potential release of radioactivityfrom a fuel element during a handling accident closely follow the "Fuel Failure during ReactorOperation" calculations for personal exposure due to a release of fission products. The objectiveof these calculations is to present a worst-case dose assessment for a person who remains in thecontainment 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 ofelements. Paired elements are always loaded opposite each other in the core. All eight (8) fuelelements are replaced every refueling. MURR has averaged refueling the core more than 52 timesa year since 1977. This type of accident has never occurred at MIJRR during any of these fuelhandlings.The two outer fuel plates of a fuel element, number-i and -24, are the plates most likely to bedamaged during fuel handling. The number-i fuel plate contains 19.26 grams of U-235 beforeirradiation. The highest peak power density in the various MURR core configurations occurs infuel plate number-i of a previously unirradiated fuel element, which has a power peaking factor of4.116 -located between 14.75 to 15.75 inches down from the top of the fuel plate. The number-24fuel plate has the most surface area to be damaged; however, it has a lower peak power density andcontains 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 offuel meat from a plate that is 1 inch square and 5 mils thick. A power peaking factor of 4.116 isalso applied.61 of 86 The following radioiodine, krypton and xenon activities will be present in the MURR core 30minutes after shutdown from 10 MW full power operation. Refuelings typically occur no soonerthan an hour after shutdown. This takes into account the time required to shut down the reactor, tosecure the primary coolant system (required to stay in operation a minimum of 15 minutes after thecontrol blades are fully inserted), and to remove the reactor pressure vessel head. For the purposeof the FHA calculations, a conservative assumption of 30 minutes is used.Radioiodine and Noble Gas Activities in the Core after 30-Minute Decay131I -9.93 x 10+04 Ci 85Kr -2.47 x 10+01 Ci 133Xe -2.73 x 10+05 Ci132I -2.68 x 10+o Ci 85mKr -1.29 x 10+05 Ci 135Xe -1.13 x i0o Ci1331 -- 5.65 x i0+0 Ci 87Kr -1.67 x 10+&deg; Ci 135mXe -- 4.79 x 10+04 Ci134I -- 5.80 X i0+0 Ci 88Kr -2.73 x 10+05 Ci '37Xe -2.37 x 10+0 Ci1351 -- 5.07 x 10+07 Ci 89}r -5.5 x 10+02 Ci 138X -1.22 x 10+05 Ci9&deg;Kr -6.66 x 10-12 Ci 139Xe -8.33 x 10-09 CiFission products released into the reactor pool will be detected by the pool surface and ventilationsystem exhaust plenum radiation monitors. However, for the purposes of this analysis, it isassumed that an actuation of the containment building isolation system occurs by action of the poolsurface radiation monitor. Actuation of the isolation system will prompt Operations personnel toensure that a total evacuation of the containment building is accomplished promptly, usually withintwo (2) to two and a half (2.5) minutes. A conservative 5-minute evacuation time is used as thebasis for the stay time in the dose calculations for personnel that are in containment during theFRA.The following radioiodine and noble gas activities from 0.125 grams of U-235 from the peak powerposition of fuel plate number-i in the peak power density fuel element are assumed toinstantaneously and homogenously distribute in the reactor pool.Example calculation of 1311 released into the reactor pool:= (1311 in fuel / 235U in core) x 235U 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 1iCiExample calculation of 85Kr released into the reactor pool:= (85Kr in fuel / 235U in core) x 235U exposed x PPF x 10+06 iCi/Ci= (2.47 x 10+&deg;' 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 Pool131I -- 9.33 x 10+06 kCi 85Kr -2.32 x 10+03 gCi 133Xe -- 2.56 x 10+07 giCi132I -2.52 x 10+0 giCi 85mKr -1.21 x 10+07 gxCi '35Xe -1.06 x 10+07 /iCi133I -5.31 x 10+07 g.Ci 87Kr -1.57 x 10+0 1iCi l35mXe -4.50 x 10+06 jiCi1341 -- 5.45 X 10+0 gtCi 88Kr -2.56 x 100 jiCi 137Xe -2.22 x 10+05 gCi35-- 4.76 x 10+07 ptCi 9K~r -5.25 x 10+0 ptCi '38Xe -1.15 x 10+07 plCi9&deg;~Kr -6.26 x 10-'&deg; ItCi 139Xe -7.83 x 10-07 pCiThe radioiodine released into the reactor pooi over a 5-minute interval is conservatively assumed tobe instantly and uniformly mixed into the 20,000 gallons (75,708 1) of bulk pooi water, which thenresults in the following pool water concentrations for the radioiodine isotopes. The water solubilityof the krypton and xenon noble gases released into the pool over this same time period are ignoredand they are assumed to pass immediately through the pool water and evolve directly into thecontainment building air volume where they instantaneously form a uniform concentration in theisolated structure.Radioiodine Concentrations in the Pool Water131I -- 4.67 x 10+02 pCilgal 1331 -- 2.66 x 10+03 giCi/gal 135I -2.38 x 10+03 PCi/gal13I -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, theevaporation rate from the reactor pool is approximately 80 gallons (302.8 L) of water per day. Forthe purposes of this calculation, it is assumed that a total of 20 gallons (75.7 L) of pool watercontaining the previously listed radioiodine concentrations evaporates into the containment buildingover the 5 minute period. Containment air with a temperature of 75 0F (23.9 &deg;C) and 100% relativehumidity contains H20 vapor equal to 40 gallons (151.4 L) of water. Since the air in containment isnormally at about 50% relative humidity, thus containing 20 gallons (75.7 L) of water vapor, theassumed addition of 20 gallons (75.7 L) of water vapor will not cause the containment air to besupersaturated. It is also conservatively assumed that all of the radioiodine activity in the 20gallons (75.7 L) of pool water instantaneously forms a uniform concentration in the containmentbuilding air. When distributed into the containment building, this would result in the followingradioiodine concentrations in the 225,000 ft3 (6,371.3 in3) air volume:Example calculation of 131I released into containment air:= 131I concentration in pooi water x 20 gal x 1/225,000 ft3 x 35.3 147 ft3/m3-4.67 x 10+02 ptCi/gal x (3.14 x 10.03 gal/in3)-1.46 pCi/m3(1.46 pCi/in3) x (1 m3/106 ml) =1.46 x 10.06 gxCi/mlNote: 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 theconcentrations after 5 minutes decay divided by 2.Average Radioiodine Concentrations in the Containment Building Air during the 5 Minutes31-- 1.46 x 10-06 pCi/mi 1331 -- 8.32 x 10&deg;06 gtCi/ml 135I -- 7.44 x 10.06 gCi/ml132j -3.91 x 10.06 gtCi/ml 14-- 8.28 x i0-&deg;5 gCi/mlAs 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 andenter the containment building air volume where they instantaneously form a uniform concentrationin the isolated structure. This assumption is extremely conservative since it ignores the knownsolubility of krypton and xenon noble gases in the 100 0F (37.8 &deg;C) pool water, which would reducetheir release into the containment building. Based on the 225,000-ft3 volume of containmentbuilding 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 wouldbe as follows:Example calculation of 85Kr released into containment air:-85Kr activity x 1/225,000 ft3 x 35.3 147 ft3/m3= 2.32 X 10+03 j.Ci x (1.60 x i004 1/in3)= 3.64 x 1001 g.Ci/m3(3.64 x 10&deg;1 gxCi/m3) x (1 m3/106 ml) = 3.64 x i0-07 plei/mlNote: Same calculation is used for the other isotopes listed below.The average noble gas concentrations are the sum of the initial concentrations and theconcentrations after 5 minutes decay divided by 2.Average Noble Gas Concentrations in the Containment Building Air during the 5 MinutesKr -3.64 x 10-07 FCi/ml 133Xe -4.02 x 10-03 pCi/miS5m~x -1.89 x 10-03 gCi/ml '35Xe -1.66 x 1003 jiCi/ml87r- 2.41 x 10.03 gCi/ml 13smXe -6.35 x 10-04 iCi/ml88r- 3.98 x i0-03 jiCi/ml 137Xe -2.45 x i0-&deg;5 gCi/mI89Kr -5.49 x 10-0 gtCi/ml '38Xe -1.61 x 10-&deg;3 pCi/mi9&deg;Kr -4.92 x 10-20 gCi/ml 139Xe -6.18 x10-17 Ci/mlThe objective of this calculation is to present a worst-case dose assessment for an individual whoremains in the containment building for 5 minutes following the ERA. Therefore, as notedpreviously, the radioactivity in the evaporated pool water is assumed to be instantaneously anduniformly 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 thyroidfrom radioiodine and the dose to the whole body resulting from submersion in the airborne noblegases and radioiodine inside the containment building. As previously noted, the exposure time forthis dose assessment is 5 minutes.Because the airborne radioiodine source is composed of five different iodine isotopes, it will benecessary to determine the dose contribution from each individual isotope and to then sum theresults. 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 25mrem/DAC-hr. Additionally, 5 minutes of one DAC-hr is 8.33 x 10-&deg;2 DAC-hr.1311 concentration in containment = 1.46 x 10-o6 g.Ci/ml131j DAC (10 CFR 20) = 2.00 x 10.o gCi/mlDose Multiplication Factor =(1311 concentration) / (131I DAC)= (1.46 x 10-&deg;6 gCi/ml) / (2.00 x 10.o8 gCi/ml)= 73Therefore, 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+&deg;2 mremNote: 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 thesame manner as the radioiodines, and the dose contribution from each individual radionuclide mustbe calculated and then added together to arrive at the final noble gas dose. Because the dose fromthe noble gases is only an external dose due to submersion, and because the DACs for theseradionuclides are based on this type of exposure, the individual noble gas doses for 5 minutes incontaimnment were based on their average concentration in the containment air and thecorresponding 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, 5minutes of one DAC-br is 8.33 x l0&deg;2 DAC-hr.85Kr concentration in containment = 3.64 X 10-07 gtCi/mnl65 of 86 85Kr DAC (10 CFR 20)Dose Multiplication Factor= 1.00 x 10"04 pCi/ml= (85Kr concentration) / (85Kr DAC)= (3.64 x 10-07 ptCi/ml) / (1.00 x 10-04 jiCi/ml)= 0.00364Therefore, 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)= 7.58X10-04 mremNote: 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 -RadioiodineRadionuclide131113211331134j1351Derived Air Concentration2.00 x 10-&deg;8 3.00 x 10-06 /xCi/ml1.00 x 10-07/Ci/ml2.00 x i0.0 gCi/inl7.00 x 10-07 iiCi/ml5-Minute Exposure1.52 x 10+o2 mrem2.71 x 10+00 mrem1.73 x 10+02 mrem8.62 x 10-&deg;1 mrem2.21 x 10+&deg;1 mremTotal = 351.44 mremDerived Air Concentration Values and 5-Minute Exposures -Noble GasesRadionuclide85Kr85m}r87KrS88K89Kr9OKr133XeI35Xe135mXe138Xe139XeDerived Air Concentration1.00 x 10-&deg;4 iCi/ml2.00 x 10.o jtCi/ml5.00 x 100o6 gCi/ml2.00 x 100o6 xtCi/ml6.00 x 10.o6 JiCi/m16.00 x 10-&deg;6 pCi/ml1.00 x 10-04 pCi/mi1.00 x 10-&deg;5 pCi/ml9.00 x 10-06 pCi/ml6.00 x 10-06 gCi/ml4.00 x 10-06 jiCi/ml6.00 X 10-06 .tCi/ml5-Minute Exposure7.58 x 10-04 mrem1.96 x 10+&deg;1 me1.00 x 10+02 mrem4.14 x 10+02 mrem1.91 x 10-&deg;1 mrem1.71 x 10-15 mrem8.36 x 10+00 mrem3.45 x 10+&deg;1 mrem1.47 x 10+&deg;1 mrem8.49 x 10-&deg;1 mrem8.37 x 10+&deg;1mre2.14 x 1012 mremTotal = 676.45 mrem66 of 86 To finalize the occupational dose in terms of Total Effective Dose Equivalent (TEDE) for a5-minute exposure in the containment building after a FHA, the doses from the radioiodines andnoble gases must be added together, and result in the following values:5-Minute Dose from Radioidines and Noble Gases in the Containment BuildingCommitted Dose Equivalent (Thyroid) 351.44 mremCommitted Effective Dose Equivalent (Thyroid) 10.54 mremCommitted Effective Dose Equivalent (Noble Gases) 676.45 mremTotal Effective Dose Equivalent (Whole Body) 687.00 mremBy comparison of the maximum TEDE and Committed Dose Equivalent (CDE) for thoseoccupationally-exposed during a FHA to applicable NRC dose limits in 10 CFR 20, the final valuesare shown to be well within the published regulatory limits and, in fact, lower than 15% of anyoccupational limit.Radiation shine through the containment structure was also evaluated when considering accidentconditions and dose consequences to the public and MUIRR staff. Calculation of exposure ratefrom a FHA was performed using the computer program MicroShield 8.02 with a RectangularVolume -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 comprehensivephoton/gamma ray shielding and dose assessment program that is widely used by industry fordesigning radiation shields.The exposure rate values provided below represents the radiation fields at 1 foot (30.5 cm) from a12-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 theexposure rate values are identical to those used for determining the dose to a worker withincontainment from noble gases. For radioiodine, the total iodine activity from the FHA was used forthe dose calculations, not the amount that evaporated in 5 minutes. The source term also assumes ahomogenous mixture of nuclides within the containment free volume.Radiation Shine through the Containment BuildingExpo sure Rate at 1-Foot from Containment Building Wall: 54.79 mrem/hrExposure Rate at Emergency Planning Zone Boundary (150 meters): 0.371 mremn/hrA confirmatory analysis of the accident condition yielding the largest consequence was validatedindependently by the use of the MCNP code. This analysis yielded a result 21% less than theMicroshield method and results provided above.As noted earlier in this analysis, the containment building ventilation system will shut down and thebuilding itself will be isolated from the surrounding areas. A FHA will not cause an increase inpressure inside the reactor containment structure; therefore, any air leakage from the building willoccur as a result of normal changes in atmospheric pressure and pressure equilibrium between the67 of 86 inside of the contaimnment structure and the outside atmosphere. It is highly probable that there willbe no pressure differential between the inside of the containment building and the outsideatmosphere, and consequently there will be no air leakage from the building and no radiation doseto members of the public in the unrestricted area. However, to develop what would clearly be aworst-case scenario, this analysis assumes that a barometric pressure change had occurred inconjunction with a FHA. A reasonable assumption would be a pressure change on the order of 0.7inches of Hg (25.4 mm of Hg at 60 which would then create a pressure differential of about0.33 psig (2.28 kPa above atmosphere) between the inside of the isolated containment building andthe 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 ratelimit [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 feetper minute (scfm) as a function of containment pressure can be expressed by the followingequation:LR = 17.85 x (CP-14.7)l"2;where:LR = leakage rate from containment (scfmn); andCP = containment pressure (psia).The minimum Technical Specification free volume of the containment building is 225,000-ft3 atstandard temperature and pressure. At an initial overpressure of 2.0 psig (13.8 kPa aboveatmosphere), 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 airvolume to 230,051 scf. The above equation describes the leakage rate that results in this drop ofcontained air volume over 1,440 minutes (24 hours).When applying the Technical Specification leakage rate equation to the assumed initialoverpressure condition of 0.33 psig (2.28 kPa above atmosphere), it would take approximately 16.5hours 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 periodwould be about 5.2 scfm.Several factors exist that will mitigate the radiological impact of any air leakage from thecontainment building following a FHA. First of all, most leakage pathways from containmentdischarge into the reactor laboratory building, which surrounds the containment structure. Since thelaboratory building ventilation system continues to operate during a FHA, leakage air captured bythe ventilation exhaust system is mixed with other building air, and then discharged from thefacility through the exhaust stack at a rate of approximately 30,500 cflm. Mixing of containment airleakage with the laboratory building ventilation flow, followed by discharge out the exhaust stackand subsequent atmospheric dispersion, results in extremely low radionuclide concentrations andvery small radiation doses in the unrestricted area. A tabulation of these concentrations and doses68 of 86 is given below. These values were calculated following the same methodology stated in Section5.3.3 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 tothe behavior of radioiodine, which has been studied extensively in the containment mockup facilityat 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% iodinedeposition in the containment building, each cubic meter of air released from containment has aradioiodine concentration that is 25% of each cubic meter within containment building air, then theradioiodine concentrations leaking from the containment structure into the laboratory building, inmicrocuries per milliliter, will be:Example calculation of 1311 released through the exhaust stack:-1311 activity / (30,500 ft3/min x 16.5 hr x 60 mir/hr x 28,300 ml/ft3)-9.33 x 10+06 1iCi / 8.55 x 10+11 ml= 1.09 x 10.05 pCi/mi(1.09 x 10-&deg;5 gtCi/ml) x (0.25) =2.73 x 10.06 pCi/mlNote: Same calculation is used for the other radioiodines listed below.Radioiodine Concentrations in Air Leaking from Containment and Exiting the Exhaust Stack13I- 2.73 x 10.06 gtCi/ml 133I -1.55 x 10&deg;5~ gCi/ml 3I-- 1.39 x 10&deg;s~ pCi/ml132I --7.37 x 10.06 giCi/ml 134I -- 1.59 x 10-&deg;s pCi/mlExample calculation of 85Kr released through the exhaust stack:-85Kr activity / (30,500 ft3/min x 16.5 hr x 60 mmn/hr x 28,300 ml/fl3)-2.32 x 10+03 ptCi / 8.55 x 10+1n ml-2.71 x 10.09 jCi/mlNote: Same calculation is used for the other noble gases listed below.Noble Gas Concentrations in Air Leaking from Containment and Exiting the Exhaust Stack85Kr -2.71 x 10.09 jCi/ml 87Kr -1.84 x 10-05 gxCi/ml 89Kr -6.14 x 10.08 jCi/ml8Smjr -1.42 x 10.05 iCi/ml 88Kr -3.00 x 10-05 jCi/ml 9&deg;Kr -7.33 x 10-22 pCi/ml33e- 3.00 x 10-0 pCi/ml 135mXe -5.27 x 10-&deg;6 gxCi/ml 138Xe -1.35 x 10.0 gxCi/ml3Xe- 1.24 x 10.05 pCi/ml 137Xe -2.60 x 10-o7 1iCi/ml 139Xe -9.16 x 10-'9 pCi/mlAssuming, as stated earlier, that (1) the average leakage rate from the containment building is 5.2scfm, (2) the leak continues for about 16.5 hours in order to equalize the containment buildingpressure with atmospheric pressure, (3) the flow rate through the facility's ventilation exhaust stack69 of 86 is 30,500 scfmn, (4) the reduction in concentration from the point of discharge at the exhauststack to the point of maximum concentration in the unrestricted area is a factor of 292 and (5) thereis no decay of any radioiodines or noble gases, then the following concentrations of radioiodinesand noble gases with their corresponding radiation doses will occur in the unrestricted area. Thevalues listed are for the point of maximum concentration in the unrestricted area assuming auniform, semi-spherical cloud geometry for noble gas submersion and further assuming that themost conservative (worst-case) meteorological conditions exist for the entire 16.5-hour period ofcontainment leakage following a FHA. Radiation doses are calculated for the entire 16.5-hourperiod. Dose values for the unrestricted area were obtained using the same methodology that wasused to deternine doses inside the containment building, and it was assumed that an individual waspresent at the point of maximum concentration for the full 16.5 hours that the containment buildingwas leaking.A worst-case scenario dilution factor of 292 for effluent dilution using the Pasquill-Guifford Modelfor atmospheric dilution is used in this analysis. We assume that all offsite (public) dose occursunder these atmospheric conditions at the site of interest, i.e. 760 meters North of MURR. In ourcase at 760 meters it occurs only during Stability Class F conditions; which normally only occur11.4% 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. AnEffluent Concentration Limit of 2.0 x 10.08 iiCi/ml is used for the "unlisted" isotopes, which equalsthe DAC/300 when using the DOE Part 835 Default DAC limits of 6.0 x 10.06 giC/ml. Exposure at1 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 8760hours per year. Thus, the time assumed to be exposed to the effluent concentration limit is a factorof 4.38 longer than the 2000 hours per year that defines a DAC. The isotopes in question are basedon a default DAC limit of 6.0 x 1 006 for short-lived (< 2 hour half-lives) submersion DAC'sin Appendix C of Part 835. No credit is taken for transit time from the stack to the receptor pointnor is credit taken for decay inside containment until release. In the case of Kr-89 and Xe-1 37 thetransit 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 theDAC value to establish the effluent concentration limit is warranted. This reduction factor of 300 isconsistent, in fact more conservative, than 10 CFR 20 Appendix B, as the NRC calculates EffluentConcentration 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-&deg;3 mrem/hr131I concentration = 9.35 x 10.0 pCi/ml131I effluent concentration limit =2.00 x 10-'&deg; pCi/ml1311 Conversion Factor = 5.71 x 10.03 mnrem/hr70 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-'&deg; x 5.71 x 10-03 mremlhr x 16.5 hrs)=4.40 x 10+&deg;&deg; mremNote: Same calculation is used for the other isotopes (radioiodines and noble gases) listed below.Effluent Concentration Limits. Concentrations at Point of Maximum Concentrationand Radiation Doses in the Unrestricted Area -RadioiodineRadionuclide1311I321133113411351Effluent Limit2.00 x 10-1&deg; p.Ci/ml2.00 x 10.08 pCi/ml1.00 X 10-0 jiCi/ml6.00 x 10-0 p.Ci/ml6.00 x 10.09 tCi/mlMaximum Concentratior2.52 x 10.0 pCi/ml5.32 x 10-&deg; jiCi/ml5.46 x 10-0 jiCi/mla' Radiation Dose4.40 x 10+&deg; mrem1.19 x 10-&deg;1 mrem5.01 x 10+00 mrem8.57 x 10-02 mrem7.49 x 10-&deg;1 mremTotal = 10.37 mremNote 1: Maximum Concentrations are radioiodine and noble gas concentrations leaking from the containment buildingand exiting the exhaust stack reduced by a dilution factor of 292.Effluent Concentration Limits, Concentrations at Point of Maximum Concentrationand Radiation Doses in the Unrestricted Area -Noble GasesRadionuclide85Kr87Kr8tKr89Kr90Kr13smXeEffluent Limit7.00 x 10.0 pCi/ml1.00 x 10.07 pCi/mi2.00 x 10.08 xCi/ml9.00 x 10-09 pCi/ml2.00 x 10.08 pCi/mi2.00 x 10-&deg; jxCi/ml5.00 x 10-0 gCi/ml7.00 x 10.08 xCi/ml4.00 x i0.0 gCi/ml2.00 x 10.0 gCi/ml2.00 x 10.08 gCi/ml2.00 x 10-08 jxCi/mlMaximum Concentration19.30 x 10-12 igCi/ml4.85 x 10-&deg;8/.tCi/ml6.29 x 10-08 pCi/ml1.03 x 10-07 pCi/ml2.51 x 10-24 pCi/ml1.03 x 10-0 pCi/mI4.25 x 10-0 pCi/ml1.80 x 10-08 itCi/ml4.61 x 10-0 pCi/ml3.14 x 10-21 p.Ci/mlRadiation Dose1.25 x 10-06 mrem4.57 x 10.02 mrem2.96 x 10.01 mrem1.07 x 10+0&deg; mrem9.91 x 10-04 mrem1.18 x 10-'7mrem1.93 x 10-02nmrem5.72 x 10.02 mrem4.25 x 10.02 mrem4.19 x 10-0 mrem2.17 x 10-&deg;1 mrem1.48 x 10"14 mremTotal = 1.76 mremNote 1: Maximum Concentrations are radioiodine and noble gas concentrations leaking from the containment buildingand 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 dosesfrom the radioijodines and noble gases must be added together, and result in the following values:Dose from Radioidines and Noble Gases in the Unrestricted AreaCommitted Effective Dose Equivalent (Radioiodine) 10.37 mremCommitted Effective Dose Equivalent (Noble Gases) 1.76 mremTotal Effective Dose Equivalent (Whole Body) 12.13 mremSumming the doses from the noble gases and the radioiodines simply substantiates earlierstatements regarding the very low levels in the unrestricted area should a FHA occur, and shouldthe containment building leak following such an event. Because the dose values are so low, thedose from the noble gases becomes the dominant value, but the overall TEDE is still only 12.13mrem, 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 licenseesanalyze the consequences of a failed fueled experiment. SAR Section 13.2.6.2 describes that limitingfueled experiments to 150 curies of radioiodine will result in a projected dose well within the limitsof 10 O CFR Part 20. The response to RAJ 13.9.a (ADAMS Accession No. ML103060018) providesradioiodine and noble gas activities for a 5-gram low-enriched fuel target. The response uses amethod similar to that used in the MHA analysis and lists the gaseous fission products to bereleased into the pool cooling system. The occupational dose calculation assumes a 2-minuteevacuation time. The NRC staff notes that the submersion dose calculations were performed usingthe DAC values, but the DAC data for isotopes with half-lives of less than 2 hours that are notlisted 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-minuteevacuation time assumed in the MHA analysis, or the SAR Section 13.2.1.2 statement that it takesthe operations staff approximately 5 minutes to secure the PCS and verify containment isolationfollowing a containment isolation signal.a. Please clarify the sequence of events, state which alarms are expected to provide indicationthat evacuation is required, justify' the evacuation time, and use that time to revise the doseassessment employing consistent DAC values, or justify why no additional information isneeded.Following the response to RAI 10O.c is the revised fueled experiment failure analysis that replacesthe 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). Aspreviously 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 takeapproximately 5 minutes for Operations personnel to secure the primary coolant system and verifythat the containment building has been evacuated following a containment building isolation. Forthe 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 fueledexperiment or for a fuel handling accident (FHA). The only required action for Operationspersonnel is to verify that the containment building has been evacuated following a containmentbuilding isolation, which will occur during both of these accidents. MURR perforns an evacuationdrill every year and the typical time for all personal to evacuate the containment building, includingverification by Operations personnel, is two (2) to two and a half (2.5) minutes. For the purposesof 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 downand 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 thecontrol room.The Derived Air Concentration (DAC) values used for the dose calculations for each accidentscenario -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 used73 of 86 whereas for the "unlisted" isotopes the DACs of 10 CFR 835 are used (published in the FederalRegister, 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-131through Iodine-135 greater than 1.5 curies or Strontiunm-90 greater than 5 millicuries shallbe vented to the facility ventilation exhaust stack through high efficiency particulate air andcharcoal filters which are continuously monitored for an increase in radiation levels." This isinconsistent with TS 3.8.o which states that a fueled experiment can be encapsulated orvented. C'larfify whether fueled experiments are vented or not and revise the TS if required, orjustify why no additional information is needed.License Amendment No. 34, issued to MURR on October 10, 2008, by the NRC, revised Technicalcurrent Specification (TS) 3 .6.o (relicensing TS 3.8.o) such that fueled experiments containinginventories 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 meetthe internal pressure design requirements specified in TS 3.6.i. TS 3.6.i states that "Irradiationcontainers to be used in the reactor, in which static pressure will exist or in which a pressurebuildup is predicted, shall be designed and tested for a pressure exceeding the maximum expectedpressure by at least a factor of two (2)."Until then, fueled experiments containing inventories of I-131 through 1-135 greater than 1.5 curiesor inventories of Sr-90 greater than 5 millicuries had to be vented to the facility ventilation exhauststack through high efficiency particulate air (H7EPA) and charcoal filters which were continuouslymonitored for radiation levels.Since Amendment No. 34 was issued after the SAR was submitted in August 2006 as a part ofrelicensing, SAR Section 13.2.6.2 is now outdated. The third bullet on page 13-67 should nowread, "Fueled experiments containing inventories of iodine-13 1 through iodine-i135 greater than 1.5curies or strontium-90 greater than 5 millicuries shall be in irradiation containers that satisfy therequirements of Specification 3.8 .i or be vented to the facility ventilation exhaust stack throughhigh efficiency particulate air (HEPA) and charcoal filters which are continuously monitored for anincrease in radiation levels."c. If such venting is permitted then explain why those contributions are not included in theinventory of normally released material ('such as Ar-41,), or justify why no additionalinformation 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 datedJune 18, 2015), all air exiting the facility through the ventilation exhaust system is monitored forairborne 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 thecontainment 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 radiationmonitoring 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 theventilation exhaust system; therefore, Ar-4 1 was used to determine the radiological impact ofairborne effluents during normal reactor operation. In addition to At-4 1, all other isotopes greaterthan 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 dischargedto the environs beyond the effective control of the licensee as measured at or prior to the point ofsuch release or discharge."~Attachment 13 (also included in the responses, dated July 31, 2015) provides the last 10 years, andaverage, of air releases from the facility per isotope in percentage of the Technical Specificationlimit. As you will note, with the exception of argon-4 1, all other isotopes discharged are less than0.6% 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 asthe basis for the source term for these dose calculations. A 5-gram LEU target irradiated for 150hours (normal weekly operating cycle) at a thermal neutron flux of 1.5 x 10+13 nlcm2-sec willproduce the following radioiodine, krypton and xenon activities (additionally, approximately 1.40 x10+04 jiCi of Strontium-90 will be produced):Radioiodine and Noble Gas Activities in a 5-Gram LEU Target131I -- 8.400 Ci 85Kr -0.002 Ci 133Xe -18.900 Ci1321 -18.600 Ci 85m~r -7.580 Ci 135Xe -13.600 Ci133I -- 39.900 Ci 87Kr -15.400 Ci l35mXe -6.760 Ci134I -- 45.400 Ci 88Kr -21.700 Ci 137Xe -35.800 Ci13I- 37.700 Ci 89Kr -27.740 Ci 138Xe -37.400 Ci9&deg;Kr -27.400 Ci 139Xe -30.700 CiTotal Iodine -150.00 Ci Total Krypton -99.822 Ci Total Xenon -143.160 CiA complete failure of the target is unrealistic for many reasons. The worst that can be expected ispartial melting; however, in order to present a worst-case dose assessment for an individual thatremains in thle containment building following target failure, 100% of the total activity of the targetis assumed to be released into the reactor pool.Fission products released into the reactor pool will be detected by the pool surface and ventilationsystem exhaust plenum radiation monitors. However, for the purposes of this analysis, it isassumed that a reactor scram and actuation of the containment building isolation system occurs byaction of the pool surface radiation monitor. Actuation of the isolation system will promptOperations personnel to ensure that a total evacuation of the containment building is accomplishedpromptly, usually within two (2) to two and a half (2.5) minutes. A conservative 5-minuteevacuation time is used as the basis for the stay time in the dose calculations for personnel that arein containment during target failure.The radioiodine released into the reactor pool over a 5-minute interval is conservatively assumed tobe instantly and uniformly mixed into the 20,000 gallons (75,708 1) of bulk pool water, which thenresults in the following pool water concentrations for the radioiodine isotopes. The water solubilityof the krypton and xenon noble gases released into the pool over this same time period areconservatively ignored and they are assumed to pass immediately through the pool water andevolve directly into the containment building air volume where they instantaneously form a uniformconcentration in the isolated structure.76 of 86 Radioiodine Concentrations in the Pool Water1311 -- 4.20 x 10+02 gCi/gal 31-- 2.00 x 10+0 pCi/gal 1351 -- 1.89 x 10+0 gtCi/gal1321 -- 9.30 x 10+02 pCi/gal 1341 -- 2.27 x 10+03 pCi/galWhen the reactor is at 10 MW and the containment building ventilation system is in operation, theevaporation rate from the reactor pooi is approximately 80 gallons (302.8 L) of water per day. Forthe purposes of this calculation, it is assumed that a total of 20 gallons (75.7 L) of pool watercontaining the previously listed radioiodine concentrations evaporates into the containment buildingover the 5 minute period. Containment air with a temperature of 75 0F (23.9 &deg;C) and 100% relativehumidity contains H20 vapor equal to 40 gallons (151.4 L) of water. Since the air in containment isnormally at about 50% relative humidity, thus containing 20 gallons (75.7 L) of water vapor, theassumed addition of 20 gallons (75.7 L) of water vapor will not cause the containment air to besupersaturated. It is also conservatively assumed that all of the radioiodine activity in the 20gallons (75.7 L) of pool water instantaneously forms a uniform concentration in the containmentbuilding air. When distributed into the containment building, this would result in the followingradioiodine concentrations in the 225,000 ft3 (6,371.3 in3) air volume:Example calculation of 131I released into containment air:=1311 concentration in pool water x 20 gal x 1/225,000 ft3 x 35.3 147 ft3/m3-4.20 x 10+02 pCilgal x (3.14 x 10.03 gal/in3)-1.32 jiCi/m3(1.32 pCi/in3) x (1 m3/106 ml) = 1.32 x 10.06 gCi/mlNote: Same calculation is used for the other isotopes listed below.The average radioiodine concentrations are the sum of the initial concentrations and theconcentrations after 5 minutes decay divided by 2.Average Radioiodine Concentrations in the Containment Building Air during the 5 Minutes1311 -- 1.32 x 10-&deg;6 ixtCi/ml 133I -- 6.26 x 10-&deg;6 giCi/ml 1351 -5.89 x 10-&deg;6 giCi/ml1321 -2.88 x 10-06 pCi/ml 1341 -6.90 x 10-&deg;6 ptCi/mlAs 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 immediatelythrough the pool water and enter the containment building air volume where they instantaneouslyform a uniform concentration in the isolated structure. Based on the 225,000-ft3 volume ofcontainment building air, and the previously listed curie quantities of these gases released into thereactor pool, the maximum noble gas concentrations in the containment structure at the end of 5minutes would be as follows:77 of 86 Example calculation of 85Kr released into containment air:= 85Kr activity x 1/225,000 ft3 x 35.3 147 ft3/m3= 1.71 x 10+0 x (1.60 x 10.0 1/mn3)-2.69 x 10-&deg;1 jiCi/m3(2.69 x 10-&deg;1 iiCi/m3) x (1 m3/106 ml) = 2.69 x 10-07 pCi/mlNote: Same calculation is used for the other isotopes listed below.The average noble gas concentrations are the sum of the initial concentrations and theconcentrations after 5 minutes decay divided by 2.Average Noble Gas Concentrations in the Containment Building Air during the 5 MinutesKr -2.69 x 10-0 pCi/mI 133Xe -2.97 x 10-03 pCi/ml8mr- 1.18 x 10-03 gCi/ml '35Xe -2.13 x 10.03 pCi/ml87Kr -2.36 x 10-0 jiCi/ml l35mXe -9.54 x 10-&deg; ~tCi/ml88Kr -3.37 x 10.0 pCi/mi 137Xe -3.95 x 10-0 iiCi/ml89r- 2.90 x 10.03 xiCi/ml '38Xe -5.23 x 10-0 jiCi/ml9&deg;Kr -2.15 x i0-0 kCi/ml '39Xe -- 2.42 x 10-0 jiCi/mlThe objective of this calculation is to present a worst-case dose assessment for an individual whoremains in the containment building for 5 minutes following target failure. Therefore, as notedpreviously, the radioactivity in the evaporated pool water is assumed to be instantaneously anduniformly 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 thyroidfrom radioiodine and the dose to the whole body resulting from submersion in the airborne noblegases and radioiodine inside the containment building.Because the airborne radioiodine source is composed of five different iodine isotopes, it will benecessary to determine the dose contribution from each individual isotope and to then sum theresults. 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 25mrem/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'311DAC (10 CFR 20) =2.00 x 10-&deg;8 pCi/ml78 of 86 Dose Multiplication Factor= (1311 concentration) /(1311 DAC)= (1.32 x 10-06 pCi/ml) / (2.00 x 10-08 ptCi/ml)= 66Therefore, 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+02 mremNote: Same calculation is used for the other radioiodines listed below.Derived Air Concentration Values and 5-Minute Exposures -RadioiodineRadionuclide13111321133I134I1351Derived Air Concentration2.00 x 10.08 3.00 x 10-06 gCi/ml1.00 X 10-0 giCi/ml2.00 x 10-&deg; iiCi/mnl7.00 X 10-7/Ci/ml5-Minute Exposure1.37 x 10+02 mrem2.00 x 10+&deg;&deg; mrem1.30 x 10+02 mrem7.18 xlO0-1mrem1.75 x 10+01 mremTotal = 287.80 mremDoses from the kryptons and xenons present in the containment building are assessed in much thesame manner as the radioiodines, and the dose contribution from each individual radionuclide mustbe calculated and then added together to arrive at the final noble gas dose. Because the dose fromthe noble gases is only an external dose due to submersion, and because the DACs for theseradionuclides are based on this type of exposure, the individual noble gas doses for 5 minutes incontainment were based on their average concentration in the containment air and thecorresponding DAC.Example calculation of whole body dose due to KrThe DAC can also be defined as 5,000 mrem/2,000 hrs, or 2.5 mremiDAC-hr. Additionally, 5minutes of one DAC-hr is 8.33 x 10.02 DAC-hr.85Kr concentration in containment85Kr DAC (10 CFR 20)Dose Multiplication Factor= 2.69 X 10.07 pxCi/ml= 1.00 x 10.04 pxCi/ml= (85Kr concentration) / (85Kr DAC)= (2.69 X 10-07 g.Ci/ml) / (1.00 x 10.04 pCi/ml)= 0.00269Therefore, a 5 minute whole body exposure from 85Kr 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  
: 10. NUREG-153 7, Section 13.1.6, "Experiment Malfunction" provides guidance that the licenseesanalyze the consequences of a failed fueled experiment. SAR Section 13.2.6.2 describes that limitingfueled experiments to 150 curies of radioiodine will result in a projected dose well within the limitsof 10 O CFR Part 20. The response to RAJ 13.9.a (ADAMS Accession No. ML103060018) providesradioiodine and noble gas activities for a 5-gram low-enriched fuel target. The response uses amethod similar to that used in the MHA analysis and lists the gaseous fission products to bereleased into the pool cooling system. The occupational dose calculation assumes a 2-minuteevacuation time. The NRC staff notes that the submersion dose calculations were performed usingthe DAC values, but the DAC data for isotopes with half-lives of less than 2 hours that are notlisted 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-minuteevacuation time assumed in the MHA analysis, or the SAR Section 13.2.1.2 statement that it takesthe operations staff approximately 5 minutes to secure the PCS and verify containment isolationfollowing a containment isolation signal.a. Please clarify the sequence of events, state which alarms are expected to provide indicationthat evacuation is required, justify' the evacuation time, and use that time to revise the doseassessment employing consistent DAC values, or justify why no additional information isneeded.Following the response to RAI 10O.c is the revised fueled experiment failure analysis that replacesthe 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). Aspreviously 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 takeapproximately 5 minutes for Operations personnel to secure the primary coolant system and verifythat the containment building has been evacuated following a containment building isolation. Forthe 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 fueledexperiment or for a fuel handling accident (FHA). The only required action for Operationspersonnel is to verify that the containment building has been evacuated following a containmentbuilding isolation, which will occur during both of these accidents. MURR perforns an evacuationdrill every year and the typical time for all personal to evacuate the containment building, includingverification by Operations personnel, is two (2) to two and a half (2.5) minutes. For the purposesof 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 downand 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 thecontrol room.The Derived Air Concentration (DAC) values used for the dose calculations for each accidentscenario -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 used73 of 86 whereas for the "unlisted" isotopes the DACs of 10 CFR 835 are used (published in the FederalRegister, 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-131through Iodine-135 greater than 1.5 curies or Strontiunm-90 greater than 5 millicuries shallbe vented to the facility ventilation exhaust stack through high efficiency particulate air andcharcoal filters which are continuously monitored for an increase in radiation levels." This isinconsistent with TS 3.8.o which states that a fueled experiment can be encapsulated orvented. C'larfify whether fueled experiments are vented or not and revise the TS if required, orjustify why no additional information is needed.License Amendment No. 34, issued to MURR on October 10, 2008, by the NRC, revised Technicalcurrent Specification (TS) 3 .6.o (relicensing TS 3.8.o) such that fueled experiments containinginventories 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 meetthe internal pressure design requirements specified in TS 3.6.i. TS 3.6.i states that "Irradiationcontainers to be used in the reactor, in which static pressure will exist or in which a pressurebuildup is predicted, shall be designed and tested for a pressure exceeding the maximum expectedpressure by at least a factor of two (2)."Until then, fueled experiments containing inventories of I-131 through 1-135 greater than 1.5 curiesor inventories of Sr-90 greater than 5 millicuries had to be vented to the facility ventilation exhauststack through high efficiency particulate air (H7EPA) and charcoal filters which were continuouslymonitored for radiation levels.Since Amendment No. 34 was issued after the SAR was submitted in August 2006 as a part ofrelicensing, SAR Section 13.2.6.2 is now outdated. The third bullet on page 13-67 should nowread, "Fueled experiments containing inventories of iodine-13 1 through iodine-i135 greater than 1.5curies or strontium-90 greater than 5 millicuries shall be in irradiation containers that satisfy therequirements of Specification 3.8 .i or be vented to the facility ventilation exhaust stack throughhigh efficiency particulate air (HEPA) and charcoal filters which are continuously monitored for anincrease in radiation levels."c. If such venting is permitted then explain why those contributions are not included in theinventory of normally released material ('such as Ar-41,), or justify why no additionalinformation 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 datedJune 18, 2015), all air exiting the facility through the ventilation exhaust system is monitored forairborne 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 thecontainment 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 radiationmonitoring 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 theventilation exhaust system; therefore, Ar-4 1 was used to determine the radiological impact ofairborne effluents during normal reactor operation. In addition to At-4 1, all other isotopes greaterthan 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 dischargedto the environs beyond the effective control of the licensee as measured at or prior to the point ofsuch release or discharge."~Attachment 13 (also included in the responses, dated July 31, 2015) provides the last 10 years, andaverage, of air releases from the facility per isotope in percentage of the Technical Specificationlimit. As you will note, with the exception of argon-4 1, all other isotopes discharged are less than0.6% 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 asthe basis for the source term for these dose calculations. A 5-gram LEU target irradiated for 150hours (normal weekly operating cycle) at a thermal neutron flux of 1.5 x 10+13 nlcm2-sec willproduce the following radioiodine, krypton and xenon activities (additionally, approximately 1.40 x10+04 jiCi of Strontium-90 will be produced):Radioiodine and Noble Gas Activities in a 5-Gram LEU Target131I -- 8.400 Ci 85Kr -0.002 Ci 133Xe -18.900 Ci1321 -18.600 Ci 85m~r -7.580 Ci 135Xe -13.600 Ci133I -- 39.900 Ci 87Kr -15.400 Ci l35mXe -6.760 Ci134I -- 45.400 Ci 88Kr -21.700 Ci 137Xe -35.800 Ci13I- 37.700 Ci 89Kr -27.740 Ci 138Xe -37.400 Ci9&deg;Kr -27.400 Ci 139Xe -30.700 CiTotal Iodine -150.00 Ci Total Krypton -99.822 Ci Total Xenon -143.160 CiA complete failure of the target is unrealistic for many reasons. The worst that can be expected ispartial melting; however, in order to present a worst-case dose assessment for an individual thatremains in thle containment building following target failure, 100% of the total activity of the targetis assumed to be released into the reactor pool.Fission products released into the reactor pool will be detected by the pool surface and ventilationsystem exhaust plenum radiation monitors. However, for the purposes of this analysis, it isassumed that a reactor scram and actuation of the containment building isolation system occurs byaction of the pool surface radiation monitor. Actuation of the isolation system will promptOperations personnel to ensure that a total evacuation of the containment building is accomplishedpromptly, usually within two (2) to two and a half (2.5) minutes. A conservative 5-minuteevacuation time is used as the basis for the stay time in the dose calculations for personnel that arein containment during target failure.The radioiodine released into the reactor pool over a 5-minute interval is conservatively assumed tobe instantly and uniformly mixed into the 20,000 gallons (75,708 1) of bulk pool water, which thenresults in the following pool water concentrations for the radioiodine isotopes. The water solubilityof the krypton and xenon noble gases released into the pool over this same time period areconservatively ignored and they are assumed to pass immediately through the pool water andevolve directly into the containment building air volume where they instantaneously form a uniformconcentration in the isolated structure.76 of 86 Radioiodine Concentrations in the Pool Water1311 -- 4.20 x 10+02 gCi/gal 31-- 2.00 x 10+0 pCi/gal 1351 -- 1.89 x 10+0 gtCi/gal1321 -- 9.30 x 10+02 pCi/gal 1341 -- 2.27 x 10+03 pCi/galWhen the reactor is at 10 MW and the containment building ventilation system is in operation, theevaporation rate from the reactor pooi is approximately 80 gallons (302.8 L) of water per day. Forthe purposes of this calculation, it is assumed that a total of 20 gallons (75.7 L) of pool watercontaining the previously listed radioiodine concentrations evaporates into the containment buildingover the 5 minute period. Containment air with a temperature of 75 0F (23.9 &deg;C) and 100% relativehumidity contains H20 vapor equal to 40 gallons (151.4 L) of water. Since the air in containment isnormally at about 50% relative humidity, thus containing 20 gallons (75.7 L) of water vapor, theassumed addition of 20 gallons (75.7 L) of water vapor will not cause the containment air to besupersaturated. It is also conservatively assumed that all of the radioiodine activity in the 20gallons (75.7 L) of pool water instantaneously forms a uniform concentration in the containmentbuilding air. When distributed into the containment building, this would result in the followingradioiodine concentrations in the 225,000 ft3 (6,371.3 in3) air volume:Example calculation of 131I released into containment air:=1311 concentration in pool water x 20 gal x 1/225,000 ft3 x 35.3 147 ft3/m3-4.20 x 10+02 pCilgal x (3.14 x 10.03 gal/in3)-1.32 jiCi/m3(1.32 pCi/in3) x (1 m3/106 ml) = 1.32 x 10.06 gCi/mlNote: Same calculation is used for the other isotopes listed below.The average radioiodine concentrations are the sum of the initial concentrations and theconcentrations after 5 minutes decay divided by 2.Average Radioiodine Concentrations in the Containment Building Air during the 5 Minutes1311 -- 1.32 x 10-&deg;6 ixtCi/ml 133I -- 6.26 x 10-&deg;6 giCi/ml 1351 -5.89 x 10-&deg;6 giCi/ml1321 -2.88 x 10-06 pCi/ml 1341 -6.90 x 10-&deg;6 ptCi/mlAs 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 immediatelythrough the pool water and enter the containment building air volume where they instantaneouslyform a uniform concentration in the isolated structure. Based on the 225,000-ft3 volume ofcontainment building air, and the previously listed curie quantities of these gases released into thereactor pool, the maximum noble gas concentrations in the containment structure at the end of 5minutes would be as follows:77 of 86 Example calculation of 85Kr released into containment air:= 85Kr activity x 1/225,000 ft3 x 35.3 147 ft3/m3= 1.71 x 10+0 x (1.60 x 10.0 1/mn3)-2.69 x 10-&deg;1 jiCi/m3(2.69 x 10-&deg;1 iiCi/m3) x (1 m3/106 ml) = 2.69 x 10-07 pCi/mlNote: Same calculation is used for the other isotopes listed below.The average noble gas concentrations are the sum of the initial concentrations and theconcentrations after 5 minutes decay divided by 2.Average Noble Gas Concentrations in the Containment Building Air during the 5 MinutesKr -2.69 x 10-0 pCi/mI 133Xe -2.97 x 10-03 pCi/ml8mr- 1.18 x 10-03 gCi/ml '35Xe -2.13 x 10.03 pCi/ml87Kr -2.36 x 10-0 jiCi/ml l35mXe -9.54 x 10-&deg; ~tCi/ml88Kr -3.37 x 10.0 pCi/mi 137Xe -3.95 x 10-0 iiCi/ml89r- 2.90 x 10.03 xiCi/ml '38Xe -5.23 x 10-0 jiCi/ml9&deg;Kr -2.15 x i0-0 kCi/ml '39Xe -- 2.42 x 10-0 jiCi/mlThe objective of this calculation is to present a worst-case dose assessment for an individual whoremains in the containment building for 5 minutes following target failure. Therefore, as notedpreviously, the radioactivity in the evaporated pool water is assumed to be instantaneously anduniformly 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 thyroidfrom radioiodine and the dose to the whole body resulting from submersion in the airborne noblegases and radioiodine inside the containment building.Because the airborne radioiodine source is composed of five different iodine isotopes, it will benecessary to determine the dose contribution from each individual isotope and to then sum theresults. 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 25mrem/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'311DAC (10 CFR 20) =2.00 x 10-&deg;8 pCi/ml78 of 86 Dose Multiplication Factor= (1311 concentration) /(1311 DAC)= (1.32 x 10-06 pCi/ml) / (2.00 x 10-08 ptCi/ml)= 66Therefore, 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+02 mremNote: Same calculation is used for the other radioiodines listed below.Derived Air Concentration Values and 5-Minute Exposures -RadioiodineRadionuclide13111321133I134I1351Derived Air Concentration2.00 x 10.08 3.00 x 10-06 gCi/ml1.00 X 10-0 giCi/ml2.00 x 10-&deg; iiCi/mnl7.00 X 10-7/Ci/ml5-Minute Exposure1.37 x 10+02 mrem2.00 x 10+&deg;&deg; mrem1.30 x 10+02 mrem7.18 xlO0-1mrem1.75 x 10+01 mremTotal = 287.80 mremDoses from the kryptons and xenons present in the containment building are assessed in much thesame manner as the radioiodines, and the dose contribution from each individual radionuclide mustbe calculated and then added together to arrive at the final noble gas dose. Because the dose fromthe noble gases is only an external dose due to submersion, and because the DACs for theseradionuclides are based on this type of exposure, the individual noble gas doses for 5 minutes incontainment were based on their average concentration in the containment air and thecorresponding DAC.Example calculation of whole body dose due to KrThe DAC can also be defined as 5,000 mrem/2,000 hrs, or 2.5 mremiDAC-hr. Additionally, 5minutes of one DAC-hr is 8.33 x 10.02 DAC-hr.85Kr concentration in containment85Kr DAC (10 CFR 20)Dose Multiplication Factor= 2.69 X 10.07 pxCi/ml= 1.00 x 10.04 pxCi/ml= (85Kr concentration) / (85Kr DAC)= (2.69 X 10-07 g.Ci/ml) / (1.00 x 10.04 pCi/ml)= 0.00269Therefore, a 5 minute whole body exposure from 85Kr 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-&deg;4mremNote: Same calculation is used for the other noble gases listed below.Derived Air Concentration Values and 5-Minute Exposures -Noble GasesRadionuclide85Kr85m~r87Kr88Kr89Kr90Kr133mXe135Xe'39XeDerived Air Concentration1.00 x 10"0 pCi/ml2.00 x 10.05 pCi/ml5.00 x 10.06 pCi/ml2.00 x 10-06 gCi/ml6.00 x 10-&deg;6 pCi/ml6.00 x 10-06 pCi/ml1.00 X 10-04 pCi/ml1.00 x i0-05 pCi/ml9.00 x 10-06 pCi/ml6.00 x 10-06 pCi/mi4.00 x 10-&deg;6 jiCi/ml6.00 X 10-&deg;6 ptCi/ml5-Minute Exposure5.59 x 10"0 mrem1.23 x 10+&deg;a mrem9.85 x 10+&deg;1 mrem3.51 x 10+02 mrem1.01 x 10+02 mrem7.48 x 10+&deg;1 mrem6.18 x 10+&deg;&deg; mrem4.43 x 1&deg; mrem2.21 x 10+&deg;1 mrem1.37 x 10+02 mrem2.72 x 10+02 mrem8.41 x 10+&deg;1 mremTotal =1203.80 mremTo finalize the occupational dose in terms of Total Effective Dose Equivalent (TEDE) for a5-minute exposure in the containment building after target failure, the doses from the radioiodinesand noble gases must be added together, and result in the following values:5-Minute Dose from Radioidines and Noble Gases in the Containment BuildingCommitted Dose Equivalent (Thyroid)Committed Effective Dose Equivalent (Thyroid)Committed Effective Dose Equivalent (Noble Gases)Total Effective Dose Equivalent (Whole Body)287.80 mrem8.63 mrem1203.80 mrem1212.44 mremNote: The addition of Strontium-90 (9&deg;Sr) will increase the above stated TEDE (whole body) by9.15 mrem (<1%).By comparison of the maximum TEDE and Committed Dose Equivalent (CDE) for thoseoccupationally-exposed during target failure to applicable NRC dose limits in 10 CFR 20, the finalvalues are shown to be well within the published regulatory limits and, in fact, lower than 25% ofany occupational limit.Radiation shine through the containment structure was also evaluated when considering accidentconditions and dose consequences to the public and MUIRR staff. Calculation of exposure ratefrom the target failure was performed using the computer program MicroShield 8.02 with a80 of 86 Rectangular Volume -External Dose Point geometry for the representation of the containmentstructure (Attachment 12). MicroShield 8.02 is a product of Grove. Software and is acomprehensive photon/gamma ray shielding and dose assessment program that is widely used byindustry for designing radiation shields.The exposure rate values provided below represents the radiation fields at 1 foot (30.5 cm) from a12-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 theexposure rate values are identical to those used for determining the dose to a worker withincontainment from noble gases. For radioiodine, the total iodine activity of the target was used forthe dose calculations, not the amount that evaporated in 5 minutes. The source term also assumes ahomogenous mixture of nuclides within the containment free volume.Radiation Shine through the Containment BuildingExposure Rate at 1-Foot from Containment Building Wall: 68.87 mrem/hrExposure Rate at Emergency Planning Zone Boundary (150 meters): 0.467 mrem/hrA confirmatory analysis of the accident condition yielding the largest consequence was validatedindependently by the use of the MCNP code. This analysis yielded a result 21% less than theMicroshield method and results provided above.As noted earlier in this analysis, the containment building ventilation system will shut down and thebuilding itself will be isolated from the surrounding areas. Target failure will not cause an increasein pressure inside the reactor containment structure; therefore, any air leakage from the buildingwill occur as a result of normal changes in atmospheric pressure and pressure equilibrium betweenthe inside of the containment structure and the outside atmosphere. It is highly probable that therewill be no pressure differential between the inside of the containment building and the outsideatmosphere, and consequently there will be no air leakage from the building and no radiation doseto members of the public in the unrestricted area. However, to develop what would clearly be aworst-case scenario, this analysis assumes that a barometric pressure change had occurred inconjunction with the target failure. A reasonable assumption would be a pressure change on theorder of 0.7 inches of Hig (25.4 mm of Hg at 60 0C), which would then create a pressure differentialof about 0.33 psig (2.28 kPa above atmosphere) between the inside of the isolated containmentbuilding and the inside of the adjacent laboratory building, which surrounds most of thecontainment structure. Making the conservative assumption that the containment building will leakat the Technical Specification leakage rate limit [10% of the contained volume over a 24-hourperiod from an initial overpressure of 2.0 psig (13.8 kPa above atmosphere)], the air leakage fromthe containment structure in standard cubic feet per minute (scfm) as a function of containmentpressure 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); andCP -= containment pressure (psia).The minimum Technical Specification free volume of the containment building is 225,000-ft3 atstandard temperature and pressure. At an initial overpressure of 2.0 psig (13.8 kPa aboveatmosphere), 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 airvolume to 230,051 scf. The above equation describes the leakage rate that results in this drop ofcontained air volume over 1,440 minutes (24 hours).When applying the Technical Specification leakage rate equation to the assumed initialoverpressure condition of 0.33 psig (2.28 kPa above atmosphere), it would take approximately 16.5hours 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 periodwould be about 5.2 scfm.Several factors exist that will mitigate the radiological impact of any air leakage from thecontainment building following target failure. First of all, most leakage pathways fromcontainment discharge into the reactor laboratory building, which surrounds the containmentstructure. Since the laboratory building ventilation system continues to operate during targetfailure, leakage air captured by the ventilation exhaust system is mixed with other building air, andthen 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 bydischarge out the exhaust stack and subsequent atmospheric dispersion, results in extremely lowradionuclide concentrations and very small radiation doses in the unrestricted area. A tabulation ofthese concentrations and doses is given below. These values were calculated following the samemethodology 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 tothe behavior of radioiodine, which has been studied extensively in the containment mock'up facilityat 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% iodinedeposition in the containment building, each cubic meter of air released from containment has aradioiodine concentration that is 25% of each cubic meter within containment building air, then theradioiodine concentrations leaking from the containment structure into the laboratory building, inmicrocuries per milliliter, will be:Example calculation of 1311 released through the exhaust stack:= 13aI activity / (30,500 ft3/min x 16.5 hr x 60 mmn/hr x 28,300 mi/fl3)= 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-&deg;6 jiCi/ml82 of 86 Note: Same calculation is used for the other radioiodines listed below.Radioiodine Concentrations in Air Leaking from Containment and Exiting the Exhaust Stack1311 -2.46 x 10-o6 gtCi/ml 133I -- 1.17 x 10.05 ptCi/mi 35-- 1.10 x 10.05 jiCi/ml132I -- 5.44 x 10.0 jiCi/ml 14-- 1.33 x 10-&deg;5 gCi/mlExample calculation of 85Kr released through the exhaust stack:= 85Kr activity / (30,500 ft3/min x 16.5 hr x 60 mir/hr x 28,300 ml/ft3)= 1.71 x 10+03 gtCi / 8.55 x 10+11 ml-2.00 x 10.o9 pCi/miNote: Same calculation is used for the other noble gases listed below.Noble Gas Concentrations in Air Leaking from Containment and Exiting the Exhaust Stack8SKr -2.00 x 10.09 pCi/ml 87Kr -1.80 x 10-&deg; pCi/ml 89Kr -3.25 x 10-05 pCi/ml85mJr -8.87 x 10.o6 gCi/ml 88Kr -2.54 x 10.o pCi/ml 9&deg;'Kr -3.21 x 10-o5 jiCi/mI3Xe- 2.21 x 100 gCi/ml l35mXe -7.91 x 10.06 pCi/ml '38Xe -4.38 x 10.o pCi/ml'35Xe -1.59 x 10&deg;5 gtCi/ml '37xe -4.19 x 10-&deg; pCi/ml 139Xe -3.59 x 10&deg;05 pCi/miAssuming, as stated earlier, that (1) the average leakage rate from the containment building is 5.2scfmn, (2) the leak continues for about 16.5 hours in order to equalize the containment buildingpressure with atmospheric pressure, (3) the flow rate through the facility's ventilation exhaust stackis 30,500 scfm, (4) the reduction in concentration from the point of discharge at the exhauststack to the point of maximum concentration in the unrestricted area is a factor of 292 and (5) thereis no decay of any radioiodines or noble gases, then the following concentrations of radioiodinesand noble gases with their corresponding radiation doses will occur in the unrestricted area. Thevalues listed are for the point of maximum concentration in the unrestricted area assuming auniform, semi-spherical cloud geometry for noble gas submersion and further assuming that themost conservative (worst-case) meteorological conditions exist for the entire 16.5-hour period ofcontainment 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 thatwas used to determine doses inside the containment building, and it was assumed that an individualwas present at the point of maximum concentration for the full 16.5 hours that the containmentbuilding was leaking.A worst-case scenario dilution factor of 292 for effluent dilution using the Pasquill-Guifford Modelfor atmospheric dilution is used in this analysis. We assume that all offsite (public) dose occursunder these atmospheric conditions at the site of interest, i.e. 760 meters North of MUJRR. In ourcase at 760 meters it occurs only during Stability Class F conditions; which normally only occur11.4% 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. AnEffluent Concentration Limit of 2.0 x 1 0-&deg; pCi/mi is used for the "unlisted" isotopes, which equalsthe DAC/300 when using the DOE Part 835 Default DAC limits of 6.0 x 10-&deg;6 Exposure at1 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 8760hours per year. Thus, the time assumed to be exposed to the effluent concentration limit is a factorof 4.38 longer than the 2000 hours per year that defines a DAC. The isotopes in question are basedon a default DAC limit of 6.0 x 10-o6 for short-lived (< 2 hour half-lives) submersion DAC'sin Appendix C of Part 835. No credit is taken for transit time from the stack to the receptor pointnor is credit taken for decay inside containment until release. In the case of Kr-89 and Xe-i137 thetransit 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 theDAC value to establish the effluent concentration limit is warranted. This reduction factor of 300 isconsistent, in fact more conservative, than 10 CFR 20 Appendix B, as the NRC calculates EffluentConcentration 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 mremihrI31 concentration =8.42 x 10.09 iiCilml131I effluent concentration limit = 2.00 x 10-'&deg; gCi/ml1311 Conversion Factor = 5.71 x 10.0 mrenm/hrTherefore, 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&deg; gCi/ml x 5.71 x i0-0 mremihr x 16.5 brs)=3.96 x 10+&deg;&deg; mremNote: 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 Concentrationand Radiation Doses in the Unrestricted Area -Radio iodineRadionuclide13111321133I13411351Effluent Limit2.00 x 10-10 p.Ci/ml2.00 x 10-&deg; xtCi/ml1.00 x 10-09 pCi/ml6.00 x 10-&deg; gCi/ml6.00 x 10-0 pCi/mlMaximum Concentration18.42 x 10-09 pxCi/ml1.86 x 10-08 pCi/ml4.00 x 10-&deg; pCi/ml4.55 x 10.0 pCi/ml3.78 x 10.08 pCi/mlRadiation Dose3.96 x 10+00 mrem8.78 x 10.02 mrem3.77 x~ 10+00 mrem7.14 x 10-02 mrem5.93 x 10-&deg; mremTotal = 8.48 mremNote 1: Maximum Concentrations are radioiodine and noble gas concentrations leaking from the containment buildingand exiting the exhaust stack reduced by a dilution factor of 292.Effluent Concentration Limits. Concentrations at Point of Maximum Concentrationand Radiation Doses in the Unrestricted Area -Noble GasesRadionuclide85Kr85mIKr57Kr85Kr89Kr9O0r135Xel3smXe139XeEffluent Limit7.00 x 10"07 jiCi/ml1.00 X 10-0 xtCi/ml2.00 x 10-0 pCi/ml9.00 x 10-09 2.00 x 10-08 pCi/ml2.00 x 10-08 ptCi/ml5.00 X 10-07 iiCi/ml7.00 x 10&deg;08 iiCi/ml4.00 x 10-0 ptCi/ml2.00 x 10-0 pCi/ml2.00 x 10.0 iiCi/ml2.00 x 10-08 pCi/mlMaximum Concentration'6.85 x 10-12 pCi/ml3.04 x 10.08 gCi/ml6.17 x 10-08 gtCi/ml8.70 x 10-08 pCi/ml1.11 x 10.07 giCi/ml1.10 x 10.o pCi/mi5.45 x 10-08 pCi/ml2.71 x 10-08 gtCi/ml1.43 x 1 0"07 pCi/ml1.50 X 10.07 ptCi/ml1.23 x 10.07 pCi/mlRadiation Dose9.22 x 10-07 mrem2.86 x 10.02 mrem2.91 x 10.01 mrem9.10 x 10.01 mrem5.24 x 10.01 mrem5.17 x 10-1torero1.43 x 10.02 mrem7.34 x 10.0 mrem6.38 x 10-02 mrem6.76 x 10.01 mrem7.06 x 10.01 mrem5.80 x 10.01 mremTotal = 4.38 mremNote 1: Maximum Concentrations are radioiodine and noble gas concentrations leaking from the containment buildingand 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 dosesfrom the radioiodines and noble gases must be added together, and result in the following values:Dose from Radioidines and Noble Gases in the Unrestricted AreaCommitted Effective Dose Equivalent (Radioiodine)Committed Effective Dose Equivalent (Noble Gases)Total Effective Dose Equivalent (Whole Body)8.48 mrem4.38 mrem12.87 mrem85 of 86 Sunuming the doses from the noble gases and the radioiodines simply substantiates earlierstatements regarding the very low levels in the unrestricted area should a failure of a fueledexperiment occur, and should the containment building leak following such an event. Because thedose values are so low, the dose from the noble gases becomes the dominant value, but the overallTEDE is still only 12.87 mremn, a value far below the applicable 10 CFR 20 regulatory limit for theunrestricted area.References:1Hlazards Summary Report, Addendum 3, Section 5.3.3, University of Missouri Research ReactorFacility, August 1972 (as revised by the 1989-1990 Operations Annual Report).2Hlazards Summary Report, Addendum 4, Appendix C, University of Missouri Research ReactorFacility, October 1973.86 of 86  
= 5.59 x 10-&deg;4mremNote: Same calculation is used for the other noble gases listed below.Derived Air Concentration Values and 5-Minute Exposures -Noble GasesRadionuclide85Kr85m~r87Kr88Kr89Kr90Kr133mXe135Xe'39XeDerived Air Concentration1.00 x 10"0 pCi/ml2.00 x 10.05 pCi/ml5.00 x 10.06 pCi/ml2.00 x 10-06 gCi/ml6.00 x 10-&deg;6 pCi/ml6.00 x 10-06 pCi/ml1.00 X 10-04 pCi/ml1.00 x i0-05 pCi/ml9.00 x 10-06 pCi/ml6.00 x 10-06 pCi/mi4.00 x 10-&deg;6 jiCi/ml6.00 X 10-&deg;6 ptCi/ml5-Minute Exposure5.59 x 10"0 mrem1.23 x 10+&deg;a mrem9.85 x 10+&deg;1 mrem3.51 x 10+02 mrem1.01 x 10+02 mrem7.48 x 10+&deg;1 mrem6.18 x 10+&deg;&deg; mrem4.43 x 1&deg; mrem2.21 x 10+&deg;1 mrem1.37 x 10+02 mrem2.72 x 10+02 mrem8.41 x 10+&deg;1 mremTotal =1203.80 mremTo finalize the occupational dose in terms of Total Effective Dose Equivalent (TEDE) for a5-minute exposure in the containment building after target failure, the doses from the radioiodinesand noble gases must be added together, and result in the following values:5-Minute Dose from Radioidines and Noble Gases in the Containment BuildingCommitted Dose Equivalent (Thyroid)Committed Effective Dose Equivalent (Thyroid)Committed Effective Dose Equivalent (Noble Gases)Total Effective Dose Equivalent (Whole Body)287.80 mrem8.63 mrem1203.80 mrem1212.44 mremNote: The addition of Strontium-90 (9&deg;Sr) will increase the above stated TEDE (whole body) by9.15 mrem (<1%).By comparison of the maximum TEDE and Committed Dose Equivalent (CDE) for thoseoccupationally-exposed during target failure to applicable NRC dose limits in 10 CFR 20, the finalvalues are shown to be well within the published regulatory limits and, in fact, lower than 25% ofany occupational limit.Radiation shine through the containment structure was also evaluated when considering accidentconditions and dose consequences to the public and MUIRR staff. Calculation of exposure ratefrom the target failure was performed using the computer program MicroShield 8.02 with a80 of 86 Rectangular Volume -External Dose Point geometry for the representation of the containmentstructure (Attachment 12). MicroShield 8.02 is a product of Grove. Software and is acomprehensive photon/gamma ray shielding and dose assessment program that is widely used byindustry for designing radiation shields.The exposure rate values provided below represents the radiation fields at 1 foot (30.5 cm) from a12-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 theexposure rate values are identical to those used for determining the dose to a worker withincontainment from noble gases. For radioiodine, the total iodine activity of the target was used forthe dose calculations, not the amount that evaporated in 5 minutes. The source term also assumes ahomogenous mixture of nuclides within the containment free volume.Radiation Shine through the Containment BuildingExposure Rate at 1-Foot from Containment Building Wall: 68.87 mrem/hrExposure Rate at Emergency Planning Zone Boundary (150 meters): 0.467 mrem/hrA confirmatory analysis of the accident condition yielding the largest consequence was validatedindependently by the use of the MCNP code. This analysis yielded a result 21% less than theMicroshield method and results provided above.As noted earlier in this analysis, the containment building ventilation system will shut down and thebuilding itself will be isolated from the surrounding areas. Target failure will not cause an increasein pressure inside the reactor containment structure; therefore, any air leakage from the buildingwill occur as a result of normal changes in atmospheric pressure and pressure equilibrium betweenthe inside of the containment structure and the outside atmosphere. It is highly probable that therewill be no pressure differential between the inside of the containment building and the outsideatmosphere, and consequently there will be no air leakage from the building and no radiation doseto members of the public in the unrestricted area. However, to develop what would clearly be aworst-case scenario, this analysis assumes that a barometric pressure change had occurred inconjunction with the target failure. A reasonable assumption would be a pressure change on theorder of 0.7 inches of Hig (25.4 mm of Hg at 60 0C), which would then create a pressure differentialof about 0.33 psig (2.28 kPa above atmosphere) between the inside of the isolated containmentbuilding and the inside of the adjacent laboratory building, which surrounds most of thecontainment structure. Making the conservative assumption that the containment building will leakat the Technical Specification leakage rate limit [10% of the contained volume over a 24-hourperiod from an initial overpressure of 2.0 psig (13.8 kPa above atmosphere)], the air leakage fromthe containment structure in standard cubic feet per minute (scfm) as a function of containmentpressure 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); andCP -= containment pressure (psia).The minimum Technical Specification free volume of the containment building is 225,000-ft3 atstandard temperature and pressure. At an initial overpressure of 2.0 psig (13.8 kPa aboveatmosphere), 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 airvolume to 230,051 scf. The above equation describes the leakage rate that results in this drop ofcontained air volume over 1,440 minutes (24 hours).When applying the Technical Specification leakage rate equation to the assumed initialoverpressure condition of 0.33 psig (2.28 kPa above atmosphere), it would take approximately 16.5hours 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 periodwould be about 5.2 scfm.Several factors exist that will mitigate the radiological impact of any air leakage from thecontainment building following target failure. First of all, most leakage pathways fromcontainment discharge into the reactor laboratory building, which surrounds the containmentstructure. Since the laboratory building ventilation system continues to operate during targetfailure, leakage air captured by the ventilation exhaust system is mixed with other building air, andthen 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 bydischarge out the exhaust stack and subsequent atmospheric dispersion, results in extremely lowradionuclide concentrations and very small radiation doses in the unrestricted area. A tabulation ofthese concentrations and doses is given below. These values were calculated following the samemethodology 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 tothe behavior of radioiodine, which has been studied extensively in the containment mock'up facilityat 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% iodinedeposition in the containment building, each cubic meter of air released from containment has aradioiodine concentration that is 25% of each cubic meter within containment building air, then theradioiodine concentrations leaking from the containment structure into the laboratory building, inmicrocuries per milliliter, will be:Example calculation of 1311 released through the exhaust stack:= 13aI activity / (30,500 ft3/min x 16.5 hr x 60 mmn/hr x 28,300 mi/fl3)= 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-&deg;6 jiCi/ml82 of 86 Note: Same calculation is used for the other radioiodines listed below.Radioiodine Concentrations in Air Leaking from Containment and Exiting the Exhaust Stack1311 -2.46 x 10-o6 gtCi/ml 133I -- 1.17 x 10.05 ptCi/mi 35-- 1.10 x 10.05 jiCi/ml132I -- 5.44 x 10.0 jiCi/ml 14-- 1.33 x 10-&deg;5 gCi/mlExample calculation of 85Kr released through the exhaust stack:= 85Kr activity / (30,500 ft3/min x 16.5 hr x 60 mir/hr x 28,300 ml/ft3)= 1.71 x 10+03 gtCi / 8.55 x 10+11 ml-2.00 x 10.o9 pCi/miNote: Same calculation is used for the other noble gases listed below.Noble Gas Concentrations in Air Leaking from Containment and Exiting the Exhaust Stack8SKr -2.00 x 10.09 pCi/ml 87Kr -1.80 x 10-&deg; pCi/ml 89Kr -3.25 x 10-05 pCi/ml85mJr -8.87 x 10.o6 gCi/ml 88Kr -2.54 x 10.o pCi/ml 9&deg;'Kr -3.21 x 10-o5 jiCi/mI3Xe- 2.21 x 100 gCi/ml l35mXe -7.91 x 10.06 pCi/ml '38Xe -4.38 x 10.o pCi/ml'35Xe -1.59 x 10&deg;5 gtCi/ml '37xe -4.19 x 10-&deg; pCi/ml 139Xe -3.59 x 10&deg;05 pCi/miAssuming, as stated earlier, that (1) the average leakage rate from the containment building is 5.2scfmn, (2) the leak continues for about 16.5 hours in order to equalize the containment buildingpressure with atmospheric pressure, (3) the flow rate through the facility's ventilation exhaust stackis 30,500 scfm, (4) the reduction in concentration from the point of discharge at the exhauststack to the point of maximum concentration in the unrestricted area is a factor of 292 and (5) thereis no decay of any radioiodines or noble gases, then the following concentrations of radioiodinesand noble gases with their corresponding radiation doses will occur in the unrestricted area. Thevalues listed are for the point of maximum concentration in the unrestricted area assuming auniform, semi-spherical cloud geometry for noble gas submersion and further assuming that themost conservative (worst-case) meteorological conditions exist for the entire 16.5-hour period ofcontainment 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 thatwas used to determine doses inside the containment building, and it was assumed that an individualwas present at the point of maximum concentration for the full 16.5 hours that the containmentbuilding was leaking.A worst-case scenario dilution factor of 292 for effluent dilution using the Pasquill-Guifford Modelfor atmospheric dilution is used in this analysis. We assume that all offsite (public) dose occursunder these atmospheric conditions at the site of interest, i.e. 760 meters North of MUJRR. In ourcase at 760 meters it occurs only during Stability Class F conditions; which normally only occur11.4% 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. AnEffluent Concentration Limit of 2.0 x 1 0-&deg; pCi/mi is used for the "unlisted" isotopes, which equalsthe DAC/300 when using the DOE Part 835 Default DAC limits of 6.0 x 10-&deg;6 Exposure at1 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 8760hours per year. Thus, the time assumed to be exposed to the effluent concentration limit is a factorof 4.38 longer than the 2000 hours per year that defines a DAC. The isotopes in question are basedon a default DAC limit of 6.0 x 10-o6 for short-lived (< 2 hour half-lives) submersion DAC'sin Appendix C of Part 835. No credit is taken for transit time from the stack to the receptor pointnor is credit taken for decay inside containment until release. In the case of Kr-89 and Xe-i137 thetransit 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 theDAC value to establish the effluent concentration limit is warranted. This reduction factor of 300 isconsistent, in fact more conservative, than 10 CFR 20 Appendix B, as the NRC calculates EffluentConcentration 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 mremihrI31 concentration =8.42 x 10.09 iiCilml131I effluent concentration limit = 2.00 x 10-'&deg; gCi/ml1311 Conversion Factor = 5.71 x 10.0 mrenm/hrTherefore, 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&deg; gCi/ml x 5.71 x i0-0 mremihr x 16.5 brs)=3.96 x 10+&deg;&deg; mremNote: 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 Concentrationand Radiation Doses in the Unrestricted Area -Radio iodineRadionuclide13111321133I13411351Effluent Limit2.00 x 10-10 p.Ci/ml2.00 x 10-&deg; xtCi/ml1.00 x 10-09 pCi/ml6.00 x 10-&deg; gCi/ml6.00 x 10-0 pCi/mlMaximum Concentration18.42 x 10-09 pxCi/ml1.86 x 10-08 pCi/ml4.00 x 10-&deg; pCi/ml4.55 x 10.0 pCi/ml3.78 x 10.08 pCi/mlRadiation Dose3.96 x 10+00 mrem8.78 x 10.02 mrem3.77 x~ 10+00 mrem7.14 x 10-02 mrem5.93 x 10-&deg; mremTotal = 8.48 mremNote 1: Maximum Concentrations are radioiodine and noble gas concentrations leaking from the containment buildingand exiting the exhaust stack reduced by a dilution factor of 292.Effluent Concentration Limits. Concentrations at Point of Maximum Concentrationand Radiation Doses in the Unrestricted Area -Noble GasesRadionuclide85Kr85mIKr57Kr85Kr89Kr9O0r135Xel3smXe139XeEffluent Limit7.00 x 10"07 jiCi/ml1.00 X 10-0 xtCi/ml2.00 x 10-0 pCi/ml9.00 x 10-09 2.00 x 10-08 pCi/ml2.00 x 10-08 ptCi/ml5.00 X 10-07 iiCi/ml7.00 x 10&deg;08 iiCi/ml4.00 x 10-0 ptCi/ml2.00 x 10-0 pCi/ml2.00 x 10.0 iiCi/ml2.00 x 10-08 pCi/mlMaximum Concentration'6.85 x 10-12 pCi/ml3.04 x 10.08 gCi/ml6.17 x 10-08 gtCi/ml8.70 x 10-08 pCi/ml1.11 x 10.07 giCi/ml1.10 x 10.o pCi/mi5.45 x 10-08 pCi/ml2.71 x 10-08 gtCi/ml1.43 x 1 0"07 pCi/ml1.50 X 10.07 ptCi/ml1.23 x 10.07 pCi/mlRadiation Dose9.22 x 10-07 mrem2.86 x 10.02 mrem2.91 x 10.01 mrem9.10 x 10.01 mrem5.24 x 10.01 mrem5.17 x 10-1torero1.43 x 10.02 mrem7.34 x 10.0 mrem6.38 x 10-02 mrem6.76 x 10.01 mrem7.06 x 10.01 mrem5.80 x 10.01 mremTotal = 4.38 mremNote 1: Maximum Concentrations are radioiodine and noble gas concentrations leaking from the containment buildingand 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 dosesfrom the radioiodines and noble gases must be added together, and result in the following values:Dose from Radioidines and Noble Gases in the Unrestricted AreaCommitted Effective Dose Equivalent (Radioiodine)Committed Effective Dose Equivalent (Noble Gases)Total Effective Dose Equivalent (Whole Body)8.48 mrem4.38 mrem12.87 mrem85 of 86 Sunuming the doses from the noble gases and the radioiodines simply substantiates earlierstatements regarding the very low levels in the unrestricted area should a failure of a fueledexperiment occur, and should the containment building leak following such an event. Because thedose values are so low, the dose from the noble gases becomes the dominant value, but the overallTEDE is still only 12.87 mremn, a value far below the applicable 10 CFR 20 regulatory limit for theunrestricted area.
 
==References:==
1Hlazards Summary Report, Addendum 3, Section 5.3.3, University of Missouri Research ReactorFacility, August 1972 (as revised by the 1989-1990 Operations Annual Report).2Hlazards Summary Report, Addendum 4, Appendix C, University of Missouri Research ReactorFacility, October 1973.86 of 86  


'ATTACHMENT 2COPYMODIFICATION RECORDModification NumberModification Title72-7 .. ..... ......,Page 1Storage BasketPage Required DateNumber Page Title Yes No Completed By1234S6Modification Record xSystem Proposal (including a detailed xhazards analysis)Crew Evaluation xSafety Evaluation (OSHA) xSafety Subcommittee Review _Reactor Advisory Committee ReviewABC Review Modification Approved 81011121314Parts RequirementInstallation RecordBlueprints, spare parts, tech manualsPre-op testSOP changesConmpliance checks and PM'sMH cards AervisorYesNo7_iL_I/616 79~Li4) Z.2 --'I T~,C /7 730 teDate'C) 15 73I -p~'-~39 -('-1~vsDaZefByZJAm9,Reactor S ervisorModification CompletedCOPY ATTACHMENT 2Analysis of an Auxiliary Spent Fuel Storage Rackin the MURR PoolThe Missouri University Research Reactor (MURR) has asa part of its pool a deep pit with two storage racks capableof holding sixteen spent fuel elements or two full coreloadings. This is inadequate for the present MURR fuelcycle and with the advent of 10 megawatt operation, thesituation will be even worse. It is proposed that an addi-tional eight element rack be installed between the existingtwo. Figure 1 is a sketch depicting a top view of proposedconfiguration. The existing two racks are hung from thepool wall along with their respective gamma shields. So asnot to stress these supports further, it is proposed that theauxiliary rack have an integral stand to support it 2 feet offthe pool floor and level with the existing racks. The newrack will attach to the present ten element rack by brack~etsthat engage underneath as shown in Figure 1. The rack isessentially self-supporting but this attachment will lendextra stability. The fully loaded rack will weigh approximately300 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 betweenthe elements in the new rack to insure that each element isseparated from every other by boral.t This is well within themaximum Keff limit of 0.8 presented in the MURR license R-103.Thus the fully loaded rack will be far subcritical.  
'ATTACHMENT 2COPYMODIFICATION RECORDModification NumberModification Title72-7 .. ..... ......,Page 1Storage BasketPage Required DateNumber Page Title Yes No Completed By1234S6Modification Record xSystem Proposal (including a detailed xhazards analysis)Crew Evaluation xSafety Evaluation (OSHA) xSafety Subcommittee Review _Reactor Advisory Committee ReviewABC Review Modification Approved 81011121314Parts RequirementInstallation RecordBlueprints, spare parts, tech manualsPre-op testSOP changesConmpliance checks and PM'sMH cards AervisorYesNo7_iL_I/616 79~Li4) Z.2 --'I T~,C /7 730 teDate'C) 15 73I -p~'-~39 -('-1~vsDaZefByZJAm9,Reactor S ervisorModification CompletedCOPY ATTACHMENT 2Analysis of an Auxiliary Spent Fuel Storage Rackin the MURR PoolThe Missouri University Research Reactor (MURR) has asa part of its pool a deep pit with two storage racks capableof holding sixteen spent fuel elements or two full coreloadings. This is inadequate for the present MURR fuelcycle and with the advent of 10 megawatt operation, thesituation will be even worse. It is proposed that an addi-tional eight element rack be installed between the existingtwo. Figure 1 is a sketch depicting a top view of proposedconfiguration. The existing two racks are hung from thepool wall along with their respective gamma shields. So asnot to stress these supports further, it is proposed that theauxiliary rack have an integral stand to support it 2 feet offthe pool floor and level with the existing racks. The newrack will attach to the present ten element rack by brack~etsthat engage underneath as shown in Figure 1. The rack isessentially self-supporting but this attachment will lendextra stability. The fully loaded rack will weigh approximately300 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 betweenthe elements in the new rack to insure that each element isseparated from every other by boral.t This is well within themaximum Keff limit of 0.8 presented in the MURR license R-103.Thus the fully loaded rack will be far subcritical.  
Line 52: Line 70:
"' "ATTACHMENT 3Criteria: Th~e heat contributed to the pool by the added 14 elements awaiti~ng ship-ment shall not cause an appreciable pool temperature increase overperiods when the pool system is secured.For this calculation it is conservatively assumed that none ofthe 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 willcontribute a majority of the decay heat load, initially 20 KW. The12 remaining elements are assumed to have a decay history of only 30days. These 1:2 contribute 15 KW. To simplify calculations it willbe assumed that the decay heat load of the newly removed elements isconstant at the 20 KW value for the first 24 hours at which time itis reduced to the heat load level 8.5 KW for 2 elements with one daydecay for the remainder of the weekend period. Thus, the total heatload for the 14 spent fuel elements will be 35 KW for the first 24hours and 23.75 KW for the remaining 48 hours. Using the formulaq = MCpAT one 8an determine that the temperature rise over the first24 hours is 19v F, while that o.ver the remaining 48 hours is 260, F.The total temperature increase in the pool water over the weekendperiod will be less than 450 F since this aT would result if no heatwere transferred to the surroundings. The degree of conservatism inthis result is illustrated by the fact that at present the pool tempera-ture increase over the weekend resulting from 24 elements stored in theZ basket and 8 elements in the core is approximately 150 F.Standard Operating Procedures require that following a shutdownthe pool system shall remain in operation for a minimum of five minutes.Data from pool temperature charts for various times of the year indicatesa maximum temperature of 800 F after system shutdowns. Thus, evenwith the extremely conservative assumptions made, the final temperatureof the pool water will be 140U.F (800.+ 15OF + 45OF) which is .wellbelow the saturation temperature of 2120 F.
"' "ATTACHMENT 3Criteria: Th~e heat contributed to the pool by the added 14 elements awaiti~ng ship-ment shall not cause an appreciable pool temperature increase overperiods when the pool system is secured.For this calculation it is conservatively assumed that none ofthe 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 willcontribute a majority of the decay heat load, initially 20 KW. The12 remaining elements are assumed to have a decay history of only 30days. These 1:2 contribute 15 KW. To simplify calculations it willbe assumed that the decay heat load of the newly removed elements isconstant at the 20 KW value for the first 24 hours at which time itis reduced to the heat load level 8.5 KW for 2 elements with one daydecay for the remainder of the weekend period. Thus, the total heatload for the 14 spent fuel elements will be 35 KW for the first 24hours and 23.75 KW for the remaining 48 hours. Using the formulaq = MCpAT one 8an determine that the temperature rise over the first24 hours is 19v F, while that o.ver the remaining 48 hours is 260, F.The total temperature increase in the pool water over the weekendperiod will be less than 450 F since this aT would result if no heatwere transferred to the surroundings. The degree of conservatism inthis result is illustrated by the fact that at present the pool tempera-ture increase over the weekend resulting from 24 elements stored in theZ basket and 8 elements in the core is approximately 150 F.Standard Operating Procedures require that following a shutdownthe pool system shall remain in operation for a minimum of five minutes.Data from pool temperature charts for various times of the year indicatesa maximum temperature of 800 F after system shutdowns. Thus, evenwith the extremely conservative assumptions made, the final temperatureof the pool water will be 140U.F (800.+ 15OF + 45OF) which is .wellbelow the saturation temperature of 2120 F.
ATTACHMENT 3Criteria: Safety in moving fuel.Prior to spent fuel shipping the second level of baskets mustbe moved to the weir area so that elements to be shipped may betransferred to the upper level of baskets. The fuel movement sequenceshall be written so that at any time that the baskets are moved therewill be no more than six elements contained in the baskets. The sixelements shall be secured in the basket, (see design drawings). Sixelements contain insufficient fuel for criticality.  
ATTACHMENT 3Criteria: Safety in moving fuel.Prior to spent fuel shipping the second level of baskets mustbe moved to the weir area so that elements to be shipped may betransferred to the upper level of baskets. The fuel movement sequenceshall be written so that at any time that the baskets are moved therewill be no more than six elements contained in the baskets. The sixelements shall be secured in the basket, (see design drawings). Sixelements contain insufficient fuel for criticality.  
"' ATTACHMENT 3 RTP-l5Revised 8-6-76PROCEDURE FOR INSTALLATION OF SPENT FUEL STORAGE BASKETI. After completion of shop work and prior to installation in pool, scan eachS element storage box with Pu-Be neutron source and detector to insure presenceof boral, record data.2. Manipulate fuel as per sequence to place the 14 elements awaiting shipmentin the east two rows of the present Z basket (Zll, to Z24)3. Install shields and basket, secure to stand. Install SRM detector and takea 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 asper sequence. A 1/rn plot will be maintained as each element is loaded intothe basket.5. Upon completion of transfers to additional Z storage baskets, return thenon-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 managerfor inclusion in mod package.Caudle ~JulianReactor ManagerDate ''6'- ?7  
"' ATTACHMENT 3 RTP-l5Revised 8 76PROCEDURE FOR INSTALLATION OF SPENT FUEL STORAGE BASKETI. After completion of shop work and prior to installation in pool, scan eachS element storage box with Pu-Be neutron source and detector to insure presenceof boral, record data.2. Manipulate fuel as per sequence to place the 14 elements awaiting shipmentin the east two rows of the present Z basket (Zll, to Z24)3. Install shields and basket, secure to stand. Install SRM detector and takea 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 asper sequence. A 1/rn plot will be maintained as each element is loaded intothe basket.5. Upon completion of transfers to additional Z storage baskets, return thenon-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 managerfor inclusion in mod package.Caudle ~JulianReactor ManagerDate ''6'- ?7  
&deg; ATTACHMENT 3SAFETY SUBCOMMITTEEMinutes of Meeting of April 8, 1976Members 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 lastmeeting, expressed desire to see more details of the proceedings in the subcommitteeminutes.3. The subcommittee reviewed the circumstances of the March 2, 1976 abnormal occurrencereport regarding the failure of vent tank level controller 925 B. C. Juliansummarized the situation and answered questions. The subcommittee unanimouslyconcurred with the action taken.4. The subcommittee reviewed the abnormal occurrence report of March 24, 1976 regardingjumpering of the rod run-in functions on regulating blade position. C. Juliandiscussed the cause and corrective action. The subcommittee unanimously approvedof the action taken.5. The subcommittee reviewed Reactor Utilization Request Number 243 submitted byM. Janghorbani of the Environmental Trace Substances Research Center. The sub-committee suggested editorial changes and D. McKown noted that the RUR limitationswere based on actual in-practice experience at the MURR. After discussion, thesubcommittee unanimously recommended approval of the RUR as modified.6. The subcommittee began discussion of proposed modification package 76-3 for theinstallation of additional spent fuel storage in the MURR pool. C. 1Julian,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 chairover to T. Storvick. After questions and explanation, the subcommittee unanimouslyreconimended approval of the design concept and recommended that the projectproceed, providing that any safety related problems or design changes be reportedback to the subcommittee for further review.7. The subcommittee reviewed proposed modification 76-4 for the replacement of poollow level rod run-in switch 910. C. Julian explained that this change is an up-grade of originally installed equipment. The subcommittee unanimously recommendedapproval of the modification.  
&deg; ATTACHMENT 3SAFETY SUBCOMMITTEEMinutes of Meeting of April 8, 1976Members 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 lastmeeting, expressed desire to see more details of the proceedings in the subcommitteeminutes.3. The subcommittee reviewed the circumstances of the March 2, 1976 abnormal occurrencereport regarding the failure of vent tank level controller 925 B. C. Juliansummarized the situation and answered questions. The subcommittee unanimouslyconcurred with the action taken.4. The subcommittee reviewed the abnormal occurrence report of March 24, 1976 regardingjumpering of the rod run-in functions on regulating blade position. C. Juliandiscussed the cause and corrective action. The subcommittee unanimously approvedof the action taken.5. The subcommittee reviewed Reactor Utilization Request Number 243 submitted byM. Janghorbani of the Environmental Trace Substances Research Center. The sub-committee suggested editorial changes and D. McKown noted that the RUR limitationswere based on actual in-practice experience at the MURR. After discussion, thesubcommittee unanimously recommended approval of the RUR as modified.6. The subcommittee began discussion of proposed modification package 76-3 for theinstallation of additional spent fuel storage in the MURR pool. C. 1Julian,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 chairover to T. Storvick. After questions and explanation, the subcommittee unanimouslyreconimended approval of the design concept and recommended that the projectproceed, providing that any safety related problems or design changes be reportedback to the subcommittee for further review.7. The subcommittee reviewed proposed modification 76-4 for the replacement of poollow level rod run-in switch 910. C. Julian explained that this change is an up-grade of originally installed equipment. The subcommittee unanimously recommendedapproval of the modification.  
:: ..ATTACHMENT 3Page twoSafety Subconhmittee MinutesApril 8, 19768. New staff members Charles McKibben, Reactor Operations Engineer and Chester Edwards,Reactor Plant Engineer were introduced to the subcommittee. The meetingwas adjourned at 1610..,. ".."..Prepared By:Caudle JulianSecretaryApproved By:Dr. Walter MeyerChai rmanCJ:Id ATTACHMENT 3REACTOR SAFETY EVALUATIONPage 4..Modification Number________________Does this change involve changes to the Technical Specifications or anunreviewed safety hazard as described in 1.0 C&#xa3;R, secti~on 50.59A proposed change, test, orexperiment shall be deemed to involve an unreviewedsafety question (i) if the probability of occurrence or the consequences of anaccident or malfunction, of equipment important to safety previ~ously evaluated inthe safety analysis report may be increased; or (ii) if a possibility for an accidentor malfunction of a different type than any evaluated previously in the safety analysisreport may be created: or (iii) if the margin of safety as defined in the basis forany technical specification is reduced.YesSignature !l* #EVALUATION,~d~zA -~ t,~ ~ ~. dA ~a, , , .,.. .. i "?_,,Form Revised 10/31/75 jATTACHMENT 3REACTOR SAFETY ANALYSISPa-ge 4-MODIFICATION NUMBER 7,,7-3Does this change involve a change to the reactor facility as defined in the HazardsSummary 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 3Modification 76-3: Upper 2 spent fuel storageCrew evaluation of this proposal was initiated 4/6/76.No constructive suggestions were forthcoming.Caudle JulianReactor Manager  
:: ..ATTACHMENT 3Page twoSafety Subconhmittee MinutesApril 8, 19768. New staff members Charles McKibben, Reactor Operations Engineer and Chester Edwards,Reactor Plant Engineer were introduced to the subcommittee. The meetingwas adjourned at 1610..,. ".."..Prepared By:Caudle JulianSecretaryApproved By:Dr. Walter MeyerChai rmanCJ:Id ATTACHMENT 3REACTOR SAFETY EVALUATIONPage 4..Modification Number________________Does this change involve changes to the Technical Specifications or anunreviewed safety hazard as described in 1.0 C&#xa3;R, secti~on 50.59A proposed change, test, orexperiment shall be deemed to involve an unreviewedsafety question (i) if the probability of occurrence or the consequences of anaccident or malfunction, of equipment important to safety previ~ously evaluated inthe safety analysis report may be increased; or (ii) if a possibility for an accidentor malfunction of a different type than any evaluated previously in the safety analysisreport may be created: or (iii) if the margin of safety as defined in the basis forany technical specification is reduced.YesSignature !l* #EVALUATION,~d~zA -~ t,~ ~ ~. dA ~a, , , .,.. .. i "?_,,Form Revised 10/31/75 jATTACHMENT 3REACTOR SAFETY ANALYSISPa-ge 4-MODIFICATION NUMBER 7,,7-3Does this change involve a change to the reactor facility as defined in the HazardsSummary 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 3Modification 76-3: Upper 2 spent fuel storageCrew evaluation of this proposal was initiated 4/6/76.No constructive suggestions were forthcoming.Caudle JulianReactor Manager  
Line 58: Line 76:
* ecs~xIN ~ ~ w~ti~.4~-46~ALQ 2 /Q?~,1~<-~y~9a~  
* ecs~xIN ~ ~ w~ti~.4~-46~ALQ 2 /Q?~,1~<-~y~9a~  
"' ' ATTACHMENT 3(b p Brooks & Perkins, IncorporatedMateiials Handling Division.* P.O. Box 650. Cadillac, Michigan 40001
"' ' ATTACHMENT 3(b p Brooks & Perkins, IncorporatedMateiials Handling Division.* P.O. Box 650. Cadillac, Michigan 40001
* 616 776-9715March 26, 1976Ref: Certification of ConformanceUniversity of MissouriPurchasing Dept.General Services BuildingColumbia, Mo. 65201Attention: .Chester Edwards1 sheet Boral 1/4" x. 48" x 120", 35% B4CWe hereby certify that the core section of the composite materialcontains 35% by weight of boron carbide.Reference: Invoice #78,139Sheet #977Brooks & Perkins, Inc.Charles S. Timmons "-.. "..Quality AssuranceEVERT L HANCOCK,Nlotary 'Public,. Wexford CountyMy Commnision expires October I 4 1 197B,4....:-' ._d". .
* 616 776-9715March 26, 1976Ref: Certification of ConformanceUniversity of MissouriPurchasing Dept.General Services BuildingColumbia, Mo. 65201Attention: .Chester Edwards1 sheet Boral 1/4" x. 48" x 120", 35% B4CWe hereby certify that the core section of the composite materialcontains 35% by weight of boron carbide.
 
==Reference:==
Invoice #78,139Sheet #977Brooks & Perkins, Inc.Charles S. Timmons "-.. "..Quality AssuranceEVERT L HANCOCK,Nlotary 'Public,. Wexford CountyMy Commnision expires October I 4 1 197B,4....:-' ._d". .
ATTACHMENT 3 ae16elementas nunmb~er must be visually confirmed.Step From To Position and E~lement NurniNu-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/_. ____1Y7 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- Ia I S I I--I " o I .. ---I I I I I ,. ,______ , .a 3 _____, __ , __________.______--- a_ a- a ....a a a II a*/_. & / , z ..
ATTACHMENT 3 ae16elementas nunmb~er must be visually confirmed.Step From To Position and E~lement NurniNu-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/_. ____1Y7 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- Ia I S I I--I " o I .. ---I I I I I ,. ,______ , .a 3 _____, __ , __________.______--- a_ a- a ....a a a II a*/_. & / , z ..
6ve$wAk?41'i4I~ to ~jl ~v(yX45G3.21775 7"F65 F5 7.'I4-6 ATTACHMENT 3 Date iNote: EaCh.step involves the transfer of a single element.Dungahst 4?2  eelement's number must be. Visually confirme~d.Step From To Position and Element NumbNwr.,ber : Move Element Number : Position ,, Position ', Time ', Confirmed by (initials*1 I .i~I I a " " ' I.I I 7 I--"I a I7 ?..' F ." ,_ _ __-I, _,....i 7' " r: , .,L 3L4 .Ii I 7"7EF67 -_'___- ",i%_L! 775 F7,/ uLTL.7' .,<)l.J23PA k,Iq. l ..-7 7~5 F I .I7 " ..16~L : 7v P 7'6 L VAL? 7_3S"1 735 LI _,a i h,,T- .', -d !/-- .* I __ .-
6ve$wAk?41'i4I~ to ~jl ~v(yX45G3.21775 7"F65 F5 7.'I4-6 ATTACHMENT 3 Date iNote: EaCh.step involves the transfer of a single element.Dungahst 4?2  eelement's number must be. Visually confirme~d.Step From To Position and Element NumbNwr.,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 __ .-
Line 66: Line 87:
ATTACHMENT 3..........I-3* 3.... ... .i. .............. ..7/31$ii.. ~4 Q-J .....L... .. ...... 1. ...../4) 132.73 X!0-~Q U~4~P~Q- 1 LoLz(
ATTACHMENT 3..........I-3* 3.... ... .i. .............. ..7/31$ii.. ~4 Q-J .....L... .. ...... 1. ...../4) 132.73 X!0-~Q U~4~P~Q- 1 LoLz(
VATTACHMENT 4MODIFICATION RECORDO RIGINALPage 1M.odification Nlumber. J-" .co'1 Modification .y& 6 -/II/PageNo.-1.*2.3.4.5.6.7.8.Page TitleModification RecordSystem ProposalPreop Test ProceduresReactor Safety EvaluationCrew EvaluationSafety Subcommittee ReviewReactor Advisory Committee ReviewAEC ReviewRequiredYes NoxxXIL_xDateCompl eted_3Q/f7:I0LLZ1LB6Modification AproeDate of Completion 2 # TModification Completed J :_"/Reactor ManagerDateForm Revised10-31-75 ATTACHMENT 4REVISION TO MODIFICATION PACKAGE 76-3INSTALLATION OF *14 ELEMENT SPENT FUEL STORAGE BASKETAs required by the Safety Committee Meeting of April 8, 1976, implementationof 10 element storage basket addition shall be reported back to the Subcommitteefor review. The following report is submitted to meet this requirement.10 Element Z Basket InstallationThe present fuel storage capacity at MURR is 38 elements4 elements in the X and V baskets. The remaining 8 spaces inare required to defuel the core if the situation arises. Anfuel storage capacity is necessitated by:in the Z basket andthe X and V basketsincrease in the1. NRC regulation of having less than 5Kg of unirradiatedfuel in the fuel vault.2. 120 plus days of decay time required per element beforespent fuel shipment.3. Operating schedule and unirradiated fuel inventory hasincreased the number of fuel elements involved in ourfuel cycle.The 10 element basket size is dictated by space available in the Z basketstorage area. Construction and material of the 10 element basket is similarto the previous 14 element basket. The support stand and shielding for the new10 element basket was incorporated in the initial construction for ModificationPackage 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. restingon 3 brackets or 1,200 lbs. per bracket. Each bracket is design rated to carrya vertical load of 2,000 Ibs; therefore, load is 60% of rated vertical loaddesign.Safety ConsiderationsPriora neutronservice,than 0.8r to loading fuel in the basket, all boral sections wisource and neutron detector to verify boral present.a subcritical measurement will be performed to ensurewith 10 elements loaded.II be scanned withWhen placed inthat Keff is less ATTACHMENT 42Decay Heat Build UpDecay heat build up in pool for 3-dayupper Z basket.14 element2 elements retired12 elements, 30 day decayperiod for 24 element versus 14 element24 element2 elements retired22 elements, 30 day decay2 elements12 elements20KW15KW2 elements22 elements20KW27. 5KWEnd of 24 hrs.35KW = 19&deg;Fincrease per1 st dayEnd of 24 hrs.47.5KW = 25.8&deg;Fincrease per1st day2 elements12 elements8.5KW1 5KW23.5KW = 12.8&deg;Fincrease per2nd & 3rd day2 elements 8.5KW22 elements 27.5KW36.0KW = 19.5&deg;Fincrease per2nd & 3rd dayEnd of 3 days450F increase versus66&deg;F increase*Calculations are conservative since they assume no heat is transferred out ofthe pool. The degree of conservatism is illustrated by the fact that during5-day week operations; pool temperature increase over the weekend resulting from24 elements stored in the lower Z basket and 8 elements in the core was approxi-mately 15 F. Maximum pool temperature after shutdow8 from 8ast experience was80 F following shutdown procedures. Thus, 1610F (80 F + 15 F + 66 F) wouldbe maximum pool temperature after 3 days which is well below saturation temperatureof 212 F.Surface Dose RateDose rate increase should be less than 2.Omr/hr at surface of pool water withpool 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 ModificationPackage 76-3 since the design model incorporated a 24 element upper Z basketaddition versus the 14 element basket installed.Submitted byApproved byDave McGinty /Reactor sth/(farl ie McKibbehSReactor ManagerK ATTACHMENT 4RTP-l 5BOctober 3, 1978PROCEDURE FOR INSTALLATION OF UPPER ZFUEL STORAGE BASKET1. Prior to installation for fuel element loading, a scan will be performed on eachelement storage box with Pu-Be-neutron source and detector to insure presenceof 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 SRMdetector and take a series of base line counts. Record dose rate at thistime.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 loadedinto the basket. If Keff from graph reaches 0.8, theprocedure will be stopped and Reactor Manager informed.5. After completion of transfer sequence and verification, that Keffis less than 0.8; (EstabliSh does rate with Z basket storage arearefuel the core according to applicable sequence.6. Compile data generated in steps, 1, 3, and 4. Complete forms arethe Reactor Manager for inclusion in mod package.of assemblycompletely loaded. )to be given toSubmitted byDave McGi ntyReactor Physici stApproved byCharl ie McKi bbenReaCtor Manager ATTACHMENT 40357/0'37I/6/, 5 05.55I,3 qf/.7qo/47g. '.&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'../3.13C/z .... .....S5,/,'4 ,A-I.-, .'7IC m1.0*OI9*.87II .'4Siiiilii' -ll~ nlt rT .lll rlrlrlil 1 l riiilli!!li!l!i!llil ]liil!l!llli!lll i i !!Iii++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 Illlllllllllltl!lll!lilil!lil!, .-+" +...$ .....'+"'++'++'+,, ;++ ,, , l + l Njil l f Hji~ l i..... II,. 1.1! I..I, I.., .*j* 4** lily tr II -TiijlP1IlI.'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~~tTI~j~,:I Ii *~ l*! I Ed i~ Ip~1J~yj 3/4 ~ P ii~. .iv i4 ~ j .t1~~it  ! Ij~ h 41 F I ]~ jhI L ~ *1I II ~ I'1P b *~. [ I IIm,3.1]Iit:ilii!i+' iL + , I ", I,+ I + i l ++, ,I,+LI III i. I.I.IllI'.Iil IN!lil!lltq!li IMI.ttll_ ! lli!!ll!l!llllititltlllltl!!ill!lil! 11 tllhillt!llli.i!ii+ fill j-:i;1 ju- w + ~ ! .i+" ;++ +/-4 JL 14~.IU44PII +/-$2j jj 11......il'.4.. --.-..........-............-.-.........- ,Ij4t+ .,7 I ~-'45 ~, 7 g r~ /0 ,.*1 5J:kI.4 ~'-ATTACHMENT 4REFUELIN'G SEQUENhCENote': Ea&h.step involves the transfez of a single elei~ent. During each step1 the transfs evi, onfiTried.. ,l*Step-Ib31aI-lore Eleiaert Euniber775 F 8o&deg;,aS,aaiPositionIIIaPositionzglTimeIiIiCos~i$.od .(imitniaI7A/f7". I --..... -_ _ , _I a3 Il.... ., , J k. 33./ _ __ _ _________ __* u __ .~l ,_' ... --......5_ -? 7 5 F /... _ F_ ' __ ! -___ _-__....___-____ _ _ _________ ....a ;* -..a-'a-t/1 77$ fK1.I " -J[ i.-3ID 61% :* : , _...
VATTACHMENT 4MODIFICATION RECORDO RIGINALPage 1M.odification Nlumber. J-" .co'1 Modification .y& 6 -/II/PageNo.-1.*2.3.4.5.6.7.8.Page TitleModification RecordSystem ProposalPreop Test ProceduresReactor Safety EvaluationCrew EvaluationSafety Subcommittee ReviewReactor Advisory Committee ReviewAEC ReviewRequiredYes NoxxXIL_xDateCompl eted_3Q/f7:I0LLZ1LB6Modification AproeDate of Completion 2 # TModification Completed J :_"/Reactor ManagerDateForm Revised10-31-75 ATTACHMENT 4REVISION TO MODIFICATION PACKAGE 76-3INSTALLATION OF *14 ELEMENT SPENT FUEL STORAGE BASKETAs required by the Safety Committee Meeting of April 8, 1976, implementationof 10 element storage basket addition shall be reported back to the Subcommitteefor review. The following report is submitted to meet this requirement.10 Element Z Basket InstallationThe present fuel storage capacity at MURR is 38 elements4 elements in the X and V baskets. The remaining 8 spaces inare required to defuel the core if the situation arises. Anfuel storage capacity is necessitated by:in the Z basket andthe X and V basketsincrease in the1. NRC regulation of having less than 5Kg of unirradiatedfuel in the fuel vault.2. 120 plus days of decay time required per element beforespent fuel shipment.3. Operating schedule and unirradiated fuel inventory hasincreased the number of fuel elements involved in ourfuel cycle.The 10 element basket size is dictated by space available in the Z basketstorage area. Construction and material of the 10 element basket is similarto the previous 14 element basket. The support stand and shielding for the new10 element basket was incorporated in the initial construction for ModificationPackage 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. restingon 3 brackets or 1,200 lbs. per bracket. Each bracket is design rated to carrya vertical load of 2,000 Ibs; therefore, load is 60% of rated vertical loaddesign.Safety ConsiderationsPriora neutronservice,than 0.8r to loading fuel in the basket, all boral sections wisource and neutron detector to verify boral present.a subcritical measurement will be performed to ensurewith 10 elements loaded.II be scanned withWhen placed inthat Keff is less ATTACHMENT 42Decay Heat Build UpDecay heat build up in pool for 3-dayupper Z basket.14 element2 elements retired12 elements, 30 day decayperiod for 24 element versus 14 element24 element2 elements retired22 elements, 30 day decay2 elements12 elements20KW15KW2 elements22 elements20KW27. 5KWEnd of 24 hrs.35KW = 19&deg;Fincrease per1 st dayEnd of 24 hrs.47.5KW = 25.8&deg;Fincrease per1st day2 elements12 elements8.5KW1 5KW23.5KW = 12.8&deg;Fincrease per2nd & 3rd day2 elements 8.5KW22 elements 27.5KW36.0KW = 19.5&deg;Fincrease per2nd & 3rd dayEnd of 3 days450F increase versus66&deg;F increase*Calculations are conservative since they assume no heat is transferred out ofthe pool. The degree of conservatism is illustrated by the fact that during5-day week operations; pool temperature increase over the weekend resulting from24 elements stored in the lower Z basket and 8 elements in the core was approxi-mately 15 F. Maximum pool temperature after shutdow8 from 8ast experience was80 F following shutdown procedures. Thus, 1610F (80 F + 15 F + 66 F) wouldbe maximum pool temperature after 3 days which is well below saturation temperatureof 212 F.Surface Dose RateDose rate increase should be less than 2.Omr/hr at surface of pool water withpool 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 ModificationPackage 76-3 since the design model incorporated a 24 element upper Z basketaddition versus the 14 element basket installed.Submitted byApproved byDave McGinty /Reactor sth/(farl ie McKibbehSReactor ManagerK ATTACHMENT 4RTP-l 5BOctober 3, 1978PROCEDURE FOR INSTALLATION OF UPPER ZFUEL STORAGE BASKET1. Prior to installation for fuel element loading, a scan will be performed on eachelement storage box with Pu-Be-neutron source and detector to insure presenceof 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 SRMdetector and take a series of base line counts. Record dose rate at thistime.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 loadedinto the basket. If Keff from graph reaches 0.8, theprocedure will be stopped and Reactor Manager informed.5. After completion of transfer sequence and verification, that Keffis less than 0.8; (EstabliSh does rate with Z basket storage arearefuel the core according to applicable sequence.6. Compile data generated in steps, 1, 3, and 4. Complete forms arethe Reactor Manager for inclusion in mod package.of assemblycompletely loaded. )to be given toSubmitted byDave McGi ntyReactor Physici stApproved byCharl ie McKi bbenReaCtor Manager ATTACHMENT 40357/0'37I/6/, 5 05.55I,3 qf/.7qo/47g. '.&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'../3.13C/z .... .....S5,/,'4 ,A-I.-, .'7IC m1.0*OI9*.87II .'4Siiiilii' -ll~ nlt rT .lll rlrlrlil 1 l riiilli!!li!l!i!llil ]liil!l!llli!lll i i !!Iii++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 Illlllllllllltl!lll!lilil!lil!, .-+" +...$ .....'+"'++'++'+,, ;++ ,, , l + l Njil l f Hji~ l i..... II,. 1.1! I..I, I.., .*j* 4** lily tr II -TiijlP1IlI.'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~~tTI~j~,:I Ii *~ l*! I Ed i~ Ip~1J~yj 3/4 ~ P ii~. .iv i4 ~ j .t1~~it  ! Ij~ h 41 F I ]~ jhI L ~ *1I II ~ I'1P b *~. [ I IIm,3.1]Iit:ilii!i+' iL + , I ", I,+ I + i l ++, ,I,+LI III i. I.I.IllI'.Iil IN!lil!lltq!li IMI.ttll_ ! lli!!ll!l!llllititltlllltl!!ill!lil! 11 tllhillt!llli.i!ii+ fill j-:i;1 ju- w + ~ ! .i+" ;++ +/-4 JL 14~.IU44PII +/-$2j jj 11......il'.4.. --.-..........-............-.-.........- ,Ij4t+ .,7 I ~-'45 ~, 7 g r~ /0 ,.*1 5J:kI.4 ~'-ATTACHMENT 4REFUELIN'G SEQUENhCENote': Ea&h.step involves the transfez of a single elei~ent. During each step1 the transfs evi, onfiTried.. ,l*Step-Ib31aI-lore Eleiaert Euniber775 F 8o&deg;,aS,aaiPositionIIIaPositionzglTimeIiIiCos~i$.od .(imitniaI7A/f7". I --..... -_ _ , _I a3 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# -1-, P6 _____5 l l il C __ __ __ __ __-'i 77S ,Fio .F7 ________-I__~~ i ?Z- ........ L#+/-I t* --iJ A .... .~, i '1lV U'#..--.-______* o .5I 6--~ 4F~I C; I/'):,S I IIi I. S.... L --_____ I ..ai .SIIS..... I .. .. II I I ATTACHMENT 4, 6" ,,, ;,-sk,;-of Sevae ;~ A~*'/2O5;U,35S,7SI15*/6q's2-o'7K['0I,13//6zz-203o2j33034,6Ac4~&,i vgc~2~~j4-G-%~~
* __ _ 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--~ 4F~I C; I/'):,S I IIi I. S.... L --_____ I ..ai .SIIS..... I .. .. II I I ATTACHMENT 4, 6" ,,, ;,-sk,;-of Sevae ;~ A~*'/2O5;U,35S,7SI15*/6q's2-o'7K['0I,13//6zz-203o2j33034,6Ac4~&,i vgc~2~~j4-G-%~~
ATTACHMENT 4REACTOR SAFETY EVALUATIONPage 4 __ __Modification Number 76 -3C 'Does this change involve changes to the Technical Specifications or anunreviewed safety hazard as described in 10 CFR, section 50.59A proposed change, test, or experiment shall be deemed to involve an unreviewedsafety question (i) if the probability of occurrence or the consequences of anaccident or malfunction of equipment important to safety previously evaluated inthe safety analysis report may be increased; or (ii) if a possibility for an accidentor malfunction of a different type than any evaluated previously in the safety analysisreport may be created: or (iii) if the margin of safety as defined in the basis forany technical specification is reduced.Yes_____No ASignature_________EVALUAT IONForm Revised 10/31/75 ATTACHMENT 4REACTOR SAFETY ANALYSISPage 4-MODIFICATION NUMBER 74- 3. Does this change involve a change to the reactor facility as defined in the HazardsSummary and its addenda?Yes No. / Signature ,zd ...,&If yes, make an analysis below, if no. outline the basis for the. &#xa3;1 *Form Revised I0/31/75 ATTACHMENT 4REVISION TO MODIFICATION PACKAGE 76-3INSTALLATION OF 14 ELEMENT SPENT FUEL STORAGE BASKETAs required by the Safety Committee Meeting of April 8, 1976, implementationof 10 element storage basket addition shall be reported back to the Subcommitteefor review. The following report is submitted to meet this requirement.10 Element Z Basket InstallationThe present fuel storage capacity at MURR is 38 elements4 elements in the X and Y baskets. The remaining 8 spaces inare required to defuel the core if the situation arises. Anfuel storage capacity is necessitated by:in the Z basket andthe X and Y basketsincrease in the1. NRC regulation of having less than 5Kg of unirradiatedfuel in the fuel vault.2. 120 plus days of decay time required per element beforespent fuel shipment.3. Operating schedule and unirradiated fuel inventory hasincreased the number of fuel elements involved in ourfuel cycle.The 10 element basket size is dictated by space available in the Z basketstorage area. Construction and material of the 10 element basket is similarto the previous 14 element basket. The support stand and shielding for the new10 element basket was incorporated in the initial construction for ModificationPackage 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. restingon 3 brackets or 1,200 lbs. per bracket. Each bracket is design rated to carrya vertical load of 2,000 Ibs; therefore, load is 60% of rated vertical loaddesign.Safety Considerationsa neservthanPrior to loading fuel in the basket, all boral sections willeutron source and neutron detector to verify boral present.vice, a subcritical measurement will be performed to ensure tln0.8 with 10 elements loaded.be scanned withWhen placed inhat Keff is less ATTACHMENT4 2Decay Heat Build UpDecay heat build up in pool for 3-day period for 24 element versus 14 elementupper Z basket.14 element2 elements retired12 elements, 30 day decay24 el ement2 elements retired22 elements, 30 day decay2 elements12 elementsEnd of 24 hrs.20KW15KW2 elements22 elements20KW27.5KW35KW = 19&deg;Fincrease per1st dayEnd of 24 hrs. 47.5KW = 25.8&deg;Fincrease perIst day2 el ements12 elements8. 5KW1 5KW23.5KW = 12.8&deg;Fincrease per2nd & 3rd day2 elements 8.5KW22 elements 27.5KW36.0KW = 19.5&deg;Fincrease per2nd & 3rd dayEnd of 3 days45&deg; increase versus66&deg; increaseCalculations are conservative sinc~e they assume no heat is transferred out ofthe pool. The degree of conservatism is illustrated by the fact that during5-day week operations; pool temperature increase over the weekend resulting from24 elements stored in the lower Z basket and 8 elements in the core was approxi-mately 15 F. Maximum pool temperature after shutdow8 from 8ast experience was80 F following shutdown procedures. Thus, 161 F (80 F + 15 F + 66 F) wouldbe max~mum pool temperature after 3 days which is well below saturation temperatureof 212 F.Surface Dose RateDose rate increase should be less than 2.0mr/hr at surface of pool water withpool 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 ModificationPackage 76-3 since the design model incorporated a 24 element upper Z basketaddition versus the 14 element basket installed.Submitted byDave McGint'y /1Reactor stApproved byc$narlie ~McKibbefiReactor Manager9-29-78 a.ATTACHMENT 4Crew EvaluationM~odification Number Page 5-All crew members are asked to comment in some manner oni this proposal.Name Remarks10/18/73 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~z3/.... .. ... -.-........ ....... .z .../6.. ... ._o. .. .i. ...... .........3..3 / ... .... ... .............. .- .......  
ATTACHMENT 4REACTOR SAFETY EVALUATIONPage 4 __ __Modification Number 76 -3C 'Does this change involve changes to the Technical Specifications or anunreviewed safety hazard as described in 10 CFR, section 50.59A proposed change, test, or experiment shall be deemed to involve an unreviewedsafety question (i) if the probability of occurrence or the consequences of anaccident or malfunction of equipment important to safety previously evaluated inthe safety analysis report may be increased; or (ii) if a possibility for an accidentor malfunction of a different type than any evaluated previously in the safety analysisreport may be created: or (iii) if the margin of safety as defined in the basis forany technical specification is reduced.Yes_____No ASignature_________EVALUAT IONForm Revised 10/31/75 ATTACHMENT 4REACTOR SAFETY ANALYSISPage 4-MODIFICATION NUMBER 74- 3. Does this change involve a change to the reactor facility as defined in the HazardsSummary and its addenda?Yes No. / Signature ,zd ...,&If yes, make an analysis below, if no. outline the basis for the. &#xa3;1 *Form Revised I0/31/75 ATTACHMENT 4REVISION TO MODIFICATION PACKAGE 76-3INSTALLATION OF 14 ELEMENT SPENT FUEL STORAGE BASKETAs required by the Safety Committee Meeting of April 8, 1976, implementationof 10 element storage basket addition shall be reported back to the Subcommitteefor review. The following report is submitted to meet this requirement.10 Element Z Basket InstallationThe present fuel storage capacity at MURR is 38 elements4 elements in the X and Y baskets. The remaining 8 spaces inare required to defuel the core if the situation arises. Anfuel storage capacity is necessitated by:in the Z basket andthe X and Y basketsincrease in the1. NRC regulation of having less than 5Kg of unirradiatedfuel in the fuel vault.2. 120 plus days of decay time required per element beforespent fuel shipment.3. Operating schedule and unirradiated fuel inventory hasincreased the number of fuel elements involved in ourfuel cycle.The 10 element basket size is dictated by space available in the Z basketstorage area. Construction and material of the 10 element basket is similarto the previous 14 element basket. The support stand and shielding for the new10 element basket was incorporated in the initial construction for ModificationPackage 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. restingon 3 brackets or 1,200 lbs. per bracket. Each bracket is design rated to carrya vertical load of 2,000 Ibs; therefore, load is 60% of rated vertical loaddesign.Safety Considerationsa neservthanPrior to loading fuel in the basket, all boral sections willeutron source and neutron detector to verify boral present.vice, a subcritical measurement will be performed to ensure tln0.8 with 10 elements loaded.be scanned withWhen placed inhat Keff is less ATTACHMENT4 2Decay Heat Build UpDecay heat build up in pool for 3-day period for 24 element versus 14 elementupper Z basket.14 element2 elements retired12 elements, 30 day decay24 el ement2 elements retired22 elements, 30 day decay2 elements12 elementsEnd of 24 hrs.20KW15KW2 elements22 elements20KW27.5KW35KW = 19&deg;Fincrease per1st dayEnd of 24 hrs. 47.5KW = 25.8&deg;Fincrease perIst day2 el ements12 elements8. 5KW1 5KW23.5KW = 12.8&deg;Fincrease per2nd & 3rd day2 elements 8.5KW22 elements 27.5KW36.0KW = 19.5&deg;Fincrease per2nd & 3rd dayEnd of 3 days45&deg; increase versus66&deg; increaseCalculations are conservative sinc~e they assume no heat is transferred out ofthe pool. The degree of conservatism is illustrated by the fact that during5-day week operations; pool temperature increase over the weekend resulting from24 elements stored in the lower Z basket and 8 elements in the core was approxi-mately 15 F. Maximum pool temperature after shutdow8 from 8ast experience was80 F following shutdown procedures. Thus, 161 F (80 F + 15 F + 66 F) wouldbe max~mum pool temperature after 3 days which is well below saturation temperatureof 212 F.Surface Dose RateDose rate increase should be less than 2.0mr/hr at surface of pool water withpool 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 ModificationPackage 76-3 since the design model incorporated a 24 element upper Z basketaddition versus the 14 element basket installed.Submitted byDave McGint'y /1Reactor stApproved byc$narlie ~McKibbefiReactor Manager9-29-78 a.ATTACHMENT 4Crew EvaluationM~odification Number Page 5-All crew members are asked to comment in some manner oni this proposal.Name Remarks10/18/73 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~z3/.... .. ... -.-........ ....... .z .../6.. ... ._o. .. .i. ...... .........3..3 / ... .... ... .............. .- .......  
,ATTACHMENT 5Page 1 of 15OkI1GNALRevised: 2/18/86App' djReactor ManagerMODIFICATION RECORDMODIFICATION NO. ___!-,, _Modification Temporary Additional in-pool fuel storage basketsPageN__o.RequiredYES NODateCompletedPage Title123456789Modification RecordSystem ProposalReactor Safety AnalysisReactor Safety EvaluationCrew EvaluationMURR SOP Review CompleteCompliance or P.M. Revision "Parts Requirement SheetPrints, Technical Manual, SpareParts Change Requirementcation Approved: XXXxxx,__ 0 17.-I1llBy(Initials)r3 IModifi(DatffDate of Completion:ModificationCompleted: i~; (PJReactor ManagerItemNo.___REVIEW AND FOLLOW-UP ACTIONRequiredYESS NODate DocumentedCompleted By (Initials) ]_____________________1 Safety Subcommittee Review2 Reactor Advisory Committee Review3 U.S. NRC ReviewII4 MURR Drawings Updated ATTACHMENT 5Page 2of 15 flD1(S NAL eie: /88MODFIATONNO_____ J\~)" ,"App'dtA~Reactor ManagerSYSTEM PROPOSALThe inability of MURR to establish spent fuel shipping capability since the GE-700 cask wasremoved from service in September 1989 has created the need for temporary additional in-pool fuel storage. This modification package documents the evaluations performed to showthat the use of two shipping baskets designed for use in the MHIA cask as temporary in-poolstorage facilities does not present an unreviewed safety question. Each MHIA shipping baskethas twelve fuel element storage positions in a three by four matrix with a boral sheet betweeneach row of four elements(see page 13). These baskets will be attached by brackets to the deeppool"X" and"Y" basket fuel element storage to provide stability and lateral support. Thesebrackets are made of 0.25" aluminium angle(see page 15) and provides a position for the OSbasket if additional in-pool storage is needed.The evaluation performed for each MHIA basket will include a criticality analysis(KENO), aboral plate verification, thermal analysis and 1/M determination when it is first loaded. Aseparate evaluation will be made of the OS basket if used as deep pool storage in conjunctionwith the two MHIA baskets.
,ATTACHMENT 5Page 1 of 15OkI1GNALRevised: 2/18/86App' djReactor ManagerMODIFICATION RECORDMODIFICATION NO. ___!-,, _Modification Temporary Additional in-pool fuel storage basketsPageN__o.RequiredYES NODateCompletedPage Title123456789Modification RecordSystem ProposalReactor Safety AnalysisReactor Safety EvaluationCrew EvaluationMURR SOP Review CompleteCompliance or P.M. Revision "Parts Requirement SheetPrints, Technical Manual, SpareParts Change Requirementcation Approved: XXXxxx,__ 0 17.-I1llBy(Initials)r3 IModifi(DatffDate of Completion:ModificationCompleted: i~; (PJReactor ManagerItemNo.___REVIEW AND FOLLOW-UP ACTIONRequiredYESS NODate DocumentedCompleted By (Initials) ]_____________________1 Safety Subcommittee Review2 Reactor Advisory Committee Review3 U.S. NRC ReviewII4 MURR Drawings Updated ATTACHMENT 5Page 2of 15 flD1(S NAL eie: /88MODFIATONNO_____ J\~)" ,"App'dtA~Reactor ManagerSYSTEM PROPOSALThe inability of MURR to establish spent fuel shipping capability since the GE-700 cask wasremoved from service in September 1989 has created the need for temporary additional in-pool fuel storage. This modification package documents the evaluations performed to showthat the use of two shipping baskets designed for use in the MHIA cask as temporary in-poolstorage facilities does not present an unreviewed safety question. Each MHIA shipping baskethas twelve fuel element storage positions in a three by four matrix with a boral sheet betweeneach row of four elements(see page 13). These baskets will be attached by brackets to the deeppool"X" and"Y" basket fuel element storage to provide stability and lateral support. Thesebrackets are made of 0.25" aluminium angle(see page 15) and provides a position for the OSbasket if additional in-pool storage is needed.The evaluation performed for each MHIA basket will include a criticality analysis(KENO), aboral plate verification, thermal analysis and 1/M determination when it is first loaded. Aseparate evaluation will be made of the OS basket if used as deep pool storage in conjunctionwith the two MHIA baskets.
Line 75: Line 96:
,,, ATTACHMENT 5Page 9 Of 15MODIFICATION NO.C?!>?0OP G tN ALRevised: 2/18/86Reactor ManagerPARTS REQUIREMENT SHEET_ Parts DescriptionPart No.Purchase Order No. Date ReceivedPurchase Order No._Ordered From_DatePurchase Order No.__Ordered From_DatePurchase Order No. ___Ordered From_____Date _ _ _ _ _ _ _ _Purchase Order No. _____Ordered From _______Date  
,,, ATTACHMENT 5Page 9 Of 15MODIFICATION NO.C?!>?0OP G tN ALRevised: 2/18/86Reactor ManagerPARTS REQUIREMENT SHEET_ Parts DescriptionPart No.Purchase Order No. Date ReceivedPurchase Order No._Ordered From_DatePurchase Order No.__Ordered From_DatePurchase Order No. ___Ordered From_____Date _ _ _ _ _ _ _ _Purchase Order No. _____Ordered From _______Date  
... ', ATTACHMENT 5Page 10 of 15MODIFICATION NO.Revised: 2/18/86Reactor Manager0ORG;IN AkLBLUE PRINTS -- SPARE PARTS -TECHNICAL MANUALSBLUE PRINTS:Print No. __Print Title New Print2306 MH1A fuel holders XSPARE PARTS:SPart Description Part No. PurRev, of Old PrintN/A'chase Order No.Rev. No.N/AS.P. No.Date Rev.Date Re'd.Purchase Order No.Ordered From ________DatePurchase Order No.Ordered From ________DateTECHNICAL MANUALSManual Title Ordered FromDate Ordered Dt e'.Mna oDate Rec'd.Manual No.  
... ', ATTACHMENT 5Page 10 of 15MODIFICATION NO.Revised: 2/18/86Reactor Manager0ORG;IN AkLBLUE PRINTS -- SPARE PARTS -TECHNICAL MANUALSBLUE PRINTS:Print No. __Print Title New Print2306 MH1A fuel holders XSPARE PARTS:SPart Description Part No. PurRev, of Old PrintN/A'chase Order No.Rev. No.N/AS.P. No.Date Rev.Date Re'd.Purchase Order No.Ordered From ________DatePurchase Order No.Ordered From ________DateTECHNICAL MANUALSManual Title Ordered FromDate Ordered Dt e'.Mna oDate 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 : :-' "0U')-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~&sect;t2ifL7~Jcw1 Aif4-4~- ~4I7i i :Di_::i- ri-lTq-7:71: bt-~--r----... ...... .... :.. -I ... -= i =.- .. -~--i ... .. ---LL .. ....... ..l.. ..I.. ..!: " F..-- : :':l=:= ': :-f= :&deg; 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 .Fi -7:. :-,.. *- ..-- :*..i.,-i --.: --- :I " : -* :':-'I ". ::I:::: ..:":[. .:: .: -i ' ::._!.:.. .77-:i: 5-!=__i.::,' ::! !.:i- ::-g:-,"t : f.i:L_-:l : ! i i::.t. ;- :-: :::l..4::-.:"" : ii I : ! .! .. !:-T T -t::-:i:L .m T I I -, , i I .i T I -i * , .__:2--: .... ! --"=.!* l:z.:: i! : " ,,'i :* ..:--! i .... t !
... 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 : :-' "0U')-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~&sect;t2ifL7~Jcw1 Aif4-4~- ~4I7i i :Di_::i- ri-lTq-7:71: bt-~--r----... ...... .... :.. -I ... -= i =.- .. -~--i ... .. ---LL .. ....... ..l.. ..I.. ..!: " F..-- : :':l=:= ': :-f= :&deg; I : : i.. : :... ...___ -.-:_____ .___.__"_._.___-_-__._.____ V -, ., : : - . 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 .Fi -7:. :-,.. *- ..-- :*..i.,-i --.: --- :I " : -* :':-'I ". ::I:::: ..:":[. .:: .: -i ' ::._!.:.. .77-:i: 5-!=__i.::,' ::! !.:i- ::-g:-,"t : f.i:L_-:l : ! i i::.t. ;- :-: :::l..4::-.:"" : ii I : ! .! .. !:-T T -t::-:i:L .m T I I -, , i I .i T I -i * , .__:2--: .... ! --"=.!* l:z.:: i! : " ,,'i :* ..:--! i .... t !
ATTACHE76 --w)l3--0J-4}iI--....8__.2&deg;1T-I9-RV0In00o7--311IO$174-ci-JSILLI iii IILI I I i !Ii i I i i I ii I(0 1.n ~.. ~OO''Id0I0.0V-f e(~a~) 4?~A ee~v~ 4~I42-  
ATTACHE76 --w)l3--0J-4}iI--....8__.2&deg;1T-I9-RV0In00o7--311IO$174-ci-JSILLI iii IILI I I i !Ii i I i i I ii I(0 1.n ~.. ~OO''Id0I0.0V-f e(~a~) 4?~A ee~v~ 4~I42-  
...ATTACHMENT 5I,PooL'I , ITIs dr2/ 5/'ET e//1 3,s" -ro FR O' :z" if!5ORIGINAL'p.
...ATTACHMENT 5I,PooL'I , ITIs dr2/ 5/'ET e//1 3,s" -ro FR O' :z" if!5ORIGINAL'p.
ORIGINALtr~QFRrlNT VIEIW~ ~B TVT ]3 F- VTF-W  
ORIGINALtr~QFRrlNT VIEIW~ ~B TVT ]3 F- VTF-W  
' "' ATTACHMENT 5September 29, 1993Attachment to Modification Package 91-3Storage of Irradiated Fuel in OS Basket (1/30/93)The use of the OS basket for temporary fuel storage adjacent to the "Xv basketwas first implemented in 1979. The current arrangement of the OS basket with theMlix and MHY temporary storage positions not in the pooi does not represent anunreviewed safety question. If the OS basket is used in conjunction with the iVHIYand MIIX baskets an additional evaluation will be performed.Elements stored in OS basket all have greater than 177 days decay. The edge ofOS basket nearest the pooi wall is approximately 12 inches from the wall. Table 4-Aof Design Data, Vol. I, .Thermal Shielding Reauirements for Spent Fuel StorageFacilities, indicates 14"' of water needed for fuel with 106 seconds of decay. Theseelements had 177 days (min) of decay [1.5 x 107 sec] so adequate thermal shielding isavailable. The thermal shielding of the 1/4" thick stainless steel bottom and side ofthe OS basket are not considered, but would further reduce the water thicknessrequired as thermal shield.
' "' ATTACHMENT 5September 29, 1993Attachment to Modification Package 91-3Storage of Irradiated Fuel in OS Basket (1/30/93)The use of the OS basket for temporary fuel storage adjacent to the "Xv basketwas first implemented in 1979. The current arrangement of the OS basket with theMlix and MHY temporary storage positions not in the pooi does not represent anunreviewed safety question. If the OS basket is used in conjunction with the iVHIYand MIIX baskets an additional evaluation will be performed.Elements stored in OS basket all have greater than 177 days decay. The edge ofOS basket nearest the pooi wall is approximately 12 inches from the wall. Table 4-Aof Design Data, Vol. I, .Thermal Shielding Reauirements for Spent Fuel StorageFacilities, indicates 14"' of water needed for fuel with 106 seconds of decay. Theseelements had 177 days (min) of decay [1.5 x 107 sec] so adequate thermal shielding isavailable. The thermal shielding of the 1/4" thick stainless steel bottom and side ofthe OS basket are not considered, but would further reduce the water thicknessrequired as thermal shield.
ATTACHMENT 6@&#xa9;[PYAP-RO-l 115Revision IMODIFICATION RECORD:: SHORT FORMFOR: 1) Addenda to existing Modification Records (e.g., modifications of same nature as onespreviously reviewed and approved).2) Significant modification~s to the facility or facility systems that are not described in theHazards Summar~y Report.3) Modifications that require engineering decisions/implementation in a time frame thatprecludes 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 OperatorRequalification Program.The Reactor Safety Subcommittee will review these modifications.Modification Number: 9 1-3, Addendum 1Modification Title: .Replacement of the Existing X. Y. MH-X, and MH-Y Fuel Storaqie Baskets WithNew X and Y BasketsByPage No.2346page TitleModification Record: Short FormModification Description(Why. Short Form is appropriate)Hazards Summary Report EvaluationReactor Safety EvaluationOP, PM, CP, and Print EvaluationSpare Parts RequirementsRequiredYes Noox __ _DateCompletedXXXX'Cx___-_-2-ByiiasC- 50.59 Screen Completed: ,/' N.t o~i.m&66..Z //d,- /Q./(Asst. Reactor Manager -E&#xa3;'gineering)Reactor Safety Subcommittee Review:___________________(Asst. Reactor. Manager -Engineering)Modification A~pp~roved: erDate:________Date: 2 Date:________Modification Completed:(Reactor Manager)Attachment 8.1]
ATTACHMENT 6@&#xa9;[PYAP-RO-l 115Revision IMODIFICATION RECORD:: SHORT FORMFOR: 1) Addenda to existing Modification Records (e.g., modifications of same nature as onespreviously reviewed and approved).2) Significant modification~s to the facility or facility systems that are not described in theHazards Summar~y Report.3) Modifications that require engineering decisions/implementation in a time frame thatprecludes 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 OperatorRequalification Program.The Reactor Safety Subcommittee will review these modifications.Modification Number: 9 1-3, Addendum 1Modification Title: .Replacement of the Existing X. Y. MH-X, and MH-Y Fuel Storaqie Baskets WithNew X and Y BasketsByPage No.2346page TitleModification Record: Short FormModification Description(Why. Short Form is appropriate)Hazards Summary Report EvaluationReactor Safety EvaluationOP, PM, CP, and Print EvaluationSpare Parts RequirementsRequiredYes Noox __ _DateCompletedXXXX'Cx___-_ ByiiasC- 50.59 Screen Completed: ,/' N.t o~i.m&66..Z //d,- /Q./(Asst. Reactor Manager -E&#xa3;'gineering)Reactor Safety Subcommittee Review:___________________(Asst. Reactor. Manager -Engineering)Modification A~pp~roved: erDate:________Date: 2 Date:________Modification Completed:(Reactor Manager)Attachment 8.1]
ATTACHMENT 6AP-RO-] 115Revision IModification Number: 91-3. Addendum 1MODIFICATION DESCRIPTIONProvide a concise description of the system change. InclUde any proposed PRE-OPERATIONAL TESTS required for this change. (If additional pages are necessary, insert afterthis 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 locationscan 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 tothe following specifications:(a) A geometr~y such that the calculated Keff is less than 0.9 under all conditions of moderation andirrespective 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, orinterchange 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 MURRDesign Data, Volume I. Thermal shielding or appreciable water thickness must be provided around thespent fuel storage baskets to protect the magnetite concrete from damage due to thermal stresses andexcessive temperatures. The thermal shielding requirements are based radiation heating in themagnetite concrete and the resulting conditions within the concrete. The design criterion employed is thatthe 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) new20 element fuel baskets that will be designated X and Y. The new X basket will replace the old X and MH-Xbaskets and the new Y basket will replace the old Y and MH-Y baskets. Each new basketwill haveessentially 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 Xand Y baskets that will provide a storage location for either the OS basket or a Be reflector ring. A separateevaluation 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 anexisting, previously reviewed and approved Modification Record (9 1-3), "Temporary Additional In-Pool Fuel.Storage.Baskets."Attachment 8.1]
ATTACHMENT 6AP-RO-] 115Revision IModification Number: 91-3. Addendum 1MODIFICATION DESCRIPTIONProvide a concise description of the system change. InclUde any proposed PRE-OPERATIONAL TESTS required for this change. (If additional pages are necessary, insert afterthis 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 locationscan 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 tothe following specifications:(a) A geometr~y such that the calculated Keff is less than 0.9 under all conditions of moderation andirrespective 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, orinterchange 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 MURRDesign Data, Volume I. Thermal shielding or appreciable water thickness must be provided around thespent fuel storage baskets to protect the magnetite concrete from damage due to thermal stresses andexcessive temperatures. The thermal shielding requirements are based radiation heating in themagnetite concrete and the resulting conditions within the concrete. The design criterion employed is thatthe 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) new20 element fuel baskets that will be designated X and Y. The new X basket will replace the old X and MH-Xbaskets and the new Y basket will replace the old Y and MH-Y baskets. Each new basketwill haveessentially 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 Xand Y baskets that will provide a storage location for either the OS basket or a Be reflector ring. A separateevaluation 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 anexisting, previously reviewed and approved Modification Record (9 1-3), "Temporary Additional In-Pool Fuel.Storage.Baskets."Attachment 8.1]
V 1ATTACHMENT 6AP-RO-1 15Revision IModification Number: 9 1-3, Addendum 1MODIFICATION DESCRIPTION (con't)The MH--X and M-H-Y baskets were installed in 1991 as additional temporary fuel storage locationsduring a period when the facility was unable to ship. fuel .because twvo spent fuel shipping casks thatwere 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. TheMI-I-X and MH-Y baskets were designed and built for the MH IA shipping cask and were not intendedfor the everyday use that they have endured at MURR. Over the years, some of the boral andaluminum plates have swelled or warped making certain storage locations unusable. To ensure thatwe maintain m~axinmum fuel storage capability during periods of shipment uncertainties, the newbaskets were designed and constructed for everyday use, similar to that of the original X, Y, and Zstorage baskets. Additionally, the newly designed baskets will increase storage capacity from 36 to 40at these locations.Each design specification listed on page 2 will be addressed individually ini the applicable sections ofthis Modification Record. Specification (a) will be addressed in the Hazards Summary Report andReactor Safety Evaluation sections. Specification (b) will be addr'essed in the Reactor SafetyEvaluation section. Specification (c) will be addressed in the Hazards Sununary Report Evaluationsection. 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 ReportEvaluation 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 interchangeof fuel elements." The closest fuel storage location from the new baskets to the reactor pool wall isabout 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 thisModification 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 Zstorage 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 ofspecification (a). The materials of construction meet the requirements of specification (e), whichstates, "Fabrication from materials compatible with the fuel elements." Included in this ModificationRecord are the design and construction prints for the new baskets (MURR Drawing No. 2640). Alsoattached 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 forfuture reference.Neutron radiography of eight randomly selected boral sheets that was performed at the University ofCalifornia-Davis reactor indicated even dispersion of boron in the plates. These radiographs will also.,.,. ..*be i~naintained in Document Control.2aAttachment 8.1 ATTACHMENT 6AP-RO-l 115Revision 1Modification Number: 91-3, Addendum 1HAZARDS SUMVMARY REPORT EVALUATIONDoes this change involve a modification to the reactor facility as defined in the Hazards SummaryReport?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, outlinethe basis for the decision.This modification does not involve a change to the reactor facility as defined in the Hazards Summary Reportand 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. Shieldrequirements for fuel storage in the pool are calculated to meet the dose rate criteria of the bulk shieldinglisted 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 primaryreactor 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 theoutside Surface of the biological shield is five (5) feet. Additional design features that are more conservativethan those assumed in Section 6.4 include: (1) the closest section of the new fuel baskets is locatedapproximately 12-inches from the reactor pool wall (tapered section) and not immediately adjacent, (2) thenew baskets are in a configuration less than an eight element array, and (3) the current MURR fuel cycleresults irn irradiated elements with a much lower activity than the design basis fuel cycle of forty days ofcontinuous operation at 10 MW. Elements stored in the new X and Y baskets for greater than 24 hours willhave greater 1E6 seconds of decay (11.6 days). This storage time requirement will be administrativelycontrolled 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 theMonte 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 valueestimated by IMCNP for the new configuration is 0.635, with a standard deviation of 0.002 -well belowTechnical Specification 3.8.d limit of 0.9. Using the most conservative approach and assumptions, thebaskets were modeled using twenty (20) "fresh" 775 gram U-235 fuel elements -a far greater number ofelements than what we are allowed to possess under our current inventory license limits. Additionally, thevalue of boral used to model the baskets was 0.0624 grams of B-10 atems/cm2. None of the boral sheetsthat'were used in the construction of the baskets had a value less than 0.0709 gms/cm2, and the averagevalue of all sheets was 0.0740 gms/cm2. A I/M criticality determination will also be made upon installation ofthe baskets to verify the results of the MCNP modeling. A Keff value of 0.635 easily meets the requirementsof specification (a).Attachment 8.1 I ' IATTACHMENT 6AP-RO-1 15Revision 1Modification Number: 91-3. Addendum 1HAZARDS 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 shieldfingthicknesses .for spent fuel storage. Thermal shielding or appreciablk Water thickness must be providedaround 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 radiationheating in the magnetite concrete and the resulting conditions within the concrete. The designcriterion 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 fuelelements wiflh a decay time of 1E6 seconds (11.6 days) require a minimum of 14-inches of themtlawater shielding. The thickn~ess requirements presented in this table are based on a configuration of arow of eight elements stored adjacent to the biological shield. Page 5 of TM-RI-ID-62-9 also statesthat "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 shieldingwith 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 storedunless they have greater than 3E7 seconds (one year) of d~ecay. Fuel elements with this decay timehave a fission product activity, and hence gamma heating source, of approximately 1/20 of the activityof a fuel element 1E6 seconds of decay: This number was obtained from J. Huang's MasterThesis, 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 depictfuel elenment decay are included in Modification Record 91-3.3aAttachment 8.1 4 4ATTACHMENT 6AIP-RO-1 15Revision 1Modification Number: 9 1-3, Addendum 1REACTOR SAFETY EVALUATIONDoes this change involve a revision(s) to the Technical Specifications or a safety hazard as describedin 10 CFR 50:597? .NOTE: A licensee may make changes to the facility as described in the I-SR without obtaining a license amaendmentonly 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 inthe HSR;2. More than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important tosafety 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 safetypreviously 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 anypreviously 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 orthe safety analyses.Yes: ___No: ___ S igna tur e: _________-, ___________ Date: 2 If YES, the change must be performed igaLong Form Modification Record. If NO, outline thebasis forthe decision.This modification does not involve a ch'ange to the Technical Specifications or a safety hazard as describedin 10 CFR 50.59. A.50.59 Screen is attached and shows that the proposed activity does not have thepotential 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.dand 3.8.e.Technical Specification 3.8.d states that "All fuel elements or fueled devices outside the reactor core shall bestored in a geometry such that the calculated Keff is less than 0.9 under all conditions of moderation." Thebasis for this Specification states that this limit is conservative and assures safe fuel storage. The MCNPmodel was used to calculate a Keff value of 0.635 for one fuel basket fully loaded with twenty (20) "fresh" 775gram U-235 fuel elements. This predicted value is well below the Technical Specification limit of 0.9. Thisvalue 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 permitsufficient 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 withregard to natural convection cooling. This satisfies the requirements of specification (b) stated in theModification Description.*Attachment 8.1 ATTACHMENT 6AP-RO-1 15Revision 1Modification Number: 9 1-3. Addendum IREACTOR SAFETY EVALUATION (con't)Furthermore, the Safety Evaluation (SE) performed by the Test & Power Reactor Safety Branch of theDivision of Reactor Licensing, documented by letter dated July 27, 1966, was in response to therequest by the University of Missouri to operate the MURR at a power level of 5 MW. The SEidentified the safety criteria for fuel storage and handling, as providing assurance of not having acritical fuel con'figuration, even with th~e unlikely mishap that might occur during fuel handling. TheSE performed by the Directorate of Licensing, documented by letter dated May 24, 1974, supportedMURR's request to operate at the higher power level of 10 MW. This SE did not elaborate any furtheron 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 storageof spent fuel in accordance with the Technical Specifications."4aAttachment 8.1'I 1[ I *ATTACHMENT 6AP-RO- l 15Revision 1Modification Number: 91-3. Addendum 1OPERATING, PREVENTATIVE MAINTENANCE, AND COMPLIANCE PROCEDURE,AND PRINT EVALUATIONDoes this charnge require a revision(s) to any 0pcrating, Preventative Maintenance, o~r ComplianceProcedure, 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 ComplianceProcedure. 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 theseprocedures and form are listed below. New prints associated with the design and construction of the new fuelstorage 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 bestored in the following deep pool storage positions for longer than 24 hours:X-15, 16, 17, 18, 19, 20, andY-11, 16, 17, 18, 19, 202. 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 hours."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 bestored in the following deep pool storage positions for longer than 24 hours:X-15, 16, 17, 18, 19, 20, andY-11, 16, 17, 18, 19,202. 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 hours."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 slorageconfigurations.Attachmuent 8.1 11 tATTACHMENT 6AP-RO-1 115Revision IModification Number: 91-3, Addendum 1SPARE PARTS REQUIREMENTS EVALUATIONDoes 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.6Attachment 8.1 ATTACHMENT 650.59 SCREENAP-RR-003Revision 1Page 1 of 2Activity Screening Number: 0e/- /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 tworedesigned 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? "YEIf this question is answered yes, do not__ continue with this procedure. Identify and report the concern tothe Reactor Manager.50.59 Screening Questions:I. Does the proposed activity involvze a change to an SSC that adversely affects a design functiondescribed in the HSR?2. Does the proposed activity involve a change to aprocedure that adversely affects how HSRdescribed 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 NOYES NOYES, No4. Does the proposed activity involve a test or experiment not described in the HSR, where an SSC isutilized or controlled in a manner that is outside the reference bounds of the design for that SSC or YESis 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. ...INO/NOA LicenseIf 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. 3NOTE: 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 whererelevant information was found. Include section/Ipage numbers. Use page 2 of this fo omrtyouzra ements.SPrint Name. ."______e_-__-__DatPreparer: Edward L. Murphy kA bj //Reactor Manager: Les Foyto ) /-/- Attachment 1 A T A T~ ITENI tAP -RR-003R~evision 150.59 SCREEN (Cont.)Activity Screening Number: 0 O Y- Page 2of 2Title: REDSIGNED DEEP POOL FUEL STORAGE BASKETSIf t~he conclusion of the five.(5) Screening QUestions is that a 50.59 EvaIluation is no__t required, providejustification 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 theHSR?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 animproved manner, the same traction as those currently installed. No other systems or components will beby this modification .. ..2. Does the proposed activity involve a change tO a procedure that adversely affects how HrsR described SSC designfunctions 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 HSde.scribed RRC"desi g funnctions ,re. perfo rmned. cnntrolled or- tested3. Does the proposed activity involve revising or replacing an described evaluation methodology that is used inestablishing the design bases or used in the safety analyses?ino The _modi ca'tion to the deep pool fuel prnoro~ed designed using establishedT4qTRdescie~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 controlledin a manner that is outside tlie reference bounds of the design for that SSC, or-is inconsistent with analyses ordescriptions presented in the HSR?No. The redesigned deep pool fuel storage baskets are functionally and operationally the same as thosecurrently installed, and will be used and controlled only in a manner within design boundaries. All testsrequired for the proposed change are covered by, and are consistent with, all analyses and descriptionspresented in the tHSR.List the documents (HISR, Technjc?,l ...p e..[i~.c~atiopns, and other Licensing.-.asis..documents) reviewed whererelevant information was found. [Ihclude section / page numbers. ]1HSR Section 6.4 "Spent Fuel Transfer and Storage". HSR Section7..1 "Fuel Handling System". HSR Section13.2.1t1 "Refueling Accident". Technical Specification 3.8.d "Fuel Element Storage Geobmetry", TechnicalSpecification 3.8.e "Cooling Requirements for Fuel Element Storage". OP-RO-250 'Fuel Handling". RP-RO-100 "Fuel Movement"1 ATTACHMENT 6From: Das Kutikkad ..To: Les Foyto, Acting Reactor manager, MURRDate: June 05, 2003Re: Results of Calculations Performed to Estimate the Keff of the New Deep-Pool" Fuel Storage Bask~etCalculations were performed to estimate the criticality of the newly designed deep pool fuelstorage basket slated to replace the X, MHX, Y and MRY baskets. The model used and the resultsobtained are surmmarized in this memo.For the purpose of simplicity, only one of the new 20-element basket (on one side of the pool) wasmodeled. A drawing of the new basket is attached to this report. One such basket is expected toreplace the combined X & MHX or the combined Y & MHY storage locations. Since the two sidesof the pool are fairly decoupled neutronically (especially with the amount of boron in the storagebaskets), this modeling should be adequate to establish the safe storage requirement specified inthe Tech :Specs.Monte Carlo simulation program MCNP was used to model the new fuel storage basket and toestimate the criticality. Several conservative assumptions were used in the modeling such as usingall fresh fuel elements (no burn up credit taken) and using a reduced thickness for boral in theoutermost 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 usedis approximately 35 w% of B4C in boral (rest Al). For the new bask~et, we purchased boral that hasless boron content. The boral used has 0.0624 grams of B-10 atoms/cm2. For a boral sheet of0.265" thick (approx 0.67 cm), this translates to a B4C value of roughly 24 w%. The boron used isnatural and not enriched in B-10. The dimensions of the basket and the wall thickness are shownthe attached drawing.The Keff value (hie MCNP for this fuel stora ge coii~figiiration (loaded with fresh 775 gU235 fuel elements) was 0.635 with a standard deviation of 0.002. This result shows th~at it is safeto store fuel in the new basket with th~e predicted Keff well below the Tech Spec limit of 0.9.
V 1ATTACHMENT 6AP-RO-1 15Revision IModification Number: 9 1-3, Addendum 1MODIFICATION DESCRIPTION (con't)The MH--X and M-H-Y baskets were installed in 1991 as additional temporary fuel storage locationsduring a period when the facility was unable to ship. fuel .because twvo spent fuel shipping casks thatwere 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. TheMI-I-X and MH-Y baskets were designed and built for the MH IA shipping cask and were not intendedfor the everyday use that they have endured at MURR. Over the years, some of the boral andaluminum plates have swelled or warped making certain storage locations unusable. To ensure thatwe maintain m~axinmum fuel storage capability during periods of shipment uncertainties, the newbaskets were designed and constructed for everyday use, similar to that of the original X, Y, and Zstorage baskets. Additionally, the newly designed baskets will increase storage capacity from 36 to 40at these locations.Each design specification listed on page 2 will be addressed individually ini the applicable sections ofthis Modification Record. Specification (a) will be addressed in the Hazards Summary Report andReactor Safety Evaluation sections. Specification (b) will be addr'essed in the Reactor SafetyEvaluation section. Specification (c) will be addressed in the Hazards Sununary Report Evaluationsection. 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 ReportEvaluation 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 interchangeof fuel elements." The closest fuel storage location from the new baskets to the reactor pool wall isabout 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 thisModification 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 Zstorage 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 ofspecification (a). The materials of construction meet the requirements of specification (e), whichstates, "Fabrication from materials compatible with the fuel elements." Included in this ModificationRecord are the design and construction prints for the new baskets (MURR Drawing No. 2640). Alsoattached 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 forfuture reference.Neutron radiography of eight randomly selected boral sheets that was performed at the University ofCalifornia-Davis reactor indicated even dispersion of boron in the plates. These radiographs will also.,.,. ..*be i~naintained in Document Control.2aAttachment 8.1 ATTACHMENT 6AP-RO-l 115Revision 1Modification Number: 91-3, Addendum 1HAZARDS SUMVMARY REPORT EVALUATIONDoes this change involve a modification to the reactor facility as defined in the Hazards SummaryReport?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, outlinethe basis for the decision.This modification does not involve a change to the reactor facility as defined in the Hazards Summary Reportand 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. Shieldrequirements for fuel storage in the pool are calculated to meet the dose rate criteria of the bulk shieldinglisted 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 primaryreactor 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 theoutside Surface of the biological shield is five (5) feet. Additional design features that are more conservativethan those assumed in Section 6.4 include: (1) the closest section of the new fuel baskets is locatedapproximately 12-inches from the reactor pool wall (tapered section) and not immediately adjacent, (2) thenew baskets are in a configuration less than an eight element array, and (3) the current MURR fuel cycleresults irn irradiated elements with a much lower activity than the design basis fuel cycle of forty days ofcontinuous operation at 10 MW. Elements stored in the new X and Y baskets for greater than 24 hours willhave greater 1E6 seconds of decay (11.6 days). This storage time requirement will be administrativelycontrolled 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 theMonte 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 valueestimated by IMCNP for the new configuration is 0.635, with a standard deviation of 0.002 -well belowTechnical Specification 3.8.d limit of 0.9. Using the most conservative approach and assumptions, thebaskets were modeled using twenty (20) "fresh" 775 gram U-235 fuel elements -a far greater number ofelements than what we are allowed to possess under our current inventory license limits. Additionally, thevalue of boral used to model the baskets was 0.0624 grams of B-10 atems/cm2. None of the boral sheetsthat'were used in the construction of the baskets had a value less than 0.0709 gms/cm2, and the averagevalue of all sheets was 0.0740 gms/cm2. A I/M criticality determination will also be made upon installation ofthe baskets to verify the results of the MCNP modeling. A Keff value of 0.635 easily meets the requirementsof specification (a).Attachment 8.1 I ' IATTACHMENT 6AP-RO-1 15Revision 1Modification Number: 91-3. Addendum 1HAZARDS 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 shieldfingthicknesses .for spent fuel storage. Thermal shielding or appreciablk Water thickness must be providedaround 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 radiationheating in the magnetite concrete and the resulting conditions within the concrete. The designcriterion 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 fuelelements wiflh a decay time of 1E6 seconds (11.6 days) require a minimum of 14-inches of themtlawater shielding. The thickn~ess requirements presented in this table are based on a configuration of arow of eight elements stored adjacent to the biological shield. Page 5 of TM-RI-ID-62-9 also statesthat "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 shieldingwith 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 storedunless they have greater than 3E7 seconds (one year) of d~ecay. Fuel elements with this decay timehave a fission product activity, and hence gamma heating source, of approximately 1/20 of the activityof a fuel element 1E6 seconds of decay: This number was obtained from J. Huang's MasterThesis, 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 depictfuel elenment decay are included in Modification Record 91-3.3aAttachment 8.1 4 4ATTACHMENT 6AIP-RO-1 15Revision 1Modification Number: 9 1-3, Addendum 1REACTOR SAFETY EVALUATIONDoes this change involve a revision(s) to the Technical Specifications or a safety hazard as describedin 10 CFR 50:597? .NOTE: A licensee may make changes to the facility as described in the I-SR without obtaining a license amaendmentonly 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 inthe HSR;2. More than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important tosafety 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 safetypreviously 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 anypreviously 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 orthe safety analyses.Yes: ___No: ___ S igna tur e: _________-, ___________ Date: 2 If YES, the change must be performed igaLong Form Modification Record. If NO, outline thebasis forthe decision.This modification does not involve a ch'ange to the Technical Specifications or a safety hazard as describedin 10 CFR 50.59. A.50.59 Screen is attached and shows that the proposed activity does not have thepotential 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.dand 3.8.e.Technical Specification 3.8.d states that "All fuel elements or fueled devices outside the reactor core shall bestored in a geometry such that the calculated Keff is less than 0.9 under all conditions of moderation." Thebasis for this Specification states that this limit is conservative and assures safe fuel storage. The MCNPmodel was used to calculate a Keff value of 0.635 for one fuel basket fully loaded with twenty (20) "fresh" 775gram U-235 fuel elements. This predicted value is well below the Technical Specification limit of 0.9. Thisvalue 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 permitsufficient 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 withregard to natural convection cooling. This satisfies the requirements of specification (b) stated in theModification Description.*Attachment 8.1 ATTACHMENT 6AP-RO-1 15Revision 1Modification Number: 9 1-3. Addendum IREACTOR SAFETY EVALUATION (con't)Furthermore, the Safety Evaluation (SE) performed by the Test & Power Reactor Safety Branch of theDivision of Reactor Licensing, documented by letter dated July 27, 1966, was in response to therequest by the University of Missouri to operate the MURR at a power level of 5 MW. The SEidentified the safety criteria for fuel storage and handling, as providing assurance of not having acritical fuel con'figuration, even with th~e unlikely mishap that might occur during fuel handling. TheSE performed by the Directorate of Licensing, documented by letter dated May 24, 1974, supportedMURR's request to operate at the higher power level of 10 MW. This SE did not elaborate any furtheron 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 storageof spent fuel in accordance with the Technical Specifications."4aAttachment 8.1'I 1[ I *ATTACHMENT 6AP-RO- l 15Revision 1Modification Number: 91-3. Addendum 1OPERATING, PREVENTATIVE MAINTENANCE, AND COMPLIANCE PROCEDURE,AND PRINT EVALUATIONDoes this charnge require a revision(s) to any 0pcrating, Preventative Maintenance, o~r ComplianceProcedure, 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 ComplianceProcedure. 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 theseprocedures and form are listed below. New prints associated with the design and construction of the new fuelstorage 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 bestored in the following deep pool storage positions for longer than 24 hours:X-15, 16, 17, 18, 19, 20, andY-11, 16, 17, 18, 19, 202. 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 hours."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 bestored in the following deep pool storage positions for longer than 24 hours:X-15, 16, 17, 18, 19, 20, andY-11, 16, 17, 18, 19,202. 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 hours."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 slorageconfigurations.Attachmuent 8.1 11 tATTACHMENT 6AP-RO-1 115Revision IModification Number: 91-3, Addendum 1SPARE PARTS REQUIREMENTS EVALUATIONDoes 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.6Attachment 8.1 ATTACHMENT 650.59 SCREENAP-RR-003Revision 1Page 1 of 2Activity Screening Number: 0e/- /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 tworedesigned 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? "YEIf this question is answered yes, do not__ continue with this procedure. Identify and report the concern tothe Reactor Manager.50.59 Screening Questions:I. Does the proposed activity involvze a change to an SSC that adversely affects a design functiondescribed in the HSR?2. Does the proposed activity involve a change to aprocedure that adversely affects how HSRdescribed 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 NOYES NOYES, No4. Does the proposed activity involve a test or experiment not described in the HSR, where an SSC isutilized or controlled in a manner that is outside the reference bounds of the design for that SSC or YESis 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. ...INO/NOA LicenseIf 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. 3NOTE: 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 whererelevant information was found. Include section/Ipage numbers. Use page 2 of this fo omrtyouzra ements.SPrint Name. ."______e_-__-__DatPreparer: Edward L. Murphy kA bj //Reactor Manager: Les Foyto ) /-/- Attachment 1 A T A T~ ITENI tAP -RR-003R~evision 150.59 SCREEN (Cont.)Activity Screening Number: 0 O Y- Page 2of 2Title: REDSIGNED DEEP POOL FUEL STORAGE BASKETSIf t~he conclusion of the five.(5) Screening QUestions is that a 50.59 EvaIluation is no__t required, providejustification 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 theHSR?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 animproved manner, the same traction as those currently installed. No other systems or components will beby this modification .. ..2. Does the proposed activity involve a change tO a procedure that adversely affects how HrsR described SSC designfunctions 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 HSde.scribed RRC"desi g funnctions ,re. perfo rmned. cnntrolled or- tested3. Does the proposed activity involve revising or replacing an described evaluation methodology that is used inestablishing the design bases or used in the safety analyses?ino The _modi ca'tion to the deep pool fuel prnoro~ed designed using establishedT4qTRdescie~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 controlledin a manner that is outside tlie reference bounds of the design for that SSC, or-is inconsistent with analyses ordescriptions presented in the HSR?No. The redesigned deep pool fuel storage baskets are functionally and operationally the same as thosecurrently installed, and will be used and controlled only in a manner within design boundaries. All testsrequired for the proposed change are covered by, and are consistent with, all analyses and descriptionspresented in the tHSR.List the documents (HISR, Technjc?,l ...p e..[i~.c~atiopns, and other Licensing.-.asis..documents) reviewed whererelevant information was found. [Ihclude section / page numbers. ]1HSR Section 6.4 "Spent Fuel Transfer and Storage". HSR Section7..1 "Fuel Handling System". HSR Section13.2.1t1 "Refueling Accident". Technical Specification 3.8.d "Fuel Element Storage Geobmetry", TechnicalSpecification 3.8.e "Cooling Requirements for Fuel Element Storage". OP-RO-250 'Fuel Handling". RP-RO-100 "Fuel Movement"1 ATTACHMENT 6From: Das Kutikkad ..To: Les Foyto, Acting Reactor manager, MURRDate: June 05, 2003Re: Results of Calculations Performed to Estimate the Keff of the New Deep-Pool" Fuel Storage Bask~etCalculations were performed to estimate the criticality of the newly designed deep pool fuelstorage basket slated to replace the X, MHX, Y and MRY baskets. The model used and the resultsobtained are surmmarized in this memo.For the purpose of simplicity, only one of the new 20-element basket (on one side of the pool) wasmodeled. A drawing of the new basket is attached to this report. One such basket is expected toreplace the combined X & MHX or the combined Y & MHY storage locations. Since the two sidesof the pool are fairly decoupled neutronically (especially with the amount of boron in the storagebaskets), this modeling should be adequate to establish the safe storage requirement specified inthe Tech :Specs.Monte Carlo simulation program MCNP was used to model the new fuel storage basket and toestimate the criticality. Several conservative assumptions were used in the modeling such as usingall fresh fuel elements (no burn up credit taken) and using a reduced thickness for boral in theoutermost 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 usedis approximately 35 w% of B4C in boral (rest Al). For the new bask~et, we purchased boral that hasless boron content. The boral used has 0.0624 grams of B-10 atoms/cm2. For a boral sheet of0.265" thick (approx 0.67 cm), this translates to a B4C value of roughly 24 w%. The boron used isnatural and not enriched in B-10. The dimensions of the basket and the wall thickness are shownthe attached drawing.The Keff value (hie MCNP for this fuel stora ge coii~figiiration (loaded with fresh 775 gU235 fuel elements) was 0.635 with a standard deviation of 0.002. This result shows th~at it is safeto store fuel in the new basket with th~e predicted Keff well below the Tech Spec limit of 0.9.
IIIIUBII19.365II II UB11HB!I/
IIIIUBII19.365II II UB11HB!I/
I IJYCF('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"long4 24 smallboral 0.265" 3oB4C boral 4.4375"x 30" J3L1-c C52AL sheet2 1 / :!:3003H 14 Alum. 19.615" x 33.25" x 1/8"6 48 Aluminum stock 1-3/8" Aluminum stock7 10 Aluminum stock 1/4" x24.14" x 1-3/8" Aluminum stock2314,)ir,, ,. .." ",,.i .... >'"" " "* :K.v. ". E',II /irdl ii ii ii ii  
I IJYCF('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"long4 24 smallboral 0.265" 3oB4C boral 4.4375"x 30" J3L1-c C52AL sheet2 1 / :!:3003H 14 Alum. 19.615" x 33.25" x 1/8"6 48 Aluminum stock 1-3/8" Aluminum stock7 10 Aluminum stock 1/4" x24.14" x 1-3/8" Aluminum stock2314,)ir,, ,. .." ",,.i .... >'"" " "* :K.v. ". E',II /irdl ii ii ii ii  
*ATTACHMENT 6'rodelling of the new 20-element deep pool fuel storage basket6c this first run is a case with just the new 20-element basketc modelled (as a replacement for the existing mhy and y baskets).c subsequent runs will add the old beryllium next to this storagec basket (in place where os basket was before) to see its effect.c the core is not modelled in this case, so the storage basketc -is a stand alone basket filled with fresh fuel elements.c- " " &deg;c a single fuel el'ement is defined and the "repeated structure" featurec of mcnp is used to construct the storage positions (bins).c some conservatiSm is used during the initial runs. some of these willc he removed during later runs if the keff is found to be unacceptablec (i.e, >0.9) -some of the conservative assumptions are listed below:cc 1) all fresh fuel considered -i.e., no burnup ctedit takenc 2) less boral thickness for the outermost layers.cc individual "bins" of the new basket are described in an auxiliaryc coordinate system. the origin of this auxiliary coordinate systemc is at the center of individual bins. These are then tranformedc into the main system centered at one corner of the basket. all thec bins are filled with the same 'universe" (i.e., one fresh fuel-c element plus water, aluminum and boral surrounding the fuel).cc ** histories tracked = 100,000 for this case ***c1 1 -1.0 (-l40:-146:150:144:-148:149) 130 -151 132 -153 154 -135imp:n=l $ water surrounding the new basket (approx 30 cm thick)S the following four cells are created since mcnp doesn't like toc complicate any one cell too much. to avoid that problem, the newc 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 onec basket (thereby complicating that one cell too much).c2 2 -2.7 140 -150 146 -141 148 -149 #20 #21 #22 #23 #24imp:n=l $ basket that contins t "bins" along x-axis3 2 -2.7 140 -150 141 -142 148 -149 #25 #26 #27 #28 #29imp:n=l $ basket that contins 5 "bins" along x-axis4 2 -2.7 140 -150 142 -143 148 -149 #30 #31 #32 #33 #34imp tn=l ** basket that cont-ins-.5 ''bins" along x-axis 5 2 -2.7 "--150 143 -144 ...48 ...1349 #35 #36 #37 #38 #39imp:n=l $ basket that contins 5 "bins" along x-axisc6 0 -130:-132:151:153:-154:135 imp:n=0 $ outside worldc7 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 bins9 2 -2.7 209 -205 206 -207 u=1 imp:n=1 $ al of the bins10 2 -2.7 208 -209 206 -210 u=1 imp:n=l $ al of the bins11 2 -2.7 208 -209 211 -207 u=l imp:n=l $ al of the binscc although i~nfinite in dimension, banal-..thickness will be limitedc by the'.diiuensions;.of the cell isrille'd with this "universe i".c12 0 208 -209 210 -211 u=1 imp:n=l fill=2 (-11.00 0 0)c20 0 200 -201 202 -203 148 -149 imp:n=l trcl=20 fill=1i above is the definition of a single "bin" that is repeated 20 timesc iATTACHMENT 62124252627282930313233343536373839clikelikelikelikelikelikelikelikelikelikelikelikelikelikelikelikelikelikelike20202020202020202020202020202020202020butbutbutbutbutbutbutbutbutbutbutbutbutbutbutbutbutbutbuttrcl=21trcl=22trcl=2 3trcl=24trcl=25trcl=2 6trcl=27trcl=2 8trcl=30trcl=3 1trcl=32trcl=33trcl=34trcl=35trcl=36trcl=37trcl=38trcl=39c description-of the fuel plates of a single element starts from herec this is the "universe-2" that fills the new storage basket bins.c1011021031041051061.07110283284285286287288289290291c292293294295ccC296297c-2.7 -4-2.7 -5-2.7 -5-2.7 -6-2.7 -6-3.88 -5-2.7 -6-1.00 -7-2.7 -8-2.7 -9-2.7 -94-1.00 -95-2.7 -96-2.7 -97-2.7 -97-2.7 -98-2.7 -98-3.88 -97-2*7. -98-1:.0 -98-1. 0 -98+0.0803 -98+0.0803 -983443345678-93949596969595969733331021021021011041,02102104102102102104102102102101104102102103106105108-101 -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...12.5imp :n=!imp :n1limp :n=1imp :n=1imp:limp:n=1imp :nfl1imnp:n=limp:n=limp: n~lamp :nfl1im~p~n=limp:n=1imp:n=1imp: n=limp:n=limp:n=1imp:n=1imp:n=limp:n=1imp:n=limp:n=1imp: n 1u=2u= 2u=2u=2U=2U=2U=2u= 2U= 2U=2U=2U=2U=2U=2U=2U=2U=2U=2U=2u= 2U= 2U=2$ p1$ p1$ p1$ p1$ p1$ p1$ p1$ p1$ p1$ p1$ p1$ p1$ pl$ p1$ p1$ p1$ p1$ p1$ p11111111222232424242424242424cladcladcladclad on topclad on hotfuelcladwg cladcladcladSwgScladScladScladSclad on topSclad on botSfuellclad4-k-105-104-107-106-124-124.-124-124125125125125$$$$fuel top waterfuel hot waterfuel top hangerfuel hot hangerside plates of the element are described next22-2 .7-2.7-98 3 108 -107 -122 124-98 3 108 -107 -125 123imp:n=l u~=2imp:n=1 u=2$ side plate noAl$ side plate no.2c the water surrounding the f-uel described next, this willc become a. fini~e amount of watet 'this "universe 2" (singlec fresh fuel element plus the water surrounding it) is filled in cellsc representing the "bins" of the new storage basket.c2981 -1.0 107:98:-123:-3:122:-108 u=2 imp:n=l $water surrounding fuelc * *
*ATTACHMENT 6'rodelling of the new 20-element deep pool fuel storage basket6c this first run is a case with just the new 20-element basketc modelled (as a replacement for the existing mhy and y baskets).c subsequent runs will add the old beryllium next to this storagec basket (in place where os basket was before) to see its effect.c the core is not modelled in this case, so the storage basketc -is a stand alone basket filled with fresh fuel elements.c- " " &deg;c a single fuel el'ement is defined and the "repeated structure" featurec of mcnp is used to construct the storage positions (bins).c some conservatiSm is used during the initial runs. some of these willc he removed during later runs if the keff is found to be unacceptablec (i.e, >0.9) -some of the conservative assumptions are listed below:cc 1) all fresh fuel considered -i.e., no burnup ctedit takenc 2) less boral thickness for the outermost layers.cc individual "bins" of the new basket are described in an auxiliaryc coordinate system. the origin of this auxiliary coordinate systemc is at the center of individual bins. These are then tranformedc into the main system centered at one corner of the basket. all thec bins are filled with the same 'universe" (i.e., one fresh fuel-c element plus water, aluminum and boral surrounding the fuel).cc ** histories tracked = 100,000 for this case ***c1 1 -1.0 (-l40:-146:150:144:-148:149) 130 -151 132 -153 154 -135imp:n=l $ water surrounding the new basket (approx 30 cm thick)S the following four cells are created since mcnp doesn't like toc complicate any one cell too much. to avoid that problem, the newc 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 onec basket (thereby complicating that one cell too much).c2 2 -2.7 140 -150 146 -141 148 -149 #20 #21 #22 #23 #24imp:n=l $ basket that contins t "bins" along x-axis3 2 -2.7 140 -150 141 -142 148 -149 #25 #26 #27 #28 #29imp:n=l $ basket that contins 5 "bins" along x-axis4 2 -2.7 140 -150 142 -143 148 -149 #30 #31 #32 #33 #34imp tn=l ** basket that cont-ins-.5 ''bins" along x-axis 5 2 -2.7 "--150 143 -144 ...48 ...1349 #35 #36 #37 #38 #39imp:n=l $ basket that contins 5 "bins" along x-axisc6 0 -130:-132:151:153:-154:135 imp:n=0 $ outside worldc7 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 bins9 2 -2.7 209 -205 206 -207 u=1 imp:n=1 $ al of the bins10 2 -2.7 208 -209 206 -210 u=1 imp:n=l $ al of the bins11 2 -2.7 208 -209 211 -207 u=l imp:n=l $ al of the binscc although i~nfinite in dimension, banal-..thickness will be limitedc by the'.diiuensions;.of the cell isrille'd with this "universe i".c12 0 208 -209 210 -211 u=1 imp:n=l fill=2 (-11.00 0 0)c20 0 200 -201 202 -203 148 -149 imp:n=l trcl=20 fill=1i above is the definition of a single "bin" that is repeated 20 timesc iATTACHMENT 62124252627282930313233343536373839clikelikelikelikelikelikelikelikelikelikelikelikelikelikelikelikelikelikelike20202020202020202020202020202020202020butbutbutbutbutbutbutbutbutbutbutbutbutbutbutbutbutbutbuttrcl=21trcl=22trcl=2 3trcl=24trcl=25trcl=2 6trcl=27trcl=2 8trcl=30trcl=3 1trcl=32trcl=33trcl=34trcl=35trcl=36trcl=37trcl=38trcl=39c description-of the fuel plates of a single element starts from herec this is the "universe-2" that fills the new storage basket bins.c1011021031041051061.07110283284285286287288289290291c292293294295ccC296297c-2.7 2.7 2.7 2.7 2.7 3.88 2.7 1.00 2.7 2.7 2.7 -94-1.00 -95-2.7 -96-2.7 -97-2.7 -97-2.7 -98-2.7 -98-3.88 -97-2*7. -98-1:.0 -98-1. 0 -98+0.0803 -98+0.0803 -983443345678-93949596969595969733331021021021011041,02102104102102102104102102102101104102102103106105108-101 -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...12.5imp :n=!imp :n1limp :n=1imp :n=1imp:limp:n=1imp :nfl1imnp:n=limp:n=limp: n~lamp :nfl1im~p~n=limp:n=1imp:n=1imp: n=limp:n=limp:n=1imp:n=1imp:n=limp:n=1imp:n=limp:n=1imp: n 1u=2u= 2u=2u=2U=2U=2U=2u= 2U= 2U=2U=2U=2U=2U=2U=2U=2U=2U=2U=2u= 2U= 2U=2$ p1$ p1$ p1$ p1$ p1$ p1$ p1$ p1$ p1$ p1$ p1$ p1$ pl$ p1$ p1$ p1$ p1$ p1$ p11111111222232424242424242424cladcladcladclad on topclad on hotfuelcladwg cladcladcladSwgScladScladScladSclad on topSclad on botSfuellclad4-k-105-104-107-106-124-124.-124-124125125125125$$$$fuel top waterfuel hot waterfuel top hangerfuel hot hangerside plates of the element are described next22-2 .7-2.7-98 3 108 -107 -122 124-98 3 108 -107 -125 123imp:n=l u~=2imp:n=1 u=2$ side plate noAl$ side plate no.2c the water surrounding the f-uel described next, this willc become a. fini~e amount of watet 'this "universe 2" (singlec fresh fuel element plus the water surrounding it) is filled in cellsc representing the "bins" of the new storage basket.c2981 -1.0 107:98:-123:-3:122:-108 u=2 imp:n=l $water surrounding fuelc * *
* end of cell definitions * *
* end of cell definitions * *
* need the following blank line.
* need the following blank line.
ATTACHMENT 6Cc34595969798Cc101.102103104105106107108cC120121c122,Q3'&#xa3;24125126127cc1301 31132133134135c1401411421.43144145146147148149c150151152153.54cthe following surfaces define the single fuel element (origin of thiscoordinate system different from the main one and from the auxiliarycoordinate system # 1 that defines the bins of *the new basket).czczczczcz"czcz99i00pzpzpzpzpzpzpzpz7.036 $ plate 17.074 ,7.12514.630 $ plate14.66814.71914.757cz 14.986.pz 030.48-30.4832.385-32.38538.10-38.1041,275-41.27524p -0.4142 1 0p 0.4142 1 0PppppppxpxPYpypz.pzpxpypyPYpypxpypypzp7pxpxPYPYpz-0.41420.4142-0.41420.4142-0.41420.4142-30.068.10-30.080.80-80.080.00.013 .0026.0039.0052 .0038.100.050. 80-42.2542 .2564. 5094. 5055.5082.00-50.001"11110000000.0 $0.0 .$-0 .05500.0550-0.46740.4674-0 .65980.6598$$$$$$partpartpartpartpartpartofofofofofof+2'2 .5-22.5+22 .5-22.5+22.5-22 :522.5 degree plane-22.5 degree planedegreedegreedegreedegreedegreedegreesidesidesidesidesidesideplateplateplateplateplateplate ATTACHMVLENT 6c the following surfaces define the auxiliary coordinate system thatdefines the individual bins of the new basket, its origin is at thec" center of each bin, a coordinate transformation then connectsc this to the main coordinate system -which is centred at one of thec corners of the new basket.c2O00201202203204205206207208209210211cpxpxpypypxpxpyPYpxpxPYPY-6.386.38z.6 386.38-6.076 .07-6.076 .07-5.755.75-5 .755.75Rodeccckcodecmlrmt1m2m3c,cccCccm6mt6rotm9cctr2Otr2 1tr2 2tr2 3tr2 4t r2 5tr26tr2 7tr2 8tr29tr30tr31tr3 2r3 3cr34nimp:n 1 18r2000 0.8 51001.5Cc .6667lwtr. 0113027.50c 113027.50c *-.60092235.50c -.37292238.50c -.028mn4 4009.50cmt4 be. Olm5 6012.50c1001. 50c8016.5Cc13027.5Ccmut5 grph. Oi1001.5Cc 0.4170lwtr. Olin8 -.. 26000.5Cc5010.50c -0.0356000.50c -0.0520.01 197r55 3000 08016.50c .33331-.85-.002-.016-.1328016 .50c0.20813027.50c 0.37500.24.,... ,000..50lc 0.2 28000.50c 0.15011.56c -0.15013027.50c -0.763 $ boral with 24 w% b4c6.4019.2032.0044.8057.606.4019.2032.0044.8057.606.4019.2032.0044.8057.606.50 0.06.50 0.06.50 0.06.50 0.06.50 0.019.50 0.019.50 0.01.9.50 0.0 ..."19.50 0.0"19.50 0.032.50 0.032.50 0.032.50 0.032.5C 0.032.50 0.0 tr35tr38tr39Cphys : nprintctinecut :fl1prdinpksrcATTACHMENT 66.40 45.50 0.019.20 45.50 0.032.00 45.50 0.044.80 45.50 0.057.60 45.50 0.0S15.0 0.0 $cross Sections above 15.0 mev will be expunged40 50 60 905000.j o .o -0.5j 2010.0 4.0 0.08.0 16.5 0.03.0 30.0 0.03.0 3.50 0.0110 120 126-0.118.021.017 .04.002.020.933 .02 .000.00.00.'00.034.030.032.0-3.504.518.029.0-3 .500.00.. 00.00.055.047 .059.01.508.022.034.0-1i.000.00.00.00.0 S IATTACHMENT 6OAAR CAR GO SYSTEMSa division of AAR Manufacturing, Inc..Jeff Moore, Sr. Manager -Nuclear Products12633 lnkster Road, Livonia, MI 48150-2272 USAPhone: (734) 522.2000 Direct: (734) 466-8110FAX: (734) 622-2240"email: jmoore@aarcorp.comI. CUSTOMER:A. NAME:B, REQUEST DATE:II. DATESA. CURRENT DATE:B. QUOTATION VALID FOR:III. CONTACTSA. AAR CONTACTI. NAME:2. TITLE:3; PHONE:4. FACSIMILE:B. CUSTOMER CONTACT1. NAMvE:2 PHONE:3. FACSIMILE:IV. sPECIFICATION AND PRICINGUniversity of MissouriApril 8, 2003April 30, 200390 daysJeff MooreSr. Manager, Nuclear Products(734) 522-2000 x 8110(734) 522-2240.Mr. Jeff Attebery1-573-882-52691-573-882-6360+ Shipping to Univ. of Missouri 4i%V. DELIVERIES TO COMMENCE:60 days AROVI. TERMSA.B.DELIVERY POINT:PAYMENT:FOB University of MissouriNet 30 da~isVII. SPECIAL INSTRUCTIONS None ATTACHMENT 6AAR? CARGO SYSTEMS0A division of AAR Manufacturing Group, INC.CERTIFICATE OF COMPLIANCECUSTOMER: University of MissouriQTY. SHIPPED: 68 pcs.DATE OF SHIPMENT: July 18, 2003CUSTOMER P.O. NUMBER: COO000009743AAR CARGO SYSTEMS SALES ORDER NUMBER: 5053667This is to certify' that the material supplied hereunder has been finspected and tested inaccordance with AAR-1 1002 QAP, Revision 23 dated November 7, 2002, and A.AR-10012QAP, Revision 18 dated April 9, 2003, and Nuclear Quality Program Manual, Revision 29and meets the requirements of the. purchase order. The Code of Federal Regulations10OCFR5O Appendix B and 10OCFR2 1 are applicable to the material on this order.SIGNATURE:TITLE:DATE:Phill PusiloLab ManagerJuly 18. 2003Appendix CAAR-1 0012 QAPPage 1 of 1...systems, components & more12633 Inkster Road Livonia Michigan 48150-2272 USATelephone 1-734-522-2000 Faxc 1-734-522-2240 ATTACHMENT 6AA R CA RGO0 S YSTEMSA div'ision of AAR Manufacturing Group, INC.BORAL DATA PACKAGE RECORD CHECKLISTSPECIFICATION: AAR-10012 OAP. REV. 17 D OCUMENTCHECKEDBYDATE"Record ChecklistCertificate of ComplianceInspection Data SheetsMaterial CertificationsJP/KEJP/KiEJP/KEJP/KEJP/KEJPIKEJP/KE7-18-037-1 8-037-12-037-1 8-037-18-037-18-037-18-03-BoraI Summary Report,Boxing ListCalibrated Equipment Data SheetREVIEWED BY:TITLE:DATE:Phill PusiloLab ManaaerJuly 18, 2003APPENDIX DAAR-10012 QAIPPAGE 1 OF I..systems, components & more12633 Inkster Road Livonia Michigan 48150-2272 USATelephone 1-734-522-2000 Fax 1-734-522-2240 tATTACHMENT 6Boral Summary Report (Pass)Job Name: University5O 5053667Serial NumberWM010013-3A 'WM010014-IBwM010015-3A.WM010016-IAWM010017-2AY(MO1001 8-2AYM0 10019-1BYM010020-8BYM010O21 -SBYMI f00022-SBYM010023 ,8AYM01]0024 -8 BLot NumberM-21 SM-2 ISM-2 I8M-21814-218M4-220M-220M-220M-220M-220M-220M-22010B gmns/em20.07400.07210,07090.07660.07260.0754,0.07580.07310.07480.07500.07330.0742Density,2.57312.55072.54742.54842.54272.65822.56322.57812.57282.56232.58472,5873Reviewed By: Phill PusiloTitle: Lab ManagerDate: 8/11/2003Appendix-AAAR10012QAPPagc: 1Ptss ATTACHMENT 6MATERIAL TRACEABILETY bY B ORAL SERIAL NUMBERS.0. # 50536.67 University of Missouri--..~ k-.- WM010013 through WM010017 M-218YM010018 through YM010024 M-220WM010013 through YM010024 "AL03-03WM010013 through YM0 10024 "3-045-C ATTACHMENT 7Volume of the Primary Coolant SystemIn-Pool Portion Mechanical Equipment Room 114 PortionScin Area Length Volume Scin Area Length VolumeScin (ft2) (ft) (ft3) Scin (ft2) (ft) (ft3)135(5) 0.7773 3.828 2.976 133(2-3) 0.7773 10.194 7.924135(6) 0.7773 3.708 2.882 135(1) 0.7773 2.000 1.555135(7) 0.7773 3.708 2.882 0.7773 22.374 17.391137 0.7773 3.250 2.526 133(7-5) 0.7773 22.374 17.391139 0.7773 2.500 1.943 133(4-3) 0.7773 14.290 11.108501 0.6048 5.937 3.591 133(2-1) 0.7773 15.584 12.113575 0.6048 2.269 1.372 132 0.6948 6.000 4.169100(2) 0.7773 4.917 3.822 131(3-2) 0.6948 4.969 3.452100(3) 0.7773 4.917 3.822 115(3-2) 0.6948 14.968 10.400101 0.7773 1.000 0.777 115(1) 0.4948 2.000 0.990102(1) 0.7773 1.000 0.777 111(7) 0.6948 4.189 2.911102(2) 0.7773 3.806 2.958 111(6) 0.6948 6.667 4.632102(3) 0.7773 3.806 2.958 111(2-5) 0.6945 16.264 11.295102(4) 0.7773 3.806 2.958 111(1) 0.6948 2.167 1.506102(5) 0.7773 5.097 3.962 105(9) 0.7773 2.167 1.684401(1) 0.2006 3.975 0.797 105(7-8) 0.7773 17.312 13.457401(2) 0.2006 3.975 0.797 105(5-6) 0.7773 15.542 12.081405(1) 3.2150 0.500 1.608 105(1-4) 0.7773 31.832 24.743405(2) 0.1389 4.708 0.654 102(7) 0.7773 2.000 1.555405(3) 0.2006 9.163 1.838 102(5-6) 0.7773 10.194 7.924460 1.3960 4.242 5.922 Total Piping Volume (ft3) 223.860406 0.7773 2.500 1.943407 0.7773 2.333 1.813 Total Piping Volume (gallons) [1,674.585Fuel Region (gallons)7.176Primary Circulation Pumps (gallons) 25.000Primary Heat Exchangers (gallons) 150.000Pressurizer (gallons) 150.000Total Volume of PCS (gallons) 2,006.761 eCFR -- tode of Federal Regulations" ATTACHMENT 8http://www.ecfr'.gov/cgi-birltext-idx?SlD=a6ddafde7f67322376d64cb..ELECTRONIC CODE OF FEDERAL REGULATIONSe-CFR data is current as of September 21, 2015Title 10 .- Chapter III --, Part 835 --. Subpart N -* Appendix-Title 10: EnergyPART 835--OCCUPATIONAL RADIATION PROTECTIONSubpart N-Emergency Exposure SituationsAPPENDIX C TO PART 835-DERIVED AIR CONCENTRATION (DAC) FOR WORKERS FROM EXTERNAL EXPOSURE DURINGIMMERSION IN A CLOUD OF AIRBORNE RADIOACTIVE MATERIALa. The data presented in appendix C are to be used for controlling occupational exposures in accordance with&sect;835.209, identifying the need for air monitoring in accordance with &sect;835.403 and identifying the need for posting ofairborne radioactivity areas in accordance with &sect;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) peryear. 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. Thedata are listed by radionuclide in order of increasing atomic mass. The air immersion DACs were calculated for acontinuous, nonshielded exposure via immersion in a semi-infinite cloud of airborne radioactive material. The DACs listedin 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 ofthe ratio of the observed concentration of a particular radionuclide and its corresponding DAC for all radionuclides in themixture. If this sum exceeds unity (1), then the DAC has been exceeded. For unknown radionuclides, the most restrictiveDAC (lowest value) for those isotopes not known to be absent shall be used.AIR IMMERSION DACRadlonlucllde aflf piL)(.qm)Ar-37 __5._2_d_____00__,_____At-39 269___yr___E-__3__ E+__ 7Ar-41 157h3-6IEOK~r-74 1. an3-6lE0Kr-76 __4____h____-__5____+_ 5Kr-79 __5.____h___E-__5______ 5Kr-81 __.__+05_y______-____E_07Kr-83m 1.83 h 7E-02 2E+09Kr-85 10.72 yr 7E-04 2E+07Kr-85m 4.48 h 2E-05 IE+06Kr-87 76.3 mni 4E-06 1E+05Kr-88 2.84 h 1E-06 7E+04.Xe-120 40.0 min 1E-05 4E+05X(e-121 40.1 mai 2E-06 BE+04Xe-122 20.1 h 8E-05 3E+06Xe-123 2.14 h 6E-06 2E+05Xe-125 16.8 h lE-05 SE+05Xe-127 36.406 d 1 E-05 8E&#xf7;05Xe-129m 8.86 d 2E-04 7E+06Xe-1 31m 11.84 d 5E-04 1E+07Xe-133 5.245 d 1 E-04 5E+08Xe-133m 2.19 d 1 E-04 5E+06Xe-135 9.11 hi 1E-O5 6E+05Xe-1 35m 15.36 rni 1 E-05 3E+05Xe-138 14.13 min 3E-06 IE+05For any single radlonuclide not listed above with decay mode other than alpha emission or spontaneous fission andI of 2I of291231201 PM eCFR -Code of Federal RegulationsATTACHMENT 8http://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/m3).[72 FR 31940, June 8, 2007, as amended at 76 FR 20489, Apr. 13, 2011]Need assistance?!of 29/23/2015 4:37 PM Case Summary of Containment ShineATTACHMENT 9Page 1 of 3~MicroShield 8.02Nathan Hogue (8.00-0000)Date I By ChkeFilename IRun Date I Run Time I DurationContainl1.msd September 29, 2015 1:21:55 PM 00:00:00Project InfoCase Title Containment ShineDescription IFuel Accident AnalysesGeometr 13 -Rectangular VolumeSource DimensionsLength 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 zSin) in) in)___________ShieldsShield N J ~Dimension Material Density ____________Source 6.12e+09 cm3  I Air 0.00122Shield 1 j 30.5 cm I Concrete 2.35Air Gap j Air 0.00122Source Input: Grouping Method -Standard IndicesNumber of Groups: 25Lower Energy Cutoff: 0.015Photons < 0.015: Included_______________________ Library: Grove _______ _______Nuclide Ci Bq 3  B q/cm3I- 131 8.9329e+000 3.3052e+01 1 1.4600e-003 5.4020e+001I- 132 2.4168e+001 8.9421 e+011 3.9500e-003 1.4615e+002I- 133 5.0905e+001 1.8835e+012 8.3200e-003 -3.0784e+002I- 134 5.2252e+001 1 .9333e+01 2 8.5400e-003 3.1598e+0021-135 4.5644e+00 1 1 .6888e+0 12 7.4600e-003 2.7602e+002IK"-85 2.2271 e-003 8.2403e+007 3.6400e-007 1 .3468e-002Kr-85m 1.1625e+001 4.3013e+011 1.9000e-003 7.0300e+001Kr-87 1.5051 e+001 5.5690e+011 2.4600e-003 9.1020e+001Kr-88 2.4596e+001 9.1 006e+01 1 4.0200e-003 1 .4874e+002Kr-89 5.0416e-002 1.8654e+009 8.2400e-006 3.0488e-001Kr-90 6.0083e-01 6 2.2231 e-005 9.8200e-020 3.63 34e-0 15file:///C:/Program%2OFiles%20(x86)/MicroShield%208/Examples/CaseFiles/HTML/Cont... 9/29/2015 Case Summary of Containment ShineATTACHMENT 9Page 2 of 3Xe-1332.4596e+001I9.1006e+01 14.0200e-0031 .4874e+002Xe- 135 1.0157e+001 I 3.7579e+011 1.6600e-003 6.1420e+001Xe-135m 4.3196e+000 j 1.5983e+011 j 7.0600e-004 2.6122e+001Xe-137 2.1292e-001 j 7.8781e+009 3.4800e-005 1.2876e+000Xe- 138 1.1013 e+001 4.0749e+011 [ 1.8000e-003 6.6600e+001Buildup: The material reference is Shield 1Integration ParametersX Direction 10Y Direction I 20Z Direction 20____________ Results -Dose Point # 1 -(1890,914,914) cm ______Fluence Rate Fluence Rate Exposure Rate Exposure RateEnergy (MeV) iActivity (Photons/sec) MeV/cm1/sec MeV/cm2/sec mR/hr mR/hr_________No Buildup With Buildup No Buildup With Buildup0.015 1.997e+11I 8.577e-253 2.641e-24 7.357e-254 2.266e-250.03 5.752e+11I 6.299e-35 2.648e-23 6.242e-37 2.624e-250.08 3.426e+ 11 1.284e-04 3.213e-03 2.031 e-07 5.084e-060.1 1.795e+09 8.335e-06 _3.]31e-04 1.275e-08 4.791e-070.15 4.738e+l11 3.531 e-02 1.778e+00 5.814e-05 2.928e-030.2 6.841 e+ 11 2.179e-01 1.067e+01 3.845e-04 1.883e-020.3 3.207e+11 6.259e-01 2.298e+01 1.187e-03 4.359e-020.4 1 .055e+1 2 6.904e+00 1 .867e+02 1 .345e-02 3.638e-010.5 2.408e+12 3.898e+01 8.084e+02 7.652e-02 1.587e+000.6 1.884e+ 12 6.256e+01 1.036e+03 1.221 e-01 2.021 e+000.8 4.714e+12 4.684e+02 5.470e+03 8.909e-01 1.040e+011.0 1.860e+12 4.187e+02 3.760e+03 7.717e-01 6.931e+001 .5 1.544e+ 12 1.406e+03 _8.149e+03 2.365e+00 1.371 e+012.0 1.110e+12 2.482e+03 _1.108e+04 3.838e+00 1.713e+013.0 8.507e+10 5.850e+02 1.898e+03 7.936e-01 2.575e+004.0 8.353e+07 1.1 55e+00 3 .087e+00 1 .429e-03 3.81 9e-03Totals 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 RateEnergy (MeY) Activity (Photons/sec) MeV/cmz/sec MeV/cm2/sec mR/hr mR/hr_______No Buildup With Buildup No Buildup With Buildup0.015 1.997e+ 11 1.798e-263 1.169e-26 1.543e-264 1.003e-270.03 5.752e+11 6.868e-38 _1.172e-25 6.807e-40 1.162e-270.08 3.426e+ 11 4.039e-07 _1.139e-05 6.392e- 10 1.802e-080.1 1.795e+09 2.705e-08 _1.190e-06 4.139e-ll 1.821e-090.15 4.738e+11I 1.274e-04 _7.800e-03 2.098e-07 1.284e-050.2 6.841le+lI1 8.648e-04 5.172e-02 1.526e-06 9.128e-050.3 3.207e+11 2.857e-03 1.255e-01 5.419e-06 2.380e-040.4 1.055e+12 3.451e-02 1.091e+00 6.725e-05 2.125e-030.5 2.408e+12 2.079e-01 4.939e+00 4.082e-04 9.695e-03file:///C 9/29/2015 ShinePge3o3Page 3 of 30.61 .884e+1 23.501e-016.536e+006.834e-041 .276e-020.8 4.714e+12 2.798e+00 3.584e+01 5.322e-03 6.817e-021.0 1.860e+ 12 2.610e+00 2.523e+01 4.81 le-03 4.651 e-021.5 1.544e+ 12 9.284e+00 5.623e+01 1.562e-02 9.461 e-022.0 1.110e+l12 1.684e+01 7.709e+01 2.604e-02 1.192e-013.0 8.507e+10 4.061e+00 1.323e+01 5.510e-03 1.795e-024.0 8.353e+07 8.082e-03 2.146e-02 9.998e-06 2.655e-05Totals 1.726e+13 3.620e+01 2.204e+02 5.848e-02 3.714e-01file %208/Examples/CaseFiles/HTML/Cont... 9/29/2015 ATTACHMENT 9 Case Summary of Containment ShineATTACMENT 9Page 1 of 3MicroShield 8.02Nathan Hogue (8.00-0000)Date ByCheckedFilename IRun Date I Run Time I Duration JContainl .msd September 29, 2015 1:23:52 PM 00:00:00Project InfoCase Title Containment ShineDescription IFuel Element Failure Accident AnalysesGeometr 13 -Rectangular VolumeSource DimensionsLength 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 11.82 in) j n nShieldsShield N Dimension Material Densit ____________Source 6.12e+09 cm3  Air 0.00122Shield 1 I 30.5 cm I Concrete I 2.35Air Gap Air 0.00122Source Input: Grouping Method -Standard IndicesNumber of Groups: 25Lower Energy Cutoff: 0.0 15Photons < 0.015: Included__ _ _ _ _ _ _ _ Library: Grove .........._ __ __ __ __ __ _Nuclide Ci Bq JLCi/cm3  Bq/cm3I-131 1 .3399e-00 1 4.9578e+009 2.1 900e-005 8.1 030e-0011-132 2.6003e-001 9.6213e+009 4.2500e-005 1.5725e+000I- 133 4.0198e-001 1.4873e+010 6.5700e-005 2.4309e+0001-134 4.9682e-001 1 .8382e+0 10 8.1 200e-005 3 .0044e+000I- 135 4.0932e-001 1.5 145e+010 6.6900e-005 2.4753e+000Kr-85 6.0940e-004 2.2548e+007 9.9600e-008 3.6852e-003Kr-85m 1.4256e-001 5.2747e+009 2.3300e-005 8.6210e-001Kr-87 2.7288e-001 1 .0097e+010 4.4600e-005 1 .6502e+000Kr-88 3.8852e-001 1.4375e+010 6.3500e-005 2.3495e+000Kr-89 4.9253e-001 1 .8224e+01 0 8.0500e-005 2.9785e+000Kr-90 4.9253e-00 1 1 .8224e+0 10 8 .0500e-005 2.9785e+000file:///C:/Program%2OFiles%20(x86)/MicroShield%208/Examples/CaseFiles/HTML/Cont... 9/29/2015 Case Summary of Containment ShineATTACHMENT 9Page 2 of 3Xe-1335.4454e-0012.0148e+0108.9000e-0053 .2930e+000Xe-135 1 .2482e-001 4.6182e+009 2.0400e-005 7.5480e-001Xe.-135m j 1.2176e-001 J 4.5050e+009 j 1.9900e-005 j 7.3630e-001Xe- 137 [ 6.3632e-001 J 2.3544e+010 J 1.0400e-004 j 3.8480e+000Xe- 138 j 6.7303 e-001 Jj 2.4902e+010 J 1.1000e-004 4.0700e+000Buildup: The material reference is Shield 1Integration ParametersX Direction 10Y ireto II20ZDirection 20____________ Results -Dose Point # 1 -(1890,914,914) cm ______Fluence Rate Fluence Rate Exposure Rate Exposure RateEnergy (MeV) Activity (Photons/sec) MeV/cmn2sec MeV/cm2/sec mR/hr mR/hr_________No Buildup With Buildup No Buildup With Buildup0.015 5.847e+09 2.51 2e-254 7.735e-26 2.1 54e-255 6.635e-270.03 1.247e+ 10 1.365e-36 5.739e-25 1.353e-38 5.688e-270.08 7.524e+09 2.819e-06 7.055e-05 4.460e-09 1.116e-070.1 6.447e+09 2.994e-05 1.125e-03 4.580e-08 1.721e-060.15 6.836e+09 5.094e-04 2.565e-02 8.388e-07 4.224e-050.2 1.617e+10 5.150e-03 2.522e-01 9.089e-06 4.45 le-040.3 1.051e+10 2.051e-02 7.532e-01 3.891e-05 1.429e-030.4 2.244e+10 1 .468e-01 3.969e+00 2.860e-04 7.733e-030.5 3.752e+10 6.074e-01 1.259e+01 1.192e-03 2.472e-020.6 2.587e+10 8.590e-0I 1.422e+01 1.677e-03 2.775e-020.8 5.049e+ 10 5.017e+00 5.859e+01 9.542e-03 1.1 14e-011.0 3.060e+ 10 6.885e+00 6.184e+01 1.269e-02 1.140e-011.5 2.473e+10 2.251e+01 1.305e+02 3.787e-02 2.195e-012.0 2.762e+ 10 6.174e+01 2.755e+02 9.547e-02 4.260e-013.0 3.396e+09 2.335e+01 7.576e+I01 3.168e-02 1.028e-014.0 8.371e+08 1.158e+01 3.094e+01 1.432e-02 3.828e-02Totals 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 RateEnergy (MeV) Activity (Photons/sec) MeV/cm2/sec MeV/cm2/sec mR/hr mR/hr_________No Buildup With Buildup No Buildup With Buildup0.015 5.847e+09 5.266e-265 3.424e-28 4.517e-266 2.937e-290.03 1.247e+ 10 1.489e-39 2.540e-27 1.475e-41 2.518e-290.08 7.524e+09 8.870e-09 2.501 e-07 1 .404e- 11 3.957e- 100.1 6.447e+09 9.718e-08 4.276e-06 1.487e- 10 6.542e-090.15 6.836e+09 1.838e-06 1.125e-04 3.027e-09 1.853e-070.2 1.617e+10 2.044e-05 1.223e-03 3.608e-08 2.158e-060.3 1.051 e+l 0 9.363e-05 4.113e-03 1.776e-07 7.801 e-060.4 2.244e+ 10 7.337e-04 2.319e-02 1.430e-06 4.518e-050.5 3.752e+ 10 3.240e-03 7.695e-02 6.359e-06 1.51 le-04file:///C:/Program%2OFiles%20(x86)/MicroShield%208/Examples/CaseFiles/HTML/Cont... 9/29/2015 Cas~1jp~xp~P~aimnent Shine Page 3 of 3Page 3 of 30.62.587e+104.807e-038.974e-029.383e-061 .752e-040.8 5.049e+10 2.997e-02 3.839e-01 5.700e-05 7.301e-041.0 3.060e+10 4.292e-02 4.149e-01 7.912e-05 7.649e-041.5 2.473e+10 1.486e-01 9.003e-01 2.501e-04 1.515e-032.0 2.762e+ 10 4.1 89e-0 1 1.91 8e+00 6.477e-04 2.965e-033.0 3.396e+09 1.621e-01 5.280e-01 2.199e-04 7.163e-044.0 8.371e+08 8.099e-02 2.151e-01 1.002e-04 2.660e-04Totals 2.893e+11 8.924e-01 4.555e+OO 1.371e-03 7.339e-039/29/2015 ATTACHMENT 9 Case Summary of Containment ShineATTACHMENT 9Page 1 of 3~MicroShield 8.02Nathan Hogue (8.00-0000)Date I By ICheckedI Filename )Run Date I Run Time I Duration jI Containl .msd j September 29,2015 1:15:41 PM 00:00:01 j_________________Project InfoCase Title Containment ShineDescription Fuel Experiment Accident AnalysesGeometry 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 Y2 1.5e+4 cm (492it 1.5 914.0em (29 fi11.8 914.0Ocm (29 fi11.8.2 in) in) in)ShieldsShield N Dimension Material Density ___________Source 6.12e+09 cm3  Air 0.00122Shield 1 I 30.5 cm I Concrete I 2.35Air Gap Air 0.00122Source Input: Grouping Method -Standard IndicesNumber of Groups: 25Lower Energy Cutoff: 0.0 15Photons < 0.015: Included_______________ Library: Grove ________Nuclide Ci Bg JLCi/cm3  Bg/cm3I- 131 8.0763e+000 2.9882e+011 1.3200e-003 4.8840e+0011- 132 1.7866e+001 6.6104e+011 2.9200e-003 1.0804e+0021-133 3.8240e+001 1.4149e+012 6.2500e-003 2.3125e+0021- 134 4.3563e+001 1.611 8e+012 7.1 200e-003 2.6344e+002I- 135 3.6160e+001 1.3379e+012 5.9100e-003 2.1867e+002Kr-85 1 .6459e-003 6.0897e+007__ 2.6900e-007 9.9530e-003Kr-85m 7.2810e+000 2.6940e+011 1.1900e-003 4.4030e+001Kr-87 1 .4807e+001 5.4785e+011I 2.4200e-003 8.9540e+001Kr-88 2.0864e+001 7.7196e+011I 3.4100e-003 1 .2617e+002Kr-89 2.6676e+001 9.8703e+011 4.3600e-003 1.6132e+002Kr-90 2.6309e+001 9.7344e+011 4.3000e-003 1.5910e+002file 9/29/2015 Case Summary of Containment ShineATTACHMENT 9Page 2 of 3Xe-1331.81 72e+0016.7236e+01 12.9700e-0031 .0989e+002Xe- 135 1,I3093 e+i001 4.8446e+011 2.1400e-003 7.9180e+001iXe-I135m 6,4856e+'000 [ 2.3997e+01 1 1 .0600e-003 j 3.9220e+001Xe- 137 3.4386e+001 1 .2723 e+012 5.6200e-003 j 2.0794e+002Xe- 138 j 3.5915e+001 [ 1.3289e+012 5.8700e-003 2.1719e+002Buildup: The material reference is Shield 1Integration ParametersX Direction 10Y Direction I 20Z Direction 20__________Results -Dose Point # 1 -(1890,914,914) cmFluence Rate Fluence Rate Exposure Rate Exposure RateEnergy (MeV) Activity (Photons/sec) MeV/cm2/sec MeV/cmZ/sec mR/hr mR/hrNo Buildup With Buildup No Buildup With Buildup0.015 2.904e+11I 1.248e-252 3.842e-24 1.070e-253 3.296e-250.03 5.010e+ll 5.485e-35 2.306e-23 5.436e-37 2.285e-250.08 2.546e+ 11 9.536e-05 2.387e-03 1.509e-07 3.777e-060.1 3.444e+ 11 1.599e-03 6.008e-02 2.447e-06 9.192e-050.15 3.872e+ 11 2.885e-02 1.453e+00 4.751le-05 2.393e-030.2 1.110e+12 3.537e-01 1.732e+01 6.242e-04 3.056e-020.3 5.920e+11 1 I.156e+00 4.243 e+01 2.192e-03 8.048e-020.4 1 .346e+12 8.808e+00 2.382e+02 1.716e-02 4.640e-010.5 2.718e+12 4.399e+01 9.123e+02 8.635e-02 1.791e+000.6 1.808e+12 6.003e+01 9.936e+02 1.172e-01 1.939e+000.8 4.038e+12 4.012e+02 4.686e+03 7.631e-01 8.912e+001.0 2.157e+12 4.855e+02 4.360e+03 8.949e-01 8.037e+001.5 1.735e+ 12 1.580e+03 9.156e+03 2.658e+00 1.540e+012.0 1.593e+12 3.562e+03 1.589e+04 5.508e+00 2.458e+t013.0 1.838e+11 1.264e+03 4.101le+03 1.715e+00 5.563e+004.0 4.532e+10 6.268e+02 1.675e+03 7.754e-01 2.072e+00Totals 1.910e+13 8.033e+03 4.208e+04 1.254e+01 6.887e+01Results -Dose Point # 2 -(15000,914,914) cmF'luence Rate Fluence Rate Exposure Rate Exposure RateEnergy (MeV) Activity (Photons/sec) MeV/cm2/sec MeV/cm2/see mR/hr mRlhr_____________No Buildup With Buildup No Buildup With Buildup0.015 2.904e+ 11 2.616e-263 1.701 e-26 2.244e-264 1.459e-270.03 5.010e-il 1 5.981e-38 1.021e-25 5.928e-40 1.012e-270.08 2.546e+11 3.001e-07 8.461e-06 4.749e-10 1.339e-080.1 3.444e+ 11 5.191 e-06 2.284e-04 7.942e-09 3.494e-070.15 3.872e+11 1,041e-04 6.374e-03 1.714e-07 1.050e-050.2 1.110e+ 12 1.404e-03 8.395e-02 2.478e-06 1.482e-040.3 5.920e+11I 5.274e-03 2.3 17e-01 1.000e-05 4.394e-040.4 1 .346e+12 4.403e-02 1.392e+00 8.579e-05 2.71 le-030.5 2.718e+ 12 2.347e-01 5.574e+00 4.606e-04 1.094e-02fle :///C :/Program%2OFiles%20(x86)/MicroShield%208/Exampies/CaseFiles/HTML/Cont... 9/29/2015 Case 1;n~ainm~ent ShinePae3o3Page 3 of 30.61.808e+123 .359e-01I6.271 e+006.5 57e-041 .224e-020.8 4.038e+12 2.397e+00 3.070e+01 4.559e-03 5.839e-021.0 2.157e+12 3.026e+00 2.926e+01 5.579e-03 5.393e-021.5 1.735e+12 1.043e+01 6.318e+01 1.755e-02 1.063e-012.0 1.593e+12 2.416e+01 1.106e+02 3.737e-02 1.711e-O13.0 1.838e+11 8.774e+00 2.858e+01 1.190e-02 3.877e-024.0 4.532e+10 4.385e+00 1.164e+01 5.425e-03 1.440e-02Totals 1.910e+13 5.380e+0O1 2.875e+02 8.360e-02 4.694e-01fie:///C:/Programi%2OFi~es%20(x86)/MicroShield%208/Examples/CaseFiles/HTML/Cont... 9/29/2015 ATTACHMENT 9 ATTACHMENT 10wIND ROSE PLOT:Station #03945 -COLUMBIAIREGIONAL ARPT, MODISPLAY:Wind SpeedDirection (blowing from)NORTH15%12%9%WESTEASTWIND SPEED(mis)U >=11 IU8.8-11.11 5.7-.88* 3.6- 5.7E]2.1 -3.62.1Calms: 1.15%SOUTHDATA PERIOD: COMPANY NAME:Start Date: 11111961 -00:00End Date: 1213111969 -21:00'MODELER:CALM WINDS: TOTAL COUNT:1.15% 73020 hr..AVG. WIND SPEED: DATE: PROJECT NO.:4.70 mls 912312015WRPLOT View -Lakes Envlonmanlat Software Wind Class Frequency Distribution:z40-35-30-%/252015105mACalms0-0.5- 2.12.1- 3.6 3.6- 5.7Wind Class (mis)0.8>= 11.15.7- 8.88.8 -11.1 ATTACHMENT 11WIND ROSE PLOT:Station #03945 -COLUMBIA/REGIONAL ARPT, MODISPLAY:Wind SpeedDirection (blowing from)NORTH20%16%12%WESTEASTWIND SPEED(mis)U>= 11.11n 8.8-11.11 57- 8.83.6- 5.7m2.1 -3.6LI 0.5- 2.1Calmns: 1.83%SOUTHCOMwMENTS: DATA PERIOD: COMPANY NAME:Start Date: 1/1/1170 -00:00End Date: 1213111990 -23:00 ___________ ____________MODELER:CALM WINOS: TOTAL COUNT:1.83% 154387 hrs.AVG. WIND SPEED: DATE; PROJECT NO.:4.44 mole 912312015WRPLOT View -Lakes Envirotnmental Software Wind Class Frequency DistributionA 3-z-r'J44~1- 4* 41 I.41 I..625--20-15-10-!0.00-0.7>= 11.1Calms0.5- 2.12.1- 3.6 3.6- 5.7 5.7- 8.8Wind Class (mis)8.8- 11.1kI VIUW ! .U.U "
ATTACHMENT 6Cc34595969798Cc101.102103104105106107108cC120121c122,Q3'&#xa3;24125126127cc1301 31132133134135c1401411421.43144145146147148149c150151152153.54cthe following surfaces define the single fuel element (origin of thiscoordinate system different from the main one and from the auxiliarycoordinate system # 1 that defines the bins of *the new basket).czczczczcz"czcz99i00pzpzpzpzpzpzpzpz7.036 $ plate 17.074 ,7.12514.630 $ plate14.66814.71914.757cz 14.986.pz 030.48-30.4832.385-32.38538.10-38.1041,275-41.27524p -0.4142 1 0p 0.4142 1 0PppppppxpxPYpypz.pzpxpypyPYpypxpypypzp7pxpxPYPYpz-0.41420.4142-0.41420.4142-0.41420.4142-30.068.10-30.080.80-80.080.00.013 .0026.0039.0052 .0038.100.050. 80-42.2542 .2564. 5094. 5055.5082.00-50.001"11110000000.0 $0.0 .$-0 .05500.0550-0.46740.4674-0 .65980.6598$$$$$$partpartpartpartpartpartofofofofofof+2'2 .5-22.5+22 .5-22.5+22.5-22 :522.5 degree plane-22.5 degree planedegreedegreedegreedegreedegreedegreesidesidesidesidesidesideplateplateplateplateplateplate ATTACHMVLENT 6c the following surfaces define the auxiliary coordinate system thatdefines the individual bins of the new basket, its origin is at thec" center of each bin, a coordinate transformation then connectsc this to the main coordinate system -which is centred at one of thec corners of the new basket.c2O00201202203204205206207208209210211cpxpxpypypxpxpyPYpxpxPYPY-6.386.38z.6 386.38-6.076 .07-6.076 .07-5.755.75-5 .755.75Rodeccckcodecmlrmt1m2m3c,cccCccm6mt6rotm9cctr2Otr2 1tr2 2tr2 3tr2 4t r2 5tr26tr2 7tr2 8tr29tr30tr31tr3 2r3 3cr34nimp:n 1 18r2000 0.8 51001.5Cc .6667lwtr. 0113027.50c 113027.50c *-.60092235.50c -.37292238.50c -.028mn4 4009.50cmt4 be. Olm5 6012.50c1001. 50c8016.5Cc13027.5Ccmut5 grph. Oi1001.5Cc 0.4170lwtr. Olin8 -.. 26000.5Cc5010.50c -0.0356000.50c -0.0520.01 197r55 3000 08016.50c .33331-.85-.002-.016-.1328016 .50c0.20813027.50c 0.37500.24.,... ,000..50lc 0.2 28000.50c 0.15011.56c -0.15013027.50c -0.763 $ boral with 24 w% b4c6.4019.2032.0044.8057.606.4019.2032.0044.8057.606.4019.2032.0044.8057.606.50 0.06.50 0.06.50 0.06.50 0.06.50 0.019.50 0.019.50 0.01.9.50 0.0 ..."19.50 0.0"19.50 0.032.50 0.032.50 0.032.50 0.032.5C 0.032.50 0.0 tr35tr38tr39Cphys : nprintctinecut :fl1prdinpksrcATTACHMENT 66.40 45.50 0.019.20 45.50 0.032.00 45.50 0.044.80 45.50 0.057.60 45.50 0.0S15.0 0.0 $cross Sections above 15.0 mev will be expunged40 50 60 905000.j o .o -0.5j 2010.0 4.0 0.08.0 16.5 0.03.0 30.0 0.03.0 3.50 0.0110 120 126-0.118.021.017 .04.002.020.933 .02 .000.00.00.'00.034.030.032.0-3.504.518.029.0-3 .500.00.. 00.00.055.047 .059.01.508.022.034.0-1i.000.00.00.00.0 S IATTACHMENT 6OAAR CAR GO SYSTEMSa division of AAR Manufacturing, Inc..Jeff Moore, Sr. Manager -Nuclear Products12633 lnkster Road, Livonia, MI 48150-2272 USAPhone: (734) 522.2000 Direct: (734) 466-8110FAX: (734) 622-2240"email: jmoore@aarcorp.comI. CUSTOMER:A. NAME:B, REQUEST DATE:II. DATESA. CURRENT DATE:B. QUOTATION VALID FOR:III. CONTACTSA. AAR CONTACTI. NAME:2. TITLE:3; PHONE:4. FACSIMILE:B. CUSTOMER CONTACT1. NAMvE:2 PHONE:3. FACSIMILE:IV. sPECIFICATION AND PRICINGUniversity of MissouriApril 8, 2003April 30, 200390 daysJeff MooreSr. Manager, Nuclear Products(734) 522-2000 x 8110(734) 522-2240.Mr. Jeff Attebery1-573-882-52691-573-882-6360+ Shipping to Univ. of Missouri 4i%V. DELIVERIES TO COMMENCE:60 days AROVI. TERMSA.B.DELIVERY POINT:PAYMENT:FOB University of MissouriNet 30 da~isVII. SPECIAL INSTRUCTIONS None ATTACHMENT 6AAR? CARGO SYSTEMS0A division of AAR Manufacturing Group, INC.CERTIFICATE OF COMPLIANCECUSTOMER: University of MissouriQTY. SHIPPED: 68 pcs.DATE OF SHIPMENT: July 18, 2003CUSTOMER P.O. NUMBER: COO000009743AAR CARGO SYSTEMS SALES ORDER NUMBER: 5053667This is to certify' that the material supplied hereunder has been finspected and tested inaccordance with AAR-1 1002 QAP, Revision 23 dated November 7, 2002, and A.AR-10012QAP, Revision 18 dated April 9, 2003, and Nuclear Quality Program Manual, Revision 29and meets the requirements of the. purchase order. The Code of Federal Regulations10OCFR5O Appendix B and 10OCFR2 1 are applicable to the material on this order.SIGNATURE:TITLE:DATE:Phill PusiloLab ManagerJuly 18. 2003Appendix CAAR-1 0012 QAPPage 1 of 1...systems, components & more12633 Inkster Road Livonia Michigan 48150-2272 USATelephone 1-734-522-2000 Faxc 1-734-522-2240 ATTACHMENT 6AA R CA RGO0 S YSTEMSA div'ision of AAR Manufacturing Group, INC.BORAL DATA PACKAGE RECORD CHECKLISTSPECIFICATION: AAR-10012 OAP. REV. 17 D OCUMENTCHECKEDBYDATE"Record ChecklistCertificate of ComplianceInspection Data SheetsMaterial CertificationsJP/KEJP/KiEJP/KEJP/KEJP/KEJPIKEJP/KE7-18-037-1 8-037-12-037-1 8-037-18-037-18-037-18-03-BoraI Summary Report,Boxing ListCalibrated Equipment Data SheetREVIEWED BY:TITLE:DATE:Phill PusiloLab ManaaerJuly 18, 2003APPENDIX DAAR-10012 QAIPPAGE 1 OF I..systems, components & more12633 Inkster Road Livonia Michigan 48150-2272 USATelephone 1-734-522-2000 Fax 1-734-522-2240 tATTACHMENT 6Boral Summary Report (Pass)Job Name: University5O 5053667Serial NumberWM010013-3A 'WM010014-IBwM010015-3A.WM010016-IAWM010017-2AY(MO1001 8-2AYM0 10019-1BYM010020-8BYM010O21 -SBYMI f00022-SBYM010023 ,8AYM01]0024 -8 BLot NumberM-21 SM-2 ISM-2 I8M-21814-218M4-220M-220M-220M-220M-220M-220M-22010B gmns/em20.07400.07210,07090.07660.07260.0754,0.07580.07310.07480.07500.07330.0742Density,2.57312.55072.54742.54842.54272.65822.56322.57812.57282.56232.58472,5873Reviewed By: Phill PusiloTitle: Lab ManagerDate: 8/11/2003Appendix-AAAR10012QAPPagc: 1Ptss ATTACHMENT 6MATERIAL TRACEABILETY bY B ORAL SERIAL NUMBERS.0. # 50536.67 University of Missouri--..~ k-.- WM010013 through WM010017 M-218YM010018 through YM010024 M-220WM010013 through YM010024 "AL03-03WM010013 through YM0 10024 "3-045-C ATTACHMENT 7Volume of the Primary Coolant SystemIn-Pool Portion Mechanical Equipment Room 114 PortionScin Area Length Volume Scin Area Length VolumeScin (ft2) (ft) (ft3) Scin (ft2) (ft) (ft3)135(5) 0.7773 3.828 2.976 133(2-3) 0.7773 10.194 7.924135(6) 0.7773 3.708 2.882 135(1) 0.7773 2.000 1.555135(7) 0.7773 3.708 2.882 0.7773 22.374 17.391137 0.7773 3.250 2.526 133(7-5) 0.7773 22.374 17.391139 0.7773 2.500 1.943 133(4-3) 0.7773 14.290 11.108501 0.6048 5.937 3.591 133(2-1) 0.7773 15.584 12.113575 0.6048 2.269 1.372 132 0.6948 6.000 4.169100(2) 0.7773 4.917 3.822 131(3-2) 0.6948 4.969 3.452100(3) 0.7773 4.917 3.822 115(3-2) 0.6948 14.968 10.400101 0.7773 1.000 0.777 115(1) 0.4948 2.000 0.990102(1) 0.7773 1.000 0.777 111(7) 0.6948 4.189 2.911102(2) 0.7773 3.806 2.958 111(6) 0.6948 6.667 4.632102(3) 0.7773 3.806 2.958 111(2-5) 0.6945 16.264 11.295102(4) 0.7773 3.806 2.958 111(1) 0.6948 2.167 1.506102(5) 0.7773 5.097 3.962 105(9) 0.7773 2.167 1.684401(1) 0.2006 3.975 0.797 105(7-8) 0.7773 17.312 13.457401(2) 0.2006 3.975 0.797 105(5-6) 0.7773 15.542 12.081405(1) 3.2150 0.500 1.608 105(1-4) 0.7773 31.832 24.743405(2) 0.1389 4.708 0.654 102(7) 0.7773 2.000 1.555405(3) 0.2006 9.163 1.838 102(5-6) 0.7773 10.194 7.924460 1.3960 4.242 5.922 Total Piping Volume (ft3) 223.860406 0.7773 2.500 1.943407 0.7773 2.333 1.813 Total Piping Volume (gallons) [1,674.585Fuel Region (gallons)7.176Primary Circulation Pumps (gallons) 25.000Primary Heat Exchangers (gallons) 150.000Pressurizer (gallons) 150.000Total Volume of PCS (gallons) 2,006.761 eCFR -- tode of Federal Regulations" ATTACHMENT 8http://www.ecfr'.gov/cgi-birltext-idx?SlD=a6ddafde7f67322376d64cb..ELECTRONIC CODE OF FEDERAL REGULATIONSe-CFR data is current as of September 21, 2015Title 10 .- Chapter III --, Part 835 --. Subpart N -* Appendix-Title 10: EnergyPART 835--OCCUPATIONAL RADIATION PROTECTIONSubpart N-Emergency Exposure SituationsAPPENDIX C TO PART 835-DERIVED AIR CONCENTRATION (DAC) FOR WORKERS FROM EXTERNAL EXPOSURE DURINGIMMERSION IN A CLOUD OF AIRBORNE RADIOACTIVE MATERIALa. The data presented in appendix C are to be used for controlling occupational exposures in accordance with&sect;835.209, identifying the need for air monitoring in accordance with &sect;835.403 and identifying the need for posting ofairborne radioactivity areas in accordance with &sect;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) peryear. 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. Thedata are listed by radionuclide in order of increasing atomic mass. The air immersion DACs were calculated for acontinuous, nonshielded exposure via immersion in a semi-infinite cloud of airborne radioactive material. The DACs listedin 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 ofthe ratio of the observed concentration of a particular radionuclide and its corresponding DAC for all radionuclides in themixture. If this sum exceeds unity (1), then the DAC has been exceeded. For unknown radionuclides, the most restrictiveDAC (lowest value) for those isotopes not known to be absent shall be used.AIR IMMERSION DACRadlonlucllde aflf piL)(.qm)Ar-37 __5._2_d_____00__,_____At-39 269___yr___E-__3__ E+__ 7Ar-41 157h3-6IEOK~r-74 1. an3-6lE0Kr-76 __4____h____-__5____+_ 5Kr-79 __5.____h___E-__5______ 5Kr-81 __.__+05_y______-____E_07Kr-83m 1.83 h 7E-02 2E+09Kr-85 10.72 yr 7E-04 2E+07Kr-85m 4.48 h 2E-05 IE+06Kr-87 76.3 mni 4E-06 1E+05Kr-88 2.84 h 1E-06 7E+04.Xe-120 40.0 min 1E-05 4E+05X(e-121 40.1 mai 2E-06 BE+04Xe-122 20.1 h 8E-05 3E+06Xe-123 2.14 h 6E-06 2E+05Xe-125 16.8 h lE-05 SE+05Xe-127 36.406 d 1 E-05 8E&#xf7;05Xe-129m 8.86 d 2E-04 7E+06Xe-1 31m 11.84 d 5E-04 1E+07Xe-133 5.245 d 1 E-04 5E+08Xe-133m 2.19 d 1 E-04 5E+06Xe-135 9.11 hi 1E-O5 6E+05Xe-1 35m 15.36 rni 1 E-05 3E+05Xe-138 14.13 min 3E-06 IE+05For any single radlonuclide not listed above with decay mode other than alpha emission or spontaneous fission andI of 2I of291231201 PM eCFR -Code of Federal RegulationsATTACHMENT 8http://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/m3).[72 FR 31940, June 8, 2007, as amended at 76 FR 20489, Apr. 13, 2011]Need assistance?!of 29/23/2015 4:37 PM Case Summary of Containment ShineATTACHMENT 9Page 1 of 3~MicroShield 8.02Nathan Hogue (8.00-0000)Date I By ChkeFilename IRun Date I Run Time I DurationContainl1.msd September 29, 2015 1:21:55 PM 00:00:00Project InfoCase Title Containment ShineDescription IFuel Accident AnalysesGeometr 13 -Rectangular VolumeSource DimensionsLength 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 zSin) in) in)___________ShieldsShield N J ~Dimension Material Density ____________Source 6.12e+09 cm3  I Air 0.00122Shield 1 j 30.5 cm I Concrete 2.35Air Gap j Air 0.00122Source Input: Grouping Method -Standard IndicesNumber of Groups: 25Lower Energy Cutoff: 0.015Photons < 0.015: Included_______________________ Library: Grove _______ _______Nuclide Ci Bq 3  B q/cm3I- 131 8.9329e+000 3.3052e+01 1 1.4600e-003 5.4020e+001I- 132 2.4168e+001 8.9421 e+011 3.9500e-003 1.4615e+002I- 133 5.0905e+001 1.8835e+012 8.3200e-003 -3.0784e+002I- 134 5.2252e+001 1 .9333e+01 2 8.5400e-003 3.1598e+0021-135 4.5644e+00 1 1 .6888e+0 12 7.4600e-003 2.7602e+002IK"-85 2.2271 e-003 8.2403e+007 3.6400e-007 1 .3468e-002Kr-85m 1.1625e+001 4.3013e+011 1.9000e-003 7.0300e+001Kr-87 1.5051 e+001 5.5690e+011 2.4600e-003 9.1020e+001Kr-88 2.4596e+001 9.1 006e+01 1 4.0200e-003 1 .4874e+002Kr-89 5.0416e-002 1.8654e+009 8.2400e-006 3.0488e-001Kr-90 6.0083e-01 6 2.2231 e-005 9.8200e-020 3.63 34e-0 15file:///C:/Program%2OFiles%20(x86)/MicroShield%208/Examples/CaseFiles/HTML/Cont... 9/29/2015 Case Summary of Containment ShineATTACHMENT 9Page 2 of 3Xe-1332.4596e+001I9.1006e+01 14.0200e-0031 .4874e+002Xe- 135 1.0157e+001 I 3.7579e+011 1.6600e-003 6.1420e+001Xe-135m 4.3196e+000 j 1.5983e+011 j 7.0600e-004 2.6122e+001Xe-137 2.1292e-001 j 7.8781e+009 3.4800e-005 1.2876e+000Xe- 138 1.1013 e+001 4.0749e+011 [ 1.8000e-003 6.6600e+001Buildup: The material reference is Shield 1Integration ParametersX Direction 10Y Direction I 20Z Direction 20____________ Results -Dose Point # 1 -(1890,914,914) cm ______Fluence Rate Fluence Rate Exposure Rate Exposure RateEnergy (MeV) iActivity (Photons/sec) MeV/cm1/sec MeV/cm2/sec mR/hr mR/hr_________No Buildup With Buildup No Buildup With Buildup0.015 1.997e+11I 8.577e-253 2.641e-24 7.357e-254 2.266e-250.03 5.752e+11I 6.299e-35 2.648e-23 6.242e-37 2.624e-250.08 3.426e+ 11 1.284e-04 3.213e-03 2.031 e-07 5.084e-060.1 1.795e+09 8.335e-06 _3.]31e-04 1.275e-08 4.791e-070.15 4.738e+l11 3.531 e-02 1.778e+00 5.814e-05 2.928e-030.2 6.841 e+ 11 2.179e-01 1.067e+01 3.845e-04 1.883e-020.3 3.207e+11 6.259e-01 2.298e+01 1.187e-03 4.359e-020.4 1 .055e+1 2 6.904e+00 1 .867e+02 1 .345e-02 3.638e-010.5 2.408e+12 3.898e+01 8.084e+02 7.652e-02 1.587e+000.6 1.884e+ 12 6.256e+01 1.036e+03 1.221 e-01 2.021 e+000.8 4.714e+12 4.684e+02 5.470e+03 8.909e-01 1.040e+011.0 1.860e+12 4.187e+02 3.760e+03 7.717e-01 6.931e+001 .5 1.544e+ 12 1.406e+03 _8.149e+03 2.365e+00 1.371 e+012.0 1.110e+12 2.482e+03 _1.108e+04 3.838e+00 1.713e+013.0 8.507e+10 5.850e+02 1.898e+03 7.936e-01 2.575e+004.0 8.353e+07 1.1 55e+00 3 .087e+00 1 .429e-03 3.81 9e-03Totals 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 RateEnergy (MeY) Activity (Photons/sec) MeV/cmz/sec MeV/cm2/sec mR/hr mR/hr_______No Buildup With Buildup No Buildup With Buildup0.015 1.997e+ 11 1.798e-263 1.169e-26 1.543e-264 1.003e-270.03 5.752e+11 6.868e-38 _1.172e-25 6.807e-40 1.162e-270.08 3.426e+ 11 4.039e-07 _1.139e-05 6.392e- 10 1.802e-080.1 1.795e+09 2.705e-08 _1.190e-06 4.139e-ll 1.821e-090.15 4.738e+11I 1.274e-04 _7.800e-03 2.098e-07 1.284e-050.2 6.841le+lI1 8.648e-04 5.172e-02 1.526e-06 9.128e-050.3 3.207e+11 2.857e-03 1.255e-01 5.419e-06 2.380e-040.4 1.055e+12 3.451e-02 1.091e+00 6.725e-05 2.125e-030.5 2.408e+12 2.079e-01 4.939e+00 4.082e-04 9.695e-03file:///C 9/29/2015 ShinePge3o3Page 3 of 30.61 .884e+1 23.501e-016.536e+006.834e-041 .276e-020.8 4.714e+12 2.798e+00 3.584e+01 5.322e-03 6.817e-021.0 1.860e+ 12 2.610e+00 2.523e+01 4.81 le-03 4.651 e-021.5 1.544e+ 12 9.284e+00 5.623e+01 1.562e-02 9.461 e-022.0 1.110e+l12 1.684e+01 7.709e+01 2.604e-02 1.192e-013.0 8.507e+10 4.061e+00 1.323e+01 5.510e-03 1.795e-024.0 8.353e+07 8.082e-03 2.146e-02 9.998e-06 2.655e-05Totals 1.726e+13 3.620e+01 2.204e+02 5.848e-02 3.714e-01file %208/Examples/CaseFiles/HTML/Cont... 9/29/2015 ATTACHMENT 9 Case Summary of Containment ShineATTACMENT 9Page 1 of 3MicroShield 8.02Nathan Hogue (8.00-0000)Date ByCheckedFilename IRun Date I Run Time I Duration JContainl .msd September 29, 2015 1:23:52 PM 00:00:00Project InfoCase Title Containment ShineDescription IFuel Element Failure Accident AnalysesGeometr 13 -Rectangular VolumeSource DimensionsLength 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 11.82 in) j n nShieldsShield N Dimension Material Densit ____________Source 6.12e+09 cm3  Air 0.00122Shield 1 I 30.5 cm I Concrete I 2.35Air Gap Air 0.00122Source Input: Grouping Method -Standard IndicesNumber of Groups: 25Lower Energy Cutoff: 0.0 15Photons < 0.015: Included__ _ _ _ _ _ _ _ Library: Grove .........._ __ __ __ __ __ _Nuclide Ci Bq JLCi/cm3  Bq/cm3I-131 1 .3399e-00 1 4.9578e+009 2.1 900e-005 8.1 030e-0011-132 2.6003e-001 9.6213e+009 4.2500e-005 1.5725e+000I- 133 4.0198e-001 1.4873e+010 6.5700e-005 2.4309e+0001-134 4.9682e-001 1 .8382e+0 10 8.1 200e-005 3 .0044e+000I- 135 4.0932e-001 1.5 145e+010 6.6900e-005 2.4753e+000Kr-85 6.0940e-004 2.2548e+007 9.9600e-008 3.6852e-003Kr-85m 1.4256e-001 5.2747e+009 2.3300e-005 8.6210e-001Kr-87 2.7288e-001 1 .0097e+010 4.4600e-005 1 .6502e+000Kr-88 3.8852e-001 1.4375e+010 6.3500e-005 2.3495e+000Kr-89 4.9253e-001 1 .8224e+01 0 8.0500e-005 2.9785e+000Kr-90 4.9253e-00 1 1 .8224e+0 10 8 .0500e-005 2.9785e+000file:///C:/Program%2OFiles%20(x86)/MicroShield%208/Examples/CaseFiles/HTML/Cont... 9/29/2015 Case Summary of Containment ShineATTACHMENT 9Page 2 of 3Xe-1335.4454e-0012.0148e+0108.9000e-0053 .2930e+000Xe-135 1 .2482e-001 4.6182e+009 2.0400e-005 7.5480e-001Xe.-135m j 1.2176e-001 J 4.5050e+009 j 1.9900e-005 j 7.3630e-001Xe- 137 [ 6.3632e-001 J 2.3544e+010 J 1.0400e-004 j 3.8480e+000Xe- 138 j 6.7303 e-001 Jj 2.4902e+010 J 1.1000e-004 4.0700e+000Buildup: The material reference is Shield 1Integration ParametersX Direction 10Y ireto II20ZDirection 20____________ Results -Dose Point # 1 -(1890,914,914) cm ______Fluence Rate Fluence Rate Exposure Rate Exposure RateEnergy (MeV) Activity (Photons/sec) MeV/cmn2sec MeV/cm2/sec mR/hr mR/hr_________No Buildup With Buildup No Buildup With Buildup0.015 5.847e+09 2.51 2e-254 7.735e-26 2.1 54e-255 6.635e-270.03 1.247e+ 10 1.365e-36 5.739e-25 1.353e-38 5.688e-270.08 7.524e+09 2.819e-06 7.055e-05 4.460e-09 1.116e-070.1 6.447e+09 2.994e-05 1.125e-03 4.580e-08 1.721e-060.15 6.836e+09 5.094e-04 2.565e-02 8.388e-07 4.224e-050.2 1.617e+10 5.150e-03 2.522e-01 9.089e-06 4.45 le-040.3 1.051e+10 2.051e-02 7.532e-01 3.891e-05 1.429e-030.4 2.244e+10 1 .468e-01 3.969e+00 2.860e-04 7.733e-030.5 3.752e+10 6.074e-01 1.259e+01 1.192e-03 2.472e-020.6 2.587e+10 8.590e-0I 1.422e+01 1.677e-03 2.775e-020.8 5.049e+ 10 5.017e+00 5.859e+01 9.542e-03 1.1 14e-011.0 3.060e+ 10 6.885e+00 6.184e+01 1.269e-02 1.140e-011.5 2.473e+10 2.251e+01 1.305e+02 3.787e-02 2.195e-012.0 2.762e+ 10 6.174e+01 2.755e+02 9.547e-02 4.260e-013.0 3.396e+09 2.335e+01 7.576e+I01 3.168e-02 1.028e-014.0 8.371e+08 1.158e+01 3.094e+01 1.432e-02 3.828e-02Totals 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 RateEnergy (MeV) Activity (Photons/sec) MeV/cm2/sec MeV/cm2/sec mR/hr mR/hr_________No Buildup With Buildup No Buildup With Buildup0.015 5.847e+09 5.266e-265 3.424e-28 4.517e-266 2.937e-290.03 1.247e+ 10 1.489e-39 2.540e-27 1.475e-41 2.518e-290.08 7.524e+09 8.870e-09 2.501 e-07 1 .404e- 11 3.957e- 100.1 6.447e+09 9.718e-08 4.276e-06 1.487e- 10 6.542e-090.15 6.836e+09 1.838e-06 1.125e-04 3.027e-09 1.853e-070.2 1.617e+10 2.044e-05 1.223e-03 3.608e-08 2.158e-060.3 1.051 e+l 0 9.363e-05 4.113e-03 1.776e-07 7.801 e-060.4 2.244e+ 10 7.337e-04 2.319e-02 1.430e-06 4.518e-050.5 3.752e+ 10 3.240e-03 7.695e-02 6.359e-06 1.51 le-04file:///C:/Program%2OFiles%20(x86)/MicroShield%208/Examples/CaseFiles/HTML/Cont... 9/29/2015 Cas~1jp~xp~P~aimnent Shine Page 3 of 3Page 3 of 30.62.587e+104.807e-038.974e-029.383e-061 .752e-040.8 5.049e+10 2.997e-02 3.839e-01 5.700e-05 7.301e-041.0 3.060e+10 4.292e-02 4.149e-01 7.912e-05 7.649e-041.5 2.473e+10 1.486e-01 9.003e-01 2.501e-04 1.515e-032.0 2.762e+ 10 4.1 89e-0 1 1.91 8e+00 6.477e-04 2.965e-033.0 3.396e+09 1.621e-01 5.280e-01 2.199e-04 7.163e-044.0 8.371e+08 8.099e-02 2.151e-01 1.002e-04 2.660e-04Totals 2.893e+11 8.924e-01 4.555e+OO 1.371e-03 7.339e-039/29/2015 ATTACHMENT 9 Case Summary of Containment ShineATTACHMENT 9Page 1 of 3~MicroShield 8.02Nathan Hogue (8.00-0000)Date I By ICheckedI Filename )Run Date I Run Time I Duration jI Containl .msd j September 29,2015 1:15:41 PM 00:00:01 j_________________Project InfoCase Title Containment ShineDescription Fuel Experiment Accident AnalysesGeometry 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 Y2 1.5e+4 cm (492it 1.5 914.0em (29 fi11.8 914.0Ocm (29 fi11.8.2 in) in) in)ShieldsShield N Dimension Material Density ___________Source 6.12e+09 cm3  Air 0.00122Shield 1 I 30.5 cm I Concrete I 2.35Air Gap Air 0.00122Source Input: Grouping Method -Standard IndicesNumber of Groups: 25Lower Energy Cutoff: 0.0 15Photons < 0.015: Included_______________ Library: Grove ________Nuclide Ci Bg JLCi/cm3  Bg/cm3I- 131 8.0763e+000 2.9882e+011 1.3200e-003 4.8840e+0011- 132 1.7866e+001 6.6104e+011 2.9200e-003 1.0804e+0021-133 3.8240e+001 1.4149e+012 6.2500e-003 2.3125e+0021- 134 4.3563e+001 1.611 8e+012 7.1 200e-003 2.6344e+002I- 135 3.6160e+001 1.3379e+012 5.9100e-003 2.1867e+002Kr-85 1 .6459e-003 6.0897e+007__ 2.6900e-007 9.9530e-003Kr-85m 7.2810e+000 2.6940e+011 1.1900e-003 4.4030e+001Kr-87 1 .4807e+001 5.4785e+011I 2.4200e-003 8.9540e+001Kr-88 2.0864e+001 7.7196e+011I 3.4100e-003 1 .2617e+002Kr-89 2.6676e+001 9.8703e+011 4.3600e-003 1.6132e+002Kr-90 2.6309e+001 9.7344e+011 4.3000e-003 1.5910e+002file 9/29/2015 Case Summary of Containment ShineATTACHMENT 9Page 2 of 3Xe-1331.81 72e+0016.7236e+01 12.9700e-0031 .0989e+002Xe- 135 1,I3093 e+i001 4.8446e+011 2.1400e-003 7.9180e+001iXe-I135m 6,4856e+'000 [ 2.3997e+01 1 1 .0600e-003 j 3.9220e+001Xe- 137 3.4386e+001 1 .2723 e+012 5.6200e-003 j 2.0794e+002Xe- 138 j 3.5915e+001 [ 1.3289e+012 5.8700e-003 2.1719e+002Buildup: The material reference is Shield 1Integration ParametersX Direction 10Y Direction I 20Z Direction 20__________Results -Dose Point # 1 -(1890,914,914) cmFluence Rate Fluence Rate Exposure Rate Exposure RateEnergy (MeV) Activity (Photons/sec) MeV/cm2/sec MeV/cmZ/sec mR/hr mR/hrNo Buildup With Buildup No Buildup With Buildup0.015 2.904e+11I 1.248e-252 3.842e-24 1.070e-253 3.296e-250.03 5.010e+ll 5.485e-35 2.306e-23 5.436e-37 2.285e-250.08 2.546e+ 11 9.536e-05 2.387e-03 1.509e-07 3.777e-060.1 3.444e+ 11 1.599e-03 6.008e-02 2.447e-06 9.192e-050.15 3.872e+ 11 2.885e-02 1.453e+00 4.751le-05 2.393e-030.2 1.110e+12 3.537e-01 1.732e+01 6.242e-04 3.056e-020.3 5.920e+11 1 I.156e+00 4.243 e+01 2.192e-03 8.048e-020.4 1 .346e+12 8.808e+00 2.382e+02 1.716e-02 4.640e-010.5 2.718e+12 4.399e+01 9.123e+02 8.635e-02 1.791e+000.6 1.808e+12 6.003e+01 9.936e+02 1.172e-01 1.939e+000.8 4.038e+12 4.012e+02 4.686e+03 7.631e-01 8.912e+001.0 2.157e+12 4.855e+02 4.360e+03 8.949e-01 8.037e+001.5 1.735e+ 12 1.580e+03 9.156e+03 2.658e+00 1.540e+012.0 1.593e+12 3.562e+03 1.589e+04 5.508e+00 2.458e+t013.0 1.838e+11 1.264e+03 4.101le+03 1.715e+00 5.563e+004.0 4.532e+10 6.268e+02 1.675e+03 7.754e-01 2.072e+00Totals 1.910e+13 8.033e+03 4.208e+04 1.254e+01 6.887e+01Results -Dose Point # 2 -(15000,914,914) cmF'luence Rate Fluence Rate Exposure Rate Exposure RateEnergy (MeV) Activity (Photons/sec) MeV/cm2/sec MeV/cm2/see mR/hr mRlhr_____________No Buildup With Buildup No Buildup With Buildup0.015 2.904e+ 11 2.616e-263 1.701 e-26 2.244e-264 1.459e-270.03 5.010e-il 1 5.981e-38 1.021e-25 5.928e-40 1.012e-270.08 2.546e+11 3.001e-07 8.461e-06 4.749e-10 1.339e-080.1 3.444e+ 11 5.191 e-06 2.284e-04 7.942e-09 3.494e-070.15 3.872e+11 1,041e-04 6.374e-03 1.714e-07 1.050e-050.2 1.110e+ 12 1.404e-03 8.395e-02 2.478e-06 1.482e-040.3 5.920e+11I 5.274e-03 2.3 17e-01 1.000e-05 4.394e-040.4 1 .346e+12 4.403e-02 1.392e+00 8.579e-05 2.71 le-030.5 2.718e+ 12 2.347e-01 5.574e+00 4.606e-04 1.094e-02fle :///C :/Program%2OFiles%20(x86)/MicroShield%208/Exampies/CaseFiles/HTML/Cont... 9/29/2015 Case 1;n~ainm~ent ShinePae3o3Page 3 of 30.61.808e+123 .359e-01I6.271 e+006.5 57e-041 .224e-020.8 4.038e+12 2.397e+00 3.070e+01 4.559e-03 5.839e-021.0 2.157e+12 3.026e+00 2.926e+01 5.579e-03 5.393e-021.5 1.735e+12 1.043e+01 6.318e+01 1.755e-02 1.063e-012.0 1.593e+12 2.416e+01 1.106e+02 3.737e-02 1.711e-O13.0 1.838e+11 8.774e+00 2.858e+01 1.190e-02 3.877e-024.0 4.532e+10 4.385e+00 1.164e+01 5.425e-03 1.440e-02Totals 1.910e+13 5.380e+0O1 2.875e+02 8.360e-02 4.694e-01fie:///C:/Programi%2OFi~es%20(x86)/MicroShield%208/Examples/CaseFiles/HTML/Cont... 9/29/2015 ATTACHMENT 9 ATTACHMENT 10wIND ROSE PLOT:Station #03945 -COLUMBIAIREGIONAL ARPT, MODISPLAY:Wind SpeedDirection (blowing from)NORTH15%12%9%WESTEASTWIND SPEED(mis)U >=11 IU8.8-11.11 5.7-.88* 3.6- 5.7E]2.1 -3.62.1Calms: 1.15%SOUTHDATA PERIOD: COMPANY NAME:Start Date: 11111961 -00:00End Date: 1213111969 -21:00'MODELER:CALM WINDS: TOTAL COUNT:1.15% 73020 hr..AVG. WIND SPEED: DATE: PROJECT NO.:4.70 mls 912312015WRPLOT View -Lakes Envlonmanlat Software Wind Class Frequency Distribution:z40-35-30-%/252015105mACalms0-0.5- 2.12.1- 3.6 3.6- 5.7Wind Class (mis)0.8>= 11.15.7- 8.88.8 -11.1 ATTACHMENT 11WIND ROSE PLOT:Station #03945 -COLUMBIA/REGIONAL ARPT, MODISPLAY:Wind SpeedDirection (blowing from)NORTH20%16%12%WESTEASTWIND SPEED(mis)U>= 11.11n 8.8-11.11 57- 8.83.6- 5.7m2.1 -3.6LI 0.5- 2.1Calmns: 1.83%SOUTHCOMwMENTS: DATA PERIOD: COMPANY NAME:Start Date: 1/1/1170 -00:00End Date: 1213111990 -23:00 ___________ ____________MODELER:CALM WINOS: TOTAL COUNT:1.83% 154387 hrs.AVG. WIND SPEED: DATE; PROJECT NO.:4.44 mole 912312015WRPLOT View -Lakes Envirotnmental Software Wind Class Frequency DistributionA 3-z-r'J44~1- 4* 41 I.41 I..625--20-15-10-!0.00-0.7>= 11.1Calms0.5- 2.12.1- 3.6 3.6- 5.7 5.7- 8.8Wind Class (mis)8.8- 11.1kI VIUW ! .U.U "
ATTACHMENT 12WIND ROSE PLOT:Station #03945 -COLUMBIA/REGIONAL ARPT, MODISPLAY:Wind SpeedDirection (blowing from)NORTH20%16%12%WESTEASTWIND SPEED(mis)U = 11.1n8.8- 11.11 57- 8.8*36- 5.7D 21 -3.6EJ 0.5 -2.1Calms: 1.63%:SOUTHCOMMJENTS: DATA PERIOD: COMP~ANY NAAE:Start Date: 11111961 -00:00End Date: 12131/1990.-23:00MODELER:CALM WINDS: TOTAL COUNT:1.63% 227407 hrs.AVG. WINO SPEED: DATE: PROJECT NO,:4.52 mls 9123/2015WRPLOT View.- Lakes Environmental Software Wind Class Frequency DistributionIl ~.. I-T.lr-i40*35.L.4.301-4.27.%/25-20-15-10-5-'0-0.7>= 11.10.5- 2.12.1 -3.6 3.6 -5.7 5.7 -8.8Wind Class (m/s)8.8- 11.1view 1-rewar 7A.Uo -LX~e uwltrmnmlenltimrw/19 ATTACHMENT 13Stack Effluent Releases -Calendar Years 2005 to 20142005 2006 2007 2008 2009 2010 2011 2012 2103 2014 AverageIsotope (% 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 69.91238C-14 0.777 0.74 0.793 0.7867 0.613 0.58 0.477 0.723 0.0083 0.0079 0.55059Os-191 0.0011 0.0018 0.0066 4.1739 0.0294 0.0008 0.0003 0.0001 0.0002 0.468241-131 0.0921 0.0435 0.0401 0.0782 0.6035 0.0415 0.0506 0.0503 0.0169 0.2201 0.12368Ce-144 0.1165 0.0852 0.10085Co-60 0.0853 0.0792 0.3372 0.0784 0.0084 0.0049 0.0054 0.08554H-3 0.0732 0.052 1 0.0485 0.0527 0.0328 0.0353 0.0496 0.0426 0.0633 0.0558 0.05059Kr-79 0.0482 0.0274 0.0378Sc-46 0.0263 0.0022 0.01425K-40 0.0093 0.0164 0.01 0.01 19Cd-109 0.0112 0.01 121-125 0.0215 0.0041 0.0021 0.0073 0.0037 0.00774Fe-59 0.0038 0.0038Se-75 0.0005 0.0057 0.003 1Sb-125 0.0026 0.0026Zn-65 0.0005 0.001 0.0026 0.0009 0.00 125Htg-203 0.0002 0.001 0.0002 0.0013 0.0033 0.0012Cs-137 0.0007 0.0013 0.0006 0.0003 0.0004 0.0012 0.00075Zr-95 0.0005 0.0005 0.00051-133 0.0003 0.0001 0.0001 0.0001 0.0003 0.0001 0.0001 0.0001 0.003 0.00047Sn-i113 0.0009 0.0003 0.000 1 0.00043Au-196 0.0005 0.0003 0.0004 0.0003 0.0004 0.00038Gd-153 0.0003 0.00031 of 2 ATTACHMENT 13Stack Effluent Releases -Calendar Years 2005 to 2014Cu-67 0.0003 0.0003Pa-233 0.0002 0.0003 0.00025S-35 0.000 1 0.000 1 0.0005 0.0002 0.00023Hf-181 0.0004 0.0001 0.000 1 0.0002 0.0002Ce-141 0.0003 0.0002 0.000 1 0.0002Xe-133 0.0002 0.0002Ba-140 0.0003 0.0002 0.0001 0.0002 0.0002Nb-95 0.0003 0.000 1 0.0002Br-82 0.0002 0.000 1 0.00015Co-58 0.0001 0.0001 0.0002 0.00013As-77 0.0002 0.0001 0.0001 0.00013Ce-139 0.000 1 0.000 1 0.000 1Ru-103 0.0001 0.000 1 0.0001 0.0001Mn-54 0.0001 0.0001Be-7 0.000 1 0.000 1Co-57 0.000 1 0.000 1Hf-175 0.000 1 0.000 1 0.0001Xe- 135m 0.000 1 0.00012 of 2  
ATTACHMENT 12WIND ROSE PLOT:Station #03945 -COLUMBIA/REGIONAL ARPT, MODISPLAY:Wind SpeedDirection (blowing from)NORTH20%16%12%WESTEASTWIND SPEED(mis)U = 11.1n8.8- 11.11 57- 8.8*36- 5.7D 21 -3.6EJ 0.5 -2.1Calms: 1.63%:SOUTHCOMMJENTS: DATA PERIOD: COMP~ANY NAAE:Start Date: 11111961 -00:00End Date: 12131/1990.-23:00MODELER:CALM WINDS: TOTAL COUNT:1.63% 227407 hrs.AVG. WINO SPEED: DATE: PROJECT NO,:4.52 mls 9123/2015WRPLOT View.- Lakes Environmental Software Wind Class Frequency DistributionIl ~.. I-T.lr-i40*35.L.4.301-4.27.%/25-20-15-10 '0-0.7>= 11.10.5- 2.12.1 -3.6 3.6 -5.7 5.7 -8.8Wind Class (m/s)8.8- 11.1view 1-rewar 7A.Uo -LX~e uwltrmnmlenltimrw/19 ATTACHMENT 13Stack Effluent Releases -Calendar Years 2005 to 20142005 2006 2007 2008 2009 2010 2011 2012 2103 2014 AverageIsotope (% 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 69.91238C-14 0.777 0.74 0.793 0.7867 0.613 0.58 0.477 0.723 0.0083 0.0079 0.55059Os-191 0.0011 0.0018 0.0066 4.1739 0.0294 0.0008 0.0003 0.0001 0.0002 0.468241-131 0.0921 0.0435 0.0401 0.0782 0.6035 0.0415 0.0506 0.0503 0.0169 0.2201 0.12368Ce-144 0.1165 0.0852 0.10085Co-60 0.0853 0.0792 0.3372 0.0784 0.0084 0.0049 0.0054 0.08554H-3 0.0732 0.052 1 0.0485 0.0527 0.0328 0.0353 0.0496 0.0426 0.0633 0.0558 0.05059Kr-79 0.0482 0.0274 0.0378Sc-46 0.0263 0.0022 0.01425K-40 0.0093 0.0164 0.01 0.01 19Cd-109 0.0112 0.01 121-125 0.0215 0.0041 0.0021 0.0073 0.0037 0.00774Fe-59 0.0038 0.0038Se-75 0.0005 0.0057 0.003 1Sb-125 0.0026 0.0026Zn-65 0.0005 0.001 0.0026 0.0009 0.00 125Htg-203 0.0002 0.001 0.0002 0.0013 0.0033 0.0012Cs-137 0.0007 0.0013 0.0006 0.0003 0.0004 0.0012 0.00075Zr-95 0.0005 0.0005 0.00051-133 0.0003 0.0001 0.0001 0.0001 0.0003 0.0001 0.0001 0.0001 0.003 0.00047Sn-i113 0.0009 0.0003 0.000 1 0.00043Au-196 0.0005 0.0003 0.0004 0.0003 0.0004 0.00038Gd-153 0.0003 0.00031 of 2 ATTACHMENT 13Stack Effluent Releases -Calendar Years 2005 to 2014Cu-67 0.0003 0.0003Pa-233 0.0002 0.0003 0.00025S-35 0.000 1 0.000 1 0.0005 0.0002 0.00023Hf-181 0.0004 0.0001 0.000 1 0.0002 0.0002Ce-141 0.0003 0.0002 0.000 1 0.0002Xe-133 0.0002 0.0002Ba-140 0.0003 0.0002 0.0001 0.0002 0.0002Nb-95 0.0003 0.000 1 0.0002Br-82 0.0002 0.000 1 0.00015Co-58 0.0001 0.0001 0.0002 0.00013As-77 0.0002 0.0001 0.0001 0.00013Ce-139 0.000 1 0.000 1 0.000 1Ru-103 0.0001 0.000 1 0.0001 0.0001Mn-54 0.0001 0.0001Be-7 0.000 1 0.000 1Co-57 0.000 1 0.000 1Hf-175 0.000 1 0.000 1 0.0001Xe- 135m 0.000 1 0.00012 of 2  
}}
}}

Revision as of 05:11, 5 April 2018

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: Butler R A, Fruits J L
Univ of Missouri - Columbia
To:
Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML15275A281 List:
References
TAC ME1580
Download: ML15275A314 (202)
v • d • e


Text

UNWVERSITY of MISSOURIRESEARCH REACTOR CENTEROctober 1, 2015U.S. Nuclear Regulatory CommissionAttention: Document Control DeskMail Station P 1-37Washington, DC 20555-000 1REFERENCE: Docket 50-186University of Missouri -Columbia Research ReactorAmended Facility License R- 103

SUBJECT:

Written communication as specified by 10 CFR 50.4(b)(1) regarding responses to the"University of Missouri at Columbia -Request for Additional Information Regardingthe Renewal of Facility Operating License No. R-l103 for the University of.Missouri atColumbia Research Reactor (TACNo. ME1580)," dated April 17, 2015On August 31, 2006, the University of Missouri-Columbia Research Reactor (MURR) submitted arequest to the U.S. Nuclear Regulatory Commission (NRC) to renew Amended Facility OperatingLicense R-103.On May 6, 2010, the NRC requested additional information and clarification regarding the renewalrequest in the form of nineteen (19) Complex Questions. By letter dated September 3, 2010, MUJRRresponded to seven (7) of those Complex Questions.On June 1, 2010, the NRC requested additional information and clarification regarding the renewalrequest in the form of one hundred and sixty-seven (167) 45-Day Response Questions. By letter datedJuly 16, 2010, MURR responded to forty-seven (47) of those 45-Day Response Questions.On July 14, 2010, via electronic mail (email), MIURR requested additional time to respond to theremaining one hundred and twenty (120) 45-Day Response Questions. By letter dated August 4, 2010,the NRC granted the request. By letter dated August 31, 2010, MURR responded to fifty-three (53) of the45-Day Response Questions.On September 1, 2010, via email, 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.eduFighting 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 theremaining 45-Day Questions. By letter dated October 13, 2010, the NRC granted the extension request.By letter dated October 29, 2010, MURR responded to sixteen (16) of the remaining 45-Day ResponseQuestions and two (2) of the remaining Complex Questions.By letter dated November 30, 2010, MURR responded to twelve (12) of the remaining 45-Day ResponseQuestions.On December 1, 2010, via email, MURR requested additional time to respond to the remaining 45-DayResponse and Complex Questions. By letter dated December 13, 2010, the NRC granted the extensionrequest.On January 14, 2011, via email, MURK requested additional time to respond to the remaining 45-DayResponse and Complex Questions. By letter dated February 1, 2011, the NRC granted the extensionrequest.By letter dated March 11, 2011, MURR responded to twenty-one (21) of the remaining 45-Day ResponseQuestions.On May 27, 2011, via email, MURR requested additional time to respond to the remaining 45-DayResponse 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 andComplex Questions.On September 30, 2011, via email, MURR requested additional time to respond to the remaining theremaining 45-Day Response and Complex Questions. By letter dated November 10, 2011, the N-RCgranted the request.By letter dated January 6, 2012, MURK responded to four (4) of the remaining 45-Day Response andComplex 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 theremaining 45-Day Response and Complex Questions. By letter dated January 26, 2012, the NRC grantedthe request.On April 12, 2012, via email, MURR requested additional time to respond to the remaining the remaining45-Day Response and Complex Questions.By letter dated June 28, 2012, MURR responded to the remaining six (6) 45-Day Response and ComplexQuestions. With that set of responses, all 45-Day Response and Complex Questions had been addressed.2 of 86 On December 20, 2012, the NRC requested a copy of the current Physical Security Plan (PSP) andOperator Requalification Program.By letter dated January 4, 2013, MURR provided the NRC a copy of the current PSP and OperatorRequalification Program.On February 11, 2013, the NRC requested updated financial information in the form of four (4) questionsbecause 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) questionsregarding significant changes to the MIURR facility since submittal of the licensing renewal application inAugust 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 formoften (10) questions.*If there are any questions regarding this response, please contact me at (573) 882-5319 orFruitsJ@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 Directorxc: Reactor Advisory CommitteeReactor Safety SubcommitteeDr. Garnett S. Stokes, ProvostDr. Henry C. Foley, Senior Vice Chancellor for ResearchMr. Alexander Adams Jr., U.S. Nuclear Regulatory CommissionMr. Geoffrey Wertz, U.S. Nuclear Regulatory CommissionMr. Johnny Eads, U.S. Nuclear Regulatory CommissionAttachments: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 System8. Meteorological Data (Wind Speed and Class) -1961 to 19699. Meteorological Data (Wind Speed and Class) -1970 to 199010. Meteorological Data (Wind Speed and Class) -1961 to 199011. 10 CFR 835, Appendix C, "Derived Air Concentration (DAC) for Workers from ExternalExposure during Immersion in a Cloud of Airborne Radioactive Material"12. Micro Shield 8.02 Dose Calculations for a Fuel Handling, Fuel Failure, and Fueled ExperimentFailure Accidents13. Stack Effluent Releases -Calendar Years 2005 to 2014JACQUELINE 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 describesthe control blade drop process including the effect of the dashpot, but does not describe the methodfor determining the drop time nor does it explain the basis for the 80 percent insertion times. Thescram times and reactivity worths used or" assumed for the various analyses in the SAR are notclearly described or provided. NUREG-15 3 7, Section 4. 5.3, "Operating Limits, "provides guidancethat 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 theassociated control blade insertion reactivity per blade. Provide typical control blade fullinsertion 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 (seeAttachment 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 rodassembly, is removed by an electro-mechanical relay contact which allows the control blade to dropand 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 whentriggered, causes the electro-mechanical relay to change state stopping the blade droptimer/chronometer. The control blade drop time is then displayed on a meter on the reactor controlroom instrument panel. Table 1 provides the minimum, average and maximum drop times of allfour (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 0.48Average 0.50 0.54 0.50 0.52Maximum 0.59 0.58 0.54 0.54Current MURR Technical Specification 3 .2.c requires the capability of inserting the shim controlblades to their 20% withdrawn position (or 80% inserted) in less than 0.7 seconds. This ensuresprompt shutdown of the reactor in the event a reactor scram signal, manual or automatic, isreceived. The 20% withdrawn position is defined as 20% of the control blade full travel of 26inches measured from the fully inserted position. Below the 20% withdrawn position the controlblade fall is cushioned by a dashpot assembly. Approximately 91% of the control blade total worthis inserted at the 20% position. This is an original design feature of the reactor and its purpose hasnot been altered in 49 years of operation. The same Technical Specification will remain in therelicensing Technical Specifications.The measured and calculated values for reactor core excess reactivity and shutdown margin areprovided 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 isfully withdrawn). Some of this information, calculated using older computer programs, can also befound on Table 4-12 of the SAR.Typical MURR operations involve a core change-out every week with eight (8) xenon-free fuelelements in various stages of burnup (mixed core operation) used at startup. The reactor coreexcess reactivity and shutdown margin values are verified after the weekly core change-out. Theverification is done during reactor startup, when the cold, clean critical control blade height ismeasured. 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 reactivityand shutdown margin in comparison to the Technical Specification limit of -0.020 Ak/k.Table 2 -Summary of Key Measured Reactor DataValueParaeter(Ak/k)Typial ota shi cotro blae wrth0.1364Typical total shim control blade worth at 80% inserted -0.1127Typical shim control blade worth at 80% inserted with the highest worth -0.0787control blade excluded (or fully withdrawn)Maximum reactor core excess reactivity after weekly core change-out +0.0400One-year average of reactor core excess reactivity (over 69 core change-outs) +0.0290Typical core sub-criticality with 3 shim control blades at 80% inserted and the -038control blade excluded (or fully withdrawn)-037Minimum shutdown margin allowed by Technical Specifications -0.0200Calculated Values:Reactor core excess reactivity and shutdown margin values were also calculated using the detailedMGNP 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) andall fresh shim control blades (most conservative), and (2) with a mixed core loading and mixedburnup control blades (typical MUIRR operation). Table 3 provides the calculated values.6 of 86 Table 3 -Summary of Key Calculated Reactor DataValueParameter -All Fresh Fuel and Fresh Control Blades Case (kkReactor core excess reactivity 0.0865Total shim control blade worth 0.1740Core sub-criticality with 3 shim control blades at 80% inserted and highest -0.0324worth control blade excluded (or stuck fully withdrawn)ValueParameter -Mixed Core / Mixed Control Blades Case (Ak/k)Reactor core excess reactivity 0.0445Total shim control blade worth 0.1517Core sub-cniticality with 3 shim control blades at 80% inserted and highest -0.0580worth 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 worthshim control blade fully withdrawn), the minimum reactor core shutdown margin required by theTechnical 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 withdrawalfrom subcritical, control blade run in, insertion of excess reactivity, etc.). For each suchevent, provide the control blade motion speeds and reactivities utilized to provide the SARanalyses, 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 occurif there were a double-ended shear of the 12-inch primary coolant piping on both sides of either thecold-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 bethe 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 systeminstrumentation can have trip values set for them at the appropriate coolant loop locations. In theRELAP modeling, a 150 millisecond time delay is set between the time a scram signal is receivedand the modeling of when the "insertion" of the control blades start. The insertion is covered by aninput table of fission and gamma reactor power as per set time steps after the reactor scrams. Thecode calculates linear values between these data points.7 of 86 Table 4 below provides the power assumed by RELAP seconds after shutdown compared to thecalculated power after shutdown, assuming 30 days of full power operation, using equation 2.66from Nuclear Reactor Engineering 3rd Edition by Samuel Glasstone and Alexander Sesonske'. Theequation is given in the upper right corner of the page along with the values of variables a and b touse 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 powerexcept for the values for 8, 9 and 10 seconds. From 10 to 150 seconds, the RELAP values areconservative 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 useconservative calculated values of reactor decay heat after the scram, which would correspond toslower insertion of the control blades.See the response to RAI 6.a for control blade drop times related to Insertion of Excess Reactivityaccidents.

References:

1Glasstone, S. and Sesonske, A., Nuclear Reactor Engineering 3rd Edition, prepared underTechnical Information Center, United States Department of Energy.8 of 86 Table 4 -Comparing RELAP Decay Heat to Calculated Decay Heat(Nuclear Reactor Engineering 3rd Edition: Equation 2.66)ts00.10.30.71.02.03.05.06.07.08.09.01020304050607080901001201501802002403004004205406008001,0002,0004,0006,0008,00010,000Power MW11.0A7.99061.97 171.25971.03730.93090.848 10.73450.693 10.65760.62920.60440.58190.45060.41000.38690.36980.35630.34530.33600.32800.32100.3090A 0.30720.30530.30150.28490.26590.243 10.23950.22140.21420.19570.18240.14620.11670. 10200.09270.0860Power MW11.00.50990.45780.42010.40480.376 10.35990.34000.333 10.32730.32230.3 1800.3 141Equation 2.66P/P0=5E-3 *a *[t4-b -(To + t4)-~b]after shutdownTo = 30 days operating period prior toshutdown(s)0.1 to 1010 to 150150 to 8E80.49700.43 170.39700.37400.35690.34340.33240.323 10.3 1500.30800.296 10.2821a12.0515.3127.43b0.06390.18070.29620.32300.30500.295 10.27860.25950.23680.233 10.21500.20780.18930. 17600.13980.11030.09570.08630.0796Note A: RELAP does not have a value entered for 0.1 seconds, but the linear value between 0 and0.3 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 thelicensee provide analyses and methods to demonstrate the secure storage of new and irradiated fuelwith a criticality limit of keff < 0.90. The NRC staff's review of the MURR SAR and HazardsSummary Report could not find a criticality analysis supporting the use of any fuel storagelocations outside of the core. Identify the locations that may be used for the storage of new orirradiate fuel, and provide supporting criticality analyses, or justify' why no additional informationis needed.As stated in SAR Section 9.2.1, there are 88 in-pool storage locations for new or irradiated fuelelements. These storage locations are situated in three (3) areas within the reactor pool and aredesignated as the "X," "Y" and "Z" storage baskets. The "Z" storage basket contains 48 fuelelement storage locations; consisting of two (2) levels, referred to as "upper" and "lower," of 24locations per level. The "X" and "Y" storage baskets each contain 20 fuel element storagelocations. 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 storagelocations. The "X" and "Y" storage baskets each had only six (6) storage locations at the timewhile the "Z" storage basket consisted of 16 storage locations -two racks (6 and 10) in the lowerlevel. In 1972, due to an increase in operating schedule and with an uprate in power from 5 to 10MWs in the near future, an additional rack of eight (8) storage locations was added to the lowerlevel of the "Z" basket, thus providing a total of 36 fuel element storage locations in the pool (24 inthe "Z" basket). Modification Record 72-7, "Additional In-Pool Fuel Storage Basket," documentsthe installation of the eight (8) element rack (Attachment 2). On page 2a of the ModificationRecord, the following is stated: "To determine the safety of installing an additional fuel rackbetween the present two, the system was modeled using the Exterminator II multi-group neutrondiffusion program. The physical model consisted of three adjacent rows of eight clean 775 gramU235 fuel elements. Each fuel element was surrounded by 0.25" thick boral as is the case in theactual design. For the fully loaded rack, the calculated Keff linlit was 0. 714. "A 1/M criticality plotof 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 increasedthe overall capacity of the "Z" storage basket from 24 to 38. Modification Record 76-3, "Upper 7Spent 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 ofanother level of elements was modelled using the Exterminator II neutron diffusion code. Thepresence of 24 rather than 14 elements on the second level was used for a 'factor of safety." Thecode predicts a value for Keff of 0.748. Thus, the above criteria is satisfied for fuel storage." Al/M criticality plot of the storage basket was also performed to verify the Exterminator II coderesults.In 1978, a 10 element rack was added to the upper level of the "Z" storage basket which increasedthe 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 results10 of 86 stated in Modification Record 76-3, which conservatively modeled 24 fuel elements instead of just14 elements.Because the criticality analyses for the "Z" storage basket are somewhat dated and vaguelydocumented, 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 followingdescribes the methodology and results.The "Z" storage basket stores fuel elements that have burnups of 0 to 150 MWds. The baskets arelined with 26- to 29-inch tall sheets of 0.25- to 0.3125-inch thick BaC (BORAL) as the absorbingmaterial to prevent the stored fuel configuration from reaching criticality. Figure 1 shows thelayout (i.e. a detailed MCNP model) of the lower "Z" storage basket configuration.-Stainless Steel-Fuel Element-Pool WaterBORALFigure 1 -Detailed MCNP model of the Lower "Z" Storage Basket ConfigurationThe upper 'Z' storage basket configuration layout shown in Figure 2 is very similar to the lowerbasket with the exception of lead shields surrounding the basket instead of stainless steel, as in thelower basket.* Lead ShieldFigure 2 -Detailed MCNP model of the Upper "Z" Storage Basket Configuration(Lead shields instead of stainless steel)11 of 86 vThe active region of the fuel elements in the lower and upper baskets is separated in height byapproximately seven (7) inches. Each fuel element in every storage location is modeled in fulldetail, with all 24 aluminum clad UAlx fuel plates. Figure 3 shows very detailed MCNP modelingof an individual MURR fuel element and the elements in their lower and upper "Z" storage basketconfigurations.Upper Level of "Z" Storage BasketLower Level of "Z" Storage BasketFigure 3 -Panels Showing Detailed MCNP Modeling of the Fuel Elements; the Left PanelShowing the Axial Configuration of the Fuel Elements in the Lower and Upper "Z" StorageBaskets and the Right Panel Showing a Cross-sectional View of a MURR Fuel ElementCriticality (i.e. KCODE) calculations using MCNP version 5 with the ENDJF/B-VII.O data librarieswere performed for two detailed instances of the "Z" storage basket configuration: (1) a singlebasket (lower), and (2) both lower and upper baskets together. All calculations were performed for20 million source particles. For the two instances, the basket(s) were filled to their maximumcapacities (24 fuel elements) with fresh, highly-enriched uranium (HiEU) UAlx MURR fuelelements. 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 twoconfigurations of the "Z" storage basket.Table 1 -KIf Values for Worst-Case "Z' Storage Basket ConfigurationsConfiguration Fuel Status Storage Capacity IefLower Fresh Max -24 Fuel Elements 0.49885Lower + Upper Fresh Max -48 Fuel Elements 0.5586212 of 86 On receipt, fresh (i.e., un-irradiated fuel) fuel elements may be stored outside the reactor pool in adry, 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 drystorage configuration containing the maximum allowable number of on-site stored fresh MURRfuel 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 anyenrichment, providing that no more than 5 kilograms of this amount is unirradiated;...". SixMURR fuel elements, containing 775 grams of uranium-235 each, equals 4.65 kilograms.Air--Fuel ElementPlywoodBoric AcidFigure 4 -Detailed MCNP Model Showing Fresh Fuel Dry Storage Filled to Possession LimitTo establish a full-scope criticality safety study, in addition to the configuration described in Figure4, two other configurations were also defined to capture the worst-case scenarios: (1) a floodedconfiguration storing the maximum allowable number of on-site stored fresh fuel, i.e., six fresh fuelelements (see Figure 4 where air is replaced with water), and (2) a flooded configuration with therack filled to its maximum capacity which equals eight fresh fuel elements (see Figure 5).,...WaterFuel ElementPlywoodBoric AcidFigure 5 -Detailed MCNP Model Showing Fresh Fuel Dry Storage Filled to Physical Capacity13 of 86 Again, the criticality (i.e. KCODE) calculations performed using MCNP version 5 with theENDF/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 mostconservative, worst-cases (in terms of attaining criticality) for the fresh fuel storage configurations.Table 2 -K~ff Values for Worst-Case Fresh Fuel Storage ConfigurationsConfiguration Fuel Status Storage Capacity KfDry (Air) Fresh License Max -6 Fuel Elements 0.02344Flooded Fresh License Max -6 Fuel Elements 0.36228Flooded Fresh Storage Max -8 Fuel Elements 0.36258In 1991, due to the inability to ship spent fuel from the facility because the cask (GE-700) that wasused to ship research reactor fuel was removed from service, two (2) new fuel storage baskets werefabricated 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 theinstallation of the additional 24 fuel element storage locations (Attachment 5). On page 2 of theModification Record, the following is stated: "The evaluation performed for each MHJA basketwill 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"Xand "Y"' storage baskets, which increased the total storage capacity in these baskets from 36 to 40locations. 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 thenew "X" and "Y" storage baskets (Attachment 6). This Modification Record contains a detaileddescription of the criticality analysis performed for these two baskets using the MCNP code. Onpage 4 of the Modification Record, the following is stated: "The MCNP model was used tocalculate a Ke,'7 value of 0. 635 for one fuel basket fully loaded with twenty (20) 'fresh" 775 gramU-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 andfresh fuel storage in the vault, using conservative, worst-case assumptions of all fresh fuelelements, indicate IKfr values much less than the MURR Technical Specification Limit of 0.9 (novalue 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 CorePhysics Parameters, "provide guidance that the licensee should identify their analytical methods,including calculations of individual control blade worths, core excess reactivity, and coefficients ofreactivity, and compare the results with experimental measurements. The MURR SAR, Section 4.5states that analyses have been performed using PDQ, EXTREMINATOR, and BOLD-VENTUREcodes using RO, RZ, and ROZ models. The NRC staff noted other analyses (e.g., the RAI responsessupporting the NRC staff review of License Amendment No. 36, ADAMS Accession Nos.ML11237A088 and ML12150A052) used Estimated Critical Position (ECP) comparisons with theMonte Carlo Neutron Production code. The design code used to support the T&H analysis appearsto be DIF3D. The NRC staff is not clear as to which analytical method is the final supportinganalysis to be reviewed for the MURR license renewal application. The final supporting analysisshould 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.ML10306002 1), it is not clear how the stuck control blade was determined, what the relativereactivity worth is for the other control blades in the shutdown margin (SDM) analysis, andwhether 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, orjustify why this information is not needed.Historically, neutron physics modeling and analyses at MUJRR have been performed using severalmulti-group and multi-dimensional neutron diffusion theory codes such as PDQ,EXTERMINATOR-fl and BOLD VENTURE. Since the BOLD VENTURE core model wasbenchmarked against the destructive analysis of a highly-enriched uranium (HEU) MURR fuelelement for the license renewal application submitted to the NRC in August of 2006, MURR usedresults provided by the above set of neutronics codes.Since then, MURR core physics analyses have switched to using newer, state-of-the-art programssuch as MCNP for neutronic analysis. For a compact core such as MURR, it is preferable to use atransport 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 nowroutinely used for all calculations of core Ker, critical control blade height, detailed powerdistribution, and experimental fluxes/reaction rates.As part of the on-going collaboration, which started in 2006, between MURR staff and ArgonneNational Laboratory (ANL) analysts for the purpose of determining the feasibility of convertingMURR from HIEU to low-enriched uranium (LEU) fuel, ANL has assembled a neutronics analysiscode suite utilizing WIMS-ALNL, REBUS-DIF3D and REBUS-MCNP. Figure 1 below illustratesthe linkage of the codes in the analysis suite.The suite of programs, or codes, was used to provide detailed (radial, axial and azimuthal) fuelcomposition for partially burned fuel elements. Since MURR routinely operates with a fuel cycleutilizing a mixed burnup core, realistic experimental flux, reaction rates and power peaking valueshave to be evaluated for the typical core weekly cycles rather than for an all-fresh core. The15 of 86 detailed fuel composition data obtained is then subsequently used in a MCNP calculation to obtainthe worst-case power peaking factors and heat flux values used in the thermal-hydraulic analysis.tMCNP Runs (Outside REBUS)Produt *Detailed power distributionsLumped Fission PrdutExperimental fluxes / reaction ratesCross-Sections in 69 groupsMCNPinufi-UREBUS 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-ItIDIF3D Neutronics Solver* Cross-section interpolation* Flux solverIRegion FluxesRegion Reaction RatesI4Fi~A N~uCross-Section LibraryFigure 1 -Linkage of the Codes Used in the Analysis SuiteThe following is a brief description of each of the programs within the ANL neutronic analysissuite:WIMS-ANL: WIMS-ANL is a one-dimensional lattice physics code used to generate burnupdependent, 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 wasdeveloped by ANE based on the neutron spectrum that exists in the MURR core. This multi-groupdata 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 canmodel systems in a number of geometries. REBUS is a depletion code that utilizes neutron fluxesfrom a neutronics solver and cross-section data to solve isotopic transmutation calculations. A16 of 86 detailed O-R-Z diffusion MURR model was developed for DIF3D. The depleted corecharacteristics (plate-by-plate and axially-segmented atom densities) can be saved and passed on toMCNP for more detailed neutronics analyses.MCNP: MCNP is a continuous energy Monte Carlo neutron transport code. MCNP is capable ofmodeling the heterogeneous details of the MURR fuel elements, core structures, and experimentalfacilities 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 modelcores of depleted and fresh elements.ANL had performed extensive work to validate the above set of neutron physics codes and modelsfor application to MURR. The MCNP and DIF3D models were benchmarked against availableexperimental data [Ref. 1].In order to speed up routine neutronics calculations, where such detailed axial, radial and azimuthalfuel composition is not necessary, MURR utilizes the MONTEBURNS program. MONTEBURNSis 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 depletioncalculations and that of MCNP5 for continuous energy, flux and reaction rate as well as criticalitycalculations.MONTEBUJRNS by itself is not designed to handle transient calculations such as during the periodfrom reactor startup through critical and then on to steady-state reactor operation since it involvescontrol blade motion due to poison buildup as well as from fuel depletion. However, with the helpof in-house developed routines, a code system including MCNP and MONTEBURNS wasdeveloped 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-physicsanalysis is shown in Figure 2.17 of 86 A/
  • MONTEBURNS 2.0 -time dependent stepwiseMCNP coupled ORIGEN nuclear burnup codeSsystem (LANL)*Critical Rod search routine -adjustments tocontrol rod height based on repeated detailedMCNP KCODE calculations.*Returns a critical rod height if the KCODEk~e is 1.0000+/-0.03% and control rod is lessthan maximum travelFigure 2 -Code Suite Flow Diagram Implemented at MURR for Routine Core Physics AnalysisThese "routine" calculations utilize a very detailed MCNP MURR core model that has thefollowing key capabilities:*It can model MURR's "mixed-core" weekly fuel configuration, i.e., atom densities of variousisotopes in the fuel matrix, for a range of fuel element burnups from fresh (0 MWd or no fueldepletion) to spent status. For the Estimated Critical Position (ECP) calculations, fuelelement definitions can be individually selected from a fuel burnup database to simulate anycombination of eight (8), xenon-free fuel elements.*Similarly, it can include a mixture of four (4) independently depleted BORAL shim controlblades -each with a different axial and radial boron depletion profile based on its operationalhistory (or core residence time).*It has the ability to account for poison and gas buildup, and the reduction of beryllium atomdensity 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 accuratelyin order to reduce the error in the FCP calculation.*With the help of a critical control blade height search routine, starting from an initial estimateof the critical control blade height, a series of MCNP5 criticality (KCODE) calculations canbe 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 ofxenon-i135 and other poisons in the core during reactor operation as well as the buildup anddecay of the poisons during shutdown and restart using the isotope buildup and decay/losscapability of MONTEBURNS.The system of codes and calculation methodology described previously has been benchmarkedextensively using actual weekly core refueling and reactor startup data. The response to Question2.a, which was included in the responses, dated July 31, 2015, to a Request for AdditionalInformation 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 MissouriResearch Reactor (MURR) Core from Highly-Enriched to Low-Enriched Uranium -CoreNeutron 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 thecorresponding measurements for the ECP (or excess reactivity) for a known critical controlblade configuration at zero power, no xenon condition, or justify why this information is notneeded.The code system that is currently used for reactor physics analysis at MURR has been benchmarkedextensively. One of the methods used for the benchmarking was by comparing the EstimatedCritical Position (ECP) calculations from the detailed MCNP MUJRR model against the actualstartup critical control blade height data from several weekly reactor startups. The detailed MCNPMURR core model includes depleted control blade data, beryllium aging effect (i.e., more and moregas molecules taking up the place of beryllium atoms with increasing run time), as well as detailedsample 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 themodel's ability to predict the ECP accurately under various core states (mixed burnup) and flux trapsample conditions. The comparison was performed over an eight (8) month period. Note that thereactor startups at MURR require an occasional "strainer" startup -where initial critical controlblade 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 DataCore Actual Critical Predicated Critical Peiae lxTaControl Blade Control BladeConfiguration Height (Inches) Height (Inches) I~f ConfigurationWeek of 1/28/2013 16.79 16.67 0.99993 StrainerWeek of 2/04/2013 16.52 16.27 0.99975 SamplesWeek of 4/29/2013 15.98 15.78 1.00017 SamplesWeek of 6/10/2013 15.44 15.42 0.99995 SamplesWeek of 8/05/2013 16.74 16.74 0.99985 StrainerWeek of 8/12/2013 15.71 15.61 0.99985 SamplesWeek of 8/19/2013 15.84 15.84 1.00016 SamplesWeek of 8/26/2013 15.64 15.69 1.00029 SamplesA negative bias of ~1.5% is seen in the predictions for the early benchmarks. After the additionalrefinements 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, andRegulating blades) for a given core configuration at a low power, no xenon condition, orjustify 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, thefour BORAL shim control blades (Shim-i, Shim-2, Shim-3 and Shim-4, also referred to ascontrol blades 'A,' 'B,' 'C' and 'D') are in various stages of burnup (core residence time) rangingfrom 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 andreplaced with another rebuilt offset mechanism and a different control blade with a different burnupstatus. This schedule satisfies the Technical Specification surveillance requirement of inspectingone (1) out of four (4) control blades every six (6) months so that every blade is inspected everytwo (2) years. In this way, a given control blade is cycled in and out of the reactor multiple timesfrom 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 inchesof the control blade tip undergoes significant boron depletion with operation. Only the controlblade tip experiences burnup since during steady-state, full power operation the control blades arealmost fully withdrawn, resulting in the active neutron absorbing region of the blades being out ofany significant neutron flux. Since accurate control blade worth information is crucial for reactoroperation, every six (6) months when a control blade is replaced, a blade worth measurement of theinstalled control blade is performed.20 of 86 Using the detailed MCNP core model of MUIRR, the differential and integral worth of the four shimcontrol blades, and that of the stainless steel regulating blade, were calculated and the results areshown in Figures 3 through 6. The calculations were performed for fresh (non-depleted) controlblades using a fresh core with no xenon. In order to show the effect of control blade depletion withoperational history, the differential and integral worth curves of a single blade with a core residencetime of over 9 years are also shown in Figures 7 and 8, respectively.0.004y l E--07' -7E-0t,', 8E*05x'+ 2E-05X- 7E-06V = 9E-08,'- 5E-6'.0x 7E-05,' -6E-0OX + 3E-050.035 7E-Og, -4E-O6,,3/4+4E-O~x' ,O002O + 1E-O5% y = 1E-O7,'- 5E-O6x; 7E-05,,'. 1E-O4x- 2E-050.0025// / CN CpAA N-~shh0.0015* MCNP hp/ANt All-fresh ShimAS0.0015 /' 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-rhSimA0.0005 --Poly. (MCNP hp/AN A4ll-fresh Shim O '---PI I(CN h/ll All-fresh Shim O* ~/ .PoIy. (MCNP hp""l-fehSi0510 15 20 25 30Control Blade Height Withdrawn (inches)Figure 3 -Calculated Differential Worth Curves for All-Fresh Shim Control Bladesin an All-Fresh Fuel Core ConfigurationShim 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 onthe west side of the core, adjacent to these control blades.21 of 86 0.060000MCNP Integral Rod Worth Shim AU 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-130.020000 = 9E-07Xa + 1E-05X3+ 9EtDSx3/4 ZE-13X+r 2E-130.000000 -,,"1 -05 10 15 20 25 30Control Blade Height Withdrawn (inches)Figure 4 -Calculated Integral Worth Curves for Ali-Fresh Shim Control bladesin an Ali-Fresh Fuel Core Configuration0.000250/4' MCNP Ag/AH \a" -.-Poly. (MCNP 0.000150 /~/0o.0o01oo /\/-..../ y = -8,7729E-1D0'- 2,9792E-8x3 + 1,0668E-06xz + 7.8270E-O6x/ Rz = 9.9953E-010.0o00000 ')0 5 10 15 20 25 30Regulating Blade Height Withdrawn (inches)Figure 5 -Calculated Differential Worth Curve for a Fresh Regulating Bladein an Ali-Fresh Fuel Core Configuration22 of 86 O.0000MOWP 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. 25.1405 10 15 20 25 30Regulating Blade Height Withdrawn (inches)Figure 6 -Calculated Integral Worth Curve for a Fresh Regulating Bladein an All-Fresh Fuel Core Configuration0.0030.025', Y -4.2163E-O~x' + 3.6065E-06x3 -1.O055E-O4x7 + 9.1731E-04iq 0.02 I: 9.9715E-01* I0.0015 IQ MCNP A/AHI=/ --- Poly. (MCNP h4p/AH) \s0.000:15 ""20 25-Control Blade Height withdrawn (inches)-0).0005Control 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 0O0000.050000.04000MCNP control blade Id: 6-.05 worth after 9.0 yearcurrently in position B.= 0.02000io/* y = -8E-Ogx5 + 9E-O7x4 -3E-O5x3 + O.O005x' + tE-t3x + 2E-130.01000 A)0 5 10 15 20 25 30Regulating 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, thereactivity 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 physicscalculations (such as ECP predictions, reactor core shutdown margin, estimation of unknownsample reactivity worths, etc.). To serve as a benchmark for the calculated control blade worths, asingle blade was selected. Using a detailed mixed-core, mixed-bumup control blade model of thereactor configuration during the last Shim-4 (control blade 'D') inspection and replacement theblade 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 areshown in Figures 9 and 10.24 of 86 0.003u0.0025S0.002L.* 0.0015*n 0.001-0000500.-).0005Control Blade Height Withdrawn (inches)Figure 9 -Comparison of Measured and Calculated Differential Worth Curvesfor Shim Control Blade 'D'0.0600U, 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 )S0.03000005 01 2 53Coto ld Hih ihran (n hsFigure 10 -Comparison of Measured and Calculated Integral Worth Curvesfor Shim Control Blade 'D'25 of 86

d. Provide a calculated and measured temperature coefficient for a given core configuration ata 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) ofthe 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 computerprograms that were described in the response to Question 3.a. The results are provided in Table 2below.Table 2 -MURR Primary Coolant Temperature CoefficientMURR Technical ANL-MCNP CalculationCoefficient Specification Limit (294 to 400 K)All Fresh Core, BOC:-13.2 x 10 2.3 x 10-7Ak/k/°FAverage Core Temperature MxdCr.BCPrimary Coolant Coefficient Shall be More -12.8 x 10.5 + 2.2 x 10.7 Ak/k/°FTemperature Negative Than:(Isothermal) Mixed Core, Eqi. Xe:-6.0x 15Ak//0F-11.8 x 10s +/-- 2.2 x 107 Ak/k/°FMixed Core, BOC (ENDF7):-12.5 x 10.5 + 2.2 x 10-7 Ak/k/°FNote: BOC = Beginning of Cycle.26 of 86
4. NUREG-1537, Section 4.5.3, "Operating Limits, "provides guidance that licensees demonstratethat their facility has sufficient control blade worth to achieve the required shutdown reactivityassuming that all scrammable control blades are released upon scram, but the most reactive bladeremains in its most reactive position. The NRC staff could not find this information in the MURRSAR, but noted a reference in the 1971 Low Power Testing Program that indicated that theshutdown margin control blade reactivity was determined using 66 percent of the 4 shim bladeinsertion worth. Explain how MURR ensures adequate SDM, whether and if so, how the 66 percentfactor from the 1971 Low Power Testing Program is used, or justify why this information is notneeded.Table 4-12, on Pages 4-41 and 4-42 of the SAR, contains the value for the reactor core shutdownmargin. The Table lists the maximum K~ff with the highest worth shim control blade fullywithdrawn, or stuck, as 0.93 8. This maximum K~ff, value translates to a minimum reactor shutdownmargin value of -0.066 Ak/k. This compares with the Technical Specification minimum reactorcore 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 theresponse to Question 3.a, is -0.0875 Ak/k for an all-fresh fuel core case [Note: license possessionlimit 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 licenseeto characterize the dose for the maximally exposed individual, at the location of the nearestpermanent residence, and at any locations of special interest in the unrestricted area.a. The MURR SAR, Appendix B, contains summary information regarding the radiologicalimpacts of the MURR generated release of Argon 41 (Ar-41) during normal operations. TheMURR methodology includes an equation on SAR page B-J O that is used to alter the effectivestack height used in the dose calculations to compensate for elevation changes of the receptordue 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 theDavidson empirical model which has limited supporting data. Describe how the effectivestack height calculations are performed for the unique topography surrounding MURR, andhow the results are sufficiently conservative for the estimation of dose, or justify why noadditional information is needed.MURR calculates effective stack height, for the purposes of determining dose from radionuclideemissions, as the difference in vertical elevation between the point of emission at the end of theMURR exhaust stack and the receptor height at the point of interest, plus the effective stack heightcalculated using the Davidson equation. This equation takes into account the stack diameter andexhaust velocity of the gases leaving MURR to calculate an injection height and thus an effectivestack height into the atmosphere. Wind speed is also an input parameter into this formula as it is afunction of the particular Pasquill atmospheric stability class that is being modeled for the generalwind 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 ". .. oftengreatly underestimates observed rises,..." this underestimation would cause the offsite dosecalculations using the Pasquill-Guifford model to overestimate doses to the individual at the pointof dose calculation interest. In fact, dose estimates generated using this model are not out of linewith doses calculated using the COMPLY2 computer code which is used to determine annual doses(demonstrate compliance) to the nearest resident from MURR as part of the facility's annualNational Emission Standards for Hazardous Air Pollutants (NESHAPS) compliance report.Additionally, using Briggs' own equations for calculation of effective stack heights from the samereference book "Plume Rise," confirms that while the Davidson model underestimates effectivestack heights, these underestimated effective stack heights lead to an overestimation of dose, thusproviding a conservative approach to the offsite dose calculations. Thus, we feel that no additionalinformation is required.

References:

'Briggs, G.A., Plume Rise, AEC Critical Review Series, U.S. Atomic Energy Commission,Division of Technical Information, 1969.2COMLY is a computerized screening tool for evaluating radiation exposure from atmosphericreleases 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 andSubpart I.28 of 86

b. SAR page B-il has an equation for X/Q that includes the cy and arz dispersion factors. TheNRC 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 andare now included in the corrected Tables B-2 and B-3 below.TABLE B-2MAXIMUM ANNUAL INDIVIDUAL DOSE AT 150 METERSLocation: 150 Meters Directly NorthElevation at Man Height: 636 Feet (194 Meters) ______ ____Eff. Height ay X %s Dose with %sClas in I in (n) se/r3) (tiCi/ml or Comb. (mremly)Class (m)___ (m)_ (m)_ (sec/m3)_ Gi/m3) _________A 35 33 23 6.27E-05 3.14E-09 2.40E-04 0.00B 27 23 15 6.09E-05 3.04E-09 5.10E-03 0.08C 23 17 11 4.55E-05 2.28E-09 1.70E-02 0.19D 20 12 7 1.14E-05 5.71E-10 6.30E-02 0.18E 23 8.5 5 4.76E-08 2.38E-12 3.10E-02 0.00F 30 6 3.2 5.24E-22 2.62E-26 1.50E-02 0.00Total 0.46TABLE B-3MAXIMUM ANNUAL INDIVIDUAL DOSE AT 760 METERSLocation: 760 Meters Directly NorthElevation at ManHeight:_700 Feet (213 Meters)____________Eff. Height ay Oz x/Q X %s Dose with %sClass (in) I(in) (mn) (sec/rn3) i/mi or Comb. (inrem/y)____I 1 _____ __________ Ci/m3) _________A 16 170 270 3.30E-06 1.65E-10 2.40E-04 0.00B 8 120 85 1.04E-05 5.18E-10 5.10E-03 0.01C 4 85 52 1.71E-05 8.55E- 10 1.70E-02 0.07D 1 55 26 3.97E-05 1.99E-09 6.30E-02 0.63E 4 42 18 1.03E-04 5.13E-09 3.10E-02 0.80F 11 30 12 2.23E-04 1.12E-08 1.50E-02 0.84Total 2.3529 of 86 Note: Tbe "%s Comb." column was added to Tables B-2 and B-3 to better aid in understanding thecalculation 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 facilitydesign features, safety limits, limiting safety system settings, and limiting conditions for operationhave been selected to ensure that no credible accident could lead to unacceptable radiologicalconsequences to people or the environment. The NRC staff review examined the analyses providedin 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 theLimiting Safety System Setting setpoint in TS 2.2 (12.S MW). The temperature feedbackcoefficient used is -7.0 x lO Ak/k rather than the TS S.3.a value of -6xlO-5 Ak/k. It is unclearwhat peaking factors are employed. SAR Figure 13.2 seems to indicate that the scram timeused is faster than the value in TS 3.2.c. The acceptability of the results is based uponwhether 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 safetyconclusions 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 10MW 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 andAcceptance Criteria, for the Insertion of Excess Reactivity accident, "The accident scenarioassumes 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 wasreanalyzed 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 SafetySystem Setting (LSSS) set point of 12.5 MW, since the rod run-in system will initiate a rod run-inat 11.5 MW (Technical Specification 3.2.f.1) and shutdown the reactor prior to reaching the LSSSscram set point of 125%.For the SAR analysis of the Insertion of Excess Reactivity accident, the temperature coefficientused was -6.0 x 10.5 Ak/k and not -7.0 x 10-5 Ak/k as stated above. Third paragraph on Page 13-17of the SAR lists the various reactivity coefficients assumed for the Insertion of Excess Reactivityaccident analysis.Details regarding the power peaking factors used were not provided in that section of the SAR. Thepower peaking values used were values obtained based on the destructive analysis of a MURR fuelelement. For the updated analysis, more up-to-date power peaking values, based on the detailedMCNP MURR core model, were used.For both the SAR analyses, as well as for the updated analysis presented here, the control bladeinsertion times are based on the current and relicensing Technical Specification 3.2.c requirementof insertion to the 20% withdrawn position in less than 0.7 seconds. So the insertion rate wascalculated 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 monthly31 of 86 control blade drop time verifications performed at MURR have always yielded insertion times of0.6 seconds or less (see response to RAI 1 .a).Similar to the SAR analysis, the Reactivity Transient Analysis program PARET (V7.5), maintainedand distributed by the Nuclear Engineering Division of Argonne National Laboratory (ANL) wasused. For the Insertion of Excess Reactivity accident analysis, two channels were modeled inPARET; a hot channel representing worst-case conditions inside the core and an average channelrepresenting the rest of the core experiencing "average" conditions. The axial power profiles usedfor this 2-channel PARET reactivity transient analysis are given in Table 1 below.Table 1 -Peaking Factors in the Hot and Average ChannelsHot Channel Average Channel2.046 1.0581.971 0.9202.145 1.0182.335 1.1322.497 1.2192.672 1.3072.835 1.3602.986 1.4113.105 1.4303.164 1.4373.169 1.4203.098 1.3832.953 1.3262.775 1.2432.542 1.1402.290 0.9892.069 0.8281.888 0.7011.703 0.6151.499 0.5301.277 0.4601.080 0.3860.904 0.3290.880 0.358As indicated earlier, the transient was started from an initial power level of 11.5 MW with corecoolant flow rate as well as core coolant inlet temperatures set at their LSSS values of 3,200 gpm32 of 86 and 155 'F, respectively. Also, pressurizer pressure was at 75 psia (LSSS value). Since theInsertion of Excess Reactivity transient was analyzed from a starting power level of 11.5 MW, therod run-in that would be initiated by the rod run-in system at 11.5 MW was bypassed and only thehigh power scram set point of 12.5 MW was modeled. Also, a delay of 150 milliseconds wasincorporated into the control blade scram model so that the control blades would only start to insert0.15 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. Asexpected, due to the higher starting core power level, much lower core coolant flow rate and muchhigher than normal core coolant inlet temperature conditions assumed for this updated analysis, thepeak power during the transient momentarily reaches approximately 37.4 MW compared to a valueof approximately 33.0 MW reported in the SAR analysis for the same 600 pcm step reactivityinsertion.40.0040035.0030.001o 20.00-POWER MW-'l'dad "C-I'f maxtCS350300l0S50* 03.000.1000.50 1.00 1.50 2.00Time (seconds)2.50Figure 1 -Reactor Power, Fuel and Cladding Temperatures vs. Timefor a Positive Reactivity Step Insertion of 0.006 Ak/kSeveral SPERT tests had shown that the reactor can withstand such short duration (fewmilliseconds) power burst without sustaining any fuel damage and only sustained operation at suchhigh power levels will lead to fuel damage. The peak fuel temperature reached during the Insertionof Excess Reactivity accident in the worst (Hot) channel is only 227.4 0C -well below the newSafety Limit of 530 0C for the aluminide fuel.33 of 86
b. Regarding Loss of Primary Coolant and Loss of Primary Coolant Flow -The initial power is11 MW rather than the LSSS setpoint in TS 2.2 (12.5 MW). It is unclear what peaking factorsare employed. The acceptability of the results is based upon the peak fuel temperatureattained rather than the safety limit identified in TS 2.1. Provide additional informationjustifying and supporting the analysis and the safety conclusions or provide a justification forwhy 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 poweroperation. 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 totaldecay heat factor.As stated on page 13-7 of NUREG-1537, Part 1, Format and Content, item 13.2(1), "State theinitial conditions of the reactor and equipment. Discuss relevant conditions depending on fuelburnup, experiments installed, core configurations, or other variables. Use the most limitingconditions in the analyses." On Pages 13-9 and 13-10 of NUJREG-1537, Part 2, Standard ReviewPlan and Acceptance Criteria, for both the LOCA and LOFA, "The scenario assumes that thereactor is operating at full licensed power and has been operating long enough for the fuel tocontain fission products at equilibrium concentrations." Therefore, we believe, the assumed 11MW steady-state power level was greater than the required assumed power level as stated inNUREG-1537, Part 2.The assumed peaking factors are graphed on Figure C.6 (Page C-l13) and listed in Table C-I (PageC- 14) of Appendix C of the SAR. The cold-leg break LOCA has the highest peak fuel temperatureof 311.7 0F (155.4 0C) occurring in fuel plate number-3 within the first second. The hot-leg breakLOCA peak fuel temperature is 281.2 0F (138.4 0C) also occurring within the first second. Thesepeak temperatures occur within the first second because the events start with a loss of normalforced circulation coolant flow with a slight delay in the reactor scram trip. The peak fueltemperature for a LOFA is 280.3 0F (137.9 °C), which occurs in fuel plate number-i, 0.3 secondsinto the transient.The LOFA and LOCA analyses were redone with the other three (3) Limiting Safety SystemSetting (LSSS) variables -core coolant inlet temperature, core coolant flow rate and pressurizerpressure -at their respective set points of 155 0F, 3,200 gpm and 75 psia. The peaking factorsused in the updated analyses are the ones described in the response to RAI 7.a (By letter dated July8, 2013, the NRC issued Amendment No. 36 to Facility Operating License No. R-103, whichrevised the MURR Safety Limits). The peak fuel temperature for the cold-leg break LOCA was413.9 0F (212.2 0C) whereas the peak fuel temperature for the LOFA was 292.3 0F (144.6 °C).Well below the new Safety Limit peak fuel temperature of 530 0C.MURR feels that having core coolant inlet temperature, core coolant flow rate and pressurizerpressure are their respective LSSS and reactor power at the full licensed limit meets the guidanceof NUREG-1537, Parts 1 and 2, and is sufficiently conservative.34 of 86
c. Regarding the maximum hypothetical accident (MHA) and Failed Fueled Experiment -theseevents use a 10 minute and 2 minute evacuation time respectively. Provide additionalinformation identifying the limiting evacuation time and then use that time to justify' andsupport the analysis and the safety conclusions or provide a justification for why suchinformation is not required.The basis for the maximum hypothetical accident (MHA) evacuation time is stated on Page 13-5 ofthe SAR: "It would take approximately 5 minutes for Operations personnel to secure the primarycoolant system (PCS) and verify that the containment building has been evacuated following acontainment building isolation. For the purpose of the MIIA calculations, a conservativeassumption of 10 minutes is used." This basis will not change; however, the MiHA has now beenrenamed the "Fuel Failure during Reactor Operation" accident since the dose consequences of afailed 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), theprimary coolant system (PCS) does not have to be secured. The only required action for Operationspersonnel is to verify that the containment building has been evacuated following a containmentbuilding isolation, which will occur during both of these accidents. MUJRR performs an evacuationdrill 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 thepurposes 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 thelicensee 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 themelting of four number 1 fuel plates in a core region where the power is at a maximum. The fuelfails submerged and it is assumed that all iodine, krypton, and xenon isotopes are released into theprimary 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 oftwelve 10O-day cycles over a 300-day period. These values were then adjusted using a peakingfactor 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 islisted as 3.676. Clarify the discrepancies in the peaking factors used, and provide a revisedcalculation of the source using the peaking factors determined from the final analysis, orjustify why no additional information is needed.Table 4-14, "SUMMARY OF MUIRR HOT CHANNEL FACTORS," in Section 4.5 of the MUJRRSAR, lists a hot spot power peaking factor of 3.6765 with no engineering factors included. Thisvalue applies the product of the radial, axial and azimuthal peaking factors of a fuel plate todetermine the hot spot on the plate. The SAR provides an overall peaking factor of 4.35; the hotspot power peaking factor of 3.6765 multiplied by the engineering factors. These two peakingfactor values apply to the potential worst-case maximum power density point in the core for theSafety 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 FluxPower-related FactorsNuclear Peaking FactorsRadial 2.220Non-Uniform Burnup 1.112Local (Circumferential) 1.040Axial 1.432Overall 3.676Engineering Hot Channels Factors on FluxFuel Content Variation 1.030Fuel Thickness / Width Variation 1.150Overall Product 4.35By 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 factorto 3.4747; with no engineering factors included. Including the engineering factors, the overallpeaking factor increases to 4.116.36 of 86 From Table F.4, "SUMMARY OF MURR HOT CHANNEL FACTORS," of Appendix F ofAddendum 4 to the MURR Hazards Summary Report (as revised by Amendment No. 36):On Heat Flux From Plate-iPower-related FactorsNuclear Peaking FactorsFuel Plate (Hot Plate Average) 2.215Azimuthal Within Plate 1.070Axial Peak 1.3805Additional Allowable Factor 1.062Overall 3.4747Engineering Hot Channels Factors on FluxFuel Content Variation 1.030Fuel Thickness / Width Variation 1.150Overall Product: 4.116This 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 anoverall peaking factor of 3.863 and an enthalpy rise of 74.8%; thus producing the most limitingcombination of heat flux and enthalpy rise.This overall peaking factor of 4.116 at mesh interval 14 would apply to the 1-inch square assumedin the analysis in the response to RAT A.27. Therefore, since the ratio is 1.372 (4.116 / 3.0), thecalculated dose rates in the response to RAI A.27 increased by approximately the 37.2%. The onlyother fuel plate exposed during handling is plate number-24, which has a lower overall peakingfactor than plate number-i. The assumed peaking factor in the response to RAT A.27 has beenrevised from 3.0 to 4.116, which increased the whole body (TEDE) "60-Minute Dose fromRadioiodine and Noble Gases in Containment" from 0.79 to 1.09 mrem. This change also requireda similar revision to the response to RAI A.6 regarding the revised Technical Specificationdefinition for Irradiated Fuel, Definition 1.11, by about the same percentage.The current MIURR maximum hypothetical accident (MHA) assumes the melting of fuel platenumber-i in four (4) different fuel elements. An unirradiated fuel plate number-i contains, onaverage, 19.26 grams of U-235, so four (4) unirradiated number-i fuel plates contain 77.04 gramsinstead of the 78.58 grams assumed in the MHA. These four (4) number-i fuel plates that meltcorrespond to 1.41% of the total U-235 in the Week 58 Core that was used to determine the highpower peaking factor for the revised SLs. The Week 58 Core has a total power history of 576MWd. This power history results in a total reduced core mass of 5,474 grams of U-235 due to theprevious fuel consumption. This 1.41% of U-235 melting releases 3.42% of the core fissionproducts 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 thecontainment building are less than that for a failed fueled experiment, which is now considered theMURR MHA.b. The release is assumed to occur into the PCS with a volume of 2,000 gallons. Identif whatcomponents comprise this volume and provide information to confirm the 2,000 gallonvolume 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 ofits individual, major components, including the piping, reactor core, pressure vessels, primarycoolant circulation pumps, heat exchangers, and pressurizer. Attachment 7 is a breakdown of themeasurements and calculated volumes, where design capacities are not available, of the individualcomponents following the RELAP Model component designations listed in Appendix C of theSAR. The total calculated volume of the PCS is 2,007 gallons. However, based on the difficulty ofmeasuring some of the in-pool PCS piping and components, this volume is conservativelyunderestimated by approximately 5 to 10%, thus radionuclide concentrations in the PCS areconservative.c. The release is assumed to remain in the PCS except for the amount that will enter the poolcooling system as part of the PCS to pool cooling system leakage. Therefore, theconcentration of iodine that is released first enters the pool cooling system and is dilutedonce again. This seems to reduce the consequences of this accident to a fraction of theconsequences 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 notbe the MHA. Provide a confirmation of the dilution assumptions stated above andclarification 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 reanalyzedusing consistent methodologies and assumptions. All three (3) new analyses are included in theseresponses. 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 mremFuel Handling Accident: 687.00 mremFueled Experiment Failure: 1212.44 mremBased on these analyses, the Fueled Experiment Failure accident has been determined to be the newMURR 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 betweenthe PCS and the pool cooling system. However, the NRC staff questions whether the release38 of 86 into the PCS will collect in the vent tanks and other places in the PCS and eventually bereleased to the environment after decay. Provide an explanation for this leakage path,including assumptions and calculations of the possibility of the isotopic concentrations beingreleased 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 buildingisolation system will occur as a result of the gaseous activity collecting in the vent tank system. Atthis point the containment building is isolated. As described in Section 9.13.3 of the SAR, the venttanks will vent through an absolute and charcoal filter if enough gases collect in the vent tanks tocause the water level in the tanks to recede to a point where level controller 925A will signal valveV552A to open and vent the gases. Note: The vent path for the vent tank system, after it goesthrough the absolute and charcoal filters, is to the pool sweep system which is connected to thecontainment building 16-inch hot exhaust line (see SAR Sections 6.2.3.8 and 9.1.2.2). Thecontainment 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 wouldnot cause the system to vent. However, if for some reason the system should vent prior to the PCSbeing secured as part of the actions of Operations personnel during an MHA (vent valves 552A and552B will not open when the PCS is secured), the gases will be vented into the isolated containmentstructure 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-termrecovery actions, which will be very well planned and organized.e. In determining the offsite doses in the unrestricted areas from the releases, theconcentrations of the released isotopes are calculated using a method described in the MURRSAR, Appendix B, which used a simplified joint frequency distribution of weather data thatwas prepared in the 1960s. Given the changes in weather conditions over the last 50 years, itis not clear to the NRC staff whether the listed probabilities and wind speeds for the stabilityclasses are still applicable. Provide available current weather data, and state whetherchanges warrant reconsideration of the cited data, or justify why no additional information isneeded.In reviewing the available meteorological data for the Columbia vicinity, newer meteorological datawas found from the Columbia Regional Airport. This facility has more current meteorologicalwind data available and this data was used to generate wind roses for updated time periods closer tothe current time frame. Based on the results of the meteorological data review, we believe that theprevious data submitted is representative of current wind rose data, in and around the Columbiaarea, as there appeared to be no substantial difference in wind speed and direction during theoriginal submittal utilizing nine (9) years of data from 1961 to 1969 and subsequent data whichincluded the above 9-year period and an additional 21 years of meteorological data for a total of 30years (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 10provides 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 listedconcentrations in the cited unrestricted location, which is also not specified. The calculationof the ratio of the average concentration in the unrestricted location to the correspondingconcentration in containment results in the reduction factor for iodine twice as large as thevalue for the noble gases. For example, for Krypton-85 the ratio is 7.5x10-14/3.0x10-s or areduction of about 4.0x10s For 1-131, the ratio is 1.36 x10-14/1.1x10-s or a reduction ofabout 8.1 x104. Provide an explanation of all assumptions relating to the calculation ofaverage isotope concentrations, specify~ all locations where these concentrations aredetermined, and explain how dispersion factors are determined and used, or justify why noadditional 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 takingthe initial concentration of I-131 in the containment building of 4.4 x 10-°8 (Page 13.6 of theSAR) and multiplying it by 0.25 (plating reduction factor) and dividing into the final offsiteconcentration 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 x10-°8) =1.24 x 10-°6. The initial concentration of Kr-85 was used in this case as there is no platingor other phenomena that would hold up the noble gases.g. hn determining occupational doses, it appears that the MURR SAR calculations use acombination of dose conversion factors (DCFs). It appears that for radioiodine, thecalculation 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 submersiondoses 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 FGR11. 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 FGR11 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 thoseisotopes 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 DCFsresult 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 uselimits from either: 10 CFR 20, Appendix B or 10 CFR 835, Appendix C (Attachment 11). Whereavailable we use Derived Air Concentration (DAC) and Effluent Concentrations from 10 CFR 20Appendix 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 (T112 < 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 usedthe 10 CFR 835 Appendix C default DAC value of 6.0 x 10.06 as noted at the end ofAppendix C. From this default DAC we estimate the applicable effluent concentration limit basedon the description provided in the Table 2, "Effluent Concentrations," footnotes to Appendix B in10 CFR 20. Thus, all dose calculations now use limits based on the background and methodologyprovided in ICRP 26 and 30.40 of 86 Revised "Fuel Failure during Reactor Operation"(Formerly the MIIA)13.2.1 Fuel Failure during Reactor Operation13.2.1.1 Accident-Jnitiating Events and ScenariosMany types of accidents have been considered in conjunction with the operation of the MUJRR. Inall cases, safety systems have been designed such that the likelihood of an accident involving therelease of a significant amount of fission products has essentially been eliminated. The safetysystems take the form of automatic reactor shutdown circuits and process systems designed toensure, through redundancy, that the reactor will shut down upon a significant deviation fromnormal operating conditions. In addition, the reactor is housed within a containment building, thusproviding 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 anaccident condition has caused the melting of the number-i fuel plate in four (4) separate fuelelements (Ref. 13.11). It is further assumed that the four (4) number- 1 fuel plates are in the peakpower 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, andthe pre-operational inspection of the reactor pressure vessels and core region following any fuelhandling evolution, all prevent an accident of this type from occurring. In addition, it has beenshown that a 75% blockage of coolant flow to the hot channel is insufficient to cause claddingfailure (Ref. 13.2).13.2.1.2 Accident Analysis and ConsequencesThe fuel failure accident postulates partial fuel melting with an associated release of fissionproducts into the primary coolant system. The accident is assumed to occur with the primarycoolant system operating, resulting in a quick dispersal of the fission products throughout thesystem. With the design of the primary coolant system and its associated systems, particulateactivity will remain in the coolant, and the gaseous activity that comes out of solution will collect inthe reactor loop vent system and be retained there. Therefore, the primary coolant system reliefvalves and pressurizer are the only paths for a release of significant quantities of fission products tothe environment.The potential energy release from the melting of four (4) number-i fuel plates could occur as apossible metal-water reaction (Ref. 13.3). While hydrogen would be formed, it is highly unlikelythat in a water environment a hydrogen deflagration reaction would occur. The amount of materialwhich would be involved in a metal-water reaction under the conditions of four (4) number-i fuelplates melting is not predictable as the amount is dependent upon many conditions. For purposes ofcalculation, it is conservatively assumed that all the fuel plate aluminum cladding exposed in the41 of 86 area is involved in the reaction. The reactor core contains a total of 33.56 Kg of aluminum. Ofthis, 1.3% or 436 grams is assumed to react according to the following equation:A1 + n2-I* A1On +nI-I2 +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 103 BTU.This amount of heat would easily be transferred to the adjacent fuel elements and primary coolantin the core. Additionally, any steam that would form in the vicinity of the molten area would alsoassist in dissipating the heat. Since the fuel failure would result in a negligible release of energy tothe primary coolant system, the introduction of pressure surges, which could lift the primary reliefvalves, are not considered credible. The pressurizer is an isolated system, and since no significantpressure 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 reactorscram and actuation of the containment building isolation system by action of the pool surfaceradiation monitor. Additionally, following actuation of the anti-siphon system when the primarycoolant system is secured, gases could also collect in the anti-siphon pressure tank. The location ofthese tanks under the pool surface, and the shielding provided by the water and the biologicalshield, 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 coolantcleanup system. This cleanup procedure would be undertaken under closely monitored andcontrolled conditions.The primary coolant system does experience some coolant leakage into the. reactor pool through thepressure vessel head packing and flange gasket. This leakage is typically less than 40 gallons (1511) per week; an almost imperceptible leakage rate of approximately 4 x 1 0- gallons of primarycoolant 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 exposureto 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 6.2Kg of 235U (normal operating cycle is 6.5 days with a total of less than 700 MVWD on the core), thefollowing 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 Core1311 -- 1.7 x 10+0 Ci 85Kr -4.7 x 10+02 Ci '3Xe -4.2 x 10+°5 Ci132I -3.3 x 10+05 Ci 85m~r -1.1 x 10+05 Ci '35Xe -9.6 x 10+04 Ci133I -- 5.1 X 10o0 Ci 87Kr -2.1 x 10+05 Ci 135mXe -9.4 x 10+04 Ci14-- 6.3 x 10+° Ci 88Kr -3.0 x 10+05 Ci 137Xe -- 4.9 x 10+0 Ci13SI -5.2 x 10+05 Ci 89Kr -3.8 x 10+° Ci '38Xe -5.2 x 10+05 Ci9°Kr -3.8 x 10+05 Ci 139Xe -- 4.2 x 10"°5 CiAn 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 thefuel 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 revisedSLs. The Week 58 Core has a total power history of 576 MWd. This power history results in atotal 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 fuelplate 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 thefuel is assumed in calculating the fission product inventory in the primary coolant system. It is alsoassumed that fission products released into the primary coolant are quickly and uniformly dispersedwithin the 2,000-gallon (7,571-1) primary coolant system volume and, during a normal week'soperation, 80 gallons (7.9 x 1 0- gpm) of coolant leaks from the primary coolant system into thepool water. Therefore, the radioactivity released into the reactor pool in 10 minutes -determinedto be the maximum personnel occupancy time in the containment building after the accident fornecessary operational personnel -is as follows:(Note: It would take approximately 5 minutes for Operations personnel to secure the primarycoolant system and verify that the containment building has been evacuated following acontainment building isolation. For the purpose of the fuel failure calculations, a conservativeassumption of 10 minutes is used.)Example calculation of 131I released into the reactor pool:=131I 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 1iCiNote: 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 Minutes3I-- 2.30 x 10+05 1.tCi 85Kr -6.35 x 10+02 1iCi 133Xe -5.67 x 10+05 13I -4.46 x 10+/-05 85m~r -1.49 x 10+°05 .Ci 135~ -1.30 x 10+05 1331 -- 6.89 x 10+05 87Kr -2.84 x 10+05 1iCi l35rage -1.27 x 10+0 ptCi1341 -- 8.52 x 10+05 j.tCi 88Kr -4.04 x 10+05 gCi 137Xe -6.63 x 10+° 13I- 7.02 x 10+05 jtCi 89Kr -5.13 x 10+05 xiCi 138Xe -7.02 x 10+05 9°Kr -5.13 x 10+05 gCi 139Xe -5.67 x 10+05 1iCiFission products released into the reactor pooi will be detected by the pool surface and ventilationsystem exhaust plenum radiation monitors. However, for the purposes of this analysis, it isassumed that a reactor scram and actuation of the containment building isolation system occurs byaction of the pooi surface radiation monitor.The radioiodine released into the reactor pool over a 10-minute interval is conservatively assumedto be instantly and uniformly mixed into the 20,000 gallons (75,708 1) of bulk pool water, whichthen results in the following pool water concentrations for the radioiodine isotopes. The watersolubility of the krypton and xenon noble gases released into the pool over this same time periodare ignored and they are assumed to pass immediately through the pool water and evolve directlyinto the containment building air volume where they instantaneously form a uniform concentrationin the isolated structure.Radioiodine Concentrations in the Pool Water131 -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/ga1When the reactor is at 10 MW and the containment building ventilation system is in operation, theevaporation rate from the reactor pool is approximately 80 gallons (302.8 L) of water per day. Forthe purposes of this calculation, it is assumed that a total of 40 gallons (151 L) of pool watercontaining the previously listed radioiodine concentrations evaporates into the containment buildingover the 10 minute period. Containment air with a temperature of 75 0F (23.9 °C) and 100%relative humidity contains H20 vapor equal to 40 gallons (151.4 L) of water. Since the air incontainment 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 causethe containment air to be supersaturated. It is also conservatively assumed that all of theradioiodine activity in the 40 gallons (151 L) of pool water instantaneously forms a uniformconcentration in the containment building air. When distributed into the containment building, thiswould result in the following radioiodine concentrations in the 225,000 ft3 (6,371.3 in3) air volume:Example calculation of 1311 released into containment air:-131I concentration in pool water x 40 gal x 1/225,000 ft3 x 35.3 147 ft3/m3= 1.15 x 10+01 pCi/gal x (6.28 x 10-03 gal/mn3)44 of 86

-7.22 x 10-o2 ptCi/mn3(7.22 x 10-° /.tCi/m3) x (1 m3/106 ml) =7.22 x 10.08 pCi/mlNote: Same calculation is used for the other isotopes listed below.The average radioiodine concentrations are the sum of the initial concentrations and theconcentrations after 10 minutes decay divided by 2.Average Radioiodine Concentrations in the Containment Building Air during the 10 Minutes1311 -- 7.22 x 10-°8 pCi/ml132I -- 1.36 x 10-07 pCi/ml133I -- 2.16 x 10-07 gCi/ml134I -- 2.53 x 10-07 pxCi/ml135I -2.18 x 10-° pxCi/mlAs noted previously, the krypton and xenon noble gases released into the reactor pool from theprimary coolant system during the assumed 1 0-minute interval following the fuel failure (Note: theprimary coolant system is shut down and secured, and the leakage driving force is stopped within10 minutes), are assumed to pass immediately through the pool water and enter the containmentbuilding air volume where they instantaneously form a uniform concentration in the isolatedstructure. Based on the 225,000-ft3 volume of containment building air and the previously listedCurie quantities of these gases released into the reactor pool, the maximum noble gasconcentrations 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 ft3 x 35.3 147 ft3/m3-6.35 x 10+02 uCi x (1.60 x 10-o4 1/in3)-9.96 x 10.02 gxCi/m3(9.96 x 10.02 3) x (1 m3/106 ml) = 9.96 x i0.0 ptCi/mlNote: Same calculation is used for the other isotopes listed below.The average noble gas concentrations are the sum of the initial concentrations and theconcentrations after 10 minutes decay divided by 2.Average Noble Gas Concentrations in the Containment Building Air during the 10 MinutesKr -9.96 x 10-°8 jiCi/ml85m~r -2.30 x 10-°5 iCi/ml87Kr -4.27 x 10°5 pCi/ml88r- 6.22 x 10°5 ptCi/ml89r- 4.47 x 10"°5 p.Ci/ml9°r- 4.03 x 10"°5 pCi/ml133Xe -l3SlXe _'38Xe -'39Xe -8.90 x 10-° pCi/ml2.02 x 10-° gtCi/ml1.63 x 10-°5 gCi/ml6.05 x 10.05 jgCi/ml8.88 x 10.05 pCi/ml4.45 x 10.05 pCi/ml45 of 86 The objective of this calculation is to present a worst-case dose assessment for an individual whoremains in the containment building for 10 minutes following fuel failure. Therefore, as notedpreviously, the radioactivity in the evaporated pool water is assumed to be instantaneously anduniformly 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 thyroidfrom radioiodine and the dose to the whole body resulting from submersion in the airborne noblegases and radioiodine inside the containment building. As previously noted, the exposure time forthis dose assessment is 10 minutes.Because the airborne radioiodine source is composed of five (5) different iodine isotopes, it will benecessary to determine the dose contribution from each individual isotope and to then sum theresults. 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 25mrem/DAC-hr. Additionally, 10 minutes of one DAC-br is 1.67 x 10-01 DAC-br.1311 concentration in containment1311 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)= 3.61Therefore, 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+°lmremNote: Same calculation is used for the other radioiodines listed below.Derived Air Concentration Values and 10-Minute Exposures -RadioiodineRadionuclide13111321133I13411351Derived Air Concentration2.00 x 10-°8 1Ci/ml3.00 x 10-06 gtCi/ml1.00 x 1 007 ptCi/ml2.00 x 10-0 jiCi/ml7.00 x 10.o iiCi/ml10-Minute Exposure1.51 x 10401 mrem1.89 x 10-°' mrem8.99 x 10+°° mrem5.27 x 10.02 mnrem1.30 x 10+° mremTotal =25.58 mrem46 of 86 Doses from the kryptons and xenons present in the containment building are assessed in much thesame manner as the radioiodines, and the dose contribution from each individual radionuclide mustbe calculated and then added together to arrive at the final noble gas dose. Because the dose fromthe noble gases is only an external dose due to submersion, and because the DACs for theseradionuclides are based on this type of exposure, the individual noble gas doses for 10 minutes incontainment were based on their average concentration in the containment air and thecorresponding DAC.Example calculation of whole body dose due to KrThe DAC can also be defined as 5,000 mrenm/2,000 hrs, or 2.5 mnremiDAC-hr. Additionally, 10minutes of one DAC-hr is 1.67 x 1001 DAC-hr.85Kr concentration in containment = 9.96 x 1 0.0 85Kr DAC (10 CFR 20) = 1.00 x i0"° pCi/mlDose Multiplication Factor = (85Kr concentration) / (SSKr DAC)= (9.96 x 10.08 pCi/mI) / (1.00 x 10o4~ pCi/mi)= 0.001Therefore, 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 xl10°4mremNote: 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 GasesRadionuclide85Kr85m~r87Kr88Kr89Kr90Kr133Xe135Xel35rage137Xe138Xe139XeDerived Air Concentration1.00 x 10"° xtCi/ml2.00 x 1005 iiCi/ml5.00 x 10-°6 itCi/ml2.00 x 100o6 p.Ci/ml6.00 x 100o6 gCi/ml6.00 x 10.o6 iCi/ml1.00 x 10"04 pCi/ml1.00 x 10-05 jCi/ml9.00 x 10-06 xiCi/ml6.00 x 10.06 gCi/ml4.00 X 10-06 pCi/mi6.00 x 10°6 jiCi/ml10-Minute Exposure4.51 x 10-°4 mrem4.80 x 10-°" mrem3.56 x 10400 mrem1.30 x 10+01 mrem3.11 x 10400 mrem2.80 x 10+00°mrem3.71 x 10-°1 mrem8.43 x 10-°1 mrem7.54 x l0-°' mrem4.20 x 10+° mrem9.25 x 1040° mrem3.09 x 10+00 mremTotal = 41.42 mrenlTo finalize the occupational dose in terms of Total Effective Dose Equivalent (TEDE) for a1 0-minute exposure mn the containment building after target failure, the doses from the radioiodinesand noble gases must be added together, and result in the following values:10-Minute Dose from Radioidines and Noble Gases in the Containment BuildingCommitted Dose Equivalent (Thyroid):Committed Effective Dose Equivalent (Thyroid):Committed Effective Dose Equivalent (Noble Gases):Total Effective Dose Equivalent (Whole Body):25.58 mrem0.77 mrem41.42 mrem42.18 mremBy comparison of the maximum TEDE and Committed Dose Equivalent (CDE) for thoseoccupationally-exposed during fuel failure to applicable NRC dose limits in 10 CFR 20, the finalvalues are shown to be well within the published regulatory limits and, in fact, lower than 1% ofany occupational limit.Radiation shine through the containment structure was also evaluated when considering accidentconditions and dose consequences to the public and MUJRR staff. Calculation of exposure ratefrom the target failure was performed using the computer program MicroShield 8.02 with aRectangular Volume -External Dose Point geometry for the representation of the containmentstructure (Attachment 12). MicroShield 8.02 is a product of Grove Software and is acomprehensive photon/gamma ray shielding and dose assessment program that is widely used byindustry for designing radiation shields.The exposure rate values provided below represents the radiation fields at 1 foot (30.5 cm) from a12-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 theexposure rate values are identical to those used for determining the dose to a worker withincontainment from noble gases. For radioiodine, the total iodine activity of the target was used forthe dose calculations, not the amount that evaporated in 10 minutes. The source term also assumesa homogenous mixture of nuclides within the containment free volume.Radiation Shine through the Containment BuildingExpo sure Rate at 1-Foot from Containment Building Wall: 1.074 mrem/hrExposure Rate at Emergency Planning Zone Boundary (150 meters): 0.007 mrem/hrA confirmatory analysis of the accident condition yielding the largest consequence was validatedindependently by the use of the MCNIP code. This analysis yielded a result 21% less than theMicroshield method and results provided above.As noted earlier in this analysis, the containment building ventilation system will shut down and thebuilding itself will be isolated from the surrounding areas. Fuel failure will not cause an increase inpressure inside the reactor containment structure; therefore, any air leakage from the building willoccur as a result of normal changes in atmospheric pressure and pressure equilibrium between theinside of the containment structure and the outside atmosphere. It is highly probable that there willbe no pressure differential between the inside of the containment building and the outsideatmosphere, and consequently there will be no air leakage from the building and no radiation doseto members of the public in the unrestricted area. However, to develop what would clearly be aworst-case scenario, this analysis assumes that a barometric pressure change had occurred inconjunction with the target failure. A reasonable assumption would be a pressure change on theorder of 0.7 inches of Hig (25.4 mm of Hig at 60°C), which would then create a pressure differentialof about 0.33 psig (2.28 kPa above atmosphere) between the inside of the isolated containmentbuilding and the inside of the adjacent laboratory building, which surrounds most of thecontainment structure. Making the conservative assumption that the containment building will leakat the TS leakage rate limit [10% of the contained volume over a 24-hour period from an initialoverpressure of 2.0 psig (13.8 kPa above atmosphere)], the air leakage from the containmentstructure in standard cubic feet per minute (scfmn) as a function of containment pressure can beexpressed by the following equation:LR = 17.85 x (CP-14.7)112;where:LR = leakage rate from containment (scfmn); andCP = containment pressure (psia).The minimum Technical Specification free volume of the containment building is 225,000-ft3 atstandard temperature and pressure. At an initial overpressure of 2.0 psig (13.8 kPa aboveatmosphere), 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 air49 of 86 volume to 230,051 scf. The above equation describes the leakage rate that results in this drop ofcontained 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 initialoverpressure condition of 0.33 psig (2.28 kPa above atmosphere), it would take approximately 16.5hours 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 periodwould be about 5.2 scfln.Several factors exist that will mitigate the radiological impact of any air leakage from thecontainment building following target failure. First of all, most leakage pathways fromcontainment discharge into the reactor laboratory building, which surrounds the containmentstructure. Since the laboratory building ventilation system continues to operate during targetfailure, leakage air captured by the ventilation exhaust system is mixed with other building air, andthen 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 bydischarge out the exhaust stack and subsequent atmospheric dispersion, results in extremely lowradionuclide concentrations and very small radiation doses in the unrestricted area. A tabulation ofthese concentrations and doses is given below. These values were calculated following the samemethodology 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 tothe behavior of radioiodine, which has been studied extensively in the containment mockup facilityat 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% iodinedeposition in the containment building, each cubic meter of air released from containment has aradioiodine concentration that is 25% of each cubic meter within containment building air, then theradioiodine concentrations leaking from the containment structure into the laboratory building, inmicrocuries per milliliter, will be:Example calculation of 1311 released through the exhaust stack:-131I activity / (30,500 ft3/min x 16.5 hr x 60 min/hr x 28,300 ml/ft3)= 2.30 x 10+05 /8.55 x 10+'11ml-2.69 X 10-07 iiCi/ml(2.69 x 10-07 iiCi/ml) x (0.25) = 6.73 x 10.08 iiCi/mlNote: Same calculation is used for the other radioiodines listed below.Radioiodine Concentrations in Air Leaking from Containment1311 -- 6.73 x 10-°8 jtCi/ml 133I -- 2.02 x 10-07 gtCi/ml 13I- 2.05 x 10-07 jiCi/ml1321 -1.30 x 10-07 pCi/ml 34 -2.49 x 10-07 pCi/ml50 of 86 Example calculation of 85Kr released through the exhaust stack:= 85}~. activity / (30,500 ft3/min x 16.5 hr x 60 mmn/hr x 28,300 mi/fl3)= 6.35 x 10+02 / 8.55 x 10+11 ml= 7.4 x1-w° ItCi/mlNote: Same calculation is used for the other noble gases listed below.Noble Gas Concentrations in Air Leaking from Containment and Exiting the Exhaust Stack85Kr -7.43 x 10-1 pCi/ml 87Kr -3.32 x 10-0 jtCi/ml 89Kr -6.00 x 10-0 giCi/ml85m~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/ml33e- 6.64 x 10-07 pCi/mil 3smXe -- 1.49 x 10-07 ptCi/ml 138Xe -8.22 x 10-07 pCi/ml135Xe -1.52 x 10-0 pCi/ml 137Xe -7.76 x 10-07 pCi/mI '39Xe -6.64 x 10-0 ptCi/mlAssuming, as stated earlier, that (1) the average leakage rate from the containment building is 5.2scfrn, (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 buildingpressure with atmospheric pressure, (3) the flow rate through the facility's ventilation exhaust stackis 30,500 scfrn, (4) the reduction in concentration from the point of discharge at the exhauststack to the point of maximum concentration in the unrestricted area is a factor of 292 and (5) thereis no decay of any radioiodines or noble gases, then the following concentrations of radioiodinesand noble gases with their corresponding radiation doses will occur in the unrestricted area. Thevalues listed are for the point of maximum concentration in the unrestricted area assuming auniform, semi-spherical cloud geometry for noble gas submersion and further assuming that themost conservative (worst-case) meteorological conditions exist for the entire 16.5-hour period ofcontainment 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 thatwas used to determine doses inside the containment building, and it was assumed that an individualwas 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 containmentbuilding was leaking.A worst-case scenario dilution factor of 292 for effluent dilution using the Pasquill-Guifford Modelfor atmospheric dilution is used in this analysis. We assume that all offsite (public) dose occursunder these atmospheric conditions at the site of interest, i.e. 760 meters North of MUIRR. In ourcase at 760 meters it occurs only during Stability Class F conditions; which normally only occur11.4% 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. AnEffluent Concentration Limit of 2.0 x 10.08 itCi/ml is used for the "unlisted" isotopes, which equalsthe DAC/300 when using the DOE Part 835 Default DAC limits of 6.0 x 10-06 p.C/ml. Exposure at1 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 8760hours per year. Thus, the time assumed to be exposed to the effluent concentration limit is a factorof 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 based51 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 'sin Appendix C of Part 835. No credit is taken for transit time from the stack to the receptor pointnor is credit taken for decay inside containment until release. In the case of Kr-89 and Xe-i137 thetransit 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 theDAC value to establish the effluent concentration limit is warranted. This reduction factor of 300 isconsistent, in fact more conservative, than 10 CFR 20 Appendix B, as the N4RC calculates EffluentConcentration 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 mremihr1311 concentration131i effluent concentration limit1311 Conversion Factor= 2.30 x 10-1° pCi/ml=2.00 x 10-10 gCi/ml= 5.71 X 10-° mremihrTherefore, 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 i010-1 gCi/ml x 5.71 x 10-03 mremihr x 16.5 brs)= 1.09 x10°01mremNote: Same calculation is used for the other isotopes (radioiodines and noble gases) listed below.Effluent Concentration Limits. Concentrations at Point of Maximum Concentrationand Radiation Doses in the Unrestricted Area -RadioiodineRadionuclide1311132I133I134I1351Effluent Limit2.00 x 10"1° giCi/ml6.00 x 10.08 iiCi/ml6.00 x 10-09 Maximum Concentration12.30 x 10-10 giCi/ml4.47 x 10.1° pCi/ml6.90 x 10-10 iiCi/ml8.54 x 10-l0 jiCi/ml7.03 x 10-1° jCi/mlRadiation Dose1.09 x 10-01 mrem2.11 x 10.03 mrem6.50 x 10-02 mrem1.34 x 10-° mrem1.10 x 10.02 mremTotal = 0.19 mremNote 1: Maximum Concentrations are radioiodine and noble gas concentrations leaking from the containment buildingand exiting the exhaust stack reduced by a dilution factor of 292.52 of 86 Effluent Concentration Limits. Concentrations at Point of Maximum Concentrationand Radiation Doses in the Unrestricted Area -Noble GasesRadionuclide85Kr~~87Kr88Kr89Kr90Kr133Xel35mXe135mXe137XeEffluent Limit7.00 x 10.0 ptCi/ml1.00 x 10.07 pCi/ml2.00 x 10-°8 jiCi/ml9.00 x 10-09 pCi/ml2.00 x 10.08 jiCi/ml2.00 x 10.08 pCi/ml5.00 X 10-07 iiCi/ml7.00 x 10.08 aCi/ml4.00 x 10.08 !iCi/ml2.00 x 10.08 aCi/m12.00 x 10.08 iiCi/ml2.00 x 10.0 iiCi/mlMaximum Concentration12.54 x 10-12 pCi/ml5.97 x 10-1°/Ci/ml1.14 x 10.0 pCi/ml1.62 x 1 0-0 gCi/ml2.06 x 10.0 ptCi/ml2.06 x 10-0 2.27 x 10.0 pCi/ml5.21 x 10"1° pCi/ml5.09 x 10-40 pCi/ml2.66 x 10-0 aiCi/ml2.81 x 10.0 pCi/mi2.27 x 10.0 pCi/mlRadiation Dose3.43 x 10-0 mrem5.63 x 10-°4 mrem5.36 x i0.0 mrem1.69 x 10.02 mrem9.69 x 10°3 mrem9.69 x 1O0-3mrem4.28 x 10-0 mrem7.01 x 10-o4 mrem1.20 x 10.03 mrem1.25 x 10-02 mrem1.33 x 10°2mrem1.07 x 10-02 mremTotal = 0.08 mremNote 1: Maximum Concentrations are radioiodine and noble gas concentrations leaking from the containment buildingand 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 dosesfrom the radioiodines and noble gases must be added together, and result in the following values:Dose from Radioidines and Noble Gases in the Unrestricted AreaCommitted Effective Dose Equivalent (Radioiodine)Committed Effective Dose Equivalent (Noble Gases)Total Effective Dose Equivalent (Whole Body)0.19 mrem0.08 mrem0.27 mremSumming the doses from the noble gases and the radioiodines simply substantiates earlierstatements regarding the very low levels in the unrestricted area should a failure of a fueledexperiment occur, and should the containment building leak following such an event. Because thedose values are so low, the dose from the noble gases becomes the dominant value, but the overallTEDE is still only 0.19 mrem, a value far below the applicable 10 CFR 20 regulatory limit for theunrestricted area. Additionally, leakage in mechanical equipment room 114 from such items asvalve packing, flange gaskets, pump mechanical seals, etc. was also considered in the fuel failureanalysis. 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 partof the control room operator's actions. The additional contaminated water vapor and associatedisotopes added to the facility ventilation exhaust system made a minimal (<1%) contribution to thetotal dose of an individual located in the facility. Therefore, the dose contribution to theunrestricted area would be expected to be approaching zero.53 of 86 13.2.1.3 ConclusionsGenerally, the most severe condition which is analyzed with regard to reactor accidents is either aloss of primary coolant or a loss of primary coolant flow during reactor operation. Both of theseaccidents are analyzed in this chapter and the results show no core damage. In addition, there areno other accidents that will result in a release of fission products from the reactor fuel, which isassumed in the fuel failure analysis. Even if such an event were to occur, the anti-siphon andreactor loop vent systems are designed such that any released radioactivity would be contained inthe primary coolant system.System design and operational procedures reduce the likelihood of any foreign material beingintroduced into the reactor core that could cause a partial flow blockage. Calculations have beenperformed which indicate that even partial flow blockage to a fuel element will not result incladding failure (Ref. 13.2). A considerable margin of safety has been designed into the system inthis regard. Also, considering the results of the analyses which show no core damage in the eventof a loss of primary coolant or a loss of primary coolant flow accident (See Sections 13.2.3 and13.2.4), and in view of the design of the anti-siphon and reactor loop vent systems, it is concludedthat there is no radiation risk to personnel in the reactor containment building or in the unrestrictedarea 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 theconsequences of a loss of coolant accident (LOCA). MURR SAR Section 13.2.3.2 describes theLOCA event for the loss of the PCS integrity, and states that the accident of greatest consequence isa rupture in the short section of the PCS piping (either the cold leg or the hot leg) between thereactor pool and either isolation valves (507B or 507A4). The SAR describes the consequences of acold leg break between the isolation valve 507B and the reactor pool in significant detail. The hotleg break discussion is more succinct. The SAR also states that how "the anti-siphon systemensures 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 breaksequence. It is our understanding that after isolation occurs the coolant surrounding the core heatsup, and because of natural buoyancy it flows upward and out of the reactor pressure vessel into thein-pool heat exchanger. After passing through the heat exchanger, the cooled water may then flowdownward through what is normally the upward flow path of the inverted loop and then into thebottom of the pressure vessel. As this process continues, the water will fill up the downwardinverted loop to the bottom of the core reaching to the inverted loop creating an open condition forreleasing 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 fueltemperatures, or justify why no additional information is needed.In the second paragraph of the above question, the NRC staff's stated understanding is closer towhat occurs during a Loss of Flow Accident (LOFA) but not for the hot-leg break Loss of CoolantAccident (LOCA). The difference is that during the hot-leg break LOCA some of the primarycoolant that is lost from the primary coolant system (PCS5) piping is located in the reactor pool, butno primary coolant is lost during a LOFA. So, in the LOFA, the natural convention flow pathdescribed above is established and provides more than sufficient cooling for the reactor core aftershutdown. During a hot-leg break LOCA, the anti-siphon system actuates and injects air into thePCS vertical 12-inch diameter piping above the inverted loop to the level of the in-pool heatexchanger outlet. The expanding air quickly voids the upper section of the potential PCS naturalconvention flow path.Key PCS components for the LOFA and LOCA are described in Table 1 along with their RELAPModel component number. These components are also indicated in the vertical cross-sectionalview 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 SystemRELAP No. Component Description405-1 hn-Pool Heat Exchanger Upper Header405-2 hn-Pool Heat Exchanger Vertical Finned Tubes405-3 hn-Pool Heat Exchanger Lower Header401-2 Last 4 feet of 6-inch Diameter Inlet Piping to hn-Pool Heat Exchanger UpperHeader407 PCS Vertical 12-inch Diameter Pipe above hn-Pool Heat Exchanger Outlet toFlanged Natural Circulation Piping406 PCS Vertical 12-inch Diameter Pipe Above hnverted Loop to hn-Pool HeatExchanger Outlet139 Horizontal PCS Inlet Piping to Upper Section of Pressure Vessel501 Pressure Vessel Above the Core to Pressure Vessel Head100-3 Last 4.917 feet of Vertical Hot-Leg Piping Before Joining Pipe No. 101101 PCS Horizontal 12-inch Diameter Pipe (Section of inverted Loop)102 PCS Downward Vertical 12-inch Diameter Pipe from the Normal Outlet End ofNo. 101 Towards the PCS Hot-Leg Outlet Isolation ValveWith no pipe break occurring in the PCS during a LOFA, all of the above sections of the PCS stayfilled with primary coolant. This results in the development of the natural circulation flow pathdescribed in Paragraph 2 of the above question. However, for the hot-leg break LOCA, a doubleshear on the inlet and outlet sides of the hot-leg isolation valve is assumed, such that the hot-legisolation valve is functionally eliminated. With anti-siphon system air being injected intocomponent 406, voiding starts at the higher elevated connected PCS components as indicated inTable 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 exchangercomponents (405-1, 405-2, 405-3) start draining. Also, the following components aredraining:* 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 thereactor core, drain downward with the primary coolant which is flowing down the pressure vesselthrough the core and then up through the PCS hot-leg outlet piping until their upper level is in57 of 86 equilibrium with the water level in component 101 or 100-3. With the in-pool portion of the PCSdrained to this level, the natural circulation flow path through the in-pool heat exchanger iseliminated.However, the hot-leg break LOCA RELAP analysis shows that the highest peak fuel center linetemperature of 281.2 0F (138.4 °C) occurs in fuel plate number-i, 0.2 seconds after the LOCAbegins. After this initial peak temperature at the start of the transient, the next highest fuel platecenterline temperature of 231.7 °F (110.9 °C) occurs in plate number-22 at 22 seconds as shown inSARk Figure 13.20. The highest coolant channel temperature 219.0 0F (103.9 °C) occurs in channel7 at 123.3 seconds and in channel 6 at 123.4 seconds as shown in SARk Figure 13.21. There issufficient heat transfer from the PCS to the pooi coolant due to conduction through the PCS pipingto avoid any fuel damage.58 of 86
9. NUREG-1537, Section 13.1.5, "Mishandling or Malfunction of Fuel" provides guidance that thelicensee analyze the consequences of a mishandled fuel event. MURR SAR Section 13.2.5.2.1describes damage to a fuel element due to mishandling. It states that the mishandling could occurduring movement and packaging of the irradiated fuel, damage could only occur to the inner or theouter fuel plate, and could only occur during fuel element relocation activities. Because thisaccident 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 anoccurrence assuming that the fuel element has decayed for 60 days as part of the spent fuelmovement from storage to a shipping container. However, the NRC staff questions whether thisevent could also occur during the initial stages of refueling which would invalidate the assumptionof 60 days of decay. The NRC staff also performed a confirmatory calculation based on thisinventory using the cited values for the MHA analysis, and it results in an inventory that is sevenpercent larger than reported by MURR.a. Explain the possibility of this event occurring during the initial stages of refueling, and theapplicability of using the stated decay time in the dose calculation. Also, describe anyradioactivity release alarms that are expected to actuate, and whether containment isolationis expected, including the time required to verify containment isolation, or justify why noadditional 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 andmethodologies as used in the Maximum Hypothetical Accident (MHA) (now referred to as the"Fuel Failure during Reactor Operation" accident) and Fueled Experiment Failure. The onlyexceptions are the source term, which is explained in the accident analysis, as well as the decayprior to the accident (which is once again explained in the analysis). As discussed in the responseto RAI 10O.a, the primary coolant system does not have to be secured for a failed fueled experimentor for a FHA. The only required action for Operations personnel is to verify that the containmentbuilding has been evacuated following a containment building isolation, which will occur duringboth of these accident scenarios. MUJIRR performs an evacuation dr-ill every year and the typicaltime period for all personal to evacuate the containment building, including verification byOperations personnel, is two (2) to two and a half (2.5) minutes. For the purposes of the failedfueled experiment and FRA calculations, a conservative assumption of five (5) minutes is used forboth accident scenarios. Additionally, verifying that the reactor has shut down and containment hasisolated only takes a few moments -all control blade positions, reactor power meters, andcontainment isolation valve and door indications are in clear view of the reactor operator in thecontrol room.b. Provide the details of how the source term is determined, or justify' why no additionalinformation is needed.As described in the FHA analysis above, the two most outer fuel plates of a fuel element, number-land -24, are the plates most likely to be damaged during fuel handling. The number-i fuel platecontains 19.26 grams of U-235 before irradiation. The highest peak power density in the various59 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 ofthe fuel plate. The number-24 fuel plate has a lower peak power density and contains 45.32 gramsof U-235, and has the most surface area to be damaged. To be conservative, the analysis assumesthat 0.125 grams of U-235 is exposed from plate number-i during the FHA, which corresponds toremoving a section of fuel meat from a plate that is 1 inch square and 5 mils thick. A powerpeaking 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 andAccounting Procedures as outlined in the Operations Procedures. Irradiated fuel is handled with aspecially 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 fuelhandling sequence. Fuel elements are always handled one at a time so that they are maintained in acriticality-safe configuration. New or irradiated fuel may be stored in any one of 88 in-pool fuelstorage locations (not including the core). These storage locations are designed to ensure ageometry such that the calculated Keff is less than 0.9 under all conditions of moderation, thusallowing sufficient convection cooling and providing sufficient radiation shielding.So the fuel handling system provides a safe, effective and reliable means of transporting andhandling 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 andmagnetic particle testing, as appropriate. Therefore, no specific accidents regarding the handling offuel have been identified for the MUIRR. The probability of dropping a fuel element whileunderwater and damaging it severely enough to breach the fuel cladding was considered. Aconservative potential radionuclide release and calculation of the occupational exposure areincluded below.The following calculations determining the postulated dose from a potential release of radioactivityfrom a fuel element during a handling accident closely follow the "Fuel Failure during ReactorOperation" calculations for personal exposure due to a release of fission products. The objectiveof these calculations is to present a worst-case dose assessment for a person who remains in thecontainment 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 ofelements. Paired elements are always loaded opposite each other in the core. All eight (8) fuelelements are replaced every refueling. MURR has averaged refueling the core more than 52 timesa year since 1977. This type of accident has never occurred at MIJRR during any of these fuelhandlings.The two outer fuel plates of a fuel element, number-i and -24, are the plates most likely to bedamaged during fuel handling. The number-i fuel plate contains 19.26 grams of U-235 beforeirradiation. The highest peak power density in the various MURR core configurations occurs infuel plate number-i of a previously unirradiated fuel element, which has a power peaking factor of4.116 -located between 14.75 to 15.75 inches down from the top of the fuel plate. The number-24fuel plate has the most surface area to be damaged; however, it has a lower peak power density andcontains 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 offuel meat from a plate that is 1 inch square and 5 mils thick. A power peaking factor of 4.116 isalso applied.61 of 86 The following radioiodine, krypton and xenon activities will be present in the MURR core 30minutes after shutdown from 10 MW full power operation. Refuelings typically occur no soonerthan an hour after shutdown. This takes into account the time required to shut down the reactor, tosecure the primary coolant system (required to stay in operation a minimum of 15 minutes after thecontrol blades are fully inserted), and to remove the reactor pressure vessel head. For the purposeof the FHA calculations, a conservative assumption of 30 minutes is used.Radioiodine and Noble Gas Activities in the Core after 30-Minute Decay131I -9.93 x 10+04 Ci 85Kr -2.47 x 10+01 Ci 133Xe -2.73 x 10+05 Ci132I -2.68 x 10+o Ci 85mKr -1.29 x 10+05 Ci 135Xe -1.13 x i0o Ci1331 -- 5.65 x i0+0 Ci 87Kr -1.67 x 10+° Ci 135mXe -- 4.79 x 10+04 Ci134I -- 5.80 X i0+0 Ci 88Kr -2.73 x 10+05 Ci '37Xe -2.37 x 10+0 Ci1351 -- 5.07 x 10+07 Ci 89}r -5.5 x 10+02 Ci 138X -1.22 x 10+05 Ci9°Kr -6.66 x 10-12 Ci 139Xe -8.33 x 10-09 CiFission products released into the reactor pool will be detected by the pool surface and ventilationsystem exhaust plenum radiation monitors. However, for the purposes of this analysis, it isassumed that an actuation of the containment building isolation system occurs by action of the poolsurface radiation monitor. Actuation of the isolation system will prompt Operations personnel toensure that a total evacuation of the containment building is accomplished promptly, usually withintwo (2) to two and a half (2.5) minutes. A conservative 5-minute evacuation time is used as thebasis for the stay time in the dose calculations for personnel that are in containment during theFRA.The following radioiodine and noble gas activities from 0.125 grams of U-235 from the peak powerposition of fuel plate number-i in the peak power density fuel element are assumed toinstantaneously and homogenously distribute in the reactor pool.Example calculation of 1311 released into the reactor pool:= (1311 in fuel / 235U in core) x 235U 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 1iCiExample calculation of 85Kr released into the reactor pool:= (85Kr in fuel / 235U in core) x 235U 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 Pool131I -- 9.33 x 10+06 kCi 85Kr -2.32 x 10+03 gCi 133Xe -- 2.56 x 10+07 giCi132I -2.52 x 10+0 giCi 85mKr -1.21 x 10+07 gxCi '35Xe -1.06 x 10+07 /iCi133I -5.31 x 10+07 g.Ci 87Kr -1.57 x 10+0 1iCi l35mXe -4.50 x 10+06 jiCi1341 -- 5.45 X 10+0 gtCi 88Kr -2.56 x 100 jiCi 137Xe -2.22 x 10+05 gCi35-- 4.76 x 10+07 ptCi 9K~r -5.25 x 10+0 ptCi '38Xe -1.15 x 10+07 plCi9°~Kr -6.26 x 10-'° ItCi 139Xe -7.83 x 10-07 pCiThe radioiodine released into the reactor pooi over a 5-minute interval is conservatively assumed tobe instantly and uniformly mixed into the 20,000 gallons (75,708 1) of bulk pooi water, which thenresults in the following pool water concentrations for the radioiodine isotopes. The water solubilityof the krypton and xenon noble gases released into the pool over this same time period are ignoredand they are assumed to pass immediately through the pool water and evolve directly into thecontainment building air volume where they instantaneously form a uniform concentration in theisolated structure.Radioiodine Concentrations in the Pool Water131I -- 4.67 x 10+02 pCilgal 1331 -- 2.66 x 10+03 giCi/gal 135I -2.38 x 10+03 PCi/gal13I -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, theevaporation rate from the reactor pool is approximately 80 gallons (302.8 L) of water per day. Forthe purposes of this calculation, it is assumed that a total of 20 gallons (75.7 L) of pool watercontaining the previously listed radioiodine concentrations evaporates into the containment buildingover the 5 minute period. Containment air with a temperature of 75 0F (23.9 °C) and 100% relativehumidity contains H20 vapor equal to 40 gallons (151.4 L) of water. Since the air in containment isnormally at about 50% relative humidity, thus containing 20 gallons (75.7 L) of water vapor, theassumed addition of 20 gallons (75.7 L) of water vapor will not cause the containment air to besupersaturated. It is also conservatively assumed that all of the radioiodine activity in the 20gallons (75.7 L) of pool water instantaneously forms a uniform concentration in the containmentbuilding air. When distributed into the containment building, this would result in the followingradioiodine concentrations in the 225,000 ft3 (6,371.3 in3) air volume:Example calculation of 131I released into containment air:= 131I concentration in pooi water x 20 gal x 1/225,000 ft3 x 35.3 147 ft3/m3-4.67 x 10+02 ptCi/gal x (3.14 x 10.03 gal/in3)-1.46 pCi/m3(1.46 pCi/in3) x (1 m3/106 ml) =1.46 x 10.06 gxCi/mlNote: 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 theconcentrations after 5 minutes decay divided by 2.Average Radioiodine Concentrations in the Containment Building Air during the 5 Minutes31-- 1.46 x 10-06 pCi/mi 1331 -- 8.32 x 10°06 gtCi/ml 135I -- 7.44 x 10.06 gCi/ml132j -3.91 x 10.06 gtCi/ml 14-- 8.28 x i0-°5 gCi/mlAs 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 andenter the containment building air volume where they instantaneously form a uniform concentrationin the isolated structure. This assumption is extremely conservative since it ignores the knownsolubility of krypton and xenon noble gases in the 100 0F (37.8 °C) pool water, which would reducetheir release into the containment building. Based on the 225,000-ft3 volume of containmentbuilding 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 wouldbe as follows:Example calculation of 85Kr released into containment air:-85Kr activity x 1/225,000 ft3 x 35.3 147 ft3/m3= 2.32 X 10+03 j.Ci x (1.60 x i004 1/in3)= 3.64 x 1001 g.Ci/m3(3.64 x 10°1 gxCi/m3) x (1 m3/106 ml) = 3.64 x i0-07 plei/mlNote: Same calculation is used for the other isotopes listed below.The average noble gas concentrations are the sum of the initial concentrations and theconcentrations after 5 minutes decay divided by 2.Average Noble Gas Concentrations in the Containment Building Air during the 5 MinutesKr -3.64 x 10-07 FCi/ml 133Xe -4.02 x 10-03 pCi/miS5m~x -1.89 x 10-03 gCi/ml '35Xe -1.66 x 1003 jiCi/ml87r- 2.41 x 10.03 gCi/ml 13smXe -6.35 x 10-04 iCi/ml88r- 3.98 x i0-03 jiCi/ml 137Xe -2.45 x i0-°5 gCi/mI89Kr -5.49 x 10-0 gtCi/ml '38Xe -1.61 x 10-°3 pCi/mi9°Kr -4.92 x 10-20 gCi/ml 139Xe -6.18 x10-17 Ci/mlThe objective of this calculation is to present a worst-case dose assessment for an individual whoremains in the containment building for 5 minutes following the ERA. Therefore, as notedpreviously, the radioactivity in the evaporated pool water is assumed to be instantaneously anduniformly 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 thyroidfrom radioiodine and the dose to the whole body resulting from submersion in the airborne noblegases and radioiodine inside the containment building. As previously noted, the exposure time forthis dose assessment is 5 minutes.Because the airborne radioiodine source is composed of five different iodine isotopes, it will benecessary to determine the dose contribution from each individual isotope and to then sum theresults. 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 25mrem/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/ml131j DAC (10 CFR 20) = 2.00 x 10.o gCi/mlDose Multiplication Factor =(1311 concentration) / (131I DAC)= (1.46 x 10-°6 gCi/ml) / (2.00 x 10.o8 gCi/ml)= 73Therefore, 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 mremNote: 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 thesame manner as the radioiodines, and the dose contribution from each individual radionuclide mustbe calculated and then added together to arrive at the final noble gas dose. Because the dose fromthe noble gases is only an external dose due to submersion, and because the DACs for theseradionuclides are based on this type of exposure, the individual noble gas doses for 5 minutes incontaimnment were based on their average concentration in the containment air and thecorresponding 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, 5minutes of one DAC-br is 8.33 x l0°2 DAC-hr.85Kr concentration in containment = 3.64 X 10-07 gtCi/mnl65 of 86 85Kr DAC (10 CFR 20)Dose Multiplication Factor= 1.00 x 10"04 pCi/ml= (85Kr concentration) / (85Kr DAC)= (3.64 x 10-07 ptCi/ml) / (1.00 x 10-04 jiCi/ml)= 0.00364Therefore, 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)= 7.58X10-04 mremNote: 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 -RadioiodineRadionuclide131113211331134j1351Derived Air Concentration2.00 x 10-°8 3.00 x 10-06 /xCi/ml1.00 x 10-07/Ci/ml2.00 x i0.0 gCi/inl7.00 x 10-07 iiCi/ml5-Minute Exposure1.52 x 10+o2 mrem2.71 x 10+00 mrem1.73 x 10+02 mrem8.62 x 10-°1 mrem2.21 x 10+°1 mremTotal = 351.44 mremDerived Air Concentration Values and 5-Minute Exposures -Noble GasesRadionuclide85Kr85m}r87KrS88K89Kr9OKr133XeI35Xe135mXe138Xe139XeDerived Air Concentration1.00 x 10-°4 iCi/ml2.00 x 10.o jtCi/ml5.00 x 100o6 gCi/ml2.00 x 100o6 xtCi/ml6.00 x 10.o6 JiCi/m16.00 x 10-°6 pCi/ml1.00 x 10-04 pCi/mi1.00 x 10-°5 pCi/ml9.00 x 10-06 pCi/ml6.00 x 10-06 gCi/ml4.00 x 10-06 jiCi/ml6.00 X 10-06 .tCi/ml5-Minute Exposure7.58 x 10-04 mrem1.96 x 10+°1 me1.00 x 10+02 mrem4.14 x 10+02 mrem1.91 x 10-°1 mrem1.71 x 10-15 mrem8.36 x 10+00 mrem3.45 x 10+°1 mrem1.47 x 10+°1 mrem8.49 x 10-°1 mrem8.37 x 10+°1mre2.14 x 1012 mremTotal = 676.45 mrem66 of 86 To finalize the occupational dose in terms of Total Effective Dose Equivalent (TEDE) for a5-minute exposure in the containment building after a FHA, the doses from the radioiodines andnoble gases must be added together, and result in the following values:5-Minute Dose from Radioidines and Noble Gases in the Containment BuildingCommitted Dose Equivalent (Thyroid) 351.44 mremCommitted Effective Dose Equivalent (Thyroid) 10.54 mremCommitted Effective Dose Equivalent (Noble Gases) 676.45 mremTotal Effective Dose Equivalent (Whole Body) 687.00 mremBy comparison of the maximum TEDE and Committed Dose Equivalent (CDE) for thoseoccupationally-exposed during a FHA to applicable NRC dose limits in 10 CFR 20, the final valuesare shown to be well within the published regulatory limits and, in fact, lower than 15% of anyoccupational limit.Radiation shine through the containment structure was also evaluated when considering accidentconditions and dose consequences to the public and MUIRR staff. Calculation of exposure ratefrom a FHA was performed using the computer program MicroShield 8.02 with a RectangularVolume -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 comprehensivephoton/gamma ray shielding and dose assessment program that is widely used by industry fordesigning radiation shields.The exposure rate values provided below represents the radiation fields at 1 foot (30.5 cm) from a12-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 theexposure rate values are identical to those used for determining the dose to a worker withincontainment from noble gases. For radioiodine, the total iodine activity from the FHA was used forthe dose calculations, not the amount that evaporated in 5 minutes. The source term also assumes ahomogenous mixture of nuclides within the containment free volume.Radiation Shine through the Containment BuildingExpo sure Rate at 1-Foot from Containment Building Wall: 54.79 mrem/hrExposure Rate at Emergency Planning Zone Boundary (150 meters): 0.371 mremn/hrA confirmatory analysis of the accident condition yielding the largest consequence was validatedindependently by the use of the MCNP code. This analysis yielded a result 21% less than theMicroshield method and results provided above.As noted earlier in this analysis, the containment building ventilation system will shut down and thebuilding itself will be isolated from the surrounding areas. A FHA will not cause an increase inpressure inside the reactor containment structure; therefore, any air leakage from the building willoccur as a result of normal changes in atmospheric pressure and pressure equilibrium between the67 of 86 inside of the contaimnment structure and the outside atmosphere. It is highly probable that there willbe no pressure differential between the inside of the containment building and the outsideatmosphere, and consequently there will be no air leakage from the building and no radiation doseto members of the public in the unrestricted area. However, to develop what would clearly be aworst-case scenario, this analysis assumes that a barometric pressure change had occurred inconjunction with a FHA. A reasonable assumption would be a pressure change on the order of 0.7inches of Hg (25.4 mm of Hg at 60 which would then create a pressure differential of about0.33 psig (2.28 kPa above atmosphere) between the inside of the isolated containment building andthe 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 ratelimit [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 feetper minute (scfm) as a function of containment pressure can be expressed by the followingequation:LR = 17.85 x (CP-14.7)l"2;where:LR = leakage rate from containment (scfmn); andCP = containment pressure (psia).The minimum Technical Specification free volume of the containment building is 225,000-ft3 atstandard temperature and pressure. At an initial overpressure of 2.0 psig (13.8 kPa aboveatmosphere), 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 airvolume to 230,051 scf. The above equation describes the leakage rate that results in this drop ofcontained 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 initialoverpressure condition of 0.33 psig (2.28 kPa above atmosphere), it would take approximately 16.5hours 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 periodwould be about 5.2 scfm.Several factors exist that will mitigate the radiological impact of any air leakage from thecontainment building following a FHA. First of all, most leakage pathways from containmentdischarge into the reactor laboratory building, which surrounds the containment structure. Since thelaboratory building ventilation system continues to operate during a FHA, leakage air captured bythe ventilation exhaust system is mixed with other building air, and then discharged from thefacility through the exhaust stack at a rate of approximately 30,500 cflm. Mixing of containment airleakage with the laboratory building ventilation flow, followed by discharge out the exhaust stackand subsequent atmospheric dispersion, results in extremely low radionuclide concentrations andvery small radiation doses in the unrestricted area. A tabulation of these concentrations and doses68 of 86 is given below. These values were calculated following the same methodology stated in Section5.3.3 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 tothe behavior of radioiodine, which has been studied extensively in the containment mockup facilityat 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% iodinedeposition in the containment building, each cubic meter of air released from containment has aradioiodine concentration that is 25% of each cubic meter within containment building air, then theradioiodine concentrations leaking from the containment structure into the laboratory building, inmicrocuries per milliliter, will be:Example calculation of 1311 released through the exhaust stack:-1311 activity / (30,500 ft3/min x 16.5 hr x 60 mir/hr x 28,300 ml/ft3)-9.33 x 10+06 1iCi / 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/mlNote: Same calculation is used for the other radioiodines listed below.Radioiodine Concentrations in Air Leaking from Containment and Exiting the Exhaust Stack13I- 2.73 x 10.06 gtCi/ml 133I -1.55 x 10°5~ gCi/ml 3I-- 1.39 x 10°s~ pCi/ml132I --7.37 x 10.06 giCi/ml 134I -- 1.59 x 10-°s pCi/mlExample calculation of 85Kr released through the exhaust stack:-85Kr activity / (30,500 ft3/min x 16.5 hr x 60 mmn/hr x 28,300 ml/fl3)-2.32 x 10+03 ptCi / 8.55 x 10+1n ml-2.71 x 10.09 jCi/mlNote: Same calculation is used for the other noble gases listed below.Noble Gas Concentrations in Air Leaking from Containment and Exiting the Exhaust Stack85Kr -2.71 x 10.09 jCi/ml 87Kr -1.84 x 10-05 gxCi/ml 89Kr -6.14 x 10.08 jCi/ml8Smjr -1.42 x 10.05 iCi/ml 88Kr -3.00 x 10-05 jCi/ml 9°Kr -7.33 x 10-22 pCi/ml33e- 3.00 x 10-0 pCi/ml 135mXe -5.27 x 10-°6 gxCi/ml 138Xe -1.35 x 10.0 gxCi/ml3Xe- 1.24 x 10.05 pCi/ml 137Xe -2.60 x 10-o7 1iCi/ml 139Xe -9.16 x 10-'9 pCi/mlAssuming, as stated earlier, that (1) the average leakage rate from the containment building is 5.2scfm, (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 buildingpressure with atmospheric pressure, (3) the flow rate through the facility's ventilation exhaust stack69 of 86 is 30,500 scfmn, (4) the reduction in concentration from the point of discharge at the exhauststack to the point of maximum concentration in the unrestricted area is a factor of 292 and (5) thereis no decay of any radioiodines or noble gases, then the following concentrations of radioiodinesand noble gases with their corresponding radiation doses will occur in the unrestricted area. Thevalues listed are for the point of maximum concentration in the unrestricted area assuming auniform, semi-spherical cloud geometry for noble gas submersion and further assuming that themost conservative (worst-case) meteorological conditions exist for the entire 16.5-hour period ofcontainment leakage following a FHA. Radiation doses are calculated for the entire 16.5-hourperiod. Dose values for the unrestricted area were obtained using the same methodology that wasused to deternine doses inside the containment building, and it was assumed that an individual waspresent 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 buildingwas leaking.A worst-case scenario dilution factor of 292 for effluent dilution using the Pasquill-Guifford Modelfor atmospheric dilution is used in this analysis. We assume that all offsite (public) dose occursunder these atmospheric conditions at the site of interest, i.e. 760 meters North of MURR. In ourcase at 760 meters it occurs only during Stability Class F conditions; which normally only occur11.4% 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. AnEffluent Concentration Limit of 2.0 x 10.08 iiCi/ml is used for the "unlisted" isotopes, which equalsthe DAC/300 when using the DOE Part 835 Default DAC limits of 6.0 x 10.06 giC/ml. Exposure at1 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 8760hours per year. Thus, the time assumed to be exposed to the effluent concentration limit is a factorof 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 basedon 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'sin Appendix C of Part 835. No credit is taken for transit time from the stack to the receptor pointnor is credit taken for decay inside containment until release. In the case of Kr-89 and Xe-1 37 thetransit 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 theDAC value to establish the effluent concentration limit is warranted. This reduction factor of 300 isconsistent, in fact more conservative, than 10 CFR 20 Appendix B, as the NRC calculates EffluentConcentration 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/hr131I concentration = 9.35 x 10.0 pCi/ml131I effluent concentration limit =2.00 x 10-'° pCi/ml1311 Conversion Factor = 5.71 x 10.03 mnrem/hr70 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+°° mremNote: Same calculation is used for the other isotopes (radioiodines and noble gases) listed below.Effluent Concentration Limits. Concentrations at Point of Maximum Concentrationand Radiation Doses in the Unrestricted Area -RadioiodineRadionuclide1311I321133113411351Effluent Limit2.00 x 10-1° p.Ci/ml2.00 x 10.08 pCi/ml1.00 X 10-0 jiCi/ml6.00 x 10-0 p.Ci/ml6.00 x 10.09 tCi/mlMaximum Concentratior2.52 x 10.0 pCi/ml5.32 x 10-° jiCi/ml5.46 x 10-0 jiCi/mla' Radiation Dose4.40 x 10+° mrem1.19 x 10-°1 mrem5.01 x 10+00 mrem8.57 x 10-02 mrem7.49 x 10-°1 mremTotal = 10.37 mremNote 1: Maximum Concentrations are radioiodine and noble gas concentrations leaking from the containment buildingand exiting the exhaust stack reduced by a dilution factor of 292.Effluent Concentration Limits, Concentrations at Point of Maximum Concentrationand Radiation Doses in the Unrestricted Area -Noble GasesRadionuclide85Kr87Kr8tKr89Kr90Kr13smXeEffluent Limit7.00 x 10.0 pCi/ml1.00 x 10.07 pCi/mi2.00 x 10.08 xCi/ml9.00 x 10-09 pCi/ml2.00 x 10.08 pCi/mi2.00 x 10-° jxCi/ml5.00 x 10-0 gCi/ml7.00 x 10.08 xCi/ml4.00 x i0.0 gCi/ml2.00 x 10.0 gCi/ml2.00 x 10.08 gCi/ml2.00 x 10-08 jxCi/mlMaximum Concentration19.30 x 10-12 igCi/ml4.85 x 10-°8/.tCi/ml6.29 x 10-08 pCi/ml1.03 x 10-07 pCi/ml2.51 x 10-24 pCi/ml1.03 x 10-0 pCi/mI4.25 x 10-0 pCi/ml1.80 x 10-08 itCi/ml4.61 x 10-0 pCi/ml3.14 x 10-21 p.Ci/mlRadiation Dose1.25 x 10-06 mrem4.57 x 10.02 mrem2.96 x 10.01 mrem1.07 x 10+0° mrem9.91 x 10-04 mrem1.18 x 10-'7mrem1.93 x 10-02nmrem5.72 x 10.02 mrem4.25 x 10.02 mrem4.19 x 10-0 mrem2.17 x 10-°1 mrem1.48 x 10"14 mremTotal = 1.76 mremNote 1: Maximum Concentrations are radioiodine and noble gas concentrations leaking from the containment buildingand 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 dosesfrom the radioijodines and noble gases must be added together, and result in the following values:Dose from Radioidines and Noble Gases in the Unrestricted AreaCommitted Effective Dose Equivalent (Radioiodine) 10.37 mremCommitted Effective Dose Equivalent (Noble Gases) 1.76 mremTotal Effective Dose Equivalent (Whole Body) 12.13 mremSumming the doses from the noble gases and the radioiodines simply substantiates earlierstatements regarding the very low levels in the unrestricted area should a FHA occur, and shouldthe containment building leak following such an event. Because the dose values are so low, thedose from the noble gases becomes the dominant value, but the overall TEDE is still only 12.13mrem, 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 licenseesanalyze the consequences of a failed fueled experiment. SAR Section 13.2.6.2 describes that limitingfueled experiments to 150 curies of radioiodine will result in a projected dose well within the limitsof 10 O CFR Part 20. The response to RAJ 13.9.a (ADAMS Accession No. ML103060018) providesradioiodine and noble gas activities for a 5-gram low-enriched fuel target. The response uses amethod similar to that used in the MHA analysis and lists the gaseous fission products to bereleased into the pool cooling system. The occupational dose calculation assumes a 2-minuteevacuation time. The NRC staff notes that the submersion dose calculations were performed usingthe 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 notlisted 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-minuteevacuation time assumed in the MHA analysis, or the SAR Section 13.2.1.2 statement that it takesthe operations staff approximately 5 minutes to secure the PCS and verify containment isolationfollowing a containment isolation signal.a. Please clarify the sequence of events, state which alarms are expected to provide indicationthat evacuation is required, justify' the evacuation time, and use that time to revise the doseassessment employing consistent DAC values, or justify why no additional information isneeded.Following the response to RAI 10O.c is the revised fueled experiment failure analysis that replacesthe 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). Aspreviously 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 takeapproximately 5 minutes for Operations personnel to secure the primary coolant system and verifythat the containment building has been evacuated following a containment building isolation. Forthe 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 fueledexperiment or for a fuel handling accident (FHA). The only required action for Operationspersonnel is to verify that the containment building has been evacuated following a containmentbuilding isolation, which will occur during both of these accidents. MURR perforns an evacuationdrill every year and the typical time for all personal to evacuate the containment building, includingverification by Operations personnel, is two (2) to two and a half (2.5) minutes. For the purposesof 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 downand 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 thecontrol room.The Derived Air Concentration (DAC) values used for the dose calculations for each accidentscenario -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 used73 of 86 whereas for the "unlisted" isotopes the DACs of 10 CFR 835 are used (published in the FederalRegister, 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-131through Iodine-135 greater than 1.5 curies or Strontiunm-90 greater than 5 millicuries shallbe vented to the facility ventilation exhaust stack through high efficiency particulate air andcharcoal filters which are continuously monitored for an increase in radiation levels." This isinconsistent with TS 3.8.o which states that a fueled experiment can be encapsulated orvented. C'larfify whether fueled experiments are vented or not and revise the TS if required, orjustify why no additional information is needed.License Amendment No. 34, issued to MURR on October 10, 2008, by the NRC, revised Technicalcurrent Specification (TS) 3 .6.o (relicensing TS 3.8.o) such that fueled experiments containinginventories 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 meetthe internal pressure design requirements specified in TS 3.6.i. TS 3.6.i states that "Irradiationcontainers to be used in the reactor, in which static pressure will exist or in which a pressurebuildup is predicted, shall be designed and tested for a pressure exceeding the maximum expectedpressure by at least a factor of two (2)."Until then, fueled experiments containing inventories of I-131 through 1-135 greater than 1.5 curiesor inventories of Sr-90 greater than 5 millicuries had to be vented to the facility ventilation exhauststack through high efficiency particulate air (H7EPA) and charcoal filters which were continuouslymonitored for radiation levels.Since Amendment No. 34 was issued after the SAR was submitted in August 2006 as a part ofrelicensing, SAR Section 13.2.6.2 is now outdated. The third bullet on page 13-67 should nowread, "Fueled experiments containing inventories of iodine-13 1 through iodine-i135 greater than 1.5curies or strontium-90 greater than 5 millicuries shall be in irradiation containers that satisfy therequirements of Specification 3.8 .i or be vented to the facility ventilation exhaust stack throughhigh efficiency particulate air (HEPA) and charcoal filters which are continuously monitored for anincrease in radiation levels."c. If such venting is permitted then explain why those contributions are not included in theinventory of normally released material ('such as Ar-41,), or justify why no additionalinformation 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 datedJune 18, 2015), all air exiting the facility through the ventilation exhaust system is monitored forairborne 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 thecontainment 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 radiationmonitoring 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 theventilation exhaust system; therefore, Ar-4 1 was used to determine the radiological impact ofairborne effluents during normal reactor operation. In addition to At-4 1, all other isotopes greaterthan 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 dischargedto the environs beyond the effective control of the licensee as measured at or prior to the point ofsuch release or discharge."~Attachment 13 (also included in the responses, dated July 31, 2015) provides the last 10 years, andaverage, of air releases from the facility per isotope in percentage of the Technical Specificationlimit. As you will note, with the exception of argon-4 1, all other isotopes discharged are less than0.6% 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 asthe basis for the source term for these dose calculations. A 5-gram LEU target irradiated for 150hours (normal weekly operating cycle) at a thermal neutron flux of 1.5 x 10+13 nlcm2-sec willproduce the following radioiodine, krypton and xenon activities (additionally, approximately 1.40 x10+04 jiCi of Strontium-90 will be produced):Radioiodine and Noble Gas Activities in a 5-Gram LEU Target131I -- 8.400 Ci 85Kr -0.002 Ci 133Xe -18.900 Ci1321 -18.600 Ci 85m~r -7.580 Ci 135Xe -13.600 Ci133I -- 39.900 Ci 87Kr -15.400 Ci l35mXe -6.760 Ci134I -- 45.400 Ci 88Kr -21.700 Ci 137Xe -35.800 Ci13I- 37.700 Ci 89Kr -27.740 Ci 138Xe -37.400 Ci9°Kr -27.400 Ci 139Xe -30.700 CiTotal Iodine -150.00 Ci Total Krypton -99.822 Ci Total Xenon -143.160 CiA complete failure of the target is unrealistic for many reasons. The worst that can be expected ispartial melting; however, in order to present a worst-case dose assessment for an individual thatremains in thle containment building following target failure, 100% of the total activity of the targetis assumed to be released into the reactor pool.Fission products released into the reactor pool will be detected by the pool surface and ventilationsystem exhaust plenum radiation monitors. However, for the purposes of this analysis, it isassumed that a reactor scram and actuation of the containment building isolation system occurs byaction of the pool surface radiation monitor. Actuation of the isolation system will promptOperations personnel to ensure that a total evacuation of the containment building is accomplishedpromptly, usually within two (2) to two and a half (2.5) minutes. A conservative 5-minuteevacuation time is used as the basis for the stay time in the dose calculations for personnel that arein containment during target failure.The radioiodine released into the reactor pool over a 5-minute interval is conservatively assumed tobe instantly and uniformly mixed into the 20,000 gallons (75,708 1) of bulk pool water, which thenresults in the following pool water concentrations for the radioiodine isotopes. The water solubilityof the krypton and xenon noble gases released into the pool over this same time period areconservatively ignored and they are assumed to pass immediately through the pool water andevolve directly into the containment building air volume where they instantaneously form a uniformconcentration in the isolated structure.76 of 86 Radioiodine Concentrations in the Pool Water1311 -- 4.20 x 10+02 gCi/gal 31-- 2.00 x 10+0 pCi/gal 1351 -- 1.89 x 10+0 gtCi/gal1321 -- 9.30 x 10+02 pCi/gal 1341 -- 2.27 x 10+03 pCi/galWhen the reactor is at 10 MW and the containment building ventilation system is in operation, theevaporation rate from the reactor pooi is approximately 80 gallons (302.8 L) of water per day. Forthe purposes of this calculation, it is assumed that a total of 20 gallons (75.7 L) of pool watercontaining the previously listed radioiodine concentrations evaporates into the containment buildingover the 5 minute period. Containment air with a temperature of 75 0F (23.9 °C) and 100% relativehumidity contains H20 vapor equal to 40 gallons (151.4 L) of water. Since the air in containment isnormally at about 50% relative humidity, thus containing 20 gallons (75.7 L) of water vapor, theassumed addition of 20 gallons (75.7 L) of water vapor will not cause the containment air to besupersaturated. It is also conservatively assumed that all of the radioiodine activity in the 20gallons (75.7 L) of pool water instantaneously forms a uniform concentration in the containmentbuilding air. When distributed into the containment building, this would result in the followingradioiodine concentrations in the 225,000 ft3 (6,371.3 in3) air volume:Example calculation of 131I released into containment air:=1311 concentration in pool water x 20 gal x 1/225,000 ft3 x 35.3 147 ft3/m3-4.20 x 10+02 pCilgal x (3.14 x 10.03 gal/in3)-1.32 jiCi/m3(1.32 pCi/in3) x (1 m3/106 ml) = 1.32 x 10.06 gCi/mlNote: Same calculation is used for the other isotopes listed below.The average radioiodine concentrations are the sum of the initial concentrations and theconcentrations after 5 minutes decay divided by 2.Average Radioiodine Concentrations in the Containment Building Air during the 5 Minutes1311 -- 1.32 x 10-°6 ixtCi/ml 133I -- 6.26 x 10-°6 giCi/ml 1351 -5.89 x 10-°6 giCi/ml1321 -2.88 x 10-06 pCi/ml 1341 -6.90 x 10-°6 ptCi/mlAs 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 immediatelythrough the pool water and enter the containment building air volume where they instantaneouslyform a uniform concentration in the isolated structure. Based on the 225,000-ft3 volume ofcontainment building air, and the previously listed curie quantities of these gases released into thereactor pool, the maximum noble gas concentrations in the containment structure at the end of 5minutes would be as follows:77 of 86 Example calculation of 85Kr released into containment air:= 85Kr activity x 1/225,000 ft3 x 35.3 147 ft3/m3= 1.71 x 10+0 x (1.60 x 10.0 1/mn3)-2.69 x 10-°1 jiCi/m3(2.69 x 10-°1 iiCi/m3) x (1 m3/106 ml) = 2.69 x 10-07 pCi/mlNote: Same calculation is used for the other isotopes listed below.The average noble gas concentrations are the sum of the initial concentrations and theconcentrations after 5 minutes decay divided by 2.Average Noble Gas Concentrations in the Containment Building Air during the 5 MinutesKr -2.69 x 10-0 pCi/mI 133Xe -2.97 x 10-03 pCi/ml8mr- 1.18 x 10-03 gCi/ml '35Xe -2.13 x 10.03 pCi/ml87Kr -2.36 x 10-0 jiCi/ml l35mXe -9.54 x 10-° ~tCi/ml88Kr -3.37 x 10.0 pCi/mi 137Xe -3.95 x 10-0 iiCi/ml89r- 2.90 x 10.03 xiCi/ml '38Xe -5.23 x 10-0 jiCi/ml9°Kr -2.15 x i0-0 kCi/ml '39Xe -- 2.42 x 10-0 jiCi/mlThe objective of this calculation is to present a worst-case dose assessment for an individual whoremains in the containment building for 5 minutes following target failure. Therefore, as notedpreviously, the radioactivity in the evaporated pool water is assumed to be instantaneously anduniformly 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 thyroidfrom radioiodine and the dose to the whole body resulting from submersion in the airborne noblegases and radioiodine inside the containment building.Because the airborne radioiodine source is composed of five different iodine isotopes, it will benecessary to determine the dose contribution from each individual isotope and to then sum theresults. 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 25mrem/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'311DAC (10 CFR 20) =2.00 x 10-°8 pCi/ml78 of 86 Dose Multiplication Factor= (1311 concentration) /(1311 DAC)= (1.32 x 10-06 pCi/ml) / (2.00 x 10-08 ptCi/ml)= 66Therefore, 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+02 mremNote: Same calculation is used for the other radioiodines listed below.Derived Air Concentration Values and 5-Minute Exposures -RadioiodineRadionuclide13111321133I134I1351Derived Air Concentration2.00 x 10.08 3.00 x 10-06 gCi/ml1.00 X 10-0 giCi/ml2.00 x 10-° iiCi/mnl7.00 X 10-7/Ci/ml5-Minute Exposure1.37 x 10+02 mrem2.00 x 10+°° mrem1.30 x 10+02 mrem7.18 xlO0-1mrem1.75 x 10+01 mremTotal = 287.80 mremDoses from the kryptons and xenons present in the containment building are assessed in much thesame manner as the radioiodines, and the dose contribution from each individual radionuclide mustbe calculated and then added together to arrive at the final noble gas dose. Because the dose fromthe noble gases is only an external dose due to submersion, and because the DACs for theseradionuclides are based on this type of exposure, the individual noble gas doses for 5 minutes incontainment were based on their average concentration in the containment air and thecorresponding DAC.Example calculation of whole body dose due to KrThe DAC can also be defined as 5,000 mrem/2,000 hrs, or 2.5 mremiDAC-hr. Additionally, 5minutes of one DAC-hr is 8.33 x 10.02 DAC-hr.85Kr concentration in containment85Kr DAC (10 CFR 20)Dose Multiplication Factor= 2.69 X 10.07 pxCi/ml= 1.00 x 10.04 pxCi/ml= (85Kr concentration) / (85Kr DAC)= (2.69 X 10-07 g.Ci/ml) / (1.00 x 10.04 pCi/ml)= 0.00269Therefore, a 5 minute whole body exposure from 85Kr 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-°4mremNote: Same calculation is used for the other noble gases listed below.Derived Air Concentration Values and 5-Minute Exposures -Noble GasesRadionuclide85Kr85m~r87Kr88Kr89Kr90Kr133mXe135Xe'39XeDerived Air Concentration1.00 x 10"0 pCi/ml2.00 x 10.05 pCi/ml5.00 x 10.06 pCi/ml2.00 x 10-06 gCi/ml6.00 x 10-°6 pCi/ml6.00 x 10-06 pCi/ml1.00 X 10-04 pCi/ml1.00 x i0-05 pCi/ml9.00 x 10-06 pCi/ml6.00 x 10-06 pCi/mi4.00 x 10-°6 jiCi/ml6.00 X 10-°6 ptCi/ml5-Minute Exposure5.59 x 10"0 mrem1.23 x 10+°a mrem9.85 x 10+°1 mrem3.51 x 10+02 mrem1.01 x 10+02 mrem7.48 x 10+°1 mrem6.18 x 10+°° mrem4.43 x 1° mrem2.21 x 10+°1 mrem1.37 x 10+02 mrem2.72 x 10+02 mrem8.41 x 10+°1 mremTotal =1203.80 mremTo finalize the occupational dose in terms of Total Effective Dose Equivalent (TEDE) for a5-minute exposure in the containment building after target failure, the doses from the radioiodinesand noble gases must be added together, and result in the following values:5-Minute Dose from Radioidines and Noble Gases in the Containment BuildingCommitted Dose Equivalent (Thyroid)Committed Effective Dose Equivalent (Thyroid)Committed Effective Dose Equivalent (Noble Gases)Total Effective Dose Equivalent (Whole Body)287.80 mrem8.63 mrem1203.80 mrem1212.44 mremNote: The addition of Strontium-90 (9°Sr) will increase the above stated TEDE (whole body) by9.15 mrem (<1%).By comparison of the maximum TEDE and Committed Dose Equivalent (CDE) for thoseoccupationally-exposed during target failure to applicable NRC dose limits in 10 CFR 20, the finalvalues are shown to be well within the published regulatory limits and, in fact, lower than 25% ofany occupational limit.Radiation shine through the containment structure was also evaluated when considering accidentconditions and dose consequences to the public and MUIRR staff. Calculation of exposure ratefrom the target failure was performed using the computer program MicroShield 8.02 with a80 of 86 Rectangular Volume -External Dose Point geometry for the representation of the containmentstructure (Attachment 12). MicroShield 8.02 is a product of Grove. Software and is acomprehensive photon/gamma ray shielding and dose assessment program that is widely used byindustry for designing radiation shields.The exposure rate values provided below represents the radiation fields at 1 foot (30.5 cm) from a12-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 theexposure rate values are identical to those used for determining the dose to a worker withincontainment from noble gases. For radioiodine, the total iodine activity of the target was used forthe dose calculations, not the amount that evaporated in 5 minutes. The source term also assumes ahomogenous mixture of nuclides within the containment free volume.Radiation Shine through the Containment BuildingExposure Rate at 1-Foot from Containment Building Wall: 68.87 mrem/hrExposure Rate at Emergency Planning Zone Boundary (150 meters): 0.467 mrem/hrA confirmatory analysis of the accident condition yielding the largest consequence was validatedindependently by the use of the MCNP code. This analysis yielded a result 21% less than theMicroshield method and results provided above.As noted earlier in this analysis, the containment building ventilation system will shut down and thebuilding itself will be isolated from the surrounding areas. Target failure will not cause an increasein pressure inside the reactor containment structure; therefore, any air leakage from the buildingwill occur as a result of normal changes in atmospheric pressure and pressure equilibrium betweenthe inside of the containment structure and the outside atmosphere. It is highly probable that therewill be no pressure differential between the inside of the containment building and the outsideatmosphere, and consequently there will be no air leakage from the building and no radiation doseto members of the public in the unrestricted area. However, to develop what would clearly be aworst-case scenario, this analysis assumes that a barometric pressure change had occurred inconjunction with the target failure. A reasonable assumption would be a pressure change on theorder of 0.7 inches of Hig (25.4 mm of Hg at 60 0C), which would then create a pressure differentialof about 0.33 psig (2.28 kPa above atmosphere) between the inside of the isolated containmentbuilding and the inside of the adjacent laboratory building, which surrounds most of thecontainment structure. Making the conservative assumption that the containment building will leakat the Technical Specification leakage rate limit [10% of the contained volume over a 24-hourperiod from an initial overpressure of 2.0 psig (13.8 kPa above atmosphere)], the air leakage fromthe containment structure in standard cubic feet per minute (scfm) as a function of containmentpressure 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); andCP -= containment pressure (psia).The minimum Technical Specification free volume of the containment building is 225,000-ft3 atstandard temperature and pressure. At an initial overpressure of 2.0 psig (13.8 kPa aboveatmosphere), 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 airvolume to 230,051 scf. The above equation describes the leakage rate that results in this drop ofcontained 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 initialoverpressure condition of 0.33 psig (2.28 kPa above atmosphere), it would take approximately 16.5hours 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 periodwould be about 5.2 scfm.Several factors exist that will mitigate the radiological impact of any air leakage from thecontainment building following target failure. First of all, most leakage pathways fromcontainment discharge into the reactor laboratory building, which surrounds the containmentstructure. Since the laboratory building ventilation system continues to operate during targetfailure, leakage air captured by the ventilation exhaust system is mixed with other building air, andthen 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 bydischarge out the exhaust stack and subsequent atmospheric dispersion, results in extremely lowradionuclide concentrations and very small radiation doses in the unrestricted area. A tabulation ofthese concentrations and doses is given below. These values were calculated following the samemethodology 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 tothe behavior of radioiodine, which has been studied extensively in the containment mock'up facilityat 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% iodinedeposition in the containment building, each cubic meter of air released from containment has aradioiodine concentration that is 25% of each cubic meter within containment building air, then theradioiodine concentrations leaking from the containment structure into the laboratory building, inmicrocuries per milliliter, will be:Example calculation of 1311 released through the exhaust stack:= 13aI activity / (30,500 ft3/min x 16.5 hr x 60 mmn/hr x 28,300 mi/fl3)= 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/ml82 of 86 Note: Same calculation is used for the other radioiodines listed below.Radioiodine Concentrations in Air Leaking from Containment and Exiting the Exhaust Stack1311 -2.46 x 10-o6 gtCi/ml 133I -- 1.17 x 10.05 ptCi/mi 35-- 1.10 x 10.05 jiCi/ml132I -- 5.44 x 10.0 jiCi/ml 14-- 1.33 x 10-°5 gCi/mlExample calculation of 85Kr released through the exhaust stack:= 85Kr activity / (30,500 ft3/min x 16.5 hr x 60 mir/hr x 28,300 ml/ft3)= 1.71 x 10+03 gtCi / 8.55 x 10+11 ml-2.00 x 10.o9 pCi/miNote: Same calculation is used for the other noble gases listed below.Noble Gas Concentrations in Air Leaking from Containment and Exiting the Exhaust Stack8SKr -2.00 x 10.09 pCi/ml 87Kr -1.80 x 10-° pCi/ml 89Kr -3.25 x 10-05 pCi/ml85mJr -8.87 x 10.o6 gCi/ml 88Kr -2.54 x 10.o pCi/ml 9°'Kr -3.21 x 10-o5 jiCi/mI3Xe- 2.21 x 100 gCi/ml l35mXe -7.91 x 10.06 pCi/ml '38Xe -4.38 x 10.o pCi/ml'35Xe -1.59 x 10°5 gtCi/ml '37xe -4.19 x 10-° pCi/ml 139Xe -3.59 x 10°05 pCi/miAssuming, as stated earlier, that (1) the average leakage rate from the containment building is 5.2scfmn, (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 buildingpressure with atmospheric pressure, (3) the flow rate through the facility's ventilation exhaust stackis 30,500 scfm, (4) the reduction in concentration from the point of discharge at the exhauststack to the point of maximum concentration in the unrestricted area is a factor of 292 and (5) thereis no decay of any radioiodines or noble gases, then the following concentrations of radioiodinesand noble gases with their corresponding radiation doses will occur in the unrestricted area. Thevalues listed are for the point of maximum concentration in the unrestricted area assuming auniform, semi-spherical cloud geometry for noble gas submersion and further assuming that themost conservative (worst-case) meteorological conditions exist for the entire 16.5-hour period ofcontainment 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 thatwas used to determine doses inside the containment building, and it was assumed that an individualwas 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 containmentbuilding was leaking.A worst-case scenario dilution factor of 292 for effluent dilution using the Pasquill-Guifford Modelfor atmospheric dilution is used in this analysis. We assume that all offsite (public) dose occursunder these atmospheric conditions at the site of interest, i.e. 760 meters North of MUJRR. In ourcase at 760 meters it occurs only during Stability Class F conditions; which normally only occur11.4% 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. AnEffluent Concentration Limit of 2.0 x 1 0-° pCi/mi is used for the "unlisted" isotopes, which equalsthe DAC/300 when using the DOE Part 835 Default DAC limits of 6.0 x 10-°6 Exposure at1 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 8760hours per year. Thus, the time assumed to be exposed to the effluent concentration limit is a factorof 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 basedon 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'sin Appendix C of Part 835. No credit is taken for transit time from the stack to the receptor pointnor is credit taken for decay inside containment until release. In the case of Kr-89 and Xe-i137 thetransit 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 theDAC value to establish the effluent concentration limit is warranted. This reduction factor of 300 isconsistent, in fact more conservative, than 10 CFR 20 Appendix B, as the NRC calculates EffluentConcentration 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 mremihrI31 concentration =8.42 x 10.09 iiCilml131I effluent concentration limit = 2.00 x 10-'° gCi/ml1311 Conversion Factor = 5.71 x 10.0 mrenm/hrTherefore, 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+°° mremNote: 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 Concentrationand Radiation Doses in the Unrestricted Area -Radio iodineRadionuclide13111321133I13411351Effluent Limit2.00 x 10-10 p.Ci/ml2.00 x 10-° xtCi/ml1.00 x 10-09 pCi/ml6.00 x 10-° gCi/ml6.00 x 10-0 pCi/mlMaximum Concentration18.42 x 10-09 pxCi/ml1.86 x 10-08 pCi/ml4.00 x 10-° pCi/ml4.55 x 10.0 pCi/ml3.78 x 10.08 pCi/mlRadiation Dose3.96 x 10+00 mrem8.78 x 10.02 mrem3.77 x~ 10+00 mrem7.14 x 10-02 mrem5.93 x 10-° mremTotal = 8.48 mremNote 1: Maximum Concentrations are radioiodine and noble gas concentrations leaking from the containment buildingand exiting the exhaust stack reduced by a dilution factor of 292.Effluent Concentration Limits. Concentrations at Point of Maximum Concentrationand Radiation Doses in the Unrestricted Area -Noble GasesRadionuclide85Kr85mIKr57Kr85Kr89Kr9O0r135Xel3smXe139XeEffluent Limit7.00 x 10"07 jiCi/ml1.00 X 10-0 xtCi/ml2.00 x 10-0 pCi/ml9.00 x 10-09 2.00 x 10-08 pCi/ml2.00 x 10-08 ptCi/ml5.00 X 10-07 iiCi/ml7.00 x 10°08 iiCi/ml4.00 x 10-0 ptCi/ml2.00 x 10-0 pCi/ml2.00 x 10.0 iiCi/ml2.00 x 10-08 pCi/mlMaximum Concentration'6.85 x 10-12 pCi/ml3.04 x 10.08 gCi/ml6.17 x 10-08 gtCi/ml8.70 x 10-08 pCi/ml1.11 x 10.07 giCi/ml1.10 x 10.o pCi/mi5.45 x 10-08 pCi/ml2.71 x 10-08 gtCi/ml1.43 x 1 0"07 pCi/ml1.50 X 10.07 ptCi/ml1.23 x 10.07 pCi/mlRadiation Dose9.22 x 10-07 mrem2.86 x 10.02 mrem2.91 x 10.01 mrem9.10 x 10.01 mrem5.24 x 10.01 mrem5.17 x 10-1torero1.43 x 10.02 mrem7.34 x 10.0 mrem6.38 x 10-02 mrem6.76 x 10.01 mrem7.06 x 10.01 mrem5.80 x 10.01 mremTotal = 4.38 mremNote 1: Maximum Concentrations are radioiodine and noble gas concentrations leaking from the containment buildingand 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 dosesfrom the radioiodines and noble gases must be added together, and result in the following values:Dose from Radioidines and Noble Gases in the Unrestricted AreaCommitted Effective Dose Equivalent (Radioiodine)Committed Effective Dose Equivalent (Noble Gases)Total Effective Dose Equivalent (Whole Body)8.48 mrem4.38 mrem12.87 mrem85 of 86 Sunuming the doses from the noble gases and the radioiodines simply substantiates earlierstatements regarding the very low levels in the unrestricted area should a failure of a fueledexperiment occur, and should the containment building leak following such an event. Because thedose values are so low, the dose from the noble gases becomes the dominant value, but the overallTEDE is still only 12.87 mremn, a value far below the applicable 10 CFR 20 regulatory limit for theunrestricted area.

References:

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

'ATTACHMENT 2COPYMODIFICATION RECORDModification NumberModification Title72-7 .. ..... ......,Page 1Storage BasketPage Required DateNumber Page Title Yes No Completed By1234S6Modification Record xSystem Proposal (including a detailed xhazards analysis)Crew Evaluation xSafety Evaluation (OSHA) xSafety Subcommittee Review _Reactor Advisory Committee ReviewABC Review Modification Approved 81011121314Parts RequirementInstallation RecordBlueprints, spare parts, tech manualsPre-op testSOP changesConmpliance checks and PM'sMH cards AervisorYesNo7_iL_I/616 79~Li4) Z.2 --'I T~,C /7 730 teDate'C) 15 73I -p~'-~39 -('-1~vsDaZefByZJAm9,Reactor S ervisorModification CompletedCOPY ATTACHMENT 2Analysis of an Auxiliary Spent Fuel Storage Rackin the MURR PoolThe Missouri University Research Reactor (MURR) has asa part of its pool a deep pit with two storage racks capableof holding sixteen spent fuel elements or two full coreloadings. This is inadequate for the present MURR fuelcycle and with the advent of 10 megawatt operation, thesituation will be even worse. It is proposed that an addi-tional eight element rack be installed between the existingtwo. Figure 1 is a sketch depicting a top view of proposedconfiguration. The existing two racks are hung from thepool wall along with their respective gamma shields. So asnot to stress these supports further, it is proposed that theauxiliary rack have an integral stand to support it 2 feet offthe pool floor and level with the existing racks. The newrack will attach to the present ten element rack by brack~etsthat engage underneath as shown in Figure 1. The rack isessentially self-supporting but this attachment will lendextra stability. The fully loaded rack will weigh approximately300 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 betweenthe elements in the new rack to insure that each element isseparated from every other by boral.t This is well within themaximum Keff limit of 0.8 presented in the MURR license R-103.Thus the fully loaded rack will be far subcritical.

  • ATTACHMENT 2' Figure IProposed Fuel Storage Rack ConfigurationAuxiliary FuelStorage RackS. 5" gammashield(TYP)Existing FuelStorage RacksReactor Pool OutlineScale: t/8"= I" ATTACHMENT 2Spent fuel elements with long operating history emitintense decay gamma radiation which produces heat in theconcrete pool walls when attenuated. To prevent the con-crete from damage due to thermal stresses and excessivetemperatures, gamma shields are placed around the storageracks 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 concretewall from pool water temperature. Figure 1 indicates thatto meet this criterion, the racks as constructed have 3-inchthick gamma shields along the sides and 1.5-inch shields oneach end. The new rack will have similar 1.5-inch thickshields 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 spentelements in the storage rack area, the strongest contributionto the gamma radiation field will be from the eight elementswith the most recent operating history. For example, a spentfuel element with several days decay after its last operationrepresents less than 10% of the decay gammas that an adjacentelement will emit with only two hours decay.Thus it may be concluded that the proposed auxiliary fuelrack may be safely used to extend the MURR spent fuel storagecapabilities by one complete core loading.Caudle JulianReactor Physicist ATTACHMENT 2Safety 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 _______ DisapprovedPlant Safety Coordinator Date-J ATTACHMENT 2............. Reactor Safety SubcommitteeMinutes of Meeting of February 1, 1972Members present: Partain, Jacovitch, Kuntz, Marriott, SlivinskyAlso present: Alger, JulianThe meeting was called to order at 1:35 p.m. by Dr. Partain.The minutes were accepted as read. Mr. Alger reported that thetest annunciator circuit approved at the December 15, 1971 meetinghad not been installed since the reason for prior reactor scramshad been located in a faulty relay. It will be used to locatesites of future scrams. Also, it was reported that the stainlesssteel fuel tank has been installed.Reactor utilization request 191 was discussed by Mr. Algerand Mr. Julian. The committee considered sample cooling, reactivityworth 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 bemade to assure it is in compliance with license limits.2. Calculations of thermal effects be made on spring insample container.3. A temperature monitor be used on outer wall of containerin initial irradiations to verify calculations. Furtherrecommendations will be based on these results.4. A preliminary week-long experiment in a low flux positionafter which the container is opened and examined for damage.5. A~fission product monitoring system for poo1 water beconsidered 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 thereactor poo1 to handle increased spent fuel elements expected when10 MW operation is in effect. These were approved.Meeting adjourned at 2:45 p.m.Robert R. KuntzSecretary ATTACHMENT 2Parts Requirement SheetModification Number Page 8- FDatePurchase Order No._________Ordered from________ _____Purchase Order No.Ordered fromDate*DatePurchase Order No. _________Ordered from________ _____Purchase Order No.Ordered fromDateDateDate Date ATTACHMENT 2Installation RecordModification Number -Page 9-ftDate Description of Work Accomplished .,Percent Completed4.J. 44. I+ 14. 44 *4 I_______ I4. I4 4.4 I9 9*_______ 1~I 4.I 4L 4.4- t4- T4. T

, ATTACHMENT 2BlueprintsSpare PartsTech ManualsModification Number Page 10-PNewRev. of*~~ir1ILn IsO. rr+/-Hn.. iitle vrrniiic virrint rKev. P40. uat~ety Part Description Part No. Purchase Order No. SP No. Date... .._ _,_ _,,_ _,, ,, ,, ,Purchase Order No.Ordered from_JA]Purchase Order No.Ordered fromA/8qDateDatePurchase Order NO.Ordered fromyd h4Purchase Order No.Ordered fromDateDateManual TitleOrdered fromate OrderedDate Rec'dManua No.i1~ I.-4XA' i4.,,= "4. f -4 , , i,,I. 4. 4. 44. 4. 4. 41* I I __

ATTACHMENT 2PROCEDURE FOR Z BASKET MULTIPLICATION MEASUREMENT1. Scan new baskets with source and a detector to insure boralplate composition.2. Install source, two detectors, and thermocouples as directedby reactor physicist.3. Defuel reactor as per refueling procedure. Wait for evaluationof l/M before transfer of each element.4. Affter all elements are transferred, remove all detectors. Afterthe pool level is returned to normal, store the source in thedeep pool (tag rope).,s. ) .A Reactor PhysicistReactor Supervisor UNIVERSITY OF" MISSOURI* ATTACHMENT 2COLUMBIA

  • ROLLA
  • ST. LOUISINTER-DEPARTMENT CORRESPONDENCEit-5February 27, 1973TODon AlgerSUBJECTZ-Basket Subcriticality MeasurementOn February 23, 1973 an experimentthe degree of subcriticality of the newtwenty-four 775-gram elementswas conducted to determineZ-basket configuration ofBefore any fuel was transferred to the new baskets, the plateswere 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 SchiapperReactor Physicistkp ATTACHMENT 2SOP ChangesModiification Number V Page 12- [____.For each change cite volume, section, part, and paragraph. Include a copy ofeach change..S&L,~ , ~-AL -,; I u i- ."2. I I.I tI 1~a +I II _ ___ ___ ___ ___ ___ ___ ___ ___ ___ ___ ___ ___ __....__ II l ATTACHMENT 2MBt CardsModification Number Page 14-RComplete the following data for the system and each major component.Manufacturer __ __ __ __ __ __ _ U. of Mo. No. _ _ _ _ _ _ _ _ _ _ _ _ _Ref. Dwg. and Manual No. 15 , 6<c) 1Specs A ~(ei -A L~~~-Date Incorporated into System LJ) Card No. f~- 5 i OItem __ _ _ _ _ _ _ _ _ _ _ _ _ _ Serial Number ______________Manufacturer U. of Mo. No.__ ____________Ref. Dwg. and Manual No.__________________________SpecsDate Incorporated into System _ ______ Card No. ______________Item _________________ Serial Number _______________Manufacturer U. of Mo. No. ______________Ref. Dwg. and Manual No. __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _SpecsDate Incorporated into System _ ______ Card No. ______________Item ________________ Serial Number _______________Manufacturer U. of Mo. No.__ ____________Ref. Dwg. and Manual No. _________________ ___________SpecsDate Incorporated into System Cr oCard No.

ATTACHMENT 3~JcoPYMODIFICATION RECORDModification Number "I Modification ....... ..=.r ....Page 1IIPage1.2.3.4.5.6.7.8.Page TitleModification RecordSystem ProposalPreop Test ProceduresReactor Safety EvaluationCrew EvaluationSafety Subcommittee ReviewReactor Advisory Committee ReviewAEC ReviewRequiredYes NoxXxxxKxDate,Comp)l et~ed',I-I / (eI sl.s~fModification ApprovedL v/A1DateDate of Completion 76Modification Completed Reactor Form Revised10-31-75 w ° ATTACHMENT 3Mod 76-3 Spent Fuel StorageThe 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 thereactor with the present fuel cycle plus the 120 day decay time per element beforeshipment causes the fuel inventory to exceed this capacity. The criteria fordetermining what the capacity should be is based on projections of inventoriessuch that at least 8 spaces will always be available to defuel the coreProposal:Install a 14 element storage basket located in the fuel storage area behindthe weir. This will be accomplished by the installation of a permanent supportstand located above the existing baskets, which rests on the weir floor. The standis spaced to permit the same access to the existing fuel storage baskets. Thesupport stand will provide the vertical guides for the side lead shields (MURRPrint #1170) and will contain lead shields at the end of each row. The side leadshield will rest on top of the existing stainless steel shields used for thelower storage baskets. The fuel basket cradles in a resting pocket which hasguide pins mounted on the stand for positioning. The basket is then secured to thestand by a threaded bolt at each end.Weight Considerations:The existing fuel baskets and stainless steel shields hang on brackets mountedon the pool liner wall of the spent fuel storage. Each bracket is design ratedto carry a vertical load of 2,000 lbs. The large shield and ten element storagebasket has three (3) brackets for a total capacity of 6,000 lbs. The largestainless steel shield and ten element basket loaded with elements have a totalweight of 1,826 lbs. The added lead shield for the proposed storage will have atotal weight of 1,440 lbs. This totals 3,266 lbs of supported weight resting onthree 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. Theadded 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 loaddesign.The weight distribution is well within tolerance of safety margin for thevertical load support.

  • : ATTACHMENT 3Materials: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 whichhas with it a letter of certification of conformance from the manufacturer thatit contains 35% by weight boron carbide.The lead for the shields conforms to A.S.T.M. designation B29-55 and will bepoured into the shields prior to sealing closed.Construction Considerations:The stand, fuel basket and shields are all welded to insure adequate strengthexcept the two 3/8 inch thick spacers (Part 1) mounted on the back of the supportstand. These are attached with machine screws so the thickness may be adjustedto insure a proper fit. Extension hooks (part 5) were added at the ends of theten element row for the placement of future shields. Mounting holes and guidepins (parts 6 and 7) were also incorporated for future use. Consideration isbeing given to constructing this ten element basket to facilitate transferringelements to necessary locations.Initial Operation:Prior to loading fuel in the basket all boral sections will be scanned witha neutron source and neutron detector.d yws, 1f 1UJR O8wib) 1THE FOLLOWING CRITERIA OUTLINE SAFETY CONSIDERATIONS:

ATTACHMENT 3Criteria: Limit dose rate outside biological shield to levels not exceeding thoseat present.The second level of element storage has lead shields whoseattenuation equals or exceeds the present solid stainless steelshields. In addition, the concrete block construction adjacent tothe current [ basket area is less dense than the poured concretewhich 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 approximately15 feet of water shielding above the second level of spent fuelstorage. Conservatively assuming a decay time of I03 secondsFigure 8 of the MURR design data indicates a dose rate of 0.1 mr/hr pernewly stored per element should be expected. The same figure indicateda per element dose rate of less than 0.1 mr/hr for elements storedfor a period of one week. Thus the total added dose rate due to 8elements just removed from service and 6 elements that had beenstored for one week would be less than 1.4 mr/hr at the surface ofthe 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 0.8Criteria: Sufficient thermal shielding or appreciable water thickness must beS provided so that the temperature rise in the concrete shall not exceed30° F..._.This criteria was addressed in the original design of the spentfuel storage racks (Design Data, Volume I, TM-RKD-62-9). Results indicateda requirement of 2.0 inches of lead to shield 8 adjacent elements justremoved from the reactor (conservatively assumed iO3 seconds decay time).The shields manufactured contain 2 inches of lead. Thus this criteriais satisfied.

"' "ATTACHMENT 3Criteria: Th~e heat contributed to the pool by the added 14 elements awaiti~ng ship-ment shall not cause an appreciable pool temperature increase overperiods when the pool system is secured.For this calculation it is conservatively assumed that none ofthe 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 willcontribute a majority of the decay heat load, initially 20 KW. The12 remaining elements are assumed to have a decay history of only 30days. These 1:2 contribute 15 KW. To simplify calculations it willbe assumed that the decay heat load of the newly removed elements isconstant 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 itis reduced to the heat load level 8.5 KW for 2 elements with one daydecay for the remainder of the weekend period. Thus, the total heatload for the 14 spent fuel elements will be 35 KW for the first 24hours 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 formulaq = MCpAT one 8an determine that the temperature rise over the first24 hours 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 weekendperiod will be less than 450 F since this aT would result if no heatwere transferred to the surroundings. The degree of conservatism inthis result is illustrated by the fact that at present the pool tempera-ture increase over the weekend resulting from 24 elements stored in theZ basket and 8 elements in the core is approximately 150 F.Standard Operating Procedures require that following a shutdownthe pool system shall remain in operation for a minimum of five minutes.Data from pool temperature charts for various times of the year indicatesa maximum temperature of 800 F after system shutdowns. Thus, evenwith the extremely conservative assumptions made, the final temperatureof the pool water will be 140U.F (800.+ 15OF + 45OF) which is .wellbelow the saturation temperature of 2120 F.

ATTACHMENT 3Criteria: Safety in moving fuel.Prior to spent fuel shipping the second level of baskets mustbe moved to the weir area so that elements to be shipped may betransferred to the upper level of baskets. The fuel movement sequenceshall be written so that at any time that the baskets are moved therewill be no more than six elements contained in the baskets. The sixelements shall be secured in the basket, (see design drawings). Sixelements contain insufficient fuel for criticality.

"' ATTACHMENT 3 RTP-l5Revised 8 76PROCEDURE FOR INSTALLATION OF SPENT FUEL STORAGE BASKETI. After completion of shop work and prior to installation in pool, scan eachS element storage box with Pu-Be neutron source and detector to insure presenceof boral, record data.2. Manipulate fuel as per sequence to place the 14 elements awaiting shipmentin the east two rows of the present Z basket (Zll, to Z24)3. Install shields and basket, secure to stand. Install SRM detector and takea 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 asper sequence. A 1/rn plot will be maintained as each element is loaded intothe basket.5. Upon completion of transfers to additional Z storage baskets, return thenon-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 managerfor inclusion in mod package.Caudle ~JulianReactor ManagerDate 6'- ?7

° ATTACHMENT 3SAFETY SUBCOMMITTEEMinutes of Meeting of April 8, 1976Members 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 lastmeeting, expressed desire to see more details of the proceedings in the subcommitteeminutes.3. The subcommittee reviewed the circumstances of the March 2, 1976 abnormal occurrencereport regarding the failure of vent tank level controller 925 B. C. Juliansummarized the situation and answered questions. The subcommittee unanimouslyconcurred with the action taken.4. The subcommittee reviewed the abnormal occurrence report of March 24, 1976 regardingjumpering of the rod run-in functions on regulating blade position. C. Juliandiscussed the cause and corrective action. The subcommittee unanimously approvedof the action taken.5. The subcommittee reviewed Reactor Utilization Request Number 243 submitted byM. Janghorbani of the Environmental Trace Substances Research Center. The sub-committee suggested editorial changes and D. McKown noted that the RUR limitationswere based on actual in-practice experience at the MURR. After discussion, thesubcommittee unanimously recommended approval of the RUR as modified.6. The subcommittee began discussion of proposed modification package 76-3 for theinstallation of additional spent fuel storage in the MURR pool. C. 1Julian,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 chairover to T. Storvick. After questions and explanation, the subcommittee unanimouslyreconimended approval of the design concept and recommended that the projectproceed, providing that any safety related problems or design changes be reportedback to the subcommittee for further review.7. The subcommittee reviewed proposed modification 76-4 for the replacement of poollow level rod run-in switch 910. C. Julian explained that this change is an up-grade of originally installed equipment. The subcommittee unanimously recommendedapproval of the modification.

..ATTACHMENT 3Page twoSafety Subconhmittee MinutesApril 8, 19768. New staff members Charles McKibben, Reactor Operations Engineer and Chester Edwards,Reactor Plant Engineer were introduced to the subcommittee. The meetingwas adjourned at 1610..,. ".."..Prepared By:Caudle JulianSecretaryApproved By:Dr. Walter MeyerChai rmanCJ:Id ATTACHMENT 3REACTOR SAFETY EVALUATIONPage 4..Modification Number________________Does this change involve changes to the Technical Specifications or anunreviewed safety hazard as described in 1.0 C£R, secti~on 50.59A proposed change, test, orexperiment shall be deemed to involve an unreviewedsafety question (i) if the probability of occurrence or the consequences of anaccident or malfunction, of equipment important to safety previ~ously evaluated inthe safety analysis report may be increased; or (ii) if a possibility for an accidentor malfunction of a different type than any evaluated previously in the safety analysisreport may be created: or (iii) if the margin of safety as defined in the basis forany technical specification is reduced.YesSignature !l* #EVALUATION,~d~zA -~ t,~ ~ ~. dA ~a, , , .,.. .. i "?_,,Form Revised 10/31/75 jATTACHMENT 3REACTOR SAFETY ANALYSISPa-ge 4-MODIFICATION NUMBER 7,,7-3Does this change involve a change to the reactor facility as defined in the HazardsSummary 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 3Modification 76-3: Upper 2 spent fuel storageCrew evaluation of this proposal was initiated 4/6/76.No constructive suggestions were forthcoming.Caudle JulianReactor Manager
"ATiTACHMENT 3Eva luationModification Number 5Page 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~xIN ~ ~ w~ti~.4~-46~ALQ 2 /Q?~,1~<-~y~9a~

"' ' ATTACHMENT 3(b p Brooks & Perkins, IncorporatedMateiials Handling Division.* P.O. Box 650. Cadillac, Michigan 40001

  • 616 776-9715March 26, 1976Ref: Certification of ConformanceUniversity of MissouriPurchasing Dept.General Services BuildingColumbia, Mo. 65201Attention: .Chester Edwards1 sheet Boral 1/4" x. 48" x 120", 35% B4CWe hereby certify that the core section of the composite materialcontains 35% by weight of boron carbide.

Reference:

Invoice #78,139Sheet #977Brooks & Perkins, Inc.Charles S. Timmons "-.. "..Quality AssuranceEVERT L HANCOCK,Nlotary 'Public,. Wexford CountyMy Commnision expires October I 4 1 197B,4....:-' ._d". .

ATTACHMENT 3 ae16elementas nunmb~er must be visually confirmed.Step From To Position and E~lement NurniNu-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/_. ____1Y7 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- Ia 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(yX45G3.21775 7"F65 F5 7.'I4-6 ATTACHMENT 3 Date iNote: EaCh.step involves the transfer of a single element.Dungahst 4?2 eelement's number must be. Visually confirme~d.Step From To Position and Element NumbNwr.,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 __ .-

XC~13,2)%-,-V454-i3-E1 234578:, 910"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 3Ji4'ii.il-i):

.1ATTACHMENT 3LL~~.dAC.4~~1, "_ _.. ._ _ ..... _ ----_ _... .. _ _ __-_I .,.?.,.7; .*ii~i .. liii ... 4 , sI,I _________ __________ ___________________________________________________________________

9. 2-c107. -. .. ..~2-3.. ... ....... 4 tc 5... ........... ............. o ................ ....-&--.4 .................. .... ...... .... ..... ...$ ..... .. ... ....... ................ .... ..... .. ... ..... ...-72e4>"7 .Q .JI~, % ...

ATTACHMENT 3..........I-3* 3.... ... .i. .............. ..7/31$ii.. ~4 Q-J .....L... .. ...... 1. ...../4) 132.73 X!0-~Q U~4~P~Q- 1 LoLz(

VATTACHMENT 4MODIFICATION RECORDO RIGINALPage 1M.odification Nlumber. J-" .co'1 Modification .y& 6 -/II/PageNo.-1.*2.3.4.5.6.7.8.Page TitleModification RecordSystem ProposalPreop Test ProceduresReactor Safety EvaluationCrew EvaluationSafety Subcommittee ReviewReactor Advisory Committee ReviewAEC ReviewRequiredYes NoxxXIL_xDateCompl eted_3Q/f7:I0LLZ1LB6Modification AproeDate of Completion 2 # TModification Completed J :_"/Reactor ManagerDateForm Revised10-31-75 ATTACHMENT 4REVISION TO MODIFICATION PACKAGE 76-3INSTALLATION OF *14 ELEMENT SPENT FUEL STORAGE BASKETAs required by the Safety Committee Meeting of April 8, 1976, implementationof 10 element storage basket addition shall be reported back to the Subcommitteefor review. The following report is submitted to meet this requirement.10 Element Z Basket InstallationThe present fuel storage capacity at MURR is 38 elements4 elements in the X and V baskets. The remaining 8 spaces inare required to defuel the core if the situation arises. Anfuel storage capacity is necessitated by:in the Z basket andthe X and V basketsincrease in the1. NRC regulation of having less than 5Kg of unirradiatedfuel in the fuel vault.2. 120 plus days of decay time required per element beforespent fuel shipment.3. Operating schedule and unirradiated fuel inventory hasincreased the number of fuel elements involved in ourfuel cycle.The 10 element basket size is dictated by space available in the Z basketstorage area. Construction and material of the 10 element basket is similarto the previous 14 element basket. The support stand and shielding for the new10 element basket was incorporated in the initial construction for ModificationPackage 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. restingon 3 brackets or 1,200 lbs. per bracket. Each bracket is design rated to carrya vertical load of 2,000 Ibs; therefore, load is 60% of rated vertical loaddesign.Safety ConsiderationsPriora neutronservice,than 0.8r to loading fuel in the basket, all boral sections wisource and neutron detector to verify boral present.a subcritical measurement will be performed to ensurewith 10 elements loaded.II be scanned withWhen placed inthat Keff is less ATTACHMENT 42Decay Heat Build UpDecay heat build up in pool for 3-dayupper Z basket.14 element2 elements retired12 elements, 30 day decayperiod for 24 element versus 14 element24 element2 elements retired22 elements, 30 day decay2 elements12 elements20KW15KW2 elements22 elements20KW27. 5KWEnd of 24 hrs.35KW = 19°Fincrease per1 st dayEnd of 24 hrs.47.5KW = 25.8°Fincrease per1st day2 elements12 elements8.5KW1 5KW23.5KW = 12.8°Fincrease per2nd & 3rd day2 elements 8.5KW22 elements 27.5KW36.0KW = 19.5°Fincrease per2nd & 3rd dayEnd of 3 days450F increase versus66°F increase*Calculations are conservative since they assume no heat is transferred out ofthe pool. The degree of conservatism is illustrated by the fact that during5-day week operations; pool temperature increase over the weekend resulting from24 elements stored in the lower Z basket and 8 elements in the core was approxi-mately 15 F. Maximum pool temperature after shutdow8 from 8ast experience was80 F following shutdown procedures. Thus, 1610F (80 F + 15 F + 66 F) wouldbe maximum pool temperature after 3 days which is well below saturation temperatureof 212 F.Surface Dose RateDose rate increase should be less than 2.Omr/hr at surface of pool water withpool 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 ModificationPackage 76-3 since the design model incorporated a 24 element upper Z basketaddition versus the 14 element basket installed.Submitted byApproved byDave McGinty /Reactor sth/(farl ie McKibbehSReactor ManagerK ATTACHMENT 4RTP-l 5BOctober 3, 1978PROCEDURE FOR INSTALLATION OF UPPER ZFUEL STORAGE BASKET1. Prior to installation for fuel element loading, a scan will be performed on eachelement storage box with Pu-Be-neutron source and detector to insure presenceof 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 SRMdetector and take a series of base line counts. Record dose rate at thistime.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 loadedinto the basket. If Keff from graph reaches 0.8, theprocedure will be stopped and Reactor Manager informed.5. After completion of transfer sequence and verification, that Keffis less than 0.8; (EstabliSh does rate with Z basket storage arearefuel the core according to applicable sequence.6. Compile data generated in steps, 1, 3, and 4. Complete forms arethe Reactor Manager for inclusion in mod package.of assemblycompletely loaded. )to be given toSubmitted byDave McGi ntyReactor Physici stApproved byCharl ie McKi bbenReaCtor Manager ATTACHMENT 40357/0'37I/6/, 5 05.55I,3 qf/.7qo/47g. '.&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'../3.13C/z .... .....S5,/,'4 ,A-I.-, .'7IC m1.0*OI9*.87II .'4Siiiilii' -ll~ nlt rT .lll rlrlrlil 1 l riiilli!!li!l!i!llil ]liil!l!llli!lll i i !!Iii++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 Illlllllllllltl!lll!lilil!lil!, .-+" +...$ .....'+"'++'++'+,, ;++ ,, , l + l Njil l f Hji~ l i..... II,. 1.1! I..I, I.., .*j* 4** lily tr II -TiijlP1IlI.'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~~tTI~j~,:I Ii *~ l*! I Ed i~ Ip~1J~yj 3/4 ~ P ii~. .iv i4 ~ j .t1~~it  ! Ij~ h 41 F I ]~ jhI L ~ *1I II ~ I'1P b *~. [ I IIm,3.1]Iit:ilii!i+' iL + , I ", I,+ I + i l ++, ,I,+LI III i. I.I.IllI'.Iil IN!lil!lltq!li IMI.ttll_ ! lli!!ll!l!llllititltlllltl!!ill!lil! 11 tllhillt!llli.i!ii+ fill j-:i;1 ju- w + ~ ! .i+" ;++ +/-4 JL 14~.IU44PII +/-$2j jj 11......il'.4.. --.-..........-............-.-.........- ,Ij4t+ .,7 I ~-'45 ~, 7 g r~ /0 ,.*1 5J:kI.4 ~'-ATTACHMENT 4REFUELIN'G SEQUENhCENote': Ea&h.step involves the transfez of a single elei~ent. During each step1 the transfs evi, onfiTried.. ,l*Step-Ib31aI-lore Eleiaert Euniber775 F 8o°,aS,aaiPositionIIIaPositionzglTimeIiIiCos~i$.od .(imitniaI7A/f7". I --..... -_ _ , _I a3 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--~ 4F~I C; I/'):,S I IIi I. S.... L --_____ I ..ai .SIIS..... I .. .. II I I ATTACHMENT 4, 6" ,,, ;,-sk,;-of Sevae ;~ A~*'/2O5;U,35S,7SI15*/6q's2-o'7K['0I,13//6zz-203o2j33034,6Ac4~&,i vgc~2~~j4-G-%~~

ATTACHMENT 4REACTOR SAFETY EVALUATIONPage 4 __ __Modification Number 76 -3C 'Does this change involve changes to the Technical Specifications or anunreviewed safety hazard as described in 10 CFR, section 50.59A proposed change, test, or experiment shall be deemed to involve an unreviewedsafety question (i) if the probability of occurrence or the consequences of anaccident or malfunction of equipment important to safety previously evaluated inthe safety analysis report may be increased; or (ii) if a possibility for an accidentor malfunction of a different type than any evaluated previously in the safety analysisreport may be created: or (iii) if the margin of safety as defined in the basis forany technical specification is reduced.Yes_____No ASignature_________EVALUAT IONForm Revised 10/31/75 ATTACHMENT 4REACTOR SAFETY ANALYSISPage 4-MODIFICATION NUMBER 74- 3. Does this change involve a change to the reactor facility as defined in the HazardsSummary 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 4REVISION TO MODIFICATION PACKAGE 76-3INSTALLATION OF 14 ELEMENT SPENT FUEL STORAGE BASKETAs required by the Safety Committee Meeting of April 8, 1976, implementationof 10 element storage basket addition shall be reported back to the Subcommitteefor review. The following report is submitted to meet this requirement.10 Element Z Basket InstallationThe present fuel storage capacity at MURR is 38 elements4 elements in the X and Y baskets. The remaining 8 spaces inare required to defuel the core if the situation arises. Anfuel storage capacity is necessitated by:in the Z basket andthe X and Y basketsincrease in the1. NRC regulation of having less than 5Kg of unirradiatedfuel in the fuel vault.2. 120 plus days of decay time required per element beforespent fuel shipment.3. Operating schedule and unirradiated fuel inventory hasincreased the number of fuel elements involved in ourfuel cycle.The 10 element basket size is dictated by space available in the Z basketstorage area. Construction and material of the 10 element basket is similarto the previous 14 element basket. The support stand and shielding for the new10 element basket was incorporated in the initial construction for ModificationPackage 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. restingon 3 brackets or 1,200 lbs. per bracket. Each bracket is design rated to carrya vertical load of 2,000 Ibs; therefore, load is 60% of rated vertical loaddesign.Safety Considerationsa neservthanPrior to loading fuel in the basket, all boral sections willeutron source and neutron detector to verify boral present.vice, a subcritical measurement will be performed to ensure tln0.8 with 10 elements loaded.be scanned withWhen placed inhat Keff is less ATTACHMENT4 2Decay Heat Build UpDecay heat build up in pool for 3-day period for 24 element versus 14 elementupper Z basket.14 element2 elements retired12 elements, 30 day decay24 el ement2 elements retired22 elements, 30 day decay2 elements12 elementsEnd of 24 hrs.20KW15KW2 elements22 elements20KW27.5KW35KW = 19°Fincrease per1st dayEnd of 24 hrs. 47.5KW = 25.8°Fincrease perIst day2 el ements12 elements8. 5KW1 5KW23.5KW = 12.8°Fincrease per2nd & 3rd day2 elements 8.5KW22 elements 27.5KW36.0KW = 19.5°Fincrease per2nd & 3rd dayEnd of 3 days45° increase versus66° increaseCalculations are conservative sinc~e they assume no heat is transferred out ofthe pool. The degree of conservatism is illustrated by the fact that during5-day week operations; pool temperature increase over the weekend resulting from24 elements stored in the lower Z basket and 8 elements in the core was approxi-mately 15 F. Maximum pool temperature after shutdow8 from 8ast experience was80 F following shutdown procedures. Thus, 161 F (80 F + 15 F + 66 F) wouldbe max~mum pool temperature after 3 days which is well below saturation temperatureof 212 F.Surface Dose RateDose rate increase should be less than 2.0mr/hr at surface of pool water withpool 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 ModificationPackage 76-3 since the design model incorporated a 24 element upper Z basketaddition versus the 14 element basket installed.Submitted byDave McGint'y /1Reactor stApproved byc$narlie ~McKibbefiReactor Manager9-29-78 a.ATTACHMENT 4Crew EvaluationM~odification Number Page 5-All crew members are asked to comment in some manner oni this proposal.Name Remarks10/18/73 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~z3/.... .. ... -.-........ ....... .z .../6.. ... ._o. .. .i. ...... .........3..3 / ... .... ... .............. .- .......

,ATTACHMENT 5Page 1 of 15OkI1GNALRevised: 2/18/86App' djReactor ManagerMODIFICATION RECORDMODIFICATION NO. ___!-,, _Modification Temporary Additional in-pool fuel storage basketsPageN__o.RequiredYES NODateCompletedPage Title123456789Modification RecordSystem ProposalReactor Safety AnalysisReactor Safety EvaluationCrew EvaluationMURR SOP Review CompleteCompliance or P.M. Revision "Parts Requirement SheetPrints, Technical Manual, SpareParts Change Requirementcation Approved: XXXxxx,__ 0 17.-I1llBy(Initials)r3 IModifi(DatffDate of Completion:ModificationCompleted: i~; (PJReactor ManagerItemNo.___REVIEW AND FOLLOW-UP ACTIONRequiredYESS NODate DocumentedCompleted By (Initials) ]_____________________1 Safety Subcommittee Review2 Reactor Advisory Committee Review3 U.S. NRC ReviewII4 MURR Drawings Updated ATTACHMENT 5Page 2of 15 flD1(S NAL eie: /88MODFIATONNO_____ J\~)" ,"App'dtA~Reactor ManagerSYSTEM PROPOSALThe inability of MURR to establish spent fuel shipping capability since the GE-700 cask wasremoved from service in September 1989 has created the need for temporary additional in-pool fuel storage. This modification package documents the evaluations performed to showthat the use of two shipping baskets designed for use in the MHIA cask as temporary in-poolstorage facilities does not present an unreviewed safety question. Each MHIA shipping baskethas twelve fuel element storage positions in a three by four matrix with a boral sheet betweeneach row of four elements(see page 13). These baskets will be attached by brackets to the deeppool"X" and"Y" basket fuel element storage to provide stability and lateral support. Thesebrackets are made of 0.25" aluminium angle(see page 15) and provides a position for the OSbasket if additional in-pool storage is needed.The evaluation performed for each MHIA basket will include a criticality analysis(KENO), aboral plate verification, thermal analysis and 1/M determination when it is first loaded. Aseparate evaluation will be made of the OS basket if used as deep pool storage in conjunctionwith the two MHIA baskets.

ATTACHMENT 5Page 3of 156>f tC \LtRvsd2/86MODIFICATION NO.______ evsd: 2/18/8-Reactor ManagerREACTOR SAFETY ANALYSISDoes this change involve a change to the reactor facility as defined in the Hazards Summaryand its addenda?Yes ___No _X_ Signature: ', ,-_j If YES, make an analysis below and attach a suggested revision to the HSR. If NO, outline thebasis for the decision.The Hazards Summary Report (HSR) describes in-pool fuel storage in three parts of theOriginal HSR. Section 6.4, Spent Fuel Transfer and Storage (p.6-5) and Section 7.1.8, FuelHandling System (p.7-8). describe irradiated fuel storage in the context of radiation doseoutside the biological shield. Section 13.2.11, Refueling Accident provides accident analysisfor irradiated fuel transfers within the pool and states "all storage racks have been designed tobe 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 thereactor facility as defined in HSR. These storage positions meet the dose rate criteria ofSection 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 primaryreactor shield (with 40 days continuous operation at 10 MW and fission product decay time of105 seconds) the dose rate at one foot from the outside of the reactor shield would beapproximately lmr/hr. This is well within the criteria of 2.5mr/hr at one foot from shieldsurface required by HSR.The minimum thickness of magnetite concrete between MHX or MHY baskets to the outsideof the biological shield is five feet. A further margin from the dose rate criteria is provided bythe 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 arrayadjacent to the pool wall and the fadt that MURR fuel cycle produces irradiated elements with ATTACHMENT 5Page 4of 15 ( MO I IC T ON N . ~UA pp'd 2 ~ LReactor Managera 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 thatthese storage positions are safe with regard to criticality to meet the requirements of OriginalHSR Section 13.2.11.Design data volume I section TM-RKD-62-9 Thermal Shielding Requirements for Spent FuelStorage Facilities provides thermal shielding requirements for spent fuel storage facilities.Thermal shielding or appreciable water thickness must be provided around the spent fuelstorage racks to protect the biological shield concrete from damage due to thermal stresses andexcessive temperatures. Thermal shielding requirements are based on radiation heating inthe concrete and resulting temperature conditions within the concrete. The design criterion isthat 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 adecay time less than 106 seconds(11.6 days), all storage positions except 1,8,9,and 10 in MH-Xand 1,8,9 and 2 in the MHY have greater than the minimum water thickness for thermalshielding of Table 4-A of Design Data Volume 1.(See page 13) The thickness requirementspresented in the table are based on a configuration of a row of eight elements stored adjacentto 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 theminimum thickness for water thermal shielding in Table 4-A and will be administrativelylimited to fuel elements with greater than one year of decay(3 x 10 seconds). Elements withthis decay represent fission product activity and hence gamma heating source, about onetwentieth(I /20) of the activity(and gamma heating source) of a fuel element with 106 secondsof decay [I- Huang M.S thesis, 300 days on cycle, 120 days of irradiation, 180 days out of corealternating in and out of core(See page 11 and 12)1

.'ATTACHMENT 5Page 5 of 15 9/-5Revisd: 2/1/8Reactor ManagerREACTOR SAFETY EVALUATIONDoes this change involve changes to the Technical Specifications or an unreviewedsafety hazard as described in 10 CFR, section 50.59?A proposed change, test, or experiment shall be deemed to involve an unreviewedsafety question (i) if the probability of occurrence or the consequences of an accident ormalfunction of equipment important to safety previously evaluated in the safety analysisreport may be increased; or (ii) if a possibility for an accident or malfunction of a different typethan any evaluated previously in the safety analysis report may be created; or (iii) if themargin of safety as defined in the basis for any technical specification is reduced.Yes___ No X Signature EVALUATIONThe safety evaluation of MURR by the Division of Reactor Licensing dated July 27, 1966identified the safety criteria for fuel storage and handling, as providing assurance of nothaving a critical fuel configuration, even with the unlikely mishap that might occur duringfuel handling. The safety evaluation by the Directorate of Licensing dated May 24, 1974supporting the MURR power upgrade to 10 MW did not elaborate further on spent fuelstorage. The most recent amendments to MURR reactor license R-103 dated May 8, 1991states: "There are no specific accidents in this type of research reactor associated with thestorage of spent fuel in accordance with Technical Specifications. The maximum hypotheticalaccident of complete fission product release of four fuel plates is not affected by increasing theamount of stored fuel. Because the fuel will be stored in accordance with Technicalspecifications, accidents previously evaluated are not changed and no new or different kind ofaccident is created. Therefore, staff concludes that the temporary increase in the possessionlimit of U-235 is acceptable."Technical specification 3.8.d states that all fuel elements stored outside the reactor core will bestored in a geometry such that calculated Kef is less than 0.9 under all conditions. This willbe met by first verifying that the two boral plates are installed in each MHIA basket(seeattached results), a computer criticality analysis(KENO) will be performed with the "Y" basketand MHY basket full of new elements, the "X" basket and MHX baskets full of new elementsand then an analysis with all baskets full of elements(see attached). When fuel is loaded ineach basket for the first time a 1/in plot will be performed (see attached). This will assurecompliance with tech. specs and demonstrate that use of these in-pool storage baskets doesnot present an unreviewed safety hazard as defined in 10 CFR 50.59.

ATTACHMENT 5Page 6 of 15MODIFICATION NO.ORIGINALRevised: 2/18/86Reactor ManagerCREW EVALUATIONAll crew members are asked to comment in some manner on this proposal.N am e"q'. A'O'7 Remarks04'-.0 &

'ATTACHMENT 5Page 7 Of 15MODIFICATION NO.?/-3OR/ 31i,,IALRevised: 2/18/86App'd v-Reactor ManagerSTANDARD PERATINGPrOCEDrECaNGEFor each change cite, section, part, and paragraph. Include a copy of each change.SectionPartParagraphI ATTACHMENT 5Page 8 of 15MODIFICATION NO.~iK7COPIGIHALRevised: 2/18/86App'djReactor ManagerCOMPLIANCE CHECK REVISIONS/PREVENTIVE MAINTENANCE REVISIONSAttach a copy of all new or modified compliance checks to this section.Compliance CheckP.M. DescriptionFreq umencDate Incorporated into System]P.M. Number Date Incoroorated into SystemDate Incorporated into System

,,, ATTACHMENT 5Page 9 Of 15MODIFICATION NO.C?!>?0OP G tN ALRevised: 2/18/86Reactor ManagerPARTS REQUIREMENT SHEET_ Parts DescriptionPart No.Purchase Order No. Date ReceivedPurchase Order No._Ordered From_DatePurchase Order No.__Ordered From_DatePurchase Order No. ___Ordered From_____Date _ _ _ _ _ _ _ _Purchase Order No. _____Ordered From _______Date

... ', ATTACHMENT 5Page 10 of 15MODIFICATION NO.Revised: 2/18/86Reactor Manager0ORG;IN AkLBLUE PRINTS -- SPARE PARTS -TECHNICAL MANUALSBLUE PRINTS:Print No. __Print Title New Print2306 MH1A fuel holders XSPARE PARTS:SPart Description Part No. PurRev, of Old PrintN/A'chase Order No.Rev. No.N/AS.P. No.Date Rev.Date Re'd.Purchase Order No.Ordered From ________DatePurchase Order No.Ordered From ________DateTECHNICAL MANUALSManual Title Ordered FromDate Ordered Dt e'.Mna oDate 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 : :-' "0U')-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~§t2ifL7~Jcw1 Aif4-4~- ~4I7i i :Di_::i- ri-lTq-7:71: bt-~--r----... ...... .... :.. -I ... -= i =.- .. -~--i ... .. ---LL .. ....... ..l.. ..I.. ..!: " F..-- : :':l=:= ': :-f= :° I : : i.. : :... ...___ -.-:_____ .___.__"_._.___-_-__._.____ V -, ., : : - . 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 .Fi -7:. :-,.. *- ..-- :*..i.,-i --.: --- :I " : -* :':-'I ". ::I:::: ..:":[. .:: .: -i ' ::._!.:.. .77-:i: 5-!=__i.::,' ::! !.:i- ::-g:-,"t : f.i:L_-:l : ! i i::.t. ;- :-: :::l..4::-.:"" : ii I : ! .! .. !:-T T -t::-:i:L .m T I I -, , i I .i T I -i * , .__:2--: .... ! --"=.!* l:z.:: i! : " ,,'i :* ..:--! i .... t !

ATTACHE76 --w)l3--0J-4}iI--....8__.2°1T-I9-RV0In00o7--311IO$174-ci-JSILLI iii IILI I I i !Ii i I i i I ii I(0 1.n ~.. ~OOId0I0.0V-f e(~a~) 4?~A ee~v~ 4~I42-

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

ORIGINALtr~QFRrlNT VIEIW~ ~B TVT ]3 F- VTF-W

' "' ATTACHMENT 5September 29, 1993Attachment to Modification Package 91-3Storage of Irradiated Fuel in OS Basket (1/30/93)The use of the OS basket for temporary fuel storage adjacent to the "Xv basketwas first implemented in 1979. The current arrangement of the OS basket with theMlix and MHY temporary storage positions not in the pooi does not represent anunreviewed safety question. If the OS basket is used in conjunction with the iVHIYand MIIX baskets an additional evaluation will be performed.Elements stored in OS basket all have greater than 177 days decay. The edge ofOS basket nearest the pooi wall is approximately 12 inches from the wall. Table 4-Aof Design Data, Vol. I, .Thermal Shielding Reauirements for Spent Fuel StorageFacilities, indicates 14"' of water needed for fuel with 106 seconds of decay. Theseelements had 177 days (min) of decay [1.5 x 107 sec] so adequate thermal shielding isavailable. The thermal shielding of the 1/4" thick stainless steel bottom and side ofthe OS basket are not considered, but would further reduce the water thicknessrequired as thermal shield.

ATTACHMENT 6@©[PYAP-RO-l 115Revision IMODIFICATION RECORD:: SHORT FORMFOR: 1) Addenda to existing Modification Records (e.g., modifications of same nature as onespreviously reviewed and approved).2) Significant modification~s to the facility or facility systems that are not described in theHazards Summar~y Report.3) Modifications that require engineering decisions/implementation in a time frame thatprecludes 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 OperatorRequalification Program.The Reactor Safety Subcommittee will review these modifications.Modification Number: 9 1-3, Addendum 1Modification Title: .Replacement of the Existing X. Y. MH-X, and MH-Y Fuel Storaqie Baskets WithNew X and Y BasketsByPage No.2346page TitleModification Record: Short FormModification Description(Why. Short Form is appropriate)Hazards Summary Report EvaluationReactor Safety EvaluationOP, PM, CP, and Print EvaluationSpare Parts RequirementsRequiredYes Noox __ _DateCompletedXXXX'Cx___-_ ByiiasC- 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: erDate:________Date: 2 Date:________Modification Completed:(Reactor Manager)Attachment 8.1]

ATTACHMENT 6AP-RO-] 115Revision IModification Number: 91-3. Addendum 1MODIFICATION DESCRIPTIONProvide a concise description of the system change. InclUde any proposed PRE-OPERATIONAL TESTS required for this change. (If additional pages are necessary, insert afterthis 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 locationscan 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 tothe following specifications:(a) A geometr~y such that the calculated Keff is less than 0.9 under all conditions of moderation andirrespective 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, orinterchange 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 MURRDesign Data, Volume I. Thermal shielding or appreciable water thickness must be provided around thespent fuel storage baskets to protect the magnetite concrete from damage due to thermal stresses andexcessive temperatures. The thermal shielding requirements are based radiation heating in themagnetite concrete and the resulting conditions within the concrete. The design criterion employed is thatthe 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) new20 element fuel baskets that will be designated X and Y. The new X basket will replace the old X and MH-Xbaskets and the new Y basket will replace the old Y and MH-Y baskets. Each new basketwill haveessentially 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 Xand Y baskets that will provide a storage location for either the OS basket or a Be reflector ring. A separateevaluation 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 anexisting, previously reviewed and approved Modification Record (9 1-3), "Temporary Additional In-Pool Fuel.Storage.Baskets."Attachment 8.1]

V 1ATTACHMENT 6AP-RO-1 15Revision IModification Number: 9 1-3, Addendum 1MODIFICATION DESCRIPTION (con't)The MH--X and M-H-Y baskets were installed in 1991 as additional temporary fuel storage locationsduring a period when the facility was unable to ship. fuel .because twvo spent fuel shipping casks thatwere 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. TheMI-I-X and MH-Y baskets were designed and built for the MH IA shipping cask and were not intendedfor the everyday use that they have endured at MURR. Over the years, some of the boral andaluminum plates have swelled or warped making certain storage locations unusable. To ensure thatwe maintain m~axinmum fuel storage capability during periods of shipment uncertainties, the newbaskets were designed and constructed for everyday use, similar to that of the original X, Y, and Zstorage baskets. Additionally, the newly designed baskets will increase storage capacity from 36 to 40at these locations.Each design specification listed on page 2 will be addressed individually ini the applicable sections ofthis Modification Record. Specification (a) will be addressed in the Hazards Summary Report andReactor Safety Evaluation sections. Specification (b) will be addr'essed in the Reactor SafetyEvaluation section. Specification (c) will be addressed in the Hazards Sununary Report Evaluationsection. 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 ReportEvaluation 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 interchangeof fuel elements." The closest fuel storage location from the new baskets to the reactor pool wall isabout 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 thisModification 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 Zstorage 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 ofspecification (a). The materials of construction meet the requirements of specification (e), whichstates, "Fabrication from materials compatible with the fuel elements." Included in this ModificationRecord are the design and construction prints for the new baskets (MURR Drawing No. 2640). Alsoattached 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 forfuture reference.Neutron radiography of eight randomly selected boral sheets that was performed at the University ofCalifornia-Davis reactor indicated even dispersion of boron in the plates. These radiographs will also.,.,. ..*be i~naintained in Document Control.2aAttachment 8.1 ATTACHMENT 6AP-RO-l 115Revision 1Modification Number: 91-3, Addendum 1HAZARDS SUMVMARY REPORT EVALUATIONDoes this change involve a modification to the reactor facility as defined in the Hazards SummaryReport?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, outlinethe basis for the decision.This modification does not involve a change to the reactor facility as defined in the Hazards Summary Reportand 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. Shieldrequirements for fuel storage in the pool are calculated to meet the dose rate criteria of the bulk shieldinglisted 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 primaryreactor 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 theoutside Surface of the biological shield is five (5) feet. Additional design features that are more conservativethan those assumed in Section 6.4 include: (1) the closest section of the new fuel baskets is locatedapproximately 12-inches from the reactor pool wall (tapered section) and not immediately adjacent, (2) thenew baskets are in a configuration less than an eight element array, and (3) the current MURR fuel cycleresults irn irradiated elements with a much lower activity than the design basis fuel cycle of forty days ofcontinuous 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 /> willhave greater 1E6 seconds of decay (11.6 days). This storage time requirement will be administrativelycontrolled 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 theMonte 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 valueestimated by IMCNP for the new configuration is 0.635, with a standard deviation of 0.002 -well belowTechnical Specification 3.8.d limit of 0.9. Using the most conservative approach and assumptions, thebaskets were modeled using twenty (20) "fresh" 775 gram U-235 fuel elements -a far greater number ofelements than what we are allowed to possess under our current inventory license limits. Additionally, thevalue of boral used to model the baskets was 0.0624 grams of B-10 atems/cm2. None of the boral sheetsthat'were used in the construction of the baskets had a value less than 0.0709 gms/cm2, and the averagevalue of all sheets was 0.0740 gms/cm2. A I/M criticality determination will also be made upon installation ofthe baskets to verify the results of the MCNP modeling. A Keff value of 0.635 easily meets the requirementsof specification (a).Attachment 8.1 I ' IATTACHMENT 6AP-RO-1 15Revision 1Modification Number: 91-3. Addendum 1HAZARDS 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 shieldfingthicknesses .for spent fuel storage. Thermal shielding or appreciablk Water thickness must be providedaround 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 radiationheating in the magnetite concrete and the resulting conditions within the concrete. The designcriterion 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 fuelelements wiflh a decay time of 1E6 seconds (11.6 days) require a minimum of 14-inches of themtlawater shielding. The thickn~ess requirements presented in this table are based on a configuration of arow of eight elements stored adjacent to the biological shield. Page 5 of TM-RI-ID-62-9 also statesthat "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 shieldingwith 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 storedunless they have greater than 3E7 seconds (one year) of d~ecay. Fuel elements with this decay timehave a fission product activity, and hence gamma heating source, of approximately 1/20 of the activityof a fuel element 1E6 seconds of decay: This number was obtained from J. Huang's MasterThesis, 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 depictfuel elenment decay are included in Modification Record 91-3.3aAttachment 8.1 4 4ATTACHMENT 6AIP-RO-1 15Revision 1Modification Number: 9 1-3, Addendum 1REACTOR SAFETY EVALUATIONDoes this change involve a revision(s) to the Technical Specifications or a safety hazard as describedin 10 CFR 50:597? .NOTE: A licensee may make changes to the facility as described in the I-SR without obtaining a license amaendmentonly 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 inthe HSR;2. More than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important tosafety 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 safetypreviously 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 anypreviously 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 orthe safety analyses.Yes: ___No: ___ S igna tur e: _________-, ___________ Date: 2 If YES, the change must be performed igaLong Form Modification Record. If NO, outline thebasis forthe decision.This modification does not involve a ch'ange to the Technical Specifications or a safety hazard as describedin 10 CFR 50.59. A.50.59 Screen is attached and shows that the proposed activity does not have thepotential 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.dand 3.8.e.Technical Specification 3.8.d states that "All fuel elements or fueled devices outside the reactor core shall bestored in a geometry such that the calculated Keff is less than 0.9 under all conditions of moderation." Thebasis for this Specification states that this limit is conservative and assures safe fuel storage. The MCNPmodel was used to calculate a Keff value of 0.635 for one fuel basket fully loaded with twenty (20) "fresh" 775gram U-235 fuel elements. This predicted value is well below the Technical Specification limit of 0.9. Thisvalue 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 permitsufficient 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 withregard to natural convection cooling. This satisfies the requirements of specification (b) stated in theModification Description.*Attachment 8.1 ATTACHMENT 6AP-RO-1 15Revision 1Modification Number: 9 1-3. Addendum IREACTOR SAFETY EVALUATION (con't)Furthermore, the Safety Evaluation (SE) performed by the Test & Power Reactor Safety Branch of theDivision of Reactor Licensing, documented by letter dated July 27, 1966, was in response to therequest by the University of Missouri to operate the MURR at a power level of 5 MW. The SEidentified the safety criteria for fuel storage and handling, as providing assurance of not having acritical fuel con'figuration, even with th~e unlikely mishap that might occur during fuel handling. TheSE performed by the Directorate of Licensing, documented by letter dated May 24, 1974, supportedMURR's request to operate at the higher power level of 10 MW. This SE did not elaborate any furtheron 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 storageof spent fuel in accordance with the Technical Specifications."4aAttachment 8.1'I 1[ I *ATTACHMENT 6AP-RO- l 15Revision 1Modification Number: 91-3. Addendum 1OPERATING, PREVENTATIVE MAINTENANCE, AND COMPLIANCE PROCEDURE,AND PRINT EVALUATIONDoes this charnge require a revision(s) to any 0pcrating, Preventative Maintenance, o~r ComplianceProcedure, 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 ComplianceProcedure. 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 theseprocedures and form are listed below. New prints associated with the design and construction of the new fuelstorage 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 bestored 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, andY-11, 16, 17, 18, 19, 202. 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 bestored 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, andY-11, 16, 17, 18, 19,202. 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 slorageconfigurations.Attachmuent 8.1 11 tATTACHMENT 6AP-RO-1 115Revision IModification Number: 91-3, Addendum 1SPARE PARTS REQUIREMENTS EVALUATIONDoes 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.6Attachment 8.1 ATTACHMENT 650.59 SCREENAP-RR-003Revision 1Page 1 of 2Activity Screening Number: 0e/- /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 tworedesigned 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? "YEIf this question is answered yes, do not__ continue with this procedure. Identify and report the concern tothe Reactor Manager.50.59 Screening Questions:I. Does the proposed activity involvze a change to an SSC that adversely affects a design functiondescribed in the HSR?2. Does the proposed activity involve a change to aprocedure that adversely affects how HSRdescribed 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 NOYES NOYES, No4. Does the proposed activity involve a test or experiment not described in the HSR, where an SSC isutilized or controlled in a manner that is outside the reference bounds of the design for that SSC or YESis 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. ...INO/NOA LicenseIf 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. 3NOTE: 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 whererelevant information was found. Include section/Ipage numbers. Use page 2 of this fo omrtyouzra ements.SPrint Name. ."______e_-__-__DatPreparer: Edward L. Murphy kA bj //Reactor Manager: Les Foyto ) /-/- Attachment 1 A T A T~ ITENI tAP -RR-003R~evision 150.59 SCREEN (Cont.)Activity Screening Number: 0 O Y- Page 2of 2Title: REDSIGNED DEEP POOL FUEL STORAGE BASKETSIf t~he conclusion of the five.(5) Screening QUestions is that a 50.59 EvaIluation is no__t required, providejustification 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 theHSR?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 animproved manner, the same traction as those currently installed. No other systems or components will beby this modification .. ..2. Does the proposed activity involve a change tO a procedure that adversely affects how HrsR described SSC designfunctions 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 HSde.scribed RRC"desi g funnctions ,re. perfo rmned. cnntrolled or- tested3. Does the proposed activity involve revising or replacing an described evaluation methodology that is used inestablishing the design bases or used in the safety analyses?ino The _modi ca'tion to the deep pool fuel prnoro~ed designed using establishedT4qTRdescie~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 controlledin a manner that is outside tlie reference bounds of the design for that SSC, or-is inconsistent with analyses ordescriptions presented in the HSR?No. The redesigned deep pool fuel storage baskets are functionally and operationally the same as thosecurrently installed, and will be used and controlled only in a manner within design boundaries. All testsrequired for the proposed change are covered by, and are consistent with, all analyses and descriptionspresented in the tHSR.List the documents (HISR, Technjc?,l ...p e..[i~.c~atiopns, and other Licensing.-.asis..documents) reviewed whererelevant information was found. [Ihclude section / page numbers. ]1HSR Section 6.4 "Spent Fuel Transfer and Storage". HSR Section7..1 "Fuel Handling System". HSR Section13.2.1t1 "Refueling Accident". Technical Specification 3.8.d "Fuel Element Storage Geobmetry", TechnicalSpecification 3.8.e "Cooling Requirements for Fuel Element Storage". OP-RO-250 'Fuel Handling". RP-RO-100 "Fuel Movement"1 ATTACHMENT 6From: Das Kutikkad ..To: Les Foyto, Acting Reactor manager, MURRDate: June 05, 2003Re: Results of Calculations Performed to Estimate the Keff of the New Deep-Pool" Fuel Storage Bask~etCalculations were performed to estimate the criticality of the newly designed deep pool fuelstorage basket slated to replace the X, MHX, Y and MRY baskets. The model used and the resultsobtained are surmmarized in this memo.For the purpose of simplicity, only one of the new 20-element basket (on one side of the pool) wasmodeled. A drawing of the new basket is attached to this report. One such basket is expected toreplace the combined X & MHX or the combined Y & MHY storage locations. Since the two sidesof the pool are fairly decoupled neutronically (especially with the amount of boron in the storagebaskets), this modeling should be adequate to establish the safe storage requirement specified inthe Tech :Specs.Monte Carlo simulation program MCNP was used to model the new fuel storage basket and toestimate the criticality. Several conservative assumptions were used in the modeling such as usingall fresh fuel elements (no burn up credit taken) and using a reduced thickness for boral in theoutermost 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 usedis approximately 35 w% of B4C in boral (rest Al). For the new bask~et, we purchased boral that hasless boron content. The boral used has 0.0624 grams of B-10 atoms/cm2. For a boral sheet of0.265" thick (approx 0.67 cm), this translates to a B4C value of roughly 24 w%. The boron used isnatural and not enriched in B-10. The dimensions of the basket and the wall thickness are shownthe attached drawing.The Keff value (hie MCNP for this fuel stora ge coii~figiiration (loaded with fresh 775 gU235 fuel elements) was 0.635 with a standard deviation of 0.002. This result shows th~at it is safeto store fuel in the new basket with th~e predicted Keff well below the Tech Spec limit of 0.9.

IIIIUBII19.365II II UB11HB!I/

I IJYCF('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"long4 24 smallboral 0.265" 3oB4C boral 4.4375"x 30" J3L1-c C52AL sheet2 1 / :!:3003H 14 Alum. 19.615" x 33.25" x 1/8"6 48 Aluminum stock 1-3/8" Aluminum stock7 10 Aluminum stock 1/4" x24.14" x 1-3/8" Aluminum stock2314,)ir,, ,. .." ",,.i .... >'"" " "* :K.v. ". E',II /irdl ii ii ii ii

  • ATTACHMENT 6'rodelling of the new 20-element deep pool fuel storage basket6c this first run is a case with just the new 20-element basketc modelled (as a replacement for the existing mhy and y baskets).c subsequent runs will add the old beryllium next to this storagec basket (in place where os basket was before) to see its effect.c the core is not modelled in this case, so the storage basketc -is a stand alone basket filled with fresh fuel elements.c- " " °c a single fuel el'ement is defined and the "repeated structure" featurec of mcnp is used to construct the storage positions (bins).c some conservatiSm is used during the initial runs. some of these willc he removed during later runs if the keff is found to be unacceptablec (i.e, >0.9) -some of the conservative assumptions are listed below:cc 1) all fresh fuel considered -i.e., no burnup ctedit takenc 2) less boral thickness for the outermost layers.cc individual "bins" of the new basket are described in an auxiliaryc coordinate system. the origin of this auxiliary coordinate systemc is at the center of individual bins. These are then tranformedc into the main system centered at one corner of the basket. all thec bins are filled with the same 'universe" (i.e., one fresh fuel-c element plus water, aluminum and boral surrounding the fuel).cc ** histories tracked = 100,000 for this case ***c1 1 -1.0 (-l40:-146:150:144:-148:149) 130 -151 132 -153 154 -135imp:n=l $ water surrounding the new basket (approx 30 cm thick)S the following four cells are created since mcnp doesn't like toc complicate any one cell too much. to avoid that problem, the newc 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 onec basket (thereby complicating that one cell too much).c2 2 -2.7 140 -150 146 -141 148 -149 #20 #21 #22 #23 #24imp:n=l $ basket that contins t "bins" along x-axis3 2 -2.7 140 -150 141 -142 148 -149 #25 #26 #27 #28 #29imp:n=l $ basket that contins 5 "bins" along x-axis4 2 -2.7 140 -150 142 -143 148 -149 #30 #31 #32 #33 #34imp tn=l ** basket that cont-ins-.5 bins" along x-axis 5 2 -2.7 "--150 143 -144 ...48 ...1349 #35 #36 #37 #38 #39imp:n=l $ basket that contins 5 "bins" along x-axisc6 0 -130:-132:151:153:-154:135 imp:n=0 $ outside worldc7 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 bins9 2 -2.7 209 -205 206 -207 u=1 imp:n=1 $ al of the bins10 2 -2.7 208 -209 206 -210 u=1 imp:n=l $ al of the bins11 2 -2.7 208 -209 211 -207 u=l imp:n=l $ al of the binscc although i~nfinite in dimension, banal-..thickness will be limitedc by the'.diiuensions;.of the cell isrille'd with this "universe i".c12 0 208 -209 210 -211 u=1 imp:n=l fill=2 (-11.00 0 0)c20 0 200 -201 202 -203 148 -149 imp:n=l trcl=20 fill=1i above is the definition of a single "bin" that is repeated 20 timesc iATTACHMENT 62124252627282930313233343536373839clikelikelikelikelikelikelikelikelikelikelikelikelikelikelikelikelikelikelike20202020202020202020202020202020202020butbutbutbutbutbutbutbutbutbutbutbutbutbutbutbutbutbutbuttrcl=21trcl=22trcl=2 3trcl=24trcl=25trcl=2 6trcl=27trcl=2 8trcl=30trcl=3 1trcl=32trcl=33trcl=34trcl=35trcl=36trcl=37trcl=38trcl=39c description-of the fuel plates of a single element starts from herec this is the "universe-2" that fills the new storage basket bins.c1011021031041051061.07110283284285286287288289290291c292293294295ccC296297c-2.7 2.7 2.7 2.7 2.7 3.88 2.7 1.00 2.7 2.7 2.7 -94-1.00 -95-2.7 -96-2.7 -97-2.7 -97-2.7 -98-2.7 -98-3.88 -97-2*7. -98-1:.0 -98-1. 0 -98+0.0803 -98+0.0803 -983443345678-93949596969595969733331021021021011041,02102104102102102104102102102101104102102103106105108-101 -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...12.5imp :n=!imp :n1limp :n=1imp :n=1imp:limp:n=1imp :nfl1imnp:n=limp:n=limp: n~lamp :nfl1im~p~n=limp:n=1imp:n=1imp: n=limp:n=limp:n=1imp:n=1imp:n=limp:n=1imp:n=limp:n=1imp: n 1u=2u= 2u=2u=2U=2U=2U=2u= 2U= 2U=2U=2U=2U=2U=2U=2U=2U=2U=2U=2u= 2U= 2U=2$ p1$ p1$ p1$ p1$ p1$ p1$ p1$ p1$ p1$ p1$ p1$ p1$ pl$ p1$ p1$ p1$ p1$ p1$ p11111111222232424242424242424cladcladcladclad on topclad on hotfuelcladwg cladcladcladSwgScladScladScladSclad on topSclad on botSfuellclad4-k-105-104-107-106-124-124.-124-124125125125125$$$$fuel top waterfuel hot waterfuel top hangerfuel hot hangerside plates of the element are described next22-2 .7-2.7-98 3 108 -107 -122 124-98 3 108 -107 -125 123imp:n=l u~=2imp:n=1 u=2$ side plate noAl$ side plate no.2c the water surrounding the f-uel described next, this willc become a. fini~e amount of watet 'this "universe 2" (singlec fresh fuel element plus the water surrounding it) is filled in cellsc representing the "bins" of the new storage basket.c2981 -1.0 107:98:-123:-3:122:-108 u=2 imp:n=l $water surrounding fuelc * *
  • end of cell definitions * *
  • need the following blank line.

ATTACHMENT 6Cc34595969798Cc101.102103104105106107108cC120121c122,Q3'£24125126127cc1301 31132133134135c1401411421.43144145146147148149c150151152153.54cthe following surfaces define the single fuel element (origin of thiscoordinate system different from the main one and from the auxiliarycoordinate system # 1 that defines the bins of *the new basket).czczczczcz"czcz99i00pzpzpzpzpzpzpzpz7.036 $ plate 17.074 ,7.12514.630 $ plate14.66814.71914.757cz 14.986.pz 030.48-30.4832.385-32.38538.10-38.1041,275-41.27524p -0.4142 1 0p 0.4142 1 0PppppppxpxPYpypz.pzpxpypyPYpypxpypypzp7pxpxPYPYpz-0.41420.4142-0.41420.4142-0.41420.4142-30.068.10-30.080.80-80.080.00.013 .0026.0039.0052 .0038.100.050. 80-42.2542 .2564. 5094. 5055.5082.00-50.001"11110000000.0 $0.0 .$-0 .05500.0550-0.46740.4674-0 .65980.6598$$$$$$partpartpartpartpartpartofofofofofof+2'2 .5-22.5+22 .5-22.5+22.5-22 :522.5 degree plane-22.5 degree planedegreedegreedegreedegreedegreedegreesidesidesidesidesidesideplateplateplateplateplateplate ATTACHMVLENT 6c the following surfaces define the auxiliary coordinate system thatdefines the individual bins of the new basket, its origin is at thec" center of each bin, a coordinate transformation then connectsc this to the main coordinate system -which is centred at one of thec corners of the new basket.c2O00201202203204205206207208209210211cpxpxpypypxpxpyPYpxpxPYPY-6.386.38z.6 386.38-6.076 .07-6.076 .07-5.755.75-5 .755.75Rodeccckcodecmlrmt1m2m3c,cccCccm6mt6rotm9cctr2Otr2 1tr2 2tr2 3tr2 4t r2 5tr26tr2 7tr2 8tr29tr30tr31tr3 2r3 3cr34nimp:n 1 18r2000 0.8 51001.5Cc .6667lwtr. 0113027.50c 113027.50c *-.60092235.50c -.37292238.50c -.028mn4 4009.50cmt4 be. Olm5 6012.50c1001. 50c8016.5Cc13027.5Ccmut5 grph. Oi1001.5Cc 0.4170lwtr. Olin8 -.. 26000.5Cc5010.50c -0.0356000.50c -0.0520.01 197r55 3000 08016.50c .33331-.85-.002-.016-.1328016 .50c0.20813027.50c 0.37500.24.,... ,000..50lc 0.2 28000.50c 0.15011.56c -0.15013027.50c -0.763 $ boral with 24 w% b4c6.4019.2032.0044.8057.606.4019.2032.0044.8057.606.4019.2032.0044.8057.606.50 0.06.50 0.06.50 0.06.50 0.06.50 0.019.50 0.019.50 0.01.9.50 0.0 ..."19.50 0.0"19.50 0.032.50 0.032.50 0.032.50 0.032.5C 0.032.50 0.0 tr35tr38tr39Cphys : nprintctinecut :fl1prdinpksrcATTACHMENT 66.40 45.50 0.019.20 45.50 0.032.00 45.50 0.044.80 45.50 0.057.60 45.50 0.0S15.0 0.0 $cross Sections above 15.0 mev will be expunged40 50 60 905000.j o .o -0.5j 2010.0 4.0 0.08.0 16.5 0.03.0 30.0 0.03.0 3.50 0.0110 120 126-0.118.021.017 .04.002.020.933 .02 .000.00.00.'00.034.030.032.0-3.504.518.029.0-3 .500.00.. 00.00.055.047 .059.01.508.022.034.0-1i.000.00.00.00.0 S IATTACHMENT 6OAAR CAR GO SYSTEMSa division of AAR Manufacturing, Inc..Jeff Moore, Sr. Manager -Nuclear Products12633 lnkster Road, Livonia, MI 48150-2272 USAPhone: (734) 522.2000 Direct: (734) 466-8110FAX: (734) 622-2240"email: jmoore@aarcorp.comI. CUSTOMER:A. NAME:B, REQUEST DATE:II. DATESA. CURRENT DATE:B. QUOTATION VALID FOR:III. CONTACTSA. AAR CONTACTI. NAME:2. TITLE:3; PHONE:4. FACSIMILE:B. CUSTOMER CONTACT1. NAMvE:2 PHONE:3. FACSIMILE:IV. sPECIFICATION AND PRICINGUniversity of MissouriApril 8, 2003April 30, 200390 daysJeff MooreSr. Manager, Nuclear Products(734) 522-2000 x 8110(734) 522-2240.Mr. Jeff Attebery1-573-882-52691-573-882-6360+ Shipping to Univ. of Missouri 4i%V. DELIVERIES TO COMMENCE:60 days AROVI. TERMSA.B.DELIVERY POINT:PAYMENT:FOB University of MissouriNet 30 da~isVII. SPECIAL INSTRUCTIONS None ATTACHMENT 6AAR? CARGO SYSTEMS0A division of AAR Manufacturing Group, INC.CERTIFICATE OF COMPLIANCECUSTOMER: University of MissouriQTY. SHIPPED: 68 pcs.DATE OF SHIPMENT: July 18, 2003CUSTOMER P.O. NUMBER: COO000009743AAR CARGO SYSTEMS SALES ORDER NUMBER: 5053667This is to certify' that the material supplied hereunder has been finspected and tested inaccordance with AAR-1 1002 QAP, Revision 23 dated November 7, 2002, and A.AR-10012QAP, Revision 18 dated April 9, 2003, and Nuclear Quality Program Manual, Revision 29and meets the requirements of the. purchase order. The Code of Federal Regulations10OCFR5O Appendix B and 10OCFR2 1 are applicable to the material on this order.SIGNATURE:TITLE:DATE:Phill PusiloLab ManagerJuly 18. 2003Appendix CAAR-1 0012 QAPPage 1 of 1...systems, components & more12633 Inkster Road Livonia Michigan 48150-2272 USATelephone 1-734-522-2000 Faxc 1-734-522-2240 ATTACHMENT 6AA R CA RGO0 S YSTEMSA div'ision of AAR Manufacturing Group, INC.BORAL DATA PACKAGE RECORD CHECKLISTSPECIFICATION: AAR-10012 OAP. REV. 17 D OCUMENTCHECKEDBYDATE"Record ChecklistCertificate of ComplianceInspection Data SheetsMaterial CertificationsJP/KEJP/KiEJP/KEJP/KEJP/KEJPIKEJP/KE7-18-037-1 8-037-12-037-1 8-037-18-037-18-037-18-03-BoraI Summary Report,Boxing ListCalibrated Equipment Data SheetREVIEWED BY:TITLE:DATE:Phill PusiloLab ManaaerJuly 18, 2003APPENDIX DAAR-10012 QAIPPAGE 1 OF I..systems, components & more12633 Inkster Road Livonia Michigan 48150-2272 USATelephone 1-734-522-2000 Fax 1-734-522-2240 tATTACHMENT 6Boral Summary Report (Pass)Job Name: University5O 5053667Serial NumberWM010013-3A 'WM010014-IBwM010015-3A.WM010016-IAWM010017-2AY(MO1001 8-2AYM0 10019-1BYM010020-8BYM010O21 -SBYMI f00022-SBYM010023 ,8AYM01]0024 -8 BLot NumberM-21 SM-2 ISM-2 I8M-21814-218M4-220M-220M-220M-220M-220M-220M-22010B gmns/em20.07400.07210,07090.07660.07260.0754,0.07580.07310.07480.07500.07330.0742Density,2.57312.55072.54742.54842.54272.65822.56322.57812.57282.56232.58472,5873Reviewed By: Phill PusiloTitle: Lab ManagerDate: 8/11/2003Appendix-AAAR10012QAPPagc: 1Ptss ATTACHMENT 6MATERIAL TRACEABILETY bY B ORAL SERIAL NUMBERS.0. # 50536.67 University of Missouri--..~ k-.- WM010013 through WM010017 M-218YM010018 through YM010024 M-220WM010013 through YM010024 "AL03-03WM010013 through YM0 10024 "3-045-C ATTACHMENT 7Volume of the Primary Coolant SystemIn-Pool Portion Mechanical Equipment Room 114 PortionScin Area Length Volume Scin Area Length VolumeScin (ft2) (ft) (ft3) Scin (ft2) (ft) (ft3)135(5) 0.7773 3.828 2.976 133(2-3) 0.7773 10.194 7.924135(6) 0.7773 3.708 2.882 135(1) 0.7773 2.000 1.555135(7) 0.7773 3.708 2.882 0.7773 22.374 17.391137 0.7773 3.250 2.526 133(7-5) 0.7773 22.374 17.391139 0.7773 2.500 1.943 133(4-3) 0.7773 14.290 11.108501 0.6048 5.937 3.591 133(2-1) 0.7773 15.584 12.113575 0.6048 2.269 1.372 132 0.6948 6.000 4.169100(2) 0.7773 4.917 3.822 131(3-2) 0.6948 4.969 3.452100(3) 0.7773 4.917 3.822 115(3-2) 0.6948 14.968 10.400101 0.7773 1.000 0.777 115(1) 0.4948 2.000 0.990102(1) 0.7773 1.000 0.777 111(7) 0.6948 4.189 2.911102(2) 0.7773 3.806 2.958 111(6) 0.6948 6.667 4.632102(3) 0.7773 3.806 2.958 111(2-5) 0.6945 16.264 11.295102(4) 0.7773 3.806 2.958 111(1) 0.6948 2.167 1.506102(5) 0.7773 5.097 3.962 105(9) 0.7773 2.167 1.684401(1) 0.2006 3.975 0.797 105(7-8) 0.7773 17.312 13.457401(2) 0.2006 3.975 0.797 105(5-6) 0.7773 15.542 12.081405(1) 3.2150 0.500 1.608 105(1-4) 0.7773 31.832 24.743405(2) 0.1389 4.708 0.654 102(7) 0.7773 2.000 1.555405(3) 0.2006 9.163 1.838 102(5-6) 0.7773 10.194 7.924460 1.3960 4.242 5.922 Total Piping Volume (ft3) 223.860406 0.7773 2.500 1.943407 0.7773 2.333 1.813 Total Piping Volume (gallons) [1,674.585Fuel Region (gallons)7.176Primary Circulation Pumps (gallons) 25.000Primary Heat Exchangers (gallons) 150.000Pressurizer (gallons) 150.000Total Volume of PCS (gallons) 2,006.761 eCFR -- tode of Federal Regulations" ATTACHMENT 8http://www.ecfr'.gov/cgi-birltext-idx?SlD=a6ddafde7f67322376d64cb..ELECTRONIC CODE OF FEDERAL REGULATIONSe-CFR data is current as of September 21, 2015Title 10 .- Chapter III --, Part 835 --. Subpart N -* Appendix-Title 10: EnergyPART 835--OCCUPATIONAL RADIATION PROTECTIONSubpart N-Emergency Exposure SituationsAPPENDIX C TO PART 835-DERIVED AIR CONCENTRATION (DAC) FOR WORKERS FROM EXTERNAL EXPOSURE DURINGIMMERSION IN A CLOUD OF AIRBORNE RADIOACTIVE MATERIALa. 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 ofairborne 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) peryear. 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. Thedata are listed by radionuclide in order of increasing atomic mass. The air immersion DACs were calculated for acontinuous, nonshielded exposure via immersion in a semi-infinite cloud of airborne radioactive material. The DACs listedin 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 ofthe ratio of the observed concentration of a particular radionuclide and its corresponding DAC for all radionuclides in themixture. If this sum exceeds unity (1), then the DAC has been exceeded. For unknown radionuclides, the most restrictiveDAC (lowest value) for those isotopes not known to be absent shall be used.AIR IMMERSION DACRadlonlucllde aflf piL)(.qm)Ar-37 __5._2_d_____00__,_____At-39 269___yr___E-__3__ E+__ 7Ar-41 157h3-6IEOK~r-74 1. an3-6lE0Kr-76 __4____h____-__5____+_ 5Kr-79 __5.____h___E-__5______ 5Kr-81 __.__+05_y______-____E_07Kr-83m 1.83 h 7E-02 2E+09Kr-85 10.72 yr 7E-04 2E+07Kr-85m 4.48 h 2E-05 IE+06Kr-87 76.3 mni 4E-06 1E+05Kr-88 2.84 h 1E-06 7E+04.Xe-120 40.0 min 1E-05 4E+05X(e-121 40.1 mai 2E-06 BE+04Xe-122 20.1 h 8E-05 3E+06Xe-123 2.14 h 6E-06 2E+05Xe-125 16.8 h lE-05 SE+05Xe-127 36.406 d 1 E-05 8E÷05Xe-129m 8.86 d 2E-04 7E+06Xe-1 31m 11.84 d 5E-04 1E+07Xe-133 5.245 d 1 E-04 5E+08Xe-133m 2.19 d 1 E-04 5E+06Xe-135 9.11 hi 1E-O5 6E+05Xe-1 35m 15.36 rni 1 E-05 3E+05Xe-138 14.13 min 3E-06 IE+05For any single radlonuclide not listed above with decay mode other than alpha emission or spontaneous fission andI of 2I of291231201 PM eCFR -Code of Federal RegulationsATTACHMENT 8http://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/m3).[72 FR 31940, June 8, 2007, as amended at 76 FR 20489, Apr. 13, 2011]Need assistance?!of 29/23/2015 4:37 PM Case Summary of Containment ShineATTACHMENT 9Page 1 of 3~MicroShield 8.02Nathan Hogue (8.00-0000)Date I By ChkeFilename IRun Date I Run Time I DurationContainl1.msd September 29, 2015 1:21:55 PM 00:00:00Project InfoCase Title Containment ShineDescription IFuel Accident AnalysesGeometr 13 -Rectangular VolumeSource DimensionsLength 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 zSin) in) in)___________ShieldsShield N J ~Dimension Material Density ____________Source 6.12e+09 cm3 I Air 0.00122Shield 1 j 30.5 cm I Concrete 2.35Air Gap j Air 0.00122Source Input: Grouping Method -Standard IndicesNumber of Groups: 25Lower Energy Cutoff: 0.015Photons < 0.015: Included_______________________ Library: Grove _______ _______Nuclide Ci Bq 3 B q/cm3I- 131 8.9329e+000 3.3052e+01 1 1.4600e-003 5.4020e+001I- 132 2.4168e+001 8.9421 e+011 3.9500e-003 1.4615e+002I- 133 5.0905e+001 1.8835e+012 8.3200e-003 -3.0784e+002I- 134 5.2252e+001 1 .9333e+01 2 8.5400e-003 3.1598e+0021-135 4.5644e+00 1 1 .6888e+0 12 7.4600e-003 2.7602e+002IK"-85 2.2271 e-003 8.2403e+007 3.6400e-007 1 .3468e-002Kr-85m 1.1625e+001 4.3013e+011 1.9000e-003 7.0300e+001Kr-87 1.5051 e+001 5.5690e+011 2.4600e-003 9.1020e+001Kr-88 2.4596e+001 9.1 006e+01 1 4.0200e-003 1 .4874e+002Kr-89 5.0416e-002 1.8654e+009 8.2400e-006 3.0488e-001Kr-90 6.0083e-01 6 2.2231 e-005 9.8200e-020 3.63 34e-0 15file:///C:/Program%2OFiles%20(x86)/MicroShield%208/Examples/CaseFiles/HTML/Cont... 9/29/2015 Case Summary of Containment ShineATTACHMENT 9Page 2 of 3Xe-1332.4596e+001I9.1006e+01 14.0200e-0031 .4874e+002Xe- 135 1.0157e+001 I 3.7579e+011 1.6600e-003 6.1420e+001Xe-135m 4.3196e+000 j 1.5983e+011 j 7.0600e-004 2.6122e+001Xe-137 2.1292e-001 j 7.8781e+009 3.4800e-005 1.2876e+000Xe- 138 1.1013 e+001 4.0749e+011 [ 1.8000e-003 6.6600e+001Buildup: The material reference is Shield 1Integration ParametersX Direction 10Y Direction I 20Z Direction 20____________ Results -Dose Point # 1 -(1890,914,914) cm ______Fluence Rate Fluence Rate Exposure Rate Exposure RateEnergy (MeV) iActivity (Photons/sec) MeV/cm1/sec MeV/cm2/sec mR/hr mR/hr_________No Buildup With Buildup No Buildup With Buildup0.015 1.997e+11I 8.577e-253 2.641e-24 7.357e-254 2.266e-250.03 5.752e+11I 6.299e-35 2.648e-23 6.242e-37 2.624e-250.08 3.426e+ 11 1.284e-04 3.213e-03 2.031 e-07 5.084e-060.1 1.795e+09 8.335e-06 _3.]31e-04 1.275e-08 4.791e-070.15 4.738e+l11 3.531 e-02 1.778e+00 5.814e-05 2.928e-030.2 6.841 e+ 11 2.179e-01 1.067e+01 3.845e-04 1.883e-020.3 3.207e+11 6.259e-01 2.298e+01 1.187e-03 4.359e-020.4 1 .055e+1 2 6.904e+00 1 .867e+02 1 .345e-02 3.638e-010.5 2.408e+12 3.898e+01 8.084e+02 7.652e-02 1.587e+000.6 1.884e+ 12 6.256e+01 1.036e+03 1.221 e-01 2.021 e+000.8 4.714e+12 4.684e+02 5.470e+03 8.909e-01 1.040e+011.0 1.860e+12 4.187e+02 3.760e+03 7.717e-01 6.931e+001 .5 1.544e+ 12 1.406e+03 _8.149e+03 2.365e+00 1.371 e+012.0 1.110e+12 2.482e+03 _1.108e+04 3.838e+00 1.713e+013.0 8.507e+10 5.850e+02 1.898e+03 7.936e-01 2.575e+004.0 8.353e+07 1.1 55e+00 3 .087e+00 1 .429e-03 3.81 9e-03Totals 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 RateEnergy (MeY) Activity (Photons/sec) MeV/cmz/sec MeV/cm2/sec mR/hr mR/hr_______No Buildup With Buildup No Buildup With Buildup0.015 1.997e+ 11 1.798e-263 1.169e-26 1.543e-264 1.003e-270.03 5.752e+11 6.868e-38 _1.172e-25 6.807e-40 1.162e-270.08 3.426e+ 11 4.039e-07 _1.139e-05 6.392e- 10 1.802e-080.1 1.795e+09 2.705e-08 _1.190e-06 4.139e-ll 1.821e-090.15 4.738e+11I 1.274e-04 _7.800e-03 2.098e-07 1.284e-050.2 6.841le+lI1 8.648e-04 5.172e-02 1.526e-06 9.128e-050.3 3.207e+11 2.857e-03 1.255e-01 5.419e-06 2.380e-040.4 1.055e+12 3.451e-02 1.091e+00 6.725e-05 2.125e-030.5 2.408e+12 2.079e-01 4.939e+00 4.082e-04 9.695e-03file:///C 9/29/2015 ShinePge3o3Page 3 of 30.61 .884e+1 23.501e-016.536e+006.834e-041 .276e-020.8 4.714e+12 2.798e+00 3.584e+01 5.322e-03 6.817e-021.0 1.860e+ 12 2.610e+00 2.523e+01 4.81 le-03 4.651 e-021.5 1.544e+ 12 9.284e+00 5.623e+01 1.562e-02 9.461 e-022.0 1.110e+l12 1.684e+01 7.709e+01 2.604e-02 1.192e-013.0 8.507e+10 4.061e+00 1.323e+01 5.510e-03 1.795e-024.0 8.353e+07 8.082e-03 2.146e-02 9.998e-06 2.655e-05Totals 1.726e+13 3.620e+01 2.204e+02 5.848e-02 3.714e-01file %208/Examples/CaseFiles/HTML/Cont... 9/29/2015 ATTACHMENT 9 Case Summary of Containment ShineATTACMENT 9Page 1 of 3MicroShield 8.02Nathan Hogue (8.00-0000)Date ByCheckedFilename IRun Date I Run Time I Duration JContainl .msd September 29, 2015 1:23:52 PM 00:00:00Project InfoCase Title Containment ShineDescription IFuel Element Failure Accident AnalysesGeometr 13 -Rectangular VolumeSource DimensionsLength 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 11.82 in) j n nShieldsShield N Dimension Material Densit ____________Source 6.12e+09 cm3 Air 0.00122Shield 1 I 30.5 cm I Concrete I 2.35Air Gap Air 0.00122Source Input: Grouping Method -Standard IndicesNumber of Groups: 25Lower Energy Cutoff: 0.0 15Photons < 0.015: Included__ _ _ _ _ _ _ _ Library: Grove .........._ __ __ __ __ __ _Nuclide Ci Bq JLCi/cm3 Bq/cm3I-131 1 .3399e-00 1 4.9578e+009 2.1 900e-005 8.1 030e-0011-132 2.6003e-001 9.6213e+009 4.2500e-005 1.5725e+000I- 133 4.0198e-001 1.4873e+010 6.5700e-005 2.4309e+0001-134 4.9682e-001 1 .8382e+0 10 8.1 200e-005 3 .0044e+000I- 135 4.0932e-001 1.5 145e+010 6.6900e-005 2.4753e+000Kr-85 6.0940e-004 2.2548e+007 9.9600e-008 3.6852e-003Kr-85m 1.4256e-001 5.2747e+009 2.3300e-005 8.6210e-001Kr-87 2.7288e-001 1 .0097e+010 4.4600e-005 1 .6502e+000Kr-88 3.8852e-001 1.4375e+010 6.3500e-005 2.3495e+000Kr-89 4.9253e-001 1 .8224e+01 0 8.0500e-005 2.9785e+000Kr-90 4.9253e-00 1 1 .8224e+0 10 8 .0500e-005 2.9785e+000file:///C:/Program%2OFiles%20(x86)/MicroShield%208/Examples/CaseFiles/HTML/Cont... 9/29/2015 Case Summary of Containment ShineATTACHMENT 9Page 2 of 3Xe-1335.4454e-0012.0148e+0108.9000e-0053 .2930e+000Xe-135 1 .2482e-001 4.6182e+009 2.0400e-005 7.5480e-001Xe.-135m j 1.2176e-001 J 4.5050e+009 j 1.9900e-005 j 7.3630e-001Xe- 137 [ 6.3632e-001 J 2.3544e+010 J 1.0400e-004 j 3.8480e+000Xe- 138 j 6.7303 e-001 Jj 2.4902e+010 J 1.1000e-004 4.0700e+000Buildup: The material reference is Shield 1Integration ParametersX Direction 10Y ireto II20ZDirection 20____________ Results -Dose Point # 1 -(1890,914,914) cm ______Fluence Rate Fluence Rate Exposure Rate Exposure RateEnergy (MeV) Activity (Photons/sec) MeV/cmn2sec MeV/cm2/sec mR/hr mR/hr_________No Buildup With Buildup No Buildup With Buildup0.015 5.847e+09 2.51 2e-254 7.735e-26 2.1 54e-255 6.635e-270.03 1.247e+ 10 1.365e-36 5.739e-25 1.353e-38 5.688e-270.08 7.524e+09 2.819e-06 7.055e-05 4.460e-09 1.116e-070.1 6.447e+09 2.994e-05 1.125e-03 4.580e-08 1.721e-060.15 6.836e+09 5.094e-04 2.565e-02 8.388e-07 4.224e-050.2 1.617e+10 5.150e-03 2.522e-01 9.089e-06 4.45 le-040.3 1.051e+10 2.051e-02 7.532e-01 3.891e-05 1.429e-030.4 2.244e+10 1 .468e-01 3.969e+00 2.860e-04 7.733e-030.5 3.752e+10 6.074e-01 1.259e+01 1.192e-03 2.472e-020.6 2.587e+10 8.590e-0I 1.422e+01 1.677e-03 2.775e-020.8 5.049e+ 10 5.017e+00 5.859e+01 9.542e-03 1.1 14e-011.0 3.060e+ 10 6.885e+00 6.184e+01 1.269e-02 1.140e-011.5 2.473e+10 2.251e+01 1.305e+02 3.787e-02 2.195e-012.0 2.762e+ 10 6.174e+01 2.755e+02 9.547e-02 4.260e-013.0 3.396e+09 2.335e+01 7.576e+I01 3.168e-02 1.028e-014.0 8.371e+08 1.158e+01 3.094e+01 1.432e-02 3.828e-02Totals 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 RateEnergy (MeV) Activity (Photons/sec) MeV/cm2/sec MeV/cm2/sec mR/hr mR/hr_________No Buildup With Buildup No Buildup With Buildup0.015 5.847e+09 5.266e-265 3.424e-28 4.517e-266 2.937e-290.03 1.247e+ 10 1.489e-39 2.540e-27 1.475e-41 2.518e-290.08 7.524e+09 8.870e-09 2.501 e-07 1 .404e- 11 3.957e- 100.1 6.447e+09 9.718e-08 4.276e-06 1.487e- 10 6.542e-090.15 6.836e+09 1.838e-06 1.125e-04 3.027e-09 1.853e-070.2 1.617e+10 2.044e-05 1.223e-03 3.608e-08 2.158e-060.3 1.051 e+l 0 9.363e-05 4.113e-03 1.776e-07 7.801 e-060.4 2.244e+ 10 7.337e-04 2.319e-02 1.430e-06 4.518e-050.5 3.752e+ 10 3.240e-03 7.695e-02 6.359e-06 1.51 le-04file:///C:/Program%2OFiles%20(x86)/MicroShield%208/Examples/CaseFiles/HTML/Cont... 9/29/2015 Cas~1jp~xp~P~aimnent Shine Page 3 of 3Page 3 of 30.62.587e+104.807e-038.974e-029.383e-061 .752e-040.8 5.049e+10 2.997e-02 3.839e-01 5.700e-05 7.301e-041.0 3.060e+10 4.292e-02 4.149e-01 7.912e-05 7.649e-041.5 2.473e+10 1.486e-01 9.003e-01 2.501e-04 1.515e-032.0 2.762e+ 10 4.1 89e-0 1 1.91 8e+00 6.477e-04 2.965e-033.0 3.396e+09 1.621e-01 5.280e-01 2.199e-04 7.163e-044.0 8.371e+08 8.099e-02 2.151e-01 1.002e-04 2.660e-04Totals 2.893e+11 8.924e-01 4.555e+OO 1.371e-03 7.339e-039/29/2015 ATTACHMENT 9 Case Summary of Containment ShineATTACHMENT 9Page 1 of 3~MicroShield 8.02Nathan Hogue (8.00-0000)Date I By ICheckedI Filename )Run Date I Run Time I Duration jI Containl .msd j September 29,2015 1:15:41 PM 00:00:01 j_________________Project InfoCase Title Containment ShineDescription Fuel Experiment Accident AnalysesGeometry 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 Y2 1.5e+4 cm (492it 1.5 914.0em (29 fi11.8 914.0Ocm (29 fi11.8.2 in) in) in)ShieldsShield N Dimension Material Density ___________Source 6.12e+09 cm3 Air 0.00122Shield 1 I 30.5 cm I Concrete I 2.35Air Gap Air 0.00122Source Input: Grouping Method -Standard IndicesNumber of Groups: 25Lower Energy Cutoff: 0.0 15Photons < 0.015: Included_______________ Library: Grove ________Nuclide Ci Bg JLCi/cm3 Bg/cm3I- 131 8.0763e+000 2.9882e+011 1.3200e-003 4.8840e+0011- 132 1.7866e+001 6.6104e+011 2.9200e-003 1.0804e+0021-133 3.8240e+001 1.4149e+012 6.2500e-003 2.3125e+0021- 134 4.3563e+001 1.611 8e+012 7.1 200e-003 2.6344e+002I- 135 3.6160e+001 1.3379e+012 5.9100e-003 2.1867e+002Kr-85 1 .6459e-003 6.0897e+007__ 2.6900e-007 9.9530e-003Kr-85m 7.2810e+000 2.6940e+011 1.1900e-003 4.4030e+001Kr-87 1 .4807e+001 5.4785e+011I 2.4200e-003 8.9540e+001Kr-88 2.0864e+001 7.7196e+011I 3.4100e-003 1 .2617e+002Kr-89 2.6676e+001 9.8703e+011 4.3600e-003 1.6132e+002Kr-90 2.6309e+001 9.7344e+011 4.3000e-003 1.5910e+002file 9/29/2015 Case Summary of Containment ShineATTACHMENT 9Page 2 of 3Xe-1331.81 72e+0016.7236e+01 12.9700e-0031 .0989e+002Xe- 135 1,I3093 e+i001 4.8446e+011 2.1400e-003 7.9180e+001iXe-I135m 6,4856e+'000 [ 2.3997e+01 1 1 .0600e-003 j 3.9220e+001Xe- 137 3.4386e+001 1 .2723 e+012 5.6200e-003 j 2.0794e+002Xe- 138 j 3.5915e+001 [ 1.3289e+012 5.8700e-003 2.1719e+002Buildup: The material reference is Shield 1Integration ParametersX Direction 10Y Direction I 20Z Direction 20__________Results -Dose Point # 1 -(1890,914,914) cmFluence Rate Fluence Rate Exposure Rate Exposure RateEnergy (MeV) Activity (Photons/sec) MeV/cm2/sec MeV/cmZ/sec mR/hr mR/hrNo Buildup With Buildup No Buildup With Buildup0.015 2.904e+11I 1.248e-252 3.842e-24 1.070e-253 3.296e-250.03 5.010e+ll 5.485e-35 2.306e-23 5.436e-37 2.285e-250.08 2.546e+ 11 9.536e-05 2.387e-03 1.509e-07 3.777e-060.1 3.444e+ 11 1.599e-03 6.008e-02 2.447e-06 9.192e-050.15 3.872e+ 11 2.885e-02 1.453e+00 4.751le-05 2.393e-030.2 1.110e+12 3.537e-01 1.732e+01 6.242e-04 3.056e-020.3 5.920e+11 1 I.156e+00 4.243 e+01 2.192e-03 8.048e-020.4 1 .346e+12 8.808e+00 2.382e+02 1.716e-02 4.640e-010.5 2.718e+12 4.399e+01 9.123e+02 8.635e-02 1.791e+000.6 1.808e+12 6.003e+01 9.936e+02 1.172e-01 1.939e+000.8 4.038e+12 4.012e+02 4.686e+03 7.631e-01 8.912e+001.0 2.157e+12 4.855e+02 4.360e+03 8.949e-01 8.037e+001.5 1.735e+ 12 1.580e+03 9.156e+03 2.658e+00 1.540e+012.0 1.593e+12 3.562e+03 1.589e+04 5.508e+00 2.458e+t013.0 1.838e+11 1.264e+03 4.101le+03 1.715e+00 5.563e+004.0 4.532e+10 6.268e+02 1.675e+03 7.754e-01 2.072e+00Totals 1.910e+13 8.033e+03 4.208e+04 1.254e+01 6.887e+01Results -Dose Point # 2 -(15000,914,914) cmF'luence Rate Fluence Rate Exposure Rate Exposure RateEnergy (MeV) Activity (Photons/sec) MeV/cm2/sec MeV/cm2/see mR/hr mRlhr_____________No Buildup With Buildup No Buildup With Buildup0.015 2.904e+ 11 2.616e-263 1.701 e-26 2.244e-264 1.459e-270.03 5.010e-il 1 5.981e-38 1.021e-25 5.928e-40 1.012e-270.08 2.546e+11 3.001e-07 8.461e-06 4.749e-10 1.339e-080.1 3.444e+ 11 5.191 e-06 2.284e-04 7.942e-09 3.494e-070.15 3.872e+11 1,041e-04 6.374e-03 1.714e-07 1.050e-050.2 1.110e+ 12 1.404e-03 8.395e-02 2.478e-06 1.482e-040.3 5.920e+11I 5.274e-03 2.3 17e-01 1.000e-05 4.394e-040.4 1 .346e+12 4.403e-02 1.392e+00 8.579e-05 2.71 le-030.5 2.718e+ 12 2.347e-01 5.574e+00 4.606e-04 1.094e-02fle :///C :/Program%2OFiles%20(x86)/MicroShield%208/Exampies/CaseFiles/HTML/Cont... 9/29/2015 Case 1;n~ainm~ent ShinePae3o3Page 3 of 30.61.808e+123 .359e-01I6.271 e+006.5 57e-041 .224e-020.8 4.038e+12 2.397e+00 3.070e+01 4.559e-03 5.839e-021.0 2.157e+12 3.026e+00 2.926e+01 5.579e-03 5.393e-021.5 1.735e+12 1.043e+01 6.318e+01 1.755e-02 1.063e-012.0 1.593e+12 2.416e+01 1.106e+02 3.737e-02 1.711e-O13.0 1.838e+11 8.774e+00 2.858e+01 1.190e-02 3.877e-024.0 4.532e+10 4.385e+00 1.164e+01 5.425e-03 1.440e-02Totals 1.910e+13 5.380e+0O1 2.875e+02 8.360e-02 4.694e-01fie:///C:/Programi%2OFi~es%20(x86)/MicroShield%208/Examples/CaseFiles/HTML/Cont... 9/29/2015 ATTACHMENT 9 ATTACHMENT 10wIND ROSE PLOT:Station #03945 -COLUMBIAIREGIONAL ARPT, MODISPLAY:Wind SpeedDirection (blowing from)NORTH15%12%9%WESTEASTWIND SPEED(mis)U >=11 IU8.8-11.11 5.7-.88* 3.6- 5.7E]2.1 -3.62.1Calms: 1.15%SOUTHDATA PERIOD: COMPANY NAME:Start Date: 11111961 -00:00End Date: 1213111969 -21:00'MODELER:CALM WINDS: TOTAL COUNT:1.15% 73020 hr..AVG. WIND SPEED: DATE: PROJECT NO.:4.70 mls 912312015WRPLOT View -Lakes Envlonmanlat Software Wind Class Frequency Distribution:z40-35-30-%/252015105mACalms0-0.5- 2.12.1- 3.6 3.6- 5.7Wind Class (mis)0.8>= 11.15.7- 8.88.8 -11.1 ATTACHMENT 11WIND ROSE PLOT:Station #03945 -COLUMBIA/REGIONAL ARPT, MODISPLAY:Wind SpeedDirection (blowing from)NORTH20%16%12%WESTEASTWIND SPEED(mis)U>= 11.11n 8.8-11.11 57- 8.83.6- 5.7m2.1 -3.6LI 0.5- 2.1Calmns: 1.83%SOUTHCOMwMENTS: DATA PERIOD: COMPANY NAME:Start Date: 1/1/1170 -00:00End Date: 1213111990 -23:00 ___________ ____________MODELER:CALM WINOS: TOTAL COUNT:1.83% 154387 hrs.AVG. WIND SPEED: DATE; PROJECT NO.:4.44 mole 912312015WRPLOT View -Lakes Envirotnmental Software Wind Class Frequency DistributionA 3-z-r'J44~1- 4* 41 I.41 I..625--20-15-10-!0.00-0.7>= 11.1Calms0.5- 2.12.1- 3.6 3.6- 5.7 5.7- 8.8Wind Class (mis)8.8- 11.1kI VIUW ! .U.U "

ATTACHMENT 12WIND ROSE PLOT:Station #03945 -COLUMBIA/REGIONAL ARPT, MODISPLAY:Wind SpeedDirection (blowing from)NORTH20%16%12%WESTEASTWIND SPEED(mis)U = 11.1n8.8- 11.11 57- 8.8*36- 5.7D 21 -3.6EJ 0.5 -2.1Calms: 1.63%:SOUTHCOMMJENTS: DATA PERIOD: COMP~ANY NAAE:Start Date: 11111961 -00:00End Date: 12131/1990.-23:00MODELER:CALM WINDS: TOTAL COUNT:1.63% 227407 hrs.AVG. WINO SPEED: DATE: PROJECT NO,:4.52 mls 9123/2015WRPLOT View.- Lakes Environmental Software Wind Class Frequency DistributionIl ~.. I-T.lr-i40*35.L.4.301-4.27.%/25-20-15-10 '0-0.7>= 11.10.5- 2.12.1 -3.6 3.6 -5.7 5.7 -8.8Wind Class (m/s)8.8- 11.1view 1-rewar 7A.Uo -LX~e uwltrmnmlenltimrw/19 ATTACHMENT 13Stack Effluent Releases -Calendar Years 2005 to 20142005 2006 2007 2008 2009 2010 2011 2012 2103 2014 AverageIsotope (% 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 69.91238C-14 0.777 0.74 0.793 0.7867 0.613 0.58 0.477 0.723 0.0083 0.0079 0.55059Os-191 0.0011 0.0018 0.0066 4.1739 0.0294 0.0008 0.0003 0.0001 0.0002 0.468241-131 0.0921 0.0435 0.0401 0.0782 0.6035 0.0415 0.0506 0.0503 0.0169 0.2201 0.12368Ce-144 0.1165 0.0852 0.10085Co-60 0.0853 0.0792 0.3372 0.0784 0.0084 0.0049 0.0054 0.08554H-3 0.0732 0.052 1 0.0485 0.0527 0.0328 0.0353 0.0496 0.0426 0.0633 0.0558 0.05059Kr-79 0.0482 0.0274 0.0378Sc-46 0.0263 0.0022 0.01425K-40 0.0093 0.0164 0.01 0.01 19Cd-109 0.0112 0.01 121-125 0.0215 0.0041 0.0021 0.0073 0.0037 0.00774Fe-59 0.0038 0.0038Se-75 0.0005 0.0057 0.003 1Sb-125 0.0026 0.0026Zn-65 0.0005 0.001 0.0026 0.0009 0.00 125Htg-203 0.0002 0.001 0.0002 0.0013 0.0033 0.0012Cs-137 0.0007 0.0013 0.0006 0.0003 0.0004 0.0012 0.00075Zr-95 0.0005 0.0005 0.00051-133 0.0003 0.0001 0.0001 0.0001 0.0003 0.0001 0.0001 0.0001 0.003 0.00047Sn-i113 0.0009 0.0003 0.000 1 0.00043Au-196 0.0005 0.0003 0.0004 0.0003 0.0004 0.00038Gd-153 0.0003 0.00031 of 2 ATTACHMENT 13Stack Effluent Releases -Calendar Years 2005 to 2014Cu-67 0.0003 0.0003Pa-233 0.0002 0.0003 0.00025S-35 0.000 1 0.000 1 0.0005 0.0002 0.00023Hf-181 0.0004 0.0001 0.000 1 0.0002 0.0002Ce-141 0.0003 0.0002 0.000 1 0.0002Xe-133 0.0002 0.0002Ba-140 0.0003 0.0002 0.0001 0.0002 0.0002Nb-95 0.0003 0.000 1 0.0002Br-82 0.0002 0.000 1 0.00015Co-58 0.0001 0.0001 0.0002 0.00013As-77 0.0002 0.0001 0.0001 0.00013Ce-139 0.000 1 0.000 1 0.000 1Ru-103 0.0001 0.000 1 0.0001 0.0001Mn-54 0.0001 0.0001Be-7 0.000 1 0.000 1Co-57 0.000 1 0.000 1Hf-175 0.000 1 0.000 1 0.0001Xe- 135m 0.000 1 0.00012 of 2

Retrieved from "https://kanterella.com/w/index.php?title=ML15275A314&oldid=566465"