ML20213G288

Forwards Comments on Reactor Operator & Senior Reactor Operator Exams Administered on 870327

ML20213G288
Person / Time
Site:	05000054
Issue date:	04/02/1987
From:	Ruzicka W CINTICHEM, INC.
To:	Coe D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML20213G260	List: IR 05000054/1987002 ML20213G258 ML20213G288
References
NUDOCS 8705180220
Download: ML20213G288 (17)

Text

Arrach ment .5 1

.; }

ClNTICHEM, INC.

a whoHy owrwed subsdory of Medi-Physics, Inc. P.O BOX 81G. TUXEDO. NEW YORK 109G7 1914)3 51 2131 AprIl 2, 1987 Mr. Douglas Coe U.S. Nuclear Regulatory Commission 631 Park Avenue-King of Prussia, PA 19406

Dear Mr. Coe:

We have quickly reviewed the R.O. and S.R.O. exams you administered on March 27th. Aitached are our comments. Some of-these comments were discussed with you after the test. Call me lf you have questions.

Sincerely,

', ,, Q -- ,,. " X WIIIlam G. Ruzicka Manager, Nuclear Operations WGR:Jb Attachment 9705180220 870 PDR ADOCK 050 54 Y PDR

R.O. and S.R.O. Exam March 27, 1987 Cintichem Comments A.02/H.01 The question is poorly worded. Two possible Interpretations could be made by a knowledgeable candidate.

Interpretation 1: Assumption that a K excess measurement was made in error and then redetermined correctly at a 10'F higher temperature. This leads to an answer of K excess decreases.

Interpretation 2: Assumption that aK excess measurement was made and then a second K excess measurement was made at a 10'F lower temperature. This leads to an answer of K excess increases.

To be more clear, the question should have been defined as K excess 1 and temperature 1 versus K excess 2 and temperature 2 The question as written is more one of how to Interpret the time frame and "had been" logic of the question than of understanding the negative temperature coef ficient. If an assumption of higher or lower temperature leads one to the correct calculation, the NRC should give full credit.

A.03/H.03 The question could be easily misinterpreted. The NRC answer wanted the candidate to assume that the "Inaccurately low reading" meant that the reading is lower than the actual flow. If the candidate Interpreted the question to mean the actual flow was lower than the reading and then proceeded to perform the calculation correctly, he should be given credit. The question should have more clearly spelled out the starting paramters.

A.04/H.04 The answer key answer is inaccurate. The small 10

• F temperature change would be negligible and would not be apparent on the instrumentation. Also,tthe answer says that fast neutrons are measured. Our boron coated chambers measure thermal neutrons.

A.06/H.07 The question does a poor Job of measuring a candidate's knowledge of Xenon poisoning. An essay type question asking about Xenon production and decay would more accurately gauge a candidate's knowledge. Answer C, for example, could have been chosen by a knowledgeable candidate as lodine production is greatly reduced and Xenon quantity is greatly increased due to the reduction of neutron flux. Xenon, of course, goes up with reduced flux because of less burn up.

The words " production" and " quantity" are similar and could lead a knowledgeable candidate to choose answer C.

'A.07 The answer. key states that " moving the fission chamber changes the reactivity of the core". This change is negligible and has not been discussed .In our training sessions concerning this procedure.

B.04 A knowledgeable candidate reading this question could assume that "no trips or Interlocks" included the phlenum leak alarm. The phlenum leak is a trip device. in fact, the NRC answer key states phlenum leak alarm will be " actuated" or tripped. If.the candidate assumes that "no trips" includes the phlenum leak alarm, he should be given credit for other answers such a r. core boiling effects seen on nuclear instruments and/or an erratic Delta T. the question should be reworded to remove this ambiguity.

C.05 The answer key includes inaccurate information. The bulk pool temperature sensor is too close to the pool surface to be.affected by any shttdown reactor cooling ef fects. The core Delta T is read out in Delta T after subtracting core T '

Inlet from core T outlet. No inlet temperature reading by Itself is provided to the operator.

I f, insuf f icient core heat removal caused in core boiling, this would be seen on the Instrumentation. Following a loss of electrical power, though, gravity flow is still provided to the core, with the flapper shut, for many minutes until the hold up tank is filled. Therefore, the reactor would be shut down for many minutes before downward gravity flow stopped, and by this time, the decay heat of the core is greatly reduced therefore reducing the likelihood of boiling.

The NRC's ES202 D16 states that: " Vague open ended questions should be avoided. if a specific number of responses are required, the question should clearly state that expectation so the candidate will know when the answer is complete." Question C5 did not specifically call for two answers.

D.02 The answer key lists all reverses as one of the four answers. If reverses are listed separately, they could consititute four of the four required answers. The question does not spect fy that " reverses" can be used only as one answer.

0.04/J.05 The answer key is incorrect:

Delete ">.001 log N" after 30 sec period log N.

i 05-01-1 also includes "LCR in calibrate ". i-E.01 All scrams are operable obove 2NL including:

l

a. Guide tube lift; f

i

. 12" low pool level;

b. l Bridge unlocked; c.

d. -Console key switch.

Also, answer four in 'the answer key should state ">2NL" not

"<2NL".

F.03/L.01a. The answer key is correct. Another correct answer is that a shutdown is required as we are not allowed to be critical with rods-below 12 inches (RM-03-5).

G.02a/J.03.a,b Another consideration not mentioned, but which could impact the quarterly dose limit, is Beta dose to the eyes, if safety glasses were not assumed to be worn, Beta to the eyes could be a factor. A 3R/ quarter versus a 1.25R/ quarter starting assumption should also be allowed.

G.02b/l.03c Another correct answer not mentioned in the answer key is that GM tubes saturate in high radiation fields. This will result in inaccurate readings, and therefore, GMs should not be used In these situations.

G.04/l.05 The answer key reference, RM-07-3, states that authority to activate the emergency vent system is not limited to specif ic operations in HP personnel . Also, EP-03-02 states that any reactor operations personnel and the HP supervisor has the authority to initiate an evacuation.

H.02 The NRC's ES202-D16 states: " Vague open ended questions should be evolded. If a specific number of responses are required, the question should clearly state that expectation so the candidate will know when the answer is complete."

The third answer was not clearly solicited.

Another consideration, not mentioned in the answer key, Is the change a ref lected versus a non reflected core would have on the Technical Specification peak to average 'f lux ratio.

A final comment on this question is that it poorly assesses the candidate's knowledge of the required shutdown nergin considerations the NRC wanted as per the ansewr key. The excess reactivity basis in the original Technical Specifications and in Section C of the Final Hazard Summary Report (attached) do not discuss shutdown margins. Instead they discuss the reactivity required for operation including beam tube and fuel effects. If the NRC desired to determine a candidate's knowledge level on the Technical Specification 3.1 shutdown margin requiremnts, a more stralghtforward question should have been asked. This would have been more appropriate than requiring an answer on the Technical Specification basis of excess reactivity which, in our Technical Specifications, is underdeveloped and imprecise.

4

. l.04 The answer key ref erence RM-06-4 lists three tasks. The question asks for four. Also, there are numerous tasks which require HP coverage besides the three Iisted in RM 4, any of which should be considered correct.-

J.03 The answer key is incorrect. Change the word " downward" In answer 1 to " forward". See RM-06-4.

J.04b The NRC's ES202-D16 states that: " Vague open ended questions should be evolded. If a specific number of responses are required, the question should clearly state that expecation so the candidate will know when the answer

j. Is complete." J.04b did not specifically ask for two answers. Also, there is another method to change rod worths and that is changing out a rod with one made of a different material.

J.06 The f ive actions listed in the answer key are correct but they are not the only correct answers. The Xenon of f gas vent would also close and the thermal column exhaust fan is also the bean tube exhaust fan.

K.02 The answer key is Incorrect. RM-10-6 only requires an SRO to be in direct supervision, not necessarily on the bridge.

K.05 The NRC's ES202-016 states that: " Vague open ended questions should be evolded. If a specific number of responses are requireed, the question should clearly state that expectation so the candidate will know when the answer 4 is complete." Question K.05 did not specifically ask for two answers and two should not be required. There are other correct answers also, such as our fuel has only been tested and qualIf led to certein iImits.

K.06 The answer key referred to Question Bank K14. This is incorrect; 11 should be K11. The question asks for "Immediate ections". Many of the immediate actions listed, such as reentering the building, determining hazards, and c colling people on the emergency call list, are subsequent actions, not immediate actions.

i

. An exam candidate also should not be required to memorize i l all the answers in a bank of questions. A candidate should use his knowledge and experience to respond generally to an i action question such as K6, and there should be latitudo used in the grading of the answer.

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s SECTION C REACTOR DESIGN 1 CORE PHYSICS

a. _I_ntroduc tion The UCNC reactor will be a 5 MW (heat) pool-type reactor employing The water reflected MTR type fuel elements, as described in Section B.S.

core will contain up to 38 fuel elements, depending on the degree of burn-up.

When the graphite elements are used, the reflector will be both graphite and demineralized light water. The coolant will be demineralized light water.

The reactor core will have two operating positions; one in the open water of the pool, and the other in the stall position into which a thermal columnWhe end six beam tubes converge.

tended that the reactor will operate continuously for ten day periods between shut-downs.

In the following sections the thermal and nuclear characteristics of the reactor core are presented in detail. The method of calculation in-volved in these evaluations is given in Appendix 1

b. Table of Thermal Characteristics for Water Reflected Core 5,000 KW, (heat)

Total Power 1,090 KW/Kg U-235 Specific Power (average)

Average Water Temperature:

68 Cold 108 0 Design 13 n/cm2 sec.

Therumi Neutron Flux (Average) 3.1 x 10 (Stall Position) l L .

c.

Core Rsactivity Effects TABLE I NUCLEAR CHARACTERISTICS, Stall Pool Item A K/K AK/K M rea'd, Temperature effect 0044 0044 49g Control 0030 0030 34g ExperLmental Facilities (in core) 0200 0200 226g Beam Tubes (6) (2 rabbit tubes) 0180 204g Burn-up + Low cross section fission products (10 days) 0076 0076 86g Equilibrium Xenon 135 .0338 0338 382g Partial Samarium 149 (10 days) 0076 0076 Total excess reactivity required ._ ,

for operation 0944 0764 Effect of additional fuel above requirements (see item 9) __.0076 0057 Max, excessliK/K to be controlled j by 5 shim safety rods and 1 regulating rod .1020 0821 oK/K Equilibrium Samarium 149 0095 Temperature coefficient (average for range from 200 C to 42 C) -2.0x10 -4 6 K/K per O C.

Cantrol rod worths tsK/K I 5 shim safety rods i Average worth / rod 0328 "otal worth .164 Regulating rod 006 Thermal column worth 0334 (mass equivalent =

378g U-235)

I Mass Equivalence - one g U-235 8. 84 x10-5 Neutron lifetime 5.1x10-5 seconds

, Water Reflected Graphite Reflected (3")

( Id Clean Critical Mass:

Stall 3.45 Kg U-235 2.12 Kg U-235 l Pool 3.83 Kg U-235 2.35 Kg U-235 l Operating Critical K1ss with beam tubes air filled:

Stall 4.52 Kg U-235 Pool 4.64 Kg U-235 l

l

(1) Temp;rcture end void Coefficients of Reactivity The temperature coefficient of reactivity was calculated 0

to be

-2.0 x 10-4 21 K/K per degree C for the temperature range of 20 C to 42 C.

This value compares favorably with measured temperature coefficients for the BSR, Borax I, and the IRL as given below:

Reactor Tempernture Range 6 K/K 0C. Reference Borax I 34.40C-550C -1.1x10-4 6 4 19 BSR 200C-550C -0.8x10 4 IRL 22.80C-25.1 C -3.9x10' 20 At the IRL, measurements of the temperature coefficient were made within a narrow temperature range at near room temperature. This measurement point was chosen to determine the coef ficient accurately at the lower temperatures of operation, where the temperature coefficient would be cxpected to be less negative. For this test, transient conditions were cimulated by elevating the temperature of the core at a more rapid rate than that of the bulk pool water. It was found that at room temperature the reactivity change with rising core temperature remained substantially negative and at a constant value over the narrow temperature range investi-gated. It is expected that over the entire operating range the temperature coefficient would remain the same or be more negative than the value given cbove. Similar tests over a wider temperature range will be made with the UCNC reactor during initial phases of operation.

For a temperature rise of 220C in the UCNC reactor (from cold to design water temperature), the required excess reactivity to compensate for temperature was calculated to be 0.44% oK/K. The required addition of U-235 to give this excess reactivity was calculated to be 49 grams of U-235.

The void coefficient of reactivity for the UCNC reactor is esti-mated to be between -0.27% kegf/% void, the value calculated for the Borax I Reactor and -0.44% keff/% void, the value measured for the IRL Reactor.

The UCNC void coefficient should be nearer the IAL value due to the greater similarity between these two reactors.

(2) Neutron Lifetime and Reactor Period The effective neutron lifetime for the UCNC core has been cal-culated to be 5.1x10-5 seconds. The computed infinite lifetime was 5.2x 10-5 seconds. Figure 13 is a plot of the stable reactor period response to otep changes in reactivity.

(3) Control and Experimental Reactivity Allowance For reasons of safety, the total excess reactivity allowance for experiments in the core has been limited to 2.0% d6K/K. This is equivalent to approximately 226 grams of U-235 To allow for the best control by the regulating rod, it was desired to have this rod in its most effective position during reactor operation.

With this regulating rod inserted approximately half way, an excess reactivity of .3% AK/K would be required. This is equivalent to approximately 34 grams of uranium 235

l .

(4) Bram Tubt Requirem nts The UCNC reactor has four 6-inch and two 8-inch diameter beam tubes. The beam tubes converge on the core in the stall position. These beam tubes can be either flooded with light water or filled with air. The excess reactivity requirement of the four 6-inch beam tubes when air filled

- is calculated as +0.90%6K/K, and +0.90%6K/K for the two 8-inch beam tubes.

The equivalent mass required to give the excess reactivity required for the beam tubes is 204 grams of uranium 235.

l There are in addition to the beam tubes two 1-1/2 inch diameter rabbits. The rabbits run tangent to the side faces of the core.

The total excess reactivity allowance for all beam tubes and two rabbit tubes is 1.80%6 K/K. The mass which must be added to give 1.80%6K/K is 204 grams of uranium 235.

IRL measurements indicate that the reactivity effects of the beam tubes and pneumatic tubes in the UCNC reactor may be smaller than predicted by design calculations. The dif ference between the calculated and the mea-sured values depend on the core loading and the relative positions of the fuel elements and the experimental facilities at the time of measurement.

IRL measurements for the water-air effect of all beam tubes for two separate loadings were -0.954% AK and -1.005%6 K respectively. The effect on re-activity of several materials placed in the pneumatic tubes was too small to be measured.

IRL also measured the reactivity effects of the beam tubes when the core was graphite reflected. The reactivity loss due to the beam tubes going from the water filled to the air filled condition was measured to be

-0.52%AK or about one-half the ef fect measured for the water reficcted core.

(5) Uranium-Burn-Up and Low Cross Section Fission Products For ten days operation at a heat output of 5000 KW, an excess reactivity of 0.76%LK/K will be required to compensate for the loss of uranium 235 atoms by burn-up, and the accumulation of low cross section j fission products. This excess reactivity is equivalent to 86 grams of uranium 235 (6) Xenon 135 noisoning Allowance l At a heat will be about 3.1x10gutput of 3 n/cm2 5000 KW sec. At this the average flux 1cvel, thermal flux in the the excess core reactivity to compensate for equilibrium xenon is calculated to be 3.38%6 K/K. This value compares very well with the value of 3.6% AK/K measured at IRL af ter 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> operation at 5000 KW. It is estimated that it would require three days, with no initial xenon present, to reach equilibrium level. The mass of uranium required to compensate for equilibrium xenon is calculated

! to be 382 grams.

Af ter three days of continuous operation at 5000 KW, the maximum xenon over-ride Icvel is calculated to be approximately 8.01%bK/K occur-ring at about nine hours after shut-down.

26-

Although no exceso reactivity for xenon over-ride is provided, there is sufficient reactivity available for control purposes to anow a minimum " grace" period of about 15 minutes after shut-down.

(7) Samarium 149 Poison Allowance Samarium 149 is a stable isotope fomed during the fission pro-cess; its equilibritan level is independent of the core neutron flux. It is determined that the excess reactivity required for equilibritan samaritan 149 during operation is 0 95% M/K. During the first ten days of operation, however, only approximately 0 76% M/K cxcess reactivity win be required.

After the first period (10 days), an additional O.19% M/K vill be required for further operation.

(8) '1hemal Column Effect In the stall position, the presence of the graphite thermal column reducco the cold clean critical mass below that required for a totally light water reflected core. It is eniculated that the thermal column has the effect of increasing the core reactivity by about 3 34% M/K, or decreasing the critical core loading by 378 grams of U-235 IRL measurements indicate that the positive effects of the themal column on reactivity is smaller in the UCNC reactor than predicted by design calculations. The measurements were made by assembling identical lattices in the open pool and the stall positions and observing the change in core re-activity. Depending on the core loading and configuration, the reactivity change going from open pool to stall position varied from +1.47% M/K to

+176% M/K.

The effect on the reactivity of the core by motion of the tower with respect to the themal column vill be eliminated by placing the core support tower under slight strain against the thermal column when the core is placed in the stall operating pocition.

(9) critical Maca and core Ioading The cold clean critical maan of a light water reflected core con-taining 10.89 grams of uranium 235 per plate, in calculated to be 3830 grams of uranium 235 The reactor will operate in either the open pool or stall position with different reactivity requirements, however, the effect of the thermal coltann in the otall position maken the operating mass in the two positions approximately the came.

The minimum operating loading for the stall position with an ex-cess reactivity of 9 44% M/K vill be 4 52 kilograma of uranium 235 The smallest increment of fuel addition which can be added to the core vill be a partial element containing 98 grams of uranitan 235; the actual initial core loading, for stnll pocitions, vill consist of six (6) l control elements and tventy (20) standard elements, plus one partial fuel element to give a total loading of 4.606 kilograms of uranium 235 For the open pool position, the cold clean critical mass was e

calculated to be 3 830 kilograms of uranium 235 For an excess reactivity requirement of 714% AK/K, the minimum operating loading vill be 4.64 kilograms of uranium-235 The core in the open end position vill consist l

of six (6) control elements, twenty-one (21) standard elements, or a total loadin8 of 4 704 kilograms of uranium-235 The fuel requirements for the two operatin6 positions are sie rized below:

i Position l Stall Pool End Item Cold cican mass 3 45 Kg. U-235 3 83 K6 U-235 Execos Requirements 1.068 .806 Minimum Operating mass 4 52 4.64 Actual loading 4.606 4 704 4

The above figures are for the initial loading with a fresh core.

The total fuel content of the core vill increase as additional fuel is added to compensate for the additional samarium-149, uranium burn-up, and r

low cress section fission pmduct build-up. The total excess reactivity at any one time vill not exceed the values given in Table I.

l (10) Control Rod Benetivity Effectiveness l The UCNC core vill have five B4C shim-cafety rods and one stain-l less steel regulating rod. The effectiveness of these rods has been cal-culated for a typical core and rod arrangment.

l The chim-cafety rods as located in Figure 16 have an average

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calculated worth of -3 26% AK/K, or a total effectiveness for the five j mas of -16.4% AK/K. The stainless steel regulating red vill be limited l to an effectiveness of about -0.6% AK/K. The total worth of all six rods in then -17 0% AK/K. This is about 67% greater than the maximum require-ment for the core in the ctall position, and about 121% greater than the l

maximum requirement for the core in the pool position.

The following table gives the individual rod worths for the l

j positions shown in Figure 16, for a typical load.

Bod Position Worth % AK/K

) 1 C-2 -3 22 l

2 n-3 -3 79 3 D-3 -2 38 l C-4 -3 79 4

5 B-S -3 22 Ecgulating rod A-4 -0.6

]

i Total -17 00% oK/K i

I l

i 1

(11) Fuel Requirements

'Ihe UCNC reactor was designed on the assmption that it vould operate continuously for ten day periods between shut-downs. During these shut-down periods, additional fuel vould be added to the core to ccupensate

• for the uranium 235 burned up, and for accumulation of low cross section fission products. The actual fueling cycle vill vary as a func, tion of the type and number of experiments to be perfomed.

To detemine the fuel requirement for long tem operation, a burn-up of 20% for the average fuel element in the final core was assumed.

Although a bum-up of 30% per element is achievable, the 20% figure makes and allowance for higher than for the time that average someinelements burn-out vill be out some elements. To of the reactor,%

achieve the 20 burn-up, it has been determined that in addition to the mass required for other reactivity effects (i.e., temperature, Xe, Sm, control and beam tubes), that 1 5 Yg of U-235 was used for burnout and buildup of low cross section fission products. This additional fuel corresponds to an increase in the core size by about 8 standard fuel elements.

For this burn-up, the reactor would operate for about 170 days at 5000 kilowatts.

Since only a fraction of this additional fuel vill be added to the core during each chut-down period, the excess reactivity allowed for burn-up vill never exceed that amount required for ten days operation at full power or +.76% AK/K.

A summary of the fuel requirements for the UCNC vater reflected reactor core are given below:

Fuel Mass Kg. U-235 Stall Open Pool Initial loading (including Xe, Sm, etc.) 4.6 4.7 Additional fuel for burn-up* 15 15 Total 6.1 6.2 Average cycle time assumed 170 days 170 days Homal operating procedures vill require that the beam tubes that do not contain experiments vill be filled with water prior to loading the core. The loading requirements vould then be reduced to 4 3 kg for the stall position. The large reactivity effect of the themal column vill be allowed for by removing a minimum of 5 elements from the core prior to mov-ing it to the stall position.

• A burn-up to 20% for the average fuel element in the 11nal loaded core.

2<) .

2. SHIELDING {

Based ca measurements made at the IRL at a power level of 5 }W  :

the radiation levels at the UCNC reactor in areas nomally occupied by personnel vill be well within limits prescribed in Part 20 of Title 10 of the Code of Federal Regulations. The shielding and expected radiation

• 1evels are as fonovs: ,

,. a. Main Concrete Shield

' I Figures 4, 5, 6, and 7 show a plan view and cross sections of the main concrete shield. The main biological shielding consists of water, magnetite concrete, and ordinary concrete in varying proportions. The maximum magnetite thickness is 5 8 ft. Radiation levels from the core e ,' vill be reduced to below 1 mr/hr at all points outside the shield.

Icad themal shielding is utilized in the stan to reduce ga:=na

(

, flux incident on the concrete, thereby relieving themal stresses caused -

by radiation heating.

b. Pool Surface The surface of the pool is 24 feet above the top of the core.

i This depth of water vin reduce direct radiation at the surface of the pool to below 1 mr/hr.

The total gn=ma level at the pool surface, directly over the core, is expected to be less than 15 mr/hr. The water-dispersed activity, which accounts for nearly all of the Sa=ma radiation above the core, is calculated in Appendix 2. Calculations were based on IRL measurements with allovance for an increased demineralizer flow rate in the UCNC facility.

c. Beam Tubes and Themal Column The beam tubes are shielded with concrete plugs followed by a lead door at tne outer end. Radiation levels at the ends of the beam tubes vill be less than 1 mr/hr.

The themal column chield consists of a lead block between the themal column and the core, and the lead shield follived by a magnetite deor at the outer end of the column. Since the original shield design, shiciding was added on the sides of the themal column to reduce radiation levelsoutsideofthecolumntobelov3mr/hr. This design change clinin-ated steps in the shield near the themal column.

During nomal operation, a negative pressure vin be maintained in the themal column, thereby preventing the escape of radioactive A41 into the reactor room.

d. Cooling System Components The cooling water drain lines, holdup tank, and pump room vill be chicided to reduce radiation to tolerance levels. If required, shield-ing vill be installed around the dcmineralizer to provide maintenance acccca to the pump room durine, reactor operation.

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e. Spent Puel Elements

'Ihe spent fuel elements vill be transferred undervater and stored in the pool and, in the canal. After an appropriate cooling time has elapsed, the spent fuel elements vill be removed in lead casks and transported for reprocessing.

f. Handling of Radioactive Materials The reactor pool is connected to the Hot Imboratories by a vater-filled canal. This canal vill be used for the transfer of radioactive -

materials. Handling casks and dollies vill be utilized for radioactive materials being moved outside the vnter shield.

~

3 SINMARY OF RADIATION LEVEIS

a. D'tring Nomal Operation Outaide main biological shield -below1mr/hr.

Pool surface, directly above core - below 15 mr/ar.

Outside of beam tubes - below 1 mr hr.

Vieir.ity of thermal colten - belov 3 mr hr.

b. After Shut-down During operation, reactor components exposed to a high neutron flux and the pool concrete become activated. In the stall, the major sources of induced radiation are the pool floor and valls, the beam tubes, the cool-ant components, and the ther=al coltun. In the open end, only the neutron exposed coolant co=ponents and the pool floor and valls are expected to become activated.

Calculations have shown that after the pool has been drained, the radiation in the stall and open end can be sufficiently reduced to pemit maintenance for controlled periods of time.

(1) Reactor Stall By removing the beam tubes and graphite from the themal colt =n and by adding temporary shielding in front of the themd column cover plate and over the caposed coolant components, the radiation level in the stall can be reduced to below 10 nr/hr.

(P) Open End If a shield is placed over the coolant pipe flange and cam in the open end, the dose rate due to induced radiation is reduced to below 1 mr/hr.

AnGcbmenf Y ATTACHMENT 4 NRC RESPONSE TO CINTICHEM COMMENTS ON NRC ADMINISTERED REQUAllFICATION EXAMINATION Question NRC Response A.02/H.01 Comment not accepted. Neither of the licensee's interpretations can be inferred from this question.

Nothing in the question indicates that an error was made calculating Kexcess. Secondly, the question clearly states that temperature is to be increased.

Only one out of ten operators had any serious dif fi-culty with this question, so interpretation does not appear to be a problem.

A.03/H.03 Comment not accepted. There is only one possible interpretation for "inaccurately low reading."

A,04/H.04 Comment partially accepted. An essentially equivalent question was used at Cintichem in an NRC examination given January 1985. At that time only one minor cor-rection was made, which is incorporated into the present question. The temperature change of 10 F is significant according to the licensee provided refer-ence question A14. The answer key was clarified in that NI's measure thermalized fast neutrons.

A.06/H.07 Comment not accepted. A knowledgable operator is expected to know the dif ference between " production" (rate) and " quantity" (amount).

A.07/(H.08) Portions of the answer in parentheses are editorial and not required for full credit. Answer key clar-ified to include power and neutron changes as well as reactivity. ,

B.04 No change. This question is taken verbatim from the licensee questions included with the reference mate-rial and is, therefore, consistent with the licensee's terminology.

C.05 Question deleted.

Attac6hent 4 2 Question NRC Response 0.02 If reverses are itemized, they will be accepted as any of the four required answers.

D.04/J.05 Comment accepted. Licensee should update reference question 0.13.

E.01 Comment accepted per RM-04-2.

F.03/L.01a Comment accepted per RM-03-5.

G.02a/I.03a, b Will accept 3 Rem quarterly limit if 5(N-18) and NRC Form 4 assumptions are made.

G.02b/I.03c Answer clarified to include saturation as a reason for inaccuracy.

G.04/I.05 Noted. No change to answer key required.

H.02 Comment not accepted. The NRC disagrees with the licensee's comment that the question is vague and open ended. There is only one basis as described in Tech-nical Specifications. The reason for the difference between stall and pool values was clearly solicited in the question.

I.04 Other tasks which clearly involve high radiation or contamination levels will be considered on a case b sis.

J.03 Typographical error corrected.

J.04b Comment not accepted. The NRC disagrees with the licensee comment that the question is vague and open ended. An identical question appears in the licen-see's reference question (J.3). The method of chang-ing rod materials to influence worth will be added as an acceptable answer.

J.06 Comment was not supported by reference material and therefore is not accepted.

K.02 Comment accepted.

Attaclhent 4 3 Question NRC Response K.05 Comment not accepted. This question is identical to

{

one on the licensee's 1986 SRO requal exam (K6) in {

which the required complete answer consisted of the I two identified elements. The licensee proposed addi- '

tional answer was not supported by reference material and, therefore, is not accepted.  ;

l K.06 Comment accepted. Non-immediate actions deleted.

This question was also taken from the facility supplied questions (K11).

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