ML20248D796

Forwards Request for Addl Info Re 880819 Application for Renewal of License R-120.Response Requested within 60 Days. Requested Info Includes Operator Training & Meteorological Data

ML20248D796
Person / Time
Site:	North Carolina State University
Issue date:	09/20/1989
From:	Alexander Adams Office of Nuclear Reactor Regulation
To:	Poulton B North Carolina State University, RALEIGH, NC
References
NUDOCS 8910040405
Download: ML20248D796 (14)

Text

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• September 20,.1989 s- ..

Docket No. 50-297 p

Dr. Bruce R. Poulton Chancellor North Carolina State University Box 7001 Raleigh, North Carolina 27695-7001

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION

'We are continuing our review of the documentation submitted in support of your application for renewal of your operating . license for the North Carolina 88. State

. University PULSTAR Research Reactor that was submitted on August 19, 19

'During our review of your submittal, questions have Pleasearisen for which provide we a response require additional information and clarification.

'to the enclosed Request for Additional Information within 60 days of the date of this letter. Following receipt of the additional information If you havewe anywill

, continue our evaluation of-your renewal application.

questions regarding this review, please contact me at (301) 492-1121.

The reporting and/or recordkeeping requirements contained in under P. L.96-511.

Sincerely,

/s/

Alexander Adams, Jr., Project Manager Non-Power Reactor, Decommissioning and Environmental Project Directorate Division of Reactor Projects III, IV, Y and Special Projects Office of Nuclear Reactor Regulation

Enclosure:

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[;E September 20,-1989

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. Dbcket No. 50-297-Dr., Bruce R..Poulton Chancellor.

North' Carolina State University Box,7001 . .

Raleigh, North Carolina 27695-7001

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Dear Dr. Poulton:

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SUBJECT:

REQUEST.FOR ADDITIONAL INFORMATION

- We are continuing our review of the documentation submitted in support of your application for renewa_1 of your operating license for the North Carolina State

" University PULSTAR:Research Reactor that was submitted on August 19, 1988.

During_ our review of your- submittal, questions have arisen for which we require additional information and clarification. Please provide a response to the enclosed Request for Additional Information within 60 days of the date of this letter. .Following receipt of the additional information we will

-continue our evaluation of your renewal application. If you have any questions regarding this review, please contact me at'(301) 492-1121.

The reporting and/or recordkeeping requirements contained in this letter affect fewer than ten respondents; therefore; OMB clearance is not required under P. L.96-511.

Sincarely, I

$8m+& O .

Alexander Adams, Jr., Pr

• t Manager Non-Power Reactor, Decommissioning and Environmental Project Directorate Division of Reactor Projects III, IV, Y and Special Projects Office of Nuclear Reactor Regulation

Enclosure:

As stated cc w/ enclosure:

See next page V

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I. North Carolina State University- Docket No. 50-297 cc: Office of Intergovernmental Relations 116 West Jones Street Raleigh, North Carolina 27603 Dr. Paul J. Turinsky, Head t ;- Department of- Nuclear Engineering North Carolina State University Box 7909 Raleigh, North Carolina' 27695-7909 l

Mr. Garry.D. Miller Nuclear Reactor Program-Department of Nuclear Engineering

. North Carolina State University L.' P. O. Box 7909-Raleigh, North ~ Carolina 27695-7909 l.

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L. ENCLOSURE l I l

REQUEST FOR ADDITIONAL INFORMATION ,

NORTH CAROLINA STATE UNIVERSITY DOCKET NO. 50-297 SAFETY ANALYSIS REPORT 1.. Section 1.3.2 Licensees must comply with 10 CFR Part 20, it is not only a guide.

Furthermore, current practice is also to implement an ALARA program, and you are expected to address that in all radiological considerations. Please address this comment.

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2. Please clarify in your SAR what design features you have considered to be Engineered Safety Features.

3. Section 1.4.4 What are the abnormal conditions referred to in this criterion and how is ,

control rod insertion accomplished?

4. Section 1.4.7, Criterion 7 10 CFR Part 100 is not applicable to research reactors. Because there is no comparable regulation for research reactors, NRC expects you to compare with 10 CFR Part 20, and the ICRP recommendations on which it is based, with ALARA considerations. Please address this in all applicable parts of your SAR.

S. Section 1.4.11, Criterion 11 You are requested to discuss in quantitative terms, in the appropriate sections of your SAR just what is monitored, where, by what instruments, etc. Discuss instrument efficiency, ranges, calibration methods and

. frequencies. Discuss radiation monitoring in event of accidents.

6. Table 1.1 Table 1.1 indicates some inconsistencies; e.g., on page 5 you say that ,

only a 5x5 array of fuel elements is considered, but the Table 1.1 )

footnote says a "5x4 Standard Core." However, the number of pins is 625.

There are several other inconsistencies; e.g. WNYNRC is no longer its name, height of pellets of BMRC is wrong, number of pins in BMRC is not I consistent with fuel element dimensions, Hp 0/UO 2 ratios seem inconsistent, capture-to-fission rate, perhaps others. Please edit and correct. I

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7. . Section 2.'1.3.1, 2.1.3.2 Please discuss (or reference) in more detail the directions and. distances from the reactor to the nearest permanent residences, the nearest.

occupied campus. building, the nearest' occupied student housing, etc., and the quantitative. radiological implications. 'What is the approximate

. population of Raleigh and the Raleigh metropolitan area?

8.- What. is'the location (birection and distance) of the ' closest in use railroad'line to the facility? What volume.of rail traffic does this line carry? What plans does the. University have in pince to respond to a railroad accident? - To what extent would these plans involve the' reactor-L f acility? '

9. 'Please provide infonnation on the reactor room wall thickness and details :

L of roof aesign. How many penetrations does the reactor room have and how

'is air. leakage through these penetrations controlled?

10.- All of Chapter 2 -

-Please discuss the' quantitative radiological implications of this chapter, and why-there is deemed to be no substantial reason that the reactor location does not-continue to be acceptable. If you take credit for building characteristics not yet discussed in detail in your SAR, give cross references to the appropriate sections, e.g. the consequences if an

- earthquake were to occur.

11. Section 2.2 Please describe your method for obtaining meteorological data in the event of an accident that' involves releases. If this involves interaction with an agency outside of the University, letters of agreement should be provided.

-12. -Seetion 2.2 l' Please provide updated meteorological data, for example, section 2.2.1.3

. discusses the time period 1916.to 1958.

13. Section 2.3.2

' Discuss drainage between the reactor site and Lakes Raleigh, Johnson, and 1 .Bonson.-

14.- Section 3.1.1 The pulsing achieved doesn't seem consistent with information in Table 1.1.

Because of possible implications to inadvertent transient accidents, please discuss. One of the objectives of an FSAR is to compare actual observations with initially projected characteristics. Please do that.

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, ] - 15. Section 3.1.2.1

~ (a) 'Please' analyze.the adequacy and accuracy of a shutdown margin of

.004 delta k/k.

(b) Now that the Pulse Mode-is no longer to be used, the role of the so-called pulse rod is changed. Explain its role in the definition

. of shutdown margin. ~ Justify the steps taken to " disable pulsing capability."

16; J Section 3.2.1-

" Vacant core positions"~are mentioned. Please be 2..re specific and provide the analyses necessary to justify such use. Include in the analysis the accidental. dropping of- a fuel assembly into the highest

- worth vacant core position possible.

.17. Section 3.2.1

Please provide discussion of the effects of the following: heat capacity, thermal diffusivity, retention of fission products, " benefits" of sintered pellets, physical supports that prevent bowing, relative core sizes of SPERT and NCSU PULSTAR (or relative power and energy densities).

18. Section 3.2.2.2 Explain why a failure of the Pneumatic System cannot syphon water out of the pool.

19. Section 3.2.2.5

=(a) Please give.more justification for gang withdrawal of control rods.

It is more conservative and realistic to assume the maximum rate of reactivity insertion not the average permitted by techhEal specifications. 4 (b) Are there credible actions by which withdrawal speeds might increase?

20. Section 3.2.2.6  !

Please discuss the design ar.d use of irradiation brackets.

21. Section 3.2.2.7 i Please provide additional explanation of the 75 MW-sec energy release j statement.

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22. Section 3.2.2.7, page 18 (a) Please explain better the relationship of. strength of zircaloy-2 at-

.1600*F and at room temperature, and the conclusion stated in the last sentence.-

(b) Pleasediscusstheeffectsofradiationexposure(damage), corrosion /

-erosion of:the cladding, and pellet growth due to burn-up on the' analysis and conclusions. Discuss your fuel conditions now and for the next 20 years.

23. Section 3.2.3.5.1.3' Please explain the reasons:and the implications of calculations and measure-ments being different, bottom of page 25.

24. (a)'LPlease discuss the method of control rod fabrication. Discuss methods used to assess integrity of the control rod cladding.

(b) - Discuss .the " safety feature" limiting repositioning control rods.

25.- Section 3.2.3.5.3 (all sections)

(a). Please discuss differences between NCSil and BMRC, and between predictions and actual measurements.

(b) 3.2.3.5.3.5 Please explain .whether these fuel assembly worths are for fresh fuel, and explain the changes.

26. .Section 3.2.4.1.1, page 33, Steady State (a) Arecriteria(2).and(3)alsoforforcedconvectiveflow? Discuss.

(b) An additional design basis should be no fuel melt in case of LOCA or any credible transient, or other accident. Please address this.

27. Section 3.2.4.1.3.1, pages 35 and 36 (a)- Fuel loading and pin spacing; discuss whether these variations were found to exist in the as-built reactor, and implications.

(b) Instrument Error: is this 7 percent some sort of limit, root-mean-square, or what? How is it verified?

28. Sect 1on 3.2.4.1.3.2., page 38 Please discuss your reasoning that limiting IO N release is the fundamental

- reason for no bulk boiling in the reactor core.

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.29.fSection 3.2.4.1.3.2, page 39 and 40

.Please provide'a quantitative discussion of the " chimney effect."

30. . Section 3.2.4.2.3, page 45

'(a) 'Was onset of DNB observed at 58 MW-sec?

(b) NRC would take a' dim view of planning to operate with DNB. Discuss probable.DNB ratio.

31. Table'3.2 (a) " Savings" for-a H2O reflector is given. Please compare with a graphite reflector.

(b) What fraction of the quoted temperature coefficient of -3.9 pcm/*F is prompt?

(c) .How is the void coefficient characterized? Is the value listed a core average, or is it for a specific core location? Please discuss.

.(d) K,ff-core (1) Does " rods out" mean all rods out including the pulse rod?

-(2) What is the maximum variation in actual U-235 burnup among individual fuel assemblies?

32. Table 3-4 Please discuss the reasons for differences between predicted and measured vc1d effects. What implications are applicable to other reactivity calculations performed with the same methods and computer programs?

33. ~Section 4.2.3 Discuss whether methods to prevent or avoid corrosion were fully successful. Discuss implications for the next twenty years.

34. Section 4.2.1 l- ,

Under what conditions is the water pressure higher in the primary side or i secondary side of the heat exchanger? Discuss in relation to possible leakage across the barrier, and control of radioactivity. If leakage to the secondary side is possible, discuss activity and isotopes that could be released to the environment.

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, 35. Fig. 4.1D What method is used to prevent primary coolant from entering the water supply system of the engineering building under any pressure conditions or valve leakages?

36. Section 4.2.3~

Provide a quantitative discussion of the statement in the last sentence of Section 4.2.3 about fuel assembly leakage.

37. Section 5.2.1, 5.2.2 Please provide a detailed quantitative discussion of the levels of radiation that trigger events, and their bases. . Please state and justify any assumptions used.

38. Section 5.2.2 What indication does the operator have that dampers are fully closed?

.Will the operator know if the damper sticks while closing?

39. Section 5.2.2 Can the dampers close automatically without any power assistance?

40. Section 6.1 Is there an indication to the operator that the flapper is fully open?

Is it possible to operate the reactor above 150 Kw with the flapper not fully closed?

41. Section 7.1.1 Please provide the ranges and uses of the period meters in your control )

console. j

42. Section 7.1.1 l What is the function of the "high voltage supply," and what is the purpose of the interlock? j j

43. Section 7.4.1, What are the actions that are permitted and/or inhibited by the various )

positions of the Gang Drive Switch?

44. Section 7.4.2 Page 19, Paragraph 1 How is " Low Shutdown Margin" monitored and interpreted by the instruments?

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. 45. Table 7-1 Please be certain that all of the protective channel settings are appropriately included in your technical specifications.

(1) Items 5 and 6: - text seems to say 1.2 MW, not 1.3 MW.

(2) Are items 10 and 11 correlated?

(3) Items 12 and 13: text seems to say 116*F.

46. Chapter 8 What area or effluent monitors are on auxiliary electric power? Discuss reasons.

47. Section 9.1.4 How is the water level in the liquid waste holding tanks determined?

48. Chapter 9 Please discuss and describe your fire protection program and provisions, including training and maintenance.

49.- Seetion 10.1.2 Please provide a more quantitative discussion of releases and dilution of liquid wastes before and after release.

50. Section 10.1.3 Additional parts of 10 CFR apply to waste disposal other than Part 20.

Discuss your reason for only citing 10 CFR Part 20.

51. Section 10.1.4

! Chapter 5, this section of Chapter 10, and your environmental report does not provide sufficient treatment of 4I Ar and 16 N. Please provide a detailed, I quantitative analysis.

52. Section 10.2.1 Please provide a quantitative discussion of the radiological conditions associated with operating at full power with 14 ft. 2 in. of water above the core.

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53. Chapter 10 Describe in more detail the organizational structure of the personnel protection organization, and discuss qualifications, training, origin of procedures, relationship to reactor manager, etc.

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I- 54. Chapter 10 Please provide a summary of the radiation exposure histories at the NCSU Reactor Facilities, in a format outlined in 10 CFR 20.407(b).

55. Section 11.1 and its subparts Please provide a docunent that establishes an ALARA philosophy and policy in the University to assure compliance by all staff and users of the reactor facilities.

56. Section 11.2.1 and 11.2.4 Discuss the need for reactor operator training to meet the requirements of 10 CFR Part 55.

57. Section 13.2.1.1 For your 4I Ar releases, please provide a detailed quantitative analysis of potential exposures and doses to personnel in unrestricted areas, including the most-exposed offsite permanent residents and any affected campus buildings. Please include the dispersion /dif fusion factors you have used in the analysis. Also address potential exposures in the reactor room.

Clearly state and justify any assumptions used.

58. Section 13.2.1.2.

What is the affect of a failure of the flapper to open in the loss of flow accident?

59. Section 13.2.1.3 Please provide a quantitative analysis of waterlogging, because inadvertent transients are credible events.

60. Section 13.2.1.4 Please describe and discuss in more detail the penetrations in the pool liner, and the primary coolant system by which pool drainage could occur accidentally. Include an anlysis of a primary pipe failure that can not be isolated.

61.. Section 13.2.1.4 (a) Please provide more details about your use of the LPTR loss of coolant data in application to PULSTAR fuel. Are there more recent and more relevant data?

(b) Please give more details about the direct and scattered radiation above the pool and near (or in) the control room. Have you included back-scattering from the roof? Where is the 250 mrem per hour location?

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-62.- Section 13.2.1.5~

What.would be the results of your fuel pin failure analysis assuming the maximum fuel burnup that may occur during the relicensing period?

.63. Section-13.2.1.5-Fuel Pin Clad Failure

-(a) What is the source'of the information in Table 13-1 and 13-27 Discuss and compare with TID 14844,-for example. What percentages of various radiologically important radionuclides migrate to and accumulate in the pellet-clad gap?'-

(b) . Instead 'of- MPC, please convert these values into the likely doses to the most exposed person in the_ unrestricted area for the event you.

-have postulated. Also, discuss potential exposures to staff within the reactor room. State these doses for the entire event for both whole bocty and thyroid. . Clearly state and justify any assumptions used.

(c) Page 9 Please justify the use of 97 percent retention.

(d)'Page13 Please provide the details of your assertions, especially for the sentence beginning: "An eight hour average ...." Clearly state and

-justify any assumptions used.

64. Seetion.13.2.2.1 How is accidential removal or ejection of a' control rod precluded?

65. Appendix 3c, pages 9-12 Please provide a similar analysis with the following initial conditions:

(a) Fuel at the normal temperatures for 1 MW steady state operation, and 400 gpm coolant flow. and (b) Fuel at the normal temperatures for either 250 kW, or at ambient water temperature, but no coolant flow.

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'66. -Page 2, Section 2, Paragraph 2 Give a quantitative comparison between the two cooling towers:-

(a) physical dimensions

'(b) heat-dissipation capacity

.(c) use of chemicals to control corrosion and residue accumulation.

67. Page 3, Section 2, Paragraph 2 If not all of the radioactive gaseous effluent is Ar-41, what else is there? Give quantitative answers and discuss.

68. Page 4, Section 4, Paragraph 2 Give quantitative discussion of. the changes in radioactive effluents when changing from. air to nitrogen purging.

69. Page 5, Paragraph 1 It is stated that federal regulations are met. Please provide the concentrations of radioactivity released, and discuss your ALARA considerations. Quantities may be expressible as MPC-hrs / year.

'70. Page 6, Section 6-Discuss-storage and disposition of spent fuel.

71. You take credit for both 10,000 cfm and 12,500 cfm airflows. Please

' discuss what provisions assure that both are operating when required.

TECHNICAL SPECIFICATIONS

72. Please submit revised technical specifications based upon the discussions that occurred during our site visit.

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, OPERATOR REQUALIFICATION PROGRAM a

73. Section 8  !

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

Please clarify which of the control manipulations addressed in 10 CFR 55.59(c)(3)(1) apply to your. facility and include this list in Section 8.

74. Section 9 The requirements.of 10 CFR 55.59(a)(2)(11) 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. 1 comprehensive sample of items specified in 10 CFR 55.45(a)(2) through (13). inclusive to the extent applicable to the facility. Please clarify which of these items are applicable to your facility and include them in the list of items to be included in the annual operating examination.

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