ML20236S299
| ML20236S299 | |
| Person / Time | |
|---|---|
| Site: | 05000131 |
| Issue date: | 07/20/1998 |
| From: | Alexander Adams NRC (Affiliation Not Assigned) |
| To: | Claassen J DEPT. OF VETERANS AFFAIRS MEDICAL CENTER, OMAHA |
| References | |
| TAC-M88345, NUDOCS 9807240250 | |
| Download: ML20236S299 (13) | |
Text
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July 20, 1998' L
Mr.' John P. Classson-Reactor Manager Veterans Administration Medical Center 4101 Woolworth Avenue Omaha, Nebraska 68105
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION (TAC NO. M88345)
Dear Mr. Cleassen:
..We are continuing our review of your request for rene'wal of Facility Operating License
. No. R-57 for the Omaha Veterans Administration Medical Center TRIGA Research Reactor which you submitted on May'10,1993, as supplemented. During our review of your renewal request, questions have arisen for.which we require' additional information and clarification. Please provide responses to the enclosed Request for Additional Information within 90 days of the date of this letter. Following receipt of the additionalinformation, we will continue our evaluation of your request.'
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Lin accordance with 10 CFR 50.30(b), your response must be executed in a signed original J
under oath or affirmation. If you have any questions regarding this review, please contact
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me at (301).415-1127.
Sincerely,
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' ORIGINAL SIGNED BY:,
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Alexander Adams, Jr., Senior Project Manager
- Non-Power Reactors and Decommissioning Project Directorate Division of Reactor Program Management.
Office of Nuclear Reactor Regulation Docket No. 50-131 '
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July 20, 1998 Mr. John P. Claassen Reactor Manager Veterans Administration Medical Center 4101 Woolworth Avenue
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Omaha, Nebraska 68105
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION (TAC NO. M88345)
Dear Mr. Claassen:
We are continuing our review of your request for renewal of Facility Operating License No. R-57 for the Omaha Veterans Administration Medical Center TRIGA Research Reactor which you submitted on May 10,1993, as supplemented. During our review of your renewal request, questions have arisen for which we require additional information and clarification. Please provide responses to the enclosed Request for Additional information within 90 days of the date of this letter. Following receipt of the additionalinformation, we will continue our evaluation of your request.
In accordance with 10 CFR 50.30(b), your response must be executed in a signed original under oath or affirmation. If you have any questions regarding this review, please contact i.
me at (301) 415-1127.
Sincerely, WM Alexander Adams, Jr., Seni r P ject Manager Non-Power Reactors and De commissioning Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket No. 50-131
Enclosure:
As stated cc w/ enclosure:
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See next page
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I Veterans Administration Docket No. 50-131 Medical Center i
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Mayor l
City of Omaha Omaha, Nebraska 68102 l-Dr. Mark B. Horton, M.S.P.H., Director Nebraska Department of Health P. O. Box 950070 Lincoln, Nebraska 68509-5007 I
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REQUEST FOR ADDITIONAL INFORMATION OMAHA VETERANS ADMINISTRATION MEDICAL CENTER (OVAMC)
DOCKET NO. 50131 A.
Radiation Protection Program l
1.
Section 2; please provide a copy of an OVAMC dr,cument (letter, memo, l
etc.) that delegates ALARA responsibility to the reactor manager.
l 2.
Section 4.3.1; the table provided is very similar to Table 1 of Regulatory Guide 1.86, which was developed to apply to residual radiation levels upon license termination. However, the applicable criteria for disposal or release i
of components and materials from a licensed facility are provided in 10 CFR Part 20, subpart K. Please replace the table on pages 5 and 6 by applicable specific criteria for routine monitoring and unrestricted release (see Q #B.23).
3.
Section 5.2, page 7; the last sentence on this page seems to be incomplete.
Please address this.
4.
Section 5.3, page 8; are there any other general topics besides " Features for External Radiation Control?" For example, internal radiation control.
B.
Safety Analysis Report, December 17,1997 1
L 1.
Section 1.2: this section refers to reactivity insertions as large as $2.00 with resulting peak power of 250 kW. Should this be 250 MW? Please address.
1 2.
Section 2.1; please provide additional discussion about industrial and transportation activities that may impact the reactor site. Is there any nearby industrial activities that may pose a hazard to the site? Figure 2.2 shows a-rail line next to the site. If the line is active, please describe the types of cargo that pass the hospital. What plans are in place to deal with l
derailments? Are there any airports closer to the reactor than Offutt Air Force Base? Do any airways pass over the reactor site?
3.
Section 2.4.1; Groundwater Hydrology; j
(a) Please refer to Appendix D of the SAR, or explain why it might not be l
applicable.
(b) Page 2-12; is the Missouri River the only source of drinking water in the region? Are there any groundwater sources such as wells? Please discuss this. Can any surface water impact reactor operation at the site?
4.
Section 2.5, Earthquakes; in addition to historic information about earthquakes, please briefly discuss possible radiological consequences of an earthquake, including the results of the applicable sections of the accident analyses and appendices.
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2 5.
Section 2.6, Tornadoes; your data stops in 1962. Please consider adding l
more data to bring your information up to date. You may refer to NUREGICR-4461, " Tornado Climatology of the Contiguous United States."
6.
Section 3.2.1, Reactor Pit, last paragraph; please provide a basis or reference l
for the measured reactivity quoted for the loaded storage pits. Discuss the reactivity change due to flooding.
7.
Section 3.2.9, Reactor Water and Purification System; can a failure occur in the chiller that can result in the release of primary coolant to the environment? If so, what are the radiological consequences of such a leak?
l Please discuss how makeup water is supplied to the primary coolant system.
l-Can primary coolant be introduced into the city water supply by a malfunction?
l l-8.
Section 3.2.10.2.5, Effects of Fuel Aging; has the fuel currently in use at your reactor only been used in the VA reactor? If not, please describe the l
history of the fuel and why your discussion of fuel aging is applicable to those elements, i
l 9.
Section 3.2.12, Limiting Design Basis; because NUREG-0988 is an NRC
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evaluation of licensee-provided information, it is not an appropriate reference for primary information. It is recommended that you review and reference the most recent applicable primary publications, for example, your reference
- 2.
10.
Section 3.2.13, Dynamic Behavior of Reactor; (a) Page 3-23; second paragraph; please check the magnitude of the prompt neutron generation time; l
(b) Subsection a); this is the first time in the SAR that you have mentioned the phase-change of the zirconium hydride. Please discuss this 'n more detail, including which type of TRIGA fuelis affected, and revyw and make any appropriate changes to section 3.2.12 of this SAR. Because l
these sections contain your discussions of the OVAMC reactor " safety limit" they should provide substantial basis for your TS value.
11.
Figure 4.1, Block diagram of instrumentation; (a) Please show more explicitly where the key switch and the " manual scram" power interrupters are located; (b) Show where the " loss of high-voltage" scram devices are located.
12.
Table 4.1, Minimum reactor safety channels; please correct the spelling of
" analog" in the footnote.
13.
Section 4.2, second paragraph on page 4-4; because failure or malfunction of the recorder could lead to control rod withdrawal, please provide an analysis
3 of the resulting reactor transient initiated by such an accident (al) event in Chapter 8. Alternatively, you could propose a remedy that would eliminate the possibility of such an event.
14.
Section 4.3.1.1, Nuclear Instrumentation; subsection (1) mentions "hard-wired bar graph " What does this mean, how do you define "hard-wired," and discuss any safety or protective action implications. Also address the same issues in Section 4.3.1.2.
15.
Figure 4.2, Functional Diagram NM-1000; please provide a copy of this figure on which all printed information is readable.
i 16.
Section 4.3.1.2, Reactor Power Safety Channel, third paragraph; in the
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middle of this paragraph, you mention the " requirements for a scram channel
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for the TRIGA." Please state these requirements and discuss how the OVAMC system addresses them.
l 17.
Section 4.3.1.3, Internal Diagnostics, third sentence discusses the watchdog
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timer that will reset the NM-1000 software if it is not reset periodically.
Please explain. The second paragraph on page 4-9; please explain what is 4
meant by the phrase " key soft tasks."
18.
Section 4.3.2, Process Instrumentation:
f a) Subsection (1); please compare the quoted sensitivity with the likely concentration of fission products resulting from the postulated event analyzed in Appendix B, and discuss implications.
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b) Subsection (2); please discuss in more detail the nature of the
" experience" mentioned, including references.
c) Subsection (5)b.; please explain the statement that the radiciodine monitor is always operable while the NMC CAM is in operation, however,
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in case of failure of the NMC CAM it is not required for reactor operation.
What is not required for reactor operation? Please explain and justify.
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Please explain the meaning of the last term in this paragraph, and compare with section 7.4. Does it mean a concentration of 10~8 Ci/cm lasting for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, a concentration increasing up to 1 x 108 Ci/cm3 i
during a 24-hour time interval, or something else?
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19.
Section 5.2, Void coefficient, last sentence; please give the units of the 10%, and explain the safety implications.
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20.
Chapter 6, Conduct of Operations; a) NRC regulations also require such additional operational documents as the Emergency Plan, Operator Requalification Plan, and Physical Security Plan. This section of the SAR should at least mention them.
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4 b) The organization chart shows the Reactor Safeguards Committee. Please provide information about its function, charter, membership, etc.
c) The organization chart does not show the Radiation Safety Committee.
Please show its organizational location, and clarify its relationship to the Reactor Safeguards Committee.
d) Please discuss reactor staffing requirements including those evolutions that require the presence of a senior reactor operator.
e) Please discuss reporting requirements, reportable events, other reports to be submitted to NRC, and record retention.
21.
Please discuss and justify material possession limits that will be the possession limits in the reactor license.
22.
Section 7.1.6; Radioactive Waste; please discuss the storage, processing, and release from the facility of any liquids disposed into the specific sinks designated for radioactive materials. How do you ensure that only soluble materialis released to the sewer as required by 10 CFR 20.2003. How is the gaseous waste vented in the two laboratory fume hoods monitored to ensure compliance with regulations for release of radioactive material to the environment?
23.
Section 7.1.7, Other Radioactive Material; as noted above (O #A.2), Table 7.1 is not applicable to the release of radioactive or contaminated materials.
Neither is an exposure rate at one meter of 5 rnrem/hr. Please review 10 CFR Part 20, arid address these is',ues. In this regard, review the caveats and reminders of the following sections of ANS/ ANSI 15.11: (1) Forward; (2)
Section 5.3, last sentence; and (3) Section 6.2.3(4).
24.
Section 7.2.1, second paragraph; please provide the basis for the 500 disintegrations per minute, and discuss the action that it would initiate.
25.
Section 7.2.3, Management Surveillance; please describe how you ensure compliance with the requirements of 10 CFR 50.59 when evaluating experiments for the reactor. Please discuss and justify materiallimits on experiments and failure and malfunctions of experiments.
26.
Section 7.4, Evaluation of Monitoring Systems; the information provided here is still not clear. Please review this entire section, along with your response to question #13 of RAI #2. Ensure that there is consistency, and respond to the following:
a) Explain how Ar-41, a noble gas, can cause a buildup of counting rate in the continuous air monitors.
b) Give the units (or parameters) that apply to the " calculated efficiencies."
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5 c) Explain how a " count rate buildup of 180 cpm /hr is equivalent to an air concentration of 2.28 x 10~8uCi/cm'", a " count rate buildup of 180 cpm /hr is equivalent to an air concentration of 1.5 x 104 Ci/cm'," "NMC Recorder rise of 180 cpm /hr = 1 x 1040 Ci/cm " "Eberline Racorder rise of 185 cpm /hr = 1 x 10 ' Ci/cm'", and ".... 1 x 10' uCi/cm' is equivalent to CAM rise of greater than 150 crm in one hour on both of our CAMS" can all simultaneously be valid. The exact numerical values are not as important to us as is a clearer understanding of the principles of operation and use of the systems for both radioactive particulate and noble gases.
27.
Section 7.4.1, Particulate Air Monitor:
a) Table 7-2; this table is not entirely clear. Are quantities of fission products dependent on the history of reactor operations? How would the reactor operator use this table to evaluate increasing counting rate in a CAM?
b) Last paragraph; how would personnel exposures due to an accidental release from an irradiated sample be evaluated? How and why would averaging dc.es over a year be accomplished for an exposure of the order of a day or less?
28.
Section 7.5, page 7-10, second paragraph; please explain how you deduced the 200 r/hr at 6 ft from a single fuel element.
29.
Chapter 8, Accident Analysis a) Section 8.1, third and fourth paragraphs; these paragraphs intermingle the concepts of " credible," " Design Basis Accident," cladding failure within the pool water, and cladding failure in air in the reactor laboratory.
As a reminder, NRC has designated the failure of cladding of a single fuel element in air as the " maximum hypothetical accident" for TRIGA reactors. This replaces the concept of a DBA, still being beyond credible, yet not directly relevant to reactor " design" and maximum or " limiting" in relation to the radiologicalimpact of the facility on the public. Please try to include these concepts, without deleting the informative potential accident scenario you have also postulated of cladding failure within the pool water.
b) Section 8.1.1; (1) First paragraph; please correct the magnitude of uranium-zirconium hydride dissolution, and ensure that the subsequent calculations are based on the correct value. (2) Explain why you have not used the same mechanism in this scenario as used in the fuel handling scenario for releasing iodines (into the water). That is, the amount accumulated in the fuel-cladding gap. Which mechanism is predominant?
Explain. (3) Please compare the magnitudes of concentrations in water and air with the same parameters in your Appendix B, and correct the appropriate ones. (4) Please review all material in both Chapter 8 and the corresponding sections of Appendix B and correct the several
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6 typographical types of errors. Also, please ensure consistency between l
l Chapter 8 and Appendix B, when responding to specific questions about i
the appendices to your SAR (5) First paragraph on page 8-3; compare i
the stated corrosion rate referred to in question 29(b)(1) abova. (6) Last paragraph of section 8.1.1 on page 8-3; please review the statements about the concentration of 1.6 x 10 mci /cm-8 of water, and also discuss 7
the safety implications, and comparisons with Appendix B of l
i 30.
Section 8.1.3, Handling Irradiated Fuel; the comments and questions here may also apply to corresponding information in Appendix B.
l a) Because this analysis should treat the limiting postulated accident, the information of most relevance is failure "at shutdown."
b) Please discuss the operation of the ventilation / confinement system during accident conditions. Is this system an engineered safety feature?
c) Please ensure that changes to tables B-1 and B-2 are reflected in Chapter 8.
d) Because of the possibility that your appendices could become detached from the main body of the SAR, it is recommended that at least the calculated doses at important locations in the unrestricted area be included in Chapters 8 and 9.
C.
Appendix A 1.
The title of this appendix mentions only Argon, but section A.3 includes l
discussions of N-16. Please address this.
2.
Section A.1; paragraph 2 states that the calculations show that the Ar-41 decays while in the water. At the appropr; ate place in the calculations, this shou;d be noted and discussed because it seems to be an important point for low-power reactors.
3.
Page A-4, section A.1, first full para;raph: please justify using the concept of i
a small electric field gradient to estimate ion mobilities. Please discuss the alternative rationale of assuming the ions to be in thermal equilibrium in the water, and obtaining random speeds using a Boltzman fluid (gas) concept.
What other parameters of the water-argon system determine the rate of escape of the argon atoms from the water surface?
4.
Page A-5, Section A.1, list of variables; please confirm that these are correct, and address whether any errors, especially the argon-40 neutron absorption cross section, cause errors in the calculated results.
5.
Page A-5, Section A.1, middle of the page; mention is made of the inhalation i
DAC listed in 10 CFR Part 20, Appendix B. However, this appendix notes l
that Ar-41 values are derived for external dosage due to immersion in a o
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1 cloud. Please address this, and also verify the numerical values you have provided.
6.
Page A-6, Section A.1, list of variables; please review the definition, units, and the numerical value for the factor "g." Also, confirm that the numerical value for the linear absorption coefficient for Ar-41 gamma rays is correct.
(We raise this point again fater, in Appendix B.)
7.
Page A-6, Section A.2, last paragraph; (a) please give the distances for the MMP and NPR, and any approximations you used in conjunction with the nomograms in NUREG-0851 (vs 08457); (b) provide an estimate of the curie strength of the source of Ar-41 released to the unrestricted environment in an average, and in a maximum, operating year; (c) discuss why the projected doses for MMP and NPR seem to be larger than the projected doses to the most exposed worker (MEW); (d) discuss the air concentration of Ar-41 and what dilution factors or conditions you derived for the locations of MMP and NPR, and compare with the calculated Ar-41 concentration in the reactor l
room; (e) please clarify whether the listed doses are for a year, or for an hour. Also, discuss the reactor operating schedule assumed for the calculations.
8.
Section A.3, page A-8, next-to-last paragraph; the same concerns we have with your discussion of argon ion speeds in the water persist here with N-16.
Please reconsider this, as you address question #3 above, and treat escape of Ar-41 and N-16 by similar methods, as appropriate.
9.
Section A.3, page A-9; please review the concept and units of equation #23.
Please make necessary changes and discuss the effects on the calculated doses above the pool related to N-16, and the conclusion you have reached at the end of section A.3.
10.
Section A.4, No.1, page A-12, equation 28; upon substitution of numbers into the equation, it seems that the last term in the numerator was not correctly substituted. Please address this, and discuss whether the result, equation 29, is correct as shown.
11.
Section A.4, No.1, page A-13, first paragraph; the scenario and method for calce!ating the quantity of Ar-41 due to pneumatic tube (PT) operation seem l
to be fundamentally flawed. The amount of Ar-41 produced and released should be primarily dependent on the duration of reactor operation per year, not on the duration of PT blower operation. The method you have used I
could be valid if the reactor were started up and operated only during the time the blower is on. The method used ignores the Ar-41 accumulated in l
the stagnant air in the PT with the reactor operating but the PT blower not I
operating. This Ar-41 is also forced out each time the blower is turned on, and could be much larger in quantity than the Ar-41 produced during the 30-second operation. A more realistic scenario could assume continuous reactor operation for 6 or 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> per day, for 250 days per year, with 10 to 12 PT 30-second flushes per day. It will be found that the number of flushes is much more relevant than their duration. Please revise this section l
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on the PT, and also the discussion and analyses applicable to the rotarf specimen rack, as appropriate.
12.
Section A-4; while revising this section of Appendix A, please verify all l
numerical quantities, for example the volume of the PT air subjected to the l
quoted neutron flux density and the volume of the rotary specimen system.
13.
The calculation of Ar-41 concentration in the air exhaust streams should not l
neglect the initial concentration as it exits the irradiation facility. Please be i
sure that factor is included in the discussion.
14.
Section A.4, No. 2, page A-14, mid-page; please review the units and values for the " concentration of Ar-41, exhausted to the outside and in the room,"
and make any necessary changes.
15.
Section A.4, No. 2, page A-15 equation (32); (a) please review the units and values of the parameters substituted into this equation; (b) please provide applicable additionalinformation about MMP and NPR, as requested in question 7 about PT operation. Revise Table #1 (page A-17) as appropriate.
D.
Appendix B i
The following comments and questions pertain to Appendix B. It is appropriate to analyze both the clad failure in the pool, and the clad failure in air in the reactor room. However, the limiting case should be treated with most detail and discussion.
This has usually been found to be the latter, which has been designated as the maximum hypothetical accident (MHA) for TRIGA-fueled reactors.
1.
Tables 8-1 and B-2; unless you can justify using two different fission-product inventories, please choose only the most applicable, giving the basis.
2.
Please review the method used to derive fission product inventories from i
referenced publications. Have you verified their validity?
3.
Review and address the assumptions that inventories of relatively short-lived fission products are proportional to integrated energy production (i.e.,
megawatt days) (compare with NUREG/CR-2387) rather than to reactor power level or neutron flux density. Refer to Tables B-1, B-2, and B-3, and correct as necessary. Please review whether inventories of relatively long-lived fission products have been derived appropriately.
4.
Because clad failure could occur at times much shorter than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, that scenario should be downplayed unless you provide justification.
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Page B-1; please clarify whether Blomeke and Todd performed these calculations for you, or you used their publication as a basis for your calculations.
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For both clad failure in water and in air, please address the fate and radiological effects of any airbome radionuclides, including escape or release to the unrestricted environment.
7.
In calculating the consequences of an accident event, unless you can justify l
estimating annual average radionuclides concentrations, annual doses, j
long-term meandering winds, etc., please use methods appropriate to a j
short-term release, such as two hours, or until the transient cloud passes the l
locations of interest, if shorter. See the OVAMC SAR, Appendix B,1995.
The offsite locations of interest should still include the most exposed person in the unrestricted area, and the nearest permanent residence.
8.
For the summary on the bottom of page B-6; (a) give results for the accident event under consideration, not for a year; (b) discuss what gase.s other than iodine and noble gases (if any) are being considered, and the mechanism by which they impart dose; (c) give the values of the relevant parameters leading to whole body dose, for example, radionuclides dispersion and diffusion, radionuclides concentrations as a function of position or location, size and shape of the plume considered, duration of the postulated exposure, l
etc.
9.
Please review all of Appendix B for typo, grammatical, and other errors.
l There are several that make it difficult to evaluate. For example: (a) review units and values for "g;" (b) capital "M" usually stands for "mega;" (c) compare values for gamma ray attenuation coefficients for the gamma ray energy assumed for fission products with the ones assumed for the Ar-41 gamma rays.
E.
Appendix C 1.
Appendix C, page C-1; please use more specific and quantitative terms in place of the subjective undefined " conservative" and " optimistic," and justify the last sentence on page C-1.
2.
Appendix C, page C-2; (a) please provide a drawing showing the geometry and distances for both of these calculations; (b) please correct the errors, typo and otherwise, in the equations on this page; (c) justify assuming one Mev photons in this appendix, and one-half Mev photons in Appendix B; (d) please use consistent nomenclature and symbols where applicable on pages C-1, C-2, and C-3.
3.
Appendix C; please discuss or reference your discussion of the temperature rise in fuel elements following rapid loss of pool water and of pool cooling (loss of coolant accident [LOCA] and loss of coolant flow [LOCF]).
4.
Table C-1, page C-1; after making changes / corrections required by the comments above, please change or reconfirm the values in this table, as applicable. Please discuss the likely uncertainties or probable errors of the dose values in this table.
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Appendix D 1.
The scenario you have postulated, calculating the volume of water required to dilute the calculated radionuclides down to 10 CFR Part 20 release concentrations is difficult to visualize and evaluate. Instead of reducing the M
quant ies of radionuclides to units of fractions of a curie, would it be reasonaLM to use the concentration of each radionuclides in the average volume of soil adjacent to the reactor vessel, and assuming that adding water does not increase. but only decreases the concentration further, deduce the amount of external water necessary to attain 10 CFR Part 20 concentration limits. Please discuss this, or propose a more realistic scenario.
2.
Also, please discuss briefly the likely amount by which the results of this scenario are over-estimates of the possible radiological consequences to the public.
G.
Technical Specifications Additional questions about the facility Technical Specifications, dated March 1,1995, will be discussed during our site visit.
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