ML20151T504
| ML20151T504 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 01/22/1986 |
| From: | Rogers T, Wilson B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20151T493 | List: |
| References | |
| 50-280-OL-85-02, 50-280-OL-85-2, NUDOCS 8602100287 | |
| Download: ML20151T504 (93) | |
Text
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-/ o, NUCLEAR REGULATORY COMMISSION
[ g REGION 11 g 101 MARIETTA STREET.N.W.
* ' *j ATLANTA, GEORGI A 30323 s l ENCLOSURE 1 EXAMINATION REPORT 280/0L-85-02 Facility Licensee: Virginia Electric and Power Company Richmond, VA 23261 Facility Name: Surry Power Station Facility Docket Nos. 50-280 and 50-281 Written and operating requalification examinations were administered at the Surry Power Station near Gravel Neck, Virginia. . Chief Examiner: 3Arud. da, b '
ha2 ffl. Thodas Rogers 1 ~ Date ' 51gr.ed Approved by:- /Lo.u_ , t[2.L/74 BrucvA. Wilson, Section Chief Date Signed Summary:
-Examinations on December 4-6, 1985 - Operating examinations were administered to twelve operators; all of whom passed.
Twelve operators were administered written examinations; eight candidates passed. l
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REPORT DETAILS
- 1. Facility Employees Contacted:
- Bailey, J., Superintendent of Nuclear Training
- Christian, D. A., Superintendent of Operations
*Gardner, L., Senior Instructor, Nuclear
*Gardner, R. D. , Supervisor, Training (Simulator)
*Gwaltney, R., Senior Instructor, Nuclear
- Henry, W. G. , LORP Coordinator
*McCallum, H., Supervisor, Training-PS0
*Saunders, R. F., Station Manager
- Shriver, B., Director, Nuclear Training
- Attended Exit Meeting
- 2. Examiners:
Lawyer, L. L. , USNRC Picker, R., EG&G Idaho
- Rogers, T., USNRC
- Chief Examiner
- 3. Examination Review Meeting At the conclusion of the written examinations, the examiners provided Mr. H. McCallum with a copy of the written examinations and answer keys for review.: The following comments were made by the facility reviewers:
- a. Reactor Operator Examination (1) Question 1.12 Comment: It is unclear from the question that a more detailed explanation as that required by the answer key is needed for full credit.
Recommendation: Either accept stating that fuel pellet swell and clad creep cause the doppler only power coefficient to become less negative over core life for full credit or weight the response so that a much greater proportion of the credit is given on the answer key for that portion of the response.
Reference:
No additional reference is required. NRC Resolution: Recommendation is not accepted. Clad creep and fuel pellet swell do not directly change the doppler coefficient. Without the temperature relationship between clad creep / pellet swell and doppler, full credit will not be awarded.
2 (2) Ouestion 2.02, Part 3 Comment: The answer given on the answer key for the normal power supply to 4160v bus IA breaker control power is not correct. The 125VDC vital bus (selection "e") is the correct choice. Recommendation: Change answer key to require selection "e" for Part 3 of this question.
Reference:
System Description Manual, Volume 2, Station Service 4160 Distribution System, page 2-2 (Attachment S-1). NRC Resolution: Recommendation accepted. The answer key has been revised. (3) Question 2.04 Comment: The normal power source for the EHC System depends on turbine speed, which was not given in the question. Recommendation: Accept PMG or semi-vital bus as correct answers.
Reference:
Main Turbine and Support System Lesson Plan, pages 6.17and7.12(Attachments-2). NRC Resolution: Recommendation is not accepted. The additional reference clearly identifies the semi-vital bus as an emergency power supply and is used only when turbine speed is less than 80% speed. It is not reasonable to assume that the question is asking for the normal power source when the turbine is operating abnormally. (4) Question 2.07 Comments: (a) -There a,re five sources for AFW supply to each unit, whereas the antwer key only lists three. Since the question requested listing three, two additional acceptable selections need to be added to the key. (b) An additional method for aligning the AFW System without operator action outside the Control Room is available, if listed in part (a) of the question. Recommendations: (a) Add the following two possible AFW System supplies as two of the possible three requested: 1 300,000 gal. Condensate Storage Tank (CN-TK-2).
.3 t- 2 Aux feed cross-tie from the other unit (b). When the aux feed cross-tie from the other unit is listed as response to part (a), it should be included in part (b), and should be given credit.-
Reference:
EP-1, Step 45, response not obtained column (Attachments-3). NRC Resolution: Recommendations accepted. The Surry Power Station Steam Generator Auxiliary Feed System Description and Qualification Manual is incomplete. (5) Question 2.10 Comment: The question does not ask for the conditions under which the seal bypass is necessary. Recommendation: Delete the 0.5 point deduction for not including the condition under which the seal bypass is necessary.
Reference:
No additional reference is required. NRC Resolution': Recommendation is not accepted. The plant conriition specified on the answer key is the reason the seal bypass valve is necessary. (6) Question 2.12 Comment: The maximum listed blowdown rate of 70 gpm can increase to approximately 75 gpm before the blowdown PCV closes. Recommendation: Accept any value between 70 - 75 gpm as the correct response.
Reference:
Steam Generator Blowdown Cooler design change descrip-tion, see lines 20, 21, and 199 through 204 (Attachment S-4). NRC Resolution: Any value of 70 5 gpm is considered acceptable for full credit. (7)-Question 2.13 Comment: This question could be answered true or false based on the statement given. Hydrazine is added by two motor driven, positive displacement pumps to the Condensate System only; there-fore, false could also be chosen. Recommendation: Delete the question.-
Reference:
No additional reference is required.
4
~ NRC Resolution: Recommendation is not accepted. The statement is.
-always TRUE since no other system adds hydrazine to the feedwater system. The statement did not -include the point of connection to the secondary system.
(8) Question 2.16 Comment: The first part of the answer given in the key was essentially stated in the question which precludes an individual addressing it as part of the expected reply. The individual
. responding is going to address those items that should have been removed but were not.-
Recommendation: Allow full credit for replies similar to that of the answer key's second part.
Reference:
No additional reference is required. NRC Resolution: Recommendation is accepted, in part. The first part has been deleted and the question has been reduced in point value to 0.5 points. Partial credit will be awarded for contain-ment isolation and to initiate / isolate SI to the cold leg. The answer key has been revised. (9) Question 2.18 Coment: Expecting operators to recall from memory the design basis for this piece of equipment does not support performance based training. During-the conditions listed on the answer key, an operator can take no actions that would affect the performance of this equipment. Additionally,.the question should have stated that three responses are required so that the person evaluated does not have to guess what the grader is expecting. Recommendation: Delete the question. 4
Reference:
No additional reference is required. NRC Resolution: Although we believe this question is relevant knowledge for an operator, NUREG-1122 does not give it an importance F factor. The utility's recommendation is therefore accepted and~
- the question has been deleted.
l (10) Question 3.06, Part 1 Consnent: This question has two possible correct answers based on the assumed time after the detector failure. Immediately at time 4 of failure it goes to 72 S/M; or later it goes to O S/M. ' Assuming that those being examined would envision a longer period of-time
5 based on the length of time following the failure in which the procedure requires the control fuses to be removed is not a safe assumption. The relationship to the procedure could be discarded when it is noted that rods are in AUTO just prior to pulling the fuses, which is not in'accordance with procedure. Recommendation: Accept either (a) 0 S/M, or (e) 72 S/M. i
Reference:
No additional reference is required. NRC Resolution: Recommendation is not accepted. The operator would not know to pull the fuses unless he observed the failed indicator. With the indicator in the fully failed position, the power mismatch rate signal would no longer exist. (11) Question 3.12 Comment: Remote EHC load dispatch is not an allowable method of unit operation. Recommendation: Change answer key to: (a) Blocks auto rod withdrawal (b). Initiates load reference runback (c) Initiates load limit runback
Reference:
Student handout on permissives (Attachment S-5) NRC Resolution: Recommendation is accepted. The answer key has been revised. (12) Question 3.15 Comments: (a) TheLquestions did not specifically ask the power supplies to the RMP as required on the answer key. The blue lights being lighted, as given in the question statement, indicates that power is available to the RMP from each unit. (b) Committing to memory the significance of the amber and white light on the RMP is unnecessary, since it is readily apparent at the panel which unit is selected to supply power to the RMP. Recommendation: Accept power is avaijable from each unit, and one of the power sources has been selected to energize the RMP as satisfactory responses.
Reference:
No additional reference is required.
6 NRC Resolution: Recommendation is accepted. The answer key has been revised. (13) Question 3.16 Comment: The question did not request the specific values listed on tne answer key. Additionally, memorization of the exact values is not warranted as they are specified in the annunciator response procedure. Recommendation: Accept the following as adequately " explaining" the meaning of the sensor status lights: (a) Same as stated on the answer key (b) Margin to saturation is satisfactory (c) Alert that RCS is approaching margin to saturation (d) Conditions in the RCS have reached saturation
Reference:
No additional reference is required. NRC Resolution: Recommendation is accepted. The setpoints were not intended to be required. (14) Question 3.18 Comment: No answer was given on the answer key for Surry. Recommendation: Require as answer: (a) Condenser vacuum available (b) 1/4 condenser CW outlet valves not fully closed
Reference:
Instrumentation Manual, Chapter 6, Figure 6 (Attachment S-6). NRC Resolution: Recommendation is accepted; however, the 2/2 logic for condenser vacuum is required for full credit since it is required by the question. (15) Question 4.05 Comments: (a) There are other available correct responses for the first two blanks.
7 (b) 5 Rem per year is now the administrative limit as set by the Vice President, Nuclear Operations. Recommendation: (a) Accept any of the following combinations as satisfactory for ~ the first two blanks: 1 Supervisor, Health Physics - 1.25 Y Station Manager - 1.75 7 Vice President, Nuclear Operations (Corporate Management) - 2.25 4 Senior Vice President, Power Operations (Corporate Management) - 2.75 (b) Accept 5 R/yr for third blank.
References:
Radiation Exposure Extension Forms (HP-12), pages 2, 3, 4, and 5 of 11 (Attachment S-7). Memorandum - Policy, Occupational Radiation Exposures, March 15, 1985, from W. L. Stewart (Attachment S-8). NRC Resolution: Recommendations (a)3, (a)4, and (b) are accepted. (a)1 and (a)2 are not accepted because tTey do not have the autliority to extend the yearly dose. (16) Question 4.09 Comment: Emergency boration is not an immediate operator action by the ATWT procedure. Even if it were, the four responses required by the answer key are not necessary. Emergency boration is contained in substep 6.c. of ECA-1. Only two actions are - necessary to perform the task. Recommendation: Accept for full credit the two steps listed in the procedure: (a) Switch BATP to FAST speed (b) Open M0V-()350
Reference:
ECA-1, pages 2, 3, and 4 of 5 (Attachment S-9). NRC Resolution: Recommendation is accepted. The answer key has been revised. (17) Question 4.11, Part (b) Comment: Only MOV- numbers are listed on the answer key whereas verbal valve functions may also suffice.
8 Recommendation: Accept valve function as stated below as addi-tional satisfactory response: (a) MOV-102B . Can be stated as SW pump supply header or CC HX supply header isolation valve. (b) MOV-106 Condenser CW inlet isolation valve (s) (c) M0V-100 Condenser CW outlet isolation valve (s)
Reference:
No additional reference is required. i NRC Resolution: Recommendation is accepted since they are equiva-lent answers. (18) Question 4.12 ,. Comment: Inclusion of containment conditions normal as a Post-LOCA SI termination criteria should be deleted from the answer key. Although this item erroneously appears on the foldout page, it has been corrected in the body of the LOCA procedure. Recommendation: Remove " containment conditions nonnal" from the answer key and redistribute points to the five criteria as listed !- - on the answer key.
Reference:
EP-2.00, page 4 of 16 (Attachment S-10).
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! NRC Resolution: Recommendation is accepted. The answer key has l been revised. l -(19) Question 4.14 i Conunent: This question could be answered.either true or false based on the examinees' interpretation of the presented statement. Since a safety injection signal does not start a containment spray pump, some examinees might assume the statement as false. If it. is assumed that a CLS signal was either automatically or manually initiated in addition.to the SI signal stated in the question,
- then the reqeired actions for the failure of a containment spray
! pump to start are those stated, and the questions assumed true. ! Recommendations: Delete the' question.
Reference:
No additional reference is required. f NRC Resolution: Recommendation is accepted. The question has t been deleted. I i .
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- b. Senior Reactor Operator Exam (1) Question '5.06, -Part a Comment: The detector located too far from the source results in a conservative. approach to criticality vs. nonconservative .as stated in the answer key.
Recommendation: Change answer to Part (a) to conservative.
Reference:
Lesson Plan ND-86.2-LP Reactor Start-Up, pages 7.48 through 7.51; illustrations 7.16 and 7.17 from lesson plan (Attachments-11). NRC Resolution: NOTE: This comment applies to Question 5.07, Part a. The answer-given.on the answer. key is correct as stated in Westinghouse Nuclear Training Operations, pages I-4.19 and - I-4.20. This was the reference supplied by the facility to write the examinations. However, the additional reference supplied with the utility's ' comments contradicts this. Since there ' are conflicting references, Part a. is deleted from the examination. (2) ' Question 5.12 Comment: See comment on question 1.12 of R0 Exam. Reconsnendation: See reconinendetion on question 1.12 of R0 Exam.
Reference:
See reference on question 1.12 of R0 Exam. NRC Resolution: See resolution on question 1.12 of R0 Exam. (3) Question 6.07, Part 1 Comment: See coninent on question 3.06 of R0 Exam.
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Recommendation: See recommendation on question 3.06 of R0 Exam.
Reference:
See reference on question 3.06 of R0 Exam. NRC Resolution: See resolution on question 3.06 of R0 Exam. (4) Question 6.08, Part 3
- Coninent
- See comment on question 2.02 of R0 Exam.
Recommendation: See recommendation on question 2.02 of R0 Exam. l
Reference:
See reference on question 2.02 of R0 Exam.
; NRC Resolution: See resolution on question 2.02 of R0 Exam.
. _ .~ . _ _ - _
, 10 (5). Question 6.10
. Comment: See comment on question 2.04 of R0 Exam. <
Recommendation: See recommendation on question 2.04 of R0 Exam.
Reference:
See reference on question 2.04 of R0 Exam. NRC Resolution: See resolution on question 2.04 of R0 Exam.
-(6) Question 6.18 Comment: See comment on question 3.18 of R0 Exam.
Recommendation: See recommendation on question 3.18 of R0 Exam.
Reference:
See reference on question 3.18 of R0 Exam. NRC Resolution: See resolution on question 3.18 of R0 Exam. (7) Question 7.05 Comment: See comment on question 4.05 of R0 Exam. ! Recommendation: See recommendation on question 4.05 of R0 Exam.
Reference:
See reference on question 4.05 of R0 Exam. NRC Resolution: See resolution on question 4.05 of R0 Exam. (8) Question 7.06 Comment: It is inappropriate to expect memorization of in-the-body steps of detailed operating procedures of this nature. The i information requested .is an extraction of steps contained on one ,- page (page 17) of a detailed refueling operating procedure, which in total is in excess of 200 pages. ! Recommendation: Delete the question.
Reference:
No additional reference is required. NRC Resolution: Question asks for knowledge of important param-eters checked during the head lift procedure, not memorization of steps. Review of the procedure (0P-4.1) revealed two additional items that are checked or verified. These items are radiation - measurements and source range monitoring in the control room. The answer key has -been changed to check any three of the five possible answers for full credit. The utility's recommendation that the question be deleted is not accepted. 1
r 11 (9) Question 7 08 Comment: See comment on question 4.09 of R0 Exam. Recommendation: See recommendation on ques' tion 4.09 of R0 Exam.
Reference:
See reference on question 4.09 of R0 Exam. NRC Resolution: See resolution on question 4.09 of R0 Exam.
-(10) Question 7.10, Part b Comment: See comment on question 4.11 of R0 Exam.
Reconnendation: See reconnendation on question 4.11 of R0 Exam.
Reference:
See reference on question 4.11 of R0 Exam. NRC Resolution: See resolution on question 4.11 of R0 Exam. (11) Ouestion 7.12 Connent: See connent.on question 4.12 of R0 Exam. Reconnendation: See recommendation on question 4.12 of R0 Exam.
Reference:
See reference on question 4.12 of R0 Exam. NRC Resolution: See resolution on question 4.12 of R0 Exam. (12) Question 7.13 Connent: See comment on question 4.14 of R0 Exam. Reconnendation: See recommendation on question 4.14 of R0 Exam.
Reference:
See reference on question 4.14 of R0 Exam.
- ,_ NRC Resolution
- See resolution on question 4.14 of R0 Exam.
(13) Question 8.03 Comment: Selection (d) is as correct as selection (b). Reconnendation: Accept either (b) or (d) as correct responses.
Reference:
No additional reference is required. Review TS 3.1-5, item 6.c. NRC Resolution: Selection (d) is neither correct nor as correct as selection (b). Removal of power from the block valve does not
12 completely satisfy 6.c. and therefore the reactor must be put in at least hot shutdown within the next six hours. To do this, keff must be reduced .to .9823 in accordance,with the definition of hot shutdown and subsequently Tave must be reduced to less than 200 F. No change is required. (14) Question 8.04 Comment: Selection (c) is as correct as selection (a). As worded, selection.(c) can be interpreted as if one rod is, in fact, inoperable and the plant is being operated below tee Tech Spec limit for this condition. The examiner meant for the opera-tor to- assume that all rods were operable and bank D was only below the more restrictive inoperable- Tech Spec curve. The poor wording of selection (c) allows for the alternate interpretation. Recommendation: Accept either (c) or (a) as correct responses.
Reference:
No additional reference is required. Review TS 3.12, c.5. NRC Resolution: Recommendation is accepted. The answer key has been revised. (15) Question 8.05 Comment: Selections for additional correct responses to sections F and G. are contained in column B and should be accepted as satisfactory. Recommendations: F. Accept-3. as an acceptable response. G. Accept 1. as an acceptable response.
Reference:
No additional reference is required. NRC Resolution: Recommendation is accepted. The answer key has been-revised. (16) Question 8.12 Comment: It is inappropriate to expect memorization of in-the-body steps of detailed procedures of this nature. The immediate general actions contained in EPIP-1.01 should more than suffice concerning expected. memorized steps contained in the Emergency Plan. Recommendation: Delete the question.
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Reference:
No additional reference is required. NRC Resolution: ' Recommendation is accepted. .The question has been deleted. (17) Question 8.14 Comment: Due to the .ery non-restrictive nature of this specifi-cation, it seems totally inappropriate to expect memorization of such a low priority item. This was a very poor choice of ques-tions on Technical Specifications when such a large quantity of more pertinent material is required knowledge and was immediately available. Recommendation: Either delete the question or reduce its point value.
Reference:
No additional reference is required. NRC Resolution: Recommendation is accepted. The question has been deleted from the examination. [18) Question 8.15 Comment: This is an example of a trick question. The question solicits a response to "... how they communicate with the SS." The expected response is that they do not. However, communica-tions :are _ established with the SS by one of several means to notify him of personnel availability. Recommendation: Change part (2) of- the answer key to allow for any of the following correct responses: (a) PBX
- (b) Ringdown phone (c) Station radio
Reference:
No additional reference is required. NRC Resolution: The second part of the question has been deleted from the examination. (19) Question 8.16
~Connent: More than three responses exist for this question. The
, question could be answered in very general terms as in the answer key, or it could be answered more specifically. Recommendation: Accept any of the following as satisfactory responses:
9 14 (a) Jumper log use (b) Blue tag (special order tag) use (c). Independent verification requirement (d) Technical review by Engineering Dept. or STA
Reference:
Administrative Procedure 29.5, pages 13 and 14 (Attachments-12). NRC Resolution: Recommendation is accepted. The answer key has been revised.
- c. General Comments (1) Ensure that written examination. questions are valid when compared to actual work assignments.
Response: 'The NRC has issued NUREG-1122, " Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Pressurized Water Reactors." Still under development is the Examiners Handbook, NUREG-1121, which provides implementing guidance for NUREG-1122. These NUREGs are scheduled for pilot-test efforts in Region II in February and March 1986, although much of the guidance contained in NUREG-1122 has been used extensively on our examinations for the last six months. NUREG-1122 was used as a basis for deter-mining'if several of the questions which h~ad specific comments by VEPC0 were performance based. (2) Comply with the NRC policy restricting the number of multiple choice questions used on written examinations. Response: The present NRC policy is that each written examination-shall contain no more than 25% True-False and multiple choice questions. This policy was fully complied with on both the Surry and North Anna written examinations. On the Surry examinations, for example, the final version of the R0 examination contained l eight multiple choice and six True-False questions out of a total of 65 questions. On the SR0 examination, there were twelve multiple choice and two True-False questions from a total of 64. The percentages were 21.5 and 21.9 respectively on the R0 and SR0 examinations. Approximately 27% of the questions were also highly , -objective, i.e., matching, fill-in-the-blank, and single word
- responses. ThesJ were not included in the category of multiple choice /True-False. The remaining 50% of the questions were more subjective, i.e., explain, describe, list, etc. The North Anna
( examination wcs very similar in construction. l l i l L
15 (3) Conduct simulator evaluations using an anticipated shift complement. The simulator examinations were conducted in full compliance with applicable Examiner Standards.. The candidates . selected = for examinations represented the percentage of licensed personnel at each site, i.e., R0s, SR0s and staff personnel. Two of the three crews at North Anna and three of the four crews at Surry consisted of two R0s and one SRO. The remaining crew at each site was of necessity, comprised of three SR0s. Each of the SR0s in the latter type of crew, functioned in both an SR0 and an R0 capacity. This is consistent with their license which permits them to "... direct the licensed activities of licensed operators at, and to manipulate all controls ..." of their applicable facilities. In addition, VEPC0 staff instructors functioning in the capacity of Shift Technical Advisors, were made available to and were utilized by each of the crews at each site. We can therefore find no basis for this general coment. (4) NRC Post-Grading Review The following changes were made as a result of the NRC post-
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grading review in accordance with NUREG-1021, ES-108, Quality Assurance Program for Review of Graded Examinations. The affected examinations were regraded accordingly. (a) R0 Exam 1 Question 4.02. Choice (b) is incorrect. Either choice (a) or (d) is the correct response to this question. However, since knowledge 'of these Technical Specifications is considered beyond the requisite knowledge of Reactor Operators, this question. was deleted from the R0 exam. The SRG answer key was modified to accept either (a) or (d). 2 Question 4.07. The answer key has been changed to require "'.oss of CCW" to the reactor coolant pump for: full credit in accordance with the foldout page. (b) SR0 Exam 1_ Question 7.01. See response to Question 4.02 above. 2, Question 7.07. Same as Question 4.02 on the R0 exam. t
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- 4. - Exit Meeting At the conclusion of'the site visit the examiners met with representatives of the plant staff. to discuss the results of the examination. Those individuals who clearly passed the oral examination were identified.
There were no generic weaknesses (greater than 75 percent of candidates giving incorrect answers to one examination topic) noted during the oral - examination. The ' examiners discussed the inability of the simulator to properly respond to several of the transients / malfunctions used during the examinations. The licensee did not identify as proprietary any of the material provided to or reviewed by the examiners. f
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e . ENCLOSURE 3
. U. S. NUCLEAR REGULOTORY COMMISSION
' REACTOR OPER~ATOR LICENSE EXAMINATION FACILITY! SURRY 1&2 REACTOR TYPEt PWR-WEC3 P '
DATE ADMINISTERED! 05/12/04 EXAMINER! TOM ROGERS ,__ . 4' APPLICA,NT! 8'____________________c_____ INSTRUCTIONS TO APPLICANT! , Uso separate paper for the answers. Write answers on one side only'. Stcple question sheet on top of the an,swer sheets. Points for each qu2stion are indicated in parentheses after the question. The passing Srnde~ requires at least 70% in each category and a final, grade of at
. loost 80%. Examination papers will be picked up six (6)- hours after '
tha, examination starts. . 3
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.. ,. % 0F ,
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C ATEGORY,. % rJ~-. APPLICANT'S CATEGORY. VALUE' TOTAL SCORE VALUE CATEGORY _______________-_-___________33.___ 18 0 25 .
- 1. PRINCI,P,LES OF NUCLEAR POWdR t- ,.cr
___I__0__ ___I_00 ___JO______ _ ________ PLANT OPERATION,.THERMODiNAMI&Si HEAT TRANSFER AND FLUID FLOW S ,_ (6. 0
'" ^^ e - -
t - _15111__ _25.00 _____ ___________ ________ 2. PLANT DESIGN INCLUDING SAFETY , . AND EMERGE,NC.Y SYST, EMS , 18.00 25.00 -
.e 3.
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I'NSTRUMENIS'AND CONTROLS T'T Pit 0CEDURES - NORM AL, , ABNORMAL,"' ,' ' _1 11 __ _ I ___________ __._____ 4. EHERGENCY 'AND' RADIOLOGICAL CONTROL >
~'. '
l 68.5 72.00 9 100.00 -
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ce-
. TOTALS
________ ______ ___________ -____-__ j. * . 4 FINAL GRADE _.________________% , l All work done on this examination is my own. I have neither
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6- ! qivsn nor received aid. - l _. t .t -
, ,, ________________________w__________
l APPLICANT'S SIGNATURE l . t.
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o t .
- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION: PAGE 2
--- isERR55.iRARiEs- REsi iRARiFEE AR5 FE0i5 FE5Q QUFSTION 1.01 (1.00)
Attached Figure # 220 shows a power history and four possible samarium traces (reactivity vs time). Select (a, b, c, or d) the correct curve fcr displaying the expected samarium transient for the given power history. QUESTION 1 02 (1.00) Attached Figure # 219 shows a power history and four possibic xenon traces (reactivity vs time). Select (a, be ce or d) the correct curve for displaying the expected xenon transient for the given power history. QUESTION 1.03 (1.00) A PWh is designed to operate like the Rankine Vapor Cycle shown on Figure 4210. Which of the fo11owin3 equations could be used to calculate the cycle's thermodynamic efficiency?
- o. AREA WITHIN (a-b-e-d-a) / AREA WITHIN (a-d-f-e-a)
- b. AREA WITHIN (e-a-b-e-d-f-e) / AREA WITHIN (a-b-c-d-a)
- c. AREA WITHIN (a-d-f-e-a) / AREA WITHIN (e-a-b-e-d-f-e)
- d. AREA WITHIN (a-b-e-d-a) / AREA WITHIN (e-a-b-e-d-f-e)
(xxxxx CATEGORY 01 CONTINUED ON NEXT PAGE xxxxx) l l l
l l
- 1. ' PRINCIPLES OF NUCLEAR POWER PLANT OPERATIONe PAGE 3 iREER559AERICi- REAi iRAREFER KR5 FEUi5 FEBR .
l OllESTION 1 04 (1.00) Which of the following equations used to perform a PWR heat balance calculation is correct?
- o. Grx = M(s) Ch(s) - h(fw)] + M(bd) Eh(bd) - h(fw)3 + Grcp
- b. Orx = M(s) [h(s) - h(fw)3 + M(bd) [h(bd) - h(fw)3 - Orcp
- c. Grx = M(s) Ch(s) -
h(fw)] - M(bd) [h(bd) - h(fw)3 - Orep
- d. Grx = M(s) Ch(s) - h(fw)] - M(bd) Ch(bd) - h(fw)3 + Orep NOTE: Notation Key 0 = Power M = Mass Flow Rate tw = Feedwater rx = Reactor bd = Blowdown s = Steam h = Specific Enthalpy GUESTION- 1.05 (1.50)
Indicate whether the following statements concerning rod worth are ndUE or FALSE.
- o. One reason for overlapping rod groups is to minimi=e the effects of rod shadowing on total rod worth.
- b. Both an RCS temperature increase and a buildup of fission product poisons will DECREASE rod worth.
- c. The maximum differential rod worth occurs at the point where the integral rod. worth is maximum.
QUESTION 1.06 (1.50) Indicate whether the Total Power Coefficient gets MORE NEGATIVE, LESS NEGATIVEr or DOES NOT CHANGE for the following conditions.
- o. From low power to high power at BOL.
- b. From low power to high power at EOL.
- c. From BOL to EOL at a constant power level.
(xxxxx CATEGORY 01 CONTINUED ON NEXT PAGE xxxxx) e
+v- , - - - - - - , , . . - - - , , - - , ------.--,--,-i----v ---e-..,e-w%,---r.-- -- - ,v--,----,-.---.-ww--,. r--e-+-.- --,e- -- , - - - -e
L. ' PRINCIPLES OF NUCLEAR POWER PLANT OPERATIONe PAGE 4
~ ~~~~iU5R566 555165~~5Edi~iR505FER 5R5~FC015 FE5R ~GUFSTION 1.07 (1.50)
Indicate whether the following will INCREASE, DECREASE, or HAVE NO EFFECT on the available (actual) Net Positive Suction Head (NPSH).
- o. Increasing pump speed.
- b. Increasine pump suction temperature.
- c. Increasing system pressure.
QUESTION 1 08 (1.00) The reactor in critical at 10,000 eps when a S/G PORV fails open. Assumin3 BOL conditions, no rod motion, and no reactor tripe indicate:
- o. Whether the final steady state Tavs will be GREATER THAN, LESS THAN, or EQUAL T0 the initial Tavg.
- b. Whether the final steady state power will be GREATER THAN, LESS THAN, or EQUAL to the point of adding heat.
QUESTION 1.09 (1.50) The reactor is operating at 25% power when one RCP trips. Assuming no roactor trip or turbine load change occure indicate whether the following parameters will INCREASE, DECREASE, or REMAIN THE SAME.
- o. Flow in operating reactor coolant loops
- b. Core delta T
- c. Operating loop steam generator pressure (xxxxx CATEGORY 01 CONTINUED ON NEXT PAGE xxxxx) 6 6
- 1. ' PRINCIPLES OF NUCLEAR POWER PLANT OPERATIONe PAGE 5
--- YREER557RARICi- Riii fissiFEE 3R5 PE015 PE5s CUESTION 1.10 (1 50)
Assun'ing a symmetrical (ideal) axial flux shaper match the CONDITION in ~ Column A to the LOCATION that it would occur in Column B. COLUMN A COLUMN B
- o. MINIMUM Critical Heat Flux 1. BOTTOM
- 2. Between BOTTOM & MIDDLE
- b. MAXIMUM Actual Heat Flux 3. MIDDLE
- 4. Between MIDDLE & TOP
- c. MINIMUM DNBR 5. TOP QUESTION 1.11 (1.50)
Match the heat transfer process in Columr. A to the equation that cpplies to that process in Column B. COLUMN A COLUMN B
- c. Between cold-leg and hot les 1. 0=aeaT of reactor (normal FC flow)
- 2. 0 = 4 5 4T
- b. Across S/G tubes (primary to secondary) 3. G=UA4T
- c. Across S/G (feedwater to steam) 4. 0=Ecah
- 5. 0=sah QUESTION 1.12 (1.00)
As the core ages, the buildup of Pu-240 causes the Fuel Temperature Ccefficient (pcm/ degree F) to become more negative. With this change cccurring, why does the Doppler Only Power Coefficient (pcm/% power) b0come less negative as the core ages? RUESTION 1.13 (2.00) TWO major factors effect differential boron woth over core life. List these TWO factors AND indicate how (MORE NEGATIVE or LESS NEGATIVE) they offect differential boron worth. (maxxx CATEGORY 01 CONTINUED ON NEXT PAGE xxxxx)
~ ~ - = , - . - - - - , , ,- - ve ,- -- - --e.m,--,-en .e--, ,-,,,-y, , ,,,-.. , -y- ,-. . , , - - - , - - , - - - , , , - -
I.*PRIN'CIPLES OF NUCCEAR POWER PLANT OPERATION, PAGE 6
~ ~ ~~~~TUEE566YU55fC5I~AEIT iE5A5EER I 6"EEU56~EEUU QUESTION 1.14 (1.00)
What is the quality of a 540 degree F vapor-liquid mixture whose specific cnthalpy is 1175 BTU /lba? (***** END OF CATEGORY 01 xxxzz) i I i l i
=
1 l
2.* PLANT DESIGN INCLUDING SAFETY AND EMERCENCY SYSTEMS PAGE 7 QUESTION 2 01 (1.00) What is the maximum purification flow rate the mixed bed dominerali=ers cre sized to except? HUESTION 2.02 (1.00) Mctch the following loads to their normal power supply. The selections ccy be used more than once.
- 1. 200 kw PZR backup heaters. a. 4160V Bus'1A.
- 2. Low head SI pump 1A. b. 480V Bus 1/J.
- 3. 4160V Bus 1A breaker control power. c. 4160V Emergency Bus 1J.
- 4. Fire protection cabinets.
- d. 120VAC Vital Bus.
- 5. Condensate pump 1A.
- e. 125VDC Vital Bus.
- f. 125VDC Intake Structure Bus.
- g. 4160V 1J Stub Bus.
y* 4lG0 4 ir 6py IQ{ GUESTION 2.03 (1.00) ggg 3g yg For the following air controlled valves, indicate if their failed position 10 open or closed upon loss of control air.
- a. Pressuri=er spray valves.
- b. Pressurizer PORVs.
- c. Charging flow control valve.
- d. Steam dump valves.
G. RHR flow control valve 605. QUESTION 2.04 (1.00) What is the normal power source for the EHC System? (xxxxx CATEGORY 02 CONTINUED ON NEXT PAGE xxxxx)
2.~ PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 8 QUESTION 2.05 (1.00) The cold les accumulators are designed so that ______ accumulator (s) eill cover ______ of the core.
- o. 1, 50%
- b. 2, 50%
- c. 1, 100%
- d. 3r 90%
HUFSTION 2.06 (1.00) H3w much steam load is the steam dump system designed to dump to the ccndenser without causin3 e reactor trip? . HUESTION 2.07 (1.00) (c) What are the three sources for AFW supplies? (.75) (b) Which of these can be properly aligned without operator action outside the control room? QUESTION 2.08 (1.00) What chemical is added to the containment spray subsystem and what is its purpose?
's l (tUESTION 2.09 ( .50) {
Containment phase ( isolation will prevent any further operation of the pressurizer PORVs. TRUE or FALSE? QUESTION 2.10 (1.00) l l Why is a seal bypass necessary for the reacttv e;4alant pump 41 seal? { (xxxxx CATEGORY 02 CONTINUED ON NEXT PAGE xxxxx) 4
2.' PLCNT DESIGN INCLUDING SAFETY AND EMERCENCY SYSTEMS PAGE 9
- +
QUESTION 2 11 (1.00) Explain what happens to the auxiliary steam condensate from the various leads it is used to heat. HUESTION 2 12 (1.00) What is the maximum steam generator blowdown rate achievable from cach steam generator? HUESTION 2.13 ( .50) Hydrazine is added to the main condensate and feedwater systems by two cotor driven positive displacement pumps. TRUE or FALSE? HUESTION 2 14 (1 00) How many passes does the MSR tube bundles make with the recent codification? HUESTION 2.15 Fi . 6 E) What is the purpose of the floating platform at the high level intake of the circulating water system? HUESTION 2.16 {i."^' (0,5)1ko
,Why has the Boron Injection Tank Design Change removed the BIT inlet I valves and retained the BIT outlet valves?
( HUESTION 2.17 (1.00) 6 'Where in the ECCS flow path does the RWSTs cross connect between Units? (xxxxx CATEGORY 02 CONTINUED ON NEXT PAGE xxxxx) s L
, ?!
tj 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 10 1
------------------------------------------------------- . ;l
. l
. .l OllESTION -
'^
.00; 3deTE b '
~!
Under wh t operating conditions is the steam senerator moisture ccParating equipment designed to limit-carry over to.0.25%? j, ItuESTION 2.19 ( .50) The Lo Head Safety Injection hot les penetration is on the SG side of the roactor coolant loop isolation valves. TRUE or FALSE? T (xzzza END OF CATEGORY 02 xxxxx) 4
I
.3 .
- INSTRUMENTS AND CONTROLS PAGE 11 QUESTION 3.01 (1.00)
How does an operator disable a motor driven AFW pump from starting on cny of the automatic start signals? QUESTION 3.02 (1.00) { I The SG 1evel program uses ______ as an input for reactor pow'er. GUESTION 3.03 (1.00) What two different conditions will prevent the feedwater regualting bypass valves from being opened. Do not include control malfunctions. (tUESTION 3.04 (1.00) Explain how the Reactor Vessel Level Indicating System is compensated to maintain the required accuracy during a LOCA. QUESTION 3.05 (1.00) Which of the following malfunctions could :ause one of the over temperature delta T trip bistables to trip?
- o. Controlling turbine impulse pressure channel failing low.
- b. Power range N43 lower detector failing low.
- c. Reactor coolant flow detector failing low.
- d. Controlling pressurizer level channel failing low.
(xxxxx CATEGORY 03 CONTINUED ON NEXT PAGE xxxxx)
~.
. , - . ,- ,. _. .w . , , - _ _ _ . _ . - . , . , _ _
! . t 3.* INSTRUMENTS AND CONTROLS PAGE 12 [ QUESTION 3.06 (1.00) Match the following conditions with the expected indication provided by the rod speed indication meter. The selections may be used more than once.
- 1. Immediately before an operator removes the N44 a. O s/a. ),
fuses because of a failed high detector. Rods ! in AUTO with no temperature mismatch, b. 8 s/m.
- 2. Rods in MANUAL with a 10 F temperature c. 40 s/m.
mismatch.
- d. 48 s/m.
- 3. Rods in AUTO with a 1 F temperature mismatch.
- e. 72 s/a.
- 4. Rods in AUTO with a 4 F temperature mismatch.
- f. 88 s/m.
- 5. Rods in AUTO with one of the Tave control instruments failed low.
QUESTION 3 07 (1.00) If emergency diesel generator 1 is taken out of service for preventive cointenance when a blackout occurs on Unit 1, how will EDG 3 respond ift durin3 its loading sequencer an SI actuation signal is received from Unit 2 with EDG 2 failing to start? UUESTION 3.08 ( .50) The cold les accumulator isolation valves are sent an open signal upon en SI actuation. TRUE or FALSE? HUESTION 3.09 (1.00) What is required to clear a 'PZR NDT Overpressure System Req'd' alarm?
'GUESTION 3.10 (1.00)
Which valves get an automatic shut signal upon receipt of a process vent particulate and gas monitor alarms? (xxxxx CATEGORY 03 CONTINUED ON NEXT PAGE xxxxx)
PAGE 13 5
'.3.' INSTRUMENTS AND CONTROLS
{
' 5 QUESTION 3.11 (1.00)
Nhat plant condition must exist to allow aanval blocking of the HI Steam Flow Line SI actuation signal? RUESTION 3 12 (1.50) jl Explain the effects of a dropped rod on unit operation caused by permissive P-3. NUESTION 3.13 (1.00) s
-Indicate whether the following valves will be open or shut after depressing the LATCH button on the EHC Control Panel.
- c. Main stop valves.
- b. Governor valves.
- c. Reheat stop valves.
-d. Intercept valves.
QUFSTION 3.14 (1.00) During a shutdown with IR N35 undercompensated,
- o. both source range detectors will automatically energize at a higher power level than normal.
- b. both source range detectors will automatically energize when IR N36 f drops below the P-6 setpoint regardless of the N35 indication.
- c. source range N31 will need to be manually energized after N32 automatically energizes.
- d. both source range detectors will have to be manually energized below P-6.
(xxx** CATEGORY 03 CONTINUED ON NEXT PAGE marax) i
r.
.3 . INSTRUMENTS CND CdNTROLS PC.CE 14 QUESTION 3 15 (1.00)
Explain the status of the Remote Monitoring Panel power supplies given the following indications! Two. blue lights and one amber light is illuminated adjacent to the , Unit / Loop Selector Switch. . QUESTION 3.16- (1.00) Explain the meaning of the following status lights for the temperature consors on the Core Cooling Monitor.
- a. All lights out.
- b. Green.
- c. Yellow.
- d. Red.
QUESTION 3.17 (1.00) Haw is the control signal for the containment vacuum pumps generated? QUESTION 3.18 (1.00) What it the minimum logic that must be satisfied for the condenser to be available for the steam dumps (C9)? 4 (xxxxx END OF CATEGORY 03 xxxxx) b
4.* PROCEDURES - NORMAL, ABNORM ^Le EMERGENCY AND PAGE 15
--- Asaracacicat canizac------------------------
GUESTION 4.01 (1.00) , It is required to exit a radiation area and report to HP whenever your 0-200 nr self-reading dosimeter reaches ______ ar. v
- o. 100 [
- b. 125
- c. 150
- d. 175 N
-eUFfrT49M- h !;.00:
Which of he following statements regarding Axial Flux Difference is CORRECT?
- o. Reactor power CANNOT be increased above 50% rated thermal power unless the indicated AFD is within the target band.
- b. If the indicated AFD is outside the target band for more than 1 HOUR CUMULATIVE over a 24 HOUR period, with reactor power between 50 and 90% of rated thermal powere EITHER reduce thermal power to less tnan 50% within 30 minutes and reduce the power range Nuetron Flux High setpoint to less than 55% within 30 minutes.
- c. If indicated AFD is outside the target band and thermal power is Sreater thin 90% rated thermal powere within i HOUR AFD aust be restored within the band or power reduced to < 90%.
- d. Below 50% rated thermal powere there is NO penalty for being outside the target band due to the fact that uneven xenon buildup in the core does not have an adverse impact at low power levels.
(xxxxx CATEGORY 04 CONTINUED ON NEXT PAGE xxxxx)
ti
- 4 ." PROCEDURES - NORM'AL, CBNORMAL, EMERGENCY AND PAGE 16 j
--- a;5i5t55iExt casiRat------------------------ .
1
. \
'I GUESTION '4.03 (1.50) I:
Match the event listed in column A with the' approximate power level in i column B at which this action is taken on a unit startup to 100% power. , COLUMN A COLUMN B
- a. Place a second Main Feed pump in service 1) 15%
- 2) 30%
- b. Stop increasing power and check for a 3) 50%
chemistry hold 4) 60%
- 5) 70%
- c. Perform a calorimetrie 6) 90%
4 QUESTION 4.04 (1 00) List the 4 methods given in the S/G Tube Rupture E0P to identify which S/G is ruptured. j UUESTION 4.05 ( .75) Radiation Workers can be allowed, by authorization of __________, up to _____ Ren per quarter or _____ Rem per year provided their lifetime dose does not exceed 5(N-18), provided the proper forms (NRC-4, HP-12) are completed. , QUESTION 4.06 (1.50) . Give the locat' ion of the following support centers that are manned duri o plant emergency. c) Operations Support Center b) Technical Support Center c) Emergency Operations Facility GUESTION 4.07 (1.00)
> Following a valid reactor trip and safety injection, what are the Reactor Coolant Pump Trip Criteria?
(xxxxx CATEGORY 04 CONTINUED ON NEXT PAGE xxxxx) O t
A 4.' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY C.ND PAGE 17
~~~~~~~~~~~~~~~~~~~~~~~~
~~~~E.5656E66i6AE~66UTR L .
~
s 00ESTION 4.08 (1.50) l List all the immediate actions that the operator must take if a SINGLE i k dropped rod occurs. i 4 QUESTION 4.09 (1.00) { List the immediate operator actions to initiate emergency boration if it is required on an Anticipated Transient Without Trip condition. , N ItVESTION 4.10 (2.00) c List ALL the immediate action sub-steps from EP-1 00, ' Reactor Trip or Safety Injection' that allow you to accomplish the following immediate cetions. c) Verify if SI is actuated (1.0) Verify SI Valve Emergency Lineup (1.0) b) QUESTION 4.11 (1.50) i Answer the following questions regarding a 'Non-Recoverable Loss of Air", (AP-40) on UNIT 1: a) What are the three conditions that require the RCPs to be secured due
; to the loss of CC? (0.75) b) When the INTAKE CANAL LO LEVEL TRIP is received, what CW & SW MOVs are , manually operated? (0.75)
UUESTION 4.12 (1 50) List the SI termination criteria following a LOCA. (xxxxx CATEGORY 04 CONTINUED ON NEXT PAGE xxxxx) i, 4
- O e
h 4.* PROCEDURES - NORMAL, ABNORMAle EMERGENCY QND
~
PAGE 18 ; . --- iX5iBE55fEKE E5siiBE------------------------ } 4 1 OUESTION 4.13 (1.00) List the 4 DISTINCT hazards to which personnel are exposed when an entry { into the reactor compartment is made during reactor operations.
- 'J ';TIOM i.10
. ";0 ; d*ktL M k 3
TRUE or FALSE: # If on an SIe a containment spray pump CANNOT be started, the ' Response Not Obtained' step requires verification that its associated Chemical , Addition Tank MOVs (MOV-CS-( )03 A & C or B & D) be closed. QUESTION 4.15 ( .50) i What is the purpose of a '15 Minute Headway
- tag?
GUESTION 4.16 ( .75) What constitutes a Class II reactor trip? ' i (xxxxx END OF CATEGORY 04 xxxxx) (xxxxxxxxxxxxx END OF EXAMINATION xxxxxxxxxxxxxxx) v -
- - . , - - - - - , e , . - . . - , _ - , , - - -
, , - . , . . . , - a n -.n... , , ,-e
.- - g 5 p 19 PAGE
' 1. ' PRINCIPLES OF NUCLEAR POWER PLCNT OPERATION: s
~~~~~iUEii66iii5fC5,~0EIi~iEI55EEE~I56~EEUi6~EECU -
t' ANSWERS -- SURRY 142 -85/12/04-TOM ROGERS ; W b ANSWER 1.01 (1.00) ?
-e i d ~-
REFERENCE y EIH! L-RG-606, pp 4, 58 Fig. 4 i. BSEPt 02-2/3-Ar pp 177 - 1808 02-0G-A, pp 60 - 61 BFNP! Xenon and Samarium LP, pp 5, 68 RQ 84/03/05 ?' Westinghouse Nuclear Trainins Operations, pp. I-5.77 - 79 ANSWER 1.02 (1.00) [ c _ REFERENCE EIHf GPNTeVol VII, Chapter'10.1-83-86 i BSEPt L/P 02-2/3-A, pp 172 - 1763 02-0G-Ae pp 57 - 60 l Westinghouse Nuclear Reactor Theory, pp. I-5.77 - 79 i ANSWER 1.03 (1.00) d REFERENCE General Physicse HTEFF, pp. 137 - 142 .
.3 it ANSWER 1.04 (1 00) b REFERENCE '
General Physics, HTEFF ANSWER 1.05 (1.50)
- o. FALSE (0 5)
- b. FALSE (0.5)
- c. FALSE (0.5) i I
.1.
- PRINCIPLES OF NUCLEAR POWER PLCNT OPERATIONr PAGE 20 i
h
~
~~~~isiki66 U55565,~555T iEIU5f5R"5U6~EE 56"fE6U Iv ANSWERS -- SURRY 112 -85/12/04-TOM ROGERS REFERENCE W stinghouse Nuclear Training Operationne p. I-5.36 - 43 ANSWER 1.06 (1.50) ?!
- 0. LESS NEGATIVE (0.5)
- b. MORE NEGATIVE (0.5)
- c. MORE NEGATIVE (0.5)
REFERENCE Wastinghouse Nuclear Training Operations, p. I-5.29 ANSWER 1.07 (1.50) O. DECREASE (0.5)
- b. DECREASE to.5)
- c. INCREASE (0.5)
REFERENCE Ocneral Physics, HT&FFe p. 320 ANSWER 1.08 (1.00)
- c. LESS THAN (0.5)
- b. GREATER THAN (0.5)
REFERENCE Westinghouse Reactor Physics, Section I-Se MTC and Power Defect DPC, Fundamentals of Nuclear Reactor Engineerins 002/000-K5.02 (3.3/3.6)
=_. 3 .1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION PACE 21 . --- iRERR559EARICli Risi TRAEiFER RE5 FC0i5 FC5E .
CNSWERS -- SURRY 142 -85/12/04-TOM ROGERS ANSWER 1.09 (1.50) O. INCREASE (0.5)
- b. INCREASE (0.5)
- c. DECREASE (0.5)
REFERENCE Ceneral Physics, HT & FF - Fluid Flow Applications for Systems and Components 002/000-K5.01 (3.1/3.4) ANSWER 1.10 (1.50)
- c. 5 (0.5)
- b. 3 (0.5)
- c. 4 (0.5)
REFERENCE General Physics, HT&FFr pp. 228 - 230 ANSWER 1.11 (1 50)
- c. 1 (or 5) (0.5)
- b. 3 (Q
- c. 5 ((
REFERENCE General Physics, HT & FFr pp. 180 and 181 001/000-K5.47 (2.4/2.9)
J,, 4.' PRINCIPLES OF NUCLEAR POWER PLANT OPERATIONe PAGE 22
, --- iREER55?EERICi- Risi fRARiFER AR5 FE015 FE5R ANSWERS -- SURRY 182 -85/12/04-TON ROGERS ANSWER 1 12 (1~00)
Aa the core ages, the pellet and clad creep together causing an increase in gap conductivity (0.4 pts.). This causes a smaller delta T of the fuel for a given power change (0.3 pts.). The smaller delta T of the fuel.causes a smaller change in reactivity for a given power change (0.3 pts.). (pen /F) X F/% power) = pcm/% power more nog less less nes REFERENCE Westinghouse Nuclear Training Operationse' p. I-5.22 ANSWER 1.13 (2.00)
- 1. Baron Concentration Decreases (0.5 pts.) - NORE NEGATIVE (0.5 pts.)
- 2. Fission Product Buildup (0.5 pts.) - LESS NEGATIVE (0.5 pts.)
REFERENCE Westinghouse Nuclear Training Operations, p. I-5.31 ANSWER 1.14 (1 00) 97% (+/- 1%) F REFERENCE Steam Tables L r I e k
'.2. PLANT DESIGN INCLUDING SAFETY CND EMERGENCY SYSTEMS PCGE 23 CNSWERS -- SURRY 122 -85/12/04-TOM ROGERS i
l 1 ANSWER 2.01 (1.00) 120 gpm. REFERENCE ' SPS CVCS Sys Des. NAPS CVCS Advance Skills Training. ANSWER 2 02 (1.00) 90.2 poipts e,acht 1-be 2-g 3-/, 4-ee 5-a. ~BMM REFERENCE SPS Electrical System Lesson Plans. ANSWER 2.03 (1.00) 9 0.2 points eacht
- o. FS.
- b. FS.
- c. FO.
- d. FS.
- o. FS.
REFERENCE SPS Sys Des. for Primary systense Main Steam, and RHR. , ANSWER 2.04 (1.00) , Permanent magnet generator on the turbine shaft. REFERENCE SPS Turbine Protection Sys Des. ANSWER 2.05 (1.00) b i I s k
. ?
- 2. PLCNY DESIGN INCLODI;:C SAFETY AND ENERGENCY SYSTEMS PAGE 24 CNSNERS -- SURRY 182 -85/12/04-TDN RDGERS '
REFERENCE SPS SI Sys Des. NAPS SI or ECCS ESF Lesson Resources. ANSNER 2.06 (1.00) 40% REFERENCE SPS Main Steam Sys Des. NAPS Steam Dump Advance Skills Training. CNSNER 2 07 (1.00) O. 9 25 points each [*dy I)
- 1. CST if, jfTA/ p Gw,
- 7. Fire main y- 'M 2 , N,w o
- 3. Emergency askeup CST
- b. CST. su AAJ g y, REFERENCE SPS AFN Sys Des.
CNSWER 2.08 (1 00) , 8:dium hydroxide C0.53 to remove iodine from the containment atmosphereE.53 REFERENCE SPS Containment Spray Subsystem Sys Dez. NAPS Quench Spray Sys Lesson Resource. Ai!SNER 2.09 ( .50) FALSE. REFERENCE SPS Primary Systems Sys Des. NAPS RCS Press. Inst. Lesson Plan.
i
~ '.2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 25 C.NSWERS -- SURRY 142 -85/12/04-TOM ROGERS .
ANSWER 2.10 (1.00) To get additional flow, and therefore more cooling to the pump bearing [0.5 3 when they are operating at low RCS pressure [0.53. ; REFERENCE I SPS RCP Sys Des. ;
-ANSWER 2.11 (1.00)
Heating drips are co11ected[0 53 and then pumped back to the condensate otorage tank [0.53. REFERENCE SPS Aux. Steam Sys Des. NAPS Aux. Steam Sys. Plant Man. Vol. 1. ANSWER 2.12 (1.00) 70 spa. dkgr Spa REFERENCE SPS SG Blowdown Sys Des. ANSWER 2.13 ( .50) TRUE REFERENCE SPS Condensate Polishing Sys Des. NAPS Chen Feed Sys Deze Plant Man. Vol. 1. ANSWER 2.14 (1.00) 4 REFERENCE SPS Main Reheat Steam Sys Des.
~
f j'
?. PLCNT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 26 ;
CNSWERS -- SURRY 142 -85/12/04-TOM ROGERS , 4
'j CNSWER 2 15 (1.00) il To breakup the vortex created there and thus minimizing 02 to the condenser tubes for corrosion control.
REFERENCE SPS Cire Water Pump & Valves Sys Des. ANSWER 2.16 ' i . ^ ^ ? (e The BIT is no longer required so the inlet valves were removed (along with the recirc piping).?^. R The outlet valves remained bepause they also function as containment isloation valves . E+r53 C.Q (45') REFERENCE DC-%4*N h% b M bb + *E BIT Design Change Training. SPS. a( % h Am.r163 eda- -- 4'.' Valmed w A e p m %T Glaus S d M O **o ANSWER 2.17 (1.00) " b%I* N M*N 4 dh '3w At the suction of the charging pumps. REFERENCE SPS SI Sys Des.
,ue.en
..m.,
9 0.5 points each: plek W
- 1. Steady state operating conditions with SG 1evels at the maximum normal operating level.
- 2. Loading and unloading at 5%/ min between 15% and 100% power.
- 3. A 10% step change between 15% and 100% power.
REFERENCE SPS Primary Sys Des. NAPS RCS, SC Lesson Resource.
i
. . .- u
.?. PLCNT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS ' PAGE 27 !
-85/12/04-TOM ROGERS i CNSWERS -- SURRY 112
- ?
a CNSWER 2.19 ( .50) .
~
FALSE. ! REFERENCE SPS Primary Sys. , f 3 m j 9 b i
. . . . 3 p
4 ..
- 3. INSTRUMENTS AND CONTROLS PAGE 28 i;
...--....--.......--....---- 6 ANSWERS -- SURRY 122 - -85/12/04-TOM ROCERS !
1 CNSWER ?.01 (1.00) Place switch in STOPr Pull to lock position. j REFERENCE j SPS AFW Sys.' Des. j i ANSWER 3.02 (1.00) Turbine first stage pressure. REFERENCE SPS SGWLC Sys. Des. NAPS SGWLC & P Advance Skills Training. - ANSWER 3 03 (1 00) 0 0.5 points eacht
- 1. SCWL high. *
- 2. SI signal present.
REFERENCE SPS Main Feed Water Sys. Des. ANSWER 3.04 (1 00) 0 0.25 points eacht It receives inputs from each of the following I"
- 1. Temperature of impulse line.
- 2. RCS temperature. L
- 3. Wide range pressure.
cnd
- 4. The d/p cell is located outside of containment.
REFERENCE SPS RVLIS Sys Des. NAPS RVLIS Advance Skills Training. I
... y J. INSTRUNENTS AND CONTROLS PACE 29 ANSNERS -- SURRY 142 -
-85/12/04-TON ROCERS t
CNSWER 3.05 (1.00) b : REFERENCE SPS TS 2.3-2. } NAPS RPS Adv. Skills TraininS. , ANSWER 3 06 (1.00) S 0.2 points eacht 1.-a, 2-dr - r4-cr5-a. REFERENCE lb0 SPS Rod Control Sys Des. NAPS Rod Control Lesson Plan. ANSWER 3.07 (1.00) . 1 It will dedicate itself to the SI plant- Unit 2. REFERENCE SPS EmerSency Power Dst. Sys Des. ANSWER 3.08 ( .50) TRUE. REFERENCE SPS SI Sys Des. l NAPS SI or ECCS ESF Lesson Resources. ANSWER 3.09 (1.00) Place both PORV key switches to ENABLE Co.53 cnd both motor operated PORV isolation valves OPEN [0 53. REFERENCE SPS Primary Systems Sys Des. i i
.. r
. 3. INSTRUMENTS AND C6NTROLS PACE 30 9 i
3 CNSWERS -- SURRY 182
-85/12/04-TON ROGERS
A CNSWER 3.10 (1.00) 1 Discharge' valves from containment vacuve system C0.53 and waste decay i tanks [0.53. (RCV-CW-160e FCV-CW-260e FCV-CH-101) p REFERENCE t SPS Red Monitoring Sys Deze Table 3 7-5. i' f-CNSWER 3.11 (1.00) l Teve less than or equal to 543 F (2/3 loops)-SPS + REFERENCE SPS SCWLC Sys Des. I' NAPS SI or ECCS ESF Lesson Resources. ANSWER 3.12 (1.50) l S 0.5 points eacht ?
- 1. Blocks auto rod withdrawal.
- 2. Initiates turbine runback. Ipb W.
- 3. O!;d;1;;d dispetd f C 3 n; . v.. i t . ,, lea / li%-( % 6 g, REFERENCE SPS UFSAR 7 2.
CNSWER 3.13 9 0.25 points eacht (1.00) { O. open. L
- b. shut.
- c. open.
- d. open.
i REFERENCE SPS Turbine Contros Sys. Des. 1 l i
3 PAGE 31 I) '.3. I::STRUMENTS AND CONTROLS
-85/12/04-TOM ROCERS
)l CNSWERS -- SURRY 182 . ..
j ANSWER 3.14 (1.00) d f REFERENCE p SPS Excore Inst. Sys Des. 1 NAPS Nuclear Inst. Sys. Lesson Plan. ANSWER 3.15 (1.00) Unit 1 and 2 battery buses. C0 53 . Linit 2 supplying power. [0.53 o' de* 4 #M d $<leM. REFERENCE i SPS Remote Hon. Pnl. Sys Des. CNSWER 3.16 (1.00) ) t 0.25 points eacht { For SPSt i
- o. Temperature input bypassed or out of,ran g
- b. Margin to rat. (>15F T/C or > 25F RTO) #AJ A t o r< .
- c. Margin to sat.(between 0 and 15F T/C or 0 and 25F RTD) /Al 4wer'
- d. OF Margin to saturation.
For NAPSt -
- o. Manually or gotonatically digabled.I NWA g
- b. l8F RTDs 8F T/C.) ';
- c. Sat.
Sat Margiq Margin@ Watween 0 and 18F RTDe 0 and 8F T/C) /A/ i for*1' " d 0F Sat. Margin. REFERENCE SPS Core Cooling Monitor Sys Des. NAPS CCM Advance Skills Training. ~ .
- 3. INSTRUMENTS AND CONTROLS PACE 32 ;,
C::SWERS -- SURRY 142
-85/12/04-TOM ROCERS 4
i ANSWER 3.17 (1.00) ; Satpoint manually inserted (based on lowest air temperature measured ]' d:wnstream of containment air coolers).CO.53 Ccapares setpoint'to ' Partial Air Pressure" which is the measured h difference between total absolute pressure and sat pressure (from a I function generator based on lowest air temp down stream of containment f air coolers).CO.53 4 REFERENCE : SPS Containment Vacuum Sys Des. ! l CNSWER 3.18 (1.00) ; 9 .5 points each! , For NAPS l
- 1. 2/2 pressures. ;
- 2. 3/4 cire water pump breakers shut.
REFERENCE NAPS Steam Dump Advance Skills Training. L Sorry
*/s. Gd vawam ( re 'ng l
'/9 (sd. m voo m su. .
4 12E., ses srtne bwP hamuovr p t e t I
.4 . PROCEDURES - NORNdle ABNORMAL, EMERGENCY AND PACE 33 I
~ ~ ~II656E 655 E AE ~ E5Nii6E ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ . I ANSNERS -- SURRY 152 -85/12/04-TOM ROGERS y 4
ANSNER 4.01 (1.00) l c- j d REFERENCE Vepco CETe P.21. ; PNG-15! Radeon Knowledge (3.4/3.9) : 3 ANSNER 4 02 'i.001 W d oko b T u) I REFERENCE
- NA TS 3/4 2.1 .
StlRRY TS 3.12-B.4 001/0508 PWG-8 (3.6/4.5) . ANSWER 4.03 (1 50) o) 4 (+.5 ea) b) 2 e) 6 (5 for Surry) REFERENCE , NA OP-2.1, pp 10-13 Surry OP-2.1, pp 9/10 f i PWG-12! Perfore Integrated Plant ops (3.5/3.4) k ANSWER 4.04 (1.00)
-Unexpected rise in S/C level (+.25 ea)
-High radiation on a S/C blowdown line
-High radiation on an MS line monitor
-High radiation as determined by sampling and analysis REFERENCE Curry EP-4.00, pp 3 NA 2-EP-3, pp 2
, . . . v
- 4. PROCEDURES ~ NORMAle C8NORMAle EMERGENCY AND PAGE 34 3 1
Ri515E55f EEE E5MR5E----~~---------- [ ANSWERS -- SURRY 182 -85/12/04-TOM ROGERS EPE-0388 EA2.03 (4.4/4.6) b
,1 i11 ANSWER 4.05 ( .75) d' L Managementi 2.758 At'l (+.25 ea) {
C rporat$* v9,Wes t.si REFERENCE
- se ve,e.4sN ; s. a r ;,
Surry HP-2.Be pp 2 ,.> I l PWG-158 Knowledge of facility Radeon Requirements (3.4/3.9) i ANSWER 4.06 (1 50) , North Anna (+.5 es responsei i Sorry i ; 3rd floor conference room of Mntnce Blds 13rd floor conf rm of Hntnce Blds Records Blds IAdjacent to Station Control Room Training Blds INext to Training Complex Simul. ; I i REFERENCE Surry SEPe pp 1.5-7 NA NAEPe pp 1 5-9 PWG-36! E-Plan (2 9/4.7) ANSWER 4.07 (1.00)
- 1) Verify Charging /SI flow (+.5 ea)
- 2) RCS Pressure < 1600 psis (Surry) 3 j.e S K 1 7 1 *. "
* * < 1230 psig (NA)
'l REFERENCE SONP Foldout Page
. NA Foldout page for 2-EP-0 Sorry Foldout page for EP-1.00 003/0008 PWG-10 (4.1/4 4) m - - - - . , -, . _ _ - - . -.-_y- , _ _ ~ . _ _ . . , 7, --,. ..____,,,,,-,._m- - . - - - . . - . _ , . . . . - -
.s . ,. .I s
.4 . PROCEDURES - NORMAL, ABNORM,.Le EMERGENCY AND PAGE 35
--- R..515E5515KE 55iWR5E--------~~~---------~~-- {,
C.NSWERS -- SURRY-182
-85/12/04-TOM ROGERS 5.
I r1 al ANSWER 4.08 (1 50) r Vcrify Turbine runback (+.3 es) . , suto rod withdrawal blocked i steam dump armed and in servicer if required If rods in manuale insert to compensate for runback f Stabili=e plant conditions consistent with the new turbine load j REFERENCE Sorry 1-AP-1.4, pp 5 . I EPE-0038 PWG-11 (4.2/4.4) 4 ANSWER 4.09 (1.00) t Vari 9 SI/ Ch k RC pres pun s tu in .2 u ef px.
) a n"" --
O . .. a _ Ufc, { h y is '" ' ' ~ '
"~ "
j t)SwitchBATPtoFad,,2'^^c:i, speed (4,rtd.) y q
- 4) Open MOV-( )350 REFERENCE Surry ECA-1, pp 4 NA 2-ECA-1, pp 3 J t
EPE-0298 PWG-11 (4.5/4.7) , 4 ANSWER 4.10 (2.00) o) SI Initiated annunciator LIT (+.25 ea) Any SI First-out annunciator LIT j
~
SI Pumps running EDGs running ? b) CHG/SI Pump RWST MOVs open High Head SI to Cold Les MOVs open Low Head Pump RWST MOVs open Low Head Pump Cold Les DischarSe MOVs open REFERENCE Surry EP-1.00, pp 4 & 7 006/0003 A3.02 & A3.03 (4.1/4.1)
. u ; q
,, .e t
.4 . PROCEDURES - NORMdLe ABNORN^,Le ENERCENCY AND PAGE 36 ,
'~~~
, R A656E 66i6 AE~66MiR 6E~ ~ ~ ~ ~ ~ ~ ~ ~ ~~~ ~~ ~ ~ ~~ ~ ~ ~~ ~
fg CNSWERS -- SURRY 142 - 85/12/04-TON ROCERS i n
, 4 ANSWER 4.11 (1 50)
?
- 0) Nigh Seering Temp (+.25 ea) i Nigh Stator Winding Temp
, 7 minutes without CC b) NOV-1028 (ss.1pns*WyMehr er ee ME s& he.tv ud.M mire)
NOV-106C & D (ce4Aege (W kke svedhywkw) a NOV-100C L D i ** * ** w ~ #
) )
REFERENCE l Sorry AP-40, pp 3 ). EPE-0658 PWG-11 (3.9/3.9) ! L. 1 ANSWER 4.12 (1 50) . SURRY (+ b es) INORTH ANNA (+.3 ea) g.____ .... h 1 W
-0;.t;i.. ;nt C_..di;.. ... Wo.. l-RCS Press > 2000 Psis A increasing
-RCS Pressure > 2000 psig 1-RCS Subcooling > 50 Des F
-RCS Subcooling >. 50 Des F 1-PZR Level > 50%
-PZR Level > 50% l-SG Level > 10% or > 30% Advrs Cnte
-SC Level > 65% (WR) or > 17 % (NR) l OR ,
OR l-AFW Flow > 730 CPN i
-AFM Flow > 540 GPN I i 1
)y REFERENCE Y
Surry EP-2 00 foldout NA EP-0 Foldout page 006/0501 PWG-7 (3.8/4.2) , ANSWER 4.13 (1.00) icinizing radiationi heat stressi differential pressurei 02 deficiency (+.25 ea) REFERENCE Surry ADN 38, pp 3 NA ADN 20.9, pp 1 PNG-18! Knowledge of Safety Procedures (3.0/3.1) s'
-,,d..-,-~,ws - - , -- - , , , - - - -y3 ..-,w- -
*
- q
- 9; 3 ,
- 4. PT.0CEDURES - NORMAL, CBNORMAle EMERGENCY CND PAGE 37 ,
, ~~~~R I656[66I65[~665YR6[~~~~~~~~~~~~~~~~~~~~~ [
ANSWERS -- SURRY 142 -85/12/04-TOM ROGERS 1 L';" 0.00 .00; y
? ::- Sp {
REFERENCE Surry EP-1.00e 'pp 11 h ' [1 1 026/0208 PWC-11 (4.5/4 5) g ANSWER 4.15 ( .50) l' e' Informs operators to allow a 15 minute delay (+.25) before reclosing a breaker should it trip open (+.25) $' e REFERENCE i NA ADM 14.0, pp 9 ; Surry ADM 29.7e pp 25 i PNG-148 Knowledge of Tagging procedures (3.6/4.0) j ANSWER 4 16 ( .75) Cause not clearly understood (+.25) or safety related/important equipment cperated in an abnormal ar degraded manner (+.5) . 4 REFERENCE $ 2 Surry NA ADM ADM19.1814', PP pp i PWG-10! Recognizing abnormal indications (4.1/4.5)
) ^
3 l l. l i i(
.h(,
. ENCLOSURE 3 ,
Lt . S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: SURRY 182 REACTOR TYPEt PWR-WEC3 DATE ADMINISTEREDI 85/12/04 FXAMINERt TOM ROGERS APPLICANT! _________________________ INSTRUCTIONS TO APPLICANT: lise separate paper for the answers. Write answers on one side only. Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The PassinS Stade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be Picked up six (6) hours after the examination starts.
% OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE 'VALUE CATEGORY
,^ '
_l_1;;__ _25.00 _____ ___________ ________ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS 18.00 25.00 ' PLANT SYSTEMS DESIGN, CONTROL, ________ ______ ___________ ________ 6. AND INSTRUMENTATION g
"" ^^
25 0 _;;;;;__ ___1_0 _ ___________ ________ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL D ADMINISTRATIVE PROCEDURES, _((i$_____1__ ___________ ________ 8. CONDITIONS, AND LIMITATIONS 69.0 7^.
^
100.00 TOTALS FINAL GRADE _________________% All work done on this examination is my own. I have neither givan nor received aid. 5PPL5CEUT 5~55GUITUR5~~~~~~~~~~~~~~
~
. . # i '= . 1
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 2 1
g
----_--------- )
I UtlESTION 5.01 (1.00) ; Explain the difference between a prompt neutron and a fast neutron. ! OUESTION 5.02 (1.00) If reactor power increases from 1000 ces to 5000 cps in 30 seconds, what is l the startup rate in OPM? HUESTION 5.03 (1.00) A PWR is designed to operate like the Rankir.e Vapor Cycle shown on Figure $210. Which of the following equations could be used to calculate the cycle's thermodynamic efficiency?
- o. AREA WITHIN (a-b-e-d-a) / AREA WITHIN (a-d-f-e-a)
- b. AREA WITHIN (e-a-b-e-d-f-e) / AREA WITHIN (a-b-e-d-a)
- c. AREA WITHIN (a-d-f-e-a) / AREA WITHIN (e-a-b-e-d-f-e)
- d. AREA WITHIN (a-b-e-d-a) / AREA WITHIN (e-a-b-e-d-f-e)
GUESTION 5.04 (1.00) , Which of the following equations used to perform a PWR heat balance - calculation is correct? ,
- c. Grx = M(s) [h(s) -
h(fw)3 + M(bd) [h(bd) - h(fw)] + Orep
- b. Orx = M(s) Ch(s) - h(fw)3 + M(bd) Ch(bd) -
h(fw)3 - Drep
- c. Orx = M(s) Ch(s) -
h(fw)3 - M(bd) [h(bd) - h(fw)3 - Orep
- d. Orx = M(s) [h(s) - h(fw)3 - M(bd) Ch(bd) -
h(fw)3 + Orcp-NOTE: Notation Key G = Power M = Mass Flow Rate fu = Feeduater rx = Reactor bd = Blowdown s = Steam h = Specific Enthalpy (xxxxx CATEGORY 05 CONTINUED ON NEXT PAGE xxxxx)
l- . i'
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND -
PAGE 3 ) THERMODYNAMICS ! GUESTION 5.05 (1 00) The reactor trips from full power, equilibrium xenon conditions. Six h:ers later the reactor is brought critical at 10E-8 amps on the inter-o?diate range. If power level is maintained at 10E-8 amps which of the following statements concerning rod motion requirements for the next two hours is correct?
- o. Rods will have to be withdrawn since xenon will closely follow its normal build-in rate following a trip.
- b. Rods will have to be inserted since xenon will closely follow its normal decay rate following a trip.
- c. Rods will have to be rapidly inserted since the critical reactor will cause a high rate of burnout.
- d. Rods will have to be rapidly withdrawn since the critical reactor will cause a higher than normal rate of build-in.
QUESTION 5.06 (1 50) Indicate whether the following statements concerning rod worth are TRUE cr FALSE.
- o. One reason for overlapping rod groups is to minimize the effects of rod shadowing on total rod worth. j
- b. Both an RCS temperature increase and a buildup of fission product poisons will DECREASE rod worth,
- c. The maximum differential rod worth occurs at the point where the integral rod worth is maximum.
l l (xxxxx CATEGORY 05 CONTINUED ON NEXT PAGE xxxxx) l l a
i
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 4 GUESTION 5.07 (2.00)
Indicate whether each of the following fuel loading situations would rcsult in a 1/M plot that was CONSERVATIVE (under predicts criticality) cr NONCONSERVATIVE (over predicts criticality).
- 3. 9:t::t:r 1 ::t:d i:: f:: ft:: ::r: '; urce), dthTu TR4 (0.5)
- b. Detector located too near core (source). (0.5)
- c. Loading core from center (source) towards detector. (0.5)
- d. Loading highest worth assemblies firsti lowest worth last. (0.5)
GUESTION 5.08 (1.50) Indicate whether the Total Power Coefficient gets MORE NEGATIVE, LESS NEGATIVE, or DOES NOT CHANGE for the following conditions.
- o. From low power to high power at BOL.
- b. From low power to high power at EOL.
- c. From BOL to EOL st a constant power level.
QUESTION 5.09 (1.50) Indicate whether the following will INCREASE, DECREASE, or HAVE NO EFFECT on the available (actual) Net Positive Suction Head (NPSH).
- c. Increasing pump speed.
- b. Increasing pump suction temperature.
- c. Inc.' easing system pressure.
l (xxxxx CATEGORY 05 CONTINUED ON NEXT PAGE.xxxxx) i i l l l l t
\ ~
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 5 QUESTION 5.10 (1 00)
Which of the following actions will increase the DNBR? Assume Mode 1 and na reactor trip occurs.
- c. Tripping a reactor coolant pump.
- b. Closing reactor coolant loop stop valves in an operating loop.
- c. Closing reactor coolant stop valves in a loop that has a tripped reactor coolant pump. Q
- d. Closing a main steam trip valve.
QUESTION 5.11 (1.50) Assuming a symmetrical (ideal) axial flux shape, match the CONDITION in
-Column A to the LOCATION that it would occur in Column B.
COLUMN A COLUMN D
- o. MINIMUM Critical Heat Flux 1. BOTTOM
- 2. Between BOTTOM & MIDDLE
- b. ' MAXIMUM Actual Heat Flux 3. MIDDLE
- 4. Between MIDDLE & TOP
- c. MINIMUM DNBR 5. TOP (tL: 3 TION 5.12 (1.00) i As the core ages, the buildup of Pu-240 causes the Fuel Temperature Coefficient (pen / degree F) to become more negative. With this change accurring, why does the Doppler Only Power Coefficient (pca/% power) become less negative as the core ages?
ItVESTION 5.13 (2.00) TWO major factors effect differential boron woth over core life. List these TWO factors AND indicate how (MORE NEGATIVE or LESS NEGATIVE) they offect differential boron worth. (xxxxx CATECORY 05 CONTINUED ON NEXT PAGE xxxxx)
. i
- j
^
- 5. 'iHCORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 6 QUESTION 5.14 (1.00)
K(z) the height-dependent correction factore is used to modify the Nuclear Heat Flux Hot Channel Factor limit given in the Technical Specifications. Why is this correction factor necessary? I (xxxxx END OF CATEGORY 05 xxxxx)
,_,----c . . - , , , . , - _ _ . - - - ,.,-__-_._.-.e. . _ . _ . . ._ . , , , . . _ -
.. . . {
l, .
- 6. PLANT SYSTEMS DESIGNe CONTROL, AND INSTRUMENTATION PAGE 7 l' 1
CUESTION 4.01 (1.00) What is the maximum purification flow rate the mixed bed domineralizers [ cre sized to except? - HUESTION 6.02 (1.00) How does an operator disable a motor driven AFW Pump from starting on cny of the automatic start signals? UUESTION 6.03 (1 00) The SG 1evel program uses ______ as an input for reactor power. QUESTION 6.04 (1.00) What two different conditions will prevent the feedwater regualting bypass valves from being opened. Do not include control malfunctions. QUESTION 6.05 (1.00) Explain how the Reactor Vessel Level Indicating System is compensated to maintain the required accuracy during a LOCA. HUESTION 6.06 (1.00) Which of the following malfunctions could cause one of the over temperature delta T trip bistables to trip?
- o. Controlling turbine impulse pressure channel failing low.
- b. Power range N43 lower detector failing low.
- c. Reactor coolant flow detector failing low.
- d. Controlling pressurizer level channel failing low.
(xxxxx CATEGORY 06 CONTINUED ON NEXT PAGE xxxxx) o l
\
A M s (
- 4. PLANT SYSTEMS DESIGNe CONTROL, AND. INSTRUMENTATION PACE 8 e
GUESTION 6.07 (1.00) : M3tch the following conditions with the expected indication provided by 4 the rod speed indication meter. The selections may be used more than once.
}
- 1. Immediately before an operator removes the N44 a. O s/m.
fuses because of a failed high detector. Rods in AUTO with no temperature mismatch. b. 8 s/m.
- 2. Rods in MANUAL with a 10 F temperature c. 40 s/m.
mismatch.
- d. 48 s/m.
- 3. Rods in AUTO with a 1 F temperature mismatch.
- e. 72 s/m.
- 4. Rods in AUTO with a 4 F temperature mismatch.
- f. 88 s/m.
- 5. Rods in AUTO with one of the Tave control instruments failed low.
QUESTION 6 08 (1.00) Match the following loads to their normal power supply. The selections oay be used more than once.
- 1. 200 kw PZR backup heaters. a. 4160V Bus 1A.
- 2. Low head SI pump 1A. b. 480V Bus ipJ. ?
- 3. 4140V Bus 1A breaker control power. c. 4160V Emergency Bus 1J.
- 4. Fire protection cabinets.
' d. 120VAC Vital Bus.
- 5. Condensate pump 1A.
- e. 125VDC Vital Bus.
- f. 125VOC Intake Structure Bus.
- g. 4160V 1J Stub Bus.
h, mov cm u 17 i, , 490 V Bus IN , ~ (xxxxx CATEGORY 06 CONTINUED ON NEXT PAGE xxxxx)
-- , - . , - , a , - + - - , - - -e ,,e ,- ,,--o -- mw - m- ,- ,
1
- - . i
, 3
- 6. PLANT SYSTEMS DESIGNr CONTROL, AND INSTRUMENTATION - PAGE 9 4 -
4 CtlFSTION 6.09 (1.00) [ Fer the following air controlled valves, indicate if their failed position jl io open or closed upon loss of control air. i D. Pressurizer spray valves. [
- b. Pressurizer PORVs.
- c. Charging flow control valve.
- d. Steam dump valves.
- o. RHR flow control valve 605.
4 OUESTION 6.10 (1.00) What is the normal power source for the EHC System? QUESTION 6.11 (1.00) If emergency diesel generator 1 is taken out of service for preventive . caintenance when a blackout occurs on Unit 1, how will EDG 3 respond if, during its loading sequence, an SI actuation signal is received from Unit 2 with EDG 2 failing to start? GUESTION 6.12 (1 00) The cold leg accumulators are designed so that ___ .. accumulator (s) eill cover ._.... of the core.
- o. 1, 50%
- b. 2, 50%
- e. In 100%
- d. 3r 90%
4UESTION 6.13 (1.00) What is required to clear a ' PZR NDT Overpressure System Reg'd* alarm? (***** CATEGORY 06 CONTINUED ON NEXT PAGE xxuzz) i l
+
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- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE - 10 }
,,, \
CllESTION 4.14 (1.00) E Which valves get an automatic shut signal upon receipt of a process ;I v;nt particulate and gas monitor alarms? t s HUESTION 6.15 (1.00) List two Places the reactor cavity purification system can dewater the I roactor cavity to. ;l (tuESTION 6.16 (1.00) ,, What interlocks must be satisfied to move the fuel conveyor car between the reactor building and the fuel handling building? HUESTION 6.'17 (1 00) i What are the two sources of waste gas to the waste gas surge drum? QUESTION 6.18 (1 00) What is the minimum logic that must be satisfied for the condenser to be cvailable for the steam dumps (C9)? i (xxxxx END OF CATEGORY 06 xxxxx)
~ - -, ..,,---,-,--v, ,.--e--, - - - - - - --..,,.,,_.,.----,-,,.,---,,,,,.m..-.n-_g,..e-,.----,-, .w .
- e. n.a .-_. , -, , - , - -
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+,
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 11 !,
~~~~EI656E665CIE'66UTR6E-~~~~~~~~~~~~~~~~~~~~~~~
t GUESTION 7.01 (1.00)
+
Which of the following statements regarding Axial Flux Difference is CORRECT?
- o. Reactor power CANNOT be increased above 50% rated thermal power unless the indicated AFD is within the target band.
- b. If the indicated AFD is outside the target band for more than i HOUR CUMULATIVE over a 24 HOUR periode with reactor power between 50 and 90% of rated thermal powere EITHER reduce thermal power to less than 50% within 30 minutes and reduce the power ranSe Nuetron Flux High setpoint to less than 55% within 30 minutes.
- c. If indicated AFD is outside the target band and thermal power is Sreater than 90% rated thermal power, within 1 HOUR AFD aust be restored within the band or power reduced to < 90%.
- d. Below 50% rated thermal powere there is NO penalty for being outside the target band due to the fact that uneven xenon buildup in the core does not have an adverse impact at low power levels.
QUESTION 7.02 (1.00) , Which of the following is correct as it applies to NON-EMERGENCY procedur' deviations?
- o. The SNSOC must approve procedural deviations prior to performing the deviated step.
- b. A ' Request to Change Procedure" form is required for a temporary procedure deviation.
- e. A ' Procedure Deviation' form is NOT required for typographical errors in frequently used procedures.
- d. The SNSOC may pre-approve procedural deviations.
l (xxxxx CATEGORY 07 CONTINUED ON NEXT PAGE xxxxx)
, -- _. - - -- _. ,. -- = - - - .
V 4
. L
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- 7. PROCEDURES - NORNAL, ABNORNAL, ENERGENCY AND 'PAGE 12 }
--- Ei5iiE55iait C5Eiiat------------------------
4 l GilFSTION 7 03 (1.50). F For each of.the following, indicate whether the conditions violate critical ( cafety-function (CSF) red path criteria or not. o) Pressurizer level of 5% and RVLIS upper he:d 80% b) Total AFW flow 400 spa with all S/G 1evels < 41 { c) Containment pressure 65 psis . QUESTION 7.04 (1.00) List the 4 methods-siven in the S/G Tube Rupture E0P to identify which S/G is ruptured. ; 4
?
HUESTION 7 05 ( .75) Radiation Workers'can be allowed, by authorization of __________, up to _____ Ren per quarter or. _____ Ren per year provided their lifetime dose does not exceed 5(N-18), provided the proper forms (NRC-4, HP-12) are completed. I 40ESTION 7.06. (1.50) While removing the reactor vessel he'ade the lifting process is stopped when ' the head is at a height of 1 inch, 2 feet and 128 inches to verify three conditions are being met (or are not occurring). List these three items of interest. QUESTION 7.07 (1.00) Following a valid reactor trip and safety injection, what are the Reactor Coolant Pump Trip Criteria? RUESTION 7.08 (1.00) List the immediate aperator actions to initiate emergency boration if it is required on an Anticipated Transient Without Trip condition. (xxxxx CATEGORY 07 CONTINUED ON NEXT PAGE Exxxx) e
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- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE '13 {
l:
~~~~Rd65UL66 CAE~6UUTR5t
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b QUESTION 7.09 (2.00) j List ALL the immediate action sub-steps from EP-1 00, ' Reactor Trip or !' Safety Injection
- that allow you to accomplish the following immediate cc.tions. ;
c) Verify if SI is actuated (1.0) I b) Verify SI Valve Emergency Lineup (1.0) 4 CUESTION 7.10 (1.50) Answer the following questions regarding a 'Non-Recoverable Loss of Air's (AP-40) on UNIT 1 4 c) What are the three conditions that require the RCPs to be secured due to the loss of CC? (0.75) b) When the INTAKE CANAL LO LEVEL TRIP is received, what CW & SW MOVs are L manually operated? (0.75) GUESTION 7.11 (1.50) Liste~ in order of priority from high to low, all the personnel (by title) who may relieve the Interin Station Emer3ency Manager. QUESTION 7.12 (1.50) List the SI termination criteria following a LOCA.
':UCO TICM 7.10 ; .50; delst '%in TRUE or FALSE:
If on an SI, a containment spray pump CANNOT be started, the ' Response Not Obtained* step requires verification that its associated Chemical Addition Tank MOVs (MOV-CS-( )03 A & C or B & D) be closed. (xxxxx CATEGORY 07 CONTINUED ON NEXT PAGE xxxxx)
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- 7. PROCEDURES - NORMAL, ABNORMAle EMERGENCY AND PAGE ~14 $
--- i;515t5515st E5HiR5t------------------------ !
RUESTION 7.14 ( .50) ) 1 What is the purpose of a '15 Minute Headway
- tas? ;
HUESTION 7.15 ( .75) $ i What constitutes a Class II reactor trip? } HUESTION 7.16 (1.00) f On a steam generator tube rupture casualty, list the correct priority (in order of highest to lowest) of methods to depressurize the RCS from the choices below!
- 1) Auxiliary Spray Normal Spray i 7)
- 3) Pressurizer PORV i
1 3 7 (***** END OF CATEGORY 07 xxxxx)
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- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 15 ,
7
=
00ESTION 8.01 (1.00) { With the unit 1 reactor at 360 desFe which one of the RCS chemistry k cnalysis values given below is between the steady state chemistry limit'and V the transient limit? ! i O. Fluoride 1 4 ppa ,,
- b. Boron 2000 ppa
- c. Dissolved oxygen 1.4 ppa
- d. Chloride 0.14 ppa j i
( QUESTION 8.02 (1.00) Which one of the following conditions requires immediate action (in cddition to that stated in the choice) according to Tech Specs with unit 1 critical? !
- o. The shutdown margin is 1.8
- b. One train of heat tracing on the BAT is inoperable and being repaired.
- c. One reactor coolant pump in a non-isolated loop is left operating after the loss of one pump.
3
- d. Two of the three charging pumps are inoperable while two are inoperable on unit 2.
J 4 (xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx) e--, . - - - . - - - , . - - , . - - - w - - - - - - -
( M E s 1
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- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PACE 16 GUESTION 8.03 (1.00) ,,
-During unit 1.startup with'the reactor about 2% powere you find that the J-PORV block valve is stuck open and incapable of closing. Which of the {;
following is a CORRECT sction (see the attached LCO, Fig 8 2)? i
- 0. Continued operation is allowed provided the PORV is operable and power is removed from the block valve.
- b. With power removed from the block valve in accordance with 6.c and -
keff reduced to 0.989 in accordance with 6.ar you are still not t allowed to maintain Tave>547 desF. i
- c. The PORV must be closede power removed from the solenoid valve and the block valve must be repaired within one hour if power operation is to continue. .
- d. Since the block valve is incapable of closing, you must proceed to :
intermediate shutdown within the next 4 hours and cold shutdown within j the followins 30 hours. QUESTION 8.04 (1.00) Which one of the following situations, while the reactor is critical at 5% reactor powere 3 loop operation, exceeds a tech spec limit? .
- o. One shutdown control rod not fully withdrawn. 2
- b. With control bank C at 0.0 fraction inserted (fully withdrawn), bank D is at 0.55 fraction inserted.
- c. Control bank D is inserted further than would be permitted by tech specs if one rod were inoperable.
- d. The position indication accuracy of two group C linear variable differential transformers (LVDT) has been determined to be different than the group step demand position by 8 steps on one LVDT and 10 steps on the other LVDT.
(xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx) t
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_I_____ _ ___ _ _____ _ __ $_ _ _ $__ O_bf _ _f0 f QUESTION 8.05 (2.00) Match the positions in column A with the authority and responsibility j ctatements in column B insofar as they relate to lines of authority or ; cdministrative controls during fuel handling on either unit 1 or 2. The . parenthetical number following the column A position indicates the proper number of choices from column B. Some column B choices may be used more
~
than once or not at all. A. B. A. auxiliary operator (1) 1. has veto authority over the shift supervisore superintendent of 3 operations and superintendent , technical services in all matters concerning radiation safety. B. licensed operator (2) 2. must be in the control room during refueling. ; C. supervisor-enar servs (1) 3. must be in the containment during refueling. D. senior licensed operator (2) 4. assists the shift supervisor , E. shift supervisor (2) 5. responsible for ensuring that all routine chemical analyses and - evaluations are properly performed ? during refueling. J F. health physics section (1) 6. reports directly to the station manager. G. supervisor-health phys.(1) 7. direction over technical support f.or the entire refueling operation.
- 8. reports directly to supt. of ops.
- 9. must have direct communication with control room during refueling
- 10. directs the fuel handling operation (xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE ***xx) 6
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- 8. ADMINISTRATIVE PROCEDURES CONDITIONSr AND LIMITATIONS PAGE 18 4
6 0UESTION 8.06 (1.00) f Complete the following tech spec specifications.
- o. The quantity of radioactivity contained in each unit 1 gas storage j tank shall be limited to less than or equal to ______ curies of noble g gases. ,
- b. With the quantity of radioactive material in any gas storage tank i exceeding the above limite immediately ______ and within 48 hours 4 reduce the tank contents to within the limit.
QUESTION 8 07 (1.00) A quarterly surveillance requirement of tech specs may be extended up to ' ______ days without declaring the. component inoperable due to the curveillance testing not being performed. Assume all previous , curveillances on the component were performed exactly on their due date. QUESTION 8.08 (1.00) Fill in the blanks. The specific activity of the secondary coolant system chall be < _____ microcuries/cc Dose Equivalent I-133. The accident this is based on is a ______ . QUESTION 8.09 (1.00) According to administrative procedures, an individual shall not be permitted to work more than ______ hours in any 24 hour period, nor more than ______ hours in any 48 hour period, nor more than ______ hours in any ceven day period (excluding shift turnover time) unless authorized at the assistant station manager (ORM) level or above. QUESTION 8.10 ( .50) According to emergency plan implementing procedurer EPIP 5.03, during a cite emergency, personnel inside the protected area and unaccounted for l chall be determined within ______ minutes of declaration of the emergency. (xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx)
,- , - - - ----a -- - - , - , . - . , --, --
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- 8. ADMINISTRATIVE PROCEDURES, CONDITIONSr AND LIMITATIONS PAGE 19 [
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QUESTION 8.11 (1.50) f# Aosume the fire protection spray and sprinkler system to the cable vault is 4 inoperable. State your immediate actions to comply with tech specs including the following as applicablet personnel assignmentse changes to i reactor operating status, equipment rerouting or provision, and action i times. ! f OUCOTICM 0.12 s4.00, e@eL W List in general, the four notifications and determinations that must be ' ocde after initiating EPIP-5.05 ' site evacuation" but prior to actually sounding the emergency alarm. QUESTION 8.13 (1 00) , Is a pump operable if its control switch is in ' pull-to-lock'? Explain. I OUCOTIOM 0.11 '1.00; J.tle'.TIL N When is the unit 2 incore monitoring system required by tech specs to be cperable? l QUESTION 8 15 (1.00) During emergency situations unnecessary operational personnel and off duty ~ i SRO's and RO's should be available to the shift supervisor as needed. Describe their availability in terms of where they report and how they communicate with the SS. (tUESTION 8.16 (1.00) During the course of station operationse bypassing some safety functions cay be required in order to perform certain tests or operations. What three administrative requirements assure you (as shift supervisor) that the bypassing is properly controlled and documented? (xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx) 1 l l
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- 8. ADMINISTRATIVE PROCEDURES, CONDITIONSr AND LIMITATIONS PAGE 20 QUESTION 8.17 (1.00)
DOfiner using a major reactor system or component, what is meant by
' controlled leakage
- as that term is used in unit 2 tech specs.
1 (xxxxx END OF CATEGORY 08 xxxxx) (xxxxxxxxxxxxx END OF EXAMINATION xxxxxxxxxxxxxxx) i l l i 1 l l
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 21 l
--- y y y ANSWERS -- SURRY 182 -85/12/04-TOM ROGERS
?
C.NSWER 5.01 (1.00) Prompt refers to the time between a fission event and and the release of a neutron.[0.53 Fast refers to the energy (speed) of the neutron [0.53. REFERENCE Nils Basic Nuclear Concepts. ANSWER 5.02 (1.00) 1.4 DPM. Free P=Po10Esur(t) s REFFRENCE NUS Basic Nuclear Concepts. AN"aWER 5.03 (1.00) d REFERENCE Ceneral Physics, NTAFF, pp. 137 - 142 ANSWER 5.04 (1.00) b REFERENCE Ceneral Physics, NTAFF ANSWER 5.05 (1.00) O REFERENCE HN10, pp. I-5.63 - 77 b 9
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDSe AND PAGE 22
--- y y ANSWERS -- SURRY 142 -85/12/04-TOM ROGERS ;
t H e ANSWER 5.06 (1.50) I
- o. FALSE (0.5)
- b. FALSE (0.5) ,
- c. FALSE (0.5)
REFERENCE Westinghouse Nuclear Training Operations, p. I-5.36 - 43 ANSWER 5.07 (2.001-(l*TD)
- . MCMCCNCE."VATIVC- Dek&L "I d (0.5?
- b. NONCONSERVATIVE (0.5)
- c. NONCONSERVATIVE (0.5) ;
- d. CONSERVATIVE (0.5)
REFERENCE Westinghouse Nuclear Training Operations, pp. I-4.19 - 21 ANSWER 5.08 (1.50)
- o. LESS NEGATIVE (0.5)
D. MORE NEGATIVE (0.5) )
- c. MORE NEGATIVE (0.5) ,
REFERENCE Westinghouse Nuclear Training Operations, p. I-5.29 CNSWER 5.09 (1.50) O. DECREASE (0,5)
- b. DECREASE (0.5)
- c. INCREASE (0.5)
REFERENCE Ceneral Physics, HT&FFr p. 320 I
? 5
., p THEORY OF NUCLEAR POWER PLANT OPERATIONr FLUIDS,'AND
^
- 5. PAGE 23
--- y y y g--------------------------------------
CNSWERS -- SURRY 112 -85/12/04-TOM ROGERS j i i i CNSWER 5.10 (1.00) e (stops core bypass flow in loop.) REFERENCE , NAPS UFSARr 15.2 6.1.1. CNSWER 5.11 (1.50)
- o. 5 (0.5)
- b. 3 (0.5)
- c. 4 (0.5) .
REFERENCE Ceneral Physics, NTAFFr pp. 228 - 230 ANSWER 5.12 (1.00) As the core ages, the pellet and clad creep together causing an increase in gap conductivity (0.4 pts.). This causes a smaller delta T of the fuel for a given power change (0.3 pts.). The smaller delta T of the fuel causes a smaller change in reactivity for a given power change (0 3 pts.). (pca/F) X F/% power) = pcm/% power more nes less less nes < REFERENCE Westin3 house Nuclear Training Operations, p. I-5.22 ANSWER 5.13 (2.00)
- 1. Boron Concentration Decreases (0.5 pts.) - MORE NEGATIVE (0.5 pts.)
- 2. Fission Product Buildup (0.5 pts.) - LESS NEGATIVE (0.5 pts.)
REFERENCE Westinghouse Noelear Training Operations, p. I-5.31 [ I
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 24 $
____,g gg g gg g ______________________________________ , ANSWERS -- SURRY 142 -85/12/04-TOM ROGERS : 1 t ANSWER 5.14 (1.00) During a LOCA the core may be blown dry and reflooded by the CLAs. Since the upper half will be reflooded laste more restrictive limits are placed on the upper half of the core during normal operations. REFERENCE ~ General Physics, HTAFF, p. 249
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- 6. PLANT SYSTENS DESIGN, CONTROL, AND INSTRUNENTATION PAGi 25 -
ANSWERS -- SURRY 112 -85/12/04-TON ROGERS . ANSWER 6.01 (1.00) & 120 spa. REFERENCE SPS CVCS Sys Des. NAPS CVCS Advance Skills Training. ANSWER 6.02 (1.00) Place switch in STOP, Pull to lock position. REFERENCE - , SPS AFW Sys. Des. ANSWER 6.03 (1.00) Turbine first stage pressure. REFERENCE SPS SGWLC Sys. Des. NAPS SGWLC & P Advance Skills Training. ANSWER 6.04 (1.00) 8 0.5 points each: '
- 1. SGWL high.
- 7. SI signal present.
REFERENCE SPS Nain Feed Water Sys. Des. f
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- 6. PLANT SYSTEMS DESIGNr CONTROL, AND INSTRUMENTATION PAGE 26
______________________________________________________ f CNSWERS -- SURRY 182 -85/12/04-TOM ROGERS r 0 3 CNSWER 6.05 (1.00) i 0 0.25 points eacht f It receives inputs from each of the following! 1-
- 1. Temperature of impulse line. ,
- 2. RCS temperature. '
- 3. Nide range pressure.
cnd
- 4. The d/p cell is located outside of containment. (
REFERENCE SPS RVLIS Sys Des. f NAPS RVLIS Advance Skills Training. i ANSWER 6.06' (1.00) } b REFERENCE SPS TS 2.3-2. NAPS RPS Adv. Skills Training. ANSWER 6.07 (1.00) i G 0.2 points eachi 1 1.-ar 2-dr 3-ar4-cr5-a. och "Bu) , REFERENCE SPS Rod Control Sys Des. NAPS Rod Control Lesson Plan. ANSWER 6.08 (1.00) 90.2 po ts eacht 1-br 2 e 3 #, 4-er 5-a. g e. REFERENCE SPS Electrical System Lesson Plans. s
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PLANT SYSTEMS DESIGNr CONTROL, AND INSTRUMENT 4 TION PACE 27 A 6.
---- .. ---------------------------------------------- 1 ANSWERS -- SURRY 1&2 -85/12/04-TOM ROCERS ANSNER 6.09 (1.00)- l t 0.2 points eacht j
- o. FS. i
- b. FS. -
- c. FO.
- d. FS. }
- c. FS. ,
REFERENCE 'f SPS Sys Des. for Primary systems, Main Steam, and RHR. ?
'P ANSWER 6.10 (1.00) j Pormanent magnet generator on the turbine shaft.
REFERENCE ~ SPS Turbine Protection Sys Des. ANSWER 6.11 (1.00) It will dedicate itself to the SI plant- Unit 2. , I REFERENCE SPS Emergency Power Dst. Sys Des. 1 ANSWER 6.12 (1.00) b REFERENCE SPS SI Sys Des. NAPS SI or ECCS ESF Lesson Resources. ANSWER 6.13 (1.00) Place both PORV key switches to EhABLE [0.53 ond both motor operated PORV isolation valves OPEN [0.53. I - i-,-.. i
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- 6. PLANT SYSTEMS DESIGNr CONTROL, AND INSTRUMENTATION PAGE '28
.. ____________________________________________________ j CNSWERS -- SURRY 112 -85/12/04-TOM ROGERS ;
4 REFERENCE J
.SPS Primary Systems Sys Des. j 1
ANSWER 6.14 (1.00) l Discharge valves from containment vacuum system [0.53 and waste decay - tanks C0.53. (RCV-GW-160e FCV-GW-260, FCV-GW-101) : REFERENCE 3 SPS Rad Monitoring Sys Des, Table 3.7-5. d ANSWER 6.15 (1.00) j e 0.5 points each! j
- 1. RHST. 1
- 7. Containment sump.
{ REFERENCE t SPS Reactor Cavity & Purif. Sys Des. ANSWER 6 16 (1.00) , 9 0.5 points each! j Both lifting frames down. , Transfer tube valve open. REFERENCE SPS Fuel Handling Sys Des. ANSWER 6.17 (1.00) i f 0.5 points each!
- 1. Catalytic recombiner subsystem
- 7. Boron recovery gas stripper.
REFERENCE SPS Gaseous Waste Disp. Sys Des. I
.s - ..
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- 6. PLANT SYSTEMS DESIGNe CONTROL, AND INSTRUMENTATION -
P' AGE 29
----------------------------------------------~~------
5 ANSWERS -- SURRY 122 - 85/12/04-TOM ROGERS 'l
}
' ANSWER 6.18 (1.00) b 9 .5 points eacht J Fct NAPS j
- 1. 2/2 pressures. -j -
- 2. 3/4 cire water pump breakers shut. ?j REFERENCE -)
NAPS Steam Dump Advance Skills Training. ): 4 Fce .sps I
- h. CmJ. Um (W&
'l4 Cor d. d.d Is+ w b n No f 4A d. is.
VfF: ses store so,*ue Proom x. ' I E, h
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~~~~
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE :10 { R A6 56E ____________ 6 656 AE"_66U T 6 6E ~~ ~~ ~ ~ ~~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~, ~ CNSWERS -- SURRY 142 -85/12/04-TOM ROGERS 4
?
p ANSWER -7.01- (1.00) Jf A. eYd W I REFERENCE NA TS 3/4.2.1 -) SURRY TS 3.12-B.4 0 001/050T PWG-8 (3.6/4.5) i, ANSWER 7.02 (1.00) ? d {< 1 REFERENCE Sorry ADM 29.4r pp 17 f PNG-23: Shift Staffing and Activities (2.8/3.5)
^
ANSWER 7 03 (1.50) c) No (+.5 ea) ; b) No l c) Yes .g REFERENCE NA CSF F-0.4, F-0.5, F-0.6 PNG-10 Recognize abnormal indications for E0Ps (4.1/4.5) ANSWER 7.04 (1.00)
-Unexpected rise in S/G 1evel (+.25 ea)
-High radiation on a S/G blowdown line
-High radiation on an MS line monitor
-High radiation as determined by sampling and analysis REFERENCE Surry EP-4.00, pp 3 NA 2-EP-3, pp 2 EPE-0381 EA2.03 (4.4/4.6)
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- 7. PROCEDURES - NORMALe ABNORMAle EMERGENCY AND PAGE 31 "[
--- i;5i5E5515;E 55siE5E------------------------
ANSWERS -- SURRY 112 -85/12/04-TOM ROGERS g ( .75) i; ANSWER 7.05-S' :
~
Corporate Managementi 2.75; Adri (+.25 ea) er v!,ketoM 4 1.t T REFERENCE wa'rf,M ar 1.tr Surry NP-2.8, pp 2 PWG-15; Knowledge of facility Radeon Requirements (3.4/3.9) e ANSWER 7.06 (1.50) Surry: levelness of the head, binding of guide studs, RCC drive shafts are
- not being lifted (+.5 es response) radibhj !
; North Anna: Levelness, Upper internals are not being liftede no Control )
r od dr ive shaf ts ar e being li f ted ( + .5 ea ) ,westaf f/**/a/I PPr. nad REFERENCE
> Surry OP-4.1, PR.17 ; NA OP-4.1, pp 26 034/0001 PNG-7 (2.9/3.7)
- i ANSWER 7.07 (1.00) y g g gg gg
, 1) Verify Charging /SI flow (tL5 ea)
- 2) RCSPressure<1600psis(Surryh{,LC)
I
< 1230 psis (NA) - REFERENCE SONP Foldout Page NA Foldout page for 2-EP-0 Surry Foldout page for EP-1 00 g
003/000; PWG-10 (4 1/4.4) l. l f 1 L 4 1 O L .
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.7. PROCEDURES - NOR' Male ABNORMALe. EMERGENCY AND PAGE 32 [
~ ~~~~It3656E665 CIE~56Mirt6E---------------~~------ _ f
. ANSNERS -- SURRY 182 -85/12/04-TOM R0GERS i i
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ANSNER 7.08 ' .(1.00) Ver .I/CHG ( daps.unnipsis[(Surp)
;,. ~ ? ht-f C ek 8 pr sur <250 " ' '
I' '
)
e e e 41 o & ef w / w_ _ ..~n , *g O~ i W Switch BATP to Fast speed G.. tea) M Open MOV-( )350 N { REFERENCE j Surry ECA-le pp 4 j NA 2-ECA-1, pp 3 EPE-0297 PWC-11 (4.5/4.7) l i 4 ANSWER 7.09 (2.00) f 4
- 0) SI Initiated annunciator LIT (+.25 es)
Any SI First-out annunciator LIT SI Pumps running EDGs running b) CHG/SI Pump RNST MOVs open .. High Head SI to Cold Les MOVs open i Low Head Pump RNST MOVs open ( Low Head Pump Cold Les Discharge MOVs open REFERENCE Surry EP-1.00, pp 4 & 7 006/0008 A3.02 & A3.03 (4 1/4 1) - ANSWER 7.10 (1.50) o) High Bearing Temp (+.25 ea) , High Stator Winding Temp l 7 minutes without CC , ! b) MOV-102B (SeJ psHPS"M4kesler se et th supp4 lader isoldwk n/pe) MOV-106C L D {tondsute l'Wu'Itt is olsk *'al*eJ) MOV-100C & D ( .. " A,#t t ., - ! REFERENCE Surry AP-40, pp 3 EPE-0658 pWG-11 (3.9/3.9) l i l e J
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- 7. PROCEDURES - NORMAL, ABNORMAle EMERGENCY AND PAGE 33
--- Ei5iBE55iEEE 55EiiBE------------------------ !
CNSWERS -- SURRY 112 -85/12/04-TOM ROGERS CNSWER 7.11 (1.50) g
- 1) Station Manager (+.25 es response, +.5 for correct order) !
- 2) Assistant Station Manager i
- 3) Superintendent Operations i
- 4) Superintendent Technical Services 1 i
REFERENCE ) Surry SEPe pp 5.7 { PNG-36: E plan (2.9/4.7) i
?
1 ANSWER 7.12 (1.50) } SURRY (+. b ea) INDRTH ANNA (+.3 ea) f; _____ i__________ C ;r.t ... c r.t C ; r.f i t i : .: Bd I-RCS Press > 2000 psis & increasing
-RCS Pressure > 2000 psis I-RCS Subcooling > 50 Des F ,
-RCS Subcooling > 50 Des F l-PZR Level > 50%
-PZR Level > 50% l-SG Level > 10% or > 30% Advrs Cnto .
-SG Level > 65% (WR) or>17%(NR) 1 OR [
OR l-AFW Flow > 730 GPM
-AFW Flow > 540 GPM i l
REFERENCE j Surry EP-2.00 foldout i NA EP-0 Foldout page 006/0507 PNG-7 (3.8/4.2)
" C'!CP '.10 .5^!
True
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y (, h[ W REFERENCE Surry EP-1.00, pp 11 026/020; PNG-11 (4.5/4.5)
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~ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 34
~~~
RA656E6656AE 66UTR6E~~~~~~~~~~~~~~~~~~~~~~~~ ____________________ y ANSWERS -- SURRY 182 -85/12/04-TOM ROGERS. j ma .
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' ~
ANSWER 7.14 (~-.50)
~
' Informs operators to allow a 15 minute delay (+.25) before reclosing a ,
br'eaker should it trip open (+.25) p REFERENCE i NA ADM 14.0, pp 9 T Sorry ADM 29.7e pp 25 I i PWG-148 Knowledge of Tagging procedures (3.6/4 0) 9 ANSWER 7.15 ( .75) j Cause not clearly understood (+.25) or safety related/important equipment j cperated in an abnormal ar degraded manner (+.5) : REFERENCE Surry ADM 14, pp 2 NA ADM 19.18, pp 1 , PWG-108 Recognizing abnormal indications (4.1/4.5) H a ANSWER 7.16 (1.00) Normal Spray >>>PZR PORV>>> Auxiliary Spray (+.33 for each in correct spot) REFERENCE i Surry EP-4.00, pp 12/13 ; NA 2-EP-3, pp 9/10 EPE-038i EK 3.01(4.1/4.3) s o I 1 1
. _ , , _ . _ _ _ _ _ _ _ ___ , _ _ _ , . . , - _ - . . - . _ _ . _ . .._,._,.--,_,,__.,_,._---._.,_-.,...e_ .-.,__.-.-..,m-- --
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- 8. ADMINISTRATIVE PROCEDURES, CONDITIONSr AND LIMITATIONS' PAGE 35 $
ANSWERS -- SURRY 182 -85/12/04-TOM ROGERS = ii ANSWER 'O.01: '(1.00) I$ (o). REFERENCE j 4 NA TS 4-20 SR TS 3.1-20 ) 004/010 K5.04 (2.6/3.3) h f 4 ANSWER 8.02 (1.00) j (c) I REFERENCE .I SR TS 3/4 3.12-2, 3.2-3, 3.1-2, 3.2-3. j ANSWER 8.03 (1.00) (b) HEFERENCE [ . SR TS 3.1-5 and 1.0-2. j II ANSWER 8.04 (1 00) (0) REFERENCE h) -t SR TS 3.12-1, 3.12-1Ar 3 12-1A and -3, 3.12-10. t ANSWER 8.05 (2.00) j A4 E 8,10 B 2,4 F1 3 07 G6 ) D 3r9 REFERENCE SR ADM 34 p 1. 1
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- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PACE 36 I S ANSWERS -- SURRY 112 -85/12/04-TOM ROCERS [
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" ANSWER 5.041 (1100)~ ,
j c) SR 24,400 +or- 2,000 ! MA 25,0_00 +or- 2,000 } b) SRA'NA suspend all additions of radioactive material to the tank.
, i REFERENCE ;
f'R TS 3.11-7,8. NA TS 3/4 11-17. { ANSWER 8.07 (1.00) ; 23 +or- 2 , REFERENCE i SR TS 4.0-1 j NA TS 3/4 0-2. ANSWER 8.08 (1 00) 0 10 microCi/cc +or-25%f steam line rupture (or any accident which results in the release of the entire contents of the units steam generators to the ,
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otmosphere). REFERENCE SR TS 3.6-2r 3.6-5. NA TS 3/4 7-8, B 3/4 7-3. ANSWER 8.09 (1 00) 16, 24, 72. REFERENCE SR ADM-3 p 4. NA ADM-20.3 p 4. 4
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- 8. ADMINISTRATIVE PROCEDURES, CONDITIONSr AND LIMITATIONS PAGE 37 [l ANSWERS -- SURRY 182 -85/12/04-TOM ROGERS ~l 4
- 1
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8.10 ( .50) , 30 -
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REFERENCE
- SR EPIP 5.03 p 1 NA'EPIP 5.03 P 1 i, 5.
ANSWER 8.11 (1.50) e Establish a continous fire watch (0.33) with backup fire suppression oquipment (0.33) for the unprotected area within 1 hour.(0.33) { REFERENCE SR TS 3 21-3 j NA TS 3/4 7-69a. 3y J N';;4:R 0.1^ 1. - ^, i
- 1) determine remote assembly area hl4M W
- 2) notify station security
- 3) notify ERC
- 4) notify training center j REFERENCE SR EPIP 5 05 pp 283. (
ANSWER 8.13 (1.00) Nor not capable of performing its specified function. REFERENCE CR TS 1.0 1-1. SR TS 1 0-2 NA TS 1.0 1-1. L i t l l l
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- 8. ADMINISTRATIVE PROCEDURES, CONDITIONSe' AND LIMITATIONS PAGE 38
__________________________________________________________ 9 ANSWERS -- SURRY 112 -85/12/04-TOM ROGERS *
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y SR During recalibration of the excore j b ic3T off-set detection system. j N3 When used for! ,,
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- a. Recalibration of the .excare neutron flux detection system. )
- b. Monitoring the GRTR. I
- t. Measuremen,t cf'Tiot channel f actors. !
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REFERENCE , CR TS 3.1F1. j NA
/ ~ 78'3/4 3-42. !
ANSWER 8.15 (1.00)
- 1) They report to the emergency switchaear room
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REFEPENCE SR ADM 33 p 1. ANSWER 8.16 (1.00) , 1 Keys which operate bypass switches'will be under the control of the shift supervisor and maintained IAW ADM-29.3. g 2 Bypass of safety function will be performed only IAW writtene approve procedures. g , 3 Bypass of safety function must be performed with the knowledge and approval of the shift supervisor. ., REFERENCE SR ADM 29.5 p 11. ! ANSWER B.17 (1.00) NA Controlled leakage shall be that seal water flow supplied to the reactor coolant pump seals. SR Leakage sources such as the reactor coolant pump controlled leakage seals. ! y e A.y - pc e.a . l l ak,% 0-e) @@Y~ c) mrs ko A W eP/Ct"L$ ' sro.
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ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LINITATIONS PAGE ~ 39 # B. __________________________________________________________ 3 ANSWERS -- SURRY 112 -85/12/04-TOM ROGERS 4 o REFERENCE ' ;l NA TS DF's r 1-2. ;
'SR TS p 3 1-13.
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1 l-ENCLOSURE 4 Facility: Surry Examiner: L. L. Lawyer, R. Picker, T. Rogers Dates of Evaluation: December 4-6, 1985 Areas Evaluated: X Written X Oral X Simulator l Written Examination
- 1. Overall evaluation of Examination: Marginal
, Oral Examination L
i
- 1. Overall evaluation of examination: Satisfactory I 2. Number Conducted: 12 j Simulator Examination Overall evaluation of examination:
- 1. Satisfactory l
2 Number Conducted: 12 l Overall Program Evaluation ~ Satisfactory Marginally Acceptable X Unsatisfactory l l Submitted: Forwarded: Approved: Examiner t/ ik
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th o_ f ,<
' Section Chief Branch Chief
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