ML20203K261
| ML20203K261 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 08/01/1986 |
| From: | Cooley R, Pellet J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20203K250 | List: |
| References | |
| 50-267-OL-86-01, 50-267-OL-86-1, NUDOCS 8608070134 | |
| Download: ML20203K261 (110) | |
Text
r REQUALIFICATION PROGRAM AUDIT AND OPERATOR LICENSE EXAMINATION REPORT Licensee: Public Service Company of Colorado P. O. Box 840 Denver, Colorado 80201 Docket: 50-267 License No.: DPR-34 Operator License and Requalification Program Audit examinations at Fort St.
Vrain (FSV).
Chief Examiner: [ k 7- 3 h b John L. Pellet, Examiner Date Signed Approved by: d k- f-/ 2b Date Signed RalphJ A. Cooley, Secfion Chief Summary:
Four Senior Reactor Operator - Limited to Fuel Handling (SRO-LFH) written and i
oral examinations were administered by the NRC at FSV during the week of May 19, 1986.- Three candidates successfully completed the examinations and have been issued appropriate licenses.
Eight Requalification Program Audit examinations (3 Reactor Operator (RO), 4 Senior Reactor Operator (SRO), and 1 SRO-LFH) were administered by the NRC during the week of May 19, 1986. Four of the eight did not successfully complete the written or oral examinations and have been placed in accelerated retraining as required. Because of the high failure rate, the FSV Licensed Operator Requalification program was evaluated as unsatisfactory. Corrective actions have been agreed to and implemented. Pending NRC reevaluation of the program as satisfactory, no operator licenses are being renewed at FSV except after satisfactory performance on an NRC examination.
8600070134 860804 PDR ADOCK 05000267 V PDR
F ,
2 REPORT DETAILS
- 1. PERSONS EXAMINED License Examinations: SRO-LFH: 4 PASS: 3 - 75%
FAIL 1 - 25%
Requalification Examinations:
TYPE: SR0 SRO-LFH R0 TOTAL
- EXAMINED: 4 1 3 8 PASS: 3 - 75% 1 - 100% 0 - 0% 4 - 50%
FAIL: 1 - 25% 0 - 0% 3 - 100% 4 - 50%
- 2. EXAMINERS J. Pellet, NRC (Chief Examiner)
R. Cooley, NRC J. Whittemore, NRC
- 3. EXAMINATION REPORT Individual performance results are not included in this report because these reports are placed in the NRC's Public Document Room.
- a. EXAMINATION REVIEW COMMENT RESOLUTION In general, editorial comments or changes made during the examina-tion, review, or subsequent grading reviews are not addressed by this resolution section. This section reflects resolution of substantive comments made in Public Service Company of Colorado (PSC) letter P-86385, dated May 23, 1986. The modifications discussed below are included in the master exam key which is provided elsewhere in this report (see 3.f), as are all other changes mentioned above but not discussed herein. Note that the comments from the letter referenced are paraphrased below for brevity. The full text of the comments is available in the referenced letter, which is attached (see 3.g.).
4 1.01 The question, as written, is over valued.
Resp.: ACCEPT. For future reference.
1.02 Curve plot is inadequate. Periods listed should be negative.
Resp.: ACCEPT. For future reference.
1.03.b Accept answer due to changes in the cross-sectional area of absorption for Xe-135 as fuel temperature changes.
Resp.: ACCEPT. Key modified.
f:
3 1.04 Part 3 should also accept decay.
Resp.: ACCEPT. Key modified.
1.05 Question is invalid since Pa-233 is not a product of the fission process.
Resp.: REJECT. Facility lesson plans provided specifically class -
Pa-233 as a fission product poison. Will note for future reference.
? 06 Calculational method used is inadequate for FSV. Best assumption for enthalpy of feedwater is sat. water @ 400 F.
Resp.: ACCEPT. Key modified.
1.08 Answer c is 12.3 psia. Part e is not valid since no identi-fied task relies on knowledge of sea-level conditions. Part f has 2 set of brackets.
Resp.: REJECT. Key for c modified. Part e tests basic understand-ing of theoretical material. Additionally it is drawn directly from the lesson plans provided. Part f has 1 set of brackets and 1 set of parenthesis. Examinees were cautioned about the similarity of appearance of [] & ().
2.02 Accept as alternate answers cold reheat and auxiliary boiler feed pump.
Resp.: ACCEPT. Key modified.
2.04 Accept increasing radiation levels and pressure > 250 psig.
Resp.: ACCEPT. Key modified.
2.05.a Accept when circulator speed decreases to < 700 rpm.
Resp.: ACCEPT. Key modified.
2.06 Accept malfunction in the slack cable circuitry.
Resp.: ACCEPT. . Key modified.
_2.07 Accept system 23 for part a and system 24 for part b.
'. ' Msp.': ACCEPT. Key modified.
' 3.01 Answer 4 should also accept 2/3 low level moisture monitors trip, no circulator seal malfunction, no hot reheat activity high, and 1/2 high level moisture monitors trip.
.s Resp.: ACCEPT. Key modified.
3.02 The question is poorly worded.
Resp.: ACCEPT. For future reference.
3.05 The question for section a is poorly worded.
Resp.: ACCEPT. For future reference.
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4 3.06 The question should include the name of HC-21417 as " Loop 1 Backup Bearing Water Controller."
Resp.: ACCEPT. Exam modified and announced during exam.
3.07 Answer a.5 should accept 1.55 gain on operating circulator.
Resp.: ACCEPT. Key modified.
4.04 Emergency procedure G should be referenced for d.
Resp.: ACCEPT. Key modified.
5.01 Question is over valued.
- Resp.: ACCEPT. For future reference.
5.02 Same'as 1.04.
5.03 Same as 1.05.
5.04 Same as 1.06.
5.05 Same as 1.08.
5.07 Answer is not specifically given in reference cited. Clarify applicability to FSV. ... .
Resp.: ACCEPT. Question deals with affect on f. lux shape of various '
core alterations, for example rod height'. . Reference discuss-
- es the effect on flux shi.pe (peaking) in geseral terms. .
Application to. specific cases tests understanding AND ability to apply theoretical knowledge to specific cases.
5.08 Question is poorly worded.
Resp.: ACCEPT. For future reference.
Same as 2.02. t 6.02 3 6.03 Same as 2.04.
6.04 Same as 2.06.
6.05 Part a, asks for 1 answer but key requires 3 for full credit.
Part b point values in key are not-consistent. For part c the alarm is not a correct answer si'nce it is a result of a trip and not a trip itself.
Resp.: REJECT. Full credit requires all 3 subpart.s but they are NOT independent answers. Key modified for parts b and c.
6.06 As written, the shock absorber is not required as an answer.
Resp.: REJECT. Both features are intended for this function.
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6.07 Part b should accept first stage pressure and throttle pressure. Part f should be reactor power. FSP should be written out as first stage pressure.
Resp.: ACCEPT, Key modified.
6.08 Answer is correct per reference but current plant practice is
'- ' to use domestic water as normal and well as emergency.
Resp.: ACCEPT. Key modified.
M.02 From FH 038.00, a fourth answer is whether hot or cold S/D.
- Resp.: ACCEPT. Key modified, l.'m .
M.03 Should reference TS !_C0 4.7.3.
Resp.: ACCEPT. Key modified.
M.05 Rod worths are not required knowledge.
. Resp.: ACCEPT. For future reference.
e,- -
M.06 Reference is FH 033.00, p. 6. Knowledge is not required.
Resp.: ACCEPT. Key modified. Noted for future reference.
N.02 Reference to SD-13-1.
Resp.: ACCEPT. Key modified.
N.03 Part b is poorly worded. Answer b is system 46 aux loops.
Resp.: ACCEPT. Key modified.
N.04_ SSLO is not required to operate vacuum pumps.
Resp.: ACCEPT. For future reference.
, N.05 SSLO is not responsible for fuel storage well cooling.
, Resp.: ACCEPT. For future reference.
1 I ~.s N.06 Reference is FH 043.00, p. 9. It notes add'l answers.
Resp.: ACCEPT. Key modified.
0.02 SSLO,is not responsible for part a. Part b answer is when required per FHP, prior to refueling.
Resp.: ACCEPT. Key modified.
0.03 SSLO is not responsible for section b.
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' Resp.: ACCEPT. For future reference.
0.11 Part a answer is Tech Svces Dept. Part b is poorly worded.
Resp.: REJECT. Part a answer is directly from reference and no alternate reference was provided.
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6 P.03- SSLO is not responsible for section 2 of TS.
Resp.: REJECT. Understanding of basic definitions in TS are re-quired. SSLO should, for example, know the definition of irradiated fuel or refueling mode.
P.06 SSLO is not responsible for recovery actions for loss of well cooling.
Resp.: REJECT. Knowledge and understanding of system design and operating procedures is sufficient to answer the question.
P.07 For part c accept also hazard alarm of FHM console.
Reca.: ACCEPT. Key modified.
- b. MAY 22, 1986 EXIT MEETING
SUMMARY
At the conclusion of the site visit, the NRC examiners met with the representatives of the plant staff to discuss the preliminary results of the examinations. The following personnel were present:
NRC UTILITY J. Pellet P. Bearly R. Cooley L. Bishard J. Whittemore M. Deniston R. Hooper R. Rivera Mr. Pellet started'the discussion by noting that the examiners as a group had encountered a positive, helpful attitude in everyone concerned. The following general topics were discussed.
(1) Preliminary results for the SRO-LFH oral examinations are that all are clear passes. No results are provided for requal-ification orals.
(2) 'NRC will'make every attempt to return formal results within 30 days.
(3). The following areas of weakness were observed in more than one examinee and are presented for the use of the facility. Note that an area of weakness does not imply unacceptable perfor-
~
mance. ~It is simply an area where knowledge or skill is less completely developed than in others.
(a) General knowledge of instrumentation, especially nuclear instrumentation was poor.
(b) Key control was weak for personnel not on operating crews.
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(c) Several individuals had difficulty explaining how they were kept up to date on procedure, drawing, or TS changes or revisions.
(d) Classification of events per the RERP was difficult, which appeared to be due to the procedure being hard to use.
(e) Technical Specification (TS) interpretation was poor, for example, required actions when nuclear instrumentation disagrees or when a liquid radioactive material release is in progress and a monitor fails,
- c. GENERIC COMMENTS The generic comments provided below were generated during grading of the written examinations. They are provided for the benefit of the licensee and no response or corrective actions are required.
(1) General understanding of core reactivity was weak. This in-cludes changes due to alterations during refueling and how peaking factors may be affected during operation.
(2) Control rod behavior and instrumentation, especially during refueling, was an area of weakness.
(3) Understanding Technical Specification definitions and LCO's, and ability to apply them to specific situations where interpreta-tion was required was weak.
(4) Basic radiation protection understanding was poor. This includ-ed relation of terms like RAD and REM,10CFR20 exposure limits, and posting limits for contaminated or radiation areas.
- d. REQUALIFICATION PROGRAM EVALUATION NRC administered eight Requalification Program Audit examinations (3 R0, 4 SR0, and 1 SR0-LFH) at FSV. Four of the eight licensed indi-viduals evaluated failed the NRC-administered examinations. They have been placed in an accelerated retraining program subject to the conditions of your Licensed Operator Requalification Program. Based on these results, the Fort St. Vrain Nuclear Facility Licensed Operator Requalification Program is evaluated as unsatisfactory. The overall program evaluation is presented on page 10.
As a result of this unsatisfactory evaluation, license renewals will be issued only for those operators who pass an NRC-administered examination until corrective actions have been identified, agreed upon, implemented, and the program reevaluated as satisfactory by the NRC.
k $
8 For further information on program evaluation and the consequences of
+ an unsatisfactory evaluation, please see NUREG-1021, " Operator Licensing Examiner Standards," especially ES-601, " Administration of NRC Requalification Program Evaluation." ,
- e. CORRECTIVE ACTIONS AND IMPLEMENTATION - JUNE 23, 1986 MEETING
SUMMARY
On June 23, 1986, the NRC staff met with representatives of the plant staff to discuss the required corrective actions and implementation schedule necessary. The following personnel were present:
NRC UTILITY T H. Johnson R. F. Walker J. E. Gagliardo J. W. Gahm J. P. Jaudon F. J. Borst R. A. Cooley Steve Willford Bruce Boger R. E. Rivera J. L. Pellet After introductory remarks, Mr. Borst described the analysis performed by the licensee to determine the root causes for the unsatisfactory performance of the Operator License Requalification Program. This analysis included:
In-depth review of operator performance on the facilty-administered and NRC examinations, both from a category and individual question standpoint.
Review of the requalification program itself, including manpower, facilities, and lesson plans.
Review of the Er.1 of Course Evaluations completed by licensed personnel.
Review by thc Requalification Review Panel.
This review concluded that the weaknesses found by the NRC were present and could be corrected with the following actions:
Review basic Health Physics material. Previously the program had assumed a minimum knowledge level that did not, in fact, exist.
Add procedure review to the formal lecture series as opposed to past practice which emphasized self-study methods.
To some extent, supported by comparing past results, the formal lectures emphasized past weaknesses. While valid and i appropriate, a minimum coverage of all areas needs to be retained to assure areas of strong performance remains so.
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9 In the past, facility testing has been on sub-parts. Future testing will be more broad, to be more similar to NRC exams.
Physical facilities will be improved.
Training staff has increased almost three-fold.
NRC accepted the described program improvements. When coupled with the accelerated training being administered to operators with unsatisfactory performance, this appears adequate to correct the identified deficiencies. NRC will evaluate the results by administering additional requalification examinations to 11 license-holders in September. This group will consist of those operators who have applied for renewal plus randomly selected additional personnel.
When combined with the scheduled license examinations, 16 exams will be administered in September. Satisfactory performance on these examinations will permit NRC to accept facility certification and issue renewals based on that certification. The requalification program will be reexamined during the next scheduled facility requalification examinations.
- f. EXAMINATION MASTER COPY The master copies (questions and answers) for the SRO, SR0-LFH, and R0 requalification examinations and the SR0-LFH license examination follow the program evaluation on the next page.
- g. FACILITY EXAMINATION COMMENTS The facility examination review comments in the form of PSC letter P-86385 follow the examination master copies.
- T i
,a o A
10 OVERALL REQUALIFICATION PROGRAM EVALUATION l
Facility: FORT ST. VRAIN Exeminers: J. PELLET, R. COOLEY, J. WHITTEMORE Dates of Evaluation: MAY 20-22, 1986 Areas Evaluated: XX Written XX Oral Simulator Written Examination Evaluation of NRC Examination Results (if given): Unsatisfactory '
Oral Examination
- 1. Overall Evaluation: Satisfactory
- 2. Number Observed: None Number Conducted: EIGHT ;
Simulator Examination
- 1. Overall Evaluation: Not Applicable
- 2. Number Observed: Number Conducted:
Overall Program Evaluation Satisfactory: Marginal: Unsatisfactory: XXX REASON FOR EVALUATION
- Pass ration for NRC written examination below 60%.
Submitted: Forwarded: Approved:
6dv .k- Branch Chief Examiner Section Chief
U.S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR REQUALIFICATION EXAMINATION Facility: FT. ST. VRAIN Reactor Type: HTGR-GA Date Administered: 86/05/20 Examiner: J. L. PELLET Examinee:
INSTRUCTIONS TO EXAMINEE:
THIS EXAMINATION REPLACES THE CURRENT CYCLE FACILITY ADMINISTERED EXAMINATION FOR DEMONSTRATION OF OPERATOR PROFICIENCY AND LICENSE RENEWAL.
FAILURE OF THIS EXAMINATION WILL REQUIRE RETRAINING UNDER THE CURRENT FACILITY REQUALIFICATION TRAINING PROGRAM AND MAY REQUIRE RE-EXAMINATION BY NRC. Points for each question are indicated in parentheses after the quastion number. The passing grade requires at leas t 70% in each category end a final grade of at least 80%. Examination papers will be picked up FOUR (4) hours after the examination starts.
% of Category % of Candidates's Category Value Total Score Value Category 25.00 25.00 1. Principles of Nuclear Power Plant Operations, Fluids, and Thermodynamic
- 25.00 25.00 2. Plant Design Including Safety and Emergency Systems -
25.00 25.00 3. Instruments and Controls 25.00 25.00 4. Procedures - Normal, l
Abnormal, Emergency, und l Radiological Control 100.00 TOTALS i
Final Grade %
All work done on this examination is my own. I have neither given nor roceived aid.
Examinee's Signature t
Iz__EBINCIELES_QE_NUGLE88_EQWEB_EL8NI_QEEB8IIQNi PAGE 2 IHEBdQQ1NedIGSt_HE8I_IB8NSEEB_8ND_ELUIQ_ELQW 4
QUESTION 1.01 (3.00)
What two (2) core operating characteristics change between BOC and EOC due to the buildup of U-233 for use as fuel?
QUESTION 1.02 (3.00)
Explain the basic reasons for the shape of the areas labeled Ca), (b), and e(c) in the curve below of reactor power after a scram. (3.0) 100 -l*
1
- 1
- 10 -l
- Cu) T=20 seconds P 1
- O I
- W 1 _I
- E I
- R I * (b) Change in slope
% 0.1 -I ************ (c) T=80 seconds
- *****x******
3 ************
0.01-1 ************
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0 60 120 180 240 300 360 SECONDS Cafter scram) f QUESTION 1.03 (3.00)
Why, and how, does the relative worth of Xenon-135 change as fuel l temperature increases? (3.0)
QUESTION 1.04 (4.00)
{ Fill in the blanks below dealing with fission products. Blanks may contain j one or more words or phrases.
- c. During a reactor startup after refueling , Xe concentration will
_(1)_ to equilibrium in _(2)_ hours, primarily due to _C3)_. (2.0)
- b. Following INITIAL reactor startup, it will take about _(1)_ to reach Samarium equilibrium. Although Sm is not _(2)_, following a reactor shutdown, Sm will _(3)_ for about _(4)_ days. (2.0)
(***** CATEGORY 01 CONTINUE 0 ON NEXT PAGE *****)
Ii__EBINCIELE2_QE_NWGLE88_EPVEB_EL8HI_9EEB8IIQNi PAGE 3 IHEBdQDYNedIG2t_BE81_IB8NSEEB_8NQ_ELUID_ELQW QUESTION 1.05 (1.50)
What significant poison is produced by the f ission process besides xenon end samarium? (1.5)
QUESTION 1.06 (4.00)
At 100% power, what enthalpy is added by the Steam Generators between feedwater entry and main steam exit? (4.0)
QUESTION 1.07 (2.50)
Why does the Ideal Oes Law apply t o the Fe.-t St. Vrsin Pr imary Coalant System? (2.5)
QUESTION 1.08 (4.00)
Ecch of the statements celow is FALSE. Replace the word (s) in brackets to arke the statement TRUE. (4.0)
- e. Theoretically, at zero degrees (Fahrenheit) molecular motion stops.
- b. Pressure is the (work done) by a fluid per unit area.
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- c. Atmospheric pressure at the FSV plant is (14.7 psia),
- d. A full vacuum at FSV's altitude is 25.1 inches of (water).
- o. Sixteen (16) inches of vacuum at FSV is un absolute pressure (equal to]
16 inches of vacuum at sea level (00 NOT USE "NOT EQUAL T0" AS AN ANSWERI).
- f. An object at 1000 degrees F is twice as hot (on the thermodynamic scale) as an obj ect at (500 degrees F).
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(***** END OF CATEGORY 01 *****)
22__ELeNI_ DESIGN _ INCLUDING _SeEEIX_eND_EMEBGENGX_SISIENS PAGE 4 QUESTION 2.01 (2.00)
List two (2) of the three (3) temperatures which form the basis for orifice valve adj us tments.
QUESTION 2.02 (3.50)
List the six (6) sources of motive power to the helium circulators.
Label the source used in extreme emergency only.
QUESTION 2.03 (3.50)
- m. Why is the reheat section of the steam generator oversized? (1.0)
- b. How oversized is the reheater section (% surface area imbalance between the main and r ei cat sections? (0.5)
- c. Why is cold reheat attemperation needed? (1.0)
- d. What are two (2) events that cause the reheat steam attemperation temperature bias signal to be activated? (1.0)
QUESTION 2.04 (4.00)
- e. How long does it take to dump one loop after initiation of the Steam /
Water Dump system due to rupture of a feedwater subheader? (0.75)
- b. What are two (2) methods that can be used to confirm the proper loop has been dumped by the steam / water dump system? (2.0)
( c. What is the purpose of the steam / water dump system? (1.25)
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(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
22__EL8NI_ DESIGN _INGLUDING_S8EEIX_8HD_EMEBGENCX_SYSIEUS PAGE 5 QUESTION 2.05 (4.00)
- c. When is the shaft brake applied to the Helium Circulator shaft (which prevents shaft rotation)? (1.0)
- b. What are the two (2) functions performed by the Helium Circulator Compressor Inlet? (1.5)
- c. What provides the drive head to promote water flow from the bearing water drain into the helium water drain during helium circulator operation? (1.5)
QUESTION 2.06 (3.00)
What are three (3) conditions which may be indicated by a " Slack Cable" clarm for a region? (3.0) 1 QUESTION 2.07 (2.00)
- n. What is the normal helium supply to the helium circulator buffer? (1.0)
- b. What is the helium supply source if the normal source is lost? (1.0)
QUESTION 2.08 (3.00)
- n. On a loss of 480 VAC bus voltage, which diesel generator will be connected to which busCes)? (1.5)
- b. How will the feed breakers to affected 480 VAC buses react to a loss of all 4160 VAC supply? (0.75)
- c. If the Reserve Auxiliary transformer fails during an outage, where can auxiliary power be supplied from? (0.75)
(***** END OF CATEGORY 02 *****)
22__IN2IBudENIS_8NQ_00NIBQL3 PAGE 6 QUESTION 3.01 (4.50)
List ten (10) automatic reactor trips. Include for each trip, the basic trip logic (such as 2/4) and the trip setpoint or level.
QUESTION 3.02 (1.00)
What automatic action (s) occur on detection of a steam leak into the primary system that do NOT occur for shutdown of a loop? Actions which cccur for loop shutdown may be included but are NOT required.
QUESTION 3.03 (3.50)
- c. What is required to produce a low feedwater pressure trip in the Boiler Feed Pump Control system (include switch positions and instrumentation logic in your answer)? (1.5)
- b. What valves change position on a low feedwater pressure trip? (2.0)
QUESTION 3.04 (4.00)
Fill in the blanks below dealing with conditions that PROHIBIT withdrawal of control rods. Blanks may contain one or more words or phrases. (4.0)
- c. Rod control load sensors sense _____. (0.75)
- b. Reactor power less than _(1)_ (0.5) AND attempted rod withdrawal
- l. _(2)_ (0.75). (1.25) l
- c. Power indication from any power channel less than _(1)_ (0.5) AND
_(2)_ (0.75) in the Power position. (1.25)
- d. Low Startup Channel countrate below _____. (0.75) i I
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(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
2t__INSIBUMENI2_8NQ_G9 NIB 0L2 PAGE 7 QUESTION 3.05 (3.00)
With respect to the steam / water dump system:
- e. What parameter is used if the moisture monitor system fails? (1.0)
- b. What actions occur, with no moisture monitor, if the steam / water dump system actuates (assume both loops are operating)? (2.0)
QUESTION 3.06 (3.00)
- c. What are the two (2) possible inputs to the measured signal displayed on HC-21417 (Loop 1 Backup Bearing Water Controller)? (2.0)
- b. Which signal will be d isp laye d ? (1.0)
QUESTION 3.07 (4.00)
- o. What specific actions are perf ormed by the Plant Protective System for a single helium circulator trip from 100% power? (2.5)
- b. What are three (3) automatic actions which occur on a circulator trip due to loss of bearing water? (1.5)
QUESTION 3.08 (2.00)
- c. What are two (2) possible causes of an inadvertent SLRDIS (Steam Line Rupture Detection and Isolation System) trip actuation? (1.0)
- b. What indication would be available to verify that SLRDIS actuation was inadvertent for ONE of the causes above? (1.0)
(***** END OF CATEGORY 03 *****)
St__fBQGEQUBEf_:_NCBdeLt_8BNQBdeLt_EMEBGENCX_8NQ PAGE 8 86DIQLQQIG8L_CQNIBQL QUESTION 4.01 (3.00)
What are two (2) plant conditions which require insertion of one or both eubgroups of the Reserve Shutdown System.
QUESTION 4.02 (2.00)
What are the four (4) ei.iergency conditions or c lass i f ic at ions in order of INCREASING severity, per the Fort St. Vrain Event and Emergency Classification Overview procedure.
QUESTION 4.03 (3.00)
- o. What action should be taken by the operator after confirmation that the correct loop has been dumped by the steam / water dump system (due to high primary moisture)? (1.5)
- b. What does this action accomplish? (1.5)
QUESTION 4.04 (4.00)
Answer YES or N0 to each of the following questions.
- a. When manually driving in any CRD's, should the rod be driven past the "zero" indication if the " Rod In" light does not come on?
- b. Is a Radiation Work Permit required for removal and replacement of Hydrogen Getter Sponges?
- c. Is Health Physics coverage required for manumi operation of HV-2301/027
- d. On a loss of feedwater with the ISS in other than the " POWER" position, is the operator required to manually scram the reactor.
- o. In the event of a fire, can you verify valve alignment from the control room?
(***** CATEGORY 04 CCNTINUED ON NEXT PAGE *****)
dz__EBQGEDWBES_ _NQBdeLt_6BNQBdeLt_EMEBGENGX_8NQ PAGE 9 88DIQLQQIG8L_CQNIBQL QUESTION 4.05 (4.00)
Ccmplete the table below dealing with external exposure limits.
- c. Federal whole body limits per 10CFR20:
- 1. _____ rem /qtr without exposure history. (0.5)
- 2. _____ rem /qtr with exposure history. (0.5)
- 3. _____ rom accumulated lifetime. (0,5)
- 4. _____ % of above for minors (under 18 years of age). (0.5)
- c. 10CFR20 limit to extremities: _____
rem /qtr. (0.5)
- d. 10CFR20 limit to the general public: .____
rem /qtr. (0.5)
- o. FSV Whole Body Guideline: _____ rem / week. (0.5)
QUESTION 4.06 (2.00)
Fill in the blanks below. Blanks may contain one or more words or numbers.
- o. A Contaminated Area is any area in which there exists loose surface contamination in excess of _(1)_ and/or _(2)_. (1.0)
- b. A Radiation Area is an accessible area where a major portion of the whole body could receive greater than _(1)_ or _(2)_. (1.0)
QUESTION 4.07 (3.00)
What are the five immediate actions required following an automatic scram?
Specific steps to complete each action are NOT required. (3.0) i l
(***** CATEGORY 04 CONTINUE 0 ON NEXT PAGE *****)
di__EBQCEDUBES_=_NQBdoLt_8BNQBd8Lt_EMEBQENGl_8NQ PAGE 10 B8DIDLQQIC8L_GQNIBQL QUESTION 4.08 (2.00)
Ecch of the definitions below, from Technical Specifications, is WRONG.
State the correct definition, or describe what must be changed to make the given def inition correct. (2.0)
- o. The reactor is considered shut down for refueling purposes when the reactor mode switch is placed in the " Fuel Loading" position simultaneous with either hot shutdown or the cold shutdown reactivity conditions. (0,5)
- b. Rated thermal power is 892 Mw(th). (0.5)
- c. Irradiated fuel is fuel that has been installed in the reactor during power operation. (0.5)
- d. A channel check is a test of the functional capability of the channel to determine that it is operable. (0.5)
QUESTION 4.09 (2.00)
Por Technical Specification LCO 4.5.1, " Reactor Building, Limiting Conditions for Operation," what two conditions must be met in order to hendle irradiated fuel within the Reactor Building? (2.0) l 4
(***** END OF CATEGORY 04 *****)
R************* END OF EXAMINATION ***************)
li__EBINGIELES_QE_NUGLE8B_EQWEB_EL8HI_QEEB8IIQNt PAGE 11 IHEBdQDINedIGSt_HE8I_IB8NSEEB_8ND_ELUID_ELQW ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER 1.01 (3.00)
- 1. Decreases negative temperature coefficient (since it is a positive contributor to thermal utilization factor). (1.5)
- 2. Decreases Beta factor (from 0,0055 e BOL to 0.0048 e EOL). (1.5)
REFERENCE FSV Reactor Physics Lesson Plaa on Core Changes.
ANSWER 1.02 (3.00)
- c. T=20 seconds for the first 150 seconds due to decay of short-lived precursors and rod insertion time (by T.S. < 160 sec.).
- b. Inflection point when rods are in and short-lived precursors have decayed to negligible value.
- c. T=80 seconds due to decay of long-lived precursors (longest ~55 sec.).
REFERENCE FSV Reactor Physics Lesson Plan on Reactor Control ANSWER 1.03 (3.00)
HOW: The relative Xe worth Decreases with increasing fuel temp. (1.0)
WHY: The change in neutron energy spectrum Chardening) affects the neutron absorption since Xe-135 does not absorb all energies equally (OR Xe-135 cross-section changes with fuel temp.). (2.0)
REFERENCE LR 008.00, HTGR PHYSICS - Fission Product Po isons , p. 19
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l It__EBINCIELES_QE_NUCLEeB_EQWEB_EL8HI_QEEB6IIQNt PAGE 12 IBEBdQQ1NedIGSt_BE6I_IB8NSEEB_6ND_ELVID_ELQW ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER 1.04 (4.00)
- e. 1. increase (0.5)
- 2. 40~50 (0.5)
- 3. the 6.7 hour8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> half-life of I-135 (OR the dency of I-135) (1.0)
- b. 1. 14 days (0.5)
- 2. flux dependent (0.5)
- 3. increase (0.5)
- 4. 12 (0.5)
REFERENCE LR 008.00, HTGR - Fission Product Poisons, p. 7, 14 ANSWER 1.05 (1.50)
Pa-233 (0.75 Pa, 0.75 233)
REFERENCE LR 008.00, HTGR - Fission Product Poisons, p. 15 ANSWER 1.06 (4.00) dH = HCout) - H(in) (0.6)
T(out) = 1000 F +/- 50 F (0.6)
P(out) = 2400 psig +/- 1004 (0.3); P(in) = 3350 psig +/- 100# (0.3) (0.6)
T(in) ~= 400 F + 50 F / -0 F (0.6)
HCin) H(sat) e inlet temperature (0.6)
= 430~320 Btu /lbm (0.2)
H(out) = 1470~1450 Blu/lbm (0.2) dH = (1470-320)~(1450-430) = 1120~1020 8tu/lbm (0.2)
Use of steam tables to get 320, 420, 1450, & 1470 Btu /lbm (0.4)
REFERENCE CE Steam Tables, SOP 22-1
12__EBINQIELES_QE_NMQLE88_EQWEB_ELeNI_QEEB8IIQNm PAGE 13 IBEBdQQ1NedIQSt_BE6I_IB6HSEEB_8NQ_ELUIQ_ELQW ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER 1.07 (2.50)
Temperature is high enough (0.8) und pressure is low enough (0.8) that colecular sticactive forces may be neglected (0.9).
REFERENCE LR 007.00.01.02.1 ANSWER 1.08 (4.00)
- e. Rankine (6 answers e 0.667 ea.)
- b. force exerted
- c. 12.3 psia
- d. mercury
- o. greater than
- f. 270 F REFERENCE LR 007.00, various pages
2x__EL8NI_ DESIGN _INGLWQING_18EEII_8ND_EMEBGENGX_1XIIEUS PAGE 14 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER 2.01 (2.00)
- 1. Fuel region coolant exit temperature. (any 2/3 e 1.0 eu.)
- 2. Steam generator module reheat steam temperature.
- 3. Steam generator helium inlet temperature.
REFERENCE FSV S0P 12-04, p. 3 ANSWER 2.02 (3.50)
- 1. Emergency condensate. (0.5)
- 2. Emergency feedwater. (0.5)
- 3. Firewster - Extreme emergency source. (1.0 - 2 0 0.5 es.)
- 4. Bypass fissh tank. (0.5)
- 5. Turbine extraction / cold reheat steam. (0.5)
- 6. Auxiliary boiler steam. (0.5)
(4, 5, & 6 are from Cold Reheat. Accept also Aux Boiler Feed Pump.)
REFERENCE FSV SOP 21-01, p. 5 ANSWER 2.03 (3.50)
- c. The reheater is oversized to allow tube plugging without reducing the steam generator's ability to meet design f unction (CONCEPT). (1.0)
- b. 21% +/-5%. (0.5)
- c. To prevent excessive hot reheat steam temperatures. (1.0) i
- d. Turbine-generator runback. (0.5)
Turbine trip. (0.5)
REFERENCE R0 121.00, various pages
21__EL6HI_DE11GN_INGLUDING_S8EEII_6ND_EMEBGENGl_2XSIEd1 PAGE 15 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER 2.04 (4.00)
- o. 15 seconds (accept any answer < 30 seconds) (0.75) l
- b. 1. Dump tank pressures > ~250 psia. (1.0)
- 2. High dump tank radioactivity. (1.0)
- c. Minimize the amount of water that can leak into the primary system as a result of a steam generator tube or subheader rupture. (1.25)
REFERENCE R0 127.00, various pages ANSWER 2.05 (4.00)
- o. When the shutdown seal is set OR when circ speed < 700 rpm. (1.0)
- b. 1. Directs flow exially into the compressor binding. (0.75)
- 2. Monitoring device in the primary f low circuit. (0.75)
- c. During normal operation, main drain pressure is higher than the buffer helium pressure. This dPressure is reflected across the helium water seal and provides the driving head to promote water flow from the bearing drains to the helium water drain (CONCEPT). (1.5)
REFERENCE EO 068.00, p. 5, 13, 15 ANSWER 2.06 (3.00)
- 1. A control rod is stuck in the guide channel. (1.0)
- 2. A control rod cable is parted . (1.0)
- 3. Adsorber section(s) have dropped into the guide channels. (1.0)
(OR Malfunction in the Slack Cable circuitry.)
! REFERENCE j FSV SOP 12-01, Section 5 t
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22__ELeNI_ DESIGN _ INCLUDING _18EEII_8ND_EMEBGENC1_SISIEd3 PAGE 16 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER 2.07 (2.00)
- e. The purifiestion system (System 23). (1.0)
- b. The high pressure helium tanks (T-2402) (System 24). (1.0)
REFERENCE FSV SOP 24, Section 5 ANSWER 2.08 (3.00)
- b. Trip open. (0.75)
- c. The unit auxiliary transformer. (0.75)
REFERENCE FSV SOP 92-01, Section 5; 92-03, Section 5
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- 2i__INSIBUDENI1_eWQ_CQUIBQL1 PAGE 17 l ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
E ANSWER 3.01 (4.50) 4
- 1. Neutron countrate 1/2 10EES cps i 2. Flux rate of rise 2/3 5 decades / minute
! 3. High neutron flux 2/3 or C4/6 118~140% power
- 4. High primary coolant moisture 2/3 & 1/2 or 2/2 500 vpm/67F dwpt
- 5. High reheat steam Temperature 2/3 1055 1075F
! 6. Low primary pressure 2/3 506 below prog.
- 7. High primary pressure 2/3 506/7.5%> normal
! 8. Low press. hot reheat line 2/3 35 psig
- 9. Low superheat line pressure 2/3 1500 psig
- 10. Loss of elec power 2/3 '60% rated /30 sec
- 11. Two-loop trouble 04/12 or 2/3 lboth loops) ---
- 12. Rx b1dg high temperature 2/3 175 degrees F (eny 10 answers e 0.15/sub-part; 30 sub-parts *0.15 em.)
REFERENCE FSV Overall Plant Control & Plant Protective Sys., SD-93-2 Rev. N, p. 19-20 i
ANSWER 3.02 (1.00)
The dump valves open (HV-2215/7 or 2216/8) (dumps inventory to dump tank).
REFERENCE i
FSV Secondary Coolant System, SD-22-1, p.14 1
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i ANSWER 3.03 (3.50)
- o. " Interlock sequence switch" in the " power position." (0.75)
! 2/3 low pressure feedwater switches energized. (0.75) ,
j b. HV-3108 ("A" BFP Disch to NORMAL FW header) (0.3) i HV-3109 ("B" 8FP Disch to NORMAL FW header) (0.3)
HV-3110 ("C" BFP Disch to NORMAL FW header) (0.3)
HV-31207 ("B" BFP NORMAL Disch bypass) (0.6)
( HV-2203 (Loop I emergency FW block) (0.75)
HV-2204 (Loop II emergency FW block) (0.75)
.! NOTE Valve # or descriptions above are not req'd. Full credit answer must describe general actions taken for normal and emergency FW, 1
including bypass valve.
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. . , . - - _ - . - - - ~ _ , - , _ - , _ _ _ , - _ , , - - - - , - . - _ . - , _ , - . - - . - - . _ - , , - . _
22__INSIBUDENIS_6N0_CONIB0L1 PAGE 18 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
REFERENCE LR 001.00.01.06 l
ANSWER 3.04 (4.00) i I
- c. movement of more than 1 rod. (0.75) l
- b. 1. 5%. (0.5) I
- 2. out of pra-arranged sequence. (0.75) l
- c. 1. 5%. (0.5) j
- 2. the Interlock Sequence switch. (0.75) '
J
- d. 2.5 cps. (0.75)
ANSWER 3.05 (3.00) l o. Primary pressure (840 psig 4 100%, or 535> program) (1.0)
I b. Preselected SG is dumped. (1.0)
Other loop is depress. to just above primary pressure (~8508). (1.0) 1 REFERENCE R0 127.00 ANSWER 3.06 (3.00)
, c. A circulator cartridge dP. (1.0) i B circulator cartridge dP. (1.0)
- b. The lower of the 2 signals will be displayed. (1.0) i
! REFERENCE LR 006.01.01.04 4
4
22__INSIBudENIS_8NQ_G96IBQL2 PAGE 19 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER 3.07 (4.00)
- c. 1. Helium circ. steam drive speed control valve closed. (0.5)
- 2. Helium cire. steam drive outlet block valve closed. (0.5)
- 3. Water turbine outlet steam trap isolation valve closed. (0,5)
- 4. Turbine load reduced to 50% e 0.5%/second. (0.5)
- 5. Feed forward & prop. gain in MS changed (1.55 on op. circ.). (0.5)
- b. 1. Affected loop emergency bearing accumulator fires.
- 2. Dryer discharge valve to the cire. closes (600 valve).
- 3. Brake applied at 700 rpm decreasing.
- 4. Static seal set after 28 second time delay.
- 5. Unneeded auxiliaries are stopped. (Any 3/5 4 0.5 ea.)
REFERENCE R0 001.00, p. 4, 5, 6 ANSWER 3.08 (2.00)
- o. The possible causes are:
- hardware failure.
- localized steam leak.
- accidental trip (e.g., during surveillance).
- due to a fire.
- due to a primary coolant leak. (Any 2/5 4 0.5 ea.)
- b. Any reasonable answer explaining alternate instrumentation. (1.0)
REFERENCE LR 014.00.01.08 REFERENCE LR 014.00.01.08
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st__PBQQEQQBES_:_NQBd8Lt_8SNQBd8(t_EMEBGENQY_8NQ PAGE 20 88DIQLQQIQ8L_QQNIBQL ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER 4.01 (3.00)
- 1. Loss of forced coolant flow C0.75) for time per power history (0.75).
- 2. Sufficient rods cannot be inserted to provide 0.01 dK SDM C0.75) at hot shutdown (220 F & Xe-free) (0.75).
- 3. Operator verifies the reactor cannot be maintained shutdown following any manual or automatic scram or shutdown (1.5).
Ceny 2/3 answers e 1.5 ea.)
REFERENCE FSV SOP 12-05, Rev. 8, p. 4 ANSWER 4.02 (2.00) 4 Notification of Unusual Event, Alert, Site Area Emergency, & General Emer.
(4 answers e 0.4 eu.; correct order 0.4)
REFERENCE FSV EP CLASS, p. 11-14 t
ANSWER 4.03 (3.00)
- c. Remote-manually close the dump valves. (1.5)
- b. Stops helium leakage into the dump tank. (1.5)
REFERENCE R0 127.00, p. 20 ANSWER 4.04 (4.00)
- c. NO
- b. YES
- c. YES
- d. YES
- o. NO
dz__EBQGEQUBES_=_NQBdekt_8BNQBd8Lt_EMEBGENGY_6NQ PAGE 21 88DIQLQQIG8L_GQNIBQL ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
REFERENCE SOP 12-01, Sect. 5; SOP 23, Sect. 5; SOP 29, Sect. 5; SOP 42; EP G ANSWER 4.05 (4.00)
- c. 1. 1.25
- 2. 3.0
- 3. 5(n-18)
- 4. 10
- b. 7.5
- c. 18.75
- d. 0.125
- c. 0.100 (8 answers e 0.5 es.)
REFERENCE FH 058.00, p. 12, 13; 10CFR20 ANSWER 4.06 (2.00)
- c. 1. 1000 dpm/100 sq cm beta-gamma. (0.5)
- 2. 10'dpm/100 sq cm alpha, (0.5)
- b. 1. 2.5 mrem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. (0,5)
- 2. 100 mrem in 5 consecutive days. (0.5)
.i' REFERENCE FH 060.00, p. 3, 4 ANSWER 4.07 (3.00)
- 1. Insert a manual scram. (5 answers e 0.6 es.)
2< Place ISS on Low Power Position.
- 3. Ensure transfer of house power.
- 4. Ensure turbine trip.
- 5. Ensure / establish stable core cooling conditions.
REFERENCE /
FSV EP APP B-1,.p. 7-9
di__EBDGEDUBES_=_NDBdelt_6BN9Bd8Lt_EMEBGENGl_8NQ PAGE 22 B8DIQL90108L_GQNIB9L ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER 4.08 (2.00)
- c. Correct except "placed," should be " locked." (0.5)
- b. Correct number is 842. (0,5)
- c. ...has a radiation level >/= 100 mr/hr at 1 foot from element. (0.5)
-d. Either change " check," to " test," or . . . is a qualitative determination that the channel is operable. (0.5)
REFERENCE FSV TS, Section 2 3
ANSWER 4.09 (2.00)
- 1. Reactor Building integrity is maintained. (1.0)
- 2. 2 of the 3 reactor building exhaust fans are operable. (1.0)
REFERENCE FSV TS 4.5.1, p. 4.5-1 . ,
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U.S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR REQUALIFICATION EXAMINATION 4 s i Facility: FT. ST. VRAIN
~
Reactor Type: HTGR-GA Date Administered: 86/05/20
+
Examiner: J. L. PELLET Examinee:
INSTRUCTIONS TO EXAMINEE:
i THIS EXAMINATION REPLACES THE CURRENT CYCLE FACILITY ADMINISTERED EXAMINATION FOR DEMONSTRATION OF OPERATOR PROFICIENCY AND LICENSE RENEWAL.
FAILURE OF THIS EXAMINATION WILL REQUIRE RETRAINING UNDER THE CURRENT
-FACILITY REQUALIFICATION TRAINING PROGRAM AND MAY REQUIRE RE-EXAMINATION BY NRC. . Points for each question are indicated in parentheses after the qusstion number. The passing grade requires at least 70% in each category
? and a final grade of at least 804. Examination papers will be picked up 5'
FOUR (4) liou rs after the examination starts.
b % of Category '
% of Candidates's Category
'- Values Total Score Value Category 25.00 25.00 5. Theory of Nuclear Power Plant Operations, Fluids, l and Thermodynamics
. 25.00 25.00 6. Plant Systems Design, Control and Instruments
~25.00 25.00 7. Procedures - Normal, Abnormal, Emergency, and Radiological Control 25.00 25.00 8. Administrative Procedures Conditions, and Limits 100.00 TOTALS
, Final Grade %
,. .n LAll work done_on this examination is my own. I have neither given nor received aid.
4 Examinee's Signature
Sz__IHEQBY-QE_NVQLE88_EQWEB_EkeNI_QEEB6IIONt_ELUIDSt_6NQ PAGE 2 IHEBdQQ1NedICS QUESTION 5.01 (3.00)
What two (2) core operating character istics change between BOC and EOC due to the buildup of U-233 for use as fuel?
QUESTION 5.02 (4.00)
Fill in the blanks below dealing with fission products. Blanks may contain one or more words or phrases.
- a. During a reactor startup after refueling, Xe concentration will
_(1)_ to equilibrium in _(2)_ hours, primarily due to _(3)_. (2.0)
- b. Following INITIAL reactor startup, it will take about _(1)_ to reach Samarium equilibrium. Although Sm is not _(2)_, following a reactor shutdown, Sm will _(3)_ for about _(4)_ days. (2.0)
QUESTION 5.03 (1.50)
What significant poison is produced by the fission process besides xenon end samarium? (1.5)
QUESTION 5.04 (4.00)
At 100% power, what enthalpy is added by the Steam Generators between feedwater entry and main steam exit? (4.0) l
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
St__IHEQBI_QE_NUGLE6B_EQWEB_EL6HI_QEEB6IIQNt_ELUIQSt_8NQ PAGE 3 IBEBBQQ1NedIGS QUESTION 5.05 (4.00)
Ecch of the statements below is FALSE. Replace the word (s) in brackets to aske the statement TRUE. (4.0)
- a. Theoretically, at zero degrees (Fahrenheit] molecular motion stops.
- b. Pressure is the (work done] by a fluid per unit area.
- c. Atmospheric pressure at the FSV plant is (14.7 psia).
- d. A full vacuum at FSV's altitude is 25.1 inches of (water).
- e. Sixteen (16) inches of vacuum at FSV is an absolute pressure (equal to]
16 inches of vacuum at sea level (00 NOT USE "NOT EQUAL T0" AS AN ANSWERI).
- f. An object at 1000 degrees F is twice as hot (on the thermodynamic scale) as an object at (500 degrees F].
QUESTION 5.06 (3.00)
- a. Define Suberitical Multiplication (M). (1.0)
- b. Briefly explain why M is not used to plot a criticality prediction. (1.0)
- c. If the count rate is 100 cps at a Keff of 0.95, what will the count rate be at a Keff of 0.99? (1.0)
QUESTION 5.07 (3.00)
- a. Explain how local power peaking in the core is affected by:
- 1. The insertion of control rods at a steady power level. (1.0)
- 2. Replacing irradiated with unirradiated fuel. (1.0)
- b. What is done to reduce peaking at the core-reflector interface? (1.0)
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
St__INEQBl_9E_NM9LE88_E9 WEB _EL8NI_9EEB6II9Nt_ELUIDSt_8NQ PAGE 4 IHEBdQQ1NedIQ1 QUESTION S.08 (2.50)
What will happen to the work done by a motor driven feed pump at constant load if the following occur? The answer should be INCREASE, DECREASE, or NO CHANGE. (2.5)
- a. Fluid temperature is reduced.
- b. Deserator steam blanket pressure increases.
- c. Fluid Specific Volume decreases.
- d. Voltage to motor decreases,
- o. Frequency to motor increases.
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(***** END OF CATEGORY 05 *****)
6t__EL8HI_212IEd2_DESIGNt_GQNIBQLt_8NQ_INSIBudENI6IIQN PAGE 5 QUESTION 6.01 (4.50)
List nine (9) loads or equipment which can be supplied power by the Alternate Cooling Method diesel generator.
QUESTION 6.02 (3.50)
List the six (6) sources of motive power to the helium circulators.
Lsbel the source used in extreme emergency only.
QUESTION 6.03 (4.00)
- c. How long does it take to dump one loop after initiation of the Steam /
Water Dump system due to rupture of a feedwater subheader? (0.75)
- b. What are two (2) methods that can be used to confirm the proper loop has been dumped by the steam / water dump system? (2.0)
- c. What is the purpose of the steam / water dump system? (1.25)
QUESTION 6.04 (3.00)
What are three (3) conditions which may be indicated by a " Slack Cable" clarm for a region? (3.0)
QUESTION 6.05 (3.00)
- a. Explain why the Start Up and Wide Range Nuclear Instrument systems both contain circuitry to eliminate any signal due to gamma, but the Linear Power Range does not. (1.25)
- b. Name the trip outputs from the from the Wide Range Summing Amplifier. (0.75)
- c. List all the trip signals supplied by the Start Up Range Nuclear Instrument channels. (1.0)
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
6t__EL8NI_SYSIENS_DESIGNt_CONIBQLt_8NQ_INSIBUMENI6IION PAGE 6 QUESTION 6.06 (2.50)
- e. Describe briefly the design features of a control rod that allows it to accidentally fall from full withdrawn and still fulfill it's intended function of keeping the neutron absorbing material within the core proper. (1.0)
- b. If the operator cannot withdraw a control rod, how can he immediately differentiate between core binding and electrical failure. (0.75)
- c. What conditions and/or preparations are required to operate a control in manual ? (0.75)
QUESTION 6.07 (3.00)
Fill in the following table to complete the information about the Ft. St.
Vrain control loops. (3.0)
CONTROL: BY ADJUSTING: ANTICIPATORY OR FEED FORWARD SIGNAL
- 1. Main Steam a.______________ b._____________
Pressure
- 2. c.__________ Helium Flow d._____________
- 3. e. f. Turbine / Gen.
Load (FSP)
[ QUESTION 6.08 (1.50) l From where can makeup water be supplied to the Service Water System?
Indicate which source (s) are preferred. (1.5) i i
(***** END OF CATEGORY 06 *****)
Zi__EB9CEDUBES_ _NQBM8L,_eBNQBU8Lt_EMEBGENGl_8NQ PAGE 7 88DIQLQGIG8L_CONIBQ8 QUESTION 7.01 (3.00) .
- a. What action should be taken by the operator after confirmation that the correct loop has been dumped by the steam / water dump system (due to high primary moisture)? (1.5)
- b. What does this action accomplish? (1.5)
QUESTION 7.02 (4.00)
Ccmplete the table belcw dealing with external exposure limits.
- a. Federal whole body limits per 10CFR20:
- 1. _____ rem /qtr without exposure history. (0.5)
- 2. _____ rem /qtr with exposure history. (0.5)
- 3. _____ rem accumulated lifetime. (0.5)
- 4. _____ % of above for minors (under 18 years of age). (0.5)
- e. FSV Whole Body Guideline: _____ rem / week. (0.5)
QUESTION 7.03 (2.00)
Fill in the blanks below. Blanks may contain one or more words or numbers.
- a. A Contaminated Area is any area in which there exists loose surface contamination in excess of _(1)_ and/or _(2)_. (1.0)
- b. A Radiation Area is an accessible area where a maj or portion of the whole body could receive greater than _(1)_ or _(2)_. (1.0)
QUESTION 7.04 (3.00)
What are the five immediate actions required following an automatic scram?
Specific steps to complete each action are NOT required. (3.0)
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
t Zz__EBQGEDUBES_:_NDBM8Lt_6BN9Bdekt_EUEBGENGl_8NQ PAGE 8 B6DIQLQGIC8L_G9NIBQL QUESTION 7.05 (3.00)
What configurations of decay heat removal that are used normally can NOT be used under the following conditions and WHY?
- a. Decay heat removal with the condenser out if service. (1.25)
- b. Decay heat removal with the deserator out of service. (1.75)
QUESTION 7.06 (4.00)
- c. What should the following switch positions be prior to withdrawing
, control rods for startup?
- 1. Reactor Mode Selector Switch.
- 2. Manual Scram Switch.
- 3. Rod Group Sequence Bypass Switch.
- 4. Interlock Sequence Switch. (1.5)
- b. Upon increasing power from 20%, what thermodynamic or enthalpic balancing should the operator perform? WHY? (2.5)
QUESTION 7.07 (3.00)
Answer the following questions concerning " Loss of Helium Circulation".
i e. Explain the limits on the number of circulators that can be operated during recovery. (1.5)
- b. Explain why and how the PCRV is depressurized. (1.5)
QUESTION 7.08 (3.00)
What are two (2) plant conditions which require insertion of one or both subgroups of the Reserve Shutdown System.
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(***** END OF CATEGORY 07 *****)
At__8DdINISIB8IIVE_EBQCEDUBESt_GQNDIIIQN2t_8ND_LIMII8IIQN1 PAGE 9 QUESTION. 8.01 (3.00)
What does signing the log when starting a shif t signify (2 answers req'd)?
QUESTION 8.02 (3.00)
- a. Is the operator required to input a manual reactor scram after an automatic scram which is caused by a known and understood human error that is corrected immediately (within 5 seconds)? EXPLAIN? (1.0)
- b. For the case above, is the operator permitted to reset the scram as soon as the cause >_ known and understood? EXPLAIN? (1.0)
- c. What is the minimum elapsed time after any reactor scram that the scram may be reset? EXPLAIN? (1.0)
QUESTION 8.03 (3.50)
Fill in the blanks in the paragraph below dealing with Technical Specification administrative controls. Each blank may represent one or more words or numbers.
Members of the plant staff who perform safety-related functions should l work a(n) _Ca)_ hour day, _(b)_ hour week when the plant is _(c)_. On a i temporary basis, the following guidelines shall be followed:
(1) An individual should not be permitted to work more than _(d)_
straight.
l (2) An individual should not be permitted to work more than _(e)_ in any _(f)_ period, nor more than _(g)_ in any _Ch)_ period Call excluding _Ci)_).
l (3) A break of at least _(j)_ should be allowed between work periods.
l
! (4) Except during _(k)_, the use of overtime should be considered on j an individual basis and not for the entire staff on shift.
QUESTION 8.04 (3.00)
When do Technical Specifications require two (2) Licensed operators in the Control Room? (3 answers required for full credit) (3.0)
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
Az__eDUINISIB8IIVE_EB9CEDUBESt_CQNDIIIQNSt_8ND_LIMII6IIQNS PAGE 10 QUESTION 8.05 (3.00)
When either the fire or radiological alarm is sounded, on back shift, what porsonnel (by job title) are required to respond to the Control Room (i.e.,
the Control Room is their assigned personnel accountability station)? (3.0)
QUESTION 8.06 (3.00)
- a. Per the RERP, initial accident classif ication is the responsibility of
_(1)_, in the role of _(2)_. (1.5)
- b. After declaration of an emergency, notification to the NRC will be made within what time period? (0.75)
- c. After declaration of an emergency, notification of of f site authorities must be initiated within what time period? (0.75)
QUESTION 8.07 (2.50)
What are two (2) of the three (3) purposes of a clearance card, specified in P-2, " Equipment Clearances and Operational Deviations?" (2.5)
QUESTION 8.08 (2.00)
- e. Who (by job title) is responsible for determining if a Radiation Work Permit (RWP) is required? (1.0)
- b. What is one (1) purpose of the RWP? (1.0)
QUESTION 8.09 (2.00)
- n. Define " Safety Limit," per Technical Specifications (TS). (1.0)
- b. Define " Limiting Safety System Setting," per TS. (1.0) l l
(***** END OF CATEGORY 08 *****)
L [************* @NB 0F EXAMINATION ***************)
St__INEQBY_QE_NUGLE88_EDWEB_EL8MI QEEB8IIONt_ELUIQat_8NQ PAGE 11 IHEBdQDINedICS ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
4 ANSWER 5.01 (3.00)
- 1. Decreases negative temperature coefficient (since it is a positive contributor to thermal utilization factor). (1.5)
- 2. Decreases Beta factor (from 0.0055 t BOL to 0.0048 G EOL). (1.5) i REFERENCE FSV Reactor Physics Lesson Plan on Core Changes.
ANSWER 5.02 (4.00)
- a. 1. increase (0.5)
- 2. 40~50 (0.5)
- 3. the 6.7 hour8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> half-life of I-135 (OR the decay of I-135) (1.0)
C
- b. 1. 14 days (0.5)
- 2. flux dependent (0.5)
- 3. increase (0.5)
- 4. 12 (0.5)
REFERENCE LR 008.00, HTGR - Fission Product Poisons, p. 7, 14 i
ANSWER 5.03 (1.50)
Pc-233 (0.5 Pa, 0.5 233) l REFERENCE LR 008.00, HTGR - Fission Product Poisons, p. 15 l
l I
Si__IHE9BI_9E_NUGLE88_EQWEB_EL8HI_DEEB8IIONi_ELVIQ1t_8NQ PAGE 12 IHEBdQQIN8 DICE ANSWERS -- FT.-ST. VRAIN -86/05/20-PELLET, J.
ANSWER 5.04 (4.00) dH = H(out) - H(in) (0.6)
T(out) = 1000 F +/- 50 F C0.6)
P(out) = 2400 psig +/- 1001 (0.3); PCin) = 3350 psig +/- 1005 (0.3) (0.6)
T(in) ~= 400 F + 50 F / -0 F CO.6)
HCin) H(sat) e inlet temperature (0.6)
= 320~430 Btu /lbm (0.2)
H(out) = 1470~1450 Btu /lbm (0.2) dH = (1470-320)~C1450-430) = 1150~1020 Btu /lbm (0.2)
Use of steam tables to get 320, 430, 1450, a 1470 Btu /lbm (0.4)
REFERENCE CE Steam Tables, S0P 22-1 ANSWER 5.05 (4.00)
- a. Rankine (6 answers e 0.667 ea.)
- b. force exerted
- c. 12.3 psia
- d. mercury
- e. greater than
- f. 270 F REFERENCE LR 007.00, various pages ANSWER 5.06 (3.00) l.
- a. M: The ratio of the total number of fission and source neutrons to the total number of neutrons which would exist due to the source only. (1.0)
- b. As Keff approaches 1, M approaches infinity and cannot be used to predict criticality graphically. (1.0) i
- c. CR = 100(1-0.95 / 1-0.99) = 100(5) = 500 cps (1.0) l I REFERENCE FSV Nuclear Energy Training, Unit 4, Sec. 6.
St__INEQBl_QE_NUQLE8B_EQWEB_EL6NI_QEEB6IIQNt_ELUIQ1t_8NQ PAGE 13 IUEBMQQ1NedIQ1 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER 5.07 (3.00)
- c. 1. This results in equivalent power being produced in a smaller volume (0.5) and peaking increases throughout the power producing regions of the core. (0.5) (1.0)
- 2. The softer neutron spectrum of the older regions causes localized power peaks at the periphery of the new region. (1.0)
REFERENCE FSV Nuclear Energy Training Operations, p. 4.6-5 ANSWER 5.08 (2.50)
- a. Increases
- b. Decreases
- c. Increases
- d. No Change
- e. Increases (5 answers e 0.5 ea.)
REFERENCE FSV Plant Performance, Unit 6, Sec. 1 l
6t__EL8NI_SYSIEd2_DE219Nt_99 NIB 9Lt_8NQ_INSIBUdENI8II9N PAGE 14 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER 6.01 (4.50)
- 1. Fire Water Pump 10. Rx Plant Ex Fan
- 2. Service Wir Pump 11. OG Oil Xfer Pumps (2)
- 3. Svce Wtr Tower Fan 12. Helium Purification
- 4. Svce Wir Return Pump Cooling Water Pump
- 5. PCRV Liner Cooling 13. Firewster Pump House System Pumps (2) Vent Fans
- 6. Circ Wtr MU Pump 14. MOV's HV-2301/02
- 8. Selected Plant Light 16. Breathing Air Comp
- 9. S/U Battery for OG/DC Control (any 9/16 e 0.5 ea.)
REFERENCE FSV SOP 48-01, p. 4 ANSWER 6.02 (3.50)
- 1. Emergency condensate. (0.5)
- 2. Emergency feedwater. (0.5)
- 3. Firewater - Extreme emergency source. (1.0 - 2 e 0.5 ea.)
- 4. Bypass flash tank. (0.5)
- 5. Turbine extraction / cold reheat steam. (0.5)
- 6. Auxiliary boiler steam. (0.5)
(4, 5, & 6 are from Cold Reheat. Also accept Aux Boiler Feed Pump)
REFERENCE FSV SOP 21-01, p. 5 ANSWER 6.03 (4.00) l
- e. 15 seconds (accept any answer < 30 seconds) (0.75)
- b. 1. Dump tank pressures > ~250 psia. (1.0)
- 2. High dump tank radioactivity. (1.0)
- c. Minimize the amount of water that can leak into the primary system as a result of a steam generator tube or subheader rupture. (1.25)
REFERENCE R0 127.00, various pages l
l
62__EL8NI_SYSIEbS_DESIGNt_G9 NIB 9Lt_8NQ_INSIBudCNI8II9N PAGE 15 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER 6.04 (3.00)
- 1. A control rod is stuck in the guide channel. (1.0)
- 2. A control rod cable is parted . (1.0)
- 3. Adsorber section(s) have dropped into the guide channels. (1.0)
(OR Malfunction in slack cable circuitry.)
REFERENCE FSV SOP 12-01, Section 5 ANSWER 6.05 (3.00)
- m. The Start up and Wide Ranges will operate in an area where decay gamma can be detected and will cause an erroneous indication. (0.5) In the area where the Linear range operates contributien from gamma becomes insignificant. (0.5] and f ission gamma is proportional to reactor power level. (0.25] (1.25)
- b. 1. Start Up channel high voltage off.
REFERENCE FSV R0003.00 i ANSWER 6.06 (2.50)
- o. The rods are made of absorber container sections and the neutron absorbing material will stay within the core within it's cantainer.
(0.5) The bottom section is a bellows type shock absorber.CO.5] (1.0)
- b. Attempt to insert the rod. A slack cable warning light means that the rod is binding. (0.75)
- c. The reactor must be depressurized (0.5] with the refueling penetration blank flange removed. (0.25] (0.75) l l REFERENCE i FSV SD 12-1, P.5 & 12-2, P.3 & 12-3, P.7 l
l
Ez__EL8NI_SISIEUS_DESIONt_GQNIBQLi_8ND_INSIBudENI6IIQN PAGE 16 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER 6.07 (3.00)
- c. Feedwater Flow
- b. T/G Load / 1st Stage Press / Throttle Press
- c. Main Steam Temp.
- d. Feedwater Flow
- o. Reheat Sim. Temp.
- f. Helium Flow / Rx Power (0.5 each) (3.0)
REFERENCE FSV MS 85.00.05 ANSWER 6.08 (1.S0)
- 1. Shallow well water---Emergency
- 2. Domestic water-------Normal
- 3. Circ. Water make up--Alternate emergency (6 answers e 0.25 ea.)
REFERENCE S0-42, P.7 & SD-46, p.4 l
I
Z2__EBQCEQMBES_:_NQBU8Li_8BNQBd8Lt_EMEBQENCY_8NQ PAGE 17 88DIDLQQIG8L_GQNIBQL ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER 7.01 (3.00)
- m. Remote-manually close the dump valves. (1.5)
- b. Stops helium leakage into the dump tank. (1.5)
REFERENCE RO 127.00, p. 20 ANSWER 7.02 (4.00)
- o. 1. 1.25
- 2. 3.0 ,
- 3. 5(n-18)
- 4. 10
- b. 7.5
- c. 18.75
- d. 0.125
- o. 0.100 (8 answers e 0.5 ea.)
REFERENCE FH 058.00, p. 12, 13; 10CFR20 ANSWER 7.03 (2.00)
- u. 1. 1000 dpm/100 sq cm beta-gamma. (0.5)
- 2. 10 dpm/100 sq cm alpha. (0.5)
- b. 1. 2.5 mrem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. (0,5)
- 2. 100 mrem in 5 consecutive days. (0,5)
REFERENCE FH 060.00, p. 3, 4
}
Zi__EBQGEQUBEQ_ _NDBdekt_8BNDBdeLt_EMEBGENC1_8ND PAGE 18 B6DIDLQQIGeL_GQNIBQL ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER 7.04 (3.00)
- 1. Insert a manual scram. (5 answers e 0.6 ea.)
- 2. Place ISS on Low Power Position.
- 3. Ensure transfer of house power.
- 4. Ensure turbine trip.
- 5. Ensure / establish stable core cooling conditions.
REFERENCE FSV EP APP 8-1, p. 7-9 ANSWER 7.05 (3.00)
- c. 1. Cond. to S/G's---------------- Steam to circulators. (0.5)
- 2. Feedwater to S/G's------------ Steam to circulators (0.5)
REASON: No heat sink for circulator exhaust. (0.25)
- b. 1. Cond to S/G's----------------- Feed to circulators. (0.5)
- 2. Feed to S/G's----------------- Feed to circulators. (0.5)
- 3. Feed to S/G's----------------- Steam to circulators. (0.5)
REASON: Feedwater not available without deaerator. (0.25) l REFERENCE FSV OPOP II, p. 4,12,20 ANSWER 7.06 (4.00)
- a. 1. Run l 2. Normal
- 3. Normal
- 4. Start (0.325 ea.) (1.5)
- b. 1. Adj ust Helium Flow Orifice Valves (0,5) to equalize reheat steam outlet temperatures. (0.75) (1.25) l 2. Adj ust S/G f eedwater flow C0.5) to equalize Main Steam header outlet temperatures. (0.75) (1.25)
REFERENCE FSV OPOP III, p. 6,20,21
Zt__EB9CEDUBE2_:_N9Bd6Lt_6BNDBU6Lt_EMEBGENCY_6NQ PAGE 19 B8DIQLQGIG6L_GQNIBQL ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER 7.07 (3.00)
- e. 1. 1 circulator on condensate or boosted firewater when the PCRV is pressurized
- 2. 2 circulators on feedwater when PCRV is depressurized.
- 3. Limits are to prevent circulator stall. (0.5 ea.) (1.5)
- b. WHY: To limit the free convection heat transport from the core to the upper plenum so that load is within the capability of the liner cooling system. (0.75)
HOW: Through the helium purification system to the reactor building exhaust stack. (0.75)
REFERENCE FSV EP APP G, p. 3,7 ANSWER 7.08 (3.00)
- 1. Loss of forced coolant flow (0.75) for time per power history (0.75).
- 2. Sufficient rods cannot be inserted to provide 0.01 dK SOM (0.75) at hot shutdown (220 F & Xe-free) (0.75).
- 3. Operator verifies the reactor cannot be maintained shutdown following any manual or automatic scram or shutdown (1.5).
l Cany 2/3 answers e 1.5 ea.)
REFERENCE FSV SOP 12-05, Rev. 8, p. 4 1
l l
l I
l I
I l
l
At__8DMINISIB8IIVE_EBQGEQUBES _CQNDIIIQN1t_8NQ_LIMII8IIQN1 PAGE 20 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER 8.01 (3.00)
- 1. Acknowledge plant status. (CONCEPT) (1.5)
- 2. Accept responsibility for the shift. (CONCEPT) (1.5)
REFERENCE FSV P-1, p. 3 ANSWER 8.02 (3.00)
- a. Yes (0.5). A manual scram is required by procedure for any automatic scram regardless of cause (0.5).
- b. No (0.5). The scram reset must occur after completion of the scram procedure (which does not include reset) (0.5),
- c. 160 seconds (0.5). This is the minimum time required after scram before reset to allow all rods to fully insert (0.5).
OR After completion of SMAP-7 & S/U OPOP (concept).
REFERENCE FSV EP APP B-1, p. 7 FSV TS, SR 5.1.1, p. 5.1-1 ANSWER 8.03 (3.50)
- a. 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, f. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- b. 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />. g. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
- c. operating. h. 7 days /1 week.
- d. 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />. i. shift turnover time.
- s. 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />. J. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
- k. extended shutdown period.
(11 answers e 0.318 ea.)
REFERENCE FSV TS, Section 7.1.1, p. 7.1-5, 6 i
?
1 ANSWER 8.04 (3.00)
- 1. Reactor startup. (1.0) l 2. Reactor shutdown. (1.0)
- 3. Recovery from reactor trip. (1.0)
At__8DMINISIB8IIVE_EBQGEQMBESt_GQNDIIIQNSt_8NQ_LIMII8IIONS PAGE 21 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
REFERENCE FSV TS 7.1.1.2.b, p. 7.1-3 4
ANSWER 8.05 (3.00)
- 1. Shift Supervisor
- 2. Reactor Operators
- 3. Auxiliary Tenders
- 4. Equipment Operators
- 5. Duty Technical Advisor
- 6. Lead Security Officer (7. Off-duty Shift Supervisor)-Optional /Not required for full credit.
(6 answers required 0 0.5 ea.)
REFERENCE FSV Admin Proc G-5, Attach G-5A, Rev 18, p. 11, 12 ANSWER 8.06 (3.00)
- c. 1. the Shift Supervisor. (0.75)
- 2. Emergency Coordinator. (0.75)
- b. I hour. (0.75)
- c. 15 minutes. (0.75)
REFERENCE FSV RERP-CR, Rev 7, p. 3, 4 i
l ANSWER 8.07 (2.50)
- 1. Assure the safety (of personnel working on the equipment).
l 2. Notify operating personnel when equipment is not available.
- 3. Indicate steps necessary to restore system to normal.
Cany 2/3 e 1.25 ea.)
REFERENCE FSV Admin Proc P-2, Rev 13, p. 2
'Az__8DdINISIB8IIVE_EBOCEDMBEft_G9NDIIIQNSt_8NQ_LIMII8IIONS PAGE 22 ANSWERS -- FT. ST. VRAIN -86/OS/20-PELLET, J.
ANSWER 8.08 (2.00)
- c. The Shift Supervisor. (1.0)
- b. 1. Record of permission for work under specified controls.
- 2. Alert Supervisors about work performed in hazardous area.
- 3. Identify protective measures required.
Cany 1/3 e 1.0)
REFERENCE FSV Admin Proc P-1, Rev 13, p. 19 ANSWER 8.09 (2.00)
- a. Safety limits are limits defined to protect the fuel particle integrity and primary reactor coolant system boundaries. (1.0)
- b. LSSS are settings for instrumentation and protective devices related to process variables upon which Safety Limits are based. (1.0)
REFERENCE FSV TS 3.0, p. 3.0-1
U.S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR REQUALIFICATION EXAMINATION LIMITED TO FUEL HANDLING Facility: FT. ST. VRAIN Reactor Type: HTGR-GA Date Administered: 86/05/20 Examiner: J. L. PELLET Examinee:
INSTRUCTIONS TO EXAMINEE:
THIS EXAMINATION REPLACES THE CURRENT CYCLE FACILITY ADMINISTERED EXAMINATION FOR DEMONSTRATION OF OPERATOR PROFICIENCY AND LICENSE RENEWAL.
FAILURE OF THIS EXAMINATION WILL REQUIRE RETRAINING UNDER THE CURRENT FACILITY REQUALIFICATION TRAINING PROGRAM AND MAY REQUIRE RE-EXAMINATION BY NRC. Points for each question are indicated in parentheses after the quastion number. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up FOUR (4) hours after the examination starts.
% of Category % of Candidates's Category Value Total Score Value Category 14.00 18.54 M. Reactor and Fuel Characteristics 15.50 20.53 N. Equipment, Instrumentation and Design Description 21.00 27.81 0. Procedures and Limitations 12.00 15.89 P. Emergency Systems and Safety Devices 13.00 17.22 Q. Health Physics and Radiation 75.50 TOTALS Final Grade %
All work done on this examination is my own. I have neither given nor roceived aid.
Examinee's Signature
Uz__BE6CIQB_eNQ_EMEL_GH888CIEBISIIGH PAGE 2 4
QUESTION M.01 (3.00)
Explain the three (3) parameters or occurrences that can affect Shutdown Margin (SDM) after shutdown for refueling. (3.0)
QUESTION M.02 (2.00)
Explain how an adverse reaction in irradiated fuel stored in fuel storage walls is prevented. (2.0)
QUESTION M.03 (3.50)
- a. Explain how and why removing spent fuel may either increase or decrease core reactivity. (1.5)
- b. Explain why removal and installation of reflectors affects core reactivity. (1.5) i
- c. How does removing a depleted neutron source affect reactivity? C0.5) l I
- QUESTION M.04 (2.50) l i a. Explain whether the control rods are more effective at " Refueling" or
" Operating" conditions. (1.25)
I l b. What accounts for the change in total control bank worth from Beginning i of Cycle to equilibrium core? (1.25) l l
l l
QUESTION M.05 (3.00)
Dascribe and explain the benefits of AXIAL and RADIAL fuel zoning for the reactor core. (3.0) i
(***** END OF CATEGORY M *****)
Nz_ EQUIEMENIt_INSIBudENI8IIQN_8NQ_QE210N_QE20BIEIIQN PAGE 3 QUESTION N.01 (2.00)
- e. Describe the seals on the top and bottom of a Reactor Isolation Valve (RIV). (1.0)
- b. Describe four (4) RIV internal interlocks. (1.0)
QUESTION N.02 (2.00)
- a. What piece of equipment is used to install the refueling sleeve? (0.5)
- b. How is any offset between a top reflector element and the Fuel Handling Machine grapple mechanism automatically eliminated before picking up the reflector? (1.5)
QUESTION N.03 (2.50) ,
- c. Describe the capacity of a fuel storage well for fuel elements and reflectors. (1.0)
- b. What are the three (3) cooling supplies for the fuel storage wells? (0.75)
- c. What are the two (2) backup cooling supplies for the fuel storage wells? (0.75)
QUESTION N.04 (3.00)
- c. Describe the method used to remove contaminated gases if both vacuum pumps of the Fuel Handling Purge System should fail. Describe system design features that allow this method to be utilized. (1.0)
- b. Describe two (2) unique design protective features for the Fuel Handling Purge System Vacuum Pumps. (1.0)
- c. What generally limits the vacuum that can be attained on a system within a short period of time? (0.5)
- d. Why is it undesirable to remove large volumes of moisture laden air l
with the vacuum pumps? (0.5) l l
l
(***** CATEGORY N CONTINUE 0 ON NEXT PAGE *****)
Nz__E991EMENIt_INSIBMMENI8IIQN.8NQ_QESIGN_QESGBIEIIQN PAGE 4 QUESTION N.05 (3.00)
- e. What instrumentation is available to aid in the detection of a leaking tube or a loss of cooling to a fuel storage well. (1.0)
- b. How are the fuel storage facility internals cleaned? (1.0)
- c. What would happen to the fluid from a leaky tube in the annular interspace between a fuel well and the outer containment tank? (1.0)
OUESTION N.06 (3.00)
- c. What major items are normally decontaminated in the Hot Service Facility? (1.5)
- b. Describe the treatment and disposition of solutions collected at the bottom of the Hot Service Facility or Storage Wells (1.5) i i
l l
(***** END OF CATEGORY N *****)
y ,
92__EB99EQUBES_eNQ_LIMII6II9NS PAGE 5 QUESTION 0.01 (2.00)
- c. What action (s) should personnel in the Reactor Building take if Reactor Building activity monitors indicate that an abnormal radioactive gas release is taking place? (1.0)
- b. How would personnel working in the Reactor Building be informed that an abnormal release was taking place? (1.0)
QUESTION- 0.02 (2.00)
- c. During Shutdown Margin (SDM) determination, what is the criteria used to determine that a control rod is fully inserted? (1.0)
- b. When would the Refueling Supervisor normally inform the control room to perform a Shutdown Margin? (1.0)
QUESTION 0.03 (2.00)
- a. Other than control rod withdrawal, what are two refueling evolutions that can cause erratic oscillation or increase on the Start-Up Nuclear instrument channels? (1.0)
- b. What are the limits on control rod withdrawal for CRDOA removal?(1.0) f QUESTION 0.04 (2.00)
When (2 conditions) should a complete startup of the Fuel Handling Machine bo performed? (2.0)
QUESTION- 0.05 (2.00)
What two (2) precautions should be observed during operation of the Fuel H:n,dling Machine handwheels? (2.0) f
(***** CATEGORY 0 CONTINUED ON NEXT PAGE *****)
91__EB0GEDUBES_8NQ_LIMII6IIONS PAGE 6 QUESTION 0.06 (3.00)
What actions must be taken if the conditions for Technical Specification LCO 4.5.1, " Reactor Vessel Internal Maintenance," cannot be met? (3.0) ,
QUESTION 0.07 (3.00)
What are the precautions for refueling as given in the FHP Manual? (3.0) m QUESTION 0.08 (2.00)
What four (4) departments are involved in new fuel receipt? (2.0)
QUESTION 0.09 (3.00)
- c. What department is responsible for coordinating spent fuel shipping activities? (0.5)
- b. How many fuel elements are stored in a fuel shipping cask? (0.5)
- c. Fill in the blanks concerning LCO 4.7.4, " Spent Fuel Shipping l Container." Blanks may contain one or more words or phrases. (2.0) l
_(1)_ (0.5) of spent fuel prior to _(2)_ (0.5) of decay time is l
l allowable provided that _C3)_ (1.0).
l l
l l
l l (***** END OF CATEGORY 0 *****)
22__EUEB9ENGY_SYSIEMS_8NQ_S8EEIY_DEVIQES PAGE 7 QUESTION P.U1 (2.00)
What are the four (4) emergency conditions or classifications in order of INCREASING severity, per the Fort St. Vrain Event and Emergency Classification Overview procedure?
QUESTION P.02 (1.50)
<5cch of the definitions below, from Technical Specifications, is WRONG.
State the correct definition, or describe what must be changed to make the g iven def init ion cor r ect. (1.5)
- c. The reactor is considered shut down f or refueling purposes when the reactor mode switch is placed in the " Fuel Loading" position timultaneous with either hot shutdown or the cold shutdown reactivity conditions. (0.5)
- b. Irradiated fuel is fuel that has been installed in the reactor during power operation. (0.5)
- c. A channel check is a test of the functional capability of the channel to determine that it is operable. (0,5)
,s a QUESTION P.03 (2.50)
- n. What process parameter is monitored, at what setpoint, to assure that irradiated fuel elements being handled with the Fuel Handling Machine are maintained less than 750 F? (1.0)
- b. What is the potential problem associated with excessive fuel element temperatures? (0.75)
- c. What non-normal condition, besides temperature, must be present for a problem to occur? (0.75)
- QUESTION P.04 (1.00)
How is adequate cooling maintained to a fuel storage well that has suffered a complete loss of both cooling coils? (1.0) 4 i
(***** CATEGORY P CONTINUED ON NEXT PAGE *****)
Et__EMEBGENGY_SYSIEMS_8NQ_18EEII_ DEVICES PAGE 8 1
QUESTION P.05 (2.50)
- b. How would such a scram affect fuel handling operations? (0.5)
- c. If you were involved in operations from the refueling floor, what would make you aware of the scram? C0.5)
QUESTION P.06 (2.50)
- n. What is the purpose of the grapple interlock? CO.5)
- b. What conditions and logic are measured or detected to satisfy the interlock? (1.25)
- c. What condition (s) cannot be bypassed to satisfy the interlock? (0.75) l l
l
(***** END OF CATEGORY P *****)
9t__HE8LIH_EH111Gl_eNQ_B8DISIION_EB9IEGIIQU PAGE -9 QUESTION Q.01 (4.00)
Complete the table below dealing with external exposure limits.
- u. Federal whole body limits per 10CFR20:
- 1. _____ rem /qtr without exposure history. (0.5)
- 2. _____ rem /qtr with exposure history. (0.5)
- 3. _____ rem accumulated lifetime. (0.5)
- 4. _____ % of above for minors (under 18 years of age). (0.5)
- e. FSV Whole Body Guideline: _____ rem / week. (0.5)
QUESTION Q.02 (2.00)
Fill in-the blanks below. Blanks may contain one or more words or numbers.
- a. A Contaminated Area is any area in which there-exists loose surface contamination in excess of _(1)_ and/or _(2)_. (1.0)
- b. A Radiation Area is an accessible area where a major portion of the whole body could receive greater than _(1)_ or _(2)_. (1.0)
QUESTION Q.03 (3.00)
Answer the following que1tions dealing with Radiation Work Permits (RWP's).
- a. What does your signature, as a worker, signify on an RWP? (1.0)
- b. What action should you take if you are performing work covered by an RWP and the job is not completed at the end of your shift, but will be resumed on your shift the next day? (1.0)
- c. What type of radiation instrument is used exclusively for setting dose rates for personnel at FSV? (0.5) l d. When entering a radiation area with a portable survey instrument, which range on the detecto'r should be selected prior to entry? (0.5)
(***** CATEGORY Q CONTINUED ON NEXT PAGE *****)
91__HE8 LIB _EHISIGS_8NQ_86DI6110N_EBQIEGIIQN PAGE 10 QUESTION Q.04 (4.00)
What are three (3) of the four checks which must be performed before you can use an RO - 2 to perform a radiation survey? (3.0)
QUESTION Q.05 (1.00)
When moving an irradiated control rod drive assembly in the ATC, where should you, the operator, stand in relation to the ATC to maintain your exposure ALARA? WHY? (1.0)
(***** END OF CATEGORY Q *****)
R************* END OF EXAMINATION ***************)
Mi__BE8CIDB_eND_E9EL_GH888CIEB11IIQ1 PAGE 11 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER M.01 (3.00)
- c. Due to Negative Temperature Coefficient, the lower core temperature, the more reactive the core. (1.0)
COR Primary temperature - whether e hot (220 F) or cold (80 F) shutdown.)
REFERENCE FSV LESSON NO. FH 038.00.01.16.1 ANSWER M.02 (2.00)
- a. The 02 content of cover gas is kept below 5 % to eliminate any chance of combustion (He atmosphere is maintained). (1.0)
- b. The surface temp. of spent fuel is kept below 750 deg's to prevent the occurrence of oxidation. (1.0)
REFERENCE FSV SD 14-1, P. 3 ANSWER M.03 (3.50)
- a. 1. Decreases reactivity by removing f iss ionab le mater ials. (0.75)
- 2. Increases reacti vity by removing f ission product poisons (including Xenon and Samarium). (0.75)
- b. 1. Removing reflector decreases reactivity by allowing increased neutron leakage. (0.75)
- 2. Installing reflectors increases reactivity by returning more neutrens to the core. (0.75)
- c. Decreases reactivity. (0.5)
Ez__BE8CIDB_8ND_EMEL_GBeBeGIEBISIIGS PAGE 12 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
REFERENCE FSV LESSON NO. 39.00, P. 3 ANSWER M.04 (2.50)
- e. The rods have a higher relative worth hot than cold (0.5) because the absorption cross section of the fuel goes down more rapidly with temperature than rod worth. (0.75) (1.25)
- b. The Rod worth at equilibrium is slightly less (0.75) due to depletion of burnable poison. (0.5) (1.25)
REFERENCE LESSON NO. FH 038.00, p. 7,8 ANSWER M.05 (3.00)
AXIAL: There are 2 axial zones with 55% of the fuel in the upper half of the core and 45% in the lower half (0.75). This takes advantage of the cooler helium temperature in the upper half of the core (0.75).
RADIAL: There are 5 radial zones in concentric circles (0.375). Th-232 loading is higher &U loading lower in the outer zone (0.375).
Radial zoning helps flatten the flux profile (0.375) and the changes in Th/U loading helps minimize peaking at the outer edges of the core (0.375).
REFERENCE FH 033.00, p. 11, 12
Nz__EQUIEUENIt_INSIBUUENI8110N_8HQ_ DESIGN _DESGBIEIIQN PAGE 13 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER N.01 (2.00)
BOTTOM - An inflatable seal on bottom of valve seal the RIV to the refueling penetration or fuel storage well. (0.5)
- b. 1. Differential pressure - Gate will not open unless dp above and below gate is < 1 psid.
- 2. Differential atmosphere - Gate will not open unless atmospheres are identical.
- 3. Seals inflate on loss of power.
- 4. Seal must be deflated to open gate.
- 5. Alarm if gate is opened.
- 6. Alarm if seul pressure < 20 psig. (Any 4/6 0 0.25 ea.)
REFERENCE FSV, FH.O.49.00.01.05.5 ANSWER N.02 (2.00)
- e. Auxiliary Transfer Cask. (0,5)
- b. When the tapered end of the pickup probe enters the element, it displaces a horizontal translator mechanism. (0.75) This offset data
! inputs to the control center which auto positions the head until the error signal is zero. (0.75) (1.5)
REFERENCE SD-13-1, p.7,8 l
l
! I i
, - - , , -,,e- --- -- -- , , - . , - - - - - , - - . - - - - - - - - - - - -~
Nz__EDUIEMENIz_INSIBubENI8IION_8HQ_ DESIGN _DESGBIEII9N PAGE 14 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER N.03 (2.50)
- e. 1. Each well can store 14 elements in each column for a total of 56. (0.5)
- 2. One column can hold 14 full length elements or 28 half length elements. (0.5)
- b. 1. Helium purge.
- 2. Cooling water from R.P. Cooling Water system.
- 3. Reactor Bldg. HVAC. (0.25 ea.) (0.75)
- c. 1. Fire water.
- 2. Booster fan and dampers. (0.375 ea.) (0.75)
REFERENCE FSV FH 041.00.01.03.1, 2 FH 041.00.01.05.1, 2 ANSWER N.04 (3.00)
- e. In the unlikely event that both pumps fail, a pump bypass line (0.5) permits the equipment and storage wells to be purged with cleen helium and discharged directly to the gas waste system. (0.5) (1.0) i
, b. 1. A pressure controlled bypass auto opens if discharge pressure l reaches 3 psig to prevent overload. (0.5) l
- 2. A temperature switch causes the pump suction valve to shut if pump oil reservoir temp. should reach (170 deg's) which will prevent pump l damage. (0,5)
- c. The saturation pressurization of any moisture in the system. (0.5) l d. High water removal rates result in water emulsif ication with the pump oil and subsequent pump downtime. (0.5)
REFERENCE FSV SD , 2, p.2,4,5 l
l l
Nz__EQUIEUENIt_INSIBWdENI8IIQN_8NQ_QESIGN_QESCBIEIIQN PAGE 15 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER N.05 (3.00)
- n. Well thermocouples, moisture monitor, and flow switches. (1.0)
- b. They are back purged with purified helium. (1.0)
- c. Individual drain lines from each well are combined and routed to the Reactor Plant Drain system. (1.0)
REFERENCE FSV S0-14-2, p. 3,4 ANSWER N.06 (3.00)
- e. 1. Control rod drives Refueling Sleeve
- 2. Fuel shipping cask RIV He Circ (Any 3 4 0.5 ea. - 1.5)
- b. Solutions are pumped by a recycle pump through a filter for removal of particulates. (0.51) Fluids are then stored for reuse, (0.33) pumped to liquid waste (0.33), or put in drums (0.33) (1.5)
, REFERENCE FH 043.00, p. 9 l
l l
l l
l
92__EBQGEQUBES_8HD_LIMII6II9N1 PAGE 16 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER 0.01 (2.00)
- c. Evacuate the RB. (1.0)
- b. Sounding of the Radiological Alarm. (1.0)
REFERENCE FSV EP APP H-2, Rev. 51, p. 3 ANSWER 0.02 (2.00)
- o. " Rod In Light in on with the " Slack Cable light is off. (1.0)
- b. Normally, prior to refueling a region when two control rod pairs are retracted. (1.0)
REFERENCE CMG-13, p. 4, 5 ANSWER 0.03 (2.00)
- e. Removing reflector blocks (0.5] or installing sources. (0.5] (1.0)
- b. 1. Remain < 1.5 dpm on both Start-up channels
- 2. 5 times CRo(NI) from Data sheet 1
- 3. No more than 20" increments. (0.33 es.] (1.0)
REFERENCE FSV CMG-13C, p. 1 & CMG-13, p. 2 ANSWER 0.04 (2.00)
- 1. After an extended shutdown (1.0)
- 2. After any system modification / calibration / repair. (1.0)
REFERENCE FH 051.00.01.04.1
Q___EBQGEQUBES_8NQ_LIMII8IIQN1 PAGE 17 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER 0.05 (2.00)
- 1. Don't move the handwheels too quickly (a servo error may result, causing the system to unclutch). (1.0)
- 2. Don't operate the handwheels when the machine is in the OFF mode. (1.0)
REFERENCE FH 051.00.01.12.1 ANSWER 0.06 (3.00)
- 1. Suspend any internal maintenance (including fuel handling). (1.0)
- 2. Retract any remotely operated mechanisms. (1.0)
- 3. Close the RIV & FHM Cask Valve (ASAP). (1.0)
REFERENCE FH 051.00.04.02.1 ANSWER 0.07 (3.00)
- 1. Maintain 2-way comm. between FHM CR & Ops CR during refueling. (0.75)
- 2. FHM CR & Ops CR coord. rod move. to ensure adequate SDM C.01 dp).(0.75)
- 3. Comply with TS (4.7.1, 4.7.2, 4.7.3). (0.75)
- 4. FHM & ATC remain attached to the overhead crane when located over the HSF Cover the FSW, LP, or Reactor, the overhead crane may be detached if the FHM or ATC is bolted to an RIV). (0.75)
REFERENCE FH 051.00.04.05.1
Oz__EB0GEDUBES_8NQ_LIMII8IIONS PAGE 18 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER 0.08 (2.00)
- 1. Maintenance. (0.5)
- 2. Health Physics. (0.5)
- 3. Quality Assurance. (0.5) ,
- 4. Technical Services. (0.5)
REFERENCE FH 053.00.01.01.
ANSWER 0.09 (3.00)
- e. Mechanical Maintenance. (0.5)
- b. 6. (0.5)
- c. 1. Loading AND/OR Shipping. (0.5)
- 2. 100 days. (0.5)
- 3. less than 200 mr/hr on surface of shipping cask. (1.0)
REFERENCE FH 056.00.01.01.1, FH 056.00.01.02.1, FH 056.00.01.03.1 l
l l
i l
i q
Ez__EUEB9EN91_111IEd2_8NQ_28EEIX_QEVICES PAGE 19 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER P.01 (2.00)
Natification of Unusual Event, Alert, Site Area Emergency, & General Emer.
(4 answers e 0.4 es.; correct order 0.4)
REFERENCE FSV EP CLASS, p. 11-14 ANSWER P.02 (1.50)
- o. Correct except "placed," should be " locked." (0.5)
- b. ...has a radiation level >/= 100 mr/hr at 1 foot from element. (0.5)
- c. Either change " check," to " test," or . . . is a qualitative determination that the channel is operable. (0.5)
REFERENCE FSV TS, Section 2 ANSWER P.03 (2.50)
- a. FHM cooling water outlet temperature (0.5) </= 150 F (0.5).
- b. Graphite oxidation. (0.75)
- c. Air inleakage must also have occurred. (0.75)
REFERENCE FSV TS 4.7.2, p. 4.7-3 ANSWER P.04 (1.00)
Vontilation flow is increased until either cooling can be restored. (1.0)
REFERENCE FSV TS 4.7.1-3, p. 4.7-2, 3, 4 l
i
Ei__EMEBGENGl_111IEB1_8NQ_18EEII_QEVICES PAGE 20 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER P.05 (2.50)
- c. Startup count rate high, 1 out of 2 logic, >/= 1.0E+5 (3 answers e 0.5)
- b. Cease all internal maintenance /f uel handling until saf e. (0.5)
.i c. Operator would advise COR Hazard Alarm on FHM Console). (0.5)
REFERENCE FSV EP APP B-1, p. 2, 11 ANSWER P.06 (2.50)
- c. Protects against dropping of a grappled element. (0.5) l b. All of the following conditions must be met: (0.5)
, 1. At least i dowel pin engaged. (0.25)
- 2. The plate is raised. (0.25)
- 3. Probe weight < 50 lbs. (0.25) (1.25)
- c. The probe weight condition cannot be bypassed. (0.75)
! REFERENCE FSV LESSON NO. FH 051.00, p. 66,67 l
l 2
- ,. .---,-%---.,7.,-y.. . --y -.r..-,--y-,-,,,,---., ,,,--,y.--.---,--.,w---,- ,ym..,- _,v---. ,, - -
,,,,..,,,,,--.e.--,---,w. w..- w -,--w._---
l 94__HE8LIH_EBISICS_eNQ_B8DI8IIQN_EBQIECIIQN PAGE 21 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
,i ANSWER Q.01 (4.00)
- o. 1. 1.25
- 2. 3.0
- 3. SCn-18)
- 4. 10
- b. 7.5
- c. 18.75
- d. 0.125
- o. 0.100 (8 answers e 0.5 ea.)
REFERENCE FH 058.00, p. 12, 13; 10CFR20 ANSWER Q.02 (2.00)
- o. 1. 1000 dpm/100 sq cm beta-gamma. (0.5)
- 2. 10 dpm/100 sq cm alpha. (0.5)
- b. 1. 2.5 mrem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. (0.5)
- 2. 100 mrem in 5 consecutive days. (0.5)
REFERENCE FH 060.00, p. 3, 4 ANSWER Q.03 (3.00)
- e. That you understand the job requirements (CONCEPT). (1.0)
- b. Return white copy to SS. (1.0)
- c. Ion chamber. (0.5)
- d. Start on the highest range (a range down ...). (0.5)
REFERENCE FH 058.00, p. 14; FH 059.00, p. 6, 11
91__HE8LIH_EHISICS_eND_BeQI8IIQN_EBQIECIIQN PAGE 22 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER Q.04 (3.00)
- 1. Inspect for physical damage.
- 2. Check for current calibration.
- 3. Perform battery check.
- 4. Verify response to a known source. Cany 3/4 e 1.0 em.)
REFERENCE FH 059.00.0.03.1 ANSWER Q.05 (1.00)
Stand near the ATC - its shielding design generally produces lower dose rotes there. (1.0)
REFERENCE FH 060.00.01.01.1 I
I I
I
,--,.n-, - ~ n,n,m._, ,,,--a.r-. - . - - - , , - - - - - - .------n.,, - , - ~ . , - - . _ _ . _ - - - -- - - - , - , - - - - , _ , - _ . -,--,,n.,w -
U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION LIMITED TO FUEL HANDLING FACILITY: _EIz_SIz_YB81N___________
REACTOR TYPE: _HIGB:08_________________
DATE ADMINISTERED:_QhlR142Q________________
EXAMINER: _EELLEIt_Jz______________
APPLICANT: _________________________
- -lN118UGIl0NE_IQ_8EELIQ8 nil 1
~
Uso separate paper for the answers. Write answers on one side only.
Sicple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grede requires ,it least 70% in each category and a final grade of at lotst 80%. Examination papers will be picked up six (6) hours after tho examination starts.
% OF CATEGORY % OF APPLICANT'S CATEGO3Y 4
__V8LUE_ _IQI8L ___fGQBE___ _V8LUE__ ______________Q8IEQQBI_____________
_12100__ _10100 ___________ ________ M. REACTOR AND FUEL CHARACTERISTICS
_2DADD__ _2D200 ___________ ________ N. EQUIPMENT, INSTRUMENTATION AND
- DESIGN DESCRIPTION i
_ 2ZiDD__ _2Z200 ___________ ________ 0. PROCEDURES AND LIMITATIONS i _1ZiDD__ _1ZADD ___________ ________ P. EMERGENCY SYSTEMS AND SAFETY DEVICES
_ 1ZiDD__ _1ZiDD ___________ ________ Q. HEALTH PHYSICS AND RADIATION PROTECTION i
190100__ 100100 ___________ ________ TOTALS FINAL GRADE _________________%
All work done on this examination is my own. I have neither l givon nor received aid.
i l
APPLICANT'S SIGNATURE l
l I
Ut__BE8CIDB_8NQ_EWEL_GH888GIEBISIIG1 PAGE 2 QUESTION M.01 (3.00) 00 scribe three (3) ways that Boron is used as a control poison at Fort St. Vrain. (3.0)
QUESTION M.02 (3.00)
Explain the three (3) parameters or occurrences that can affect Shutdown
- Mergin.(SOM) after shutdown for refueling. (3.0)
- QUESTION M.03 (2.00)
Explain how an adverse reaction in irradiated f uel stored in fuel storage wolls is prevented. (2.0) i
- QUESTION M.04 (3.50)
- e. Explain how and why removing spent fuel may either increase or decrease core reactivity. (1.5)
- b. Explain why removal and installation of reflectors affects core reactivity. (1.5)
- c. How does removing a depleted neutron source affect reactivity? (0.5) l
-QUESTION M.05 (2.50)
( m.- Explain whether the control rods are more effective ut " Refueling" or
" Operating" conditions. (1.25) i b. What accounts for the change in total control bank worth from Beginning of Cycle to equilibrium core? (1.25)
I
(***** CATEGORY M CONTINUE 0 ON NEXT PAGE *****)
Mz__BEeGIQB_6NQ_EMEL_GH888CIEBISIIGA PAGE 3 QUESTION M.06 (2.00) -
Why is the reflector material loaded with boron? (2.0)
QUESTION M.07 (3.00) 00 scribe and explain the benefits of AXIAL and RADIAL fuel zoning for the roactor core. (3.0)
I l
l (***** END OF CATEGORY M *****)
Ni__EQUIEUENIi_INSIBUUENI8IIQN_8NQ_QESIGN_DESCBIEIIQN PAGE 4 QUESTION N.01 (3.00)
- c. Describe the seals on the top and bottom of a Reactor Isolation Valve (RIV). (1.0)
- b. Beside sealing, what are 2 functions served by the RIV? (1.0)
- c. Describe four (4) RIV internal interlocks. (1.0)
QUESTION N.02 (3.00)
- c. What is the purpose of the refueling sleeve? (1.0)
- b. What piece of equipment is used to install the sleeve? (0,5)
- c. How is any offset between a top reflector element and the Fuel Handling Machine grapple mechanism automatically eliminated before picking up the reflector? (1.5)
QUESTION N.03 (2.50)
- c. Describe the capacity of a fuel storage well for fuel elements and reflectors. (1.0)
- b. What are the three (3) cooling supplies for the fuel storage wells? (0.75)
- c. What are the two (2) backup cooling supplies for the fuel storage wells? (0.75) l l
(***** CATEGORY N CONTINUED ON NEXT PAGE *****)
Nz__E991EMENIi_INSIBudENI8IIDN_8HD_ DESIGN _DESGBIEI19N PAGE 5 QUESTION N.04 (3.00)
- c. Describe the method used to remove contaminated gases if both vacuum pumps of the Fuel Handling Purge System should fail. Describe system design features that allow this method to be utilized. (1.0)
- b. Describe two (2) unique design protective features for the Fuel Handling Purge System Vacuum Pumps. (1.0)
- c. What generally limits the vacuum that can be attained on a system within a short period of time? (0.5)
! d. Why is it undesirable to remove large volumes of moisture laden air with the vacuum pumps? (0.5)
QUESTION N.05 (3.00)
- e. What instrumentation is available to aid in the detection of a leaking tube or a loss of cooling to a fuel storage well. (1.0)
- b. How are the fuel storage facility internals cleaned? (1.0)
- c. What would happen to the fluid from a leaky tube in the annular interspace between a fuel well and the outer containment tank? (1.0)
QUESTION N.06 (3.00) r
- a. What maj or items are normally decontaminated in the Hot Service Facility? (1.5)
- b. Describe the treatment and disposition of solutions collected at the bottom of the Hot Service Facility or Storage Wells (1.5)
QUESTION N.07 (2.50)
Dcscribe fully the NORMAL and ABNORMAL flow paths of fuel deck generated ges waste. (2.5) l
(***** END OF CATEGORY N *****)
02__EB9CEDUBEl_8NQ_LIMII8IIQN3 PAGE 6 QUESTION 0.01 (2.00)
- o. What action (s) should personnel in the Reactor Building take if Reactor Building activity monitors indicate that an abnormal radioactive gas release is taking place? (1.0)
- b. How would personnel working in the Reactor Building be informed that an abnormal release was taking place? (1.0)
QUESTION 0.02 (2.00)
- o. During Shutdown Margin (SDM) determination, what is the criteria used to determine that a control rod is fully inserted? (1.0)
- b. When would the Refueling Supervisor normally inform the control room to perform a Shutdown Margin? (1.0)
QUESTION 0.03 (2.00)
- c. Other than control rod withdrawal, what are two refueling evolutions that can cause erratic oscillation or increase on the Start-Up Nuclear instrument channels? (1.0)
- b. What are the limits on control rod withdrawal for CRDOA removal?(1.0)
QUESTION 0.04 (2.00)
What are four (4) of the steps in the Fuel Handling Machine abbreviated Startup Procedure? (2.0)
QUESTION 0.05 (2.00)
I When (2 conditions) should a complete startup of the Fuel Handling Machine bo performed? (2.0) l t
l (***** CATEGORY 0 CONTINUED ON NEXT PAGE *****)
92__EB9CEDUBES_6NQ_LIMII8IIONS PAGE 7 QUESTION 0.06 (2.00)
What two (2) precautions should be observed during operation of the Fuel Hcndling Machine handwheels? (2.0)
QUESTION 0.07 (3.00)
What actions must be taken if the conditions for Technical Specification LCO 4.5.1, " Reactor Vessel Internal Maintenance," cannot be met? (3.0)
QUESTION 0.08 (3.00)
What are the precautions for refueling as given in the FHP Manual? (3.0)
, QUESTION 0.09 (2.00)
What four (4) departments are involved in new fuel receipt? (2.0) i QUESTION 0.10 (4.00)
Ocscribe the process used reflector blocks undergo as preparation for shipment of f site. Include in your answer where the blocks are stored onsite and the path and equipn.ent used to prepare them for shipment? (4.0)
QUESTION 0.11 (3.00)
- c. What department is responsible for coordinating spent fuel shipping activities? (0.5) i b. How many fuel elements are stored in a fuel shipping cask? (0.5)
- c. Fill in the blanks concerning LCO 4.7.4, " Spent Fuel Shipping
- Ccntainer." Blanks may contain one or more words or phrases. (2.0)
! _(1)_ (0.5) of spent ruel prior to _(2)_ (0.5) of decay time is allowable provided that _C3)_ (1.0).
(***** END OF CATEGORY 0 *****)
22__EDEBGENGI_11SIEMS_8NQ_18EEII_ DEVICES PAGE 8 QUESTION P.01 (3.00)
Por Technical Specifications, what three (3) conditions must be met in order to perform any reactor vessel internal maintenance with irradiated fuel in the vessel which requires removal of both the primary and secondary closures? (3.0)
QUESTION P.02 (2.00)
What are the four (4) emergency conditions or classifications in order of INCREASING severity, per the Fort St. Vrain Event and Emergency Classif ication Overview procedure?
, QUESTION P.03 (1.50)
Ecch of the definitions below, from Technical Specifications, is WRONG.
State the correct definition, or describe what must be changed to make the given definition correct. (1.5)
- e. The reactor is considered shut down f or refueling purposes when the reactor mode switch is placed in the " Fuel Loading" position simultaneous with either hot shutdown or the cold shutdown reactivity conditions. (0.5)
- b. Irradiated fuel is fuel that has been installed in the reactor during power operation. (0.5)
- c. A channel check is a test of the functional capability of the channel to determine that it is operable. (0.5) i
{ QUESTION P.04 (2.00)
Por Technical Specification LCO 4.5.1, " Reactor Building, Limiting Conditions for Operation," what two conditions must be met in order to hcndle irradiated fuel within the Reactor Building? (2.0) t 1
i
(***** CATEGORY P CONTINUED ON NEXT PAGE *****)
- . _ . . ~ . _ _ _ _ _ _ . . , _ _ _ _ _ _ _ _ . _ . _ , _ . . , . . _ _ _ _ _ _ . . _ . _ _ , _ . _
Ez__EMEBGENGl_S12IEM2_6NQ_S8EEII_ DEVICES PAGE 9 QUESTION P.05 (2.50)
- o. What process parameter is monitored, at what setpcini, to assure that irradiated fuel elements being handled with the Fuel Handling Machine are maintained less than 750 F? (1.0)
- b. What is the potential problem associated with excessive fuel element temperatures? (0.75)
- c. What non-normal condition, besides temperature, must be present for a problem to occur? (0.75)
QUESTION P.06 (1.00)
How is adequate cooling maintained to a fuel storage well that has suffered o complete loss of both cooling coils? (1.0)
QUESTION P.07 (2.50)
- b. How would such a scram affect fuel handling operations? (0.5)
- c. If you were involved in operations from the refueling floor, what would make you aware of the scram? C0.5)
QUESTION P.08 (2.50)
- a. What is the purpose of the grapple interlock? (0.5)
- b. What conditions and logic are measured or detected to satisfy the
! interlock? (1.25)
I
! c. What condition (s) cannot be bypassed to satisfy the interlock? (0.75)
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(***** END OF CATEGORY P *****)
}
92__HE6LIH_EBX11C1_880_B6018IION_EBQIECIION PAGE 10 QUESTION Q.01 (4.00)
Ccaplete the table below dealing with external exposure limits.
- c. Federal whole body limits per 10CFR20:
- 1. _____ rem /qtr without exposure history. (0.5)
- 2. _____ rem /qtr with exposure history. (0.5)
- 3. _____ rem accumulated lifetime. (0.5)
- 4. _____ % of above for minors (under 18 years of age). (0.5)
- o. FSV Whole Body Guideline: _____ rem / week. (0.5)
QUESTION Q.02 (2.00)
Fill.in the blanks below. Blanks may contain one or more words or numbers,
- m. A Contaminated Area is any area in which there exists loose surface contamination in excess of _(1)_ and/or _(2)_. (1.0)
- b. A Radiation Area is an accessible area where a major portion of the whole body could receive greater than _(1)_ or _(2)_. (1.0)
QUESTION Q.03 (3.00)
Answer the following questions dealing with Radiation Work Permits (RWP's).
- o. What does your signature, as a worker, signify on an RWP? (1.0) i
- b. What action should you take if you are performing work covered by an i RWP and the job is not completed at the end of your shift, but will be resumed on your shift the next day? (1.0)
- c. What type of radiation instrument is used exclusively f or setting dose rates for personnel at FSV? C0.5)
- d. When entering a radiation area with a portable survey instrument, which range on the detector should be selected prior to entry? (0.5) i
(***** CATEGORY Q CONTINUE 0 ON NEXT PAGE *****)
- -,--.__, ..-..-_.,,,., . ,__ .,-. --,___,,.--,,_,,________.__..._..,._m.__... , . , ._,_._%.._,m.-_..._. . , . , , . _ .
o t .
92__HEeLIH_EHISIGS_8ND_B8016IIQN_EBQIEGIIQN PAGE 11
~
$% ' er QUESTION Q.04 (1.00)
State a formula which describes the relationship between Radiation *bsorbed Dese, Roentgen Equivalent Man, and Quality Factor. (1.0) x QUESTION Q.05 '(3.00)
What are three (3) of the four checks which must be performed before you ccn use an RO,- 2 to perform a radiation survey? (3.0)
QUESTION Q.06 (1.00)
When moving an irradiated control rod drive assembly in the ATC, where chould you, the operator, stand in relation to the ATC to maintain your oxponure ALARA? WHY? (1.0)
< QUESTION Q.07 (3.00)
MATCHING. Match each radiation type with ALL associated characteristics.
Ecch character istic may apply to one or more types of radiation. (3.0)
a. Alpha 1. Most penetrating radiation.
___ b. Beta 2. Particle.
___ c. Gamma 3. Wave.
___ d. Neutron 4. Primary hazard to eyes and skin.
- 5. Least penetrating.
- 6. Quality factor of 1.
- 7. Lead is a good shield.
- 8. Water is a good shield.
- 9. Smallest particulate radiation.
- 10. Electrically charged.
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I R************* END OF EXAMINATION ***************)
Uz__BE6CI98_eND_EWEL_GH888CIEBISIIGS , PAGE 12 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER M.01 (3.00)
- c. As main control poison in the form of control rods. (1.01
- b. As a reserve shutdown system
- independent of the control rod drive system. ,
(1.0)
- c. As a burnable poison placed in the fuel elements to reduce the amount of control rods (0.5) and control excess reactivity at the beginning of life (0.5) (1.0)
REFERENCE FSV LESSON NO. FH 038.00.01.07.1 ANSWER M.02 (3.00)
- e. Decay of Xenon isotopes decreases the SOM with time. (1.0)
- c. Due to Negative Temperature Coefficient, the lower core temperature, the more reactive the core. (1.0)
COR Whether at hot (220 F) or cold (80 F) shutdown.)
REFERENCE FSV LESSON NO. FH 038.00.01.16.1 ANSWER M.03 (2.00)
- e. The 02 content of cover gas is kept below 5% to eliminate any chance of combustion (He atmosphere is maintained) . (1.0)
- b. The surface temp. of spent fuel is kept below 750 F to prevent oxidation. (1.0)
REFERENCE FSV SD 14-1, P. 3
'4 r di__BEeGI9B_eND_EMEL_GHeBeGIEBISIIGS PAGE 13 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER M.04 (3.50)
- c. 1. Decreases reactivity by removing fissionable materials. (0.75)
- 2. Increases reactivity by removing f ission product poisons (including Xenon and Samarium). (0.75)
- b. 1. Removing reflector decreases reactivity by allowing increased neutron leakage. (0.75)
! 2. Installing reflectors increases reactivity by returning more neutrons to the core. (0.75)
- c. Decreases reactivity. (0.5)
REFERENCE FSV LESSON NO. 39.00, P. 3 i
i ANSWER M.05 (2.50)
- c. The rods have a higher relative worth hot than cold (0.5) because the ubsorption cross section of the fuel goes down more rapidly with temperature than rod worth. (0.75) (1.25) l l '. .b' . The Rod worth at equilibrium is slightly less (0.75) due to depletion of burnable poison. (0.5) (1.25)
REFERENCE
, FH 033.00, p. 6 ANSWER M.06 (2.00)
Minimizes thermal neutron heating (1.0) of core core barrel and vessel components (1.0).
REFERENCE FH 038.00, p. 7 s
, - . , . . - - - - . - , .n ., - - - - - - . - , ~ . -
Uz__BE8CISB_6HD_E9EL_GH888CISBISIIG2 PAGE 14 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER M.07 (3.00)
AXIAL: There are 2 axial zones with 55% of the fuel in the upper half of the core and 45% in the lower half C0.75). This takes advantage of the cooler helium temperature in the upper half of the core (0.75).
RADIAL: There are 5 radial zones in concentric circles (0.375). Th-232 loading is higher & U loading lower in the outer zone (0.375).
Radial zoning helps flatten the flux profile (0.375) and the changes in Th/U loading helps minimize peaking at the outer edges of the core (0.375).
REFERENCE FH 033.00, p. 11, 12
Nt__EQUIEMENIi_INSIBUDENI8IIQN eND_ DESIGN _QESGBIEIIQN PAGE 15 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER N.01 (3.00)
BOTTOM - An inflatable seal on bottom of valve seal the RIV to the refueling penetration or fuel storage well. (0.5)
- b. Provides shielding (0.5) and serves as a platform for either the FHM or the ATC. (0.5) (1.0)
- c. 1. Differential pressure - Gate will not open unless dp above and below gate is < 1 psid.
- 2. Differential atmosphere - Gate will not open unless atmospheres are identical.
- 3. Seals inflate on loss of power.
- 4. Seal must be deflated to open gate.
- 5. Alarm if gate is opened.
- 6. Alarm if seal pressure < 20 psig. (Any 4/6 9 0.25 ea.)
REFERENCE FSV, FH.O.49.00.01.05.5 ANSWER N.02 (3.00)
- e. Provide accurate guidance and alignment for proper functioning of the FHM while in the PCRV penetration. (1.0)
- b. Auxiliary Transfer Cask. (0.5)
- c. When the tapered end of the pickup probe enters the element, it displaces a horizontal translator mechanism. (0.75) This offset data inputs to the control center which auto positions the head until the error signal is zero. (0.75) (1.5)
REFERENCE SD-13-1, p.7,8
Ni__EQUIEMENIt_INSIB9dENIaIIQN_8NQ_QE110N_DESGBIEIIQU PAGE 16 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER N.03 (2.50)
- e. 1. Each well can store 14 elements in each column for a total of 56. (0.5)
- 2. One column can hold 14 full length elements or 28 half length elements. (0.5)
- b. 1. Helium purge.
- 2. Cooling water from R.P. Cooling Water (Sys 46 aux loops).
- 3. Reactor Bldg. HVAC. (0.25 ea.) (0.75)
- c. 1. Fire water.
- 2. Booster fan and dampers. (0.375ea.) (0.75)
REFERENCE FSV FH 041.00.01.03.1, 2 FH 041.00.01.05.1, 2 ANSWER N.04 (3.00)
- c. In the unlikely event that both pumps fail, a pump bypass line (0.5) permits the equipment and storage wells to be purged with clean helium and discharged directly to the gas waste system. (0.5) (1.0)
- b. 1. A pressure controlled bypass auto opens if discharge pressure reaches 3 psig to prevent overload. (0.5)
- 2. A temperature switch causes the pump suction valve to shut if pump oil reservoir temp. should reach (170 deg's) which will prevent pump damage. (0.5)
- c. The saturation pressurization of any moisture in the system. (0.5)
- d. High water removal rates result in water emulsif ication with the pump oil and subsequent pump downtime. (0.5)
REFERENCE l FSV SD-13-2, p.2,4,5 l
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Ni__EQUIEUENIt_INSIBWHENI8IIQN_8NQ_QE210N_DEIGBIEIIQN PAGE 17 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER N.05 (3.00)
- a. Well thermocouples, moisture monitor, and flow switches. (1.0)
- b. They are back purged with purified helium. (1.0) 4
- c. Individual drain lines from each well are combined and routed to the Reactor Plant Drain system. (1.0) 1 REFERENCE FSV SD-14-2, p. 3,4 ANSWER N.06 (3.00)
- e. 1. Control rod drives Refueling Sleeve
- 2. Fuel shipping cask RIV He Circ (Any 3 8 0.5 ea. - 1.5)
- b. Solutions are pumped by a recycle pump through a filter for removal of particulates. (0.51) Fluids are then stored for reuse, (0.33) pumped to liquid waste (0.33), or put in drums (0.33). (1.5)
REFERENCE FH 043.00, p. 9 l
l ANSWER N.07 (2.50)
NORMAL: Enters the inlet header through a pre-filter (0.25) and then l passes through a normal system filter. (0.25) If gas is within limits it i is routed through a disch. f1ter (0.25) via waste gas blowers (0.25).(1.25)
! ABNORMAL: Highly radioactive gas is piped directly to the waste vacuum tank. (0.25) From here it is pumped to a surge tank (0.25) by the waste l gas compressor. (0.25) It is released through exhaust filters when within limits. (0.25) (1.25)
REFERENCE FSV FH045.00, P.6
92__EBQCEQMBES_8NQ_LIMIIBIIQNS PAGE 18 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER 0.01 (2.00)
- c. Evacuate the RB. (1.0)
- b. Sounding of the Radiological Alarm. (1.0)
REFERENCE FSV EP APP H-2, Rev. 51, p. 3 ANSWER 0.02 (2.00)
- a. " Rod In Light in on with the " Slack Cable light is off. (1.0)
- b. Normally, prior to refueling a region when two control rod pairs are retracted. (1.0)
REFERENCE CMG-13, p. 4, 5 ANSWER 0.03 (2.00)
- m. Removing reflector blocks (0.5).or installing sources. (0,5) (1.0)
- b. 1. Remain < 1.5 dpm on both Start-up channels
- 2. S times CRo(NI) from Data sheet 1
- 3. No more than 20" increments. (0.33 ea.) (1.0)
REFERENCE FSV CMG-13C, p. 1 & CMG-13, p. 2 ANSWER 0.04 (2.00)
- 1. Turn on the Master Power (Key) Switch. (Any 4/6 4 0.5 ea.)
- 2. Wait ~1 minute for intial warmup.
- 3. Select machine location from the master control panel.
- 4. Put the grapple into " Receive" condition (using bypass switch).
- 5. Check that servo null indicators are on (if not then null).
- 6. Select either " Manual Mode" or start " Automatic Mode" selection proc.
92__EBQGEDUBES_8HD_LIMII8IIQNS PAGE 19 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
REFERENCE FH 051.00.01.03.1 ANSWER 0.05 (2.00)
- 1. After an extended shutdown (1.0)
- 2. After any system modification / calibration / repair. (1.0)
REFERENCE FH 051.00.01.04.1 ANSWER 0.06 (2.00)
- 1. Don't move the handwheels too quickly (a servo error may result, causing the system to unclutch). (1.0)
- 2. Don't operate the handwheels when the machine is in the OFF mode. (1.0)
REFERENCE FH 051.00.01.12.1 ANSWER 0.07 (3.00)
- 1. Suspend any internal maintenance Cincluding fuel handling). (1.0)
- 2. Retract any remotely operated mechanisms. (1.0)
- 3. Close the RIV a FHM Cask Valve (ASAP). (1.0)
REFERENCE FH 051.00.04.02.1
02__EBQQEDUBES_eND_LIMII6IIQNS PAGE 20 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
l l
ANSWER 0.08 (3.00) l l
- 1. Maintain 2-way comm. between FHM CR & Ops CR during refueling. (0.75) :
- 2. FHM CR & Ops CR coord. rod move. to ensure adequate SDM C.01 dp).(0.75)
- 3. Comply with TS (4.7.1, 4.7.2, 4.7.3). (0.75)
- 4. FHM & ATC remain attached to the overhead crane when located over the HSF Cover the FSW, LP, or Reactor, the overhead crane may be detached if the FHM or ATC is bolted to an RIV). (0.75)
REFERENCE FH 051.00.04.05.1 ANSWER 0.09 (2.00)
- 1. Maintenance. (0.5)
- 2. Health Physics. (0.5)
- 3. Quality Assurance. (0.5)
- 4. Technical Services. (0,5) i REFERENCE l
FH 053.00.01.01.
l ANSWER 0.10 (4.00) l
- 1. Blocks are picked up from the FSW by the FHM. (1.0)
- 2. The FHM is moved to the RIV installed on the HSF. (1.0)
- 3. Each block (or 2- 1/2 size) is put in a 55 gal. drum and capped. (0.5) l i 4. HSF Manipulator & hoist loads capped drum on a pallet (0.5) i l S. Pallet is xferred from the HSF to the truck bay via the RB crane (0.5)
- 6. When 2 pallet loads are in the cask, it is sealed & ready. (0.5) l 1
Qi__EBQGEQUBE3_8NQ_LIMIIeIIQN3 PAGE 21 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
REFERENCE FH 057.00, p. 3, 4 ANSWER 0.11 (3.00)
- c. Mechanical Maintenance. (0.5)
- b. 6. (0.5)
- c. 1. Loading AND/OR Shipping. (0.5)
- 2. 100 days. (0.5)
- 3. less than 200 mr/hr on surface of shipping cask. (1.0)
REFERENCE FH 056.00.01.01.1, FH 056.00.01.02.1, FH 056.00.01.03.1 4
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E2__EUEBGENGl_11SIEd2_eND_28EEI1_QEYICES PAGE 22 ANSWERS -- FT. S T, . VRAIN -86/05/20-PELLET, J.
ANSWER P.01 (3.00)
- 1. The reactor vessel is depressurized to at or below atmospheric pressure. (1.0)
- 2. The reactor average helium gas inlet temperature is </= 165 F. (1.0)
- 3. The reactor is maintained in a shutdown or refueling condition AND the reactivity of the core is monitored continuously by at least 2 neutron flux monitors. (1.0)
REFERENCE FSV TS 4.5.2, p. 4.5-3 ANSWER P.02 (2.00)
Nstification of Unusual Event, Alert, Site Area Emergency, & General Emer.
(4 answers e 0.4 es.; correct order 0.4)
REFERENCE FSV EP CLASS, p. 11-14 ANSWER P.03 (1.50)
- c. Correct except "placed," should be " locked." (0.5)
- b. ...has a radiation level >/= 100 mr/hr at 1 foot from element. (0.5)
- c. Either change " check," to " test," or . . . is a qualitative determination that the channel is operable. (0.5)
REFERENCE FSV TS, Section 2 ANSWER P.04 (2.00)
- 1. Reactor Building integrity is maintained. (1.0)
- 2. 2 of the 3 reactor building exhaust fans are operable. (1.0) i i
t
Ez__EMEEGENGl_SYSIEMS_8NQ_18EEII_QEVIGES PAGE 23 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
REFERENCE FSV TS 4.5.1, p. 4.5-1 ANSWER P.05 (2.50)
- e. FHM cooling water outlet temperature (0.5) </= 150 F (0.5).
- b. Graphite oxidation. (0.75)
- c. Air inleakage must also have occurred. (0.75)
REFERENCE FSV TS 4.7.2, p. 4.7-3 ANSWER P.06 (1.00)
Vontilation flow is increased until either cooling can be restored. (1.0)
REFERENCE FSV TS 4.7.1-3, p. 4.7-2, 3, 4 ANSWER P.07 (2.50)
- a. Startup count rate high, 1 out of 2 logic, >/= 1.0E+5 C3 answers e 0.5)
- b. Cease all internal maintenance / fuel handling until safe. [0.5)
- c. Operator would advise OR Hazard alarm on FHM console. (0,5)
REFERENCE FSV EP APP B-1, p. 2, 11
..-m- . - , . - - - - . - , . - , ,
-Ez__EMEBGENC1_112IEMS_8NQ_S8EEII_ DEVICES PAGE 24 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER P.08 (2.50)
- e. Protects against dropping of a grappled element. (0.5)
- b. All of the following conditions must be met: (0.5)
- 1. At least 1 dowel pin engaged. (0.25)
- 2. The plate is raised. (0.25)
- 3. Probe weight < 50 lbs. (0.25) (1.25)
- c. The probe weight condition cannot be bypassed. (0.75)
REFERENCE FSV LESSON NO. FH 051.00, p. 66,67 l
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91__HE6LIH_EHYSICS_8NQ_B8DI8IIQN_EBQIECIIQN PAGE 25 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER Q.01 (4,00)
- e. 1. 1.25
- 2. 3.0
- 3. 5(n-18)
- 4. 10
- b. 7.5
- c. 18.75
- d. 0.125
- o. 0.100 (8 answers e 0.5 ea.)
REFERENCE FH 058.00, p. 12, 13; 10CFR20 ANSWER Q.02 (2.00)
- e. 1. 1000 dpm/100 sq cm beta-gamma. (0.5)
- 2. 10 dpm/100 sq cm alpha. (0.5)
- b. 1. 2.5 mrem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. (0.5)
- 2. 100 mrem in 5 consecutive days. (0.5)
REFERENCE FH 060.00, p. 3, 4 ANSWER Q.03 (3.00)
- a. That you understand the job requirements (CONCEPT). (1.0)
- b. Return white copy to SS. (1.0) i c. Ion chamber. (0.5)
- d. Start on the highest range (a range down ...). (0.5) l REFERENCE FH 058.00, p. 14; FH 059.00, p. 6, 11 1
1
.=-
91__Hg8(IH_EHIgIgg_8NQ_B8QI8IIQN_EBQIEGIIQN PAGE 26 ANSWERS -- FT. ST. VRAIN -86/05/20-PELLET, J.
ANSWER Q.04 (1.00)
Red'* QF = REM (1.0)
REFERENCE FH 058.00.01.04.1 ANSWER Q.05 (3.00)
- 1. Inspect f or phys ical damage.
- 2. Check for current calibration.
- 3. Perform battery check.
- 4. Verify response to a known source. Cany 3/4 e 1.0 ea.)
REFERENCE FH 059.00.0.03.1 ANSWER Q.06 (1.00)
Stand near the ATC - its shielding design generally produces lower dose rates there. (1.0)
REFERENCE FH 060.00.01.01.1 i
ANSWER Q.07 (3.00)
- e. 2, 5, 7, 8, 10 (5 answers e 0.15 ea.)
- b. 2, 4, 6, 7, 8, 9, 10 (7 answers e 0.107 ea.)
- c. 1, 3, 6, 7, 8 (5 answers e 0.15 ea.)
- d. 2, 8 (2 answers e 0.375 ea.)
i REFERENCE FH 058.00.01.01.1 w,- - - - - - - . - - , - - - , ~ -
, -n.--- - - - - - - , , , e, - - rwr-,--- - - - -m-,- . - - - - - , , ,---s------. -,--n.~- , + - - - - - -
, . . n azms o F~TiiMM!My }
o8 1
. -;i A i 0 Pubh.c Service
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Company of Colorado 16805 WCR 19 1/2, Platteville, Colorado 80651 May 23, 1986 Fort St. Vrafn Unit No. 1 P-86385 Mr. Ralph Cooley, Section Chief Operating Licensing Group U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 Docket No. 50-267
SUBJECT:
NRC Administered Examinations
Dear Mr. Cooley:
The purpose of this letter is to provide comments to the NRC on the Licensed Requalification and Special Senior Licensed Operator (Fuel Handler) examinations administered at Fort St. Vrain the week of May 19, 1986.
In general, the examinations were fair and applicable to the Fort St.
Vrain facility. However, we have a few concerns which are listed in attachments 1 and 2.
Any comments o r- questions concerning this tratter should be made directly to Steve R. Willford at Fort St. Vrain, (303) 571-7436, extension 450.
Sincerely,
[f.qlt?!ltbh J. W. Gahm Manager, Nuclear Production Fort St. Vrain Nuclear Generating Station JWG/sl Attachments
ATTACMMENT 1 REVIEW OF REACTOR OPERATJR REQUALIFICATION EXAMINATION 1.01: The question is a valid question; however, the question asks only to identify the two characteristics and not how or why they vary. The question as written is over valued.
1.02: The question is valid. The method used to plot the curve is inadequate. On the diagram: (a) should be T = -20 seconds, (b) is adequate, (c) should be T = -80 seconds.
~
1.03: The question is valid. Part b. of the answer key should also accept: This is result of the changes that occur in the cross-sectional area of absorption for Xe-135 as fuel temperature changes. See the following graph.
se.
~"
.T 1.04: Part 3 should also accept " decay".
1.05: This question is invalid. Protactinium 233 is a product of neutron absorption by Thorium 232 and is not produced by the fission process. However, we feel that the correct answer will be given by the trainees simply by the process of elimination of possible answers. We suggest that this question be rewritten prior to reuse in the future.
l
1.06: Calculation of the enthalpy of feedwater for HTGR conditions is not possible using: h(sat) - 1 BTU /LBM/F Subcooling because HTGR is at 3350- psi and is supercritical. Exact enthalpy determination would require compressed water tables.
The best that can be done using the C-E Steam Tables is to use the enthalpy of saturated water at 403 F, which is 379.4 BTU /LBM. The actual feedwater enthalpy is 382.4. This is well within the tolerances of the problem. The problem would then be:
db = h(out) - h(in)
T out ='1000 F P
out = 2400 psi h
out = 1461 BTU /LBM T
in = 400 F P
in = 335 psi h = 379.4 BTU /LEM Ah = 1461 - 379.4 = 1081.6 BTV/LBM 1.07: Acceptable 1.08: The answer for c. is 12.3 psia. The question part e. is not valid as no tasks have been identified for operations that uses sea level atmospheric pressures. The question f. is not a valid question as there appears to be more than one set of brackets for the response.
2.01: Acceptable 2.02: The question is valid. The answers 4, 5 and 6 utilize a common line on the cold reheat header and as such just cold reheat would be acceptable. The answer should.also accept Auxiliary Boiler Feed Pump.
2.03: Acceptable 2.04: The answer should accept " Determine leaking loop dumped by increasing radiation indication and/or greater than 250 psig in dump tank per Emergency Procedure A."
2.05: The correct answer for section (a) is "When circulator speed decreases below 700 RPM". Answers for section (b) and (c) are acceptable.
2.06: The answer should also accept " Malfunction in the Slack Cable Circuitry" or words to this effect.
2.07: The answer for section (a) should also accept " System 23."
The answer for section (b) should also accept " System 24."
2.08: Acceptable.
3.01: Answer 4 should also accept 2/3 low level moisture monitors trip, no circulator seal malfunction, no hot reheat activity high, and one of two high level moisture monitors trip.
3.02: The question is poorly worded. The question should read "On a steam leak into the primary coolant system, what automatic actions occur in addition to a loop shutdown?" The answer should also require a Reactor Scram on Two Loop Trouble.
3.03: -Acceptable 3.04: Acceptable 3.05: The question for section (a) is poorly worded. The question should read "What parameter is used for steam / water dump system actuation if the moisture monitor system fails to detect high moisture in the primary coolant system?"
3.05 The question for section (b) is poorly worded. The question should read "What actions occur if the moisture monitor system fails to detect moisture in the primary coolant system?" The answer should be "The preselected loop shutdown, steam water dump and other loop is depressurized to approximately 850 psig. (850 should not be required as it is a controller setpoint that is subject to change.)
3.06: The question section (a) should include the name of HC-21417
" Loop 1 Backup Bearing Water Controller". Question section b is acceptable.
3.07: The answer a.5 for this question should also accept "1.55 gain on operating circulator."
3.08: Acceptable 4.01: Acceptable 4.02: Acceptable 4.03: Acceptable 4.04: Emergency Procedure G should be referenced on answer d.
4.05: Acceptable 4.06: Acceptable 4.07: Acceptable 4.08: Acceptable 4.09: Acceptable
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REVIEW 0F SENIOR REACTOR OPERATOR REQUALIFICATION EXAMINATION 5.01: The question is a valid question but the question asks only to identify the two characteristics and not how or why they vary. The question as written is over valued.
5.02: Same comment as 1.04 5.03: Same comment as 1.05 5.04: Same comment as 1.06 5.05: Same comment as 1.08 5.06: Acceptable 5.07: The answer for 1. and 2. could not be found in the specified reference (FSV Nuclear Energy Training Operations, 4.6-5).
We need a clarification as to the applicability of this question to the Fort St. Vrain facility. Section (b) is acceptable.
5.08: For the answers given, the question should be reworded as follows: "What will happen to work done by a motor driven feed pump working against a constant differential pressure if the'following changes occur?" The words " constant load" could lead the licensee to say that the work done is a constant.
6.01: Acceptable 6.02: Same comment as 2.02 6.03: Same comment as 2.04 6.04: Same comment as 2.06 6.05: Question (a) as written is valid but the question asks for one answer and the answer key specifies three answers. The answer snocid be adjusted accordingly. The total point values for question (b) for the individual responses does not add up to the total point value specified for the question.
Answer 4 for Question (c) is not a valid answer. This alarm is a result of a trip and is not a trip in itself.
6.06: As the question is written, the shock absorber in the answer ,
should not be required.
6.07: Answer (b) should also accept 1st stage pressure and throttle pressure. Answer (f) should be reactor power. In addition Turbine / Gen. Load (FSP) should be written out as "First Stage Pressure".
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6.08: The answer is correct by the specified reference. Plant conditions at the present time utilize domestic water as the normal makeup. The answer should also accept domestic water as normal makeup and shallow well water as emergency makeup.
7.01: Same comment as 4.03 7.02: Acceptable 7.03: Acceptable 7.04: Acceptable 7.05: Acceptable 7.06: Acceptable 7.07: Acceptable 7.08: Acceptable 8.01: Acceptable 8.02: Acceptable 8.03: Acceptable 8.04: Acceptable 8.05: Acceptable -
8.06: Acceptable 8.07: Acceptable 8,08: Acceptable 8.09: Acceptable i
2
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. ATTACHMENT 2 REVIEW 0F SPECIAL SENIOR LICENSED OPERATOR (FUEL HANDLER) EXAMINATION NOTE: Comments apply to SSLO Requalification Exam also.
M.01 Acceptable M.02 From FH 038.00, page 13, a fourth answer is possible; whether it is a " hot" (220 F) or " cold" (80 F) shutdown.
M.03 Reference should be made to LCO 4.7.3. A helium atmosphere is maintained. However, monitoring or maintenance of fuel storage well conditions are not SSLO responsibility.
M.04 Acceptable M.05 Rod worths are not required knowledge for SSLO since SSLO knowledge is limited to reactor shutdown condition. The referenced material FH 038.00, page 7, 8 was provided as background information and was not part of objective FH 038.00.01.07.
M.06 Reference should be FH 033.00, page 6, not FH 038.00, page 7.
Per FH 033.00.01.05, SSLO's are only responsible for the location of the boronated reflector since only specific reflector is boronated.
Question should read: "Why is certain reflector material loaded with boron?" Again, the boronated reflector has little application to reactor shutdown condition.
M.07 Acceptable N.01 Acceptable N.02 Reference should be 50-13-1 not SD-31-1.
N.03 Section (b) question implies any three cooling supplies, but could state: ...three normal cooling supplies... Section (b),
answer #2 should be ' system 46 auxiliary loops.' Reactor plant cooling water system covers any cooling water used in the reactor building.
N.04 SSLO is not responsible for operation of the fuel handling purge system vacuum pumps. No FH objective identified which could cover these questions, since this was not trained as SSLO duty.
N.05 SSLO is not responsible for fuel storage well operation or l monitoring.
I N.06 No reference for answer. Should be FH 043.00, page 9.
Section (a), objective FH 043.00.01.02 requires 3 facilities - HSF, l ESW, FSW, not 3 major items as in answer. A multitude of items could l be deconned in HSF and all are major, such as refueling sleeve, FHM, i
ATC, RIV, SA, Helium Circulator, CRD, Fuel Shipping Cas, or even the HSF itself. Section (b), objective FH 043.00.01.03 only requires on-line diagram. SSLO not responsible for operation /flowpaths of decontamination system.
N.07 Acceptable 1
i
.. ,: : ~2' 0.01 Acceptable 0.02 Section (a), SSLO not responsible for performance, determination, or verification of SDM; only responsibility is for notification of necessity. Section (b) answer is "when required by Fuel Handling Procedure, prior to refueling a region."
0.03 Section (a) acceptable. Section (b), SSLO not responsible for these limits.
0.04 Acceptable 0.05 Acceptable '
O.06 Acceptable 0.07 Acceptable 0.08 Acceptable 0.09 Acceptable 0.10 Acceptable 0.11 Section (a), Technical Services Department has overall responsibility for coordinating spent fuel shipping activities. To obtain answer of " Mechanical Maintenance," question must state:
"... activities with other involved departments." Section (b),
appropriate word might be ' loaded' or ' shipped' instead of ' stored' since fuel shipping cask is'not used as a storage location. Section (c) is acceptable.
P.01 Acceptable P.02 Acceptable P.03 SSLO not responsible for referenced Technical Specifications (Section 2).
P.04 Acceptable P.05 Acceptable P.06 SSLO responsible for knowledge of FSW cooling methods, but not responsible for determination or performance of recovery actions for loss of cooling.
P.07 Section (a) acceptable. Section (b) acceptable. Section (c),
answer should be " Facility hazard alarm on FHM console. Operator (Reactor) . advisement is determined by Operations Department Procedures."
P.08 Acceptable Q.01 through Q.07 acceptable.