ML20245G839

Informs of Forthcoming Visit to Facility Re Application for Renewal of License.Forwards Preliminary Questions to Be Discussed During Visit

ML20245G839
Person / Time
Site:	North Carolina State University
Issue date:	04/25/1989
From:	Admas A Office of Nuclear Reactor Regulation
To:	Poulton B North Carolina State University, RALEIGH, NC
References
NUDOCS 8905030246
Download: ML20245G839 (26)

Text

, _ _.

3

. 'o , * .

April 2S, 1989 k

' Docket No. 50-297 Dr. Bruce R. Poulton Chancellor North Carolina State University Box 7001 Raleigh, Horth Carolina 27695-7001

SUBJECT:

REVIEW 0F LICENSE RENEWAL APPLICATION The Nuclear Regulatory Commission 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 North Carolina State University PUSTAR Research Reactor. We plan to conduct a visit to your reactor facility to discuss your application and to increase our familiarity with your facility. It is anticipated that a total of four people will participate in the visit. The date of our visit will be scheduled with the reactor staff.

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

Please review the preliminary questions and contact me at (301) 492-1121 for that a visit date can be finalized.

Sincerely,

/s/

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

Enclosure:

As stated cc: See next page i g DISTRIBUTION:

3 Docket File; EJordan AAdams NRC & Local PDRs BGrimes PDSNP R/F OGC Ellylton ACRS(10) 1 (LTR TO DR. BRUCE P0ULTON)  !

L NP PM:P P D:PD E h'* on AAd cw CMiller 044j/89 04/20/89 04h f/89 ,

8905030246 890425 PDR ADOCK 05000297 P PDC

e kQ Cf 0p Cf ^o UNITED STATES 8" t NUCLEAR REGULATORY COMMISSION  !

$ n$ WASHINGTON, D. C. 20555 f April 25, 1989

• .. +

Docket No. 50-297 Dr. Bruce R. Poulton Chancellor North Carolina State University Box 7001 Raleigh, North Carolina 27695-7001

SUBJECT:

REVIEW 0F LICENSE RENEWAL APPLICATION The Nuclear Regulatory Commission 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 North Carolina State University PUSTAR Research Reactor. We plan to conduct a visit to your reactor facility to discuss your application and to increase our familiarity with your facility. It is anticipated that a total of four people will participate in the visit. The date of our visit will be scheduled with the reactor staff.

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

Please review the preliminary questions and contact me et (301) 492-1121 for that a visit date can be finalized.

Sincerely, khGte0 h ,,

Alexander Adams, Jr. Pro' ct Manager Standardization and o ower Reactor Project Dir ctorate Division of Reactor Projects - III, IV, V and Special Projects Office of Nuclear Reactor Regulation

Enclosure:

As stated cc: See next page

1 ., . .  :

9 North Carolina State University Docket No. 50-297  ;

1 cc: Office of Intergovernmental Relations 116 West Jones Street Raleigh, North Carolina 27603 Dr. Paul J. Turinsky, Head  :

Department of Nuclear Engineering '

North Carolina State University i Box 7909 .i Raleigh, North Carolina 27695-7909 j Mr. Garry D. Miller i Nuclear. Reactor Program i Department of Nuclear Engineering )

North Carolina State University i P. O. Box 7909 I Raleigh, North Carolina 27695-7009 l S

1

.1 l

1 i

l l

l j

' t. ,. ': ,

ENCLOSURE'

~ PRELIMINARY QUESTIONS

' NORTH CAROLINA STATE UNIVERSITY DOCKET NO. 50-297

. SAFETY ANALYSIS REPORT

1. Section 1.2.3 Please provide additional information on the engineered safeguards'found in the Reactor Building. Are these the engineered safeguards discussed in Section 1.3.27

2. Section 1.3.2 (a) Licensees must comply with 10 CFR 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.

(b) Items (a) and (b) would not properly be considered ~" Engineered Safety Features," according to current usage. Please review and discuss.

3. Section 1.4.1 What are the features of reactor design essential to the prevention of accidents?

4. Section 1.4.4 What are the abnormal conditions referred to in this criterion and how is control rod insertion accomplished?

5. Section 1.4.5 How does this criterion (No. 5) fit into your Emergency Plan? From what hazards and how does the control room protect its occupants?

6. Section 1.4.7, Criterion 7 10 CFR 100 is not applicable to research reactors. Because there is no comparable regulation for research reactors, NRC expects you'to compare witi 10 CFR 20, and the ICRP recommendations on wnich it is based, with ALARA considerations. Please address this in all applicable parts of-1 your SAR.

7. Section 1.4.11, Criterion 11 1

, 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, etc. Discuss radiation monitoring in eve.it of accidents.

8. 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. WN NRC is no longer its name, height of pellets of BMRC is wrong, number of pins in BMRC is not consistent with fuel element dimensions, H2 0/U0 ratios seem inconsistent, 2

capture-to-fission rate, perhaps others. Please edit and correct.

9. Section 2.1 Page 1, last paragraph; please cross reference where in your SAR we can find the quantitative discussion and consequences of release and diffusion of radioactive effluents from the stack to potential exposed people, both on campus and off.

10. 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.

11. All of thapter 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.

12. Section 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 es m ment should be provided.

13. Section 2.2 Please provide updated meteorological data, for example, section 2.2.1.3 discusses the time period 1916 to 1958.

l e

1 J

14. Section 2.3.2 Discuss drainage between the reactor site and Lakes Raleigh, Johnson, and 1 Bonson. 1

)

15. Section.3.1.1. 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.

16. Section 3.1.2.1 I

(a) Reference where in this SAR you have analyzed 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 t'e n ]

so-called pulse rod is changed. Explain its role in the definition of shutdown margin. Discuss the reasons for not changing its name now. Justify the steps taken to " disable."

17. Section 3.1.2.2 The use of the verb "shall" implies a. requirement. Do you intend that all of these requirements be added to your technica1' specifications? If so, provide a proposed specification with bases.

18. Section 3.1.2.3 Most of this e action on fuel mechanical limits and total burnup " allowed" should be in technical specifications, Section 5. Some should also remain in technical specifications, Section 4, for surveillance. Please address and propose changes to technical specifications.

19. Section 3.2.1

" Vacant core positions" are mentioned. Please be more specific and provide the analyses necessary to justify such use. Provide a technical specification change that controls " vacant core positions."

20. Section 3.2.1 I

Most of these paragraphs are evidently intended to be summaries of the reactor's characteristics. This implies detailed and specific discussion and analyses elsewhere in the SAR. Please provide such 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), etc.

__m__ _ _ _ _ _ _ _ _ . _ . _ _

4

- 4.-

21. Section 3.2.2.1, page 9 Please refer to where you discuss the potential loss of pool water, because the water pipes are below'the core. Because this .section of your .

SAR addresses shielding, include an analysis of radiation exposure rates J in the reactor room, including top of pool, and the control room, .due to LOCA.. Furthermore, include such an analysis if the water level were to fall to 14 feet.

1 I

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

23. Section 3.2.2.5, top of page 12 It is. stated that for the control rods: "new positions will conform to the Technical Specifications." That seems appropriate,~but please refer to the particular_" specification" that controls. locations. Provide such specifications if not already included as stated.

24. Section 3.2.2.5 'q P

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

.It is more. conservative and realistic to assume the maximum rate of l reactivity insertion not the average permitted by technical '

specifications.

(b) Pulse rod Can the pulse rod be dropped by a SCRAM signal of some sort? Does i it automatically drive in upon a SCRAM signal? How does the pulse rod enter into considerations of shutdown margin and excess i reactivity?

(c) Are there credible actions by which withdrawal speeds might increase?

25. Section 3.2.2.5 Will the former pulse rod continue to be left completely out of the core  !

for most operations? l i

26. Section 3.2.2.6, page 14 (a) Please discuss, or refer to your SAR sections that discuss and analyze " insertion of samples--into the core region."

I (b) Clarify what is meant by a nonsecured experiment; where is it j defined, and analyzed for safety significance? j 4

i l

27. Section 3.2.2.6 Please discuss the design and use of irradiation baskets.

28. Section 3.2.2.6 Please refer to where you explicitly discuss the 4l Ar production and control related to the thermal column system. Please be quantitative.

29. Section 3.2.2.7 Has there been any indication of fuel pellet fracture due to pulsing. If indications of fracture exists, discuss the implications for safe operation of the reactor.

30. Section 3.2.2.7 Give a reference to and provide additional explanation of the 75 MW-sec energy release statement.

31. 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.

32. Section 3.2.3.5.1 (all subsections)

Discuss briefly which of these computer programs and input data have been superseded, and what changes in results and conclusions have occurred since the early 1960s.

33. Section 3.2.3.5.1.3 (a) Please explain the reasons and the implications of calculations and measurements being different, bottom of page 25.

(b) We do not have access to all of these references, either provide them or expand the discussion.

34. Section 3.2.3.5.2 Are calibration curves made for a particular core based on a time interval (such as every year) to reflect rod worth changes due to burnup?

l l

I, l

35. Section 3.2.3.5.2 (a) Please provide the maximum rate of change of reactivity with the ganged-rod arrangement.

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

36. Section 3.2.3.5.3 (all sections) q (a) Please discuss differences between NCSU 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 l

fuel, and explain the changes. 1 l -!

37. Section 3.2.4.1.1, page 33, Steady State l

l

.l (a) Are criteria (2) and (3) also for forced convective flow? Discuss.

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

38. 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 17 percent some sort of limit, root-mean-  !

square, or what? How is it verified?

l

39. Section 3.2.4.1.3.2., page 38 ,

Shouldn't the fundamental and relevant reason for no bulk boiling be to avoid DNB? Please discuss your reasoning that limiting I6 N release is the fundamental reason. Also give quantitative discussion of the 16 N problem and consequences, or cross reference, for ease of_ following your arguments.

l

40. Section 3.2.4.1.3.2, page 39 and.40 l

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

41. Section 3.2.4.1.3.3., page 41, second paragraph >

Please provide a quantitative justification for " assuming nucleate boiling effects." What does this mean?

42. Section 3.2.4.2.3, page 43, first paragraph 1

is it correct that DNB will not occur in the NCSU PULSTAR under natural convection conditions for pulse energies of less than 49.3 MW-sec? l

O

43. 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 79eate with DNB. Discuss probable DNB ratio.

(c) Compare 58 MW-sec with the 75 MW-sec mentioned on pages 3-18

44.

References:

(a) Many of these' references are more than 20 years old. Since then, much work has been done relating to power reactor fuel. Because you have implied the PULSTAR fuels similarity to power reactor fuel has some safety significance, please provide a discussion of the implications to PULSTAR of more recent publications and analytical developments.

(b) Were all references explicitly used somewhere for Chapter 3 of the SAR? For example, for what analysis did you use references 3-19 through 3-237

45. Table 3.2 (a) " Savings" for a H2 O reflector is given. Please compare with a graphite reflector.

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

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

(d) Keff-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?

46. Table 3-3 Are tFere any neasured values for neutron fluxes in the reactor?

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

~

.a a -

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

49. Section 4.2.1 Under what conditions is the water pressure higher in.the primary side or 'l secondary side of the heat exchanger? Discuss in relation to a.possible leakage across the barrier, and control of radioactivity.

50. Section 4.2.1 i

Discuss possible rupture or other failure of the primary coolant system' in relation to loss of coolant and fuel decay heat, and decay radiation-in the reactor room.

51. Fig. 4.1D 1

~

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?

52. Section 4.2.3  ;

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

53. Section 4.4

.Please be.sure that all required surveillance tests are included as specifications in your technical specifications. Give bases.-

54. Chapter 5 t It would be helpful if you would more specifically reference Chapters 10 and 13, and your environmental report where radiological ~ controls and considerations are ' expected to be developed quantitatively in detail..

55. Section 5.2 Why is the ventilation system not considered an engineered safety feature?

56. Section 5.2.1, 5.2.2 Please provide a more quantitative, discussion of the levels of radiation that trigger events, and their bases.

l

.g-

57. 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?

58. Section 5.2.2 What are the requirements for bypassing radiation alarms?

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

60. Chapter 6 Please review the definition of Engineered Safeguards and provide

, sufficient comment to assure that you and NRC use the same definition.

For example, what about Section 6.1.4 and 6.1.57

61. 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?

62. Section 6.2 Please be sure that any necessary tests are included in your technical specifications, " Surveillance".

63. Section 7.1.1 The SAR states that there is a circuit that tests the period circuit at

+10 second and +30 second periods. However, the startup rate scale is

-100 to +10 seconds. Please explain.

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

65. Section 7.4.1.

l What are the actions that are permitted and/or inhibited by the various positions of the Gang Drive Switch?

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

67. Section 7.4.2 If confinement integrity is a limiting condition of operation, why is an "open door" not a cause for operator reaction? Discuss this.

68. Table 7 Pleasebecertainthatalloftheprot6btivechannel.settingsare 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.

69. Figure 7-3: Magnet Power / Control Circuit  ;

i Does this solid state switch o)en the ground side?- Please discuss implications of failures and s1 ort circuits.

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

71. Section 9.1.4 i flow is the water level in the liquid waste holding tanks determined? i J

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

73. Section 10.1.2  ;

Please provide a more quantitative discussion of releases and dilut' ion of I liquid wastes before and after release.

74. 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 20.

! 75. Section 10.1.4 '

Chapter 5, this section of Chapter 10, and your environmental report does i not provide sufficient supplementary treatment of 41 Ar and 16 N. Please provide a detailed, quantitative analysis.  ;

i

11 -

76. 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.

77. 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.

78. 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).

79. Section 11.1 In addition to the text, the Associate Director (or his management) is responsible to manage reactor facility operations in conformance with applicable NRC regulations and the Operating License. This should be stated in the technical specifications.

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

81. Section 11.1.1.5.1 What are the responsibilities and qualifications of the Director of the Nuclear F.eactor Program?

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

83. Section 13.2.1.1 This radionuclides is released during normal operations. It should be discussed quantitatively in detail in both Chapter 10 and in your Environmental Report, rather than in Chapter 13.

84. Section 13.2.1.1 For your 4I Ar releases, please provide a more accurate and quantitative analysis of potential exposures and doses to personnel in unrestricted areas, including the most-exposed offsite permanent residents. Please include the dispersion / diffusion factors you have used in the analysis, Also address potential exposures in the reactor room.

85. Section 13.2.1.2 What is the affect of a failure of the flapper to open in the loss of flow accident?

86. Section 13.2.1.3 ,

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

l

87. Section 13.2.1.4 Please describe and discuss in more detail the penetrations in the pool 1 liner, and the primary coolant system by which pool drainage could occur i accidentally.

88. Section 13.2.1.4 (a) In order to avoid later confusion, LITR are not the initials for the Livermore Reactor, LITR is the Low Intensity Testing Reactor at Oak

.l 1

{

Ridge; LPTR stands for the Livermore Pool Type Reactor. I

\

(b) 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?

(c) Please give more details about the direct and scattered radiation above the pool and near (or in) the control room. Have you included 4 back-scattering from the roof? Where is the 250 mrem per hour  !

location? i

89. 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? i l

90. Section 13.2.1.5 Fuel Pin Clad Failure (a) Reference 13-6 is quite old. What later information can be used to confirm those early results for your situation?

(b) What is the source of the information in Table 13-1 and 13-2?

Discuss and compare with TID 14844, for example. What percentages of various radiologically important radionuclides migrate to and accumulate in the pellet-clad gap?

I i l

4

f l

l (c) Table 13-4 What is the source of the information in Table 13-47 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 reactor room.

4 (d) Page 9 Please justify the use of 97 percent retention. There must be more l current information available. .

(e) Page 10 l

Please justify the 3 x 10-2 value.

(f) Page 13 f a

Please provide the details of your assertions, especially for the j sentence beginning: "An eight hour average ...."

91. Section 13.2.2.1 (a) Discuss the Nplict tions if your excess reactivity, and also the maximum worth of ar added fuel assembly when limited by technical specifications to 1130 pcm.

(b) Compare this analys is to total loaded excess reactivity.- Place and )

justify this maximun allowed excess reactivity in your TS.

(c) How is accidential removal or ejection of a control rod precluded?

92. Appendix 3c, pages 9-12 Please provide a similar analysis with the following initial conditions: 1 (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.

93. Appendix 3C What maximum reactivity insertion (inadvertent) would be possible without initiating any film beiling at the fuel hot spot? '

c_ ._____m_ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ . _ _ _

. 4 OVESTIONS NCSU Environmental Report

1. Page 2r S_setion 2, Paragraph 2 Give a quantitative comparison between the N cooling towers:

a. physical dimensions

b. heat-dissipation capacity l

)

i

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

2. 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.

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

(b) The statement is made that the PULSTAR Fac'lity keeps the .

dlAr releases as low as practical. - Furthermore, it is i stated that the annual releases of radioactive effluents have been and will be at or below established federal limits. Please provide in some detail the quantitative analyses by which you have arrived at the ALARA condition. As part of this, please l

provide your best estimates, with bases, for the doses per year j to the maximum exposed off-site resident. Include all sources I

of d!Ar, including the Thermal Column.

J. ,

Page 5, Paragraph 1 4.

It is stated that federal regulations are met. Please provide the l l

concentrations of radioactivity released, and discuss your ALARA considerations. , Quantities may be expressible as MPC-brs/ year.

5. Page 5, last sentence -

Please provide as much quantitative evidence as necessary to support the statement " expected to be small". What does "small" mean?

l

6. Page 6, Section 5 10 CFR 100 is not applicable to research reactors. Give quantitative

^

results here, or refer explicitly to your SAR, and compare with 10 CFR 20 and its ICRP bases.

7. Page 6, Section 6 Discuss storage and disposition of spent fuel.

8. Accidents If there are accident events postulated that could impact the environment, they should at least be discussed in this report, and references to where they are analyzed. -

9. You take credit for both 10,000 cfm and 12,500 cfm airflows. Please discuss what provisions assure that both are operating when required. This question also applies to Chapters 10 and 13.

Questions and Comments Technical. Specifications

1. T.S. 1.8 Now that. pulsing is no longer an approved mode of operation, why not.

rename the " pulse rod," and include it in with the other rods?

2. 'T.S. 1.17, Reportable Event: see question _about T.S. 6.7.1.

Definition of Reportable Events should follow ANS 15.1, and include:

l (a) Reaching or exceeding a Safety Limit. 1 (b) An uncontrolled release of radioactivity off-site. {

(c) An uncontrolled or unanticipated exposure to radiation of some one within the restricted area. 1

3. T.S. 1.18(b)

Does the definition of Secured Experiment consider forces on the experiment under both normal and abnormal conditions?

4. T.S. 1.18 Experiments: you have deft..cd a " Secured Experiment." Are all others considered to be nonsecured by definition?

5. T.S. 1.20, Steady State (a) Not only should the switch be in the steady state position,'but more importantly, the reactor power must be maintained at some constant average value.

(b) The last sentence about power increases is inappropriate in this i definition.  !

6. T.S.1.21,ShutdownMargin--seecommentaboutT.S.3.2(b).  !

7. T.S. 2.1, S.L.: Figure 2.1.1 (a) What is basis of curve exceeding 100% of full flow?

(b) What is basis of note about loss of flow? Is it necessary?

8. T.S. 2.2.1 (a) Many reactors have a LSSS = 1.10 times Licensed Power. What is basis for 1.3 times Licensed Power?

(b) Water height of 14.ft. 2 in, may be acceptable for fuel safety, but how about personnel safety / core shielding? ,

I

- _ _ _ _ _ _ _ _ _ _ . - _ _ - - - - _ _ _ - - M

9

-2~

9. T.S.'2.2.2 Height of Water 41 Analyze radiation shielding and diffusion release of Ar and 16 N into the room, and consequences in the SAR, and summarize and reference'it here.

10. T.S. 3.1 This is a LCO, yet there doesn't seem to be an automatic protective action (SCRAM). Why not, and what provisions could be made to include that?

11. T.S. 3.2 (a) For " shutdown margin," this condition should exist "with.the highest worth and any nonscrammble rods fully withdrawn......under operating condition."

(b) Rod drop time Most licensed nonpower reactors have a specification of "not greater than 1.0 sec." The argument being that the free-fall (into water) time will certainly be less than 1.0 sec, and any significant excess time (even 1.5 sec. total) indicates some kind of mechanical inter-ference or hang-up. Please justify the 1.5 seconds. Please discuss drop-time measurements in the SAR.

(c) ilost licensed nonpower reactors have an upper limit on total excess reactivity. Please justify your not having one or proposed one.

~

(d) Most nonpower reactors do not use pcm as a unit of reactivity. In the interest of clarity, why not define it in your set of definitions?

(e) You have reactivity limits on movable experiments but you do not define.

(f) Why is the reactivity worth of the pneumatic rabbit different from movable experiment?

(g) The second sentence of basis "e" includes the statement " reactor power will be reduced....." This seems to be an operational requirement that should not be mentioned only in a basis. It should be a " Specification" or provided for in " Operating Procedures".

Please explain where it is located and why.

(h) Basis "f" includes the statement that " Specification 3.2f places a reasonable limit on the worth....". Please provide additional' quantitative reasoning that justifies the Specification..

1 i

l

12. T.S. 3.3 The information here appears to be a repeat of information found in T.S. 3.4.

Please comment.

13. T.S. 3.4 l (a) With few exceptions, these specifications only require operability, i and do not explicitly. include the settings for Alarm, SCRAM, or l other action. These. limits should be an explicit component of the S) edification. Among other reasons for displaying the settings in t11s table, there would be less change of a wrong setting, and.less ']

change of a difference of opinion by an NRC inspector.

(b) The actions required by the two pool temperatures should be explicitly-spelled out and provided for, either in technical specifications or specific operating procedures. Where are they?

1 (c) The use of the over-the-pool radiation monitors might be more j important to avoid or warn of excess radiation exposure rates than '

to avoid exceeding the safety limit. 'Please address this use, including the 41 Ar and IO N factors.

(d) The basis relating to initiating evacuation / confinement seems to include a " requirement" for a VAMP not included in the table of Section 3.4. A " basis" section is not appropriate for an additional requirement. Where is this VAMP required? Why not in this Specification? Justify its not being part of the specifications.

(e) Item f, Primary Coolant Flow, has the cperating mode listed as Steady State. What other mode is there?

(f) When the bridge monitor is bypassed, is the entire monitor removed from service, or are just the alarm circuits?

14. T.S. 3.5 (a) T.S.(b)(2)

The temporary instrument is to be "...kept under visual observation by the operator on duty."

(b) The settings for the alarm level and the range of operability for these monitors should be specified by T.S. 3.5.

(c) T.S.b(2)

Is there any limit to the replacement?

M) Is there any requirement to record the output of the radiation monitoring equipment? If not, why not?

i j ,

..~

(e) What provision is made for a functioning radiation conitor in the.

event of a severe radiological accident?

i

15. Section 3.6 1 (a) Footnote (2)shouldincludeanupperlimitontime,'suchasthe  !

5minutesoffootnote(1). l (b) Confinement: it should be made clear that the filter train includes the charcoal filter, and how it is tested.

(c) Please justify' footnote (3), which allows reactor operation' with the reactor confinement.out of service.

(d) Concerning footnote' (5),- can the' reactor public address systsi be heard in all areas that the building evacuation system can?-

16. T.S. 3.7 (a) This Specification does not include " fueled experiments." Either T.S. 3.8 should be included within 3.7, or 3.7 should disclaim q control of fueled experiments.

(b) T.S. 3.7c (Explosives) seems to be inconsistent. . If there is no way of distinguishing between " explosive"--first sentence, and "potentially.

explosive"--second sentence, the two sentences are not self-consistent.

Furthernere, if sentence 2 is retained, it should say, as a minimum:

" ..to maintain seal integrity even if detonated to. prevent damage...".

Please define potentially explosive and explosive.

(c) Consider adding to the last sentence of the bases,'on-page 20: "

...and thereby become an unreviewed safety question."

17. T.S. 3.8 (a) Is 3.8 (1)(a) consistent with minimum water height above the core given in T.S. 2.1.1 and 2.1.2?

(b) Consider moving Section 6.8.3 limitations to this section on fueled experiments.

(c) Specification 3.8 (3) and (4) seem to be applicable to a11' experiments, so should also be in Specification 3.7.

l (d) Relate your T.S requirements for thermal power to fission product concentrations.

(e) T.S.3.8(4)(d). Define flansnable, highly toxic, and cryogenic -liquids.

l

. \

. q

~.*

}

(f) It seems that the first paragraph of the Bases, page 22, has the impact of a requirement, and therefore should be a Specification, not a basis for one.

18. You should consider the need for T.S. requirements concerning emergency power.

19. Consideration should be given to T.S. limits and surveillance requirements on coolant pH and conductivity.

20. T.S. 4.0 i (a) The intent of the inspection interval is to maintain an average, with occasional extensions for good reason. Therefore, in order to avoid ambiguity this should state: " ...: inspected biennially 1 on the average, but with no interval to exceed thirty months."  !

There are several other places in your technical specification where'  !

this same type of ambiguity exists and should be remedied. j (b) Pleasedefineclosely'associatedsystemasusedinfootnote(1).

21. T.S. 4.1 (a) What are fuel assemblies inspected for? What are the pass / fail criteria?

(b) What is the total maximum burnup allowed on your fuel? 1

22. T.S. 4.2 (a) Section4.2(c)

Either in the Specification itself, or in the basis, more should be included about the purposes of inspection, including looking for signs of deterioration, corrosion, flaking, etc. of the absorber-material.

(b) There shculd also be a Specification that assures checks on Shutdown Margin and Excess Reactivity.

23. T.S. 4.3 ,

(a) What type of checks, tests, or calibrations are required if equipment is taken out of service for repair?

(b) The bases should explicitly include the fundamental reason for item (c) such as, "to assure operation within the limits of the Operating License."

{

5 24 T.S. 4.5 Do you perform surveillance on the charcoal filter to determine its continued ability to sorb iodine?

25. T.S. 5.0 Consideration should be given to describing design features such as the area of the building to be under the reactor license, control rods, cooling system, etc.

26. T.S. 6.1 (a) Your organizational chart shows a director of the nuclear reactor program. However, it appears that the director has no responsibility for the safe operation of the facility. Please explain.

(b) In T.S. 6.1.3, because NRC Regulations predominate over the License in case of conflict, this should also say: " ... within the limits prescribed by the facility license,. all applicable NRC Regulations, and the provisions...".

(c) What are the duties of the reactor operators and senior reactor' operators? In particular what duties must be carried out by a SR07

27. T.S. 6.2 (a) Please see ANS 15.1 for additional duties your council should have.

(b) How often must review and audit occur?

(c) T.S.6.2.2(d)

NRC regulations do not seem to exclude your RPC from participating on your Physical Security Plan reviews, so that part of 6.2.2(d) is not required by NRC. Please address this.

(d) T.S.6.2.2(e) l The audit function should also include how well radiation controls-and radiological protection are functioning, how well the emergency plan and operator requalification plan are working, and how well l

corrective actions put in place to correct identified weaknesses are L working.

(e) T.S. 6.2.5 Is there any time limits on when RPC meeting minutes must be distributed?

l

/

l

~

1 l

1

(

1 l

28. T.S.6.3(a)(10) j The " Plans" to and Training be reviewed Plans," should and similar probably Facility include to Documents "Op(erator Requalification account for possible

' future things you might want to include). Please consider.

1

29. T.S. 6.5 (a) What happens to the reactor if a reportable event occurs?

(b) Who approves the resumption of operations if the reactor is shutdown?

30. T.S. 6.7 (a) Please consider including 6.7.1 (a) and (b) and 6.7.2(b) as part of the definition of Reportable Event?

(b) T.S. 6.7.1 The NRC Operations Center should also be informed.

(c) Written reports should be sent to the Region II Regional Administrator and the Document Control Desk, USNRC, Washington.

(d) Section 6.7.4 Please consider deleting the words ".. 120 days until October 30, 1973, then ... thereafter...".

31. T.S. 6.8.1 What are the tried experiment criteria?

1 l

l 1

l i

l j

j

. V t

REQUEST FOR ADDITIONAL INFORMATION.

NORTH CAROLINA STATE PULSTAR REACTOR FACILITY OPERATOR REQUALIFICATION PROGRAM BACKGROUND l

By letter request dated August 19, 1988, the North Carolina Sta.te Reactor Facility (Licensee), requested that the staff review the Licensee's 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 the criteria of 10 CFR 55.59, "Requalification". Our review and )

request for additional information follows. j i

SECTION 8 The requirements of 10 CFR 55.59(c)(3) state that the requalification program l must include on-the-job training addressing control manipulations. Licensee J 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 Section 8. {

j SECTION 9 The requirements of 10 CFR 55.59 (a)(2)(ii) state that operators and senior operators must be examined such that they demonstrate an understanding of the j ability to perform the actions necessar '

of items specified in 10 CFR 55.45 (a)(y to accomplish a comprehensive sample 2) throu 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.

.