ML20214T063

Forwards Response to 861107 Request for Addl Info Re Application for Renewal of License.Encl 2 Includes Missing Page from Original Sar,Per Question 1.Revised Operator Requalification Program Also Encl,Per Question 26

ML20214T063
Person / Time
Site:	Purdue University
Issue date:	12/02/1986
From:	Clikeman F PURDUE UNIV., WEST LAFAYETTE, IN
To:	Berkow H Office of Nuclear Reactor Regulation
References
NUDOCS 8612080465
Download: ML20214T063 (61)

Text

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9 PURDUE UNIVERSITY

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SCHOOL OF NUCLEAR ENGINEERING December 2, 1986 Herbert N. Berkow, Director Standardization and Special Projects Directorate Division of PWR Licensing-B Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.

20555

SUBJECT:

Additional Information as Requested

Dear Mr. Berkow:

In response to your letter of November 7,

1986, requesting additional information in support of our application for for renewal of the operating license for our research reactor facility, we are submitting the enclosed information.

Included with this letter are three enclosures.

Enclosure I

consists of our response to the questions from your letter. Enclosure 2 includes the missing pages from the original SAR as requested in Question

1. 1.

Enclosure 3

is made up of a revised Operator Requalification Program which answers Question #26, and the concerns expressed in your Enclosure 2.

If there are any additional questions that may arise, or if we can be on any assistance in any manner, please feel free to call upon us.

Sincerely, N

=;-f F. M. Clikeman Professor and Acting Head hj.o.2./1b zus, d d_

Enclosures - 3 MARILYN B. BROWN, NOTARY PUBLIC 86120a0465 861202 RESIDENT OF TIPPECANOE COUNTY INDIANK PDR ADOCK 05000182 P

PDR MY COMMISSION EXPIRES MAY 15,' 1388

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e-d Nuclear Engineering Laboratories EXPERIMENTAL FACILITY LOCATIONS

b FACILITY DESCRIPTION The Nuclear Engineering Laboratory facility is located in the basement of the Dancan Annex of the Electrical Engineering Building at Pardse University. This area consists of the Reactor Room (B-70A),

a class-room laboratory (B-76), a wet lab (B-77), a counting room (B-81),

the Subcritical Pile Room (B-84), a library (B-85),

a heat transfer laboratory (B-86), and support facilities, such as a shop, offices, and storage areas. The area is classified as a restricted area with exterior doors remaining locked except during scheduled class periods.

The reactor and its control console are located in Room B-70A, a

high bay area. A crane is mounted on the ceiling beams and is used for moving large obj ects into and out of the pool.

It is located off-center of the pool,.away from the reactor core. Two transformer vanits have access off this high bay area. A large lectare hall is located on the floor over the reactor room.

The reactor pool has an 8 foot diameter, and is 17 feet, 4

inches deep.

It is made of a carbon steel tank with a stainless steel liner.

The tank rests on a 15 inch thick concrete pad, and is surrounded by a steel casing with compacted magnetite sand filling the space between.

Outside the steel casing is the regular gravel upon which the building sits.

The grid plate upon which the reactor core is positioned, and two fuel storage racks are located in the pool. All three are bolted to the bottom of the pool. The fuel storage rackr are located on the opposits side of the pool from reactor core as shown in Drawing 4 of the Hazards j

Summary Report.

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

ANSWER Drawings and schematics of the ventilation and water process systems are included as enclosures.

The reactor pool has no drain. A break in s ay water line in the water process system would lower the water level in the pool to more than three feet, leaving no less than ten feet of water above the reactor core.

No open floor drain is located in the reactor room.

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N REACTOR WATER PROCESS.iYSTEM

QUBSTION 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 committee approval for inserting experiments into the reactor.

ANSWER An approved operating manual describes the general, normal operation of the reactor.

Specific procedures, approved by CORO before being implemented, are written to cover approach to critical, shutdown, placement, operation, and removal of experiments, certain surveillance tests, and fuel handling operations.

Both are readily available at the console for operators reference.

A system of dating is used to distinguish between proposed and CORO approved procedures. The proposed procedure should have the author's name and date in the upper right hand corner when it is submitted for initial review. Af ter 00R0 has reviewed and approved the procedure, the word approved and the date of approval will be added in the upper right hand corner under the original date. An experiment is not inserted into the reactor until the reactor supervisor has a copy of the approved procedure with these designations in his possession.

The use of reactor produced radioactive materials and associated wastes are handled under the University's broad scope license. All wastes are placed in special waste containers for dry contaminated wastes that are available is all laboratories using unsealed radioisotopes. Normally, metal waste cans with a step-podal operated lid and plastic bag liners are used. All dry waste containers must be conspicuously labeled with a ' Caution Radioactive Material' sign in accordance with the requirements of 10 CFR 20.

Inside liner bags, when filled, are to be sealed and properly labeled with the standard radioactivity caution label and should bear the following additional information:

(1) user's name, department, and authorization number; (2) isotope (s); (3) approximate quantity of activity; and (4) date.

QUESTION 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 pennanent bases for the licensing of your reactor.

ANSWER The Hazards Summary Report was written in 1962 before the reactor was built and when reactor technology was fairly new.

Experimental values were reported after going critical but no final Safety Report was written. The Technical Specifications were written in 1977 incorporating safety limits and limiting safety system settings that were either established by the HSR or that evolved from 15 years experience in operations. The 1986 SAR was written to comply with current regulations using experimental values and conservative assumptions. All of those documents form the bases for the licensing of the PUR-I.

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QUESTION 5 Comments Pertaining to Table 4.1 Table 4.1 was mislabeled in the renewal application and the correct title is Table 3.1.

Since a namber of changes were requested in the table, a revised table is included with the answers for question 5.

I a.

On page 3-3, ' average thermal neutron lifetime' =

-5 i

77.2 x 10 g,,,,,7 7,

,,,,,,g,,,g,,,,,,,,,gg,,

time' = 77.2 psec. Resolve what is the magnitude of this number, what is the origin of this numerical value, and what it should be labeled.

ANSWER The correct term is the ' prompt neutron lifetime' and is equal

~0 to 77.2 10 sec. This is a calonisted value.

b.

We cannot reconcile the maximum or average rate of reactivity 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.

ANSWER The speed of the regulating rod as stated is in error and should be 17.7 in/ min. The reported vaines for the rate of reactivity for the rods came from the Hazards Summary Report and were calculated before the reactor was built. The correct vaines for the rates of reactivity, as measured, are given in the corrected table.

c.

The void coefficient should be stated in the more common terms

)

of percent void, i.e. (Ak/k % void), or a comparison made.

Is the void coefficient given an average, maximum, or other? Is the void coefficient nowhere positive?

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ANSWER The void coefficient restated is -2.6 Ak/k % void. This is a j

measured value averaged over the vertical height of one of the

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coolant channels in one of the four central fuel elements of I

the reactor.

Measurements made in one of the 12 outer fuel

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assemblies indicate that the void coefficient is negative for all regions of the reactor.

i 4

d.

Discuss flax and/or power peaking and hottest fuel channels.

ANSWER Calonistions show that the fuel assembly with the highest flax is located in position F4.

The peak to average finx ratio for this assembly is 1.21.

The fuel plate with the highest finz is located adjacent to the skin safety rod #1 and would only have the highest finz when the rod is at its apper limit.

Under these conditions, the rod channel is filled with water and the ratio of the flax in the fuel plate to the average flax in the L

reactor is 1.45.

e.

What is the temperature dependence of the temperature coefficient? A carve of temperature coef ficient versas moderator temperature wonid be helpful.

Is it negative at all credible temperatures?

ANSWER The reported temperature coefficient is the average value determined from calculations from 27 *C to 80 *C.

The coefficient is almost zero from 27 to 30

• C and then is almost a constant to 80 *C.

In the range below 26 *C, the temperature coefficient is positive. Heat transfer into and out of the pool due to ovaporation of the water and conduction into and out of the surrounding areas maintain the pool temperature between 26 and 27 *C, The water chiller in the water process system is not operated.

f.

What are the ancertainties in the various parameters in Table 4.17 ANSWER Uncertainties are included in the revised Table 3.1 i

Uncertainties in measured valaes are based on the reproducibility of the measurements.

Uncertainties in the calculated vaines are based on estimates of the input data and the model.

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Maximum power level 1 kW Geometry lf t. x! 1ft. x 2ft.

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Maximum excess reactivity s

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Number of plates per standard as'sembly 10 g.

Number of plates per control assembly 6

Plato dimensions (inches) l.

2.76 x 25.12 x 0.060 Active feel length (inches) 23 3/8 i

Enrichment 93%

l.,1 U-235 per plate 16.! su Water gap

.207 inch, Cladding 0.020 alominam Reflector Material on sides Gesphite; q

Number of graphite assemblies 20 Control rods and drives Number of regn1 sting rods 1

Number of shin safety rods 2

Total number of control rods 3

Rod worths (measured)

Regulating rod 2.6 1 06 10' Ak/k Shim safety rod #1 5.0 1 2 10~

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Shim safety rod #2 2.4 1 1 10 Ak/t Rod speed-oct Regulating rod 17.7 in./ min.

Shim safety rods 4.4 in./ min.

Scram-time for complete insertion of Shis rods 1 sec.

Material Regulating rod hollow stainless steel Shim safety rods solid borated staisless steel N

Table 3.1 - 1

l Size Regulating rod (inches) 1/2 x 2 1/4 x 25 1/2 Shim safety rods (inches) 1/ 2 x 2 1/4 x 25 1/ 2 Maximum rate of reactivity change (Withdrawing rods)

Regalating rod 0.6 1 01 10~ Ak/k/sec.

l Shim safety rod #1 3.1 1 1 10~

Ak/k/sec.

Shis safety rod #2 1.3 + 1 10~

Ak/k/sec.

Average rate of reactivity charge (Withdrawing rods)

-4 Regulating rod 0.3 1 005 10 Ak/k/sec.

-4 Shim safety rod #1 1.5 1 06 10 Ak/k/sec.

-4 Shim safety rod #2 0.7 1 03 10 Ak/ k/ se c.

Reactivity effects Temperature coefficient

-4 (calculated)

-2.1 1 5 10 Ak/k per *C.

(measured)

-3.4 10~ Ak/k per C

-4 Void coefficient (measured)

-2.6 1 1 10 Ak/k % void Process water resistivity

>330,000 DEM-CM pH 5.5 1 1 Flow rate 30 GPM r

Table 3.1 - 2

f QUESTION 6 What is the ancertainty and reproducibility in the true valna of the control rod position (s)?

kNSWER The ancertainty in the tr'ne value of the control rod positions is i 1.0 mm, based on the nonlinearity in the measured values of the resistivity of the multiturn potentiometers used in the drive control systems to indicate the rod positions and the measured travel distance between the limit switches. The reproducibility of a given rod position is 10.2 mm and is limite'd by the digital voltmeter used in the system.

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QUESTION 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?

ANSWER-Fuel element accidents are discussed in 7.1.

The accident analyzed concerns physical damage to the fuel assembly and 1

possible exposure of the radioactive material (fission products) to the water. Because of the relatively small mass 4

and weight of the fuel assemblies in water, 2.0 kg, it is unlikely that any damage would occur to the f uel assemblies.

Since the fuel is inside of the fuel cans, it is even less likely that any damage would occur to the fuel plates whereby the fuel seat would be exposed to the water. Assuming that l

2 such an event did occur, and a gash in the clad exposed 1 cm of the weat to the water, and all of the Cs-137 and Sr-90 in the exposed fuel were to dissolve in the water, the concentrations in the water would be:

Cs-137 3.01 10-8,,,,g,,j,,3 Sr-90 2.97 10 p,,,g,, f,,3

-8 1

Both of these concentrations are at least a factor of ten below the limits established in 10 CFR 20 for drinking water.

If a fuel element were to drop into a voter-filled fuel element j

position, there would be no problem since the all fuel handling i

is to be performed with the control rods inserted in the reactor, or the reactor has been made subcritical by removing two or more fuel assemblies.

In either of these two cases, the reactor would still be subcritical.

In the event that all of j

the administrative procedures were violated and the control rods were either withdrawn to their upper limits or removed from the reactor, then the reactor would become supercritical with a maximum reactivity of 0.6 % Ak/k. The consequences of this accident are covered in Sec 7.5 where the step insertion of the maximum reactivity is discussed.

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i QUESTION 8 What are the appropriate limits for non-faeled experiments?

Please discuss the bases for these limits.

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l ANSWER The limitations on experiments are established in Sec 3.5 of the Technical Specifications.

Sections e and f establish the limits on the levels of activity to be produced in experiments Placed in the reactor.

Specific numerical vaines are not stated, but the limits are based on the possible radiation dose i

that might be received either by personnel in the reactor room or by people in adj acent unrestricted areas.

Section 4.5 states that calculations shall be made on samples of known composition to assure that the limits of specification 3.5 f and 3.5.s are not exceeded.

In the event of an sample of anknown composition such as an activation analysis experiment, the sample size shall not exceed 10 grams.

The basis for the 10 gram limitation is the analysis of the hazards of different elements activated for a period of 30 minutes in the irradiation position wi

the maximum flax. The irradiation l

l time of 30 minutes is three time longer than the maximum recommended time for unknown staples.

Calculations show that elemental iodine presents the greatest hazard of non-radioactive elements.

Iodine has only one stable isotope, I-127 which produces I-128 when activated.

I-128 combines a relatively short half-life (25.0 mins.) with a hikh average beta energy per decay, high concentration factor in the body in a small organ, the thyroid, and a relatively large activation cross section.

Using the assumptions that a 10 gram sample breaks and is instantly uniformly dispersed in the reactor room as either a gas or small particles and that the time to vacate the room is 90 seconds, the committed dose equivalent to the thyroid is 1.5 rom, 10% of the annual dose as stated in 10 CFR 20.

Using the assumption that the ventilation system is turned off as the personnel exit the room, the leakage from the roon j

to unrestricted areas will not result in a committed dose equivalent exceeding.15 rem during a two hour period.

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

ANSWER Hard-wired interlocks require individual selection of the rod to be raised.

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

ANSWER Hard-wired interlocks are on both the source drive and the fission chamber drive system, so that only one can be raised at a time.

In order to maintain the minimum count rate on the start-up channel and complete the rod raise circuit, both the source and fission chamber need to be near their lower limits.

No interlock requires this, but by custom, both are left on their respective lower limits.

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QUESTION 11 P. 3-13 Last sentence of section 3.7.6.2 says ' acknowledge j

botton may be used.'

Can operation resume or continue without resetting the abnormal function? Discuss.

l ANSWER Normal operation cannot be resumed natil the fanit is corrected and the acknowledge button silences the buzzer and the annunciator button reset. The setback will continue until the annunciator is reset.

=

QUESTION 12 Environmental Report Section a.

Explain in more detail what is meant by 'not detectable' Ar, 10, and Tritina.

What methods were 41 N

in the cases of used, what are the limits of detection of these methods, and how do they compare with regulatory limits as given in 10 CFR 207 Re 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.

ANSWER a.

Area monitors located in the reactor room have never recorded any unexplainable radiation levels above the lower se t points. The reactor room air is sampled continuously for particulate by a Constant Air Monitor (CAM). Filters are currently changed semi-monthly and area analyzed for gross alpha and beta activity using a windowless 2n sac flow proportional counter.

Our typical MDA for alpha activity is 3 x 10-16* Ci/cc which is well below the most restrictive MPC for an alpha emitter of

-13 6 x 10 C1/ c c.

The most likely source of H-3 would be the pool water. Water semples are taken monthly. The typical MDA is 4 x 10" *pCi/cc which is well below the

-3 most restrictive MCP, for H-3 of 3 x 10 pC1/ c c.

b.

Reactor produced materials and the associate waste is handled under the University's broad scope license. The volume of contaminated material produced by reactor operations generally does not exceed 1 f t / year.

i

• Calculations por Regulatory Guide 4.14.

QUESTION 13 Provide a document.(letter, memo, etc.) expressing the University's commitment to ALARA.

How is it implemented and assured?

ANSWER The University's commitment to ALARA was established in 1951 by the Radiological Control Committee.

Purdue University is committed to a policy of making overy reasonable effort to keep radiation exposures as far below the specified regalatory limits as readily achievable.

Thus, the maderlying philosophy of the Radiological Control operations of the University will be to maintain radiation exposures 'es low as reasonably achievable.'

This philosophy, referred to as 'ALARA,' is in keeping with the recommendations of the National Council on Radiation Protection and Measurements, the National Academy of Sciences-National Research Council, and other independent scientific organizations. The principle of ALARA is also codified as part of the Nuclear Regulatory Commission regulations in Section 20.1(a) of Title 10, Part 20, Code of Federal Regulations, which states that licensees should '----make every reasonable effort to maintain radiation exposures, and releases of radioactive materials in affluents to unrestricted areas, as far below the limits specified in the part as practicable.'

To implement compliance with ALARA, all personnel at the reactor are film badged. To assure compliance, the monthly results are reviewed.

Personnel receiving greater than 100 mrom during any one month are required to explain how the exposure occurred. All exposures greater than 100 mrom are evaluated by the staff of Radiological Control to assure that exposure is kept ALARA.

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6 PURDUE U NIV ERSITY

!s...-d MjL.s November 19, 1986 sicven c. semaino PRESiOENT U.S. Nuclear Regulatory Commission Washington. D.C.

20555 To Whom It May Concern:

The University's commitment to ALARA was established in 1951 by the Radiolog-ical Cor. trol Committee. Purdue University is committed to a policy of making every reasonable effort to keep radiation exposures as far below the specified regulatory limits as readily achievable. Thus, the underlying philosophy of the Radiological Control opera-tions of the University will be to maintain radiation exposures "as low as reasonably achievable". This philosophy, referred to as "ALARA."is in keeping with the recommen-dations of the National Council on Radiation Protection and Measurements, the National Academy of Sciences-National Research Council, and other independent scientific organi-zations. The principle of ALARA is also codified as part of the Nuclear Regulatory Commission regulations in Section 20.1(c) of Title 10. Part 20. Code of Federal Regulations, which states that licensees should "_make every reasonable effort to main'ain radiation exposures, and releases of radioactive materials in effluents to unrestricted areas. as far below the limits specified in the part as practicable".

Sincerely, Q_

>?...4 E

WEST LAFAYETTE. lNDIAN A 479o7

• 317 494 97o8

W QUESTION 14 Who are ' reactor-related personnel' (see section 5.2.3)?

Does this include students, possible experimenters, etc?

ANSWER Reactor-related personnel refers to the reactor supe rvisor, all licensed reactor operators and the electronic technician.

It does not include students or experimenters.

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QUESTION 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 l

provide and discuss.

ANSTER TLD or film badges have been used for continuous area monitoring in the reactor room. The only potential neutron exposure reported was 40 mrea in October 1978.

For the same exposure period, no reactor personnel film badge showed any exposure to neutron, samma or x-ray.

Typical annual exposures (all listed are gamma or X-ray doses)

Reactor Room Class Room (restricted)

(unrestricted) 1985 30 aren 20 mrom 1984 10 mrom M*

1983 M*

M*

1982 130 mrom**

110*

1981 50 mrom 30 mrom

• M:

< 10 mrom x, gamma or thermal neutron

< 20 mrom fast or mod. neutron

< 40 aren beta During 1982, background was not subtracted from the results.

Exposure recorded on 3 badges in nonradiation areas 129 aren,150 arem and 50 mesa respectively, were l

i

QUESTION 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 mor.itors, especially the CAM, are performed.

4 ANSWER The Remote Area Monitor's (RANs) are set to slarm at 7.5 mR/ hr.

This was calculated using the limit set in 10 CFR 1

20 of 3 res per quarter maximum whole body dose and assuming 10 weeks per quarter and 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per work week.

There is no real reason for the selection of set points on the Constant Area Monitor (CAN) as it is intended to be j

used as a yes (there is radiation) or no (there is not radiation) detector. The CAM was not intended to be a quantitative instrument.

Calibration of the instruments is performed semi-annually by Radiological Control. The RANs are calibrated for exposure rate and the CAM is calibrated for detection efficiency.

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QUESTION 17 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 difficnit or impossible, and discuss this. This section also states that a fuel plate inspector's finger badge may indicate

>100 mRes.

What apper limit is accepted, and what is the basis?

ANSWER It is not impossible to estimate whether whole body doses were or were not from the reactor f acility. Doses could be estimated using pocket dosimeters. However, it would be difficult and is not considered necessary as the whole body dose to reactor personnel is generally less than detectable with film badges.

The 100 mRom finger ring dose referred to in the text is considered reactor related as the dose occurs only one time per year during the month when fuel plates are inspected. The acceptable dose is 7.5 rem which is 10% of the current annual allowable dose to hands and forearms per 10 CFR 20.

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QUESTION 18 Section 5.6.1, states that 'no airborne offluents have been released--.' Assaming this means radioactive effluents, what limits and accuracy do you ascribe to this

'zero?'

ANSWER The CAM has never detected an unexplainable level of activity from either the scommalation of particulates or immersion exposure (i.e. immersion gas).

So it is assumed that no ef fluent released to the environment has been contaminated with radioactive materials.

Analysis of the CAM filter generally have an MDA of

~10 3 x 10 pci/cc for particulates which is well below the most restrictive MPC for an unknown alpha emitter of

-13 6 x 10 pC1/ cc.

i

QUESTION 19 Section 5.8, please show how you arrive at the 'less than 1 mRom/yr' figure.

l ANSWER The maximum potential dose to persons not related to reactor operations is (1 mrom per year. This is based on film badge data of remotor personnel and the classroom above the reactor (estimated weekly time / person less than 5 hrs./ week).

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

ANSWER Section 6.2 should be reworded to indicate that CORO will review to determine if an item is an unreviewed safety question and needs to be approved by the NRC.

t

QUESTION 21 P. 7-2, How was the value of 0.M Ak/k determined for the 5

drop tabe?

ANSWER The 0.M Ak/k for the 5

drop tabe was determined experimentally, i

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QUESTION 22 What maximme temperature rise in the f uel 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.

ANSWER Assuming a loss of coolant accident immediately following a long (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) reactor run, the power level of the reactor due to the decay of the fission products is about 65 watts. The power level would drop rapidly and reach a power of.87 watts in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Assaming no heat transfer from the f uel plates, the temperature rise would be 9.5 C in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

If heat a

conduction to the aluminam fuel assembly cans is assumed, the temperature rise would be even smaller.

The radiation from the reactor following a loss of coolant accident would be confined to a broad beam pointed toward the ceiling. Radiation levels in the classroom above the reactor are calculated to be less than 4 mR/hr.

In such an event, the classroom would be evacuated and would remain empty until the problem had been resolved.

The radiation level would drop rapidly as the activity of the fission products decayed and reach levels below 1 mR/hr within a few hours.

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QUESTION 23 Section 7.5.3 s.

In the first paragraph you state no credit for negative temperature feedback is taken.

Please estimate the magnitude of this feedback during the postulated scena rio, and its likely effect oa the consequences.

ANSWER The temperature feedback was not taken into accoast in the event that the excursion is tenminated by the action of the safety systems because of the short time of the event.

The major component of the temperature coefficient is moderator expansion and during the short excursion there is no significant heat transferred to the moderator.

b.

You mention a design of PUR-I for 10kW.

Please explain the basis of the design limitation, and the impact on the consequences of this postulated accident.

ANSWER The PUR-I reactor is very conservatively designed. The design limit for 10kW operation was based, not upon performance of the fuel, but rather on the radiation shielding provided and the radiation exposure to both the operating personnel and the classroom above the reactor.

The average heat finz of the PUR-I reactor is about 0.01 watts /cm. Fuel plates of this type are routinely operated at average heat finz levels of 1.6 watts /cm with natural convection cooling (U of Missouri, Rolla).

2 An average heat finx of 1.6 watts /cm would correspond to a power level of about 160 kW for the PUR-I reactor, i

Therefore, events where the safety system initiates a scram will result in an event that is well within the design of the reactor.

c.

In the second paragraph you indicate a maximum power level of 360 kilowatts while Table 7.3 gives 380 kW.

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 trae value, (3) mention the direct radiation levels, for example in adjacent classrooms, resulting from this hypothetical accident.

ANSWER The 360 kW 1evel mentioned in the second paragraph is the power level reached three minutes into the excarsion.

The power level would continae to rise and level at about 3 80 kW.

At a power level of 380 kW, calonistions indicate that the centerline temperature of the fuel would reach about l

140 'C with the clad temperature only slightly less at about 135 'C.

This is well below the reported clad i

melting temperature of 660 'C.

At these temperatures, nucleate boiling could be expected in the coolant, introducing voiding which wonid introduce an additional negative reactivity to the reactor.

Two temperature coefficients were given; one, a measured value and one, a calculated value. The measured value is not typical of the temperature coefficient applicable to the accident being analyzed since the experiment involved raising the temperature of the reactor pool. Thus the

' measured' temperature coef ficient is for the situation where both the reactor and pool water are at a uniform temperature.

In the accident analyzed, the reactor pool is assumed to be at a constant temperature and only the reactor fuel plates and the water in the reactor is heated and increasing in temperature. The calculated temperature coefficient used in the analysis is assumed to be the correct one.

Only a calculated temperature coefficient was used in the transient calculations since the experiments required to make the correct measurements wonid involve power levels exceeding the licensed power level.

With respect to the radiation level in the adj acent classroom, the radiation level in the classroom over the reactor wonid reach about 15 mR/hr at a reactor power of 380 kW.

This assumes that the water level in the pool is maintained at a level of 13 feet over the

core. Assuming that the building could be completely evacasted in 15 minutes or less, the maximos dose equivalent to people in non-restricted areas is less than 4 mRes.

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QUBSTION 24 Section 7.6 a.

What is the basis of choosing 100m as the ' release point' in the unrestricted area?

ANSWER The calculation of the dose equivalent was for a person in the plume,100 m from the point of release. The valas of 100 m was chosen as the typical distance that is required for the dispersion of a plume to reach values predicted by the equations used in the calculations.

It was also assamed that although this distance may involve areas outside the boundaries of the University cmapas, these areas could be readily controlled and evacuated by the University Police.

)

b.

Please clarify whether the thyroid doses shown are l

delivered within the time of exposure to the airborne radionuclides, or are they committed doses on into the

?

future.

What do you mean by thyroid dose rate?

ANSWER The thyroid dosee shown are committed radiation dose equivalents that would be received by being in the radiation plane or the reactor room for the assmaed times.

c.

For non power reactors,10 CFR 100 is generally not 4

applicable.

What is meant by 'the determination of an excInsion area' (page 7-13)?

l ANSWER The use of the term ' exclusion area' and reference to 10 CFR 100 was used only to provide a reference for i

comparison to the relatively low dose equivalent calculated for a person at this distance.

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d.

The last sentemos os page 7-13 seems to be imoonplete.

Please clarify.

ANSWER The last sentence on page 7-13 should read: It is conoladed that the experiments using fissile material osa be radiated in the PUR-I within the power limits (1 watt) i smalysed in this section.

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QUESTION 25 Please provide information on your program for preventive maintenance, which might help give ass 2rance of operability of systems and components.

ANSWER Periodic checks of key voltages and an annual electronic calibration by the electronic technician is the main preventive maintenance for the reactor instrumentation.

In addition, during the prestartup check certain voltages, set points, and calibrations are again checked.

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QUESTION 26 Please provide responses to the omolosed qsestion related to your revised Operator Regnalification Program, transmitted to as by letter dated September 19, 1986.

l ANSWER A revised Operator Regaalification Program dated November 1986 is submitted as Enclosure 3.

It addresses the comoeras ir. your Enclosure 2.

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QUESTION 27 Technical Snacification 2333 1 What is the basis for choosing more than approximately 0.5% Ak/k addition to a sabcritical reactor as the catoff for a Reportable Ocontrence?

I ANSWER The basis for choosing 'more than approximately 0.5%

Ak/k addition to a subcritical reactor' was the definition for a Reportable Occurrence from the draf t of the publication Guidance for Non-Power Reactors, Section 6.0, Administrative Controls, supplied by the N.R.C. at the time the Technical Specifications were being written.

It was a limitation that was not too restrictive for Purduo's operation.

233a 1 50 KW is an arbitrary ' Safety Limit.'

A more quantitative Technical basis for the Safety Limits i

should be given.

I ANSWER The average heat flux of the PUR-I reactor is about 2

0.01 watts /cm. Fuel plates of this type are routinel operated at average heat flux levels of 1.6 watts /cm with natural convection cooling (U of

+

l Missouri, Rolla). An average host flux of 1.6 2

watts /cm would correspond to a power level of about 160 kW for the PUR-I reactor.

Since the PUR-I is licensed of only 1 kW, it can be predicted with much confidence that no damage to the fuel elements will occur, which is the objective of this specification.

2331 1 What is the reactivity worth of most worthy fuel element? Which position in the core, and what is the source of these data?

ANSWER The reactivity worth of the most worthy fuel element l

is 5.5% Ak/k. This is not only due to its geometric l

position in the core as one of three central fuel elements, but also the fact that it has one more fuel l

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2333 H Calos1sted experiment worth priorjto insertion 12 the teactor shon1d also satisfy ez';eriment re'sotiv,ity criterion.

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ANSWER Specification 4.1.c on page.18 shunid r,ead:

J The calculated vaine of 's, movable and ansecured experiment shall not exceed 0.003 Ak/k and the total worth of all secured experiments shall not exceed 0.005 Ak/k. The reactivity worth of esseriments will' be experimentally verifiad daring the first startup i

after insertion.

t Zaga n 15 months is a long interval between drop time check for the Shim-Safety rods. % Basis'is:

' consistent for 14 years since the PUR-I was built.'

Page 3-1 of the SAR states the reactor'eas built 24 years ago.

Some other check intervals need to be reviewed too.

l

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s.

ANSWER Shin-safety rod drop times were measured quarterly; [

astil 1977. Approximately 15 years of rod drop tines showed no trends toward longer times, so Purdue was

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advised that it would its permissable to measste them after the annual inspection or any control rod maintenance that might af fect-the drop ' times.

Purdae is watching for any trends aoward longer drop times.

The apparent discrepancy in t.ine intervals has to do with the fact that the Technical Specif1 actions were written in 1976 and 1977, while the Safety Analysis t

Report was written in 1906.

Zaga H The basis for limitin't the mass of samples of unknown composition to 10 grams is 'Past Experience.' This past experience should be referenced, and a technical

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basis also given.

ANSWER The basis for the 10 gram limitation is the analysis of the hazards of different elements activated for a period of 30 minates in the irradiation position with the maximum fluz. The irradiation time of 30 minates is three times longer than the maximas recommended time for unknown samples.

Calonistions show that elemental iodine presents the greatest hazard of non-radioactive elements.

Iodine has only one stable isotope, I-127, which produces I-128 when activated.

I-128 combines a relatively short half-life (25.0 mins.) with a high average beta energy per decay, high concentration factor in the body in a small orgen, the thyroid, and a relatively large activation cross section. Using the assamptions that a 10 gram sample breaks and is instantly uniformly dispersed in the reactor room as either a gas or sen11 particles and that the time to vacate the room is 90 seconds, the committed dose to the thyroid is 1.5 rom, 10% of the annual dose as stated in 10 CFR 20.

Using the assumption that the ventilation system is turned off as the personnel exit the room, the leakage from the room to unrestricted areas, combined with the short half-life will not result in committed doses exceeding.15 rem during a two-hour period.

2333 11 If only a quorou 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.

ANSWER The minimum number of members of CORO is seven with a q:orum consisting of not less than four members.

At present, it is impossible for a majority of any quorum to :onsist of personnel directly concerned with administratica or operation of the reactor.

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r 2111 11 In addition to a log book entry (step 6.47) it is assnaed that the person performing the procedure step initial that the step has been successfully conpleted.

Reference your written procedure that provides for this action.

ANSWER Operators do not initial each step as it is conpleted. The operator signs in as his/her l

operation starts, and all entries in the same handwriting is attributed to that operator antil I

another operator signs in.

E111 11 Step J.5.2e.

When an updated drawing has been placed

=/

in the file, what is the fate of the now outdated drawing?

ANSWER Pnedue has few changes to drawings, but at present the changes are narked and dated on the current drawing.

If a conplete drawing becones obsolete, it would be prominently noted, dated, and retained in an obsolete file.

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Missing Pages From Original SAR

Prevailing winds are from the west or southwest during the winter and from the soutt daring the summer.

Wind velocity is highest in February and lowest in August.1 1

2.5.2 Weather Laf ayette uses weather data from Weir Cook Municipal Airport, Indianapolis, which is about 60 miles distant, and has similar conditions. Table 2.3 shows the wind conditions at Weir Cook and indicates an annual mean wind speed of 7.7 miles per hour and a maximum wind speed of 49 miles per hour. According to the Unified Building Code,1985 edition, the Purdue University lies in the maximas wind zone of 80 miles per hour, which translates to a wind load of 17 pounds per squa:e foot. Buildings at Purdue University are designed to withstand this wind load.

2.5.3 Severe Weather This region of the United States is subj ected to tornado activity, primarily during the late Spring and early Sammer months. Table 2.4 shows the frequency of tornados for the period from 1950 through 1985. Twenty-five tornados occurred over a thirty-five year period, which averages to less than one per year. The probability of damage due to a tornado is minimal.

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2.6 References I

t 1.

Maarouf, Abdelrahman M., and Melhorn, Wilton N., ' Technical Report No.61', Purdue University Water Resources Research Center, June, i

1975.

2.

Ulrich, H.P.,

Barnes, T.E.,

and Krantz, B.A.,

' Soil Survey, 2-6

c Tippesanoe Cosaty. Indiana Series 1940, No.22',1959.

3.

Reseasheim, Joseph S., 'Grosad-Tater Resources of Tippecanoe Cosaty, Indiana', State of Indiana, Indiana Department of Conservation. Division of Water Resources Balletin No.8,1958 4.

Braile, L.T. Professor, Pardse University, private cosassication.

G 2-7

1 Table 2.1 Population Distribution Around Purdue University Radius (miles)

Total Population Inoramental Increase with Increased Radius 1.25 29,550 29,550 2.5 58,973 29,423 5

100,277 41,304 10 117,7i2 16,935 1 Values in this table were supplied by the Tippecanoe County Area Plan Commission, and are based on the 1980 Census Summary.

i 2-8

Table 2.2 MOH71tl.Y NEAN VAI.UES OF CLINATIC EI.ENENTS FOR PERIOD 1953-1970 Temp _erainre (*F)J F.

.N

,A N

J.

.J A

S 0

N D

Year Daily. max.imum 31.5 36 45.3 60.9 71.2 80.6 8_4. 5 82.8 77.4 65.9 49.9 36.5 60.2 Dolly mipipsum 14.4 18.1 27.0 40.3_

49.4 58.5_

6.2. 0 59.7 52.2 42.0 31.1 20.3 39.6 Ne_a_m.of the. day _

23_.0 27.1 36_.2 50.6 60.3_

69.6 73.3 71.3 64.8 54.0 40.5 28.4 49.9 Absolute maximum 65 70 80 85 92 98 1.6 96 98 90 78 66 1 06 Year 1967 1954 1963 1960 1964 1954 1954 1954 1954 1963 1961 1966 July 1954 Absolute minimum

-18

-23

-12 20 26 39 43 35 32 20

-1

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-23 Y e s.r.

1.970 1963 1960 1969_

1968 J9.66 1963 1965 195.9 1962 1958 1963 Feb. 1963 Total precipitation 1,81 1.41 2.22 4.31 3.94 3.85 4.74 3.38 2.62 2.69 2.20 2.01 35.68

_ inches).

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'Y Snow sad sleet 4.8 5.9 3.5 0.7 0

0 0

0 0

T 1.7 3.9 20.5

_ 1.nches)

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% Relative Humi-46, 70 59 55 50 52 52 50 52 49 49 64 69 56 Da_11y Minimum *,,

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8.6 38 February W

9.0 43 March W

9.1 38 April SW 9.0 42

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6.6 43 October S

7.0 36 November W

8.2 43 December T

8.5 49 Annual SW 7.7 49 2-10

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' fable 2.4 Tornado Frequency, Tippecanoe County, Indiana 1950-1985 Year Number of Tornadoes 1953 2

1956 1

1961 1

1963 2

1965 2

1967 1

1968 1

1971 1

1973 3

1974 2

1976 2

1978 5

1980 1

1981 1

. Total 25 2-11

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PRESIDENT PuRoue IJNivERSITY Ii_________________

RADIOLOGICAL CONTROL DEAN SCll00LS OF ENGINEERING j

COMMITTEE I

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RADIOLOGICAL llEAD i

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CONTROL OFFICER

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Scil 00L OF iluCLEAR ENGINEERING g

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COMMITTEE ON 8

REACTOR OPERATIONS Y

l REACTOR i

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.SilPERVISOR 8

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REACTOR OPERATIONS PriniariIy Adsninisiration Prlinarily Safety Figure 6.1 Organizational Structure

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l Revised Operator Requalification Program

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'O November 1986 I

OPERATOR RFAUALIFICATION PROGRAM for the PUR-1 REACIOR FACILITY This program is designed to comply with the intent of 10 CFR 55, Appendix A,

concerning the continned training and regnalification of l

operators for the PUR-1 reactor.

It will be mandatory for all operators 1

licensed on the PUR-1 reactor to participate in the program.

l The regnalification program will consist of the following parts:

i A.

A series of eight meetings will be held over a two year period, l

during which all topics listed below in part A.1.b. will be l

covered.

1.

Each meeting will consist of:

a.

A review of reactor operations and modifications, if

any, b.

A lecture of one or more of the following topics:

l 1.

Theory and principles of operation.

ii.

General and specific plant operating characteristics, iii. Plant instrumentations and control systems.

iv.

Plant protection systems.

v.

Engineered safety systems, vi.

No rma l, abnorm al, and emergency operating procedures.

vii.

Radiation control and safety.

B 2

viii. Technical specifications.

ix.

Applicable portions of Title 10, Chapter I, Code of Federal Regulations.

2.

The lectures will be given by the reactor operators, senior operators, aniversity radiation control officers, or facnity members of the School of Nuclear Engineering.

B.

Completion of the biennial operator requalification program will consist of a written examination and a demonstration of operator proficiency in reactor operation.

1.

Written examination:

a.

One of the senior operators will be exempt from taking the examination.

This senior operator will make up and administer the examination to all other operators and senior operators. The senior operator may receive assistance for making up questions on the topics in part A.1.d.

from the instractor for each topic.

The senior operator exemption will rotate through the entire senior operator roster.

b.

Any person who scores less than 70%, overall, on the examination will be relieved from his licensed duties and enrolled in an accelerated program antil such time as he can satisf actorily pass an examination covering the material. The coarse content and duration will depend upon the individual's deficiencies.

2.

Operator proficiency:

a.

The exempt senior operator will also administer an operator proficiency examina tion to all other operators and senior operators.

b.

Any person who could not demonstrate proficient operation of the reactor would be relieved of his licensed duties antil such time as proficient operation could be demonstrated.

e e

C.

Each operator will be required to make a minimum of 10 reactor startups or power level changes during the two year period covered

]

by his license, j

l l

D.

Each reactor operator and senior operator will annas 117 review the contents of the operating manual, technical specifications, and the emergency procedures. A signed statement to this fact will be kept in the regnalification file for a period of four years.

E.

Records will be maintained to document each instructor, each topic discussed, each licensed operator's and senior operator's participation in the requalification program.

The records will contain copies of each written exam, answer sheets, results of evaluation, and the biennial operator proficiency demonstration.

Documentation of additional training and test required for individuals exhibiting deficiencies will also be included in the files. All records will be maintained by the training coordinator for a period of four years.

F.

During intervals when the licensed operations crew consists only of senior operators who are instructors for topics in part A.1.b.,

the roqualification program will be modified to exempt those senior operators from parts A and B.1.

Parts B.2, C, D, and E will remain in effect.

When the licensed operations crew increases to include those who do not instruct in the program, the program will revert to its initial content.

_