ML20213G408

Forwards Request for Addl Info in Order to Continue Review of Documentation Submitted in Support of Application for Renewal of License R-87.Responses Should Be Provided by 861121

ML20213G408
Person / Time
Site:	Purdue University
Issue date:	11/07/1986
From:	Berkow H Office of Nuclear Reactor Regulation
To:	Clikeman F PURDUE UNIV., WEST LAFAYETTE, IN
References
NUDOCS 8611180085
Download: ML20213G408 (9)

Text

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November 7, 1986 Docket No. 50-182 Dr. F. M. Clikeman Head, School of Nuclear Engineering Purdue University W. Lafayette, Indiana 47907

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION We are continuing our review of documentation that has been submitted in support of your application for renewal of the operating license of your research reactor facility. An additional review was performed during our visit to your facility the week of October 6,1986. During these reviews, several questions have arisen for which we require answers. You are requested to provide written responses to the enclosed questions no later than November 21, 1986. Following receipt of this infonnation we will continue our j safety evaluation.

If you have any quoi. ions concerning this request, please contact our Project Manager for your facility, Robert Carter, at (301) 492-8206.

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.

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

Enclosures:

As stated cc w/ enclosures:

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Docket No. 50-182 Dr. F. M. Clikeman Head, School of Nuclear Engineering Purdue University W. Lafayette, Indiana 47907

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION We are continuing our review of documentation that has been submitted in support of your application for renewal of the operating license of your research reactor facility. An additional review was performed during our visit to your facility the week of October 6, 1986. During these reviews, several questions have arisen for which we require answers. You are requested to provide written responses to the enclosed questions no later than November 21, 1986. Following receipt of this information we will continue our safety evaluation.

If you have any questions concerning this request, please contact our Project Manager for your facility, Robert Carter, at (301) 492-8206.

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.

Sincerely,

< t .

f Herbert N. Berkow, Director tandardization and Special Projects Directorate Division of PWR Licensing-8 Office of Nuclear Reactor Regulation

Enclosures:

As stated cc w/ enclosures:

See next page

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Purdue University Docket No. 50-182

-1 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

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9 Enclosure 1 4

Formal Questions - Purdue University Operating License Renewal

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 location,

c. Neutron start-up source location,

d. Control detector locations,

e. Reflector locations and geometry,

f. Experiment location,

g. Block diagram of the safety circuitry,

h. Experimental facility locations and descriptions.

2. Please provide drawings or schematic diagrams for ventilation and drain systems and the water process system.

3. Describe any procedures (operating manuals) used in the operation of the reactor and associated systems as well as for handling of wastes.

Describe procedures for obtaining safety comittee approval for inserting experiments into the reactor.

4. 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 and analyzed in the Safety Analysis Report, and/or HSR. These documents form the permanent bases for the licensing of your reactor.

5. Canents 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) We cannot reconcile the maximum or average rate of reactivity change from the rod worths and linear speeds on Table 4.1. The average

' rate of reactivity change for the regulation rod is missing the exponent.

c) The void coefficient should be stated in the more common terms of percent void, i.e.,  % void), or a comparison made. Is the void coefficientgivenan@ average, maximum,orother? Is the void coefficient nowhere positive?

d) Discuss flux and/or power peaking and hottest fuel channels, e) What is the temperature dependence of the temperature coefficient?

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

f) What are the uncertainties in the various parameters in Table 4.1?

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

7. 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?

8. What are the appropriate limits for non-fueled experiments? Please discuss the bases for these limits.

9. P. 3-12 What is the basis for the conclusion that control rods cannot be gang-raised?

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

11. 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.

12. Environmental Report Section

a. Expginif6m re detail what is meant by "not detectable" in the cases 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 20? 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. That is, what average annual quantities were disposed of, and how?

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

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

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

15. Section 5.3.1 should give more quantitative information about reactor-related radiation levels in various restricted and unrestricted areas. Some 25 years of operation should have produced actual data.

Please provide and discuss.

16. 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 perfonned.

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17. Section 5.5.2, among other items, states that "it is not possible to determine" the fraction of raactor-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?

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

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

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

20. Section 6.2 seems to be inconsistent with regulations; please review 10 CFR 50.59 and provide a revised discussion.

21. P. 7-2, How was the value of 0.3% Ak/k determined for the 5" drop tube?

22. What maximum temperature rise in the fuel would you expect following a loss of coolant accident. Discuss the quantity of decay heat expected.

Also, consider the consequences of direct radiation to the surrounding areas in the event of loss of coolant.

23. Section 7.5.3

a. In the first paragraph you state 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.

24. Section 7.6

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

b. 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?

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

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

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

26. Please provide responses to the enclosed questions (Enclosure 2) related to your revised Operator Requalification Program, transmitted to us by letter dated September 19, 1986,

27. Technical Specification

a. Page 3 What is the basis for choosing more than approximately 0.5% Ak/k addition to a subcritical reactor as the cutoff for a Reportable Occurrence?

b. Page 7 50 KW is an arbitrary " Safety Limity". A more quantitative Technical basis for the Safety Limit should be given.

c. Page 9 What is the reactivity worth of most worthy fuel element?

Which position in the core, and what is the source of these data?

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

e. 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.

f. 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.

g. Page 31 If only a quorum of CORO is present, it is 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.

h. 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.

1. 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?

4 c' ' Enclosure 2 4

PURDUE UNIVERSITY REVISED OPERATOR REQUALIFICATION PROGRAM REQUEST FOR ADDITIONAL INFORMATION SECTION A - LECTURE TOPICS This section of the licensee's submittal states that eight meetings, i.e.,

classes, will be held over a 2-year period. Each meeting will consist of a review of reactor operations and modifications, if any, and a lecture on one or more of the topics listed in Part 2 of 10 CFR 55 Appendix A. Although this satisfies the requirements of that review document, the words "one or more"donotmeetthecriteriacontainedinSection6.1(2)(b)of ANSI /ANS 15.4. This section of the standard states:

The schedule shall be such that the entire program covering the topics listed in (Section) 5 is completed over a 2-year period.

SECTION 8 - EXAMINATIONS The licensee states in this section that operators will take an annual written examination emphasizing the material in the lectures for that year.

Since Section A does not make it clear that all topics listed will be covered in the 2-year cycle, this section of the program does not meet the criteria of Section 6.1(2)(b) of ANSI /ANS 15.4 which states:

An examination shall be administered at the end of each l segment, or a final examination given at the conclusion of the program covering all topics shall be substituted for individual examinations.

The licensee should clarify its intent in Sections A and B with respect to the inclusion of all the topics listed in ANSI /ANS 15.4. Section 5, as follows:

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(1) Nuclear Theory and Principles of Operation (2) Design and Operating Characteristics (3) Facility Instrumentation and Control Systems (4) Facility Safety Systems and Engineered Safety Feature (5) Normal, Abnormal, and Emergency Procedures (6) Radiation Control and Safety (7) Technical Specifications and Bases SECTION E - RECORDS This section of the program description should include a requirement for documenting the subject matter taught by senior operators who are exempt from any portion of the requalification training program. This documentation should be retained as part of the training records of these individuals.

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