ML20010H058
| ML20010H058 | |
| Person / Time | |
|---|---|
| Site: | Armed Forces Radiobiology Research Institute |
| Issue date: | 09/02/1981 |
| From: | John Miller Office of Nuclear Reactor Regulation |
| To: | Tyler P DEFENSE, DEPT. OF, DEFENSE NUCLEAR AGENCY |
| References | |
| NUDOCS 8109230112 | |
| Download: ML20010H058 (8) | |
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SEP 2 1981 O
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]7,, u.s,Ddgren Captain Paul E. Tyler, Director Armed Forces Radiobiology Research e
Institute s[
Defense Nuclear Agency s,
Bethesda, Maryland 20014 y1
Dear Captain Tyler:
By latter dated July 10, 1981 you transmitted to the Nuclear Regulatory Commission a Safety Analysis Report (SAR) for your TRIGA reactor which you operate under NRO license R-84.
In accordance with your request, we are reviewing this report in connection with your application for renewal of your facility operating license.
Questions have arisen for which we require answers before we can complete that review. Therefore, please provide written responses to the enclosed list of questions within 15 days of receipt of this letter.
If you have any questions concerning this matte.
please contact your Project Managep, Robert E. Carter on (301) 492-8106.
Sincerely.
ORIGINAL SIGygg g James R. Miller, Chief Standardization & Special Projects Branch Division of Licensing
Enclosure:
List of Questions DISTRIBUTION:
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SEP 2 1981 Arbid F; ces Radiobi31ogy 'Research Institute cc w/ enclosure Director Department of State Robert L. Brittigan, Esq.
Planning General Counsel 301 West Preston Street Defense Nuclear Agency Baltimore, Maryland Washington, D.C.
20305
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Director
~ Elizabeth B. Entwisle, Esq.
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Department of Natural Resources 8118 Hartford Avenue Power Plant Siting Program Silver Spring, MD 20910 Energy & Coastal Zone Administration
- Tawes State Office Building Atomic Safety and Licensing Board o
- Annapolis,sMaryland 21401 Panel
- U.S. Nuclear Regulatory Commission The Honorable Michael D. Barnes' Washington, D.C.
20555 Congress of the United States
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House of Representatives Atomic Safety and Licensing Appeal 1607 Longworth House Office Building Panel (5)*
s Washington, D.C.
20515 U.S. Nuclear Regulatory Commission Washington, D.C.
20555 County Executive
. Montgomery County Government Docketing and Services Section (7)*
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Rockville, Maryland 20850 Office of the Secretary U.S. Nuclear Regulatory Commission Louis'J. Carter, Esq., Chairman Washington, D.C.
20555 Administrative Judge Atomic Safety and Licensing Board '
Major Ronald Smoker 23 Wiltshire Road Armed Forces Radiobiology Research Philadelphia, PA 19151 Institute National Naval Medical Center Mr. Ernest E. l!ill Bethesda, Maryland 20014 Administrative Judge Lawrence Livermore Laboratory University of California P. O. Box 808, L-123 Livermore, CA 94550 Dr. David R. Schink
-Administrative Judge Department of Oceanography Texas A&M University College Station, TX 77840
- Through deposit in the Nuclear Regulatory Commission's internal mail system.
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l QUESTIONS ON THE AFRRI SAR A
MAY ~ 12, 1981 i
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- 1. ' Describe the radiation protection staff.
Identify the number,. level, responsibilities of personnel, and the lines of communication.
2.- Outline the minimum qualifications (training and/or previous experience) for each of your Health-Physics-related positions.
3; - Describe any Health Physics training for non-Health Physics staff.
If c
possible, provide a -topic outline of the course, indicating the level and duration of each course.
4.
Summarize your general radiation safety procedures.
Identify the minimum frequency of survey, action points, and appropriate responses.
5.
Describe the program to ensure that personnel radiation exposure and r -
releases of radioactive material are maintained at a level that is "as low as reasonably achievable" (ALARA).
6.
For your fixed-position radiation monitors. nas monitor, and effluent particulate monitor, specify the type of irs and their efficiericies and the operable ranges.
7.
For.the fixed-position monitors, describe the methods and frequency of
- instrument calibrations and the routine operational checks.
8.
For monitors that are alarmed, specify the alarm set-points and indicate the staff response to each alarm.
9.
Iuentify the type, number, and operable range of all of the portable Heal th Physics instrumentr.
Specify the frequency and methods of calibration.
10.
If you anticipate that additional or specialized instrumentation may be readily available from other NNMC facilities, indicate the type, number, and range of the available equipnent.
11.
Describe your personnel monitoring program, including bio-assay and in 1
vivo counting, if used.
12.
Identify any adeinistrative exposure limits and the anticipated actions if these levels are exceeded. Also, identify the operf tional constraints that are placea on personnel entering potential radiation /hich radiation or. contaminated areas.
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- 13. What is the Health Physics review and exposure control of one-of-a-kind, short-term, low-to intennediate-risk tasks, such as simple but nonroutine maintenance activities and one-shot experimental measurements? If Special Work Permits (SWPs) are used for these events, please discuss the applicable requirements, limitations, and approvals.
- 14. Are the check valves in the purification and cooling system regularly checked for leakage?
15.-
Provide a summary of the AFRRI annual whole body radiation exposures (the number of persons receiving total annual exposure within the designated exposure intervals; that is, less than 0.5 rem, 0.5 to 1.0 rem, and so on) for past 10 years of operation.
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- 16. What type of neutron dosimeter is used for personnel dosimetry? Describe briefly the neutron calibration of the dosimeter.
- 17. What checks are made during the readout of personnel dosimeters to assure s
the proper operation of read-out device?
18.
Describe the radwaste handling systems and the applicable procedures, including methods for monitoring and measuring activity levels in radioactive waste before final disposal.
- 19. What are the technical specifications of the conductivity cells used to monitor the water demineralizer?
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- 20. What functional tests are conducted on the reactor room ventilation
-isolation system and the overpressure relief damper? How frequently are these tests performed?
21.
Describe the procedures for monitoring and changing the filters in the ventilation aad water purification systems.
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22.
List and discuss the facility components that are on emergency back-up power.
Describe the test and maintenance schedule for the back-up power supplies.
L 23.
How many used fuel elements are in the pool storage racks? Wha t problems will arise if the water level drops? If there is a total loss of coolant, how will the fuel elements be moved to repair the
-leak? Include the estimated exposures of operations personnel.
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,. 24. - Describe the techniques'and instrumentation used to = insure the accurate determination of the reactor power level in both steady-state and pulse modes. - Include the calibration methods and the frequency of calibration in the description.
- 25. How is the reactor tank constant measured?
- 26. Consider the overhead crane used in the regular operations.
a.
How often and by whom is it proof loaded?
b.
What is the safety factor between the proof loading o
and (1) the workind load rating? (2) the maximum anticipated load?
- 27. Are there formal plans to place instrumentation in the ventilation system so that draw-fan failures can be readily detected? If so, describe; if not, justify.
28.
Describe the' techniques and procedures used to document reactor facility configuration, and changes in configuration.
- 29. Are the thermocouple channels that are used to measure the fuel elsnent temperature calibrated? How often are they calibrated?
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- 30. Under what' conditions is operation of the N diffuser system required?
- 31. Are the CCTV cameras routinely used for surveillance of the exposure rooms before reactor operation? Explain
- 32. What water, soil, and vegetation samples are taken from the area surrounding the liquid waste tanks?
- 33. What measurements are made to determine that the reactor areas exhausted into the AFRR1 stack are maintained at a negative pressure?
(See Fig.
'3-5) 34.
Describe the procedures and techniques used to prevent an inadvertent release of the contents of the radioactive liquid waste tanks to the public sewer.
35.
Provide details of the communication system (s) linking the operator in the control room with the personnel operating the access doors to the exposure rooms, s
- 36. How do you insure that any let kage frem the primary coolir.g water circulating and demineralizing system goes into the hot water drain system?
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-37.
How are the reactor rcom exhaust dampers checked for leakage when they are closed?
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38.
How are the ' warm" and " hot" liquid waste tanks protected from corrosion?
.-Who is responsible for the operation, inspection, and service of. this Are the warm waste and hot waste storage tanks routinely system?3 inspected-for leaks? For structural integrity?
39.
Consider. the effluent air u.anitoring systems.
a.
Describe the air particulate sampling probe in the reactor effluent line.
b.
What is the flow rate in the effluent line at this. point and what is the sampling rate?
c.
Describe _the gas sampling probe at the top of the effluent stack.
-40.
Are all of the operable thermocouples in each instrumented fuel element tied into the scram logic? Describe this portion of the scram system in detail.
The following questions are based on Chapter 6 of the 1981 Safety Analysis Report.
3 41.
P. 6-6.
You obtain a value of 0.08 heV/(cm.s) for S. We believe that it g
3 should be 0.18 MeV/(cm.s).
Explain and verify the subsequent computations.
42.
P. C-8 Table 2.
For postulated critical excursions the Regulatory
. Guides recommend the consideration of more nuclides than you consider in your postulated accident. Justify your reasons for considering only the nuclides which yoa have.
- 43. ' P. 6-13, Para. 3.
Justify the assumption that 99.8% of the iodine
-dissolves in the water?
44.
P. 6-16. Ea. 3.
Conversion factors are needed to yield " rad."
-4 45.
P. 6-17, Eq. 4.
If BR = 3.47 x 10 m s is the breathing rate for
-Standard Man, it should be so s+3ted.
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46.
P. ti-7 last paragraph. How did you obtain the N activity of 1.2 x 10~
pCi/cmb? Are the units correct?
47, Please provide a copy of the document stating the administrative policy
' that requires that a fuel element not be removed from fhe pool for at least 2 weeks followiaq its use in the core.
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l 48.
P. 6-3, last sentence. Have there been experiments to check the 550' C " maximum' average" fuel temperature for insertion of 2.8% ak/k?
How can you be certain that the fuel temperature will not exceed the technical. specifications limiting safety system setting of 600 C for fuel temperature?
!49. "P.
6-5, Eq..l.
It is stated that Eq. I gives the activity production assuming a saturation condition. We believe that the equation is o
incorrect.
Explain the presence of t in the numerator following E '
a and clarify the meaning of " saturation condition."
50.
P. 6-7, Dara. 2.
We'believe that the macroscopic 160 (n,p)l6N cross J
-section gf oxygen in water is incorrect. We believe that it is 6.2 x 10- 7 cm- '.
Please verify this and the following calculations that depend on the cross section.
Furthermore, we remind you that he reaction is not prodeced by thermal neutrons.
- 51. ~ P. 6-10, last sentence.
Define thermal ratcheting.
52.
P. 6-2, Para. 2.
Please explain the significar.:e of the distance of "25 meters from the AFRRI facility." What is the specific definition used in this document for the "AFRRI facility"?
53.
P. 6-3, Sec. 6.2.2, Para. 2.
Unless justified, please use consistent nomenclature for reactivity changes.
- 54.. P. 6-3, Sec. 6. 2.2, Pa ra. 3.
We believe that General Dynamics divested themselves of General Atomics before 1975.
Please change as appropriate.
55.
P. 6-3, S ec. 6.2.2, Para. 4.
Please clarify the meaning of "a maximum average fuel temperature."
If this means the maximum temperature in an average fuel element, please justify why you have not evaluated the maximum fuel temperature at any location in the reactor core.
56.
P. 6-4, Sec. 6.2.3, Para. 1.
Please clarify the meaning of "slightly cri tical."
57.
P. 6-5, Sec. 6. 2.4, Pa ra. 2.
What fuel loading is assumed in order to obtain 1 x 1013 for the average thermal leutron flux density?
I obtain 0.6 x 1013 for your core.
Please explain.
58.
P. 6-5, last paragraph.
It is stated that the assumed Yonditions represent "a highly conservative assumption for this facility". Please explain why you consider these conditions to be " highly caservative",
and give a quantitative estimate of the magnitude.
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59.
P. 6-9, first line.
Please justify the use of a distance of 25 meters from the "AFRRI. facility."
60.
P. 6-9,~ Sec. 6.3.1, Para.1.
Please discuss the radiological consequences of the demineralizer resins collecting all of the radionuclides from a
- fuel element cladding. failure.
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61 '. P 6-9, Sec. 6.3.1, Para. 3 last line. "Reac tivi ty" presumably refers to chemical reactions. Since " reactivity" is also a nuclear term, it is sugge-ted that you clarify.
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62.
P. 6-10, Sec. 6.3.2.1, last sentence. Please ciarify the meaning of
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"previously used aluminum cladding".
63.
P. 6-11, Sec. 6.3.2.1, last sentence. Since the prior operating conditions were not given for the examples of failed clad fuel elements, it is not clear that the conclusion "would not constitute an undue risk..." is warranted under all conditions ~ of possible operations.
Please justify.
s Are these measurements consistent with the computations, paragraph one, page 6-13? - Discuss.
64.
P. 6-12, Sec. 6.3.2.2, Calculated Fission Product Inventory, Para.1, Ultimate and penultimate sentences; please explain the sigrificance to this computation of the terms " prompt fission" and " prompt inventories".
3 65.
P. 6-15,. Sec. 6.3.3, last paragraph.
Please justify using a distance of 20 meters from the "AFRRI facility".
- 66. iP. 6-19, reference 2.
In addition to the inter'nal General Atomics document, the following reference treats much of the same information, and in addition was published in a publicly available referead journal:
M. T. Simnad, F. C. Faushee, and G. B. West, Fuel Elements for Pulsed TRIGA Research Reactors, Nuclear Technology, 28, 31, (1976).
67.
P. 6-12, Sec. 6.3.2.2, Determination of Fission Products in Gap, Para. 2.
Please give the technical justification for using "0.1 percent gap activity" in the subsequent analyses.
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