ML20247B950

Forwards Preliminary Questions & Comments Re Univ of Arizona Application for Renewal of OL for Triga Reactor Facility. Visit to Facility Planned for 890502 to Discuss Application & to Increase Familiarity W/Facility

ML20247B950
Person / Time
Site:	05000113
Issue date:	03/21/1989
From:	Michaels T Office of Nuclear Reactor Regulation
To:	Nelson G ARIZONA, UNIV. OF, TUCSON, AZ
References
NUDOCS 8903300082
Download: ML20247B950 (24)

Text

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March ~21, 1989 I h ~ Docket No. : 50-113 . -  !

lDr. Geor.ge W.) Nelson, Director.

' Nuclear Reactor Laboratory -

F . University'of Arizona '

a. Tucson; Arizona 85721  !

Dear Dr.:

Nelson:

SUBJECT:

. REVIEW OF LICENSE RENEWAL APPLICATION l I The NRC staff and 'our contractor, the Idaho National Engineering Laboratory, ,

are continuing.our. review of the documentation submitted in support of your I application for.. renewal of your operating ~. license for the University of Arizona . l TRIGA Reactor' Facility. . We have also'. planned a two or three day visit to your 3

reactor . facility beginning on May 2,1989 to discuss'your application and to "

increase.our familiarity with your facility.

During our visit we want also to d'iscuss the information indicated by the enclosed set of Preliminary Questions and Comments. Responses to these ..  ;

. questions should not be submitted formally; instead, the information should be. j made available in draft form, as appropriate, for discussions while we are at.  !

your reactor facility.- Following our return we will-develop and send you a j r . formal set of questions and request written responses.for our files and review. j i

., LIf you have any questions, please call me. at (301) 492-1102. I Sincerely, I

/s/ l 8903300082 890321 g Theodore S. Michaels, Project Manager l PDR ADOCK 05000113 , Standardization and Non-Power  ;

P PNU P.cactor Project Directorate h

Division of Reactor Projects - III,.IV, Y and Special Projects  ;

Office of Nuclear Reactor Regulation i Enclosure As stated cc: See next page DISTRIBUTION L_

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Docket No. 50-113 Dr. George W. Nelson, Director Nuclear Reactor Laboratory i University of Arizona Tucson, Arizona 85721

SUBJECT:

REVIEW OF LICENSE RENEWAL APPLICATION The NRC staff and our contractor, the Idaho National Engineering Laboratory, are continuing our review of the documentation submitted in support of your application for renewal of your operating license for the University of Arizona TRIGA Reactor Facility. We have also planned a two or three day visit to your reactor facility beginning on May 2,1989 to discuss your application and to increase our familiarity with your facility.

During our visit we want also to discuss the information indicated by the enclosed set of Preliminary Questions and Comments. Responses to these questions should not be submitted formally; instead, the information should be made available in draft form, as appropriate, for discussions while we are at your reactor facility. Following our return we will develop and send you a formal set of questions and request written responses for our files and review.

If you have any questions, please call me at (301) 492-1102.

Sincerely, 5)

Theodore S. Michaels, Project Manager Standardization and Non-Power Reactor Project Directorate Division of Reactor Projects - III, IV, V and Special Projects Office of Nuclear Reactor Regulation

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Enclosure As stated cc: See next page

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,G-University of Arizona Docket No. 50-113 cc: Office of the Mayor P. O. Box 27210

-Tucson, Arizona 85726-7210 ,

' Arizona Radiation Regulatory 4-Agency .

4814 S. 40; Street Phoenix, Arizona 85040 j

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e PRELIMINARY QUESTIONS AND COMMENTS UNIVERSITY OF ARIZONA LICENSE RENEWAL I. SAR Questions

1. How close is the nearest residence to the reactor and where is it located?

2. Is there any heavy industry in the vicinity of the University of Arizona campus and if so, where is it located? How far is Interstate 10, Southern' Pacific and AMTRAK from the reactor?

3. The hydrology section, section D, pg. 7, should contain information that quantifies the flood potential at the site. For example, provide the maximum historic flood elevation on nearby streams and compare j them to the elevation of the site. Provide judgements as to whether or not there is any potential to flood the site. If there is a potential to flood the site and reactor. building or other safety related equipment, then discuss the effects of flooding, such as the concentration of radionuclides at the secure boundary and how they compare to 10 CFR Part 20 limits.

4. When the UARR is operating at 100 Kw and the trip set point is at 110%, it is not set below the licensed power level. This is not in accordance with the licens.e conditions which limits your power to 100 Kw. There are at least four options which can be used to correct this condition as follows:

a. Operate the reactor at a reduced power level (e.g. 90% of licented power limit) sufficiently beinw the licensed power level to allow:

(i) normal operation and (ii) testing of reactor trip setpoints without exceeding power levels specified in the existing license condition.

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b. Submit justification for-'a power. increase (typically about ten percent) that would permit reactor trip testing and power level setpoint calibration without exceeding the amended license condition.

c. Modify the reactor protection system circuitry to accommodate-electrical simulation of the overpower trip channel signal.

The modification is required only for those functions:of the reactor protection system for which reactor trip credit is taken in the licensee's safety analysis report.

. d. Modify the neutron detector position system to allow repeatable detector or moderator position changes to vary the neutron flux by a predetermined amount to initiate a reactor trip and verify calibration of the overpower trip setting. This method, while acceptable, is not preferred by the staff.

Other options will be considered by the staff. Please be prepared to discuss the option you propose to choose.

5. What is the mass of U-235 in the standard fuel element and the control rod follower element? Does the transient rod have a fuel follower?

6. What is the maximum and minimum fuel element worth in the UARR and where in the core are they located?

7. What is placed in the vacant (unfueled) core positions?

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8. Wnat is the maximum Technical Specification allowed core excess reactivity? (Core excess reactivity must be established in order to define shutdown margin.)

9. Control Rods:

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l (a) What is the maximum ok/k/sec? Which rod would produce tho maximum?

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• p (b.) -What is the length and diameter of the poison section?

(c). Does the control rod or the transient rod run in a guide tube?

(d) Vertical travel (distance) of the control rods?

(e) Where is the control rod and transient rod position disp' layed?

10. Has the Thermal Column described on page 9 ever been installed.in the reactor tank?

11. What is the value of peff and the neutron lifetime.

12. What is the maximum pulse allowed? The Safety Analysis states 2.44$; while the Technical Specifications seem to allow 2.50$.

13. .Please describe in detail the use of the transient rod during normal (nonpulse) operations. Please include. its position in the core, its use as a control rod, the rod speed, and scram capability. Does the transient rod fall back into the core upon loss of electric power?

Note: If the TR does not respond quickly to emergencies, shutdown margin concerns will arise.

14. Is there a period scram? If not, what limits period in manual control?

15. What is the gallon per minute (GPM) rating of the purification system pump?

16. How is the purification system filter checked for radioactivity, and what dispor,al method is used?

17. Are there any credible paths for reactor pool water to get into the campus water or the sewer system?

18. Please describe the monitorir.g and disposal of the resins in the ion exchanger.

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19. 'What is the technical or safety reason for the 45'C pool water temperature limit? Is there an alarm when pool water temperature reaches 45 C? How is the water temperature measured? Where is the water temperature displayed?

20. Is reactor start-up always in the manual mode?

21. Are there any emergency power supplies? If so, what system do they power. If not, what procedures do you follow in case of a loss of off-site power?

22. Please provide a description of the operation of the ventilation system including:

(a) The flow rate into and out of the reactor, control, and storage rooms.

(b) The location of.the 500, 1000, and 1250 cfm fans.

(c) The direction of airflow during normal operations in the reactor room. What is the relative pressure in the reactor room compared to the surrounding rooms?

(d) The ability to isolate the reactor room.

(e) The room used for experiment handling. Does it have separate ventilation, e.g., a fume hood?

(f) The possibility of a (partial failure of the ventilation system causing air to flow from thc reactor room into the control room or other roorts in the Engineering Bui? ding.

23. Fire Protection j (a) How does fire department receive an alarm?

(b) What radiation training do the firemen receive?

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24. Please describe the Communication System in Reactor Laboratory and Engineering Building.

25.' Fuel Storage (a) How many 30 position racks are in the reactor pool?

(b) How many 13 position holsters are in the reactor, pool?

(c) Is fuel stored in the " LOCKED PITS" in the reactor room floor?

If fuel is stored in the area, what limits are placed on this area and what would the K,ff value be.if flooding occurs?

26. What is the Void worth of the Lazy Susan and the Central Thimble?

27. Please describe ~the process at the UARR facility including the handling of solid and liquid waste.

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28. Please provide a description of the ALARA program at the UARR 1acility. Also, a definite ALARA statement signed by high University officials is necessary.

29. Lhat plans have been made regarding the ultimate disposal of the LARR TRIGA fuel?

30. Argon-41 (a) Assuming zero escape of the AR-41 from the pool water in your analysis is nonconservative, assume a conservative and defend-able release fraction and recalculate dose rates in the restricted and unrestricted areas.

(b) What are the setpoints for the radiation alarms?

(c) What restrictions are there on students being next to the Engineering Building near the exhaust fan when the reactor is operating?

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(d) What is the predicted AR-41 exposure t'o the nearest residence?

(e) What results have you obtained from monitoring for Ar-41 in the reactor room air during and following long power runs?

31. How many fixed radiation and constant air monitors are in'use at the UARR?

(a) What type?

(b) Where are'they located?

(c) Is their readout in the control room?

(d) Are there audible or visual alarms?

(e) Is there an alarm in the Reactor Room on the Exhaust? If so, does it automatically secure the ventilation system?

.32. Table 4.1 EIS:

(a) In 1983-84 one person received a total dose in the range 101-500 mR, what was the source of this radiation?

l (b) Please provide,the'information requested in the Table in 10 CFR 20.407 for the years 1984, 85, 86, 87, and 1988.

33. What is the natural background radiation level in the University of Arizona area?

34. The NRC position, on the basis of the Columbia University hearings, is that the maximum hypothetical accident for a TRIGA is the instan-taneous failure in air and consequent release into the reactor room air of all the fission products in th9 fuel element gap immediately following operation at the maximum authorized power level of sufficient length for all these fission products to reach their saturated activity 03/15/89 6 ARIZONA LICENSE RENEWAL

levels. Assuming this scenario and using defendable conservative techniques:

(a) Calculate the Whele-body Immersion Dose and the Thyroid-Committed Dose in the restricted area over a time span sufficient to evacuate this restricted area (1 to 5 min) and (b) Calculate the same doses for the public exposure immediately outside the restricted area over a time span sufficient to evacuate this area (1 to 2 hr).

35. What precautions are in place to restrict access to the reactor room-during operations?

36. Describe the surveillance requirements on the mechanical block and positioning switches to assure the transient rod cannot add more than 2.50$.

37. The Safety analysis states that a minimum of 14 ft of water above the core is required.

(a) What is the technical basis for 14 ft?

(b) Is this a Technical Specification requirement?

(c) Is there a low water level alarm?

38. Please discuss the potential for a LOCA at the UARR including any way water could be pumped out of the vessel and catastrophic vessel failure. Also justify 50 gallons per minute resulting in core uncovery in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

39. What is the maximum fuel temperature during the UARR LOCA?

40. Assuming an instantaneous loss of vessel water:

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4 (a) What is maximum fuel temperature?

(b) What is resultant dose rate, i.e., what would Table 7.1 values-be at 10 sec.?

41. Page 11, last paragraph Give details of the history of the used stainless-steel-clad elements. When were they new, where were they used, how were they used- pulses, steady-state power level, any history of failures from among the " lot?"

42. Page 24,'first paragraph Give an upper limit (realistic and quantitative) of the radioactivity in the Freon.

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43. Page 28, bottom of page

.(a) On what channel and what detector is the LSSS for pulsing set?

What limitation on pulse size or integrated pulse energy is imposed by the LSSS?

44. Fast Irradiation Facility (fir):

l (a) What are the use limitations?

l l (b) What is the effect of flooding?

(c) If flooded, what is the effect on neighboring fuel during pulsing?

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45. Page 35, Section C.

Regarding the use of neutron radiography tube, please discuss the radiation conditions (personnel protection), control of access, both 03/15/89 8 ARIZONA LICENSE RENEWAL

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How is shielding integrity assured?

when in use and when secured.

L Is there an Interlock?

46. Page 36, Section D Please discuss quantitatively the reactivity changes in the control thimble during flooding. If flooded, what is the cffect on neighboring fuel, during maximum pulses? Discuss any limitations on the type of experiments permitted including fueled experiments.

47. Page 37, top paragraph Discuss criteria and bases for use of the demountable fuel element.'

48. Pages 39 and 40 Control system; are there any instrumented fuel elements? If fuel I

temperature is a safety limit', how is it protected? Discuss for both pulse and steady state operations.

49. Page 44-Please explain the safety reasons for the pulse mode permissive.

50. Page 50, last line Should the word " minimum" be replaced by the ward " maximum"?

51. Page 51 and top of page 52 Please justify the implication that peak-to-average fuel temperatures and peak-to-average power densities are equal. Please compare with peak-to-average neutron flux densities in your reactor.

Compare with GA provided ratios, and discuss.

52. This SAR presumably becomes the definitive document for your facility, perhaps for the next 20 years. Please provide a more complete list 03/15/89 9 ARIZONA LICENSE RENEWAL

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[ of, primary references on'which the operating characteristics and.the' l , . safety considerations are based. It is' recommended that you review the proceedings of the hearings.for the~ Columbia. University TRIGA if that'has not-been done recently..

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. PRELIMINARY. QUESTIONS AND COMMENTS.

UNIVERSITY OF ARIZONA LICENSE RENEWAL II. . Technical Specifications

1. Page 3, definition for Time. Interval; Please revise to read as follows:

Time Interval - The average over any extender' period for each surveillance time interval shall be'the normal surveillance time, e.g. for two year interval the average shall be two years.

a) Biennially - at two year intervals (interval not ' exceed 30 months) b) Annually - at one year intervals (interval not to exceed 15 months) c) Semiannually'- at 6-month intervals (interval not to exceed seven and'one-half months) d) Quarterly - at 3 month intervals (interval not to exceed four months) e) Monthly - at one-month intervals (interval not to exceed six weeks).

Any extension of these intervals shall be occasional and for a valid reason and shall not affect the average as defined.

2. Page 5, Specification 2.1, Safety limit (1000 C not kW); NRC has accepted GA's proposal that " fuel temperature" be the safety limit, and all other TRIGAs use that parameter. Please review and address this. You have already relied on that parameter in your bases and you do have an instrumented fuel element which could be used for pro-tective purposes. Also, notice that GA proposed a different temperature in the absence of water (see e.g. Simmad, et al, Nuclear Technology 28, p. 31, 1976).

3. Page 6, Specification 2.2; Because the stated objective is to limit the fuel temperature, this is consistent with fuel temperature being 03/15/89 11 ARIZONA LICENSE RENEWAL

.the " safety limit," as noted above. Secondly, the LSSS should be on a system directly related to that, namely fuel temperature ~and you do have an instrumented fuel element which could be used for' I protective purposes. Please address these comments.

4. Page 7, Limiting Condition for Operation (LCO); because of various conditions presumed or assumed in your SAR, you should have LCOs on the following:

(a) Coolant Conductivity (b) Coolant pH (c) Coolant Temperature (d) Coolant Height Above Core (e) Maximum excess reactivity (2.70$ ?) (Is this consistent w/SAR?)

5. Page 7, LCO (a) 'In your previous Technical Specifications, 3.1f also included a  !

provision limiting the ramp or oscillating rod to <10$/sec.

Why has this been omitted from the present Technical Specification?

(b) Please provide a basis for Specification 3.li.

(c) Please state the UARR maximum allowed core excess reactivity (under any conditions).

6. Page 11, Table; Please address the following:

I (a) Why the reactor period channel in the flux regulator is not required to be operable and read out at the console.

l (b) Why the area radiation monitors are not clearly specified to be l observable by the operator at the console.

l (c) Why the fuel temperature is not required to be read out at the console in all modes.

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7. Page 12, Table; Please aadress the following:

(a) Why a Fuel Temperature Scram is not' required.

(b) Why a reactor period scram and interlock is not required.

(c) Why the scram setpoints'are not specified.

(d) Why the Safety Analysis Report addresses ten scram functions Land the Technical Specification only addresses four of these.

Furthermore, the stated objective should include assurances that the LCOs and the license are not exceeded.

8. Page 13, Bases; Please explain the meaning of the last line of the last sentence.

9. Page 14, Specification 3.6, Ventilation System (a) From the Safety Analysis, the Argon-41 dose rate in the unrestricted area depends on blower operating at 1250 cfm.

This should be a technical specification requirement.

10. Page 15; Specification a; Please address the following:

. (a) Your Specifications 3.1, 3.7, and 6.8 relating to experiments, should all be in one section. Show the analysis and justification for the values in 3.7(a) and 6.8(d).

(b) What are the values for I-131 through I-135. Are they 1.5 curies or 1.5 milli Ci?

(c) A fueled experiment worth 3$ would contain a large amount of fissile material. Show that during the maximum planned or inadvertent pulse no credible mechanism (e.g. self heating) exists which could cause the experiment to fail.

11. Specification b; For potentially explosive materials you must have more limiting requirements than this. Typical TS for this statement are as follows: "Known explosive materials shall not be irradiated in the reactor in quantities greater than 25 milligrams. In 03/15/89 13 ARIZONA LICENSE RENEWAL

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addition, the pressure produced in the experiment container upon-detonation of the explosive shall have been determined experimentally, or by' calculations, to be less than the design pressure of the container." If this statement is acceptable, please modify TS accordingly.

12. Page 16, (a) What'is'the justit; ration for Specification 4.1.a?

(b) Specification b; Please justify the 1/4-inch elongation limit for stainless steel fuel and compare with gas recommended elongation limit.

(c) There is no provision for periodic inspection of your fuel.

Because of its projected age before the requested license extension would expire, please provide an inspection speci- ,

fication. Consider the following proposal to your Technical Specifications:

4.1d. All fuel elements shall be removed from the core and visually inspected for evidence of deterioration of cladding, including at least corrosion, erosion, wear, cracking, welds integrity at least once every five years. (This could include inspecting a few every year, not necessarily all at one shut-down.)

13. Page 19; Are the setpoints verified by radiation test sources?

14. Page 20; Specifications 4.5 should be more spe d fic; for example:

Specifications Any additions, modifications, or maintenance to the ventilation system, the core and its associated support structure, the pool or its penetrations, the pool coolant system, the rod drive mechanism, 03/15/89 14 ARIZONA LICENSE RENEWAL L

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l or,the reactor safety system shall be made and tested in accordance l-with the specifications to which the systems were originally-l designed and fabricated or to specifications approved by the Reactor Safety Committee. A system shall not be considered operable until after it is successfully tested. A licensed reactor operator shall be present during maintenance of.the reactor control and safety system.

Bases This specification relates to changes.in reactor systems which could directly affect the safety of the reactor. As long as changes or replacements to'these systems continue to meet the original design specifications, then it can be assumed that they meet.the presently accepted operating criteria.

If this statement is acceptable, please modify TS accordingly.

15. It is suggested that two' additional surveillance specifications be added; for example:

4.6 Coolant Conditions The pH and conductivity shall be verified to be within specified limits at least weekly.

4.7 Ventilation System i Operation of the ventilation system within the specified limits shall be verified at least monthly.

16. Page 21; Specification 5.la. Enrichment is to be <20%, see 10 CFR 73 l

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~1 7. Page 22; Specification 5.2c Since the basis for the Technical Specification is the SAR, where is it.shown that release at a point 12 ft above ground level is acceptable?-

18. Page 23; Specification 5.3; Please show us your analysis and L justification for an 800 C temperature for both a fuel element and a fueled device. Describe the " device".

p 19. Page 24, Specification 6.1 (a) 6.lb. The reactor facility must be operated within all applicable NRC regulations, which always takes precedence over the license in case of a conflict.

(b) 6.1d. Please be more specific about the locations of the operator and the second person. As presently stated, disagreements could arise.

20. Page 25, Specification 6.2.c; Reference to the "present Reactor Committee" is not sufficiently descriptive. Should state 1) who besides Health Physicist should be on committee and 2) Committee members should be competent in the field of reactor operations, radiation science or reactor / radiation engineering.

21. Page 26, Specification 6.3.2, ALARA; Please specify at what level of management this is to be established, and provide us with a copy of the directive that does require it.  ;

22. Page 30, Specification 6.7a; change to read as follows:

(a) A report within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and telegraph to the U.S.

NRC Region V.

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(b) Specification 6.7a.3; change.to read as follows:

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'Any reportable occurrences as defined in Section 1.0 (Reportable Occurrence) of these specifications in writing.  !

(c) ' Specification 6.7b; change to read as follows:

A report within ten days (in writing to the U.S. NRC Region V, with copies to the Director'of Nuclear Reactor Regulation, U.S.

NRC, Washington, D.C. 20555) of:

(d) Specification 6.7.b.3; change ~to read as follows:

Any reportable occurrences as defined in Section 1.0 (Reportable Occurrence) of these specifications; l

(e) Add a new specification 6.7.b.5 to read as follows.

0 Any accidental offsite release of radioactivity.above permissible ' limits, whether or not the release resulted in property damage, personal injury, or exposure.

(f) Specification 6.7c; change to read as follows:

A report within 30 days (in writing to the U.S. NRC Region V with copies to the Director of Nuclear Reactor Regulation, U.S. )

NRC, Washington, D.C. 20555) of:

23. (a) Page 31 Specification 6.7d; change to read as follows:

A report within 60 days after completion of startup testing of the reactor (in writing to the U.S. NRC, Region V, with copies to the Director of Nuclear Reactor Regulation, Washington, D.C.

20555) upon receipt of......

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i (b) Specification 6.7a; ' change to read as 'follows: jl

.An annual report within 60 days following the 30th of June each {

year (in writing to the U.S. NRC, Region V,'with. copies to the )

Director of Nuclear Reactor Regulation, Washington, D.C. 20555):-

providing the following--information:

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s Preliminary Questions and Comments

. University of Arizona License Renewal III. Emergency Plan-The review of the emergency plan indicates that the proposed changes in two areas decrease the effectiveness of the currently approved emergency plan as follows:

1) One proposed revision eliminates the designation of a specific location for the emergency support center (ESC). The proposed plan states that "A room on the first floor shall serve as the ... ESC".

The proposed plan revision provides no justification for the elimi-nation of the former ESC nor does it discuss who has the authority to designate an ESC or when an ESC should be designated. The effective-ness of emergency response could be decreased by a delay necessitated by the need to designate an ESC. You should predesignated a specific .

i location for the ESC so that prompt emergency response can be assured.

The emergency kit should be in the ESC or easily accessible to the ESC.

2) The second proposed revision would change the frequency of operation checks for monitoring instruments from quarterly to semiannually and of calibration from semiannually to annually. The frequencies of operation checks and calibration of health physics instruments in the current plan are in accordance with accepted practices in the industry and should be maintained.

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2 Preliminary Questions and Comments University of Arizona License Renewal IV. Operator and Senior Operator Requalification

Background

By letter dated October 17, 1988, the University of Arizona (the Licensee) requested that the staff review the Licensee's operator requalification program. While the program meets the criteria of ANSI /ANS 15.4, " Selection and Training of Personnel for Research Reactors," it does not meet '.he criteria of 10 CFR 55.59 "Requalification". Our review and request for additional information follows.

Section 1 Written Examinations (1.5)

The requirements of 10 CFR 55.59 (a)(2)(i) state that the annual written examination sample the items specified in 10 CFR 55.41 and 10 CFR 55.43 to the extent applicable to its facility. Licensee must clarify which of these items are applicable to its facility and include this list of items

.to.be included in the annual written examination.

Operating Test (1.4)

The requirements of 10 CFR 55.59 (a)(2)(ii) state that operators and senior operators must be examined such that they demonstrate an under-standing of the ability to perform the actions necessary to accomplish a comprehensive sample of items specified in 10 CFR 55.45 (a)(2) through (13) inclusive to the extent applicable to the facility. Licensee must clarify which of these items are applicable to its facility and include them in the list of items to be included in the annual operating examination.

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Reactor Operations-The requirements of 10 CFR 55.59(c)(3) state'that the requalification-program must incidde on-the-job training addressing control manipulations.

.' Licensee must clarify which of the control manipulations addressed in 10 CFR 55.59_ (c)(3)(i) apply to its-facility and include this list in the section on Reactor Operations.

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