ML16207A426

Purdue University Responses to Request for Additional Information Re PUR-1 License Renewal and Power Uprate

ML16207A426
Person / Time
Site:	Purdue University
Issue date:	07/19/2016
From:	Townsend C Purdue University Research Reactor
To:	Cindy Montgomery Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML16207A426 (26)

Text

+

PURDUE UNIVERSITY COLLEGE OF ENGINEERING July 19, 2016 Document Control Desk ATIN: Cindy Montgomery U.S. Nuclear Regulatory Commission One White Flint North 50- \Kl.

11555 Rockville Pike Rockville, MD 20852-2738 RE: Purdue University* Request For Additional Information- ML15328A314 Dear Ms. Montgomery-Enclosed are responses to the Request for Additional Information sent by the NRC to Purdue University in your letter dated January 19, 2016 under ADAMS Accession Number ML15328A314 in support of the PUR-1 License Renewal and Power Uprate.

Please let me know if you have any questions or require further information at the contacts listed below.

I hereby certify under penalty of perjury with my signature below that the information contained in this submission is true and correct to the best of my knowledge.

Sincerely, Clive Townsend PUR-1 Reactor Supervisor clive@purdue.edu 765-494-5764 Attachments: As described.

1. TS 1.49: PUR-1TS1.49 provides a definition for an Unsecured Experiment. This definition refers to the definition in PUR-1 TS 1.36 for a secured experiment. The definition for "Secured Experiment," is provided in TS 1.37. Update TS 1.49 to reference the definition for secured experiment in TS 1.37, or explain why it is not necessary.

The definition for an Unsecured Experiment has been updated to reference the appropriate section in the PUR-1 Technical Specifications. Amendment #13 to the Tech Specs as revised is enclosed in the RAI response . Note : The reference to TS 1.49 is revised to TS 1.50 following the solution to RAI 3.

2. TS 2.2: The basis in PUR-1 TS 2.2, states that a steady-state power level of 94.2 kW is required to initiate the onset of nucleate boiling (ONB) with a maximum fuel temperature of 49 degrees Celsius. The response to RAI 5 in your letter dated July 24, 2015, indicates that the PUR-1 ONB power level is 98.6 kW with a maximum fuel temperature of 43.20 degrees Celsius. Clarify the ONB power level and maximum fuel temperature and provide updates to the PUR-1TS2.2 basis that correctly reflects the ONB power and maximum fuel temperature under the renewal of the PUR-1 facility license.

The onset of nucleate boiling power level has been clarified to show that ONB will start at 98.6 kW. The maximum fuel temperature at this power level has also been clarified in Section 2.2 of the PUR-1 Tech Specs. The fuel temperature of 43.2 degrees centigrade remains well below the Safety Limit of 530 degrees centigrade. Amendment #13 to the Tech Specs as revised is enclosed in the RAI response.

3. TS 3.2: PUR-1 TS 3.2, states that the measured value of the power level scram shall be no higher than 12.0 kW. In the license renewal application as supplemented (Purdue University Research Reactor, "Application for Relicense of License Number R-87 with Power Uprate, Safety Analysis Report, " dated July 07, 2008, ADAMS ML083040443), PUR-1 requested to operate the PUR-1 facility at 12.0 kW normal power. TS 3.2 Table I identifies the safety channels required for operation that includes the Log N and period and safety channels. The setpoint for the Log N and period slow scram function is 120% power. The setpoints for the safety channel setback function is 110% and for the fast scram is 120% power. The setpoint values are not consistent with the TS 2.2 power level scram of 12.0 kW, since the requested normal power operating level is 12.0 kW. Describe and explain the relationship between the power level scram and the setpoint values for the Log N and period and safety channels.

Minor ambiguities have persisted in the interpretation of the requested maximum power level since the "Application for Relicense of License Number R-87 with Power Uprate, Safety Analysis Report," dated July 07, 2008, ADAMS ML083040443). TS Definition 1.26 has been added and states: "Power Level - There are three important and separately defined power levels.

Instantaneous Power Level is the power level of the reactor at any given moment, as indicated by the reactor instrumentation . Setback and Scram set points will trip based on the instantaneous operating power level. The Steady State Power Level is the four hour average power level from pg. 1

which setpoints for scram and setback shall be calculated . The steady state operating power level should be 10 kW or less. Transient deviations above 10 kW are allowed provided the Maximum Licensed Power Level is not exceeded . The Maximum Licensed Power Level is the maximum instantaneous power level allowed by the PUR-1 License. The Maximum Licensed Power Level is 12 kW."

The Limiting Safety System Setting specification (TS 2.2) states "The measured value of the power level scram shall be no higher than 12.0 kW." The LSSS is satisfied by the definition of Instantaneous Power Level. Table I in Technical Specification 3.2 specifically now references that the setpoint for the Log N channel shall be 120% of the steady state operating power level. 120%

of 10 kW is 12 kW, the licensed maximum power level. Similar statements can be made about the rest of Table I. Amendment #13 to the Tech Specs as revised is enclosed in the RAI response.

4. TS 3.5(g): PUR-1 TS 3.5(g) states that The radioactive material content, including fission products, of any double encapsulated experiment or vented experiment should be limited so that the complete release of all gaseous, particulate, or volatile components from the encapsulation or confining boundary of the experiment could not result in (1) a dose to any person occupying an unrestricted area continuously for a period of two hours starting at the time of release in excess of 0.5 Rem to the whole body or 1.5 Rem to the thyroid or (2) a dose to any person occupying a restricted area during the length of time required to evacuate the restricted area in excess of 5 Rem to the whole body or 30 Rem to the thyroid."

These dose limits do not meet the requirements in 10 CFR Part 20. Propose revised dose limits in accordance with 10 CFR 20.1201, "Occupational dose limits for adults," and 20.1301, "Dose limits for individual members of the public, "or explain why the above limits are acceptable.

TS 3.S(g) has been amended to read:

The radioactive material content, including fission products, of any double encapsulated experiment or vented experiment should be limited so that the complete release of all gaseous, particulate, or volatile components from the encapsulation or confining boundary of the experiment cou ld not result in (1) a total effective dose equivalent exceeding 0.1 rem or a dose exceeding 2 mrem in any one hour to a member of the public in any unrestricted area or (2) a total effective dose equivalent to any person occupying a restricted area during the length of time required to evacuate the restricted area in excess of 5 Rem or the sum of the deep-dose equivalent and the committed dose equivalent to any individual organ or tissue other than the lens of the eye being equal to 50 rems.

Amendment #13 to the Tech Specs as revised is enclosed in the RAI response.

5. TS 5.2.4: NUREG-1537, Part 1, Section 5.1, "Summary Description," and 5.2 "Primary Coolant System," provides guidance for licensees to provide the design bases and the functional requirements of the primary and secondary cooling system. In Section 5.3, "Secondary Coolant pg. 2

System," of the safety analysis report (SAR), the licensee states that when operating the PUR-1 at a power level of 10 kW, the calculated reactor pool temperature rise would be 0.465 degree Celsius per hour (taking no credit for heat loss to the surrounding sand and gravel or loss by evaporation). Further, the licensee states that since the heat-removal capacity of the heat exchanges is 10,550 Watts, the heat exchanger will maintain the pool temperature at an acceptable level.

PUR-1 TS 5.2.4, describes the Primary Coolant Chiller System and states that this heat removal capacity is sufficient to maintain the pool temperature at 75 degrees Fahrenheit (23.09 degrees Celsius) at continuous operation at 10 kW. However, no technical specification is provided for the maximum allowable pool temperature when operating. In addition, the PUR-1 license renewal application requests a continuous operating power level of 12.0kW, which is above the temperature at which analyses were performed. The PUR-1 SAR, Section 4.6, Table 4-18, "Model Dimensions for the Thermal Hydraulic Models," provides the conditions for the performed thermal hydraulics safety analysis and showed that when operating the reactor core at 10 kW, with a pool bulk temperature at 30 degrees Celsius, the maximum fuel temperature is well below the fuel temperature safety limit.

Provide an analysis of the cooling system, including pool temperature and operating temperature such that the reactor operates within analysis conditions. Provide a technical specification that ensures that the reactor pool bulk temperature is always maintained below the SAR-analyzed temperature under all operating conditions, with steady-state operating reactor core power of 12 kW.

The heat exchanger is rated to be capable of remov ing up to 10,550 Watts. Supposing that the reactor was run at a power level of 18 kW (12 kW + 50% instrumentation uncerta inty), the additional heat being added to the pool beyond the heat removal capacity of the heat exchanger would be 18 kW -10.55 kW= 7.45 kW or 7450 Watts For conservat ism, suppose the heat add ition is 7.5 kW. The relevant heat transfer equation is Q = mc!J.T where m is the mass of water, Q is the generation of heat given by Q = Power

• Time c is the specific heat capacity and !J.T is the change in temperature. For a 6400 gallon tank, the mass of water is approximately liters cm 3 g m = 6400 gallons

• 3.785 ll

• 1000-.-

• 1 - =2.422x10 7 grams H2 0 ga on 1iter cm 3 pg . 3

and the specific heat capacity for water is 4.179-1-. Solving for the change in temperature per g**c hour Pt

=>!:J.T=-

mc 7500 U/s)

• 60 m in/hr

• 60 sec/min °C

/:J.T = 2.422x10 7 (g)*4.179U/g°C)

= 0.267 -

hr In a four hour instrument check interval, the expected temperature rise would therefore be 1.07 °C . Supposing the heat exchanger was not operating at all, the rise in temperature would be 18000 Ujs)

• 60 min/hr* 60 sec/min °C

/:J.T = 2.422 x 10 7 (g)

• 4.179 U/g°C)

= 0.640 -hr which corresponds to a 2.56 °C temperature rise over the course of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Using the NATCONV code previously implemented to find the onset of nucleate boiling, the above calculated potential change in pool temperature at 18 kW without the heat exchanger operating and no benefit taken for heat loss to the environment, an increase of the pool temperature of 2.56 °C decreases the reactor power at which the ONB occurs by 2.23 kW. As the licensed power level requested is 12 kW, a decrease from 98.6 kW to approximately 96 kW does not significantly change the overall safety margin of the facility or affect the risk to the public.

Technical specification 3.3.d has been added to read "The primary coolant (bulk pool volume) shall be maintained below 30 °C while the reactor is at the steady state operating power."

Additionally, the "Bases" for Tech Spec 3.3 has been amended to read " ... Maintaining the primary coolant temperature will ensure the margin to the onset of nucleate boiling is maintained and analyses shown in the Safety Analysis Report remain valid ." Amendment #13 to the Tech Specs as revised is enclosed in the RAI response .

6. NUREG-1537, Part 1, Section 13.1.2, "Insertion of Excess Reactivity, "provides guidance for the analysis of insertion of excess reactivity. In Section 13.2.2, of the SAR, the licensee presents two reactivity insertion scenarios, one with and one without a scram. The initial power assumed by PUR-1 at the initiation of the reactivity insertion accident is 10.0 kW. Provide a discussion and supporting analyses that shows that the 10 kW initial power level provides the most limiting transient results. If 10 kW is not limiting, provide the limiting analysis. Include consideration for power level measurement uncertainty as well as the power level scram at 12.0 kW.

Alternatively, demonstrate that the analysis results are insensitive to initial power.

The main equation in power transient analysis is pg.4

It has been shown that a 0.6 % !J.k/k reactivity insertion (maximum excess reactivity) will induce a one second period in the PUR-1 core. If the initial reactor power is 12 kW plus a 50% instrument uncertainty, the maximum power during a transient without a period trip and a one second delay in full control rod insertion would be P1 = 18 e 111 = 48.9 kW At this power level, the onset of nucleate boiling has not started yet and there is no threat to any system in the PUR-1.

7. Guidance in NUREG-1537 states that the format and content of the TSs follow ANSI/ANS 15.1-2007, Section 3.8.2, provides guidance for experiments involving fissionable, explosive, reactive, or corrosive materials.

RAI 8(c) in NRC letter dated August 29, 2014 states:

NUREG-1537, Part I, Section 13.1.1 provides guidance in identifying an acceptable MHA for non-power reactors. The PUR-1 MHA accident analysis for "Failure of a Fueled Experiment " is stated to be based upon a 1 W power deposition in the fueled experiment as consequence of the reactor operating at 1 kW. Please provide ... a safety analysis that provides the details of the energy deposition determination in the fueled sample with the reactor operating at the maximum requested licensed reactor power including the power level measurement uncertainty of 50% stated in SAR, Section 13.1.2.

The response to RAI 8(c) in your letter dated July 24, 2015, indicates that a limit of 0.5 Ci of radio-iodine is specified in the PUR-1 TSs that is half the amount analyzed in the MHA. Your analysis indicates that this limit corresponds to 1.10 gram of fissile material at the proposed operating power of 12 kW (plus 50% margin and assuming 3% enriched uranium). The analysis further indicates that a potential failure of an experiment with the maximum allowable fissile material content results in dose rates which are well below the dose rates of a maximum hypothetical accident (MHA) event both for an occupational worker and also a member of the public.

The requirement of a maximum allowable limit of 0.5 Ci radio-iodine for experiments containing fissile material does not seem to appear in the PUR-1 TSs. Amend PU-1 TSs to include this requirement, provide a reference to where this requirement exists either in the TSs or other controlling procedure, or provide a explanations describing your reason(s) for not including the maximum allowable limit of 0.5 Ci radio-iodine for experiments containing fissile material.

Technical Specification 3.5.h has been added to the PUR-1 TS document and reads "A fueled experiment shall not produce more than 0.5 Curies of radio-iodine." The corresponding technical pg. 5

basis reads, "Limiting the amount of radio-iodine levels in a fueled experiment will ensure that the Maximum Hypothetical accident analyzed in the Safety Analysis Report remains the bounding incident which could occur at the PUR-1." Amendment #13 to the Tech Specs as revised is enclosed in the RAI response .

8. NUGEG-1537, Part 1, Section 13.1.6, "Experiment Malfunction," and 13.2, "Accident Analysis and Determination of Consequences, provides guidance for analyzing experiment failures including the evaluation of the potential radiological consequences. The radiological consequences should include external and internal exposure and an occupational worker and a member of the public. The dose conversion calculations may use the Environmental Protection Agency's Federal Guidance report (FGR)-11 and FGR-12 dose conversion coefficients or another equivalent methodology to account for inhalation/ingestion and submersion exposures. Your analysis seems to define dose values by combing whole-body submersion with thyroid inhalation doses. Provide an updated analysis and discuss the potential maximum whole-body TEDE radiological dose estimate due to the failure of an experiment with the maximum allowable fissile material content or provide justification why it is not necessary.

Technical Specification 3.5.h as included in the enclosed revision of the PUR-1 Tech Specs limits radio-iodine production to 0.5 Ci and specifies the bases for this technical specification is to ensure the failure of a fueled experiment does not become the Maximum Hypothetical Accident. The whole-body total effective dose equivalent (TEDE) is a Nuclear Regulatory Commission term which is used to quantify the effects of external and internal radiation exposures. The deep dose equivalent (DDE) from external gamma radiation and the committed effective dose equivalent (CEDE) are summed to find the TEDE. The CEDE is the internal dose received by the body in the 50 years following a radioactive inhalation and is weighted by the risk of fatal cancer in organs which are affected.

TEDE = CEDE+ DDE Suppose that 0.5 Ci of radio-iodine was immediately released into the air from the failure of a fueled experiment. The reactor room has a volume of 424 m 3 which would yield a radio-iodine concentration of

. 6 µCi .

0. 5 Ct* 10 Ci µCi

---

~3 = 0.00118-424 m3

• 1Q6 cm cm3 m3 FGR-11 estimates the volume of air breathed by a worker with a normal breathing rate to be 0.02 m 3 /min . In the table below, are the calculations for the TEDE dose received by an occupational worker who evacuates the room following the release. A sample calculation follows below for 1-135.

The volume of air breathed in per second is pg.6

m3 cm 3 1 minute cm 3 0.02 *10 6 - * =333--

minute m 3 60 seconds sec If it takes 60 seconds for the worker to evacuate the reactor room, he will breathe in cm 3 333-

• 60 sec= 20,000 cm 3 sec With the radio-iodine of concentration from above, the total µCi of radio-iodine breathed in is

µCi 0.00118 -

• 20,000 cm 3 = 23.6 µCi cm 3 The Evaluated Nuclear Data File Database of December 7, 2015 (https://www-nds.iaea.org/exfor/endf.htm) gives a yield of 82.9% of 135 1 from the thermal fission of Uranium-235 with respect to all the other isotopes of Iodine (The fractional yield of 1-135 is 0.0293 with respect to all isotopes from thermal 235 U fission .) Therefore, the radioactivity of 135 1 breathed in is 23.6 µCi * .829 (1 3 sI) = 19.56 µCi of 13 sI FGR-11 shows an inhalation committed effective dose equivalent of 1.23 mRe~ for 135 1 which

µCi yields an inhalation dose of mRem 19.56 µCi

• 1.23 -C-.- = ~ 24 mRem

µ l 135 Similarly, the submersion dose for 1is calculated from FGR-12. If there is 0.5 Ci of radio-iodine released, that would correspond to 0.415 Ci of 135 1 which is dispersed around the room giving a concentration of 0.415 Ci

--- = 9.78 x 10- 10 µCi/cm 3 424m 3 135 FGR-12 gives a conversion for submersion in 1of mRem/sec 296----

µCi/cm3 Given the concentration of 135 1, the dose received while evacuating the room in 60 seconds is mRem/sec µCi 296 3

• 9.77 x 10 -3

• 60 seconds= 1.74 x 10-s mRem

µCi/cm cm pg. 7

135 The TEDE from 1is the sum of the dose from inhalation and submersion yielding 24.06 mReminhaltion + 1.73 X 10-s mRemsubmersion = -24 mRem in 60 seconds The table below sums the dose for all isotopes of iodine to facility personnel.

Isotope CEDE Percent Yield Inhalation CEDE Submersion (mRem/µCi) from Fission Dose (mRem) Submersion Dose (mRem)

(mRem/ hr)

µCi/cm 3 1-126 44.40 0.0% 0.00 79.6 2.71E-16 1-128 0.05 0.0% 0.00 15.4 3.21E-13 1-130 2.64 0.0% 0.00 385.2 1.20E-09 1-131 32.89 0.1% 0.86 67.41 5.29E-09 1-132 0.38 0.3% 0.02 56.67 1.04E-08 l-132m 0.30 0.3% 0.02 414.8 7.59E-08 1-133 5.85 2.3% 3.22 108.9 1.80E-07 1-134 0.13 14.2% 0.44 481.5 4.82E-06 1-135 1.23 82.9% 24.01 295.6 1.73E-05 Inhalation 28.57 Submersion 0.00002 Dose (mRem) Dose (mRem)

Total Effective Dose Equivalent -2s.6 mRem The dose of 28.6 mRem is far less than the 5 Rem limitation as regulated by 10 CFR Part 20.

The following paragraphs will analyze the potential risk to a member of the public in an unrestricted area during the failure of a fueled experiment containing 0.5 Ci of radio iodine. The PUR-1 Facility has an air exhaust system which allows for the release of air from the reactor room through a stack which is at least 50 feet in the air above the Electrical Engineering building in which the reactor is housed. The reactor room itself is required by TS 3.4.a.1 to have a negative air pressure with respect to the atmosphere of 0.05 inches of water for operation. Therefore, any release of airborne radioactivity will be through the exhaust rather than any of the other perimeter points.

Doses to persons outside the bu ilding will come from submersion in a cloud of released radioiodines. Because 10 CFR Part 20.1301 requires that the dose to a person in an unrestricted area be no more than 2 mRem in any one hour, the analysis is slightly different due to the length of time under which exposure is examined .

The submersion and inhalation dose results from the diluted radioiodine stream from the exhaust fan . An analysis for the activity concentration release from the building can be performed using the equation below as specified in NUREG/CR-2260. The concentration of activity of a radioiodide is given as pg.8

where Ai is the fractional release of activity of the element, x/Q is the atmospheric dispersion factor with units seconds/m , and V 3

(m 3 /sec) is the volumetric release rate from the fan. The dispersion factor is calculated by taking the most appropriate value of x/Q from the equations below (1)

(2)

(3) where rr = 3.14159, u 10 is the wind speed at ten meters above plant grade, O"y is the lateral plume spread (a function of atmospheric stability and distance), O"z is the vertical plume spread (a function of atmospheric stability and distance), Ly is the lateral plume spread with meander and building wake effects, and A is the smallest vertical-plane cross-sectional area of the reactor building in square meters. To find the best version of the dispersion factor to use, the larger dispersion factor from equations (1) and (2) should be compared with the value from equation (3). The lesser of the first result and equation (3) should be the final dispersion factor.

This analysis will conservatively consider moderately stable atmospheric conditions meaning low dispersion. Specifically, a Class F atmospheric stability is used. A wind speed of 1 m/s, which is much lower than the average wind speed as referenced in the Safety Analysis Report, will be used.

The distance from the smoke stack will be considered to be 100 meters as that is the minimum value available in Regulation Guide 1.145, Figures 1 and 2. At 100 meters for a moderately stable atmosphere, O"y = 4 m, O"z = 2.6 m. The value for Ly for distances less than 800 meters is given as where Mis a correction factor based on atmospheric stability and wind speed . With a wind speed of 1 m/s and Class F stability, M = 4 and therefore Ly= 4

• 4 = 16 Finally, the cross sectional area is A = 288 m 2

• The values of the atmospheric dispersion factor is then pg . 9

1 sec x/Q = m( 288 m2) = 0.00566 m3 1 s rr

• 4 m

• 2.6 m + 2 (1) 1 sec x/Q = m = 0.0102 - 3 s

• 3

• rr

• 4 m

• 2.6 m 1- m (2) 1 sec x/Q = m = 0.00765 - 3 s

• rr

• 16 m

• 2.6 m 1- m (3)

The larger of equations (1) and (2) is equation (2)'s value of 0.0102 sec/m 3 and the smaller of equation (3) and the previous result is equation (3)'s solution of 0.00765 sec/m 3 . Therefore, the value of the dispersion factor is taken to be sec x/Q = 0.00165 -

m3 Given that the exhaust fan expels approximately 0.2 m 3 /sec of contaminated air (in a scenario where the exhaust fan errantly remains running), the room also brings in 0.2 m 3 /sec of clean air.

The amount of radio-iodine present at any given time is therefore dC(t) m3 C(t)

- - = -0.2-* - - -

dt sec 424 m3 The solution to this first order differential equation is where the concentration at t = 0 is 0.00118 cm µc~ giving a value for k 1 of and finally the concentration as a function of time is C(t) = 0.00118

• e-4.717 x 10-4t The concentration has been reduced to 50% of the original concentration after pg . 10

C(t)

Co= 0.5 = e-4.717 X10

-4 t

ln(0.5)

=> t = -4.717 x 10- 4 = 1469 seconds = 24.5 minutes Using the concentration in the plume with the appropriate dispersion factor, the plume concentration initially at time zero (neglecting the time for the plume to travel out of the exhaust) 3 sec m [Ci]

Cinitial =Ai

• 0.00765 - 3

• 0.2- = 0.00153 *Ai - 3 m sec m 3

µCi) sec m _6 [µCi]

Cinitial = 0.00118 (- -

cm 3

• 0.00765 - 3 m

• 0.2-sec

= 1.8054 X 10 - -3 cm It would be more appropriate to express the concentration as a function of time. Combining the plume concentration equation and the concentration of the radio-iodine in the plume being exhausted from the facility, C(t) = 0.00118

• e- 4 .717 xio- 4 t

• 0.00153 = 1.8054 x 10- 6

• e- 4 .717 x10- 4 t [µCi]3 cm Suppose then, that the concentration of the radio-iodine concentrations are broken into four time increments using the concentration at each time period. That is to say the concentration for the first 15 minutes is the initial concentration, the concentration for the next 15 minutes is the concentration at t = 15 min, and so on. The resultant concentrations are then C(O) = 1.81x10- 6 µCi/cm 3 C(900 seconds) = 1.8054 x 10- 6

• e- 4.71 7 x io- 4

• 9 oo = 1.18 x 10-6 µCi/cm 3 C(1800 seconds) = 1.8054 x 10- 6

• e- 4 .717 x io - 4

• 1800 = 7.72 x 10- 1 µCi/cm 3 C(2700 seconds) = 1.8054 x 10- 6

• e- 4 .717 xio- 4

• 2100

= 5.05 x 10- 7 µCi/cm 3 The TEDE dose equivalent for each time step is CEDE DDE TEDE = - - *time+ - - *time time time In the first 15 minutes with the initial concentration, the dose received to a member of the public is 0.656 mRem. The second 15 minutes give a dose of 0.281 mRem, and so on.

The table below shows the total effective dose equivalent to a member of the public in each of the above time steps pg. 11

Total Radio-iodine TEDE Dose Time Step Concentration (µCi/cm 3 ) Received (mRem) 0-15 minutes 1.81x10- 6 0.656 15-30 minutes 1.18 x 10- 6 0.429 30-45 minutes 7.72 x 10- 7 0.281 45-60 minutes 5.05 x 10- 7 0.200 Total Dose in the first hour 1.566 mRem This final conservative total effective dose equivalent to a member of the public of 1.566 mRem in the first hour after an accident scenario following the complete failure of a fueled experiment meets those requirements set forth in 10 CFR Part 20.1301.

9. NUREG -1537, Part 1, Section 13.1.6 and 13.2 (7) provides guidance for analyzing experiment failures and includes the evaluation of the potential radiological consequences. The radiological consequences should include external and internal exposure whole-body TEDE values for the duration of the accident. The response to RAI S(c) in your letter dated July 29, 2015, indicates that the dose analysis was based on several exposure periods. Provide the total radiological whole-body TEDE (thyroid does, if more limiting) to (1) an occupational worker, allowing for a realistic evacuation process, and (2) to the maximally exposed member of the public considering an emergency evacuation plan.

Please reference response to RAI 8 above regarding the TEDE to a facility worker, allowing for a realistic evacuation process, and to a maximally exposed member of the public, considering an emergency evacuation plan.

10. RAI 7 in NRC letter dated August 29, 2014 stated:

The requirements of 10 CFR 20.1101 states that each licensee shall develop, document, and implement a radiation protection program commensurate with the scope and extend of licensed activities in order to limit the total effective dose equivalent to facility workers (annual occupational dose less than 5 rem [roentgen equivalent man])

and the total effective dose equivalent to individual members of the public (annual public dose less than 100 mrem). Please provide a safety analysis that explains all analyses, assumption and conclusions at the requested license power level for the maximum potential estimate of the total annual production of argon-41 from PUR-1 normal operations. In addition, please evaluate and discuss the potential maximum dose to a facility worker and to a member of the public (i.e., classrooms, hallways adjacent rooms, nearest dormitories, offices, etc.) due to this bounding yearly production and release of argon-41 from the facility.

Your response to RAI 7 by letter dated July 24, 2015, provided an analysis of Ar-41 production in the PUR-1 facility during normal operation due to air being absorbed in the reactor coolant and activated in the reactor core. The maximum allowable Ar-41 concentration for occupational pg . 12

workers is 3 x 10-6 µCi/cm 3 and the public 1x10-8 µCi/cm 3 established in 10 CFR Part 20, appendix B, Table 2. The PUR-1 analysis estimates substantially higher concentration levels.

Provide an updated Ar-41 safety analysis that explains all analysis and assumptions, and how it conforms to the requirements of 10 CFR Part 20 for occupational workers and members of the public.

The buildup of Ar is due to the absorption of a thermal neutron by 40Ar. Gasses are naturally 41 present in fluid like water and the amount is dependent on temperature as well as the partial pressure ofthe surrounding volume. Henry's Law dictates the amount of various gasses which are dissolved in fluid and carries a Henry's constant of 1.4 x 10- 5 [;~ a] at standard temperature 1

and pressure conditions for natural argon. (Sander, 2014). Because solubility decreases with room temperature, this is a conservative estimate as the temperature of the pool for the rest of the analysis has been at or above 20 °C.

The amount of argon in air is approximately 1%. The partial pressure of argon in the atmosphere is then Pargon = Patmosphere

• 0.01 Pargon = 101,325 Pa* 0.01 = 1,013.25 Pa With a partial pressure of 1,013.25 Pa at the top of the pool, the number of Ar atoms dissolved per cubic centimeters is

~ol ]

• 1013.25 Pa* 6.022 x 10 2 3 3

1 1.4 x 10-s [ [atoms] * [m ]

m Pa mol 1x106 cm 3 atoms]

= 8.543 x 10 15 [- -

cm3 40 Ar makes up 99% of all natural Argon, so number of 40 Ar atoms per cubic centimeter would be corrected as 8.543 x 101s [ato~s] * .99 [A40Ar] = 8.457 x 101s [ato~s]

cm rnat cm The amount of time that the coolant spends within the core will be its irradiation time. The NATCON code, used in response to RAI #5 in the letter dated July 24, 2015, predicted a mass flow 3

rate of 6.86 [ h grams ] or 6.86 [ h cm ]

c anne 1*sec c anne 1*sec . The 13 standard flow elements have 15 channels and the 3 control elements have 11 which yields 228 channels and a total volumetric flow rate of 1564.08 [cm sec 3

]. The recirculation time of this portion of the pool water through the core will be given by Vpool Tcirc =-.-

l'core pg. 13

The pool has a radius of 4 [ft] or 121.92 [cm] and a height of 17 [ft] or 518 [cm]. The volume of 2.42 x 10 7 [cm 3] gives a recirculation period of

_ 2.42 x 10 7 [cm 3] _ 4 _

Tcirc - [ 3] - 1.54 7 x 10 [sec] = 4.3 [hours]

1564 08 cm

• sec Using the MCNP6 model, a thermal neutron flux is predicted as being = 2.66 x 10 11 [neutrons]

cm2 sec within the core volume. This value is conservative as the thermal neutron flux is usually considered linear with power increase which would suggest a flux of = 10 11 [neutrons].

sec The saturation activity of 41Ar will be that which would be normally produced and decayed while in the reactor volume, reduced by that which decays while circulating throughout the pool volume.

Here, N is the number of atoms found within the core at any given time, ath is the thermal neutron absorption cross section for 40 Ar (ath = 6.1 x 10- 25 cm 2), A. is the decay constant (1.0566 x 10- 4 [sec- 1], and t is the transit time through the core. With a core height of 60.96 [cm] and a coolant velocity of 1.916 [cm], the time in the core is sec 60.96 [cm]

t = [cm]= 31.82 [sec]

1.916 sec 8.457 x 101s [_!!__] 6.1x10-2s[cm2] 2.66 x 1011 [-#-] (1 - e-(1.0S66x10-4 [s-1]*31.82 [s] )

cm 3 cm 2s Asat=~~~~~-='"""'-~(-1~--e---(-1.-05_6_6_ x _10---

4 -[s-ec---1-]*-(3 . 8~2=

[s~ec~]~+-1-.5-47_x_1_04-[-se_c_])_)~~~~~~~-

decays ]

Asat=5.717 [ 3 sec* cm The number of 41 Ar atoms is 5 714 [ decays ]

- Asat - . sec* cm3 - 4 [atoms]

NAr- 41 - 1 - l - 5.411X10 3

/l 1.0566 x 10-4 [-] cm sec The exchange of a gas in water with atmosphere can be modeled as S = 0.93 BNAr-41Asurf where B is an exchange coefficient reported as 5.7 x 10- 3 [;:] and A surf is the area of the surface of the pool as calculated from a radius of 121.92 [cm] .

S = 0.93

• 5.7 x 10- 3 [-cm]

• 5.411x10 4 [atoms] -

• rr(l21.92) 2 [cm 2]

sec cm pg. 14

s = 1.34 x 10 7 [atoms]

--

sec Multiplying the source value obtained above with the decay constant would yield the activity emitted from the pool surface per second.

[atoms]* 1.0566 x 10- [2-] = 1.41x10 [Bq]

• 1 S= 1.34 x 10 7 sec 4

sec 3 [µCi]

sec 3.7 x 10 Bq 4

c s= 0.03825 [!:...!:.]

sec The time radioactive air rema ins in the reactor room is a factor of the pumping rate of ventilation systems and the decay of the argon. The half-lives can be added as resistances and would be 1 1 1

-=--+-

• ett T:Ar-41 *air If the reactor room has a volume of 4.24 x 10 8 [m 3 ] and the fan removes air at a rate of 2 x 10 5 [cm sec 3

], the air has a lifetime of Vtotal 4.24 X 10 8 [cm 3] .

• air= . = [ 3] = 2120 [sec]= 35.3 [mm]

Vremoval 2 x 10s cm sec Using this as the half-life of air in the room which accounts for mixing and parts of the air staying 41 over time, the effective half-life of Ar in the room is 109.34

• 35.3 . 3

• etf = . + . = 26.7 [mm] = 1.602 x 10 [sec]

109 34 35 3 and the decay constant is ln(2) [ 1 ]

Aetf =- - = 4.326 x 10- 4 -

• etf sec Considering the room now as the entire source term, the activity at saturation is Spool Aroom(t) = Ji.V 0.03825 [µCi]

sec

=~~~~~~....,-~~~~~~-

4.326 x 10- 4 [s;c]

• 4.24 x 10s [cm3]

_

Aroom,sat - 2.085 X 10

_7 [µCi]

Crn"3 Referencing FGR-11, the dose conversion factor for someone submerged in argon is pg. 15

mRem/hr]

DCFAr- 41 = 8.029 x 10s [ µCi/cm 3 yielding a dose rate to a worker of D. = 2.08 x 10- 7

• 8.029 x 10s = 0.1674 [mRem]-,;.;:--

This is an incredibly conservative estimate for the steady state derived air concentration of the 41 Ar effluent. It assumes the pool water has reached a saturation of 41Ar, which has then diffused completely into the reactor facility, as it is continually evacuated. To simply make the second pass of water through the reactor core would take more than four hours. An extremely long run time would be 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> at which time the pool would start to approach the saturation point, a time far less than that of the saturation time for the entire reactor facility.

The reactor facility operates at a negative pressure and air is expelled to the outside of the building through an exhaust fan . This is 15 meters above the ground. Following a similar analysis to that in RAI 8 regarding the failure of a fueled experiment but assuming the exhaust rate of Argon-41 to be constant with time, the dispersion factor is 0.00765 3

[s/m 3

]. With an exhaust rate of 0.2 [m /s]

Ci =A (x/Q)ir = 2.085 x 10- 7

• 0.00765

• 0.2 = 3.190 x 10- 10 [µCi]

cm 3 Multiplying this by the dose conversion factor for someone perfectly under the plume yields b= 8.029 x 10s

• 3.190 x 10- 10 = 2.56 x 10-4 [m~;m]

This dose rate is far less than that cited in 10 CFR 20.1301(a).(2) of 0.002 [rem/hr].

11. RAI 12(a) in NRC letter dated August 29, 2014 stated:

Provide an MHA safety analysis that explains all analyses, assumptions, and conclusion at the requested licensed power level for the maximum potential estimate of the total radioactive fission product release after the failure of one side of one fuel plate. Discuss methodological assumptions associated with the following analytical steps:

(a) Derivations of fission product atmospheric dispersion factor, x/Q using either the methodology suggested in Regulator Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments At Nuclear Power Plants," Revision 1, issued February 1983, or another equivalent method.

Your response to RAI 12(a) by letter dated July 24, 2015, provide an analysis of the atmospheric dispersion factor, x/Q using Eq. (1) from RG 1.145. However, RG 1.145 suggests that the final x/Q value should be selected through an evaluation procedure by comparing Eq. (1) with (2)

(select higher}, and then compare with Eq. (3) (select lower). Please discuss whether your pg . 16

selection of x/Q based on Eq. (1) in RG 1.145 followed the RG methodology and is a conservative assumption and provides a bounding value for the atmospheric dispersion factor.

The total amount of radionuclides released from the failure of a fuel plate will be found by presuming that the entire face of one side of the plate is instantaneously removed . No credit is taken for the reduction in activity resulting from radioactive decay during the time of the release, i.e. an instantaneous release of the radioiodine that can escape the fuel is assumed . Complete and perfect mixing of the available radionuclide inventory with the reactor facility volume is also assumed .

The activity via the production rate of the ith radioiodine isotope in a plate is determined by the following:

Ai = AiNi = RfissFi where R is the rate of fissions in the plate of interest, Fi is the fraction fission yield for each radioiod ine, Ai is the decay constant for each radioiodine and Ni is the saturation number of atoms of each specific radioiodine . The constants Fi, Ai, and the results of calculations for AiNi and Ni for plate 1348, are shown assuming 12 kW+ 50% operating power uncertainty.

It is assumed that not all of the nuclides produced will be released from the plate. As suggested by NUREG/CR-2079, only the fission fragment gases within recoil range of the surface of the fuel (1.37 x 10- 3 [cm] for aluminum matrix fuels) will escape. The thickness of the fuel meat in a PUR-1 plate is 0.0508 cm. The total volume of the fuel meat is Vtot = 60.01

• 5.96

• 0.0508 = 18.170 [cm 3 ]

and the volume released from recoil of the fission gasses is Vrelease = 60.01

• 5.96

• 1.37 X 10- 3 = 0.490 [cm 3 ]

The fractional fission product gas release is therefore Vrelease 0.490 Frelease = Vtot = 18.170 = 0.027 The number of fissions per second in Plate 1348 with a fission power of 157 [Watts] is given by 157 w = 157 w * -

1 f

• 6.2415 x 10 12 MeV 1 fission

• = 4.90 x 10 12 fissions 1W l]oule 200 MeV sec As an example, the activity of 1-135 is then the product of these number of fissions and the fractional yield per fission of this element which is 6.4%.

. . [fissions] [decays] 1 c Actwity1_ 135 = 4.90 x 10 12 d

• 0.064 f. .

• secon isswn 3.7 x 1010 [decay/ ;econd]

• 0.027 {Fractional Release}

Activity1_ 135 = 0.229 (Ci]

pg . 17

The activity produced from the fuel plate breach is then dispersed throughout the immediate reactor facility air. The facility has a volume of 424 m 3 giving the activity per volume of air is given as Activity A/V=---

Volume 0.229 [Ci] 10 6 [µCi] _4 [µCi]

~ (A/V)i-i 3 s = 4.24 x 10 8 [cm 3 ]

• 1 [Ci] = S.40 x lO cm 3 In the event of a maximum hypothetical accident such as the one discussed, reactor facility workers would be required to identify that an alarm has occurred, notify others in the immediate vicinity to evacuate, and disable the ventilation system to attempt to contain as much of the fission products as possible . This process would take approximately one minute in a very conservative setting. The analysis of the exposure due to the failure of a fueled experiment in RAI 8 above showed the volumetric breathing rate is 333 cm 3 /sec and in a 60 second evacuation 3

scenario, the volume of air breathed will be 20,000 cm .

The total effective dose equivalent is calculated with the same method which is outlined in RAI 8.

CEDE CEDE Submersion Inhalation Percent Factor Factor Yield from Inhalation Submersion (mRem/hr)

Isotope (mRem/µCi) Fission Dose (mRem) µCt/cm 3 Dose (mRem) 1-131 32 .89 2.9% 160.87 67.41 0.989 1-132 0.38 4.3% 2.76 56.67 1.23 1-133 5.85 6.5% 64.09 108.89 3.58 1-134 0.13 8.0% 1.77 481.48 19.5 1-135 1.23 6.4% 13.26 295.56 9.57 Kr-85m 1.lOE-01 1.30% 0.242 4.89E+01 0.322 Kr-87 5.25E-01 2.50% 2.22 5.07E+02 6.42 Kr-88 1.33E+OO 3.60% 8.09 5.00E+02 9.11 Xe-131m 5.48E-03 2.90% 0.0268 1.79E+01 0.262 Xe-133 2.25E-05 6.50% 0.000246 1.84E+01 0.605 Xe-133m 1.99E-02 6.50% 0.218 3.85E+OO 0.127 Xe-135 1.73E-01 6.40% 1.87 1.16E+02 3.74 Xe-135m 2.79E-01 6.40% 3.01 1.10E+02 3.56 Total Total Inhalation 258.42 Submersion 59.0 Dose Dose TEDE 317.43 mRem pg. 18

Doses to persons outside the building will come from submersion in a cloud of re leased radionuclides and from radiation em itted from the reactor building. The dispersion factor is given as (x/Q) = 0.00765 [~ 3 ] from previous analyses and the exhaust is assumed to be closed giving a leakage rate from the reactor room of 0.02 m 3 /sec .

Similarly breaking down the dose rates into 15 minute increments, the table below shows the dose to a person in an unrestricted area in increments of 15 minutes.

Total Radio-nuclide TEDE Dose Time Step Concentration (µCi/cm 3 ) Received (mRem) 0-15 minutes 8.28 x 10- 7 0.728 15-30 minutes 5.42 x 10- 7 0.476 30-45 minutes 3.54 x 10- 7 0.312 45-60 minutes 2.32 x 10- 7 0.204 Total Dose in the first hour 1.72 mRem

12. RAI 12(b) in NRC letter dated August 29, 2014, stated:

(b) Dose conversion calculation using the Environmental Protection Agency's Federal Guidance report (FGR)-11 and FGR-12 dose conversion coefficients or another equivalent methodology to account for inhalation/ingestion and submersion exposures.

Your response to RAI 12(b) letter dated July 24, 2014, seems to provide occupational and public does values by combing thyroid inhalation and whole-body submerged dose estimates. The radiological consequences should include external and internal exposure and provide an estimate of the whole-body TEDE values for an occupational worker and a member of the public. The dose conversion calculations may use the Environmental Protection Agency's Federal Guidance Report (FGR)-11 and FGR-12 dose conversion coefficients or another equivalent methodology to account for inhalation/ingestion and submersion exposures.

Provide an updated analysis and discuss the potential maximum whole-body TEDE radiological dose estimate due to the MHA.

Please reference RAI 11 above for the inclusion of TEDE values in the analysis of a maximum hypothetical accident.

13. NUREG-1537, Part 1, Section 13.2 provides guidance for analyzing the MHA event including the evaluation of the potential radiological consequences. The radiological consequences should include external and internal exposure whole-body TEDE values for the duration of the accident.

The response to RAI 12(b) in your letter dated July 24, 2015, indicates that the dose analysis was based on several exposure periods. Provide the total radiological whole-body TEDE (thyroid dose, if more limiting) dose estimate to an occupational worker allowing for a realistic evacuation process and also to the maximally exposed member of the public considering any emergency evacuation plan.

pg . 19

Please reference RAI 11 above for the inclusion of TEDE estimates to an occupational worker allowing for a realistic evacuation process as well as the maximally exposed member of the public considering an emergency evacuation plan.

14. NUREG-1537, Partl, Section 13.2 provides guidance to licensees to systematically analyze and discuss credible accidents in each accident category. A postulated MHA even may result in gamma-ray radiation levels in the class room areas above the reactor room. Provide an analysis estimating the consequent maximum dose rates in the class room areas including accumulated doses to the maximally exposed member of the public considering procedures required by your Emergency Plan. The results should show compliance with the regulations in 10 CFR Part 20.

The PUR-1 facility is in the basement of the Electrical Engineering building on Purdue's campus.

In the event of a maximum hypothetical accident, it is conceivable that a member of the public was directly on the other side of a door adjacent to the reactor facility and could be exposed to the shine from the dispersed radionuclides originating from the failure of a fuel element. The table below shows the em itted radionuclides, their concentrations, the gamma energies of each nuclide, and the effective dose for a semi-infinite cloud as taken from FGR-12.

Gamma Concentration Energy Effective Dose Isotope (µCi/cm 3 ) (MeV) (mRem/hr) 1-131 2.45E-04 0.382 1.61E-09 1-132 3.63E-04 2.28 1.59E-08 1-133 5.48E-04 0.607 5.82E-09 1-134 6.75E-04 2.625 3.47E-08 1-135 5.40E-04 1.576 1.58E-08 Kr-85m 1.lOE-04 0.158 2.94E-10 Kr-87 2.llE-04 0.793 2.96E-09 Kr-88 3.04E-04 1.955 1.13E-08 Xe-131m 2.45E-04 0.02 7.87E-12 Xe-133 5.48E-04 0.046 2.64E-10 Xe-133m 5.48E-04 0.041 2.09E-10 Xe-135 5.40E-04 0.249 2.31E-09 Xe-135m 5.40E-04 0.429 4.00E-09 Total Effective Dose Equivalent 9.52 x 10-s (mRem/hr)

It is clear that even without taking benefit for the walls, doors, and other room features, the dose rate to a member of the public is far below the limits as set forth in 10 CFR Part 20. The dose to a member of the public from shine of a dispersed cloud in the reactor room is near zero.

15. RAI 14 in NRC letter dated august 29, 2014, stated:

pg. 20

10 CFR Part 20, "Standards for Protection against Radiation," provides the regulatory framework and NUREG-1537, Part 1, Section 13.1.3 provides the guidance for licensees to systematically analyze and discuss credible accidents in each accident category.

Section 13.1.3 of the updated PUR-1 SAR, describes the loss of coolant accident (LOCA) scenario. The updated PUR-1 SAR does not include an estimate for radiation levels in the reactor floor and the roof areas, due to the unshielded reactor core, after a postulated large LOCA event. The SAR should provide the consequent maximum dose rates at various locations on the reactor floor and outside on the reactor building roof.

In accordance with 10 CFR Part 20, provide the accumulated dose to the reactor building occupants and the maximally exposed member of the public, considering evacuation procedures and potential residence time for staff. In addition, provide an estimate when facility staff may enter the reactor building to start recovery operations.

Your response to RAI 14 by letter dated July 24, 2015, provides a dose rate estimate of 6.6 Rem/hr for a member of the public in the class room area. Provide an analysis and discuss the maximum radiological dose estimate due to the LOCA gamma-ray shine considering any evacuation procedure and potential residence time and demonstrating compliance with 10 CFR Part 20.

The response to this RAI is intended to supersede all previous Loss of Coolant Accident analyses.

The maximum possible conce ivable LOCA accident in the PUR-1 reactor would be that of having a hole or crack develop in the reactor pool li ner and the water dra ining from there. The only other option would be that of pumping or evaporating water from the top of the pool which would not be a bounding case.

The figure below shows the design of the tank (with dimensions in feet/inches) . Note that the stainless steel outer wall is made up of individual cylinders which are then welded together to for the full height tank. Below the tank is a fifteen inch thick, 3000 psi concrete footing which is anchored to the ground floor concrete approximately fourteen feet above. Surrounding the tank is a corrugated metal outer wall. The space between the tank wall and the outer sleeves is filled with compacted magnetite sand fill.

The most probable event initiating a LOCA would be that of having a crack develop between one of the welds on the wall of the pool wall. Supposing the crack was at the lowest level of the tank gives a conservative analysis as the water will drain the fastest and for the longest time. The crack in the tank wall would seep out to fill the small voids between sand particles. The density of sand is commonly cited to be somewhere between 1400 kg/m 3 and 2000 kg/m 3

• Assuming a conservative value of 1000 kg/m 3 (the same density as water), at least half of the available space for the water is taken by the sand. Additionally the flow rate will be greatly reduced because the tank is not draining to empty space but rather to and around the sand in-fill.

pg. 21

\o ' -

.

10 '- 0"0 1~c.

The flow rate out of a free standing tank is given by (Streeter et al., 1998, p. 467; Daugherty et al.,

1985, p. 413)

A tank with constant cross sectional geometry will also follow dh Q = Atank dt Integrating h from initial height to final height and solving for time pg . 22

An analysis is done for the most credible scenario of the formation of a crack at the joint between the tank wall and its floor. The circumference of the tank is 7.66 meters. If this crack were 1 mm w ide, the area of the hole would be Ahole = 0.001 [m]

• 7.66[m] = 7.66 X 10- 3 [m 2 ]

This area of a hole is equivalent to an instantaneous hole in the bottom of the reactor pool with a diameter of -lO[cm].

2 2 Ahole = rr ( 2d) = rr (0.10)

- -

2

=7.9 X 10 -3

[m 2 ]

The area of the tank is Atank = rr

• r 2 = rr * (121.92 2 ) = 4.67 m 2 t

4.67 /2

= 7.66 x 10-3 (~ - fZiJ * ~9:81 = 610 (~ -Fr)

As shown in the diagram above, there is a corrugated metal tank encasing the compacted magnetite sand fill. This will contain the water from the draining pool, further restricting the coolant and shield drain. The volume of the tank is approximately 24 m 3 . The lower portion of the corrugated tank has a volume of approximately 13 m 3 . Therefore, there are 11 m 3 of water remaining to fill up the upper, outer portion . The final height of the water will be 2.7 meters (5.1 fee t above the core) . The total t ime to reach this height is (z1 = 2.7 meter s) t = 610('-13.96-.../2.7) =210 seconds The dose rate in the air above the reactor pool is given by D = 4>(z)µ air The flux of photons is the primary question in determining the dose rate. The flux is essentially the attenuated dose at a distance r [cm] from the core, multiplied by a buildup factor.

Joo ~17(E)B(µr, 4>(z)

J o oo 4>(z, E)dE 0

S 4rrr E)e-µr dE A sample calculation for a 2 MeV gamma shielded by 395 cm of water follows . The photon attenuation coefficient is a function of energy and for water at 2 MeV, µ = 0.0494 L~]

(Attenuation data taken from the National Institute of Standards and Technology). For a shielding t hickness of 395 [cm), the quantity µR is 19.5 and represents the number of mean free paths the photon must travel through to be emitted through the top of the pool. Note that this is the dose directly above the reactor and would be further reduced at the edges of the pool as there is more attenuation.

pg. 23

The buildup factor B(µr, E) has multiple forms one of which was reported by J.J. Taylor and has the form of a summed exponential a second form is a cubic formula as shown below The appropriate equation to use for buildup is dictated by the fitted parameters and is shown in the Taylor paper. The coefficients A(E), a 1 (E), and a 2 (E) have no physical meaning and are evaluated in tables reported by multiple sources and the value for f3 is a function of the number of mean free paths (µr). For the 2 [MeV] photon emitted in 395 [cm] water, B(µr, E) = 22.082

• e4*21 xio- 2

• 27 *9 + [1 - 22.082]e- 6 xio- *27 *9 = 50.7 4

The number of fissions in the reactor is bounded by a 12 kW operating power plus 50% of power uncertainty. The number of fissions at full power is 1000 W 1{ 6.2415 x 1012 MeV 1 fission 14 fissions 18 kW= 18 kW* 1 kW *1W

• l]oule

• 200 MeV = 5 *617 x lO sec At 2 [MeV] , there are 1.8 [-.-Y-.-]. The 2 [MeV] flux is therefore fission 5 _617 x 1014 [fissions]

<1>(395 [cm], 2 [MeV]) = 4rr

• 3952

[

cm s:~

• 1.8 [risswn Y. ]

• 50.7

• e- 19.5

<1>(395, 2) = 89 [  ;

cm sec

]

The dose rate is then the product of this flux and the attenuation in air. This calculation is then done for every gamma energy at each water level to give the instantaneous dose rate. The dose rate at each water level is then integrated over time to give the total dose delivered to a worker and a member of the public.

Using this process to find the dose rate at the top of the reactor following the initiation of a LOCA event, the reactor will scram in 80 seconds. The operator would then follow up with a manual scram and evacuate the reactor room. This analysis assumes it takes the reactor operator 30 seconds to evacuate. This time frame is adequate for an operator to become aware of the situation, recognize a scram and begin an evacuation .

The figure below shows the total dose received by a member of the public and the radiation worker as a function of time after the LOCA. The primary source of the worker dose is following the loss of shielding but before the reactor is scrammed.

pg . 24

Total Dose Received After LOCA Event 0.60 ose Rat e Retu rns E

0.5

---***111:111111r -/1 o Background

~ 0.40

.§.

~ 0.30 0

• Worker Integral Dose Q

ii 0.20

• Pu blic Integral Dose

~

0.10 0 00 200 300 400 Time A ter OCA (seconds}

The calculated total dose to an operator is 0.49 mRem and the total dose to a member of the public is 0.15 mRem. These dose levels bound those that would be encountered during fuel handling operations as fuel remains at the bottom of the pool until it has cooled to levels which are acceptable within bounds of the TS. These dose levels are bounded those analyzed in the MHA.

16. TS 5.3: TS 5.3, Reactor Core and Fuel, "describes normal core configurations. TS 5.3.6 states "Representative fuel assemblies shall be inspected annually, with no interval to exceed 15 months." However there is no definition of "representative" within the TS. Further, although it is recommended by ANSI/ANS 15.1, there is no description of specifications that specifically describe requirements for inspection of the fuel as well and conditions for operating the reactor with damaged fuel elements or provide an explanation why this is not necessary.

Technical Specifications 5.3.g and 5.3 .h have been added to outline the definition of representative as well as actions to be taken following the identification of compromised fuel plates. Amendment #13 to the PUR-1 Tech Specs is enclosed for reference .

17. 10 CFR 50.36 (a)(l), Technical Specifications," states "A summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the technical specifications." Provide bases for the technical specifications in Section 5.

Please reference the revised Technical Specifications Amendment #13 for the addition of technical bases to Section 5.

pg. 25