ML20195B578
| ML20195B578 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 10/26/1988 |
| From: | Bjorgen J, Hare E, Jordan M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20195B575 | List: |
| References | |
| 50-155-OL-88-01, 50-155-OL-88-1, NUDOCS 8811020081 | |
| Download: ML20195B578 (150) | |
Text
{{#Wiki_filter:. - _ _ _ _ n U.S. NUCLEAR REGULATORY COMISSION REGION III , t Report No. 50-155/0L-88-01(DRS) Docket No. 50-155 -
' License No. DPR-6 Licensee: Consumers Power Company l 212 West Michigan Avenue Jackson, MI 49201 Facility Name: Big Rock Point Examination Administered At: Charlevoix, Michigan Examination Conducted: August 30, and October 12, 1988 Examiners: J. C. Bjorgen 7
///%5/88 D' ate E.A.Hareh /*[.2['[
Date f[f d I[ Approved By: M. J. Jordan, Chief , Operator Licensing Sectio Date Examination Summary Examination administered on August 30 and October 12, 1988 Report No. 50-155/0L-88-01(DR5)) Written examinations were administered to two Senior Reactor Operator and one Reactor Operator candidates. Operational examinations were administered to one Senior Reactor Operator candidate. , Results: One Senior Reactor Operator candidate passed all portions of the examinations. One Senior Reactor Operator candidate and one Reactor Operator candidate failed the written examination ands consequently, were not given an l operating examination. l l I l l 1 l
0 REPORT DETAILS
- 1. Examiners J. C. Bjorgen, Region III, Chief Examiner E. A. Hare, Region III
- 2. Exit Meeting An exit meeting was held on September 1, 1988. The following personnel were present at this meeting.
Consumers Power Corpany f T. W. Elward, Plant manager R. R. Frisch, Senior Licensing Analyst R. Buckner, Nuclear Training Administrator D. G. La Creix, Senior Nuclear Instructor - C. R. Abel, Production and Performance Superintendent P. Donnelly, Nuclear Safety Administrator l T. R. Fisher, Senior Quality Assurance Consultant ' H. Scheels, Shift Supervisor E US NRC J. C. Bjorgen, Chief Examiner l E. A. Hare, Examiner t G. C. Wright, Chief, Operations Branch t N. R. Williamson, Resident Inspector The following topics were discussed at the exit meeting: f i
- a. The examiners expressed concern that the examinations could not be completed during the week of August 29 1988, due to the lack of L adequate reference material being provIded for preparation of the i examinations. The operating examinations had to be postponed because the operational scenario's were developed using material .
that was no longer valid for the plant. This included the ' Emergency Operating Procedures and the Nuclear Instrumentation , System Description. > t An additional meeting was held with the plant manager on October 12, i
, 1988, to briefly discuss the findings of the operating examination. I L
No generic weaknesses were identified.
- 3. Examination Review Copies of the written examinations and answer keys were given to the facility for review at the conclusion of the examination. The facility coments were provided to the examiners on September 9,1988. The coments and resolution are enclosed as Attachment 1 to the report.
t 2 i
ATTACHMENT 1 l i General Coments Coment: Several questions were not directly related to Big Rock Point. Although the information reflected in those questions was, in fact, based on study material for the Reactor Operator / Senior Reactor Operator Program, it was provided as supplemental information on which a student can base his understanding of the concept and apply that understanding to Big Rock Point. In most cases, a question directly applicable to Big Rock Point could have been asked. It is felt that .he student can more fully comprehend specific Big Rock Point concepts if some of this additional information is made available. Examination questions in those areas should be avoided since they do not, for the most part, aid in determination of the student's understanding and knowledge of this plant. Response: Enclosure 1 to the 90 day letter, dated May 12 1988, provided specificguidanceontheneedforplantspecifIcreference material. Such guidance is provided to avoid the situation identified in the above coment. Nonrelevant material should be clearly indicated as such in the material furnished for exam developement. Comment: In general, the Nuclear Regulatory Commission's existing policies on comunications with the facility during development of the Hot License Examinations, and review of the examination after it has been administered, have led to problems that could have been avoided. Communications with the facility while developing examinations on any uncertain points is the best tool to eliminate examination discrepancies. Response: The staff is aware of the licensees' concerns as noted in the coment. Examiner Standard ES-201, Revision 5, Attachment 2, Enclosure 4, "Requirements For Facility Review of Written Examination," will allow the Chief Examiner the option to have the facility review the written examination. Whenever this option is utilized, a Pre-Examination Security Agreement will be signed. Regardless of whether this option is utilized or not, the submittal of thorough, accurate, and appropriate reference material is essential to minimizing problems with examination content. Coment: Numerous questions require "setpoints" to be included in the answer. However, the guidance of ES-202 states that setpoints should be avoided. Because of this, setpoints are not normally required to be memorized and student answers may reflect this. Response: Specifically, six questions required numerical responses in the answer: 1
(1) Question 2.04. The numerical value enclosed in parenther.es in the answer key is for information and is not required for full credit. (2) Question 2.10. Same disposition as Question 2.04 above. Parentheses inadvertently omitted from Part b of the answer key. (3) Question 2.11. The 150 psig value requested in the question relates to the shutdovr. cooling system operational design limitations as providcd in the lesson plan and system operating procedure. Expected candidate knowledge of this value is consistent with the examiner standards and NUREG 1123. The candidate d.monstrated this knowledge by correctly answering the question. (4) Question 3.01. This question requested the candidate to provide a safety system interlock (when a scram is bypassed). This is consistent with the examiner standards and NUREG 1123. (5) Question 3.03. This question requested the control settings for the safety related liquid poison system heaters and the length of time (Design Criteria) that the system remains operable upon loss of the heaters. These knowledges are consistent with the examiner standards and NUREG 1123. l (6) Question 3.11. This question tests for the operational
- interlocks for the Circulating Water Pump and it's discharge l valve. This knowledge is consistent with the examiner standards and NUREG 1123.
l l Comment: Several answers in the examination keys contained parameter values I which included tolerances. These tolerances, in all cases, are ! absent from the references given and appeared to be somewhat l arbitrarily assigned. Additionally tolerances were assigned to I parameters for which none exist. Several of these were cases where the reference and plant procedures given parameter values as approximate pressures, times, etc. Clearly, no tolerance applies to these and, in fact, no tolerance applies to any of the values where they appear in the answer key using the reference provided. Response: In general, when numerical responses are requested, examination policy and common sense necessitate that a nominal tolerance be assigned. This allows the candidates to demonstrate their knowledge of the important system numerical values in magnitude without being penalized if they miss the correct answer by a small amount. When none is present in the facility material, the examiners assign a value that is subject to facility review and comment. This approach was used in two cases on Questions 2.11 and 3.11. The candidate is not required to state these nominal tolerances in his/her answer. l l 2 k
t t l Two other questions were also assigned tolerances based on the same l general guidance. In these two cases, Questions 3.01 and 3.03, the !' tolerance assigned was based on the conflicting information in the provided reference material. ! Question 3.01. Two values were given for the scram bypass i setting; less than 500 psig and 450 psig. Hence the tolerance ! of 500 + 0 - 50 psig. P Question 3.03. Regarding liquid poison heater settings the material I referred to two heaters, one set at 140*F, the other set at 150 F. , Hence, the 140 - 150'F tolerance. ! i l I i 1 4 1 l 3
I 9 SPECIFIC COMENTS ; Comment: Questions 1.01, 5.01. Big Rock Point (BRP) does not have "IRMs." Several hours later leads the operator to a point past the peak (four hours + time to startup + several (?) hours). l Actual xenon concentration equals rate of production - rate of removal. Mathematical calculation is not required on the job. Actual time at which peak is reached and xenon concent*ation decreases and adds reactivity would be f,omething less than 7 to 11 hours. The multiple choice answers do not reflect the actual condition. This question and answer does not solicit the operator's knowledge ofthesubjectmatter.
Reference:
BRP Lesson Plant BNA 31A and 31A. BRP Reactor Theory, Page 45 (Peak 7-11 hours af ter scram). BRP Reactor Theory, Pages 42 to 45. Resolution: Agree in part. Clarification to the candidates was made during exam that the reactor was operating at system pressure and temperature in lieu of "Range % of the IRM." The Xenon ! being added due to reactor re-start is also adding negative ' reactivity. In order to maintain reactor at system pressure and temperature, control rods would have to be withdrawn to compensate for Xenon build-in. This question tests basic understanding of how Xenon effects reactor power. The answer l key was not changed. I l Comment: Questions 1.03, 5.03. The answer to this question is dependcat on I exactly what temperature and pressure the start up is at when the j rod drop occurs as to whether temperature coefficien'. will affect the accident. Students may quality, t .
Reference:
Reactor Theory Handout, Pages 28, 29, 30, 31, 32, 33 v.usolution: Agree with the comment. Student may qualify. "NRC Rules l and Guidelines for License Examinations," used to brief . the candidates prior to taking the test states, "Show l l all calculations, methods, or assumptions used to obtain
~
1 l an answer . . . ." If based upon a candidate's assumption ' I that was correct, he was graded accordingly. Answer key unchanged. t Comant: Questions 1.04, 5.04. Inswer "c" increase. Referent.e material indicates that an increase or decrease may occur. r
Reference:
Reactor Theory Handout, Page 38 (J.3.a.6.a). ; l _ - - - . _____. __ _ . _ ~ . _ _ _ _ _ _ _ _ _ . . _ _ . . . . , _ _ . . _ _ _ _ _ _ _ - , _ _ _ , _ _ _ _ _ . . _ . ,
Resolution: Agree with comment. Part e of this question was no longer discriminating (i.e. , did not discriminate between a knowledgeable vs unknowledgeable candidate). Therefore, Question 1.04/5.04 c. was deleted.
-Comment: Questions 1.06, 5.06. Sha=N nnd deep rods concept not used.at BRP. This information useu -
_ oadground information. Resolution: Comment noted. LearningObjecti . 4 of Control Rod section states "Discuss Flux shaping wi' etrol Rods." During ' discussions with the site "eact . engineer, it was noted that the concept of flux shaping using control rods is used to some extent at Big Rock Point. The concepts utilized at the plant and the specific question asked in the exam was consistent with the training material and requires the type of general knowledge expected of licensed operators regarding control rod theory. Answer key unchanged. Comment: Questions 1.08, 5.08. Critical heat flux is used only as it applies to critical power.
Reference:
BRP Technical Specifications, P. ige 43. Resolution: Comment noted. Since no specific recommendation was made, no modification of answer key was made. Comment: Questions 1.10, 5.10. BRP also has flow indicators in the control room.
Reference:
P&ID S-001-A. Resolution: Agree with Comment. Answer key was changed to accept recirculation flow indicator in control room. Recommend reference material be updated to reflect this.
- Coment: Questions 1.11, 5.11. Barrier fuel not used at BRP. This concept is used only to illustrate possible sclutions to PCI.
Reference:
BRP Material Science Less Plan. Resolution: Comment noted. Enabling Objective No. 5 states, "Describe the method incorporated to remedy pellet cladding interaction as described in Chapter 10," (NUS Reactor Materials Manual, Chapter 10). Listed under PCI Remedies is a discussion on "barrier fuel." The way the lesson plans at Big Rock are written makes it very difficult for the examiner to differentiate between what is required knowledge of an operator (per enabling / learning objectives) and what is actually used at Big Rock. It is recommended that all training material ba revised to clarify between background information and the required information a candidate can be examined on. Answer key unchanged, since it did not change the pass / fail decision for the candidates. This cuestion will not be uploaded to the exam bank to prevent it's future use. 2
Comment: Question 1.14. Equation sheet on R0 Exam has 1 x 10 -5 sec = A*. Student used this number for calculation. Answer should reflect this number. Resolution: Comment noted. Equation sheet is considered to be in error. Answer key changed to allow the use of 1 x E-5 or 1 x E-4. Comment- Questions 2.01, 6.01. Question should lead students to ONP if that is the specific answer requested. As this question is stated, other answers would be acceptable, i.e., Loss of Station power, MG set supply breaker open, etc. Resolution: Comment noted. ONP 2-36 does not list specific causes for a loss of MG set. P&ID WD 740, Sheet 11 does not show any direct causes related to RPS MG set loss. ONP 2-35 is the only document tchich specifkally states possible causes for loss of a RPS MG sat. Loss 01 Station Power is an indirect cause of loss of MG set. The direct cause would be due to low voltage of MG set. Answer key was changed to accept low voltage as an alternative answer for loss of generator voltage and to accept supply breaker and output breaker trip. Comment: Questions 2.03, 6.03. Pump can also be started locally at the diesel fire pump from the magnetic switch levers.
Reference:
50P-26, Section 6.3.3. Resolution: Comment accepted. Answer key changed. If this information was provided initially then questions like these would not occur. Comment: Question 2.04. Setpoints are not required to be memorized in the training and should be avoided in written exvinations. Question 6.04 did not ask for setpoints.
Reference:
NUREG 1021, Revision e, Page 3 of 6 E. General Guidance No. 6. Resolution: Comment noted. Specific values, when not mandatory in the answer, are indicated by enclosing them in parentheses in the answer key. Parentheses added to the Part b answer key. Comment: Questions 2.05, 6.05. M0-7067, Bypass Valve Isolation Valve closes on isolation scram.
Reference:
P& ids - Drawing No. 7040-G30743, Sheet 3, Reactor Protection System - Drawing No. 7040-G30119, Sheet 1, Turbino Bypass Isolation Valve. P Resolution: Comment accepted. Answer key was changed to acca e i Valve isolation Valve (M0 7067). Recommend refet .ial be updated to include this valve. 3
q Comment: Questions 2.06, 6.07. RDS Modifications, c. Air amplifier not presently being used. This modification was made to ensure valve closure anytime operated at operating system pressure including during testing. Resolution: Comment noted. However comment implies.that this modification is still in place. Answer key unchanged. Comment: Questions'2.07,6.08. Question does not lead the student to the system description. Further information is given in the. SOP. and the Technical Specifications.
Reference:
Standard Operating Procedure SOP-8, Post Incident System; and Technical Specifications, Page 133. Resolution: Comment noted. 50P-8 basically rewords the primary function of the Post Incident / Emergency Core Cooling System as defined in the system description. Exact wording is not necessary for full credit. BRP Technical Specifications does not clearly state the primary functions. It discuss the Bases for the requirement for operation. Answer key was not changed. , Comment: Question 2.10. Setpoint not required. Setpoint in answer key wrong (75 mr/hr). Actual setpoint at present (72 mr/hr). Setpoint by procedure is (10 mr/hr above background).
Reference:
ALP 1.3.33. Resolution: Comment noted. The 75 mr/hr valte is enclosed in parentheses in the answer key indicating that it is not required for full credit. Comment: Question 2.11.(a). Shutdown System Design pressure = 300 psig. Shutdown System placed in service when Reactor pressure is less than 280 psig. Turbine seals ned only very low pressure s9 psig. 150 psig is not required to establish seals. Turbine seals are placed in service by proceauce at 100 psig.
Reference:
SOP-5, Section 2.2.4, 50P-5, Section 3.0.3, SOP-1, Section 6.1.1.15. Question 2.11.b; Panel number (P-29) should not be required. Resolution: (a) Comment not addressed since the candidate corNCt}y answered the question per the answer key. (b) Comment noted. Panel number not required provided candidate demonstrates sufficient knowledge of panel location (i.e., inside or outside containment). Comment: Question 3.01. S-4 switch commonly referred to at BRP as "mode switch." Actual settings of switch 450 psig. 4
Reference:
'Page 55, Section 6.1.'3, 50P 1, Section 3.0.11.
Resolution: Comment ~noted. Answer key changed to add (mode) to switch identification. Tile 450 psig switch setting was included in the original unswer key tolerance. w Comment: Question 3.03(a). Answer of approximately 150*F is acceptable.
Reference:
Comment noted. Answer key unchanged.- Comment: Question 3.05. Students should be given credit for answer in accordance with DC Wide Range Monitor handout.
Reference:
Lesson Plan, BNA-31B, Section 6, Trips. Resolution: Coment noted. _The lesson plans received for the new instrumentation were essentially instructor guidelines with insufficient information to adequately develop exam questions. cl.ue to the differences between the System Description Manual and Lesson Plan information received, Question 3.05 was deleted. Comment: Question 3.11. Answer key (a) SOM says 33*, SOP-21 says 30%, ALP 1.4.44 says N 1/3; (b) SDM says 30 , 50P-21 says 30%, ALP 1.4.44 says N 1/3. Resolution: Coment noted. Review noted that ALP 1.4.44 is incorrect. ALP 1.4.42 (Revision 139, Page 37) refers to a Trip setting of 30% open. 30%openwouldbe27 degrees (90 degrees travel x 0.3). Answer key for Part d changed to 27 to 30 degrr.es " Coment: Quen. ion 3.07.(b). Fire pumps do not start on a signal of 2'9" Reactor water level. Fire pumps start on decaying system pressure, or on low steam drum level of -17 inches. Core spray valves open when: Reactor level reaches 2'9" above the core and Reactor pressure drops to < 200 psig. 200 psi;; is not considered an interlock. Check valves are installed in the system to prevent oserpressurizing fire system.
Reference:
SOP-8, Page 5 of 11, Section 6.2,1 and SDM Chapter 8, Page 2 of 8, Section 8.3.1. Resolution: Coment noted. Wording of Part b. answer chan start . . ." to "the system actuates . . . ." ged The fromremainder "the pumps of the coment was not addressed since the candidate answered correctly per the answer key. Coment: Question 3.09. Answer No. 2, Student may answer down stream of condensate demins or between condensate demins and LP heater. 5
Reference:
P&ID Drawing No. 0740-G44011, Sheet 1 of 2, S-011-A, Condensate System Valve Line-up Diagram. Resoli: tion: Comment noted. Answer key unchanged since candidate gave a location consistent with the answe? key. Comment: Question 4.02. Answer key, Value appears to have been derived from an interpolation of the graph. The procedure is to find the line associated with the approximate Beta energy range in this case >.55 MEV. The student should have used the >0.5-0.6 MEV line and the answer should have been Q.4 mci /sec. Interpolation ' not required.
Reference:
SOP-34, Page 2 of 10, Section 6.1.1. . Resolution: Comment accepted. Answer key changed to 2.3 - 2.5 millicuries per second. For Part b an additional answer of "verify stack gas monitor switched to high range" and poliit distribution changed to-any 5 at 0.4 points each. Comment: Question 4.04. Student did not answer on direction from examiner. Question should be eliminated. Question 4.05. Student did not answer on direction from examiner. Question should be eliminated. Question 4.06. Student answered in accordance with , Big Rock Point E0Ps.
Reference:
E0P-1: Immediate operator action "scram the Reactor" E0P-2: Immediate cperator action would be in response only to further information than is given, i.e: Centainment temp >100 F C(ntainment Pressure >l psig Containment water level 574 feet l Only answer "scra0 the reactor." Resolution: Comments noted. Since the E0Ps were not provided in the ; reference material supplied by the l!censee, Question 4.04 thru 4.06 were deleted. Comment: Question 6.06. Students answered this question in accordance with BRP's existing DC Wide Range Monitors. Reference DRP L.sson Plan, BNA-31B, Section 6, Trips. Resolution: Comment noted. Question was deleted since it did not address the new system installed at Big Rock. Recommend that the , i training material (SDM) be updated to reflect this system change. l l t t 6 l
-, _ _ j
Coment: Question 6.10. Hi Condenser Pressure same as Low Condenser Vacuum. MSIV same as Main Steam backup Isolation (only one Main Steam Isolation valve at BRP).
Reference:
P&ID Drawing No. 0740-G30743, Sheet 2, M0-7050, Main Steam Isolation Valve. Resolution: Comment accepted. Answer key was changed to accept "or Low condenser Vacuum" and MSIV will be accepted the same as steam line Backup Isolation valve. Coment: Question 7.02. The purpose of these interlocks is so you don't move fuel while a control rod is withdrawn from the core.
Reference:
Final Hazard Sumary Report, Section 7.7.2, Page 13. Resolution: Coment noted. F.H.S.R. Section 7.7.2 states the following: l Rod Withdrawal During Refueling l Interlocks are provided to prevent all motion with any of the refueling cranes (namely, jib crane, transfer cask winch and monorail crane), which are positioned over the reactor whenever any control rod is not fully inserted in the core during fuel loading or fuel manipulation. This is identical to the BRP Technical Specifications. Answer key was not changed. Comment: Question 7.03. EMP-3.5A deleted. Examiners told student not to answer. Students did answer in accordance with E0P-1 and con *.ingency 4 (ATWS), i Reactor scram setpoint exceeded and Reactor power >5% or cannot be determined, and Reactor pressure >1360 psig and rising; or l Drum level <-8" and falling; or Containment Pressure >l psig. j
Reference:
E0P-1 Resolution: Coment. noted. Question was deleted since the reference material was no longer current. Coment: Question 7.04. This information is the same in the the BRP E0Ps and the answer is correct for either the E0Ps or the EMPs. 7
Reference:
E0P-1 Cautions. Resolution: Comment noted. Question was deleted since reference to EMP-3.3 was not longer current. Comment: Question 7.07(a). This question is in reference to general informatian given in the ONP 'and major headings in the procedure. Are students required to memorize this information? , Students may give more specific information covered under major headings.
Reference:
ONP-2.7, Sections 2.0 and 3.0. , Resolution: Comment noted. Sectius 2.0 and 3.0 basically reword the answers , L contained in the answer key. Exact wording is not required for l full credit. Answer key unchanged. i Comment: Question 7.15. EMP-3.5A has been' deleted. Delete answer 3. E0P-1 RC/Q gives requirements for injecting poison. Include (3). 1 Reactor scram condition and power >5% or cannot be determined; and Reactor pressure >1360 and rising; or Drum level <-8" and ft.liing; or Containment pressure >l psig.
Reference:
E0P-1 RC/Q Resolution: Comment noted. Answer key was changed to delete a.3 since EMP 3.5 is no longer in use. Comment: Question 8.09. DC Wide Range Monitors have the same requirements as the PICOs had. 4 Answer key is correct. t
Reference:
BRP Technical Specification, Page 57. } Resolution: Comment noted. From the reference material supplied by the licensee it is not clear that the DC Wide Range Monitors have the same requirements as the PICOs. This question was deleted. Comment: Question 8.13. Memorizing definitions is avoided where they do not apply to plant conditions. At BRP, it is felt to be adequate knowing that the definitions in Technical Specifications exist. It is a better operational practice to look up the definition when required to do so. This operator action tends to eliminate errors in memory.
Reference:
NUREG 1021, Revision 4, ES-202, Page 3 of 6, Section E.5. 8
. c ',' .
O. Reso1'ution: Comment noted. In this specific case, memorization of the required definitions is substantiated by learning objective No. 3 which states "Define the terms presented in Technical-Specification, Section 12." Answer key was not changed. E' t i A 4 4 9
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, NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS Durimg the administration of this examination the f ollowing rules appl y: ' 1 :. Cheating on tha examinatiors meens an automatic denial of your application and could result i t; more severe penalties.
- 9. 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 possibility of cheating.
- 3. Use black ink or dark pencil only to facilitate legible reproductions.
- 4. Pr i tit your riame in the blank pr ovided on the cover sheet of the examination.
- 5. Fill in the date on the cover sheet of the enemination (if necessary).
- 6. Uso only the paper provided for answers.
- 7. Print your name f ri the upper right-hand corner of the first page of each '
section of the answer sheat. O. Consecutively riumber each enswer sheet, write "End of Category __" an appropriate, start each category on a new page, write only on one side of the paper, and write "L~ast Page" on the last answer sheet. Number each answer as to category and number, for example, 1.4, 6.3. 9.
- 10. Skip at least three lines between each answer.
- 11. Separate onswer shoetu f r oin p .4 d and place finished answer sheets face ;
duwre on your desk or table. r l '2. Uce abbrevi ations oril y 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 usud as a guide for thu depth of answer required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer l to mathematical problems whether indicated i n the' questi on or not. l
?
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE DUESTION AND DO NOT LEAVE ANY ANSWER DLANK.
- 16. If parts of the examination are not clear as to intent, ask questions of i the examinor oril y.
- 17. You must sign the statement on the c.over sheet that indicates that the work is your own and you have not received or been given assistance in compl et t rig the> exami nation. 1his must be dune after the examination has been completed.
i 4 i
- 10. When you complete your examinstion, you bha 11:
'a . Assnmble your examination as follows:
(1) Exem questions on top. (2) Exam aida - figuros, tables, etc. (3) Aritwer pagen including figures which are part of the answer .
- b. lurn in your copy of the exarnination and all pagou used to answer tiee eneminct i on questi ons,
- c. lurri in all scrap paper and the balance of the paper that you did not uno tor answering the questions.
- d. Leove the examiriation area, as defined by the examiner. If efter l eevi ng , you are found in this area while the examination is utill i n pr ogr ecs, your license may be denied or revoked.
O
Pego 4
' It__EBJ U9J E!.E@_QE_UpQLg66_E9 WEB _EL6MI_QEgB811993 # 10EBOOpfyOU1GS 3 _UEOI_IBOUSEEB_9dp_ELUlp_Elgy OUESTION 1.01 (1.00)
The reactor trips from full power, equilibrium xenon conditions. Four (4) hours later tg g egor is brought critical and power level is maintained & r; - - m,m M;M f or several hours. Which of the following statements is CORRECT concerning rod motion during this period?
- a. Rods will have to be withdrawn due to xenon build-in.
- b. Rode will have to be rapidly inserted since the critical reactor will cause a high rate of xenon burnout.
- c. Rods will have to bra inserted since aenon will closely follow its normal decay rate.
- d. Rods will approximately remain as is as the xenon establishes its equilibrium value for the power level.
QUESTION 1.02 (2.00) The effect~ive/ decay constant in the wei ghted average of t h e d c-c ay constants for the six precursor groups. Typical values are 0.1 for up power tr ansi ents and 0.05 for down power transients. What is the basis for the varying values? QUESTION 1.03 '. 2. 50 ) Absume that the reactor is being sterted up from Cold Shutdown and a rod drop accident occurs early in the startup. Of the void, doppler, and temperaturo coefficients, which will act first, second, and third to limit the rapid power rise? EXPLAIN YOUR AN3WER. (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
1x__EBIN91 ELE 9_9E_ NUCLE 86_EQWEB_ELGNI_QEEBOligN2 Pcg2 5
- ' IUE6099XN00195.i_UE01_IBONSEEB_099_EL91D_Eb99
.f r fO QUESTION 1.04 CLM4 FILL IN THE DLANK r The react *vity worth of a single control rod will __________. (For each sstatervnt below, indicato INCREASE OR DECREASE.)
- a. If the void content arouted the rod INCREASES.
- b. If the moderator temperature DECREASES. j m it _e , - , , , ,,..._i .m, e 4m u t , m , ,, , ,
d4[4[ch
- d. If Xe-135 concentration around the rod DECREASES.
QUESTION 1.05 (1.bO) LIST three b) operational methods that will lengthen the cycle of plant operation. r 4 OUESl10N 1.06 ( 1. 00 ) . '. J For each of the f ollowing types of corit r ol rods "a" and "b", choor.e t-hose characteri sti cs 1 through 3 which describe that type. (More than orm char acteri stic MAY epply to each type. ) Control Rod lype 4
- a. Shallow Coritrol Rod
- b. Doup Control Rod r
Char ar i er s ut 2 c t, [
- 1. A one-or-two not ch withdr awal can result in reverstr power offect. ,
l
- 2. Substantially affocts radial power diutribution.
- 3. Also I:nown as Shaping Rods. l OUES1 ION 1. O'/ (1.SO) ;
HOW doen xenon concentretiori affect peripheral rod worth following a scram from high powur and Why does this occur'? (***** CATEGORY 1 CON 1INUED ON NEX1 PAGE *****) i
.Jz__P81Ug1PLES_Gr NUCLEAR POWER _ PLANT OPERATION 3 'Pcgt 6
.' s IUE6009XUeUIg_,_UgeI_IBeNEEE8_eUQ_E691D_EL9H OUESTIOt1 1.08 (1.50) Concerning heat transfer in a boiling water reactor, DEFINE the following terms:
- a. . Departure from Nucleate Boiling (DNU)
- b. Critical Hoat Flux (CHF)
- c. Critical Heat Flun Ratio'(CHFR)
OUES110t1 1.09 (2.00) A reactor heat bal ance was perf ornied by br.nd during the last shift. Answer the following questions TRUE or FHl.SE.
- a. If the feedwater flow rate used in the hest balance calculated was LOWER than thu actual f eedwater flow rate, the actual power is HIGHER than the currently calculated power. (0.5)
- b. If the reactor recirculation pump heat input used in the host balance calculated was OMllTED, then the actual power is LOWER than the currently calculated powor. (u.5)
- c. If the steani flow used in the heat balance calculation war LOWEH than .he actual steam flow, then the actual power is LOWER then the curr entl y calculated power . (0.L)
- d. If the Clean-up return tempera *.ure used in the heat balance calculation web LOWER than the actual Clean-up return temperature, then the actual powee is HIGHER than the currently calculated power. (0.5)
OUESTION 1.10 (2.00) List four (4) possible indications of reatur al circulation flow < reo recirculation pumps running). QUESTION 1.11 (1.00) EXPLAIN how berrier fuel difft Ihf om the previous types of fuel drWAgn and/or material makeup) (1.0) Icaded in the core. (i . e (***** CAIEGORY 1 CON 11NUED ON NEXT PAGE *****)
12;_ERJU91ELgS'OF NUCbgeB_E9 WEB _ELeNI_gEgBAI1QN3 Pago 7 I 'IUEBd99YUedics,_HgeI_IBeugEEB_eND_ELUIp_E69W OUEST10t1 1.12 (1.50) lhe attached-Figure i shows a T-S diagram for a steam cycle.
.a. What phenomenon is taking place between points 1 and 27
- b. How does this process benefit plent operation?
- c. What is the disadvantage associated with this process?
.UUESilON :.13 (1.00)
When the flow rate through a centrifugal pump decreases, availeblo NPSH (INCREASES or DECREASES) ond required NPSH (INCREASES or DECREASEG). Choose correct recponses. (1.0) QUES 110N 1.14 (2.50) Calcul ate the reactor period for the given condi tions: Beff = .007
- a. Meff a 1.000) for a prompt critical reactor
- b. Keff = 1.0001 for a delayed critical reactor (Show all work)
QUESTION 1.15 (1.00) With the plant operating at 100*/. power the feedwater reguleting valve mal +uncti ons cauuing an 1NCHEASE in feedwater FLOW. this causes core inlet subcooling to (INCREASE, DECREASE, REMAIN THE SAME). As a result the reactor will trip on high (LEVEL, POWER, PRESSURE) early into the event. ( Two parts G 0,5 pts each) (***** END OF CATEGORY 1 *****) 1'
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- T- S hyam Qu e erIoN 1,12 ;
c ;--- -,----- , Page 8 L. 2t_mEbeUI_QEElGU_1GGbue1Ng_S'EgIy_6ND_EMEB@EUQY e synlEUS OUES110N 2.01 (1.50) A 1oss of Reactor Protection Motor Generator Set i has just occurred.
~
List 1HREE possible causes for the' Motor Generator loss. QUESlION 2.02 (2.25) The Reactor Cleanup Pump Selector Switch, HS-6410, provides three modes of op er.3 tion for the cleanup pump and valve CV-4042. STATE the THREE modes of operation of HS-6410. Include puanp and valve CV-4042 status. QUESilON 2.03 (3.50) Regarding ihe Diesel Fire Pumps
- a. State TWO automatic initiation signals for the pump, INCLUDING BE1 POINTS. ( 1. O)
~
- b. STATE ALL locations WHERE the pump can be started MANUALLY. (1.0)
- c. HOW does the control system respond when the engine fails to start after 90 seconds f rom the first crank cycle? (0.5)
- d. The alarm "DIESEL FIRE PUMP 1 ROUBLE" will annunciate if one of four causes. has occurred. STATE 1HREE of these causes. (1.0)
QUESTION 2.04 (2.00) 51 A1E t he pur pose /f unction of the f ollowing components in the Stetson Power Battery Chargers, including SETPOINTS:
- a. Inservice Charger (1.0)
- b. Autosnatic Current Limi t control (1.0)
OUESIION 2.05 (1.75) 51A1E SEVEN of the eight containment isolation valves that automatically CLOSE when an isolation SCHAM occurs. (****a CATEGORY Z LON11NUED ON NEXT PAGE *****)
Pega 9
<3,._.EL9NI pgS19U,1Ng6UDIUQ_@OEEIy_@UD_EdgBGENQy ;
QY91EUS QUESlION 2.06 (2.00) The Reactor Depressurization System (RDS) has had' codifications installed. STATE the REASON for the following installed modifications. ,
- s. Are isolatiori transf ormer was installed in AC3 to isolate the fire pump start. circuitry from the RDS. (0.5) j
- b. CV-4184, RDS bypecs velve was rotated 180 degrees ( 2 reasons). (1.0)
- c. An air amplifier was installed in the instrument air supply to boost ,
control afi pressur e. (0.5) 4
. 6 OUESTION 2.07 (1.50) i STATE the 1HREE primary furictions of the Post Incident / Emergency '
Core Cooling System. . li OUEG110N 2.08 (2.50)
- Regarding the Feedwater Contrel System:
- a. State which input si griol (s) ar e prescut e compensated. (1.0)
- b. State the effect that rapid depressuri:ation will have on Steam Drum ,
level indication, tricluding the required operator action (s) f or level control. (1.5) 1 i t OUEB110N 2.09 (3.00) i
- a. 6141E the 1HREE sources of cooling water for the Emergency Diesel Generator, in their order of preference. (1.0) ,
[
- b. 14rief ly e>tpl ein how backup emergency power is obtained during a
- Losu of Offsite Power situation IF the Emergency Diesel Generator
- fails. Include the equipment identification, the location, how I power is provided to the bus (os), and the time allowed to have the equip.aent in service. (2.0) :
I i f I (***** CATEGORY 2 CON 11NUED ON NEXT PAGE *****) f r l _, - - - - - , _ - , - , _ , _ - , , . _ , . . . _ . - , _ . - . . - - l
.[q Is._ ELBUI_DE9198.INGLUDIN9_90EEIY_BNp_gdggggggy Peg 2 10 SYSIBd9 QUEST 10N 2.10 (1.00)
.S1A1E the purpose of the Emergency Condenser Vent Monitoring System.
l
-QUESTION 2.11 (2.25)
- a. State the prewbure at which the S'hutdown Cooling System is required to be placed in service, by design, including the reason why. (1.0) ,
- b. During a shutdown using the Alternate Shutdown System, des,cribe ;
which Shutdown Cooling pump is to be used and whure it receives power. l
. (l.25) i f i
OUES1!ON 2.12 (1.50) .
~
State the THREE functions of the Control Rod Drive System. 4 e d i s i V r l f I t . I F i I !, h l t i I (***** END OF CATEGORY 2 *****) 4 l 1 i i
; ---3---------_--
Pags 11 , . 13t__1NB189dENIS eUG_GQNIGQb@ 1 QUESTION. 3.01 (1.50) Briefly STATE the conditions under which the HIGH' CONDENSER PRESSURE scram is BYPASSED. , F I OUE9110N 3.02 (2.50) Regarding the Reactor Cleanup systems
- a. S1 ATE the three Cleanup putnp trips (setpoints not required).
(1.5)
- b. STATE the two conditions that cause a Cleanup Demin 1 rouble alarm (setpoints not required). (1.0) '
t
.UUESilON 3.03 - (1.00) .
The liquid poi son tank is supplied with heater s to maintain the liquid ' poison at (a) _,____ degrees F. The poison tank is also insulated so that, upon loss of the heaters, the solution will be maintained above saturation , tempuretur e f or (b) ,, ,,, _ ,, .. _h o u r s . d QUESIlON 3 04 (2.00) Bri ef ly expl ain how the water level is maintained in the shell side of the Emerguncy Condenser. Include normal and backup water supplies and the TYPE and LOCATION of the controls. (Val ve number NU1 required). QUEGTION 3.05 (3.00) r% g - Briefly EXPLAIN how rua or power is monitored by the nuclear-instrumentation system during a startup f.om Cold Shutduwn to 'ull power. INCLUDE the approximate effective power -ange of each instrument range and the automatic protective features of each rance, if applicable. (***** CAlEGORY 3 CON 11NUED ON NE AT PAGE *****)
or .j
.Page 12 4 3 _,1NSIBUMENJg_SUp_ggyIBgLS t
OUESTION 3.06 (2.60) briefly describe the sequence of events required for AUTOMATIC initiation of the Reactor Depressurization System. INCLUDE the required parameters, timers, and approximate instrument setpoints. QUESTION 3.07 (2.00)
- a. ST ATE how the fire protection system Deluge Isolation valve (CV-4101) responds to automatic initiation of the core upray system.
(0.5)
- b. Briefly EXPLAIN how the core spray system AU10MA11CALLY actuates.
. Include any INTERLOCKS, including SE1 POINTS. (1.5)
UUESTION 3.00 - (2.00) .
- a. CHOUSE the correct type of detector used ire the Process Liquid Monitor systems (0.5)
- 1. Geiger-Mueller detector
- 2. luokinetic Probe detector
- 3. Gamma Scintillation dcetector
- b. Answer the following TRUE or FALSE: ( 3 at 0.5 each)
- 1. the radiation levels of all five process liquid streams are recorded on a multipoint recorder in the Control Room.
- 2. lechnical Specifications do not limit procesc liquid radioactivity levels and associated releases.
- 3. the procets liquid monitor s only alarm when Technica) Specification limits are uncueded.
UUESTION 3.09 (3.00) Condonner hot well level is automatically maintained by the operation of three valves. IDENTIFY the THREE valves and BRIEFLY DESCRIBE each valve's function in level control. INCLUDE the LEVEL BAND and FLON PATH for each valve. ( 1.0 pt for each valve) (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
. ,- o .
?t__IUSISUMEU @_SUp_CgNIB9LS 'Pega 13 e
OUESTION 3.10 (2.50) Regarding the Reactor Cooling Water system:
- a. Upon receipt of a "Reactor Cooling Water Pumps Low Discharge
' Pressure" al arm, br s efly ST ATE the required condition (s) that must be satisfied for the second pump to automatically start. (0.5)
- b. After the second pump automatically starts, STATE how the pump
, is secured. (0.5)
IL c. STA1E how the makoup to the Reactor Cooling Water tank is maintained !' ( manually or eutomati cal l y) (0,5) and BRIEFLY EXPLAIN why (two f reasons). (1.0) OUESTION 3.11 (2.00) j 1 Regard 2ng operataan of the Cir cul ating Water pumps, conipl ete the following: a.~When the corstral handle of the Circulating Water pump is placed { to tho ______,,__.,__ position, the discharge valve of that pump will start-to .open. 10.5)
- b. When the di schar ge val ve has opened appron a matel y ______ degrees, the pump will stort. (0.5) l c. If the discharge "S10F'" pushbutton is depressed and then released
. with the pump running, the valve will (REMAIN OPEll, CLOSE, RETURN 10 FULL OPEN). ( 0. G) j d. If the discharge velve of an operatin0 pump is closed to less than i ______
degrees, the pump will trip. (0.5) i OUESilON 3.12 (2.00) l Rogarding the Alternate Shutdown System 4
- e. Driefly describe HOW power from the Emergency Diesel Gonter at or is transferred from bus 2B to thu Alternate Shutdown System, INCLUDING the l ocet t ore of the appropriate equipment /devicets). (1.0)
- b. S1 ATE the LOCATION and PURPOSE of the Main Alternate Shutdown Fused Disconnect Switch, Disc 1441. (1.0) i i
(***** END OF CA1EGORY 3 ****<> I
4t__EBQGEDUBES_ _UQBUGLi_GBUQBUGLi_gdERQENQY Paga 14
'000_B09196901G06_G991096 DUESTION 4.01 (1.00)
During a Recirculation pump startup, procedure SDP 29, "Nuclear Steam Supply System", directs the operator to keep the pump discharge shutoff valves, MO-NOO1A and B closed for two minutes f ollowing pump startup. 514tE the reason for keeping these valves closed for two minutes. QUES 11UN 4.02 (3.00)
- a. Usi rig the attached curve 15.b.D.3 from the Technical Data book and the following data, DETERMINE the noble gas activity being released from the stack in mC)/uct (millicuries per second). (1.0)
Chemistry reportn the average maximum Beta activity is >.55 Nev T ht. gross reading from,the normal range noble gas monitor R1 8327 1s 9x10E3 cpm (counts per minute).
- b. If a value determitied using the method of part a. was noted to be 60 millicuries per second, identify the IMMEDIATE operator actions required. (2.0)
OUES110N 4.03 (2.00)
- With the plant oper at i ng at 80% power, the f ollowing cymtoms are obcer vc d:
Fundwater flow exceeds steam flow Steam drum level is steadily increasing Reactor power is increasing
- e. Stat e the r equi r ed i mmedi at e oper ator actions. (1.0)
- b. What prncedure should be referred to to verify proper actions once t hre plant hee stabili::ed? (Title only required) (0.5)
OUE3110N 4.04 (2.00) Emergency Procedure EMF Q.3, Lojs # Rosctor Geol ant " , pr ovi dos TWO conditions when automatic rar r rs can be placed in the MANUAL made durano en emergency satuati r. 4TE these TWO conditions. (***** CATEGORY 4 CONTINUED ON NEXT PAGE ****+)
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St__ESQCEQUBEQ_ _UQSdOL1_6hugBd661_EUEBGEUQX Pcge 15 SUU_B00196001GOL_G9dIB96 QUESTION 4.05 (1.50) . Emergency Procedure EMP 3.3, "Loss of Reactor Coolant", requires the operator to isolate Instrument and Service Air to containment if the containment environment is degraded,
- a. State the containment pressure limit by which air isol ation must be completed. (.75)
- b. State the REASON why Instrument and Service Air are isolated. (.75)
UUESI10N 4.06 (2.00) Wa t h ther p1 ant operating a AY 1007. r a't{e, power, you note the fol1owing conditionu: Rec.ctor power increenes to over 120% Reactor p essure increaset to over 1400 psig No contr.1 rod niovement is observed Steam dium safety relief valves start lifting with associated alarms S'i n1 E F OUR r equi r ed IMMEDIAlE operator actions. OUESTION 4.07 (2.00) With the plant operating at 90*/. power, condenser vacuum is noted to be rapidly decreasing. STA1E FOUR of the five IMMEDIA1E operator actions 1 requared by UNP 2.24, " Lono of Condenser Vacuum". ( e*** CA1EGURV 4 CONT INUED ON NEx T F AGE * * * * * )
zj =; .2 : . Paga 16 ' St 'E8QGEEW6EE_ _UO6 debt _6EUQBU6bt_EME8@EUQY cNu_60919699190b_GOU1BQb OUES110N 4.08 (2.50) Federa; regulationc 10CFR 55.53 currently requires operators to maintain their license in an ACTIVE status. COMPLETE the followings
- a. To maintain AC11VE status, the licensee must actively perform the functions of an operator on a minimum of (1) .____ eight hour shifts or (2)______ twelve hour shifts per calendar quarter. (1.0)
- b. If your license were to become inactive, you would have to complete a minimum of _________ hours on shift under the direct supervision of an active 11 comme holder. (0,5)
- c. The onbhift timo noted in part b. above must include the completion of TWO specific tasks. Briefly STATE these TWO tasks. (1.0)
OUESIION 4.09 - (1.50) . State WHO is normally responsible f or managing the control room in the absence of the Shift Supervisor according to Administrative Proceduru 2.1.1, "Shift Operations". (0.75)
- b. WHO is authorized to ref use entry into or direct personnel to leavo the control room if their presence intor f er es with plant operetionc? (0.70)
OUESTION 4.10 (1.50) Per GOP-1, "Pl.<nt Startup From Cold Shutdown", during the approach to rated system pressure, three temperature limito are to be monitored. ST ATE the three temperature restrictions placed during plant startup. QUEG110N 4.11 (2.00) Entlowing the r e cei pt. of a r ee . tor ucrem, pr ocedur e ONP 2.31, " R ac a c t o r Scram", requiron the operator to verify EIGH1 immediate actions occur. STATE these LIGHT IMMEDIAlE actions. (***** CATEGORY 4 CONTINUED ON NEX1 P AGli * * * * * )
Pegg 17 St_,EBQGEQUBES_:_U9BdO6t_0B00BdOLi_EdgBGEUQY 000_B00196001906_GQNIBQL DUESTION 4.12 (1.00) SOP-1,"Reactor Operation", requires reactor power adjustments to be made by control rod movemont in the sequence provided in the Technical Data book. If it becomes necessary to deviate from the sequence, what action (s) must be taken? DUESlION 4.13 (1.00) SOP-0,"Reactor Shutdown System"., states that the system should be operated with full cooling water flow and the reactor flow throttled. S'I A~I E t.he r easori for this precaution. QUESTION 4.14 (0.50) A visitter comes on site and does not take the required Gener al Employee Training. Nr procedure,5.1, "Radiation Protection Policy", WHO can author i::o L a person to enter a HIGH RADIATION area? UUESI'10N 4.15 (z.00) Due to vacations and required training, you have been a k k red to wori. the tollowing nchedule neitt wook. Identify the overtime guidelines, as outlined in procedure 1.6, "Overtime Limitations", that you would be violating. Assume times DO NOT include TURNOVER and the plant is operating at rated power. Sunday 0000-2000 Monday 0800-1900 luesday 0800-2400 Wednesday 0000-0100 0000-1700 Thur sday 0000-2100 Friday 0800-2100 , Satur day 0800-1900 l (***** END OF CATEGORY 4 *****) (********** END OF L x AM I N Al lON **********)
- c :
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. t = ne v = s/c "*'k t)
- 2 Cycle efficiency = -
w = us s = v,e + hat E = aC a = (vg - y )/t g ,g ,-it EE = %=v 2 v g,v+,{ g , gg PE = mah . = e/t 3 = in 2/cq = 0.693/eg W = v&P-
*g(*ff) " (g,:){gv) .
az = 93ta, , (fg+t)b 6=imC,a7 , , 3 , g',-rx " ' 4 = UAAT g , g*,-yx - - w = wi e , 2 - 1 ,i,-x/ m F = F, 10 III- TV1. = 1.3/u BVI. e 0.693/u F = F, e" T _ - ;
'SUR = 26.06/T ,
T = 1.44 DT SCR = S/(1 - E,gg) sua = 26 f1'trh o Cx,= s/(1 - x,gg,) 3,,
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' * ~' aff' M = (1 . gg ,) f(g . g ,)
8 " IEeff'IIIEsff * #efflEeff st3 = (1 . g ,g)/g gg p= [1*/TK,'gg-) + [I/(1 + 1,gg ),T) , 1* = 1 x 10 seends
' = E(V/(3 x 1010) 1,ggA= 0.1 seconds"I
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VATtr. PARAMETA's Idf=Id g 2 R/hr = (0.5 CE)/d g,,,,,,) 2 1 541. = 8.345 1ha R/hr = 6 CE/d2 (g,,,) , 1 sal. = 3.78 liters N15crt.t. ant 005 coNytRSIO3 . . 1 ft3 = 7.48 set. Density = 62.4 lbs/ft 3 1 Curie = 3.? m 10 10g ,, Density = L as/ca l 1 ks = 1.21 1ha 1 hp = 2.54 4 10 STV/hr West of vat ortsutiou = 970 ttu/lba Heat of fusien = 144 Stu/lba 1 N = 3.41 1 100 5tu/hr
! Ata = 14.7 Psi = 29.9 in. Ig. 1 Stu = 778 f t-lbf 2 g inch = 2.54 cm 1 f t. H 2O = 0.4333 lbf /in T = 9/5'c + 32 C = 5/9 ('r - 32)
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b ANSWER- 1.01 (1.00)
- a. (1.0)
REFERENCE Dig Rock Point Reactor lheory Lesson Plan, Flusion Products Poisons l-section pg 46 K/A 292000 K1.07 (3.2/3.2), K1.OB (2.8/3.2) 292OO2K108 292OO6K107 ..(KA's) ANSWER 1.U2 (2.00) F C'r up power trenst ents the shor ter lived precursors ers dorninetit, and [ as reactivity in added, the shorter lived precursors give birth to i neutr ons wh!.ch wi 11 cause more fibeions sooner and increase the rate ( 1. O ', at which power i nc r e at,e s. , F or dow power transients, the rate of power change is limited to the rate of decay of the longeut lived precursor. 'lhey will continue to supply neutrons at a value dependent on the previous power level end (1.0) thus retard the rate of power decrease. REFERENCE i Big Rock Point Reactor 't heor y Lessore Plen, Recctor Kinetics section pg , 21 r
. K/A 292003 K1.36 (3.7/3.7) 292OO3K106 ..(KA's)
I l l r l ( I i l I h I i i (***t* CATEGORY ! CONI 1NUED ON NEX1 PAGE *****) 1 1 i
- ~_ . ___,,,m,
11__EB1U91ELEG_9E_N9GLE66_E9 WEB _ELONI_9EEB6119Na Pe9o 19 l' 'IUEBU99YN601GS2_UE8I_IBBNSEEB_0ND ELVID_EL99 s ANSWER 1.03 (2.50)
- a. Doppler (.34) ,is the first to add ..egative reactivity. The increaue in power level causes a rise in fuel temperature and a correspos. ding addition of negative reactivity due.to the doppler effect (0.5).
- b. Moderator lemperature Coefficient (.33) is the next as there is a time delay (fuel time constant) for the heat generation in the fuel pc.11et to reach the coolant in the channel and cause the temperature to increase (0.5).
- c. Void Coefficient (.33) is the laut as the moderator temperature has to increase to the point of saturatton beforo voidu are formed in appreciable quantity. (Also accept if there are no voids early in startup) (0.5)
REFERENCE Big Rock Point Reactor ThLory Lesson Plan, Coefficients of Reettivity section pg 28, 32, and 33 K/A 292004 K1.14 (3.3/3.3) 292OO4K114 ..(KA's)
- 1. 5 O HNEWER 1.04 (W
- a. DECREASE DECREnSE
. b.
- w INCREASE 1 - d tI&A d.
(4 at 0.5 each = 2.0) hEFERENCE Big Rock Point Reactor Theory Lesson Plon, Control Rods section pg 37 and 3G. K/A 292000 K1.09 (2.L/2.6) 292OO5K109 ..(LA's) (***** CATEGORY 1 CONTINt'ED ON NEXT PAGE *****)
,_.; 9 . - .. . ..
Pcgo 20
. 1z__EBJUGIELED_DE_UVGLEBB E9 WEB _EL601.9EEB911964
'IOGBd90YU601994_UEGI_I60NHEE6_809_EL91D_EL99 Af45WER 1.05 (1.50) .
- 1. Derating (continued operation at a lower but constant power level)
- 2. Coastdown (Load is allowed to drop while operating with all 'rceds out)
- 3. Feedwater temperature reduction (Results in decre sed plant officiency but allows maintenance of full turbine load provided licensed rcactor power is not exceeded)
'4. Encess co'.. flow (Reaching 100*/. power on a less then 100% flow control line b/ exceeding 100% cor e f l ow)
(any 3 at U.S each = 1.5) REFCREtJCC Dig Rock Point Reactor Thpory Lesson Plan, Operational Summary section pg 53 R/A 292002 K1.09 (2.4/2.6), R1.11 (3.2/3.3), kl.14 (2.6/2.9)
- 292OO7K114 292002K111 292OO2K109 ..(RA's)
ANSWER 1.06 (1.00)
- a. 1 (0.5), 3 (0. 5 )
- b. 2 (0.5)
REFERENCE Fig Rock Point Reactor Theory Lesbon Plan, Control Rods section pg 39 h/A 292005 M1.11 (2.4*/2.5*), K1.12 (2.6/2.9) 292OODU112 292OO5K111 ..(KA'c) ANSWER 1.07 (1.50) Per i pher al r od worth will increase (O.S). High Xe concentration to the center of the core (highest power before scram) will depress the t hermal neutron flux in thet region c ausing the relative neutron tlux in the peripheral bundles to be htgher, thus increase the relativeII.5) rod wor th a n per ipheral rods (1.0). (***** CATEGORY 1 CONTINUED ON NEAT PAGE *****)
7 ., Pcq> 21
..it_.EB1001 ELE 9_QE_UUGLE06_EQWEB_ELBU1_QEE6011 Qui IUEBU991UGU1Ges UEGl_1606SEEB_009_ELVID_Eb9W f.
REFERENCE big Hoci; Point Reactor Theory Lesson Plan, Fission Products Poisons section pq 45 h/A 292006 K1.07 (3.2/3.2) 292OO6K107 ..(KA's) ANSWER 1.08 11.SO)
- a. The heat flux et which the change of heat trensfer coef f icient / del ta 1 devt etes f rom a straioht line function.
- 0 7< -
T he locat flon at which si trannition froni nucleate boiling to film (O. b) botItno occurs.
- b. liie heat flux at wlisch DNH occurs. (ur that heat flux at which the host transter coeff1c1 ort in drast1cel1y reduced for any (0.5) further increase in delte 1.)
- c. The ratio of critical fieat flux to actual heat flun. (0.5)
REFERENCE Din Roci: Putnt Heat T r ensf er . 1 her mod enenit t.ti. . end Fluid Flow, t her n.el Ltmats wetion pg 8 > U/H 293000 K1.09 (3.0/3.2), bl.10 (2.9/3.0) 29300GK110 293OObK109 ..(MA's) ANSWER 1.09 (2.00)
- a. FALSk
- b. 1 RUE
- c. F f4LSE j
- d. FAlbE REFERENCE l
Dig Rock Foint Heat Tr nsfer, lhermodynamics, end Fluid Flow, Heat Balence sect 1on pg 15 L/A 29300~/ K1.13 (2.3/2.9) 293007K113 ..(k4's) (+++++ CATE60RY 1 CONlINUED ON NEXT FAGE *****>
lt__EB1991 ELE 9_9E_UUCLE0B_E9 WEB _ELONI_9EEBOI190, ecg2 22
- 'lUE609DXU00195a_UEDI_IBOUSEEB_00D_ELVID EL9H ANSWER 1.10 (2.00)
- 1. Slight core delta P
- 2. Blight reci rcul ation ru.mp del ta P
- 3. Temperature differentials between various points on the reactor veusel and the steam drum
- 4. Indicatiens on recit culation pump flowchar ts (4 at O.S each = 2.0) g g g ggg4 REFERENCE 1419 Roci: l'oint Heat Transfor, T her modynami c s , arid Fluid Flow, lhermal Analysis section pg 15
, K/A 293008 R1.37 (3.2/3.4) 203OO8K137 ..(HA's)
ANSWEH 1.11 . (1.00> 0I 7
- 1. Copper pleting on the n du r cIffuel clad
- 2. Coertruded zirconium (2 at 0.5 eette n .1. 0 ) iser(/
REFERENCE liig Ract Point Materi a) Science Lesson Plan, PCI section pg 7 ) K/A 293009 K1.31 (3.0/3.4)
- 293OO9h131 ..(KA*n)
ANSWER 1.12 (1.50)
- a. Coridensate depression or subcool i tig l
- b. Helps provide NPSH to condensate pumps to prevent cavitatiori
- c. Reducub plant efficiency - (this is additional heat ene r oy that must be provided by tne reactor)
(3 at O.D cach = 1.5) (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
o-.- ~ . Pcgs 23
, 12__EBIN91ELEW_9E_U99LEBB_E9 WEB _ELBUI_9EEB811962
- *1UEBd9DXUBDIGW4_UE81 1680DEEB_8BD_ELVID_EL9W
~
REFERENCE Property of Steams section pg 4 GE Academics Series, Heat Transfer and Fluid Flow, pg 4-59 and 4-60 K/A 293003 K1.16 (2.0/2.8), 293006 K1.10 (2.7/2.0) 293OO6K110 293OO3K116 ..(KA's) ANSWER 1.13 (1.00) (0.5 vach) increasow, decreasec HEFERENCE Dig Rock Point Heat Transfer, lhermodynamics, and Fluid Flow, FF-5, p9 7 K/A 291004 K1.06 (3.3/3.3) 291004K106 . ..(KA's) . ANSWER 1.14 (2.50) p = Foff - 1/V.off
= (1.0001 -
1)/1.U001
= 0.0001 dk/k (0.5)
- a. Reac t or period 1 = 1+/p (0.5)
= (1 x E-4 sec)/(1 x E-4 dk/k) 64 [/ g 6 fses)//x E f-4
= 1 sec og ,/ g o (0.b)
- b. i = (Duff ~ p)/ (Lp) (0.5)
= (.007 - 0.0001)/t(0.1)(0.0001)3
= 690 see w (0.5)
REFEREllCE lii g kor k Point Peactor t aeory Lesson Plan pg 14 and 20 K/A 292002 K1.11 (3.2/3.3), 292003 K1.OS (3.7/3.7) 292OO3K105 292OO2K ',11 ..(KA's) ANSWER 1.15 (1.00)
- 1) increase
- 2) power
(-++++ CATEGORY 1 LONTINUED ON NEXT FAGE *****)
g , - - - - - - - - - -
. 1.t _ _ E6199.1 EL-E S _9E _ UWG LE OS_ E9WE B_ Eb eUI _9ER B B119 th Pc9a 24 1Uh609D.tUDb1GS.s_UEDI_16BUSE5B 909_E69.1D_E69W i REFI'RENCE Tr an s 1 erit and Accident Analysis. " I n c.r eas e in Feedwater Flow", pg 32 295008Gv 1 295000K201 2c500GK302 s . . (k A 's) 4
(***** END OF CATLGORY t *****)
Pogo 25 L..._PbeUI DES 190_10GLUDIN9_EGEEIY_0ND_Et!EB9ENGY 81616t ' ANSWER 2.01 (1.50)
- 1. Loss of gerierator regulation
- 2. A voltago of >140 volts (will cause generator output trip) M Co sf cir vs g
- 3. Low voltage to the prime mover (will cause a generator output trip) g 3 0 0.5 pts each)
- q, 9pfpcy gnegg,z yp-p I' d #! M N REFERENCE ONP-2.35, Loss of Reactor Protection Motor benerator Sets 212000A201 212OOOA110 ..(R4's)
ANSWER 2.02 (2.25) N' t
- 1. Fump UFF (O.25): Punip i s OF F (O.25): Valvu is CLOSED (O.25)
- 2. Pump ON (0.25); Pump running (0.25); Valve is OPEN (0.25) {
- 3. Pump bypass (0.25); Pump is off (0.25); Valve is OPEN (0.25)
REFERENCE t t SDH, Ruoctor Cleanup Systeni pg 3 h 204000A401 . ..(M A 's ) 1 ANSWER 2.03 (3.DO) 4{
- 1. Firo heador pressure of 60 pnty (0.5) 0
- a. Y
- 2. Receives a start signal from 9DS (0.25) when steam druni level reachen -17 incheu below centor line. (0.25) ,yp
- b. Froni the RDS control corisole in the Control Room (0.Jef and f rom the control panel EU-9 in the Screenhouse. (0 - /4NF NMde s/~/ 4- O
- c. T he control util c.rese f ur t her craniing, lock i s.el f out of s cr s i c.e .
and sound an alarm. (0.5)
- d. 1. Lube oil pressuru failure
- 2. high cooltog water temperature
- 3. Datter1es disconnected r,r dead
- 4. Poinp control switch in "MANUAL" or "OFF" (any 3 di O.34 ftrst plus two t O.33 each, REFEhENCE SDM Fire Protection System pg 15 and 16 206000 GOO 8 286000A405 286000A301 ..(KA'u)
(***** CATEGORY 2 LONTINUED ON NEXT FAGE +++++)
*~
Pego 26
.Ez_sELBUI_ DEB 199 10GLWPIU9_EDEEIY_0UD_gtjESQEUQy SYSIEtJS At45WER 2.04 (2.00)
- a. Meinteins a coristant voltage across the batteries (0.5) and also supplios the dc loads up to it's rated load (50 ampo). (0.6)
- b. Protects, the char ger ego nst over current tripu(0.5) by limiting output (0.5) current (to 110% of rate REFEREf4CE SDt1 Station Powt r Syntune pg 95 263OOOGv 7 . 0%
- u)
ANSWER 2.Ob (1.75)
- 1. (1alV
- 2. Enclosure clean sunip tuolation valves
- 3. Enclosure dirty sump ipolation valves
- 4. Reactor and fuel pit drain isolation valven
- 5. Reactor cleanup resin slutce valvec
- 6. Cont a t nnierit verit I ati on supply end eyhaust velves
- 7. Ireated waste valve
- 8. Vent 11 at i on pr obe iboletion voives (any 7 e O.25 e,4ch) 9e $ Y/'A SS V8 l W /S 0 f.WR dd t/d ( y<. .
FEF EREt4CE SDt1 Eont at omerit Vesnel pg 4 223OOlh101 ..(h4'S) ANbWER 2.06 (2.00)
- a. To protect RDL from circutt faults that may occur to the diest1 4are pump circuitry. (U.S)
- b. 1) 'Io decrease the probebi1ity of the RDb valveb popping when CV-4184 1s opened.
- 2) heduce corr ost ori in the system. ( two @ O.5 pts t acts)
- c. Io enuaire va!vus will c1 cso ductog eri snadvertent blowdown. (0.5)
RE F E RENCE SDM fc.r Rub pg 15 end 16 210000h401 210Govh201 . 0;A's)
*'18000h404
(***** CATEGORY 2 cot 4T itJUED 014 tJEX1 PAGE *****)
Ex.. EbeUI_DE6100_lNGLUDIUQ SOEllY_660_EdESQEUGX Pc92 27 E191605 ANSWER 2.07 (1.50)
- 1. Prevent gross clad f ailure f ollowing a LOCA by providing malteup through core spray. (O.S)
- 2. Limit containment temperature rise and pressurn rise through the use of enclosure spray. (0.5)
- 3. Following a LOCA, to provide aniergency cooling water to the spent fuul pool. (0.5)
REFERENCE SDM Post Incident System pg 1 2090010004 ..(KA'u) ANSWER /.00 U/.bO)
- a. Steam drune l evel (0.5) and Steam flow (0.5)
- b. Level indication will go upscale (due to reference leg fiashing) (0.3)
Required actions are to tal:e manual control (0.5) and open the feed water regulating valve (0.5) REFF.kENCE BOP-16 F eedwat er bysten. , SDit f-eedweter systen, 294001L109 ..(Ks) ANSWER 2.09 (3.00) ,
- a. 1. House service water
- 2. Fire protection systen.
- 3. Doniest i c water bycteni (3 @ 0. 25 ecch pl us 0. 25 f or licted or der )
- b. A standby portable diesel gennrator (250 kw) is provided (U.5);
norniall y' Lept in the f enced in area of the doniestic water well house i (0.b): manually connected to bus 2B vna a 480/2#-00 volt transformer and bus tie line with breaters (0.5) required to be oper able 24 hours +ollowing a LOCA. (U.5) REFERENCE SDM Chepter 20. section 15 SDP 20 pera 2.2.3 264000LbOb 264000K407 264000K104 264000K101 . . O s s) t+**** CATEGORY 2 CONTINUED ON NEXT PAGE +++++)
Pc93 20
. 3 __ELOUl_9ES100 10GLUDIU9_E9 EELY _009_EUEB9EUGY
. *EYSIEUW 1 l
. 1
- l Ar45WER 2.10 (1.00) lo provide an al ar m (0.25) on high radiation levels (75mr/hr) (0.25) to give indication of an Emorgency Condensor Tube Leak (0.5)
REFEREr4CE SDM Chapter 34 pg i 207000K405 ..(MA's) AtJ5WLR 2.11 (2.20)
- a. 100+or -10 psig (0.0) The turbine seals (air ejectors) become ineffective and the bypass valve must be closed. (0. 5) tJo, 2 pump (0.5) power ced + r one e cable f rom penul P-29 inside b,
containment. (0.75) REFEREt4CE GDM Chapter 5 pg 1 snd 4 20L0016v10 20 GOO 1K201 " 205001K101 ..O A's) ANSWER 2.12 (1.00)
- 1. Ic anst rt or withdrow rods to r at su or 1ower power 1 ti response l
, f roin siunal c f r o<n the control room.
- 2. lo initieto e steam operation.
to provide cooling water flow to the drives. (3 at 0.5 pts each) 3. REFEREt4CE l 1 1 1 SDM Chapter 30 pg 1 2OlOO1L107 . . ( V A ' t. ) ( 201001L403 201001K100 l . i t L b l (***** Et4D OF CATEGORY 2 *****) l
Poge 29
. h__10E16WrJEUIE_6UD.G99169L6 2
AtJSWER 3.01 (1.50) (M09L)
- 1. When the reactor operationiswatch, S4, is in the refuel (.30) and bypass
- duicp tank positiono(.37). (.75)
- 2. When steam drum pressur e is below 500 +0 -50 psig. (.75)
REFERENCE SDM Chapter 1.3.6 212OOOK412 ..(KA's) AtJCWER 3.02 (2.50)
- a. 1. Undervoltage
- 2. Ove:r c ur r ent
- 3. H1gh tempor atur u (O. 'd each )
- b. 1. H1ch da f f t>r ent a al prensure
- 2. High temperature (0.5 each)
REFEREhCE Cleonop SDM Appendnu 11 204 Ovum 4v3 2040006401 ..O.A's) ANSWER 3.03 (l.00)
- a. 140-150 F (0.5)
- b. 40 plus or mi nus 2 hr t,. (0. 5 )
REFERENCE SDM Chapter 4 211000L403 .. O A's) ANSUER 3.04 (2.00) Weter level 4 or nnstiel fi11 and automatsc level control 1s obtanned from the duint nerai t ;:ed water system witn a condenser mounted level switch end a bolenosd operated control valve. (1.O) The bacLup supply as f e om the ftre header through a remote manually operated valve, operable from the control rcom or the Al t er net e Shutdown E4 u i l d i n y . ( 1. O ) (***** CATEGORY 3 CONTINUED ON NEXY PAGE *****)
Pega 30 j
. Et_,1NSISWdEUIS 00D_GQNlBQLS l
i REFERENCE l SDM Chapter 6, page 4 ' 207000K407 207000K403 207000K106 207000K105 ..(KA's) i I ANSWER 3.05 (3.00) g # i dy
- 1. Startup range channels 6 and 7 Source level to about 10E-4% of full l powar ( . 34) ; includes period meters that monitor reactor period of -100 l' set to sufinity to +10 sect.33): no protective f eatur es, other then a ,
short period alarm (+20 suc.)(.33). (1.0) ,
- 2. Intermediate rango Channels 4 and Sg range from 10E-5% power to.100% C power ( . 34 ) ; includes ported instrumentat. ion that alarms at +15 sec.(.33) l and scramb the reactor on a short period of +10 sec.(.33). (1.0) i
- 3. Power rango Channels 1, 2, and 33 range from 125x10E-8% power to (
125% power ( .34 ) : 19 range switches and meters that provide a scram at ! 9 6*/. o f motwr scale (38% power and 120% power, depend'ng on meter i scal e t .33) ; downuccle rod withdrawal blocits at 5% scale s . 33) . (1.0) ! i REFERENCE - i SDM Chapter 31 215004K402 215004K401 215003K401 210005K402 215005K401 i
..(KA's) )
i ANbWER 3.06 (2.Su)
)
(When 6tcam drum level reaches -4 inches this actuetes a containment j evacuation ala'cm): when steam drum level reaches -17 inches (0.5) both fir e pumps start (0.5) and the two minute timers start (0.5); when j at least one fire pump has been confirmed running ( 100 psig discharge l pressure) and a two minute timer has timed out (0.5) and reactor level ! reachen 2 feet 9 inches above the core (.5); auto blowdown ocenen. REFERENCE f SDM Chapter 18 2180000001 218000K403 ..(KA's) I I l l l l i (*t*** CATEGORY 3 CON 11NUED ON NEXT PAGE *****) l I
,v ,
y Pago 31
. Jt_,1HEIBWjEUIE_00D_GQUIP96S 9
ANSWER 3.07 (2.00) a. b. 1he The gelg (0.5) low reactor level of 2 feet 9 inches above the core [ (0.5) and reactor pressure about 200 psig (0.5) . This also opens the i isolation valves, but flow to the vessel does not occur until reactor l pressure gets down below fire main pressure of about 120 psig(0.51. l REFERENCE CDM Chapter 8, page B-2 and B-3 i 209001K609 209001K400 209001k401 ..(KA's) l [ ANSWER 3.08 (2.00)
- a. 3 (0.5)
- b. 1. True
- 2. F al se
- 3. Falce (3 at 0.5 ea.)
REFERENCE SDM Chapter !? 272OOOA101 272OOOG011 272OOOGOOO 272OOOGvO7 ..(KA's) ANSWER 3.09 (3.00) M.he up valve (0,5); normally oper ates to provide water to the hotwell from the Condensate Storage TanL(.25) operates to open at approg. 35 inches and closes again at approx. 35. 5 i nches ( . 20) . (1.0)
- 2. Reject valve (0.b); opens on increasing level of approx. 37.5 or 30.5 inches and rejects water from the condensate line to the LP heater to the Condensate Storage Tank (.25); the valve closes at approx. 37.6 inches decreasing level t.25). (1.0)
- 3. Fill valvuto.5); opens on low l evel of appron. 33 inches and peovides water f rom the Coridensate Storage Tank to the hot we11 ( . 25) : closes at approx. 34.5 inches (.25) (1.0)
REFERENCE SDM Chapter 15 attechment 15.1 256000Alu4 2b6000A403 2S6000K507 ..(KA's) (***** CA1EbORY 3 CONTINUED ON NE11 FAGE *****)
Pcgo 32
. 2t_,10SIEVUUUl&_00D_C991896h e
AtJSWER 3.10 (2.60)
- a. The second punip 's control switch must be in STAf4DBY. (0.5)
- 6. Manually taking the control switch to OFF. (0.5)
- c. Manually (0.5), to keep track of possible 3eakage(0.5) and note makeup ,
amounts to perform required sempling for potassium chr omate (i nhi bi tor ) (0. 5) . (1.5) REFEREtiCE SDit Chapter 7, pp 2 2050004411 205000A410 205000A200 ..(UA's) ANSWER 3.11 (2.00)
- a. closed
- b. 33 plus or minus 1 degree
- c. return to full open
- d. sWM w inttme- 1 degI , - 2 ~[ O 30 #OAN REFEREtJCE SDt1 Chepter 21, pg 21-6 293OO2A107 ..(hA's)
Y ANSWER 3.12 (2.00)
- a. Dy operetang the Emergericy D>ebel Generator Transfer Switch (TAS-1401)
(0.5) located in the D/G room. (0.5)
- b. Provideu isolation and fuse protection for ASD circuitry (0.5) located in 8) / Li room adjacent to the b/G tr.ansfer switch. (0.5)
REFERENCE EDh thepter 36 pg 3 295016410/ 2Vbv16L202 ..(F6's) l l [ f l l t I (***** END OF CA1EGORV 3 +++++) ! l
e c. . . _ - .
> St. ,E60GEDUBES_:_U9Bd6La_6EU9BdeL2_EdEB9ENGy Pcos 33 BUD _BBD196091GOL_G961696 e
ANSWER 4.01 (1.00) To allow time f or temperature equalization of the ioops. Will also accept to prevent a high flux scram due to a cold water ti'ansient. REFERENCE SOP-29 page O step 16,1r ansient and Acci dent Ana7 ysis Sect. 3.8 202001A113 ..(KA's) ANdWER 4 . O'.' (3.00) 2.3*&$ Ze}-b$ a . M;.:k1 H 10E3 mi cr ocur i es / sec ond ( . 75) or M-3ci mil 11 curies /second(1.0)
- b. Will accept all reasonable ainwers that recognise that the valuo given exceedu the offgas isolatfor criteria and required procedural plant shutdown criterial appropriate actions given by ONP-2.17 include:
- 1. Vuri f y or initiate of f gao isolation j 2. If offoaw monitors do not reflect the same high reading, isolete the Reactor building ventilation valves.
l
! 3. Close cppropriate containment ventilation valves (ref uel accident).
- 4. Attes.pt to i denti f y and isolate the source of leakago.
b at 0.4 eacht c1 ternate wording 4 5. Initiale a plant shutdown. 8jV accepted) G. VMuff .ffMCK. $AF6 k*N e f6LO .SN/ff 70 N/W $N& 4 REFERENL'E SC. 34 para 6.1.2 ONP 2.17 para 2.2,2,3, and 2.4 272000K402 272000G015 272000A405 ..(MA's) 4 272000G014 4 ANSWER 4.03 (2.00)
- n. 1. Trip one rectrculating pump
- 2. Place feedwater cont r ol in manual
- 3. Reduce feedflow to match steamflow (3 0 0.5 each)
- b. Loss of F eedwa t er (ONP 2.20) (0.5)
REFERENCE ONP 2.20 para 1.3 259002G014 259002A103 259002A102 25Y002A101 259002G015
..(KA's) i
(***** CATEGORY 4 CON 11NUED ON NEXT PAGE *****)
Va . .- . . EBOGEDUBED_ _U9Bd8La_eENQBdg(4_gdggggygy Pegg 34 l , .ft "BU9_B8DJ96991GOL_G991696 j
'!( l ANSWER 4.04 (2.00) i
- 1. Hisoperation inautomat]sceIerer sa confirmed by at least two independent 1 j
indications. (1.0)
- 2. Core cooling in assured and the procedures state specifically to do so.
f 1 (1.0) j REFEPENCE EMP 3.3, pagu 2, pr ecaution i t em 4 , 29dO31 GOO 7 ..(LA's) l t s i i ANSWER 4.05 (1.50)
- a. 10 pstg (+0,-1 poig) (.75) DA.Ie1ed i I
- b. To prevent overpressurizing the containment. (.75)
~
REFERENCE , t EMP 3.3 page 3. precaution item 9 ; 2960246007 ..(UA*n) l
^
l ANSWLR 4.06 (0.00) h
- 1. Manually scram the reactor g[CIM f
{
- 2. Trip both recirculation pumps
- 3. If the Channel 1 and 2 scram clarms are not ing trip the RPS i undervoltage breaters g
- 4. !( pressure stayb above 1360 pbig, inject liquid poison )
- 5. Stop cleanup systein f low ( any four at 0.5 each) p I
t
- REFSRENCE i
) i EMP 3.54 Anticipated 1ransients Wathout Scram, par a 1.0-1.3 t 1 295037G011 295037G010 29503/A104 29SO37A101 ..(hn'u) l l l 1 ; l I I 1 l f (***** CATEGORY 4 CONllNUED ON NEXT PAGE *****)
o . v -. ... - :. _ . . . - - - - ------- - : - - - - 1
.s I .-
Pcgs 3s 4 t__EB9GEDWBE9_:_N9Bb6L4_ePN9Bdek2_EMEBQEUGX
*eND_B69196991GOL_G2NIBOL t
ANSWER 4.07 (2.00) L Any f our of the following at 0.5 pts eacht
- 1. Check SJAE steam supply (250 psig) ,
- 2. Check circulating water pumps and valves for proper operation ;
- 3. Check offgas isolation valves open ,
- 4. Check turbine steam seal pressure normal 5..If the reactor has not scrammed, trip one recirculation pump l t
h/' REFERENCE , ONP 2-24 para 1.3 - 295002G010 ..(MA*w) t i i ANSWER 4.08 82.50) I L
- a. 1. seven 30.5) * !
- 2. five (0. 5) i
- b. 40 (0.5)
(0. 5 ) 'l
- c. 1. A complete plant tour
- 2. All required shift turnover proceduren (0.U) l i
i REFERENCE 10CFR 55.63 e cod i 294001A109 20100XGOO1 ..(KA's) i , f
?
ANSWER 4.09 (1.50) i
- a. Control Room Operator I or Reactor Operator (.75)
[
- b. Shift Supervisor . 70)
I REFERENCE Administrat i ve Procedure 2.1.1 para 5.L.1 and 5.7 f 294001A103 294001A109 20100AGOO1 ..(kA's) t h I I l t 2 (***** CATEGORY 4 CON 11NUED ON NE AT PAGE * * * * * ) i
e Pcga 36
. St_c,BB9CEDWB69_ _U9600b4_8phQBdO62_hdgSggggy T009.60D196991906 G901696 ANSWER 4.10 (1.50)
- 1. 100 degrees F/hr reactor temperature change
- 2. 160 degrees F reactor vennel temperature differential of any two points on the vessel
- 3. 100 degrees F steem drum temper ature dif f erential of any two points.
(3 9 0.5 each) fiEFERENCE DRP GDP-1, pg b 290002R506 ..O:4*s) ANSWER 4.11 (2.00)
- 1. Out of core instrumentation decay.
- 2. Scr am valved open .
- 3. Dump tank valves close
- 4. All control rods fully inserted
$. Vuntilation system supply and exhaunt isolation valveu close
- 6. Turbine trips
- 7. 116 OCb openn G. Sercond r od dr t ve pump t.t er t b (B at 0.25 e.<.)
REFERENCE ONP 2.31, page 2 201002G014 . 0: A 's) At1SWER 4.12 (1.00) lhe deviation must be cl erer e d t hr ough the reactor enu l tieer tv.0), eno documented on a temporary procedure chance form. (0.5) REFE REf 4CE SUP-1, pg 11 201001G013 ..OJA'6) (+++++ CATEGORY 4 CONTINUED ON NEAT PAGE *****) i . .. . . . _ _
EB99E99BEW_: U9BUOb4_6DUgBb863_gt1QQgNgy Pcga 37
. ds.
[GU9.609196991G06G9UIB96 0 ANSWER 4.13 (1.00) This subcools the* reactor water to the poirit where "flashing" of the reactor coolant wtIl not occur and cause vibration and water hammer in the heat unchanger tubes. REFERENCE SOP-5, pg 1 205000G010 ..(VA's) ANSWER 4.14 (U.50) The Chemi str y and Heel t h Phycies Super t ritendent REFERENCE Adm1 n procedur e S.1, pg 3' 294001kt03 ..(LA's) ANGWER 4.1D (2.00) i
- 1. No mor e than 16 hourb ti a 24 hour per1od tTuen)
- 2. No more than 24 hours t ri e 48 hour per l od ( M-T, T-W. 1-F)
- 3. At least 8 hours reset between sh14tb (1-W)
. 4. No more then 72 hours in a 7 day period
- 5. No mor e then 16 hours straight (Tues)
REFERENCE Admin proc.. 1.6, lec h Specs pg 103a 2940014103 ..(LA's) (***** END OF CATEGORY 4 *****) (********** END OF EXAMIN4 TION **********)
y , , TEST CROSS REFEAENCE Pcgo 1
- ~
Wf&I19H VALug BgEgegggg_
, 1.01 1.00 ZZZOOOOOO1 1.02 2.00 ZZZ0000002 1.03 2.50 2Z70000003 1.04 2.00 2Z20000004 1.05 1.50 ZZ20000005 1.06 1.50 ZZZOOOOOO6 1.07 1.50 2Z20000007 1.00 1.50 ZZZ0000000 1.09 2.00 2ZZ0000009 1.10 2.00 2220000010 1.11 1.00 ZZZOOOOO11 1.12 1.50 ZZZOOOOO12 1.13 1.00 Z2Z0000013 1.14 2.50 ZZ70000014 1.15 1.00 ZZZ0000039 24.60 2.01 1.50 Z2Z0000043 2.02 2.25 22Z0000044 2.03 3.50 Z2Z0000045 2.04 2.00 ZZZ0000046 2.05 1.75 ZZZ0000047 2.06 2.00 Z :' 70000048 2.07 1.50 Z Z Z Ov;"TO49 2.00 2.50 2ZZ0000050 2.09 3.00 Z270000051 2.10 1.00 22Z0000052 2.11 2.25 ZZZ0000053 2.12 1.50 ZZZOOOOO54 24.75 3.01 1.50 2ZZ0000015 3.02 2.50 2ZZ0000016 3.03 1.00 ZZZ0000017 3.04 2.00 ZZZ0000010 3.05 3.00 ZZZ0000019 3.06 2.60 ZZZ0000020 3.07 2.00 ZZZ0000021 3.08 2.00 ZZZ0000022 3.09 3.00 ZZZ0000032 3.10 2. 's0 ZZZOOv0040 3.11 2.00 ZZZOOOOO41 3.12 2.00 ZZZ0000042 26.00 4.01 1.Ou ZZZ0000023 4.02 3.00 ZZ20000024 4.03 2.00 ZZZ0000025 4.04 2.00 ZZZ0000026 4.05 1.50 ZZZOOOOO27 4.06 2.00 ZZZ000000G 4.07 2.60 ZZZOOvov.'9
r TEST CROIS REFERENCE Pcgo 2 OUhrrI199 _Y6!,UE BEEEBEUGE_
*y ,4.00 2.50 ZZZOOOOO30 e
4.09 1.50 ZZZ0000031 4.10 1.50 ZZZOOOOO33 4.11 2.00 2ZZ0000034 4.12 1.00 ZZZOOOOO35 4.13 1.00 2ZZ0000036 4.14 0.50 ZZ20000037 4.15 2.00 Z2Z0000030 25.50 100.7 e
J* U. MASTEls S. NUCLEAR REGULATORY COMMISSION COP--'Y SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _QIQ.BQCK_EQIUI _________ REACTOR TYPE: _BWB:QEl________________. DATE ADMINISTERED:_SalDB230________________ EXAMINER: .66 bet _Et_8t__.._________ CANDIDATE: ...________________.._____ INSIBUCIIQUS IQ_C86DIDeIEi Upe separate paper for the answers. Write answers on one side only. Staple question sheet on top of the answer sheets. Points for each quoction are indicated in parenthuses after the question. The passing 9rade requires at least 70% in each category and a final grade of at losot 80%. Examination papers will be picked up six (6) hours after tho examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY
_.VeLUE. _IDI6L __ SCQBE___ _Y6LUE__ _______.______CeIEQQBI_____________ t *&. s
.-Ol&L 2 . hhf'22 ...________ ..._____ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS 1Lt. O PLAN' SYSTEMS DESIGN, CONTROL,
_Ftrf0__ _Lhr49 ...._______ ________ 6. ANs INSTRUMENTATION PROCEDURES - NORMAL, ABNORMAL,
- T __ .2fTfC ___._______ ______._ 7.
EMERGENCY AND RADIOLOGICAL CONTROL C f.O ADMINISTRATIVE PROCEDURES, _itrfD._ .Streb ._____..... ._______ 8. CONDITIONS, AND LIMITATIONS ________% Totals isstas__ __________. g y, y Final Grade All work done on this examination is my own. I have neither given nor received aid. I . [ Y, - f i g / 8 m ,,,
).- d ate 5~ Signature
~~~~~~~~~~
y,6 .
~
M STER u.
I
- NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules applyt
- 1. Cheating on the examination means an automatic denial of your application and could result in more severe pensities. ,
- 2. Rcstroom 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 possibility of cheating.
- 3. Use b1cck ink or dark pencil gDly to facilitate legible reproductions.
- 4. Print your name in the blank provided on the cover sheet of the examination.
S. Fill in the date on the cover sheet of the examination (if necessary).
- 6. Une only the paper provided for answers.
- 7. Print your name in the upper right-hand corner of the first page of aggb ,
Doction of the answer sheet.
- 8. Consecutively number each answer sheet, write "End of Category _." as appropriate, start each category on a Ogg page, write gDIY 90 SDR Ridt of the paper, and write "Last Page" on the last answer sheet.
1.4, 6.3.
- 9. Number each answer as to category and number, for example,
- 10. Skip at least ibtti lines between each answer.
- 11. Separate answer sheets f rom pad and place finished answer sheets face !
down on your desk or table. t
- 12. Uce abbreviations only if they are commonly used in facility 111st319tg.
l 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 en answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTG OF THE QUESTION AND 00 NOT LEAVE ANY ANSWEP BLANK. )
- 16. If parts of the examination are not clear as to intent, ask questions of the RESTIURC only.
- 17. You must sign the statement on the cover sheet that indicates that the l uork 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.
l J
- 18. When you complete your examination, you shall:
G. Assemble youti examination es follows: (1) Exam questions on top. (2) Exam sids - figures, tables, etc. l (3) Answer pages including figures which are part of the answer.
- b. Turn in your copy of toe examiastion and all pages used to answer the examiration questions. >
- c. Turn in all scrap paper and the belance of the paper that you did -
not use for snswering 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 revoked.
I I I i l r i i
PAGE 2 ; II_INEQBl_QENWQLEeB_EQWEB_ELANI_QEEBeIIQNt_ELWIQ1t_6NQ IBEBdQQXSedIG1 QUESTION 5.01 (1.00) Tho reactor trips from full power, equilibrium xenon conditions. Four (4) hours later the reactor is brought critical and power level is maintained on range 5 of the IRMs for several hours. Which of the following statements is CORRECT concerning rod motion during this ; poriod?
- s. Rods will have to be withdrawn due to menon build-in. [
- b. Rods will have to be rapidly inserted since the critical reactor [
will cause a high rate of xenon burnout. j
- c. Rods will have to be inserted since xenon will closely follow .
its normal decay rate.
- d. Rods will approximately remain ss is as the xenon establishes its equilibrium value for the power level.
l i
- (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
l i
' PAGE 3 li__IWEQBI_QE NWGLE68.EQWEB.ELANI QEE8eIIQS&_ELVIQ1&_6NQ IHEBdQDIWeb1G1 QUESTION 5.02 (2.00)
The effective decay constant is the weighted everage of the decay constants for the six precursor groups. Typical values are 0.1 for up t pouer transients and 0.05 for down power transients. What is the basis for the varying values? 1 l k l t 4 l r l a b i i t i l i I i (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) f w
PAGE 4 It__IHEQBl_QE_WUGLEeB_20 WEB _EL881_QEE86I1QWi ELVIQ1t.$dQ 18EBdQQ1NedIG3 QUESTION 5.03 (2.50) . Ascume that the reactor is being started up from Cold Shutdown and a
<*od drop accident occurs early in the startup. Of the void, doppler, and temperature coefficiente, which will act first, second, and third to limit the rapid power rise? EXPLAIN YOUR ANSWER.
)
L i I t i [ l l F (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
5 1 $:._I W E Q B l_ Q E _N W Q L E 8 B _E Q W E B _ E L 8 HI _ Q E E B & Il 0 N tPAGE _ EL W lQ i t _ & IHEBdQQ186 DIG 1 QUESTION 5.04 (4 M \ .6 FILL IN.THE BLANK The reactivity worth of a single control rod will __________. (For oech statement below, indicate INCREASE OR DECREASE.)
- a. If the void content around the rod ,1NCREASES.
- b. If ths moderator temperature DECREASES.
t.- _ 18 :r -;dfw+*Wa4 eol-cod-is WI T H9AAWN, dd[CQ
- d. If Xe-135 concentration around the rod DECREASES.
4 Y l e (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
PAGE 6 54 _IHEQBX.QE.UUQLEeB_EQWER_tLeNI_QEEB&IIQWi ELUIQ1t_eWO IHE800DXWeblG1 QUESTION 5.05 (1.50) LIST three (3) operational methods that will lengthen the cycle of l plant operation. , l (***** s' AT E C O RY 05 CONTINUED ON NEXT PAGE *****)
PAGE 7 It__IHEQBl_QE_NUGLEeB_EQWEB_tL&NI_QEEB&I1QNt_ELVIQ1t_eND
. IWE8dQQ1Ned1G1 l
i 1 QUESTION 5.00 (1.50) , For each of the following types of control rods "a" and "b", choose those characteristics 1 through 3 which describe that type. (More , than one characteristic MAY apply to each type.) i Control Rod Tvoe i i
- c. Shallow Control Rod
- b. Deep Control Rod !
Characteristics
- 1. A one-or-two notch withdrawal can result in reverse power effect. [
- 2. Substantially affects radial power distribution.
- 3. Also known as Shaping Rods.
; i f
r i I s k l I t i t ; t i 4 L I f l (***** CATEGORY 05 CONTINUED ON NE,XT PAGE *****) f l
PAGE 8 It;_IWEQBl_QE_WWQLE8B_t0 WEB _tL8HI_QEEB8110Wi_ELW1Q1t_8WQ
, IHEBBQQ1W85101 QUESTION 5.07 (1 50)
HOW does xenon concentration affect peripheral rod worth following a ocram from high power and Why does this occur? 1 (***** CATEGORY 05 CONTINUE 0 ON NEXT PAGE *****)
PAGE 9 Atl_IHEQBl_QE_WWCLE8B_EQWEB_EL&WI_QEEB8110Nt_ELW101t_6NQ
. IBE800016651G1 QUESTION 5.08 (1.50)
Concerning heat transfer in a boiling water reactor, DEFINE the following terms:
- a. Departure f. rom Nucleate Boiling (DNB)
- b. Critical Heat Flux (eHF) ;
- c. Critical Heat Flux Ratio (CHFR) i I
i i i' i l i l i 1 i i t l L L i (se*** CATEGORY 05 CONTINUED ON NEXT PSGE *****) , ) I ,1
- r l
h
Ei; IHEQBY_QE_UWCLE68_EQWEB_EL8NI_QEE88IIQWi_ELWIQ1t_6SD PAGE 10 1 IHEBdQDINedIGS QUESTION 5.09 (2.00) A reactor heat balance was performed by hand during the last shift. Answer the following questions TRUE or FALSE.
- a. If the feedwater flow rate used in the heat balance calculated was LOWER than the actual feedwater flow rate, the the actus!
power is HIGHER than the currently calculated power. (0.5)
- b. If the reactor recirculation pump heat input used in the heat balance calculated was OMITTED, then the actual power is LOWER than the currently esiculated power. (0.5)
- c. If the steam flow used in the heat balance calculation was LOWER tr.en the actual steam flow, then the actual power is LOWER than the currently calculated power. (0.5)
- d. If the Clean-up return temperature used in the heat balance calculation was LOWER than the actual Clean-up return temperature, then the actual power is HIGHER than the currently 1
calculated power. (0.5) 1 4 (sess CATEGORY 05 CONTINUED ON NEXT PAGE *****)
PAGE. 11 I __INEQBI.QE.WWCLE&B.EQWEB_EkeNI_QEEB8Il0Nt_ELVIQ15.6NQ '
, IHEBUQQ1W851G1 QUESTION 5.10 (2.00)
List four (4) possible indications of natural circulation flow (no recirculation pumps running). 1 I e ] (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
PAGE 12 f It._IBEQBI_QE_WWCLE65_20WEF_tL6WI_QEEB&Il0Nt_ELUIQ1A_eWQ
. IBEB500XW851Q1 QUESTION 5.11 (1.00) f(oh he,p e tous types of fuel ;
EXPLAIN how barrier(i.e.,fuel differs i Iceded in the core. design 4hd/ rffB terial makeup) (1.0) 1 0N vu . i L I T L i i i i l I l (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) i i
600 PAGE 13 54..T.UEQB1 QE_UUQLE88 EQWEB.EL88I_QEEB&Il006_ELul01 _ 1BEBUQQ168d101 h QUESTION 5 12 (1.50) The attached Figure i shows a T S diagram for a steam cycle.
- a. What phenomenon is taking place between points 1 and 27
- b. How does this process benefit plant operation?
- c. What is the disadvantage associated with this process?
?
(***** CATEGORY 05 CONTIiiUED ON NEXT PAGE *****)
I [ t i ceWic. \ 7*'i pt gory + e d Vo?*e s ys
/
i T, . T / t
/ wet
' 1 Ste.am / 7*
/ \ /
/ N /
/ '
Ta - - r - __ _ _ . L._ 1 J~ j
- / 1 .
4 9 e S 3 i
- Figure :_
- T-S heyam 1
4 i
PAGE 14 52..IHEQBY.QE.NWCLE68.20 WEB.EL6WI.QEEB6IIQNt.ELVIQ15.68Q IWEBBQQ166 DIGS QUESTION 5.13 (1.00) When the flow rate through a centrifugal pump decreases, available NPSH (INCREASES or DECREASES) and required Choose correct responses. (1.0) NPSH (INCREASES or DECREASES). (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) O
II;_IMEQBX_QENUCLEeB_tQWEB_tLANIQtEB6IIQNt_ELul08t_6NQ PAGE 15 IMEB500XN851GI QUESTION 5.14 (2.50) Calculate the reactor period for the given conditions: Beff = .007
- s. Keff = 1.0001 for a prompt critical reactor
- b. Keff = 1.0001 for a delayed critical reactor (Show all work)*
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
PAGE. 16
$ .'.IBEQBI_QE. NUCLE 68.tQWEB.tkeNI.QtEBeIIQNi_ELVIQ1t_6NQ
, IHEBdQQINedici QUESTION 5.15 (1.00)
With the plant operating et 100% power the feedwater reguisting velve I ce1 functioning causing en INCREASE in feedwater FLOW. This causes core inlet subcooling to (INCREASE, DECREASE, REMAIN THE SAME). As e ! l result the reactor will trip on high (LEVEL, POWER, PRES $URE) early : into the event. (Two parts et 0.5 ench) f i l i I l
, i l
L f l I l I I i t l (***** END OF CATEGORY 05 *****) l J i
i PAGE 17 h.'_tLANI_1XIIEUI QE AISN.s CON'J80Lt 6ND.It'U8WUENIeI1QN l 1 l f QUESTION 6.01 (1.50) , l A loss of Reactor Protection Motor Generator Set 1 has just occurred. , ! LIST the three (3) possible causes for the Motor Generator loss. (1.5) , l l 1 l i l l i 1 l , l l ! t I L I i a 1 \ l l i i I ( l i t f i I (***** CATE30RY 06 CONTINUED ON NEXT PAGE *****) i i t I
PACE 18 65." tL6WI l!!IEdl QEIIANs QQWIBQLt.eND.INIIBWBENI6Il0N 9 QUESTION 6.02 (2.25) The Reactor Clean-up pump selector switch MS-6410 provides three modes of operation for the clean-up pump and valve CA-4042. STATE the three (3) modes of operation of HS-6410. Include pump and valve CV-4042 Otetus. (2.25) l l l (***** CATEGORY 06 CONTINUED CN NEXT PAGE *****) s
PAGE 19 ti. 2L6SI_IIIIEU1_QE11tWi_QQUIBQLa_eNQ.INIIBWUENI8IIQN I QUESTION 6.03 (3.50) , Answer the following questions concerning the diesel fire pump in the fire protection system: l List two (2) automatic initiation signals for the pump. l e. (INCLUDE SETPOINTS) (1.0)
- b. WHERE can the pump be started manually. (Provide all locations) ;
(1.0) ,
- c. What happens when the engine fails to start efter 90 seconds from the first crank cycle? (0.5)
~
- d. "DIESEL FIRE PUMP TROUBLE" will ennuciate s trouble storm it one of four causes has occurred. List three (3) of these causes.
(1.0) ; i l t t l l i l (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) i I i
1 ' PAGE 20 ti .EL6NI.111IEUI.QE119NA_QQNIBQL&_eNQ_INIIBudENIellQU QUESTION 6.04 (2.00) l STATE the purpose and/or function for the following componente in the i Stetton Power Bettery Chargere, and how they accomplish their l function.
- a. Inservice charger (1.0)
- b. Automatic current limit control (1.0) l l
l l (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) i L
PAst 21 42.:2L4WI.111IEul.0511tWoG0 NIB 0Lo.4WD.1W11MW5EWIAIlos QUESTION 6.05 (1.75) STATE seven of the eight containment isolation velves which automatically close when en isolation scram occurs. (1.75) (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
PAGE 22 6[.*.tL6WI.1XIIEdi.0E119Wa_00 NIB 0Lo.6W0.INII5WDENIAIIDW t QUE$ TION 6.06 & 0 0 T-- Lib The power range instrumentation has.three trip units. STATE when each of the trip units are actuated and the resulting control response.(3.0) I t 1 1 I
- i I
1 i f ! (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) I i l l t
PAGE 23 li._tLeNI.111IEdi_DE110NA.CQUIB0La_eNQ.INIIBWWENI&Il0N , 1 l QUESTION 6.07 (2.00) ! l l Tho Reactor Depressurization System (RDS) has had modifications ) instelled in its system. STATE the reason (besis) for the following installed modifications.
- 1. An isolation transformer was instelled in AC3 to isolete the fire pump start circuitry from the RDS. (0.5) l
- 2. CV-4184, RDS bypass velve was rotated 180 degrees l (2 reasons) (1.0) '
f
- 3. An air esplifier was installed in the instrument sir supply.(0.5) l l
l 1 i i i (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) , I l - - - . - . . _ -
PAGE 24 62.:.t L 6WI. IIII E U A .Q E 11 t W s.Q Q WI B Q L t.6WQ.1N SIB WUEWI611Q W t QUESTION 6.08 (1.50) i STATE the three (3) primary functions of the Post I nc ide nt /Eme rg e ticy Core cooling System. (1.5) , I I l I t i i
?
I s l t t l I i i t l l l t I t I t I (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) ! I t I l
PAGE 25 6A.'.2L&WI.IIIIEdl.QEllids-QQUIBQLA 8WQ.lWIIBWdENIsllQW , i QUESTION 6.09 (2.50) An0wer the following questions regarding the Feedwater Control System
- o. What input signals are pressure compensated? (1.0)
- b. What effect will e rapid steem system depressurtretion have on 3 the Feedwater Control System inputs? (0.5)
- c. What corrective action is required by the operator following e Rapid Steam System depressurizetion? (1.0) '
l i f I I i. I r I r i l I I i l l l [ (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
. l 46..tL&WI l!IIEul 9E119Wo GQUI30La 4WQ.INII8W5EWI&I1QW PAst 26 i i
i i
. QUESTION 6.10 (2.00) i WHAT Trio Functions are bypassed when the Mode Selector Switch ,
10 in the Refuel position? , (2.0) l r i l r [
- i r
I l 1 i i { l
?
t I t 1 ! t ) 4 1 1 l l 1 4 (***** CATEGORY 06 CONTINUED ON NEXT PA6E *****) i i
- - - , - - - - ,.--w,_.,,-..--.,,-..n_.,,,.nn-
._v----,.n__,
PA6E 27 6 .'_th6WI.111IEdi_Q 11985-QQUIBQLA 680.1911BudEWI61108 QUES!!ON 6.11 (2.001 An3wer the following rego ding the shutdown of the reactor and it's cubsequent cooldown. .
- 1. HOW is lockage past the inboard (cloGest to reactor coolant) 4HA CCCkA'#3 isolation valves detected? (0.5)
- 2. When cooling down the reactor with the mein condenser, EXPLAIN why it is necessary to close the turbine bypass valve ano discontinue this mode of cooldown prior to 150 psig. (0.5) 1 The Emergency condenser serves as a backup to the main condenser for reactor cooldown. How long con the Emergency condense * (0.5) function (remove heet) on a loss of A/C power?
4 What are the normal and beckup water supplies to the Emergency condenser 1 (Shell side) (0.5) l A i i (esses CATEGORY 06 CONTINUED ON NEXT PAGE
*****)
PAGE 28 it_'_EL68I_1IIIE01_DE11GNt_QQUIBQLt_88Q_INSIBudENI&Il0N QUESTION 6.12 (1.00)
'Accume 1 hour after a large break LOCA, the core spray syrstem is lost.
Diagnosis reveals a passive failure in the underground pipe which normally provides water to the core spray system. What provision (s) (1.0) hcve been made for this contingency? t
, (***** END OF CATEGORY 06 *****)
i
PAGE 29 ZE_'_EBQGEDWBE1_=_NQBdekt_eBNQBd8L&_EUEBGEdGY_eNQ B6DIQLQGIG8L_CQNIBQL QUESTION 7.01 (3.00) The core shutdown margin verification shall be performed prior to otart-up after any shutdown in which the system has cooled cufficiently to be opened to atmospheric pressure and if any of one of four situations exist.
- c. STATE the four (4) situations. (2.0)
- b. HOW is core shutdown margin verified? (1.0)
I
*****)
1 (***** CATEGORY 07 CONTINUED ON NEXT PAGE l
PAGE 30 ZA__EBQQEQUBEl_=_NQBd8LA_8ENGBd8LA_EdEBGENC1_86Q B8DIQLQQIQ8L_CQUIBQL QUESTION 7.02 (1.00) Tho refueling operation controls contain position interlocks. WHAT is the purpose of these interlocks as stated in Technical Specifications? (1.0) L l [ i ! i i l l-(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) _ _ _ _ _ _ . -_ _ i
PAGE 31 Zi__EBQQEQUBEl_:_UQBd8Lt_6BNQBd8L&_EdEBGEUQ1 8ND
,86019LQQlC8L_CQNIBQL QUE3 h 3 - (2.50) DMck STATE the four (4) symptoms that could indicate that there is an Anticipated Transient Without Scram present, as described in EMP 3.5A, "Anticipated Transients Without Scram". (2.0) l i
*****)
l ( : ** CATEGORY 07 CONTINUED ON NEXT PAGE l
PAGE 32 Z4 _EBQGEQUBES_:_NQBd8Li_8tNQBd8Li_EdEBGENGY_8NQ B8DIQLQQ1G8L_GQUIBQL
%3rgT!nN-7,c4 r 2. co t kG Ecorgency Procedure EMP-3.3, "Loss of Reactor Coolant", cautions on piccing automatic controls in manuel mode. Under what two (2) conditions should these controls be pieced in manual mode? (2.0)
(**'** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
PAGE 33 l Zi__EBQGEQUBEl_:_NQBd8Lt_8BNQBd8Lt_EdEBGENGX_8NQ B801QLQQ198L_GQUIBQL QUESTION 7.05 (1.50) Por GOP-1, "Plant Start-up From Cold Shutdown", during the approach to rated system pressure, three temperature setpoints are to be monitored. STATE the three (3) temperature restrictions placed during plant start-up. (1.5) (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
o PAGE 34 Zi _EBQGEQWBES_:_NQBdelt_8BNQBdelt_EdEBGENGl_8NQ
.,B&QlDLQQ1 GAL GQUIBQL QUESTION 7.06 (2.00)
Following receipt of a reactor scram, procedure ONP 2.31, "Reactor Scram", requires operator to verify WHAT eight (8) immediate actions cuot be verified? (2.0) e I l i l l l l l (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) i L
PAGE 35 ZA__EBQCEQWBES_=_NQBdeL&_eBNQBdekt_EdEBGENG1.88Q B8DIQLQGICeL_GQWIBQL QUESTION 7.07 (2.00) Anawer the f ollowing questions with regard to ONP-2.7, "Mispositioned Control Rods":
- c. A mispositioned control rod can result from what three (3) things? (.75)
- b. What are the SYMPTOMS that a control rod withdrawal out-of-sequence may have occurred? (1.25) i
(***** CATEGORY 07 CONTINUE 0 ON NEXT PAGE *****)
- .~ .
PAGE 36 2t__EBkCEQWBEl_=_NQBdelt_etNQBd8Lt_EMEBGENGl_eNQ
,BeQ10LQGIC8L GQNIBQL QUESTION 7.08 (1.00)
Rocctor power output adjustments will be made by control rod movement in the sequence prescribed in the Technical Data Book. If it becomes necessary to deviate from the prescribed sequence, what action must be done? (1.0) f l 1 (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
- ' .~
PAGE 37 Za__EBQQEQWBEl_ _NQBdekt_8BNQBdekt_EMEBGENQ1_&NQ
.B&Q1QLQQlC8L_GQUIBQL QUESTION 7.09 (1.00)
Whenever performing work that requires the use of hand tools over en epon reactor vessel, HOW ore the tools accounted for? (1.0) (2 methods) (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) ) ,
PAGE 38 2t__EBQQEQWBEl_:_NQBd6L&_8BNQBdeL&_EdEBGENQ1_88Q
'.BeQ10LOGIQ8L_QQUIBQL QUESTION 7.10 (1.00)
SOP-5, "Reactor Shutdown System", states the system should be operated eith full cooling water flow and tne reactor water flow throttled. What is the reason for this precaution? (1.0) (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
PAGE 39 . Zi__EBQGEQUBEl_:_NQBdeLi_8BNQBd8L&_EDEBQENQ1_68Q l
' .B&QIQLQQIG8L GQNIBQL l
QUESTION 7.11 ( .50) A visitor comes on site and does not take the required General Ecployee Training. If authorized by the Chemistry and Health Physics Superintendent, may this visitor be allowed to enter a high radiation arom? (YES or NO) (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
~
PAGE 40 Zi_$_EBQGEQUBES__NQBdeLi_8tWQBd8Lt_EdEBEENGl_eNQ 38010LQQIGAL_GQUIBQL QUESTION 7.12 ( .50) Big Rock Point has several locking doors which are used to control Gocess to high radiation areas which may not be opened from the inside. What precaution (s) should be taken when entry into these cross are required? (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
PAGE 41 Z4__BBQGIOUNEl_:_dQBd8kt_8BNQBd8ks_EdEBQENG1.88Q 88Q1DLQQ1G8L_GQUIBQL QUESTION 7.13 (1.50) WHEN [0.50] and under WHAT conditions (1.03 may a Shift Supervisor authorize en entry into a newly identified airborne activity eres which does not have a preexisting RWP, per Administrative Procedure 5.5, "Radiation Work Permit". i (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
PAGE 42 Zt__EBQCEDUBE1_=_NQBd8Lt_8BNQBd8L&_EdEBGENCY_6HD
,88QIQLQGIG8L_GQUIBQL QUESTION 7.14 (1.50)
During en emergency situation, you decide that it is necessary to
.- oxceed the 10 CFR 29 dose limits. WHAT three (3) items should be kept under consideration when choosing the rescue personnel per EPIP-4A "Site Emergency Director". (1.5) 8 i
i i I
! l I
i I i l l I (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) i i I
PAGE 43 Zi__EBQQEQWBE1_ _NQBd6Lt_eBNQBd8L&_EdEBGENQ1_8NQ
,88Q1QLQQIQ8L_QQUIBQL QUESTION 7.15 (M) QOD)
Answer the following questions in regard to the Liquid Poison System:
- a. When is the liquid poison required to be injected? (include i requirements from all thcae procedures dealing with liquid poison ,.
inj ect ion, i.e. SOP 4, ONP 2.21, Enr 4 . ii G+rt)(l.C)
- b. When can the liquid poison injection be terminated? (Include both inadvertent and intentional initiation conditions.) (1.0)
')C Y ,
i l 1 I (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
e PAGE' 44 l Zi__EBQQEQUBEl_=_NQBdelt 6HNQBdelt_EdEBGESQY_860
.B&Q19LQE10&L_QQUIBQL
. i QUESTION 7.16 (1.50)
Oporating requirements in GOP-2, "Stert-up from Hot Shutdown", limit tho average rate-of-change of reactor power during power operations. STATE the rate-of change limits and their respective power levels. (1.5) I (user
- END OF CATEGORY 07 *****)
PAGE 45 Ai_'_8DDIN11IB8IIVE_EBQQEQWBES&_CQUQ1110 Nit _eSQ_LidII6IIQH1 QUESTION 8.01 (2.00) Match the following Emergency Classifications (a - d) with the cppropriate description of that event (1 - 4). (2.0)
- a. General Emergency 1. Events which involve actus1 or imminent substantial core degradation or melting with potential for loss of containment integrity. ,
- b. Site Emergency 2. Any condition that involves an actus1 or potential substantial degradation of the level of safety of the plant.
- c. Alert 3. Events which involve likely or
- actus1 major failures of plant functions needed for the protection of the public.
- d. Unusual Event 4. Any station reisted event which indicates a potentisi degradstion of the level of safety of the plant, but which is not likely to affect onsite personnel or the public or result in radioactive releases requiring offsi'e monitoring.
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
PAGE 46 Ai__8001N11188IIVE_EBQQEQUBEli_GQNQ1Il0Nii eND_L10116IIQN1 I QUESTION 8.02 (1.50) The Shift Supervisor is responsible for determining t instelling a Jumper, Link or Bypass (JLB) will have or. .Ifected cc ponent and system operation. The JLB is considered a modification Ond will be controlled by Administrative Procedure 2.1.4, "Plant i Status and Equipment Control", if it meets what three (3) conditions. j i i (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
.; 1 o .
I PAGE 47 A4__eQdIN11IB&IIVE_EBQQEQUBEft_GQNDII1QNit_eNQ_ lid 1IAIIQN1 QUESTION 8.03 (1.50) Anawer the following question per Administrative Procedure 2.1.4 "Plant Status and Equipment Control":
- o. The requirement to have a second individual verify an equipment / system lineup should be waived under what two (2) conditions? (1.0)
- b. What should be done in lieu of the independent v'erification to verify operability? (0.5)
(ess** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
PAGE 48 A'i__eQUINISIBeIIVE_tBQQEQUBEli_00801I1981t_8N0_L1d118I10N1 QUESTION 8.04 (2.00) Duo to vacations and required training, you have been asked t's work tho following schedule next week. Review the schedule and identify the overtime guidelines of BRP Administrative Procedure 1.6, "Overtime Licitations", that you would be violating. (NOTE: All times exclude Shift turnover time.) Assume the plant is operating at 100% power. Sunday 0800 - 2000 Monday 0809 - 1g00 Tuesday C800 - 2400 Wednesday 0000 - 0100 0800 - 1700 Thursday 0800 - 2100 Friday 0800 - 2100 Saturday 0800 - 1g00 r (s**** CATEGORY 08 CONTINUE 0 ON NEXT PAGE *****)
PAGE 49 A t _ _ eQulb 1 S I B 8IIV E _ E B Q Q E D W B E 14._G Q U Q 1IIQ N 14._eU Q _L id lI&IlQ81 QUESTION 8.05 ( .50) (TRUE or FALSE) At the end of each day, the operating records, including ALL 24-hour
'rocorder charts, are assembled and checked for completeness by the A-Shift Supervisor.
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
)
PAGE- 50 St_.eQdIN11IB8IIVE_EBQGEDWBEft_QQUQ1IIQNit_8NQ_ Lid 1IeIIQN1 QUESTION 8.06 (1.00) Shift Operating Personnel shall follow approved procedures in the Operating Manual at all times with several exceptions. What are THREE instances / situations where deviations from operating proceoures are , allowed per procedure 2.1.2 - Operations Documents ? . i l I (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
" v
PAGE 51 RA__8DdlN1118811VE EBQGEQWBEIA.GQNQll10NIA 6NQ_Lidil&IlQN1 QUESTION 8.07 (2.00) A Technical Specification Quarterly (92 days) surveillance was pCrformed on the Emergency Core Cooling S y s t eit. on January 1, 1988, and April 15 1988. The surveillance prior to January 1, 1988, was completed ninety (90) days before January 1, 1988.
- a. WHAT is the last MONTH end DAY the next quarterly su veillance con be performed on? (Show your work) (1.0)
- b. WHAT is the consequence of failing to perform o surveillance in the specified time intervel? (1.0) 1988 calender ettsched
] i 1 . ) l l (ese** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
1 1
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PAGE 52 54__8QdINIIIB8IIVE_2800EQUBElt_QQNDIIIQNit_6N0_LIMII8IIONI QUESTION 8.08 (2.50)
- o. WHAT is the minimum crew composition for power operation es identified in the Technical Specifications? (Include numbers required for licensed and non-licensed etc.) (1.5)
- b. Under WHAT conditions, for WHAT length of time, and HOW many positions may the shift crew complement be below ;'e minimum?
(1.0) ( (***** CATEGORY 08 CONTINUED ON NEXT PAGE ***<*)
PAGE 53 Bl _8 QUIN 1118811VE.EBQGEQWBEIA_QQUQlllQNit 88Q_LidlI611QN1
. (
QUESTION 8.09 ,-{-2,40-) D i G.tc. Any of the three power range flux monitors may be taken oJt of service : for maintenar.ce during reactor operation per Technical Specification ' Oporationsi Requirements section 6.1.5. If one nonitor is out of corvice, WHAT two (2) conditions or restrictions must be followed per ; f Tochnical Specif ications? (2.0) t i 1 1 l I I i t i' I i 8 I I i (eses
- CATEGORY 08 CONTINUED ON NEXT PAGE ***")
d
PAGE 54 Et__eQUINISIB8IIVE_ERQCEDUBESt_GQNDIIIQNSt_eNQ lib 1IeIIQNS QUESTION 8.10 (2.00) Duo to valve manipulations, the situation resulted in isolating a section of the Exterior cable penetration eres sprinkler system and the Interior cable penetration area firt hose stations for the last four hours . From a Technical Specifications standpoint this creates a problem. For the above situation, describe what actions should be taken per your Technical Specifications when the equipment was found(2,0) to be inoperable. i 4 o t + I 1 l J L 1 i i i i i (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) i
; .;* 55 PAGE 1 1..__8DdIN11IB8IIVE_280GEDWBEft_GQN91Il0 Nit _8NQ_ lid 1I8Il0Ni I .
QUESTION 8.11 (1.50) Section 4.1.2.b of Technical Specifications addresses Primary Coolant Rocirculation System operating requirements. Answer the following questions regarding this section
- 1. The primary coolant Conductivity has a Maximum and a Maximum Transient value. WHY is there a Maximum T.-ansient conductivity limit and WHEN can this limit be used. (1.0)
- 2. If this maximum Transient value is exceeded, what action (s) should be taken? (0.5)
{ r t l (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) '
..,7 PAGE 56 St_IeQUIN11IB8IIVE_EBQQEQWBElt_QQNDIIIQNit_880_LIMIIeIIONS QUESTION 8.12 ( .50)
(TRUE or FALSE) Wh0n purging the Generator, Hydrogen is purged directly by Air for the j purposes of personnel safety. l e i i r i P t l t i I i i , l (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) j r i
o PAGE 57 Bi__6Qd1W11IB8IIVE_EBQGEQWBEft_QQNQlIl0 Nit _860_LIdlI8110N1 QUESTION 8.13 (2.25) DEFINE the following:
- 1. CHANNEL CALIBRATION (0.5)
- 2. CHANNEL CHECK (0.5)
- 3. CHANNEL FUNCTIONAL CHECK (analog and bistable channels) (0.75)
- 4. SOURCE CHECK (0.5) i j
j (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
o PAGE 'i It__8Q51611IB6IIVE.EBQQEQWBEl&_CQNDIIIQNit_6NQ_LIdlI6IIQN1 QUESTION 8.14 (2.00) , Technical Specification 7.5.7 states "It shall be permissible to recove a control rod drive from the reactor vessel when ... the mode One control solector switch is locked in the "Shutdown" position. ... rod drive pun 3 shall be operating during the removal and reins rtion and during th time the control rod drive is outside the react <r." WHAT are the r esons for having the mode switch in "Shutdown" and one control rod drive pump running? (2.0) (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) 4
g,_ . . - - - - _ - _ _ _ _ _ _ _ _ _ - ._. e PAGE 59 it__AQUIN11IB6IIVE_EBQGEQWBEht_GQNQ1110 Nit _6NQ_ Lid 1IAIIQN1
. t QUESTION 8.15 f,1.50)
What are the Technical Specification beste for providing the core opray recirculation syetem? I I i 1 i I l c f L (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
o __ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ PAGE 60
- 89. 8DdlN1118&IIVE_EBQQEQUBEIA_QQNQlllQN14_8ND LidlI811QN1 QUESTION 8.16 (1.25)
What are the Technical Specifice. tion Basis for the cepecity of the Station bettery? l l I l t l-i ,1 l [ i I i 1
- I l i
(***** END OF CATEGORY 08 *****) (************* END OF EXAMINATION ***************)
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=CO' Volume
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A=A e A = AN O CQUVE851QNE: 3 1 ge/cm = 62.4 lbm/ft Density of water (20 C) = 62.4 lbm/ft 1 gal = 8.345 lbm 23 1 ft = 7.48 gal Avogadro's Number = 6.023 x 10 1 gal = 3.78 liters Heat of Vapor (H O) = 970 Stu/lbm 2 1 lbm = 454 grams Heat of Fusion (!CE) = 144 Stu/lbm 1 AMU = 1.66 x 10
~* grams e = 2.72 n = 3.14159 Mass of Neutron = 1.008665 AMU 1 KW = 738 ft-lbf/sec Mass of Proton = 1.007277 AMU i 1 KW = 3413 Btu /hr Maws of Electron = 0.000549 AMU 1 HP = 550 ft-lbf/sec One atmosphere = 14.7 psi a = 29.92 in. Hg 1 HP = .746 KW 'F = 9/5 'C + 32 1 HP = 2545 Btu /hr 'C = 5/9 (*F - 32) 1 Stu = 778 ft-lbf 'R = 'F + 460
-16 ' K = ' C + 273 1 NEV = 1.54 x 10 Btu
' ~ h = 4.13 x 10 M-sec 10 jg,,gOn,j,,, 1 W = 3.12 x 10 g g = 32.2 lbm-ft/lbf-sec c = 931 MEV/ AMU 1 inch = 2.54 cm C=3x 10 m/sec
-8 R l r = 0.1714 x 10 Stu/hr ft l
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- 17. Carbon Steel 25.00 1*/. Chrome Steel 35.00 Aluminum 118.00 Copper 223.00 Silver 235.00 W0ter (20 psia, 200 degrees F) 0.392 Steam (1000 pais, 550 degrees F) 0.046 Uranium Dioxide 1.15 Helium 0.135 Zircaloy 10.0 d15CELLONEQUS_1NEQ8dGIl0N:
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# /////////////////
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10 CFR 20 Appendix B !
- - - Table 1 Table !!
Gamma Energy Col I Col 11 Col 1 Col 11 MEV per Air Water Air Water Material Half-Life Disintegration uc/mi uc/mi uc/ml uc/ml
-6 ~8 Cr-41 1.84 h 1.3 Sub 2x10 ----- 4x10 ------
~7 ~3 ~
~3 Co-60 5.27 y 2.5 S 3x10 1x10 1x10 ' 5x10
~3D
~3 3x10 ~7
~
1-131 8.04 d 0.36 S 9x10 ' 6x10 1x10
-3 ~7 Kr-85 10.72 y 0.04 Sub 1x10 ----- 3x10 ------
~3 ~
Ni-65 2.52 h 0.59 S 9x10
~7 4x10 ~3 3x10 1x10 *
~32 6x10 ~3' 5x10 ~6
~
4 1x10
- Pu-239 2.41x10 y 0.000 S 2x10
~
~3 ~33 ~7 Sr-90 29 y ----- S 1x10 ' 1x10 3x10 3x10
~6 ~#
Xe-135 9.09 h 0.25 Sub 4x10 ----- 1x10 ------ Any single radionuclide with T which does not decay by alpha br 2 > 2 he ~Y -5 1x10 -10
~
3x10 9x10 3x10 ' cpontaneous fission i i 2 Average flux to deliver Neutron Energy (MEV) Neutrons per cm equivalent to 1 rem 100 mrem in 40 hours thermal 970x10'6 670 0.02 400x10 280 (neutrons) 0.5 43x10 30 -- 3--- x s e-c 10 24x10 17 cm Linear Absorption Coefficients p (cm-3) Energy (MEV) Water Concret e Iron Lead 0.5 0.090 0.21 0.63 1.7 1.0 0.067 0.15 0.44 0.77 1.5 0.057 0.13 0.40 0.57 l 2.0 0.048 0.11 0.33 0.51 i 2.5 0.042 0.097 0.31 0.49 0.30 0.47 I 3.0 0.038 0.088 i l l I 1
' PAGE 61
. It__IWEQBY QE_NWQLEeB_EQWER 2L&NI.QtEB8I1QSi_ELWIQ1t_8WQ
$,fMEBBQQ1Ned1Q1
-88/08/30-HARE, E. A.
ANSWERS -- BIG ROCK POINT ANSWER 5.01 (1.00)
- o. (1.0)
REFERENCE Big Rock Point Reactor Theory Lesson Plan, Fission Products Poisons ocction pg 45 K/A 292006 K1.07 (3.2/3.2), K1.08 (2.8/3.2) 292002K108 292006K107 ...(KA'S) ANSWER 5.02 (2.00) Fcr up power transients tne shorter lived precursors are dominant, and so reactivity is added, the shorter lived precursors give birth to noutrens which will cause more fissions sooner and increase the rate (1.0) at which power increases. For down power transients, the rate of power change is limited to the rate of decay of the longest lived precursor. They will continue to supply neutrons at a value dependent on the previous power level and thus retard the rate of power decrease. (1.0) REFERENCE Big Rock Point Reactor Theory Lesson Plan, Reactor Kinetics section pg 21 K/A 292003 K1.06 (3.7/3.7) 292003K106 ...(KA'$) 9
y , _ , PAGE 62 la_ IHEQBY.QE.WWGLE88_E0 WEB.2L6HI.QtEB8Il0Nt_ELWlQ1t_&WQ
-IHEBdQQXNedlG1 ANSWERS -- BIG ROCK POINT -88/08/30-HARE, E. A.
ANSWER 5.03 (2.50)
- e. Doppler (.34) is the first to add negative reactivity. The increase in power level causes a rise in fuel temperature and a corresponding addition of nogettve reactivity due to the doppler effect (0.5).
- b. Moderator Temperature Coefficient (.33) is the next as there is a time delay (fuel time constant'. for the heat generation in the fuel pellet to reach the coolant in the channel and cause the temperature to increase (0.5).
- c. Void Coefficient (.33) is the last as the moderator temperature has to increase to the point of saturation before voids are formed in appreciable quantity. (Also accept if there are no voids early in startup) (0.5)
REFERENCE Big Rock Point Reactor Theory Lesson Plan, Coefficients of Reactivity cection pg 28, 32, and 33 K/A 292004 K1.14 (3.3/3.3) 292004K114 ...(KA'S) ANSWER 5.04 L2<et)
, fs'
- a. DECREASE
- b. DECREASE
- e. IMCREASE
- d. INCREASE (4' at 0.5 each = Frt )
- J5 RSFERENCE tig hock Point Reactor Theory Lesson Plan, Control Rods secti^n pg 37 cnd 38.
K/A 292005 K1.09 (2.5/2.6) 292005K109 ...(KA'S)
' PAGE 63 li__IBEQB1_9E_HWCLE88_EQWEB_tL8HI_QEEB8IIgst_ELVIQis_8NQ JEEBdQQ1NedlG1
_88/08/30-HARE, E. A. ANSWERS -- BIG ROCK POINT ANSWER 5.03 (1.50)
- 1. Derating (continued operation at a lower but constant power leval)
- 2. Coastdown (Load is allowed to drop while operating with all rods out)
- 3. Feedwater temperature reduction (Results in decreased plant efficiency but allows maintenance of full turbine load provided licensed reactor power is not exceeded)
- 4. Excess core flow (Reaching 100% power on a less than 100% flow control line by exceeding 100% core flow)
(any 3 at 0.5 each = 1.5) REFERENCE Big Rock Point Reactor Theory Lesson Plan, Operational Summary section og 53 K/A 292002 K1.09 (2.4/2.6), K1.11 (3.2/3.3), K1.14 (2.6/2.9) 292002K109 292002K111 292002K114 ...(KA'S) ANSWER 5.06 (1.50)
- a. 1 (0.5), 3 (0.5)
- b. 2 (0.5) l REFERENCE tig Rock Point Reactor Theory Lesson Plan, Control Rods section pg 39 K/A 292005 K1.11 (2.4*/2.5*), K1.12 (2.6/2.9) 292005K111 292005K112 ...(KA*S)
ANSWER 5.07 (1.50) Peripheral rod worth will increase (0.5). High Xe concentration in thw center of the core (highest power before scram) will depress the thermal neutron flux in that ragion causing the coletive neutron flux in the peripheral bundles to be higher, thus increase the relative rod corth in peripheral rods (1.0). (1.5) REFERENCE Big Rock Point Reactor Theory Lesson Plan. Fission Products Poisons soction og 45 K/A 292006 K1.07 (3.2/3.2)
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'IHEBdQQIUedIG1 l
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. ANSWERS -- BIG ROCK POINT l 292006K107 ...(KA'S) ANSWER 5.08 (1.50) ;
- 0. The heat flux at which the change of heat tronafer coefficient / delta T deviates from a straight line function.
-OR-The heat flux et which a transition from nucleate boiling to film boiling occurs. (0.5)
- b. The hect flux et which DNB occurs. (or that heat flux et which the nest transfer coefficient is drastically reduced for any further increase in delta T.) (0.5)
- c. The ratio of critical heat flux to actual heat flux. (0.5)
REFERENCE Big Rock Point Heat Transfer, Thermodynamics, and Fluid Flow, Thermal Liaits section pg 8 K/A 293006 K1.09 (3.0/3.2), K1.10 (2.9/3.0) 293008K109 293008K110 ...(KA*S) ANSWER 5.09 (2.00)
- n. FALSE
- b. TRUE
- c. FALSE
- d. FALSE REFERENCE Big Rock Point Heat Transfer, Thermodynamics, and Fluid Flow, H e a't Solence section pg 15 K/A 293007 K1.13 (2.3/2.9) 293007K113 ...(KA'S)
PAGE 65 inu_IHEQB1_QE_NWQLE65 EQWE8_EL6NI.QEEB&Il0Nt_ELW101&_6HQ IMEBdQQ1Ned1G1
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ANSWERS -- BIG ROCK POINT ANSWF. 5.10 (2.00)
- 1. Slight core delta P
- 2. Slight recirculation pump delta P
- 3. Temperature dif f erentials between various points on the r e actor vessel and the steem drum 4 Indications on recirculation pump flowcharts .
(4 et 0.5 each = 2.0) g %Q.gc.Jrth yto (N iCMM RE7kRENpE Big Rock Point Heat Transfer, Thermodynamics, and Fluid Flow, Thermal Analysis section pg 15 K/A 293008 K1.37 (3.2/3.4) 203008K137 ...(KA*S) ANSWER 5.11 (1.00) fuel clad
- 1. Copperplatingonthejn de u(ffkeof
- 2. Coextruded zirconium'ipser{jk (2 at 0.5 each = 1.0)
(EFERENCE Pig Rock Point Material Science Lesson Plan, PCI section pg 7
./A 29300P K1.31 (3.0/3.4) 293009K131 ...(KA'S)
ANSWER 5.12 (1.50)
- o. Condensate depression or subcooling
- b. "alps provide NPSH to condensate pumps to prevent cavitetton
- c. Reduces plant efficiency - (this is additional heat energy that must be provided by the reactor)
(3 et 0.5 each = 1.5) REFERENCE Big Rock Point Heat Transfer, Thermodynamics, and F1 Lid Flow, Property cf Steams section pg 4 GE Academics Series, Heat Transfer end Fluid Flow, pg 4-59 and 4-60 K/A 293003 K1.16 (2.8/2.8), 293006 K1.10 (2.7/2.8) 293003K116 293006K110 ...(KA'S)
PAGE 66
- li _IBEQBY_QE.NWQhE85_EQWEB ELeNI_Q2EB611QNi ELW1016.8NQ 1HE50001680101 ANSWERS -- 810 ROCK P0 INT -88/08/30-HARE, E. A. .
ANSWER 5.13 (1.00) (0.5 each) t r.c r e as e s , decreases REFERENCE Big Rock reint Heat Transfer, Thermodynamics, and Fluid Flow, FF-5, pg 7 K/A 291004 K1.06 (3.3/3.3) 291004K106 ...(KA*S) ANSWER 5.14 (2.50)
.po Keff - 1/Keff o (1.0001 - 1)/1.0001 o 0.0001 dk/k (0.5)
- e. Reactor period T = 1*/p (0.5)
= (1 x E-4 sec)/(1 x E-4 dk/k)
= 1 sec (0.5)
- b. T= (Beff - p)/ (Lp) (0.5)
= (.007 - 0.0001)/((0.1)(0.0001)) (0,5)
= 690 sec REFERENOE tig Rock Point Reactor theory Lesson Plan pg 14 and 20 K/A 202002 K1.11 (3.2/3.3), 292003 K1.05 (3.7/3.7) 292002K111 292003K105 ...(KA'S)
ANSWER 5.15 (1.00)
- 1) increase
- 2) power REFEREf4CE BRP Transient and Accident Analysis, "Increase in Feedwater Ficw", pg 32 K/A 295008 K3.02 (3.6*/3.95), k2.01 (3.7/:.u), G011 (J.9/4.18) 295008G911 295008K201 295008K302 ...(KA'S)
' PAGE 67 6t_.EL&WI_11SIEUI-QE110Nt_CQWI80La_8SD_IUlIBWWENI&IIQN
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ANSWERS -- BIG ROCK P0 INT ANSWER 6.01 (1.50) lo loss of generetor regulation \gg goht
- 2. a voltage of >140 volts (will cause generator output trio)
- 3. Iow voltage to the prime mover (will cause a generator output (3 at each = 1.5) N GA4pd 6, bfhb DhbLdE 3/
REFERENCE ONP-2.35 - Loss of Reactor Protection Motor Generator Sets K/A 21200u A1.10 (2.88/2.g), A2.01 (3.7/3.g), 212000A110 212000A201 ...(KA'$) ANSWER 6.02 (2.25)
- 1. PUMP OFF (.25) - Clean-up pump is off (.25), CV-4042 closed (.25) 20 FUMP ON (.25) - Clean-up pump running (.25), CV-4042 open (.25)
- 3. PUMP BYPASS (.25)- Clean-up pumo is off C.25), CV-4042 open (.25)
REFERENCE BRP Syster Description Manual, Reactor Clean-up System pg 3 K/A 204000 A4.01 (3.1/3.0) 204000A401 ...(KA'S)
PAGE 68 le .ELeWI.IIIIEdi DESIGN, QQWIBQLt.8ND.INIIBWDENI6IIQN
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ANSWERS -- BIG ROCK POINT I ANSWER 6.03 (3.50)
- a. 1. Header pressure of 60 psig (0.5) l
- 2. R4ceives a start signal from the RDS (0.25) when steam drum .
level reaches -17" below center line. (0.25) I
- b. Can be manually started f rom the RDS console in the control !
l room EtrF)dend)from the control panel EC-9 in ,the screen U ' 3.b))
. w$ . OECGcfL.,cufTA. CLUxffCu jaANW hosae (
- c. (Tte control will cease further cranking), lock itself out of service (and sound its alarm bell) (0.5).
- d. 1. 'ube oil prersure failure
- 2. ccoling water temperature too high
- 3. Bav<eries are disconnected or dead
- 4. pump control switch is in a "MANUAL" or "0FF" position (1 et 0.34, 2 at 0.33 each = 0.66) ,
f REFERENCE BRP System Description Manual, Fire Protection System pg 15 and 16 K/A 286000 A3.01 (3.4/3.4), A4.05 (3.3/3.3), G006 (3.7/3.6) 286000A301 286000A405 286000G006 ...(KA*S) j ANSWER 6.04 (2.00) ,
- a. Maintains a constant voltage across the station battettes (0.5) ,
and also supply the station de loads up to its rated current output (of 50 emps) (0,5)
- b. Protects the charger against overcurrent conditions by limiting j the maximum charger output current (0.5)(to approximately 110% of l the rated charge current), thus preventing tripping of the chargers input or output breakers (0.5) !
REFERENCE BRP System Description Manual, Station Power System (SPS) pg 95 l K/A 263000 0007 (3.3/3.5) j 2630000007 ...(KA'S) l I I i
PAGE 69 [
.6t'..tL6HI.IIIIEdl.QE119No.QQNIBQLa_eUQ_IN1189dENIellQU ANSWERS -- BIG ROCK POINT -88/08/30-HARE, Eo A.
ANSWER 6.05 (1.75) to Main steem isolation valve
- 2. Enclosure clean sump isolation valves
- 3. Enclosure dirty sump isolation valves
- 4. Reactor and fuel pit drain isolation valves
- 5. Reactor clean-up resin slutce velves
- 6. Containment ventilation supply and exhaust volves
- 7. Trested waste valve
- 8. Ventilation probe isolation valves (any 7 at 0.25 each = 1.75) g$ n caht 7740@ Yebt.
REFERENCE ORP System Description Manual, Containment Vessel section pg 4 K/A 223001 K1.01 (3.7/3.9) 223001K101 ...(KA*9) ANSWER 6 .\ 6 .(3.00) ddA Trip unit 3: Actuated when unit has a reading downscale of <5% of scale (O. ).circuit When 2 units are actuated, the control rod is opened, (no control rod withdraws!) (0.5). withdrawa Trip unit 2: A usted when the upscale reading is above 105% on the 125% scale o 34% on the 40/5 scale (0.5) This will cause a high-level f1 x alarm, (any 1 unit operation will actuate annuniciator "Meutron Flux-Hi" alarm) (0.5) Trip unit 1: Actuethd when flux level causes an up scale reading above 120% When one ofon thethe\125% scale trip units or about operates, 38% the on theannuncistor atetton 40% scale (0.5).
"Neutron Flux Hi Sc)em" alarms operates but neither protection channel operates un1 ss at least 2 of the 3 units actuated thus causing a reactor ser m (0.5).
REFERENCE BRP System Dercription Manue y Nuclear Instrumentation pg 7 K/A 215005 K4.01 (3.7/3.7), KA.02 (4.1*/4.2), A3.04 (3.2/3.2) 215005A304 215005K401 15005K402 ...(KA's)
PAGE 70
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ANSWERS -- BIG ROCK POINT ANSWER 6.07 (2.00)
- 1. To protect RDS from circuit faults that may occur in the diesel fire pump circuitry. (0.5)
- 2. a. Decrease the probability of the depressurization valves popping when CV-4184 is opened. (0.5)
- b. Reduce corrosion in the system. (0.5)
- 3. Boosted the control air (from 90 to 130 psig) to ensure the valves will close during an inadvertent blowdown. (0.5)
REFERENCE BRP System Description Manual, RDS pg 15 and 16 K/A 218999 K2.10 (3.1*/3/38), K4.01 (3.7/3.9), K4.04 (3.5/3.6) 218000K201 218000K401 218000K404 ...(KA'$) ANSWER 6.08 (1.50)
- 1. Prevent gross cled failure following Loss of Coolant Accident (LOCA) by providing emergency makeup through core spray.
- 2. Limit containment temperature /oressure rise following a LOCA through the use of enclosure spray.
- 3. Following a LOC A, provide emergency cooling water to the spent fuel pool.
(3 at 0.5 each = 1.5) REFERENCE BRP System Description Manual, Post-Incident System pg i K/A 209001 G004 (3.7/3.7) 2090010004 ...(KA'$)
PAGE 71
'ti..EL&WI 1XIIEUA.QE1198t_QQUIBQLa_6N0 IN1IBWBENI8I1QN ANSWERS -- BIG ROCK POINT -88/08/30-HARE, E. A.
ANSWER 6.09 (2.50) ,
- c. Steam drum level (0.5) and steam .!;w (0.5) are pressure compensated (to correct for density changes)
- b. The steam drusa level indication will go full scale high (0.5) "
(because of reference leg flashing).
- c. Take manual control of the feedwater control valve (0.5) and open it (0.5). .(The reference leg flashin9 will be er,uivalent to a high level in the steam drum and the feedwater valve will close if left in automatic.)
REFERENCE BRP System Description Manual, Feedwater System
$0P-16 Feedwater System P110 M-121 K/A 294001 K1.09 (3.4/3.8) 294001K109 ...(KA'S)
ANSWER 6.10 (2.00)
- 1. Low Steam Drum Water level
- 2. Recirculation Water line Valves Closed
- 3. Steam Line Backup Isolation Valve Closed (/Tji/V' f
- 4. High Condenser Pressure ce L o c C o d n n e r \' a u % ,3 (4 at 0.5 each = 2.00)
REFERENCE 9RP Tech Spec p9 55o JJf1 C)fntfr$. K/A 212000 K4.12 (3.9/4.1) SQ 9 212000K412 ...(KA*S)
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ANSWERS -- 816 ROCK POINT ANSWER 6.11 (2.00) to A bleed off orifice with a pressure switch located between the isolation volves. (0.5)
- 2. The turbine seals become ineffective. (0.5)
- 3. 4 hours (0.5)
- 4. Dominere11 red water pump (0.25)
Fire system (0.25) REFERENCE Final Hazarde Summary Report chapter 5, pg 8, 9, and to K/A 207000 K5.01 ((2.6*/3.0*), 0007 (3.8/4.0) 20700G007 207000K501 ...(KA's) ANSWER 6.12 (1.00) Capability exists to bypass the effected portion of the piping with a (1.0) fire hose. REFERENCE BRP Tech Spec Bases p 133 K/A 209001 A2.05 (3.3/3.6) 209001A205 ...(KA'S)
PAGE 73 Zi..EBQQEQWBEl.:.NQBdebt etNQBdels.EdGBGENQ1.600 B&Q1QLQ91G8L.GQUIBQL ANSWERS -- 816 ROCK POINT -88/08)30-HARE, E. A. t ANSWER 7.01 (3.00)
- e. 1. Fuel has been added and/or repositioned in a way which is not definitely known to reduce reactivity
- 2. Any steel channels have been replaced by 21rceloy channels during the shutdown
- 3. A control rod has been changed and presence of poison has not been verified
- 4. 35,000 mwd thermal have been generated by the plant since the previous margin demonstration.
(4 at 0.5 each = 2.0)
- b. It shall be verified by a demonstration that the reactor is subcritical with the most valuable reactivity worth control blade fully withdrawn (0.5), plus en immediately adjacent blade withdrawn to a position known to contribute 0.003 keff or more to the ef f ective multiplication (0.5)
REFERENCE BRP Tech Spec 5.2.2(b) K/A 201001 0005 (3.3/3.9') 2010010005 ...(KA'S) ANSWER 7.02 (1.00) Provent all motion with any of the refueling cranes (namely: jib cranes, and transfer cask winch), which are positioned over the recctor vessel whenever any control rod is not fully inserted in the core and the mode selector switch is in the "ref uel" position. REFERENCE 8RP Tech Spec pg 60 K/A 234000 K5.02 (3.1/3.7) 234000K502 ...(KA'S) l I t
PAGE 74
.Zc..tBQGEQWBEl_ .WQBd&L&_6BWQBd6LA.EdEBGEWQ1.6NQ B6Q10LQtIQ6L.QQUIBQL ANSWERS -- 816 ROCK POINT -88/08/30-HARE, E. A.
ANSWER 7.03 (2 00) 3 _r$ d C d
- 1. Reactor scram set p int exceeded (with or without annunciation)
- 2. All control rod driv s not et "00" position
- 3. Flux level still abov 4% power
- 4. Primary system pressur et or above 1360 psig (4 at 0.5 each a 2.0) ,
REFERENCE BRP EMP 3.5A, pg i ! K/A 295037 G011 (4.4*/4.7*) 295037G011 ...(KA's) ANSWER 7.04 (2.0 )
- 1. Misoperation inautomkticmode is confirmed by at least two independent process panameter indications. (1.0)
- 2. Core cooling is assured nd these guidelines state specifically !
to do otherwise. (1.0) i l REFERENCE BRP EMP-3.3 pg 2 l K/A 295031 K1.01 (4.6*/4.78), G00 g(3.7/4.0) , 1 2950310007 295031K101 ...(RA*S) ! l l ANSWER 7.05 (1.50) , I 1. 100 degrees F/hr. reactor temperature change. I l 2. 150 degrees F reactor vessel temperature differentist of any two ! points on the vessel. ! 3, 150 degrees F steem drum temperature differential of any two ! points. [ (3 at 0.5 each =1.5) l REFERENCE : BRP 60P-1, pg 5 K/A 290002 K3.06 (2.6/3.2) 290002K506 ...(KA*S) i I i
.y 7 - ---
PAGE 75 ZA tB9CEQWBEl WQBU6Lt etWQBdekt EUERGEWQ1.8WQ BAD 10LQ91G8keGQUIt9L
-88/08/30-HARE, E. A.
ANSWERS -- BIG ROCK POINT ANSWER 7.06 (2.00)
- 1. Out of core instrumentation decay (.25)
- 2. Scram volves open (.25), dump tank isolation valves closed (.25)
- 3. All control rods fully inserted (.25)
- 4. Ventilation System exhaust and supply valves close (.25) 50 Turbine trips (.25) and 116 0C8 opens (.25)
- 6. Second rod drive pump starts (.25)
REFERENCE ONP-2.31 pg 2 K/A 201002 6014 (3.8*/3.4*) 2010026014 ...(KA'$) ANSWER 7.07 (2.00)
- e. 1. Rod withdrawal out of normal sequence
- 2. Rod insert out of a normal sequence
- 3. An uncoupled rod which does not follow the drive as the drive is withdrawn (3 et 0.25 each = .75)
- b. 1. Increased wifs range monitor readings
- 2. Increased steem flow
- 3. Increased generator load
- 4. Neutron flux high annuniciator
- 5. Control rod not et its intended position (5 et 0.25 each =1.25)
REFERENCE 8RP ONP.2.7, pg i K/A 201001 G015 (3.6/3.8) 2010016015 ...(KA*5)
Zs..EBQCEQUBEl. NQB5eLt.8tNQBdekt EdERGENCX.eWQ PAGE 76 88D10LQGICAL.CQUIBQL ANSWERS -- BIG ROCK POINT -88/08/30-HARE, E. A. ANSWER 7.08 (1.00) Prcscribed sequence deviation must be cleared through the Reactor Engineer (0,5) and documented on a Temporary Procedures Change Form s'0.5) REFERENCE BRP SOP-1, og 11 K/A 201001 0013 (3.6/3.3) 2010010013 ...(KA'S) 4 l ANSWER 7.09 (1.00) All tools will be accounted for by one or both of the following cethodst
- 1. The use of a tool log sheet to keep track of tools used in a tool control area.
- 2. Certain common hand tools will be kept on a "controlled tool" board.
2 et 0.5 each = 1.0) REFERENCE ( BRP SOP-2 pg 6 K/A 234000 0013 (3.1*/3.3) 2340000013 ...(KA*S) ANSWER 7.10 (1.001 This subcools the reactor water to the point where fleshing will not occur and cause vibration and hommering of the shutdown best Ouchenger tubes. (1.3) REFERENCE
< BRP $0P-5, pg 1 K/A 205000 G010 (3.2/3.3)
- 2050000010 ...(KA'S) 1
PAGE 77 Zi_.EBQQEQWBEl_ _NQBeels 8tNQBdeLe_EMERGEWQ1.6WD B6019LQ9108L.QQUIBQL ANSWERS -- 816 ROCK POINT -88/08/30-HARE, E. A. ANSWER 7.11 ( .50) YES REFERENCE BRP Admin Proc 5.1, pg 3 K/A 294001 K1.03 (3.3/3.8) ; 294001K103 ...(KA'S) ANSWER 7 .1 ', ( .50) A door watch is required at these entrances. (to prelude locking an , individual into the high radiation area.) REFERENCE ORP Admin Proc 5.1, pg 4 K/A 294001 K1.05 (3.2/3.7) 294001K105 ...(KA*S) ANSWER 7.13 (1.50) If immediate action ta required (0.50) Provided:
- 1. Dedicated Radiation Technician coverage should be provided (if individual is advanced radiation worker there may be en exception from dedicated coverage) (0.50) ,
- 2. The entry shall be documented on an RWP after completion.
(0.50) ; REFERENCE BRP Admin Proc 5.5, pg i K/A 294001 Kl.03 (3.3/3.8) ; 294001K103 ...(KA'$) t e i 1 r 9
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ANSWERS -- B1G ROCK POINT . ANSWER 7.14 (1.50)
- 1. Rescue personnel should be volunteers or professional rescue personnel who are broadly familiar with the consequences of exposure.
- 2. Women capable of reproduction should not take part in these actions.
3 '. Volunteers above the age of 45 should be selected (3 at 0.5 each = 1.5) REFERENCE BRP EPIP-4A pg 5 K/A 294001 K1.04 (3.3/3.6) 294001K104 ...(KA'S) ANSWER 7.15 (-2r60) Gkt /, $
- o. 1. Anytime the Control Operator believes it is necessary to maintain the plant in a safe condition
?- A a y-t-i m e r e s c. t c, r subor-it-icality-cannot be as s u r e MueMo-fail _ure_of_ normal-shutdown mechanism
.-3,- -Anyt-imsm a cto r p r e s s u r e -is- g r e at e r t h e n-13 6 0 ps ig nnd no t dooressing LF at 0.5 each = 1
$ o.
Inadvertant _ A ytime
- b. 1.
- 2. Intentional - After all the poison has been injected (2 at 0.5 each = 1.0)
REFERENCE BRP SOP 4 sec 2.2, Ot4P 9-?i, EMP-375 X/A 211000 0014 (3.9*/3.9*) 2110000014 ...(KA'S)
n .- PAGE 79 24 _EBQQEQUBES_:_NQBdeL&_eHNQBd8Lt_EdEBGENQ1_8NQ 88D10LQE198L_QQNIBQL
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AESWERS -- BIG ROCK POINT ANSWER 7.16 (1.50)
<50 MW thermal / minute when power is <120 MW thermal
<20 MW therme1/ minute when power is between 120 and 200 MW thermal
<10 MW thermal / minute when power is between 200 and 240 MW thermal (3 at 0.5 each = 1.5)
REFERENCE BRP GOP-2, pg i BRP Tech Spec pg 14 K/A 201001 G006 (2.6*/3.5) 2010010006 ...(KA'S) l k i l l .
PAGE 80 At__8DdIN11IB8IIVE_EBQQEDWBEft_CONQ1Il0 Nit _8NQ_LIMII&IlQUE
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ANSWERS -- BIG ROCK POINT ANSWER 8.01 (2.00)
- o. 1
- b. 3
- c. 2
- d. 4 (0.5 each)
REFERENCE BRP Site Emergency Plan, Chap 3, pg 6, 7, 9, and 11 X/A 294001 A1.16 (2.98/4.7*) 294001A116 ...(KA*S) ANSWER 8.02 (1.50) . O. is temporary
- b. Disrupts the normal function of an electrical circuit or piping system
- c. Affected equipment / system will be in service or operable.
(3 at 0.5 each = 1.5) REFERENCE BRP Admin Proc. 2.1.4 pg 9 K/A 294001 K1.02 (3.9/4.5*) 294001K102 ...(KA'S) ANSWER 8.03 (1.50)
- a. 1. The individual would be subjected to significant radiation '
exposures. (0.5)
- 2. Extreme environmental conditions or other personnel hazards exist.
(0.5) A functional test should be performed. (0.5) b. REFERENCE BRP Admin Proc 2.1.4 pg 14 K/A 294001 K1.01 (3.7/3.7) 294001K101 ...(KA'S) ~ - _ - ___ _
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ANSWERS -- BIG ROCK POINT ANSWER 8.04 (2.00)
- 1. No more than 16 hours in a 24 hour period (Tues. 17 hrs.)
- 2. No more then 24 hours in a 48 hour period (Mon-Tues, Tues-Wed, Thur-Fri)
- 3. At least 8 hours rest between work periods (Tues-Wed)
- 4. No more than 72 hours in any 7 day period (84 hours Sun-Sat)
- 5. No more then 16 hours straight (Tues)
(5 at 0.4 each = 2.0) REFERENCE BRP Tech Spec pg 103a K/A 294001 A1.03 (2.7/3.7) . 294001A103 ...(KA'S) ANSWER 3.05 ( .50) TRUE REFERENCE. BR? Admir. Proc 2.1.2 p 3 K/A 294001 A1.06 (3.4/3.6) 294001A106 ...(KA'S) ANSWER 8.06 (1.00)
- 1. Prevent injury to personnel. (0.34)
- 2. Prevent demoge to equipment. (0.33)
- 3. Prevent damage to the environment. (0.33)
REFERENCE BRP Admin Proc 2.1.2 p 11 , K/A 294001 A1.03 (2.7/3.7) 294001A103 . . . ( K A ' S ,8
PAGE 82 Ela_CQUQlIl0 Nit _6NQ_LidlI6110N1 A. __6DdlN1*IB611VE_EBQQEQWd -88/08/30-HARE, E.
- i. .
NSWERS - SIG ROCK POINT R 8.07 (2.00) NSWER
- o. 3 25 x 92 = 299 days Surv #1 = 90 days ,
Surv #2 = 105 days days Therefore Surv #3 = 104 day the surveillance can be performed (1,0) July 28, 1988 is the last gg a surveillance constitutes(1.0)a
- b. Failure to meet the time interval the operability for requirement of the LCO.
failure to meet REFERENCE 8RP Tech Spec pg i K/A 209001 G005 (3.3/4.2)
...(KA'S) 2090010005 8.08 (2.50)
ANSWER
- c. 1 Shift Supervisor 2 Licensed RO 2 Non-licensed Operators (3 at 0.5 each =1.5) for a period of May be one less than minimum requirements (.25)in order to accommoda time not to exceed 2 hours (.25) crew members ft creuf 25) b.
composition (1.0) unexpected absence of on-duty (.25). shiftimmediate action la to within the t.11nimum requirenants REFERENCE 106 BRP Tech Speu pg (3.3/4.25) K/A 294001 A1.09...(KA'S) 294001A109
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#,NSWERS _- BIG ROCK POINT -88/08/30-HARE, E. A.
6 l f ANSWER 8.09 4W DAN 1
- 1. A trip on either of the two remaining monitors shall scram the i reactor (or the effecte channel placed in a tripped condition)
- 2. When maintenance is necess ry, no major changes in power level, flux distribution or contro rod pattern shall be unde (2 at 1.0 each = 2.0)
REFERENCE BRP Tech Spec pg 57
- K/A 215005 G006 (2.9 /3.8) 2940050006 ...(KA'S)
ANSWER 8.10 (2.00) Por Tech Spec., if a sprinkler system is not operable, estab1; 2 a continuous fire watch with backup fire suppression equipment for the unprotected ares (within one hour) (1.0), and with a fire hose station inoperable, provide an additional hose for the unprotected area at an cporable hose station (within one hour) and designate the new hose and otation as safety related (1.0). REFERENCE BRP Tech Spec pg 154 and 155 K/A 286000 A1.05 (3.2/3.2), G005 (3.1/3.9), K(.01 (3.4/2.6) 286000A105 286000C005 286000K401 ...(KA'S) ANSWER 8.11 (1.50)
- 1. The conductivity of the primary coolant tends to increase (0,5) temporarily after startups from cold shutdown.
Applies only to the period subsequent to a cold shutdown between criticality and 24 hours after reaching 20% power. (0.5) l
- 2. Reactor shall be shutdown. (0.5)
REFERENCE BRP Tech Spec C.1.2.b p 18 K/A 202001 G006 (3.08/4.1*) i 2020010006 ...(kA'S)
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/.WSWERS -- BIG ROCK POINT ANSWER 8.12 ( .50)
Folse REFERENCE Losson Plen BNA-138 p 6 K//. 294001 K1.15 (3.4/3.8) 294001K115 ...(KA'S) ANSWER 8.13 (2.25)
- 1. Adj ustments, es necessary, of the channel cutput such that it responds to the required accuracy. The calibration shall encompass the entire channel and includes the Channel Functional Test. (0.5)
- 2. The qualitative assessment of channel behavior during operation by observation and wherever possible comparison with an instrument measuring the same parameter. (0.5)
- 3. Analog Channels - The injection of a signal into the channel as close to the sensor as practicable (0.25) to verify operability including alarm and/or trip functions. (0.25)
Bintablo Channels - The injection of a signal into the sensor (J.25) (to verify operability including alarm and/or trip functions.)
- 4. The qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity. (0.5) 9EFERENCE ORP Tech Spec pg 2a Loarning objective #3 K/A 21500% 6036 (2.98/3.8) 2150050006 ...(KA'S)
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#NSWERS -- BIG ROCK POINT -88/08/30-HARE, E. A.
ANSWER 8.14 (2.00) Tho mode selector switch shall be locked ir, the "shutdown position b0cause this inserts a scram into the saf ety system (0.5) and prevents rod withdrawal. (o,5) One control rod drive pump shall be operating during the evolution of drive change out. This insures that adequate CRD header pressure is avc11able (0.5) event the possibility of any ortves frcm drifting out of the core while the scram is in (0.5). REFERENCE BRP Tech Spec section 7.5.7 pg 75 BRP LER 87-005-00 K/A 201001 G006 (2.6*/3.5) 2010010006 ...(KA'S) ANSWER 8.15 (1.50) Prevent excessive water buildup in the containment sphere. (0.5) (0,5) Provide for long-term, post-accident cooling. REFERENCE BRP Tech Spec Bases p 133 { K/A 209001 G006 (3.0*/4.0) 209001G006 . . . (K A' S ) ANSWER 8.16 (1.251 The s+ation bettery has adeouste capacity to curry normal loads (0.25) plus an assumed failure (locked rotor current) of the d.c lube oil pump (0.25) for 54 minutes (0.25) without the battery enarger (0.25) cnd still provide sufficient power for equipment required to operate after a LOCA. (0.25) REFERENCE BRP Tech Spec Basis p 149 K/A 262002 G006 (2.6*/3.7) 2620020006 ...(KA'S)}}