ML20234D144
| ML20234D144 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 06/09/1987 |
| From: | Elin J, Morrill P, Royack M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20234D114 | List: |
| References | |
| 50-528-OL-87-02, 50-528-OL-87-2, NUDOCS 8707070113 | |
| Download: ML20234D144 (155) | |
Text
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Examination Report No. 50-528/0L-87-02 Decket Nos. 50-528/529/530 Licensee: Arizona Public Service Company P. O. Box 2166 Phoenix, Arizona 85036 _
Facility Name: Palo Verde 1, 2 and 3 Examination at: Wintersburg, Arizona Examination conducted: April 8 - April 16, 1987
- #' # # R ####'
Chief Examiner: 9 P. Morrill, Chief Examiner DaTE 61gned
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Examiner: # $/ '
M. Royack7 Examina9 7
Date Signed Appro d @ -
dh Summary: /0 Elin, Sfction Chief Datepigned Examinations were conducted from April 8 through April 16, 1987. The Written Exam was administered on April 10, 1987 to ten SR0 and ten R0 candidates.
Operating exams were conducted on April 8, 9, 14, 15 and 16, 1987 to ten SR0 and ten R0 candidates. Nine SR0 candidates and ten R0 candidates passed the written examinations. Nine SR0 candidates and ten R0 candidates passed the operating examinations.
1 8707070113 870617 I PDR ADOCK 05000528 l V PDR
DETAILS
- 1. Examiners
- P. Morrill, RV (Chief Examiner)
R. Pate, RV (Branch Chief)
- M. Royack, RV L. Defferding, PNL R. Gruel, PNL
- 2. ANNP Persons Contacted
- J. Driscoll, Assistant Vice President
- R. Henry, Licensing Training
- R. Badsgard, Nuclear Engineering
- R. Gouge, Operations
- W. F. Fernow, Training Manager
- J. R. Bynum, Plant Manager
- J. Stanely, Simulator Coordinator
- R. Baron, Compliance Supervisor
- D. Craig, Training Supervisor, License Operators W. Rudolph, Lead Instructor
- Attended Exit Meeting on April 17, 1987.
- 3. Examination Review At the conclusion of the written examinations the examiners gave the facility staff copies of the two examination keys for review.
The facility staff reviewed the keys and discussed their comments with the examiners. The facility comments and the resolution of the facility comments are included as Attachment 1 to this report. Based on the comments ar/ the stated resolutions, the key was modified and the candidates' tests were graded.
During the review of the examinations the examiners observed that most of the f6cility comments were due to' incomplete or incorrect reference material sent by the facility to the NRC. The Chief Examiner told the facility personnel that this situation was unacceptable and could not be repeated.
The examiners copied several pages from the facility Training Articles to demonstrate the problem to the facility management. The following is a partial list of examples of inadequate reference material (which were given to facility personnel at the exit meeting).
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Condensate System, Figure PGS-9A-3 and 4 - Illegible drawing l Essential Spray Pond, Figure PGS-8A Illegible drawing Circulating Water System, Figure PGS-7A Illegible drawing i Reactor Coolant System, Figure NS-1A RCPs illegible j CVCS, Figure NS-A-5b - Illegible' drawing i Main Turbine, Figure PGS-3A Illegible drawing of control board .
i Fuel Pool Cooling, Figure AS-1D-4b - Illegible drawing Main Steam, Figure PGS-1A Illegible - ,
CRVAS actuated devices list, Table NS-7B-VIII..-4 Table incomplete ,
CREFAS actuated devices list, Table NS-7B-VII - Table incomplete i ESFAS alarms, page NS-7A Table incomplete .
ESFAS Annunciators and Setpoints, Table NS-7A-I - Table incomplete and j incorrect ESFAS Trip Values, Table NS-7A-III - Setpoints incorrect.
Additionally, the Chief Examiner pointed out that the P& ids sent to the NRC were very difficult to read (for example AS-9-3, Gaseous Radwaste) and did not represent all of the important plant systems. The Chief Examiner also stated that the selection of procedures sent to the NRC was less than he expected and frequently material was incomplete. For ,
example, one section of the Technical Specifications as well as pages 131 through 181 of Alarm Response Procedure 41AL-1RK6B were missing. In s the future he would expect to receive additional procedures and to review i the material received as soon as possible to determine its acceptability. j The Chief Examiner pointed out that in the future the problems could be j eliminated by having someone check the material sent to the NRC to verify j that it was legible, accurate, and complete. l l
t, . Operating Examinations !
Simulator and oral examinations were conducted on April 8, 9, 14, 15, )
and 16, 1987. During the simulator portion of the operating examination the examiners observed many anomalous events and several obvious simulator modeling errors. These problems resulted in longer than normal simulator and oral examinations to ensure fair evaluation of the candidate's perfonnance.
Some of the observed simulator anomalous events, shortcomings, and malfunctions are listed below (and were presented to the facility management at the exit meeting).
Boration flow meter failed for no apparent reason.
Dilution flow meter failed for no apparent reason.
Containment spray flow indicator indicated flow for over one minute while the spray valves were shut.
Charging flow indicated 94 GPM with one charging pump running.
An MSIV closed for no apparent reason.
Containment isolation valve UV-24 stayed open on a CIAS.
Condensate header pressure indicated 500 PSIG vice 100 PSIG.
Valve HV-612 (Nitrogen to SITS) could not be closed after opening.
About 50% of PAMS out of service.
GRID failed and required simulator operator to reset.
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Alarm typewriter failed several items. !
With sequencer inoperable the 480 V buses could not be rereergized and j reloaded.
T (cold) recorder above turbine controls failed for no apparent ,
reason.
CEA positions unavailable from CPCs. 3 QSPDS not modeled.
RVLMS not modeled.
Subcooling meter not modeled. j Cannot fail channel B middle detector.
Start-up channel could not be failed low.
Cannot model middle of core life.
Cannot model end of core life.
Reactivity changes due to boration and/or dilution are not properly modeled (reactivity change too small in both cases).
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Following a feed line break in containment, with the steam generator indicating 0 level, 0 pressure, and feed isolated the containment pressure continued to rise and " spike" thereby initiating a CSAS.
Following a steam line break outside containment, the faulted generator level decreased, increased to normal, then decreased and emptied.
The Chief Ext. miner told the facility personnel that this many problems associated with the simulator.was unacceptable. This-is especially significant since the same examiner has observed various simulator problems of the same magnitude over the last two years with no general improvement. Individual problems appear to be fixed, however, the general problems remain and new problems appear to the occurring.
- 5. Exit Meeting On April 17, 1987 an exit meeting was conducted with licensee representative listed in paragraph 2 of this report.
During this meeting, the Chief Examiner described the inadequacies of the written material provided by the facility to the NRC (see paragraph 3 of this repert). He stated that the NRC examiners minimized the impact of poor reference material by extra research into the reference material and by review of the key with facility personnel.
The Chief Examiner then went on to describe the observed problems related to the simulator (see paragraph 4 of this report) and pointed out that two years ago the simulator had also had significant problems which almost caused the examiners to cancel the simulator portion of the operating tests. Although individual problems appear to have been solved,' there are new problems which are at least as significant as before. Apparently little real progress has been made in correcting the simulator's underlying problems. He stated that the examiners had lengthened the oral and simulator portions of the examinations to ensure that the candidates evaluations were not unduely affected by the simulator _ problems.
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l Based on these problems the Chief Examiner posed two questions:
(1) How does the facility avoid negative training with the out of date 1 and incomplete reference material? j (2) How does the facility avoid nesative training with simulator anomalous behavior and modeling problems? {
In response to the first question, the facility management stated that the Training Articles sent to the NRC were not used for retraining. The Training l Articles were last updated in 1982 and are currently undergoing extensive i revision for use in the INPO Accreditation Program. Above 65% of them will be updated to be end of June 1987. The facility management went on to state that handouts and lesson plans are kept current and that the handouts are given to the trainees. In response to questions by the Chief Examiner, the facility management stated that the Training Articles had been specifically requested by the NRC. The Chief Examiner and Senior Resident Inspector both pointed out that even if this material had been requested it was the facilities l responsibility to identify that the material was incorrect and that more up-to- i date material existed, j l
Regarding the simulator, the facility management stated that they l realized there were significant deficiencies and that they had a plan to !
rorrect these problems. The facility has taken interim steps and is !
working hard on the reactor and steam / generator trainsient responses. The !
facility staff felt that the anomalous events were due to the inability of the simulator operator to input several commands oat the same item.
The facility training staff explained that when they get to a point beyond the simulator capabilities they conduct a walkthrough and explain the differences between the simulator and the actual units. Li kewise, incorrect simulator responses are discussed during the critique. The final reactor model will be in the simulator by the next NRC examination at PVNGS, although the reactor coolant system and steam generator models may not be completed. The facility management stated they were firmly behind the simulator and that it will be financed and improved.
The Chief Examiner also described the apparent generic weaknesses observed during the oral examinations. A few candidates demonstrated poor knowledge in the following areas:
(1) containment airlock mechanical interlocks; (2) spent fuel handling equipment and refueling procedures; (3) location of "non-essential" auxiliary feed pump suction at the consendate storage tank; (4) the types of radiation detection devices used for process and area monitors; (5) purpose of the SRO review of Radiation Exposure Permits; (6) SR0 control of keys; and (7)shutdownversusshutdownmargin.
7-ATTACHMENT (1)
FACILITY COMMENTS / RESOLUTIONS 1.02 Question Comment:
Part "a" asks how each of 2 parameters will vary but does not specify that a quantitative response is required. The key includes both a qualitative and quantitative response.
(The quantitative response should not be. required.) The relationship between speed and head and speed and power should be enough for full credit.
1.02 Question Resolution:
Facility comment accepted.
The quantitative portion of question a.1 and a.2 will be considered as optional i and be parenthesize, indicating that the quantitative rer.ponse is optional.
1.08.a Question Congnent:
Solution CR y C -Ky ) = CR2(1~K) 2 81 = -6.4% A K/K = - 6400 pcm)
=
.06400 A K/K K = 1 = 1 = .93985 I
1-fy 1-( .064)
(1-K2 ) = 5_ (1 .93985) 25 l
= .2 (.06015) !
1
= .01203
-K = .01203-1 K
2
= .98747
- 2* K 2-1 = .01218 2
= -1.22% A K/K Rod Worth =8 final O-/ initial
=
-1.22 - (-6.4)
=
5.18% A K/K a
. H 1
OR CR y 81 = CR 2 #2 p2 =5/25(-6.4) q
= 1.28%'A K/K Rod Worth = 5.12% A K/K Both answers are close to key value of 5% A K/K and should be accepted.
1.08.b' Question Comment:
This value for shutdown margin violates PVNGS Technical
' Specifications (T.S.) and candidates may be confused.
They may infer that " Shutdown riargin" refers to the actual amount shut down and give the answer as -1.22%
A K/K.
The' numbers used have complicated the problem. Even the initialconditiongnet = -6.4% A K/K violates T.S.
1.08 Question Resolution (a) Facility comment accepted.
- 1. The answer key will be corrected to show the worth of the shutdown banks to be 5.2% A K/K.
- 2. A rod worth of 5.12 A K/K will be added to the answer key as ,
acceptable. l (b) The " Shutdown Margin" is the amount the reactor is or would be sub critical, at current conditions, with the most reactive rod stuck out.
The answer key will be modified to accept the following as a correct I response:
The reactor is shutdown by -6.4% A K/K and will not meet the Technical Specification limit of -5% A K/K shutdown margin due to stuck rod =
3.5% A K/K.
2.05 Question Comment:
The two electrical systems are 125VDC and 120VAC power.
See elementary 01-E-NNA-001.
2.05 Question Resolution Facility comment accepted, 125 VAC should read 120 VAC.
Answer Key change not required.
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2.07.c Question Comment:
Valves (UV-65 and UV-145) are manually opened by the operator, the valves do not open automatically on a LOP.
See elementary Cl-E-EWB-003.
2.07 Question Resolution
.(c) Corrected reference material, supplied after the examination, shows that UV-65 and UV-145 do not operate automatically. '
The answer key will be modified to accept..."no automatic action"...
as a correct response.
2.09.b Question Comment:
Question asks for three systems that must be in operation.
The systems required are:
- 1. Plant Cooling Water.
- 2. Gland Steam seals and drains.
The answer key.had separated the second of these into two systems. This may confuse candidates trying to name three systems when only two exist.
2.09 Question Pesolution Facility comment accepted.
Full credit will be given for..." Gland Steam Seal System" and "(Gland Seal) i Drain System" or " Gland Steam Seal and Drain System"...
2.10.b Question - Comment:
Turbine driven Auxiliary Feedwater Pump has two feed regulation valves.
HV-32 powered from PKA-M41 HV-33 powered from PKC-M43
Reference:
PGS-ISD page 10 PGS-11 page 16 2.10 Question Resolution (b) Facility comment accepted.
125 VDC PKA-M43 will be added to the answer key. The new point distribution will be:
M-41 (0.66 pts.) since PMA-M41 supplies power to four of the six valves; and M-43 (0.33 pts.) since PKA-M43 supplies power to two of the six' valves.
2.13 Question Comment:
(a)Thenumberthreechargingpumpissetupwithtwo power supplies PGA-L35 and PGB-L36, the use of the term " normal" and " alternate" are just a matter of convenience used in the electrical line-up verification check list of the procedure. It doesn'.t signify that the charging pump will be powered preferentially l from L36. It will be powered from the bus that allows I l
the greatest flexibility of operation.
l (b) We don't require the cperator to memorize electrical .
I power supplies when the information is easily obtained i from the control board labels for the equipment. An l acceptable answer would be: .
i 480VAC load centers from either Train "A" or Train "B" essential busses.
2.13 Question Resolution j (a) Facility comment accepted.
Since either bus PGA-L36 or PGA-L35 could be the normal or alternate power supply the words " normal" and " alternate" will be deleted from the answer key.
(b) Facility comment not accepted. i The question does ask for specific busses which an operator familiar with the CVCS control board or the electrical distribution system should be familiar with. Therefore no change will be made to the answer key.
, 2.14 Question Comments:
1 The reference material used for the above answer is not ;
all inclusive and a copy of Procedure 410P-1CH04 (Gas I Stripper) has been supplied to the examiner. The procedure in steps 5.2.6 states that the following systems are capable of supplying the requirements of the GS:
- 1. Nuclear Cooling Water
- 2. Auxiliary Steam
- 3. Instrr: ment and Service Air
- 4. Gaseous Radwaste
- 5. 480V Non Class IE Power An appropriate answer should be any two of the five support systems for full credit.
i i
2.14 Question Resolution ,
l Facility comments not accepted.
The requested change is based on a procedure not previously provided to the examiners. Revising the answer from two out of two required systems to two out of five systems substantially dilutes the original questions' difficulty. l The additional systems listed, in the facility comments, include 480 Volt non-class IE power and Instrument and Service air both of which are needed for the operation of the facility. Similarly the Gaseous Radwaste system is ,
obviously a necessary system. 1 Therefore, the answer key will not be revised to reficct the facility comments.
3.02.c Question Comment: j RCP speed, rather than RCS flow, may be the response.
RCP speed is the parameter measured to determine flow. l l
3.02 Question Resolution I (c) Facility comment accepted, i Reactor coolant pump speed is used to determine reactor coolant flow, therefore, the key will be revised to accept " Reactor Coolant Pump Speed" as a parameter.
3.04 Question Comment:
AWP is only generated on positive deviation of T
ave Tref > +6'F, not 16 F.
Reference:
NS-90-9, 26 and 27 3.04 Question Resolution (b) Facility comment accepted.
The answer key will be revised to show:
(Tave Tref) 2 +6 F 3.06.b Question Comment:
Variable overpower trip setpoint:
Rate: 10.6% min Ceiling: 110%
Band: 9.8%
Reference:
T.S. Table 2.2-1
)
3.06 Question Resolution (b) Facility comment accepted.
i Original reference material incorrect. The answer key will be corrected j to read.. 110% or 9.8% w/10.6%/ min.
3.07.a Question Connent: i The correct answer should include " Low RCS temperature following reector trip".
3.07.b Some plant conditions still affect SBCS:
- 1. Condenser interlock
- 2. Emergency off 3.07 Question Resolution (a) Facility comment accepted.
l The answer key will be changed to read... Low RCS temperature following j a reactor trip and loss of feed pump.
I (b) Facility comment accepted, however, the comment does not change the required answer.
The answer key will be changed to include..."unless a loss of vacuum occurs or an emergency off signal is generated"...inside the parentheses.
3.09 Question Connent:
Answer key implies there is a difference between backup and proportional heaters; both function off of the same bistable and both trip on high pressure and low level.
3.09 Question Resolution (a) Facility comment accepted.
The answer key will be revised to include, in parentheses,..."one bistable is utilized".
4.03 Question Comment:
The answer is appropriate if the operator used 41A0-1ZZ27, Shutdown Outside Control Room. However, 4 if the operator took a more conservative approach and used 41A0-1ZZ44, assuming that the bomb threat was real and could result in fire and or smoke in the control room, they would take a different set of immediate actions. These actions are:
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- 1. Manually trip the reactor.
- 2. If conditions permit; Manually trip all reactor coolant pumps.
- 3. If conditions permit; Manually trip the turbine.
- 4. If conditions permit; Trip the main feedwater j pumps and the "A" essential auxiliary feedwater pump.
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- 5. . If conditions permit; Manua'.ly initiate a MSIS.
i The above set of actions should also be an acceptable answer. i l
This question is valid and within the scope of Training l Department objectives. Deletion of this question may impact this section of the test such that a candidate may not receive an adequate score, although he knew information pertinent the subject matter. In other i words based on Training Department priorities, this )
question requires knowledge that is as important as other questions in this section.
4.03 Question Resolution I (a) Facility comments not accepted.
Procedure 41A0-1ZZ44 was not supplied in the original reference I material. There is a note in procedure 41A0-1ZZ27 which requires the operator to enter 41A0-1ZZ44 if the control room is being evacuated due to fire or smoke. The use of the " bomb threat" was chosen to avoid having to enter 41A0-12Z44. The question specifically states that the control room is being evacuated due .
to a " bomb threat" and makes no mention of fire or smoke, therefore, the key will remain unchanged.
l 4.07 Question During the grading of the examination, the examiners noted that the responses to part b of the question were not consistent with the answer key. The facility was requested to supply the latest revision of procedure 40AC-9ZZ15, Station Tagging and Clearance Revision 1.
Based on the updated information from 40AC-92Z15 Revision I, the ,
answer key was revised to accept " System Operations Supervisor", '
which is equivalent to S0C Supervisor, and Transmission Control Center Supervisor.
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4.09 Question Comment:
T.S. 3.4.5.2 lists RCS leakage as follows:
- b. 1 GPM unidentified leakage
- c. 1 GPM total primary to secondary through all S/Gs and 720 GPD through any one S/G
- d. 10 GPM identified leakage from RCS
- e. 1 GPM at RCS pressure of.2250 1 20 psia from any valve listed in T.S. Table 3.4-1 The answer of "10.2 gpm total leakage exceeds 10 gpm total leakage limit" is not an LCO. The actual LC0 says 10 gpm identified leakage from RCS.
0.6 gpm leak in one S/G is not an LCO. LC0 of 720 GPD will be exceeded if leak remains for 1 day.
The only LC0 exceeded without some assumption is "No pressure boundary leakage", by 0.2 GPM through the socket weld.
4.09 Question Resolution Facility comments regarding identified leakage vs. total leakage is accepted. The answer key will be revised to delete the "10.2 GPM total leakage exceeds 10 GPM total leakage limit".
Facility comments regarding the 0.6 GPM steam generator leakage vs.
720 GPD Technical Specification LCO limit is accepted. The answer key will be revised to give full credit for a response that provides the following information "720 GPD is the limit and/or 0.6, if maintained, will exceed the 720 GPD limit". ,
5.03 Question Comment:
5.03(F) In regards to the term " Xenon Precluded" the keyed answer assumes that the only precluded condition _j is a reactor restart during " Peak Xenon" conditions. '
To solicit this response the question should have used the term " Xenon Precluded Startup". Other transients may result in the Xenon Response
" Precluding" normal control of the reactor. For example, recent communication between ANPP Reactor Engineering and Operations has identified the following:
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Near the end of corelife a power reduction to 20%
will create a xenon transient of sufficient rate that the ability of CVCS to adjust RCS Boron concentration will be exceeded. Thus, desired RCS 'l temperatures will not be inaintainable.
The grader should accept any answer which identifies a condition where the xenon transient exceeds the {
ability of the control system to compensate for rate '
I or magnitude of xenon insertion.
5.03 Question Resolution Facility comment accepted.
Based on the new material provided by the facility the answer key will be revised to include the words ..."and the rate of change of Xenon precludes continued operation in the normal operating band of parameters".
5.05 Question Comment:
Problem does not provide suction pressure. If a suction pressure is assumed, full credit should be given for a correct answer since pump delta pressure follows pump ~
law, not discharge pressure.
Example: Assume 5 psig suction pressure.
delta P at 1800 RPM = 45 psi delta P at 1200 PRM = 45 psi * (.666)2
= 20 psi Discharge Pressure at 1200 RPM = 20 + 5 psig
= 95 psig 5.05 Question Resolution Facility comment has been reviewed.
The answer key will not be revised. The pumps take suction from the bottom of the spray pond, therefore, suction pressure does not change. However, the grader will evaluate the candidates' response with respect to the facility comment.
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~5.09.a Question ' Comment:
Question is vague. The value of power coefficient does not depend on RRS unless a power change (change in steam ;
demand) occurs.without compensating reactivity actions by the operator. If T . differs from the programmed .
value for a given powefVTevel, then the power coefficient will change. The question does not: lead the candidate to assume that T is going.to differ.
ave Example: Even if RRS is in manual, ~a power change can be H
[and usually is] accompli.shed by a change-in RCS boron concentration and the power coefficient will not. change.
~5.09 Question Resolution :;
Facility comment not' accepted.
.The introduction to the question states that Power Coefficient = Moderator Temperature Coefficient + Fuel Temperature Coefficient. Part a'of'the-question does sta.te that'there is N0 rod. motion, this fact was also' pointed-out to all SR0 candidates in a clarification statement, by an examiner, j during the examination. The candidates should know that:
- 1. T will be constant with no rod motion, ave
- 2. Power Coefficient = Moderator Temperature coefficient + Fuel Temperature Coefficient, and
- 3. Moderator Temperature Coefficient is negative at power.
Therefore, a change to the answer key will not be made.
5.11 Question Comment:
Answer key subtracts losses'to' ambient. These should-be added. Correct solution is:
3692 MWt Q) 17 MWt RCP)
+- 7 MWt ambientlosses)
T682 MWt
)
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V o 3 3
5.11 Question Resolution Facility comment accepted.
Answer key will be corrected to show a +7 MWt (ambient losses), which will revise final answer to:
3692 - 17 + 7 = 3682 MWt (+/- 4) 1 6.02.b Question Comment:
The question is related to the Gaseous Radwaste System and N is added to the Gaseous Radwaste System whenhighcobcentrationsofoxygenaredetected.
Gases added to the VCT are H, to control oxygen concentrations-in the RCS coblant and N9 to purge H Due tbgasduringdegasingoperationsintheRCS, the confusion of part "b", request deletion of part "b" and that full credit be given for part "a" of the question.
6.02 Question Resolution Facility comment not accepted.
Procedure 410P-1CH01 states that nitrogen is used to dilute the gas in the VCT whenever oxygen is above 2%. The use of hydrogen as a dilutant is not described in the referenced procedure. Step 7.2.10 directs the operator to Section 6.0, " Establish a Nitrogen Blanket," on the VCT, whenever oxygen exceeds 2%. Additionally, in step 7.3 the operator is cautioned NOT to add hydrogen to the VCT when oxygen concentration is greater than 2%.
Therefore the answer key will not be changed.
6.03.a Question Comment:
i CSAS signal will start the ch'illers in the same way SIAS does. This should be an acceptable answer. J
Reference:
drawing J104-66 (or electrical 342-0100).
6.03 Question Resolution Facility comment accepted.
Based on additional reference material the facility comment is accepted, response a.1 will be revised to include "or CSAS".
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..._._._.m._.
6.04.a Question Comment:
Different terminology should be accepted: 1
- 1. " Select switch" may be'" Mode select switch".
- 3. May mention that proper group must be selected, l
6.04 Question Resolution Facility comment accepted.
The switches on the CEDMCS are labled differently than described in the i reference material. Based on the examiners. observation of the CEDMCS control console the key will be changed as the facility has requested.
I 6.05.a.1 Question Comment:
" Peak linear heat rate" may also be worded " linear power density".
6.05 Question Resolution Facility comment accepted.
" Linear power density" is equivalent to " peak linear heat rate". The key will be changed to reflect this equivalence.
6.06.b.1 Question Comment: j AWP is only generated on positive deviation of T
ave
-T ref + 6 F.
Reference:
01-J-SFE-053.
6.06 Question Resolution Facility comment accepted.
b.1 Facility comment correct; reference rnaterial alarms are not consistent with the AWP signal Reference Drawing 01-J-SFE-053, Revision 0, which was supplied to the examiners after the examination.
The answer key will be revised to read:
(T ave -Tref) 3 + 6'F.
6.08.d Question Comment:
Correct answer is MSIS, but candidate may say that during the transient, AFAS-1 may occur, but would be eventually locked out due to the S/G 2 being greater than 185 psid from S/G 1.
6.08 Question Resolution Facility comment accepted.
The answer key will be revised for 6.08.d to give full credit for MSIS.
AFAS will NOT be accepted unless candidate explains that it is locked out due to DELTA pressure.
6.12 Question Comment:
Answer key does not list generator trips, only Diesel trips. Question asks for Diesel-Generator trips.
All. trips of Diesel and generator should be accepted as correct responses.
6.12 Question Resolution Facility comment not accepted.
The facility comment is correct. There are a total of sixteen trip signals listed on page 61 of the reference material. The intention of the question was to test the candidates knowledge of the trip signals generated by the diesel engire itself. Because of the ambiguity of the l question and the large numbe; of potentially correct responses (16),
this question is not a good discriminator. Therefore this question will be dropped from this examination.
6.13.b Question Comment:
A " condenser interlock or emergency off" will inhibit the manual permissive.
Reference:
01-J-SFE-058.
6.13.c Comment:
Loss of feedwater pump (not pumps).
I 6.13 Question Resolution 6.13.b Facility comment accepted.
The facility comment is correct, however, it has no effect on the answer. 1 The phrase...,"unless a loss of vacuum occurs or an emergency off signal is generated." will be added to the response in parentheses. )
6.13.c Facility comment accepted. !
1 The facility comment is correct. The answer key will be revised to..." pump".
I 6.15 Question Comment:
Question is misleading. During normal operation air removal is not going "thru-filter", moisture would not normally be removed. The answer is correct as to the reason this is done.
6.15 Question Resolution Facility comment is accepted as an alternative answer to part "a".
The answer key for part "a" will be revised to accept...."or normal air removal does not go through the charcoal filters therefore no moisture removal is necessary", full credit will be given for this response.
6.16.b Question Comment:
i A low level in the EW surge tank will also isolate EW from NC.
Reference:
01-E-EWB-003.
6.16 Question Resolution i
Facility comment accepted. l Based on additional facility reference material an additional response will be added to 6.16.b, low surge tank level (L-91). Each response will l be worth 0.25 pts. each. !
7.04 Question Comment:
Confusion may)have by dose (mrem insteadoccurred from(mrem of dose rate areas/hr being)given
1 7.04 Question Resolution Facility comment not accepted.
The question provides appropriate information for the candidate to evaluate and correctly respond to the question. Therefore, the answer key will not be revised.
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7.10.b Question Comment:
i Diesels are cooled by Essential Spray Pond Water.
7.10 Question Resolution Facility comment accepted.
Essential spray pond water will be added to the answer key.
7.12.b Question Comment: '
On flow chart, page 2 of Appendix E, seven items are ,
evaluated, not six.
- 1. Th stable or decreasing
- 2. Tc stable or decreasing
- 3. Core AT less than full power AT of 57 F
- 4. CET's and Th trending consistently
- 5. RCS subcooled margin 28 F
- 6. RV outlet plenum full
- 7. RCS voiding suspected We do not verify Tc at S/G T (#2 of answer key). #5 of original answer key not eit$btly correct since there are two 0-100% ranges on RVLMS. The plenum must indicate 100%.
7.12 Question Resolution Facility consent accepted.
The answer key will be revised to accept the "seven" items, as listed in the facility comment.
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8.05 Question Comment:
Question asks "What is the minimum Licensed Crew l composition in accordance with T.S. 6.3.1 for operation in Mode 5". The T.S. li::ts an A.0. as part of the required crew. No points should be taken off if the A.0. is included in the answer. l 8.05 Question Resolution Facility comment reviewed. The comment overlooks the fact that the question 1 asks for the " minimum licensed crew" not the " minimum crew". This was done )
to avoid ambiguity. No points will be taken off if the candidates include !
the "A0" (Auxiliary Operator) and state that the A0 is unlicensed, i i
8.08 Question Comment:
Question 8.08 of the Senior Reactor Operator License Examination states, "You have entered a Site Area Emergency, with minimum Technical Specifications Shift manning."
(b) Who is the person responsible for making notifications over the unit page and the satellite technical support center (STSC) or control room radio circuit of transients and ESFAS actuations.
Exam Key Answer:
- b. The Operation Technician (OPS. TECH)
The procedure reference used by the examiner does state their answer on page 20a PCN05, however, the questions stated that the shift was at minimum levels of staffing, by the Technical Specifications. There is no mention of the OPS. TECH position in Table 6.2-1 of Tech. Specs.
Due to the way the question was asked an accepted answer could be any of the following: Shift Supervisor, Control Room Supervisor / ASS, Reactor Operator, Nuclear Operator or STA. !
8.08 Question Resolution Facility comment reviewed.
The facility comment is correct, since there is no single clear correct answer to the question part "b" of the question will be deleted from the examination.
8.13 Question Comment:
During the grading of the examination, the examiners noted that the responses to'part a.1 were not consistent with the answer key. The facility was requested to supply the latest revision of procedure 40AC-9ZZ15, Station Tagging and Clearance, Revision 1. Based on the updated information from 40AC-9ZZ15, Revision 1, the answer key was revised to accept " Systems Operations Supervisor which is equivalent to 50C Supervisor and Transmission Control Center Supervisor.
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U.S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION
. Facility: Palo Verde NGS Reactor Type: PWR -
CE80 Date Administered: April 7,1987 Examiner: Michael J. Royack Candidate: ----- m -
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 indi-cated 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 Cet:: gory Value Total Score Value Category J
s T1
. 25, t'.r 5. Theory of Nuclear
\
i
- Power Plant Operation, l Fluids, and Therr- l 22,g 23.20 dynamics 20 20 i
- 6. Plant Systems Design, '
Control, and
~ Instrumentation 1577 25 6 _
- 7. Procedures - Normal, 4 Abnormal. Emergency, and Radiological 24,5 25,2G
- -+5- 8. Administrative Pro-cedures, Conditions, and Limitations
-MG- . Key Totals "
Final Grade All work done on this examinatfort is my own, I have neither given nor received aid.
KEi--------
Candidate's Signature s.
j 1 NR'C RULES AND GUIDELINES FOR LICENSE EXAMINATIONS
[l During the administration of this examination the following rules apply-
- 1. Cheatin
~' ~"'""''''''''''"'"~on the examination means an automatic denial of your appl l!
L, p.
- 2. 'Restroom trips are to be Ifmited 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. ,
- 1
- 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.
, , l 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~ tie last answer sheet. .
.- 1.
Number each answer as to category and number, for example,1.4, 6.3.
- 10. Skip at least three Ifnes 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 4 . question and can be used as a guide for the depth of answer. required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the. question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE -
QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
- 16. the If parts of the'only.
examiner examination are not clear as to intent, ask questions of i
- 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 been completed.the examination. This must be done after the examination has '
( .
.. ..'4
'g Dh $
q
1 4
.e
- 18. When you complete your examination, you shall:
- a. Assemble your examination as follows:
(1) Exam questions on top.
(2) Exam aids - figures, t' ables, 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 anrwer 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 stili in progress, your license may be denied or revoked.
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EQUATION SHEET f = ma v = s/t '.
w ". 28
~ '
2 s=vt+ 'a t Cycle efficiency = Net' Work (out)
E = mC ~ y a = (v'g i v,)/t' KE = mv vg A = AN t
= v, + at , A = A,e
, PE = mgh a = .,0/t . A = In 2/tg = 0.693/tg W = vaP AE = 931Am ' tg(eff) = (t, )(ts) '
, (t +t) I
- q r a,C aT ..
k e M Ah
.. . P - r . r eo - I.x.:. -
4 " "'T Pur = Wg m
-6 = RAC%s.Tir,,) : I = I o= "*~
I.1 9 to / M
-x i P-p 10 SUR(t) " _
TyL 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) {
, SUR = 2'6 /A*ff p) CR
,p = S/(1 - K,ggx)
~
T - D*/p ) + [(s- 'p)/A,ffp ] 1( ~K*ff}l = CR 2(l ~ Keff)2 T = t*/ (p _ y; M = 1/(1 - K,gg) = CR y/CR0 T = (3 - p)/ Aeffp !
M = (1 - K,gg)0!II - Eeff)1
- " ( eff-1)/K,gg = AK,gg/K,gg , ,
~
eff eff p=
[1*/TK,:gg..] + [H/(1 + Aeff)] T t* = 1 x 10 seconds *
{._.- -
P = I(V/(3 x 10 0) geff = 0.1 seconds I = No ,
Idgy=Id22 -
~
WATER PARAMETERS Idg =1d2 I gal. = 8.345 lbm 2 i gal. = 3.7B liters R/hr = (0.5 CE)/d (meters)
I ft = 7.48 gal.
R/hr = 6 CE/d (feet)
MISCELLANEOUS CONVERSIONS Density = 62.4 lbm/ft 3 _
1 Curie = 3.7 x 1010dps
- Density = 1 gm/cm I kg = 2.21 lbm Heat of va;orization = 970 Etu/lbm I hp = 2.54 x 103 BTU /hr Heat of fusica = 144 Beu/lbm 1 Hw = 3.41 x 100 Beu/hr 1 Atm = 14,7 psi = 29.9 in. I'g. ,
1 Btu = 778 ft-lbf
. 1 ft. H 2O = 0.433~, ibf/in 2 1 inch = 2.54 cm F = 9/5 C + 32 C = 5/9 ( F. 32) .
I Properties of Saturated Steam and Saturofed Water
- Absolute Pressure Vacuum Temper. Heat of Lbs. per Inches Latent Heat Total Heat Specific Volume
~
inches ature the, Sq.In. of Hg of Hg Liquid of ,
Evaporation of Steam y
P' %, p.
t na Water i s ons. _ _ s,o m,. seu m,. co. ii. m, w I co. Steamti. u, m_
0.00&7 0.02 29.90 32.018 ;
0.10 0.0003 1075.5 1075.5 0.016022 0.20 29.72 35.023 3.026 3302.4 0.15 1073.8 1076.8 0.016020 2945.5 0.31 29.61 45.453 13.498 0.20 0.41 1067.9 1081.4 0.016020 ~ 2004.7 29.51 53.160 21.217 1053.5 0.25 1984.7 0.016025 1526.3 0.51 29.41 59.323 27.382 0.30 0.61 1060.1 1087.4 0.016032 1235.5 29.31 64.484 32.541 1057.1 0.35 0.71 1089.7 0.016040 1039.7 29.21 68.939 36.992 1054.6 109).6 0.40 0.81 29.11 72.869 0.016048 898.6 0.45 40.917 1052.4 1093.3 0.92 29.00 76.387 0.016056 792.1
- 44.430 1050.5 1094.9 0.016063 0.50 1.02 28.90 708.8 79.566 47.623 1048.b 1096.3
.0.60
~
1.22 28.70 85.218 0.016071 t>41.5 0.70 53.245 1045.5 1098.7 0.016085 1.43 28.49 90.09 58.10 540.1 0.80 1042.7 1100.8 0.016099 466.94 1.63 28.29 94.38 62.39 0.90 -
1.83 1040.3 1102.6 0.016112 411.69 28.09 98.24 66.24 1038.1 1.0 1104.3 0.016124 368.43 2.04 27.88 101.74 69.73 1.2 2.44 1036.1 1105.8 0.016136 333.60 27.48 107.91 75.90 1032.6 1.4 2.85 27.07 1108.5 0.016158 280.95 113.26 81.23 1029.5 1110 7
" . 1.6 3.26 26.66 117.98 0.016178 243.02 .
1.8 85.95 1026.8 1112.7 0.016196
~ 3.66 26.26 122.22 90.18 214.33 1024.3 1114.5 0.016213 191.85 2.0 4.07 25.65 126.07 2.2 94.03 1022.1 1116.2 0.016230 4.48 25.44 129.61 97.57 173.76 l 2.4 4.89 1020.1 1117.6 0.016245 158.87 25.03 132.88 100.84 1018.2 l
2.6 5.29 24.63 1119.0 , 0.016260 146.40 135.93 103.88 1016.4 1120.3 t
2.8 7.70 24.22 0.016274 135.80 138.78 106.73 1014.7 1121.5 !
3.0 6.11 0.016287 126.67 23.61 141.47 109.42 l 3.5 7.13 1013.2 1122.6 0.016300 118.73 '
22.79 147.56 115.51 1009.6 4.0 8.14 21.78 152.96 1125.1 0.016331 102.74 4.5 120.92 1006.4 1127.3 0.016358 i
- 9.16 20.76 157.82 125.77 90.64 i 5.0 10.18 1003.5 1129.3 0.016384 83.03 19.74 161.24 130.20 5.5 1000.9 1131.1 0.016407 73.532
, 11.20 18.71 166.29 134.26 6.0 12.22 998.5 1132.7 0.016430 17.70 170.05 - 138.03 67.249 6.5 13.23 16.69 996.2 1134.2 0.016451 61.984 7.0 173.56 141.54 994.1 1135.6 14.25 15.67 176.84 0.016472 57.506 7.5 15.27 144.83 992.1 1136.9 14.65 179.93 0.016441 53.650 147.93 990.2 1138.2 8.0 16.29 13.63 0.016510 50.294 8.5 182.56 150.87 988.5 17.31 12.61 1139.3 0.016527 47.345 185.63 153.65 986.8 9.0 18.32 11.60 188.27
!!40.4 0.016545 44.733 '
9.5 19.34 156.30 985.1 1141.4 10.58 190.80 158.84 0.016561 42.402 10.0 20.36 983.6 1142.4 0.016577 9.56 193.21 161.26 982.1 1143.3 40.310 l'
!!.0 22.40 7.51 197.75 0.016592 38.420 t 12.0 24.43 165.82 979.3 1145.1 5.49 201 96 170.05 0.016622 35.142 ;
. 13.0 26.47 3.45 976.6 1146.7 0.016650 205.88 174.00 32.394 14.0 38.50 1.42. 974.2 1148.2 0.016676 30.057
{
209.56 177.71 971.9 !
- 1149.6
- 0.016702 28.043 1 i
b*
- Pressure Temper. Heat of Latent Heat Total Heat a
Lbs. per Sq. In. sture Specific Volume the of of Steam !
Absolute Gage t
Liquid Evaporation y .
P' P he Water omm r. s.o ns. n., n s. 3,o ns. co. e.. n, u 6 team 14.696 0.0 212.00 co. ,.. y, m.
15.0 160.17 970.3 1150.5 0.3 213.03 181.21 0.016719 26.799 16.0 2.3 969.7 1850.9 0.016726 216.32 184.52 26.290 17.0 2.3 219.44 967.6 1152.1 0.016749 187.66 965.6 24.750 18.0 3.3 222.41 190.66
!!53.2 0.016771 23.385 19.0 4.3 963.7 a 1154.3 0.016793 225.24 193.52 961.8 22.168 20.0 5.3 1155.3 0.016814 21.074 i
227.96. 196.27 .
21.0 6.3 9b0.1 115b.3 0.016834 230.57 198.90 20.087 22.0 7.3 958.4 1157,3 0.016854 233.07 201.44 956.7 19.190 23.0 8.3 235.49 1158.1 0.016873 18.373 24.0 203.88 955.1 1159.0 9.3 237.82 206.24 0.016891 17.624 25.0
'953.6 1159.8 0.016909 10.3 240.07 208.52 16.936 26.0 11.3 952.1 1160.6 0.016927 242.25 210.7 950.6 16.301 27.0 12.3 244.36 1161.4 0.016944 15.7138 212.9 949.2 1162.1 28.0 13.3 246.41 214.9 0.01b961 - 15.168a 29.0 14.3 947.9 1162.8 0.016977 248.40 217.0 946.5 14.6607 1163.5 0.016993 l 30.0 15,3 250.34 14.1869 218.9 945.2 lib 4.1 1
31.0 16.3 252.22 220.8 0.017009 11.7436 32.0 943.9 1164.8 17.3 254.05 222.7 942.7 0.017024 '13.3280'
- 33.0 18.3 255.84 1165.4 0.017039 {
224.5 941.5 1166.0 12.9376 34.0 19.3 257.58 226.3 0.017054 12.5700 940.3 1166 6 <
0.017069 12.2234 i i
r p . ;
I
j
.== a )
2 Properties of Saturated Steam and Saturofed Water-continued _
Pressure Temper.
, Lbs. per Sq. In.
- Heat of Latent Heat Total Heat Specific Volume atur, the of of Steam ]
--- A bsgute ,- -Gage .,.i--. ~~. I"I4"d b' P*
- I*" y 1 "r P o..,m r. he Wat
~'
eiuns. niuno. neunb. Steam 35.0 10.3 cu. ti. o.er
, ih. ca.ri.i~ ,e6.
259.29 228.0 939.1 {
36.0 21.3 260.95 A167.1 0.017083 11.8959 37.0 229.7 938.0 1167.7 l 22.3 262.58 o ~ 231.4 0.017097 11.5860
== 936.9
-~38.0 ~'23.3 -*
" 1168.2
' ~ 264.17 "233.0 0.017111 11.2923 39.0 24.3 935.8 1168.8 0.017124 '
'265.72 234.6 934.7 11.0136 40.0 25.3 1169.3 0.017138 10.7487
.267.25 236.1 933.6
< 41.0 g26.3 . 1169.8 0.017151 10.4965 J.237.7 42.0 - -27.3- - . 268.74 1 270.21 ~ . '239.2 * -. 1 932.6 931.5 I170.2 0.017164 10.2563 (3.0 18.3 271.65 240.6 1170.7 0.017177 10.0272 44.0 29.3 930.5 1171.2 0.017189
- 273.06 242.1 929.5 9.8083 45.0 30.3 1171.6 0.017202 9.5991 v.46,0 274.44 243.5 928.6
.31.3 d275.80 1172.0 0.017214 9.3988 4720 -
. . u244.9 ./ M27.6 -
32:3- v.277.14
- PS ;246 2 - -"7,M26.6'-"I*' Tl172.'9T Jll72.5 . . N- ~.0.017226 9.2070 "48.0 33.3
. 278.45 0.017238 949.0 i 34.3 ? $ 279.74 '
i 247.6 I
- m,248.9 -
.,2 925.7 . i . . .:J 173.3 ". '.
, 0.017250 *.e5 9.0231 8.8465
, y 924.8 c50.0 45.3 e 281.02 -
1 c -1173.71.1 0.017262 *. . $ 8.6770 51.0 v250.2 <
- 923.9 -
1174.1 36.3 282.27 0.017274 8.5140 52.0 - 251.5 . 923.0 1174.5 0.017285
--- 37.3 8.3571 53.0 38.3 -- - -283.50284.71
- " - ~252.8 ~~ ' 922.I~~ ~ 1174.9 254.0 0.017296 8.2061 54.0 39.3 921.2 1175.2 0.017307 h" )
285.90 255.2 920.4 8.0606 55.0 40.3 1175.6 0.017319 .-
7.9203 - j 287.08 256.4 ar. * * $6.0 41.3 288.24 257.6 919.5 1175.9 0.017329 7.7850 q 57.0 918.7 42.3 289.38 258.8 917.8 1176.3 ~ 0.017340 7.6543 $
58.0 43.3 290.50 1176:6 0.017351 7.5283 59.0 259.9 917.0 1177.0 44.3 =
60.0 45.3 291.62 292.71 261.1 262.2 916.2 1177.3 0.017362 0.017372 7.4059 7.2879 )
61.0 46.3 915.4 1177.6 j 293.79 263.3 0.017363 7.1736 62.0 47.3 914.6 1177.9 0.017393 j
294.86 264.4 913.8 7.0630 63.0 48.3 295.91 1178.2 0.017403 6.9558 64.0 265.5 913.0 1178.6 49.3 296.95 266.6 0.017413 6.8519 65 0 912.3 1178.9 50.3 297.98 0.017423 6.7511 66.0 267.6 911.5 1179.1 51.3 298.99 268.7 0.017433 6.6533 67.0 52.3 910.8 1179.4 0.017443 299.99 269.7 910.0 6.5584 68.0 - 53.3 300.99 1179.7 0.017453 6.4662
- 69.0 270.7 909.3 1180.0 54.3 301.96 "271.7 0.017463 6.3767
{
70.0 908.5 .
1180.3 55.3 302.93 272.7 0.017472 6.2696 {
71.0 56.3 907.8 11b0.6 303.89 273.7 407.1 0.017482 6.2050 72.0 57.3 304.83 1180.8 0.017491 73.0 274.7 906.4 6.1226 58.3 305.77 1181.1 0.017501 74.0 275.7 905.7 6.0425 59.3 306.69 . ,, 1181.4 0.017510
- C276.6 905.0 .1181.6 5.9645 75.0 60.3 307.61 0.017519 5.8885 277.6 904.3 ;
u76.0 61.3 308.51 . 278.5
- 1181.9 0.017529 5.8144
- -- 77.0 - 62.3 - 309.41 . . 903.6 902.9
~ 1182.1 l 0.017538 5.7423 )
h8.0 79.0
~63.3 .310.29 i 279.4 *, ';**m
(. U280.3 . 9,902.3 ... . II82.4 : 1.'
e
- 1182.6
.l 0.017547
. a. . 5.6720 *
. l
- 64.3 M1.17 *281.3 0.017556 4 7. I 5.6034 -
80.0 J 901.6 1182.8 '
65.3 312.04 0.017565 5.5364 M 81.0 - . 282.1 900.9 1183.1 ]
I 82.0
-- 66.3 -- -312.90 - - 283.0 ' ' ' 900.3 ~ 0.017573 5.4711 67.3 313.75 283,9 ~1183.3 ~ ~~ " 0.017582 ~ .
.- 83.0 - 899.6 1183.5 5.4074
. 68.3 314.60 284.8 0.017591 5.3451 84.0 -
69.3 899.0 1883.8 315.43 285.7 898.3
- 0.017600 "g 5.2843
. J5.0 _ .- 70.3 . 1184.0 0.017608 " ,5.2249 316.26 286.5 897.7 j 86.0 71.3 317.08 287.4 lit 4.2 0.017617 5.1669 87.0 72.3 897.0 1184.4 317.89 288.2 896.4 0.017625 5.!!01 88.0 73.3 318.69 1184.6 89.0 289.0 8 95.8 -
1184.8 0.017634 ,,. 5.0546 74.3 319.49 289.9 0.017642 .5.0004 90.0 895.2 1185.0 75.3 320.28 0.017651 4.9473-91.0 290.7 894.6 1185.3 76.3 321.06 29t.5 0.017659 92.0 77.3 893.9 1185.5 4.6953 321.84 292.3 0.017b67 4.8445 91.0 78.3 893.3 1185.7 322.61 293.I 892.7 0.017675 4.7947 94.0 79.3 323.37 1I85.9 0.017634 293.9 892.1 1886.0 4.7459 95.0 40.3 J24.!J 0.017692 4.6982
%.0 294.7 691.5 libb.2 81.3 324.88 295.5 0.017700 4.6584 97.0 82.3 891.0 1186.4
'325.63 296.3 890.4 0.017708 4.6055 94.0 A3.3 326.36 1156.6 0.017716 49.0 297.0 BR9.8 1886.8 4.5606 A4.3 327.10 297.8 889.2 1887.0 0.0lifM 4.5166 100.0 85.3 J27.82 0.01773i 4.4734 101.0 298.5 sas.6 1187.2 86.3 32N.54 299.3 0.017;40 4.4}io 102.0 M7.3 888.1 1187.3 J29.26 300.0 8&7.5 0.01775 ' 4.3895 103.0 HN.3 329.97 1187.5 'O.01776
- 104.0 300.8 886.9 1187.7 4.3487 M9.3 330.67 301.5 0.01776 AB6.4 1187.9 4.3087 .
105.0 90.3 JJa.J7 0.01777 4.2695 106.0 J02.2 6s5.5 lib 5.0 91.3 332.0b 30J.0 0.017;g 4.2309-107.0 92.3 885.2 !!88.2 332.75 3J3.7 844.7 0.01779 4.1931 108.0 93.3 J3J.44 304.4 1188.4 0,0];79 toto BM4.1 1188.5 4.1560 94.3 334.11 305.1 883.6 0.01780 4.1195 1188.7 0.01781 4.0837
'l,
- roperties of Saturated Steam and Saturated Water-continued l
- Pressure Temper- Heat of Latent Heat i Total Heat Specific volume Lbs. per Sq. In. the of of Steam Absolute Gage sture II9" b
- F "I p t h, Water P' P cu.Steam 110.0 95.3 o,.m. r.
J34.79
- niu m,.
305.6 aiu nk 683.1 neunk 1185.9 cem* n. m, w l
U.01782 4.0484 111.0 96.3 335.46 306.5 882.5 1889.0 0.01782 4.0138 112.0 97.3 336.12 307.2 882.0 1189.2 0.01783 113.0 98.3 3.9798 336.78 307.9 881.4 1189.3 0.01784
- 3.9464 114.0 99.3 337.43 308.6 880.9 1189.5 0.01785 3.9136 !
115.0 100.3 338.08 309.3 boo.4 1189.6 0.01785 116.0 101.3 3.8813 338.73 309.9 879.9 1189.8 0.01786 3.8495 117.0 102.3 339.37 '310.6 879.3 1189.9 0.01787 118.0 103.3
- 3.8183 340.01 311.3 878.8 1190.1 0.01787 3.7875 119.0 104.3 340.64 311.9 878.3 1190.2 0.01788 3.7573 120.0 105.3 341.27 312.6 877.8 1190.4 0.01789 121.0 106.3 3.7275 i 341.89 313.2 877.3 1190.5 0.01790 3.6983 122.0 107.3 342.51 313.9 876.8 1190.7 0.01790 3.6695 123.0 108.3 343.13 314.5 876.3 1190.8 0.01791 3.6411 124.0 109.3 343.74 315.2 875.8 1190.9 0.01792 3.6132 125.0 110.3 344.35 315.8 875.3 1891.1 0.01792 3.5857 126.0 111.3 344.95 316.4 874.8 1191.2 0.01793 3.5586 127.0 112.3 345.55 317.1 874.3 1191.3 0.01794 3.5320 128.0 113.3 346.15 317.7 873.8 1191.5 0.01794 3.5057 129.0 114.3 346.74 318.3 873.3 1191.6 0.01795 3.4799
- ** 130.0 115.3 347.33 319.0 872.8 1191.7 131.0 0.01796 3.4544 116.3 347.92 319.6 872.3 1191.9 0.01797 132.0 117.3 348.50 3.4293 320.2 871.8 1192.0 0.01797 3.4046 133.0 118.3 349.08 323.8 871.3 1192.1 134.0 0.01798 3.3802 119.3 349.65 321.4 870.8 1192.2 0.01799 3.3562
- 135.0 120.1 350.23 322.0 670.4 1192.4
- 136.0 0.01799 3.3325 121.3 350.79 322.6 869.9 1192.5 137.0 122.3 0.01800 3.3091 351.36 323.2 869.4 1192.6 0.01801 138.0 123.3 351.92 3.2861 323.8 868.9 1192.7 0.01801 3.2634 139.0 124.3 352.48 324.4 868.5 1192.8 0.01802 3.2411
, 140.0 125.3 353.04 325.0 868.0 141.0 1193.0 0.01803 3.2190 !
126.3 353.59 325.5 867.5 1193.1 142.0 127.3 0.01803 3.1972 354.14 326.1 867.1 1193.2 0.01804 143.0 128.3 354.69 3.1757 326.7 866.6 1193.3 0.01805 3.1546 144.0 129.3 355.23 327.3 866.2 1193.4 0.01805 3.1337 I 145.0 130.3 355.77 327.8 863.7 146.0 131.3 356.31 328.4 1193.5 0.01606 3.!!30 865.2 1193.6 0.0!$06 3.0927 147.0 132.3 356.84 329.0 864.8 148.0 1193.8 0.01807 3.0726 )
133.3 357.38 329.5 864.3 1193.9 149.0 0.01808 3.0528 134.3 357.91 330.1* 863.9 1194.0 0.01808 3.0332 150.0 135.3 358.43 330.6 863.4 152.0 1194.1 0.01809 3.0139 !
137.3 359.48 331.8 862.5 1194.3 '
154.0 139.3 0.01810 2.9760 .
36D.51 332.8 861.6 1194.5 156.0 141.3 0.01812 2.9391 1
- 361.53 333.9 860.8 1194.7 0.01813
. 158.0 143.3 362.55 2.9031 1 335.0 - 859.9 1194.9 0.01814 2.8679 160.0 145.3 363.55 336.1 859.0 1195.1 -
0.01815 2.8336 162.0 147.3 364.54 337.1 858.2 1195.3 0.01817 2.8001
- 164.0 149.3 365.53 338.2 166.0 168.0 151.3 153.3 366.50 339.2 857.3 856.5 1195.5 1195.7 0.01818 0.01819 2.7674 2.7355 367.47 340.2 855.6 1195.8 0.01820 170.0 155.3 2.7043 368.42 341.2 854.8 1196.0 0.01821 172.0 157.3 369.37 1196,2 2.6738 342.2 853.9 0.01823 174.0 159.3 370.31 2.6440 l 343.2 853.1 1196.4 0.01824 176.0 161.3 371.24 1.6149 '
344.2 852.3 1196.5 0.01825 178.0 163.3 372.36 2.5864 345.2 851.5 1196.7 0.01826 2.5585 f 180.0 165.3 373.08 346.2 850.7 1896.9 0.01827 181.0 157.3 373.98 347.2 2.5312 !
184.0 849.9 1197.0 0.01828 2.5045 169.3 374.88 348,1 849,1 1197,2 {
186.0 171.3 0.01830 2.4783 375.77 349.1 848.3 1197.3 0.01831 188.0 173.3 37MS 350.0 2.4527 l 847.5 1197.5 0.01832 2.4276 ;
190.0 175.3 377.53 350.9 192.0 846.7 1897.6 0.01833 2.4030 177.3 378.40 351.9 845.9 1197.8 i
194.0 179.3 0.01834 2.3790 379.26 352.8 845.1 1197.9 l 196.0 181.3 0.01835 2.3554 380.12 353.7 844.4 1198.1 0.01836 2.3322 198.0 183.3 380.% 354.6 843.6 1198.2 0.01838 2.3095 l 200.0 165.3 351.60 355.5 642.8 205.0 1898.3 0.01839 2.28728 190.3 383.88 357.7 840.9 1
210.0 1198.7 0.01841 2.23349 I 195.3 385.91 359.9 839.1 1199.0 215.0 200.3 0.01844 2.18217 387.91 362.1 837.2 !!99.3 220.0 205.3 389.88 0.01847 2.13315 364.2 835.4 1199.6 0.01850 225.0 210.3 2.08629 391.80 366.2 833.6 1899.9 230.0 215.3 0.01852 2 04443 393.70 368.3 831.8 1200.1 '
?35.0 220.3 395.56 0 OltiS5 1.99846 3?0.3. 830.1 1200.4 0.011(57 240.0 225.3 397.39 372.3 828.4 1.95725 l 245.0 1200.6 0.01 A60 1.91769 230.3 399.19 374.2 826.6 i 1200.9 0.01863 1.87970 l 1
. l
- *r.
I Pressure Properties of Saturated Steam and Satumted Water-conduded e Lbs. per Sq. In. - Temper. Heat of Latent Heat Total Heat j Specific Volume sture the of of Steam Absolute Gage t Liquid Evaporation g y P, - P- om.. r. 'seu na. seu nd. smns. Water $ team 250.0 235.3 400.97 cu.re. u n n. cu. re. ros.
i 376.1
~
.255.0 240.3 .. 625.0 1201.1 !
402.72 378.0 0.01665 1.84317 260.0 .
245.3 823.3 1201.3 I
404.44 379.9 821.6 - . 0.01868 a 1.80802 265.0 250.3 406.13 1201.5
- 0.01870 270.0 381.7 820.0~ 1201.7 1.77418
" 255.3 407.80 383.6 0 01873 818.3 1201.9 1.74157 275.0 260.3 409.45 0.01875 1.71013 280.0 385.4 816.7 265.3 411.07 1202.1 0.01878 285.0 387.1 815.1 1.67978 270.3 . '412.67 - 388.9 -
1202.3 0.01880 1.65049 290.0 275.3 J 813.6 1202.4 0.01882 414.-25 9 1 390.6 " .
- R12.0V 1.62218 295.0 280.3 ' 415.81 1202.6 0.01885 300.0 285.3 392:3 - 810.4 ' 1202.7 0.01887 1.59482 417.35 394.0 808.9 1.56835 320.0 305.3 423.31 1202.9 0.01859 340.0 400.5 802.9 1203.4 1.54274
- s. 325.3 428.99 0.01899 360.0 i 345.3 , 406.8 797.0 1203.8 , 1.44801
~380.0 ~ '" 365.3 434.41 - 412.8 - 791.3- 0.01908 7 ; 1.36405-439.61 1204.1 0.01917 418.6 785.8 1.28910 --
400.0 385.3 1204.4 0.01925 444.60 424.2 780.4 1.22177 420.0 405.3 . 449.40 1204.6 0.01934 440.0 429.6 775.2 1204.7 1.16095 425.3 .
454.03 434.8 0.01942 1.10573 460.0 .c, . 445.3i.. 770.0 '
1204.8 458.50 - .._439. 8 .. 0.01950 1.05535
~480.0 ; r
- ~ ~465.3 5 ,462.82 444.7* * .
765.0 v. _1204.8 W, . - 0.01959 , . .,-. - - 1.00921-500.0 760.0 -
1204.8 er- 4b5.3 467.01 0.01967 0.96677 -
520.0 449.5 755.1 1204.7 505.3 471.07 454.2 0.01975 0.92762 540.0.. . 525.3 . 750.4 1204.5 0.01982 475.01 458.7 s 0.89137 560.0 545.3 745.7 1204.4 0.01990 478.84 463.1 0.85771 580.0 565.3 482.57 741.0 1204.2 0.01998' 467.5 736.5 1203.9 , 0.02006 O.82637 600.0 - 585..b 620.0
- 486.20 471.7 732.0 0.79712 605.3 489.74 1203.7 - 0.02013 640.0 475.8 727.5 0.76975 625.3 493.19 479.9 1203.4 0.02021 660.0 723.1 0.74408 645.3 i 496.57 483.9 1203.0 0.02028 0.71995 680.0 665.3 718.8 1202.7
- 499.86 487.8 0.02036 0.69724 700.0 714.5 1202.3 0.02043
- 685.3 503.08 491.6 0.67581 720.0 705.3 710.2 1201.8 740.0 506.23 495.4 0.02050 0.65556 725.3 509.32 706.0 1201.4 . 0.02058 760.0 745.3 499.1 701.9 1200.9 0.63639 512.34 502.7 0.02065 0.61822 1780~.0 " 765.3" 515.30 506.3
. 697.7 -
693.6 1200.4 0.02072 0.60097 )
800.0 785.3 1199.9 0.020S0 1 518.21 509.8 0.58457 820.0 805.3 521.06 689.6 !!99.4 840.0 513.3 0.02087 0.56596 825.3 685.5 1198.8 860.0 523.86 516.7 0.02094 0.55408 845.3 526.60 681.5 1198.2 880.0 - 520.1 0.02101 0.53988 i
865.3 '529.30 677.6 1197.7 --
-0.02109 i 900.0 523.4 673.6 1197.0 0.52631 '
885.3 531.95 526.7 0.02116 0.51333 920.0 905.3 669.7 1196.4 534.56 530.0 0.02123 0.500V1 940.0 925.3 a.. 665.8 1195.7 537.13 533.2 0.02130 960.0 -i- 945.3 539.65
- 661.9 1895.1 . 0.02137 0.48901 980.0 536.3 658.0 0.47759 965.3 542.14 1194.4 0.02145 539.5 654.2 0.46662 -
1000.0 985.3 1193.7 0.02152 544.58 1050.0 1542.6 650.4 0.45609 '
- 1035.3 550.53 ' 550.1 >
!!92.9 .. ~ 0.02159 0.44596 -
1100.0 1085.3 640.9 1191.0 -
e T-~
. . 1150.0 1135.3 556.28 561.82 557.5 631.5 1189.1 0.02177 0.42224 1200.0 564.8 622.2 0.02195 0.40058 1185.3 567.19 571.9 1187.0 0.02214 1250.0 613.0 1184.8 0.38073 1235.3 572.38 578.8 0.02232 0.36245 1300.0 1285.3 603.8 !!B2.6 1350.0 577.42 585.6 594.6 0.02250 0.34556 1335.3 582.32 592.2 1180.2 0.02269 ig .
1400.0 1385.3 585.6 1177.8 0.32991 1450.0 587.07 598.8 0.02288 1435.3 567.5 1175.3 0.31536 591.70 605.3 567.6 0.02307 0.30178 1500.0 1455.3. 1172.9 0.02327 1600.0 596.20 611.7 558.4 0.28909 1585.3 604.87 1870.1 0.02346 1700.0 . + 624.2 540.3 1164.5 0.27719 1685.3 613,13 .
0.02387 1800.0 636.5 - 522.2 !!58.6 '0.25545 1785.3. 621.02 648.5
- 0.02428 1900.0 1885.3 503.8 1152.3 .
0.23607 628.56 660.4 485.2 0.02472 2000.0 19MS.J 1145.6 0.21861 6J5.30 . 672.1 0.02517 0.20278 2100.0 20M5.3 466.2 li3'8.3 2200.0 642.76 683.8 446.7 0.02565 2185.3 1130.5 0.18831 2300.0 649.45 695.5 426.7 0.02615 0.17501 2285.3 655.89 707.2 1122.2 0.02669 2400.0 23M5..I 406.0 1113.2 0.16272 662.11 719.0 384.8 0.02727 0.15133 2500.0 cc 2MS.3r , hbM.s I .
1103.7 0.02790 ' -
2600.0 - 731.7 Jb l .6 0.14076 25K5.3 673.91 1093.3 2700.0 744.5 337.6 0.02859 ,3 0.33 pen 26MS.3 674.53 1082.0 2800.0 757.3 312.3 0.02938 0.12110 2783.3 644.96 1069.7 770.7 0.03029 2900.0 2HN5.3 285.1 1055.8 0,11194 690.22 7A5.1 254.7 0.03134 0.10305 Joon.0 29MS.J ~1039.8 0.03262 3100.0 695.33 801.5 218.4 0.09420 3085.3 700.28 824.0 1020.3 0.03428 3200.0 31M5.3 169.3 993.3 0.0s500
- 705.08 875.5 0.03681
_ 3208.2 319,1.5 705.47
. 56.1 931.6 0.07452 906.0 0.0 0.04472 0,05663 906.0 0.05078 0.05078
. .j . ,
1
. l SECTION FIVE THEORY NUCLEAR POWER OPERATION.
FLUIDS AND THERMODYNAMICS l l
l I
5.01 QUESTION (2.0)
A plutonium-beryllium neutron source i s installed in a reactor core for initial start-up and an antimony-beryllium source is installed f or subsequent start-ups.
- a. What are the two principle reasons an installed (1.0) neutron source is required ? j i
- b. Why is the plutonium-beryllium source used only f or (0.5) initial start-up ? ,
I
- c. How long af ter shutdown will it take f or the (0.5) antimony-beryllium source to decay to about half of its' original strength ?
5.01 ANSWER
- a. Provide verification of startup range instrument (1.0) operability and allow monitoring of neutron !
popul ati on changes.
l 1
- b. The plutonium depletes under a high neutron flux (0.5) l (such as that which occurs when the reactor is at l power)
- c. About 60 days (source will be useless after about (0.5) 240 days)
REFERENCE:
Atomic and Nuclear Structure, NLC55-OOXC-OO1, Pg. 5.5-5.7
5.02 OUESTION (2.5)
With shutdown and partial length CEA's withdrawn and control bank CEA's inserted, the reactivity of the reactor is -6950 PCM. The START UP instrumentation indicates a count rate of 25. After the control CEA's are moved out to 20 inches on bank 4 the count rate increases to 250.
s-
- a. Nhat is K when the count rate is 2507 (1.5) eff
- b. After criticality is achieved, how much positive (1.0) reactivity is required f or a START UP RATE of 0.7 DPM?
(assume BETA = 0.007 and LAMDA = 0.1) 5.02 ANSWER
- a. -6950 PCM = -0.0695 ; K = 1/1 p = 0.935 (0.5) effo CR O - eff0'* 1 ~ eff1' '
25 (1-0.935) = 250 (1-Keff1) 1.625 = 250 - 250(Kg4,1) 248.375 = 250(Keff1}
K effi =248.375 /250 K
eff1 = 0.9935 (+ or - 0.001) (0.5)
- b. SUR = 26*( h*p)/(b - p) SUR = +0.7 DPM (0.5) l
- 0. 7 = 26 (0.1) (p) / (0. OO7 p)
{
O.7(0.007 p) = 2.6*p O.0049 - 0.7p = 2.6p 3.3p = 0.0049 therfore p = 0.00149 0.0015 = /0 = 150 PCM (+/- 2 PCM) (0.5)
REFERENCE : Palo Verde Nuclear Generating Station " licensed Operator Training Reactor Theory" sections 8 and 10 and PVNGS Operating Procedure No.410P-12203 Rev.4 .
i i
1 1
r t
5.03 OUESTION (3.O)
The two most significant fission products which insert negative reactivity into a core are samarium and xenon. To answer the f ollowing questions please refer to attached Figures 5.1 8< 5. 2 Samari um Concentrate on ' and Figure 5. 2 Xenon Concentrate on.
- a. On Figure 5.1, (Point "A") why does samarium cont:.nue (0.5) to increase after shutdown 7
- b. On_ Figure 5.2, (Point "B") why does samarium drop below the (0.5) equilibrum valve of -630 pcm after start-up 7
- c. On Figure 5.3, (Point "C") why is equilibrum xenon dependent (0.5) upon power level 7
- d. On Figure 5.3, (Point "D") why does xenon drop after the (0.5) power increase ?
- e. How long does it take for xenon to peak following (0.5) shutdown ?
- f. What does the term " xenon precluded" mean ? (0.5) 5.03 ANSWER
- a. Samarium increases because removal has become zero (0.5) and production (from promethium decay) continues.
- b. Samarium burns out when at power and removal (0,5) exceeds production until samarium precursor (promethium) builds up,
- c. Increasing power (f l ux ) increases both production (0.5) and removal, however removal also depends on decay of xenon.
- d. Removal (from burnout) is greater than production until (0.5) xenon precursor (iodine) builds up in the core.
- e. 6 to 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> (0.5)
- f. The negative reactivity asscociated with xenon exceeds (0.5) the the avaiable excess reactivity, or the rate of change of Xe precludes continued operation.
REFERENCE : PVNGS LICENSED OPERATOR TRAINING REACTOR THEORY,-NLC-55.page 15.4, 16.3-16.4b ,
ANPN -OO148-DFH-96.21 (NEW MATERIAL DATED NOV. 20, 1986.).
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...m COMPLETE SAMARIUM CYCLE
~
Reactor Startup e c r To 1007. Power Recet'or 880 St rtup To Shutdown g /
3
',. '3 100% Power y
.-630- -
-Sm Reactivity l-12.5d J lQl l -
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g_ . . .
Days Of Power Operation S
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O 5.04 OUESTION (3.0)
The reactor is initiall y operating at 50% power, begining of life, 120 j inches withdrawn on control group 5, baron concentration at 650 ppm. T 1 is programmed normally. Use the attached Figures 5.4, 5.5, 5.6, 5.7 an8#0.8 to answer the f ollowing questions. Consider each situation separately.
(a) .With no rod motion what change in boron concentration (1.5)'
is necessary to increase power to 100% ?
- (b) What will be the rod insertion immediatly af ter a power (1.5) reduction to 25% without any boron concentration change ? l 5.04 ANSWER (a) At 50% BOL power defect = - 750 (+/-20) pcm, at 100% it is- 1380 (+/-20) pcm. (0.5) l Change in reactivity required = 630 pcm. g {
Boron worth = -11.4 (+/-0.2) pcm/ ppm G 579 - 593 F (0,5)
{
l 630 pcm/(-11.4 pcm/ ppm) = - 55 (+/-5) ppm I baron must be decreased 55 ppm - (0. 5)
J (b) At 50% BOL power defect is -750 (+/-20) pcm At 25% it is -410 (+/-20) pcm; reactivity change is -340 pcm. (0.5)
Rod group 5 at 120 In. is 20 (+/-10) pcm on CEA Worth curve. 1 20 + 340 = 360 pcm (0.5)
This is 15 (+/-10) In. on group 5 Rods must be inserted to 15 inches on group 5 (0.5) l i
REFERENCE:
Palo Verde, Licensed Operator Training, Section One, NLC55, Chapter 13, 14, 15, and 16.
i
=
FIGURE 5.y PVNGS/ UNIT 1 CYCLE 1 ,
POWER DEFECT VS POWE.R LEVEL (BOC, MOC, EO
-2 0 0 0,'
, . c .- . '
j'
-1750-j i .[l
- I / -
-1500~
i # /l "
5 l '
+ll
-1250 -
) 3/ Y' 5
[ ,/h[ l -
i -
fi/ l P -1000
-i y / AA f- ; / ,f i
-250f f//
i lhf -
' ff7
-500-5 Af i / -
T
-250'
_ / .
i / ,
0 ]*
l -l 0 10 20 30 40 50 60 4 i-r 70 80 90 100 f(. FUt.1. POWER LEGEND: EFPD 0 M -e 208.3 - 455'.7 '
et t
"r
FIGURE 5.5 '
\
i BORON WORTH V5 T-A VG: l (70 - 565 F, 0% P, B O C) _
-1s. 0 ' ';< ' ' '
I IN _
l l l l l i I II l
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l g l l l g g, 50 100 150 200 250 300 350 400 450 s00 550 T-AVG (DEGREES F)
L EGEN5': PPM -
600 '
0-0 -e 1000 h e 1500 -
2000 e
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FIGURE 5.L PVNGS/ UNIT 1 CYCLE 1
~
l BORON WORTH VS T-A VG \
.' (505* - 615 F, 100% P, BOC) _ !
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i 5.05 DUESTION (2.0)
The spray pond pump normal parameters are li sted below: ,
Discharge pressure 50 PSIG Discharge flow 16000 GPM Motor Amps 50 Amps Motor speed 1800 RPM During an outage the motor i s replaced with a new motor. Unknown to PVNGS personnel the vendor delivered a 1200 RFM motor.
When the pump is started f or testing, what are the f ollowing parameters 7
- a. Discharge pressure (0.5)
- 6. Discharge flow (0.5)
- c. Motor amps (1.0) 5.05 ANSWER
- b. N g /N v /v 2 1 2: (1200/1000) x 36000 = 10700 GPM +/-100 (0.5)
- c. (N /N ) = Hp /Hp :
2 3 2 (0.33)
Since voltage is fixed Hp or work is proportional to amps (0.33)
(1200/1800) x 50 = 15 +/-1 amps. (0,33)
REFERENCE : PVNGS LICENSED OPERATOR TRAINING, THERMDHYDRULICS REVIEW, page 19 and 20 4
I 5.06 DUEST1ON (2.O)
A reactor is operating at 75% indicated power. 100% power is 3800 MWT. The following plant data is observed.
Feedwater Temperature........... 440 F Feedwater Enthalpy(hf).......... 420 DTU/LBM 6
Total Feedwat er Fl ow. . . . . . . . . . . . 12. 5x 10 LBM/HR Steam Pressure.................. 965 PSIG Other System Losses / gains....... 10 MW (RCP power-losses to ambient)
What is the actual power output of the reactor, in percent of full power ?
(Show Cal cul ati ons , use the attached Steam Tables) 5.06 ANSWER (0.5)
O=MdH
= 12.5x10 x(1195-420) = 9. 69;t 10 DTU/HR BTU /MW-HR = 28a0 MWT (0.5) 9.69x10 BTU /HR/3.41x10 (0.5) 2840 - 10 = 2830 MWT from reactor (0.5) 2830/3800 x 100 = 74.8 % +/-0.6%
REFERENCE : PVNGS LICENSED OPERATOR TRAINING, THERMOHYDRAULICS.page 16 and 17 i
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1 5.07'OUESTION (1.5) l l
Technical Specifications state limits' on dissolved oxygen in the reactor i coolant system.
- a. What chemicals are used to control dissolved oxygen in the (1.0) I primary coolant system 7 -
5.07 ANSWER
- a. Hydra:ine and hydrogen (1.0)
- b. O.1 ppm and 1.0 ppm respectively (0.5)
REFERENCE : PVNGS Technical Specification 3/4.4.6 and Training Artical NS-2A, Pg. NS-2A-66.
Technical Specifications 3/4 4-23 l
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t 5.08 OUESTION (1.0) 4 Answer each of the following statements TRUE or FALSE. (0.25 each)*
- a. The flow of a positive displacement pump can be varied by
_ adjusting a pressure control discharge valve,
- b. A centrifugal pump will draw less current if it is started with the discharge valve shut, than it would with the discharge valve open.
- c. Net positive suction head' decreases with decreasing flow and decreasing temperature.
- d. The flow of a centrifugal pump system decreasec when the head loss in the cystem decreases.
5.8 ANSWER
- a. FALGE
- b. TRUE l
- c. FALSE 1
- d. FALSE
REFERENCE:
PVNGS THERMOHYDRAULICS REVIEW PUMP LAWS.
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l 5.09 OUESTION (1.0)
The power coef ficient of reactivity and the isothermal temperature coefficient of reactivity are due to the effects of the fuel temperature coefficient of reactivity and the moderator temperature coef ficient of
~~
reactivity.
(0.5)
- a. Assuming that the Reactor Regulating System was in manual (no rod motion), why would the power coefficient of reactivity be less negative 7
- b. Why is the isothermal' coefficient of reactivity more negative (0.5) for increased temperatures 7 5.09 ANSWER The negative contribution due to moderator temperature (0.5) a.
coefficient with increased Tave would not be present.
- b. The moderator de'sity n changes f aster with increasing temperature. (0.5)
I REFERENCE : TRAINING ARTICALE NS-1C REV.1, NLC55-XC-13 pages 13.11, and i 13.10 Note: The candidates were given a clear explanation (verbal) of this question during the examination.
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DUESTION 5.10 (1.5)
During an outage 15 % of all steam generator tubes are plugged. While at full power, prior to the outagg, the steam generator pressure was 1062 PSIA. T ave is to remain at 593 after the tube plugging.
What is the expected full power steam pressure 7 5.10 ANSWER
. l 0 = UA6T = constant = UA(Tave - Tstm) (0.5) j i
I Ux 100% x (593 - 552) =Ux 85% x (593 - T) (0.25) 48.2 = 593 - T T = 545 (0.25)
T(sat) is 545 therefore P(sat) is 1000 PSIA (0.5) I REFERENCE; Palo Verde, Licensed Operator Training, Section Two, Thermohydraulics Review, page 14 -15 l
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j QUESTION 5.11 (2.5)
A primary calorimetric is being conducted using Th, Tc, and RCS fldw rate l aftnr plugging 1250 tubes in the Loop 2 steam generator. The f ollowing {
readings are recorded.
Flow rate Th Tc Loop 1 6 82x10 LB/HR 621.3 562.8 Loop 2 6 80x10 LB/HR 620.9 564.3 I Assume the following plant data.
Cp of water = 1.35 BTU /LB- F RCP power = 17 MW Losses to ambient = 7 MW
- a. What is the total reactor power in MWT ? (2.0) !
- b. If Tave is decreased 10 F at the same reactor power level (0.5) what will happen to indicated RCP motor amps ?
5.11 ANSWER
- a. 0 = MCpAT (0.5)
=
(82 x (621.3 - 562.8) + 80 x (620.9 - 564.3)) x 10 6 x 1.35 (0.5)
= 6 (4797 + 4528) x 10 1.35 = 12.589 x 10 BTU /HR 12.589 x 10 BTU /HR/3.41 x 10 6 BTU /MW-HR = 3692 MWT (0.5) 3692 - 17 + 7 = 3682 MWT (+/- 4) (0.5) l I
- b. RCP amps will increase (water is denser at lower Temp.) (0.5) {
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REFERENCE:
Palo Verde, Licensed Operator Training, Section Two, page 5 - 6 Training Arti cl e, NS-4, Excore nuclear Instrumentation System, page 21 NS-1A, Reactor Coolant System, pages 7 and 14 l
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'DUESTION 5.12 (1.0)
A moisture separator reheater relief valve is leaking to atmospheric pressure. The moisture separator reheater pressure is 120 PSIA and the discharged steam i s at 350 F.
Using the ' attached steam tables, what is the superheat of the steam in the moisture separator reheater ?
5.12 ANSWER 350 steam @ 14.7 PSIA has 1216 BTU /LB (0.2)
Constant enthalpy (0.2)
At 120 PSIA Tsat is 341 F (0.2) 1216 BTU /LB is 385 F (+/-5) (0.2) 44 F super' 'at (+/-5) (0.2)
REFERENCE:
Palo Verde, Licensed Operator Training, Section Two, page 12 Steam Tables 1
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5.13 QUESTION (2.0)
Shutdown margin and reactivity limits vary with plant operati onal mode.
1
- a. What is the shutdown margin limit for modes 1 through 4 7 (0.25)
- b. What worst case accident sets thi s limit ? _ (0.75)
- c. What is the shutdown margin limit for mode 5 ? (0.25)
- d. What accident is the mode 5 SDM limit based upon ? (0.75) 5.13 ANSWER 1
- a. 6% AK/K modes 1-4 (0.25)
- b. A steam line break at EOL with Tc at no-load operating (0.75) l temperature
- c. 4% AK/K modes 5 (0.25) '
- d. Single CEA withdrawl accident (0.75)
REFERENCE:
PVNGS TECHNICAL SPECIFICATIONS, page 1-9, B3/4 1-1, and 3/4 1-1 END OF SECTION FIVE l CONTINUE ONTO SECTION SIX l
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SECTION SIX PLANT SYSTEMS
- DESIGN, CONTROL, INSTRUMENTATION 6.01 OUESTION (1.0)
The reactor vessel level monitoring system (RVLMS) is used to determine and display the water inventory in the reactor vessel above the fuel alignment plate.
- a. What type of detector is used to measure the reactor vessel level? _( 0.25)
- b. How many channels of the RVLMS are there? (0.25)
- c. What is the basic principle of operation of the RVLMS device?
(Do not include materials or specific minor' components. Be brief.) (0.5) 6.01 ANSWER
- a. (Heated junction) thermocouple (HJTC) (0.25)
- b. Two ( The HJTCS is a class 1E safety grade system, therefore, it has two independent channels.) (0.25) l
- c. The basic principle of system operation is the measured temperature difference bctween adjacent heated and unheated thermocouple. (If a steam environment is present around the sensor a large delta T will be indicated, if a liquid inventory is present a small temperature difference will be detected.) (0.5)
REFERENCE : PVNGS TRAINING ARTICLE-NS-1C, Pgs. 16 & 26.
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4 6.02 OUESTION (1.0)
When venting the VCT to the Gaseous Radwaste decay tanks the maximum allowable oxygen concentration is limited to 2%. j
)
- a. Why is the concentration of oxygen of concern in the radwaste l system waste gas decay tanks? (0.5)
or greater? (0.5)
)
6.02 ANSWER
- a. A large portion of the waste gas from the VCT is hydrogen, if it is mixed with a suf ficient amount of oxygen a hydrogen explosion may occur. (0.5) 1
- b. Nitrogen (0.5)
REFERENCE:
PVNGS AS-9, Pgs. 4 & 5, 410P-1CHO1 Pgs. 17, 18, 19, 21.
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P-6.03 OUESTION (2.0)
The essential chilled water system supplies chilled water to the essential air cooling and air handling units in the control and i
auxiliary buildings.
- a. What are four of the five automatic signals that will start the Essential Chilled Water System (EC)? (1.0)
- b. What are the power supplies (Busses and Voltages) to chiller ;
package ECA-E01 and ECB-E01 compressor motors? (0.5) 1 I
- c. What is the normal and the alternate sources of makeup water )
to the Essential Chilled Water Expansion Tanks? (0.5) i 6.03 ANSWER
- a. (0.25 pts. ea. maximum 1.0 pts.)
- 1. Saf ety injection actuation signal (SIAS) or Containment Spray Actuation Signal (CSAS).
- 2. Control room essential filtration actuation signal (CREFAS).
- 3. Control room ventilation isolation actuation signal ..
(CRVIAS).
- 4. Loss of off-site power signal (LDP).
- 5. Auxiliary feedwater actuation signal (AFAS). 1 l
- b. For ECA-E01 Class 1E 4.16 Kv Bus PBA-SO3 (0.25)
For ECB-E01 Class 1E 4.16 - Kv Bus PBB-SO4 (0.25) c.
- 1. Normal source of makeup water is Demineralized Water. (0.25)
- 2. Alternate source is the condensate storage tank. (0.25)
REFERENCE : PVNGS Training Article AS-11B Pgs AS-118-4,88<14, Drawing J104-66 and Electrical Single Line Diagram 342-0100.
t 6.04 QUESTION (2.O)
It is necessary to move a single regulating CEA to return it to its proper position.
- a. What switches are used for the Manual Individual (MI) mode(1.0) of CEA operation?
- b. What three interlocks will inhibit CEA movement in the MI (1.0) mode?
6.04 ANSWER
- a. Select switch (0.33)
CEA-TENS switch (0.33)
CEA UNITS switch (0.33) or i
Group Select j
Mode Select Individual Select
- b. UEL (Upper Electrical Limit) (0.33)
LEL (Lower Electrical Limit) (0.33)
CWP (CEA Withdrawal Prohibit) (0.33)
REFERENCE:
PVNGS TRAINING ARTICLE NS-9E, Pgs. NS-9E-7,11.
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6.05 OUESTION (2.0) i j
Besides Total Core Power what are the other FOUR parameters that the Core Operating Limit Supervisory System (COLSS) CONTINUALLY calculates and compares to their respective LCOs? (2.0) 6.05 ANSWER !
- a. (0.5 pts. ea. Max. 2.0 pts.)
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- 1. Peak Linear Heat Rate (LPD)
- 2. Margin to DNB (Thermal Margin).
3 Animuthal Tilt.
- 4. Axial Shape Index.
REFERENCE : PVNGS TRAINING ARTICLE NS-6C, Pgs. 28<3.
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P 6.06 QUESTION (1.5)
When equipment failure or abnormal plant conditions are sensed, Automatic Motion Inhibit (AMI) and Automatic Withdrawal Prohibit
-(AWP) signals are sent to the-Control Element Drive Mechanism Control System (CEDMCS) from the Reactor Regulating System (RRS).
- a. What three parameters are compared for deviations by the Reactor Regulating System to generate an AMI? (1.0)
(Setpoints not required.)
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- b. What two Reactor Coolant System Parameters will generate an Automatic Withdrawal Prohibit? (Include setpoints.) (0.5) j 6.06 ANSWER ,
f a.
( 1. 0) . 4
- 1. T g deviation, i
- 2. Turbine Load Index (TLI) or turbine first stage pressure j devi ati on , J
- 3. Reactor Power deviation.
- b. (0.5) J 1
- 1. High T ae
>T gg, by 6 F error
- 2. High T , 575 F.
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REFERENCE:
TRAINING ARTICLE NS-9C, REACTOR REGULATING SYSTEM, l Pgs. 4 - G and 29 - 30 Drawing 01-J-SFE-053, Rev. O.
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6.07 OUESTION (1.0)
A small main steam line break has occured inside the containment.
The primary system is shutdown and stable, however the
~
containment atmosphere is greater than 200 F with a loss of pressurizer cubicle ventilation.
j
- a. Will this condition cause the indicated pressurizer level to (0.25) j increase or decrease?
- b. Why does an increase in the containment temperature to 200 F cause the indicated pressurizer l evel to be inaccurate? (0.75) 6.07 ANSWER (0.25)
- a. Increase
- b. Since the level transmitter reference leg is external to,the pressurizer and not' insulated, its temperature will also increase. (An elevated reference leg temperature will result in an artificially high pressurizer level i ndi cati on. ) (0.75)
REFERENCE : PVNGS TRAINING ARTICLE NS-9F, Pg. NS-9F-10.
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6.08 DUESTION (2.O)
For each of the f ollowing sets of conditions, . state which (one or more) of the following systems should actuate. Consider each event one at a tire for Unit 1. l SIAS CSAS CIAS MSIS RAS AFAS
- a. A feedwater line break occurs which causes containment pressure to increase to 7 PSIG, steam generator pressures to to drop to 1000 PSIS and 900 PSIG, and RCS pressure to drop to (0.5) 1900 PSIG.
- b. A LOCA occurs which causes containment pressure to increase to 15 PSIG and RCS pressure to drop to 1800 PSIG. (0.5)
- c. A hi gh l evel over-ride occurs as steam generator levels go to 93% narrow range and RCS pressure drops to 1800 psig. (0.5)
- d. Following a reactor trip, steam generator i falls to 800 PSIG and 20% wide range while steam generator 2 falls to 1000 PSIG and 50% wide range. RCS pressure f alls to 1900 PSIG. (0.5) 6.08 ANSWER i
(0,5)
- d. MSIS (AFAS-1 may occur but will be locked out) (0.5)
)
REFERENCE:
TRAINING ARTICLE NS-7A, Engineered Safety Features Actuation System, Pgs. 31-34 and 48-50, and TECHNICAL SPECIFICATION Pg. 3/4 3-25.
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- 1 6.09 QUESTION (1.5)
What are the lift set points for the main steam line safety valve groups 1, 2 and 3? (1.5)
I 6.09 ANSWER Grcup 1 Grcup 2 Grcup 3 Lift Setting 1250 psig 1290 psig 1315 psig (1.5)
REFERENCE : PVNGS TRAINING ARTICLE PGS-1A, Pg. PGS-1A-44, and Technical Specification Table 3.7-1.
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f 6.10 QUESTION (1.5)
The Main Steam Isolation Valves (MSIVs) are hydraulic-actuated, double disk, wedge valves.
(0.5)
- a. What system provides power for the MSIV hydraulic pump? I
- b. What feature of the hydraulic system provides fast closure(0.5) of ]
the MSIVs? J
- c. What is the " slow" mode of operation of the MSIVs used for?
(0.5) 1 4
6.10 ANSWER f 1
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- a. (Air from the ) Instrument and Service Air System. (0.5)
(0.5) l
- b. The hydraulic accumulators. )l (0,5)
- c. Normal operations and testing.
REFERENCE:
PVNGS, TRAINING ARTICLE PGS-1A, MAIN STEAM SYSTEM, Pgs. 11 -
14.
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6.11 OUESTION (2.0)
The Main Turbine Generator mechanical-hydraulic trip sub-system combines mechanical signals available at the Turbine Front Standard with electrical signals f rom the Electrical Trip and Monitoring sub-system (T&M) to control the Emergency Trip System (ETS).
The Main Turbine Generator is rolling at 1800 RPM:
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- a. Besides manual, what are the two other mechanical trips (non- {
electrical) to the mechanical-hydraulic sub-system? (1.0)
- b. Besides the master trip button and loss of 24 volt power trip, what are the two other 24 volt trip signals? (1.0) J l
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I 6.11 ANSWER Turbine overspeed. (0.5)
- a. 1.
i (25 PSI) bearing oil pressure. (0.5) 2.
(0.5)
- b. 1. Loss of both speed signals trip, Backup overspeed trip (DOST), (0.5) 2.
REFERENCE : PVNGS TRAINING ARTICLE PGS-3B, Pgs. PGS-3B-49&50.
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6.12 QUESTION (2.5)
- Question Deleted Generator differential, low engine lube oil pressure and engine overspeed are three signals that will trip the diesel in the '
emergency mode of operation.
What are five of the seven additional signals that will trip (2.5) the diesel engine in the TEST mode?
6.12 ANSWER *0uestion Deleted!*
(0.5 pts. each, Max 2.5 points)
- 1. High jacket water temperature,
- 2. Low turbocharger lube oil pressure,
- 3. Turbocharger bearing failure,
- 4. E>:cessive er.gine vibration ,
- 5.
Dearing high temperature,
- 6. Crankcase pressure high,
- 7. Incomplete sequence.
REFERENCE:
PVNGS LESSON NO. NLC22-OO-RC-033-OA REV.A, Pg.42.
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6.13 QUESTION (1.5)
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The Steam Bypass Control System (SBCS) is designed to maximize unit availability by accommodating load rejections.
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- a. Why is an Automatic Withdrawal Prohibit (AWP) signal generated 1 by the SBCS when there is a demand signal to open the turbine bypass valves? (0.5)
- b. How will the SDCS control system be affected by placing a SBCS
" valve permissive switch "
in " manual"? (0.5) l
- c. Besides a reactor trip with low reactor coolant temperature what other event will cause a quick-opening block of the j turbine bypass valves? (0.5) j j
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6.13 ANSWER
- a. The AWP is generated to prevent the potential increase in !
reactor power at a time when excess NSSS energy exists with
~
l respect to steam demand. (0.5)
]
- b. A valve permissive signal is generated regardless of the plant conditions. (Valve can be controlled during startup and shutdown) (unless a loss of vacuum occurs or an emergency off signal is generated). (0.5)
Loss of feedwater pump. (0.5)
- c. ]
REFERENCE:
PVNGS TRAINING ARTICLE NS-9B, Pgs. 10, 11 and 21.
Instrument Loop Diagram 01-J-SFE-058 Rev. O.
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6.14 DUESTION (1.5)
The function of the 125 VDC Class 1E Power System is to distribute power to 125 VDC Class 1E vital loads. !
- a. What. Motor Control Center supplies power to backup battery (0.5) charger A-C (PKA-H15)?
}
- b. On a loss of power to battery charger "A" (PKA-H21) with the back-up battery selected to channel "C", what will be the ]
sources of power to the Channel "A" 120 VAC distribution j (1.0) l panel?
Note: Candidates were instructed during the e>tamination to add l the word " charger" after ...back-up battery... . 3 i
1 6.14 ANSWER -l PHA-M33 (480 V MCC). (0.5) a.
- b. Inverter "A" and/or battery "A", (0.5) l 480/120 Volt regul ated power. (0.5)
REFERENCE:
PVNGS TRAINING ARTICLE PGS-15D, FIG. PGS-15D-2, PGS-15E, PG. 8, PGS-15C, PG. 13.
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6.15 OUESTION (1.O)
The Condenser Air Removal System removes air and non-condensable gasses from the condenser shell during startup and power operations. Moisture separators and electric heating coils are used to remove moisture from the condenser air removal system !
exhaust.
- a. Why is moisture removed from the air removal system exhaust?
(0.5)
- b. What automatic action occurs if a High Radiation alarm is received from SON-RU-117 (0.5) i Note: The candidates were inf ormed during the examination to delete SON-RU-11 and replace it with "the condenser air !
removal discharge monitor". f 1
6.15 ANSWER l
)
- a. Moi stur e is removed to prevent charcoal filter damage. (0.5) j (During normal operation air removal does not go through the j charcoal filters therefore moisture removal is not necessary.)
i
- b. A high radiation alarm will cause the exhaust flow to divert (0.5) i to the exhaust filter unit.
REFERENCE:
PVNGS TRAINING ARTICLE PGS-9B,Pgs. PGS-9B-12, 13, 23.
l
i 6.16 OUESTION (1.0)
The Nuclear Cooling Water System (NC) provides cooling water to safety and non-safety related components. It is cooled by the Plant Cooling Water System (PW).
- a. What method of system operation is used to prevent contaminating the circulating water system with NC? (0.5)
- b. What signal will automatically isolate the NC from the EW (0.5) system if they are cross connected?
6.16 ANSWER
- a. The operating pressure of the NC is less than the PW system(0,5) preventing leakage from the NC to the PW system.
6.
- 1. A SIAS will automatically isolate the EW system f rom the NC system. (0.25)
- 2. Low Surge Tank Level. (0.25)
REFERENCE:
PVNGS TRAINING ARTICLE AS-1B, Pgs. 5 8< 25, Drawing 01-E-EWB-OO3, Rev. 3.
END OF SECTION SIX CONTINUE ONTO SECTION SEVEN l
l SECTION 7 PROCEDURES NORMAL, ABNORMAL, EMERGENCY, AND RADIOLOGICAL CONTROL i
l 7.01 QUESTION (1.5)
Procedure 410P-1CHOi "CVCS Normal Operations" governs the operation of the CVCS for baron addition to the Reactor Coolant 1 System.
- a. How do the operators ensure that all of a large boron addition is made into the RCS - ather than remaining in the Charging Systems? ( 0. 7,5 ) - q i
- b. If a baron addition is made that changes the RCS boron j concentration by more than 75 ppm, what action must be l s
taken? (0.75) 7.01 ANSWER l
- a. By flushing the line from the blending tee to the suction of j the charging line (approximately 30 gal.). (0.75) ]
- b. Initiate pressurizer spray to equalize the Pressurizer to the RCS boron concentration (to within 10 ppm.). (0. 75)
REFERENCE : 410P-1CHO1 pages 27 and 28.
i
t 7.02 OUESTION (3.0)
Procedure 41RO-12ZO5 " Loss of Feedwater" cautions the operator _s when instrumentation indicates the Steam Generator is dry.
- a. Why is 0% Wide Range level indication not a good indicator by itself of dryout conditions in a Steam i Generator? (1.0)
- b. What other indication must be utilized to determine i f a single Steam Generator is dryed out? (1.0)
- c. With both Steam Generators dry, why is the operator cautioned to initiate f eed very slowly to a dry Steam l 1
Generator? (1.0) 7.02 ANSWER
- e. 0% Wide Range still leaves 1/3 of tube length covered. (1.0)
- b. Tc approaching Th for the affected loop. (1,0)
- c. Feed should be initiated very slowly to minimize thermal !
stress on the Steam Generators (and associated piping).
(1.0)
REFERENCE : 41RO-1ZZO5 pgs. 6, 7, and 8.
.1 l
l i
7.03 QUESTION (1.O)
During a roll of the Main Turbine the. procedure 410P-;MTO2 " Main Turbine" cautions that the Turbine should not bn held at speeds ,
of less than 800 RPM for more than 5 minutes.,What two concerns arise from holding the Turbine to_ speeds under 800 RPM for ;
i greater than 5 minutes? (1.0) !
7.03 ANSWER Increased possibility of packing rubs. (0,5)
Lower than actual vibration readings. (0.5) , j i
, -4
'{ '
REFERENCE:
PVNGS PROCEDURE 410P-1MTO2, Pg.36. ;.
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7.04 QUESTION (3.5) 4 A survey of radiation and contamination levels has been made by Radiation Protection personnel . The following page has a layout of the area around the three Charging pumps. For each of the Charging pump rooms, what posting and controls have to be f instituted per facility procedures at the entrance to the rooms? l (2.5) ]
(Pertinent portions of 10CFR2O are also provided.) i 7.04 ANSWER Charging Pump No. 1 I Room must be posted as High Contamination Area; (0.5) '
and Radiation Area. (0.5)
Must also be barricaded. (0.5)
Charging Pump No. 2 l
Room must be posted as High Radiation Area and; (0,5) j locked with key controlled by Unit Radiation (0.5)
Protection Supervisor.
Charging Pump No. 3 Room must be posted as Radi ation Area; (0,5) and Airborne Contamination Area. (0.5)
REFERENCE : 75AC-9ZZO1 and 75AC-9ZZO3 l
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PALO VERDE NUOLEAR GENERATING TROCENRE NO.
STATION MANOAL ,
pacendix A
. .~
TsAC-szzes page i of 1 -
,. - RADIOACTIVE REVISION CONTAMINATION CONTROL 4 I Page 10' of 10 fl0 1,
EVNQ$ CONTAMINATION LIMITS (LCTS #010670) -
SKIN AND CLOTHING CONTAMINATION LIMITS TYPE OF LOCATION . -
CONTAMINATION CONTAMINATION LIMITS (
REMOVABLE Skin FIXED
- j 3
Beta-Gamma
! Alpha None detectable None Detectable 1000 dpm**
Clothina Beta-Gamma
- \I Alpha 1000 dem/100cm2 '
20 dem/100cm2 1000 dem
- (
TOOLS, !
- ))
' TYPE OF - EQUIPMENT AND MATERIAL CONTAMINAT S }
CONTAMINATION CONTAMINATION LIMITS )
REMOVAELE !
Beta-Gamma- FIXED
- T Alpha 1000 dem/100cm2 J 3 1000 dpm 20 dem/100cm2 # - '
AREA CONTAMINATION LIMITS t
\
LOOSE SURFACE CONTAMINATION LIMI!
BETA-GAMMA Clean Area (No costing) ALPHA Less than
! 1000 dpm/100cm2 Less .than Contamination Area 20 dpm/100em2 11000 dpm/100cm2 High Contamination Area '
120 dpm/100cm2 150000 dpm/100cm2 o
11000 dpm/100cm2 As measured o* Personnel with HP-210 probe or ecui unable shall be valent.
Suoervisor.
to evaluated be decontaminated by the Unit to below 0 .All Radiation this level Protection -
further evaluated by Radiation Protecticases und shallwhere be fixed Al on Supervision.
I
~ = -
FOR INFORMATION ON,LY O%
I 7.05 DUESTION ( 2. 0 ).
Regarding operation of the Essential Auxiliary Feedwater System' in accordance with procedure 410P-1AF01 " Essential Auxiliary -
Feedwater System":
- a. Why must the minimum speed of 1000 RPM be maintained when operating the Steam D. riven Essential Auxiliary ,
Feedwater Pump? (1.0) {
- b. When aligning the Essential Auxiliary Feedwater System !
to the Reactor Makeup Water. Tank (RMWT), what wil1 occur if venting the supply line with CHN-VMO69. is not done? (1.0) 7.05 ANSWER
- a. Ensure adequate (oil) lubrication (of the pump's sleeve) bearings and (water lubrication) of the pump internals.
l (1.0)
- b. If this action is not taken loss of suction can occur if suction has to be taken from the RMWT. (1.0)
REFERENCE : PVNGS PROCEDURE 410P-1AF01, Pgs. 12 and 13. )
l l
1 I
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l 7.06 OUESTION (2.0)
You have declared Unit 1 control room uninhabitable as the Shift Supervisor. Shutdown is to be conducted in accordance with j procedure 41AO-12Z27 " Shutdown Outside Of Control Room". Aside from tripping the Reactor what are the other four (4) initial operator actions that must be taken while evacuating the Control j i
Room? (2.0) 7.06 ANSWER
- 1. Manually trip Two (2) reactor coolant pumps (RCP's 1A and l 1B or 2A and 2B). (0.5) l l
Manually trip the turbine (at B06). (0.5) j 2.
- 3. Place the letdown control valves select switch '
(CHN-HS-110-1) in the "Both" position. (0.5)
- 4. Place the backpressure control valves select switch (CHN-HS-201) to the "Both" posi ti on.
REFERENCE:
PVNGS PROCEDURE 41AO-12Z27, Pgs. 6 84 8.
l' 4
4 7.07 OUESTION ( 1. O)
Procedure 41AO-12229 " Reactor Coolant Pump and Motor Emergency" specifies certain actions to be taken immediately if certain' pump parameters are out of their limits.
1A IB 2A 2B RCP No.
Thrust Bearing 225 233 218 243 U
Temperature F i 177 180 181 195 Upper Journal Bearing Temp. F I
233 230 231 229 No. 1 Seal Pressure psig What actions are you required to take based on the information provided above? (1.0) 7.07 ANSWER Trip the Reactor. (0.5)
Trip RCP 2B (Upper journal bearing temperature high.) (0.5)
REFERENCE:
PVNGS PROCEDURE 41AO-12Z29, Pgs. 8,.10, and 11.
1
i 7.08 OUESTION (2.0)
Unit 1 operating at 100% power has just e>tperienced a Containment Spray Actuation. During the diagnoses per 41AO-1ZZ30 " Inadvertent CSAS" it is determined that the event was inadvertent and will not reset. The Containment Spray Header Isolation valves did not open. Step 3.1.1 of the procedure requires the Containment Spray Pumps be stopped. The procedure further cautions that the pumps will not start automatically after they are stopped with a CSAS present. Theref ore the system is INOPERABLE.
- a. What is the purpose of not allowing.the pumps to restart automatically if required? (1.0)
- b. What action needs to be taken in accordance with the Technical Specifications? (1.0) 7.08 ANSWER
- a. Stopping the containment spray (CS) pumps while a CSAS ;
~
signal is present will inhibit pump operation until the CSAS signal is reset to prevent the breakers from cycling. (1.0) l
- b. (The system would appear not to meet the operability J requirements of the LCO's (f or 3.3.2 and 3.6.2.1) therefore Tech. Spec. 3.0.3 applies.) Action must be taken within ONE hour to place the plant in a MODE in which the specification does not apply. (1.0)
REFERENCE:
PVNGS PROCEDURE NO. 41AO-1ZZ30 Pgs. 4 8< 6 and 17.
T. S.'s 3.3.2, 3.6.2.1, and 3.0.3 I
1 1
i l
5 7.09 OUESTION (2.0)
I
, You are in the'"Small Loss of Coolant Accident" procedure 41RO-12208 following the rupture of several instrument lines on l the Reactor Coolant System. What four criteria must be met in order for the safety injection flow to be throttled? (2.0) ,
7.09 ANSWER Reactor coolant system subcooled greater than 28 F, ( . 5) 1.
- 2. Reactor vessel level indicates voids restricted to upper head, (0.5)
Pressurizer level greater than 33% and controllable, (0.5) 3.
- 4. One steam generator capable of maintaining heat removal. (0.5)
REFERENCE:
PVNGS PROCEDURE 41RO-1ZZO8, Pgs. 10 84 35.
I l
I
~
l i
~ ' . 10 QUESTION (2.5)
Unit 2 is operating at 90% power when a station blackout occurs, )
procedure 41RO-1ZZO9 " Blackout" has been entered. It has been determined that the Emergency Diesel Generators have not closed in on their respective busses, but are running unloaded.
- a. How much time do the operators have to restore the Emergency Diesel Generators to their busses if the engines are lef t running? (0.5)
- b. Why are you limited for time? (0.5)
- c. If the Blackout will continue for a considerable time, I what conditions must be maintained to maximize the duration of Subcooled Natural Circulation (Three responses required).? (1.5) 7.10 ANSWER
- a. 60 minutes. (0.5)
- b. No Essential Cooling Water, (Essenti a11 Spray Pond) (0.5)
- c. 1. Safety valves must not be allowed to cycle. (0.5)
- 2. RCS heat removal must be minimized. (0.5)
- 3. RCS inventory loss must be minimized. (0.5) l
REFERENCE:
PVNGS PROCEDURE 41RO-1ZZO9, Pgs. 5 and 7, PE< I D 13-M-SPP-OO2 Rev. 11.
f I
7.11 QUESTION (1.5)
~The reactor has been manually tripped due to a loss of DC bus-PKA-M41 as required by Procedure 41AO-1ZZ17, " Loss Of 125 VDC Class 1E Electrical Power". This causes Instrument Air containment isolation valve IA-UV-2 to fail closed.
- a. How is pressurizer pressure controlled when pressurizer 4 I
spray valves become inoperable once the air pressure bleeds off? (1.0)
- b. In what position will Auxiliary Feedwater-Regulating valve AFA-HV-32 fail (open, closed, or as is)? (0.5) 7.11 ANSWER
- a. Maintain pressuri er pressure using pressuri er Auxiliary Spray. (1.0) .
- b. As is. (0.5)
REFERENCE:
PVNGS PROCEDURE 41AO-1ZZi7, Pg. 7 of 48.
I i
3: I I
1 7.12 OUESTION (3.0)
- Due to a Loss of Offsite Power, you are establishing natural {
l circulation per procedure 41AO-1ZZ13 " Natural Circulation Cooldown" and are verifing that it exists with 41EP-12201 )
j
" Emergency Operations" appendix E.
- a. Approximately how long after loss of forced flow will it f be before indications of natural circulation will be apparent? (0.5)
- b. What are f our (4) of the six parameter values used for verification of adequate Natural Circulation cooling? l (2.0) l
- c. If you notice subcooling margin decreases to 20 F what f action is required? (0.5) f 7.12 ANSWER I
- a. 5 to 15 minutes (0.5)
- b. (0.5 pts. ea. any 4) I i
- 1. Hot leg temperatures stable or decreasing (observe core exit thermocouple if T ff "C#l*}*
hat
- 2. Cold Leg temperatures stable or decreasing.
- 3. Cold leg temperatures close to (at) T g of S/G.
- 4. Core Delta T less than full power Delta T of 57 F. ;
- 5. Consistency between hot leg temperature and core exit thermocouple (ERFDADS or QSPDS).
- 6. RVLMS greater than ero (outlet plenum indicates full).
- 7. RCS greater than 28 F subcooled.
- c. Initiate SIAS. (0.5)
REFERENCE:
PROCEDURES 41EP-12ZO.1 Appendix E pgs. 41, and 42.
Page 34 paragraph 5.1 specifies knowledge required for the Operators in the use of the flow charts.
END OF SECTION SEVEN CONTINUE ON TO SECTION EIGHT
I SECTION 8 ADMINISTRATIVE PROCEDURES, ~-
CONDITIONS, AND LIMITATIONS 8.01 QUESTION (1.0)
Conduct of Shift Operations procedure 40AC-9ZZO2 assigns different duties to the Unit 1, 2, and 3 Shift Supervisors
- a. Which Shift Supervisor has the additional ;
responsibility of the 525 KV switchyard?
(0.25)
- b. When is the Unit i shift supervisor responsible for station security? (0.75) 8.01 ANSWER
- a. Unit 1 (0.25) .
- b. 1. Backshifts, (0.25) I J
- 2. Weekerids , (0.25)
- 3. Holidays. (0.25)
REFERENCE:
PVNGS CONDUCT OF OPERATIONS 40AC-92ZO2, Pg. 9 of 45.
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l 8.02 OUESTION (1.5) i l
In accordance with Technical Specification 3/4 7.11 Fire ;
Suppression Water System operability. i a, How many fire suppression pumps are required for the fire suppression system to be operable? (0.5)
- b. How many water supply tanks are required for(0.5) the fire supression system to be considered operable?
- c. What is the minimum volume (or feet) of water required in each tank for the system to be considered operable?
(0.5) l l
8.02 ANSWER (0.5) .)
- a. Three (fire suppression pumps) . \
J I
(0.5)
- b. Two separate water supply tanks.
(0.5)
- c. Volume of 300,000 gallons each (23 feet).
REFERENCE:
PVNGS TECNICAL SPECIFICATION 3/4.7.11, FIRE SUPPRESSION WATER SYSTE11.
I
t 8.03 OUESTION (3.5) l While moving fuel in MODE 6: 'l j
- a. What conditions must met f or containment integrity in j accordance with Technical Specification 3/4.9.47 (2.0) i flux
- b. What are the requirements f or source range neutron monitors in accordance with Technical Specification 3/4.9.2? (1.5) 8.03 ANSWER a.
- 1. The equipment door closed and held in place by a minimum(0.5) of four bolts, l (0.5)
- 2. A minimum of one door in each airlock is closed,
- 3. Each penetration providing direct access from the be containment atmosphere to the outside atmosphere shall either:
1.
c?osed (by an isolation valve, blind flange, or manual (0,5) valve or),
ii. be capable of being closed by an operable automatic containment purge valve. (0.5) b.
- 1. Two source range neutron flux monitors shall be operable(0,5) and operating, indication in the control room,
- 2. each with continuous visual (0.5)
- 3. one source range neutron flux monitor audible indication(0.5)in the control room and containment.
REFERENCE:
PVNGS TECHNICAL SPECIFICATION 3/4.9.4 AND 3/4.9.2.
8.04 QUESTION (2.5)
Technical Specification 3/4.7.2. places limits on Steam Generator secondary coolant pressure and temperature.
- a. What i s the minimum secondary coolant - temperature f or a Steam Generator pressure above 230 psig? (0.5)
- b. What immediate action is required if the minimum temperature for 230 psig steam ganerator pressure is not met? (1.0)
- c. What is the basis f or the Technical Specifications Steam Generator pressure / temperature limitations?
(1.01 8.04 ANSWER
- a. 120 F (minimum secondary coolant temperature for 230 psig (0.5) steam generator pressure).
to
- b. Reduce the steam generator pressure to less than or equal (1.0) 230 psig within 30 minutes.
j
- c. Prevent brittle fracture of the Steam Generator vessel . (The limitation on steam generator pressure and temperature l ensures that the pressure induced stresses in the steam cenerators do not exceed the maximum allowable fracture toughness stress limits. The limitations to 120 F and 230 psig are based on steam generator RTNDT f 40 F and are sufficient to prevent brittle fracture.) (1.0)
REFERENCE:
PVNGS TECHNICAL SPECIFICATION 3/4.7.2, Pg. 3/4 7-11 0 TECHNICAL SPECIFICATION BASES 3/4.7.2, Pg. B 3/4 7-3.
.g
t 0.05 OUESTION (1.0) dance with -
(1.0)
What Technical is the Minimum Licensed Crew Composition in accorS 8.05 ANSWER ,
(0.5) 1 Shift Supervisor, (0,5)
{
i 1
Reactor Operator
, ADMINISTRATIVE
REFERENCE:
PVNGS TECHNICAL SPECIFICATION CONTROLS, Pgs. 6-1 L 6-5.
l i
1 1
i
)
8.07 QUESTION (2.0)
Techni c al Specification 3.1.1.4 provides minimum temperatures f or criticality.
- a. What is the minimum temperature of T cold f r criticality? (1.0)
- b. What ection must be taken in made 1 if you are below this l i mi t? (1.0) 8.07 ANSWER (1.0)
- a. 552 F
- b. Restore T to within its limit within 15 minutes (0.5) or be 1 in HOT STkNDBY within the ne>:t 15 minutes (0.5).
REFERENCE:
TECHNICAL SPECIFICATION LCO 3.1.1.4, Pg. 3/4 1-6 AND BASES 3/4.1.1.4, Pg. B 3/4 1-2.
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8.08 OUESTION (1.0)
- Delete Part b of Question!*.._
You have entered a Site Area Emergency, with minimum Technical Specification Shift manning, Who can fill the position of Control Room Supervisor I
a.
during an emergency? (0.5) ) 1 i'
Del ete
- b. Who is the person responsible f or making notifications over the unit page and the satellite technical support )
center (STSC) or control room radio circuit of transients and ESFAS actuations? (0.5) f 8.08 ANSWER i I
(0.5)
- a. The assistant shift supervisor or the shift supervisor.
(0.5)
- b. The Oper ations Technician (OPS TECH)
- Delete
- Pgs, 20 8< 20a of 45.
REFERENCE:
PVNGS PROCEDURE NO. 40AC-9ZZO2 i
1 l
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)
i 8.09 OUESTION (3.0)
In accordance with procedure 40AC-9ZZ15 Station Tagging and Clearance Procedure.
What are the purposes f or hanging the f ollowing types of tags? l Red Danger Tags. (1.0) a.
Yellow Caution Tag. (1.0) b.
Blue Men at Work Tag. (1.0) c.
0.09 ANSWER a.
A red danger tag shall be placed on equipment (mechanical would or el ectri cal ) which, if the equipment was operated, endanger personnel or damage equipment. (1.0)
A yellow caution tag is utilized where no danger to personnel is involved, state specific information which 6.
shall be understood and observed before operating equipment. !
)
(1.0) c.
A blue men at work tag is to be used only for trouble (maintenance work / repair is not shooting or testing, allowed.) (1.0) system components to l
( A blue men at work tag is placed onis to be operated only by the l indicate that the equipment to whom the tag i s i ss'u ed , after j direction of thehas individual obtained permission f rom the resposible that individual supervisor in order to ensure operation of the equipment does not affect the systems operation.) i PVNGS PROCEDURE No 40AC-9ZZ15, Pgs. 6A 8< 6B.
REFERENCE:
(:
1 8.10 QUESTION (3.0) i l
Technical Speci f i cation 3/4.4.5.2 pl aces limits on Reactor j Coolant Leakage : i j
- a. What is the bases f or limiting one steam generator leakage to (1.0) 720 gallons per day? 1 i
- b. What is the limit f or reactor coolant system identified (0.5) l leakage? l l
- c. What is the basis f or Reactor Coolant identified leakage?(1.0) l (RCS)
- d. What is the Leakage limit through Reactor Coolant 2250system psia? (0.5) pressure isolation valves at RCS pressure of 8.10 ANSWER The 720 GPD leakage limit per steam generator ensures that
- a. of steam generator tube integrity is maintained in the event (1.0) a main steam line rupture or under LOCA conditions.
(0.5)
- b. 10 gpm
- c. The 10 GPM identified leakage limit provides allowances for a limited amount of leakage from known sources whose presence will not interfere with the detection of unidentified leakage (1.0) ,
by the leakage detection systems. l (0.5)
- d. 1.0 GPM
REFERENCE:
PVNGS TECHNICAL SPECIFICATION BASES 3/4.4.5.2, Pg. B 3/4 4-4.
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1 8.11 DUESTION (3.5)
I I
Unit 1 is operating at 100% power. The "A" Diesel Generator is l l
tagged out because of a voltage regulator failure that occured l during a recent surveillance test. The facility experiences a loss of offsite power due to lightning strikes in the switchyard. l l
After the reactor trip it becomes apparent that a Steam Generator tube is ruptured and that one Charging pump is just keeping up l with the inventory loss. The Containment Mini-Purge dampers were open for pressure equalization when the event occured and the supply dampers won't close.
USING the attached appendix A of EPIP-02 " Emergency
)
Classification" :
- a. Which items would be checked when evaluating the event above? (1.5)
I
- b. How would you classify the event? (1.0)
How would you classify the event, if in addition 2 Core Exit Thermocouple were reading 735 F during the event?
c.
}
(1.0)
(Page 1 of Appendix A to EPIP-02 pr ovided. )
0.11 ANSWER (0.5)
- a. Physical Breach of Containment 10 GPM primary to secondary leak concurrent with loss (1.0) of condenser vacuum.
(1.0)
- b. Site Area Emergency (1.0)
- c. General Emergency REFERENCE : EPIP-02 Appendix A and C.
FO R I v e " C '"~"! n " ' ^ N LY . '
WORKING COPY , SRO EXAM 1
QUESTION 8.11 PVNGS EMERGENCY PLAN QCEDURE APPENDI.X A IMPLEMENTING PROCEDURE EPIP-02 Pay,e 2 of 4 REVISION EMERGENCY CLASSIFICATION 5 Page 12 of 29 I
i Appendix A Table 1 - Barrier Cha11ence/ Failure Classification Criteria ..
i 1.0 ~
Make s, Checks by any of the.following conditions that currently exist.
, , . ' CLAD CONTAINHEAT-
.RVI2!S indicates ATk'S
, voiding in upper Physical breach plenum of containment RCS pressure Excessive RCS E . l
> 2750 psia Activity (> 300 CIAS required but uc/gm dose equiva- not completed (i.e.
Uncontrolled loss lent I-131)
- of RCS inventory both automatic j
> 50 gpm valves in a pene- !
CET > 700 F tration fail to close)
H 2* "**"*#8*I "
> 3.5'.' by volume Containment pressure
. > 50 psig )
i Vital Auxiliaries / Radiation Release h~
Total loss of offsite and onsite AC power (Blackout)
Total than 60loss of offsite and onsite AC power (Blackout) for longer minutes .
. Loss of all Class IE DC power. .
Loss of all Class.IE DC power for longer than 15 minutes.
Failure of ESF Safety Systems (both trains) .to actuate when requir (Diesel Generator start failures excluded)
> 10gpm primary / secondary leakage concurrent with loss of condenser vacuum g . .
> 10 gpm primary /secoridary leakage concurrent with loss of secondary coolant outside containment
'^\
1 4
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8.12 OUESTION (1.0)
You are informed that the CD, systems in the battery room train A ones BA and BB are (channel A). and train D (channel B) inoperable due to a maintenance error.
What action must be taken to remain in conformance with Technical ]
Specification 3.7.11.3, CO 2 Systems?
8.12 ANSWER Within one hour establish a contiuous fire watch with backup fire systems f
j suppression equipment f or those areas in which redundant or components could be damaged. (For other areas establish an (1.0) hourly fire watch).
j
REFERENCE:
pvt 4GS TECHNICAL SPECIFICATION 3.7.11. 3, Pg.3/4 7-35.
i 1
9
0.13 OUESTION (1.0) _
Procedure 40AC-92Z15. Station Tagg2ng and Clearance, governs clearances on equipment and systems in the plant.
- a. Who i s the Responsible Supervisor for tagging and clearances of equipment in the following areas:
- 1. 525 KV Switchyard? (0.25)
- 2. Technical Support Center? (0.25)
- 3. Site Security Computer? (0.25)
- b. Who is the Responsible Supervisor for tagging and clearances for the Diesel Driven Fire Pump, FPN-Poib. (0.25) 8.13 ANSWER a.
- 1. (Used to be Systems Operation Supervisor, SOC Supervisor)
Transmission Control Center Supervisor. (0.25)
- 2. Stations Service Supervisor. (0.25)
- 3. Unit 1 Shift Supervisor. (0.25)
- b. Unit i Shift Supervisor. (0.25)
REFERENCE:
40AC-9ZZ15, Pgs. 7'AND 14c, Station Tagging and Clearance, Rev. O and Rev. 1 (Revi si on 1 issued January 29,1987, was supplied as a reference after the e>:ami nati on ) .
END OF EXAMINATION
i
',' U.S. Nuclear Regulatory Commission Reactor Operator License Examination Facility: PALO VERDE NGS ,
Reactor Type: PWR - CE80 ,
Date Administered: APR/L soy /487 l Examiner: PHILIP MORRIll Candidate: - - - - K E V- .
. j 3
INSTRUCTIONS TO CANDIDATE: J 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 l will be picked up six (6) hours after the examination starts. -
% of Category X of Candidate's Category i Value Total Score Value Category 15.25 25 09- _ 1. Principles of ;L: lear Power .
Plant Operation, Thermodynamics, Heat Transfer and Fluid Flow 19.15 -
25 M- 2. Plant Design including Safety and Emergency Systems -
25.15 .
i 25 Fr 3. Instruments and Controls M SMs 46 05 4. Procedures - Normal, Abnormal, 1 Emergency, and Radiological Control W -
TOTALS Final Grade % I l
All work done on this examination is my own. I have neither given nor received aid.
~ ~ - - KE y - --
i' Candidate's Signature I
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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 s' evere , penalties.
- 2. Restroom trips are to be Ifmited 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 fac1Ittate legible repro ~duYtfins7,
'~ ,
- 4. Print your name in the' blank provided on the cover sheet of the J' 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.
Consecutive appropriate 1y' number each answer sheet, write "End of Category " as "
of the paper,, and write "Last Page" on the lastn,answer sheet. start. each c 1.
Number each answer as to category and number, for example,1.4, 6.3.
- 10. Skip at least three lines between each answer. .
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
, {
- 12. Use abbreviations only if they are commonly used in facility literature.
- 13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
~
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE -
QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
- 16. If parts of the examination are not clear as to intent, ask questions of the examiner only. *
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given t.ssistance in !
completing the examination;. Jhis must be done after the examination has l been completed. j l
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1B.WhenyoucoEileteyour' examination,youshall: .
- n. . , , .
- a. Assemble your examination as follows:
_,. (1) Ex,am questiens.on, top. ,,
~
~
(2) Exam aids . figures, tables, etc.
(3) Answer pages including figures which are part of the ansver. .
, b. Turn in your copy of the examination and all pages used to answe.r the examination questions. ~ -* *-- - ~- '~'- " *--'
i's c. Turn in all scrap paper and the balance of the paper that you d!d f
not use for answering the questions. - -
c : .
.. ^..s .. . c o .-
- d. Leave the ex.amination"a... .. rea,'. as . . , ., defined b'y the,..... exam iner. If after leaving, you are found in this area while the examination is stili -
in progress, your ifcense may be denied or revoked. .
- 7. . . . . . . ycer rior:... . . .a:.r i .f ;'-, . , , . . , ...
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EQUATION SHEET f: f = ma v = s/t w = mg 2 Cycle efficiency = ** # "
a = v,t + at Energy (in)
E = mC a = (vg - v9 )/t KE = my vg=v + at A = AN A=Aeg PE = mgh a = 0/t A = In 2/1 = 0.693/tg W = vaP AE = 931Am t g(eff) = (t,)(ts) i (t +t) b Q = [nc AT Qch 6b. I=Ie -IX Q=U T k Qkke'Yum) I.= I e -UX Pwr = Wf In I=I O lo -XM
, P=P 10 SUR(t)' TVL = 1.3/u p=p o
et/T HVL = 0.693/u SUR = 26.06/T T = 1.44 DT SCR = S/(1 - K,ff)
SUR = 26 /Aeffp) CR = S/(1 - K,ff )
g T = '(1*/p ) + [(f 'p)/xeff) D '
T = L*/ (p - D N
- IIII ~ Ecff) = CR /CRg g
"( ~ E)! eff M = (1 - K gg)O ( ~
eff 1 p = (K,gg-1)/K,gg = Eegg/Keff SDM = (1 - K,gg)/K,ff p=
[L*/TK,gg) + (E/(1 + AdfT )] .
f,* = 1 x 10 sec ds P = I4V/(3 x 1010) A, g = 0.1 seconds I = No Idlg=1d22 WATER PARAMETERS Id =Idyy g
1 gal. = 8.345 lbm 2 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 .
Density = 62.4 lbm/ft 10 1 Curie = 3.7 x 10 dps Density = 1 gm/cm 3 1 kg = 2.21 lbm Heat of va;orization = 970 Etu/lba 1 hp = 2.54 x 10 3BTU /hr Heat of fusica = 144 Btu /lbm 1 Hw = 3.41 .x 10 Btu /hr 6
1 Atm = 14,7 psi = 29.9 in, l~g. I BLu = 778 ft-Ibf 1 ft. H O = 0,433*i lbf/in 2 1 inch = 2.54 cm .
F = 9/5 C + 32 "C = 5/9 ( F - 32)
1 l
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, Properties of Saturated Steam and Saturofed Water
- Absolute Pressure Vacuum Temper- Heat of Latent Heat Total Heat Specific Volume Lbs. per inches Inches ature the of of Steam p Sq.In. of Hg of Hg Liquid Evaporation P, t b Water i Steam ru,.. r. siuns neuns. nions. cu. vi. c., is. I cu ri . ,,, is.
0.0087 0.02~ 29.90 32.018 0.0003 1075.5 1075.5 0.016022 3302.4 0.10 0.20 29.72 35.023 3.026 1073.8 1076.8 0.016020 2945.5 0.15 0.31 29.61 45.453 13.498 1067.9 1081.4 0.016020 2004.7 0.20 0.41 29.51 53.160 21.217 1053.5 1084.7 0.016c25 1526.3
) 0.25 0.51 29.41 59.323 27.382 1060.1 1087.4 0.016032 1235.5 l 0.30 0.61 29.31 64.484 32.541 1057.1 10* 7 0.016040 1039.7 0.35 0.71 29.21 68.939 36.992 1054.6 It A - 0.016048 898.6 0.40 0.81 29.11 72.869 40.917 1052.4 109L 0.016056 792.1 0.45 0.92 29.00 76.387 44.430 1050.5 10945 0.016063 708.8 0.50 1.02 28.90 79.566 47.623 1048.6 1096.3 0.016071 641.5 0.60 1.22 28.70 85.218 53.245 1045.5 1098.7 0.016065 540.1 0.70 1.43 28.49 90.09 58.10 1042.7 1100.8 0.016099 466.94 0.80 1.63 28.29 94.38 62.39 1040.3 1102.6 0.016112 411.64 0.90 1.83 28.09 98.24 66.24 1038.1 1104.3 0.016124 368.43 1.0 2.04 27.88 101.74 69.73 1036.1 1105.8 0.016136 333.60 1.2 2.44 27.48 107.91 75.90 1032.6 1108.5 0.016158 280.96 1.4 2.85 1 ?7.07 113.26 81.23 1029.5 1110.7 0.016178 243.02
. 1.6 3.26 '
26.66 117.98 85.95 1026.8 1112.7 0.016196 214.33 1.8 3.66 26.26 122.22 90.18 1024.3 1114.5 0.016213 191.85 2.0 4.07 25.65 126.07 94.03 1022.1 1116.2 0.016230 ' 173.76 2.2 4.48 25.44 129.61 97.57 1020.1 111f.6 0.016245 158.87 h
2.4 4.89 23.03 132.88 100.84 1018.2 1119.0 , 0.016260 146.40 2.6 5.29 24.63 135.93 103.88 1016.4 1120.3 0.016274 135.80 2.8 7.70 24.22 138.78 106.73 1014.7 1121.5 0.016287 126.67 3.0 6.11 23.81 141.47 109.42 1013.2 1121 5 0.016300 118.73 3.5 7.13 22.79 147.56 115.51 1009.6 1125.1 0.016331 102.74 4.0 8.14 21.78 152.% 120.92 1006 4 1127.3 0.016358 90.64 4.5 9.16 20.76 157.82 125.77 1003.5 1129.3 0.016384 83.03 5.0 10.18 19.74 162.24 130.20 1000.9 1131.1 0.016407 73.532 5.5 11.20 18.72 166.29 134.26 998.5 1132.7 67.249 6.0 0.016430 12.22 17.70 170.05 138.03 9%.2 1134.2 0.016451 61.984 6.5 13.23 16.69 173.56 141.54 994.1 1135.6 0.016472 57.506 7.0 14.25 15.67 176.84 144.83 992.1 1136.9 53.650 7.5 0.0l M 91 15.27 14.65 179.93 147.93 990.2 1138.2 0.016510 50.294 8.0 16.29 13.63 182.86 150.67 988.5 1139.3 8.5 0.014527 47.345 .
17.31 12.61 185.63 153 b5 986.8 1140.4 0.016545 44.733 I 9.0 18.32 11.60 188.27 156.30 985.1 1141.4 8 9.5 0.016561 42.402 19.34 10.58 190.80 158.84 983.6 1142.4 0.016577 40.310 10.0 20.36 l 9.56 193.21 161.26 982.1 1143.3 0.016592 38.420 11.0 22.40 7.52 197.75 165.82 979.3 1145.1 0.016622 l
12.0 35.142 .
24.43 5.49 201.% 170.05 976.6 1146.7 0.016650 32.394 13.0 26.47 3.45 205.88 174.00 974.2 1148.2 14.0 0.016676 30.057 38.50 1.42. 209.56 177.71 971.9 1149.6 0.016702 28.043 Pressure Temper. Heat of Latent Heat Total Heat Specific Volume !
Lbs. per Sq. In. atur the Absolute Gage Liquid of Evaporation of Steam y .-
j P' P t
rwem. r. nio ns.
b Water Steam niuris. neuris. cu. vi. per is. co. re. ,,, is.
14.696 " 0.0 212.00 180.17 970.3 1150.5 0.016719 26.799 15.0 0.3 213.03 181.21 %9.7 1150.9 0.016726 16.0 26.290 1.3 216.32 184.52 %7.6 1152.1 0.016749 17.0 24.750 2.3 219.44 187.66 %5.6 1153.2 0.016771 23.385 18.0 3.3 222.41 190.66 %3.7
- 1154.3 0.01(793 19.0 22.168 4.3 225.24 193.52 961.8 11?5.3 0.016814 21.074 20.0 5.3 227.96 196.27 %0.1 1156.3 0.016634 20.087 21.0 6.3 230.57 198.90 95F 4 1157.3 0.016854 22.0 19.190 7.3 233.07 201.44 956.7 !!58.1 0.016873 18.373 23.0 8.3 235.49 203.88 l 955.1 1159.0 0.016891 17.624 !
24.0 9.3 237.62 206.24 953.6 1159.8 0.016909 16.936 l 25.0 10.3 240.07 26.0 11.3 242.25 208.52 210.7 952.1 950.6 1160.6 0.01692? - 16.301 1161.4 0.016944 15.7138 27.0 12.3 244.36 212.9 949.2 1162.1 0.016 % 1 -
28.0 15.1684 13.3 246.41 214.9 947.9 1162.8 0.016977 14.6607 29.0 14.3 248.40 217.0 946.5 1163.5 0.016993 14.1869 30.0 15.3 250.34 218.9 945.2 lib 4.1 0.017009 31.0 13.7436 16.3 252.22 220.8 943.9 1164.8 0.017024 13.3280 32.0 17.3 254.05 222.7 942.7 1165.4 0.017039 33.0 12.9376 18.3 255.84 224.5 941.5 1166.0 0.017054 12.5700 34.0 19.3 257.58 226.3 940.3 1166.6 0.017069 12.2234
'. Prbpsrtits cf St.turetad Stam end Scturntad Watsr-centinusd
~
Pressure Temper. Heat of Latent Heat Total Heat g Ll>s. per Sq. In. Specific volume ature the of 5 .
Absol,ute Gage t. Liquid Evaporation of Steam p 3 i
P P- o n=. r. neu ns. niu ns. nions. cu. n. weis. cu. n. m, m 35.0 20.3 259.29 228.0 939.1 36.0 1167.1 0.017063 11.8959 21.3 260.95 229.7 938.0 37.0 22.3 1167.7 0.017097 11.5860 262.58 231.4 936.9 . 1168.2 38.0 23.3 0.017111 11.2923 264.17 233.0 935.8 1168.8 39.0 24.3 0.017124 11.0136 265.72 234.6 934.7 1169.3 40.0 0.017138 10.7487 25.3 267.25 236.1 933.6 1169.8 41.0 26.3 268.74 0.017151 10.4965 237.7 932.6 1170.2 0.017164 42.0 27.3 270.21 239.2 10.2563 931.5 1170.7 0.017177 10.0272 43.3 28.3 271.65 240.6 930.5 44.0 29.3 1171.2 0.017189 9.8083 273.06 242.1 929.5 1171.6
~ ~ 45.0 0.017202 9.5991 3
30.3 274.44 143.5 928.6 1172.0 i
46.0 31.3 275.80 0.017214 9.398b I 244.9 927.6 1172.5 0.017226 47.0 32.3 277.14 246.2 9.2070 926.6 1172.9 0.017238 48.0 33.3 278.45 247.6 9.0231 49.0 925.7 1173.3 0.017250 B.8465 34.3 279.74 248.9 - 924.8 1173.7 0.017262 8.6770 50.0 35.3 281.02 250.2 923.9 51.0 1174.1 0.017274 8.5140 36.3 282.27 251.5 923.0 52.0 37.3 1174.3 0.017285 8.3571 283.50 252.8 922.1 1174.9 53.0 38.3 284.71 0.017296 8.2061 254.0 921.2 1175.2 0.017307 54.0 39.3 285.90 255.2 8.0606 920.4 1175.6 0.017319 7.9203 55.0 40.3 287.08 256.4
- 919.5 1175.9 0.017329 7.7850
- 56.0 41.3 288.24 257.6 918.7 57.0 1176,.3 0.017340 7.6543 42.3 289.38 258.8 917.8 54.0 43.3 1176.6 0.017351 7.5280 290.50 259.9 917.0 1177.0 s 59.0 44.3 291.62 0.017362 7.4059 j 261.1 916.2 1177.3 60.0 45.3 0.017372 7.2879 292.71 262.2 915.4 1177.6 61.0 46.3 293.79 0.017383 7.1736 263.3 914.6 1177.9 0.017393 62.0 47.3 294.86 264.4 7.0630 63.0 913.8 1178.2 0.017403 6.9558 48.3 295.91 265.5 913.0 6d.0 49.3 1178.6 0.017413 6.8519 2 % .95 266.6 912.3 1178.9 0.017423 6.7511 65.0 66.0 50.3 ~ 297.98 267.6 911.5 1179.1 0.017433 6.6533 Nl 51.3 298.99 268.7 910.8 i 67.0 52.3 1179.4 0.017443 6.5584 299.99 269.7 910.0 1179.7 1'
68.0 53.3 300.99 0.017453 6.4662
- 270.7 % 9.3 1180.0 0.017463 69.0 54.3 301.% 271.7 6.3767 908.5 !!80.3 0.017472 6.2896 70.0 55.3 302.93 272.7 !
71.0 907.8 1180.6 0.017482 6.2050 56.3 303.89 273.7 907.1 72.0 57.3 1180.8 0.017491 6.1226 304.83 274.7 9%.4 1181,1 73.0 58.3 305.77 0.017501 6.0425 275.7 905.7 1181.4 74.0 59.3 306.69 276.6 0.017510 5. % 45 905.0 1181.6 0.017519 5.8885 )
75.0 60.3 307.61 277.6
~ t 76.0 904.3 1181.9 0.017529 61.3 308.51 278.5 903.6 5.8144 77.0 62.3 1182.1 0.017538 5.7423 309.41 279.4 902.9 1182.4 78.0 63.3 310.29 0.017547 5.6720 280.3 902.3 1182.6 0.017556 79.0 64.3 311.17 281.3 5.6034 '
901.6 1182.8 0.017565 5.5364 80.0 65.3 312.04 282.1 j 81.0 900.9 1183.1 0.017573 66.3 312.90 283.0 900.3 5.4711 .
82.0 67.3 1183.3 0.017582 5.4074 313.75 283.9 899.6 !!B3.5 83.0 68.3 314.60 0.017591 5.3451 284.8 899.0 1183.8 84.0 69.3 315.43 285.7 0.017600 5.2843 898.3 1184.0 0.017608 85.0 70.3 316.26 5.2249 266.5 897.7 1184.2 0.017617 86.0 71.3 317.08 5.1669 87.0 267.4 897.0 1184.4 0.017625 i 72.3 317.89 288.2 8%.4 5.1101 88.0 73.3 1184.6 0.017634 5.0546 318.69 289.0 895.8 1184.8 89.0 74.3 319.49 0.017642 5.0004 289.9 895.2 1185.0 90.0 75.3 0.017651 4.9473 320.28 290.7 894.6 1185.3 91.0 76.3 321.% 0.017659 4.6953 92.0 291.5 893.9 1185.5 0.017667 77.3 321.84 292.3 893.3 4.8445 93.0 78.3 322.61
!!85.7 0.017675 4.7947 94.0 293.1 892.7 1185.9 79.3 323.37 0.017684 4.7459 293.9 .
892.1 1886.0 95.0 80.3 0.017692 . 4.6982 324.13 294.7 891.5 1886.2
%.0 81.3 324.88 0.017700 4.6514 97.0 295.5 891.0 1186.4 82.3 325.63 2%.3 0.017708 4.6055 98.0 890.4 1186.6 0.017716 83.3 326.36 297.0 8B9.8 4.5606 99.0 84.3 1186.8 0.017724 4.5166
- 327.10 297.8 889.2 !!87.0 100.0 65.3 0.0177J2 4.4734 327.82 298.5 688.6 1187.2 101.0 86.3 328.54 0.017740 4.4380 102.0 199.3 888.1 1887.3 87.3 329.26 300.0 0.01775 4.3895 103.0 887.5 1187.5 'O.01776
- 88.3 329.97 300.8 886.9 ' 4.3487 104.0 H9.3 1187.7 0.01776 4.3087 330.67 301.5 RB6.4
+
105.0 1187.9 0.01777 4.2695 90.3 331.37 302.2 885.8 106.0 91.3 1188.0 0.01778 - 4.2309 332.06 303.0 885.2 1188.2 107.0 92.3 332.75 0.01779 4.1931 108.0 303.7 R84.7 1888.4 93.3 333.44 304.4 0.01779 4.1560 109.0 884.1 1188.5 0.01780 94.3 334.11 305.1 883.6 4.1195 1188.7 0.01781 -
4.0837
I Praptrtiss cf Saturatzd Stam end Saturat:d Watzr-continusd Pressure Temper. Heat of 2 Latent Heat i Total Heat Specifie Volume
), Lbs. per Sq. In. sture the of of Steam p Absolute Gage I Liquid . Evaporation p's 5tcam P, P tw.. . r. neurin. niu rin. niurin. cu Wep.,r ate sw. t.u ri. n., in.
!!0.0 95.3 . 334.79 305.6 883.1 1868.9 0.01762 4.0484 111.0 %.3 335.46 306.5 882.5 1189.0 0.01782 4.0138 112.0 97.3 336.12 307.*., 882.0 1189.2 0.01783 3.9798 P 113.0 98.3 336.78 307.9 881.4 1189.3 0.01784 3.9464 114.0 99.3 337.43 308.6 880.9 1189.5 0.01785 3.9136 115.0 100.3 318.08 309.3 880.4 1869.6 0.01785 3.8813 116.0 101.3 338.73 309.9 - 879.9 1889.8 0.01786 3.8495 l 117.0 102.3 339.37 310.6 879.3 '
1189.9 01:1787 3.8183 118.0 103.3 340.01 311.3 878.8 1190.1 0.01787 3.7875 119.0 104.3 340.64 311.9 878.3 1190.2 0.01788 3.7573 120.0 105.3 341.27 312.6 877.8 1190.4 0.01789 3.7275 121.0 106 3 341.89 313.2 877.3 1190.5 0.01790 3.6983 122.0 . 107.3 342.51 313.9 876.8 1190.7 0.01790 3.6695 t 123.0 108.3 343.13 314.5 876.3 1190.8 0.01791 3.6411 124.0 109.3 343.74 315.2 875.8 1190.9 0.01792 3.6132
]
125.0 !!0.3 344.35 315.8 875.3 1891.1 0.01792 3.5857 126.0 111.3 344.95 316.4 874.8 1191.2 0.01793 3.5586 I
127.0 112.3 345.55 317.1 874.3 1191.3 0.01794 3.5320 128.0 113.3 346.15 317.7 873.8 1191.5 0.01794 3.5057 l 129.0 114.3 346.74 318.3 873.3 1191.6 0.01795 3.4799
, 130.0 115.3 347.33 319.0 872.8 1191.7 0.017 % 3.4544 131.0 116.3 347.92 319.6 872.3 1191.9 0.01797 3.4293 132.0 117.3 348.50 320.2 871.8 -1192.0 0.01797 3.4046 133.0 118.3 349.08 320.8 871.3 !!92.1 0.01798 3.3802 134.0 119.3 349.65 ._ 321.4 870.8 1192.2 0.01799 3.3562
- 135.0 120.3 350.23 312.0 870.4 1192.4 0.01799 3.3325 t 136.0 121.3 350.79 322.6
- 869.9 1192.5 0.01800 3.3091 137.0 122.3 351.36 323.2 869.4 1192.6 0.01801 3.2861 138.0 123.3 351.92 323.8 868.9 - 1192.7 0.01801 3.2634 139.0 124.3 352.48 324.4 868.5 1192.8 0.01802 3.2411 140.0 125.3 353.04 315.0 868.0 . .~
1193.0 0.01803 3.2190 1 141.0 126.3 353.59 325.5 867.5 1193.1 0.01803 3.1972 142.0 127.3 354.14 326.1 867.1 1193.2 0.01804 3.1757 143.0 128.3 354.69 326.7 866.6 1193.3 0.01805 3.1546 144.0 129.3 355.23 327.3 866.2 1193.4 0.01805 3.1337 145.0 130.3 355.77 327.8 865.7 1193.5 0.01606 3.1130 146.0 131.3 356.31 328.4 865.2 1193.6 0.01806 3.0927 147.0 132.3 356.84 329.0 864.8 1193.8 0.01807 3.0726 '
148.0 133.3 357.38 329.5 864.3 1193.9 0.01808: 3.0528 149.0 134.3 357.91 330.1" 863.9 1194.0 0.01808 3.0332 150.0 135.3 358.43 330.6 863.4 !!94.1 0.01809 3.0139 152.0 137.3 359.48 331.8 862.5 1194.3 0.01810 2.9760 ,
154.0 139.3 36D.51 332.8 861.6 1194.5 0.01812 2.9391
- 156.0 141.3 361.53 333.9 860.8 1194.7 0.01813 2.9031 , !
158.0 143.3 362.55 335.0 859.9 1194.9 0.01814 2.8679 i 160.0 145.3 363.5!, 336.1 859.0 1195.1 0.01815 2.8336
- l 162.0 147.3 364.54 337.1 858.2 1195.3 0.01817 2.8001 . i 164.0 149.3 365.53 338.2 857.3 1195.5 0.01818 2.7674 166.0 151.3 366.50 339.2 856.5 1195.7 0.01819 2.7355 168.0 153.3 367.47 340.2 855.6 1195.8 0.01820 2.7043 170.0 155.3 368.42 341.2 854.8 11 % .0 0.01821 2.6738 172.0 157.3 369.37 342.2 453.9 11 % .2 0.01823 2.6440 174.0 159.3 370.31 343.2 853.1 11 %.4 0.01824 2.6149 176.0 161.3 371.24 344.2 852.3 11 % .5 0.01825 2.5864 178.0 163.3 372.16 345.2 851.5 11 % .7 0.01826 2.5585 180.0 165.3 373.08 346.2 850.7 11 % .9 0.01827 2.5312 182.0 167.3 373.98 347.2 849.9 1197.0 0.01828 2.5045 184.0 169.3 374.88 348.1 849.1 1197.2 0.01830 2.4783 186.0 111.3 375.77 349.1 848.3 1197.3 0.01831 2.4527 188.0 173.3 376.65 350.0 847.5 1197.5 0.01832 2.4276 190.0 175.3 377.53 350.9 846.7 1197.6 0.01833 2.4030 192.0 177.3 378.40 351.9 845.9 1197.8 0.01834 2.3790 194.0 179.3 379.26 352.8 845.1 1197.9 0.01835 2.3554 1%.0 181.3 380.12 353.7 844.4 1198.1 0.01836 2.3322 198.0 183.3 380.% 354.6 843.6 1198.2 0.01838 2.3095 200.0 1b5.3 381.80 355.5 842.8 1198.3 0.01839 2.28728 205.0 190.3 383.88 357.7 840.9 1198.7 0.0184} 2.23349 210.0 195.3 385.91 359.9 839.1 1199.0 0.01844 2.18217 215.0 200.3 387.91 362.1 837.2 1199.3 0.01847 2.13315 220.0 205.3 389.88 364.2 835.4 1199.6 0.01850 2.08629 225.0 210.3 391.80 366.2 833.6 1199.9 0.01852 2.04143 230.0 215.3 393.70 368.3 831.8 1200.1 0.01855 1.99846 235.0 220.3 395.56 370.3 830.1 1200.4 0.01857 1.95725 240.0 225.3
- 397.39 372.3 828.4 1200.6 0.01860 1.91769
. 245.0 230.3 399.19 374.2 826.6 1200.9 0.01863 1.87970
_ Pr:pirti:s cf S::tur tad Staam and Saturated Water-conduded Pressure Temper- Heat of )
- 4. Lbs. per Sq. In. Latent Heat Total Heat Specific Volume sture the of of Steam l Absolute Gage Uqu hap ratbn p t gp P' P tw..... r. niofis. Water 5 team niurie. neurin. cu. re. r , ib. cu ri. ceris.
250.0 235.3 400.97 376.1 815.0 255.0 1201.1 0.01b65 1.b4317 240.3 402.72 378.0 813.3 1201.3 260.0 245.3 0.01868 1.80802 404.44 379.9 821.6 1201.5 0.01870 265.0 250.3 4 % .13 1.77418 381.7 820.0 1201.7 0.01873 1.74157 270.4 255.3 407.80 383.6 818.3 1201.9 0.01875 1.71013 !
275.0 260.3 409.45 385.4 816.7 280.0 1202.1 0.01878 1.67978- !
265.3 411.07 387.1 815.1 1202.3 285.0 270.3 0.01880 1.65049 412.67 388.9 813.6 1202.4 0.01882 290.0 275.3 414.25 1.62218 390.6 812.0 1202.6 0.01885 1.59482 295.0 280.3 415.81 392.3 810.4 l 1202.7 0.01887 1.56835 300.0 285.3 417.35 394.0 808.9 3
320.0 1202.9 0.01869 1.54174 '
305.3 423.31 400.5 802.9 1203.4 340.0 325.3 0.01899 1.44801 428.99 406.8 797.0 1203.8 0.01908 360.0 345.3 1.36405 434.41 412.8 791.3 1204.1 0.01917 380.0 365.3 439.61 1.28910 418.6 785.8 1204.4 0.01925 1.22177 400.0 385.3 444.60 424.2 780.4 420.0 1204.6 0.01934 1.16095 405.3 449.40 429.6 775.2 1204.7 440.0 425.3 0.01942 1.10573 454.03 434.8 770.0 1204.8 0.01950 460.0 445.3 458.50 1.05535 439'.8 765.0 1204.8 0.01959 1.00921 480.0 465.3 462.82 444.7 760.0
- 1204.8 0.01967 0.96677 500.0 485.3 467.01 449.5 755.1 I204.7 0.01975 0.92762 520.0 505.3 471.07 454.2 750.4 540.0 525.3 1204.5 0.01982 0.89137 475.01 458.7 745.7 1204.4 0.01990 560.0 545.3 478.84 0.85771 463.1 741.0 1204.2 , 0.01998 0.82637 580.0 565.3 482.57 467.5 736.5 1203.9 0.02006 0.79712 600.0 585.3 486.20 471.7 620.0 732.0 1203.7 0.02013 0.76975 605.3 489.74 475.8 727.5 640.0 625.3 1203.4 0.02021 0.74408 493.19 479.9 723.1 1203.0 M0.0 645.3' 4 %.57 0.02028 0.71995 483.9 718.8 1202.7 5 0.02036 0.69724 680.0 665.3 499.86 487.8 -
714.5 1202.3 0.02043 0.67581 700.0 685.3 503.08 491.6 720.0 710.1 1201.8 0.02050 0.65556 705.3 506.23 495.4 706.0 1201,4 l 740.0 725.3 509.32 0.02058 0.63639 499.1 701.9 1200.9 0.02065 i
760.0 745.3 512.34 502.7 0.61822 1 780:0 697.7 1200.4 0.02072 0.60097 765.3 515.30 506.3 693.6 !
1199.9 0.02080 0.58457 800.0 785.3 518.21 509.h i
820.0 689.6 1199.4 0.02087 0.56896 805.3 521.06 513.3 685.5 840.0 825.3 1198.8 0.02094 0.55408 523.86 516.7 681.5 1198.2 860.0 845.3 526.60 0.02101 0.53988 -
880.0 520.1 677.6 1197.7 0.02109 865.3 529.30 523.4 0.52631 673.6 1197.0 0.02116 0.51333 900.0 685.3 531.95 526.7 j 920.0 669.7 11 % .4 0.02123 905.3 534.56 530.0 0.50091 ,
940.0 665.8 1195.7 0.02130 0.48901 925.3 537.13 533.2 661.9 960.0 945.3 1195.1 0.02137 0.47759 539.65 536.3 658.0 1I94.4 i 980.0 %5.3 542.14 0.02145 0.46662 '
539.5 654.2 1193.7 0.02152 1000.0 985.3 0.45609 544.58 542.0 650.4 1192.9 1050.0 1035.3 550.53 0.02159 0.446 %
550.1 640.9 1191.0 0.02177 1100.0 1085.3 556.28 0.42224 557.5 631.5 1189.1 0.02195 1150.0 1135.3 561.82 564.8 0.40058 1200.0 622.2 1187.0 0.02214 0.38073 1185.3 567.19 571.9 613.0 1184.8 0.02232 0.36245 1150.0 1235.3 572.38 1300.0 578.8 603.8 1182.6 0.02250 1285.3 577.42 585.6 0.34556 1350.0 594.6 1180.2 0.02269 0.32991 1335.3 582.32 592.2 585.6 1400.0 1385.3 1177.8 0.02288 0.31536 587.07 598.8 567.5 1175.3 .
1450.0 1435.3 591.70 0.02307 0.30178 605.3 567.6 1172.9 0.02327 1500.0 1485.3 0.28909 596.20 611.7 558.4 1170.1 1600.0 1585.3 604.87 0.02346 0.27719 1700.0 624.2 540.3 1164.5 0.02387 1685.3 613.13 636.5 0.25545 1800.0 522 2 1158.6 0.02428 0.23607 1785.3 621.02 640.5 503.8 1900.0 1885.3 1152.3 0.02472 0.21861 628.56 660.4 485.2 1145.6 2000.0 19MS.3 0.02517 0.20278 635.b0 . 672.1 466.2 1138.3 2100.0 2085.3 642.76 0.02565 0.18831 2200.0 683.8 446.7 1130.5 0.02615 2185.3 649.45 695.5 0.17501 2300.0 426.7 1122.2 0.02669
- 2285.3 655.89 707.2 0.16272 2400.0 23R53 406.0 1113.2 0.02727 0.15133 662.11 719.0 384.8 1103.7 2500.0 24k5.3 0.02790 0.14076
% 8.81 731.7 361.6 1093.3 2600.0 2585.3 673.91 0.02859 . 0.13068 2700.0 744.5 337.6 1082.0 2685.3 679.53 757.3 0.02938 0.12110' 2800.0 2785.3 312.3 1069.7 0.03029 0.11194 684.% 770.7 285.1 1055.8 2900.0 2885.3 690.22 0.03134 0.10305 785.1 254.7 1039.8 0.03262 0.09420 3000.0 2985.'J 695.33 3100.0 801.8 218.4 1020.3 3085.3 700.28 824.0 0.03428 0.08500 3200.0 31H5.3 169.3 993.3 0.03681 0.07452 705.08 875.5 56.1 3208.2 3I03.5 931.6 'O.04472 0.05663 705.47 906.0 0.0 906.0 0.05078
- 0.05078
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l
".' I SECTION 1 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER, AND FLUID FLOW l i
1 oQUESTION 1.01 (2.5) i Refer to Figure 1.1 which shows a reacti vity insertion into an already critical reactor core (starting at time t = 0). Source neutrons are not significant and the reactor is well below the point of adding heat. (Assume b = 0.007 and A = 0.1.) )
(a) On Figure 1.1 draw the resulting reactor start-up rate as a function of time. (1.0) l (b) On Figure 1.1 draw the resulting reactor log power as a function of j time. (0.5) '
(c) If the total positive reactivity insertion increased k(eff) to 1.003, what is the start-up rate as time approaches t = 5 7 j (All RO candidates were inf ormed during the exam to delete "as time j approaches t = 5 ?" and to replace it with "after the reactivity change ?"
l cANSWER (a) t< (b) See attached: (1.5)
(c) p= (K gg4 - 1)/K = (1.003 -1)/1.003 = .0030 (0.3) eff T= (b - p)/gp = (0.007 - 0.003)/(0.1 x 0.003) (0.3)
= 13.3 Sec.
SUR = 26/T = 1.95 Decades / Minute (0.4)
OREFERENCE Palo Verde, Licensed Operator Training, Reactor Theory, Lessons 10 and 11 -
Reactor Period.
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60UESTION 1.02 (2.0)
.A motor driven centrifugal pump is used in c. cl osed cooling system. The
-pump has two speeds, 1200 RPM and 1800 RPM.
(a) How does each of the f oll owing parameters vary when the pump is switched from fast to slow speed:
(1) . Pump dif f erential pressure ? (0.5)
(2) Pumping power ? -(0.5)
(b) For constant speed, why will an increase in.the temperature (1.0) of the fluid being pumped decrease the indicated motor Amps ?
OANSWER (a) (0.5 each)
( 1.) Dif f erenti al pressure is proportional to speed squared, (therf ore pump differential pressure will go down by a factor of 0.44).
(2) Pumping power i s proportional to speed to the third power, cubed, (therf ore power will go down by a factor of 0.30).
(b) As the temperature of the fluid being pumped increases its density and viscosity will decrease. This will decrease pumping power and therfore motor Amps will decrease. (1.0) oREFERENCE 1
Palo Verde, Licensed Operator Training, Section Two, Thermohydraulics Review, pges 18 - 20
l ODUESTION 1.03 (1.O)
As the core ages the ratio'of Fu-239 to U-235 atoms increases. For a constant positive reactivity addition, over the life of the core:
1 (a) What will be the effect of the increased Pu/U ratio on (0,5) start-up rate ?
(b) What term in the the start-up rate rel ationship (equation) (0.5) causes this ef f ect ?
{
i oANSWER
)
(a) Start-up rate (f or the same reactivity) will increase over (0.5) l core life.
(b) b or beta (Decreases over core age. Since reactor period is (0.5) proportional to b -p, for a given p the period will be i smaller and the SUR higher. I
- REFERFNCE Palo Verde, Licensed Operator Training, Section One, Lesson 17 , page 4 - 5 I
I l
l i
- )
kOUESTION 1.04 (2.0)
After the reactor is shutdown, the fuel can continue to generate subtantial amounts of heat.
(a) Right after a SCRAM why does nuclear power decrease at (1.0) about -1/3 DPM ? _
.(b) Several hours after a SCRAM, with indicated nuclear power (1.0) low in the source range, what is the principle source of heat I generation in the core 7 OANSWER (a) Right after a SCRAM the principle source of neutrons is the longest lived delayed neutron precursor (Br-87) which decays at -1/3 DPM.
(1.0)
(b) After several hours decay heat is due to the decay of f i ssi on products. (1.0)
OREFERENCE Palo Verde, Licensed Operator Training, Section One, Lessons 1 8< 2 l
l l
I e
. . . \m
4 oOUESTION 1.05 (2.0) l During an outage 15 % of all steam generator tubes are plugged. While at full power, prior to the outagg, the steam generator pressure'was 1062 PSIA. T is to remain at 593 after the tube plugging.
What is the (new) expected full power steam pressure 7 oANSWER O = UA AT = constant K).5)
Ux 100% x (593 - 552) = U x 85% x (593 - T) (0.5) 48.2 = 593 T T = 545 (0.5)
T(sat) is 545 therefore P(sat) is 1000 PSIA (0.5)
OREFERENCE Palo Verde, Licensed Operator Training, Section Two, Thermohydraulics i Review, page 14 -15 i
i
A 60UESTION 1.06 (2.0)
A reactor startup is being conducted 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after a reactor trip. The reactor had been operating at 100% power for 60 days prior to the trip. An i estimated cri ti cal rod position of 30 inches on group three has been calculated. . _ _
How would each of the f ollouing events (occuring one at a time) affect the actual critical rod position ?
(a) Steam oenerator level is raised significantly with (0,5) feedwater at 200 F during the start-up rod withdrawl.
(b) The startup is delayed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. (0.5)
(c) The steam dump pressure setpoint is increased. (0.5)
(d) Two reactor coolant' pumps are restarted just prior to rod (0.5)-
withdrawl ?
1 OANSWER I
)
(a) Rod position lower (due to decreased primary temperature) (0.5)
(b) Rod position higher (>:enon buil d-up ) (0.5)
I (c) Rod position higher (inc. eased primary -temperature) (0,5) j (d) No effect (0,5)
I OREFERENCE i 1
1 Palo Verde, Licensed Operator Training, Section One, NALO9, Chapter 10 on i ECP l l
I
~ 6 9
oDUESTION 1.07 (3.5)
The reactor is initially operating at 50% power, begining of life, 120 inches withdrawn on control group 5, baron concentration at 650 ppm. T l is programmed normally. Use the attached Figures 1.2, 1.3, 1.4, 1.5 an8#T.6 l to answer the following questions. Consider each situation separately. 'i (a) With no rod motion what change in baron concentration (1.5) 1 is necessary to increase power to 100% 7 l (b) What will be the rod insertion immediatly after a power (1.5) reduction to 25% without any boron concentration change ? i (c) What is equilibrum xenon at 100% power 7 (0.5) )
(be sure to indicate if it i s + or - reactivity. ) ;
OANSWER (a) At 50% BOL power defect = - 750 (+/-20) pcm, at 100% it is - 1380 (+/-20) pcm. (0.5)
Change in reactivity required = 630 pcm. g Baron worth = -11.5 (+/-0.1) pcm/ ppm @ 57C F (0.5) 630 pcm/(-11.5 pcm/ ppm) = - 55 (+/-4) ppm boron must be decreased 55 ppm (0.5)
(b) At 50% BOL power defect is -750 (+/-20) pcm At 25% it is -410 (+/-20) pcm; reactivity change is -340 pcm. (0.5)
Rod group 5 at 120 In. is 20 (+/-10) pcm on CEA Worth curve.
20 + 340 = 360 pcm (0.5) ;
I This is 15 (+/-10) In. on group 5 l Rods must be inserted to 15 steps on group 5 (0,5)
(c) 100% equilibrum xenon is - 2675 (+/-25) pcm. (0.5)
OREFERENCE
)
1 Palo Verde, Licensed Operator Training, Section One, NLC55, Chapter 13, 14, 15, and 16. j 1
9
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1 POWER DEFECT VS P0iFE.R LEVEL (BOC, MOC, EOC)
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- /. FULL PON8R LEGEND: EFPD 0 o+G 209.3 m 455.7 O
. FIGURE I.3 BORON WORTH VS T-A VG:
(70 -
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' CORE DATA BOOK UNIT 1 CYCLE 1 PVNGS/ NUCLEAR SECTION .
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- QUESTION 1.08 (3.O)
A startup is being conducted for PVNGS No. 2. Total rod worth is 15% K/K and the most reactive rod is worth 3.0%d K/K. With all rods fully inserted the reactor is shutdown by - 6.4%. The source range instruments indi cate 5 CPS.
When shutdown control rods are fully withdrawn the count rate is 25 CPS.
(a) What is the worth of the shutdown banks in %AK/k 7 (2.0)
(b) What is the " shutdown margin" of the reactor in %AK/K ? (1.0)
- ANSWER (a) CR(1)/CR(0) = p(o) >: (100% - p (1) ) /p (1) >: (100 p (o) ) (0. 5) p(o) = - 6.4%
25/5 = -6. 4 (100 - p ) /p (106. 4 ) : 532 >: p= 6. 4 >: p - 640 (1.0) p=- 1.22% therfore shutdown banks = 6.4 - 1. 22 = 5. 2% AK/K (0,5)
Alternate Method CR (1) p (1) = CR(2)p(2) p (2) = 5/24 (-6.4%) i p(2) = -1.28% A K/K :
Therefore Shutdown banks are worth: 6.4 - 1.28 = 5.12% 6 K/K (b) (The shutdown margin is the amount the reactor could be shutdown for a given condition and the most reactive rod stuck out.)
6.4% - 3% = 3.4% shutdown margin (1.0) i
- REFERENCE Palo Verde, Licensed Operator Training, Section One, NLC55, Chapter 1 and 8
'1 l
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\l "oOUESTION 1.09 (2.0)
A pressuri er relief valve and a steam generator relief' valve are each !
leaking to atmospheric pressure. The pressurizer i s at 2250 PSI A . (enthalpy is 1115 Stu/Lb) and the ster.m generator is at 900 PSIA (enthalpy is 1196L .
Btu /Lb). I i
(a) Show the leakage process for each valve on Figure 1.7 (1.0) l (Mollier Chart) Clearly label each valve. {
1 (b) For each valve, is the leakage superheated or saturated ? (1.0) oANSWER
.(a) See attached Figure: (1.0)
(b) Pressuri:er leakage is saturated, Steam generator leakage is (1.0) superheated.
OREFERENCE Palo Verde, Licensed Operator Training, Section Two, page 12
' Steam Tables
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1 1000 1335 1650 1965 2.300
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'oOUESTION 1.'10 (2.O)
Refer to attached Figure 1.8 which shows the cross section of a fuel rod.
(a) Draw the temperature profil e f rom the f uel pin center to the (1.0)
_ adjacent bulk coolant.
(b) Assuming the same power generation rate, why will the fuel (1.0) pin center line temperature decrease with core burnup ?
OANSWER (a) See attached Figure: s. 4 (b) As the reactor is operated the cladding creeps and the fuel (1 (')
pellets swell. This reduces the gap between the fuel and tne cladding, which reduces the gap delta T. Since moderator temperature is unchanged, the fuel pellet temperature decreases.
OREFERENCE Palo Verde, Licensed Operator Training, Section Two, page 5 - 6 Palo Verde, Licensed Operator Training, NLC55, Chapter 17 1
l 1
FICrVRE I.T d
L Fuel !8 8
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- 1. I, I g
i Radiol Distence ii 1 1I I
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8 Generation ,
38 8
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h Fuel I:
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60UESTION 1.11 (2.0)
A calorimetric is being conducted to calibrate the encore nuclear instruments. Consider ef f ects of each of the f oll owing si tuati onr. or errors one at a time.
-(a)- What will be the effect on calculated power if the feedwater (1.0) flow venturies are eroded (l arger internal diamater) 7 (b) What will be the ef f ect on calculated power if steam generator (1.0) blowdown is not considered 7 cANSWER (a) Calculated power will be lower than actual . (1.0)
(Indicated flow will be lower than actual.)
(b) If blowdown is not considered, calculated power will be too- (1.0) high. (Steam flow has higher enthalpy than blowdown)
OREFERENCE Pal o Verde, Licensed Operator Training, Dection Two, page 5 - 6 Training Article, NS-4, Excore nuclear Iristrumentation System, page 21 i
l
'oDUESTION 1.12 (1.0)
Lithium hydroxide is used as a PH control agent for the Reactor Coolant system. The purification ion exchangers are used +.o remove lithium early in core life.
Where does excess lithium come f rom early in core lif e ?
OANSWER Neutrons interact with the boron in the RCS to form lithium.. (Boron levels and hence lithium levels are high early in core lif e. )
OREFERENCE Training Article, NS-2A, Chemical and Volume Control System, page 66 END OF SECTION 1 GD ON TO SECTION 2 a
I w
4 SECTION 2 PLANT DESIGN, INCLUDING SAFETY AND EMERGENCY SYSTEMS oDUESTION 2.01 ('2. 0 ) .
A CREFAS or a CRVIAS signal shifts the control room ventilation mode f rom
" normal " to " essential ventilation".
(a) How is the control room habitability preserved by maintaining (1,0) a positive pressure in the control room f ollowing a CREFAS 7 (b) 'How is the control room ventilation system line-up dif f erent (1.0) following a CRVIAS-?
OANSWER (a) A positive pressure prevents potentially contaminated or (1.0) unfiltered air from entering.
(b) Following a CRVI AS the outside air isolation dampers (1.0)
(HJA/HJB - MO2/MO3) are closed to isolate the control room from the outside atmosphere.
OREFEREENCE PVNGS, Training Article AS-5A, Control Room Ventilation System.
i i
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)
oOUESTION 2.02.(2.5) 1 The Fire Protection Water System is required to be operable by Technical Specifications. ,
(a) ..What is the source of cooling _ water supplied to the ,.
(0,5) diesel driven fire pumps ?
\
l (b) With the fire water pumps in standby, what are the (1.0) {
two normal ways fire protection water header pressure , i can be maintained ? '
(c) How is an automatic preaction sprinkler system activated ? (1.0)
(Two steps are required.)
OANSWER (a) The dAscharge of the cssociated fire pump supplies the (0.5) cooling water.
(b) Well water booster pumps ;or domestic water' system) or the (1.0) jockey pump maintain system pressure, b
(c) Smoke detcctors activate a deluae valve which fills and (1.0) pressurizes the piping. Melting the fusable link in the {
spri nkl er head (s) will then activate water sprayi I oREFEREENCE ,
PVNGS, Trai ni ng Arti cl e AS-20B, Fire Protection - Water System l
v 5 I 9 q I
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a 60UESTION 2.03 (2.0) I
- The Main Feedwater System utilizes the econominer valves, downtommer valves, and f eed pumps to maintain stable steam generator levels. '
(a) Why are the downcommer valves opened above 50% paper 7 (1.0)
'1
-(b) What percent of total flow is supplied through the (0,5) j downcommer valves above 50% power ? )
1 (c) What is the capacity of-one main fecd pump 7 (0,5) j (In terms of 0% to 100% system rated flow) !
OANSWER (a) This provides sub-cooling f or the recirculating water (1.0) and promotes increased recirculation ratio. (Also reduces carry-under.) i (b) 10% of total flow. (0.5)
(c) 65% of system rated flow ( 0. 5) ;
OREFERENCE i 1
PVNGS, Training Article PGS-10A, Main Feedwater System, page 11 and 21 '
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'oDUESTION 2 04 (3.0)
Following a reactor trip, plant design f eatures operate automatically to maintain forced RCS flow for as long as possible. 1 (a) After a unit trip how does the 13.8 KV electical system (1.0) automatically maintain power to the reactor cool ant pumps 7 (b) After a unit trip with loss of off-site power, what sequence (1,0) of automatic actions by the 525 KV and 13.8 KV electrical systems provide coastdown power to the reactor coolant pumps ? ,
(c) Following loss of power to all reactor coolant pumps, what (1.0) mechanical component of the RCPs increases pump coastdown time 7 0 ANSWER (a) The 13.8 KV buses are automatically transf ered from the unit (1.0) auxiliary transformers to off site power.
(b) The 525 KV unit breakers are tripped and all loads on the (1.0) 13.5 KV buses are tripped except the RCPs.
(c) The fl ywheel on the RCPs increases pump coastdown time. (1.0) oREFERENCE PVNGS, Training Article PGS-13A, Main Generation System, page 43 and NS-1A, Reactor Cooolant System, page 9 i
i
e UQUESTION 2.05 (2.0)
The 125 VDC and 125 VAC systems distribute electrical power to 125 V vital
.l oad r, .
(a) If the normal battery ~ charger fails, what are two sources (1.0)
- of power to the 125 VDC "B" distribution pannel ?
(b) What are the two sources of regulated 125 AC' power to 125 VAC (1.0) distribution pannel "C" ?
OANSWER (a) The B-D (backup) battery charger and'the "B" battery. (1.0)
(b) The "C" inverter and the 480/120 VAC voltage regulator. (1.0)
OREFERENCE PVNGS, Training Article PGS-15D and E, 125 VDC/125VAC Class IE Power System, Elementry Drawing 01-E-nna-OO1.
1 1
i
- . , , .s. , -- . u. . _ . . -
bOUESTION 2.06 (1.0)
The Essential Spray Pond System is operated during normal shutdown and emergency conditions.
What is the, purpose of the Spray Pond bypass headers ?
oANSWER
-The bypasses are used to minimize evaporation and drift. (Minimine losses when spray pond temperatures are suf ficiently low)
OREFERENCE PVNGS, Training Article PGS-BA, Essential. Spray Pond System, page 13, 17 and 26 I
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'50UESTION 2.07 (2.5)
The Essential Cooling Water (EW) System removes heat from selected safety related components during normal shutdown and emergency conditions.
1 (a)
Why is the EW system pressure maintained lower than the (0.5)
Essenti al Spray Pond System pressure ? .
(b) What are two sources of make-up water to the EW system ? (1.0)
(c) How does the 'End system automatically provide cooling to (1.0) ,
the Reactor Coolant Pumps after a loss of off-site power ?
OANSWER (a) This design prevents leakage from the EW to the SP system (0.5) J in the event of a EW heat exchanger leak. (minimize- )
environmental contamination) {
(b) Demineralized water. system and condensate _ storage tank (1.0) are sources of make-up.
(c) ( Af ter a LOP the loop A cross tie isolation valves (UV-65 (i.0) and UV-145) used to open to provide cooling to the RCPs.) 4 There are NO automatic provisions, an operator is required to manually manipulate the valves. 1 l
PVNGS, Training Article PGS-8B, Essential Cooling Water System, page 6, 7, and 23, Elementry Drawing C1-E-EWB-OO3.
t l
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.e oDUESTION 2.00 (2.0)
The Main Steam Isolation Val ves are hydrauli c-actuated , double disk, wedge
- valves.
~
(a) What system provides power for the MSIV hydraulic pum'p 7 (0.5)
(b) What feature of the hydraulic system provides fast closure (0.5) of an MSIV ?
(c) What is the " slow" mode of operation of the MSIVs used for 7- (1.0) 0 ANSWER (a) (Air from the) Instrument and Service Air System. (0.5)
(b) The hydraulic accumulators (0.5)
(c) Normal operation of the MSIVs and to test the MSIVs. (1.0)
OREFERENCE PVNGS, Training Article PGS-1A, Main Steam System, page 11 - 14
(
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Q oOUESTION 2.09 (1.5) ]
The Condenser Air Removal System removes non-condensable gasses from the ]
condenser. 1 1
(a) What automatic action occurs if high radiation is detected (0.5 in the normal condenser exhaust flow ?
(b) What three other plant systems must'be in. operation prior to '( 1. 0 )
starting the. Condenser Air Removal System ?
OANSWER (a) High radiation in the exhaust flow path will cause the (0.5) exhaust flow to divert to the exhaust filter unit.
1 (b) The Plant Cooling Water, and the Gland Steam Seal and Drain (1.0)
Systems must be in operation.
- REFERENCE PVNGS, Training Article PGS-9B, Condenser Air Removal System, page 27 - 28 1
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- QUESTION 2.10 (1.5) )
)
The motor driven and turbine driven auxiliary feedwater pumps are used I during startup, shutdown, hot standby, and emergency operations.
1 l
(a)' What is the back-up supply of water for the auxiliary f eed (0.5) pumps ? ;
I (a) What electrical system (voltage and control center) powers (1.0) j the turbine driven pump's steam supply and feed regulating '
valves ?
- ANSWER (a) The back-up water supply is the reactor make-up water tank (0.5) j (b) 125 VDC M-41 Control Center (0.66) 125 VDC M-43 Control Center (0.33) l PVNGS, Training Article PGS-11, Auxiliary Feedwater System, page 9-84 28 PGS iSD, 125 VDC IE Power Supply, page 10 I
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89 oQUESTION 2.11 (1.O) k The Chemical and Volume Control System letdown heat enchanger discharge temperature control system is out of service. You observe that the letdown ;
. temperature, however, is now 145 F and rising.
What three components should be isolated or bypassed to avoid damage ?
OANSWER (0.33 each)
The ion exchangers, the process radiation monitor, and the baronometer should be isolated.
OREFERENCE PVNGS, Training Article NS-2A, Chemical and Volume Control System, pages 74, 68, 69, and 7 l
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bDUESTION 2.12 (1.0)
A connection is provided f rom the discharge side of the CVCS charging. pumps to the High Pressure Safety Injection header and hot leg injection piping. l Other than an alternate charging flow path, what are the two main uses of this piping connection 7 oANSWER This connection is used to (1) check operation of the saf ety injection check valves while at power (0.5) and (2) to adjust boron concentration in the-safety injection tanks.(0.5)
OREFERENCE PVNGS, Training Article NS-34, Safety Injection and Shutdown Cooling Systems, page 22 i
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t h0UESTION 2.13 (1.0)
What are the normal and alternate power supplies (buses and voltage)' to charging-pump number three ?
OANSWER 480 VAC PGB-L36 and PGA-L35.
OREFERENCE Procedure 410P-1PG01, 480 VAC IE Switichgear, page 41 and'44, 410P-1CH 04 , Step 5.2.6.
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50UESTION 2.14 (2.0)
The Chemical and Volume Control System controls RCS chemistry and inventory.
(a) What is the purpose of the backpressure control valves (1.0)
_ located downstream of the let-down flow control valves ?
(b) Which two systems are needed to operate the gas. stripper ?' (1.0)
I cANSWER (a) The backpressure control valves prevent flashing in the (1.C) letdown heat exchanger.
(b) The Auxiliary steam and the Nuclear Service Water systems (1.0) I are needed. J OREFERENCE PVNGS, Training Article NS-2A, Chemical and Volume Control' System, pages i 74 9 68, 69, and 7
,1 END OF SECTION 2 GO ON TO SECTION 3 i
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SECTION 3 INSTRUMENTS AND CONTROLS l l
~~
- OUESTION 3.01 (2.5) ,
l Refer to Figure 3.1. With CEDMCS in manual and all other systems.in automatic the reactor is operating at 50% power when control bank 5 is wi th-drawn si>: inches. Sketch, on attached Figure.3.1, reactor power, T (f uel ) , T(hot), T(cold), and T(steam) versus time until new ~ final values are reached. Assuming no reactor trip occurs, show the peaks and trends of the paramaters. Numerical answers are not desired.
oANSWER See attached Figure 3.1 OREFERENCE Training Article, NS-9C, Reactor Regulating System
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-t bOUESTION 3.02 (3.0)
The Core Protection Calculators initiate reactor trips to prevent cladding damage or fuel melting.
(a) Which CPC trip has a projected calculation ? (0.5)
(b) What are the two " projected" parameters ? (1.0)
(c) What are six of the seven pl ant parameters f rom the plant (1.5) process systems required by the CPC calculations 7 oANSWER (a) Low DNBR (0.5)
(b) RCS pressure and RCS flow. (1.0)
(c) (0.25 each for any six)
Reactor power RCS flow or RCP speed T(cold)
T (hot )
Pressurizer pressure CEA position Deviation Penalty f actors OREFERENCE Training Article f4S-6B, Core Protection Calculator, page 1 -3
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oQUESTION 3.03 (3.0)
For each of the f ollowing sets of conditions, state which.(one or more) of the f ollowing systems should actuate. Consider each event one at a time for-PVNGS Unit 1.
SIAS ,
CSAS CIAS l MSIS RAS AFAS (A) A feedwater line break occurs which causes containment pressure (0.75) to increase to 7 PSIG, steam generator pressures to drop to 1000 PSIG and 900 PSIG, and RCS pressure to drop to 1800 PSIG.
(b) A LOCA occurs which causes containment pressure to increase to (0.75) 15 PSIG and RCS pressure to drop to 1000 PSIG.
(c) A High Level Over-ride occurs as steam generator levels go to (0.75) 93% narrow range and RCS pressure drops to 1900 PSIG.
(d) Following a reactor trip, steam generator 1 drops to 800 PSIG (0.75) and 55% wide range while steam generator 2 is at 1000 PSIG and 20% wide range. RCS pressure falls to 1750 PSIG.
OANSWER (a) SIAS, CIAS, and MSIS (0.75)
(b) SIAS, CIAS, MSIS, and CSAS (0.75)
(c) MSIS (0.75)
(d) AFAS, SIAS, CIAS, and MSIS (0.75)
OREFERENCE Training Article NS-7A, Engineered Safety Features Actuation System, pages 40-50 and 31-34 Technical Specifications page 3/4 3-25
'oOUESTION 3.04 (2.5)
When equipment f ailure or abnormal plant conditions are sensed, Automatic l Motion Inhibit (AMI) and Automatic Withdrawl Prohibit (AWP) signals are sent to the Control El ement Drive Mechanism Control System (CEDMCS) from h the_ Reactor Reguating System (RRS), y (a) What three parameters are compared for deviations by the (1.5) =
Reactor Regulating System to generate an AMI ? f (setpoints are not required)
(b) What two Reactor Ceolant System parameters will generate (1.0) en AWP ? (include setpoints) 1
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oANSWER (a) T(ave) deviation (1.5)
TLI or Turbine first stage pressure deviation Reactor power devi ati on (b) T(ave) > T(ref) by 6 F ' ( 1. 0)
High T(c), 575 F
- REFERENCE Trai ni ng Arti c.' e NS-9C , Reactor Regulating System, pages 4-8 and 29-30 Instrument Loop Diagram Reactor Coolant System 01-J-SFE-053 Rev. O.
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.s oOUESTION 3.05 (3.O)
The excore nuclear instrumentation system uses three types of detectors and four channels. Outputs from these detectors are used for plant control and saf ety f unctions.
(a) How does the start-up channel differentiate betwe'en gamma (1.0) and neutron flux ?
(b) 'Which channels' outputs are sent to the core protection (1.0) calculators 7 (c) Why do the "A" and "B" log-saf ety channels have their (1.0) preamplifiers outside containment ?
OANSWER (a) A pulse height discriminator is used to eliminate gamma flux. (1.0)'
(b) Linear safety channels (1.0)
(c) This allows the "A" and "B" channels to be qualified f or (1.0) post accident monitoring.
OREFERENCE Training Arti cle NS-4, Excore Nuclear Instrumentation System, pages 25-26
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- QUESTION 3.06 (3.5)
The Reactor Protection System initiates automatic protective actions during anticipated operational events and accidents. .
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(a) What reactor trip protects the reactor from a dilution (1.0) {
accident at low power ? (trip name and setpoint) {
l (b) What reactor trip protects the r eactor f rom a rod ejection (1.0) l accident ? (trip name and setpoint)
(c) What signals do the CEA Calculators send to the Reactor (0.5)
Protection System ?
(d) Which two reactor trips use a rate limited setpoint ? (1.0)
(setpoints not required)
- ANSWER (a) High log power trip at 1% (1.0)
(b) Varriable overpower trip at 110% or 9.8% w/10.6%/ Min. (1.0)
(c) Deviation penalty f actors (0.5)
(d) Low steam generator flow and variable over-power (1.0)
- REFERENCE Training Article NS-6, Reactor Frotection System, pages 10, 15, 23, 24, Technical Speci f i cati on Tabl e 2. 2-1.
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s bOUESTION 3.07 (2.0)
The Reactor Power Cutback System (RPCS) is designed to mitigate the ef f ect of a loss of load or loss of feedwater p, imp event. The Steam Bypass Control System (SBCS) is designed to maximize unit availability by accommodating load rejections.
(a) What two events will caune a quick-opening block of the (1.0) turbine bypass valves ?
(b) How is the SBCS control system affected by placing a SBCS (0.5)
" valve permissive switich" in " manual" ?
(c) Above what power level will CEAs be dropped into the core (0.5) by the RPCS assuming a load rejection to house loads occurs ?
OANSWER (a) Low RCS temperature following a reactor trip and loss of feed pump. (1.0)
(b) A valve permissive signal is generated regardless of the (0.5) plant conditions.(valve can be controlled during startup 8< shutdown or unless a loss of vacuum occurs or an emergency off.
signal is generated)
(c) above 75% power CEAs are dropped into the core (0.5)
OREFERENCE i Training Article NS-9B, Steam Bypass Control System, page 21 t< 15 and NS-9D, Reactor Power Cutback System, page 10-11 l
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l s i oDUESTION 3.08 (2.5) l Fol l owi ng a LOCA which actuates the Safety I n.i ec t i on and Containment Spray l systems, a spurious RAS signal i s recieved.
(a) What effect does the RAS signal have on the LPSI pumps ? (0.5) i (b) What effect does the RAS have on the CS pumps ? (0.5) J 1
(c) What effect does the RAS have on the HPSI minimum flow - ( 0. 5) I recirculation valves ?
1 (d) What two signals will actuate a RAS ? (1.0)
- ANSWER (a) The LPSI pumps will stop (0.5)
I (b) The CS pumps remain operating (no effect) (0,5)
(c) The recirculation valves go shut (0.5)
(d) Manual initiation and low refueling water tank level ( .1. 0 )
OREFERENCE Training Article NS-7A, Engineered Safety Features Actuation System, page 33 and Al arm Response Procedure 41 AL-1RK5B page 15 and 46 i
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bDUESTION 3.09 (3.0)
The Pressurizer Pressure and Level Control Systems automatically maintain pressurizer pressure and level by controlling spray flow, letdown flow, proporti onal heaters, backup heaters, and charging pumps.
(a) What three immediate automatic actions will occur - (1.0) if the controlling pressurizer pressure transmitter fails high ?
(b) What three immediate automatic actions will occur, (1.0) with heater-cutout switch set to "both", if the selected ,
pressurizer level control channel f ails low ? I (c) What effect does a loss of instrument air have on (1.0) the pressurizer l evel and pressure control systems ?
oANSWER >
i (a) Spray valves open, proportional heaters go off, and (1.0) backuo heaters are disabled (one bistable is utilized).
(b) The standby charging pump will start, letdown will (1.0) go to minimum, and the pressurizer heaters will trip (c) Letdown control valves fail closed, and the pressurizer. (1.0) spray valves fail closed.
OREFERENCE Training Article NS-9F, Pressurizer Level Control System, page 11 Training Article NS-9G, Pressurizer Pressure Control System, page 12-13 END OF SECTION 3 GD ON TO SECTION 4
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s SECTION 4 PROCEDURES- NORMAL, ABNORMAL, EMERGENCY, AND RADIOLOGICAL CONTROL oQUESTION 4.01 (2.5)
Following a-loss of all AC power, procedure 41RO-12ZO9 " Blackout" is being followed. Natural circulation has been established.
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(a) What two of the three conditions must be maintained (1,0) l to maximice the duration of subcooled natural circulation ? ]
(b). According to procedure,under blackout conditions, what (0.75) is the lifetime of the essential batteries ?
(c) If one emergency diesel generator can be started, (0.75) but cannot be closed in on the bus, how long can the diesel be left running ? l OANSWER (a) Do not allow saf ety valves to cycle, minimize RCS heat (1.0) removal, and minimice RCS inventory loss (b) 90 minutes (0.75)
(c) 60 minutes (0.75)
OREFERENCE
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Procedure 41RO-12ZO9, Blackout, pages 7, 21, and 22
oOUESTION 4.0. (3.0)
Following a loss of primary coolant accident in containment SIAS and CSAS have been reset. The RCS is at 350 degres and 175 PSIA.The containment is at 1.5 PSIG with 1.5 % H 2 Reinitiation of SI and CS is being considered.
(a) What are the criteria f or SI reinitiation ? (f ouF items) (2.0)
(b) How is natural convection in the containment determined ? (1.0)
(numeri al values or specific instrument numbers not required)
+ ANSWER (a) Subcooling less than 28 degrees (2.0)
Void not restricted to upper head (per RVLMS)
Pressurizer less than 33% (or not controllable)
Heat removal capacity of both S/Gs lost (b) Natural convection can be demonstrated by observing that (1.0) temperatures of lower portions of the containment are higher that temperatures of the upper portions of the containment.
OREFERENCE Procedure 41EP-12201, Emergency Operations, Appendix P page 2'and Appendin S page 2.
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I s l ODUESTIDN 4.03 ( 3. O ',
The control room is being evacuated due to a bomb' threat.
(a) What are the five immediate action to be taken while (2.05 evacuating the control room ?
(b) Following evacuation with a loss of f orced flow, (1.0) how long should you wait to verif y natural circulation ?
OANSWER (a) Manually trip the reactor Trip two reactor coolant pumps Manually trip the turbine Place the letdown control valve select switich to the "Both" position q Place the backpressure control valve select switich to the "Both" position (b) After 10 minutes (5 to 15 acceptable)
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- REFERENCE i i
Procedure 41AD-12227, Shutdown Dutside Control Room, page 6 and 9 l l
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s oOUESTION 4.04 (3.0)
Procedure 74AC-92ZO1, Radiation Exposure and Access Control, expl ainc .the framework by which personnel radiation exposures are controlled. ]
(a) What are the whole body exposure administrative limits (1.0) i per year and per week ?
(b) What two criteria must be met for an individuals whole body (1.0) exposure limit to be raised to the maximum of 3 Rem /Qtr ?
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(c) What are the Regulatory Limits f or maximum permissible (1.0) 1 quarterly exposure to the skin and to the extremities ?
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OANSWER (a) 4 Rem /Yr and 300 Mr/ week (1.0)
(b) 5(N-18) will not be exceeded and NRC-4 (or Equi v. ) is available (1.0)
(c) 7.5 Rem / skin and 18.75 Rem / extremities quarterly (1.0)
OREFERENCE Procedure 75AC-9ZZO1, Radiation Exposure and Access Control , page 24-26 I
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WDUESTION 4.05 (2.5) l J
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Control of contamination on personnel and equipment is included in procedure 7 SAC-9ZZO3, Radioactive Contamination Centrol as well as procedure 74AC-92ZO1, Radiation Exposure and Access Control.
.s (a) Whil e . f ri st.i ng , how many counts above bar;kg'rcund i re d i cate (0.5) contamination is present ?
(b) What attions do you take if the radiation ' portal monitor - ( 1. O) alarms as you pcss through ?
i (c) At the exit of the Radiological Control Area, under what (1.0) circumstances are you required to f risk only your hands and feet ? (two si tuations)
OANSWER (a) 100 counts above background '
(0.5)
(b) Pass through the monitor once m3re, i ft.it still alarms (1.0) notif y Radi ation Protection and await help 6 (c)' If you have not entered a contaminated area or if you m (1.0) already have done a whole body frisk efter leaving a ;
contaminated area.
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OREFERENCE '
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Procedure 75AC-92ZO1, Radiation Exposure and Access Controlf.q pqge 13a-14a, 18a r ,
a Procedure 75AC-92203, Radioactive Contamination Control, phge(y )
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ODUESTION 4.06 (2.0)
In procedure EPIP-02, Emergency Classification, events are classified by the number of barrier cha11anges or f ailures as well as plant vital auxili ari es and radiati on release conditions.
(a) What are the three barriers that are used to determine event _ (1.0) classification ?
(b) How many cha11anges and/or f ailures are necessary ' to qualif y (1.0) an event as a site area emergency 7 OANSWER I (a) RCS, clad, and containment -(1.0)
(b) two (two challangec, two failures, or one of each) (1.0)
OREFERENCE Procedure EPIP-02, Emergency Classification, Appendix C page 2 and 5
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- DUESTION.4.07'(3.0)
Proper station tagging and clearance procedures are used to prevent 'l personnel injury an equipment damage.
1 (a) -What are " mini-tags" and where are they used ? (1.0) l (b). Who is the responsible supervisor" who can authorize (1.5) j clearances f or each of the items listed below ?
(consider one clearance at a time)
Start-up Transformer XO2 !
Swi ti chgear in the 525 KV Switchyard (c) What kind.of tag (Danger, Caution, or Men-at-Work) is most (0.5) appropri ate f or ' i nservi ce (flow and dischrage pressure) testing of an-auxiliary feed pump ?
CANSWER (a) " Mini-tags" are smaller versions of red, blue, or yellow (1.0) tags and are.used in limited space areas.
(b) (0.5 each) (1.5)
S/U transformer -- Unit 1 Shif t supervisor TSC -- Stati on Services Supervi sor 525 KV switchgear -- (Used to be System Operations Supervisor, SOC Super vi sor) Transmi ssi on Control Center Supervisor.
(c) " Men-at Work" tags (are used for testing) (0.5) cREFERENCE Procedure 40AC-9ZZ15 Rev O, Station Tagging and Clearance, page 6a, 66, and 7, and 40AC-92Z15 Rev. 1 (Rev. 1, issued January 29, 1987, was supplied as a ref erence af ter the examination. )
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'oOUESTION 4 .08 (2.0)
Shutdown cooling uses the LPSI and CS systems. The procedure for Shutdown-Cooling Initiation, 41DP-1SIO1, has several precautions for the operator.
For the f ollowing questions assume " Train A" is being used for shutdown !
cooling. __
(a) Why must the LPSI recirculation valve be closed bef ore the (1.0) shutdown cooling loop i s aligned to the RCS ?
(b) Why.must the LPSI-CS cross tie valve'(SI-HV687) from the (1.0) shutdown cooling heat exchanger be shut prior to initiating shutdown cooling ?
- ANSWER (a) The recirculation valve is shut.to avoid-putting reactor (1.0) coolant into the RWT.
(b) The valve is shut to avoid spraying hot reactor coolant into - ( 1. 0 ) .
the reactor containment in the event of a containment spray actuation.
OREFERENCE i Procedure 41DP-1SIO1, Shutdown Cooling Initiation, page 12 and 15 i e
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- DUESTION 4.09 (2.0)
Techni cal . Speci f icati on 3.4.5.2 states the limi ts f or Reactor Coolant Gyctem leakage. Upon watch relief you are informed of the following RCS leakages.
O.5 GPM through the loop 1 A check valve 3.5 GPM through a valve stem seal 4.5 GPM through leaking pressurizer relief s 0.6 GPM steam generator 1A tube leaks 0.2 GPM RCS leakage through pressurizer socket weld 0.9 GPM unidentified leakage What leakage (s) e>:ceed Techni cal Specification LCD limits ? (2.0)
(Include the leakage specification f or each out of LCO condition)
- ANSWER (1.0 each) 0.2-GPM through the socket weld, no leakage is allowed 0.6 GPM through one steam generator,.720 GPD equivalent to 0.5 GPM is the limit per st eam generator.
- REFERENCE Technical Specifications page 3/4 4-19 1
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- OUESTION 4.10 (1.0)
During " Mode 1" operation, which one of the f ollowing conditions (i f not 1 corrected) does NOT require taking the plant to hot standby within one hour 7 (A) RCS T is 548 degrees F c I q
(D) One RCP is lost (C) One pressurizer code saf ety valve is inoperable (D) A SIT isolation valve is closed
- ANSWER
"(C)" (you have six hours)
- REFERENCE Technical Specifications pages 3/4.1-6, 3/4.4-1, 3/4.4-0, and 3/4.5-1 END OF SECTION 4 END OF EXAMINATION l
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