Text

University of Missouri - Columbia Research Reactor, Response to Request for Additional Information on License Renewal Amendment Request
	ML16110A164
Person / Time
Site:	University of Missouri-Columbia
Issue date:	04/15/2016
From:	Rhonda Butler; Univ of Missouri
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	TAC ME1580
	Download: ML16110A164 (133)
	v • d • e

{{#Wiki_filter:UNIVERSITY of MISSOURI RESEARCH REACTOR CENTER April 15, 2016 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Mail Station Pl-37 Washington, pc 20555-0001

REFERENCE:

Docket 50-186 University of Missouri-Columbia Research Reactor Amended Facility Operating License No. R-103

SUBJECT:

Written communication as specified by 10 CPR 50.4(b)(l) regarding responses to the "University of Missouri at Columbia - Clarifications Needed to Nuclear Regulatory Commission Staff Request for Additional Information Regarding the Renewal of Facility Operating License No. R-103 for the University of Missouri at Columbia Research Reactor (TAC No. ME1580)," dated March 23, 2016 On August 31, 2006, the University of Missouri-Colum.bia Research Reactor (MURR) submitted a request to the U.S. Nuclear Regulatory Commission (NRC). to renew Amended Facility Operating License No. R-103. By letter dated May 6, 2010, the NRC requested additional information and clarification regarding the renewal request in the form of nineteen (19) Complex Questions. By letter dated September 3, 2010, MURR responded to seven (7) of those Complex Questions. By letter dated June 1, 2010, the NRC requested additional information and clarification regarding the renewal request in the form of one hundred and sixcy-seven (167) 45-Day Response Questions. By letter dated July 16, 2010, MURR responded to forty-seven (47) of those 45-Day Response Questions. On July 14, 2010, via electronic mail (email), MURR requested additional time to respond to the remaining one hundred and twenty (120) 45-Day Response Questions. By letter dated August 4, 2010, the NRC granted the request. By letter dated August 31, 2010, MURR responded to fifty-three (53) of the 45-Day Response Questions. On September 1, 2010, via email, MURR requested additional time to respond to the remaining twelve (12) Complex Questions. By letter dated September 27, 2010, the NRC granted the request. 1513 Research Park Drive Columbia, MO 65211 Phone: 573-882-4211 Fax: 573-882-6360 Web: www.murr.missouri.edu Fighting Cancer with Tomorrow's Technology

On September 29, 2010, via email, MURR requested additional time to respond to the remaining sixty-seven (67) 45-Day Response Questions. On September 30, 2010, MURR responded to sixteen (16) of the remaining 45-Day Questions. By letter dated October 13, 2010, the NRC granted the extension request. By letter dated October 29, 2010, MURR responded to sixteen (16) of the remaining 45-Day Response Questions and two (2) of the remaining Complex Questions. By letter dated November 30, 2010, MURR responded to twelve (12) of the remaining 45-Day Response Questions. On December 1, 2010, via email, MURR requested additional time to respond to the remaining 45-Day Response and Complex Questions. By letter dated December 13, 2010, the NRC granted the extension request. On January 14, 2011, via email, MURR requested additional time to respond to the remaining 45-Day Response and Complex Questions. By letter dated February 1, 2011, the NRC granted the extension request. By letter dated March 11, 2011, MURR responded to twenty-one (21) of the remaining 45-Day Response Questions. On May 27, 2011, via email, MURR requested additional time to respond to the remaining 45-Day Response and Complex Questions. By letter dated July 5, 2011, the NRC granted the request. By letter dated September 8, 2011, MURR responded to six (6) of the remaining 45-Day Response and Complex Questions. On September 30, 2011, via email, MURR requested additional time to respond to the remaining the remaining 45-Day Response and Complex Questions. By letter dated November 10, 2011, the NRC granted the request. By letter dated January 6, 2012, MURR responded to four (4) of the remaining 45-Day Response and Complex Questions. Also submitted was an updated version of the MURR Technical Specifications. On January 23, 2012, via email, MURR requested additional time to respond to the remaining the remaining 45-Day Response and Complex Questions. By letter dated January 26, 2012, the NRC granted the request. On April 12, 2012, via email, MURR requested additional time to respond to the remaining the remaining 45-Day Response and Complex Questions. By letter dated June 28, 2012, MURR responded to the remaining six (6) 45-Day Response and Complex Questions. With that set ofresponses, all 45-Day Response and Complex Questions had been addressed. 2 of54

By letter dated December 20, 2012, the NRC requested a copy of the current Physical Security Plan (PSP) and Operator Requalification Program. By letter dated January 4, 2013, MURR provided the NRC a copy of the current PSP and Operator Requalification Program. By letter dated February 11, 2013, the NRC requested updated financial information in the form of four (4) questions because the information provided by the September 14, 2009 response had become outdated. By letter dated March 12, 2013, MURR responded to the four (4) questions. By letter dated December 3, 2014, the NRC requested additional information in the form of two (2) questions regarding significant changes to the MURR facility since submittal of the licensing renewal application in August 2006. By letter dated January 28, 2015, MURR responded to the two (2) questions. By letter dated April 17, 2015, the NRC requested additional information in the form of ten (10) questions. On May 29, 2015, via email, MURR requested additional time to respond to the ten (10) questions. By letter dated June 18, 2015, the NRC requested additional information in the form of two (2) questions. By letter dated July 31, 2015, MURR 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 fifteen (15) questions regarding the PSP. By letter dated September 15, 2015, MURR provided the NRC a copy of the current EP. By letter dated October 1, 2015, MURR responded to the ten (10) questions from the April 17, 2015 request. By letter dated December 2, 2015, MURR responded to the fifteen (15) questions from the September 14, 2015 request regarding the PSP. By letter ciated December 17, 2015, the NRC requested additional information in the form of thirteen ( 13) questions regarding follow-up information from MURR's October 1, 2015 responses to the NRC's April 17, 2015 request for additional information. 3 of54

By letter dated February 8, 2016, MURR responded to the thirteen (13) questions from the December 17, 2015 request. By letter dated February 8, 2016, the NRC requested updated :financial information in the form of four (4) questions because the information provided by the March 12, 2013 response had become outdated. By letter dated March 23, 2016, the NRC requested additional information in the form of twenty-one (21) questions regarding follow-up information from MURR's February 8, 2016 responses to the NRC's April 17, 2015 request for additional information. By letter dated April 8, 2016, MURR responded to the four (4) financial information questions. The March 23, 2016 questions, and MURR's responses to those questions, are attached. If there are any questions regarding this response, please contact me at (573) 882-5319 or FruitsJ@missouri.edu. I declare under penalty of perjury that the foregoing is true and correct. ENDORSEMENT: s~~JZS" Reviewed and Approved, Jacqueline m, Notary PubliC My ConvnisslOll Expires: t.Wch 26, 2019 John L. Fruits Ralph A. Butler, P.E. Reactor Manager Director JACQUELINE L,,BOHM Notary Public-Notary Seal ST ATE OF MISSOURI xc: Reactor Advisory Committee Commissioned for 1lowaai County My Commissi\Hl Expires: March 26, 2019 Reactor Safety Subcommittee commission # 15634308 Dr. Garnett S. Stokes, Provost Dr. Mark Mcintosh, Vice Chancellor for Research, Graduate Studies and Economic Development Mr. Alexander Adams Jr., U.S. Nuclear Regulatory Commission Mr. Geoffrey Wertz, U.S. Nuclear Regulatory Commission Mr. Johnny Eads, U.S. Nuclear Regulatory Commission Attachments:

1. Fuel Failure during Reactor Operation and Fuel Handling Accident Source Terms -

MONTEBURNS Output File

2. Fuel Failure during Reactor Operation - Excel Spreadsheet Dose Calculations - Restricted Area

3. Fuel Failure during Reactor Operation - Excel Spreadsheet Dose Calculations - Unrestricted Area

4. Calculated Derived Air Concentration Values for Kr-89, Kr-90, Xe-137 and Xe-139

5. Containment Building- Excel Spreadsheet Leakage Rate 4 of54

6. Fuel Failure during Reactor Operation - MicroShield 8.02 Dose Calculations

7. Fuel Handling Accident - Excel Spreadsheet Dose Calculations - Restricted Area

8. Fuel Handling Accident - Excel Spreadsheet Dose Calculations - Unrestricted Area

9. Fuel Handling Accident-MicroShield 8.02 Dose Calculations

10. Fueled Experiment Failure - Excel Spreadsheet Dose Calculations - Restricted Area

11. Fueled Experiment Failure - Excel Spreadsheet Dose Calculations - Unrestricted Area

12. Fueled Experiment Failure - MicroShield 8.02 Dose Calculations 5 of54

Note: The numbering in the column labeled "RAJ No. " corresponds to the RAJs issued by NRC letter dated April 17, 2015. The brackets [page} at the end of each item indicate the page number on the RAJ response provided by MURR letter dated February 8, 2016. RAJ No. 3.c. - Information/Clarification Needed

Provide control blade D total reactivity worth - measured and calculated in numerical form (not graphical) [page 9].

In order to benchmark the MCNP core and control blade model, the measured total control blade worth for control blade 'D' was compared against the MCNP calculated total control blade worth. The results are shown in the Table below. Control Total Reactivity Worth -Measured Total Reactivity Worth - Calculated Blade (Ak/k) (Ak/k) D -0.0355 -0.0364 RAJ No. 7.g. - Information/Clarification Needed

Explain origin ofsource term data provided in RAJ response [page 19}.

For operation at 10 MW for 1,200 MWd in twelve 10-day cycles over a 300-day period with 6.2 Kg 235 of U (normal operating cycle is 6.5 days with a total of less than 700 MWd on the core), the radioiodine, krypton and xenon activities listed in the "Fuel Failure during Reactor Operation" accident analysis will conservatively be present in the core. Due to the complex nature of the MURR mixed core fuel cycle, the source term was determined using the computer program MONTEBURNS instead of simplistic ORIGEN runs. MONTEBURNS is a coupled MCNP-ORIGEN code system developed by Los Alamos National Laboratory (LANL). Within the MONTEBURNS program, MCNP calculations are used to obtain accurate one-group fluxes and one-group cross sections that are then utilized by ORIGEN for fuel depletion and fission product activity calculations. The use of MONTEBURNS and the flow diagram utilized for various neutronic computations at MURR were described in detail in the response to Request for Additional Information (RAI) Question 3 .a (MURR letter dated October 1, 2015). Using MONTEBURNS to simulate the burnup of all eight (8) fuel elements for the core configuration beginning with an all-fresh core for the aforementioned hypothetical operational cycle, the source term for the "Fuel Failure during Reactor Operation" analysis was derived from summing the noble gases and iodine fission products inventories from all eight (8) fuel elements at the end of 300 days (i.e., at the End-of-Irradiation of cycle 12). The source term for the "Fuel Handling Accident (FHA)" analysis was similarly derived using the same radionuclides by extending the previous 300-day 6 of54

MONTEBURNS simulation and decaying it for 30 minutes (i.e., 0.0208 days after the End-of-Irradiation of cycle 12).

Provide Attachment 4 in landscape mode to eliminate data wrapping around the next line, remove transport edits, and provide an explanation to translate calculated radioisotope inventories into the RAJ response inventory.

Attached is the MONTEBURNS output file (Attachment 1) which provides the sources terms for both the "Fuel Failure during Reactor Operation" and "Fuel Handling Accident (FHA)" analyses.

Provide the equations (or spreadsheets) for calculations from Attachments 5 and 6.

Attached are the new spreadsheets (Attachments 2 and 3) for the "Fuel Failure during Reactor Operation" restricted and unrestricted area doses. Additionally, the equations are better defined in the revised accident analysis, which is attached at the end of the Fuel Failure during Reactor Operation section. RAJ No. 7.g.iv. -Information/Clarification Needed

Correct Kr-90 dose by eliminating the negative dose contribution.from the curve fit [page 20].

The negative dose contribution from the curve fit has been eliminated; therefore, the new Kr-90 Derived Air Concentration value is 2. 8 x 1o-06 µCi/ml (see Attachment 4). This value now reflects that the whole body is the limiting organ and not the skin as in the previous analysis.

Review/correct the gamma source terms.

The gamma source terms have been corrected (see Attachment 4).

Correct reference to Attachment 7 (not Attachment 5 as indicated).

Attachment 5 was improperly referenced in the previous MURR response, by letter dated February 8, 2016. It should have referenced Attachment 7 in the "Fuel Failure during Reactor Operation" accident analysis. The revised "Calculated Derived Air Concentration Values for Kr-89, Kr-90, Xe-137 and Xe-139" is now Attachment 4 to this response.

Provide an updated Attachment 7.

Attachment 7 has been updated; it is now labeled Attachment 4 in this response. 7 of54

RAJ No. 9.b.ii. - Information/Clarification Needed

Explain the source term for the 30 minutes decay [page 38].

As described in the response to RAI No. 7.g. above, for operation at 10 MW for 1,200 MWd in twelve 10-day cycles over a 300-day period with 6.2 Kg of 235U (normal operating cycle is 6.5 days with a total of less than 700 MWd on the core), the radioiodine, krypton and xenon activities listed in the "Fuel Failure during Reactor Operation" accident analysis will conservatively be present in the core. Due to the complex nature of the MURR mixed core fuel cycle, the source term was determined using the computer program MONTEBURNS instead of simplistic ORlGEN runs. MONTEBURNS is a coupled MCNP-ORlGEN code system developed by Los Alamos National Laboratory (LANL). Within the MONTEBURNS program, MCNP calculations are used to obtain accurate one-group fluxes and one-group cross sections that are then utilized by ORlGEN for fuel depletion and fission product activity calculations. The use of MONTEBURNS and the flow diagram utilized for various neutronic computations at MURR were described in detail in the response to Request for Additional Information (RAI) Question 3.a (MURR letter dated October 1, 2015). Using MONTEBURNS to simulate the burnup of all eight (8) fuel elements for the core configuration beginning with an all-fresh core for the aforementioned hypothetical operational cycle, the source term for the "Fuel Failure during Reactor Operation" analysis was derived from summing the noble gases and iodine fission products inventories from all eight (8) fuel elements at the end of 300 days (i.e., at the End-of-Irradiation of cycle 12). The source term for the "Fuel Handling Accident (FHA)" analysis was similarly derived using the same radionuclides by extending the previous 300-day MONTEBURNS simulation and decaying it for 30 minutes (i.e., 0.0208 days after the End-of-Irradiation of cycle 12). Attachment 1 is the MONTEBURNS output which provides the sources terms for both the "Fuel Failure during Reactor Operation" and "Fuel Handling Accident (FHA)" analyses. 8 of54

Fuel Failure During Reactor Operation Analysis [page 22] -Information/Clarification Needed

Explain the origin of core inventory of isotopes (which we now understand is not from Attachment 2)

[page 24]. As described in the response to RAJ No. 7.g. above, for operation at 10 MW for 1,200 MWd in twelve 10-day cycles over a 300-day period with 6.2 Kg of 235 U (normal operating cycle is 6.5 days with a total ofless than 700 MWd on the core), the radioiodine, krypton and xenon activities listed in the "Fuel Failure during Reactor Operation" accident analysis will conservatively be present in the core. Due to the complex nature of the MURR mixed core fuel cycle, the source term was determined using the computer program MONTEBURNS instead of simplistic ORIGEN runs. MONTEBURNS is a coupled MCNP-ORIGEN code system developed by Los Alamos National Laboratory (LANL). Within the MONTEBURNS program, MCNP calculations are used to obtain accurate one-group fluxes and one-group cross sections that are then utilized by ORIGEN for fuel depletion and fission product activity calculations. The use of MONTEBURNS and the flow diagram utilized for various neutronic computations at MURR were described in detail in the response to Request for Additional Information (RAJ) Question 3.a (MURR letter dated October 1, 2015) .

 . Using MONTEBURNS to simulate the bumup of all eight (8) fuel elements for the core configuration beginning with an all-fresh core for the aforementioned hypothetical operational cycle, the source term for the "Fuel Failure during Reactor Operation" analysis was derived from summing the noble gases and iodine fission products inventories from all eight (8) fuel elements at the end of 300 days (i.e., at the End-of-Irradiation of cycle 12). The source term for the "Fuel Handling Accident (FHA)"

analysis was similarly derived using the same radionuclides by extending the previous 300-day MONTEBURNS simulation and decaying it for 30 minutes (i.e., 0.0208 days after the End-of-Irradiation of cycle 12). Attachment 1 is the MONTEBURNS output which provides the sources terms for both the "Fuel Failure during Reactor Operation" and "Fuel Handling Accident (FHA)" analyses.

The mass transport equation is not clear, bracketing appears inconsistent - provide step-by-step description of the transport process with supporting equations for clarity [page 26].

The "Fuel Failure during Reactor Operation" analysis has been revised to provide a step-by-step description of the transport process with supporting equations.

Clarify how the current MURR TS 4.2.c containment building leakage rate is being used as the assumption for the leakage in the analysis [page 30].

As described in detail in the revised "Fuel Failure during Reactor Operation" accident analysis, the containment building ventilation system will shut down and the building itself will be isolated from the surrounding areas. Fuel failure will not cause an increase in pressure inside the reactor 9 of54

containment structure; therefore, any air leakage from the building will occur as a result of normal changes in atmospheric pressure and pressure equilibrium between the inside of the containment structure and the outside atmosphere. It is highly probable that there will be no pressure differential between the inside of the containment building and the outside atmosphere, and consequently there will be no air leakage from the building and no radiation dose to members of the public in the unrestricted area. However, to develop what would clearly be a worst-case scenario, this analysis assumes that a barometric pressure drop has occurred in conjunction with fuel failure. An extreme assumption would be a pressure change on the order of 0.7 inches of Hg (25.4 mm of Hg at 22 °C) from an initial atmospheric pressure of 15.0333 psia. This would then create a pressure differential of about 1/3 psig (2.28 kPa above atmosphere) higher on the inside of the isolated containment building than on the inside of the adjacent laboratory building, which surrounds most of the containment structure. With an initial internal pressure in the containment building of 15.0333 psia, it would contain 230,102 standard cubic feet (scf) of air. The conservative assumption is made that the containment building will leak at a rate slightly greater than the Technical Specification (TS) leakage rate limit. The TS leakage rate limit shall not exceed either 16.3 ft3/min (STP) with an overpressure of one pound per square inch gauge or 10% of the contained volume over a 24-hour period from an initial overpressure of two pounds per square inch gauge. Additionally, the minimum TS free volume of the containment building is 225,000-ft3 at standard pressure and temperature. The following equation represents the air leakage rate from the containment building in standard cubic feet per minute (scfm) as a function of containment pressure which at 1 psi over pressure would corresponds to 17.68 ft3/minute. This would correspond to a leakage rate 8.4 % greater than the TS limit of 16.3 ft 3/minute at 1.0 psig. LR 17.68 x (CP-14.7) 112 where: LR leakage rate from containment (scfm); and CP containment pressure (psia). Using this equation for the assumed initial overpressure condition of 0.333 psig (2.28 kPa above atmosphere), it would take approximately 16.5 hours for the leakage rate to decrease to zero from an initial leakage rate of approximately 10.25 scfm, which would occur at the start of the event. The average leakage rate over the 16.5-hour period would be approximately 5.15 scfm. Attached is an Excel spreadsheet (Attachment 5) that provides the leakage rates from the point of when the accident occurs (t = 0) to the point when the isolated containment building has equalized in pressure with the adjacent laboratory building (t = 16.5 hrs). 10of54

The analysis uses an iodine reduction factor of 75 percent. The NRC staff noted (from Attachment

13) that MURR previously used an iodine reduction factor of 50 percent in accordance with the guidance in Regulatory Guide 1.3 in support of MURR License Amendment No. 8. The NRC staff needs additional information to understand the technical basis for use ofan iodine reduction factor of 75 percent in the current fuel failure analyses [page 31}.

The analysis has been revised using the guidance of Regulatory Guide 1.3, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors," which employs a 50% reduction of radioiodines from plate-out and deposition.

The analysis indicates that there was no decay applied to the isotopes, but it appears that the decay of source isotopes was performed Explain [page 32].

No decay is applied to the isotopes from the time they exit the ventilation exhaust stack to when they reach the receptor point in the unrestricted area. Decay is only calculated for the time the isotopes are in the isolated containment building, not after they leak out of the containment structure due to an assumed worst-case pressure differential.

Derived Air Concentration values for Kr-89, Kr-90, Xe-137, and Xe-139 were divided by 300. Title JO of the Code ofFederal Regulations Part 20, Appendix B, uses a factor o/219, which the NRC staff will also use. Consider revision using a factor of219 [page 35].

The "Fuel Failure during Reactor Operation" accident analysis has been revised using the 10 CFR 20 AppendixB factor of219, instead of300. 11of54

Revised "Fuel Failure during Reactor Operation" (Formerly the MHA) 13.2.1 Fuel Failure during Reactor Operation 13 .2.1.1 Accident-Initiating Events and Scenarios Many types of accidents have been considered in conjunction with the operation of the MURR. In all cases, safety systems have been designed such that the likelihood of an accident involving the release of a significant amount of fission products has essentially been eliminated. The safety systems take the form of automatic reactor shutdown circuits and process systems designed to ensure, through redundancy, that the reactor will shut down upon a significant deviation from normal operating conditions. In addition, the reactor is housed within a containment building, thus providing further protection against a significant release of radioactive material to the environment. In the "Fuel Failure during Reactor Operation" accident for the MURR, it is assumed that an accident condition has caused the melting of the number-I fuel plate in four (4) separate fuel elements (Ref. 13.11). It is further assumed that the four (4) number-I fuel plates are in the peak power region of the core. While one might postulate that this accident could result from a partial flow blockage to the fuel, mitigating features such as the primary coolant system strainer, the fuel element end-fittings, and the pre-operational visual inspection of the reactor pressure vessels and core region following any fuel handling evolution, all prevent an accident of this type from occurring. In addition, it has been shown that a 75% blockage of coolant flow to the hot channel is insufficient to cause cladding failure (Ref. 13.2). 13 .2.1.2 Accident Analysis and Consequences The fuel failure during reactor operation accident postulates partial fuel melting with an associated release of fission products into the primary coolant system. The accident is assumed to occur with the primary coolant system operating, resulting in a quick dispersal of the fission products throughout the system. With the design of the primary coolant system and its associated systems, particulate activity will remain in the coolant, and the gaseous activity that comes out of solution will collect in the reactor loop vent system and be retained there. Therefore, the primary coolant system relief valves and pressurizer are the only paths for a release of significant quantities of fission products to the environment. The potential energy release from the melting of four (4) number-I fuel plates could occur as a possible metal-water reaction (Ref. 13.3). While hydrogen would be formed, it is highly unlikely that in a water environment a hydrogen deflagration reaction would occur. The amount of material which would be involved in a metal-water reaction under the conditions of four (4) number-I fuel plates melting is not predictable as the amount is dependent upon many conditions. For purposes of calculation, it is conservatively assumed that all the fuel plate aluminum cladding exposed in the area I2of54

is involved in the reaction. The reactor core contains a total of 33 .56 Kg of aluminum. Of this, 1.3% or 436 grams is assumed to react according to the following equation: Al + nH20 ---? AlOn + nH2 + 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 coolant in the reactor core. Additionally, any steam that would form in the vicinity of the molten area would also assist in dissipating the heat. Since the fuel failure would result in a negligible release of energy to the primary coolant system, the introduction of pressure surges, which could lift the primary relief v&lves, are not considered credible. The pressurizer is an isolated system, and since no significant pressure surges are anticipated, it will not be subject to mixing with the primary coolant system. Any significant gaseous radioactivity entrapped in the reactor loop vent tank will cause a reactor scram and actuation of the containment building isolation system by action of the pool surface radiation monitor. Additionally, following actuation of the anti-siphon system when the primary coolant system is secured, gases could also collect in the anti-siphon pressure tank. The location of these tanks under the pool surface, and the shielding provided by the water and the biological shield, will significantly reduce any radiation exposure to the reactor staff, visitors, or researchers. Fission products entrapped in the primary coolant system can be removed by the reactor coolant cleanup system. This cleanup procedure would be undertaken under closely monitored and controlled conditions. The primary coolant system does experience some coolant leakage into the reactor pool through the pressure vessel head packing and flange gasket. This leakage is typically less than 40 gallons (1511) per week; an almost imperceptible leakage rate of approximately 4 x 10-3 gallons of primary coolant per minute into the pool. However, for purposes of calculation, a leakage rate of 80 gallons (303 1) per week is used. Based on this assumed conservative leakage rate, the radiation exposure to personnel in the containment building following fuel failure is calculated below. Due to the complex nature of the MURR mixed core fuel cycle, the source terms for the "Fuel Failure during Reactor Operation" and "Fuel Handling Accident" were determined using the computer program MONTEBURNS instead of simplistic ORIGEN runs (Attachment 1). MONTEBURNS is a coupled MCNP-ORIGEN code system developed by Los Alamos National Laboratory (LANL). Within the MONTEBURNS program, MCNP calculations are used to obtain accurate one-group fluxes and one-group cross sections that are then utilized by ORIGEN for fuel depletion and fission product activity calculations. The use of MONTEBURNS and the flow diagram utilized for various neutronic computations at MURR were described in detail in the response to Request for Additional Information (RAI) Question 3 .a (MURR letter dated October 1, 2015). 13of54

Radioiodine and Noble Gas Activities in the Core 131 1 - 2.20 x 10+05 Ci 85Kr - 4.63 x lO+oi Ci 133 Xe - 3.85 x 10+05 Ci 132 1 - 3.08 x IO+o5 Ci 85 mKr - I .3 I x I 0+05 Ci 135 Xe - 7.56 x I0+04 Ci 133 1 - 5.42 x IO+o 5 Ci 87Kr - 2.05 x I0+05 Ci 135 mxe - 3 .62 x I 0+04 Ci 1341 - 6.1 Ix I0+05 Ci 88Kr - 2.9I x 10+05 Ci 137 Xe - 4.8 I x I 0+05 Ci 1351 - 5.06 x IO+o5 Ci 89Kr - 3.69 x 10+05 Ci 138Xe - 5.0I x IO+o5 Ci 90Kr - 3.68 x I0+05 Ci 139Xe - 4.07 x I0+05 Ci An unirradiated fuel plate number-I contains, on average, I9.26 grams of mu, so four (4) unirradiated number-I fuel plates contain 77.04 grams of 235 U instead of the 78.58 grams assumed in the 2006 Safety Analysis Report analysis. These four number-I fuel plates that melt correspond to I .4 I% of the total mu in the Week 58 Core that was used to determine the high power peaking factor for the revised Safety Limits. The Week 58 Core has a total power history of 576 MWd. This power history results in a total reduced core mass of 5,474 grams of mu due to the previous fuel consumption. This l.4I% of 235 U melting releases 3.42% of the core fission products due to the highest power density fuel plate number-I overall power peaking factor of 2.423, which is conservatively assumed to apply to all four (4) number-I fuel plates (1.41%x2.423 = 3.42%). A very conservative value of a 100% release of the radioiodine and noble gas fission products from the fuel is assumed in calculating the fission product inventory in the primary coolant system. It is also assumed that fission products released into the primary coolant are quickly and uniformly dispersed within the 2,000-gallon (7,571 I) primary coolant system volume and, during a normal week's operation, 80 gallons (7.9 x 10-3 gpm) of coolant leaks from the primary coolant system into the pool water. Therefore, the radioactivity released into the reactor pool in IO minutes - determined to be the maximum personnel occupancy time in the containment building after the accident for necessary operational personnel - is as follows: (Note: It would take approximately five (5) minutes for Operations personnel to secure the primary coolant system and verify that the containment building has been evacuated following a containment building isolation. For the purposes of the fuel failure calculations, a conservative assumption of 10 minutes is used.) Example calculation of 131 1 released into the reactor pool during the 10-minute leakage period: 131 1 in fuel x 0.0342 x 1/2,000 gal x (7.9 x 10-03 gpm) x 10 min 2.20 x 10+05 Ci x 0.0342 x 1/2000 gal x 0.079 gal 0.297 Ci 2.97 X 10+05 µCi Note: Same calculation is used for the other isotopes listed below. 14of54

Radioiodine and Noble Gas Activities Released Into the Pool after 10 Minutes 131 1 - 2.97 x 10+05 µCi 85Kr - 6.25 x 1o+oz µCi 133 Xe - 5.20 x 10+05 µCi 132 1 - 4.16 x 10+05 µCi 85 mKr - 1. 77 X 1O+os µCi 135 Xe - 1.02 x 1o+o5 µCi 133 1 - 7.32 x 10+os µCi 87Kr - 2. 77 X 10+05 µCi 135mxe - 4.89 x 10+04 µCi 1341 - 8.25 X 10+05 µCi 88Kr - 3 .93 X 10+05 µCi 137 Xe - 6.50 x 1o+o 5 µCi 1351 - 6.84 X 1O+os µCi 89Kr - 4.98 x 10+05 µCi 138 Xe - 6.77 x 10+05 µCi 90Kr - 4.97 x 10+05 µCi 139Xe - 5.50 x 10+05 µCi Fission products released into the reactor pool will be detected by the pool surface and ventilation system exhaust plenum radiation monitors. For the purposes of this analysis, it is assumed that a reactor scram and actuation of the containment building isolation system occurs by action of the pool surface radiation monitor. The radioiodine released into the reactor pool over a 10-minute interval is conservatively assumed to be instantly and uniformly mixed into the 20,000 gallons (75,708 I) of bulk pool water, which then results in the following pool water concentrations for the radioiodine isotopes. The water solubility of the krypton and xenon noble gases released into the pool over this same time period are ignored and they are assumed to pass immediately through the pool water and evolve directly into the containment building air volume where they instantaneously form a uniform concentration in the isolated structure. Radioiodine Concentrations in the Pool Water at 10 Minutes 131 1 - 1.49 x 10+01 µCi/gal 133 1 - 3.66 x 10+01 µCi/gal 1351 - 3.42 x 10+01 µCi/gal 132 1 - 2.08 x 10+01 µCi/gal 1341 - 4.13 x 10+01 µCi/gal When the reactor is at 10 MW and the containment building ventilation system is in operation, the evaporation rate from the reactor pool is approximately 80 gallons (302.8 L) of water per day. For the purposes of this calculation, it is assumed that a total of 40 gallons (151.4 L) of pool water containing the previously listed radioiodine concentrations evaporates into the containment building over the 10-minute period. This assumption results in about seventy times more radioiodine in the containment building air than would be present at the end of the 10 minutes of evaporation. In addition, air with a temperature of 75 °F (23.9 °C) and 100% relative humidity contains H 20 vapor equal to 40 gallons (151.4 L) of water. Since the air in containment is normally at about 50% relative humidity, thus containing approximately 20 gallons (75.7 L) of water vapor, the assumed addition of 40 gallons (151.4 L) of water vapor over estimates by a factor of two the amount of water that would make the containment air saturated. It is also conservatively assumed that all of the radioiodine activity in the 40 gallons (151.4 L) of pool water that evaporates instantaneously forms a uniform concentration in the containment building air. When distributed into the containment building, this would result in the following radioiodine concentrations in the 225,000 ft3 (6,371.3 m3) air volume: 15of54

131 Example calculation of average 1 in containment air during the tenth minute: 131 1 concentration in pool water at 10 minx (9.5 min/10 min) x 40 gal x (9.5 min/10 min) x 1/225,000 :ft3 x 1 ft3/28,317 cc x EXP[(-0.693 x 9.5 min) I (8.02 day x 1440 min/day)] (1.49 x 10+01 µCi/gal) x 0.95 x 38 gal x 1/225,000 :ft3 x 1 ft3/28,317 ml x 0.9994 8.41 x 10-08 µCi/ml The average radioiodine concentrations are determined for the first and last minute and four 2-minute concentrations for the other eight (8) minutes. The overall average concentration is then calculated. Average Radioiodine Concentrations in the Containment Building Air during the 10 Minutes 131 1 - 3.08 x 10*08 µCi/ml 133 1 - 7.56 x 10-08 µCi/ml 135 1 - 7.00 x 10-08 µCi/ml 132 1 - 4.16 x 10-08 µCi/ml 134 1 - 7.77 x 10-08 µCi/ml As noted previously, the krypton and xenon noble gases released into the reactor pool from the primary coolant system during the assumed 10-minute interval following fuel failure are assumed to pass immediately through the pool water (ignoring their solubility in water) and enter the containment building air volume where they instantaneously form a uniform concentration in the isolated structure. (Note: Within 10 minutes the primary coolant system is shut down and secured; therefore, the leakage driving force is stopped.) Based on the 225,000-:ft3 (6,371.3 m 3) volume of containment building air and the previously listed Curie quantities of these gases released into the reactor pool, the average noble gas concentrations in the containment building for each time interval of the 10 minutes would be as follows: 85 Example calculation of average Kr released into containment air during the 2-minute time interval between three (3) to five (5) minutes: 85 Total Kr activity x (4 min/10 min) x 1/ (225,000 ft3 x 28,317 cc/ft3) x EXP[(-0.693 x 4 min) I (10.76 yr x 365.25 days/yr x 1,440 min/day)] 6.25 x 10+oz uCi x 0.4 I (6.371x10+ 09 ml) x 1.0 3 .93 X 10-os µCi/ml The 10-minute average noble gas concentrations are calculated from the average of the first and last I-minute time intervals and the other four (4) 2-minute time intervals. Average Noble Gas Concentrations in the Containment Building Air during the 10 Minutes 85 133 Kr - 4.91 X 10-os µCi/ml Xe - 4.08 x 10*05 µCi/ml 85 mKr - 1.3 7 x 10*05 µCi/ml 135 Xe - 7.95 x 10-06 µCi/ml 87 Kr - 2.05 x 10*05 µCi/ml 135 mxe - 2.86 x 10-06 µCi/ml 88 Kr - 3.00 x 10*05 µCi/ml 137 Xe - 1.69 x 10*05 µCi/ml 89 Kr - 1.05 x 10*05 µCi/ml 138 Xe - 3.86 x 10*05 µCi/ml 139 90 Kr - 4.84 x 10-07 µCi/ml Xe - 8.08 x 10*07 µCi/ml 16 of 54

The objective of this calculation is to present a worst-case ~ose assessment for an individual who remains in the containment building for 10 minutes following fuel failure. Therefore, as noted previously, the radioactivity in the evaporated pool water is assumed to be instantaneously and uniformly distributed into the building once released into the air. Based on the source term data provided, it is possible to determine the radiation dose to the thyroid from radioiodine and the dose to the whole body resulting from submersion in the airborne noble gases and radioiodine inside the containment building. As previously noted, the exposure time for this dose assessment is 10 minutes. Because the airborne radioiodine source is composed of five (5) different iodine isotopes, it will be necessary to determine the dose contribution from each individual isotope and then sum the results. Dose Multiplication Factors were established using the Derived Air Concentrations (DACs) for the "listed" isotopes in Appendix B of 10 CFR 20 and calculated values for the four (4) "unlisted" submersion isotopes (Kr-89, Kr- 90, Xe-137 and Xe-139). The submersion DAC values that were calculated were done in accordance with the data and methodology as supplied in Federal Guidance Report (FOR) No.12 (Attachment 4). 131 Example calculation of thyroid dose due to 1: The DAC can also be defined as 50,000 mrem (thyroid target organ limit)/2,000 hrs, or 25 mrem/DAC-hr. Additionally, 10 minutes of one (1) DAC-hr is 1.67 x 10-01 DAC-hr. 131 1 concentration in containment 3.08 x 10-08 µCi/ml 131 1DAC (10 CFR20) 2.00 x 10-08 µCi/ml Dose Multiplication Factor = (131 1 concentration) I (1311 DAC) (3.08 x 10-08 µCi/ml) I (2.00 x 10-08 µCi/ml)

                                                =      1.54 131 Therefore, a 10-minute thyroid exposure from         1 is:

Dose Multiplication Factor x DAC Dose Rate x 10 minutes 1.54 x (25 rnrem/DAC-hr) x (1.67 x 10-01 DAC-hr) 6.42 mrem Note: Same calculation is used for the other radioiodines listed below. 17of54

Derived Air Concentration Values and 10-Minute Exposures -Radioiodine Radionuclide Derived Air Concentration 10-Minute Exposure 2.00 x 10-08 µCi/ml 6.42 x 1o+oo mrem 06 3.00 x 10- µCi/ml 5.78 x 10-02 mrem 1.00 x 10-01 µCi/ml 3.15 x lO+oo mrem 2.00 x 1o-05 µCi/ml 1.62 x 10-02 mrem 7.00 x 10-07 µCi/ml 4.17 x 10-01 tnrem Total = 10.07 mrem Doses from the kryptons and xenons present in the containment building are assessed in much the same manner as the radioiodines, and the dose contribution from each individual radionuclide must be calculated and then added together to arrive at the final noble gas dose. Because the dose from the noble gases is only an external dose due to submersion, and because the DACs for these radionuclides are based on this type of exposure, the individual noble gas doses for 10 minutes in containment were based on their average concentration in the containment air and the corresponding DAC. Example calculation of whole body dose due to 85Kr: The DAC can also be defined as 5,000 mrem I 2,000 hrs, or 2.5 mrem/DAC-hr. Additionally, 10 minutes of one (1) DAC-hr is 1.67 x 10-01 DAC-hr. 85

         .Kr concentration in containment           4.91x10-08 µCi/ml 85 Kr DAC (10 CFR 20)                         1.00 x 1o-04 µCi/ml 85                    85 Dose Multiplication Factor                   ( Kr concentration) I ( Kr DAC)

(4.91x10-08 µCi/ml) I (l.00 x 10-04 µCi/ml) 4.91x10-04 Therefore, a 10-minute whole body exposure from 85Kr is: Dose Multiplication Factor x DAC Dos~ Rate x 10 minutes

   =   4.91x10-04 x (2.5 mrem/DAC-hr) x (1.67 x 10-01 DAC-hr) 2.05 x 1o-04 mrem Note: Same calculation is used for the other noble gases listed below.

18of54

Derived Air Concentration Values and 10-Minute Exposures -Noble Gases Radionuclide Derived Air Concentration 10-Minute Exposure 85.Kr 1.00 x 10*04 µCi/ml 2.05 x 10*04 mrem 85m.Kr 2.00 x 10*05 µCi/ml 2.85 x 10*01 mrem 87.Kr 5.00 x 10-06 µCi/ml 1.71 x lO+oo mrem 88.Kr 2.00 x 10-06 µCi/ml 6.26 x lO+oo mrem 89.Kr 1.90 x 1o-06 µCi/ml 2.31 x 1o+oo mrem 90.Kr 2.80 x 10-06 µCi/ml 7 .20 x 10-01 mrem 133 Xe 1.00 x 10*04 µCi/ml 1.70 x 10*01 mrem 135 Xe 1.00 x 10*05 µCi/ml 3 .31 x 10*01 mrem J35mxe 9.00 x 10-06 µCi/ml 1.32 x 10*01 mrem 137 Xe 2.00 x 10*05 µCi/ml 3 .53 x 10*01 mrem 138 Xe 4.00 x 1o-06 µCi/ml 4.03 x 10+00 mrem 139 Xe 3.70 x 10-06 µCi/ml 9.10 x 10-02 mrem Total = 16.38 mrem To finalize the occupational dose in terms of Total Effective Dose Equivalent (TEDE) for a 10-minute exposure in the containment building after fuel failure, the doses from the radioiodines and noble gases must be added together, and result in the following values: 10-Minute Dose from Radioiodines and Noble Gases in the Containment Building Total Iodine-Committed Dose Equivalent (CDE) 10.07 mrem Iodine - Committed Effective Dose Equivalent (CEDE) (CEDE= CDE x 0.03) 0.302 mrem Noble Gas - Committed Effective Dose Equivalent (CEDE) 16.38 mrem Total Dose-Total Effective Dose Equivalent (TEDE) 16.69 mrem By comparison of the maximum TEDE and CDE for those occupationally-exposed during fuel failure to applicable NRC dose limits in 10 CFR 20, the final values are shown to be well within the published regulatory limits and, in fact, lower than 1% of any occupational limit. Radiation shine through the containment structure was also evaluated when considering accident conditions and dose consequences to the public and MURR staff. Calculation of exposure rate from a fuel failure was performed using the computer program MicroShield 8.02 with a Rectangular Volume - External Dose Point geometry for the representation of the containment structure (Attachment 6). MicroShield 8.02 is a product of Grove Software and is a comprehensive photon/gamma ray shielding and dose assessment program that is widely used by industry for designing radiation shields. The exposure rate values provided below represents the radiation fields at 1 foot (30.5 cm) from a 12-inch (30.5 cm) thick ordinary concrete containment wall and at the Emergency Planning Zone (EPZ) boundary of 150 meters (492.1 ft). The airborne concentration source terms used to develop the exposure rate values are similar to those used for determining the dose to a worker within 19of54

containment from noble gases. For radioiodine, the total iodine activity that leaked to the pool system in the 10-minute period was used for the dose calculations. The source term also assumes a homogenous mixture of nuclides within the containment free volume. Radiation Shine through the Containment Building Exposure Rate at 1-Foot from Containment Building Wall: 1.374 mrem/hr Exposure Rate at Epiergency Planning Zone Boundary (150 meters): 0.0094 mrem/hr As noted earlier in this analysis, the containment building ventilation system will shut down and the building itself will be isolated from the surrounding areas. Fuel failure will not cause an increase in pressure inside the reactor containment structure; therefore, any air leakage from the building will occur as a result of normal changes in atmospheric pressure and pressure equilibrium between the inside of the containment structure and the outside atmosphere. It is highly probable that there will be no pressure differential between the inside of the containment building and the outside atmosphere, and consequently there will be no air leakage from the building and no radiation dose to members of the public in the unrestricted area. However, to develop what would clearly be a worst-case scenario, this analysis assumes that a barometric pressure drop has occurred in conjunction with fuel failure. An extreme assumption would be a pressure change on the order of 0.7 inches of Hg (25.4 mm of Hg at 22 °C) from an initial atmospheric pressure of 15.0333 psia. This would then create a pressure differential of about 1/3 psig (2.28 kPa above atmosphere) higher on the inside of the isolated containment building than on the inside of the adjacent laboratory building, which surrounds most of the containment structure. With an initial internal pressure in the containment building of 15.0333 psia, it would contain 230, 102 standard cubic feet (set) of air. The conservative assumption is made that the containment building will leak at a rate slightly greater than the Technical Specification (TS) leakage rate limit. The TS leakage rate limit shall not exceed either 16.3 ft3/min (STP) with an overpressure of one pound per square inch gauge or 10% of the contained volume over a 24-hour period from an initial overpressure of two pounds per square inch gauge. Additionally, the minimum TS free volume of the containment building is 225,000-ft3 at standard pressure and temperature. The following equation represents the air leakage rate from the containment building in standard cubic feet per minute (scfm) as a function of containment pressure which at 1 psi over pressure would corresponds to 17.68 ft3/minute. This would correspond to a leakage rate 8.4% greater than the TS limit of 16.3 ft 3/minute at 1.0 psig. LR 17.68 x (CP-14.7) 112 ; where: LR leakage rate from containment (scfm); and CP containment pressure (psia). 20 of 54

Using this equation for the assumed initial overpressure condition of 0.333 psig (2.28 kPa above atmosphere), it would take approximately 16.5 hours for the leakage rate to decrease to zero from an initial leakage rate of approximately 10.25 scfm, which would occur at the start of the event. The average leakage rate over the 16.5-hour period would be approximately 5.15 scfm. This conservatively over calculates the actual amount of activity that would leak out of the containment building and potentially expose someone in this assumed accident. Several factors exist that will mitigate the radiological impact of any air leakage from the containment building following fuel failure. First of all, most leakage pathways from containment discharge into the laboratory building, which surrounds the containment structure. Since the laboratory building ventilation system continues to operate during fuel failure, leakage air captured by the ventilation exhaust system is mixed with other building air, and then discharged from the facility through the exhaust stack at a rate of approximately 30,500 scfm. Mixing of containment air leakage with the laboratory building ventilation flow, followed by discharge out the exhaust stack and subsequent atmospheric dispersion, results in extremely low radionuclide concentrations and very small radiation doses in the unrestricted area. A tabulation of these concentrations and doses is given below. A second factor which helps to reduce the potential radiation dose in the unrestricted area relates to the behavior of radioiodine, which has been studied extensively in the containment mockup facility at Oak Ridge National Laboratory (ORNL). From these experiments, it was shown that up to 75% of the iodine released will be deposited in the containment vessel. For the purposes of this analysis, MURR used the more conservative NRC-accepted value of 50% reduction of radioiodines from plate-out and deposition (Regulatory Guide 1.3, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors"). Thus, if due to this 50% iodine deposition in the containment building, each cubic foot of air released from the containment structure has a radioiodine concentration that is 50% of each cubic foot within the containment building air, then the radioiodine concentrations leaking from the containment structure into the laboratory building, in microcuries per milliliter, will be: 131 Example calculation of the average 1 released through the exhaust stack during the first hour: Note: The average cubic foot per hour leakage rate can be calculated for each time interval. ( 131 1 activity in containme~t x leakage rate in scf/hr x 0.50) I (30,500 ft3/min x 60 min/hr x 28,317 cc/ft3)

    =      (2.58 x 10*03 µCi/ft3 x 595.6 scf/hrx 0.50) I (5.18 x 10+10 ml) 1.48 x 10*11 µCi/ml Note: Same calculation is used for the other radioiodines listed below.

Radioiodine Concentrations in Air Leaking from Containment and Exiting the Exhaust Stack during the First Hour 131 1 - 1.48 x 10* 11 µCi/ml 133 1 - 3.60 x 10*11 µCi/ml 135 1 - 3.24 x 10*11 µCi/ml 132 1 - 1.79 x 10* 11 µCi/ml 134 1 - 2.78 x 10*11 µCi/ml 21of54

Example calculation of 85Kr released through the exhaust stack in the first hour: 85 ( Kr activity in containment x leakage rate in scf/hr) I (30,500 ft3/min x 60 min/hr x 28,317 cc/ft3) , (2.72 x 10* µCi/scfx 595.6 scf/hr) I (5.18 x 10+ 10 ml) 03 3.13 x 10*11 µCi/ml Note: Same calculation is used for the other noble gases listed below. Noble Gas Concentrations in Air Leaking from Containment and Exiting the Exhaust Stack in the First Hour 85 Kr - 3.13 x 10-11 µCi/ml 87 Kr - 1.05 x 10-08 µCi/ml 89 Kr - 3.39 x 10*11 µCi/ml 85 mKr - 8.19 X 10*09 µCi/ml 88 Kr - 1. 74 x 10-08 µCi/ml 90 Kr - 4.20 x 10*25 µCi/ml 133 Xe - 2.59 x 1o-08 µCi/ml 135 mxe - 6.29 x 10-10 µCi/ml 138 Xe - 7.75 x 10*09 µCi/ml 135 Xe - 4.92 x 10*09 µCi/ml 137 Xe - 1.41 x 10*10 µCi/ml 139 Xe - 6.22 x 10*22 µCi/ml The same type of calculation is used for all the radioiodines during the five (5) 1-hour intervals, the two (2) 4-hour intervals, and the one (1) 3.5-hour interval to determine the average concentration in the exhaust stack, which is listed below. Average Radio iodine Concentrations in Air Leaking from Containment and Exiting the Exhaust Stack 131 1 - 7.56 x 10*12 µCi/ml 133 1 - 1.59 x 10*11 µCi/ml 135 1 - 1.06 x 10*11 µCi/ml 132 1 - 3.37 x 10*12 µCi/ml 134 1 - 2.91 x 10*12 µCi/ml Note: Same calculation is used for the noble gases listed below. Noble Gas Concentrations in Air Leaking from Containment and Exiting the Exhaust Stack 85 Kr - 1.62 x 10* 11 µCi/ml 87 Kr - 1.3 8 x 10*09 µCi/ml 89 Kr - 2.06 x 10*12 µCi/ml 85 88 mKr - 2.26 X 10*09 µCi/ml Kr - 3.75 x 10*09 µCi/ml 90 Kr - 2.55 x 10*26 µCi/ml 133 Xe - 1.31 x 1o-08 µCi/ml 135 mxe - 4.06 x 10~ 11 µCi/ml 138 Xe - 4.94 x 10*10 µCi/ml 135 Xe - 1.80 x 10*09 µCi/ml 137 Xe - 8.53 x 10*12 µCi/ml 139 Xe - 3. 77 x 10*23 µCi/ml Assuming, as stated earlier, that (1) the average exhaust gas concentration due to the containment leakage rate applies for the 16.5-hour period in order to equalize the containment building pressure with atmospheric pressure, (2) the flow rate through the facility's ventilation exhaust stack is 30,500 scfm, (3) the reduction in concentration from the point of discharge at the exhaust stack to the point of maximum concentration in the unrestricted area is a factor of 292, and (4) after exiting the ventilation exhaust stack, conservatively there is assumed no decay of any radioiodines or noble gases. The following concentrations of radioiodirtes and noble gases with their corresponding radiation doses are calculated to occur in the unrestricted area. The values listed are for the point of maximum concentration in the unrestricted area assuming a uniform, semi-spherical cloud geometry 22of54

for noble gas submersion and further assuming that the most conservative (worst-case) meteorological conditions exist for the entire 16.5-hour period of containment leakage following fuel failure. Radiation doses are calculated for the entire 16.5-hour period. Dose values for the unrestricted area were obtained using the same methodology that was used to determine doses inside the containment building, and it was assumed that an individual was present at the point of maximum concentration for the full 16.5 hours that the containment building was leaking. A worst-case scenario effluent dilution factor of 292 using the Pasquill-Guifford Model for atmospheric dilution is used in this analysis. It is assumed that all offsite (public) dose occurs under these atmospheric conditions at the site of interest, i.e. 760 meters north of MURR. In our case, at 760 meters, it occurs only during Stability Class 'F' conditions, which normally only occurs 11.4% of the time when the wind blows from the south. Thus this calculation is conservative. 10 CFR 20 Appendix B Effluent Concentration Limits are used for the "listed" isotopes. Effluent Concentration Limits were calculated for each of the four (4) "unlisted" noble gases (Kr-89, Kr-90, Xe-137 and Xe-139) using the data and methodology contained in FGR No. 12 for submersion isotopes. The DAC value was first calculated and then a factor of 219 was applied using 10 CFR 20, Appendix B, as a reference point for DAC values from submersion isotopes. Exposure at 1 DAC equates to 5000 mrem per year whereas at the Effluent Concentration Limit it is 50 mrem per year. Thus, there is a factor of 100 times lower allowable dose for the Effluent Concentration Limit as compared to the DAC. Exposure at the Effluent Concentration Limit assumes you are in that effluent concentration for 8760 hours per year. Therefore, the time assumed to be exposed to the Effluent Concentration Limit is a factor of 4.38 times longer than the 2000 hours per year that defines a DAC. No credit is taken for transit time from the exhaust stack to the receptor point nor is credit taken for decay inside the containment building until release. In the case ofKr-89 and Xe-137, the transit time alone would be approximately one (1) half-life while the transit time for Kr-90 and Xe-139 would be at least four (4) half-lives. The DAC values were then divided by 219 to arrive at the effluent concentration limits for the calculated isotope limits, consistent with the methodology used in Appendix B of 10 CFR 20. 131 Example calculation of whole body dose in the unrestricted area due to 1: Conversion Factor: (Public dose limit of 50 mrem/yr) x (1 yr/8760 hrs)= 5.71 x 10*03 mrem/hr 131 1 concentration 7.56 x 10* 12 µCi/ml 131 1 effluent concentration limit 2.00 x 10* 10 µCi/ml 131 1 Conversion Factor 5. 71 x 10*03 mrem/hr 131 Therefore, a 16.5-hour whole body exposure from 1 is: 131 1 concentration I ( 131 1 Effluent Concentration Limit x Conversion Factor x 16.5 hrs) 7.56 x 10*12 µCi/ml I (2.00 x 10*10 µCi/ml x 5.71x10*03 mrem/hr x 16.5 hrs/292) 1.22 x 10*05 mrem 23of54

Note: Same calculation is used for the other isotopes (radioiodines and noble gases) listed below. Effluent Concentration Limits, Concentrations at Point of Maximum Concentration and Radiation Doses in the Unrestricted Area - Radioiodine Radionuclide Effluent Limit Maximum Concentration1 Radiation Dose 2.00 x 1o-io µCi/ml 2.59 x 10-14 µCi/ml 1.22 x 10-os mrem 2.00 x 1o-08 µCi/ml 1.16 x 10-14 µCi/ml 5.44 x 10-08 mrem 1.00 x 1o-09 µCi/ml 5.44 x 10-14 µCi/ml 5.13 x 10-06 mrem 6.00 x 1o-08 µCi/ml 9.97 x 10-15 µCi/ml 1.57 x 10-08 mrem 6.00 x 1o-09 µCi/ml 3.63.x 10-14 µCi/ml 5.70 x 10-07 mrem Total = 1.SOE-05 mrem Note 1: Maximum Concentrations are radioiodine and noble gas concentrations leaking from the containment building and exiting the exhaust stack reduced by a dilution factor of 292. Effluent Concentration Limits, Concentrations at Point of Maximum Concentration and Radiation Doses in the Unrestricted Area - Noble Gases Radionuclide Effluent Limit Maximum Concentration1 Radiation Dose 85Kr 7.00 x 10-01 µCi/ml 5.56 x 10-14 µCi/ml 2.19 x 10-06 mrem 85mKr 1.00 x 1o-07 µCi/ml 7.75 x 10-12 µCi/ml 2.13 x 10-03 mrem 87Kr 2.00 x 1o-08 µCi/ml 4.71x10- 12 µCi/ml 6.49 x 10-03 mrem 88Kr 9.00 x 10-09 µCi/ml 1.28 x 10-11 µCi/ml 3 .93 x 10-02 mrem 89Kr 8.60 x io-09 µCi/ml 7.04 X 10"15 µCi/ml 2.25 x 10-05 mrem 90Kr 1.20 x 1o-08 µCi/ml 8.72 x 10*29 µCi/ml 2.00 x 10*19 mrem 133 Xe 5.00 x 10-01 µCi/ml 4.48 x 10-11 µCi/ml 2.47 x 10-03 mrem 10-08 µCi/ml 6.17 x 10*12 µCi/ml 2.42 x 10*03 mrem 135 Xe 7.00 x 13smxe 4.00 x 1o-08 µCi/ml 1.39 x 10* 13 µCi/ml 9.56 x 10-os mrem 10-08 µCi/ml 2.92 x 10-14 µCi/ml 8.83 x 10-06 mrem 137 Xe 9.10 x 138Xe 2.00 x 1o-08 µCi/ml 1.69 x 10*12 µCi/ml 2.33 x 1o-03 mrem 1o-08 µCi/ml 1.29 x 10*25 µCi/ml 2.22 x 10"16 mrem 139 Xe 1.60 x Total = 5.52E-02 mrem Note 1: Maximum Concentrations are radioiodine and noble gas concentrations leaking from the containment building and exiting the exhaust stack reduced by a dilution factor of 292. To finalize the unrestricted dose in terms of Total Effective Dose Equivalent (TEDE), the doses from the radioiodines and noble gases must be added together, and result in the following values: 24of54

Dose from Radioiodines and Noble Gases in the Unrestricted Area Total Iodine - Committed Effective Dose Equivalent (CEDE) 1.80E-05 mrem Noble Gas - Committed Effective Dose Equivalent (CEDE) 5.52E-02 mrem Total Dose-Total Effective Dose Equivalent (TEDE) 5.53E-02 mrem Summing the doses from the noble gases and the radioiodines simply substantiates earlier statements regarding the very low levels in the unrestricted area should a fuel failure occur, and should the containment building leak following such an event. As can be seen in the above analysis, the doses are extremely low; less than 0.1 % of the Part 20 limits for allowable dose to the general public. Additionally, leakage in mechanical equipment room 114 from such items as valve packing, flange gaskets, pump mechanical seals, etc. was also considered in the fuel failure analysis. A realistic leakage rate of 60 milliliters within the 10-minute time interval was used - after 10 minutes the primary coolant system would be shutdown, isolated and depressurized as part of the control room operator's actions. The additional contaminated water vapor and associated isotopes added to the facility ventilation exhaust system made a minimal (<1 %) contribution to the total dose of an individual located in the facility. Therefore, the dose contribution to the unrestricted area would be expected to be approaching zero. 13 .2.1.3 Conclusions Generally, the most severe condition which is analyzed with regard to reactor accidents is either a loss of primary coolant or a loss of primary coolant flow during reactor operation. Both of these accidents are analyzed in this chapter and the results show no core damage. In addition, there are no other accidents that will result in a release of fission products from the reactor fuel, which is assumed in the fuel failure analysis. Even if such an event were to occur, the anti-siphon and reactor loop vent systems are designed such that any released radioactivity would be contained in the primary coolant system. System design and operational procedures reduce the likelihood of any foreign material being introduced into the reactor core that could cause a partial flow blockage. Calculations have been performed which indicate that even partial flow blockage to a fuel element will not result in cladding failure (Ref. 13.2). A considerable margin of safety has been designed into the system in this regard. Also, considering the results of the analyses which show no core damage in the event of a loss of primary coolant or a loss of primary coolant flow accident (See Sections 13.2.3 and 13.2.4), and in view of the design of the anti-siphon and reactor loop vent systems, it is concluded that there is no radiation risk to personnel in the reactor containment building or in the unrestricted area should one of these events occur.

References:

Same as those stated on pages v through vii of Chapter 13 of the SAR. 25of54

Fuel Handling Accident Analysis [page 39] - Information/Clarification Needed

Origin ofcore inventory not clear (similar to comment on page 24 above) [page 40}.

As described in the response to RAI No. 7.g. above, for operation at 10 MW for 1,200 MWd in twelve 10-day cycles over a 300-day period with 6.2 Kg of 235U (normal operating cycle is 6.5 days with a total ofless than 700 MWd on the core), the radioiodine, krypton and xenon activities listed in the "Fuel Failure during Reactor Operation" accident analysis will conservatively be present in the core. Due to the complex nature of the MURR mixed core fuel cycle, the source term was determined using the computer program MONTEBURNS instead of simplistic ORIGEN runs. MONTEBURNS is a coupled MCNP-ORIGEN code system developed by Los Alamos National Laboratory (LANL). Within the MONTEBURNS program, MCNP calculations are used to obtain accurate one-group fluxes and one-group cross sections that are then utilized by ORIGEN for fuel depletion and fission product activity calculations. The use of MONTEBURNS and the flow diagram utilized for various neutronic computations at MURR were described in detail in the response to Request for Additional Information (RAI) Question 3.a (MURR letter dated October 1, 2015). Using MONTEBURNS to simulate the burnup of all eight (8) fuel elements for the core configuration beginning with an all-fresh core for the aforementioned hypothetical operational cycle, the source term for the "Fuel Failure during Reactor Operation" analysis was derived from summing the noble gases and iodine fission products inventories from all eight (8) fuel elements at the end of 3 00 days (i.e., at the End-of-Irradiation of cycle 12). The source term for the "Fuel Handling Accident (FHA)" analysis was similarly derived using the same radionuclides by extending the previous 300-day MONTEBURNS simulation and decaying it for 30 minutes (i.e., 0.0208 days after the End-of-Irradiation of cycle 12). Attachment 1 is the MONTEBURNS output which provides the sources terms for both the "Fuel Failure during Reactor Operation" and "Fuel Handling Accident (FHA)" analyses.

The bracketing in several equations appears inconsistent (similar to comment on page 26 above, and again on page 47) [page 41}.

The "Fuel Handling Accident" analysis has been revised to correct for any bracketing inaccuracies in the equations.

Units used are non-standard (micro-Curies per cubic foot versus micro-Curies per cubic milliliter, similar comment on pages 47 and 48). Verify units [page 41}.

The "Fuel Handling Accident" analysis has been revised to correct for any non-standard units. 26of54

The value of 229,801 cubic feet was used versus the Technical Specification value of 225,000 cubic feet (similar comment on pages 47 and 63). Explain [page 41].

For all three (3) radiological accident analyses the barometric pressure is assumed to be 15.0333 psia at the start of the accident. At that pressure, the containment building would contain 230,102 scf of air [see "Containment Building - Excel Spreadsheet Leakage Rate" (Attachment 5)]. Assuming the slightly greater than Technical Specification allowed leakage on the table, the average volume of air in the containment structure during the first hour after a drop of 1/3 psi would be 229,801 scf. 27of54

                                     "Fuel Handling Accident (FHA)"

All fuel handling is performed in accordance with Special Nuclear Material (SNM) Control and Accounting Procedures and as outlined in the Operations Procedures. Irradiated fuel is handled with a specially designed remote tool. The fuel handling tool is designed to provide a positive indication of latching prior to movement of a fuel element. This feature is tested prior to any fuel handling sequence. Fuel elements are always handled one at a time so that they are maintained in a criticality-safe configuration. New or irradiated fuel may be stored in any one of the 88 in-pool fuel storage locations (not including the core). These storage locations are designed to (I) ensure a geometry such that the calculated l<eff is less than 0.9 under all conditions of moderation, (2) allow sufficient convection cooling, and (3) provide sufficient radiation shielding. Therefore, the fuel handling system provides a safe, effective and reliable means of transporting and handling reactor fuel from the time it enters the facility until it leaves. All cask lifting equipment, including the 15-ton capacity overhead rectilinear crane, is rigorously maintained, including preventive maintenance and testing, as appropriate. Hence, no specific accidents regarding the handling of fuel have been identified for the MURR. However, the probability of dropping a fuel element while underwater and damaging it severely enough to breach the fuel cladding was considered. A conservative potential radionuclide release and calculation of the occupational and public exposures from a Fuel Handling Accident (FHA) are included below. The following calculations determining the postulated dose from a potential release of radioactivity from a fuel element during a handling accident closely follow the "Fuel Failure during Reactor Operation" calculations and methodology for personal exposure due to a release of fission products. The objective of these calculations is to present a worst-case dose assessment for a person who remains in the containment building for five (5) minutes following the release from a breached fuel element. MURR's fuel cycle averages having approximately 40 fuel elements in the cycle - divided into 20 pairs of elements. Paired elements are always loaded opposite each other in the core. All eight (8) fuel elements are replaced every refueling. MURR has averaged refueling the core more than 52 times a year since I 977. This type of accident has never occurred at MURR during any of these fuel handlings. The two outer fuel plates of a fuel element, number-I and -24, are the plates most likely to be damaged during fuel handling. The number-1 fuel plate contains I9.26 grams of mu before irradiation. The highest peak power density in the various MURR core configurations occurs in fuel plate number-I of a fresh fuel element, which has a power peaking factor of 4. I 16 - located between 13.75 to I4.75 inches down from the top of the fuel plate. Fuel plate number-24 has the most surface area to be damaged; however, it has a lower peak power density and contains 45.32 grams of 235U. To be conservative, the analysis assumes that 0.125 grams of mu is exposed from plate number-I during the FHA, which corresponds to removing a section of fuel meat from a plate that is 1-inch square and 5 mils thick. A power peaking factor of 4. I I 6 is also applied. Due to the complex nature of the MURR mixed core fuel cycle, the source terms for the "Fuel Failure during Reactor Operation" and "Fuel Handling Accident" were determined using the computer 28of54

program MONTEBURNS instead of simplistic ORIGEN runs (Attachment 1). MONTEBURNS is a coupled MCNP-ORIGEN code system developed by Los Alamos National Laboratory (LANL). Within the MONTEBURNS program, MCNP calculations are used to obtain accurate one-group fluxes and one-group cross sections that are then utilized by ORIGEN for fuel depletion and fission product activity calculations. The use of MONTEBURNS and the flow diagram utilized for various neutronic computations at MURR were described in detail in the response to Request for Additional Information (RAl) Question 3.a (MURR letter dated October 1, 2015). The following radioiodine, krypton and xenon activities will be present in the MURR core 30 minutes after shutdown from 10 MW full power operation. Refueling typically occurs no sooner than one (1) hour after shutdown. This takes into account the time required to shut down the reactor, to secure the primary coolant system (required to stay in operation a minimum of 15 minutes after the control blades are fully inserted), and to remove the reactor pressure vessel head. For the purposes of the FHA calculations, a conservative assumption of 30 minutes is used .

                 . Radioiodine and Noble Gas Activities in the Core after 30-Minute Decay 131 1 - 2.20 x 10+05 Ci                   85
                                                   .Kr - 4.63 x 10+02 Ci        133 Xe - 3.85 x 10+05 Ci 132 1 - 3.07 x 10+05 Ci                 85 m.Kr - 1.23 x 10+05 Ci         135 Xe - 9 .11 x 10+04 Ci 133 1 - 5.39 x 10+05 Ci                   87
                                                   .Kr - 1.58 x 10+05 Ci      135 mXe - 3.62 x 10+04 Ci 134 1 - 5.49 x 10+05 Ci                   88
                                                  .Kr - 2.58 x 10+05 Ci         137 Xe - 2.24 x 10+03 Ci 135 1 - 4.80 x 10+05 Ci                   89
                                                  .Kr - 5 .28 x 10+02 Ci        138 Xe - 1.16 x 10+05 Ci 90
                                                    .Kr - 6.31 x 10-12 Ci       139 Xe - 7 .89 x 10-09 Ci Fission products released into the reactor pool will be detected by the pool surface and ventilation system exhaust plenum radiation monitors. For the purposes of this analysis, it is assumed that an actuation of the containment building isolation system occurs by action of the pool surface radiation monitor. Actuation of the isolation system will prompt Operations personnel to ensure that a total evacuation of the containment building is accomplished promptly, usually within two (2) to two and a half (2.5) minutes. A conservative 5-minute evacuation time is used as the basis for the stay time in the dose calculations for personnel that are in containment during the FHA.

The following radioiodine and noble gas activities from 0.125 grams of 235 U from the peak power position of fuel plate number- I in the peak power density fuel element are assumed to instantaneously and homogenously distribute in the reactor pool. Example calculation of 131 1 released into the reactor pool: ( 131 1 in fuel I 235 U in core) x 235 U exposed x Power Peaking Factor x (1x10+06 µCi/Ci) (2.20 x 10+05 Ci I 5,474 grams) x 0.125 grams x 4.116 x (1x10+06 µCi/Ci) 2.07 x 10+07 µCi 29 of54

Example calculation of 85.Kr released into the reactor pool: ( 85

               .Kr in fuel I 235 U in core) x 235 U exposed x Power Peaking Factor x (1x10+06 µCi/Ci)

(4.63 x 10+02 Ci/ 5,474 grams) x 0.125 grams x 4.116 x (1x10+06 µCi/Ci) 4.35 x 10+04 µCi Note: Same calculations are used for the other isotopes listed below. Radioiodine and Noble Gas Activities Released into the Pool 131 1 - 2.07 x 10+07 µCi 85

                                                    .Kr - 4.35 x 10+04 µCi        133 Xe - 3 .62 x 10+07 µCi 132 1 - 2.89 x 10+07 µCi                 85 mKr - 1.16 x 10+07 µCi          135 Xe - 8.56 x 10+06 µCi 133 1 - 5.07 x 10+07 µCi                   87
                                                    .Kr - 1.49 x 10+07 µCi      135 mxe - 3.40 x 10+06 µCi 134 1 - 5.16 x 10+07 µCi                   88
                                                    .Kr - 2.42 x 10+07 µCi        137 Xe - 2.11  X 1O+o5 µCi 135                                        89 1 - 4.51x10+07 µCi                        .Kr - 4.96 x 10+04 µCi        138 Xe - 1.09 x 10+07 µCi 90                               139
                                                     .Kr - 5.93 X 10-IO µCi            Xe - 7.42 x 10-07 µCi The radioiodine released into the reactor pool over a 5-minute interval is conservatively assumed to be instantly and uniformly mixed into the 20,000 gallons (75,708 1) of bulk pool water, which then results in the following pool water concentrations for the radioiodine isotopes. The water solubility of the krypton and xenon noble gases released into the pool over this same time period are ignored and they are assumed to pass immediately through the pool water and evolve directly into the containment building air volume where they instantaneously form a uniform concentration in the isolated structure.

Radioiodine Concentrations in the Pool Water 131 1 - 1.03 x 10+03 µCi/gal 133 1 - 2.53 x 10+03 µCi/gal 135 1 - 2:26 x 10+03 µCi/gal 132 1 - 1.44 x 10+03 µCi/gal 134 1 - 2.58 x 10+03 µCi/gal When the reactor is at 10 MW and the containment building ventilation system is in operation, the evaporation rate from the reactor pool surface is approximately 80 gallons (302.8 L) of water per day. For the purposes of this calculation, it is assumed that a total of 20 gallons (75.7 L) of pool water containing the previously listed radioiodine concentrations evaporates into the containment building over the 5-minute period. Containment air with a temperature of 75 °F (23.9 °C) and 100% relative humidity contains H20 vapor equal to 40 gallons (151.4 L) of water. Since the air in containment is normally at about 50% relative humidity, thus containing 20 gallons (75.7 L) of water vapor, the assumed addition of 20 gallons (75.7 L) of water vapor over five (5) minutes will not cause the containment air to be supersaturated. It is also conservatively assumed that all of the radioiodine activity in the 20 gallons (75.7 L) of pool water that evaporates instantaneously forms a uniform concentration in the containment building air. When distributed into the containment building, this would result in the following radioiodine concentrations in the 225,000 ft3 (6,371.3 m3) air volume: 30of54

131 Example calculation of 1 released into containment air to determine the average concentration during the first minute: 131 ( 1 concentration in pool x 4 gal/minx 0.5 min) x EXP[(-0.693 x 0.5 min) I (8.02 day x I440 min/day)] I (225,000 ft3 x 28,3 I 7 ml/ft3) (1.03 x I0+03 µCi/gal x 2 gal) x 0.9997 I (6.37I x 10+09 ml) 3 .25 x I o-07 µCi/ml The same calculation is used for the other isotopes listed below and was performed for each one (I) minute interval. The average radioiodine concentration over the 5-minute interval is the average of the five (5) I-minute intervals. Average Radioiodine Concentrations in the Containment Building Air during the 5-Minute Period 131 1 - 1.62 x 10-06 µCi/ml 133 1 - 3.97 x I o-06 µCi/ml 135 1 - 3 .52 x I o-06 µCi/ml 132 134 1 - 2.23 x 10-06 µCi/ml 1 - 3.88 x 10-06 µCi/ml As noted previously, the krypton and xenon noble gases released into the reactor pool during the 5-minute interval following the FHA are assumed to pass immediately through the pool water and enter the containment building air volume where they instantaneously form a uniform concentration in the isolated structure. This assumption is extremely conservative since it ignores the known solubility of krypton and xenon noble gases in the 100 °F (37.8 °C) pool water, which would reduce their release into the containment building. Based on the 225,000-ft3 (6,371.3 m3) volume of containment building air, and the previously listed Curie quantities of these gases released into the reactor pool, the maximum noble gas concentrations in the containment structure at the end of five (5) minutes would be as follows: Example calculation of 85Kr average concentration in containment air during the first minute*:

    =    85 Kr activity x EXP[(-0.693 x 0.5 min) I 85Kr T 112] I (225,000 ft 3 x 28,3 I 7 ml/ft3)
    =    (4.35 x 10+04 µCi x EXP[(-0.693 x 0.5 min) I (10.76 yr x 365.25 day/yr x 1440 min/day)] I 6.371 x 10+09 ml) 6.83 x I o-06 µCi/ml Likewise, the average noble gas concentration of the five (5) I-minute interval concentrations is the average concentration of the five (5) minute interval.

31 of54

Average Noble Gas Concentrations in the Containment Building Air during the 5-Minute Period 85

               .Kr - 6.83 x 10-06 µCi/ml                               133 Xe - 5.68 x 10-03 µCi/ml 85 m.Kr - 1.80 X 10-03 µCi/ml                                135 Xe - 1.34 x 1o-03 µCi/ml 87
               .Kr - 2.28 x 1o-03 µCi/ml                             135 mxe - 4.78 x 10-04 µCi/ml 88
               .Kr - 3.77 x 10-03 µCi/ml                               137 Xe - 2.17 x 1o-05 µCi/ml 89
               .Kr - 4.71x10-06 µCi/ml                                 138 Xe - 1.52 x 1o-03 µCi/ml 90
               .Kr - 1.3 5 x 10-20 µCi/ml                              139 Xe - 2.11 x 10-17 µCi/ml The objective of this calculation is to present a worst-case dose assessment for an individual who remains in the containment building for five (5) minutes following the FHA. Therefore, as noted previously, the radioactivity in the evaporated pool water is assumed to be instantaneously and uniformly distributed into the containment building once released into the air.

Based on the source term data provided, it is possible to determine the radiation dose to the thyroid [Committed Dose Equivalent (CDE)] from radioiodine and the dose to the whole body [Committed Effective Dose Equivalent (CEDE)] resulting from submersion in the airborne noble gases and radioiodine inside the containment building. Because the airborne radioiodine source is composed of five (5) different iodine isotopes, it will be necessary to determine the dose contribution from each individual isotope and then sum the results. Dose Multiplication Factors were established using the Derived Air Concentrations (DACs) for the "listed" isotopes in Appendix *B of 10 CFR 20 and calculated values for the four (4) "unlisted" submersion isotopes (.Kr-89, Kr- 90, Xe-137 and Xe-139). The submersion DAC values that were calculated were done in accordance with the data and methodology as supplied in Federal Guidance Report (FGR) No.12 (Attachment 4). 131 Example calculation of thyroid dose due to 1: The DAC can also be defined as 50,000 mrem [thyroid target organ limit-(CDE)]/2,000 hrs, or 25 mrem/DAC-hr. Additionally, five (5) minutes of one (1) DAC-hr is 8.33 x 10-02 DAC-hr. 131 1 concentration in containment 1.62 x 10-06 µCi/ml 131 1 DAC (10 CFR 20) 2.00 x 1o-08 µCi/ml 131 Dose Multiplication Factor ( 1 concentration) I (131 1 DAC) (1.62 x 10-06 µCi/ml)/ (2.00 x 10-08 µCi/ml) 81.0 131 Therefore, a 5-minute thyroid exposure from 1 is: Dose Multiplication Factor x DAC Dose Rate x 5 minutes 81.0 x (25 mrem/DAC-hr) x (8.33 x 10-02 DAC-hr) 169 mrem 32of54

Note: Same calculation is used for the other radioiodines listed below. Derived Air Concentration Values and 5-Minute Exposures -Radioiodine Radionuclide Derived Air Concentration 5-Minute Exposure 1311 2.00 x 1o-08 µCi/ml 1.69 x 10+02 mrem 1321 3 .00 x 1o-06 µCi/ml 1.55 x lO+oo mrem 1331 1. 00 x 10*07 µCi/ml 8.26 x 1o+oi mrem 05 1341 2.00 x 10* µCi/ml 4.04 x 10*01 mrem 1351 7.00 x 10*07 µCi/ml 1.05 x 1o+oi mrem Total = 264.00 mrem Doses from the kryptons and xenons present in the containment building are assessed in much the same manner as the radioiodines, and the dose contribution from each individual radionuclide must be calculated and then added together to arrive at the final noble gas dose. Because the dose from the noble gases is only an external dose due to submersion, and because the DACs for these radionuclides are based on this type of exposure, the individual noble gas doses for five (5) minutes in containment were based on their average concentration in the containment air and the corresponding DAC. Example calculation of whole body dose (CEDE) due to 85.Kr: The DAC can also be defined as 5,000 mrem/2,000 hrs, or 2.5 mrem/DAC-hr. Additionally, five (5) minutes of one (1) DAC-hr is 8.33 x 10*02 DAC-hr. 85

          .Kr concentration in containment            6.83 x 1o-06 µCi/ml 85
          .Kr DAC (10 CPR 20)                         1.00 x 10*04 µCi/ml 85                     85 Dose Multiplication Factor                     ( .Kr concentration) I ( .Kr DAC)

(6.83 x 10*06 µCi/ml) I (1.00 x 10*04 µCi/ml)

                                                 =    0.0683 Therefore, a 5-minute whole body exposure from 85 Kr is:

Dose Multiplication Factor x DAC Dose Rate x 5 min 0.0683 x (2.5 mrem/DAC-hr) x (8.33 x 10*02 DAC-hr) 1.42 x 10*02 mrem Note: Same calculation is used for the other noble gases listed below. 33of54

Derived Air Concentration Values and 5-Minute Exposures - Noble Gases Radionuclide Derived Air Concentration 5-Minute Exposure 85Kr 1.00 x 1o-04 µCi/ml 1.42 x 10-02 mrem 85mKr 2.00 x 10-05 µCi/ml 1.88 x 10+01 mrem 87Kr 5.00 x 10-06 µCi/ml 9 .49 x 10+01 mrem 88Kr 2.00 x 1o-06 µCi/ml 3 .92 x 10+02 mrem 89Kr 1.90 x 1o-06 µCi/ml 5.17x 10-01 mrem 90Kr 2.80 x 10-06 µCi/ml 1.00 x 10- 14 mrem 133 Xe 1.00 x 10*04 µCi/ml 1.18 x 10+01 mrem 1.00 x 10-05 µCi/ml 135 Xe 2. 79 X 1O+Ol mrem I35mxe 9.00 x 10-06 µCi/ml 1.11 x 10+01 mrem 2.00 x 1o-05 µCi/ml 2.26 x 10-01 mrem 137 Xe 4.00 x 1o-06 µCi/ml 7.90 x 10+01 mrem 138 Xe 139Xe 3.70 x 10-06 µCi/ml 1.19 x 10- 12 mrem Total = 636.49 mrem To finalize the occupational dose in terms of Total Effective Dose Equivalent (TEDE) for a 5-minute exposure in the containment building after a FHA, the doses from the radioiodines and noble gases must be added together, and result in the following values: 5-Minute Dose from Radioiodines and Noble Gases in the Containment Building Total Iodine - Committed Dose Equivalent (CDE) 264.00 mrem Iodine- Committed Effective Dose Equivalent (CEDE) (CEDE= CDE x 0.03) 7.92 mrem Noble Gas - Committed Effective Dose Equivalent (CEDE) 636.49 mrem Total Dose -Total Effective Dose Equivalent (TEDE) 644.41 mrem By comparison of the maximum TEDE and CDE for those occupationally-exposed during a FHA to applicable dose limits in 10 CFR 20, the final values are shown to be well within the published regulatory limits and, in fact, lower than 15% of any occupational limit. Radiation shine through the containment structure* was also evaluated when considering accident conditions and dose consequences to the public and MURR staff. Calculation of exposure rate from a FHA was performed using the computer program MicroShield 8.02 with a Rectangular Volume - External Dose Point geometry for the representation of the containment structure (Attachment 9). MicroShield 8.02 is a product of Grove Software and is a comprehensive photon/gamma ray shielding and dose assessment program that is widely used by industry for designing radiation shields. The exposure rate values provided below represents the radiation fields at 1 foot (30.5 cm) from a 12-inch (30.5 cm) thick ordinary concrete containment wall and at the Emergency Planning Zone (EPZ) boundary of 150 meters (492.1 ft). The airborne concentration source terms used to develop the exposure rate values are identical to those used for determining the dose to a worker within 34 of54

containment from noble gases. For radioiodine, the total iodine activity from the FHA was used for the dose calculations, not the amount that evaporated in five (5) minutes. The source term also assumes a homogenous mixture of nuclides within the containment free volume. Radiation Shine through the Containment Building Exposure Rate at 1-Foot from Containment Building Wall: 59.77 mrem/hr Exposure Rate at Emergency Planning Zone Boundary (150 meters): 0.41 mrem/hr As noted earlier in this analysis, the containment building ventilation system will shut down and the building itself will be isolated from the surrounding areas. The FHA will not cause an increase in pressure inside the reactor containment structure; therefore, any air leakage from the building will occur as a result of .normal changes in atmospheric pressure and pressure equilibrium between the inside of the containment structure and the outside atmosphere. It is highly probable that there will be no pressure differential between the inside of the containment building and the outside atmosphere, and consequently there will be no air leakage from the building and no radiation dose to members of the public in the unrestricted area. However, to develop what would clearly be a worst-case scenario, this analysis assumes that a barometric pressure drop has occurred in conjunction with the FHA. An extreme assumption would be a pressure change on the order of 0. 7 inches of Hg (25 .4 mm of Hg at 22 °C) from an initial atmospheric pressure of 15.0333 psia. This would then create a pressure differential of about 1/3 psig (2.28 kPa above atmosphere) higher on the inside of the isolated containment building than on the inside of the adjacent laboratory building, which surrounds most of the containment structure. With an initial internal pressure in the containment building of 15.0333 psia, it would contain 230,102 standard cubic feet (set) of air. The conservative assumption is made that the containment building will leak at a rate slightly greater than the Technical Specification (TS) leakage rate limit. The TS leakage rate limit shall not exceed either 16.3 ft3/min (STP) with an overpressure of one pound per square inch gauge or 10% of the contained volume over a 24-hour period from an initial overpressure of two pounds per square inch gauge. Additionally, the minimum TS free volume of the containment building is 225,000-ft3 at standard pressure and temperature. The following equation represents the air leakage rate from the containment building in standard cubic feet per minute (scfm) as a function of containment pressure which at 1 psi over pressure would corresponds to 17.68 ft3/minute. This would correspond to a leakage rate 8.4% greater than the TS limit of 16.3 ft 3/minute at 1.0 psig. LR 17.68x(CP-14.7) 112 ; where: LR leakage rate from containment (scfm); and CP containment pressure (psia). 35of54

Using this equation for the assumed initial overpressure condition of 0.333 psig (2.28 kPa above atmosphere), it would take approximately 16.5 hours for the leakage rate to decrease to zero from an initial leakage rate of approximately 10.25 scfm, which would occur at the start of the event. The average leakage rate over the 16.5-hour period would be approximately 5.15 scfm. This conservatively over calculates the actual amount of activity that would leak out of the containment building and potentially expose someone in this assumed accident. Several factors exist that will mitigate the radiological impact of any air leakage from the containment building following a FHA. First of all, most leakage pathways from containment discharge into the reactor laboratory building, which surrounds the containment structure. Since the laboratory building ventilation system continues to operate during a FHA, leakage air captured by the ventilation exhaust system is mixed with other building air, and then discharged from the facility through the exhaust stack at a rate of approximately 30,500 scfm. Mixing of containment air leakage with the laboratory building ventilation flow, followed by discharge out the exhaust stack and subsequent atmospheric dispersion, results in extremely low radionuclide concentrations and very small radiation doses in the unrestricted area. A tabulation of these concentrations and doses is given below. A second factor which helps to reduce the potential radiation dose in the unrestricted area relates to the behavior of radio iodine, which has been studied extensively in the containment mockup facility at Oak Ridge National Laboratory (ORNL). From these experiments, it was shown that up to 75% of the iodine released will be deposited in the containment vessel. For the purposes of this analysis, MURR used the more conservative NRC-accepted value of 50% reduction of radioiodines from plate-out and deposition (Regulatory Guide 1.3, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors"). Thus, if due to this 50% iodine deposition in the containment building, each cubic foot of air released from the containment structure has a radioiodine concentration that is 50% of each cubic foot within the containment building air, then the radioiodine concentrations leaking from the containment structure into the laboratory building, in microcuries per standard cubic foot (scf), will be: 131 Example calculation of average 1 concentration in containment during the first hour: 131 ( 1 concentration in pool x 20 gal) x EXP[(-0.693 x 0.5 hr) I 131 1 T 112] I (scf in containment building) (1.03 x 10+03 µCi/gal x 20 gal) x EXP[(-0.693 x 0.5 hr) I (8.02 days x 24 hr/day)] I (229,801 scf) (1.03 x 10+03 µCi/gal x 20 gal) x 0.9982 I (229,801 scf) 8.98 x 10*02 µCi/scf Note: Same calculation is used for the other radioiodines listed below. Radioiodine Concentrations in Containment Air during the First Hour 131 1 - 8.98 x 10*02 µCi/scf 133 1 - 2.17 x 10*01 µCi/scf 135 1 - 1.86 x 10*01 µCi/scf 132 134 1 - 1.08 x 10*01 µCi/scf 1 - 1.51 x 10*01 µCi/scf 36of54

Example calculation of the 85Kr concentration released into containment: 85 Kr activity x EXP[(-0.693 x 0.5 hr) I (10.76 yr x 365.25 days/yr x 24 hr/day)] I 229,801 ft 3 4.35 x 10+04 µCi x 0.999996 I 229,801 ft3 1.89 x 10-01 µCi/scf Note: Same calculation is used for the other noble gases listed below. Average Noble Gas Concentrations in the Containment Air during the First Hour 85 Kr - 1.89 x 10-01 µCi/scf 133 Xe - 1.57 x 10+02 µCi/scf 135 ssmKr - 4.66 X 1O+OI µCi/scf Xe - 3.59 x 10+01 µCi/scf 87 135 Kr - 4.92 X 1O+OI µCi/scf mxe - 3 .80 x 1o+oo µCi/scf 88 01 137 Kr - 9.34 x 10+ µCi/scf Xe - 3.97 x 10*03 µCi/scf 89 Kr - 2.94 x 1o-04 µCi/scf 138 Xe - 1.09 x 1O+OI µCi/scf 90 32 139 Kr - 4.36 x 10* µCi/scf Xe - 7.29 x 10*26 µCi/scf The average containment building leakage rate was calculated for each of the first five (5) hours and for the following three (3) 4-hour intervals: Hours: 0- 1 1-2 2-3 3-4 4-5 5-9 9 - 13 13 - 16.5 scf/hr: 595.6 558.7 521.8 485.0 448.1 355.8 207.9 67.9 The average concentration of the radioactive iodine's and noble gases in the facility's exhaust stack was calculated based on the average isotope concentration in containment during each of the time periods and the average leakage rate for the time interval. The iodine activities are reduced to 50% due to the previous stated 50% deposition in the containment structure in their leakage path. This leakage of activity out of containment is drawn into the facility exhaust ventilation system. The facility exhaust ventilation system has a flow rate of 30,500 scfm and results in the following average exhaust activities: Example calculation of the average concentration of 1311 released through the exhaust stack during the first hour: 131 ( 1 activity concentration in containment µCi/scf x scf leakage rate x 0.5) I (30,500 scfm x 60 min/hr x 1 hr x 28,317 ml/ft3) (8.98 x 10*02 µCi/scfx 595.6 scfx 0.5) I 5.18 x 10+10 ml 5.16 x 10* 10 µCi/ml Note: Same calculation is used for the other radioiodines listed below. 37of54

Average Radio iodine Concentrations in Air Leaking from Containment and Exiting the Exhaust Stack during the First Hour 131 1 - 5.16 X 10-lO µCi/ml 133 1 - 1.25 x 1o-09 µCi/ml 135 1 - 1.07 x 10-09 µCi/ml 132 134 1 - 6.20 x 10- 10 µCi/ml 1 - 8.69 x 10-10 µCi/ml Example calculation of the average concentration of 85Kr released through the exhaust stack during the first hour: (8 5Kr activity concentration in containment µCi/scfx scfleakage rate) I (30,500 scfm x 60 min/hr x 1 hr x 28,317 ml/ft3) (1.89 x 10-01 µCi/scf x 595.6 scf/hr) I (30,500 scfm x 60 min/hr x 1 hr x 28,317 ml/ft3 ) 2.18 x 1o-09 µCi/ml Note: Same calculation is used for the other noble gases listed below. Average Noble Gas Concentrations in Air Leaking from Containment and Exiting the Exhaust Stack during the First Hour 85 Kr - 2.18 x 1o-09 µCi/ml 87 Kr - 5.65 x 10-07 µCi/ml 89 Kr - 3.38 x 10-12 µCi/ml 85 mKr - 5.3 5 x 10-07 µCi/ml 88 Kr - 1.07 x 10-06 µCi/ml 90 Kr - 5.01x10-4° µCi/ml 133 Xe - 1.80 x 1o- µCi/ml 06 135 mxe - 4.37 x 1o-os µCi/ml 138 Xe - 1.25 x 1o-07 µCi/ml . 137 135 Xe - 4.12 x 10-01 µCi/ml Xe - 4.56 X 10-ll µCi/ml 139 Xe - 8.38 x 10-34 µCi/ml Assuming, as stated earlier, that (1) the average leakage rate from the containment building is 5.2 scfm, (2) the leak continues for about 16.5 hours in order to equalize the containment building pressure with atmospheric pressure, (3) the flow rate through the facility's ventilation exhaust stack is 30,500 scfm, (4) the reduction in concentration from the point of discharge at the exhaust stack to the point of maximum concentration in the unrestricted area is a factor of 292, and (5) there is no decay of any radioiodines or noble gases following release from the exhaust stack, then the following concentrations of radioiodines and noble gases with their corresponding radiation doses will occur in the unrestricted area. The values listed are for the point of maximum concentration in the unrestricted area assuming uniform, semi-spherical cloud geometry for noble gas submersion and further assuming that the most conservative (worst-case) meteorological conditions exist for the entire 16.5-hour period of containment leakage following a FHA. Radiation doses are calculated for the entire 16.5-hour period. Dose values for the unrestricted area were obtained using the same methodology that was used to determine doses inside the containment building, and it was assumed that an individual was present at the point of maximum concentration for the full 16.5 hours that the containment building was leaking. A worst-case scenario effluent dilution factor of 292 using the Pasquill-Guifford Model for atmospheric dilution is used in this analysis. It is assumed that all offsite (public) dose occurs under these atmospheric conditions at the site of interest, i.e. 760 meters north of MURR. In our case, at 760 meters, it occurs only during Stability Class 'F' conditions, which normally only occurs 11.4% of the time when the wind blows from the south. Thus this calculation is conservative. 38of54

10 CFR 20 Appendix B Effluent Concentration Limits are used for the "listed" isotopes. Effluent Concentration Limits were calculated for each of the four (4) "unlisted" noble gases (Kr-89, Kr-90, Xe-137 and Xe-139) using the data and methodology contained in FGR No. 12 for submersion isotopes. The DAC value was first calculated and then a factor of 219 was applied using 10 CFR 20, Appendix B, as a reference point for DAC values from submersion isotopes. Exposure at 1 DAC equates to 5000 mrem per year whereas at the Effluent Concentration Limit it is 50 mrem per year. Thus, there is a factor of 100 times lower allowable dose for the Effluent Concentration Limit as compared to the DAC. Exposure at the Effluent Concentration Limit assumes you are in that effluent concentration for 8760 hours per year. Therefore, the time assumed to be exposed to the Effluent Concentration Limit is a factor of 4.38 times longer than the 2000 hours per year that defines a DAC. No credit is taken for transit time from the exhaust stack to the receptor point nor is credit taken for decay inside the containment building until release. In the case ofKr-89 and Xe-137, the transit time alone would be approximately one (1) half-life while the transit time for.Kr-90 and Xe-139 would be at least four (4) half-lives. 131 Example calculation of whole body dose in the unrestricted area due to 1: Conversion Factor: (Public dose limit of 50 mrem/yr) x (1 yr/8760 hrs)= 5.71x10-03 mrem/hr 131 I concentration 2.63 x 10- 10 µCi/ml 131 1 effluent concentration limit 2.00 x 10- 10 µCi/ml 131 1 Conversion Factor 1.325 mrem/hr 131 Therefore, a 16.5-hour whole body exposure from 1 is: 131 [( 1 concentration I 131 1 Effluent Concentration Limit) x Conversion Factor x 16.5 hrs] I Dilution Factor [[(2.63 x 10-10 µCi/ml) I (2.00 x 10-10 µCi/ml)] x 1.325 mrem/hr x 16.5 hrs] I 292

    =   4.24 x 10-04 mrem Note: Same calculation is used for the other isotopes (radioiodines and noble gases) listed below.

39 of 54

Effluent Concentration Limits, Concentrations at Point of Maximum Concentration and Radiation Doses in the Unrestricted Area - Radioiodine Radionuclide Effluent Limit Maximum Concentration1 Radiation Dose 1311 2.00 x 10* 10 µCi/ml 9.01x10* 13 µCi/ml 4.24 x 10*04 mrem 1321 2.00 x 1o-08 µCi/ml 4.01x10* 13 µCi/ml 1.89 x 10-06 mrem 1331 1.00 x 10*09 µCi/ml 1.88 x 10*12 µCi/ml 1.77 x 10*04 mrem 1341 6.00 x 10*08 µCi/ml 3.12 x 10*13 µCi/ml 4.89 x 10*07 mrem 1351 6.00 x 10*09 µCi/ml 1.20 x 10*12 µCi/ml 1.88 x 10*05 mrem Total = 6.23E-04 mrem Note 1: Maximum Concentrations are radioiodine and noble gas concentrations leaking from the containment building and exiting the exhaust stack reduced by a dilution factor of292. Effluent Concentration Limits, Concentrations at Point of Maximum Concentration and Radiation Doses in the Unrestricted Area - Noble Gases Radionuclide Effluent Limit Maximum Concentration1 Radiation Dose 85.Kr 7.00 x I 0*01 µCi/ml 3.87 x 10*12 µCi/ml 5.21x10-01 mrem 85m.Kr 1.00 x 10*01 µCi/ml 5.06 x 10- 10 µCi/ml 4.77 x 10*04 mrem 81Kr 2.00 x 10*08 µCi/ml 2.53 x 10*10 µCi/ml 1.19 x 10*03 mrem 88.Kr 9.00 x 10*09 µCi/ml 7.93 x 10*10 µCi/ml 8.30 x 10-03 mrem 89.Kr 8.60 x 10-09 µCi/ml 7.01x10*16 µCi/ml 7.68 x 10*09 mrem 90.Kr 1.20 x 1o-08 µCi/ml 1.04 x 10*43 µCi/ml 8.17x 10-37 mrem 133Xe 5.00 x 10*07 µCi/ml 3.12 x 10*09 µCi/ml 5.88 x 10-04 mrem 1o-08 µCi/ml 5.17x 10-10 µCi/ml 6.96 x 10*04 mrem 135 Xe 7.00 x 135mxe 4.00 x 10*08 µCi/ml 9.68 x 10* 12 µCi/ml 2.28 x 10*05 mrem 9.10 x 10-08 µCi/ml 9.46 x 10*15 µCi/ml 9.80 x 10*09 mrem 137 Xe 138Xe 2.00 x I o-08 µCi/ml 2.72 x 10*11 µCi/ml 1.28 x I 0*04 mrem 139Xe 1.60 x 1o-08 µCi/ml 1.74 x 10*37 µCi/ml 1.02 x I 0*30 mrem Total = 1.14E-02 mrem Note 1: Maximum Concentrations are radio iodine and noble gas concentrations leaking from the containment building and exiting the exhaust stack reduced by a dilution factor of292. To finalize the unrestricted dose in terms of Total Effective Dose Equivalent (TEDE), the doses from the radioiodines and noble gases must be added together, and result in the following values: Dose from Radioiodines and Noble Gases in the Unrestricted Area Total Iodine - Committed Effective Dose Equivalent (CEDE) 6.23E-04 mrem Noble Gas - Committed Effective Dose Equivalent (CEDE) 1.14E-02 mrem Total Dose -Total Effective Dose Equivalent (TEDE) 1.20E-02 mrem 40of54

Summing the doses from the noble gases and the radioiodines simply substantiates earlier statements regarding the very low levels in the unrestricted area should a FHA occur, and should the containment building leak following such an event. Because the dose values are so low, a value far below the applicable 10 CFR 20 regulatory limit for the unrestricted area, MURR does not anticipate any issues regarding offsite dose during a FHA scenario. 41of54

Fueled Experiment Failure [page 56) -Information/Clarification Needed

Several equations do not appear to have been bracketed correctly. Explain.

The "Fueled Experiment Failure" accident analysis has been revised to correct for any bracketing inaccuracies in the equations.

Units used are non-standard (micro-Curies per cubic foot versus micro-Curies per cubic milliliter, similar comment on pages 47 and 48). Verify.

The "Fueled Experiment Failure" accident analysis has been revised to correct for any non-standard units. 42of54

Revised "Fueled Experiment Failure" (MURR's new Maximum Hypothetical Accident) The release of the radioisotopes of krypton, xenon and radioiodine from a 5-gram, low-enriched uranium (LEU) target is the major source of radiation exposure to an individual and will, therefore, serve as the basis for the source term for the dose calculations of a Fueled Experiment Failure. A 5-gram LEU target irradiated for 150 hours (normal weekly operating cycle) at a thermal neutron flux of 1.5 x 10+ 13 n/cm2 -sec will produce the following radioiodine, krypton and xenon activities (additionally, approximately 1.40 x 10+04 µCi of 90 Sr will be produced): Radioiodine and Noble Gas Activities in a 5-Gram LEU Target 1311 - 6.755 Ci 85Kr - 0.0020 Ci 133 Xe - 18.925 Ci 1321 - 18.635 Ci ssmKr _ 7.580 Ci 135 Xe - 13.630 Ci 133 1 - 39.875 Ci 87Kr - 15.405 Ci 13smxe _ 6.760 Ci 1341 - 45.405 Ci 88Kr - 21.660 Ci 137 Xe - 35.800 Ci 135 1 - 37.695 Ci 89Kr - 27.740 Ci 138 Xe - 37.380 Ci 90Kr - 27.410 Ci 139 Xe - 30.675 Ci Total Iodine - 148.365 Ci Total Krypton - 99.797 Ci Total Xenon - 143.170 Ci The source term for the Fueled Experiment Failure accident analysis [the new Maximum Hypothetical Accident (MHA)] was determined using the ORIGEN program. The fueled experiment target was irradiated for 150 hours at the peak flux level it is exposed to during full power reactor operation and the resulting activities were calculated using ORIGEN 2.2. A complete failure of the fueled target is unrealistic for many reasons. The worst that can be expected is partial melting; however, in order to present a worst-case dose assessment for an individual that remains in the containment building following target failure, 100% of the total activity of the target is assumed to be released into the reactor pool. Fission products released into the reactor pool will be detected by the pool surface and ventilation system exhaust plenum radiation monitors. However, for the purposes of this analysis, it is assumed that a reactor scram and actuation of the containment building isolation system occurs by action of the pool surface radiation monitor. Actuation of the isolation system will prompt Operations personnel to ensure that a total evacuation of the containment building is accomplished promptly, usually within two (2) to two and a half (2.5) minutes. A conservative 5-minute evacuation time is used as the basis for the stay time in the dose calculations for personnel that are in containment during target failure. The radioiodine released into the reactor pool over a 5-minute interval is conservatively assumed to be instantly and uniformly mixed into the 20,000 gallons (75,708 1) of bulk pool water, which then results in the following pool water concentrations for the radioiodine isotopes. The water solubility of the krypton and xenon noble gases released into the pool over this same time period is conservatively ignored. The gas bubble rise time in the reactor pool from where the target is situated in the graphite reflector region to the pool surface has been measured at 17 seconds, so it is assumed 43 of 54

the earliest any of the radioisotopes from the fueled experiment enters into the containment building air volume is after 17 seconds. It is also assumed when the radioactivity enters the containment air volume it instantaneously forms a uniform concentration in the isolated containment structure. Radioiodine Concentrations in the Pool Water 131 1 - 3.38 x 10+02 µCi/gal 133 1 - 1.99 x 10+03 µCi/gal 135 1 - 1.88 x 10+03 µCi/gal 132 1 - 9.32 x 10+02 µCi/gal 134 1 - 2.27 x 10+03 µCi/gal When the reactor is at 10 MW and the containment building ventilation system is in operation, the evaporation rate from the reactor pool surface is approximately 80 gallons (302.8 L) of water per day. For the purposes of this calculation, it is assumed that a total of 20 gallons (75.7 L) of pool water containing the previously listed radioiodine concentrations evaporates into the containment building over the 5-minute period. Containment air with a temperature of 75 °F (23.9 °C) and 100% relative humidity contains H20 vapor equal to 40 gallons (151.4 L) of water. Since the air in containment is normally at about 50% relative humidity, thus containing 20 gallons (75.7 L) of water vapor, the assumed addition of 20 gallons (75.7 L) of water vapor over five (5) minutes will not cause the containment air to be supersaturated. It is also conservatively assumed that all of the radioiodine activity in the 20 gallons (75.7 L) of evaporated pool water instantaneously forms a uniform concentration in the containment building air. When distributed into the containment building, this would result in the following radioiodine concentrations in the 225,000 ft3 (6,371.3 m3) air volume: Example calculation of the averageu 1l concentration released into containment air during the first minute: 131 1 concentration in pool water/gal x 20 gal/5 minx 0.5 minx EXP[(-0.693 x (17 + 30 sec)) I (8.02 day x 8.64 x 10+04 sec/day)] /(225,000 ft3 x 28317 ml/ft3)

    =     (3.38x10+02 µCi/gal) x 2 gal x 0.99995I(6.317x10+09 ml)
    =     1.06 x 10-07 µCi/ml Same calculation is used for the other isotopes listed below and was performed for each one minute interval. The average radioiodine concentration over the 5-minute interval is the average of the five (5) I-minute intervals.

Average Radio iodine Concentrations in the Containment Building Air during the 5 Minutes 131 1 - 5.30 x 10-07 µCi/ml 133 1 - 3 .12 x 10-06 µCi/ml 135 1 - 2.96 x 10-06 µCi/ml 132 1 - 1.44 x 1o-06 µCi/ml 134 1 - 3 .40 x 10'06 µCi/ml As noted previously, the krypton and xenon noble gases released into the reactor pool from the 5-gram LEU target during the 5-minute interval following failure are assumed to have no absorption in the pool water, rise through the pool in 17 seconds (thus slightly decaying) and enter the containment building air volume where they are assumed to instantaneously form a uniform concentration in the isolated structure. Based on the 225,000-ft3 volume of containment building air, and the previously 44 of54

listed Curie quantities of these gases released into the reactor pool, the maximum noble gas concentrations in the containment structure at the end of five (5) minutes would be as follows: Example calculation of the average 85Kr released into containment air during the first minute after the gas enters the containment air: 85 Kr activity x EXP[(-0.693 x (17 + 30 sec)) I (10.76 yrs x 3.156 x 10+07 sec/yr)] I (225,000 ft3 x 28317 ml/ft3) 2.00 x 10+03 µCi x 1.000 I (6.371 x 10+09 ml)

    =  3 .14 x 10*07 µCi/ml Note: Same calculation is used for the other isotopes listed below.

The average noble gas concentrations are the average of the five (5) 1-minute decay corrected concentrations. Average Noble Gas Concentrations in the Containment Building Air during the 5 Minutes 85 Kr - 3 .14 x 10*07 µCi/ml 133 Xe - 2.97 x 10*03 µCi/ml ssmKr - 1.18 x 10*03 µCi/ml 135 Xe - 2.13 x 10*03 µCi/ml 87 Kr - 2.36 x 10*03 µCi/ml 135 mxe - 9.37 x 10*04 µCi/ml 88 Kr - 3.36 x 10*03 µCi/ml 137 Xe - 3.50 x 10*03 µCi/ml 89 Kr - 2.48 x 10*03 µCi/ml 138 Xe - 5.13 x 10*03 µCi/ml 90 Kr - 4.33 x 10*04 µCi/ml 139 Xe - 6.50 x 10-04 µCi/ml The objective of this calculation is to present a worst-case dose assessment for an individual who remains in the containment building for five (5) minutes following target failure. Therefore, as noted previously, the radioactivity in the evaporated pool water is assumed to be instantaneously and uniformly distributed into the building once released into the air. Based on the source term data provided, it is possible to determine the radiation dose to the thyroid from radioiodine and the dose to the whole body resulting from submersion in the airborne noble gases and radioiodine inside the containment building. Because the airborne radioiodine source is coinposed of five (5) different iodine isotopes, it will be necessary to determine the dose contribution from each individual isotope and to then sum the results before calculating an effective dose equivalent for the radioiodines. These results were then added to the submersion doses calculated for the noble gases. Dose multiplication factors were established using the Derived Air Concentrations (DACs) for the "listed" isotopes in Appendix B of 10 CFR 20 and were calculated using the methodology presented in Federal Guidance Report (FGR) No. 12 for those isotopes that are "unlisted" submersion isotopes, and the radionuclide concentrations in the containment building (Attachment 4). 45of54

Example calculation of thyroid dose due to 131 l: The DAC can also be defined as 50,000 mrem (thyroid target organ limit)/2,000 hrs, or 25 mrem/DAC-hr. Additionally, five (5) minutes of one DAC-hr is 8.33 x 10-02 DAC-hr. 131 1 concentration in containment 5.30 x 10-07 µCi/ml 131 lDAC (10 CFR20) 2.00 x 1o-08 µCi/ml Dose Multiplication Factor ( 131 1 concentration) I ( 131 1 DAC) (5.30 x 10-07 µCi/ml) I (2.00 x 10-08 µCi/ml) 26.5 131 Therefore, a 5-minute thyroid exposure from 1 is: Dose Multiplication Factor x DAC Dose Rate x 5 minutes of a DAC-hr 26.5 x (25 mrem/DAC-hr) x (8.33 x 10-02 DAC-hr)

   =   55.2mrem Note: Same calculation is used for the other radioiodines listed below.

Derived Air Concentration Values and 5-Minute Exposures -Radioiodine Radionuclide Derived Air Concentration 5-Minute Exposure 1311 2.00 x 1o-os µCi/ml 55.2 mrem 1321 06 3.00 x 10- µCi/ml 0.997 mrem 1331 07 1.00 x 10- µCi/ml 65.0 mrem 1341 2.00 x 1o-os µCi/ml 0.354 mrem 1351 07 7.00 x 10- µCi/ml 8.80 mrem Total = 130.4 mrem Doses from the kryptons and xenons present in the containment building are assessed in much the same manner as the radioiodines, and the dose contribution from each individual radionuclide must be calculated and then added together to arrive at the final noble gas dose. Because the dose from the noble gases is only an external dose due to submersion, and because the DACs for these radionuclides are based on this type of exposure, the individual noble gas doses for five (5) minutes in containment were based on their average concentration in the containment air and the corresponding DAC. Example calculation of whole body dose (CEDE) due to 85Kr: The DAC can also be defined as 5,000 mrem/2,000 hrs, or 2.5 mrem/DAC-hr. Additionally, five (5) minutes of one DAC-hr is 8.33 x 10-02 DAC-hr. 46of54

85Kr concentration in containment 3.14 x 10*07 µCi/ml 85Kr DAC (10 CFR 20) 1.00 x 10*04 µCi/ml Dose Multiplication Factor (85Kr concentration) I (85Kr DAC) (3.14 x 10*07 µCi/ml) I (l.00 x 10*04 µCi/ml) 0.00314 Therefore, a 5-minute whole body exposure from 85Kr is: Dose Multiplication Factor x DAC Dose Rate x 5 minutes of a DAC-hr 0.00314 x (2.5 mrem/DAC-hr) x (8.33x10*02 DAC-hr) 6.54 x 10*04 mrem Note: Same calculation is used for the other noble gases listed below. Derived Air Concentration Values and 5-Minute Exposures - Noble Gases Radionuclide Derived Air Concentration 5-Minute Exposure 85Kr 1.00 x 10*04 µCi/ml 6.54 x 10*04 mrem 85mKr 2.00 x 10*05 µCi/ml 1.23 X 1O+Ol mrem 87Kr 5.00 x 10*06 µCi/ml 9.82 x 10+01 mrem 88Kr 2.00 x 10*06 µCi/ml 3.50 x 10+02 mrem 89Kr 1.90 x 1o-06 µCi/ml 2.71 x 10+02 mrem 90Kr 2.80 x 10*06 µCi/ml 3.22 X 1O+Ol mrem 133 Xe 1.00 x 10*04 µCi/ml 6.18 x lO+oo mrem 1.00 x 10*05 µCi/ml 135 Xe 4.44 X 1O+Ol mreni mmxe 9.00 x 10*06 µCi/ml 2.17 x 10+01 mrem 2.00 x 10*05 µCi/ml 137 Xe 3 .65 X 1O+Ol mrem 4.00 x 10"06 µCi/ml 2.67 x 10+02 mrem 138 Xe 139Xe 3.70 x 10*06 µCi/ml 3 .66 x 10+01 mrem Total = 1176.42 mrem To finalize the occupational dose in terms of Total Effective Dose Equivalent (TEDE) for* a 5-minute exposure in the containment building after target failure, the doses from the radioiodines and noble gases must be added together, and result in the following values: 5-Minute Dose from Radioiodines and Noble Gases in the Containment Building Total Iodine - Committed Dose Equivalent (CDE) 130.36 mrem Iodine - Committed Effective Dose Equivalent (CEDE) (CEDE= CDE x 0.03) 3.91 mrem Noble Gas - Committed Effective Dose Equivalent (CEDE) 1176.42 mrem

Total Dose - Total Effective Dose Equivalent (TEDE) 1180.33 mrem 47of54

Note: The addition of 90 Sr will increase the above stated TEDE (whole body) by <l %. However, due to the chemical form of the strontium, which would not be prone to evaporation from the pool into the containment atmosphere, it would remain as a residue or concentrate in the unevaporated pool water and its contribution to dose would essentially be zero. By comparison of the maximum TEDE and CDE for those occupationally-exposed during target failure to applicable NRC dose limits in 10 CFR 20, the final values are shown to be well within the published regulatory limits and, in fact, lower than 25% of any occupational limit. Radiation shine through the containment structure was also evaluated when considering accident conditions and dose consequences to the public and MURR staff. Calculation of exposure rate from the target failure was performed using the computer program MicroShield 8.02 with a Rectangular Volume - External Dose Point geometry for the representation of the containment structure (Attachment 12). MicroShield 8.02 is a product of Grove Software and is a comprehensive photon/gamma ray shielding and dose assessment program that is widely used by industry for designing radiation shields. The exposure rate values provided below represents the radiation fields at 1 foot (30.5 cm) from a 12-inch (30.5 cm) thick ordinary concrete containment wall and at the Emergency Planning Zone (EPZ) boundary of 150 meters (492.1 ft). The airborne concentration source terms used to develop the exposure rate values are identical to those used for determining the dose to a worker within containment from noble gases. For radioiodine, the total iodine activity of the target was used for the dose calculations, not the amount that evaporated in five (5) minutes. The source term also assunies a homogenous mixture of nuclides within the containment free volume. Radiation Shine through the Containment Building Exposure Rate at 1-Foot from Containment Building Wall: 74.69 mrem/hr Exposure Rate at Emergency Planning Zone Boundary (150 meters): 0.514 mrem/hr As noted earlier in this analysis, the containment building ventilation system will shut down and the building itself will be isolated from the surrounding areas. Target failure will not cause an increase in pressure inside the reactor containment structure; therefore, any air leakage from the building will occur as a result of normal changes in atmospheric pressure and pressure equilibrium between the inside of the containment structure and the outside atmosphere. It is highly probable that there will be no pressure differential between the inside of the containment building and the outside atmosphere, and consequently there will be no air leakage from the building and no radiation dose to members of the public in the unrestricted area. However, to develop what would clearly be a worst-case scenario, this analysis assumes that a barometric pressure drop had occurred in conjunction with target failure. An extreme worst-case assumption would be a pressure change on the order of 0.7 inches of Hg (25.4 mm of Hg at 60 °C), which would then create a pressure differential of about 0.33 psig (2.28 kPa above atmosphere) between the inside of the isolated containment building and the inside of the adjacent laboratory building, which surrounds most of the containment building. 48of54

The conservative assumption is made that the containment building will leak at a rate slightly greater than the Technical Specification (TS) leakage rate limit. The TS leakage rate limit shall not exceed either 16.3 ft3/min (STP) with an overpressure of one pound per square inch gauge or 10% of the contained volume over a 24-hour period from an initial overpressure of two pounds per square inch gauge. Additionally, the minimum TS free volume of the containment building is 225,000-ft3 at standard pressure and temperature. The following equation represents the air leakage rate from the containment building in standard cubic feet per minute (scfm) as a function of containment pressure which at 1 psi over pressure would corresponds to 17.68 ft3/minute. This would correspond to a leakage rate 8.4% greater than the TS limit of 16.3 ft3/minute at 1.0 psig. LR 17 .68 x (CP-14. 7) 112 ; where: LR leakage rate from containment (scfm); and CP containment pressure (psia). Using this equation for the assumed initial overpressure condition of 0.33 psig (2.28 kPa above atmosphere), it would take approximately 16.5 hours for the leakage rate to decrease to zero from an initial leakage rate of approximately 10.25 scfm, which would occur at the start of the event. The average leakage rate over the 16.5-hour period would be approximately 5.15 scfm. This conservatively over calculates the actual amount of activity that would leak out of the containment building and potentially expose someone in this assumed accident. Several factors exist that will mitigate the radiological impact of any air leakage from the containment building following target failure. First of all, most leakage pathways from containment discharge into the laboratory building, which surrounds the containment structure. Since the laboratory building ventilation system continues to operate during target failure, leakage air captured by the ventilation exhaust system is mixed with other building air, and then discharged from the facility through the exhaust stack at a rate of approximately 30,500 scfm. Mixing of containment air leakage with the laboratory building ventilation flow, followed by discharge out the exhaust stack and subsequent atmospheric dispersion, results in extremely low radionuclide concentrations and very small radiation doses in the unrestricted area. A tabulation of these concentrations and doses is given below. A second factor which helps to reduce the potential radiation dose in the unrestricted area relates to the behavior of radioiodine, which has been studied extensively in the containment mockup facility at Oak Ridge National Laboratory (ORNL). From these experiments, it was shown that up to 75% of the iodine released will be deposited in the containment vessel. For the purposes of this analysis, MURR used the more conservative NRC-accepted value of 50% reduction of radio iodines from plate-out and deposition (Regulatory Guide 1.3, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors"). Thus, due to this 50% iodine deposition in the containment building, each cubic foot of air released from 49 of54

containment has a radioiodine concentration that is 50% of each cubic foot within containment building air, then the radioiodine concentrations leaking from the containment structure into the laboratory building, in microcuries per standard cubic foot (set), during the first hour will be (average 131 1 concentration in containment):

131 Example calculation of average 1 concentration in containment during the first hour: 131 131 ( 1 activity in pool water x 20 gal) x EXP[(-0.693 x 0.5 hr) I 1 T 112]) I (scf in containment) (3.38 x 10+02 µCi/gal x 20 gal) x 0.9982 I (229,800 set) 2.94 x 1o-02 µCi/scf Note: Same calculation is used for the other radioiodines listed below. Radioiodine Concentrations in Containment Air during the First Hour 131 1 - 2.94 x 10-02 µCi/scf 133 1 - I. 70 x 10-01 µCi/scf 135 1 - 1.55 x 10-01 µCi/scf 132 1 - 6.97 x 10-02 µCi/scf 134 1 - 1.33 x 10-01 µCi/scf Example calculation of the 85Kr concentration released into containment during the First Hour: 85 Kr activity x EXP[(-0.693 x 0.5 hr) I (10.76 yr x 365.25 days/yr x 24 hr/day)] I 229,801 ft3 . 03 3 2.00 x 10+ µCi x 0.999996 I 229,801 ft 8.70 x 10-03 µCi/scf Note: Same calculation is used for the other noble gases listed below. Average Noble Gas Concentrations in the Containment Air during the First Hour 85 Kr - 8.70 x 10-03 µCi/scf 133 Xe - 8.21 x 10+01 µCi/scf 85 135 mKr - 3.05 x lO+Ol µCi/scf Xe -,. 5.71x10+01 µCi/scf 87 135 Kr - 5 .10 x 1O+Ol µCi/scf mxe - 7.56 x lO+oo µCi/scf 88 137 Kr - 8.34 X 1O+Ol µCi/scf Xe - 6.74 x 10-01 µCi/scf 8 9:Kr - 1.64 x 10-01 µCi/scf 138 Xe - 3.72 x 10+01 µCi/scf 90 139 Kr - 2.02 x 10-15 µCi/scf Xe - 3.02 x 10-12 µCi/scf The average containment building leakage rate was calculated for each of the first five (5) hours and for the following three (3) 4-hour intervals: Hours: 0-1 1-2 2-3 3-4 4-5 5-9 9-12 12 - 16.5 scf/hr: 595.6 558.7 521.8 485.0 448.1 355.8 207.9 67.9 The average concentration of the radioactive iodine's and noble gasses in the facility's exhaust stack was calculated based on the average isotope concentration in containment during each of the time 50of54

periods and the average leak rate for the time interval. The iodine activities are reduced to 50% due to the previous stated 50% deposition in the containment structure in their leakage path. This leakage of activity out of containment is drawn into the facility exhaust ventilation system. The facility exhaust ventilation system has a flow rate of 30,500 scfm and results in the following average exhaust ventilation system activities: Example calculation of the average concentration of 1311 released through the exhaust stack during the first hour: (1 31 1 activity in containment in µCi/scf x scfleakage rate x 0.5) I (30,500 ft3/min x 60 min/hr x 28,317 ml/ft3) (2.94 x 10-02 µCi/scfx 595.6 scf/hr x 1hrx0.5) I 5.182 x 10+ 10 ml/hr 1.69 x 10- 10 µCi/ml Note: A similar, but longer, calculation is used to determine the average concentration in air exiting the exhaust stack over the full 16.5 hours for the radioiodines listed below. Average Radioiodine Concentrations in Air Leaking from Containment and Exiting the Exhaust Stack Over the 16.5 Hour Period 131 1 - 8.60 x 10- 11 µCi/ml 133 1 - 4.32 x 10-10 µCi/ml 135 1 - 2.92 x 10-10 µCi/ml 132 l- 7.56 x 10- 11 µCi/ml 134 1- 8.01 x 10-11 µCi/ml Example calculation of 85Kr released through the exhaust stack during the first hour: 85 ( Kr activity/scfx 595.6 scf/hr x 1 hr) I (30,500 ft3/min x 60 min/hr x 1hrx28,317 ml/ft3) (8.70 x 10-03 µCi/scfx 595.6 set) I 5.182 x 10+1° ml 1.00 x 10- 10 µCi/ml Note: Same calculation is used for the other noble gases listed below. Average Noble Gas Concentrations in Air Leaking from Containment and Exiting the Exhaust Stack Over the 16.5 Hour Period 85 Kr - 5.19 X 10-ll µCi/ml 87 Kr - 7 .66 x 1o-08 µCi/ml 89 K.r - 1.14 X 10-lO µCi/ml 85 08 mKr - 9.69 x 10- µCi/ml 88 Kr - 2.07 x 1o- µCi/ml 07 9

                                                                                  °Kr - 1.40 x 10-24 µCi/ml 133                                                                           138 Xe - 4.76 x 10-07 µCi/ml           135 mXe - 5.61x10-09 µCi/ml                Xe - 2.73 x 10-os µCi/ml 135 Xe - 2.40 x 1o- µCi/ml 07               137                    10 Xe - 4. 70 x 10- µCi/ml          139 Xe - 2.10 x 10-21 µCi/ml Assuming, as stated earlier, that (1) the average leakage rate from the containment building is 5.2 scfm, (2) the leak continues for about 16.5 hours in order to equalize the containment building pressure with atmospheric pressure, (3) the flow rate through the facility's ventilation exhaust stack is 30,500 scfm, (4) the reduction in concentration from the point of discharge at the exhaust stack to the point of maximum concentration in the unrestricted area is a factor of 292, and (5) there is no decay of any radioiodines or noble gases following release from the stack, then the following 51of54

concentrations of radioiodines and noble gases with their corresponding radiation doses will occur in the unrestricted area. The values listed are for the point of maximum concentration in the unrestricted area assuming a uniform, semi-spherical cloud geometry for noble gas submersion and further assuming that the most conservative (worst-case) meteorological conditions exist for the entire 16.5-hour period of containment leakage following target failure. Radiation doses are calculated based on a 16.5-hour exposure period. Dose values for the unrestricted area were obtained using the same methodology that was used to determine doses inside the containment building, and it was assumed that an individual was present at the point of maximum concentration for the full 16.5 hours that the containment building was leaking. A worst-case scenario dilution factor of 292 for effluent dilution using the Pasquill-Guifford Model for atmospheric dilution is used in this analysis. We assume that all offsite (public) dose occurs under these atmospheric conditions at the site of interest, i.e. 760 meters North of MURR. In our case at 760 meters, it occurs only during Stability Class 'F' conditions, which normally only occurs 11.4% of the time when the wind blows from the south. Thus this calculation is conservative. 10 CFR 20 Appendix B Effluent Concentration Limits are used for the "listed" isotopes. Effluent Concentration Limits were calculated for each of the four (4) "unlisted" noble gases (Kr-89, Kr-90, Xe-137 and Xe-139) using the data and methodology contained in FGR No. 12 for submersion isotopes. The DAC value was first calculated and then a factor of 219 was applied using 10 CFR 20, Appendix B, as a reference point for DAC values from submersion isotopes. Exposure at 1 DAC equates to 5000 mrem per year whereas at the Effluent Concentration Limit it is 50 mrem per year. Thus, there is a factor of 100 times lower allowable dose for the Effluent Concentration Limit as compared to the DAC. Exposure at the Effluent Concentration Limit assumes you are in that effluent concentration for 8760 hours per year. Therefore, the time assumed to be exposed to the Effluent Concentration Limit is a factor of 4.38 times longer than the 2000 hours per year that defines a DAC. No credit is taken for transit time from the exhaust stack to the receptor point nor is credit taken for decay inside the containment building until release. In the case ofK.r-89 and Xe-137, the transit time alone would be approximately one (1) half-life while the transit time for Kr-90 and Xe-139 would be at least four (4) half-lives. Example calculation of whole body dose in the unrestricted area due to 13!1: Conversion Factor: (Public dose limit of 50 mrem/yr) x (1 yr/8760 hrs)= 5.71x10-03 mrem/hr 131 1 concentration 8.60 x 10- 11 µCi/ml 131 1 effluent concentration limit 2.00 x 10-10 µCi/ml 131 1 Conversion Factor 5. 71 x 10-03 mrem/hr 131 Therefore, a 16.5-hour whole body exposure from 1 is: 131 1 concentration I (13!1 effluent concentration limit x Conversion Factor x 16.5 hrs) (8.60 x 10-11 µCi/ml I 2.00 x 10-10 µCi/ml) x 5.71x10-03 mrem/hr x 16.5 hrs/292 1.39 x 10-04 mrem 52of54

Note: Same calculation is used for the other isotopes (radioiodines and noble gases) listed below. Effluent Concentration Limits, Concentrations at Point of Maximum Concentration and Radiation Doses in the Unrestricted Area - Radioiodine Radionuclide Effluent Limit Maximum Concentration 1 Radiation Dose 1311 2.00 x 10* 10 µCi/ml 2.95 x 10*13 µCi/ml 1.39 x 10*04 mrem 1321 2.00 x 10-08 µCi/ml 2.59 x 10*13 µCi/ml 1.22 x 10"06 mrem 1331 1.00 x 10*09 µCi/ml 1.48 x 10*12 µCi/ml 1.39 x 10*04 mrem 1341 6.00 x 1o-08 µCi/ml 2.74 x 10-13 µCi/ml 4.31 x 10*01 mrem 1351 6.00 x 10*09 µCi/ml 9.99 x 10-13 µCi/ml 1.57 x 10-os mrem Total = 2.95E-04 mrem Note 1: Maximum Concentrations are radioiodine and noble gas concentrations leaking from the containment building and exiting the exhaust stack reduced by a dilution factor of292. Effluent Concentration Limits. Concentrations at Point of Maximum Concentration and Radiation Doses in the Unrestricted Area - Noble Gases Radionuclide Effluent Limit Maximum Concentration 1 Radiation Dose 85Kr 7.00 x 10*07 µCi/ml 1. 78 x 10- 13 µCi/ml 2.39 x 10-08 mrem 85mKr 1.00 x 10*01 µCi/ml 3.32 X 10-IO µCi/ml 3.13 x 10*04 mrem 81Kr 2.00 x 1o-08 µCi/ml 2.62 x 10*10 µCi/ml 1.24 x 10*03 mrem 88Kr 9.00 x 10*09 µCi/ml 7.08 x 10- 10 µCi/ml 7.41x10*03 mrem 89Kr 8.60 x 10*09 µCi/ml 3.92 x 10* 13 µCi/ml 4.29 x 10-06 mrem 90Kr 1.20 x 1o-08 µCi/ml 4.81x10*27 µCi/ml 3. 78 x 10*20 mrem 133Xe 5.00 x 10*01 µCi/ml 1.63 x 10*09 µCi/ml 3.07 x 10*04 mrem 135Xe 7.00 x 10-08 µCi/ml 8.23 x 10*10 µCi/ml 1.11 x 10*03 mrem 13smxe 4.00 x 1o-08 µCi/ml 1.92 x 10*11 µCi/ml 4.53 x 10-os mrem 137 Xe 9.10 x 10-08 µCi/ml 1.61 x 10*12 µCi/ml 1.67 x 1o-06 mrem 138Xe 2.00 x 10-08 µCi/ml 9.34 x 10- 11 µCi/ml 4.40 x 10*04 mrem 139Xe 1.60 x 1o-08 µCi/ml 7.20 x 10*24 µCi/ml 4.24 x 10* 17 mrem Total = 1.09E-02 mrem Note 1: Maximum Concentrations are radioiodine and noble gas concentrations leaking from the containment building and exiting the exhaust stack reduced by a dilution factor of 292. To finalize the unrestricted dose in terms of Total Effective Dose Equivalent (TEDE), the doses from the radioiodines and noble gases must be added together, and result in the following values: 53of54

Dose from Radioiodines and Noble Gases in the Unrestricted Area Total Iodine - Committed Effective Dose Equivalent (CEDE) 2.95E-04 mrem Noble Gas - Committed Effective Dose Equivalent (CEDE) l.09E-02 mrem Total Dose - Total Effective Dose Equivalent (TEDE) 1.12E-02 mrem Summing the doses from the noble gases and the radioiodines simply substantiates earlier statements regarding the very low levels in the unrestricted area should a failure of a fueled experiment occur, and should the containment building leak following such an event. Due to low concentrations present at the modeled site of interest, dose to the public is low with the noble gases and iodines each contributing about equally. The overall TEDE is much less than 1 mrem, a value far below the applicable 10 CFR 20 regulatory limit for dose to the public in an unrestricted area. 54 of54

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ATTACHMENT 1

.$

1.. 2 3 4 5 6* 7' g 9' 'le 11* . *Ii 13 14 15 "16 17 12 19

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"22

.23 24

ATTACHMENT 2 Fuel Failure during Reactor Operation - Resricted Area Dose Time interval: t = 0-1 min t = 1-3 min t = 3-5 min t = 5-7 min t = 7-9 min t = 9-10 min 4 #1 Plates Primary Activity Pool Containment Containment Containment Containment Containment Containment Dose to Core Actvity Coolant in Pool Act. Cone. Concentration Concentration Concentration Concentration Concentration Concentration Average Workers Isotope T112 Actvity 0.0342 Activity at 10 min at 10 min (uCi/cc) w/ decay w/ decay w/ decay w/ decay w/decay Concentration DAC 10 min (Ci) (Ci) (uCi/gal) (uCi) (uCi/gal) No I evap@T0 (uCi/cc) (uCi/cc) (uCi/cc) (uCi/cc) (uCi/cc) (uCi/cc) mrem Time min: 0.5 2 4 6 8 9.5 1-131 8.02 d 2.20E+05 7.52E+03 3.76E+06 2.97E+05 1.49E+01 2.33E-10 3.73E-09 1.49E-08 3.36E-08 5.97E-08 8.41 E-08 3.08E-08 2.00E-08 6.42E+OO 1-132 2.28 h 3.08E+05 1.05E+04 5.27E+06 4.16E+05 2.08E+01 3.26E-10 5.17E-09 2.05E-08 4.56E-08 8.03E-08 1.12E-07 4.16E-08 3.00E-06 5.78E-02 1-133 20.8 h 5.42E+05 1.85E+04 9.27E+06 7.32E+05 3.66E+01 5.74E-10 9.18E-09 3.67E-08 8.25E-08 1.46E-07 2.06E-07 7.56E-08 1.00E-07 3.15E+OO 1-134 52 .6 m 6.11E+05 2.09E+04 1.04E+07 8.25E+05 4.13E+01 6.43E-10 1.01 E-08 3.93E-08 8.62E-08 1.49E-07 2.06E-07 7.77E-08 2.00E-05 1.62E-02 1-135 6.57 h 5.06E+05 1.73E+04 8.65E+06 6.84E+05 3.42E+01 5.36E-10 8.55E-09 3.41 E-08 7.64E-08 1.35E-07 1.90E-07 7.00E-08 7.00E-07 4.17E-01 1.01E+01 Kr-85 10.76 y 4.63E+02 1.58E+01 7.92E+03 6.25E+02 4.91E-09 1.96E-08 3.93E-08 5.89E-08 7.85E-08 9.33E-08 4.91 E-08 1.00E-04 2.05E-04 Kr-85m 4.48 h 1.31 E+05 4.48E+03 2.24E+06 1.77E+05 1.39E-06 5.53E-06 1.10E-05 1.64E-05 2.18E-05 2.57E-05 1.37E-05 2.00E-05 2.85E-01 Kr-87 1.27 h 2.05E+05 7.01 E+03 3.51 E+06 2.77E+05 2.16E-06 8.54E-06 1.68E-05 2.47E-05 3.23E-05 3.79E-05 2.05E-05 5.00E-06 1.71 E+OO Kr-88 2.84 h 2.91E+05 9.95E+03 4.98E+06 3.93E+05 3.08E-06 1.22E-05 2.43E-05 3.61 E-05 4.78E-05 5.64E-05 3.00E-05 2.00E-06 6.26E+OO Kr-89 3.15 m 3.69E+05 1.26E+04 6.31 E+06 4.98E+05 3.50E-06 1.01 E-05 1.30E-05 1.25E-05 1.08E-05 9.19E-06 1.05E-05 1.90E-06 2.31 E+OO Kr-90 32.3 s 3.68E+05 1.26E+04 6.29E+06 4.97E+05 2.05E-06 1.19E-06 1.81 E-07 2.07E-08 2.10E-09 3.62E-10 4.84E-07 2.SOE-06 7.20E-01 1.13E+01 Xe-133 5.243 d 3.85E+05 1.32E+04 6.58E+06 5.20E+05 4.08E-06 1.63E-05 3.26E-05 4.90E-05 6.53E-05 7.75E-05 4.08E-05 1.00E-04 1.70E-01 Xe-135 9.1 h 7.56E+04 2.59E+03 1.29E+06 1.02E+05 8.01 E-07 3.20E-06 6.38E-06 9.54E-06 1.27E-05 1.50E-05 7.95E-06 1.00E-05 3.31 E-01 Xe-135m 15.3 m 3.62E+04 1.24E+03 6.19E+05 4.89E+04 3.75E-07 1.40E-06 2.56E-06 3.51 E-06 4.27E-06 4.74E-06 2.86E-06 9.00E-06 1.32E-01 Xe-137 3.82 m 4.81 E+05 1.65E+04 8.23E+06 6.50E+05 4.66E-06 1.42E-05 1.97E-05 2.06E-05 1.91 E-05 1.73E-05 1.69E-05 2.00E-05 3.53E-01 Xe-138 14.1 m 5.01 E+05 1.71 E+04 8.57E+06 6.77E+05 5.18E-06 1.93E-05 3.49E-05 4.75E-05 5.74E-05 6.33E-05 3.86E-05 4.00E-06 4.03E+OO Xe-139 39.7 s 4.07E+05 1.39E+04 6.96E+06 5.50E+05 2.56E-06 2.12E-06 5.23E-07 9.66E-08 1.59E-08 3.91E-09 8.08E-07 3.70E-06 9.10E-02 5.10E+OO - Denotes DACs calculated using methodology as described in FGR No. 12. Total Iodine (COE) (mrem) 10.07 Iodine (CEDE) (mrem) (CDE x 0.03) 0.302 Noble Gas (CEDE) mrem 16.38 Total Dose (TEDE) mrem 16.69

ATTACHMENT3 Fuel Failure during Reactor Operation - Unresricted Area Dose 4 #1 Plates Actvity Activity Pool Containment Containment Containment Containment Containment Containment Containment Containment Containment Exhaust Total Exhaust Effluent Concentration Dose to Core Actvity in in Pool Act. Cone. Concentration Concentration Concentration Concentration Concentration Concentration Concentration Concentration Concentration 1st hour Exhaust Overall Cone . at Maximum Public Isotope T112 Actvity 0.0342 Coolant at 10 min at 10 min No I evap@T 0 w/ decay w/decay w/ decay w/ decay w/decay w/ decay w/ decay w / decay Ave . Cone. Activity Ave . Cone . Limit Dose Point 16.5 hr Ci Ci uCi/ al uCi uCi/ al uCi/sc uCi/sc uCi/sc uCi/sc uCi/sc uCi/sc uCi/sc uCi/sc uCi/sc uCi/cc uCi uCi/cc uCi/cc EC/292 mrem 1 Hr Intervals: 0.5 1.5 2.5 3.5 4.5 4 Hr Intervals: 7 11 14.75 Containment Volume scf: 229801 229224 228684 228181.0 227714.0 226733.0 225605.0 225085.0 Containment Avera e Leaka e Rate scf/hr: 595.6 558 .7 521 .8 485 448.1 355.8 207.9 67.9 1-131 8.02 d 2.20 E+05 7.52 E+03 3.76E+06 2.97E+05 1.49 E+01 2.58E-03 2.57E-03 2.56E-03 2.55 E-03 2.54E-03 2.52E-03 2.49E-03 2.45 E-03 1.48E-11 6.47E+OO 7.56E-12 2.00 E-10 2.59E-14 1.22E-05 1-132 2.28 h 3.08 E+05 1.05E+04 5.27E+06 4.16E+05 2.08E+01 3.11E-03 2.30 E-03 1.69E-03 1.25 E-03 9.22 E-04 4.31 E-04 1.28E-04 4.09E-05 1.79E-11 2.88E+OO 3.37E-12 2.00 E-08 1.16E-14 5.44E-08 1-133 20.8 h 5.42E+05 1.85E+04 9.27E+06 7.32E+05 3.66E+01 6.27E-03 6.06E-03 5.86E-03 5.67 E-03 5.49E-03 5.05 E-03 4.42E-03 3.90 E-03 3.60E-11 1.36E+01 1.59E-1 1 1.00E-09 5.44E-14 5.13E-06 1-134 52.6 m 6.11E+05 2.09E+04 1.04E+07 8.25E+05 4.13E+01 4.84E-03 2.19E-03 9.96E-04 4.52 E-04 2.05E-04 2.84E-05 1.20 E-06 6.20 E-08 2.78E-11 2.49E+OO 2.91 E-12 6.00 E-08 9.97 E-15 1.57E-08 1-135 6.57 h 5.06E+05 1.73E+04 8.65E+06 6.84E+05 3.42E+01 5.64E-03 5.08 E-03 4.57E-03 4.11 E-0 3 3.70E-03 2.84E-03 1.86E-03 1.26 E-03 3.24E-11 9.07E+OO 1.06E-11 6.00 E-09 3.63 E-14 5.70 E-07 1.80E-05 Kr-85 10.76 y 4.63E+02 1.58 E+01 7.92E+03 6.25E+02 2.72E-03 2.72E-03 2.72E-03 2.72E-03 2.72E-03 2.72E-03 2.72E-03 2.72 E-03 3.13E-11 1.39 E+01 1.62E-11 7.00E-07 5.56E-14 2.19E-06 Kr-85m 4.48 h 1.31 E+05 4.48E+03 2.24E+06 1.77E+05 7.13E-01 6.11 E-01 5.23E-01 4.48 E-01 3.84E-01 2.61 E-01 1.40E-01 7.86 E-02 8.19E-09 1.93E+03 2.26E-09 1.00 E-07 7.75E-12 2.13E-03 Kr-87 1.27 h 2.05E+05 7.01E+03 3.51E+06 2.77E+05 9.17E-01 5.32 E-01 3.08E-01 1.78E-01 1.03E-01 2.64E-02 2.98E-03 3.85 E-04 1.05E-08 1.18E+03 1.38E-09 2.00E-08 4.71E-12 6.49E-03 Kr-88 2.84 h 2.91 E+05 9.95 E+03 4.98E+06 3.93E+05 1.51 E+OO 1.19E+OO 9.29E-01 7.28 E-01 5.71 E-01 3.10 E-01 1.17E-01 4.68 E-02 1.74E-08 3.21 E+03 3.75E-09 9.00 E-09 1.28E-11 3.93E-02 Kr-89 3.15 m 3.69E+05 1.26E+04 6.31 E+06 4.98E+05 2.95E-03 5.46 E-09 1.01E-14 1.87 E-20 3.46E-26 1.61 E-40 1.89 E-63 6.01 E-85 3.39E-11 1.76E+OO 2.06E-1 2 8. :;og 7.04E-15 2.25E-05 Kr-90 32.3 s 3.68E+05 1.26 E+04 6.29E+06 4.97E+05 3.66E-17 1.04E-50 2.98E-84 8.52E-118 2.43E-151 3.35E-235 O.OOE+OO O.OO E+OO 4.20E-25 2.18E-14 2.55E-26 1.20&-08 8.72E-29 2.00 E-19 4.79E-02 Xe-133 5.243 d 3.85E+05 1.32E+04 6.58E+06 5.20E+05 2.26E+OO 2.24E+OO 2.23E+OO 2.22E+OO 2.21 E+OO 2.18E+OO 2.13E+OO 2.09E+OO 2.59E-08 1.12E+04 1.31 E-08 4.48E-11 2.47E-03 Xe-1 35 9.1 h 7.56E+04 2.59 E+03 1.29E+06 1.02E+05 4.28 E-01 3.96E-01 3.67E-01 3.40 E-01 3.15E-01 2.61E-01 1.92 E-01 1.45E-01 4.92E-09 1.54E+03 1.80 E-09 6.17E-12 2.42E-03 Xe-135m 15.3 m 3.62E+04 1.24E+03 6.19E+05 4.89 E+04 5.47E-02 3.61 E-03 2.38E-04 1.57E-05 1.04E-06 1.16E-09 2.21E-14 8.30 E-19 6.29 E-1 0 3.47E+01 4.06E-11 1.39E-13 9.56E-05 Xe-137 3.82 m 4.81E+05 1.65E+04 8.23E+06 6.50E+05 1.22E-02 2.29 E-07 4.30E-12 8.06E-17 1.51E-21 2.30 E-33 2.83E-52 5.31 E-70 1.41 E-10 7.29E+OO 8.53E-12 2.92E-14 8.83E-06 Xe-138 14.1 m 5.01E+05 1.71E+04 8.57E+06 6.77E+05 6.74E-01 3.53E-02 1.85E-03 9.70E-05 5.08E-06 3.19E-09 2.41 E-14 3.79E-19 7.75E-09 4.22E+02 4.94E-10 1.69E-12 2.33E-03 Xe-139 39.7 s 4.07E+05 1.39 E+04 6.96E+06 5.50E+05 5.41 E-14 2.76E-41 1.41 E-68 7.21 E-96 3.69 E-123 2.17E-191 1.48E-300 O.OO E+OO 6.22 E-22 3.22 E-11 3. 77E-23 1.29E-25 2.22E-16 7.32E-03 Denotes Effluent Concentrations derived from DACs calculated using methodology as described in FGR No. 12. (DAC/219) Total Iodine (CEDE) mrem 1.80E-05 Noble Gas {TEDE) mrem 5.52E-02 Total Dose (TEDE) mrem 5.53E-02

ATTACHMENT 4 MURR Health & Safety Technical Basis Report DATE: March 20, 2016 (Supersedes: January 11, 2016) BY: Nathan G. Hogue, Health Physics Manager~ TITLE: 16-01: Determination of Derived Air Concentrations for Kr-89, Kr-90, Xe-137, and Xe-139. Purpose The purpose of this Technical Basis is to document the derivation of derived air concentrations (DAC) for Kr-89, Kr-90, Xe-137, and Xe-139 as no current published values exist within 10CFR20 APP 8, Federal Guidance Report 11, or Federal Guidance Report 12. Backeround In conjunction with the MURR application for renewal of the Facility Operating License during evaluation of the maximum hypothetical accident four nuclides (Kr-89, Kr-90, Xe-137, and Xe-139) were observed to be present during a containment evacuation without an available DAC reference for dose determination. Derivations of specific DAC values corresponding to each of these nuclides were needed to determine proper dose consequences during the maximum hypothetical accident. A comprehensive review of Federal Guidance Report (FGR) - 11 and FGR- 12 was conducted to derive each value based on industry acceptable methodologies within. All the nuclides in question are common in that they are noble gases and are not metabolized to an appreciable extent by the body. The methodology for calculating a DAC for materials of this nature, established within FGR-11 and continued in FGR-12, are based on consideration of external dose due to submersion in air only. In the case of air submersion a worker is assumed to be immersed in a pure parent semi-infinite homogeneous cloud and no radiation from airborne progeny, absorbed gas within the body, or inhalation of radioactive decay products are considered. The derivations within this report also make the aforementioned assumptions. FGR-12 provides an updated table compared to those in FGR-11 of dose coefficients for air submersion based on use of the continuous energy Monte Carlo photon transport code ALGAMP. Within FGR-12 calculations were performed for twelve monoenergetic sources ranging from 0.01 to 5.0 MeV incident on a phantom with Page 1of7

ATTACHMENT 4 spectra generated by the PHO FLUX code for air submersion and normalized to a source strength to determine dose coefficients presented in Table 11.4. Coefficients for effective dose equivalent were also presented in Table 11.4 based on ICRP 60 weighting factors. Using the results of these calculations nuclide specific dose coefficients were derived taking into account the specific photon spectrum of each radionuclide and using interpolation of the determined dose coefficients from the twelve point monoenergetic source. The nuclide specific photon spectra were gathered from the Brookhaven National Laboratory National Nuclear Data Center database and Table 1 of Erdtmann/Soyka gamma ray spectroscopy reference text. The use of both references allowed for collection of a comprehensive listing. A direct repetition of the methodology of using interpolation of the data presented in Table 11.4 of FGR-12 was used to calculate photon dose coefficients for each of the nuclides in question. The skin dose and effective dose equivalent data from Table 11.4 FGR-12 were plotted in excel and assigned a best fit line used for the interpolation (See Figure 1 and 2). Figure 1 Graph of Table 11.4 FGR. 12 EDE 3.SOE-13 3.00E-13 - + - - - - - - - - - - - - - - - - - - - - - - y = 2E-15x + SE-14x - SE-16 R2 =1

   ~      2.SOE-13  + - - - - - - - - - - - - - - - - - ---- - - - -

S Ill g 2.00E-13

   ...a.>

c. 1 SOE-13 -+-----------~------------

   ~ *
   ~      1.00E- 13 + - - - - - - - -_,,...._________________

0.00E+OO ~'----.-------.-----,--------.------r----, 0.00E+OO 1.00E+OO 2.00E+OO 3.00E+OO 4.00E+OO 5.00E+OO 6.00E+OO Photon Energy (MeV) Page 2of7

ATTACHMENT 4 Figure 2 Graph of Table 11.4 FGR-12 SOE-Skin 3.SOE . - - - - - - - - - -- - - - - * - - - - - - - -- -- - - - 3.00E-13 + - - - ---,., E.... y==-........ -1..,. 6x ="'l2.....,.+oFl"4x:-61r-" R2 = 1

' 2.SOE-13 +------------------/~-----

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i

  ~"'    1.00E-13 _,__ _ _ _ _ _,,_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

0.00E+OO .,_.:___ _~---~---~---~---~--~ O.OOE+OO l.OOE+OO 2.00E+OO 3.00E+OO 4.00E+OO 5.00E+OO 6.00E+OO Photon Energy (MeV) Examination of the limiting coefficients for other isotopes of each nuclide, given by bold-face type within FGR-11Table2.3, was completed to determine if other organs should be considered in the determination of the most limiting DAC. During the review of FGR-11 and FGR-12 either the effective dose equivalent or skin was used as the limiting factor and therefore these two components will be considered to ensure the most limiting is selected in derivation of the DAC values. For the calculation of effective dose equivalent only penetrating photon radiation was considered in the derivation as delineated in FGR-12; the derivation of electrons to dose to organs and tissues of the body other than the skin need not be considered, due to the short range in tissue of electrons emitted. For the calculation of skin dose both photon and beta radiation was considered in determining the dose coefficients. The decay emission types, energies, and intensities for shallow dose derivations were gathered from the Brookhaven National Laboratory National Nuclear Data Center database. FGR-12 provides Figure 11.25, Electron skin dose coefficient for submersion in contaminated air, which was used to interpolate the dose coefficient from the beta average energies (one-third beta max) as seen in Figure 3. For emissions that result in a negative Page 3of7

ATTACHMENT 4 dose coefficient the values were excluded from the DAC derivation. This is considered acceptable as the decay energies resulting in a negative coefficient are not sufficient to contribute to the dose. Figure 3 Graph of Figure 11.25 FGR-12 SDE-Beta 8.00E-01 . - - - - - - -- - - -- - - - - - - - - - - - -

   ,.... 7.00E-01 +--- - - - - - - - - - - -- - - --=...-- - -

1= 6.00E-Ol ~- =--'-0=.0~0;...=;0..:;; 4x;.:..3_+.. . 0;:"'"'.0""""0-=4=3x:=...2_+.. .0=

                                                                                        ;: .0'-'-7...:.0=

Sxc:...--=0=.0"-"0=5= 8 ---,,.,...c-_ _ _ _ _ __ R =1 2 C"

i. 5.00E-01 + - - - - - - - - - - - -,,,,- - - - - - - - - - -

e11 3° 4.00E-01 +---------~~----------- iii

   ~ 3.00E-01       + - - - - - - ----- - - - - - - - - - - - - --

_e;

      ~   2.00E-01  + - - - - -,,,C-- - - - - - - - - - -- - - - --
   .r:."'

l.OOE-01 +--~~------------------- 0 2 4 6 8 10 12 Beta Energy (MeV) An example subset of the derivation of both the effective dose equivalent and skin dose coefficients are provided below for the readers review. The entire calculation was performed within excel and is attached in two versions; Attachment 1) with the formulas displayed and Attachments 2-5) with the values displayed. The same methodology was used for all other nuclides within an excel file and are also include as Attachments 2-5. Calculation Effective Dose Equivalent DAC Equation 1, derived from plotting the hE data from Table 11.4 of FGR-12, was used to interpolate a dose coefficient specific to each photon energy (Eqn. 1) y = 2E - 15x 2 + SE - 14x - 8E - 16 Y = hE(Gy per Bq s m- 3 ) x =Photon energy in MeV Page4of7

ATTACHMENT4 Equation 2 was then used to determine a dose rate per unit activity value specific to each photon energy (Eqn. 2) R = yxi R = Dose rate per unit activity y =interpolated dose.coefficient from Eqn.1 i =intensity of decay Equation 3 is the summation of each dose rate value determined in Eqn. 2 for each photon energy to determine the total dose coefficient specific to the nuclide (Eqn. 3) S = R1 + R2 + R3 + ... Rn S = Sum of dose rate factors per unit activity R = Dose rate per unit activity from Eqn. 2 ! Equation 4.is used to convert the units of the total dose coefficient to mRem hr-1 per

µCi mI-1 (Eqn. 4)       E = S(l.33E19)

E =EDE coefficient specific to the nuclide S = Sum of dose rate factors per unit activity from Eqn. 3 Equation 5 determines the DAC by scaling'the derived value to the annual occupational limit of 5 Rem or 2.5 mRem/hr (Eqn. 5) D = 2.s E D =EDE Derived Air Concentration in µCi mr 1 of submersed air E = EDE coefficient specific to the nuclide from Eqn. 4. Shallow Dose Equivalent DAC In the derivation ofSDE both a photon component and electron comp~nent will contribute to dose therefore both must be considered. First we will consider the photon component in Equations 6 through 10 and later the electron component in equations 11through15. Equation 6, derived from plotting the hskin data from Table II.4 ofFGR-12, was used to ,interpolate a dose coefficient specific to each photon energy (Eqn. 6) y = BE - 16x2 + 6E - 14x - 6E - 16 y = hskin(Gyper Bq sm- 3 ) x = Photon energy in MeV Page 5 of7

ATTACHMENT 4 Equation 7 was then used to determine a dose rate per unit activity value specific to each photon energy (Eqn. 7) R = yxi R = Dose rate per unit activity y = interpolated dose coefficient from Eqn. 6 i = intensity of decay Equation 8 is the summation of each dose rate value determined in Eqn. 7 for each photon energy to determine the total dose coefficient specific to the nuclide (Eqn. 8) S = R1 + R2 + R3 +... Rn S =Sum of dose rate factors per unit activity R =Dose rate per unit activity from Eqn. 7 Equation 9 is used to convert the units of the total dose coefficient to mRem hr- 1 per µCi ml* 1 (Eqn. 9) E = S(1.33El 9) E = SDE coefficient specific to the nuclide S =Sum of dose rate factors per unit activity from Eqn. 8 Equation 10 determines the DAC by scaling the derived value to the annual occupational limit of 50 Rem or 25 mRem/hr (Eqn.10) D= 2s E D = SDE Derived Air Conc{!ntration in µCi mz- 1 of submersed air

E SDE coefficient specific to the nuclide from Eqn. 9 Equation 11, derived from plotting the hskin data from Figure 11.25 of FGR-12, was used to interpolate a dose coefficient specific to each beta energy (Eqn.11) y = -0.0004x 3 + 0.0043x 2 + 0.0705x - 0.0058

y hskin(pSv s- 1 per Bq m- 3 ) x = Beta energy in MeV Equation 12 was then used to determine a dose rate per unit activity value specific to each average (one third of the maximum) beta energy (Eqn.12) R = yxi R = Dose rate per unit activity y =interpolated dose coefficient from Eqn.11 i = intensity of decay Page 6 of7

ATTACHMENT4 Equation 13 is the summation of each dose rate value determined in Eqn. 12 for each beta energy to determine the total dose coefficient specific to the nuclide (Eqn.13) S = R1 + Rz + Rg+ ... Rn S = Sum of dose rate factors per unit activity R = Dose rate per unit activity from Eqn. 12 Equation 14 is used to convert the units of the total dose coefficient to mRem hr-1 per µCi ml*1 (Eqn. 14) E = S(1.33E7) E = SDE coefficient specific to the nuclide S =Sum of dose rate factors per unit activity from Eqn.13 Equation 15 determines the DAC by scaling the derived value to the annual occupational limit of 50 Rem or 25 mRem/hr (Eqn.15) D = 2s E D = SDE Derived Air Concentration in µCi ml- 1 of submersed air E = SDE coefficient specific to the nuclide from Eqn.14 Equation 16 sums the SDE DAC from both photon emission and beta emission to determine the total SDE DAC for the nuclide (Eqn. 16) T = EsDE photon + EsDE beta T = SDErotal Derived Air Concentration in µCi ml- 1 Conclusion Based on the derivation described above the following nuclides may be assigned the following limiting derived air concentration values to determine dose consequence. Nuclide DAC C1-1Ci/ml) Kr-89 1.9 E-6 (EDE) Kr-90 2.8. E-6 (EDE) Xe-137 2.0 E-5 (EDE) Xe-139 3.7 E-6 [EDE) Page 7of7

ATTACHMENT 4 Attachment 1

ATTACHMENT4

'* I 5 PhotoriEmiss!ons:

6 Deterrnln.tlon ofDACUmlted tly EOE lntensitv lnterpelatedGammaDoseFactor DoseRate erUnitAet<vi 7 E~=

                                                    "'~'

Energy

    (keV)
                        '"'~

(%) jMode J. hE(GyperBqsm.J) l Gy/s 104701.5 =Al0/1000 0.01 11 685.6 =All/1000 0.008 _, 12 655.6 =A12/l000 0.01 13 631.5 =AB/1000 0.028 144489.2 =Al4/1000 0.135 447S.3 =AlS/1000 0.014 b-

                        =A16/l000     0.01          b-4405.l               =A17/1000     0.008         b-

~:~ 2of4tU 21 307.4 :A21/1000 0.01 224279.4 =A22/1000 0.02 23 267.7 =A23/1000 0.02 244253.3 "A2A/1000 0.014 254184.3 =A25/1000 0.05 264176.2 -A26/1000 0.012 b- =0.000000000000002°B26°B26)+0.00000CIOOOOOOOS0 (826)).(l.OOOOOOOOC:OOOOO =C26/100°E25 27 4162.6 =1>;1.7/1000 0.028 b- =(0.0000XIOOOOC(lQ(2°(827°B27))+(0.00000000000CXlS 0 (B27))-0.00COOOOOOOOO:X: =C27/100"E27

                        =A28/1000     0.016         b-     =O.OOOOOOCXlOOCI002°B28°82S)+{O.OOOOOOOOOOOCXlS 0 028)).C.OOOOOCIOOO:lOOO        =C2S/100°E28
                        =A29/l000     0-026         b-     =(0.000000000000002°/829°829))+(0.00000000000CXlS"fB29))-0.00000000COOOOOO       =C29/100°E29 eA30/1000     0.014         b-     =(0.000CXXX)00000002°(830°B30))+/0.00000CI00000005°(B30))-0.000JOOOOCOOOOOO      =C30/100°E30
                        -A31/1000     0.074         b-     =(0.000000000000002°(B31"B31))+/0.00000000000005"fB31))-0.0CO)OOOOC()COOO        =C31}100°E31
                        =A32/1000     0.117         b-     =(o.OOOOJ0000000002*ren*e32)!+10.00000000000Xls*1e32)l-O.oocooooocoooi:r         =C32/100"E32
                        =A33/1000     0.02          b-
                        =A34/1000     0.028         b-
                        -A35/1000     0.143         b-
                        =A36/1000     0.27          b-37't3977.5               "A37/1000     0.'17         b-38.J.3965.5              =A3Sf1000     0.209         b-
                        =A39/1000     OA2           b-
                        =A40/1000     0.135         b-
                        "A41/1000     0.034
                        =A42/1000     o...

coA43f1000 0.111 4413837.6 =MA/1000 0.082 lb-

                        =A45/1000     0.139       J b-
                        "A46/1000
                        =A47/1rtXJ
                        =A48/1000     0.016 49[37325                 "A49/1000     0.14        1 b-SQ.1_3721.5              =AS0/1000     0.048       J b-
                        =ASl/1000     o."
                        "AS2/1000     0.066
                        =AS3/1000     0-""
                        =A54/1000     0.058
                        =ASS/1000     0.038
                        =AS6/1000     0.038
                                                    .
                        =AS7/1000     o."'
                        =ASS/1000     0.259 1{1000   0.06          b 1/1000   0.056 n<m                   b-sil3so3.6 633463..3 "T=A62/1000
                        =A63/l000
                                      "

O.D2 o... J b-643439.6 coA54/1000 0.04 653399.9 ...i=_A6~fl000

                                     -%.*

1.05 0.042 b- "(0.0000J0000000002"{868°B68})+[0.l'.lOOOCO'.KXXl5°B68))-0.0C:OOOOOOCOOOOOB =C6Sf100°E6S 0.058 b- =(0.0000Cl0000000002"{869°B69})+(0.00000000000005°f869))-0.00COOOOOC0000008 =C69/100°E69 0.036 b- =(0.000000000'.l0002°(BWB70))+(0.0000COOOOODO:lS 0 f870))-0.00COOOOXl0000008 =C70/100°E70 0.07 b- =(0.000CXl0000000002°B71°B71+0.0CJOClCOOOOOOCO*fB71)-0.000'.IOCIOOOOOOB =C71/100"E71 7il3317.9 "T=N2/1000 0.082 b- =(O.OOOOOOOOOOCXX:Xl2°(B72°B72})+(0.000CXXlOOOOOOOS 0(B72)}-0.00Xl000000000008 "C72/tOO*E72 733300 =A73/1000 0.038 b- -(O.oooooooc:ocx::o:m*(073*e1s})+fo.00000000000005*1013))-0.00CQOOOOOXIOOOD =C73f100°E73 74 327L3 =A74/1000 0.05< b- =(o.ooocoooco:ioooo2*974*e11i,J+10.00000000000005*fB74)J-O.OOOOOOOOOClOOOOB =C74/100"E74 753257 7613219.84

                        =A7S/1000 l=A76/1000 0.052 OA3
                                                    ,_

b- :(0.0000Cl0000000002"(B75°B75))+(0.00000000000CXl5°f875)J-O.OOCOOOOOCOOOOOO

                                                           =(O.OOOOJ0000000002°(B76°B76))+(0.00000'.IOOOOOC0°f876))-0.00COOOOOCOOOOOO
                                                                                                                                            =C75/l00°E75
                                                                                                                                            =C76/100°E76

ATTACHMENT 4

A78/1000 793159.8 "A79/1000 0,062

,_

80 31S4A ::AS0/1000 0.026 ~ 1.05 0.195

'"

o."" 0.33 0.096

0.133 ~ 0.04

O.D3 0.046 0.066 0.125 O.o28 0.036 0.022 0.109 ~ b-o."" 0.05 0.05

                           ,_
                           ~

0.024

                           ,_
                           ,_
                           ~

0., b-0.2' O.o36

,_

0.052 ~ 0.046 0.21

,_

0.046

,_

0.12 0.025 0.042

                         .,_
                           ~

O.o3 0.263

'-"

0.036 0.038 0.133

,_

o." b-0.034

ATTACHMENT 4 b-0.03 b-0.034 0.197

'"'

0.05 0.35 b-b- b-b- 0.141 0.03 0.129 0.0< 0.072 009 =A190/1000 0.191 =Al91/1000 b-o"" =A192/1000 b- "A193/1000

A194/1000

=A195/1000 '*' =A196/l000 "A197/1000

"'

=A198/1000 =A199/1000 =A200/1000

'""

0."'4 =A20l/1000 =A202/1000 0.179 =A203/1000 0.123 b-

A204/1000 b-

=A205/1000 0.052 =A206/1000 0.02 "A207/1WJ

A208/1000

=A209/1000 b- =A210/1000 b- "A211/1000 "A212/1000 0.195 =A213/1000 0.13

A214/1000

=A215/1000 =A216/1000 b- =A217/l000 b- =A218/1000 b- =A219/1000 b- "A220/1000 =A221/1000 0.038 =A222/1000 o.oss =Al.23/1000 =A224/1000 0.145 =A22S/10CXJ 0.022 b- "A226/1000 b- =A227/1000

A228/1000 0.185

=A229/1000 0.167

A230/1000 0.98

=A231/1000 =A232/1000 '""'

ATTACHMENT 4 0.161 b-

                      "     b-0-

0.8 b-0."6< 0-0.24 0.068 0-0.07 b-0.03 0.052 0.41 1010.84 " 249997.37 25 974.39 251 69.7 0.058 25 9&0A2 0.32 953.18 25 944.19 0.165 0.066 257934.6 25 930.95 259917.78 0.0?4 7.2 0- ' 904.27 b-890A 0.3 " ,_, 0.64 ' 887.9 0.161 857.37 ' ....7 0.52 ' 1.11 0-b- 1.13 0.92 OA b-o.on b-0.11 4.22 279729.63 0.26 ' 716.2 0.5 671A " 66"6 " 0.078 291650.5

                       '*""

o.o:.s " 652.6 0.3' 0,088 16.6 b-576.95 =A298/1000 0-

299557.3 =A299/1000

           =AJ00/1000  0.03
           =AJOl/1000  0.03
           =A30Ul000 30 510.1   =A303/1000 509.1  =A304/1000 30 498.6   =AJOS/1000  1.15 30 97.38   =ASOS/1000  6.7     0-3 490.76   =A307/l000          b-
           =A308/1000 309465.13  =A309/1000
           =A310/1000
           =A3111000

ATTACHMENT 4 0.038 2.57 0.32 0.046 319364.88 32 356.16

1.19 O.><

**

32 318.3 ' "'*' b* 1.83

**

b* 0.22

**

b*

                  ~
                  ~
                  **

Gv/sperBq/m3 mRem/hrperBci/m3 mRem/hrperuO/ml

                                                         =2.5/F342      uO/ml DetennlnatlonofOACbySOE Intern tty
                   '"'

o.* 1.5

                *.2         b*

2.2 ~ 1 0.96 "o.s* 0.13 28

                            ****

OA7 2.23 b* b*

                            ~
                            **
               '0.85 0.31 1.99 1.so 2

b* 0.074

**

b*

ATTACHMENT4

                      =A389/1000
                      =A390/1000    =B390"0.33    0.21
                      =A391/1000    =6391*0.33    0.23
                      =A.392/1000   =6392"0.33    OM
                      =A.393/1000 J-6393°0.33   _10.49    b*
                      =A394/1000                         ;;:
                     ,=A395/1000
                     '=A.396/1000
                      =~:~~~:. 1"=6398°0.33   To.1s1 399)759.3              =A.399/1000 _rB399"0.33   10.2 582.8              =A400fl000
                      =M01{1000
                      =A402/1000    =B402"0.33   To.21s
                      =A403/1000    =8403'-0.33   O.S9 "A404/1000    =8404*0.33     O.>    b*

A405/1000 =6405°0.33 0.32 ;;:

                      "A406/1000    =8406"0.33     0.32
                     ':A407/1000    =MO?*o.33    lD.57
                      =MOS/1000     =8408°0.33 (pSvs-lperllqm-3'
                                                                                                 =G410*100*1000*50*0011ooooooocoooo mRem/hrperllq/m3
                                                                                                 =G411"37000"1000000                mRem/hrperuCl/ml
                                                                                                 =25/6412                                uCi/ml

~~!Photon Emissions:

~3 ~~~~

Energy (Mo~ ,., In~

                                                    ""'"                 Do$eRate£!!.!!nitActivl Gy/s

o.ooa b*

                                    !Ml             ;;:
                                    ~               ;;:

O.BS ;;: 0.01

                                    '0:008 0.01 Mi
                                    ~

IM!6 ~

;;:

0.117 Mi ~

                                                     ;;:
                       =A446/l000
                       =A447/1000
                       =A448/1000
                       =A449{1000
                       ;A450/1000
                       =A451/1000
                       =A452/1000 0.111            b*

45413837.6 M8i' ;;: 10.066 005' M38

ATTACHMENT 4 0.038 ~ 46 3629.2 ~AA67/1000 0.08 ~ 4 3583.9 =A468/1000 0.2S9 46 3574 =AA69/1000 o.o; 47 3567.9 =A470/1000 0.056 47 3532.88 =A47l/1000 L3S ~ 47 3503.6 cA472/1000 0.02 ~ 47 3463.3 =A.473/1000 0.042 O.OM o:ii1

                             '0:62 1.05   ~
                                     .

0.042 ~ 4791"3347.4 l=A479{1000 0.0611 b-3340.8 =A480/1000 0.036 b-4B1U321.9 =AABl/1000 O.Q7 48213317.9 =A482/1000 0.082 0.038 b-0.054 ~ 0.052 ~ OA3 b-48713213.2 =M87/1000 on32 3172.1 =A4S8/1000 0.1 3159.8 =A489/l000 0.062 49C:UlS4.4 =A490/1000 0.026 4913140.26 =A.491/1000 1.0S b-49"3107.26 =M92/1000 0.195 ~ 0.038 b-0.04 ~ 0.271 ~ 0.26 b-O.OM b-0.078 ~ 0.03 ~ 0.33 ~ 0.096 b-1.75 b-0.054 b-504l2853.3 =A504/1000 0.24 50512819.58 =ASOS/1000 0.133 SOEl2804.1 =A506/1000 0.04 50712793.75 =PS07/1000 0.68 ~

     '789.2     ::ASOS/1000   0.052  b-0.76
         .
     '782.11    =AS09/1000 srains.1         =ASl0/1000    0.03
       ,,.,

Slll?!7S0.3 :oA511/1000

                =AS12/1000 0.041 0.066
                =AS13/1000    0.125 42.3    =A.514/1000   O.Q28 21.9    -ASlS/1000    0.036 511>12703.2      =AS16/1000    0.034 51112659.1       =f.S17/1000   0.086  b-OAl    b-0.14 52   26228      l::A520/1000  1Mll 52   2597.92 52   2555.3 52   2549.9 52   2545.4      =AS24/1000    0.05 52   2534.9      =A.5251000    o."'

52 2522 =AS26/1000 0.05 52 2510,8 =A527/1000 0.24 b 52 2503 =A5213/1000 o.os b-52 2487.8 =AS29/1000 0.024 ~ 5 2467.3 =AS30/1000 0.016 53 2440,9 =AS3l/1000 0.046 53 2400.99 =AS32/1000 0.72 53 2377A ooA533/1000 0.8 5 2352.7 =/,;;34/1000 0.18 5 2335.2 =A.535/1000 0.1 5 1330 =A536/1000 0.036 53 2321.7 =A537/1000 0.052 5 218S.6 "'A.538/1000 0.046 53 2280.2 =AS39/1000 0.21 54 2249 =AS40/1000 0.08 5412239.8 ::AS4l/1000 0.05 0.024 o.046

ATTACHMENT 4 0.12 0.026 0.042

                      "'     b-
                             .

b* 0.062 b-0.94 b* b-0.245 157 0.119 0.038 0.197 0.08 0.12 b-0.086

                             *
                      """

0.03 b-0.046 b* 0.107 0.76 0.056 om b-

...

0.26 b* 0.141

""'

0.82

b* 0.072 0.09 0.191 b-b-

           ;A603/1000 0-'

5041533.68 =A604/1000 5.1 b- "' 1530.04 "A605/1000

           =A606/1000
           =A607/1000 b-b-

b*

           =A608/1000 o....
           =A609/1000 0.044
           =A610/1000
           =A611/1000
           =A612/1000
           =A613/1000         b-
           =A614/1000         b*
                             .**
           =A615/1000  0.052
           =A516/1000  0.02   b-6171421.64  =A617/1000  0.225
           =A618/1000  0.265
           =A619/1000  o.oss
           =A620/1000
           =A621/1000

ATTACHMENT4

.,

1335.4

62 1324.28 62 1302.7 62 0."'4

""

62 1278.5 6291273.73 0.024 0038 631241.5 1235.62

"'    1228.8 0.022 ~
 ""'1210.2 631195.1 0.018 0.,..
                      ~

0.185 0.167 0.98 b-1162.S 0.215 ~

   "  1152.2    O.OM  ~
   "  1131.51         ~
                      ~
""    1119.6 1116.61         b-b-

0.9

   "  1098.1    0-'M
 "" ""

1076.48 1067.7 b-

0.03 0.052 1044.4

   "
  "   1038.3 0.66 0.98 b-
                      ~

53.18 0.107 b-

   '"
  "" ,,...'9 934.6 0.165 0.066 0.038
                      ~

30.95 b-

   '                  b-0.8 OAS 67 890A 67 887.9           b-
                      ~

67 870A2 0.161 snss1.os S.9 67 844.7 ~ 835.53 b-68 783.5 0.022

                      ~
       ,,.

762.9 o.*

                      ~
                      ~

535 '0.092 ~

   "            0.11 4.22 03 026

"'" " 10.05 696.24 1.79 674.11

ATTACHMENT 4 Attachment 2 A B c D E F G H I 1 Kr-89 2 Tl/2 3.15m 3 4 Determination of DAC limited by EDE 5 Photon Emissions: 6 7 Energy Energy Intensity Decay lnterpelated Gamma Dose Factor Dose Rate per Unit Activity 8 (keV) (MeV) (%) Mode hE (Gy per Bq s m" 3 ) Gy/s 9 10 4701.5 4.7015 O.ol b- 2.78483E-13 2.78483E-17 11 4685.6 4.6856 0.008 b- 2.7739E-13 2.21912E-17 12 4655.6 4.6556 0.01 b- 2.75329E-13 2.75329E-17 13 4631.5 4.6315 0.028 b- 2.73677E-13 7.66294E-17 14 4489.2 4.4892 0.135 b- 2.63966E-13 3.56354E-16 15 4478.3 4.4783 0.014 b- 2.63225E-13 3.68515E-17 16 4448.1 4.4481 0.01 b- 2.61176E-13 2.61176E-17 17 4405.1 4.4051 0.008 b- 2.58265E-13 2.06612E-17 18 4368.4 4.3684 0.042 b- 2.55786E-13 1.0743E-16 19 4341.1 4.3411 0.105 b- 2.53945E-13 2.66643E-16 20 4321.2 4.3212 0.01 b- 2.52606E-13 2.52606E-17 21 4307.4 4.3074 0.01 b- 2.51677E-13 2.51677E-17 22 4279.4 4.2794 0.02 b- 2.49797E-13 4.99593E-17 23 4267.7 4.2677 0.028 b- 2.49012E-13 6.97232E-17 24 4253.3 4.2533 0.014 b- 2.48046E-13 3.47265E-17 25 4184.3 4.1843 0.05 b- 2.43432E-13 1.21716E-16 26 4176.2 4.1762 0.012 b- 2.42891E-13 2.9147E-17 27 4162.6 4.1626 0.028 b- 2.41984E-13 6.77557E-17 28 4146.9 4.1469 0.016 b- 2.40939E-13 3.85502E-17 29 4143 4.143 0.026 b- 2.40679E-13 6.25765E-17 30 4117.7 4.1177 0.014 b- 2.38996E-13 3.34594E-17 31 4081.4 4.0814 0.074 b- 2.36586E-13 1.75073E-16 32 4048 4.048 0.117 b- 2.34373E-13 2.74216E-16 33 4043.8 4.0438 0.02 b- 2.34095E-13 4.68189E-17 34 4004.9 4.0049 0.028 b- 2.31523E-13 6.48266E-17 35 3996 3.996 0.143 b- 2.30936E-13 3.30239E-16 36 3977.5 3.9775 0.27 b- 2.29716E-13 6.20233E-16 37 3977.5 3.9775 0.07 b- 2.29716E-13 1.60801E-16 38 3965.5 3.9655 0.209 b- 2.28925E-13 4.78454E-16 39 3923 3.923 0.42 b- 2.2613E-13 9.49745E-16 40 3901.7 3.9017 0.135 b- 2.24732E-13 3,03388E-16 41 3898.4 3.8984 0.034 b- 2.24515E-13 7.63351E-17 42 43 3882.5 3842.7 3.8825 3.8427

                                   -    0.04 0.111 b-b-

2.23473E-13 2.20868E-13 8.9389E-17 2.45163E-16 44 3837.6 3.8376 0.082 b- 2.20534E-13 1.80838E-16 45 3827.4 3.8274 0.139 b- 2.19868E-13 3.05616E-16

ATTACHMENT 5 Containment Building Leakage Rate (scf/min) = 17.68 scf x (Containment Presssure - 14.7 psia) 112 Containment Containment Containment Containment Time Pressure LR Ave . LR Ave . LR Time Pressure LR Ave. LR Ave . LR Time Pressure LR Ave . LR Ave . LR Time Pressure LR Ave . LR Ave . LR (min) (psia) (scfm) Vol. scf (scf/m) (scf/h) (min) (psia) (scfm} Vol. scf (scf/m) (scf/h} (min) (psia ) (scfm) Vol. scf (scf/m) (scf/h) (min) (psia) (scfm) Vol. scf (scf/m) (scf/h) 0 15.0333 10.2075 230102.0 305 14.8605 7.0838 227457.2 620 14.74746 3.8518 225726.5 914 14.70211 0.8125 225032.33 5 15.0300 10.1563 230051.0 310 14.8582 7.0326 227421 .7 625 14.74621 3.8004 225707.2 916 14.70201 0.7918 225030.70 10 15.0267 10.1051 230000.2 315 14.8559 6.9813 227386.6 630 14.74496 3.7490 225688.2 918 14.70190 0.7712 225029.12 15 15.0234 10.0539 229949.6 320 14.8536 6.9301 227351 .7 635 14.74374 3.6976 225669.5 9th thru 12th Hour 920 14.70180 0.7505 225027 58 20 15.0201 10.0028 229899.4 325 14.8514 6.8788 227317.0 640 14.74253 3.6462 225651 .0 3.466 207.9 922 14.70170 0.7297 225026.08 25 15.0168 9.9516 229849.4 1st Hour 330 14.8491 6.8276 227282.6 645 14.74134 3.5947 225632.8 (scf/m) (scf/h) 924 14.70161 0.7090 225024.62 30 15.0136 9.9004 229799.6 9.926 595.6 335 14.8469 6.7764 227248.5 650 14.74017 3.5433 225614.8 926 14.70152 0.6883 225023.20 13th thru 15.Sth Hour 35 15.0103 9.8492 229750.1 (scf/m) (scf/h) 340 14.8447 6.7251 227214.6 655 14.73901 3.4919 225597.1 928 14.70143 0.6676 225021 .82 1.132 67.9 40 15.0071 9.7980 229700.9 345 14.8425 6.6739 227181 .0 660 14.73787 3.4405 225579.6 930 14.70134 0.6468 225020.49 (scf/m) (scf/h) 45 15.0039 9.7468 229651 .9 350 14.8403 6.6226 227147.6 665 14.73674 3.3890 225562.4 932 14.70125 0.6261 225019.19 50 15.0007 9.6956 229603.1 355 14.8381 6.5713 227114.5 670 14.73564 3.3376 225545.5 934 14.70117 0.6053 225017.94 55 14.9976 9.6444 229554.7 360 14.8360 6.5201 227081.7 675 14.73455 3.2861 225528.8 936 14.70109 0.5845 225016.73 60 14.9944 9.5933 229506.4 365 14.8339 6.4688 227049.1 680 14.73347 3.2347 225512.3 938 14.70102 0.5637 225015.56 65 14.9913 9.5421 229458.5 370 14.8318 6.4176 227016.7 5th thru 8th Hour 685 14.73242 3.1832 225496.2 940 14.70094 0.5429 225014.43 70 14.9882 9.4909 229410.8 375 14.8297 6.3663 226984.6 5.930 355.8 690 14.73138 3.1317 225480.3 942 14.70087 0.5221 225013.35 75 14.9851 9.4397 229363.3 380 14.8276 6.3151 226952.8 (scf/m) (scf/h} 695 14.73035 3.0803 225464.6 944 14.70080 0.5013 225012.30 80 14.9820 9.3885 229316.1 385 14.8255 6.2638 226921 .2 700 14.72935 3.0288 225449.2 946 14.70074 0.4804 225011 .30 85 14.9789 9.3373 229269.2 2nd Hour 390 14.8235 6.2125 226889.9 705 14.72836 2.9773 225434.0 948 14.70068 0.4595 225010.34 90 14.9759 9.2861 229222.5 9.312 558.7 395 14.8214 6.1613 226858.8 710 14.72 739 2.9258 225419.2 950 14.70062 0.4386 225009.42 95 14.9728 9.2349 229176.0 (scf/m) (scf/h) 400 14.8194 6.1100 226828.0 715 14.72643 2.8743 225404.5 951 14.70059 0.4283 225008.98 100 14.9698 9.1837 229129 .9 405 14.8174 6.0587 226797.5 720 14.72549 2.8228 225390.2 952 14.70056 0.4180 225008.55 105 14.9668 9.1325 229083.9 410 14.8155 6.0074 226767.2 725 14.72457 2.77 12 225376.0 953 14.70053 0.4076 225008.14 110 14.9638 9.0813 229038.3 415 14.8135 5.9562 226737.1 730 14.72366 2.7197 225362.2 954 14.70050 0.3973 225007.73 115 14.9609 9.0301 228992 .9 420 14.8115 5.9049 226707.4 735 14.72277 2.6681 225348.6 955 14.70048 0.3869 225007.33 120 14.9579 8.9789 228947.7 425 14.8096 5.8536 226677.8 740 14.72190 2.6166 225335.3 956 14.70045 0.3766 225006.94 125 14.9550 8.9277 228902 .8 430 14.8077 5.8023 226648.6 745 14.72105 2.5650 225322.2 957 14.70043 0.3662 225006.57 130 14.9521 8.8765 228858.2 435 14.8058 5.7511 226619.6 750 14.72021 2.5135 225309.3 958 14.70041 0.3559 225006.20 135 14.9492 8.8253 228813.8 440 14.8039 5.6998 226590.8 755 14.71939 2.4619 225296.8 959 14.70038 0.3455 225005.85 140 14.9463 8.7741 228769.7 445 14.8021 5.6485 226562.3 760 14.71859 2.4103 225284.5 960 14.70036 0.3352 225005.50 145 14.9434 8.7229 228725.8 3rd Hour 450 14.8002 5.5972 226534.1 765 14.71780 2.3587 225272.4 961 14.70034 0.3248 225005.17 150 14.9406 8.6717 228682.2 8.697 521.8 455 14.7984 5.5459 226506 .1 770 14.71 703 2.3070 225260.6 962 14.70032 0.3144 225004.84 155 14.9377 8.6205 228638.8 (scf/m) (scf/h) 460 14.7966 5.4946 226478.3 775 14.71627 2.2554 225249.1 963 14.70030 0.3040 225004.53 160 14.9349 8.5693 228595.7 465 14.7948 5.4433 226450.9 780 14.71554 2.2038 225237.8 964 14.70028 0.2936 225004.22 165 14.9321 8.5181 228552 .9 470 14.7930 5.3920 226423.7 785 14.71482 2.1521 225226.8 965 14.70026 0.2832 225003.93 170 14.9293 8.4669 228510.3 475 14.7913 5.3407 226396.7 790 14.71411 2.1004 225216 .0 966 14.70024 0.2728 225003.65 175 14.9266 8.4156 228468.0 480 14.7895 5.2894 226370.0 795 14.71343 2.0487 225205.5 967 14.70022 0.2624 225003.37 180 14.9238 8.3644 228425.9 485 14.7878 5.2381 226343.5 800 14.71276 1.9970 225195.3 968 14 .70020 0.2520 225003.11 185 14.9211 8.3132 228384.1 490 14.7861 5.1868 226317.4 805 14.71211 1.9453 225185.3 969 14.70019 0.2416 225002.86 190 14.9184 8.2620 228342 .5 495 14.7844 5.1355 226291.4 810 14.71 147 1.8936 225175.6 970 14.70017 0.2312 225002.62 195 14.9157 8.2108 228301.2 500 14.7827 5.0842 226265.7 815 14.71085 1.8418 225166 .1 971 14.70016 0.2207 225002.39 200 14.9130 8.1596 228260.1 505 14.7810 5.0329 226240.3 820 14.71025 1.7900 225156 .9 972 14.70014 0.2102 225002.16 205 14.9103 8.1084 228219.3 4th Hour 510 14.77939 4.9816 226215.2 825 14.70967 1.7382 225147.9 973 14.70013 0.1998 225001.95 210 14.9077 8.0571 228178.8 8.083 485.0 515 14.77776 4.9302 226190.2 830 14.70910 1.6864 225139.3 974 14.70011 0.1893 225001 .75 215 14.9050 8.0059 228138.5 (scf/m) (scf/h} 520 14.77615 4.8789 226165.6 835 14.70855 1.6345 225130.8 975 14.70010 0.1788 225001 .57 220 14.9024 7.9547 228098.5 525 14.77456 4.8276 226141 .2 840 14.70801 1.5827 225122.7 976 14.70009 0.1683 225001 .39 225 14.8998 7.9035 228058.7 530 14.77298 4.7763 226117.1 845 14.70750 1.5308 225114.7 977 14.70008 0.1577 225001 .22 230 14.8973 7.8523 228019 .2 535 14.77142 4.7249 226093.2 850 14.70700 1.4788 225107.1 978 14.70007 0.1472 225001.06 235 14.8947 7.8010 227979.9 540 14.76988 4.6736 226069.6 855 14.70651 1.4268 225099.7 979 14.70006 0.1366 225000.91 240 14.8921 7.7498 227940.9 545 14.76835 4.6223 226046.2 860 14.70605 1.3748 225092.6 980 14.70005 0.1259 225000.78 245 14.8896 7.6986 227902.2 550 14.76684 4.5709 226023.1 865 14.70560 1.3228 225085.7 981 14.70004 0.1153 225000.65 250 14.8871 7.6474 227863.7 555 14.76535 4.5196 226000.2 870 14.70517 1.2707 225079.1 982 14.70003 0.1046 225000.54 255 14.8846 7.5961 227825.4 560 14.76387 4.4682 225977.6 875 14.70475 1.2186 225072.7 983 14.700028 0.0938 225000.43 260 14.8821 7.5449 227787.5 565 14.76241 4.4169 225955.3 880 14.70435 1.1664 225066.6 984 14.700022 0.0830 225000.34 265 14.8796 7.4937 227749.7 5th Hour 570 14.76097 4.3655 225933.2 885 14.70397 1.1142 225060.8 985 14.700017 0.0720 225000.25 270 14.8772 7.4425 227712 .3 7.468 448.1 575 14.75954 4.3142 225911.4 890 14.70361 1.0619 225055.2 986 14.700012 0.0610 225000.18 275 14.8748 7.3912 227675.1 (scf/m) (scf/h) 580 14.75813 4.2628 225889.8 895 14.70326 1.0096 225049.9 987 14.700008 0.0497 225000.12 280 14.8724 7.3400 227638.1 585 14.75674 4.2115 225868.5 900 14.70293 0.9572 225044 .9 988 14.700005 0.0382 225000.07 285 14.8700 7.2888 227601.4 590 14.75537 4.1601 225847.4 902 14.70281 0.9365 225042.95 989 14.700002 0.0260 225000.03 290 14.8676 7.2375 227565.0 595 14.75401 4.1087 225826 .6 904 14.70268 0.9159 225041 .07 16.5 hr = 990 14.700000 0.0121 225000.01 295 14.8652 7.1863 227528.8 600 14.75266 4.0573 225806.1 906 14.70256 0.8952 225039.24 300 14.8629 7.1351 227492.8 605 14.75134 4.0060 225785.8 908 14.70245 0.8746 225037.45 Overall Average Leak Rate = 5.154 scf/min 610 14.75003 3.9546 225765.8 910 14.70233 0.8539 225035.70 615 14.74874 3.9032 225746.0 912 14.70222 0.8332 225033.99

Case Summary of Containment Shine ATTACHMENT 6 Page 1 of 3 MicroShicld 8.02 Nathan Hogue (8.00-0000) Date By Checked Filename Run Date Run Time Duration Fuel Element Failure.msd March 29, 2016 10:27:53 AM 00:00:00 Project Info Case Title Containment Shine Description Fuel Element Failure Accident Analysis Geometry 13 - Rectangular Volume Source Dimensions Length l .9e+3 cm (60 ft 9.9 in) Width l.9e+3 cm (60 ft 9.9 in) Height l.8e+3 cm (60 ft 0.1 in) Dose Points x y z

A 914.0 cm (29 ft 11.8 914.0 cm (29 ft 11.8 -X

#1 l.9e+3 cm (62 ft 9.5 in) in)                     in) l .5e+4 cm ( 492 ft 1.5   914.0 cm (29 ft 11.8       914.0 cm (29 ft 11 .8 W2                                         in) in)                                               in)

Shields Shield N Dimension Material Density 3 Source 6.29e+09 cm Air 0.00122 Shield I 30.0 cm Concrete 2.35 Air Gap Air 0.00122 Source Input: Grouping Method - Standard Indices Number of Groups: 25 Lower Energy Cutoff: 0.015 Photons < 0.015: Included Library: Grove Nuclide Ci Bq µ.Ci/cm 3 Bq/cm 3 1-131 2.9700e-001 l.0989e+Ol0 4.7241 e-005 1.7479e+OOO I-132 4.l600e-001 l.5392e+Ol0 6.6170e-005 2.4483e+OOO 1-133 7.3200e-001 2. 7084e+Ol 0 l . l 643e-004 4.3080e+OOO 1-134 8.2500e-OO l 3.0525e+010 l .3 l 23e-004 4.8554e+OOO 1-135 6.8400e-001 2.5308e+010 1.0880e-004 4.0255e+OOO Kr-85 6.2500e-004 2.3 l 25e+007 9.9414e-008 3.6783e-003 Kr-85m 1. 7700e-00 l 6.5490e+009 2.8 l 54e-005 l.0417e+OOO Kr-87 2.7700e-001 l .0249e+Ol0 4.4060e-005 l .6302e+OOO Kr-88 3.9300e-001 1.454le+Ol0 6.2511 e-005 2.3 l 29e+OOO Kr-89 4.9800e-001 l.8426e+Ol0 7.9213e-005 2.9309e+OOO Kr-90 4.9700e-001 l.8389e+ol0 7.9054e-005 2.9250e+OOO file:/I /C:/Prograrn%20Files%20(x86)/MicroShield%208/Examples/CaseFiles/HTML/Fuel... 3/29/2016

Case Summary of Containment Shine ATTACHMENT 6 Page 2of3 Xe-133 5.2000e-OO1 1.9240e+010 8.2712e-005 3.0604e+OOO Xe-135 l .0200e-001 3.77 40e+009 l .6224e-005 6.0030e-001 Xe-135m 4 .8900e-002 1.8093e+009 7.778le-006 2.8779e-001 Xe-137 6.5000e-001 2.4050e+010 l .0339e-004 3.8254e+OOO Xe-138 6.7700e-OO 1 2.5049e+o10 l .0769e-004 3.9843e+OOO Buildup: The material reference is Shield 1 Integration Parameters X Direction 10 Y Direction 20 Z Direction 20 Results - Dose Point# 1 - (1914,914,914 cm Fluence Rate Fluence Rate Exposure Rate Exposure Rate Energy (MeV) Activity (Photons/sec) MeV/cm 2/sec MeV/cm 2/sec mR/hr mR/hr No Buildup With Buildup No Buildup With Buildup 0.015 5.920e+09 3.327e-250 7.710e-26 2.853e-251 6.613e-27 0.03 l.223e+ 10 5.l 12e-36 5.542e-25 5.066e-38 5.493e-27 0.08 7.349e+09 3.477e-06 8.553e-05 5.502e-09 1.353e-07 0.1 6.506e+09 3.682e-05 1.354e-03 5.633e-08 2.071e-06 0.15 8.408e+09 7.362e-04 3.61 le-02 l.212e-06 5.947e-05 0.2 l.611e+10 5.921e-03 2.820e-01 l.045e-05 4.978e-04 0.3 l.179e+10 2.601e-02 9.294e-Ol 4.933e-05 l .763e-03 0.4 2.973e+10 2.168e-01 5.714e+OO 4.224e-04 1.113e-02 0.5 5.015e+ 10 8.956e-01 l.813e+Ol l.758e-03 3.559e-02 0.6 3.807e+ 10 l.383e+OO 2.239e+Ol 2.700e-03 4.370e-02 0.8 7.924e+ 10 8.512e+OO 9.745e+Ol l.619e-02 1.854e-01 1.0 4.160e+ 10 l.003e+Ol 8.849e+Ol 1.849e-02 l.63 le-01 1.5 3.319e+ 10 3.190e+Ol 1.822e+02 5.368e-02 3.065e-01 2.0 3.012e+ 10 7.046e+Ol 3.103e+02 1.090e-01 4.799e-Ol 3.0 3.439e+09 2.449e+Ol 7.862e+Ol 3.323e-02 1.067e-Ol 4.0 8.464e+08 1.205e+Ol 3. I 91e+Ol 1.491e-02 3.948e-02 Totals 3.747e+11 1.600e+o2 8.364e+02 2.504e-01 1.374e+OO Results - Dose Point# 2 - 1 15000,914,914) cm Fluence Rate Fluence Rate Exposure Rate Exposure Rate Energy (Me V) Activity (Photons/sec) MeV/cm 2/sec MeV/cm 2/sec mR/hr mR/hr No Buildup With Buildup No Buildup With Buildup 0.015 5.920e+09 7.298e-261 3.473e-28 6.260e-262 2.979e-29 0.03 1.223e+ 10 5.629e-39 2.496e-27 5.579e-41 2.474e-29 0.08 7.349e+09 1.101 e-08 3.055e-07 l.742e-11 4.834e-10 0.1 6.506e+09 1.203e-07 5.189e-06 l .840e-10 7.938e-09 0.15 8.408e+09 2.675e-06 I .599e-04 4.405e-09 2.633e-07 0.2 1.61 le+ IO 2.367e-05 1.380e-03 4.178e-08 2.436e-06 0.3 l.179e+ 10 1.196e-04 5.123e-03 2.268e-07 9.718e-06 0.4 2.973e+ 10 1.092e-03 3.371e-02 2.127e-06 6.569e-05 0.5 5.015e+10 4.814e-03 1.ll 9e-01 9.449e-06 2.196e-04 file :///C:/Program%20Files%20(x86)/MicroShield%208/Examples/CaseFiles/HTML/Fuel... 3/29/2016

ATTACHMENT 6 Case Summary of Containment Shine Page 3of3 0.6 3.807e+ 10 7.802e-03 I .426e-01 I .523e-05 2.784e-04 0.8 7.924e+ 10 5.127e-02 6.446e-Ol 9.752e-05 l.226e-03 1.0 4.l 60e+ 10 6.308e-02 5.999e-01 1.163e-04 l.106e-03 1.5 3.3l9e+10 2.126e-01 l.270e+OO 3.577e-04 2.137e-03 2.0 3.012e+IO 4.826e-Ol 2.183e+OO 7.463e-04 3.376e-03 3.0 3.439e+09 l.718e-OI 5.540e-01 2.330e-04 7.516e-04 4.0 8.464e+08 8.521e-02 2.243e-01 l.054e-04 2.775e-04 Totals 3.747e+ll 1.080e+OO 5.770e+OO 1.683e-03 9.449e-03 file:///C:/Program%20Files%20( x86)/MicroShield%208/Examples/CaseFiles/HTML/Fuel... 312912016

ATTACHMENT 7 Fuel Handling Accident - Resricted Area Dose Core Activity Pool Containment Containment Containment Containment Containment Dose to 30 min Released Released Activity Concentration Concentration Concentration Concentration Concentration Average Workers Isotope T 112 Decay Activity Activity Con. No I evap@T0 w/decay w/ decay w/ decay w/ decay Concentration DAC 5min Ci Ci uCi uCi/ al uCi/cc uCi/cc uCi/cc uCi/cc uCi/cc uCi/cc mrem Time min: 0.5 1.5 2.5 3.5 4.5 1-131 8.02 d 2.20E+05 2.07E+01 2.07E+07 1.03E+03 3.25E-07 9.74E-07 1.62E-06 2.27E-06 2.92E-06 1.62E-06 2.00E-08 1.69E+02 1-132 2.28 h 3.07E+05 2.89E+01 2.89E+07 1.44E+03 4.52E-07 1.35E-06 2.24E-06 3.11 E-06 3.98E-06 2.23E-06 3.00E-06 1.55E+OO 1-133 20.8 h 5.39E+05 5.07E+01 5.07E+07 2.53E+03 7.95E-07 2.38E-06 3.97E-06 5.56E-06 7.14E-06 3.97E-06 1.00E-07 8.26E+01 1-134 52 .6 m 5.49E+05 5.16E+01 5.16E+07 2.58E+03 8.05E-07 2.38E-06 3.92E-06 5.41 E-06 6.87E-06 3.88E-06 2.00E-05 4.04E-01 1-135 6.57 h 4.80E+05 4.51 E+01 4.51E+07 2.26E+03 7.07E-07 2.12E-06 3.52E-06 4.93E-06 6.32E-06 3.52E-06 7.00E-07 1.05E+01 1.97E+02 1.97E+08 264.00 Kr-85 10.76 y 4.63E+02 4.35E-02 4.35E+04 6.83E-06 6.83E-06 6.83E-06 6.83E-06 6.83E-06 6.83E-06 1.00E-04 1.42E-02 Kr-85m 4.48 h 1.23E+05 1.16E+01 1.16E+07 1.81 E-03 1.81 E-03 1.80E-03 1.80E-03 1.79E-03 1.80E-03 2.00E-05 1.88E+01 Kr-87 1.27 h 1.58E+05 1.49E+01 1.49E+07 2.32E-03 2 .30E-03 2.28E-03 2.26E-03 2.24E-03 2.28E-03 5.00E-06 9.49E+01 Kr-88 2.84 h 2.58E+05 2.42E+01 2.42E+07 3.80E-03 3.78E-03 3.77E-03 3.75E-03 3.74E-03 3.77E-03 2.00E-06 3.92E+02 Kr-89 3.15 m 5.28E+02 4.96E-02 4.96E+04 6.98E-06 5.60E-06 4.49E-06 3.61 E-06 2.89E-06 4.71 E-06 t.90E-08 5.17E-01 Kr-90 32 .3 s 6.31E-12 5.93E-16 5.93E-10 4.89E-20 1.35E-20 3.73E-21 1.03E-21 2.84E-22 1.35E-20 2.80E-06 1.00E-14 5.08E+01 5.08E+07 506.51 Xe-133 5.243 d 3.85E+05 3.62E+01 3.62E+07 5.68E-03 5.68E-03 5.68E-03 5.68E-03 5.68E-03 5.68E-03 1.00E-04 1.18E+01 Xe-135 9.1 h 9.11 E+04 8.56E+OO 8.56E+06 1.34E-03 1.34E-03 1.34E-03 1.34E-03 1.34E-03 1.34E-03 1.00E-05 2.79E+01 Xe-135m 15.3 m 3.62E+04 3.40E+OO 3.40E+06 5.22E-04 4.99E-04 4.77E-04 4.56E-04 4.36E-04 4.78E-04 9.00E-06 1.11E+01 Xe-137 3.82 m 2.24E+03 2.11 E-01 2.11E+05 3.02E-05 2.52E-05 2.1OE-05 1.75E-05 1.46E-05 2.17E-05 2.00E-05 2.26E-01 Xe-138 14.1 m 1.16E+05 1.09E+01 1.09E+07 1.67E-03 1.59E-03 1.51 E-03 1.44E-03 1.37E-03 1.52E-03 4.00E-06 7.90E+01 Xe-139 39.7 s 7.89E-09 7.42E-13 7.42E-07 6.89E-17 2.42E-17 8.49E-18 2.98E-18 1.04E-18 2.11E-17 3.70E-06 1.19E-12 5.93E+01 5.93E+07 129.99 - Denotes DACs calculated using methodology as described in FGR No . 12. Total Iodine (COE) (mrem) 264.00 Iodine (CEDE) (mrem) (CDE x 0.03) 7.92 Noble Gas (CEDE) mrem 636.49 Total Dose (TEDE) mrem 644.41

ATTACHMENT 8 Fuel Handling Accident - Unresricted Area Dose Core Activity Pool Containment Containment Containment Containment Containment Containment Containment Containment Ave. 1st Hour Average Effluent Concentration Dose to 30 min Released Released Actvity Concentration Concentration Concentration Concentration Concentration Concentration Concentration Exhaust Exhaust Concentration at Maximum Public Isotope T112 Decay Actvity Activity wl decay wl decay wl decay wl decay wldecay wl decay wl decay Concentration Concentration Limit Dose Point 16.5 hr Ci Ci uCi uCils uCils uCils uCils uCils uCils uCils uCilcc uCilcc uCilcc ECl292 mrem 1 Hr Intervals: 0.5 1.5 2.5 3.5 4.5 4 Hr Intervals: 7 14.75 Containment Volume scf: 229801 229224 228684 228181 227714 226733 225605 225044 Containment Avera e Leaka e Rate scflhr: 595.6 558.7 521.8 485.0 448.1 355.8 207.9 67.9 1-131 8.02 d 2.20E+05 2.07E+01 2.07E+07 1.03E+03 8.98E-02 8.95E-02 8.92E-02 8.89E-02 8.85E-02 8.77E-02 8.65E-02 8.53E-02 5.16E-10 2.63E-10 2.00E-10 9.01E-13 4.24E-04 1-132 2.28 h 3.07E+05 2.89E+01 2.89E+07 1.44E+03 1.0BE-01 7.96E-02 5.87E-02 4.33E-02 3.20E-02 1.50E-02 4.43E-03 1.42E-03 6.20E-10 1.17E-10 2.00E-08 4.01E-13 1.89E-06 1-133 20.8 h 5.39E+05 5.07E+01 5.07E+07 2.53E+03 2.17E-01 2.10E-01 2.03E-01 1.96E-01 1.90E-01 1.75E-01 1.53E-01 1.35E-01 1.25E-09 5.50E-10 1.00E-09 1.88E-12 1.77E-04 1-134 52 .6 m 5.49E+05 5.16E+01 5.16E+07 2.58E+03 1.51 E-01 6.86E-02 3.11E-02 1.41 E-02 6.40E-03 8.87E-04 3.76E-05 1.94E-06 8.69E-10 9.10E-11 6.00E-08 3.12E-13 4.89E-07 1-135 6.57 h 4.80E+05 4.51E+01 4.51E+07 2.26E+03 1.86E-01 1.68E-01 1.51E-01 1.36E-01 1.22E-01 9.38E-02 6.15E-02 4.14E-02 1.07E-09 3.50E-10 6.00E-09 1.20E-12 1.88E-05 1.97E+02 1.97E+08 6.23E-04 Kr-85 10.76 y 4.63E+02 4.35E-02 4.35E+04 1.89E-01 1.89E-01 1.89E-01 1.89E-01 1.89E-01 1.89E-01 1.89E-01 1.89E-01 2.18E-09 1.13E-09 7.00E-07 3.87E-12 5.21E-07 Kr-85m 4.48 h 1.23E+05 1.16E+01 1.16E+07 4.66E+01 3.99E+01 3.42E+01 2.93E+01 2.51 E+01 1.70E+01 9.18E+OO 5.14E+OO 5.35E-07 1.48E-07 1.00E-07 5.06E-10 4.77E-04 Kr-87 1.27 h 1.58E+05 1.49E+01 1.49E+07 4.92E+01 2.85E+01 1.65E+01 9.57E+OO 5.55E+OO 1.42E+OO 1.60E-01 2.06E-02 5.65E-07 7.38E-08 2.00E-08 2.53E-10 1.19E-03 Kr-88 2.84 h 2.58E+05 2.42E+01 2.42E+07 9.34E+01 7.32E+01 5.73E+01 4.49E+01 3.52E+01 1.91 E+01 7.21E+OO 2.89E+OO 1.07E-06 2.31E-07 9.00E-09 7.93E-10 8.30E-03 Kr-89 3.15 m 5.28E+02 4.96E-02 4.96E+04 2.94E-04 5.44E-10 1.01E-15 1.86E-21 3.45E-27 1.61E-41 1.88E-64 5.99E-86 3.38E-12 2.05E-13 8.6 ;;oe 7.01E-16 7.68E-09 Kr-90 32 .3 s 6.31E-12 5.93E-16 5.93E-10 4.36E-32 1.25E-65 3.56E-99 1.02E-132 2.90E-166 4.00E-250 O.OOE+OO O.OOE+OO 5.01E-40 3.04E-41 .20E.()8 1.04E-43 8.17E-37 5.08E+01 5.08E+07 9.97E-03 Xe-133 5.243 d 3.85E+05 3.62E+01 3.62E+07 1.57E+02 1.56E+02 1.55E+02 1.54E+02 1.54E+02 1.52E+02 1.48E+02 1.45E+02 1.80E-06 9.11 E-07 5.00E-07 3.12E-09 5.88E-04 Xe-135 9.1 h 9.11E+04 8.56E+OO 8.56E+06 3.59E+01 3.32E+01 3.08E+01 2.85E+01 2.64E+01 2.19E+01 1.61 E+01 1.21 E+01 4.12E-07 1.51 E-07 7.00E-08 5.17E-10 6.96E-04 Xe-135m 15.3 m 3.62E+04 3.40E+OO 3.40E+06 3.80E+OO 2.51E-01 1.66E-02 1.10E-03 7.23E-05 8.10E-08 1.54E-12 5.78E-17 4.37E-08 2.83E-09 4.00E-08 9.68E-12 2.28E-05 Xe-137 3.82 m 2.24E+03 2.11E-01 2.11E+05 3.97E-03 7.43E-08 1.39E-12 2.61E-17 4.89E-22 7.44E-34 9.17E-53 1.72E-70 4.56E-11 2.76E-12 9.10!:09 9.46E-15 9.80E-09 Xe-138 14.1 m 1.16E+05 1.09E+01 1.09E+07 1.09E+01 5.69E-01 2.98E-02 1.56E-03 8.18E-05 5.14E-08 3.88E-13 6.11E-18 1.25E-07 7.96E-09 2.00E-08 2.72E-11 1.28E-04 Xe-139 39.7 s 7.89E-09 7.42E-13 7.42E-07 7.29E-26 3.73E-53 1.90E-80 9.73E-108 4.97E-135 2.93E-203 O.OOE+OO O.OOE+OO 8.38E-34 5.08E-35 1. 1.74E-37 1.02E-30 5.93E+01 5.93E+07 1.44E-03 Denotes Effluent Concentrations derived from DACs calculated using methodology as described in FGR No . 12. (DAC/219) Total Iodine (COE) (mrem) 6.23E-04 Noble Gas !TEDE} mrem 1.14E-02 Total Dose (TEDE) mrem 1.20E-02

Case Summary of Containment Shine ATTACHMENT 9 Page 1 of 3 MicroShield 8.02 Nathan Hogue (8.00-0000) Date By Checked Filename Run Date Run Time Duration Fuel Handling Accident.msd February 1, 2016 4:10:32 PM 00:00:00 Project Info Case Title Containment Shine Description Fuel Handling Accident Analysis Geometry 13 - Rectangular Volume Source Dimensions Length l.9e+3 cm (60 ft 9.9 in) Width l.9e+3 cm (60 ft 9.9 in) Height l.8e+3 cm (60 ft 0.1 in) Dose Points x y z __rx

A WI l.9e+3 cm (62 ft 9.5 in) 914.0 cmin)(29 ft 11.8 914.0 cm (29 ft 11.8 in) l .5e+4 cm (492 ft 1.5 914.0 cm (29 ft 11.8 914.0 cm (29ft11 .8

 #2 in)                       in)                     in)

Shields Shield N Dimension Material Density Source 6.29e+09 cm 3 Air 0.00122 Shield 1 30.0 cm Concrete 2.35 Air Gap Air 0.00122 Source Input: Grouping Method - Standard Indices Number of Groups: 25 Lower Energy Cutoff: 0.015 Photons < 0.015: Included Library: Grove Nuclide Ci Bq J.1Ci/cm 3 Bq/cm3 1-131 9.3300e+OOO 3.4521 e+Ol l l .4840e-003 5.4910e+OOI 1-132 2.5200e+001 9.3240e+Ol 1 4.0084e-003 1.483 l e+002 I-133 5.31 OOe+OOl 1.9647e+Ol2 8.4462e-003 3.125le+002 I-134 5.4500e+001 2.0165e+Ol2 8.6689e-003 3.2075e+002 I-135 4. 7700e+OO I l.7649e+012 7.5873e-003 2.8073e+002 Kr-85 2.3200e-003 8.5840e+007 3.6902e-007 1.3654e-002 Kr-85m 1.2lOOe+OO1 4.4770e+011 l .9247e-003 7.1212e+001 Kr-87 l.5700e+001 5.8090e+Ol 1 2.4973e-003 9.2399e+OOJ Kr-88 2.5700e+OOJ 9.5090e+Ol 1 4.0879e-003 l .5 l 25e+002 Kr-89 5.2500e-002 l.9425e+009 8.3508e-006 3.0898e-001 Kr-90 6.2600e-016 2.3162e-005 9.9573e-020 3.6842e-015 file ://IC :/Program%2 0Files%20(x86)/M icroShield%208/Examples/CaseFi Ies/HTML/Fuel%... 2/ I /2016

Case Summary of Containment Shine ATTACHMENT 9 Page 2of3 Xe-133 2.5700e+OO 1 9.5090e+Ol 1 4.0879e-003 l.5125e+002 Xe-135 I .0600e+OO 1 3.9220e+Ol 1 l.6861e-003 6.2384e+001 Xe-135m 4.5000e+OOO l .6650e+O 11 7.1578e-004 2.6484e+OOI Xe-137 2.2300e-001 8.251 Oe+009 3.5471e-005 l .3124e+OOO Xe-138 1.1500e+OO I 4.2550e+Ol 1 1.8292e-003 6.7681e+001 Buildup: The material reference is Shield 1 Integration Parameters X Direction 10 Y Direction 20 Z Direction 20 Results - Dose Point# 1 - (1914,914,914 cm Fluence Rate Fluence Rate Exposure Rate Exposure Rate Energy (Me V) Activity (Photons/sec) MeV/cm 2/sec Me V/cm 2/sec mR/hr mR/br No Buildup With Buildup No Buildup With Buildup 0.015 2.085e+l l 1.172e-248 2.715e-24 1.005e-249 2.329e-25 0.03 6.009e+l l 2.512e-34 2.723e-23 2.489e-36 2.699e-25 0.08 3.580e+l l l.694e-04 4.167e-03 2.680e-07 6.594e-06 0.1 l.875e+09 l.061e-05 3.902e-04 l.624e-08 5.970e-07 0.15 4.935e+l 1 4.322e-02 2.120e+OO 7.117e-05 3.491e-03 0.2 7.142e+l l 2.626e-01 1.251e+Ol 4.634e-04 2.207e-02 0.3 3.346e+ 11 7.380e-01 2.638e+Ol l.400e-03 5.003e-02 0.4 l.102e+12 8.03 le+oO 2.l l 7e+02 1.565e-02 4.125e-01 0.5 2.512e+l2 4.485e+Ol 9.082e+02 8.804e-02 1.783e+OO 0.6 1.965e+12 7.139e+Ol l.156e+o3 1.393e-01 2.255e+OO 0.8 4.917e+12 5.282e+02 6.047e+03 1.005e+OO 1.150e+Ol 1.0 l.942e+ l2 4.683e+02 4.130e+03 8.632e-01 7.613e+OO 1.5 1.613e+12 l.551e+03 8.853e+03 2.609e+OO l.490e+Ol 2.0 1.160e+ 12 2.713e+03 l.195e+04 4.196e+OO l.848e+01 3.0 8.875e+10 6.320e+02 2.029e+o3 8.575e-Ol 2.753e+OO 4.0 8.699e+07 1.238e+OO 3.280e+OO l.532e-03 4.057e-03 Totals 1.801e+13 6.019e+03 3.533e+04 9.776e+o0 5.977e+o1 Results - Dose Point # 2 - 15000.914,914 cm Fluence Rate Fluence Rate Exposure Rate Exposure Rate Energy (Me V) Activity (Photons/sec) MeV/cm 2/sec MeV/cm 2/sec mR/br mR/br No Buildup With Buildup No Buildup With Buildup 0.015 2.085e+l l 2.570e-259 l .223e-26 2.205e-260 I .049e-27 0.03 6.009e+l 1 2.766e-37 1.227e-25 2.74le-39 l.216e-27 0.08 3.580e+l l 5.363e-07 1.488e-05 8.488e-10 2.355e-08 0.1 1.875e+09 3.468e-08 l .496e-06 5.305e-l 1 2.288e-09 0.15 4.935e+ 11 1.570e-04 9.385e-03 2.586e-07 l.545e-05 0.2 7.142e+l l l .050e-03 6.119e-02 l .853e-06 l.080e-04 0.3 3.346e+ 11 3.393e-03 l .454e-Ol 6.437e-06 2.758e-04 0.4 I.I 02e+ l2 4.045e-02 l.249e+OO 7.882e-05 2.434e-03 0.5 2.512e+12 2.41 Ie-Ol 5.604e+OO 4.733e-04 l.lOOe-02 file:///C:/Program%20Files%20(x86)/MicroShield%208/Examples/CaseF iles/HTM L/Fuel%... 2/ I /2016

Case Summary of Containment Shine ATTACHMENT 9 Page 3of3 0.6 l.965e+ l2 4.027e-01 7.362e+OO 7.860e-04 l.437e-02 0.8 4.917e+ l2 3.181e+OO 4.000e+Ol 6.05le-03 7.608e-02 1.0 l.942e+12 2.944e+OO 2.800e+Ol 5.427e-03 5.161e-02 1.5 l.613e+ 12 1.033e+Ol 6.172e+Ol l .739e-02 1.038e-Ol 2.0 l.160e+ 12 1.858e+Ol 8.405e+Ol 2.874e-02 l .300e-Ol 3.0 8.875e+ 10 4.433e+OO I.430e+Ol 6.014e-03 1.940e-02 4.0 8.699e+07 8.757e-03 2.305e-02 1.083e-05 2.852e-05 Totals 1.801e+13 4.017e+Ol 2.425e+o2 6.497e-02 4.091e-Ol file :///C:/Program%2 OF i les%20(x86)/M icroShield%208/Examples/CaseFi !es/HTML/Fuel%... 2/1/2016

ATTACHMENT 10 Fueled Experiment Failure - Restricted Dose Act w/Decay 17 Second Containment Containment Containment Containment Dose to Pool 20 gal Bubble Containment Concentration Concentration Concentration Concentration Average Workers Isotope T112 Activity Activity Actvity Evap Assent Cone. w/ decay w/ decay w/ decay w/ decay Concentration DAC 5min Ci uCi uCi/ al uCi uCi/cc uCi/cc uCi/cc uCi/cc uCi/cc uCi/cc uCi/cc mrem Time min: 0.283333 0.5 1.5 2.5 3.5 4.5 1-131 8.02 d 6.755 6.76E+06 3.38E+02 6.76E+03 1.06E-07 3.18E-07 5.30E-07 7.42E-07 9.54E-07 5.30E-07 2.00E-08 5.52E+01 1-132 2.28 h 18.635 1.86E+07 9.32E+02 1.86E+04 2.91 E-07 8.69E-07 1.44E-06 2.01 E-06 2.57E-06 1.44E-06 3.00E-06 9.97E-01 1-133 20.8 h 39.875 3.99E+07 1.99E+03 3.99E+04 6.26E-07 1.88E-06 3.12E-06 4.37E-06 5.62E-06 3.12E-06 1.00E-07 6.50E+01 1-134 52.6 m 45.405 4.54E+07 2.27E+03 4.54E+04 7.05E-07 2.09E-06 3.43E-06 4.75E-06 6.02E-06 3.40E-06 2.00E-05 3.54E-01 1-135 6.57 h 37.695 3.77E+07 1.88E+03 3.77E+04 5.92E-07 1.77E-06 2.96E-06 4.14E-06 5.32E-06 2.96E-06 7.00E-07 148.365 Kr-85 10.76 y 0.0020 2.00E+03 3.14E-07 3.14E-07 3.14E-07 3.14E-07 3.14E-07 3.14E-07 3.14E-07 1.00E-04 6.54E-04 Kr-85m 4.48 h 7.580 7.58E+06 1.19E-03 1.19E-03 1.18E-03 1.18E-03 1.18E-03 1.18E-03 1.18E-03 2.00E-05 1.23E+01 Kr-87 1.27 h 15.405 1.54E+07 2.41 E-03 2.40E-03 2.38E-03 2.36E-03 2.34E-03 2.31 E-03 2.36E-03 5.00E-06 9.82E+01 Kr-88 2.84 h 21 .660 2.17E+07 3.40E-03 3.39E-03 3.38E-03 3.36E-03 3.35E-03 3.33E-03 3.36E-03 2.00E-06 3.50E+02 Kr-89 3.15 m 27.740 2.77E+07 4.09E-03 3.66E-03 2.94E-03 2.36E-03 1.89E-03 1.52E-03 2.48E-03 1.90E-06 2.71 E+02 Kr-90 32 .3 s 27.410 2.74E+07 2.99E-03 1.57E-03 4.33E-04 1.20E-04 3.30E-05 9.11E-06 4.33E-04 2.SOE-06 3.22E+01 99.797 7.64E+02 Xe-133 5.243 d 18.925 1.89E+07 2.97E-03 2.97E-03 2.97E-03 2.97E-03 2.97E-03 2.97E-03 2.97E-03 1.00E-04 6.18E+OO Xe-135 9.1 h 13.630 1.36E+07 2.14E-03 2.14E-03 2.13E-03 2.13E-03 2.13E-03 2.13E-03 2.13E-03 1.00E-05 4.44E+01 Xe-135m 15.3 m 6.760 6.76E+06 1.05E-03 1.02E-03 9.79E-04 9.35E-04 8.94E-04 8.54E-04 9.37E-04 9.00E-06 2.17E+01 Xe-137 3.82 m 35.800 3.58E+07 5.34E-03 4.87E-03 4.07E-03 3.39E-03 2.83E-03 2.36E-03 3.50E-03 2.00E-05 3.65E+01 Xe-138 14.1 m 37.380 3.74E+07 5.79E-03 5.65E-03 5.37E-03 5.12E-03 4.87E-03 4.64E-03 5.13E-03 4.00E-06 2.67E+02 Xe-139 39.7 s 30.675 3.07E+07 3.58E-03 2.12E-03 7.44E-04 2.61E-04 9.16E-05 3.21E-05 6.50E-04 3.70E-06 3.66E+01 143.170 4.12E+02 - Denotes DACs calculated with methodology as described in FGR No. 12 . Total Iodine (COE) (mrem) 130.36 Iodine (CEDE) (mrem) (CDE x 0.03) 3.91 Noble Gas (CEDE) (mrem) 1176.42 Whole Body (TEDE) (mrem) 1180.33

ATTACHMENT 11 Fueled Experiment Failure - Unrestricted Area Dose Actvity Containment Containment Containment Containment Containment Containment Containment Containment Exhaust Exhaust Effluent Concentration Dose to in Pool Concentration Concentration Concentration Concentration Concentration Concentration Concentration Concentration 1st hour Average Concentration at Maximum Public Isotope T112 Activity Actvity at 10 min (uCi/scf) w/decay w/decay w/decay wt decay wt decay w/decay wt decay Ave. Cone Concentration Limit Dose Point (16.5 hrs) (Ci) (uCi) (uCi/gal) No I evap@T 0 (uCi/scf) (uCi/scf) (uCi/scf) (uCi/scf) (uCi/scf) (uCi/scf) (uCi/scf) (uCi/cc) (uCi/cc) (uCi/cc) (EC/292) (mrem) 1 Hour Intervals: 0.5 1.5 2.5 3.5 4.5 4 hr intervals: 7 11 14.75 Containment Volume scf: 229800 229223 228682 228179 227712 226707 225580 225061 e Rate ft3/hr: 595.6 558.7 521.8 485 448.1 355.8 207.9 67.9 1-131 8.02 d 3.38E+02 2.94E-02 2.93E-02 2.92E-02 2.90E-02 2.89E-02 2.87E-02 2.83E-02 2.79E-02 1.69E-10 8.60E-11 2.00E-10 2.95E-13 1.39E-04 1-132 2.28 h 9.32E+02 6.97E-02 5.14E-02 3.79E-02 2.80E-02 2.07E-02 9.66E-03 2.86E-03 9.16E-04 4.00E-10 7.56E-11 2.00E-08 2.59E-13 1.22E-06 1-133 20.8 h 1.99E+03 1.70E-01 1.65E-01 1.59E-01 1.54E-01 1.49E-01 1.37E-01 1.20E-01 1.06E-01 9.79E-10 4.32E-10 1.00E-09 1.48E-12 1.39E-04 1-134 52.6 m 2.27E+03 1.33E-01 6.04E-02 2.74E-02 1.24E-02 5.63E-03 7.81 E-04 3.31 E-05 1.71E-06 7.65E-10 8.01 E-11 6.00E-08 2.74E-13 4.31E-07 1-135 6.57 h 1.88E+03 1.55E-01 1.40E-01 1.26E-01 1.13E-01 1.02E-01 7.82E-02 5.13E-02 3.45E-02 8.92E-10 2.92E-10 6.00E-09 9.99E-13 1.57E-05 2.95E-04 Kr-85 10.76 y 0.0020 2.00E+03 8.70E-03 8.70E-03 8.70E-03 8.70E-03 8.70E-03 8.70E-03 8.70E-03 8.70E-03 1.00E-10 5.19E-11 7.00E-07 1.78E-13 2.39E-08 Kr-85m 4.48 h 7.580 7.58E+06 3.05E+01 2.62E+01 2.24E+01 1.92E+01 1.64E+01 1.12E+01 6.02E+OO 3.37E+OO 3.51 E-07 9.69E-08 1.00E-07 3.32E-10 3.13E-04 Kr-87 1.27 h 15.405 1.54E+07 5.10E+01 2.96E+01 1.71E+01 9.93E+OO 5.75E+OO 1.47E+OO 1.66E-01 2.14E-02 5.87E-07 7.66E-08 2.00E-08 2.62E-10 1.24E-03 Kr-88 2.84 h 21 .660 2.17E+07 8.34E+01 6.54E+01 5.12E+01 4.01 E+01 3.14E+01 1.71E+01 6.44E+OO 2.58E+OO 9.59E-07 2.07E-07 9.00E-09 7.08E-10 7.41E-03 Kr-89 3.15 m 27 .740 2.77E+07 1.64E-01 3.04E-07 5.62E-13 1.04E-18 1.93E-24 8.97E-39 1.05E-61 3.35E-83 1.89E-09 1.14E-10 8.80E-o9 3.92E-13 4.29E-06 Kr-90 32 .3 s 27.410 2.74E+07 2.02E-15 5.76E-49 1.64E-82 4.70E-116 1.34E-149 1.85E-233 O.OOE+OO O.OOE+OO 2.32E-23 1.40E-24 1.20E-OS 4.81E-27 3.78E-20 99 .797 8.96E-03 Xe-133 5.243 d 18.925 1.89E+07 8.21E+01 8.17E+01 8.12E+01 8.08E+01 8.03E+01 7.92E+01 7.75E+01 7.59E+01 9.44E-07 4.76E-07 5.00E-07 1.63E-09 3.07E-04 Xe-135 9.1 h 13.630 1.36E+07 5.71E+01 5.29E+01 4.90E+01 4.54E+01 4.21E+01 3.48E+01 2.57E+01 1.93E+01 6.56E-07 2.40E-07 7.00E-08 8.23E-10 1.11 E-03 Xe-135m 15.3 m 6.760 6.76E+06 7.56E+OO 4 .99E-01 3.30E-02 2.18E-03 1.44E-04 1.61 E-07 3.06E-12 1.15E-16 8.69E-08 5.61 E-09 4.00E-08 1.92E-11 4.53E-05 Xe-137 3.82 m 35.800 3.58E+07 6.74E-01 1.26E-05 2.37E-10 4.44E-15 8.32E-20 1.26E-31 1.56E-50 2.92E-68 7.75E-09 4.70E-10 9.10E-o8 1.61E-12 1.67E-06 Xe-138 14.1 m 37 .380 3.74E+07 3.72E+01 1.95E+OO 1.02E-01 5.36E-03 2.81 E-04 1.76E-07 1.33E-12 2.09E-17 4.28E-07 2.73E-08 2.00E-08 9.34E-11 4.40E-04 Xe-139 39.7 s 30.675 3.07E+07 3.02E-12 1.54E-39 7.88E-67 4.02E-94 2.06E-121 1.21E-189 8.27E-299 O.OOE+OO 3.47E-20 2.10E-21 1. -08 7.20E-24 4.24E-17 143.170 1.90E-03 Denotes Effluent Release Limits calculated from DAC values that were calculated using FGR No. 12 methodology. (FGR 12 DAC/219) Total Iodine (CEDE) mrem 2.95E-04 Noble Gas !CEDE) mrem 1.09E-02 Total Dose (TEDE) mrem 1.12E-02

Case Summary of Containment Shine ATTACHMENT 12 Page 1 of 3 MicroShield 8.02 Nathan Hogue (8.00-0000) Date By Checked Filename Run Date Run Time Duration Fueled Experiment Failure Shine.msd February l, 2016 1:13:10 PM 00:00:00 Pro.iect Info Case Title Containment Shine Description Fueled Experiment Accident Analvsis Geometry 13 - Rectangular Volume Source Dimensions Lenllth l .9e+3 cm (60 ft 9.9 in) Width 1.9e+ 3 cm (60 ft 9.9 in) Height l.8e+3 cm (60 ft 0.1 in) Dose Points x y z

A

 ~1  l .9e+ 3 cm (62 ft 9.5 in) 914.0 cm (29 ft 11.8      914.0 cm (29 ft 11 .8                       -x in)                     in) 1.5e+4 cm (492 ft 1.5     914.0 cm (29 ft 11.8      914.0 cm (29 ft 11.8
'#2 in)                        in)                     in)

Shields Shield N Dimension Material Density Source 6.29e+09 cm 3 Air 0.00122 Shield I 30.0 cm Concrete 2.35 Air Gap Air 0.00122 Source Input: Grouping Method - Standard Indices Number of Groups: 25 Lower E nergy Cutoff: 0.015 Photons< 0.015: Included Library: Grove Nuclide Ci Bq µ.Ci/cm 3 Bq/cm 3 1-131 6.7550e+OOO 2.4994e+Ol l l .0745e-003 3.9755e+001 1-132 l .8635e+001 6.8950e+O 11 2.964le-003 l .0967e+002 I-133 3.9875e+001 1.4754e+012 6.3426e-003 2.3468e+002 1-134 4.5405e+OO I 1.6800e+012 7 .2222e-003 2.6722e+002 1-135 3.7695e+OOI l.3947e+Ol2 5.9958e-003 2.2 l 85e+002 Kr-85 2.0000e-003 7.4000e+007 3 .18 l 2e-007 1.1771 e-002 Kr-85m 7.5800e+OOO 2.8046e+Oll l .2057e-003 4.461 le+OOl Kr-87 l .5405e+OO 1 5.6999e+Ol l 2.4504e-003 9.0663e+001 Kr-88 2.1660e+OO 1 8.0142e+01 I 3.4453e-003 l .2748e+002 Kr-89 2.7740e+OOI 1.0264e+O 12 4.4 l 24e-003 1.6326e+002 Kr-90 2.7410e+001 1.0142e+Ol2 4.3599e-003 l .6 I 32e+002 file://IC :/Program%20Files%20(x86)/MicroShield%208/Examples/CaseFi Ies/HTML/Fueled ... 211/2016

Case Summary of Containment Shine ATTACHMENT 12 Page 2of3 Xe-133 l .8925e+o01 7.0023e+Ol 1 3.0103e-003 1.113 8e+002 Xe-135 1.3630e+OO 1 5.043 le+Ol l 2. I 680e-003 8.0217e+OOI Xe-I35m 6.7600e+OOO 2.50I2e+OI l l .0753e-003 3.9785e+001 Xe-137 3.5800e+001 I .3246e+Ol2 5.6944e-003 2.1069e+002 Xe-I 38 3.73 80e+OO I 1.3831e+Ol2 5.9457e-003 2. l 999e+002 Buildup: The material reference is Shield l Integration Parameters X Direction 10 Y Direction 20 Z Direction 20 Results - Dose Point# 1 - (1914,914,914 cm Fluence Rate Fluence Rate Exposure Rate Exposure Rate Energy (MeV) Activity (Photons/sec) MeV/cm 2 /sec MeV/cm 2/sec mR/hr mR/hr No Buildup With Buildup No Buildup With Buildup 0.015 3.0l 9e+ 11 1.696e-248 3.932e-24 1.455e-249 3.373e-25 0.03 5.188e+ 11 2.169e-34 2.351e-23 2.149e-36 2.330e-25 0.08 2.635e+l 1 l .247e-04 3.067e-03 l.973e-07 4.854e-06 0.1 3.588e+ l 1 2.03 le-03 7.465e-02 3.106e-06 l.142e-04 0.15 4.031e+ ll 3.530e-02 1.73 le+OO 5.812e-05 2.851e-03 0.2 1.155e+12 4.247e-01 2.023e+Ol 7.496e-04 3.570e-02 0.3 6.123e+l l 1.350e+OO 4.826e+Ol 2.562e-03 9.155e-02 0.4 1.352e+ l2 9.856e+OO 2.598e+02 1.920e-02 5.063e-01 0.5 2.832e+12 5.057e+ol 1.024e+03 9.927e-02 2.0lOe+OO 0.6 l .880e+ 12 6.829e+Ol 1.1 05e+03 l.333e-01 2.158e+OO 0.8 4.207e+ 12 4.519e+02 5. l 74e+03 8.596e-01 9.84le+OO 1.0 2.248e+12 5.420e+02 4.781e+03 9.991e-01 8.812e+OO 1.5 1.807e+ l2 l.737e+03 9.920e+03 2.923e+OO 1.669e+Ol 2.0 l.657e+12 3.878e+03 l.708e+04 5.997e+OO 2.641e+Ol 3.0 l.912e+ 11 l.362e+03 4.37le+03 1.847e+OO 5.930e+OO 4.0 4.713e+l0 6.71 le+o2 l .777e+03 8.302e-Ol 2.198e+OO Totals l.984e+13 8.772e+03 4.556e+04 l.37le+ol 7.469e+Ol Results - Dose Point# 2 - 15000,914,914) cm Fluence Rate Fluence Rate Exposure Rate Exposure Rate Energy (Me V) Activity (Photons/sec) MeV/cm 2/sec MeV/cm 2 /sec mR/hr mR/hr No Buildup With Buildup No Buildup With Buildup 0.015 3.0l 9e+ 11 3.722e-259 l.771e-26 3.l 92e-260 1.519e-27 0.03 5.l 88e+ 11 2.388e-37 1.059e-25 2.367e-39 1.050e-27 0.08 2.635e+l l 3.948e-07 l .095e-05 6.247e-10 l .733e-08 0.1 3.588e+ 11 6.634e-06 2.862e-04 1.015e-08 4.378e-07 0.15 4.03le+ l 1 l.282e-04 7.665e-03 2.l 12e-07 l.262e-05 0.2 1.155e+ l2 1.698e-03 9.897e-02 2.997e-06 l.747e-04 0.3 6.123e+ I I 6.210e-03 2.661e-OI l .178e-05 5.047e-04 0.4 l.352e+l2 4.964e-02 1.533e+OO 9.673e-05 2.987e-03 0.5 2.832e+12 2.719e-01 6.3I9e+OO 5.336e-04 l.240e-02 file:///C:/Program%20Files%20(x86)/MicroShield%208/Examples/CaseFiles/HTML/Fueled ... 211/20I6

Case Summary of Containment Shine ATTACHMENT 12 Page 3of3 0.6 1.880e+ 12 3.852e-01 7.043e+OO 7.519e-04 l.375e-02 0.8 4.207e+ l2 2.722e+OO 3.422e+Ol 5.l 77e-03 6.509e-02 1.0 2.248e+12 3.408e+OO 3.241e+Ol 6.282e-03 5.974e-02 1.5 l.807e+ l2 l.158e+Ol 6.916e+Ol l .948e-02 1.164e-01 2.0 l.657e+ 12 2.656e+ol 1.20le+02 4.107e-02 l.858e-01 3.0 l.912e+ 11 9.549e+OO 3.080e+Ol 1.296e-02 4.l 78e-02 4.0 4.713e+1 0 4.745e+OO l .249e+Ol 5.870e-03 I .545e-02 Totals 1.984e+13 5.928e+Ol 3.145e+02 9.224e-02 5.140e-01 file://IC :/Program%20Files%2 0( x86)/MicroShield%208/Examples/CaseFiles/HTML/Fueled... 21I/2016}}

v t e Letter Sequence Response to RAI
TAC:ME1580, (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request Acceptance... Administration Withholding Request Acceptance Questions 10 July 2009 3 December 2014 17 April 2015 14 September 2015 28 October 2015 17 December 2015 8 February 2016 24 August 2016 7 September 2016 Answers 28 January 2015 12 February 2015 1 October 2015 2 December 2015 8 April 2016 15 April 2016 25 July 2016 15 November 2016 Results Approval, Approval Other: ML13079A215, ML13079A216, ML13079A218, ML15154B812, ML15289A012
MONTHYEAR2015-07-21 [Table View]

ML16110A164

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