ML20213F411
| ML20213F411 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 10/27/1986 |
| From: | Elin J, Johnston G, Meadows T, Morrill P, Obrien J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20213F407 | List: |
| References | |
| 50-528-OL-86-02, 50-528-OL-86-2, NUDOCS 8611140193 | |
| Download: ML20213F411 (150) | |
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Examination Report No. 5d-528/0L-86-02 Docket Nos. .
50-528/529/530 Licensee: Arizona Public Service Company P. O. Box 21666 I._ Phoenix,. Arizona 85036 Facility Name:' ', alo' Verde 1 and 2 Examination at: Wintersburg, Arizona
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Examination conducted: September 22 - October 3, 1986 Chief Examiner: [O'22~
J. . OBrien, Chief Examiner Date Signed.
Examiners: V N~M b P. Morrill, Examiner Date Signed G.
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nst , Examiner lb27 h Dat4 Sl'gfied T.
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Endows, Examiner Fat /0 8%
Date Signed I
Approved by:
J Jm E
. Elin,'Sectron Chief
/0[77 $
Date Sigded i
i Examinations were conducted from September 22 to October 3, 1986. The Written Exam was administered on September 23, 1986 to eight SRO and eleven RO candidates.
Operating exams were conducted from September 24 to October 3, 1986 to eight SRO and eleven R0 candidates. All candidates who took t'ae written exams passed.
l All candidates passed the operating exams.
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DETAILS
- 1. Examiners
- J. P. O'Brien, RV (Chief Examiner)
+ G. Johnston, RV
- P. Morrill, RV T. Meadows, RV '
- G. Zwetzig, RV
- L. Defferding, PNL J. Upton, PNL J. D. Smith, PNL
- 2. ANNP Persons Attending. Exit Meeting
+*W. F. Fernow, Training Manager
- J. M. Allen, Manager of Operations
- J. R. Bynum, Plant Manager
- W. Rudolph, Lead Instructor
- J. Stanely, Simulator Coordinator
- R. Baron, Compliance Supervisor
- D. Craig, Training Supervisor, License Operators
- Denotes September 26, 1986 meeting.
+ Denotes October 3, 1986 meeting.
- 3. Exam Review At the conclusion of the written exams, the facility staff reviewed the exams. The comments made by the staff at the conclusion of the review, and the resolution of these comments are included in attachments (1) and (2). These comments were discussed with the staff, and appropriate changes were made to the exam keys prior to the grading of the exams.
General weaknesses were noted by the examiners in the accuracy of reference materials provided by the training staff.
- 4. Operating Examinations Simulator and oral exams were conducted September 23, 1986 to October 3, 1986. The examiners noted a significant improvement in the operability and capabilities of the simulator.
General weaknesses were identified by the examiners in the area of Radiological procedures and release points and rad waste system design.
- 5. Exit Meetings On September 26 and October 3, 1986, exit meetings were conducted with licensee representatives listed in paragraph 2. Some minor generic weaknesses and condition of the simulator were discussed. Further, the process for licensing of those candidates granted waivers: for the "480 hours0.00556 days <br />0.133 hours <br />7.936508e-4 weeks <br />1.8264e-4 months <br /> - extra person on shift" requirement was discussed. Upon completion of the remaining on-shift time, (greater than 20% power),the licensee will notify to NRC, Region V, and then those licenses can issued.
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Attachment I l
FACILITY COMMENTS / RESOLUTIONS l Answer 1.04 Facility Coment: None. As long as if they select MOC Power Defect & Rod i worth the answer is graded accordingly. 1 l
Resolution: Comment noted. Key was M0C - as identified by 300 ppm. Note was i added to answer key to accept 40-50. If candidate answered BOC or E0C,
[-0.5] was deducted.
Answer 1.06 Facility Coment: Possibly expect as an acceptable answer, because Axial Shape !
flattens and then dips in the middle (Axial Mid Plane) l Resolution: /::cepted. Answer key changed to accepted (a.) or (d.) as correct answer.
Answer 1.10 Facility Coment: Given the range of acceptable tolerance - none - however if a person assumes 10% Q-dot and 50% delta T then he should not lose full credit - but the key seems to allow this.
Resolution: Coment noted. Wide tolerance given and accepted. Answer key changed to reflect (25-100% acceptable) and (5-10% acceptable depending upon assumptions).
Question 1.11
. Facility Comment: The Question (part C) seems to ask how high the pump could l deliver water. I would expect the student to answer "not knowing pump work, I cannot answer the question".
Resolution: Coment noted - No change. Candidates did not answer question in this manner.
Answer 2.03a Facility Coment: According to Procedure 410P-ISI01, " Shutdown Cooling Initiation" the interlocks as stated in precaution and limitations section 3.12 are:
- 1) the suction valves cannot be opened when pressurizer is greater than 410 psia.
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- 2) the valves automatically close when pressurizer pressure exceeds l 500 psia.
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Resolution: Accepted. Answer key changed to read !
- 1. ....RCS pressure is less than 410 psia [+0.5]
- 2. .... pressure increases above 500 psig [+0.5]
Answer 2.04a !
I Facility Comment: The answer " SEAS in the Control Room" is not valid during I an AFAS if the 'B' AFW pump has started normally then no " SEAS" alarm will be received. This alarm is for a pump that failed to start properly when called upon by the ESFAS system.
The " alarm permissive circuitry energized" is correct which normally results in a pump low discharge pressure alarm until the pump has developed sufficient head. Should also accepted a " low discharge pressure alarm".
, Another indication which is energized is the 'B' AFW pump motor ammeter indication in the C.R. and at the switchgear and should be accepted as correct.
Resolution: Accepted. Answer key changed to read:
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- a. 1. red indicating light in control room
- 2. red indicating light on switchgear
- 3. SEAS in control room (not if B pump starts)
- 4. alarm permissive circuitry energized
- 5. Low-discharge pressure alars
- 6. Pump motor ammeter CR and switchgear
- 7. Pump pressure and flow indication Answer 2.06 Answer key was changed to read:
- a. 1600 - 1700 psig,1850 [-0.1],1950 [-0.2, > 1950 [-0.4]
- b. 150 psig, 200 [-0.1], 250 [-0.2], > 250 [-0.4]
- c. 610 psig (575 - 625 psig acceptable)
- d. 815 gpm (750 - 900 gpmacceptable), 1000 [-0.1],1100 [-0.2] > 1100 [-0.4],
650 - 750 [-0.1], 550 - 650 [-0.2], ( 550 [-0.4]
- e. 4300 gpm (4000 - 5000 gpa or 28-72% acceptable) l l
- f. 14000 gal (1800 - 1900 ft-3 acceptable) (1901 - 2000 [-0.1] deducted) l l
Question 2.07 Question was changed to read:
FILL-IN-THE BLANKS in the following parts of the questions pertaining to the feedwater system. (ASSUMEbypassisclosed.)
- a. With one string of two high pressure feedwater heaters out of service, electrical power out is reduced from full power to about %. (ASSUME bypass is closed.)
Answer 2.07a Answer key was changed to read:
- a. 75 w..To
- b. 1 - / 5" w . I f Answar 2.07b Facility Comment: According to Procedure 410P-1ZZ04, " Plant Startup Mode 2 to Mode 1" the first MFW pump is placed in service at =2% power. Step 4.3.7.
In the Feedwater pump Turbine procedure, 410P-1FT01, Section 6.0 has the Main Feedwater system placed in automatic control after the Feed pump is started, Steps 6.3.134.17, 6.3.14.18, 6.3.14.20.
Therefore an answer of between 2% power and 15% power should be accepted as the correct answer for this question.
It should be noted that the FW system is not always placed in auto at 2% and varies from crew to crew but is generally in auto prior to 15% power.
I Resolution: Accepted. Answer key changed to read:
- b. 2-15 Answer 2.08 Answer key changed to read:
- b. fuel pool coolers (or back up to nuclear cooling) l l
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Answer 2.09 Answer key changed to read:
- c. Since this leakage is collected and known it would be classified as identified leakage (or if past second seal it would be classified as boundary leakage.)
Answer 2.10b Facility Comment: According to P&ID 13-M-RCP-001 the spray flow comes from Loop 1A cold leg & Loop 1B cold leg. Your answer states just loop A & B cold legs but should specify Loop 1A & IB cold legs.
Resolution: Accepted. Answer key changed to read:
- b. cold leg 1A loop ;+0.3; cold leg 18 loop + 0.3 Answer 2.11d Facility Comment: RU-145 initiates a FBEVAS. The FBEVAS causes a cross trip of the Control Room ventilation system CREFAS. Mentioning the CREFAS is not required to answer the question and no extra points should be deducted for stating CREFAS as long as FBEVAS is mentioned.
Resolution: Accepted. The following note was added to the answer key: (Note:
Adding CREFAS does not deduct points.)
Answer 3.02b Facility Comment: AWP to CEDMCS will only be generated due to steam bypass demand, as in part a. Reactor Power Cutback is generated to RPCS (No AMI because > 15%).
Resolution: Comment noted. The statement "(AWP signal also produced)" was deleted from the answer key.
Answer 3.03c.1 Facility Comment: There may be a semantic problem as to what constitutes a
" detector". The entire assembly is the dectector. It consists of 2 dual proportional counters, or 4 BF-3 tubes total.
Resolution: Comment noted. No change.
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Answer 3.04 Facility Comment:
a.2 Answer may also be given as "RSPT's" or " target CEA positions".
a.6 Should be just "RCP speed".
a.7 Answer may appear as "CEAC penalty factors".
Resolution: Comments accepted. Answer key changed to read:
- 2. CEA positions (or RSPTs)
- 6. RCP speed
- 7. CEA (or CEAC) penalty factors
Answer 3.05 Facility Comment:
a.2 AWP will not be generated by a Tc failing low. Requires high Tc () 575) or high Tave - Tref (+6 deg F).
a.4 Channel deviation LED is actually on an enclosed back panel (RRSTD) and must be unlocked and opened to see.
- a. Correct answers may also include Control Room meter indications; TC decreasing, Tave decreasing, Pzr Level setpoint decreasing or such items as high or low rate withdrawal demand lights.
c.1 PLCS may be taken to manual or to " local setpoint". CEDMCS may be taken to " standby".
Resolution:
a.2 Accepted. Answer key changed to read:
- 1. T-ave-T-ref Hilo alarm
- 2. Tc decreasing
- 3. T-ave decreasing
- 4. PZR setpoint decreasing
- 5. PZR trouble Hi-Low rate withdrawal
- 6. AMI
- 7. Channel deviation LED on RRSTD 5
a.4 Comment noted. No change - see resolution to a.2.
c.1 Accepted. Answer key changed to read: take manual control of CEDMCS and PLCS or local setpoint.
Answer 3.06 Facility Comment: The normal CEA control items may be listed as interlocks and should not be counted as incorrect. These include upper and lower group stops (UGS, LGS), CEA limits (UCL, LCL) and electrical limits (UEL, LEL), and sequential permissives.
Rasolution: Comment noted. No change.
Answer 3.07 Facility Comment: RPCS in " auto select" should also be considered correct vice "when no preselection is made", which refers to manual select.
Resolution: Comment noted. No change.
Answer 3.08 Facility Comment: Accept also "ecomomizer M/A station".
Resolution: Comment noted. No change.
Answer 3.09 Facility Comment: The control systems may also be listed. These include PPS, which generates a CSAS, and 80P-ESFAS, which sequences on the CS pumps, because this is I & C section, not plant design. CS pumps support shutdown cooling so RCS may be listed as an interconnecting system.
Resolution: Comment noted. No change.
Answer 3.12 Facility Comment: In actuality, H-2 goes on degas of RCS. It is questionable as to whether or not this question should be section 3.
l Resolution: Comment noted. No change.
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Answer 3.13 Facility Comment: Same comment as 3.12, above on being Section 3. Actual plant efficiencies run much lower than design of 99.9%.
Resolution: Comment noted. No change.
Answer 4.03d Answer key changed to read: False or True with explanation Answer 4.05 Answer key changed to read:
- 2. ....(RCN-PIC-100) to approximately 2220 psia psia or (2230 psia with explanation also acceptable) [+0.75] (2220-2230 Answer 4.07b Facility Comment: This will cause a loss of feedwater. Loss of feedwater from loss of Condensate pumps. Loss of Condensate pumps from loss of HPSH.
Resolution: Comment noted. Will take into account for [+0.8].
Facility Comment: If he assumes that he has a loss of vacuum which could happen because a loss of Steam to header. His prime concern will be CST level and not hotwell.
Resolution: Comment noted. No change - True, but not asked for.
Question 4.08 Facility Comment: He may list sp2cific equipment which may be important.
Resolution: Comment noted. No change - Not asked for, points will not be l deducted.
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Question 4.11 Facility Comment: This cob 1d be 2 or 3 because the value may be in T. S.
3.6.3 as per table 3.6-1 4 hrs to restore.
Resolution: Accepted. Pertains to part d. Will only accept T. S. 3.6.3 as correct answer as noted by facility. Answer key corrected to read:
- d. 3 M
4 f
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Attachment'2, Part 1 Following are ANNP facility comments on the NRC examinations administered on September 23, 1986: -
Section 5 5.03 "None. Student may answer 15 psi due to difference between psia
& psig." ;
5.03 "Might have small steam bubble in top of Pzr. Heat of compression I of steam bubble might not have been disapated into fluid and small bubble may still exist."
5.04 " Exposing a partially filled fluid system to a lower pressure area.
Example: Cut drains into condenser or feed ring type problems ;
i experienced at San'Onofre."
5.06 "Subcooling is also an indication of RCS inventory. By the maintenance of safety functions proper and adequate inventory control is main-tained."
5.10 "We have de emphasized cry (1-K ) = CR2 (1-K2) and substituted i CRg Py = CR P in this case P 2 2 2 =-5000(f)=-12,500 b ' rods
= - 7500.
! While this assumption losses validity beyond the -5000 pcm mark, i- the concept still shows understanding and the examinee should receive partial credit."
- 5.11 "None. If student selects wrong curve should not loosa full credit."
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l SECTION 6 6.01.e " SIS connections - may cause some confusion they are know as
" cold leg injection"."
f "HPSI only connection - hot leg injection "
6.03 Your answer is ok however you may see the Tech. Spec answer to the three paths. It assumes gravity feed because the BAMPs are non-class powered. So for answer:
a) You might get gravity feed from RWT or SFP through V164 (filter bypass & a charger pump).
b) You may get gravity feed from RWT or SFP through V536 and 4
a charging' pump.
c) You may get gravity feed from RWT via V327 and a charging pump.
6.04.b Answer says water exiting from third seal flows to which two tanks the key says RDT & VCT - actually it might not be viewed as an
" exiting" from the seal in a since, it comes from HP side of seal.
- 6,06.a Answer 1 or 2 should be acceptable because Full Arc and Partial Arc admission have been used at 100% power to control control valves.
The question assumes that Full arc admission is turbine warming as is hinted at.in the reference material but this is not so. The question also assumes that only partial arc is used for normal steaming operations and this has not been so. If the question had been worded for turbine (chest) warming vs power operations it woald have been okay (except that only 1 Main Stop Valve (#2) has the internal bypass & servo valve).
To word the question better next time just ask which valve (s) control during che'st warming and which during normal ops, because the description of low valve gate, etc.
6.07.a- Question is talid. Answer:b might be then necessary. By projecting the effect .> .the change they project the DNBR that is compared to the trip setpoint. This is all one process, but answer is correct.
6.09.b CSAS will also start the pump. CRVIAS is no longer an automatic system so CSAS might be a more likely answer but 4'of 5 should be
.all that is necessary for full credit. We also teach any signal that start DG they are AFAS-1, AFAS-2 and LOP.
6.10.a' OK - but this is not conventionaly way of look at DG cooling..
b 1) Also acceptable as something cooled by Jacket Water Cooler is the Governor Oil Cooler.
- 6.11.a OK l b.4 Answer is redundant, but should be counted in part 3 as its use.
I.E. I doubt if many will say Reactor Power (Neutron) is used to l generate Rx power neutron. But they will say that it is used to l create power Error as in b.3 above.
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SECTI'ON 7 ,
7.01.b Also accept " Radiation Protection Supervisor".
c and d - accept shigt supervisor.
d - accept emergency coordinator in E-Plan 7.05.b greater than or equal to 552*F.
l 7.02.b look up second part of answer "and releases of radioactive materials".
If this is in 10CFR20.1 (c) see if we way the same in our texts.
7.07.a Expect reasonable responses such as:
excessive flow inadequate flow i maintaining heat removal 7.08 May expect to see validate alarm prior to taking action.
7.09 Requires prcedure memorization to answer this question.
May expect to see:
- check for obvious grounds 1
loose wires dirt components closed
- adequate ventilation
- look for closed load breakers 7.11 To say you can't continue is wrong. Conservatively you would not.
This answer assumes the leakage is from a source that will see a higher pressure as the RCS is pressurized. The Leakage could very well be in a portion of the CVCS that will not see any further increase in pressure.
i "Yes." response with logical "b" explaination should be evaluated and graded on acceptable explaination.
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.SECTION 8.0 i 8.02.c in modes 5 & 6 we are required to have only 1 RO & 1 AO.
! 8.08.b In addition to those already give the answer key should include l the actions taken under section 2.0 of Tech Specs.
1.e. a) Place the unit in Hot Standby within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, b) If 2750 psia is exceeded decrease pressure
> 2750 psia within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in modes 1 & 2 or
_ written 5 min. in modes 3, 4, or 5.
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'AITACHMENT2,PART2:
RESOLUTIONS TO FACILITY COMMENTS FOR THE SENIOR REACTOR OPERATOR EXAM:
5.03 There is no apparent basis for accepting an answer of 5 psi.
No credit will be added or deducted if a candidate postulates a small bubble in the pressurizer and explains the mechanism as in the facility coment; however, the correct and sufficient answer is the reactor vessel - as stated in the answer key.
5.04- Reasonable answers pertaining to water-hamer initiating actions -
such as those suggested - will be accepted if adequately explained.
5.06 Coment is correct, but is not germane to the question.
5.10 According to the facility staff, the stated approximation is taught to only be applicable when'k,ff > 0.95. Since the question is outside'this range, the candidate should not use the approximation.
Partial credit, however, will be given if use of the approximation cis stated.
5.11 Partial credit will be given if the only error is the use of the wrong _ curve.
,6.01.e If adequately explained by response, the term " cold leg injection" may be accepted - however, training material uses term " SIS".
6.01.f If adequately explained by response, the term " hot leg injection" may be accepted - however, training material uses term "HPSI".
6.03 The Boration path via V164 is a variation of Answer B in the answer key and will be accepted. The other paths described are equivalent to Answers b and c in the Answer Key and are therefore acceptable.
It is noted the referenced training material text does not agree with the plant P& ids in that the text suggests the borated water may
! be drawn from either line CH-A-424-GCBC-20" or line l CH-B-425-GCBG-20". The P&ID (13-M-SIP-001) indicates the borated water can only be drawn from the latter line.
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6.04.b Full credit will be given if response states water exiting seal only flows to RCDT.
I 6.06.a The Answer Key is based on the Training Material supplied to the NRC. Inasmuch as the Training Material does not mention these alternative answers, the Answer Key will not be amended.
6.07.a coment does not indicate need to revise Answer Key.
6.09.b Because the Training Material indicates LOP, AFAS-1 and AFAS-2 start i the Essential Cooling Water pumps and_the diesel generators, and.
thus, indirectly start the Essential Spray Pond Pumps, 0.5_ points l credit will be given for an answer which includes all three of these
< indirect signals.~- In' fairness to the candidates, CSAS will be substituted for CRVIAS for this examination. However, to avoid
.possible penalties to future candidates, the-facility is advised to i
provide correct training materials in future submissions.
6.10.a Comment does not indicate need to revise Answer Key.
6.10.b Because the Training Material includes a reference to the diesel generator, governor, oil cooler (in the section on the governor), this s ' component will.be'added.to the answer.
6.11.b.kAnswerneed;notbestructuredthesameastheAnswerKey._Aslong as the'use of Reactor Neutron Power is accurately described, full
. credit will be giv'en for this part of the answer.
7.01.b -The proposed alternate'a'nswer is not acceptable because it could also indicate the. Supervisor.of the Unit Radiation Protection Group
~- which is-clearly incorrect.
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l 7.01.c and d. Based on common' usage, the term " Shift Supervisor" will also be l ' . accepted. Emergency Coorindator will also be accepted for part d.
7.02.b. An'swer Key is consistent with 10CFR20.1 (c). No change needed.
7.05.b Comment accepted. Correct answer is 552'F, not 532*F.
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7.07.a Answer Key is based on verbatim text of Training Material.
i Equivalent answers will be acceptable, but suggested overly l- " general" answers are not.
, 7.08 Validating a radiation alarm prior to evacuation is not an acceptable answer because it is not in accordance with the procedure and because, if the alarm is valid, the delay required to validate-the alram could result in excessive, if not serious, radiation exposure.
l - 7.09 We do not agree the question requires procedure memorization. The l question asks SRO candidates for four prerequisites for. energizing a
- 480 Volt Class IE Switchgear-Load Center. Because the answers.in I the Answer Key are straightforward and based on common-sense, we believe an SRO should, based on experience and knowledge of the system, know most of the five prerequisites stated in reference.
Substantive functional answers equivalent to those stated in the
, Answer Key will be accepted. Overly general answers, such as those suggested (dirt, loose wires, adequate ventilation, etc.) will not be accepted.
j 7.11 Comment is accepted. A "yes" response with a logical explanation 4
will be evaluated and graded based on the validity of the j- explanation.
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8.02.c Answer Key will be modified to' add. fact only one R0 and one AO are needed during Modes 5 and 6.
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8.08.b- ' Answer Key will.be modified to include additione.1 requirements.
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A aswer< Key M ARI< co v P U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION Facility: Palo Verde Reactor Type: Pressurized Water Reactor (CE)
Date Administered: September 23, 1986 Examiner: J. O'Brien Candidate:
INSTRUCTIONS TO CANDIDATE:
Use 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 passing grade 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.
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% of Category X of Candidate's Category Value Total Score Value Category 25 25 5. Theory of Nuclear Power Plant Operation, Fluids, and Thermo-dynamics 25 25 6. Plant Systems Design, l
- Control, and l
Instrumentation .3 25 25 7. Procedures - Normal.
Abnormal, Emergency, .!
and Radiological -
Control ',#
25 25 8. Administrative Pro-cedures, Conditions, and Limitations 100 Totals Final Grade All work done on this examination is my own, I have neither given nor received sid.
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I Candidate's Signrture i Examiner Standards i
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ES-201-2 NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of,this examination the following rules apply:
- 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
- 2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination ~
room to avoid even the appearance or possibility of cheating.
- 3. Use black ink or dark pencil only to facilitate legible reproductions.
i 4. Print your name in the blank provided on the cover sheet of the 3
examination.
- 5. Fill in the date on the cover sheet of the examination (if necessary).
- 6. Use only the paper provided for answers.
- 7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
- 8. Consecutively number each answer sheet, write "End of Category " as appropriate, start each category on a new page, write only one side of the paper, and write "Last Page" on the last answer sheet.
- 9. Number each answer as to category and number, for example, 1.4, 6.3.
- 10. Skip at least three lines between each answer.
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
- 12. Use abbreviations only if they are commonly used in facility literature.
- 13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the' question or not.
. 15. Partial credit may be given.. Therefore, ANSWER ALL PARTS OF THE QUESTION l AND DO NOT LEAVE ANY ANSWER BLANK.
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- 16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
- 17. You must sign the statement on the cover sheet that indicates that the work is your own an'd you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.
Examiner Standards 12 of 18
( 18. When you complete your examination, you shall:
- a. Assemble your examination as follows:
(1) Exam questions on top. .
(2) Exam aids - figures, tables, etc.
(3) Answer pages including figures which are a part of the answer.
- b. Turn in your copy of the examination and all pages used to answer the examination questions.
- c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions.
- d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
Examiner Standards 13 of 18
s' EQUATION SHEET f = ma y = s/t 2
w = mg s = v,t + at Cycle efficiency = N{ W rk (o t) _
E = aC a = (vf - y )/t E KE = my vg = v, + a A = AN A=Aeg PE = agh to = 9/t A = In 2/tg = 0.693/tg W = vaP t q(eff) = (t,1)(ts)
AE = 931Am (tg+t)3 '
= [nCpAT ,. I.Ie -Ix
- Q = UAAT I . I UX l Pwr = Wg In I=I 10
- o P=P 10 (t) TVL = 1.3/p P=P o
et/T HVL = 0.693/u SUR = 26.06/T T = 1.44 DT SCR = S/(1 - K,gg) fA e p)
SUR = 26 g p CR x = S/(1 - Kdfx)
T = (1*/p ) + [(s .' p) /Aeff] p
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b~
T = 1*/ (p - D M = 1/(1 - Kd f) = CR g/CR0
. T = (3 - p)/ A,gg p M = (1 - Kdf)0 (1 - K,gg)g
=
p = (K,ff-1)/K,gg = R ,gg/Keff SDM = (1 - K,gg)/K,gg p= -5
[1*/TK,'gg ] + [H/(1 + A,ggT )] ,
1* = 1 x 10 s onds
~
P = E(V/(3 x 1010) g = 0.1 seconds eff I = No Idyg=Id22 WATER PARAMETERS Idg =10 22 1 gal. = 8.345 lbm R/hr = (0.5 CE)/d (meters)
I gal. = 3.78 liters R/hr = 6 CE/d (feet)
I ft = 7.48 gal. MISCELLANEOUS CONVERSIONS .
3 10 Density = 62.4 lbm/ft 1 Curie = 3.7 x 10 dps 3
Density = 1 gm/cm 1 kg = 2.21 lbm 3
Heat of varorization = 970 Etu/lbm I hp = 2.54 x 10 BTU /hr 0
, Heat of fusica = 144 Btu /lbm 1 Mw = 3.41 x 10 Btu /hr k 1 Atm = 14.7 psi = 29.9 in. I'g. 1 Btu = 778 ft-lbf I ft. H O 2
= 0.4335 lbf/in 2 1 inch = 2.54 cm F = 9/5 C + 32 C = 5/9 ( F - 32)
T4ste A-21. SnAu TAOLE' beenhe wheine Lnibins L n i r..,,y -
Te m p" Sat
~
des F Pre **- Eat. Sal. Eat. l'at.
8 Ib/eq an. hquad Evan. vapor lsquid ban. vapor hquid b.a p. Sat.
7 es ' e, as A A, er 8's S apur 32 0 08654 0 01602 330G 330G 0 ou 1075.s ~s075 8 0 Ouuo 35 0 0're'65 0 01002 2947 2947 3 02 1074 1 1077 3 0 Oin.1 2.1877 2 17trs iI"'I'ly$
2 I; .
40 0.12170 u 01002 2444 2444 8.05 1071.3 IO7ts 3 O nli.J 45 0.14752.0 01602 203G. 4 203G.4 13.06 IC0M.4 InM 5 0 fr262 2 2.1415 IIG7 i2.14 2 ly/,. .
50 0.17813 '0. 01G031703. 2 1703.2 18.07 10G5.0 1083 7 0.0J68 2 OWS 2. 32w 60 0.2563 0.01604 120G.0 3200 7 28.00 1059.9 loss 0 0 0555 2 0'193 2 Ogg 70 0.3G31 0 016061 8G7.8 667.9 38.04 1054 3 1092.3 0 074, I.9902 g gy.
80 0.5069 0 0100Sl G33.1 633.1 48 02 1048 G 109G G 0 Omt2 1.9424 2 Ow'.
90 0.6982 0 01610. 464.0 4GR.0 57 99 3042.9 Il00 9 0 1115 1. B'87J 2 0%-
100 0.94'2 J 0 01G33 350.3 350.4 67.97 1037.2 1805.2 0.3295 1.8531 1.962.'i 110 1.2748 0.01617 2G5 3 2G5 4 77.94 1031.6 1109 5 0.1471 1.e10G 1. 9 5..**
120 1.6924 0 01020 203.25 20l.27 87.02 1035.8 1133.7 0. t r.45 1 7ue4 g .g3 g .
130 2.2225 0 01625 157.32 157.34 97 90 1020 0 1I17.9 0.1810 1.729G I 911 140 2.8&b6 0 01G29 122 99 121 On 107 lr> 1014.1 1122 0 0 19v4 1.G910 8 ht. i ISO 3.718 0.0lG34 97.06 97.07 117.89 1008.2 182G.1 0 2149 3 6537 1. bGn 160 4.741 0 01639 77.27 77.29 127.89 1002.3 1130.2 0 2331 1.6174 170 8.992 0.01045 62.04 62 OG 137.90 990 3 1834.2 0.2472 1.5822 8.82.
8 64ul ISO 7.510 0 01051 50 21 60.23 147.02 990.2 1838.1 0 2Glu 1.5440 1. b lire 190 9.339 0 01G57 40 94 40.90 157.95 984.1 1142 0 0.2785 1.5147 3.7932 200 11.520 0.01663 33.62 33.64 167.99 977.9 1I45.9 0.2938 1.4824 3.77G2 210 14.123 0.01670 27.80 27.82 178 05 978.6 1849.7 0.3090 1.4508 212 14.696 0.01672 26.78 26.80 ISO 07 970.3 3150 4 0.3120 1.444G 1 75%.,
1.75o 220 17.186 0.01G77 23.13 23.15 188.13 9G5.2 1153.4 0.3239 1.4201 1.744.i 230 20.780 0.01G84 19.365 19.382 198.23 958.8 !!57 0 0 3387 l.3901 8.72n 240 24.969 0.01692 16.30G 16.323 208.34 952.2 1800.5 0.3531 1.3009 1.734.i 250 29.825 0 01700 13.804 13.821 218.48 945.5 1164.0 0.3675 1.3323 260 35.429 0.01709 11.74G II .763 228 G4 938.7 IIG7.3 0.3817 1.3043 1. 1.0656m.
270 48.858 0.01717 10.044 to OGl 238.84 931.8 1870 0 0 3958 1.27G9 1.672*
280 49.203 0.0172G 8.628 8.G45 240.00 924.7 1173 8 0.4090 1.2501 1. 65..;
290 57.55G 0.01735 7 444 7.4GI 259 31 917.5 1876.8 0.4234 1.2238 1.647.
300 67.013 0.01745 6.449 6 4G0 2G9.59 910 I 1179.7 0.4369 1.1980 1.635" 310 77.68 0.01755 5.G09 5.620 279.92 902.6 1182.5 0.4504 1.1727 1.62.4 320 89.66 0 01765 4.836 4 914 290.28 894 9 1885 2 0.4G37 1.1478 1. t,11 *,
330 103.06 0 01770 4.289 4.307 300.68 887.0 1187.7 0.47G9 1.1233 3.6w2 340 118.01 0.01767 3.770 3.788 all.13 879.0 !!90.1 0.4900 1.0992 1.5ast 350 134.63 0.01799 3.324 3.342 321.63 870.7 1192.3 0.5029 1.0754 1.57t1 360 153.04 0.01811 2.939 2.957 332.18 8G2.2 1194.4 0.5158 1.0519 1.567-370 173 37 0.01823 2.606 2.625 342.79 853.5 II96.3 0.5286 1.0287 1.557-380 195.77 0.01836 2.317 2 335 318.45 844 6 1898.8 0.8413 1.0059 1.547-390 220.37 0.01650: 2.0G51 3.0830 3G4 17 835.4 1899.6 0.5539 0.9832 1.537.
400 247.31 0.018G4 1.8447 l.8633 374 97 826 0 1208.0 0.5664 0.9608 1.52 2 410 276 75 0 01878 8.6512 3.G700 385.83 SIG 3 1202.1 0.5788 0.9386 1. 5 th 420 308.83 !.4811 1.5000 390 77 806.3 1203.1 0.5932 0.916G 1.507*
430 343.72 0Olb94l* 1.3308 0.01910 1.3494 407.79 706.0 1203.8 0 6035 0.8947 1.495.
440 381.59 0.0192G 1.1979 1.2171 418.90 785.4 1204.3 0.6158 0.8730 1.46 %
480 422.6 0.0194 1.0799 1.0993 430.1 774.8 1204 6 0.6280 0.8513 1.47fr 460 4G6.9 0.0196 0.9748 0.9944 443.4 7G3.2 1204.6 0.6402 0.8298 1.470 470 614.7 0.0198 0.8811 0.900u 452.8 751.6 1204 3 0.6523 0.8083 1.460.-
400 566.1 0.0200 0.7972 0.8172 4G4.4 739.4 1203.7 0.6645 0.7868 1.451 490 621.4 0.0202 0.7221 0.7423 476.0 726.8 1202.8 0.6706 0.7653 1.441!
800 680 8 0.0204 0.6545 0.6749 487.8 733.9 1201.7 0 6887 0.7438 820 812.4 0.0209 0.5385 0 5594 Sil.9 686 4 1898.2 0.7130 0.7006 1.432'.
1.413 540 962.8 0.0215 0.4434 0.4649 536.6 656 6 !!93.2 0 7374 0.6568 1.394 560 1133.3 0.0221 0.3647 0.3868 562.2 624.2 1186.4 0.7621 0.6128 1.374.
680 1325.8 0.0228 0.2989 0.3217 548.9 884.4 1177.3 0.7872 0.5659 1.3532 600 1542.9 0l0236 0.2432 0.2668 617.0 '848.8 1165.5 0 8131 0.617G 1.330*
620 1786.6 0 0247 0.1955 0.2201 646 7 503.6 1150.3 0.8398 0.46G4 1.3002 640 2059.7 0.0260 0.1538 0.1798 678 6 452.0 1130.8 0.8679 0.4110 1.278'.
! 860 2365.4 0.0278 0.11G5 0.1442 714.2 390.2 1804.4 0.8987 0.3485 1.2472 l 480 2708.1 0.0305 0.0810 0.1135 757.3 309.9 1067.2 0.9351 0.2719 1.2071 700 3093.7 0 0369 0.0392 0.0761 823.3 172.1 995.4 0.9905 0.1484 1.1389 705 4 3?on 2 0 05n1 0 0 0503 902 7 0 902 7 1 0580 0 1 0580
- Rennated from abridred edition of " Thermodynamic Properties of Steam." by Jueeph II. Keenan and & redenck G. Keyes. John Wiley di Sons. Inc. New York.1937 with the permienson of the authern and publesbar.
1 l
l l
CATEGORY 5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS AND THERMODYNAMICS.
- 5.01 QUESTION (2.0)
I Give two reasons based on safety considerations (i.e. NOT power shaping) why CEA insertion limits are established in the technical specifications. (1.0 each) l
- 5 01 ANSWER
- a. To ensure adequate shutdown margin. (1.0)
- b. To limit the reactivity worth of an ejected CEA. (1.0)
- REFERENCE Palo Verde Unit 1 Technical Specifications, p. B 3/4 1-6 1
e 0
5-1
- 5.02 QUESTION (2.0)
A typical large PWR has operated at full power for more than 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> (Xenon equilibrium is established). A partially inserted CEA on one side of the core is now inserted an additional twelve inches and the partially inserted diametrically opposite CEA is withdrawn as necessary to maintain the original power level (for the purpose of the question, disregard CEA alignment requirements).
- a. Will the Xenon concentration in the vicinity of the end of the Inserted CEA initially increase or decrease ? (0.5)
- b. Why will the Xenon concentration initially change in the above manner ? (1.0)
- c. How will the above channe in Xenon concentration affect the local neutron flux ? (0.5)
- 5.02 ANSWER
- a. Increase. (0.5)
- b. Reduced neutron flux in vicinity of CEA will reduce the rate of burnup of Xenon (0.5). Rate of production of Xenon [by Iodine decay] is not affected by reduced flux (0.5).
- c. Local neutron flux is reduced [due to increased Xenon concentration]. (0.5)
- REFERENCE PVNGS Licensed Operator Reactor Theory Review, NLC-55, pp. 16.5 - 16.6 5-2
- 5.03 QUESTION (2.5)
Following a reactor trip involving the loss of electric buses, the water temperature in the Pressurizer is 520*F, and the water temperature in the reactor (indicated by the core exit thermocouples) is 560*F. The Reactor Coolant System pressure is 1119 psig. Use the enclosed Steam Table, as needed, to answer the following:
- a. Is the water in the Pressurizer saturated, subcooled or superheated ? (0.5)
- b. Is the water in the vicinity of the core exit thermocouples saturated, subcooled or superheated ? (0.5)
- c. If the water in either or both of the above locations is subcooled, what is the location and the amount of subcooling in 'F 7 (0.5)
- d. If the water in either of the above locations is subcooled, what is the amount of subcooling in counds per sauare inch (usi) of pressure ? (0.5)
- e. Where would the steam " bubble", if any, be located for the l
l stated conditions ? (0.5)
I l
- 5.03 ANSWER
- a. Subcooled. (0.5)
- b. Saturated. (0.5)
- c. Pressurizer: 40 *F subcooled. (0.5)
- d. 321 2 psig. (0.5)
- e. Bubble is in the head of the reactor vessel [and p"ossibly in the RCS hot less). (0.5) No dednto'on if fosdb/c Y$5?b bubble E?t PZR is m enfo'orsed with odeyvate e y /spea ffort.
- REFERENCE Saturated Steam Tables. PVNGS Thermohydraulics Review for Licensed Operators, p. 12. Training Article NS-1B, Pressurizer, p. NS-1B-2.
5-3 l
l
- 5.04 QUESTION (2.0)
Describe two fundamentally different situations which can l produce a serious water hammer event. (1.0 each) l
- 5.04 ANSWER
^
Note: 1.0 each for any two of the following:
Rapidly closing or opening a valve on a flowing system.
( l.
- 2. Introduction of steam into a cooler system; or steam contacting cooler water.
l 3. Starting a pump to [quickly] fill an empty system
[ rapidly moving fluid strikes system components).
l '/. Othe reasortable mechanism.s derived from <3clual
- RERRENCE op,,, g., y 4 ,, ,,, , , , e .
PVNGS Thermohydraulics Review for Licensed Operators, p. 33.
I l
5-4
- 5.05 QUESTION (2.0)
One of the corrections applied by the Core Protection Calculator to the signal produced by the ex-core neutron detectors, is a l correction for the Reactor Coolant System Cold Les Temperature (T )*
c
- a. Explain why this temperature affects the measured neutron signal. (1.0)
- b. If, due to plant operations, T was lower than its normal value and no correction was made for this lower temperature, would the indicated (i.e. measured) neutron. power be Higher or Lower than the true neutron ;
power. (0.5) '
- c. Which of the following measurements of T is used to correct the measured neutron power: (1)cAverage value of T ; (2) Lowest value of T ; (3) Highest value of T .
c c (0.8)
- 5.05 ANSWER
- a. As T decreases, the water becomes more dense. (0.5)
Beca6se this water is located outside the core, more neutrons are reflected back into the core [and fewer escape to be counted by the ex-core neutron detectors].
(0.5)
- b. Lower. (0.5)
- c. (2) Lowest value of T c (0.5)
- REFERENCE i
Training Article NS-6B, Core Protection Calculator, para. 2.8.2.2.
5-5
l l *5.06 QUESTION (1.0)
Explain why it is important to maintain the Reactor Coolant System sub-cooled during cooling by natural circulation.
- 5.06 ANSWER Loss of sub-cooling would result in production of steam voids in the natural circulation loop. These steam voids would disrupt or halt natural circulation. (1.0)
- REFERENCE Training Article NS-10, Reactor Vessel and Internals, p.
NS-1C-5.
5-6
- 5.07 QUESTION (2.0) 1 i
The Low Pressure Safety Injection pumps have a design pressure l of about 710 psig but only develop a head of about 335 feet (145 psi) at design flow conditions. Explain how a pump with these characteristics can produce flow when the system pressure is greater than the head developed by the pump, or give an example of an LPSI pump being used in this manner.
- 5.07 ANSWER The pressure, or head, developed by the pump is a delta-P, or increment in pressure. The absolute discharge pressure is the sum of the pressure at the inlet of the pump and the pump head.
Thus, if the discharge pressure is greater than the system pressure, the pump can produce flow. 2r An example is provided by the LPSI pumps when they are used in the shutdown cooling mode with system pressures as high as 400 psia. (2.0)
- REFERENCE Training Article NS-3A, Safety Injection and Shutdown Cooling Systems, pp. NS-3A-10 and -47; CESSAR Table 6.3.2-1.
5-7
- 5.08 QUESTION (1.0)
For the PVNGS reactors, what is the most restrictive operational occurrence from the standpoint of challenging DNBR limits ?
- 5.08 ANSWER Four pump loss of flow. (1.0)
- REFERENCE Training Article NS-6B, Core Protection Calculator, p. 3 r
i 5-8
- - _ _ . = _ . - - - - - _ __ ._ _ .. ._- . + . - _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ . __
- 5.09 QUESTION (2.0)
The reactor is at 80% power with T = 578'F and T = 542*F.
Core flow is at 100%, which is 80 Nillion lbs/hr.c A station ,
blackout occurs, and after natural circulation is established, the core Delta T = 30*F. If the decay heat is 2% of full power, what is the mass flow rate (% of full flow) ?
- 5.09 ANSWER Q=mo p delta T (0.5) -
2% = m op 30*F 80% = 100% c p 36 *F cp = 2% / (m 30*F) =cp = 80% / (100% 36*F) (0.5) m = 2% x 100% x 36*F /(30*F x 80%)
m = 3% (1.0)
- REFERENCE PVNGS Operator Academic Fundamentals, Heat and Thermo, NLA06, Chapter 8, p. 8-24.
a 5-9
l l l
l l
- 5.10 QUESTION (3.5)
The reactivity of the core is - 5000 pcm with the Shutdown Banks fully withdrawn; and the neutron count rate is 30 counts per second. When the Shutdown Banks are fully inserted, the count l rate stabilizes at 12 counts per second. Based on the above data, calculate the reactivity worth of the Shutdown Banks in pcm. Show all work. ,
- 5.10 ANSWER
- 5000 pcm = - 5% rho. rho = (k,ff - 1)/h,ff; k,ff = 1/(1 - rho); keff0 = 1/(1 + .05) = 0.952 (1.0 point)
CR0 /CRg = (1 - k,ffy)/(1 - keff0}I h,ffy = 1 - CR0x(1 - keff0)/ cry = 1 - 30x(1 .952)/12 = 0.88 (1.5 point) rho = (k,ff - 1)/k,ff = (.88 - 1)/.88 = .136 = - 13600 pcm (0.5 point)
Worth of Shutdown Banks =
reactivity (banks inserted) - reactivity (banks withdrawn) =
- 13600 pcm - (- 5000 pcm) = - 8600 ocm. (0.5 point)
- REFERENCE l
. PVNGS Licensed Operator Training, Reactor Theory, pp. 8.9 to 1 8 .1 ?. ,
i n oTE : I.5 pain f.s if cando'da fe. uses facil.47 aperoximde l
%t form of ej ua% n and $efs co u ,c.t answer .;;;
b ased on that Six plified e gua Mo*
5-10
i
- 5.11 QUESTION (3.0)
The reactor is operating in mid-cycle at full power with equilibrium Xenon. The Boron concentration is 500 ppm, T is normal, and Group 5 of the CEAs is withdrawn to 105 inchegyg It is desired to withdraw Group 5 to 125 inches while remaining at full power. Using the enclosed figures, calculate the new Boron concentration needed to compensate for the desired rod withdrawal.
- 5.11 ANSWER From Fig. 5.11-2, the change in pcm corresponding to rod withdrawal is (- 30) = - 49 pcm. From Figure 5.11-4, the Boron worth at full power nominal T Boron concentration, is - 10.57 pcm?35m(.593* F) and 500 Therefore, ppm the new Boron concentration needed is 500 + (- 49 pcm)/(- 10.57 pcm/ ppm) = 500 + 4.63 ppm ,
= 504.6 com. Answer. (3.0) l
, 2: I
- REFERENCE PVNGS Core Data Book, Unit 1, Cycle 1, Curves 2.5.1 and l t 2.3.2.
i N ote : {-lalf credit if candidate us es avo ng c u r v e. , but gets cor r e ct answer b ased ow wrowy cu r v e .
m 5-11
CORE DATA BOOK [ g ! M g" Uhli 1 CYCLE 1 PAGE 4 JANUARY 27,198 PVNGS/ NUCLEAR SECTION REVISION 00
[ REC CEA WORTH (OVERLAP) VS WITHDRA WAL (50% P, EOC.
-600
\
\
-500 I \
\
R \
0 \
p -400 Y T l 0 l R \
l -300 \ '
' N .
Nt E N u
-200 x N
. \
_ soc
\
N
\
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. GROUP 5 , .
0 30 60 90 120 1,50 GROUP 4
, 90 120 150 CEA POSITION (INCHES WITHDRAWN)
PLANT CONDITIONS 507. POWER REFERENCE SOURCE OF DATA 455.7 EFPD V-CE- 1901C COMMENTS: C225Z5, IDENTICAL TO CURVE 3.12.5 APPROVED TOR USE BY: T) k 7 //'[
' NUCLEAR' SUPERVISOR DATE F/G, S.p 't .
1 7 ';) 77 2 R ii' T. n .'i
- W O "?
l
3 REVISION 000 PbG5/ NUCLE % ~~mmi[ R SECTIONS \} . s .
REG GROUP 5 WORTH VS WITHDRAWAL (HFP, BOC, hf0C, EOC) fA -330
- 1 ;
s .,
I I
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COMMENTS: C22521, IDENTICAL TO CURVE 3.12.1 ,
i APPROVED TOR USE BY: b['
be 2.[2.7 /FF
' DATE '
NUCLEAR SUPERVISOR FIG.s.n-2 fi".;D w ../ d 4 h.5'.f:"IOI";j a'i es l.CD
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- tFWJL@/17JL95Ly@@ SECTION,g. g s , ,a , , - = :e s 1 & 'M JANUARY 27,1Sb
,,, - r1, ,
rey StoN oct BORON WORTH VS T-A VG:
(505 * - 615
- F; 100% P, BO C) 1 ') '
t t
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500 510 520 530 5 10 550 560 570 580 590 600 61b Sh0 T-AVG (OEGREE5 F)
LEGEN0: PPM +-+-+ 600
- : : 1000 o-o-e 800 *-o -a 900 PLANT CONDITIONS
- -a--- 1200 1007. POhTR REFERENCE SOURCE 07 DATA 000 EFPD V-CE-19010 COMMENTS: C220Z1. IDENTICAL TO CURVE 3.8.1 APPROVED TOR USE BY: Dd bue L.[ .7 /pp -
HUCLEAR SUPERYlSOR DATE C
f/ G. S. Il- 3 .
- L. S.. .e$a .2 a .4 < :- - _
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JANUARY 27, PYNGS/ NUCLEAR SECTION ==7 ^ 3 1'"M umm T / 1'liO'".3 REV!SIO:
BORON WORTH VS T-AVG:
(505 615 F, 100% P, MOC)
(
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500 510 520 530 540 550 560 570 580 590 600 610 620 T-RVG (CEGREES F) i
' LEGEN0: PPM r - r 200 a-a-a 1100 o-o -6500
- : : 600 ~---- 0 0 0 PLANT CONDITIONS 1007. POWER REFERENCE SOURCE OF E 208.3 EFPD V-CE-19 COMMENTS: C22322 IDENTICAL TO CURTi. 3.8.2
~ APPROVE 0 TOR USE er: "T1. . s Idb e+- > fz 7/1'l NUCLEAR SUPERVISOR DATE F/ G. s.;1 if '
O n s V U 9 D i. 1_ T O T l Ei.l $ 2 ?
l
- w....-,, ,- . . . - - - - - - - ste.visWN Usu PVNGS/ NUCLEAR SECTION a
-w w d 'j ,
BORON WORTH VS T-AVG:
(505 * - 615
- F, 100% P, EOC)
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510 520 530 540 550 560 570 580 590 600 610 620 500 T-RVG (OEGREE5 FI LE0ENO: PPM r r r 0 0-EF4 100 6-e-4 200 +-+-+ 400 PLANT CONDITIONS REFERENCE SOURCE OF DAT V-CE-1901 1007. POWER 1 455.7 EFPD COMMENTS: C22323, IDENTICAL TO CURVE 3.8.3 c ,
APPROVED FOR USE BY: M b HUCLEAR SUPERVISOR
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- 5.12 QUESTION (1.0) l Which statement below is correct if the power range instruments have been adjusted to 100% based on a calculated calorimetric ?
- a. If feedwater temperature used was higher than actual, actual power will be less than indicated.
I
- b. If RCP pump heat is omitted from the calorimetric, actual power will be less than indicated.
- c. If steam flow used was lower than actual, actual power will be less than indicated.
I
- d. If steam pressure used was lower than actual, actual power will be less than indicated.
l
- 5.12 ANSWER l b. (1.0)
- REFERENCE PVNGS Operator Academic Fundamentals, Heat and Thermo, NLA06, p. 9-9.
i l
5-12 l
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l I
l *5.13 QUESTION (1.0)
Why is the rupture of a Main Steam line at End of Life (EOL) a much more limiting accident than at the Beginning of Life ?
END OF CATEGORY 5.
PROCEED TO CATEGORY 6.
- 5.13 ANSWER The Moderator Temperature Coefficient is less negative at BOL (0.5) than at EOL. This difference in magnitude increases the severity of the addition of positive reactivity (.25) due to the sudden cooling of the reactor coolant from the Main Steam line
> rupture (.25).
- REFERENCE PVNGS Licensed Operator Training, Reactor Theory, NLC55,
- p. 12.5.
e 5-13
1 l
l
)
CATEGORY 6. PLANT SYSTEMS: DESIGN. CONTROL. AND INSTRUMENTATION.
- 6.01 QUESTION (3.0)
Figure 6.01-1, attached, is a simplified schematic of the Reactor Coolant System. Sketch on Figure 6.01-1 the routing or point of' connection for the following. BE SPECIFIC REGARDING WHICH LOOP IS THE POINT OF CONNECTION:
- a. Pressurizer Surge Line (0.5)
- b. Pressurizer Spray Line (2 connections) (.25 each)
- c. RCS Charging connection (0.5)
- d. RCS Letdown connection (0.5)
- e. SIS connections (4) (.125 each)
- f. HPSI (only) connections (2) (.25)
- 6.01 ANSWER See Figure 6.01-1-Answer
- REFERENCE Training Article NS-1, Reactor Coolant System, Figure NS-1A-1, Rev. 1.
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- 6.02 QUESTION (2.0)
Listed below are several of the control actions related to pressurizer level control. Using the numbers 1, 2, 3, etc.,
indicate the order of these actions with respect to the value of the pressurizer level error signal. Use the number 1 to indicate the action which occurs at the largest positive value of the error signal (i.e. high pressurizer level), and the ,
number 6 to indicate the action that occurs at the largest negative value of the error signal (i.e. low pressurizer level).
(0.33 each)
- a. Backup heaters off .
- b. Stop standby charging Pump .
- c. Stop normally running charging pump .
- d. Backup heaters on .
- e. Start standby charging pump .
- f. Start normally running charging pump .
- 6.02 ANSWER
- a. Backup heaters off 4 . (.33)
- b. Stop standby charging Pump 5 . (.33)
- c. Stop normally running charging pump 1 . (.33)
- d. Backup heaters on 2 or 3. (.33)
- e. Start standby charging pump 6 . (.33)
- f. Start normally running charging pump 2 or 3. (.33)
- REFERENCE Training Article NS-1B, Pressurizer, p. NS-1B-15 and Fig. NS-1B-19.
6-2
- _ _ _ _ _ _ _ _ , _ . ~ _
0
- 6.03 QUESTION (1.5)
List or sketch three paths by which borated water from the Refueling Water Tank can be supplied to the suction header for the charging pumps.
- 6.03 ANSWER
- a. Via the Boric Acid Makeup Pumps (0.5)
- b. Via a bypass around the Boric Acid Makeup Pumps (0,5)
- c. Via a line from the pipes supplying borated water from the RWT to the Engineered Safety Features (0.5)
- REFERENCE Training Article NS-2A, CVCS I, p. NS-2A-11 l
l 6-3
- 6.04 QUESTION (4.0)
- a. Draw a sketch of the Reactor Coolant Pump hydrodynamic shaft seal system, showing the arrangement of the seals, coolers, jet pump and auxiliary impellers. Label each of these components. (2.0)
- b. Water exiting from the third seal flows to which two tanks ?
(0.5 each)
- d. Describe two functions provided by the flow through the pump lower journal bearing. (0.5 each)
- 6.04 ANSWER
- a. See Figure 6.04-1-Answer (2.0)
- b. 1. Volume Control Tank. (0.5)
- 2. Reactor Coolant Drain Tank. (0.5)
I c 1. Cools journal bearing. (Any 2, 0.5 each)
- 2. Lubricates journal bearing.
- 3. Prevents possible contaminants in the RCS from entering the seals.
- REFERENCE Training Article NS-1A, Reactor Coolant System, pp. NS-1A-7 to NS-1A-9; Figure NS-1A-10 Rev. 1 l
NOTE: F 4 cred if n' f r es gun.s e sta+es wa+ey " ev.; %g" third seat only 3ee3-to R C.D 7-6-4
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- F1 G v RE (, 0 4 A N S W E R.
- 6.05 QUESTION (1.5)
- a. What plant conditions will initiate Safety Injection ? (1.0)
- b. How many signals are needed for each condition to initiate .
)
Safety Injection 7 (0.5) l
- 6.05 ANSWER l
l a. 1. Low pressurizer pressure. (0.5)
- 2. High containment pressure. (0.5) l
- b. Two [out of four]. (0.5)
I
- REFERENCE i Training Article NS-3A, Safety Injection and Shutdown Cooling Systems, p. NS-3A-39.
l ,
l l
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l 1
6-5
l
- 6.06 QUESTION (1.0)
- a. Concerning Main Turbine FULL ARC operation, IDENTIFY the correct statement from the following selection: (0.5)
(1) The outputs of main steam pressure with limiter, load limit and load set runback, and STOP VALVE amplifier circuits are gated in a low value gate, with the lowest ~
valve opening signal controlling.
l (2) The outputs of main steam pressure with limiter, load limit and load set runback, and CONTROL VALVE amplifier circuits are gated in a low value' gate, with the lowest valve opening signal controlling.
- b. Concerning Main Turbine PARTIAL ARC operation, IDENTIFY the l correct statement from the following selection: (0.5)
(1) The outputs of main steam pressure with limiter, load limit and load set runback, and STOP VALVE amplifier circuits are gated in a low value gate, with the lowest valve opening signal controlling.
l (2) The outputs of main steam pressure with limiter, load limit and load set runback, and CONTROL VALVE amplifier circuits are gated in a low value gate, with the lowest valve opening signal controlling.
- 6.06 ANSWER
- a. (1) (0.5)
- b. (2) (0.5)
- REFERENCE Training Article PGS-3B, Main Turbine Control System, p.
PGS-3B-19.
O 6-6 s
, ~ , - - - - n- .. ,n --, , - - - - . , - - . . -. . , . , - - - , - , - , - - - - - . - - - - - . - - - , . - - - - . .---..--------n . . - - - - - . - . - - - - - - - . . , - .
- 6.07 QUESTION (2.0)
In addition to the STATIC computer program, the Core Protection Calculator also uses the UPDATE and FLOW PROJECTION programs in calculating DNBR.
- a. Why are the UPDATE and FLOW PROJECTION programs needed?
(1.0) ,
- b. How do these programs work with the STATIC program ?
(1.0)
- 6.07 ANSWER
- a. RCS parameters [ pressure and flow) can change so fast (0.5) that the STATIC program, by itself, cannot assure that a protection signal will be generated in time to respond to the transient (0.5). Or, UPDATE and FLOW PROJECTION are needed to provide protection for transients (0.5) that STATIC is too slow to handle (0.5).
I
- b. The UPDATE and FLOW PROJECTION programs project the effect of rapid changes in RCS parameters on DNBR (0,5) and project l when trip should be initiated to prevent violation of DNBR limits (0.5).
- REFERENCE Training Article NS-6B, Core Protection Calculator, pp. 2-3.
i 6-7 e-,- g----.-,-mm- - - -+---r--wy ----Ime p w,r-e s---+-.,w-, -----------m----ree-------mw-r--r e-w-+
- 6.08 QUESTION (1.0)
- a. What safety features signal causes the Containment Spray pumps to start but not deliver spray ? (0.5)
- b. What safety features signal allows borated water to be pumped to the Containment Spray nozzles ?
(0.5)
- 6.08 ANSWER l
- a. SIAS. (0.5)
- b. CSAS or containment pressure high-high; (0.5)
- REFERENCE Training Article NS-3B, Containment Spray and Iodine Removal Systems, p. NS-3B-10 l
6-8
- 6.09 QUESTION (3.0)
- a. List the two major components cooled by the Essential Spray Pond System ? (1.0)
- b. List four signals that will start both spray pond pumps.
(2.0)
- 6.09 ANSWER
- a. Diesel Generator and Essential Cooling Water Heat Exchangers. (0.5 each)
- b. (0.5 each)
- 1. Diesel Generator running.
- 2. Safety Injection Actuation Signal.
- 3. Control Room essential filtration actuation signal.
- 4. ....... avvm v o u . . . . . ... .... ..... __. ... C 5 A .5 .
- 5. $s'q nats which start Essenh'al Coo lin3 Water- OR i.oP,AFAS-_t
- REFERENCE dwa AFAS-2_.
Training Article PGS-8A, Essential Spray Pond System, pp.
PGS-BA-3 and -26.
e 6-9
- 6.10 QUESTION (3.0)
- a. Name the two separate cooling systems making up the Diesel Generator Cooling Water subsystem. (1.0)
- b. For each of the above systems, list at least two loads cooled by the system. (2.0)
- 6.10 ANSWER
- a. 1. Jacket [ Cooling] Water System.
- 2. [ Essential] Spray Pond [ Cooling] System.
- b. 1. Jacket Water System: (0.5 each)
Engine Water Jacket Turbocharger Govermot O il Coo le r'
- 2. Spray Pond System: (0.5 each for any two)
Turbocharger Intercoolers Jacket Water Cooler Lube Oil Cooler Fuel Oil Cooler
- REFERENCE Lesson No. NLC23, Rev. 5, Diesel Generator, pp. 34-36 l
, 6-10
i l
l
- 6.11 QUESTION (3.0) l 1
- a. List the measured quantities which provide input signals to l the Reactor Regulating System (Note: if the same quantity is measured in more than one location, only one need be listed). (1.0)
- b. Explain how each of the above signals is used to provide control of the reactor. (2.0)
END OF CATEGORY 6.
PROCEED TO CATEGORY 7.
- ANSWER
- a. 1. RCS Hot Leg temperature (T )
h (.25 each)
- 2. RCS Cold Leg temperature (T )
c
- 3. Turbine Load Index (TLI) or Turbine First Stage Pressure
- 4. Reactor [ Neutron] Power as measured by the ex-core detectors or e n*
- b. 1. T:h used to generate T,y, (.25)
- 2. T:c used to generate T,y, (.25)
- 3. TLI: used to generate T,,f (.25) AND [ Reactor] Turbine Power (.25). T,,f is compared with T avs to generate a
! Temperature Error (.25); and [ Reactor] Turbine Power is compared with Reactor [ Neutron] Power to generate a Power Error (.25). These errors are fed to the RRS to control the reactor (.25).
- 4. (n: used to generate Reactor [ Neutron) Power (.25).
- REFERENCE Training Article NS-90, Reactor Regulating System, pp.
NS-9C-5 to NS-9C-9.
6-11
CATEGORY 7. PROCEDURES: NORMAL. ABNORMAL. EMERGENCY AND RADIO-LOGICAL CONTROL-
- 7.01 QUESTION (2.0)
What individuals are authorized to approve the following:
- a. A whole body radiation exposure in excess of 2250 arem/ quarter or 4000 mrem / year 7 (0.5) -
- b. A whole body radiation exposure in excess of 300 mrem / week 7 (0.5)
- c. Containment entry during Operational Modes 1 and 2 ?
(0.5)
- d. An Emergency Radiation Exposure Permit ? (0.5)
- 7.01 ANSWER
- a. [PVNGS] Plant Manager (0.5)
- b. Radiation Protection Support Supervisor (0.5)
- c. Unit Shift Supervisor (0.5) OR ShMt Sogerdsor
- d. Unit Operations Shift Supervisor (O.5) OR ShiR 50pevisoR I"'*'dI" '
- REFERENCE O S S W ffe7
- PVNGS Procedure 75AC-9ZZ01, paras. 5.8.2.3, 5.8.2.1, 4.8, and 5.4.1.4 l
i l
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7-1 i
i 1
i
- 7.02 QUESTION (1.5) ]
- a. "ALARA" is an abbreviation for what phrase ? (0.5)
- b. What is the goal of the ALARA concept ? (1.0)
- 7.02 ANSWER
- a. As low as reasonably achievable. (0.5)
- b. To keep radiation exposure to individuals (0,5) and releases of radioactive materials (0.5) as low as reasonably achievable.
- REFERENCE 10 CFR 20.1(c).
I 7-2
- 7.03 QUESTION (3.0)
When responding to a plant emergency, list three conditions when use of the FUNCTIONAL RECOVERY PROCEDURE, 41RO-1ZZ10, is required.
- 7.03 ANSWER (Note: 1 point each)
- 1. When the [ Diagnostic Flow Chart (Appendix A) of the]
EMERGENCY OPERATIONS Procedure [41EP-1ZZ01] has been attempted and is unclear or leads to a diagnosis the CRS knows is incorrect [or incomplete). (1.0)
- 2. A Recovery Operation Procedure [41RO-1ZZ01 through 41RO-1ZZO9] is in use but is not mitigating the event [due to equipment failure or multiple event occurring). (1.0)
- 3. An unexplained loss of control of a safety function has occurred. (1.0)
- REFERENCE Procedure 41EP-1ZZ01, EMERGENCY OPERATIONS, para. 1.3 I
i 7-3
- 7.04 QUESTION (2.0)
While operating at full power, Unit 1 has been tripped as a result of a steam generator tube rupture. You are the Shift Supervisor and have been informed the size of the leak is about 30 rpm and the Radiation Protection Group has advised you the projected offsite dose, based on the Condenser Exhaust Monitor, is about i uci/cc.
- a. Based on the attached portions of EPIP-02, what emer-gency classification would you assign to this event ?
(1.0)
- b. Explain your answer. (1.0)
- 7.04 ANSWER I a. Alert (1.0)
- b. [Although the RCS leak rate corresponds an Unusual Event classification per Table 1, the] Projected offsite dose rate corresponds to the value for Alert classification [>4.60 E-1 uci/cc] given in Table 2 of Appendix A of EPIP-02.
[According to para. 4.1.1 of EPIP-02, if a conflict or uncertainty exists, the more conservative classification l should be used.] (1.0)
- REFERENCE EPIP-02, EMERGENCY CLASSIFICATION.
7-4
FOR INFORMATION ONLY
- ~ ; "~ PVNGS EMERGENCY PLAN PROCEWRE -
o ._ 7 IMPLEMENTING PROCEDURE EPI?-02 . .
4 REVISION IMERGENCY CLASSIyICATICN 4 Page 7 of 27 4.0 DETATTTn TRuu.uusr.
4.1. Tersmna.1 Ted--w 2.on/Respesib414t4 es m '
2he rationale and criteria used to derive Appendices A and B is given in Appendix C.
Appendix C describes the criteria used in the
, development of the emergency classifications.
It is intended for information only, not for event classification.
. IPIP-02 is not implemented for the notification of significant events unless a situation degrad's e to the point of impac. ting on a fission product barrier and compromising a safety -
- function. Notification of significant events is .
made per 71AC-9ZZ01, " Event Related Reporting".
4.1.1 If a conflict or uncertainty exists, the more conservative, higher numbered, Imple:enting Action EPIP should be initia:ad chan classifying the avanz.
, 4.1 2 yor Emergency Classifications of ALEKT or higher, the
- . Shift Supervisor of the designated unaffected unit shall ,
.. re11 ave the Shift Supervisor of the affected unit as the J . ". . *. * ._, Imergency Coordinator. Tor Nos 434-=-4= of Unusual Ivezrts
.' . ,i .it uill be the discretion of the Shift Supervisor of the affected unit, if he is to be relieved as Emergency 5,4.f..,,s -
rW4
- er by the Shift Suparvisor of t:ha dasit stad g*g , . .. -+m-4 d unit. . .
ji Y 4.3 3 %e ncz: sal assignments of designated unaffected unit Shift i Supervisors are listed.
- , .- If conditions exist which make -
the use of tha listed Shift Supervisor undesirable,
.--- another qualified individual may relieve as the Emergency -
- Coordinator at the discretion of the affected unit Shift Supervisor.
4.1.4
.If cytw4reton of barrier chs1lange or fal. lure ex.ists ~~
l which is inconsistent with the recovery procedure in use, initiate the E=ergency Classification indicated and
- rediagnose plant conditions to identify any additional prbcedure which may be necessary to address existing !
conditions. . l FOR INFORMATION ONLY'
FOR INFORMATION ONLY (P:
- , Qy
- ~ PVNGS EMERGENCY PLAN PROCEW RE .
-J, NO. . - /,
. . . . IMPLEMENTING PROCEDURE EPIP-02 ,
REVISION MGENCY CLAsdAA AG"TCN 4 Page 3 of 27
- ~
4.1.3 The rationale used to develop the classification of events i
based on 4-die.ations of ba==ier challenge or fm41nve is provided in Appendix C.
zularienship between safety This tm+m infe= Mmdaa4=---de descr.bes the
_ ; i y. ,
4.1.6 -
Responsibilities of the affected unit shift supervisor are:
- 4.1.6.1 Initial classification of the event per this procedure.
N0TE Designated Unaffected Unit Shift Supervisor to assume the role to the
- Emergency Coordinator in the Onshift E=ergency Organization are:
gg- Affected Unit SS Unaffected Unit SS Unit 1 Unit 2 Q3 Unit 2 Unit 1 \
h Unit 3 Unit 2 Entire Site Unit 1 4.1.6.2 Notificacion of the Shift Superv.isor of the d==4 ameed
- nnaffecced 1x::
Coordinavnr. tre vther individual selected as Emergrney -
,. 4.1. 6.3 Drganization of the em=h4 *r staff to placa the plant in a
, , , ,, safa candition.
4.1.6.4 s' = - Assumptian of the Imargency Cor=44nator's ra-4+4aa nutil l :
" .ra1Leved.
41.'7 -
7 ;M W== af the I-,,r.ucy Cormi4n=+or are:
l 4.1.7.1 Overall responsibility for directing the onshift
, 4 i -
emergency response organization.
4.1.7.2 Implement EPIP's based on initial classification.
4.1.7 1 Verifiestion/reclass4f 1 ( e -n = e e = ~ 4 an =_ are 4r'*4rm comple:ed.uf the evexxx af ter initial -
4.1.7.4 Honitoring plant conditions and rec.lassifying the event g%-J
- as necessary until the event is terminstad.
. 4.1.7.5 Doungrade tha avent based on plant status with all safety-functions satisfied and boundary status verified.
- ' ~
POR INFORMATION ONLY
FOR INFORMATION ONLY PVNGS EMERGENCY PLAN [ROCEDURE o
i g(:2 IMPLEMENTING PROCEDURE EPIP-02 l REVISION EMERGENCY CLASSIFICATION 4 Page 9 of 27 4.1.7.o Terminate the event taking into account that the event -
has been downgraded and the anticipated plant response is such that there should be no challenge to any fission product barriers or radiation teleases in excess of Tech l
Specs; and present plant conditions are such that there is no possibility of an adverse impact on the health or safety of the general public or plant personnel.
4.2 Prerequisites ,
4.2.1 A situation has occurred which requires the implementation of the PVNGS Emergency Plan to protect the health and safety of the public.
4.3 Instructions 4.3.1 Vnen plant conditions are such that Emergency Plan implementation may be required..the Shift f -
Supervisor / Emergency Coordinator shall perform the followin,g:
4.3.2 Cla'ssify the event using the appropriate appendix:
. Appendix A - If an event oriented Recovery Procedure or the functional Recovery Procedure is in use.
Appendix B - If a non-Rx trip event has occurred.
l -
4.3.3 If 41RO-1ZZ01, " Reactor Trip Recovery Procedure", is in use and effectively directing the maintenance of Critical Safety l Functions and plant recovery, the CRS/EC may elect to NOT classify an uncomplicated Reactor Trip as NUE. Plant-parameters must be trending as expected in order to NOT l classify a Reactor Trip' as a NUE. Appropriate notifications should be made per 71AC-92Z01, " Event Related Reporting".
4.3.4 Record the date/ time / events of initial classification. Upon l verification / reclassification of the event, record the date and time and supporting information.
l 4.3.5 Initiate and complete the implementing actions given in the l appropria e classified-event implementing procedure (i.e., p EPIP-037 4) 0~ u 05)r
~~""
FOR INFORMATION ONLY
FOR INFORMATION ONLY
' ' ~
o PVNGS EMERGENCY PLAN PROCEDURE -
" N O.
APPENDIX A IMPLEMENTING PROCEDURE EPIP-02 Page 1 of 4 .-
REVISION I
EMUENCI CI.ASSITICATICN 4 Page 10 of 27 Offsite Dese aM Barrier Challenge /Tailure Event Classifiestion 1.0 Determ2ne the event e1=4 +4-*4= as folhurs:
~- ~1.1' -W ---
-, " ~im thallenges/ failure w Appemiix 1, Mla 2. , ,
1.2 Evaluate any current offsite radioactive release per Appendix A, Table 2.
NOTE .
Protective Action recommendations are based on b plant and contain=ent conditions anc these recommendations are made to offsite officials even when no release is in progress.
- 2.0 Select the most restrictive, higher classification, from the Table 1 and Table 2 evaluations as the event classification.
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FOR INFORMATION ONLY
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l PVNGS EMERGENCY PLAN PROCEDURE NO. APPENDIX A,.
IMPLEMENTING PROCEDURE EPIP-02 Page 2 of 4 . e REVISION NmNCY C.ASSITICATION 4 Page 11 of 27
, , Appendix A
- Table 1 - Ear-ier Cha11 ente /Failura Classifiestien Criwis
'" 2 .D ' 1 hik a h L i -, vf %= following conditions shar razrently exiszi RCS CLAD
- CONTAIhWEm' l
RVIF.S indicates A'lVS Physical breach voiding in upper -
of er-" 4= ent plenum -
_OR OR RCS pressure Excessive RCS
> 2750 psia Activity (> 300 CIAS required but ue/gm dose equiva- not cc::pleted (i.e.
Uncontrolled loss lent I-131) both automatic of RCS inventory valves in a pene-
> 50 gpm CET > 700 F tration fail to close) i H2 * ***"****I*"
> 3.5% by volume Cantainment pressure
> 50 psig
'~
, Vital Auxiliaries / Radiation Release 4
,- Toss of offsita and onsite AC power *
's.):kO
~ . Toss nf offsita and ansita AC power int langer than 40 miaures
"?'
.' ."Ioss af an. Class IE .DC power. _
v.
Failure of ESF Safety Systems (both trains) to actuate when required .
, > 10gpm pricary/ secondary leakage concurrent with LOP
~~
l .
DR
. :. - .. t t - .
' ,f > 10 gpm primary / secondary leakage concurrent with loss of secondary coolant outside containment
! f.g..
l l .
l
[- -
EOR INEORMATION ONLY
FOR INFORMATION ONLY PVNGS EMERGENCY PLAN PROCEDURE
.QA NO. APPENDIX A CF IMPLEMENTING PROCEDURE EP:P-02 Page 3 of 4 f REVISION '
EMERGENCY CLASSIFICATION 4 Page 12 of 27 Appendix A Table 1 - Barrier Challenge / Failure Classification Criteria (Con't'd.)
2.0 Determine the emergency classification level for barrier chall.enge/ failure per the following guidelines:
Number of Checks made in 1.0 Barrier Status Classification 0 No barriers lost Unusual Event or challenged (EPIP-03) 1 One barrier lost Alert (EPIP-04) or challenged 2 Two barriers lost Site Area Emergency or' challenged (EPIP-g) p 3 or Three barriers lost General Energency more or challenged (EP!Pg)
O b
B FOR INFORMATION ONLY
~
FOR INFORMATION ONLY .
, PVNGS EMERGENCY PLAN [R CEDURE
($ IMPLEMENTING PROCEDURE EPIP-02 A?? D1X A Page 4 of (
REVISICN EMEF.GESCY CI.ASSITICATION 4 Page 13# of 27,-
Appendix A Table 2 - Offsite Dese Proiecti n Classifiestien Criteria Based upon infor=ation provided by the Radiatica Pre action Group and/or
, determine the emergency classification level according to the folleving:
Notification of Unusual Event (EPIP-03)
Plant Vent Menitor RU-143.Chn. 1 > 6.62 E-4 uci/cc Fuel Bldg. Exh. Menitor RU-145 Chn. 1 > 1.63 E.-3 uci/cc Condenser Exh. Monitor RU-141 Chn. 1 > 4.60 E-2 uci/cc Alert (EPI?-OL)
Plan: Vent Monitor ' .RU-143 Chn. 1 > 6.62 E-3 uci/cc
. Tual Bids Exh. Monito RU-146 Chn. 1* > 5.91 E-2 uct/cc w Ccndenser Exh. Monitor RU-142 Chn. 1 > 4.60 E-1 uci/cc
'te.
(-
t Site Area Emereenev (EPIP- M-)
Plan: Ven: Monitor RU-144 Chn. 1 30 min. G > 2.20 E-1 uci, RU-144 Chn. // 2,, min. @ > 2.20 uci/cc Fuel Bldg. Exh. Monitor RU-146 Chn. 1
- ~ ~~
30 min. G > 1.95 uci/cc RU-146 Chn. 2 2 min. G > 1.95 E+1 uci/c Condenser Exh. Monito: RU-142'Chn. 1 30 min. G > 1.53 E+1 uci, RU-142 Chn. 2 2 min G > 1.53 E+2 uci/cc General E- errency (EPIP- )
Plan: Ven: Monitor RU-144 Chn. Kt > 4.40 uci/cc Fuel Bldg. Exh. Monitor RC-146 Chn. 2 > 3.91 E+1 uci/cc Condenser Exh. Monitor
- RU-142 Chn. 2 > 3.05 E+2 uci/cc -
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_. FO_R INFO __RMATION ONLY __
- 7.05 QUESTION (2.0)
- a. The plant is in Mode 3 and the RCS cold leg temperatures of the four loops are 534*F, 533*F, 530*F and 532*F. Can the plant go critical under these conditions 7 (1.0)
- b. Explain your answer to the above question. (1.0)
- 7.05 ANSWER
- a. No. (1.0)
- b. Reactor cannot be made critical unlesn'the lowest RCS cold leg temperature is equal to or greater than 44HP4% (1.0) 551* F
- REFERENCE Palo Verde Unit 1 Technical Specification 3.1.1.4.
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- .' . 3 6 QUESTION (1.0)
According to plant operating procedures, under what condition is addition of positive reactivity by more than one method at a time (excluding xenon effects) permitted 7
- 7.06 ANSWER
~
When the reactor is generating [ sensible] heat [ greater than
-3 % power].
about 10 (1.0)
- REFERENCE Procedure 410P-12ZO3, REACTOR STARTUP, para. 3.8 i
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- 7.07 QUESTION (4.0)
Complete the following statements:
- a. When starting and stopping Reactor Coolant Pumps (RCPs),
consideration should be given to flow through the core and maintaining capability. (1.0) ,
- b. To preclude , the fourth RCP shall not be started until the Reactor Coolant System temperature is equal to or greater than 500*F. (0.5)
- c. Do not operate a RCP longer than (maximum time period) after a loss of Nuclear Cooling Water (NCW),
with seal injection flow established. (0.5) i
- d. If the plant is operating at power and loses both Seal Injection and Nuclear Cooling Water, you should restore Cooling Water within (time period) or:
- 1. ,
- 2. , and
- 3. .(2.0)
- 7.07 ANSWER
- a. When starting and stopping Reactor Coolant Pumps (RCPs),
consideration should be given to balancina the (0.5) flow through the core and maintaining nressurizer serav (0.5) capability.
- b. To preclude core unlift (0.5) , the fourth RCP shall not be started until the Reactor Coolant System temperature is equal to or greater than 500*F.
- c. Do not operate a RCP longer than 10 minutes (0.51 (maximum time period) after a loss of Nuclear Cooling Water (NCW),
with seal injection flow established.
- d. If the plant is operating at power and loses both Seal Injection and Nuclear Cooling Water, you should restore Cooling Water within one minute (0.51 (time period) or:
- 1. trin __ the remeter (0.5) ,
- 2. _ ston the raffectedl RCP(s) (0.5) , and
- 3. close associated bleed-off valves rHV-430. -431.
-432 and -4331 (0.5) .
- REFERENCES Procedure 410P-1RC01, REACTOR COOLANT PUMP OPERATION, paras.
3.12, 3.13, 3.15 and 3.19. Procedure 41AO-1ZZ29, REACTOR COOLANT PUMP AND MOTOR EMERGENCY, para. 3.1.
7-7
- 7.08 QUESTION (1.5)
During a refueling outage, fuel handling operations are in progress inside Containment when a local area radiation monitor sounds an alarm. List three immediate actions that should be taken.
- 7.08 ANSWER
- a. Evacuate all personnel from containment (0.5)
- b. Notify the Radiation Protection Section (0.5)
- c. Verify Containment Purge Valves [UV-2A&B, UV-3A&B, UV-4A&B and UV-5A&B] are closed (0.5)
- REFERENCE Procedure 41AO-1ZZ26, IRRADIATED FUEL DAMAGE, paras. 2.1 to 2.3.
7-8
- 7.09 QUESTION (2.0)
List four prerequisites for energizing a 480 Volt Class 1E Switchgear Load Center.
- 7.09 ANSWER (0.5 each for any four of the following)
- a. [ Associated] 4.16 KV bus energized
- b. [125 VDCJ Control power available to Load Center
- c. Relays reset on the 480 V Load Center Main Feeder Breaker
[at the switchgear cubicle]
- d. Relays reset on the 480 V Load Center Transformer Supply Breaker [at the switchgear cubicle]
- e. 480 V Load Center Main Feeder Breaker is racked in 1
- REFERENCE Procedure 410P-1PG01, 480 VAC CLASS 1E SWITCHGEAR, Sec. 4.2 Nde : Acid . tJ o Activ e Cl eara n c e5
- G roundin ) .D evic es Remove d
- O fedktVs Gm Leas Cen+ee opes or RaeI<ed ov1
- NO (avUS on bOS 7-9
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- 7.10 QUESTION (3.0)
- a. List three design features provided to correct a low pressure condition in the Instrument Air header. (1.5)
- b. For total loss of instrument air pressure, indicate whether the following flow paths are UNAFFECTED or ISOLATED by the loss: (0.5 each)
- 1. RCS Letdown
- 3. Pressurizer Auxiliary Spray
- 7.10 ANSWER
- a. Redundant air compressor (s), Service Air isolation and ;
Backup Nitrogen [ system] (1.5) I
- b. 1. ISOLATED (0.5)
- 2. ISOLATED (0.5)
- 3. UNAFFECTED (0,5)
- REFERENCE Procedure 41AO-1ZZ06, Loss of Instrument Air, Appendix B and P& ids 13-M-RCP-001 and 13-M-CHP-001 t
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- 7.11 QUESTION (1.5)
An RCS Water Inventory Balance has been performed during a plant heatup with the RCS at 380*F, 500 psia. The results are 0.1 gpm Unidentified Leakage and 9.8 gpm Identified Leakage.
- a. Would you continue the plant heatup? (0.5)
- b. Explain your answer to a. (1.0)
- 7.11 ANSWER (1.5)
-->a. No. (0.5)
- b. The T.S. limit for identified leakage is 10.0 spm. With Identified Leakage at 9.8 gpm at such a low pressure, ,
increasing RCS pressure will soon result in exceeding the !
LCO [ leakage flow proportional to the square root of the I (1.0) pressure difference).
- REFERENCE PVNGS Procedure 41ST-1RC02, RCS WATER INVENTORY BALANCE, Sec. 4.4.5.2.1.c; Technical Specification 3.4.5.2 NOTE
- A "3es" m fo*S* 3"fp e hA pca oplanatom wll\ be
" e valua te d ewd $raded based ow % e.
Va ll' d i b[ of the 4 y. f lR wc$Y a'0%, .
i 7-11
- 7.12 QUESTION (1.5)
List three separate conditions that require EMERGENCY BORATION. <
State all your assumptions.
END OF CATEGORY 7 PROCEED TO CATEGORY 8
- 7.12 ANSWER (Any 5, 0.5 each) -
- a. Failure of 2 or more CEAs to trip following a reactor trip.
- b. Shutdown margin is less than 6.0% [ Modes 1 to 4].
- c. Shutdown Margin is less than 4.0% with T less than or equal to 210'F. avs
- d. Reactor power not decreasing to source level following a reactor trip.
- e. During refueling:
-K greater than 0.95
-B8$$nconcentrationlessthan2150 ppm
- REFERENCE PVNGS Procedure 41AO-1ZZ01, EMERGENCY BORATION, p. 6, and Procedure 41RO-1ZZ10, FUNCTIONAL RECOVERY PROCEDURE, p. 12.
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CATEGORY 8 - ADMINISTRATIVE PROCEDURES. CONDITIONS AND LIMITATIONS
- 8.01 QUESTION (2.0)
- a. What is the pressurizer low pressure trip setpoint for safety injection in Modes 1 and 2 ? (0.5)
- b. Describe the pressurizer low pressure trip setpoint for .
safety injection in Modes 3 to 6 ? (1.5)
- 8.01 ANSWER
- a. 1837 psia or 11822 psia. (0.5)
- b. Above 400 psia, trip setpoint can be manually reduced to 100 Psia or 400 psi below pressurizer pressure - whichever is greater.(1.0) Trip can be manually bypassed when pressurizer pressure is below 400 psia.(0.25) Trip is automatically reinstated when pressurizer pressure exceeds 500 psia.(0.25)
- REFERENCE Palo Verde Unit 1 Technical Specifications, Table 3.3-4, Item I.A.2 and Footnote (1) l 8-1 I
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- 8.02 QUESTION (3.0)
- a. List the job titles of the Minimum Shift Crew required by the Technical Specifications during Modes 1, 2, 3 and 4, and l the number of personnel per Unit required in each of these positions. (1.0) l
- b. What additional personnel (excluding Security) are reouired to be on-site during these modes ? (1.0)
- c. How do the above requirements change during Modes 5 and 6 ?
(1.0)
- 8.02 ANSWER /
- a. SS-1, SRO-1, RO-2, AO-2 and STA-1. (0.2 each)
- b. Radiation Protection Technician (0.5) and a 5 man site Fire Team, not including personnel in part a, above (0,5).
O? // need i Ro eme t Ao,
- REFERENCE Technical Specification 6.2.2.1.
1 8-2
- 8.03 QUESTION (1.0)
If it should become necessary to start the Fire Pumps manually, from what locations can this be accomplished ?
- 8.03 ANSWER From the Unit 1 Control Room, and locally. (0.5 each)
- REFERENCE PVNGS Unit Differences Handout, July 10, 1986.
O 8-3
- 8.04 QUESTION (3.0)
- a. When a Clearance is issued for maintenance work, what individual is responsible that the equipment to be worked on is properly removed from service, drained, vented and/or de-energized to ensure it is placed in a condition to be worked on with safety to men and equipment ? (0.5)
- b. Name four considerations (other than the safety of personnel and equipment) the Responsible Supervisor must include in his evaluation when deciding whether a Clearance can be issued to release equipment for maintenance. (0.5 each)
- c. Describe how you would fill out a Tag Assignment Sheet for a component that has no equipment number (e.g. a temporary valve or a root isolation valve not shown on the drawing).
(0.5)
- 8.04 ANSWER
- a. The Responsible Supervisor: Shift Supervisor nr Assistant Shift Supervisor. (0.5)
- b. Effects on (0.5 each for any four of the following):
- 1. Security System
- 2. Fire Protection System
- 4. Method of Emergency Cooling
- 5. Establishment of a path for Decay Heat Removal
- 6. Technical Specification Limiting Conditions for
. Operation
- c. Attach a copy of the P&ID showing the location of the valve in RED and indicating which Tag No. is to be hung on it.
Reference attached drawing on Tag Assignment Sheet for that Tag. (0,5)
- REFERENCE Procedure 40AC-9ZZ15, STATION TAGGING AND CLEARANCE, j
sections 4.3.1, 5.1.2.2 and 5.1.4 N k '. A M m sumum Powee Capablld~) .
f allowin3 *.
Poss;ble Radioacisve Re lease.s AnclIlary a geratu'ons (wRF, Adein oldy sfc.)
Te c h , ade g vacy of Cleorance
.v:,,,,... . -- - : : +~
l dyr'stin y C le e'ra n c es 8-4
- 8.05 QUESTION (1.5)
List three actions that should be taken when a Clearance, Temporary Modification or plant condition results in disabling a safety related system for which no automatic input to the Safety Equipment Status System (SESS) panel is provided.
- 8.05 ANSWER
- 1. Initiate a manual Bypass / Inoperable signal on the SESS panel. (0.5)
- 2. Document the initiation of the manual signal in the Control Room Log. (0.5)
- 3. Inform the Shift Supervisor. (0.5)
- REFERENCE Procedure 40AC-9ZZO2, CONDUCT OF SHIFT OPERATIONS, Sec. 7.11 a
8-5
- 8.06 QUESTION (1,0)
An operating procedure directs that independent verification be performed to ensure a specified device is in the required position. Describe how this independent verification is documented.
- 8.06 ANSWER By a second verification column on the system lineups or on the Independent Verification Record. (1.0)
- REFERENCE Procedure 40AC-9ZZO2, CONDUCT OF SHIFT OPERATIONS, paras.
13.2 and 13.5.
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-b 8-6
- 8.07 QUESTION (3.0) i Due to plant conditions, it is necessary to make a temporary change to the surveillance test for a safety-related device.
List the three conditions that must be met, per the Technical Specifications, if such a change is to be made.
l l *8.07 ANSWER
- a. Change does not change intent of procedure.
- b. Change is approved by two members of [ plant] supervisory staff, at least one of which is a shift supervisor or
( assistant shift supervisor and has an SRO license on the l affected unit.
- c. Change is documented (.25), reviewed (.25), and approved by PVNGS Plant Manager or designee (.25) within 14 days of implementation.
!
- REFERENCE l
Facility Technical Specification 6.8.3.
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- 8.08 QUESTION (3.0) l l a. List the three Safety Limits (not power distribution limits) l for this facility as stated in Section 2 of the Technical l Specifications. (.33 each)
- b. List two regulatory requirements specified in the Facility ,
Technical Specifications that are applicable to the Control Room staff and plant operations if a Safety Limit is ]
violated. (1.0 each) l
- 8.08 ANSWER l
- a. 1. DNBR (.33 each)
- 2. Peak Linear Heat Rate
- 3. Reactor Coolant System Pressure
- b. 1. Notify NRC Operations Center within one hour. (1.0)
- 2. Critical operation of the unit cannot be resumed until authorized by the Commission. -(1.0)
I
- REFERENCE Facility Technical Specifications 2.1 and 6.7.
- 3. Plare (1 3 ,' t j m HoTSTAAlp3/ withis one.bo0P.
4 If A7so psiga .' s eu c ea e d , d ecrease pressure o 6 anso psa uhh a h,or is m es I and 2, OY WiYhiv S n1IMotes o' yt #100ES 3, y a n d 5.
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l *8.09 QUESTION (1.5)
- a. What is the operability requirement for Reactor Coolant Pumps when the reactor is in Operational Mode 1 7 (0.5)
- b. What Action is required in what period of time if the operability requirement for the Reactor Coolant Pumps cannot be satisfied 7 (1.0) 1 *8.09 ANSWER
- a. All four pumps must be operating. (0.5)
- b. Be in [at least] Hot Standby (0.5) within one hour (0.5).
- REFERENCE Facility Technical Specification 3.4.1.1 l
l 8-9
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- 8.10 QUESTION (2.0)
- a. Based on the facility procedure for gaseous radioactive releases, what action must be performed daily while a containment purge is in progress ? (1.0)
- b. List two reasons given in the Facility Technical Specifications for allowing the 8-inch containment purge supply and exhaust to be open. (1.0)
- 8.10 ANSWER
- a. An evaluation should be performed [ daily] to determine if the containment purge is still necessary [and if the current permit's dose projections are accurate]. (1.0)
- b. (0.5 each for any two of the following)
- 1. For pressure control.
- 2. For ALARA and respirable air quality considerations for personnel entry.
- 3. For surveillance tests that require the valve to be open.
- REFERENCE Procedure No. 75AC-9ZZO2, GASEOUS RADIOACTIVE EFFLUENT RELEASE ADMINISTRATIVE CONTROL, Sec. 5.1.3.3, and Facility Technical Specification 3.6.1.7.b.
I 8-10
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- B.11 QUESTION (2.0) !
l l While operating in MODE 1 and making repairs to one ECCS subsystem that is inoperable, it is determined the other train is also inoperable. The technical specification LCO applicable to these subsystems in MODE 1 is attached.
! a. Since the ACTION statement of the attached LCO does not ,
I address the loss of both subsystems, what initial action is required to be taken by the operators 7 (1.0) ;
l
- b. How soon must the above action be taken 7 (0.5)
- c. In what section of what document is th'e requirement for the above actions stated (be specific as to the number of the Section) 7 (0.5) ;
- 8.11 ANSWER
- a. Initiate action to place the Unit in a MODE in which the specification does not apply, OR, initiate action to shutdown the unit. (1.0)
- b. Within one hour. (0.5)
- c. Technical specifications (0.25), Section 3.0.3 (0.25).
NOTE: Answers consistently based on technical specification 3.5.3 will also be accepted.
- REFERENCE Facility Technical Specification 3.0.3.
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l EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS - T GREATER THAN OR EQUAL TO 350*F old LIMITING CONDITION FOR OP'ERATION 3.5.2~ Two independent Emergency Core Cooling System (ECCS) subsystems shall be OPERABLE with each subsystem comprised of: .
- a. One OPERABLE high pressure safety injection pump,
- b. One OPERABLE low pressure safety injection pump, and
- c. An independent OPERABLE flow path capable of taking suction from the refueling water tank on : safety injection actuation signal and automatically transferring suction to the containment sump on a recirculation actuation signal.
APPLICABILITY: MODES 1, 2, and 3*.
ACTION:
- a. With one ECCS subsystem inoperable, restore the inoperable subsystem i
! k' to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following i( 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days des-cribing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for !
i each affected injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.
l "With pressurizer pressure greater than or equal to 1837 psia.
L PALO VERDE - UNIT 1 3/4 5-3 i
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- 8.12 QUESTION (2.0)
The PVNGS Technical Specifications require that Shutdown Margin be greater than 6% while in Modes 1 through 4 and greater then 4% while in Mode 5.
- a. What is the definition of Shutdown Margin ? (1.0)
- b. What are the most restrictive accident conditions (i.e. -
Technical Specification Bases) which require the 6% Shutdown Margin limit ? (0,5)
- c. What are the most restrictive accident conditions (i.e.
Technical Specification Bases) which require the 4% Shutdown Margin limit ? (0.5)
END OF CATEGORY 8 END OF EXAMINATION
- B.12 ANSWER
- a. SDM is the amount of reactivity by which the reactor is, or would be, subcritical from its present condition assuming no change in part-length CEA position and all other CEAs are fully inserted except the single assembly of highest worth <
which is fully withdrawn. (1.0)
- b. The 6% SDM limit is based on controlling the reactivity transient associated with [an uncontrolled) RCS cooldown caused by a steam line break at the end of life and T c at no load operating temperature. (0.5)
I
- c. The 4% SDM limit is based on ensuring that reactivity transients resulting from a single CEA withdrawal event are l minimal. (0.5) !
j
- REFERENCE PVNGS Technical Specifications, Definition 1.28, and Bases ,
3/4.1.1.
! 8-12
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MhbT{S U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: PALO VERDE 1&2 REACTOR TYPE: PWR-CE80 DATE ADMINISTERED: 86/09/23 EXAMINER: DEFFERDING, L.
CANDIDATE: AflSWER KEY INSTRUCTIONS TO CANDIDATE:
Use 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 ind.fcated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 25.00 25.00 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 25.00 25.00 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 25.00 25.00 3. INSTRUMENTS AND CONTROLS 25.00 25.00 4. PROCEDURES - NORMAL,. ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 100.00 Totals Final Grade '
All work done on this examination is my own. I have neither given nor received aid.
Candidate's Signature
CD NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
- 1. Cheating on the examination means an automatic dental of your application and could result in more severe, penalties.
- 2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
- 3. Use black ink or dark pencil only to facilitate legible reproductions.
- 4. Print your name in the blank provided on the cover sheet of the examination.
- 5. Fill in the date on the cover sheet of the examination (if necessary).
- 6. Use only the paper provided for answers.
- 7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
- 8. Consecutively number each answer sheet, write "End of Category _" as appropriate, start each category on a new page, write only on one side of the paper, and write "Last Page" on the last answer sheet.
- 9. Number each answer as to category and number, for example,1.4, 6.3.
- 10. Skip at least three lines between each answer.
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
- 12. Use abbreviations only if they are commonly used in facility literature.
- 13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer. required.
, 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
l l 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE -
QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
l 16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.
b
- 18. When you complete your examination, you shall:
- a. Assemble your examination as follows:
(1) Exam questions on top.
(2) Exam aids - figures, tables, etc.
(3) Answer pages including figures which are part of the answer.
- b. Turn in your copy of the examination and all pages used to anrwar the examination questions,
- c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions.
- d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your Itcense may be denied or revoked.
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 2 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.01 (3.00)
A questionable ECP calculation reveals that criticality should be achieved when 4% delta K/K has been added to the core. A boron dilution of 200 ppm B (2% delta K/K) is planned, followed by control rod withdrawal to criticality. The initial count rate is 30 cps.
- a. WHAT is the expected count rate after the 200 ppm B boron dilution? (1.0)
- b. Following the first 100 ppm B dilution, the source ranges indicate 50 cps. WHAT amount of reactivity in % delta K/K should the original ECP calculation have derived as sufficient to reach criticality? (2.0)
QUESTION 1.02 (2.25) -
In a hypothetical reactor, power is decreasing during a control rod insertion. DESCRIBE HOW reactor power would react after rod motion is stopped if the resultant reactor condition was each of the following:
- a. supercritical 0.75 b, critical 0.75
- c. subcritical 0.75 QUESTION 1.03 (2.25)
A plot of K-eff versus the moderator-to-fuel ratio (Figure 1.1) -
peaks in the middle; the areas to either side of the peak are'
, referred to as either "undermoderated" or "overmoderated." .
- a. LIST the competing processes that cause the peak in this curve '
and HOW each of these processes is affected by INCREASING moderator temperature (less moderator). (1.'0)
- b. The Palo Verde reactor is either undermoderated or overmoderated at power operation. STATE which AND AND why it was so designed. (1.25) l l
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 3 i
THERMODYNAMICS HEAT TRANSFER AND FLUID FLOW QUESTION 1.04 (3.50)
The Palo Verde reactor is operating at 100% power with 300 ppm boron.
Reactor power is to be decreased to 85%. Initial Group 5 rod height is 130 inches withdrawn,
- a. Use Figures 1.2 and 1.3 to determine the final rod height if
- boron concentration is to remain unchanged. (2.0) i
- b. WOULD a SMALLER or LARGER rod insertion be required if the same power maneuver was performed at EOC7 WHY? (1.5)
QUESTION 1.05 (3.00)
ANSWER the following TRUE or FALSE.
I a. Equilibrium samarium concentration is a function of power. (0.5)
- b. Equilibrium xenon reactivity is dependent upon core burn up. (0.5)
, c. Samarium concentration initially increases after a reactor trip.
! (0.5)
- d. Xenon concentration initially increases after a reactor trip. (0.5)
- e. Equilibrium samarium concentration after a trip is a function of the previous power level. (0.5)
- f. Equilibrium xenon concentration after a trip is a function of the previous power level. -
(0.5) l QUESTION 1.06 (1.00) '
WHICH of the following does NOT occur over core life, i.e., from BOC to EOC? (1.0)
. axial flux distribution flattens
. effective fuel temperature decreases
. MTC becomes more negative
. reactor response time increases
(***** CATEGORY 01CONTINUEDONNEXTPAGE*****)
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- 1. PRINCIPLES OF NU; LIAR POWER PLANT OPERATION, PAGE 4 THERMODYNAMICS, h'IAT TRANSFER AND FLUID FLOW
~
QUESTION 1.07 (1.00)
A relief valve on a pipe opens at 900 psia. ;The temperature of the exhausted steam is 320 deg F. The temperature of the fluid (water or steam) within the pipe is approximately: (1.0)
- a. 540 deg F
- b. 400 deg F
- c. 320 deg F
- d. 212 deg F QUESTION 1.08 (2.50) 12.0 x 10**6 lbm/hr of feedwater at 426 deg F and 1060 psig is being delivered to the steam generators. Using steam tables CALCULATE the approximate fraction of core rated thermal power that is being produced by the reactor. SHOW your work. '
(2.5)
QUESTION 1.09 (1.00)
The margin to DNB: (1.0)
. decreases with increasing flowrate
. increases with increasing reactor power
- c. increases with increasing pressurizer pressure
- d. increases with increasing T-ave
{ -
l QUESTION 1.10 (2.00) ,
A reactor trip occurs at 100% power due to loss of offsite power.
CALCULATE the core flow, in terms of a percent of core flow at '
power, that exists when natural circulation is established.
STATE all assumptions. SHOW'your work. (2.0) l -
/
(***** CATEGORY 01CONTINUEDONNEXTPAGE*****)
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 5 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.11 (2.50)
A large vented water tank 60-ft high has a small capacity centrifugal pump taking a suction from its base. The pump requires 5 ft of net positive suction head (NPSH) to prevent cavitation and it is located at a vertical elevation corresponding to the bottom of the tank.
The tank is initially entirely full of water and is maintained at 60 deg F by heaters. ASSUME that the vent becomes totally clogged .
(fully closed) while the pump is in operation. ANSWER the following questions.
- a. To WHAT level would the tank drop to, before cavitation occurs, as the pump continues to remove water from the tank? (1.0)
- b. WHAT two (2) equipment failures could result from continued pump operation with a clogged vent? (1.0)
- c. To WHAT level could the pump continue to pump water to, before ,
cavitation occurs, if the vent was then opened? (0.5)
QUESTI 1.12 (1.00)
Two (2) id tical centrifugal pumps are connected in parallel in a
- fluid system. If initially only one (1) pump is operating, and l then the second mp is started, WHICH one of the following is
, correct: (1.0)
. total discharge pr sure remains the same
. total system flow do es
. system head loss curve nges .
. minimum required NPSH decr es 5 dM l
(***** END OF CATEGORY 01 *****)
l l
l *
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 6 QUESTION 2.01 (2.00)
ANSWER TRUE or FALSE to the following statements pertaining to the RCP seals,
- a. The maximum cooldown rate for the RCP seals is 100 deg F/ min. (0.5)
- b. Leakage through the No. 3 seal is routed to the reactor drain tank. (0.5)
- c. During normal operation, the pressure drop across each of the three seals is about 33% (+ or - 2%) of the total RCS pressure. (0.5)
- d. The seal injection heat exchanger is designed to maintain seal injection flow at 125 deg F. (0.5) i
\
QUESTION 2.02 (2.00)
- a. DESCRIBE the system design feature that ensures the MSIVs can be closed following a loss of the Instrument and Service Air system. (1.0)
- b. Fill-in-the-blanks.
- 1. During a slow close of the MSIVs hydraulic fluid is i
supplied to the actuating mechanism from the . (0.5)
I
- 2. During a fast close of the MSIVs, the hydraulic fluid is supplied to the actuating mechanism from the . (0.5)
QUESTION 2.03 (2.50)
~
- a. LIST two (2) interlocks and one (1) design feature that exist te prevent overpressurization of the shutdown cooling system '
(SCS) during plant cocidown? (1.5)
- b. HOW is the plant cooldown rate controlled with the SCS? (1.0) l l
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 7 QUESTION 2.04 (3.00)
- a. If an AFAS occurs, LIST four (4) INDICATIONS associated with auxiliary feedwater pump B that will be energized. (2.0)
- b. If, while auxiliary feedwater pump B is running, a loss of offsite power occurs; WHAT will happen to auxiliary feed pump 87 (1.0)
QUESTION 2.05 (3.00)
- a. LIST the three (3) sources of DC power to the DEI 125 DC distribution panel. (1.5) b., LIST three (3) types of class IE - 125 DC loads. (1.5)
QUESTION 2.06 (2.40)
COMPLETE the following table concerning the design of the safety injection system. (2.4)
RCS Pressure Rated Major Component Limit for Injection Flow Volume HPSI a. d.
LPSI b. e.
SIT c. f.
l
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 8 QUESTION 2.07 (2.00) ,
FILL-IN-THE-BLANKS in the following parts of the questions pertaining to the feedwater system. ( Asyme 3yPWS /S Ce 43cno )
- a. With one string of high pressure feedwater heaters out of service, electr cal power out is reduced from full nower to about %. #S$4WE 8/Y#SS /$ d c07ev 3 (0.5)
- b. During normal startup, the feedwater system is placed in automatic control above % power. (0.5)
- c. One feedwater pump alone can be expected to supply %
of system rated flow. (0.5)
- d. During normal operation above 70% power % of the feedwater enters through the downcomer. (0.5) j QUESTION 2.08 (1.50)
Assuming that the nuclear cooling system is operating,
- a. WHAT are the two (2) normal heat loads for the essential cooling water system? (1.0)
- b. WHAT is the one (1) optional heat load for the essential cooling water system? (0.5)
QUESTION 2.09 (2.00) .
- a. WHAT method is used to seal the reactor vest al closure head to the reactor vessel and HOW are these seals activated? (1.0) .
- b. HOW would a failure of one vessel head seal be detected? (0.5)
- c. HOW is leakage from vessel head seal classified according to Technical Specifications on RCS leakage? (0.5)
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 9 QUESTION 2.10 (2.10) ,
- a. Each of the two (2) normal pressurizer spray valves has a manual bypass valve which is adjusted to allow 1.5 gpm bypass flow. WHAT are the two (2) reasons for maintaining this bypass flow? (1.0)
- b. From WHAT location (s) in the RCS is the water obtained for the normal pressurizer spray? (0.6)
- c. WHAT is normally used to reduce the high pressure in the pressurizer when the RCPs are not running? (0.5)
QUESTION 2.11 (2.00)
WHAT automatic functions (if any), other than alarms, are actuated by the following upon exceeding their high setpoint?
- a. Ru-4 (steam generator blowdown monitor) (0.5)
- b. Ru-29 (control room ventilation monitor) (0.5) i
- c. Ru-141 (condenser vacuum pump monitor) (0.5)
- d. Ru-145 (fuel building ventilation exhaust monitor) (0.5)
QUESTION 2.12 (.50)
ANSWER TRUE or FALSE to the following statement. -
(0.5)
During normal power operation, the volume control tank (VCT) 1n the CVCS has a nitrogen overpressure to ensure an adequate NPSH for the charging pumps. ,
(***** END OF CATEGORY 02 *****)
- 3. INSTRUMENTS AND CONTROLS PAGE 10'
, QUESTION 3.01 (3.50) .
INDICATE WHAT (if any) ESFAS actuation signals will be energized by the following: (3.5)
- a. 2 out of 4 containment pressure greater than 6 psig
- b. 2 out of 4 steam generator (1) level less than 25%
- c. 1 out of 4 containment pressure greater than 12 psig
- d. 3 out of 4 pressurizer pressure of 1600 psia QUESTION 3.02 (3.00)
The Steam Bypass control System (SBCS), in addition to controlling bypass and atmospheric dump valves, produces control signals (inhibits, permissives or demand signals) to other equipment. For each of the following situations INDICATE: 1. the control action;
- 2. the system to which the signal is sent; and 3. the purpose of the action. -
- a. steam bypass demand is present (1.5)
- b. load rejection from 100% to 35% (1.5) i QUESTION 3.03 (3.00)
Durir.g a reactor startup, verification of proper overlap between startup and Log Safety Nuclear Instrumentation Channels is required prior to deenergizing the high voltage to the startup channel.
- a. Over WHAT range of each instrument should the overlap occur? (1.0)
- b. WHY is the high voltage to the startup channel deenergized? (1.0)
- c. 1. HOW many detectors are used for each startup channel? (0.5)
- 2. WHAT type of detector is used for startup channels? (0.5)'
~
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
- 3. INSTRUMENTS AND CONTROLS PAGE 11 QUESTION 3.04 (2.00) .
Two (2) plant protection system trips, High Local Power Density and Low DNBR, are generated by the Core Protection Calculators.
- a. LIST five (5) inputs to the core protection calculator which are used to generate these trips. (1.5)
- b. WHAT control action takes place when the pre-trip setpoints are reached? (0.5)
QUESTION 3.05 (3.00)
Assume the plant is operating at 75% power with CEDMCS in AS mode, Loop 1A Tc RTD shorts out to ground, and Tave selector is on AVG.
- a. LIST three (3) of the four (4) indications that should occur in the control room. (1.5)
- b. WHAT should happen to the PZR level setpoint? (0.5)
- c. WHAT operator actions should follow? (1.0)
QUESTION 3.06 (2.50)
EXPLAIN the difference between the "AS" and "MS" modes of "ontrol in the CEDMCS. INCLUDE any applicable interlocks. (2.5)
QUESTION 3.07 (1.00) s WHEN will the Plant Monitoring computer select the rods to be dropped by the Reactor Power Cutback System when a cutback condition exists? (1.0) l l
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
i I
- 3. INSTRUMENTS AND CONTROLS PAGE 12 QUESTION 3.08 (1.00) .
With a High Level Override condition, the economizer valve: (1.0)
(a.) cannot be operated (b.) can be operated from the Master control Station (c.) can be operated by placing the Feedwater Control System in
~
operate (d.) can be operated from the Economizer Control Station QUESTION 3.09 (2.00)
IDENTIFY four (4) systems that interface with the Containment Spray System. (2.0)
QUESTION 3.10 (1.00)
Given Pressurizer Pressure Control System:
I The manual / automatic station for the pressurizer heaters and spray is in the MANUAL mode. To increase the heat from the proportional heaters you must. (1.0)
. Increase the controller output
. adjust pressure setpoint higher
. decrease the controller output .
. adjust pressure setpoint lower .
l QUESTION 3.11 (1.00)
ANSWER TRUE or FALSE.
- a. SIAS will start the operable containment spray pumps and spray chemical addition pumps. -
(0.5)
- b. An operator can manually override initiation of containment spray pumps and the spray chemical addition pump from the control room. (0.5)
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
. . . , - . ,,__,,n., . , . , , ,, , _ . .-,_ , , _. ...-., . . - , , . . , - .--,.
- 3. INSTRUMENTS AND CONTROLS PAGE 13 QUESTION 3.12 (1.00) .
During degassification of the RCS in the gas stripper section, the dissolved hydrogen used for oxygen scavinging is removed.
EXPLAIN how the hydrogen concentration in the RCS remains relatively constant in view of this continuous removal of hydrogen by the gas stripper. (1.0)
QUESTION 3.13 (1.00)
CHOOSE the closest correct value for the efficiency of the gas stripper from the percentage values given below. (1.0) 1
. 90.0
. 95.0
. 99.0
. 99.9 -
1 l
l l
l
(***** END OF CATEGORY 03 *****)
. _ _ _ _ . -_ _ _ - . = . . . _
- 4. PROCEDURES - NORMAL, ABNORMAL. EMERGENCY AND PAGE 14 RADIOLOGICAL CONTROL
~
QUESTION 4.01 (2.25)
You need to operate a component that is tagged. STATE how operation of the component is affected by each of the following tags:
- a. blue men at work tag 0.75
- b. red danger tag 0.75
- c. yellow caution tag 0.75 QUESTION 4.02 (2.25)
Under WHAT three (3) conditions would the absence of an R0 in the 2
controls area be acceptable? (2.25)
QUESTION 4.03 (2.00)
ANSWER TRUE or FALSE to the following.
- a. When more than one dosimeter is used, they should be worn in separate locations between the thigh and head. (0.5)
- b. Personnel dostmetry can be removed from the restricted area if reentry is made within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. (0.5)
- c. If a portal monitor alarms, the individual should pass through l the portal monitor once more. (0.5)
- d. A High Radiation Area is defined as any area with a general area dose rate greater than 100 mR/hr but less than 1000 mR/hr. (0.5)
(
- QUESTION 4.04 (1.50)
Assume you have received 200 mR so far this week. You are now needed to perform work in an area with a dose rate of 200 mR/hr.
- a. Without special approvals, HOW long can you remain in the area before you expect to exceed your weekly allowable dose? (1.0)
- b. WHOSE approval is needed to exceed the normal weekly limit? (0.5)
'l
(***** CATEGORY 04CONTINUEDONNEXTPAGE*****)
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 15 RADIOLOGICAL CONTROL QUESTION 4.05 (1.50)
If RCS boron concentration is changed by more than 50 ppm, WHAT two (2) actions are taken to adjust the pressurizer pressure control system? (1.5)
QUESTION 4.06 (1.00)
Assume xenon oscillations are occurring during full power operation.
CEA insertion should begin when the xenon oscillation is: (1.0)
( . at the positive peak
. halfway down the backside of the positive peak
. at the negative peak
. halfway up the frontside of the positive peak QUESTION 4.07 (2.00)
A CAUTION in 41AO-12ZO2, Load Rejection, states that if steaming to atmosphere, ensure that hotwell level and CST level are acceptable.
j a. WHY would hotwell level decrease in this situation? (0.5)
- b. WHY is low hotwell level of concern? (1.0)
- c. WHY would CST level decrease in this situation? (0.5)
QUESTION 4.08 (2.00) ,
Appendix C to the Functional Recovery Procedure describes a sequence of events to be taken if a reactor trip is not accomplished. '
- a. WHY is it necessary to deenergize L10 and LO3 for 5 seconds, as opposed to only 1 or 2 seconds? (1.0) l b. WHY is subsequent reenergization of these load centers i required? (1.0) l .
l l (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
l o l
- 4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 16 RADIOLOGICAL CONTROL QUESTION 4.09 (2.00) 41AO-1ZZ01, Emergency Boration lists potential emergency boration flowpaths.
- a. LIST the three (3) preferred emergency boration flowpaths using the refueling water tank. (1.5)
- b. LIST one (1) emergency boration flowpath using the spent i fuel pool. (0.5)
QUESTION 4.10 (3.00)
Palo Verde Technical Specifications, Section 2.1, Safety Limits, lists three (3) safety limits, two (2) applicable to the reactor core and one (1) applicable to the RCS. LIST the three (3) i parameters of concern AND the associated safety limit. (3.0) l l QUESTION 4.11 (3.00)
For EACH of the following off-normal situations, STATE the time allowed to take required action, if any, per Technical Specifications.
1 Assume the reactor is at power (Mode 1). ANSWER:
- 1. for 15 minutes or less
- 2. for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
- 3. for greater than I hour or no action required . !
- a. one inoperable pressurizer code safety valve ,
(0.5)
- b. a safety injection tank isolation valve (UV-644) is closed (0.5) i
- c. refueling water tank temperature of 50 deg F (0.5)
- d. a containment isolation valve is stuck open (0.5) l
- e. primary containment internal pressure is 2 psig ,
(0.5)
- f. the turbine auxiliary feedwater pump is under repair, and both motor driven auxiliary feedwater pumps discharge isolation .
valves (V013 and V025) are closed. (0.5)
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 17 RADIOLOGICAL CONTROL l
l
~
QUESTION 4.12 (2.50) ;
ANSWER TRUE or FALSE.
Operation may continue for EACH of the following according to ;
Technical Specification 3.4.5.2, Operational Leakage. '
- a. 0.7 gpm primary-to-secondary leak in one steam generator (0.5) l
- b. 6 gpm leakage into a safety injection tank (0.5) l
- c. 2 gpm unidentified leakage (0.5) 1
- d. 3 gpm body-to-bonnet leak the pressurizer spray valve (0.5) l
- e. 8 gpm leakage through a pump seal (0.5) l
.~
- END OF CATEGORY 04 *****)
(********(******ENDOFEXAMINATION***************)
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, $ PAGE 18 THERMODYNAMICS HEAT TRANSFER AND FLUID FLOW ANSWERS -- PALO VERDE 1&2 -86/09/23-DEFFERDING, L.
ANSWER 1.01 (3.00)
- a. (Since the amount of reactivity inserted is one-half of the total amount needed to reach criticality,) the count rate should double, or about 60 cps. [+1.0]
OR For rho e K-1 (valid approximation)
CR1/CR0 = (1-K0)/(1-K1))/(1-(rhol
= (1-(rhoo + 1) + 1))
= rho 0/rhol
=2 CR1 = 2*CR0 = 60 cps (actual is 58.85 cps) [+1.0]
- b. For rho = K-1 CR1/CR0 = rho 0/rhol OR rho 0 = rho 1*CR1/CR0 [+0.5]
Since the change in reactivity was 1% delta K/K (one-half af the total dilution): [+0.5]
rhoo = (rho 0 + 1%)*CR1/CR0
= (rhoo + 0.01)*50/30
= 1.67 rho 0 + 0.0167
-0.67 rho 0 = 0.0167 rhoo = -0.025
= -2.5%
Therefore, the original ECP should have determined that criticality the core. [would +1.0] be achieved if 2.5% delta K/K was added to REFERENCE
- 1. PVNGS: Reactor Theory, pp. 8.9 through 8.13.
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 19 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- PALO VERDE 1&2 -86/09/23-DEFFERDING, L.
ANSWER 1.02 (2.25)
- a. supercritical - power would turn, the reactor would be on positive SUR.
- b. critical - power would level off at (or slightly below) the power level that existed at the time rod motion was stopped.
- c. subcritical - power would continue to decrease (although at a slowerrate).
[+0.75] each REFERENCE
- 1. PVNGS: Reactor Theory, p. 11.7.
ANSWER 1.03 (2.25)
- a. Resonance escape probability [+0.3] decreases [+0.2] and the thermal utilization factor [+0.3] increases [+0.2] with increasing temperature. ,
- b. Undermoderated [+0.5]. Ensures that MTC is negative [+0.75].
Note: Figure 1.1 is Figure 12-5.
REFERENCE
- 1. PVNGS: Reactor Theory, pp.12.3 and 12.4.
l
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 20 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- PALO VERDE 1&2 -
86/09/23-DEFFERDING, L.
i 4
i ANSWER 1.04 (3.50)
- a. From Figure 1.2:
defect at 100% power: -1500 pcm defect at 85% power: -1300 pcm Therefore, a reactivity change of 200 pcm is required. [+1.0]
i From Figure 1.3:
rod worth at 130 inches withdrawn: 20 i required rod worth: 220 pcm (200 + 20)pcm yals:
, This corresponds to a rod height of 48 inches withdrawn. [+1.0] 40 - 1.
"~
- b. A larger reactivity insertion would be required [+0.5]. This is because the increase in power defect at EOC outweighs the increase in rod worth at EOC [+1.0].
i Note: Figures 1.2 and 1.3 are Figures 13-13 and 14-6, respectively, i REFERENCE
- 1. PVNGS: Reactor Theory, pp.13.10 and 14.6 ANSWER 1.05 (3.00)
- a. False
- b. True
- c. True
- d. True
- e. True
- f. False
! [+0.5] each j REFERENCE
- 1. PVNGS:' Reactor Theory, pp. 15.4, 16.3-5, and Figure 16.8.
- 1 f
.,w, - - - -
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 21 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- PALO VERDE 1&2 -86/09/23-DEFFERDING, L.
ANSWER 1.06 (1.00)
(d.) [+1.0]
- ( 3 '/ ( f ,)
REFERENCE
- 1. PVNGS: Reactor Theory, pp.17.3 through 17.5.
ANSWER 1.07 (1.00)
(a.) [+1.0]
~
REFEREfCE... ..
- 1. PVNGS:
Heat and Thermo, 'pp.54 ard 44., , ~ . -.... _
ANSWER 1.08 (2.50)
Q-dot = m-dot delta-h [+0.5]
m-dot = 12 x 10**6 lbm/hr h-steam = h-g at 1075 psia = 1190 BTU /lba ;+0.5; h-feed = h-f at 426 deg F = 403.5 BTU /lba + 0.5 Therefore Q-dot = 12 x 10**6 (1190-403.5) BTU /hr
= 9.4 x 10**9 BTU /hr [+0.3]
Converting to Mw Q-dot = 9.4 x 10**9 BTU /hr x (1 Mw/3.41 x 10**6 BTU /hr)
= 2770 Mw [+0.3]
% of full power Q-dot = 2770/3800 = 73% [+0.4]
REFERENCE
- 1. PVNGS: Heat and Thermo, p. 8-25.
.,,,wn, e., --n -,- - , ,
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 22 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- PALO VERDE 1&2 -86/09/23-DEFFERDING, L.
ANSWER 1.09 (1.00)
(c.) [+1.0]
REFERENCE
- 1. PVNGS: Heat and Thermo, pp. 12-5 and R12-2.
ANSWER 1.10 (2.00)
Assumptions:
Q-dot (nc) = 7% Q-dot (100%)
delta T (nc) = 50% delta T (100%) [+0.5'[ 0.5] (25-80%I,+(2-8%
acceptable) acceptable) n.:
Calculation:
Q-dot (nc)/Q-dot (100%) = m-dot C delta-T (nc)/m-dot C delta-T (100%)
[+0.5]
Therefore m-dot (nc) = [Q-dot (nc/100%)] x [ delta-T (100's/nc)]
m-dot (100%)
= 0.07 x (1/0.5) m-dot (100%) ~
= 14% m-dot (100%) [+0.5] (2M to b acceptable depending upon assumptions)
REFERENCE
- 1. PVNGS: Heat Transfer, Fluid Flow and Thermodynamic Review,
- p. 5.
ANSWER 1.11 (2.50)
' ~
- a. 5 feet [+1.0]
- b. 1. pump damage due to cavitation [+0.5]
- 2. tank collapse due to low internal pressure [+0.5]
- c. O feet (all the water). [+0.5] (Theaddedpressureof 14.7 psia removed.) (34 ft of water) would allow all of the water to be --
REFERENCE
- 1. PVNGS: Fluids, p. 6-8.
^
l 1
l i 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 23 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- PALO VERDE 1&2 -86/09/23-DEFFERDING, L.
~
i ANSWER 1.12 (1.00)
(d.) [+1.0 REFERENCE 1
l 1. PVNGS: Fluids, p 6-9, 6-33, and 6-34.
^
l l
1 - an
. . .=-.
'l
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 24 ANSWERS -- PALO VERDE 1&2 -86/09/23-DEFFERDING, L.
I ANSWER 2.01 (2.00)
- a. False (50 deg F/ min)
- b. True
- c. False (42, 42, 16)
- d. True
[+0.5] each REFERENCE l~ 1. PVNGS: Training articles NS-2A-24, NS-1A-8&9, Operating Procedure 410P-1CH03.
ANSWER 2.02 (2.00)
- a. Two air reservoirs are provided as a backup control air supply for positioning the hydraulic fluid control valves [+1.0].
- b. 1. (hydraulicpump) [+0.5]
- 2. accumulator [+0.5]
REFERENCE
- 1. PVNGS: Training article PGS-1A-13.
- 2. PVNGS: Training article PGS-1A-22 and 23.
2 ANSWER 2.03 (2.50) 1 l a. 1. interlocks prevent opening the valves on the suction lines until RCS pressure is less than 400 psia [+0.5]
l 'I ' >
i 2. interlock will close suction valves if pressure increases
! above 435 psig [+0.5]
l 3. suction line relief valves (SIA-PSV179, SIB PSV-189) [+0.5]
i
- b. By adjusting the amount of flow through the SDCHE 0.5 hile bypassing the remainder around the heat exchanger ;++0.5;w(total
- flow remains constant (5000 gpm)).
I REFERENCE
- 1. PVNGS: Training article NS-3A 49-55.
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 25 ANSWERS -- PALO VERDE 1&2 -86/09/23-DEFFERDING, L.
i ANSWER 2.04 (3.00)
- a. 1. red indicating light in control room ./ ; ,
- 2. red indicating light on switchgear
- 3. SEAS in control room <
c ,.
7/1
- 4. alarm permissive circuitry energized
( . t r ...c. . .
[+0.5] each yi,; ,. ge. a.., .
9, e Lo r e f> Q !# y , t4 ., j . . /
- b. Pump will stop as the loads are shed [+0.5] and will restart sequentially when DG picks up the load [+0.5].
REFERENCE
- 1. PVNGS: Training article PGS-11-20.
t l ANSWER 2.05 (3.00)
! a. 1. A-batteries
- 2. A-battery charger [+0.5] [+0.5]
i 3. A-C backup battery charger [+0.5]
- b. 1. reactor trip switchgear
- 2. valves (such as aux feed, steam supply) i
- 3. inverters
- 4. controls Any three (3) [+0.5] each, +1.5 maximum.
REFERENCE
- 1. PVNGS: Training article PGS-150-10 and Figure PGS-15D-2.
ANSWER 2.06 (2.40)
- ' ~~
l a. 1600 - 1700 pf,1g ; - /-
- b. 150 psig ,,:s t 2 c. -:- .>
- c. 610 psig op(- <
- d. 815 gpm ; g- 70 ,; , _ . . , ,,- _
~
,.n _c ;,
- e. 4300 gpm v .c r -
- f. 14000 g ]c ,; 3 , g it(, ,c e _
[+0.4] each ~'- ' ',
,- l1 "
. o
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 26 ANSWERS -- PALO VERDE 1&2 -86/09/23-DEFFERDING, L.
REFERENCE
- 1. PVNGS: Training article PGS NS-3A-5, 8, 10, 13, and 15.
ANSWER 2.07 (2.00) a.
b.
.*I7f yf ; - 15
- c. 65
- d. 10
[+0.5] each REFERENCE
- 1. PVNGS: Training article PGS-10A-9,11,21.
ANSWER 2.08 (1.50)
- a. 1. essential water chiller
- 2. shutdown heat exchanger
- b. fuel pool coolers (a^v
[+0.5] each REFERENCE
- 1. PVNGS: Training article PGS-88-6.
ANSWER 2.09 (2.00)
- a. Two concentric 0-ring seals [+0.5] are used. The inner diameter of rings is slotted to allow RCS pressure to inflate or expand the ring (s) exposed to the RCS [+0.5].
- b. (Theleakpathhasasolenoidvalvewhichisclosedduring operation.) Pressure builds and a pressure transmitter sends the signal to a pressure indicator and a pressure switch hi alarm. [+0.5] , .
- c. Since this leakage is collected and known it would be classified #
itm4h as identified leakage.V [+0.5] .g%
~.-A1/.,
- i.
,)
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 27 ANSWERS -- PALO VERDE 1&2 -86/09/23-DEFFERDING, L.
REFERENCE
- 1. PVNGS: Training article NS-1C-18.
ANSWER 2.10 (2.10)
- a. 1.
keepsspray when pipes, valves,[and is used +0.5] nozzles hot to reduce thermal shock
- 2. maintain water chemistry uniform in PZR and RCS [+0.5]
- b. cold leglA loop ;+0.3; cold leg lB loop + 0.3
- c. auxiliary spray [+0.5]
, ANSWER 2.11 (2.00)
- a. no auto functions (alarm only) [+0.5]
- b. actuates the control room essential filtration units (limits flow through EFUs [+0.5]
- c. automatic initiation of filtration of condenser vacuum pump / gland seal exhaust (thru-filter mode) [+0.5]
- d. Isolates the normal ventilation and activates the essential ventilation system (FBEVAS) [+0.5] ,,,i < / ; T /- # ' -
)
REFERENCE I **
- 1. PVNGS: Radiation Monitoring System, pp. SQ 3, 4, 5, and 7.
ANSWER 2.12 (.50)
False [+0.5]
REFERENCE
- 1. PVNGS: Training article NS-2A-64.
T e w ., , , , - - - - - - - , - - - - - - - - - . , - , - . , - . . , , , , , , - . , _ - - . , . , , , - - - - . - - - - - - . , - , - , - - . . - - . - - . . - , , - - - , - - , - . - , - -
- 3. INSTRUMENTS AND CONTROLS PAGE 28 ANSWERS -- PALO VERDE 1&2 -86/09/23-DEFFERDING, L.
ANSWER 3.01 (3.50)
- a. 1. containment isolation signal (CIAS)
- 2. main steam isolation signal (MSIS)
- 3. safety injection signal (SIAS)
- b. auxiliary feedwater actuation signal -
- c. none
- d. 1. CIAS
- 2. SIAS
[+0.5] each REFERENCE
- 1. PVNGS: Training article NS-7A-44.
ANSWER 3.02 (3.00)
- a. 1. prevents CEAs from being automatically withdrawn (AWP)
[+0.5]
- 2. CEDMCS [+0.5]
- 3. prevent raising reactor power since there is excess NSSS energy [+0.5]
- b. 1.
prevents back reactor CEAs power from [+0.5 being] (AWP automatically withdrawn signal also produced) __.and cut
- 2. CEDMCS and RPCS [+0.5]
- 3. allows a quick reloading of the turbine / generator if the loss of load is due to a temporary fault. Reduces reactor power until SBCS can handle excess steam [+0.5]
REFERENCE ;
- 1. PVNGS: Training article NS-98 and NS-9C.
- 3. INSTRUMENTS AND CONTROLS PAGE 29 ANSWERS -- PALO VERDE 1&2 -86/09/23-DEFFERDING, L.
ANSWER 3.03 (3.00)
~
- a. 20-2000 cps = 2 x 10**-8 to 2 x 10**-6 percent power [+1.0]
- b. to prolong the life of the startup channel [+1.0]
- c. 1. 4 per channel (dual section each 2 BF 3) [+0.5]
- 2. BF(3) [+0.5]
REFERENCE
- 1. PVNGS: Training article NS4-4 and Figure NS4-3.
ANSWER 3.04 (2.00)
- a. 1. ex-core fission chambers
- 2. CEA positions (m R $775)
- 3. Th
- 4. Tc
- 5. PRZR pressure
- 6. RCP speed fi n.
- 7. CEA penalty factors Any five (5) [+0.3] each, +1.5 maximum.
- i
REFERENCE
- 1. PVNGS: Training article NS-6B 11 and 20-23.
I e
E
. 3. INSTRUMENTS AND CONTROLS PAGE 30 ANSWERS -- PALO VERDE 1&2 -86/09/23-DEFFERDING, L.
ANSWER 3.05 (3.00)
J. 'c #3 ~> ' r + ' 0'
- a. 1. T-ave-T-ref HILo alarm h
7 (.
~'
channel deviation LED on RRSTD ,
,,. e , w e f-- A w ,
Any three (3) [+0.5] each, +1.5 maximum
- b. PZR setpoint will be at or approaching min setpoint [+0.5]
- c. 1. take manual control of CEDMCS and PLCS ,[+0.5]
- 2. switch Tave selector to unaffected channel and return systems to auto /nonnal [+0.5]
REFERENCE
- 1. PVNGS: Training article NS-9C-16.
ANSWER 3.06 (2.50)
AS mode RRS controls CEA motion duration [+0.25], direction [+0.25] and rate [+0.25] (high or low)
CWP [+0.25], AMI [+0.25] and AWP [+0.25] interlocks are effective MS mode Operator controls CEA motion demand and direction. Low rate is not available. [+0.5]
AMI and AWP interlocks are not available.) CWP is available.
+0.5]
REFERENCE
- 1. PVNGS: Training article NS-9E-9 and 12.
ANSWER 3.07 (1.00) when no preselection ~1s made [+0.5] and when reactor power is -
greater than 75% [+0.5]
l
I
- 3. INSTRUMENTS AND CONTROLS PAGE 31 ANSWERS -- PALO VERDE 1&2 -86 / 09 /23-DEFFERDING, L.
REFERENCE
- 1. PVNGS: Training article NS-90-3.
ANSWER 3.08 (1.00)
(d.) [+1.0]
/
REFERENCE
- 1. PVNGS: Training article NS-9A, Logic Diagram FWCS #1.
ANSWER 3.09 (2.00) e
- 1. todine removal system
- 3. SDCS (SDC heat exchangers)
- 5. ECWS (SDC heat exchanger heat sink)
- 6. essential chilled water system (cooling for air cooling in ESF pump rooms) l Any four (4) [+0.5] each, +2.0 maximum.
REFERENCE
- 1. PVNGS: Training article, Vol. II, pp. NS-38-14 and 15.
l ANSWER 3.10 (1.00)
(c.) [+1.0]
REFERENCE
- 1. PVNGS: Training article NS-9G-5.
I
, . . , - - , . . - , - ,,---.--.-a - . - - - . -- - - - - - - - - - - - , - - - - - , , , , , , - - -,,1-.~._.--,---
W -
4
- 3. INSTRUMENTS AND CONTROLS PAGE 32 ANSWERS -- PALO VERDE 1&2 -86/09/23-DEFFERDING, L.
f ANSWER 3.11 (1.00)
- a. False [+0.5
- b. True [+0.5)) l
! REFERENCE
- 1. PVNGS: Training article NS-38-10 and 19.
- 2. PVNGS: Training article NS-3P-18 and 19.
ANSWER 3.12 (1.00)
Water returned to the VCT is sprayed with the hydrogen atmosphere of the VCT to redissolve hydrogen. ., [+1.0]
- REFERENCE i
- 1. PVNGS: Training article, Vol. II, p. NS-28-33.
ANSWER 3.13 (1.00)
(d.) [+1.0]
l REFERENCE
! 1. PVNGS: Training article, Vol. II, p. NS-28-33.
- l 1
- i t
1 j
-- _s._.'- - . . , , . . - _ . _ _ . , . . . _ _ - , . - , . _ _ . ..t - , . _ _m,. _ , _ . - . .
~ _. . - . _. - - . - _ _ _ _ _ _ .. - . - .
- 4. PROCEDURES - NORMAL ABNORMAL. EMERGENCY AND PAGE 33 RADIOLOGICAL CONTROL ANSWERS -- PALO VERDE 1&2 -86/09/23-DEFFERDING, L.
1 ANSWER 4.01 (2.25)
- a. the direction of the individual Operation to whom the is tag allowed only by+0.75]
is issued. I
- b. Operation is not allowed. [+0.75]
- c. Operation is allowed as long as the specific information is observed. [+0.75]
REFERENCE
- 1. PVNGS: Station Tagging and Clearance, p. 6A.
ANSWER 4.02 (2.25)
- 1. R0 is relieved by an SRO
- 2. R0 is in another area in the control room during an emergency (to verify receipt of an annunciator or initiate corrective action)
- 3. R0 is directed by the Shift Supervisor to evacuate (to the Remote Shutdown Panel)
[+0.75] each REFERENCE
- 1. PVNGS: Conduct of Shift Operations, 5.3.1, p.14.
ANSWER 4.03 (2.00)
- a. False
- b. False
- c. True #
- d. False . Tr"! + 4' " , '
[+0.5] each -
,j REFERENCE
- 1. PVNGS: Radiation Exposure and Access Control, pp. 12, 13a, 18 thrc:gh 20.
l J
. . l
l l
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 34 i RADIOLOGICAL CONTROL )
ANSWERS -- PALO VERDE 1&2 -86/09/23-DEFFERDING, L. <
ANSWER 4.04 (1.50)
- a. Weekly limit = 300 mR [+0.5]
Remaining dose = 300 mR - 200 mR = 100 mR Allowable time = 100 mR/(200 mR/hr) = 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> [+0.5]
l l b. Radiation Protection Support Supervisor [+0.5]
REFERENCE l 1. PVNGS: Radiation Exposure and Access Control, p. 26.
ANSWER 4.05 (1.50)
- 1. override and energize all pressurizer backup heaters [+0.75]
l
- 2. decrease the setpoint on the pressurizer pressure controller (RCN-PIC-100) to approximately 2220 psia [+0.75]
.w.;..
REFERENCE - - i-
- 1. PVNGS: 410P-12Z05, Power Operations, Appendix E.
l l
ANSWER 4.06 (1.00)
(b.) [+1.0]
REFERENCE
- 1. PVNGS: 410P-12Z05, Power Operation, Appendix B, p. 5.1 and Figure 2. -
ANSWER 4.07 (2.00)
- a. loss of secondary mass to atmosphere [+0.5]
- b. adequate NPSH for condensate pumps [+1.0] (to stay above condensate pump trip setpoint will be acceptable for +0.8) -
- c. makeup to hotwell demand [+0.5]
I i
)
4 . - . , , __ _ _ , - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _-
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 35 RADIOLOGICAL CONTROL ANSWERS -- PALO VERDE 1&2 -86/09/23-DEFFERDING, L.
REFERENCE
, 1. PVNGS: 41AO-1ZZ02, Load Rejection, p. 9.
ANSWER 4.08 (2.00)
- a. to allow the MG sets time to drop off [+1.0]
- b. to allow operation of other equipment on those load centers
[+1.0]
REFERENCE
- 1. PVNGS: 41RO-1ZZ10, p. 31 of 157.
ANSWER 4.09 (2.00)
- a. 1. gravity feed through CH-HV-536 to charging pumps
[+0.5] each
- b. 1. spent fuel pool through CH-V164, Boric Acid filter bypass, and CH-UV-514 to charging pumps
- 2. spent fuel pool gravity feed through CH-HV-536 to charging pumps
[+0.5] for either answer REFERENCE
- 1. PVNGS: 41AO-1ZZ01, Emergency Boration, p. 3.
AilSWER 4.10 (3.00)
- 1. DNBR [+0.5] greater than or equal to 1.231 [+0.5]
- 2. peak linear heat, rate [+0.5] less than or equal to 21 kW/ft
[+0.5]
- 3. RCS pressure [+0.5] less than or equal to 2750 psia [+0.5]
i l
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 36 RADIOLOGICAL CONTROL i l
ANSWERS -- PALO VERDE 1&2 -86/09/23-DEFFERDING, L.
REFERENCE
- 1. PVNGS: Technical Specifications, " Safety Limits," Section 2.1.
ANSWER 4.11 (3.00) i a.
- b. 1 (innediately 1 (15 minutes))
- c. 2 _
l
- d. 2- "
- e. 3 -0.3 to 2.5 psig is allowed)
- f. 1 innediately)
- s c c.3 a w w REFERENCE
- 1. PVNGS: Unit 1 Technical Specifications, 3.4.2.2, 3.5.1, 3.5.4,
.L6rirl; 3.6.1.4, 3.7.1.2.
'(,- ..
ANSWER 4.12 (2.50)
- a. False (greater than 720 gallons / day on one steam generator)
- b. False
- c. False
- d. False
- e. True
[+0.5] each REFERENCE
- 1. PVNGS: Technical Specifications, " Operational Leakage,"
Dsction 3.4.5.2.
e EQUATION FORMULA AND PARAMETER SHEET Where mg = m2 (density)t(velocity)3(area)g = (density)2(velocity)2(area)2 2
KE=y PE = mgh PE + KE +P Y l l 1 1 = PE 2+KE 2+P Y22 where V = specific volume
/ P = pressure 0" p(Tout -Tin) Q = UA (T,y,-Tsta) Q = m(hy-h2 ) ,
P = Po10(SUR)(t) p , p ,t/T SUR = 26.06 T= " = - #}
o T p p A,ff
____s______________________________________________________________________
delta K = (K,ff-1) CRg (1-K,ffg) = CR2 (1-Keff2) CR = S/(1-K,ff)
(1-K,ff1) (1-K,ff) x 100%
SDM = A'If = 0.08 sec_y t
M = (1-Keff2) E eff l
i -
decay constant = _1n (2) " 0.693 A1
- Ao' (decay constant)x(t)
_t t 1/2 1/2 Water Parameters Miscellaneous Conversions 1 gallon = 8.345 lbs 1 Curie = 3.7 x 10 10 dps 1 gallon = 3.78 liters 1 kg = 2.21 lbs 3
1 ft = 7.48 gallons I hp = 2.54 x 103Btu /hr 3 6 t Density =62.4lbg/ft 1 MW = 3.41 x 10 Btu /hr l
Density = 1 gm/cm 1 Btu = 778 ft-lbf Heat of Vaporization = 970 Btu /lbm Degrees F = (1.8 x Degrees C) + 32 Heat of Fusion = 144 Btu /lbm 1 inch = 2.54 centimeters 2 1 Atm = 14.7 psia = 29.9 in Hg g = 32.174 ft-lbm/lbf-sec ,
1 ft H 2O = 0.4335 lbf/in.2 ,,
?
___..._______________.____________________________________=.____=____======_
. - - - . - - - - - . - . , , - , -- - . - ~ - - - - - -,--7 ,, --r,-- ----y-,--r-.----n--w , , . - - , - - w,w- --,