ML20210J421

Forwards Preliminary Questions & Comments for Discussion During 861007 Site Visit Re License Renewal Application

ML20210J421
Person / Time
Site:	Purdue University
Issue date:	09/25/1986
From:	Berkow H Office of Nuclear Reactor Regulation
To:	Stansberry E PURDUE UNIV., WEST LAFAYETTE, IN
References
NUDOCS 8610010142
Download: ML20210J421 (8)

Text

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,; September 25, 1986~

Docket No. 50-182 Mr. Eldon Stansberry Reactor Manager Nuclear Engineering Department Purdue University West Lafayette, Indiana 47907

SUBJECT:

APPLICATION FOR LICENSE RENEWAL 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 Purdue University research reactor. We have also planned a two day visit and review at your reactor facility beginning on October 7,1986 to discuss your application and to increase our familiarity with your facility.

During our visit we also want 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 home 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 contract our Project Manager for your facility, Robert Carter, at (301) 492-8206.

Sincerely, original signed by O. D. T. Lynch, Jr., for Herbert N. Berkow, Director Standardization and Special Projects Directorate Division of PWR Licensing-B Office of Nuclear Reactor Regulation

Enclosure:

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September 25, 1986 Docket No. 50-182 l Mr. Eldon Stansberry Reactor Manager

, Nuclear Engineering Department Purdue University West Lafayette, Indiana 47907

SUBJECT:

APPLICATION FOR LICENSE RENEWAL 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 Purdue University research reactor. We have also planned a two day visit and review at your reactor facility beginning on October 7, 1986 to discuss your application and to increase our familiarity with your facility.

During our visit we also want to discuss the information indicated by the enclosed set of Preliminary Questions and Coments. 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 home 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 contract our Project Manager for your facility, Robert Carter, at (301) 492-8206.

Sincerely, Herbert N. B ko frector Standardization and Special Projects Directorate Division of PWR Licensing-B Office of Nuclear Reactor Regulation

Enclosure:

As stated cc w/ enclosure:

See next page

Purdue University Docket No. 50-182 cc: Mayor Sonya L. Margerum City of West Lafayette 609 W. Novajo West Lafayette, Indiana 47906 State Board of Health ATTN: Director, Bureau of Engineering 1330 West Michigan Street Indianapolis, Indiana 46206 Attorney General State House Indianapolis, Indiana 46206 O

Enclosure  !

Preliminary Questions and Coments-Purdue University

1. Please provide the figures missing from the SAR, and additional figures to help understand the reactor, core configuration, general facility layout, and geometry, to include the following:

-a. Fuel locations.

b. Regulating and shim rod locations,

c. Neutron start-up source location.

d. Control detector locations,

e. Reflector locations and geometry,

f. Experiment locations.

2. Conclusions appear to be drawn without a very well defined basis, in some instances. For examples, sectionr 7.1, 7.2, 7.3.

3. Drawings or schematics are needed for ventilation and drain systems and water process system.

4.. Describe any procedures (operating manuals) used in the operation of the '

! reactor and associated systems as well as for handling of wastes.

5. Explain the relationship between the SAR, HSR, and Tech Specs, and be sure appropriate cross references are made. Bases in the Technical Specifications should support the specification itself, and,in general, should be derived from the Final Safety Analysis Report.

6. Tech Specs do not conform to ANSI /ANS 15.1 in all respects. Please review.

7. Comments Pertaining To Table 4.1 a) On page 3-3, " average thermal neutron lifetime" = 77.2 x 10-5 sec.

On page 7-7, " prompt neutron generation time" = 77.2 usec. Resolve what is the magnitude of this number, what is the origin of this numerical value, and what it should be labeled.

b) Cannot reconcile the maximum or average rate of reactivity change 4 from the rod worths and linear speeds on Table 4.1. The average rate of reactivity, change for the regulating rod is missing the exponent.

c) The void coefficient should be stated in the more comon terms of i percentvoid,i.e.,( Is the void coefficient given an average, maximum,other? or $/%Isvoid).

the void coefficient nowhere positive?

j d) Discuss flux and/or power peaking and hottest fuel channels.

What is the temperature dependence of the temperature coefficient?

e)

A curve of temperature coefficient versus moderator temperature should be helpful. Is it negative at all credible temperatures?

f) What are the uncertainties in the various parameters in Table 4.17

8. When would the water " chiller" be used - is it "on line"?

9. What is the uncertainty and reproducibility in the true value of the control rod position (s)?

1 l 10. Fuel element handling accident - discuss what happens if one element falls  ;

on top of the core, or falls into a water-filled fuel element position?

11. Identify the written procedures that contain the " specific actions" to be taken in case of a pool leak.

12. What is the basis of assuming 1.1 gm U-235 in position F6 as the maximum l experiment to be analyzed?  !

13. Additional comments:

a Table labeled 4.1 appears in Chapter 3, b Figure 6-1, referred to on page 6-1, is missing, c Spert I fuel was 93% enriched, not 100% as shown as Table 7.4.

14. P. 3-5 Exponent is missing on reg rod reactivity change rate.

15. P. 3-12 What is the basis for the conclusion that control rods cannot

. be gang-raised?

16. P. 3-13 Are there hard-wired interlocks on the Neutron Source Drive system and on the Fission Chamber Drive System? Discuss.

17. P. 3-13 Last sentence of section 3.7.6.2 says " acknowledge button may be used". Can operation resume or continue without resetting the abnormal function? Discuss.

18. Environmental Report Section

a. Expgaininmoredetailwhatismeantby"notdetectable"inthecases of Ar,*N, and Tritium. What methods were used, what are the limits of detection of these methods, and how do they compare with regulatory limits as given in 10 CFR 207 Be as quantitative as possible, and also discuss from the perspective of the ALARA principle.

b. Provide a more quantitative discussion of solid waste disposal over the many years of operation.

c. These should be cross-referenced to sections 5.7 and 5.8.

19. What use is made of the " Radiological Control and Health Physics Handbook?" We are examining the reactor facility and operations, so is this handbook incorporated into a reactor operating Procedure?

20. Provide a document (letter, memo, etc) expressing the University's i commitment to ALARA. How is it implemented and assured?

21. Is the Health Physics staff only "available", or is it actually consulted "in matters concerning radiation safety?" By what mechanism is

~

consultation assured?

22. Who are " reactor-related personnel" (see section 5.2.3)? Does this include students, possible experimenters, etc?

23. Section 5.3.1 should give more quantitative information about reactor-related radiation levels. Some 25 years of operation should have produced actual data.

2.4. Section 5.4 of the SAR: Provide information on the bases for setting the

-adjustable alarm set points, what responses are established if the various radiation alarms sound, how,and how frequently calibrations of monitors, especially the CAM, are performed.  !

25. Section 5.5.1 mentions keeping occupational exposures below regulatory 1 limits. Please explain the mechanism of also implementing the ALARA principle. i

26. Section 5.5.2, among other items, states that "it is not possible to  ;

determine" the fraction of reactor-related exposure. Please reconsider  !

whether it is only difficult or impossible, and discuss this. This  !

section also states that a fuel plate inspector's finger badge may indicate 100 mrem. What upper limit is accepted, and what is the basis?

2f. Section 5.6.1 states that "no airborne effluents have been released -- ".

Assuming this means radioactive effluents, what limits and accuracy do j you ascribe to this "zero"?

28. P. 5-4 Where is the location of the CAM relative to the potential l pathway of radioactive effluent? Where does it read out and alarm? '

What procedures spell out the operator response upon CAM indication above acceptable level? What is that level, and on what criteria is it based?

29. Section 5.8, please show how you arrive at the "less than Imrem/yr" figure.

30. Section 6.1, paragraphs 2 and 3 indicate that the reactor supervisor has two channels for responsibilities. Please indicate more clearly how the responsibilities are agreed-upon and coordinated among the parties.

31. Section 6.2 seems to be inconsistent with regulations; please review 10 CFR 50.59 and be prepared to discuss.

32. Section 7.1, Fuel Element Handling Accident, is appropriate to consider, including administrative provisions intended to prevent such an accident. However, the nature of an accident is an event that may occur in spite of administrative limitations. Therefore, this and other accidents should be analyzed from the view point of "what would be the consequences if the postulated accident did occur."

33. P. 7-2 What is the reference to show result of 0.3% AK/K?

34. Second paragraph on page 7-3 states that the most severe problem identified in a loss of coolant accident is the removal of decay heat.

Please discuss this more quantitatively. Furthermore, also consider the consequences of direct radiation from the core if there were a loss of coolant.

i. '.

35. What " specific actions" (see page 7-3) would be taken in the event of loss of coolant? '

36. P. 7-3 What is the rationale that a water level instrument on the pool is not needed?

37. Section 7.5.3

a. In the first paragraph you state that no credit for negative temperature feedback is taken. Please estimate the magnitude of this feedback during the postulated scenario, and its likely effect on the consequences.

b. You mention a design of PUR-1 for 10kw. Please explain the basis of the design limitation, and the impact on the consequences of this postulated accident.

c. In the second paragraph you indicate a maximum power level of 360 kilowatts while table 7.3 gives 380kw. But, there are three more substantive questions raised: (1) demonstrate that the maximum fuel temperature is below the alloy melting point, (2) resolve and explain which temperature coefficient is closest to the true value, (3) mention the direct radiation levels, for example in adjacent classrooms, resulting from this hypothetical accident.

d. References 2 and 3 should be supplemented by later reports in the SPERT program.

38. Section 7.6

a. Please give a technical and operational basis for assuming a I watt

fueled experiment. If more quantitative analysis has been performed to show that Tech Spec 3.5.f. limits the source to 1 watt, please discuss it.

b. What is the basis of choosing 100m as the " release point" in the unrestricted area?

c. Please clarify whether the thyroid doses shown are delivered within the time of exposure to the airborne radionuclides, or are the committed doses on into the future. What do you mean by thyroid dose rate?

, d. For non-power reactors,10 CFR 100 is generally not applicable. what i is meant by "the determination of an exclusion area" (page 7-13)?

e. The last sentence on page 7-13 seems to be incomplete. Please clarify.

39. Technical Specifications - some canments: see page 5.

40. Please provide information on your program for preventive maintenance, which might help give assurance of operability of systems and components.

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Technical Specifications

a. Page 3 What is the basis for choosing more than approximately 0.5%

k/k addition to a suberitical reactor as the cutoff for a Reportable Occurrence?

b. -f_aC 50 Kw is an arbitrary " Safety Limit". A more quantitative Technical basis for the Safety Limit should be given.

c. Page 9 Is loading permitted with Keff=0.977 What is reactivity worth of most worthy fuel element? Which position in the core, and what is the source of these data?

d. Page 14 What is the reference (or basis) for 13 ft. water head?

e. Page 18 Calculated experiment worth prior to insertion in the reactor should also satisfy experiment reactivity criterion.

f. Page 21 15 months is a long interval between drop time checks for the

. SHIM-Safety rods. Basis is: " consistent for 14 years since the PUR-1 was built". Page 3-1 of the SAR states the reactor was built 24 years ago. Some other check intervals need to be reviewed, too.

g. Page 24 The basis for limiting the mass of samples of unknown composition to 10 grams is "Past Experience". This past experience should be referenced, and a technical basis also given.

h. Page 31 If only a quorum of COR0 is present, is it not permissible for a majority of this quorum to be made up of personnel directly concerned with the administration or operation of the reactor?

Clarify this point explicitly.

1. Page 35 In addition to a log book entry (step 6.47) it is assumed that the person performing the procedure step initial that the step has been successfully completed. Reference your written procedure that provides for this action.

J. Page 36 Step 6.5.2e. When an updated drawing has been placed in the file, what is the fate of the now outdated drawing? I