ML20235G697
| ML20235G697 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 06/25/1987 |
| From: | Johnston G, Morrill P, Obrien J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20235G659 | List: |
| References | |
| 50-528-OL-87-01, 50-528-OL-87-1, NUDOCS 8707140348 | |
| Download: ML20235G697 (47) | |
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i: Enclosure (1)- i EXAMINATION REPORT i Examination Report No.: 50-528/0L-87-01 Facility: . Arizona Nuclear Power Project Post Office Box 52034 Phoenix, Arizona. Facility License No.: NPF-41 Examinations administered at Palo Verde Nuclear Generating Stations 1,2, and 3 Wintersburg, Arizona Requalification Review 1987 i Chief Examiner: {hd b/2M((7 Date Signed PhilipQrirl Examiner: /// ck O'Errie
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Examiner: Dat'e St'gne( Gyy/Johngn Examiner: 7C m- - </uM7 ! Tom Meadows Date Signed Approved by: k O. h L Dale' Signed 57
- f. O. Elit, Section Chief Summary: ,
The review of the requalification program at Palo Verde Nuclear Generating Statior was performed in accordance with NllREG-1021. ES-601, " Administration of NRC Requalification Program Evaluations" and (for parallel graded examinations) the Pilot Test Program criteria defined in a memorandum from William T. Russel, j Director, Division of Human Factors Technology, NRR, dated May 22, 1986. Based on these criteria the requalification program at ANPP's Palo Verde Facility is evaluated as marginal. ] l l 8707140348 870626 DR ADOCK0500g8 . % . ' . i .
1 REPORT DETAILS
- 1. - Description of Requalification Evaluation The evaluation consisted of written and operating examinations involving twenty-five licensed operators.and senior operators (approximately twenty percentoftheoperatingstaff).
i Ten reactor operators and fourteen senior reactor operators were administered simulator and oral examinations by NRC personnel during the period January 28, 1 1987 through February 6, 1987. l Ten reactor operators were given an NRC prepared written examination on January 27, 1987. At the conclusion of this examination the examir.ers reviewed the answer key with the facility staff. The facility coments and i the NRC resolution to these comments are documented in Attachment (1) to this i report. The indicated changes were made to the examination key prior to grading.by the NRC. Six facility prepared senior reactor operator written requalification examinations (administered to fourteen senior reactor operators by facility personnel) were parallel graded by the NRC to complete the evaluation of the senior reactor operators. These facility tests were originally given during the period August 22, 1986 through September 25, 1986.
- 2. Program Evaluation The requalification program at the Palo Verde Nuclear Generating Station was evaluated based on the criteria of NUREG-1021, ES-601, Rev. 3, " Administration of NRC Requalification Program Evaluation" and (for parallel grading of .
examinations) the Pilot Test Program criteria defined in a memorandum from j William T. Russel, Director, Division of Human Factors Technology, NRR, .I dated May 22, 1986. j
- a. Review of Facility Written Examination Required for Satisfactory Rating:
"The facility written examination should follow the guidance contained i in ES-601 (Rev.2) with respect to exam format, content, and length." !
Six requalification written examinations were reviewed; NLR99C-011-86, i 013-86, 015-86, 017-86, 020-86, and 022-86. With the two exceptions described below, Region V found these examinations to be comprehensive and in general compliance with the Examiner Standards.
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t, (1) The Region V staff found that five of the six facility examinations had j less than 36% (average 29% HT and FF) of the theory section of the examination on heat transfer and fluid flow (HT and FF) topics. NUREG-1021 requires 40% +/-4% of the theory section to be on heat ; transfer and fluid flow. j (2) The overlap of questions used on subsequent examinations varied from 1S% up to 37% and averaged approximately 25%. There is no standard for j the allowable overlap of questions, however 25 - 37% repeat questions i from the proceeding examination appears considerably above industry _ l practice. The Region V policy for examinations written by Region V does not allow any repeat questions from the last prior examination given at a facility. Evaluation: Marginal
- b. Parallel grading of the Facility Graded Written Examination Required for Satisfactory Rating:
" Facility graded written examination results shall be within 10% per '
section of the results reached by NRC examiners independently grading the same examinations." A total of six PVNGS requalification written examinations administered by the facility staff to fourteen senior reactor operators were . independently graded , by NRC Region V examiners. In three cases the NRC graded sections were more l than 10% different than the grades determined by the facility staff. On the average the facility grades for each operator in each section were 3 to 4% higher than the NRC grades. The facility graded complete examinations were 1 ! to 8% higher than the NRC graded examinations. The Region V examiners observed the following deficiencies related to the facility examinations. (1) The facility examinations contained open-ended questions which were not consistently graded. . In some cases the answers were not graded in ' accordance with the answer keys. Some points were given for material not listed in the keys. (2) The facility examinations and answer keys contained questions which did not provide a clear assignment of points for each part to a several part answer. This was especially significant when the question was an " explain" or " discuss" type question. Evaluation: Marginal
- c. Pass / Fail Evaluation: i Required for Satisfactory Rating:
"More than 80% of the final pass / fail determinations made by both the i facility and the NRC must be in agreement." !
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. 8, The facility graded examinations passed all fourteen senior operators (100% pass rate), while the NRC grading resulted in ten of the fourteen (71% pass rate) senior operators passing the written examination. As described in paragraph 2.b above, the facility grading averaged 3 to 4%
higher than the NRC grading. This appeared to be due to occasional vague point assignments in the answer key and not consistently grading to the answer key. Evaluation: Marginal
- d. Candidates Ability to Satisfactorily Complete Requalification Required for Satisfactory Rating:
"More than 80% of the evaluated operators passed all portions of the examinations."
Of the twenty-five operators examined, one reactor operator and one senior reactor operator failed the operating examinations. All ten reactor operators passed the NRC prepared and administered written examination. Ten of fourteen senior reactor operators passed the NRC parallel graded facility prepared ! requalification examination. Consequently eighteen out of twenty-five personnel (72%) passed all portions of the examinations. Evaluation: Marginal
- 3. Program Strengths and Weaknesses The NRC administered oral, simulator, and written examinations indicate that the operators are generally knowledgeable and capable of safely operating the facility. The written requalification examinations which were reviewed and parallel graded indicate a need for additional management attention as well as improvement in the content, preparation and grading of facility administered requalification examinations. Specifically, the facility training staff needs to establish guidelines for examination administration and to improve in the following areas.
- a. Ensure that the theory part of the written requalification tests contain a sufficient number and depth of questions on heat transfer and fluid flow,
- b. Ensure that the overlap or reuse of the same questions on subsequent examinations does not become excessive. The questions reused (if any) should not allow individuals to pass the test by memorizing previous tests.
- c. Eliminate open ended type questions where there is no single correct answer or the question allows the person being tested to ramble on without a clear end to the answer.
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- d. Assign clear point break-downs for essay or multiple concept questions, both on the examination and on the key. This must be done to ensure
_ grading is consistent and fair.-
- e. , Award points only for responses that are in the key. 'The key must be corrected if it omits a correct answer.
- 4. Exit Meeting The examiners met with the licensee representatives listed below at the end of each week of the site visit to discuss the results of the evaluation to that point. During the second week of simulator examinations the examiners observed that the simulator modeling of the pressurizer and safety injection acc nnulators did not appear to perform as expected.
The following personnel were present at an exit meeting held on January 30, 1987. J. O'Brien, NRC Examiner M. Royack, NRC Examiner R. Wells, PVNGS, Lead Requalification Instructor D. Craig, PVNGS, License Training Supervisor W. Rudolph, PVNGS, lead Simulator Instructor T. Cotton, PVNGS, Acreditiation Project Manager J. Staveley, Jr., PVNGS . Lead Simulator Engineer J. Allen, PVNGS, Operations Manager The following personnel were present at an exit meeting held on February 6, 1987. G. Johnston, NRC Examiner R. Henry, PVNGS, Licensing Training W. Fernow, PVNGS, Training Manager D. Craig, PVNGS, Training Supervisor, License Operators W. Rudolph, PVNGS, Lead Instructor J. Staveley, Jr., PVNGS, Lead Simulator Engineer R. Wells, PVNGS, Lead Requalification Instructor L. Speight, PVNGS, Operations Representative I
(Attachment 1) RESOLUTION OF FACILITY COMMENTS Reactor Operators Requalification Examination Facility Comment: Question 1.01 "The actual limit for DNBR at PVNGS is 1.231 and not 1.2, this could have caused some confusion in the minds of the students." Reference; Technical Specification 2.1.1.1. Resolution: l The examiner sees no reason to change the key. The question will be changed , before it is put on the EQB. Facility Comment: Question 1.02 "May want to consider not just the initial response, but also the longer term effects on axial shape, due to Xenon, and on Axial power distribution, as the reactor power increases back to the secondary power following the cooldown caused by the dropped CEA. This should be optional material but not wrong if included." Resolution: The examiner will accept a correct response but will not change the key. Facility Comment: Question 1.04 "The students have been taught that if you half the SDM, then insert the same reactivity again the reactor's condition may be slightly supercritical. Per conversation with Bob Simmons, but proper response is still consistent with answer key and procedure." Resolution: The examiner will not change the key. ' 1 l Facility Comment: Question 1.08 i ? 1 "Part a. assuming TOTAL LOSS OF NATURAL CIRC, Delta T will remain constant based on stagnated condition of hot and cold leg, if proper explanation supports remain the same on Delta T, otherwise answer key is correct. Resolution: The key will remain the same. The examiner always considers the comments made by examinees for partial and, if warranted, full credit. I
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Facility Comment: Question 1.09 "This question needs to have a quantifier of time associated with it, if the student assumes that we made the power change according to approved station manual procedures, then the power change must have taken at least 5 hours. 10% per hour is the maximum rate of power increase that the procedures allow. At the end of a 5 hour long ramp (10% hour) the net effect of Xenon at that time will be negative or positive." Reference; Appendix. A of 410P-12Z05. Resolution: If the case of a ramped increase is specified then we will accept an appropriate answer. However, it is clear that a step increase is the subject of the question, no change will be made to the key. Facility Comment: Question 2.01
" Answers a and b are true that they will result in a RPCB signal to be generated, and answer d will cause a RPCB to be generated if it results in a turbine trip. Answer c should be the proper response to this question, it will only generate a runback signal if the RPCB system has been activated previously by any of a, b, or c above."
Resolution: The examiner will change the key, but will also add partial credit for part d as it will not directly cause a RPC signal, but will cause a turbine trip that will. Facility Comment: Question 2.02 From 0% to 15% (reactor power) the downcomer regulating valve ramps open to allow the feedwater to be preheated by the can deck, from 15% to 50% (reactor power) the economizer regulating valve ramps open to increase the efficiency of the steam cycle, meanwhile the downcomer regulating valve is closed between 15% (reactor power) and 50% (steam flow), finally at 50% (steam flow) the downcomer regulating valve " POPS" open to 10% " full feedwater flow position" this increases the recirculation ratio and reduces carryunder, meanwhile the economizer continues to ramp open from 50% (steam flow) to 100% (reactor power)." Reference; PWCS Functional Block Diagram (fortraining). Resolution: The examiner will change the key. Facility Comment: Question 2.03 Resolution: The facility comment is for information and does not affect the correct answer. No change will be made to key. { _ ____________ -______ _ __ - _ -__ _ _ O
Facility Comment: Question 2.04 Resolution: The facility wanted the following 2 items included in the key, the examiner agrees: Reference; PVNGS Dwg. 13-J-03K-025.
- 1. Reduced plant efficiency.
- 2. Decreased feedwater temperature.
Facility Comment: Question 2.05 Resolution: The facility wanted the following 6 items included in the key, the examiner agrees: Reference; PVNGS Dwg. 13-M-RCP-001
- 1. Loop drains from the cold legs.
- 2. "0" ring interspace drain.
- 3. RCP vapor seal leakage.
- 4. RCP cyclone filter drains.
- 5. Reactor head vent system.
- 6. Pressurizer venting system.
Facility Comment: Question 2.06 The facility reviewer noted that there are some steps in the procedure that were left out. Reference; 410P-1RC01, Steps 3.2, 3.8, and 4.2. Resolution: The examiner will include those specific steps from the procedure that apply. Facility Comment: Question 2.07 The facility identified two other trips for the heater drain pumps. Reference; PVNGS Dwg. 01-E-EDB-001, trip logic. Resolution: The examiner will include those trips in the key. Facility Comment: Question 3.01 The facility reviewer identified a correction to the key in that BU heaters will energize on high level error of 3% as long as pressure is less than 2350 PSIA. Reference; PVNGS Dwg. 01-E-RCB-010, 011, 012, 013. Resolution: The examiner will change the key. l l
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" Recommend that'the answer key accepts any of the four parameters as an Auxiliary trip by themselves, and not just as an overall statement on parameters." Technical Specifications Bases B2-6.
1 Resolution: The' question clearly indicates that what-is sought are the auxiliary trips. No change to the key. Facility Comment: Question 3.03
" Answer a should.say " loss of l'of the 2 Main FWPS"."'
- Resolution:
Examiner sees no need to change the key. Facility. Comment: Question 3.04 1
" Omit first part of answer - Highest Tcold does not impact power calculation because power uses'10 west'Tcold. 2nd part OK. - 3rd part wrong - aux' trip for Tcold high is 610 deg. F 40 deg. high = 605 deg. F. But fail Tcold will i impact asymmetric penalty on Tcold of 13'deg. F cause DNBR pretrip and trip."
Resolution: The examiner.will change key. Facility Comment: Question 3.05 The facility reviewer identified an' additional condition giving a CEA RWP. Reference; PVNGS Dwg. 13-M-SGP-002. Resolution: i The examiner will change the key. Facility Comment: Question 3.06
" Question should specify whether they asking about the Economizer Regulations valves (fails as-is) or the Economizer isolation valves (fails closed).
Resolution: The examiner agrees and will change key before uploading to the'EQB. The answer is not affected.
Facility Comment: -Question-3.07' The facility reviewer identified an error in the key as to the origin of the AMI and AWP. This involved .the originating signals. The reviewer also expressed a concern about part b, wanting full credit for the answer of "neither". Resolution: The examiner agrees and will change the key. Facility Comment: Question 3.08 "Part b - HLO prevents' water from entering the steam lines which could over stress lines due to weight considerations. Because we operate with high S/G level trip to protect carryover to turbine! High level trip causes MSIS which happens before HLO! At HLO Main Steam Iso valves already closed." Resolution: The examiner agrees and will change key. Facility Comment: Question 3.09 "Part b other terms that explain the 10-2% power response are Campbelling and Mean Square variance." Reference; PVNGS Training Article NS-4. Resolution: The examiner will accept those answers as they are terms descriptive of the answer key. Facility Comment: Question 4.01 The facility reviewer expressed a concern that if students allow for Xenon in their answer the correct response would be true. Reference; 410P-1ZZ03, precaution 3.8. Resolution: If the individual specifies the case he would get credit. Facility Comment: Question 4.03 "Part a #4 in your answer key is not in the bases of T.S. 3.1.1.4."
"Part b the reason for this limitation and precaution is to ensure the operators do not inadvertently lift the LTOPs, the Shutdown cooling System is protected from over pressure by the automatic close signal to the Shutdown Cooling Isolation valves, this is set for 500 psia."
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Resolution: The examiner agrees with the comment for part a. For part b the examiner ! feels that the answer is appropriate in the key and does not warrant ! changing, j i Facility Comment: Question 4.07 "Part b the RCS pressure control criteria is, pressure between 2200 to 2300 . psia and not 2275 psia." Reference; 41EP-1ZZ01. l Resolution: !
-i The examiner agrees and will change the key.
Facility Comment: Question 4.08 j The facility reviewer identified that the key scores points for steps that ! are not specifically required from the question. Reference; 41A0-1ZZ29, caution step 8.1. Resolution: The examiner agrees and will apportion points on the basis of the comment. Facility Comment: Question 4.09 "Part a of the question requests a description of " identified leakage", not the definition, the students could give many descriptions of identified ' leakage that fits into the category of " identified leakage"." Reference; Technical Specification Section 1. Resolution: The examiner points out that the individuals are responsible for knowing the definitions of tenns described in Section 1 of the Technical Specifications. This is clearly the case with this question despite the wording. The examiner feels no need to change the key. f l . I i ) !
EtJCLOS U R E 2.) l' ; TO 50 -5?8/04-87-01 a d' . U.S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR REQUALIFICATION EXAMINATION l Facility: _ PALO VERDE NGS Reactor Type: CE SYSTEM 90 Date Administered: Januacy ??. 1897 Examiner: Gary W. Johnston g Candidate: V. Kew 4(--
) J INSTRUCTIONS TO CANDIDATE Read the attached instruction page carefully. This examination replaces the current cycle facility administered requalification examination. Retraining requirements for failure of this examination are the same as for failure of a requalification examination prepared and administered by your training staff.
Points for each question are indicated in parentheses after the question. 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 Candidate's Category Value Total Score Value Category 15.0 25.0 1. Principles of Nuclear Power Plant Operation, Thermodynamics, Heat Transfer and Fluid Flow 15.0 25.0 2. Plant Design Including l Safety and Emergency Systems 15.0 25.0 3. Instruments and Controls 15.0 25.0 4. Procedures - Normal.
Abnormal, Emergency, and l Radiological Control 60.0 TOTALS Final Grade All work done on this examination is my own, I have neither given nor received aid. Candidate's Signature { i __ _ __ _ o
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.i EQUATION SHEET f = ma y s[t w = mg a=vt+-
2 Cycle efficiency = N 1 k( t) at E=mC a = (vf - y )/t -
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KE = mv vg = v, + a A = AN A=Ae PE = mgh m = 6/t A = In 2/tg = 0.693/tg W = v6P (t,)(tx) AE = 9314m % "' (tg+t)3 !
= nCpaT y ,7 ,-Ix 9 .
Q = UAAT y ,7 ,-ux
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. P=P 10 8 (t) TVL = 1.3/p t
P=P e /T HVL = 0.693/p SUR = 26.06/T T = 1,44 DT SCR =_S/(1 - K,ff) SUR = 26 [Aeffp) g CR = S/(1 - K ff ) T = (t*/p ) + {(s
- p)/xeff) E l ff T = 1*/ (p - Q M = 1/(1 - K,gg) = CR g/CR0 T = (3 - p)/ A,gf p g " (I ~ geff)0/II ~ Eeff)1 P " ( eff~I)/Eeff = AK,ff/K,gg gg , ,
p= [ L*/TK,'gg ] + [E/(1 + A,ggT )] 1*=1x1[ seconds P = E4V/(3 x 1010) ~l A,f = 0.1 seconds I = No Idg1=1d22 WATER PARAMETERS Id =Id2 g 1 gal. = 8.345 lbm R/hr = (0.5 CE)/d (meters) I gal. = 3.78 liters R/hr = 6 CE/d (feet) I ft = 7.48 gal. E SCELLANEOUS CONVERSIONS , Density = 62.4 lbm/ft 10 1 Curie = 3.7 x 10 dps Density = 1 gm/cm i kg = 2.21 lbm Heat of vaporization = 970 Etu/lbm I hp = 2.54 x 10 3BTU /hr
, Heat of fusica = 144 Btu /lbm 6 1 Mw = 3.41 x 10 Btu /hr 1 Atm = 14.7 psi = 29.9 in. I'g. 1 Btu = 778 ft-lbf 1 ft. H O = 0,4333 lbf/in 2 1 inch = 2.54 cm F = 9/5 C + 32
*C = 5/9 ( F - 32)
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ar , , .. ", + NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS I During the administration of this examination the following rules apply:
- 1. Cheating on the examination means an automatic denial of your appitcation and could result in more severe , penalties.
- 2. Restroomtripsaretobeiimitedandonlyonecandidateatatimemay -
leave. You must avoid all contacts with anyone outside the examination rnom to avoid even the appearance or possibility of cheating.
- 3. Use black ink or dark pencil only to facilitate legible reproductions.
- 4. Print v m name in the blank provided on the cover sheet of the exark s. e.
- 5. Fill in 4 on the cover sheet of the examination (if necessary).
- 6. Use on) 3per provided for answers.
- 7. Print yc . v.tx' n the upper right-hand corner of the first page of each ~
section of Ue . Swer sheet.
- 8. Consecutively number each answer sheet, write "End of Category " as appropriate, start each category on a new page, write only on o_ne side of the paper, and write "Last Page" on the last answer sheet.
- 1. Number each answer as to category and number, for example,1.4, 6.3.
- 10. Skip at least three lines between each answer.
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
- 12. Use abbreviations only if they are commonly used in facility literature.
- 13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer. required. .
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE -
QUESTION AND 00 NOT LEAVE ANY ANSWER BLANK.
- 16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.
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- 18. When you complete your examination, you ; hall:
- a. Assemble your examination as follows:
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(1) Exam questions on top. 1 (2) Exam aids - figures, tables, etc. I (3) Answer pages including figures which are part of the answer. l
- b. Turn in your copy of the examination and all pages used to anrwar the examination questions.
- c. Turn in all scrap paper and the balance of the paper that you dtd not use for answering the questions,
- d. . Leave the examination area, as defined by the examiner. If after
. leaving, you are found in this area while the examination is still in progress, your license may be denied or rev6ked. l e
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J I SEC110N 1 l 1 1 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, ~ j THERMODYNAMICS, HEAT TRANSFER AND FLU 1D FLOW ) 1 i
*OUESTION 1.01 (2.0)
Frequent mention is made in the technical specifications and elsewhere that saf ety limits are placed on Departure f rom Nucleate Dailing Ratio (DNDR) and Linear Heat Ratio (LHR).
- a. Describe what could happen if either 1imit is exceeded (DNBR less/ than 1.2, LHR greater than 21.O kw/ft.).
(1.25)
- 6. If one limit is violated, does this mean that other limit is also vialated? Exp1ain. (O.75)
- ANSWER
- a. Exceeding the DNBR 1imit could reduce heat transfer at c1ad surface and lead to localized burnout and failure of the cladding resulting in release of the fission products to the j RCS. Exceeding the LHR limit will lead to high centerline i f uel temperatures and melting of the fuel. Possible fucirod rupture could result from result:ng high internal pressure.
(Burnout)- (1.25)
- b. No, the limits and f ailure mechanisms are not related, i (0.70)
- REFERENCE
- 1. Nuclear Physics Reactor Theory notes, Volume IV, Sect. 4; CE PWR System T hermal-Hyd. Description pg. 7 and pgs.14-15.
- 2. RPS System Description pgs. 3-7
- 3. RPS System Description; CE PWR System Thermal-Hyd.
*OUESTION 1.02 (2.O)
If while operating at 50% power, a rod located very close to the middle of core (Rod 20) drops. What effect will this dropped rod have one
- a. LOCAL Radial Flux di stri buti on? (1.0)
- b. LOCAL Axial Flux distribution? (1.0)
+ ANSWER
- a. Radi al flux distribution will be skewed away from the droppro rod Peak flux will increase s'id the ilux at the point of the dropped rod will be near zero. (1.0)
- b. Axial flux distribution will decrease and shape remains the j same. (1.0)
- REFERENCE Pal o Verde React or Theory 414 i
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- QUES 1 ION 1.03 (2.O)
If pressurizer pressure control was lost, woul d the HPS1 pumpu be able to maintain the plant 28 F subcooled at no 1 cad Tav,e? WHY? (2. O)
- ANSWER Yes [0.53, no load Tave = 565 F plus 28 F for subcooling equal s 593 or 1468 psia. HPSI pumps have a shutoff head of 1775 psig and therefore will be able to maintain RCS 28 F subcooled L1.5].
(2.0)
- REFERENCE ESF system descriptions, and Steam lables
*OUESTION 1.04 (2.O)
A reactor startup is in progress and it i s necessary to dilut e 200 baron to get the critical baron concentration prior to pulling the control banks. Prior to the dilution, the source range instruments read 30 and 37 cps. After diluting 100 PPM of baron the same instrument s read 60 and 75 cps. Should the ; operator cont.inue with the planned dilution of another 100 PPM? ' Ex pl ai n. (2.0)
- ANSWER NO, if the counts are doubled the SDM i s half ed, thus the shutdown reactivity was decreased by approximately 50%. By addi ng t he same amount of reactivity again the reactor would be critical. (2.0)
- REFERENCE Pal o Verde 1esson 48
*QUECTION 1.05 i1.O) l Expl ain how and why moderator temperature coefficient changes with baron concentration. (1.0)
+ ANSWER Increased boron concentration makes moderator temperature coefficient less negative (0.5) due to greater change in boron concentration with density change of water (0,5).
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- REFERENCE Plant Physical Data Book, Standard Nuclear Theory I
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*DUE s t l ON 1.06 (1.5)
Would control rod worth be greater:
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- a. When the moderat or is at 150 degrees F or when it is at 500 degrees F? (0.5)
- b. When it is wi thdrawn with al1 other rods inserted, or when it is inserted by itself with al1 other rods withdrawn?
(0.5)
- c. When the Baron concentration is 500 ppm or when the Baron concentration is 1500 ppm"? (0.5) o
- ANSWER
- a. 500 deg F
- b. Withdrawn
- c. 500 ppm
- REFERENCE Palo Verde Requalification exam bank
*DUESTION 1. O'7 (1.0)
Choose f rom the list below the operating condition i or a typical centri 4ugal pump which would requi re the most pump work (or motor horsepower / amperage): ( 1., 0 )
- a. discharge at peak efficiency head (
b, diccharoe at design head I
- c. ditcharge at shutoff head
- d. discharge at runout head
- ANSWER "d" (1.0)
- REFERENCE St andard F 1 ui d Methani c s:,/ Pump Desi gn , Pal o Ver de thermodynamics, Heat Transfer and Fluid Flow handbook Volume XI
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+ DUE ST 10N 1.08 (2.O)
If, after oper at i ng in natural circulation for 2 hours, an operator error caus,es a complete loss of natural circulation flow, how will the f ollowing parameters change ( i nc r ease ~, decrease, or remain the same)? Bri ef l y e>:pl ai n your answer. ( Assume- no 4 urther operator attion.) (2.O)
- a. core del ta "T"
- b. steam generator pressure
- ANSWER
- a. Increase (0.5) as boiling occur s in the core. Th (0.25) will increase while Tc (0.25) remai ns rel ativel y constant.
- b. Decrease (0.5) l ess primary-secondary heat t ransf er (0,5).
- REFERENCE Heat 1ransfer, Thermodynamics, and Fluid Flow, General PhysicsCAF
*DUEST10N 1.09 (0.5)
After optarating at 50% power 4 or several days, the reactor power is increased to 100% power. Reactivity changes in the n e>: t 1 to 2 hour s will be... (Sel ect the most correct response.)
- a. Negative due t o the 6.2 hour half-1ife of Iodine 135.
- b. Posi i i ve due ta the facter burnup of Xenon at higher neutron f 1 u >: 1evels.
- c. Positive due to t he increased r at e cd decay of Xenon to Cesium,
- d. Negative due to the increased production of Xenon at the higher iinsion rate.
- ANSWER (b)
- REFERENCE Palo Verde Req ual i f i c at i on e>:am bank
- QUESTION 1.10 (1.0)
What happens t o the A>:i al Shape I n d e>: when power is reduced to 95% power from 100% power at the end of cycle operation with the insertion of rods only? (1.0)
- ANSWER Adding negative reactivity to the top of the core (0.5) causes the AS1 ta become more positive (O.5) (power is deiven to the l bott om of the core.).
- REFERENCE l Pal o Ver de Requal i f i cati on e>:ami nat i on bank l 1
l l I l J
SECTION 2 PLANT DESIGN INCLUDING SAFETY , AND EMERGENCY SYSTEMS
- QUESTION 2.01 (1.O)
Concerning the Renttor Power Cutback System; which of the following will NOT cause a RPC signal to be generated?
- a. Lo control oil pressure #2 feedpump,
- b. Large load rejection signal initiated by the operator manually,
- c. Secondary pressure biased by average reactor coolant temperature is less than setpoint, rt . Main turbine thrust bearing oil supply pressure low. (1.0)
- ANSWER Response d will cause turbine trip not RPC. (1.0)
- REFERENCE RPCS System Description pg. 11
*DUEST10N 2.02 (1.5)
Describe the posi tioning of the steam generator f eedwat er Regulating valve to the steam generator during a power level iner ease f rom 0% to 100% power. Explain the reasoning for each phase of this valve program. (1.5)
- ANSWER From 0% to 1G% the downcomer FWRV r amps open ta preheat thr feedwater in order to prevent thermal shock. From 15% to LOX the downcomer FWRV is shut to increase the ef f iciency of the steam cycle From 50% to 100% the downcomer FWRV ramps open to 10% to alleviate the af f ects of steam carryunder, and enhancing the recirculation ratio. (1.5)
+ REFERENCE RCS System Description pgs. 23-24 F eedwat er Contral Syst em Description pgs. 2-3
- QUES 1 ION 2.03 (2.O)
- a. While load testing a diesel gener ator (p ar al l el ed and loaded to 4000 kW) a SIAS is received. How will the diesel generator respond? (1.0)
- b. While running a diesel gener at or not synchronized t o the bus a loss of power (LDP) signal is received. How wi 11 the diesel gener at or respond? (1 O)
- ANSWER
- a. The diesel generator will cont 3nue to run and the output breaker opens. (1.0)
- b. Generator will close onto the bus (1.0)
- REFERENCE Diesel System Description pgs. 47, 53-59
a ' -
*DUESl1ON 2.04 (1.O)
Whi1e operating at 50% the outIet drain valve for feedwater heater 6 fails closed. What are THREE control room indic,ations of this event? (1.0)
*ANSWEk
- 1. High Ievel al arm in heater 6
- 2. Fai11ing ievel in the heater drain tank
- 3. Reduced f1ow from hreater drain pump. (O.33 each)
- REFERENCE Pal o Ver de ettam questi on bank
*DUEST1ON 2.OS (2.O)
T here ar e f i ve dir ect piping system connections bet ween the Chemical and Volume Control system and the Reactor Coolant System (RCS). What er e FOUR of these five connect 2ons? (2.O) l
- ANSWER O.25 for any four of the 4 allowing.
l Letdown line (off the loop 11J cold leg). l Char gi ng line to loop 1A cold leg l Charging line to loop 2A cold leg l Au>:i l i ar y pressurizer spray line to the pr essur i :* er . Reactor Cool nt Pump normal seel leakoff.
REFERENCE:
Training System Description, Chemical and Volume Control System, pages 1, 2, 60, and 61
~, L .,,
*OUESI10N 2.06 (2.O)
You are beginning a normal pl ant heat-up 4 o11owing ref ueling. A bubble is in the pressuri::er , the RCS has been vented, a,n d the check-off list for the reactor cool ant pumps ' electrical and valve line up has been completed. In order to start the first reactor cool ant pump cert ain i ni ti al conditions must be established or l verified. List four of the five initial conditions which must be satisfied prior to star ting the first reactor coolant pump. (2.0)
- ANSWER !
O.5 each for any four of the f ollowing: I i NPSH requirements must be met Upper and lower bearing oil reservoirs must be full Normal component cooling water flow niust be established to the motor coolers and the seal wat er heat exchanger. Controlled bleedoff flow must be established. An oil lift pump and 'an anti-reverse rotation device pump must be operating (redundant pumps in standby)
- REFERENCE Tr ai ni ng Syst em Descri pt i on , Reactor Coolant System page 45 and procedure 8023-3-1.7, Reactor Coolant Pump Operation
*DUESTION 2.07 (1.5)
What THREE conditions will cause a heater drain pump to trip, other than manual? (1.5)
- ANSL'ER
- 1. Lo-Lo l evel in t he associ ated dr ain tank
- 2. Low NPSH to the
- 3. High Steam generator feedwater pump suction head (1.D)
- REFERENCE 410P .EDO1 pgs. 9-10
( l l
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*DUEST ION '2. 08 (2.O)
Why is the deborating ion exchanger i nstal l ed in the Chemical and Volume Control Syst em'? (2.0) ~
- ANSWER Remove Boron at end of life (????? "B" conc) instead of dilution (1.0) to reduce large volume of waste generated by dilution (1.0).
- REFERENCE CVCS SYSTEM DESCRIPTION
*OUESTION 2.09 (2.O)
Explain two (2) f unctions served by the bypass orifices in the pressurizer spray lines. (2.0) I I
- ANSWER
- 1. Line warming / minimize thermal tr ansi ent of spray nozzle. (1.0) j
- 2. Turn over pressurizer water inventory (1.0) I I
- REFERENCE RCS SD
)
9
. L ,,
1 i I SECT 10N 3 I 1 I
- INSTRUMENTS AND CONTROLS ,
I I 1 l *OUEST10N 3.01 (2.0) Given that reactor power is at 100% with the Pressurizer Pressure J control system (PPCS) and Pressurizer Level control system (PLCS) { in automatic, Tc fails low with Reactor Regulatirig selected to i l Tave. What are the control actions and alarms associated with the PFCS and PLCS7 (2.0)
- ANSWER
- 1. With low Tc, setpoint program goes to minimum ('30%) which produce a high level error alarm. EO.53
- 2. Maximum letdown flow. [0.53
- 3. Energize backup heaters on high level error if less than j 2275 psia. [0.53 '
- 4. Stops normal running charging pump.EO.53 (2.0)
- REFERENCE Palo Verde exam bank
- QUESTION 3.02 (1.5)
Concerning the Core Protection Calculators (CPC's): i
- a. One of the inputs to the CPC calculators is neutron power.
What THREE corrections does this power si gnal undergo? (0.75)
- b. What are THREE of the FOUR auxiliary trips generated by the CPC. (0.75)
- ANSWER
- a. 1. Shape annealing 2. Rod shadowing 3. Temperature shadowing (0.75) [0.25 irach3
- b. 1. Core conditions outnide the analyzed parameter space 2. <
Less than 2 RCP's running 3. Maximum hot leg temperature greater than the satura- tion temperature for the pressurizer pressure. 4. Internal processor faults (0.75) Eany 3 (0.25) each]
- REFERENCE CPC system description pgs. 9, 16-18
. t -
*OUEST10N 3.03 (2.0)
Describe the two (2) events for which the Reactor Power Cutback System is designed to allow the NSSS to remain at power.,(2.0) i
- ANSWER
- 1. Loss of 1/2 main feed pumps. (1.0) i
- 2. Large loss of load. (1.0) f
- REFERENCE Palo Verde NSSS Lt?cture on RPCS, page 4
*OUESTION 3.04 (1.5)
During normal 100 percent power oper at i on , WHAT wi11 happen to RPS channel "A" if Trie channel "A" cold-leg temperature input fails 40 F high. (1.5) (Include any meter response or alarms.)
- ANSWER The high cold leg temperature will cause the calculator power to drop causing a NI-calculator power deviation EO.53. The cold leg temperature may cause a DNBR trip, since Tc is an input to the DNBR calculation 00.53; or an auxiliary trip may occur because Tc is higher than 580 degrees fshrenheit whicb would prevent the DNBR tripCO.53 (1.5)
- REFERENCE Palo Verde E.xam bani.:
- QUESTION 3.05 ( 1. O )
What THREE (3) conditions will give you a CEA withdrawal prohibt (CWP) with Reactor power >1%? (1.0)
- ANSWER 1.High pressurizer pressure
- 2. Local power density
- 3. Low DNBR (1.0) EO.33 each]
- REFERENCE CEA System description pgs. 8-10, 17 l
{ i l
e ( ,,
*0UES110N 3.06 (1.O)
- a. Following a Reactor trip due to a loss of instrument air, how is the plant decay heat removed? (0.5) ,
- b. How does the Econi mi: er valves f ail on a loss of insturment air? (0.5)
- ANSWER
- a. By the use of Atmospheric dumps (steam dumps f ail closed).
(0,5)
- b. as-is (0.5)
- REFERENCE 41AO-17206 pg. 4
*OUEST10N 3.07 (2.0)
The Control Element Drive Mechanism Control System (CEDMCS) receives Automatic Motion Inhibit (AMI) or Automatic Withdrawal Prohibit (AWP) signals from other plant systems.
- a. What are the signal (s) (AMI or AWP) from the Reactor Regul ating System to the CEDMCS and describe the plant j parameters or conditions which cause the si gnal (s) . (1.0) i
- b. List the signal (s) (AMI or AWP) from the Plant Protection System to the CEDMCS and describe the plant parameters or conditions which cause the si gnal (s) . (1.0)
- ANSWER (a) REACTOR REGULATING SYSTEM Automatic Withdrawl Prohibit signal is generated by an excessive devi ati on between T(ave) and T (re f ) (6.8 F) or by high cold leg temperature (558 F) (1.0)
(b) PLANT PROTECTION SYSTEM Neither, the Plant protection syst em generates a CWP signal. (1.0) or Control El ement Assembly Withdrawl Prohibit signal is generated by pretrips from high reactor pressure, high KW/FT, or low DNBR. (High KW/FT and low DNBR pretrips are not enabled until > 10-4 7. power)
- REFERENCE Training Syst em Description CEDMCS, pages 50 and 51
- QUEST 10N 3.08 ( 1. O )
The feedwater regulating system has two abnormal conditions, Reactor Trip Override (RTO) and High Level ~ Override (HLO). i
- a. What is a RTO designed to accomplish or prevent ? (0.5)
- 6. What is a HLO designed to accomplish or prevent ? (0.5)
- ANSWER (a) The purpose of the reactor trip override (RTO) is to prevent pressuri: er pressure and level transients resulting from overcooling
~
of the reactor coolant after a reactor trip. (0.5) (h) HLO prevents moisture carryover to the mai n turbine. (0. 5) W, Sca ornt swuis .
- REFERENCE Trai ni ng System Descripti on , Feedwater Regulating System, page 32
*OUEST10N 3.09 (2.O)
Regardingl the ex-core Nuclear Safety Channel instruments:
- a. At low power levels (approximately 10-2 to 10-8 % power) explain how neutron interactions are s elected over gamma interactions 7 (1.0)
- b. At power levels between 10-2 and 200 % power where individual pulses cannot be counted by normal methods how is the power l evel signal generated 7 (1.0)
- ANSWER (a) A pulse height di scr i mi nator is used to eliminate lower amplititude gamma culses from the higher amplititude neutron culses. (1.O)
(b) At > 10-2 % power neutron pulses pile up and produce an AC signal which fluctuates around a mean DC signal. The square of this fluctuation is pr op or ti onal to power. (Gammas have a smal l er fluctuation about the mean and, si nce this f l uc tuat i on is squar ed , the gamma-to-neutron current ratio is even larger, resulting in effective gamma d2ccrimination.) (1.0)
- REFERENCE Training System Description, Excore Neutron Monitoring System l
l l \ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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f 1 ( l l
*OUESTION 3.10 (1.0)
What are f our (4) of the five (5) limi tirig condi tions f or i operation monitored by COLSS? (1.0) , f
- ANSWER
- 1. Peak Linear Heat Rate (0.25 ea) FOR FOUR
- 2. Margin to DNB/DNBR
- 3. Total. Core power ,
- 4. Azimuthal tilt '!
- 5. A>:i al Shape I n d e>:
- REFERENCE Palo Verde NSSS Lectures on COLSS, page ;?
I i
~
t.C ,, . 2 SECT ION 4 PROCEDURES - NDRMAL, ABNORMAL, , AND RADIOLOGICAL CONTROL
*OUESTION 4.01 (2.0)
Answer the f ollowing questions TRUE or FALSE in ref erence to t he Reactor Startup procedure,
- a. All regulating CEA groups shall be withdrawn in their prescribed sequence except for physics and surveillance tests. (0.5)
- b. Positive reactivity additions by more than one methoc is acceptable below the point of adding heat. (0.5)
- c. If the estimated critical position (ECP) is considered unreliable by the Shift Supervisor, he can direct the critical approach to be guided by a 1/M plot. (0.5)
- d. Only three reactor coolant pumps are required operating when the reactor is critical. (0.5)
- ANSWER
- a. True
- b. False
- c. "Irue
- d. False (2.0)
- REFERENCE Palo Verde exam bank
- QUESTION 4.02 (1.5)
Safety Injection can be throttled from its ful1 flow configuration if certain conditions are met. What are THREE of the four conditions that must be met? (1.5)
- ANSWER
- 1. RCS subcooled >28 F
- 2. P r. level greater than 33% and controllable
- 3. One S/G capable of maintaining heat removal
- 4. RVLMS indicates void restricted to upper head. 00.5 each3 (1.5)
- REFERENCE 41RO-12ZO1 pg. ~/
l
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+
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- QUESTION 4.03 (1.5)
What ar e the reason (s) for the following limitations and pre-cautions" 7 ,
- a. The reactor shall not be made critical if RSC temperatur e is less than 552 deg. F. (TWO required) (1.0)
- b. Do not allow pressurizer pressure to exceed 410 psia when the shutdown cooling loop is in operation aligned to the RCS. (0.5)
- ANSWER
- a. 1. Ensures modulator temperature coefficient is within its analyzed temperature range.
- 2. The protective instrumentation in within its normal operating range.
- 3. To ensure consistency with the FSAR safety analysis.
- 4. The reactor vessel is above its minimum RTNDT temperature. (0.5) For any two.
- b. Overpressure protection of SDCS components. (0.5)
- REFERENCE TS Bases 3/4.1.1.4 410P-1SIO1
*0UESTION 4.04 (1.0)
Midway through an inspection tour in a radiation area, you drop your and pocket dosimeter is now reading 'off scale high'. The TLD was damaged and can not be used to determine exposure. What are the immediate actions required of you? (1.0)
- ANSWER Immediately proceed to exit point and report to Radiation Protection Section. (1.0)
- REFERENCE 75AC-9ZZO1 APPX A of 75RP-9ZZ12
*DUESTION 4.05 (1.0)
Choose from the list below the automatic protective trip or shutdown of the emergency diesel generators which is bypassed by a SIAS. (1.0)
- a. Engine Overspeed
- b. Engine Lube Oil Pressure Low
- c. Generator Differential
- d. Incomplete Sequence
- ANSWER "d"
- REFERENCE Emergency Diesel Generator Oper ati ng Procedure 410P-1DG01 l
1 1 a.
)
l *DUESTION 4.06 (2.0) According to PVNGS Procedure 41AO-12Z35, Continuous CEA Withdrawal -
- a. What actions should be taken to stop the rod withdrawal? ,
11.0)
- b. What action is required by the procedure if the continuous rod withdrawal cannot be stopped? (1.0) i
- ANSWER I
- a. Stop the withdrawal by selecting standby on the CEDMCS remote operator module. (1.0)
- b. Trip the reactor manually. (1.0) (Procedure 41EP-1ZZO1)
~* REFERENCE 41AO-1ZZ35 and Palo Verde Requlification exam bank
- QUESTION 4.07 (1.0)
Following a reactor trip you are in the process of completing ohe standard post trip actions.
- a. What is the RCS inventory control criteria ? (0.5)
- b. What is the RCS pressure control criteria ? (0.5)
- ANSWER (a) Pressurizer l evel must be controlled between 33 %
and 55 %. (0.5) (b) Pressurizer pressurizer must be maintained less than 2275 PSIA. (0.5)
- REFERENCE Emer gency Operating Instruction 41RO-12ZO1, Reactor Trip i
- _ _ _ _ . _ _ _ _ _ _ _ ____________U
~
4
*OUESTION 4.08 (2.O)
Palo Verde 1 is operating at 100% power. The f ollowing alarms come in rapidly: ,
"RCP LO NCW" "NUC CLG WTR SYS trb1" "NCWS PUMPS DISCH HDR PRESS HI-LO" What conditions does the applicable procedure (41AO-1ZZO5 "Less of Nuclear Cooling Water") require that the reactor be tripped?
And how much time can elapse before this action must be taken? (2.0)
- ANSWER If no N.C. pump can be started CO.5] or neither essenti al cooling CO.53 water train can be manually cross connected and operated properly EO.5] within ten minutes E0.5] the reactor must be i tripped. (2.0) {
- REFERENCE Pal o Verde Requali f i cati on e>:am bank 41AO-12ZO5
*OUEST10N 4.09 (2.0)
The f ollowing questions relate to primary coolant e>yst em leakage as defined in the Technical Specifications.
- a. Describe the three sources of Identified Leakage. (1.5)
- b. Define Unidentified Leakage. (0.5)
- ANSWER
- a. Leakage into closed systems, such as pump seal or valve packing leaks that are captured, and conducted to a sump or c ol l ec ti ng tank. (0.5) !
\
Leakage into the containment atmosphere from j sources that are both specifically located and j known either not to interfere with the operation j of leakage detection syst ems or not Pressure i Boundary Leakage. (0.5) l l 1 Reactor cool ant system leaketge through a steam i generator to the secondary system. (0.5)
- b. All leakage which is not Identified Leakage is Un identified Leakage. (0.5) I i
*RETERENCE Technical Specifications, Definitions, pages 1-3 and 1-6 l
1 1
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- QUESTION 4.10 (1.0)
Regarding Radiation protection l i mi t s. What are the weekly, quaterly, and annual Admi ni strati ve limits for radiation doses to
~
the whole body? (1.0) f
- ANSWER Week)y 3OOmRem Quarterly 1000 mrem Annualy 4000 mrem (0.33) each j
- REFERENCE I 75AC-9ZZO1, page 26
*END 1
l j l l 2 l 1 l 1 I i 1 a b
~
U. .. F. U.S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR REQUALIFICATION EXAMINATION Facility: PALO VERDE NGS Reactor Type: CE SYSTEM 80 Date Administered: January 27. 1897 Examiner: Garv W. Johnston Candidate: INSTRUCTIONS TO CANDIDATE ,
- Read the' attached instruction page carefully.. This examination replaces the current. cycle facility administered requalification examination. Retraining requirements for failure of this examination are the same as for' failure of a requalification examination prepared and' administered by your training staff.
Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at j
-least 80%. Examination papers will be picked up four (4) hours after the !
examination starts.
% of Category . % of Candidate's Category value Total Score value ' Category-15.0 25 . 0 ., 1.- Principles of Nuclear Power Plant Operation, Thermodynamics,-Heat Transfer and Fluid Flow 15.0 25.0 2. Plant Design Including Safety and Emergency '
Systems 15.0 25.0 3. Instruments and Controls 15.0 25.0 4. Procedures - Normal, Abnormal, Emergency, and Radiological Control 60.0' TOTALS Final Grade All work done on this examination is my own, I have neither given nor received { aid. ' i l Candidate's Signature
EQUATION SHEET f = ma v = s/t V = mg a=vt+ at 2 Cycle efficiency = et Work (out) 3 o Energy (in) E = mC" a = (v - y )/t KE = my -t v f
=v o +a A = AN A=Aeo PE = mgh to = e/t A = In 2/tg = 0.693/tg i W = vaP AE = 931Am t q(eff) = (ti:)(ts) !
(t ,t) j h=[nC4T p 7 7e -IX Q = UAAT y,7g -ux - Pwr = Wf In I = Io10 * ~
, P=P 10 SUR(t) TVL = 1.3/p t
P=P 0 e /T HVL = 0.693/p l SUR = 26.06/T .b ' T = 1,44 DT (A *ff p\ SCR = S/(1 - K ff) SUR = 26 -g l CR = S/(1 - Keffx) T = '(t*/p ) + [(f 'p)/x g,p] CR1 (1 - Keff)1 = CR 2 (1 ~Keff)2 T = 1*/ (o - D M " 1/(1 ~ Eeff) = CR /CR y 0 T = (I - p)/ A ggf p E " ( eff-1)/Keff " OEeff/Keff seg = (1 - geff)IEeff p= [ t*/TK,'ff .] + [H/(1 + Aeff)] T t* = 1 x 10~ seconds P = I4V/(3 x 10 ) A,ff = 0.1 seconds
~1 I = Na Idgy=Id22 WATER PARAMETERS Id =Id22 y
1 gal. = 8.345 lbm R/hr = (0.5 CE)/d (meters) I gal. = 3.78 liters R/hr = 6 CE/d2 (geet) 1 ft = 7.48 gal. MISCELLANEOUS CONVERSIONS , Density = 62.4 lbm/ft 10 1 Curie = 3.7 x 10 dps Density = L gm/cm 1 kg = 2.21 lbm Heat of vaporization = 970 Etu/lbm I hp = 2.54 x 10 BTU 3
/hr i , Heat of fusicn = 144 Btu /lbm 1 Mw = 3.41 x 10 Btu /hr 6
1 Atm = 14,7 psi = 29.9 in. I'g. 1 Btu = 778 ft-lbf I ft. H O = 0,4333 lbf/in 2 1 inch = 2.54 cm F = 9/5 C + 32 "C = 5/9 ( F - 32)
l J s NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS
)
During the administration of this examination the following rules apply:
- 1. Cheating on the examination means an automatic denial of your application j and could result in more severe , penalties.
- 2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibt'iity of cheating.
- 3. Use black ink or dark pencil only to facilitate bqible reproductions.
- 4. Print your name in the blank provided on the cover sheet of the examination.
- 5. Fill in the date on the cover sheet of the examination (if necessary).
- 6. Use only the paper provided for answers.
- 7. Print your name in the upper right-hand corner of the first page of each ^
section of the answer sheet.
- 8. Consecutively number each answer sheet, write "End of Category " as appropriate, start each category on a new page, write only on o_ne side of the paper, and write "Last Page" on the last answer sheet.
- 3. Number each answer as to category and number, for example,1.4, 6.3.
- 10. Skip at least three lines between each answer.
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
- 12. Use abbreviations only if they are commonly used in facility literature.
- 13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer. required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE -
QUESTION AND CD NOT LEAVE ANY ANSWER BLANK.
- 16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.
I. . . l-l
- 18. When you complete your examination, you shall:
- a. Assemble your examination as follows:
(1)' Exam questions on top. (2) Exam aids - figures, tables, etc. (3) Answer pages including figures which are part of the answer.
- b. Turn in your copy of the examination and all pages used to answer the examination questions.
- c. Turn in all scrap paper and the balance of the paper that you d.i.d not use for answering the questions.
- d. Leave the examination area, as defined by the examiner. If after
. leaving, you are found in this area while the examination is stili in progress, your license may be denied or rev6ked.
e a O 1 L
4 1- , o i 5 I l
'SECTION 1: 4
- PRINCIPLES OF' NUCLEAR POWER. PLANT' OPERATION, THERMODYNAMICS, HEAT' TRANSFER AND' FLUID FLOW
- QUESTION 1.01- (2.O)
Frequent mention - is made in_the technical specifications and-
.elsewhere that . saf ety limits are placed on ' Departure f rom
.l' Nucleate Boiling . Ratio . (DNDR) and Linear.. Heat Ratio (LHR).
a.; Describe what'could'h'ppen a if either, limit is exceeded
~ (DNBR less than 1.2, LHR greater than 21.0 kW/ft.).
j
~
'(1.25) i b ,, - I f one: limit.is violated, does this mean that other l'imit
- i s al so vi cl ated? Explain. (0.75)-
.*OUESTION 1;O2 (2.O)
If while operating at 50% power, a rod 1ocated very.close to the , middle of core (Rod 20) drops. What effect will this dropped rad ! have ont
- a. Local Radial Flux distribution'? (1.0) -i
- b. Local Axial Flux distribution? (1.0)
J
- QUESTION 1.03 12.O)
- .I f pressurizer pressure control was lost, would the HPSI pumps be able.to: maintain the plant.28 F subcooled at no' load Tave? WHY7 j (2.O) j i
*OUESTION 1.04 (2.0) j A reactor startup is in progress and it is necessary to dilute j 200 baron to get the critical baron concentration prior to. j pulling the control banks. Prior to the dilution,'the source -j range instruments read 30 and 37 cps. After diluting 100 PPM of
' boron the same instruments read 60 and 75 cps. Should the operator continue with the planned dilution of another 100 PPM? j Ex p1 ai n. (2.O) J
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1 i 1 l i ( 1 L ; 1 l~ I l l I l l i I l l 1 i n -- _ - _ _ _ _ - _ _ _ - - - _ - _ Q
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*0UESTION 1.05 (1.0) >
Explain how and why'mederator temperature' coefficient changes .i with bor'on concentration. (1.0) ]
*0UESTION'1.06 (1.5) ;
'Would control' rod worth be greater
'a.. When~the moderator is at 150 degrees F or when it is at 500 'l degrees F? (0.5) j
- b. When it is withdrawn with all other rods i nser ted , or when 4
it is inserted in/ itself with all other rods withdrayn? - (0.5) ,
- c. When~the Baron concentration is 500 ppm or'when the Baron
-concentration is 1500 ppm? (0.5) {
l
*DUERTION 1.07. (1.0) ]
Choose f rom the' list below the operating condition for a typical ]
-centrifugal _ pump which woul.d require the.most pump work (or motor ]
horsepower / amperage): ( 1. 0 ) - l La. discharge at peak efficiency head'
'l
- b. discharge at design head '
~j k !
i
- c. ' discharge at shutoff head J
- d. discharge at runout head
- QUESTION 1.08 (2.0)
If, after operating in natural circulation for 2 hours, an operator error causes a complete loss-of natural circulation flow, how will the following parameters change (increase, decrease, or remain the same)? Briefly explain your answer. i' (Assume no further operator action.) (2.0) , 4
- a. core delta "T" l l
- b. uteam generator pressure
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*OUESTION 1.09 (0,5)
Afier operating at SO% power ior several days, the reattor power is increased to 100% power. Reactivity chances in the next i to 2 hours will be... (Select the mont correct response.)
- a. Negative due to the 6.2 hour half-life of ladine 135.
- b. Positive due to the faster bur nup of Xenon at higher neutron flux levels.
- c. Positive due to the increased rate of decay of Xenon to Cesium.
- d. Negative due to the increased production of Xenon at the higher fission rate.
*DUESTION 1.10 (1.O)
What happens to the Axial Shape Index when power is reduced to 9S% power from 100% power at the end of cycle operation with the insertion of rods only? (1.0) END OF SECTION 1 GO ON TO SECTION 2 NEXT PAGE
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or ( SECTION 2 1 L o PLANT DESIGN INCLUDING SAFETY i AND EMERGENCY SYSTEMS I
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- QUESTION 2.01 (1.0) l Concer.ning;the Reactor. Power Cutback System; .which of'the j following will' NOT cause'a RPC cignal to be generated?
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- a. Lo control oil pressure #2 feedpump. . . j
- b. .Large' load rejection signal initiated by the operator j
, manually.
- c. Secondary pressure' biased by: average reactor coolant temperature is-less than setpoint. ~
- d. Main turbinesthrust'. bearing oil supply pressure low- ( 1.' O )
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*0UESTION'2.02 - ( 1. 5) ..
i Describe the positioning of the steam generator feedwater
- Regulating valve-to the steam generator during a power level
-increase from 0%-to 100% power. Explain the. reasoning for each phase'of this valve program. (1.5) !
- QUESTION 2.'03 '(2. 0) 1 J
'a.. While load testing a. diesel generator (paralleled and loaded to.4000 kw) a GIAS is received. How will the diesel generator respond? (i . 0)
- b. While running a . diesel generator not synchronized to the bus a loss of' power (LOP) signal is received. How will the diesel-generator respond? (1.0)
- QUESTION 2.04' (1.0)
While operating at 50%:the outlet drain valve for feedwater
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heater 6 fails closed. What are THREE control room indications of 1 this event? (1.0)
- QUESTION 2.05 (2.0)
There are five. direct piping system connections between the
' Chemical - and Volume Control' system and the Reactor' Cool' ant System (RCS). What are FOUR of these'five connections? (2.0)
*OUESTION 2.06 (2.0) 'l Ycu are beginning-a no'rmal plant heat-up following refueling. A !
bubble is in the pressurizer, the RCS has been vented, and the
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l check-off list for the reactor coolant pumps' electrical and valve line up ham'been completed. In order to start the first reactor l coolant pump certain i ni ti al conditions must be established or , verified. List four of the five initial conditions which must be satisfied' prior to starting the first reactor coolant pump. (2.0)
*OUESTION 2.07 (1.5)
What THREE conditi ons will cause a heater drain pump to trip, other than manual? (1.5)
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- QUESTION 2.OS- -(2.O)
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Why is the deborating ion' exchanger installed in the Chemical and g- Valume Cont.ral System? (2.O) , l* QUESTION 2.09 (2. O ) - 4
+ Explain two (2)' ';f uncti ons served by the bypass ori f i ces in the pressurizer spray lines. (2.0) i END OF SECTION 2 GO ON TO SECTION 3 NEXT PAGE ~I k
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s. SECTION 3 I'NSTRUMEN1S'AND CONTROLS
*OUESTION'3.01 (2.O)
- Given.that~ reactor power is .at' 100% with the Pressurizer Pressure control, system-(PPCB) and Pressurizer, Level control system (PLCS) in automatic, Tc fails low with Reactor Regulating selected?to Tave. What are the control. actions and alarms associated with the PPCS and PLCS? .2.O) (
*DUESTION 3.'02 (1.5)-
Concerning the Cor e Protection Calculators (CPC's): a .- One of tile inputs to the' CPC; calculators is neutron power.-
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What THREE corrections'does this power signal undergo? (0.75)
- b. What are THREE of:the FOUR auxiliary trips' generated by the j CPC. (0.75)
*DUESTION 3.03-(2.0)
Describe the two (2) events for which the Reactor Power Cutback System is designed to allow the NSSS to remain at power. (2.0) :;
- QUESTION 3.04 (1.5)
During normal 100 percent power operation, WHAT will happen to RPS channel "A" if The channel "A" cold-leg temperature input
- fails-40 F high. (1.5)
(Include any meter response or al arms. )
- QUESTION 3.05 (1.0)
Wnat THREE (3) conditions will give you a CEA withdrawal prohibt (CWP) with Reactor power >1%? (1.0) l
*DUESTION 3,06 (1.0) q
- a. Following a Reactor trip due to a loss of instrument air, how is the plant decay heat removed? (0.5)
- b. How does the Econiminer valves fail on a loss of insturment ;
air? (0.5) . I i i
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l :- *DUESTION 3.07' (2.O) ! The Contral Element Drive Mechanism.Contrel System (CEDMCS)
'recuives (Tutamatic Motion. Inhibit (AMI) or Automatic' Withdrawal Prohibit (AWP) signals from other plant systems.
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.a. ' What are the signal'(s) (AMI or'AWP) from the Reactor g Regulating System to the CEDMCS and ' describe the plant
! parameters or conditions which cause the signal (s) . (1.0) l
- b. Li st. the signal (s) (AMI'or AWP) from the Plent. Protection l System to the CEDMCS and describe the plant parameters or 1- conditions which cause the signal (s) . (1.0)
I *QUESTIGN 3.08 (1.O) The feedwater regul ating . system has two abnormal I conditions, Reactor Trip Override (RTO) and High Level Override l (HLO). 1
- a. What is a RTO designed to accomplish or prevernt 7 (0.5) l l b. What is a HLO designed to accomplish or prevent 7 (0.5)
*OUEST10N 3.09 - (2. O )
Regarding the ex-core Nuclear.Saiety-Channel instruments:
- a. .At low power levels (approximately 10-2 to 10-8.% power) explain how neutron interactions are selected over gamma
- l. interactions 7 (1.0)
- b. At power. levels between_10-2 and 200 % power where individual pulses cannot be. counted by normal methods how is the power-level signal generated ? (1.0)
*OUEST1'ON 3.~10 (1.0)
.What are four (4) of the five (5) l i mi ti ng . condi ti ons for
. operation monitored by COLSS7 (1.0)
END OF SECTION 3 GO ON TO SECTION 4 NEXT PAGE
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g - - - _ _ _ _ - - t- . . [ . / SECTION 4 PROCEDURES - NORMAL, ABNORMAL, .l L AND RADIOLOGICAL.' CONTROL ' ' 1' l
- QUESTION 4.01 (2.0) . ;
.. Answer the f ollowing questions TRUE or FALSE in ref erence to the ;
Reactor Startup procedure. ] 1 lt a. All regulating'CEA groups shall be withdrawn in their i prescribed sequence except for physics and surveillance I tests. (0.5) e . . . i
- b. Positive reactivity additions by more than one method is. ]
acceptable below the point of adding heat. (0.5) j c.- If the estimated. critical position (ECP) is considered unreliable by the Shift Supervisor, he can direct the critical approach to be guided by a 1/M plot. (0.5) ) l
- d. Only three reactor-coolant pumps are required operating when j the reactor'is critical. (0.5)
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- QUES 110N 4. 02 (1.5) i Safety Injection can be throttled from-its full flow configuration if certain conditions'are met. What are THREE of the four conditions that must be met? (1.5)
-*OUESTION 4.03 (1.5)
What are the recson (s) for the f ollowing limitations - and pre-cautions" 7
- a. The reactor shall not be made. critical if RSC temperature i s less than 552 deg. F. (TWO required) (1.0)
- b. Do not allow pressurizer pressure to exceed 410 psia when the shutdown cooling loop is in operation aligned-to the-RCS. (0.5)
*OUESTION 4.04 (1.0)
Midway through an inspection tour iri a radi ation area, you drop your'and pocket dosimeter is now reading 'off scale high'. The TLD was damaged and can not be used to c'atermine exposure. What are the immediate actions required of you? (1,0)
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- QUESTION 4.05- ( l'. O)
Choose f rom theflist below the automatic protective trip or shutdown of the. emergency diesel generators which is "bypaused by a'SIAS. (1.0) 1 a ,, Engine Overspeed i
- b. ' Engine Lube Oil Pressure. Law.
l~ .c. Generator Differential
- d. Incomplete Sequence i
'*DUESTION 4.06 (2.0) i According'to PVNGS Procedure 41AO-1ZZ35, Continuous CEA Withdrawal
- a. What actions should be taken to stop'the rod withdrawal?
(1.0) i
- b. What action is required by the procedure :if the continuous ,
rod withdrawal cannot be stopped? (1,0)
- QUESTION 4.07 ( 1. O )
. Foll owi ng a reactor, trip you are in the process of completing the standard post trip actions.
- a. What is the 'RCS inventory control criteria ?? (0.5) l
- b. What is the RCS pressure control criteria ? ( 0. 5) -
*DUESTION 4.08 (2.0)
Palo Verde 1 is operating at 1007. power. The following alarms come in rapidlyn l l
'"RCP LU NCW" ,
l "NUC CLG WTR SYS trb1" 1
"NCWS PUMPS DISCH HDR PRESS HI-LO" l .l What conditions does the applicable procedure (41AO-1ZZO5 " Loss of Nuclear Cooling Water") require that the reactor be tripped?
l And how much time can elapse before this action must be taken? l
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*DUESTION 4.09 (2.0) l The f ollowing questions relate to primary coolant system leakage ,
j as defined.in the Technical Spetiii cati ans. I' l a. Describe the three sources of Identifaed Leakage. (1.5) l
- b. Define Unidentified Leakage. (0.5) l l
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*QUEEa rION 4'.10 ' .i*.0)
Regarding Radiation protection limits. What are the weekly, -! quaterly, and annual Administrative limits f or radiation doses to :i the whole body? ( 1. O) END OF SECTION 4- l 1 END OF EXAMINATION i l l l' l [ l l > I l:- l l' l l l;
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