IR 05000315/1985002
| ML20134N979 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 08/30/1985 |
| From: | Ferrell R, Jaggar F, Reidinger T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20134N971 | List: |
| References | |
| 50-315-85-02, 50-315-85-2, 50-316-85-02, 50-316-85-2, NUDOCS 8509060096 | |
| Download: ML20134N979 (101) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Report No. 85-02 Docket Nos. 50-315; 50-316 License No. OPR-58 Licensee: American Electric Power Service Corporation 1 Riverside Plaza Columbus, OH 43216 Facility Name: Donald C. Cook Nuclear Power Plant Examination Administered At: Donald C. Cook Nuclear Power Plant Examination C nducted: July 16,-18, 1985 Examiners: e'Jdh pq A/J//Pf
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Date e 1 {.Jaggar 8/Io//1 Date L lt' - errell 8/fd/f5 Date /
Approved By: J cMillen, Chief Op[eratingLicenseSection //.f4//5
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Examination Summary Examination administered on July 16-18, 1985 (Report No. 85-02(DRS))
The applicants consisted of nine SR0, two R0 and one R0 retake candidate Results: All SR0 candidates passed the oral examination, three SRO candidates passed the written examination. Two R0 candidates passed the written examina-tion, RO retake candidate (Section 1 only) also passed written portion of examinatio RO candidate passed the oral examinatio S PDR ADOCK 050 O
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REPORT DETAILS Examiners
- T. D. Reidinger
F. Jaggar R. Ferrell
- Chief Examiner Examination Review Meeting The Attached " Resolution of Facility Comments" contains the facility comments and the examiners reolution of these comments. It should be noted that for several items, the utility supplied additional or updated material not provided to the examiners while they were developing the examination . Exit Meeting Facility representatives from Operations and Plant Management, the NRC Resident Inspector, and the chief examiner met on July 19, 1985. The excminer indicated to the management the names of all the candidates who clearly passed the oral portion of the examination. The examiner noted no specific generic weaknesse ,
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. RESOLUTION OF FACILITY COMENTS Examination Review Team i Bill Nichols j Dick'Strassor
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Bill Davidson Mike Mieran
. Lias Tatrault Dave Drape Jack Pecorgro Chuck Smith SECTION 1
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Question 1.05b
! Facility Comment:
The key listed .0585 as the correct answer. The equation needed for this cal-culation was not provided on the equation sheet:
Pk ~ 2 -k 1 KK
t However, this equation was provided on the equation sheet:
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9_ - ^Keff geq {
I The answer calculated from the equation provided with the exam is .0526. We request this also be accepted for full credi Reference ES 202-E GENERAL GUIDANCE POINT 22 0F NUREG 102 NRC Resolution:
Using the equation (s)
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(Keff2 ) (Keffs)-
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- which was supplied on the equation sheet, the correct answer can be obtained.
j The equation listed above will be removed from the equation sheet. Will accept 3 and 4 responses as correc Question 1.07
! Facility Comment:
i This information is given in Figure 8.1 of the Technical Data Book and is unit specific.
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U-1 E0C 100% Eq. Xenon = 3011 pc U-2 E0C 100% Eq. Xenon = 2860 pc Additionally, the approximate value use'd during training of Reactor Operators has been 2800 pcm without regard to time in core life (see attached thumbrules).
.The closest value given to the thumbrule is 2750 pc We request that either b. 2750 pcm, c. 2900 pcm or d. 3000 pcm can be accepted for full credi Figure 8.1 for both units are provided as an attachment to this revie NRC Resolution:
Due to the proximity of the responses to the many actual values or "thumbrules" the question will be delete Question 1.08 Facility Comment:
The solution requires the candidate to assume an initial Keff in order to calculate the reactivity additions. We request that d., not enough data given, also be accepted for full credi NRC Resolution:
If any Keff of less than 1 assumed initially the correct answer can be obtained using either "thumbrules" or equations, given on equation sheet. No change to answer ke Question 1.10 a and b Facility Comment:
Part a: Since time of core cycle will affect core axial flux profile and the concentration of fission product poisons, both of which will affect differential rod worth, we request that 4., Time of Core Cycle, also be accepted for full credi Part b: The question does not solicit an explanation. We request that TRUE without explanation be accepted for full credi NRC Resolution:
Part a: The question asked for the MOST effec This is answer choice No change to answer ke Part b: Explanation was in parenthesis indicat.ing that it was not required for credi No change to answer ke *
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Question 1.11b Facility Comment:
Since the question did not describe the rate of temperature increase as linear,
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this question could also be TRUE. The equation needed for the solution (i.e.,
heat flux as a furiction of T y -T is not provided on the equation sheet anditscomplexityisbeyondfNscoh8tof knowledge required by the Reactor Operator position. We request this question be deleted or both TRUE and FALSE be accepted for full credi NRC Resolution:
Knowledge of the boiling water curve is all that is necessary to answer this questio It is agreed that the question could have been stated clearer; therefore, since both answers could conceivably be correct, the question will be delete Question 1.14 Facility Comment:
The answer as keyed neglects mechanical moisture removal in the MS There is, essentially, no latent heat added to the steam leaving the moisture separator.
- Assuming then that the enthalpy change is due to superheating only, a. 85 BTU /
l lbm should also be accepted for full credit. The equations needed to solve this problem as keyed were not provided on the equation sheet which is in conflict with ES 2-2-E GENERAL GUIDANCE POINT 22 of NUREG 1021. We request that on future examinations, all required equations be given for calculations of this type either in the body of the question or on the equation shee NRC Resolution:
This question tests knowledge and use of the Steam Tables and Moller diagram, no equations are required. Use of Steam Tables is covered in ES-202 Section B.1 of NUREG 1021.
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It is agreed that the wording in the questions could have been more clear as to what was desired, i.e. , steam or vapor enthalpy change. Both a. and responses will be accepted for full credi .-_ .. .- ..
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SECTION 2 Question 2. Facility Comment:
The keyed answer is correct, however, this signal is also known as " flow retention." We request that this be accepted for full credi NRC Resolution:
Agree, " flow retention" added to answer key as acceptable answe Question 2. Facility Comment:
The keyed answer includes relief valve setpoints which are not elicited by the question. Request the relief valve setpoints not be required for full credi NRC Resolution:
The relief valve setpoint was included in parenthesis indicating it was not required for credi Question 2.05a Facility Comment:
The keyed answer includes safety valve numbers which are not elicited by the question nor would an operator be expected to know them. Request that the valve numbers not be required for full credi NRC Resolution:
Agree, valve numbers should have been placed in parenthesis and are not required for credi Question 2.06 Facility Comment:
The keyed answer requires (.5 each): H2 enters via natural circulation with containment ai . The air is preheated by the discharge unit . Electric heaters raise the temperature of the air where H2c mbines with Og spontaneously to recombine to form stea . Steam passes into mixing chamber with cool air and returned to
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- Parts of the keyed answer are flowpath descriptions and not elicited by the
! questio Request that the answer key be amended to read:
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"The recombiners use electric heating elements to increase the
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temperature of the containment atmosphere passing through the The l temperature is raised to the point where the hydrogen and oxygen ,
i recombine to form stea The containment atmosphere, containing
- hydrogen, is drawn into the recombiner by natural convection."
NRC Resolution:
The required responses are a description of the_" operation" of the hydrogen '
recombiner. The required answers will be modified as follows:
l j The air is preheated (by the discharge units).
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j This response not required for credit.
l Question 2.07a I
j Facility Comment:
- The keyed answer is correct for Unit 1. Request that the following alternate l correct answer relating to Unit 2 be accepted as well as the keyed answe i i
! High Steam Flow with Low Tave
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! Low Steam line pressure
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j High-high Containment pressure
l NRC Resolution:
i Agree, additional information for Unit 2 added to answer key.
j Question 2.08a and b (
! Facility Comment:
l l 1 The keyed answer references a one inch drainline on the first stage of the HP turbine. We request that the size of the line not be required for full credit.
j l The keyed answer states that windage is the heating of blading and air !
i spaces by friction. Since the turbines are only operated in steam ,
l environments, we request that no reference to " air spaces" be required l
for full credi i
! NRC Resolution: .
! i j Agree, "one inch" will be placed in parenthesis on the answer key.
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L The wording of the answer key is a direct quote from the training article (PGS-2A pg 25), however, if the response of the candidates mentions
" heating caused by friction of turbine blading moving-inside the casing",
full credit will be give Question 2.09b Facility Comment:
The question asks for the (one) feature. Since there are two features which are lost, we request that either one be awarded full credi NRC Resolution:
Agree, answer key changed to accept either feature for full credi Question 2.10b -
Facility Comment:
PGS-10 page 6 states that the Feedwater Isolation Signal causes Trip of Main Feed Pump . Closure of Feed Regulating Valve . Closure of Feed Line Isolation Valve PGS-10 page 37 does not list FW Isolation as a trip, but lists Reactor Trip las a FW Pump Tri The main feedpump trip is actually the result of the reactor trip (see attached DWG OP-1-98101-3 and OP-1-98211-6). Since the MFP trip is not a result of the Feedwater Isolation but will " occur with" a FWI (from reactor trip and low TAVE)
and the question asked for "three", we request that full credit be awarded whether the response includes " Trip of Main Feedpump" or no NRC Resolution:
Upon review of OP-1-98101-3, OP-1-982'l-6(c-4), and OP-1-98508-1(G-3), it was proved that the Main Feedpumps are tripped by Safety Injection or Steam Generator H -H, water level signals. These are two of the Feedwater Isolation signals. Tberefore,"tripofthemainfendpumps"willremainasarequired respons In addition drawing OP-1-98508-1(g-3) also lists turbine trip as a result of the same signals listed above, therefore, " turbine trip" will also be accepted as a proper respons ,
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SECTION 3 Question 3.01 Facility Comment:
For a rapid PT-505 failure, rod motion is generated by both power mismatch and the large deviation between Tref (generated by PT-505) and auctioneered high Tav The answer assumes that the rods will overshoot the no load Tavg setpoint. If this assumption is not made, the remaining 0.75 points of the answer will not be provide We request that the answer be modified to include both reasons for initial rod insertion and that the discussion of why rod motion doesn't occur after a T no-load overshoot not be required for full credi NRC Resolution:
- Agree, the fact that rods moving out is not solicited in the question and will not be counted for credit, however, since rods will drive in for approylmately 200 sec. at a decreasing rate, there will be a temperature overshoot and C-5 will be active and rods will not drive out. If mention is made that rods will drive out a .25 will be added to the grade, however, no deduction will be made if C-5 is not mentione Question 3.04 Facility Comment
The possibility exists that a candidate may arrive at a different answer from the key because Rx Makeup control is normally in " manual." If is is assumed that VCT low low level is reached early in the sequence of events, a refueling water sequence (swapover to RWST suction) may occur. This would result in pressurizer low pressure or OTAT tri The Technical Specification minimum value for the high level trip setpoint is 92%, the actual trip setpoint is slightly less, at 91%.
We request that the answer key be modified to reflect a setpoint alternative and that an alternate reactor trip scenario be accepted for full credit if additional correct assumptions are mad NRC Resolution:
A setpoint of 91% will be accepted for full credit. If it is stated by the candidate that makeup control is in manual, full credit will be given for mentioning RWST swapover sequence, however, the boron addition will not be sufficient to lower temperature and the pressure (which will be held up by pressurizer heaters) to the point of a reactor trip before pressurizer level
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increases to the trip setpoint.
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Question 3.05 Facility Comment:
The answer requires both P-10 and P-13 in addition to P-7 for those trips that are unblocked by P-7 permissive. The question does not elicit the permissives that are used in the generation of the P-7 signal. The referenced plant drawing shows only P-7 as the permissive interlocked with the reactor trips of concer We request that P-7 be a sufficient answer for full credit and that P-10 and P-13 be placed in parenthesis for the following trips:
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Pressurizer low pressur Pressurizer high leve Turbine tri loop loss of RCS flow (UV, UF, flow, and breaker position).
- NRC Resolution
Will accept P-7 as the answer for full credi P-10 and P-13 will be placed in parenthesis and will not affect the grad Question 3.07a, b, c Facility Comment: The PORVs will open together at approximately 2335 psig. For all of the PORVs, an interlock pressure of 2335 psig increasing and a set pressure of 2335 psig must be present, in addition tn the selector switch in "AUT0",
for the valve to open. The interlock pressures are not rate sensitiv (Refer to referenced drawing.)
We recommend that the answer key be changed to indicate that the valves will open together at approximately 2335 psi A single pressure channel failing high would not result in any PORV opening. The " interlock" channel would also have to fail high to unblock its respective PORV(s). (Refer to referenced drawing.)
We recommend that the answer key be changed to indicate that a single channel failure will not result in PORV openin With channels 1-2 selected (switch position #2), no PORV actuation will occur (as previously discussed in parts "a" and "b"). Heater and spray operation are only affected by pressure channel 1 in this position. The only action to occur would be that of indication and the respective annunciator alarm l
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Recomend that answer key be changed to reflect actual control circuit respons NRC Resolution:
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a, b, Material received post-exam was reviewed and acknowledged; however,
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the material supplied was first dated 11-14-79. The reference material supplied pre-exam (NS-3) was last revised 1-81 without incorporating the revision as stated in RFC-DC-12-2887. The changes (to the facility) were made ,
well in advance of the conduct of the license examination and should have been ,
placed in the reference neterial sent to the examiners so that the information I would be current.
! Exams will be graded per the answer key as written with reference to the i material originally supplied by the facilit It is requested that on future j examinations that all reference material be current and up-to-date so that
questions asked will be on current system status, i
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- The grading of this question as written on the answer key does not effect the pass / fail condition of any candidate.
- Question 3.08a i
Facility Coment:
l A candidate may choose to list "CRP-3 to the CRID" as two separate answers.
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Normal lighting via CRP-3 and standby lighting via CRP-3 are two possible
- different methods.
! We recomend that the answer key be modified to reflect the two possible
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soureces of power via CRP-3.
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NRC Resolution:
Agree, answer key amended to accept both responses for full credit, i i Question 3.08c
- Facility Coment
In order to provide the first response (600 volt AC tie breaker 21AC/2180
opens and locks out), the candidate must first assume an abnormal electrical
, lineup (21AC/2180) is present. For a normal plant lineup, only the non-
! essential 600V AC motor feeder breakers tri We request that the answer key be modified to place response No. 1 in parenthesis and only require response No. 2 for full credi NRC Resolution:
The question asked for "all the automatic actions". No change is required
, for response. Point breakdown changed to reflect a lower point value for the *
first response.
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{ SECTION 4 Question 4.03a Facility Comment:
PMI-2110 CLEARANCE' PERMIT SYSTEM, Section 3.1.1.2 reads " Minor adjustments and ,
troubleshooting on energized or pressurized systems may be conducted without ,
a clearance, provided it is performed in accordance with PMI-2290 (Job Orders)
or an approved plant procedure, and so long as there is no personnel or equipment hazard involved." While the exam item quotes the first phrase of this sentence and may be assumed to be true if the reader infers that the conditions expressed in the deleted qualifying clause have been satisfied, the statement is not entirely true as it stands. We request' that you accept either
"TRUE" or " FALSE" as a correct-- response.
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NRC Resolution:
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The question is a True or False question. If work may be performed on energized or pressurized systems-without a clearance at any time, then the ansuer is true.
. No changes to answer ke Question 4.04a 050.013 states that T.S. 4.5.1.b requires accumulator boron concentration be
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verified at each 1% volume change. This volume change corresponds to 13.45 cubic feet. In order to ensure sampling is performed as required, a volume change of >10 f t3 necessitates resampling. As the narrow range accumulator level gauges are scaled from 900 ft3 to 1000 ft 3, a 10 fta change is equiva-
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lent to a 10% change in indicated level. This is roughly equivalent to a 1%
- change in actual level. 'We are therefore requesting that you accept, in i
addition to the keyed answers of 10 cubic feet or a 10% change in indicated level, a response of a 1% change in actual: leve NRC Resolution: ,
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I Agree, answer key modified to also accept a 1% change in actual level.as a correct response for full credit.
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SECTION 5 Question Comment:
Choice "a" is a characteristic of the Source Range instrument sensitivity and installed neutron source decay not subcritical multiplicatio We request that choice "a" also be accepted for full credi Examiner's Comment:
The question was designed to elicit characteristics of subcritical multiplica-tion in regards to reactor theory. Choice "a" is not a characteristic of subcritical multiplication but of installed source decay and instrument sensitivit Choice "a" will not be accepte _ Question Comment:
The doppler coefficient becomes more negative over core life due to the buildup of resonant absorbers (Pu-240). Clad creep affects doppler defect by reducing the 0 to 100% fuel temperature rise but the doppler coefficient (pcm/ F) is not affected by clad creep. We request this question be deleted as there is no correct response in the list of choices provide Examiner's Comment:
The examiner accepts the premise presented by the utility; however, the question did solicit the Doppler defect vice the Doppler coefficient. The question was editorially corrected prior to the administration of the exam to the Doppler defect which reflects the correct answer of clad creep. The copy of the question given to utility at exam review did not reflect the updated revision to the Coppler defec Question 5.12 Comment:
When performing QPTR as a manual immediate action of OHP-4024.210.011
" Computer Alarm NIDS Tilt" (attached), corrective action is required if QPTR is greater than or equal to 1.0195. When performing QPTR under normal surveillance procedures, action is not taken unless QPTR is greater than 1.02. We request that either 1.02 or 1.0195 be accepted as correc Examiner's Comment:
Either answer will be acceptable. However, the examiner notes that Technical Specification specify a corrective action with a QPTR >1.02 and procedure 1-OHP 4021.011.003 paragraph 6.9 states that if QPTR exceeds 1.02 then apply Technical Specification 3.2.4 which states actions for QPTR >1.0 l l
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I i Question 5.13 l
Comment:
The question does not solicit an explanation of why the enthalpy rise
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changes. We request that "the enthalpy change decreases as power increases"
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be accepted for full credi ! Examiner's Comment:
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The question wishes to probe the reason of how and why the enthalpy changes ,
? across the steam generators as rector power increases. The candidate will ,
find it difficult not to explain how and why the enthalpy changes with powe >
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Partial credit will be grante :
i Question 5.14c
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Comnent:
! PZR level program is designed to maintain a constant RCS mass. As
! Auctioneered TAVG drops, the Reference PZR level tracks accordingly. No a change in CVCS control will occur since programmed level will change with actual level. We request that NO CHANGE be accepted as correct.
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Examiner's Comment: .
- The alternative answer will be acceptabl Question 5.15b
l Comment:
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The question does not solicit the level of detail found in the key. We <
- request that the following be accepted for full credit:
i Take manual control of rods to maintain WI on targe i Examiner's Comment:
The alternative answer will be accepted.
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Question 5.16 j Comment:
- Shutdown Margin is normally verified at power by ensuring rod postion above i .the rod insertion limit. We request this question be deleted or that the l following also be accepted for credit in addition to those listed in the ke . Fraction of Rated Thermal Powe ! Control Rod Demand Positio Individual Rod Position Alignment with Demand Position, Number of RCS loops operating, e .- . Boron Concentrations.
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i RCS Tav Fuel burnu i Xenon concentration.
1 Samarium concentratio Reference: Technical Specification 3/4.1.1. Examiner's Comment:
The alternative answer is acceptabl Question 5.17 Comment:
l The utility provided references relating to the Technical Data Boo ;
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i Question 5.18c
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Comment: ,
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i Nounitisspecified(U-2HZPRILis84stepsonBandC). We request that
, "above low-low insertion limit" be accepted for full credi i Reference: Step 4.10 2-0HP-4021.001.00 ,
Examiner's Comment:
The examiner does not accept the alternative answer but the examiner notes that the majority of the candidates stated 45 steps albeit incorrectl D.C. Cook system description still carries the rod insertion limit of 45 steps which was in effect.in 197 Question 5.19
Comment:
The question specifically solicits the possible effect of altering core geometry on criticality. Temperature effects are generic-not limited to the severely damaged core. The question also states that boron concentration is maintained. We request that " alteration of core geometry could cause
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recriticality" be accepted for full credit and temperature and coolant boil i
off discussion be deleted from the ke Examiner's Comment:
The alternative answer is acceptable.
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SECTION 6
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Question Comment: i
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RFC-12-2834 modified the MDAFP on Low-Low level in one steam generator, while the control switch is in neutral. Request that the answer key be modified to accept " Main Feed pumps tripped" for full credit.
Reference: Subject Memo RFC-DC-12-2834 (attached).
Examiner's Comment:
The alternative answer will be accepted although the examiner notes that the RFC-12 2834 is currently being accomplished during the outag Question 6.5a Comment:
The answer states that the trip occurs due to " indication of low RCS flow."
This is correct, however, the specific indication is that a RCP breaker indicates open (even though it remains closed) due to the loss of the CRID power supply. We request that either "RCP breaker indicates open" or the keyed answer be accepted for full credi Reference: Subject Memo RFC 32-2382 (attached).
Examiner's Comment:
Alternative answer is acceptabl Question 6.7:
Comment:
The reference given does not contain the keyed answe The purpose of the standpipe on the #2 seal leak off is to maintain sufficient pressure to ensure flow to the #3 seal. The standpipe maintains a nominal 7 ft. of water to provide the required head. The high and low level alarms are il ft. from nomina Since the question does not indicate any abnormal condition, it must be assumed that the standpipe level is between 5 and 7 ft. With the majority of system pressure being dropped across the #1 seal, changes significantly impact #1 seal leak off flow.
i We request that the following be accepted for full credit as well as the keyed answe .
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N The No.1 seal standpipe is on the outlet of the #2 seal and level changes would not affect No. 1 seal leakof Reference: NS-2 page 4 Fig. NS-2-2 Annunciator Responses 4024.107.071 4024.107.081 Examiner's Comment:
The alternative is acceptabl Question Comment:
The keyed answer is not contained in either of the indicated reference The key point for Step 6.12 states:
"6.12 Commence transfer when levels start to become unstable, or if steam and feedflow indications become unreliable."
Request the answer key be corrected to read:
"At low power levels, steam and feed flow indications become unreliable."
Examiner's Comment:
The alternative answer is acceptabl Question Comment:
The keyed answer requires the candidate to quote from the body of a **
procedure.' The ** indicates that the procedure must be in hand while being performe It is not required that operators provide specific quotations from the body of procedure The C&I Technicians are trained to perform this procedure and are available on site 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day. The SR0 only needs to instruct the Technician to set the DG up for local contro Request the answer key be amended to read:
" Ensure that the technician uses the procedure."
E/aminer's Comment:
Alternative answer will be accepted for partial credit with full credit going to the recognition that for local control certain colored wire and labels should be identified and recognize . ._ _ - _ . _ _ _ . __ _ . _ _ - . _. -
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f Question 6.11 t
Comment:
The answer key states:
" The diesel operating in test mode with a SI signal will return to standby operatio ! The emergency loads will automatically energize by offsite power."
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Since there is no diesel trip on a SI signal, it will not return to standby
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operation (shutdown in standby readiness). The DG output breaker will tri '
l Request that part "a" of the answer key be amended to state that "the diesel
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will continue to operate unloaded."
! Reference: AS-10 page 6 ,
Examiner's Comment:
The utility's comment matches exactly the definition and answer noted in
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Part A. The alternative answer is an editorially clarification of the
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examiner's answer. Full credit.
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Question 6.14
- Comment
A loss of Offsite power or all AC results in a loss of Circulating water.
1 A loss of 250V DC or vital AC power supplies de-energizes controllers and
- relays necessary for steam dump actuatio *
i Request the following be included as correct answers: ;
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Offsite Powe V A V DC Vital AC power supplie Reference: PGS-12 PGS-14 PGS-15 Examiner's Comment:
Alternative answers acceptabl *
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Question 6.15 Comment:
The keyed answer requires the examinee to assume that the cause of the dropped rod also resulted in a "possible" urgent failure.
i Because the assumption is that a failure of a control or instrument circuit is preventing rod withdrawal, we request that any one rod stop condition listed ;
- below or a general statement of a possible rod stop be accepted for full credit. OPAT, OTAT, rod uregent failure, PRNIS overpowe Examiner's Comment
Alternative answers are acceptabl Question 6.16 Comment: The wording of the question is distractive in that the pressure transducers associated with CTS and Phase B Containment Isolation are known as " Lower Containment" pressure instruments. An additional three instruments measure upper containment pressure which do not affect any safety system actuctions. We request that this be considered while grading the questio . The keyed answer contained about four defenses of the "No" answe We request that only one defense be required for full credi Examiner's Comment:
Will accept one answer minimum for full credi Question 6.17 Comment:
The keyed answer is not contained in the indicated referenc RFC 12-2663 is being installed in Unit 1 during the current outage (started May 6, 1985 and scheduled to end August 28,1985). See subject letter RFC-DC-12-2663 attached. No completion notice had been issued for the modification as of the exam date (July 16, 1985).
'The answer for the pre-outage configuration is false, for the post outage
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configuration the answer will be true. We request that either True or(False be accepted as correct answers.
. Examiner's Comment:
Alternative answer's will be acceptabl .
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. Question 6.21 Comment:
RFC-2448 has been partially installed, this results in deleting R-11, R-12 as well as adding additional correct answers. RMS Equipment Outline concerning Eberline/Victoreen monitors is attached. Request the following be accepted as additional correct answers: , Containment Area Monitor Containment Ventilation (1100, 1200) Isolation (CVI)
Lower Containment CVI Airborne (1300, 1400)
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Unit Vent Monitor (1500) Shifts to accident mod Examiner's Comment:
Alternative answers are acceptable. The examiner notes that the facility supplied additional information to amend the answer. Answer key changes are
note I
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SECTION 7 i
Question 7.03 i Comment:
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The answer says that the Shift Supervisor must obtain permission from the SRO-C Plant Manager Instruction (PMI-4050) states that the SRO-CA is responsible to the Plant Manager for refueling operations and 1 Paragraph 3.6.2.8 states that he will authorize bypassing any crane interlocks. The Refueling Shift Supervisor referred to by the reference
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NS-16-52 would be the senior person associated with a contracted refueling crew and is not mentioned in this line of authority for PMI-405 We request that only SR0-CA be an acceptable answer for full credi Reference: PMI-4050 Paragraph 3.6. Examiner's Comment:
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Alternative answer is acceptabl Question 7.04 Comment:
The answer key does not recognize that there are normally two "N" train battery chargers available for the "N" train battery. If the candidate assumes that the second charger is not available/ inoperable, then the answer l key is correct. If the other charger is assumed to be operable, then the AFWP would also be operable, and no action would be required per Technical Specification We request that either of the two possible answers be given credit based on the assumption made by the candidat Reference: DWG # OP-2-98210- Examiner's Comment:
The facility provided additional information not available prior to the examination and is acceptable.
.
Question 7.10 Comment:
The candidate may state that they would submit a " Condition Report" in lieu of answers "c and d" (submit a safety violation report and submit the report within 14 days). The responsibility for SR0 action in a Safety Violation normally is completed with submission of Condition Report.
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I We request that answers "a" and "b" be required but that " submit a condition report" be acceptable for both answers "c and d" if this cction is take Reference: PMI-7030 (Revision 6, TP-7), Page Examiner's Comment: ,
Alternative answers acceptabl !
Question 7.13
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Comment:
The procedure (OHP 4023.001.002) states four conditions that it will assist in diagnosing. These conditions are listed as the four answers in the answer ke " Spurious Actuation of Safety Injection" is not an emergency procedure but
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rather is a part of procedure OHP 4023.001.002 (Section 5.8 on page 5 of 5).
I We request that answer No. 1, Spurious Actuation of Safety Injection, not be required for full credit and that the answer key be modified to more closely parallel the actual procedure titles as follows:
!
(1) Loss of Reactor Coolan (2) Steam Line Brea (3) Steam generator Tube Ruptur Reference: OHP 4023.001, pages 1 and 5.
Comment:
The referenced procedure lists the following immediate manual actions:
(1) Reduce load if necessar (2) Check proper operation of steam seal syste (3) Check proper operations of air ejector (4) Insure adequate circulating water flo (5) Insure proper venting of condenser water boxe (6) Check vacuum breaker position and proper seal water.
f We request that the answer key be modified to more closely follow the referenced procedur '
Reference: OHP 4022.053.001, page 1 of Examiner's Comment:
Although immediate actions are noted the examiner will accept reasonable statements not as verbatim as the procedure calls for.
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SECTION 8 i
Question 8.12 Comment:
There has been a very recent change to PMI-2210 " CLEARANCE PERMIT SYSTEM" which impacts this situation. Procedure Change Sheet #3 to PMI-2110 was placed in effect on Friday, July 12th (2 working days prior to the exam date). This change was instituted as a result of NRC Generic Letter 84-15 which addresses Diesel Generator Reliability and established an objective to reduce the number of cold fast star LER 50-316/78-037 commitment referenced in the keyed answer has been rescinded via revised LER dated May 15 with a proposed affective date of June 30. Attachment 1 of PMI-2110 now states that if an EDG is to be taken out of service, the other must be proven operable per
-
Technical Specification requirements. As Technical Specifications do not require the other diesel generator to be proven operable prior to removing the affected EDG from service, PMI-2110 does not now require any prior action ! Therefore, we are requesting that you accept either "no prior requirememts" (follow applicable Technical Specification action statement for subsequent actions) or the keyed answer as correc , Examiner's Comment:
The alternative answer is acceptabl Question 8.18 Comment:
(a) We request that you change your answer to reference T.S 3.4. Overpressure Protection Systems applicability statement: "When the temperature of one or more RCS cold legs is less than or equal to (Unit 1: 188 F, Unit 2: 152 F), except when the reactor vessel head is removed." We also request that you accept the temperature associated with either uni (b) In addition to the keyed answers, we request that you include as correct responses:
. PORV must time open in < 10 seconds. (Reference: 1-0HP 4021.001.004
" Plant Cooldown form Hot Standby to Cold Shutdown" Step 6.2.7.1.)
. T WIDE RANGE must remain greater than the minimum allowable RCS thhh0rature(basedonPORVopeningtime). (Reference:
1-OHP 4021.001.004, Caution Note after Step. 6.3.5.)
Examiner's Comment:
The alternative answer is also acceptabl . - - -_ - . , .. - .. - - , - , _ . _ . . - . _ .
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c Question 8.19 Comment:
The keyed answer to c categorizes a body leak on QRV-301, the low pressure letdown valve, as pressure boundary leakag Referencing the Definitions section of Technical Specifications, " PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage)
through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall".
A QRV-301 body leak would not qualify as pressure boundary leakage because it is not a Reactor Coolant System Component and its ability to be isolate As a minimum, 2 containment isolation valves exist in the letdown flowpath upstream of QRV-30 Since the Reactor Coolant leakage criteria does not apply to QRV-301, we request that item c be deleted or both "not applicable" or "10 gpm" (since the leak location is known) be accepted for full credi Examiner's Comments:
10 gmp will be accepted for full credi Question 8.20 Comment:
The D.C. Cook Unit 1 and Unit 2 Technical Specification basis for the Reactor Core Safety Limit (2.1.1) reads as follows: "The curves of Figures 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature:
. Unit 1: For which the minimum DNBR is no less than the applicable design DNBR limit, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liqui . Unit 2: Below which the calculated DNBR is no less than the correlation DNBR limit value or the average enthalpy at the vassel exit is less than the enthalpy of saturated liqui We therefore request for part (b) that you accept " ACTUAL DNBR > DESIGN DNBR" as a correct response so as to reflect the D.C. Cook Reactor Core Safety Limit design basis wordin Examiner's Comment:
Alternative answer is acceptable.
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U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _C00E_112________________
REACTOR TYPE: _ EWE =WECA________________
DATE ADHINISTERED _B3ZO2216 _______________
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EXAMINER: _J0GGe )
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APPLICANT: _____j , q { }_ gyj-q
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IWSIEUCIIONS_IO.AEELICAWII Use separate rarer for the unuweru. Writu answeru un unu uide unl Staple uuvution shout un tue of the answer uheut Points for each ausution are indicated in parenthouus after the uuuntion. The pausind grade reuuires at Ivaut 70% in uach cateduru and a final drade of at least 80%. Examination rarers will be picked un six (6) hours after the examination start % OF CATEGORY % OF APPLICANT'S CATEGORY
__UALUE. _IDIAL ___ SCORE ___ _UALUE__ ______________CAIEGORY_____________
a.3 50 225r02:a _25.00 ___________ ________ PRINCIPLES OF NUCLEAR POWER
, PLANT OPERATIONr THERHODYNAMICSr HEAT TRANSFER AND FLUID FLOW
_23.00__ _25.00 ___________ ________ PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
_25.00__ _25.00 ___________ ________ INSTRUMENTS AND CONTROLS
_25.00__ _25.00 ___________ ________ PROCEDURr 3 - NORMAL ADNORMALr EMERGENCY bHD RADIOLOGICAL CONTROL 9f 50 400.04 _ 100.00 ___________ ________ TOTALS
-' FINAL GRADE _________________%
All work dono un this examination is au own. I have nuither diven nur received ai ___________________________________
APPLICANT'S SIGNATURE
_ .-...___.1.__ . ._ __ _ . l t s
! *1._bEIWCIELES_DE_WUCLEAE_EDWER_EleWI_DEEE6IIDW2 PAGE 2
, IWERMODYWAMICSz_WEAI_IE4WSEER_0HD ELUID_ELOW DUESTION 1.01 (2.00)
i a. If the reactor in operating in the power runder how lund will it take to raise Power from 20% to 40% with a +0.5 DPH Start-
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up rate?
- se . 21 se ,
' 3. 36 se . 54 uu (1.0)
b. How land will it take to raius power from 40% to 60% with the same +0.5 DPH Startue rate? se . se . 36 se . 34 se (1.0)
DUESTION 1.02 (1.00)
Which of the followind best describus the effect un MTC if the RCS temevrature is LOWERED 7 It becomes issu nudative becauuv burun and watur~moluculus are sweet into the cure as a result of the uutuurdu from the Pressurizure thereforce neutronu unend sure time in the ruuun a.1c e radio It becomes issu nadative becauuv the ratu of chande in the dunsitw of water eve degrew tuarurature chands is luuu at lower temperature which causes a lessor change in rato in reuunance wucare reubabilit It becomes aure nudative becauuv thermal utilization increases and reuunance escape erubabilitu decrease It becomes more nudative because au temperature is lowered the moderatur becomes muro denues this increanus the amount of
- water moleculuu in the cure therefore neutrons have a drvater
! Probabilitu of colliding with a water mulveule and this is an increased nedative reactivitu uffect.
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IWERWODYNAWICSa_WEAI_IRAWSEEE_AWD ELUID_ELOW QUESTION 1 03 (4.00)
. Using the attached Xenon worth curver Fi .1, answer the followin , Power at TO was at 70%. What was the power level between T1 and T27
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1. 90%
2. 50%
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3. 20%
[ % (1.0)
b. What was the length of time between T2 and T37
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, 1. I hour 2. 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> hours hours (1.0) What happened at T27 1. Reactor trieve . Rodu were placed in AUT0r and turbine power was raiuud to 100%. Reactor Power was reduced to 10%.
4. Turbinu ruwer remained constant, rods were in manual and inserted 50 stwes and the utvan duer valves failed uren (10% of rated Power). (1.0)
, At time T4 ... All Xenon production hau utuPPede Iudine decau to Xenon hou utuere .
3. All Xenon production remains constanta but the burnout increase . Xenon production direct 1u from fisuion has uturned, but Xenon Production from decau Iudine continue (1.0)
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QUESTION 1.04 (1 00)
Delaved neutrons elaw a muJor role in trie orvration of the core I
because thew ...
e are born at (thermal) ulow enerdu levolu (lvuu than i ev) and I therefore are more apt to cause a finuion au compared to
! beind absorbed bw a poiso are considered an epithermal neutrons and thereforu thou will not travel far enough to Iwak out of the cor c. are born so much later than the prompt nuutronu and provide controlabilitw during utvadw state oevrations and power tranuient . Provide 70% of the fiuuion neutron inventoru and have higher importance factors associated with them au compared to prompt neutron QUESTION 1.05 (2.00)
If the Source Range (SR) instruments indicatu 50 cru with Keff eaual to 0.9e what would the SR instrument indicate if rods were withdrawn to brind Keff euual to 0.937 Auuume BUL condition . 50 ces cps cps cru (1.0)
b. How much reactivitu wou added?
1. 0.0347 2. 0.0500 3. 0.0526 .0383 (1.0)
(***** CATEGORY 01 CONTINUED ON HEXT PAGE *****)
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e ' I.__ERINCIELES DE_WUCLEAE_20WER_ELeWI_DEERAIIDWz PAGE 5 IHERMOD%WeHICSz_WEAI_IRAWSEER_oWD_ELUID_ELOW QUESTION 1 06 (1.50; Compare the calculated Eutinated Critical Pouition (ECP) for a startur 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> after a trip to the actual Critical Rod Position (ACP) if the following eventu/conditionu occurre Considur wach
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indeevndentlu. Limit uuur answer to:
1 ACP higher than EC '
b. ACP lower than EC ACP would not be uidnificantlu different than EC . One Rvtur Coolant Pump is utoneud one minutu Prior to criticalit . The stuum dump rressure uuteuint is inereuuwd to a value Just below thw code safties uuteuint . The startue is delswed 2 more hour QUESTION 1.07 (1.00)
Which of the below is the approximatu value for 100% Power souilibrius Xvnon reactivitu at EOL7 pcm eca rem rem (***** CATEGORY 01 CONT 1 HUED ON HEXT PAGE *****)
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r QUESTION 1.08 (1.00)
During a reactor utartune the first reactivitu uddition causud count rate to increase from 10 ens to 16 ces. The uncond reactivitu addition caused
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count rate to incruaue from 16 cru to 32 ceu. Which of the following
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statements describing the relationship between the reactivitu valuvu of the first and second reactivitu additions is currect?
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l The first reactivitu addition was large . The second reactivitu addition wou large ' The first and second reuetivitu additions were veua There is not enough data given to determine relationship of ruuctivitw ,
value i l
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QUESTION 1.09 (1.00)
i During a Xenun-free reactor startune critical data wou inadvertent 1u taken two decades below the revuired Intermediate Range (IR) level (1xE-20 ames).
The critical data wou taken uguin ut the proper IR Invel (1xE-8 uses).
Assuming RCS tumevratures and burun concentrations were the same for wach set of datar which of the following statementu is currect?
4 The critical rod euuition taken at the proper IR Ivvv1 in LESS THAN I the critical rod euuition taken two decaden below the prueur IR level.
i The critical rod punition taken at the proper IR Invel in THE SAME AS
the critical rod puuition taken two decades below the prueur IR Ivvel.
! The critical rod punition taken at the prueur IR level iu GREATER THAN
- the critical rud Puuition taken two dveadeu below the proper IR Ivvel.
I There is not enough information given to determine the relationship between the critical rod punition taken ut the prueur IR level und the critical rod Position taken two dueudes below the proper IR level.
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- IMERHODYWAMICSa_WEeI_IkeWSEER_oWD ELUID_ELOW
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GUESTION 1.10 (1.00)
I Which Parameter below will have the HOST vffect un the shaev of a Differential Rod Worth Curve?
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1) Cure radial flux profile 2) Cure axial flux profile 3) Cure axial temperature profile l
4) Time of core cuelo (0.5)
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, TRUE or FALSE 7 The effvet of the bank overlar prudrum on the Diffuruntial Rod Wortti curve is to make the sharu of the curve more linea (0.5)
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i QUESTION 1.11 (1.00)
TRUE or FALSE 7 During 100% ruwur ururatione Durarture from Nucleatu Boiling Ratio (DNBR) is droater than the DNBR for 20% ruactor powe b. As the temperaturu diffurunce betwoun the fuel rod surface and the saturation temevrature of the coulant (Twall-Tuat) incruaucur the heat flux across the fuel surface (BTU /hr uu. ft) increases at a constant linear rat QUESTION 1.12 (1.50)
TRUE or FALSE 7 The faster a centrifudal pump rotatuur the druutur the NPSH reuuired to prevent cavitatio b. One of the ruur laws for centrifugal pumps statuu that the volumetric flow rate is inveruplu proportional to the unwed of the Pum Puer runuut is the term used to duueribe the condition of a centrifugal rume running with nu volumetric flow rat (***** CATEGORY 01 CONTINUED ON HEXT PAGE *****)
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IMERMODrWAHICSz_HEAI_IEeWSEER_eWD_ELUID_ELOW
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, QUESTION 1.13 (1.00)
I s TRUE or FALSE 7
. During a RCS heature as tumeurature dets hidherr it will
, take a smaller Ivtdown flow rate to maintain a conutant Presuurizwr leve Increasing condensate duercusion (subeculind) will cause
- BOTH a decreacw in plant efficiencu AND an inervusw in
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condensate (hotwell) rune available NPS QUESTION 1.14 (1 00)
Steam exiting thw HP turbinw is at 785 euide 90% uualit Steam enturing the LP turbine is superheated to 100 What is the
, enthalew change of thw steam?
, BTU /lbe BTU /lba BTU /lba BTU /lba
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QUESTION 1.15 (1 00)
In order to maintain a 200 F subcoulind mardin in thu RCS when reducing RCS pressure to 1600 euide stuum duneratur eruuuuru must be reduced to approximate 1w esis esis esis esis (***** CATEGORY 01 CONTINUED ON NEXT PAGE *1***)
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QUESTION 1.14 (2.00)
If utuam does through a throttlind Procesue indicate whether the following
. Parameters will INCREASEe DECREASEe or REMAIN THE SAM Enthairw Pressure i EntroPw
' Tvererature
.
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QUESTION 1.17 (2.00)
- Indicate how the following will affect Unit efficiencu (increase, decreasee no change) at a utvadv utate power Ivve (Consider each caue seraratelv.)
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a. Absolute condenser pressure changwu from i esi to 1.23 Ps b. Total S/G blowdown is chanded from 33 dem to 40 ge c. Condenser hotwell tumeurature chandvu f rom 123 F to 130 (Assume no change in condenser Preuuure.)
i Steam uualitw changen from 99.8% to 99.7%.
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(***** END OF CATEGORY 01 *****)
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f GUESTION 2.01 (2 50)
+ List the signals that will uturt the motor und turbine-driven auxiliarv feedruse Suteoints not reuuire (1.5) What automatic action takes placu if a low flow condition through the AFW Punes existu? (0.5)
i What signal will cauuv the S/G uurplu valves from the AFW Punes to close to the intermediate position? Sutpoint not reuuire (0.5) ,
QUESTION 2.02 (1.00)
From the following statementue choouw the statement that in INCORRECT concerning PRESSURIZER design criteri . The combined water and stuum volume is sufficient to provide desired Pressure response to suuten volume change The water volume in uufficient to prevent uncoverind heaters I on a rame load increase of 10% of full powe The steam voluau is lordu enough to provent a high luvul reactor trie from occurring from the uurde creatud on a decidn 100% .
load rvJuction with reuetor control or stuum dune The steam volume is larde enough to prevent water relief through the safetw valves followind a lous of loud with thu high watur Ivvel initiating the tri GUESTION 2.03 (2.00)
State the locatione in the Chemical and Volume Control (CVCS) Suutene of FOUR of the FIVE relief valveu that protvet the CVCS Intdown and charding line Include the uutpoint und the location whure vach relief valve discharde in directe (2.0)
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GUEST 10N 2.04 (2.50) ' State the three uuurces of fluidu cooled bu thu Seal Water
Heat Exchange (0.75) State the design Iwakoff rate for each of the RCP uvalu and where each dischardes. Asuume normal RCS pruuuur (0.75)
, ;
i During a Phase 'A' isolutione what in the RCP 41 uval leukoff floweath7 Include where thu water gue (1.0)
?
QUESTION 2.05 (3.00)
' Wherve in the Residual Heat Removal (RHR) suutems aru the FOUR safutu valves locatede AND what in the uutpoint of uuchi (2.0)
\
t State two design featurvu that Protvet the RHR PUMPS from damage
.i and state when each is usedT (1.0)
QUESTION 2.06 (2.00)
A Describe the ureration of a hudrogen recombiner uni Include in the descrirtion how the hwdrogen in drawn ine the procuus that takes
- elaeve end sevcificallu how the hudrouen is removu (2.0)
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QUESTION 2.07 (2.00) What signals will cause the Main Steam Iuulation Valves (HSIV)
to close automatica11uT Seteuintu not ruuuiru (1.0) What is the deuidn clouind time for an MSIV followind automatic Safetu InJaction initiationT (0.25) Describe the swstem uuud to close the HSIV in this tinu purio (0 75)
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! QUESTION 2.08 (4.00)
i How is muisture removed from the:
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I 1. High Pressure Turbinv7 and
'! . Luw Pressure Turbinv7 (1.5)
! ,' Define " windage" an it portains to a turbin (1.0)
- WHEN and WHY are Exhaust Hood Spraus uued? (1.5)
.
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l QUESTION 2.09 (3.00)
.
A fire has been detected, and the fire supervuuiun Justus huu been activated in the 4KV switchgvar roo Describe the automatic neuuence of eventu that occurs to
- deliver CO2 to this ruum. Bu unveifi (2.0)
,s b. Describe the feature that in lost if the uvutum is initiated
,
manual 197 Be seveifi (0.5)
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c. How can wou determine if CO2 has been manuallu initiated into a ruum wou are enteringT (0.5)
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! QUESTION 2.10 (3.00) List the signals (seteuintu not ruuuired) that will create a feudwater isolation uigna Explain whv feudwater is inulated in each cas (2.25) What THREE automatic actions will occur with the actuation of a feudwater isolation signal? (0.75)
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, QUESTION 3.01 (2.50)
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- The plant is at 100% Power with the rod control nuuten in
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automatic. Explain the responus of the rod control sustem
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if the turbine inrulse chamber Preuuure uidnal fundind rod i control fails low. Carru uuur explantion to the point where rod
! notion stors and ausume no crurator action or protvetive actio (2.5)
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QUESTION 3.02 (2.00) Explain the Purruse of the utuam presuuru input used in the developement of a utuam flow uidnal for the S/G water Ivvel control suste (1.0)
b. Conrare INDICATED steam flow to ACTUAL uteam flow its during a
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Power increase from 0-100%, the steam preuuure signal stuck at it's 50% valu (1.0)
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QUESTIDH 3.03 (4.00)
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, i The following concern the Steam Dump Suste t WHY is it necessaru to varu the number of utuam dune valves which i operate in resrunse to different conditions (i.e. magnitude of temperature errore turbine trip or load ruJuetion)T (0.50) Describe the function of the RESET position on the Steam Dune Mode Selector Switc (1.0) When the Steam Dume Control Selvetor switcheu are in the l OFF/ RESET position, what control uvuten uidnals will uren
.
the arming valvest (0.50) State TWO rurposes of the Steam Dume Control Suutus interlock (1.0)
, Which reuuires more utvan duer capacitue a turbine trip or a large load rvJectionT Explain wh (1.0)
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'3.__IWSIRUIEWIS_dWD_ COWIE 0LS PAGE 14
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QUESTION 3.04 (3.00) 1
With Plant load at 50% and the Chemical and Volume Control Suutum (CVCS)
,
in a normal linmure the charging swstem flow rate in manuallu uvt at 30 dem. Assuming no oeurator actione state the uvuuence of events that will Ivad to a reactor tri Include uutpoints where applicabl :
,
QUESTION 3.05 (3.00)
List ALL reactor tries that are affected either manuallw or automatica11w bw a PERMISSIVE. Include the name of the trie and ALL associated ruraissive (3.0)
QUESTION 3.06 (3.50)
I a. Now can overcuarensation of one intermediate range excore detector be recognized on the startup RATE maturu after a reactor trir? (0.75)
, Intermediate range channel N35 is reading 10 -11 aeru while
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channel N36 is reading 10 -10 amps. Using the Source Ranger how would wou verifw which channel ju reading currvetlut (0.75) . Exelain the effset of an adjustment of the 'uumming level gain adjust rot * to each of the throu Power range drawer meter (0.75) Whw should caution be uurd when adJuuting this not while at POwur7 (0.75) .
3. When is this Pot adJuuted while at powert (0.5)
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3.__INSIEUTEWIS_aWD_COWIROLS PAGE 15
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QUESTION 3.07 (3 50) If Pressure were to rise rapidlue would all of the Preuuurizer Power Operated Relief Valves (PORV) uren uimultaneouslwT EXPLAI (0.75) Is it russible for sure than one PORV tu open if a SINGLE Pressure channel fa12s hight EXPLAI (0.75) With Pressurizer Pressure channels 1 and 2 uvluetud for contro1r explain how the pressurizer ervuuure control nuuten would respond if Channel 2 failed hidh. Assume no ururatur action. Fu unecific and give aerrueriate uuteuint (2.0)
,
QUESTIDH 3.08 (3.50) List FOUR raths of Power to a Contrul Ruum Inutrumentation Distribution Panel (CRID). (2.0) Explain the oevration und purecue of the interluck auuociated with 600-volt bus-tic breaker 21A (0.75) List all the automatic actions affecting the low voltage elvetrical sustoms that occur when the diesel denerators s start as a result of a 4 KV bus undervoltad (0.75)
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(***** END OF CATEGORY 03 *****)
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'd*- EBOCEDURES_=_W3BMALa_eBWDEMAL2_ EMERGENCY _AND PAGE 16 RADIDLOGICAL_COWIROL I
%
QUESTION 4 01 (1 00)
If the minimum tumeurature for criticalitu Technical Seucification is violated, which of the below untions uhuuld be performed?
a. Resture to within its limits within 15 minutuu or be in Hot Standbw within the next 15 minute Restore to within its limits within 15 minutuu or be in Hot Standbw within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
,
c. Resture to within its limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in Hot Standbw within the next 15 minute d. Resture to within its-limits within 15 minutuu or be in Hot Standbw within the next i hou QUESTION 4.02 (4.00)
The following euncern erweautions found in the P3 ant Huutue from Cold Shutduwn tu Hut Standbu procedure 1-OHP-4021.001.00 When Pressurizer temperature is below normal unerating temperaturer whw is thw use af the but calibration auters of concernT (1.5) When the RCS is at low Prusuurv with eruuuuru being controlled with the letdown Pressure control valve (DRV-301): how will RCS
'
Pressure initiallu respond to the stoppind of an RHR puse? (1.0) Whw should Tavg be greater than 541 F before energizing the
,
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control and shutdown rudu? (1.0) TRUE or FALSE 7 According to Technical Specificationur it IS rurminuiblu to make mode changes with degraded vuuipment that fu reuuired in thu mode being entere (0.5)
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'A - CROCEDURES_=_WORZ$La_eBWORroLa_EHERGENCY_AND PAGE 17 RADIOLOGICAL _CONIROL
- QUESTION 4.03 (7 00)
,
Indicate whether each of the following are TRUE or FALSE concerning the Clearance Fermit Swstver PHI-211 .
a. Hinor adjustments and troubleshoutind on unardiaud or pruuuurized swstems saw be conducted without u clearanc b. A ' Red' tag and a ' Striped * tus cluurance rurait can be issued for a coerununt at the same tim c. All Technical Seucification vuuipment for which a clearancu eurait is obtained reuuirvu 'Operationu' verification of corruet tudgins and positioning regardivuu of the mode in which the vuuipment is reuuire d. Lust tags must have a new tag iusued and hung with erurer verification and notation, tags that full off muu bu replaced provided reuuired comrunent position is verifie ; QUESTION 4.04 (1.50)
j The following concern Standind Order 030.013: Accumulatur Level Surveillanc a. How much change in accumulutur luvel (nutification limit) ruuuires resampling of the accumulator burun concentrationT (0.5) Seccificallw how in the chande in accumulatur luvul deturmined AND how often is this verification performedT (1.0)
>
J QUESTION 4.03 ( .50) l
l Whor bw Job titiv/rosition, muut unprove withhuldind the reuut of ECS l eauipment for Safetu Injections occurring below the low preuuure SI i seteuinte according to Standind Order 050,6417 (0.5) ,
i i QUESTION 4.06 (2.00) l List 4 of the 8 immediate actions rouuired bu PHP 2080 EPP.013e
" Duties of the Individual Who Discovers an Emerduncu Condition'.
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'd.__2ROCEDURES_=_WORMALc_OBWORTALa_EHERGENCY_AWD PAGE 18 RADIOLOGICAL CONIROL QUESTION 4.07 (2.50)
The following concern 1-OHP-4023.001.011 'Ruactor Shutdown from Hot
, Standbv Panel Due to Control Room Inaccessibilitw*.
. How would the oeurstor in the control room be uavrted to a transfer switch on the Hot Standbw Panel unie.g placed in the LOCAL
. Position? (0.5)
b. How is Tavs determined at the Hot Standbu Panel? (1.0) What are the TWO manual immediate actions ruuuired at the Hot Standbw Panel following a control room evacuation? (1.0)
QUESTION 4.08 (1.00)
.
According to 1-OHP-4023.016.001 'Lous of CCW'r the RCP muut be stuered if bearing tumeurature reachuu _____ F OR 41 uval 3eakoff temperature reaches _____ F, 185 F Fe'200 F Fe 195 F F, 200 F (1.0)
QUESTION 4.09 (1.50)
The following concern the ' Loss of Power to the CRID Distribution Cabinets' procedure 1-OHP-4022.082.00 I With the reactor at 100% powers explain how a reactor trip will occur as a result of a loss of power to a CRI (1.0) If the failure affected CRID II onlue what additional automatic action saw occur 7 (0.5)
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'A.__E'ROCEDURES_=_ NORM ALa_ AB WORM ALa_ EHCRGENCY_ LWD PAGE 19 RADIOLOGICAL COWIROL
, l OUESTION 4.10 (2.00) l
- Which of the following is an indication of a druered control rod?
1. Rapid rise in RCS tvereratur . Rapid rise in Pressurizer Preuuur . Rarid drue in Pressurizer Ivve . Rapid drop in charging flow indicatio (1.0)
b. If TWO or more ruds have druered without a reactor trier what are the reuuired manual immediate action steru? (1.0)
,
QUESTION 4 11 (3.00) Assume the Plant is operating at full power and the Axial Flux Difference (AFD) has been outside the target band for the last 3 minute What are the TWO actions unecifieds une of i which must be rurformed according to the Technical
, Sevcification reauirements? (1.0) Assume that it is 0310 on 3/19/83 and the plant is presentlw at 43% rower. Considering the AFD Penaltu hiuturu belows at what date and tiac mas ruwer be incruauwd above 50%7 EXPLAI (Show all work.) Assume nu deviation uutside the band after 0310 on 5/19/8 TIME WENT GUT TIME BACK DATE OF BAND IN BAND POWER 5/18/85 0310 0318 85%
5/10/83 1557 1637 45%
5/19/83 0148 0310 45% (2.0)
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(***** CATEGORY 04 CONTINUED OH HEXT PAGE *****)
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- d*__EROCEDURES_=_WORHOLa_6BWORMALa_EMEEGENCY_aWD PAGE 20 RADIOLOGICAL _COMIROL QUESTION 4.12 (1.00)
Which of the following would be a nuortum of a Power Range instrument failing HIGHT
, Ruds stepping ou Tuve increas ' OPdT reactor tri Ruds steering i ! QUESTION 4.13 (1.00)
-: Prior to a reactor startue with normal unwruting temevruture
<
' and Pressure the following RCS Iwakagou exiut. For wuch leak rate belows indicate if wou would STARTUP or REHAIN SHUTDOW . 0.5 GPH from a cracked weld un a narrow range temevruture instrument manifol i
. .0 GPM from a manual valve racking glan . 0.4 GPM tube leakage un une Steam generato . Leak from an unknown suurev of 1.2 GP QUESTION 4.14 (2.00)
According to the D.C. Cook Radiation Prutsetion Manual, Table IV.D.-1, wour uuurterlu whole budu expouure limit is 1.25 REM. Under certain conditions it saw be raised to a higher limi To what new limit meu vuur uuurterlu limit be raised? (0,5) What three conditions must be met prior to raising it to this new limit? (1.5)
(***** END OF CATEGORY 04 *****)
j (************* END OF EXAMINATION ***************)
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EQUATZON SHEET f = ma v = s/t Cycle efficiency = (Net work cut)/(Energy in)
,
w = mg s = Vo t + 1/2 at2
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g = mc 2 ,
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I 2 A = AN A=Aeg KE = 1/2 mv a = (Vf - Vn )/t PE = mgn V7 = V, + at * = e/t x = tn2/t1/2 = 0.693/t1/2
-
y , , 3p
-
n0 2 t
l/2*" " h SI *WII Il A= 4 ((t 1/2 b I *
3.E = 931 am m = V,yAo -Ex 7,
. Q = mCoat
.' h = UA A T I=Ieo Pwe = W f d I=I 10-*/U o
'
- TVL = 1.3/u P = P010 sur(t) HVL = -0.693/u P = Po et /
SUR = 25.06/T SCR = S/(1 - K,ff)
CR x = S/(1 - K,ffx)
SUR = 26a/t= + (s - o)T CRj (1 - K,ffj) = CR2 (I ~ "eff2)
.
T = ( t*/o ) + ((a - o V Io] M = 1/(1 - K,ff) = CR;/CR ,
T = 1/(o - s) M = (1 - K,ffa)/(1 - K,ff))
T = (s - o)/(Io) SOM = ( -K,ff)/K,ff a = (K ,ff-1)/K ,ff = M ,ff/K,ff
-
L' = 10 seconds I = 0.1 seconds-I o = [(t*/(T K,ff)] + [I,ff /(1 + IT)]
Idli*Id P = (t4V)/(3 x 1010) I)d; 2 ,2gd2 22 2 t = 2N R/hr = (0.5 CE)/d (meters)
,
R/hr = 6 CE/d2 (feet) ,
Water Parameters Miscellaneous Conversions '
)
I gal. = 8.345 le curie = 3.7 x 1010 eps ;
1ga].=3.78 liters 1 kg = 2.21 lem '
1 ft4 = 7.48 ga np=2.54x103Stu/br Density = 62.4 lbs/ft3 1 mw = 3.41 x 100 Btu /hr Density = 1 gm/c9 lin = 2.54 cm .
Heat of vaporization = 970 Stu/lem 4 = 9/5'C + 32 l Heat of fusion = 144 Stu/lbm *C = 5/9 ( T-32) l 1 Atm = 14.7 psi = 29.9 l's. H BTU = 778 ft-lbf .
I ft. H O
= 0.4335 lbf/i H
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al.__ERINCIELES_DE_WUCLEAR_EDWER_ELoWI_DEEReIIDW2 PAGE 21 IHERHODYNAMICS2_HEAI_IRANSEER_oWD_ELUID_ELOW ANSWERS -- COOK 132 -85/07/16-JAGGARe ANSWER 1.01 (2.00)
a. 3
, E1.0 v REFERENCE Cook Thworve PP. I3.15-1 KA001/010sK3.37s ANSWER 1.02 (1.00) REFERENCE SONP, 6 Factor Formula lussone pr. 3-5 KA001/000eK3.26, Cook Theurve Pe I-5.2-1 ANSWER 1.03 (4.00) b. 3 c. 3 A REFERENCE SONPe Revivw of cure Puisons issuune P. 6 KA004/000iK5.20, Cook Theurve.Pr. I-3.57-7 ANSWER 1.04 (1.00) REFERENCE SONP, Rwview of Neutrun Kineticue n. 5 KA001/000eK5.49, Cook Theorve Pr. I-3.3-1 ~
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IWERMODYMAHICS4_HEAI_IRANSEER_AWD_ELUID_ELOW i ANSWERS -- COOK 112 -83/07/16-JAGGAR, .
'! ANSWER 1.03 (2.00)
a. 3 i b. 4
.
REFERENCE SONPe Suberitical Multiplication lusuone'r. 31 Ruview of Kineticue e. 3 Cook Theurge Pr. 1-4.13-1 KA001/000eK3.49, ,
ANSWER 1.06 (1.50)
1. e (same)
. a (ACP hisher)
3. b (ACP lower) CO. 3 e REFERENCE SONPe Review of Cure Poisonne re. 4-7 KA001/000eKU 18, '
Cook Theurve Pr. I-36-4 .
ANSWER 1.07 (1.00) d4.,I43ceja4sl.o n REFERENCE Shearon Harris Curvv Bouke Curve C-1 KA001/000eK3.13e Cook Curve Book- Figure K3.34, K5.17e ,
ANSWER 1.08 (1.00)
a REFERENCE HBRe Reactor Theurve Sessionu 41 and 42 DCC, WNTC Theorv twxte er. 1-4.26 -2 . , ,
b
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IM E R MO DY N AMICS a_ W E AI_IR A W S EER_ A N D_ E LUID_ E LD'd
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ANSWERS -- COOK 112 -85/07/16-JAGGAR, l i
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ANSWER 1.09 (1.00)
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b
REFERENCE NUS, NucIvar Enerdu Traininge Hudule 3, Unit 4 i Westinghcuse Reactor Phwuicut Svet. 3r Neutron Kinetics and Suet. Se Core Phusics
.
HBRe Reactor Theurve Sessiunu 20 and 24 - 31
,
s ANSWER 1 10 (1.00)
, (2) Core axial flux profil (0.5) TRUE-The curve is more linear (due to the additive uffvet of the rod worths at their luu values.) (0.5)
REFERENCE IP-3 ECI Rx Thvorva Charter 7e Pauvu 21, 22e and 27 DCC Rn Theurw Review Texte er. I-5.42 - .50
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ANSWER 1.11 (1.00) False
,
b. False c$ tid 4- / i CO.5 e REFERENCE Westinghouse Thermal Scioneer Chapter 13e Pr. 17-2 KA002/000rKU.01: J ANSWER 1.12 (1.50) True b. False i c. False C0.5 vach3 l
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- 1.__ERINCIELES_DE_WUCLEdR_EDWER_ELAMI_DEERSIIDWa PAGE 24 l IWERMODYWAWICSa WEAI IRANSEER_AND_ELUID_ELOW
ANSWERS -- COOK 112 -83/07/16-JAGGAR, ;
REFERENCE Westinghouse Thermal Scionec, Charter 10e PP. 41-49
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ANSWER 1.13 (1.00)
, FALSE
<
I TRUE [0.3 e i
REFERENCE (SGNP) General Phesics, HT1FFe er. 133 und 320 und KA036/000eKU.02 Subcuuled Liuuid Densite Tables K3.03: (Cook) Westinghouse Thermal ScienceeCharter 10e Pr 9-24 KA004/020,K3.06: Chapter 09e Pr 21-2 ANSWER 1.14 (1.00)
C REFERENCE i SDNP, HTFF texte re. 23 - 24 y Cooke Westinghouse Thurnal Sciencer CIWurter 7e Pe 32-4 ANSWER 1.13 (1.00) REFERENCE Shearon Harris-Thermu-LP-1.1 und utvun tablev KA002/000eKU.01, ,
Cook-Westinghouse Thurnal SciencueChapter 2ePr 63-70,7 /020eK3.06 3 4 ANSWER 1.16 (2.00) REHAIN THE SAME DECREASE INCREASE DECREASE REFERENCE Stvum Tables
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j I.__ERINCIELES_DE_WUCLEAE_EDWER_ELAWI_DEERSIIDWa PAGE 25 I IMERMODXWeMICSz_WEAI_IRANSEEk_AND_ELUID_ELOW ANSWERS -- COOK 132 -85/07/16-JAGGAR, ,
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ANSWER 1.17 (2.00) Decrease
- Decrease Increase
- d. Decrease CO.5 e i~ REFERENCE
} SGNPI HTFF, Page 15 WNTC There. and Hwd. Prin., Chap. 7 P e
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I 2.__ELfMI_DESIGW_IWCLUDING_EdEEIY_eWD_ EMERGENCY _SYSIEMS PAGE 26 l ANSWERS -- COOK 112 -85/07/16-JAGGARs '
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i ANSWER 2.01 (2.50)
, Motor driven- Lo-lo Invul in anu S/ .,
Trie of both MFW pump , S J1 Blackou Turbine driven- Lo-lo Ivvel in 2 S/G' '
UV on 2/4 RCP busse C.25 vach3 (1.5)
j The voerdenew leakoff valve oesn (.50) High flow through the resevetive pump (u). (.50)
. ta.1\ n\ La accep+ "(&O n&kn " s ogns\
REFERENCE Cook S.D. Charter AS-11er.42 33-3 KA061/000pK4 02e i K4.04, .
ANSWER 2.02 (1.00)
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REFERENCE
- Cook S.D. Charter NS-3, p.1 KA002/000eK1.09: >
t ANSWER 2.03 (2.00) Lutdown line downstream of orifievu. C.23 600 evid C.33 PRT C.23 Letdown line downstream of low Pres letdown valv C.23 200 puid E.13 VCT C.23
- Volume Control Tan [.23 75 puid C.13 HUT [.23 I P. D. Pune discharge [.23 2735 evidE.13 VCT E.23 Charging Pump suction [.23 220 puid E.13 PRT E.23 l C# b. Sed sJnW rekur n \ns [foUY pgy pg , $ ( , g} p g T- ( .1]
Y*UUS{*
REFERENCE Cook S.D. Charter NS-6e . KA004/000eK6.03 *
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2.__ELdWI_DESIGW IWCLUDING_SAEEIY_AWD_EMERGENCX_SYSIEMS PAGE 27
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ANSWERS -- COOK 112 -85/07/16-JAGGAR, ANSWER 2.04 (2.50) Excess Intdown (heat exchanger).
Reactor Coolant Pune seal water.
,
Minimum flow from the cent. charging nume E.25 es.]
. Leakrate Location e1 3 gen VCT
'
82 3 geh RCDT
- #3 100 cc/hr RCDT Outivt E.25 per uma13 d e-
' Through the (1509) relief in the uval return line to the PRT ag through the 42 seal to the RCD E1.0J s REFERENCE Cook S.D. NS-2, P. 47 and figure NS-2-2 KA003/000eK1.03 '
NS-6, p. 45 and figuru NS-6- KA003/000eK1.0Se2.78 NS-6e P. 43
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ANSWER 2.05 (3.00)
a. (SV-1023 Couldown return to luces 2 & 3 600 esig (SU-1031Couldown inist to RHR eumpu 450 esis (SV-104E10utlet of East HX 600 esis (SV-104W)0utivt of West HX 600 esis
' E0.25 va. response 3
' The minimum flow valves be20 rovide a lowpath protveting the e when thew are at a uhutoff hvad% n a low RWST luvul trim #yes rsutects the pumps when theu are inJucting istto the RC '
u.s/ .c., i Je p t= 44 coo Cp.a.r]
REFERENCE " 5O 9" ' " * " w g".e._cf a mw8"U'D **fA*
Cook S.D. NS-Be ,2 KA005/000eK1.04 K1.06e ,
K1.10, K1.11: K1.12e K1.13e K4.06 K4.07,3.2
-n-n.... - .
,- m-- 7e , , -- -- , , . - - - - - - - - , -
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i a 2.__ELCWI_DESIGW INCLUDING SOEEIY_oWD_ EMERGENCY _SISIEMS PAGE 28 i
ANSWERS -- COOK 112 -85/07/16-JAGGAR, t .
P
' ANSWER 2.06 (2.00)
t Hwdrogen enters viu natural circulation with the containment ai t 2. The air isPreheated(bu thedischargeunits).
i 3. Electric heaters raise the tumeurature of thu air to where hudrogen
' ' and oxwgen spontaneousiv recombine forming utea ' . The steam Passes into a mixing chambers mixed with cool air and sturnedtocontainment) - M * *P$**b E0.5 v (2.0)
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REFERENCE
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Cook S.D. Charter NS-12e . KA028/000eK6.01, .
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ANSWER 2.07 (2.00)
(Awi41
', High steam flue coincident with low uteum preuuuru or lo-lo Tav Hi-Hi containment Presuurugy 3 (1.0) second (0.25) A cwlinder Pressurized with full utcom ervsuuru iu vented across the valve seat to utsusehere creating a closing delta (0.75)
.
REFERENCE
- Cook S.D. Charter PGS-2Ae e.19-2 KA039/000eK4.05 L Un.Y A i ~ & Ska , pt a m,% \o T m .
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. w SW We. g+nsur N., 4, L b J fress w '
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. 2.__ELfMI_DESIGW_IWCLUD1HG_SAEEIY_AND_EUERGENCY_SYSIEMS PAGE 29 l
i ANSWERS -- COOK 112 -85/07/16-JACGAR, ,
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I f.NSWER 2.08 (4.00) . H.P.- muisture continouslu drains from one stude to the next until it Ivaves with the exhaunt utvaar except the first stadv which has a(une inch) drain lin (0.75)
2. L.P.- iraternal Pauvages and extraction nozzlvu drain continuous 1 (0.75)
, Windage is the heating of the turbine bludind (and air seucus)CO.53 caused bv friction of the turbine bludind [0 53 movind inside the casin (1 0)
e During startue and low loud uneration of the turbine 00.53: vindage creates high exhaust temperatures. The steam flue in not enough
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to cool the Isut turbine stadese .._ '_ 'Ndi ;"'b:
und the sp r ov u a r ePV t % -- 1.03 4 (1.5)
,
o b s4[e a m 4 % e_I4k sh3 0-REFERENCE Cook S.D. ChaPtvr PGS-34e Pp. 26e34-3 KA 045/010rK4.06e K5.02 :
ANSWER 2.09 (3.00) The which:
ruumdetectorssendasiunal(tothe relaucabinet)[0.253 1. Actuates the pre-discharde alarm timeru [0.253 2. Actuates the timer to control length of discharde E0.253
, Sunds alarms CO.253 The master valve uruns to surplu CO2 from the lluuid utorage tank to the surelv header LO.53 und the selector vulve uPens to supple the 4KV ruum from the supple houder CO.5 (2.0) The length of discharge C M erv~dischorde alarms are los E. g Cf r- (0.5)
c. Wintergreen odor in the ruum (0.5)
REFERENCE Cook S.D.eChaPter AS-18, P . KA086/000,A4.06, ,
A3.02, K4.04 3.1*
c
. . -
e.- - - - - . -
-. --4
_ _ _ _ _ . _ _ _ . _ . _ _ _ . _ .-.
,
.
! +
2.__ELAWI_DESIGW_ INCLUDING _SAEEIY_oWD_ EMERGENCY _SYSIEMS PAGE 30 i
ANSWERS -- COOK 112 -83/07/16-JAGGARe f i i ANSWER '2.10 (3.00)
,
' Safetw InJuction - maw amplifu thu effects of the accident which cauuwd the S Rx trir v/lu Tave- maw cauuv excessive couldown of the RCS and loss of SD I Hi-Hi S/G Ivvel -
maw cause excuusive moisture carruover to the turbin [.25 va signale .3 vu explanation 3 Trie MFW Puer Rapid closure of the FRV's Closure of the feedline isolution valve CO.25 each3 alIalso eccepF *pps We mainbb *
,
i REFERENCE Cook S.D. Chaeter ens 10e P. KA039/000eK4.19e i
.
.
-
--- , . - r -. -
-. - % -m, ~_ - - . . - - - - - - - . . - -
.. . - _ _ _ _ _ _ _
. - _ . - . . . . . . . . . - , . - . . . . .
i 3.__IWSIRUMEWIS_ FWD _COWIROLS PAGE 31 ANSWERS -- COOK 112 -85/07/16-JAGGAR, l t
,
I RNSWER 3.01 (2.50)
l
!
,
The low impulse Pressure uidnul will create a Power miumatch und cause
-
rods tu stor in. [0.53 (at maximum speed). In tinue the power minuutch to
'{ LO.53 and the mismatch devolured between
-
the rod control swstem will fud
', Trofe at the no-load valuwe CO 3 und Tuvde below nu-loude [%
' will attempt to take over 3 L,. will ...' ---- u ?. f ^$-_- E-_ _
^ ^^ rod motio " "
c.._'..i q?w." M (2.5)
i N .n*bu .. bM - -ar th M 4 ro A ( Wve m .
REFERENCE
,
Cook S.D. Chapter NS-4e Pp. 4,14-17e2 KA 001/000eK1.05: i f.NSWER 3.02 (2.00)
' Steam pressure is used to compunuatu the steam flow ujdnul for
'
densitu variations in the steam au euwer inervoue (1 0) Indicated steam flow will be higher than actua (1.0)
.
l REFERENCE Cook S.D. Charter PGS-7, Pr KA 059/000eK1.04, ANSWER 3.03 (4.00) To control the rate at which unurdu is removed from the RC (.50) Rosets the load reJaction bistables. (Clouvu the armind volveu) (1.0) Hun (.50) . Tu prevent overrroquurization pr the main condense will auce+ em. mg W JL e ia ce o* ile.lity,
'
i 2. To rrevent an uncontrolled couldown of the RCS following steam dump actuatio (1.0) Large load rejection C.253-bveauur un a turbine trip the ruuctor trie with it will reduce the heut duneration rotu that the steam duaru will have to hundiv. [.753 (1.0)
. . . . . . . . _ .
'
__
n _ . . . . ..- . . - . .
!
- '3.-_IWSIEUNEWIS dWD COWIEDLS PAGE 32
,
ANSWERS -- COOK 132 -85/07/16-JAGGARr REFERENCE
- Cook S.D. Chaeter 12, Pr. 7-8 18 28:29,31,4 KA041/020 KA35, ANSWER 3.04 (3.00)
Answer must include the following as a minimum for full credi Pzr. Invel will dversase due to charding < lutdow % Par. Invel isolateu letdow Pzr. Ivvel increases due to charging > Intdow At SE% Pzr. Ivvv1 r 9/ to.Il s\so e t.tyh'ah'<tu s t. r %t rie hecu rs .,sch Arvo *Stum eny AwLT SW* At y 9 ew c REFERENCE F * "**** *
Cook S.D. Chapter NS-3, r.35 KA011/000rK1.01: NS-6e P.19 K1.04: i'
NS-11er.50 K4.05:3 7*
K4.06 3.3
,
,
CNSWER 3.05 (3.00)
'
TRIPS PERMISSIVES
---------- -- ----- HI flux P-6, P-10 Hi flux 2. P-10
. Hi flux lo stet 3. P-10 . Pz Pz Iow Pressure hi level 4. P-7 5. P-7,
/P-10orP-13)
/P-10 or P-13) Single loor LOF P-8 M 7 Double loor LOF P-7 w/P-10 or P-13) M RCP undervoltage 8. P-7 /P -10 or P-13)
9. RCP uriderfre . RCP 'rvaker o trin 1 . P-7
/P-10 or F-13) Ug 1 ,,a .u. .n o s 11. P-7 P-7(w/P-10 or P-13) N 7 6/P -10 o r P-13),
Turbine /Roactortrie('e.-( a w c e)
E g vach trie namer yt"vach major perm., E _:P ""'"""O'"" "---
? D- (3.0)
M.13 2. I1 REFERENCE Cook S.D. Charter NS-11e e. 3 KA 012/000rK4.02,3.9 l
.
l -- -- -_ . - -- -, .. . -,_ _ _ _ _ . _ _ .
_ _ . . - - . . - - . . - - . - . - .
-
>
' '
t 3 __JMSIEUMEWIS_CND_CONIROLS PAGE 33 ANSWERS -- COOK 112 -85/07/16-JAGGAR, i
!
! ANSWER 3.06 (3.50) l
< 1
'
i
, Its associated SUR indication woul be more nudutive than the
'
other(when rower ischanging) und approximatulu 5-10 minutes after a trip it will indieute sure nudativu than
- -1/3 dem CO . (0.75)
t
" (0.75) I/R 10 -10 ames = 10 4 CPS on the S/R I
'
, . No effect on the detector current meterse but it will cause the power meter to move accordindl (0.75)
i 2. Caution should be excursied when udJuutind this eut us actual Protvetive funtions can be inutute (0.75)
, After a calorimetric (if reuuired). (0.5)
', REFERENCE Cook S.D. Chapter NS-9e Pr.10-22,23 28-33,3 KA015/000sK1.01, K3.01, K4.01 i K5.02, ! K5.03r2.3*
i A1.01, A1.02 A1.03r A3 03 ,
t t
&
- ~ ~ - ~ . . . - -, .
L 3
- - .. _ ._- _ _ - -_ __ - _ _ . -
9# ^*
-__- ._ _ _ ,.++.-f*W6we-.
'
W % W= .A h6 M & W 4 9 W w MM4#==i' 9
'
3.__1MSIEUNEWIS_eWD_CDWIROLS PAGE 34 ANSWERS -- COOK 182 -85/07/16-JAGGAR, t i
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l ANSWER 3.07 (3 50)
, No.C.253 One PORV's (NRV-152) ervuuure signal convu from the
Pressure controller which makes this signal rate sunnitiv This valve would open before the other two on a rapid
- Pressure rise. C.53 (0.75)
. Yes. C.253 Two PORV's (NRV-151 und 153) recieve their
- Pressure signal from the same detvetor. Should this detector i fail highe both PORV's would open. C.53 (0.75)
' ' Pressurizer Pressure would begin to drop due to 2 PORV's openingt.53 At 2250 rsig the cueling heaters would energize and be full on bw 2220 esis. C.253 i At 2210 rsig the backur heaters would energize. C.253
- At 2150 esis the oevn PORV would recieve a blocking signal
, } from another Pressure channel and would shut. C.53 i Pressure would then rise until the blocking bistable rese '
l The PORV would reoren and Pressure would then ewelv from
'
between 2150 and the bistable reset value. C.53 (2.0)
i
,
, ! REFERENCE
i Cook S.D. Charter NS-3e Pp. 30-33,49 KA 010/000rK3.01,3 8
'
- and figure NS-3- K4.03,3.8 i
- A2 03 i
.
!
i-
!
I I
I t i
,
I
}
l I
l
.
p-= e <e
. ._. . - . ,.-.- - . .-.--.1_._._- . - - - - . - - . . _ , - - . _ _ _ - , - - . . _ . .-. - - - - - -
I .
a3.__IWSIRUIENIS_CND_CDWIROLS PAGE 35 ANSWERS -- COOK 112 -85/07/16-JAGGAR !
}
\
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'ii CNSWER 3.08 (3.50)
l
- . From 400 volt AC bus through CRID inverter to CRID rane ', 2. From 600 volt AC bus through batterv charder und CRID inverter
. to CRID Pane I 3. From batterv to CRID inverter to CRID Pane . 4. From 120 volt lighting (normal or standbu) through distribution
, Panel CRP-3 to CRID Pane [.5 each2 (2.0)
. wil .4s accepts % 3 lighbg pe<J %% ALT b CRf'-l +o CRso .
, AC orens and is disabled from control ruum operation while the
.
emergenew diesel overstes from an emergenew start. L.53 Thiu prevents 5 the voorgenew dicsv1 generators from operating in parallel through this breaker. [.253 (.75) . 600 volt AC tiv breakers 21AC and 21BD (11AC and 11BD) upon and
< '
lockout.Co. Ad3
- 2. Non-essential 600 volt AC motor feeder breakers trie.Ce.Ji] (.75)
i REFERENCE Cook S.D. Chaeter PGS-15, Pr. 7-9e41e5 KA015/000eK2.01: ,
KA064/000eK4.01: ' K4.02: I
.
!
l i
i
.
'. ..
'
. - _ . _ _. - _ _ _ _ . _ _ _ _ - ____ - -
_
. . . ._ _
-.. . . - - - - . . _ - . . - - . . . . . . ...
l
'
'd*__ERDCEDURES_=_WORMOLa_ARWORMALa_ EMERGENCY _eWD PAGE 30
. R&DIOLDOICAL COWIROL i
i ANSWERS -- COOK 112 -85/07/14-JAGGARe !
1
.
t CNSWER 4.01 (1.00)
?
l #
, - REFERENCE I DCC Technical Sevcifications 3.1. SYS. GEN. 002/020 24, I l GEN. K/A 24, +
KA 002/000, K5.18, .
}
v ANSWER 4.02 (4.00)
,
!
- a. The hut calibration meters will be reading hidher than the actual levele [1.03 thereforve the heuters are sure likelv
- to be uncovered. [0.53 (1.5)
i l b. The RCS rressure vill increas (1.0)
c. Tu Prevent a rossibie luckue situation of the rudu und CRDM'u (1.0)
,
,
! d. TRU (0.5)
!
.
!
REFERENCE DCC 1-OHP-4021.001.001, er. 3-5 KA 011/000, K4.07, 29 K5.13e 3.28 i KA 004/010e K5.05, i
- KA 001/010, A1.02, ,
PLA. GEN. 9, 35
-
i
!
I i CNSWER 4.03 (2.00)
i i TRUE
!
b. FALSE c. TRUE FALSE ;
I REFERENCE DCC PMI-21109 P , 5, 8, 11 PLA. GEN. 11, 3.6*
,
l
.
! . . - . - - -
.
'
. . - - - - .- . _ _ _ _ . , _ _ _ _ _ _ - - , . . . . . _ . - - - - . - _ - . , - - - - -
-
.. . . . _ .
. . . . - . - . . . . . - . . . . - . .
-
'
.A*__GEDCEDURES_=_WORU^L4_oBNDRUdLa_ECERGENCY_AND PAGE 37 RADIDLOGICAL_COWIROL ANSWERS -- COOK 112 -85/07/16-JAGGARe !
J l
i
,I ANSWER 4.04 (1 50)
i % change in indicated level (10 cu. f t . ) o r- /N E9c sn 4 vueh0.5)
i b. Dw noting the difference in current Ivvel und the Invel recorded in dresse rencil on the buttom of the Invwl indicator cover. [0.753 Determined once euch shift. CO.253 (1.0)
i,
!,
l REFERENCE
,
'
,
DCC Standing Order 050 013 PLA. GEN. 9e SYS. GEN. 006/050, 25, : 27e i 36, !
> .'
'
l
i CNSWER 4.05 ( .50)
. Unit Survrvisor (0.5)
REFERENCE DCC Standind Order 050.061 SYS. GEN. 006/030, 22, 3.1*
. .
t
,
f
{
i I
- .- .-
, _ . ----- -- - . , - -,m - . . , - ,.-.p - . _ _ _ _ , . - _ . _ -
, - . - , , . .-
. - _ _ _ . _ .-
_ _ _ . . _ _ . . . . _ , _ . . . _ _ _ .
'
.
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d._JERDCEDURES =_NORreLa_dBWOR22La_E"ERGENCY_dND PAGE 38 RADIOLOGICAL _CONIROL ANSWERS -- COOK 112 -85/07/16-JAGGARe ,
t
, ,
3 l ANSWER 4.06 (2.00)
- , 1. Notifw control roo . 2. Warn all Personnel in the are
3. Extinguish small fires.
,
1 Chuck injured Personne . Isolate swstem in the event of a ruptur P 6. Locate Protective svar that mau be rwouire . Cordon off the are i~
8. Follow instructions issued bw S '
(4 reuuirvde 0.5 va.) (2 0) ,
REFERENCE
' FLA. GEN. 27, 4.1 l
DCC PHP 2080 EPF.013 i 28, $
i ANSWER 4 07 (2.50) Control room annunciatu (0.5) Tavd is'areroximated from S/G ereuuure using an uttachment for water Pressure vs. teureratur (1.0)
'
, . Verifw reactor tri . Initiate energenew boratio [0.5 w (1.0)
.
REFERENCE DCC 1-OHP-4023.001.011e er. 2, 3 PLA. GEN. 21e , 4.3
2e 4.3 l
l .
l
'
l --- .
- . _ _ _ __ _ _ _ _ _ _ ._ _
-. . _
_ _ _ ~ . _ _ , _ . - _ . _ _- .
I, .
ed.._CROCEDURES_=_ WOR 2dLa_ABWORMALa_ECERGENCY_CWD PAGE 39 RADIOLOGICAL _CONIROL ANSWERS -- COOK 112 -85/07/16-JAGGott e ,
i
- t t
.
i
,
,
ANSWER 4.08 (1.00)
b.
r REFERENCE
'
DCC 1-OHP-4023.016.001, SYS. GEN. 024/000 24r KA 003/000, A4.05e K1 03, l
.
i
.
ANSWER 4.09 (1.50)
. The RPS senses a loss of RCP und when >50% Power the unit will
? tri . (1.0)
i, b. Various Steam Dune Valves will full open. Until ulturnate power is selected or P-12 uctuate (0.5)
'
REFERENCE
} DCC 1-0HP-4022.082.001 KA 000/037: K3.01, ! A2.03, l A2.19, i i
ANSWER 4.10 (2.00)
,
. . . Initiate S . Trip the reactor und voerdeneu burat . Recover druered rods.
'
4. Perform order 1w shutdown to Hot Standb [0.25 v (1.0)
,
REFERENCE DCC 1-OHP-4022.012.004, pr. 1e2 KA 001/000, K3.01, 2.9*
3.02, 3 4*
A2 04, .09, 3.9 l
l l
l l
. . _ _ . . - .
{
,
.- _-. - -- -- _ _ __ _ . - -. . . . _ . - - -____
.
_ . . . _ _ _ , . _ _ _ _ _ . . . _ _ . _ _ . _ - _ . . ._ _ -
r
i
- d._JREDCEDURES =_ WORM *La_iBWORMALa_EMERGENCX_AND PAGE 40 l RADIOLOGICAL _COMIROL ,
ANSWERS -- COOK 152 -
85/07/16-JAGOARe *
'
l A
ANSWER 4.11 (3.00) <
- . Restore the indicated AFD to within the tardut band immediatelwe CO.53 or 4 2. Reduce the thermal Power to <90% of rated thurnal power. [0.33 (1.0)
, l Accumulated Penaltw over the Past 24 houru is 89 minutu [1.03
, The pensitw will be reduced to 60 minutuu at 1618 un 3/19/85 and then ruwer saw be increase E1.03 (2.0)
{
i REFERENCE HBR Tech Spec 3.10 , -6 KA 001/000 KU.34, DCC Tech Spec 3. /4 2-3 i
i ANSWER 4 12 (1.00) REFERENCE
,
'
SONP AOI-4De e. 1 of 2 KA 015/000 K3.01: DCC 1-OHP-4022.013.004 3.02, ANSWER 4.13 (1.00)
1. Remain Shutdown
'
2. Startup l
3. Remain Shutdown 4. Remain Shutdown CO.23 vach3 REFERENCE SONP Technical Seucifications 3/4.4. KA 002/020 SYS. GEN. 23, DCC Tech Spec 3.4. , ,
[
- & s' o
=r - - - - - - - - -+- - y, . - - - ... -- - - - -
t
'
l
! . ROCEDURES_=_WDEMLa_ABWDE"2La_ EMERGENCY _dWD PAGE 41
'
R&DIDLOGICAL_COWIROL ANSWERS -- COOK 112 - 85/07/16-JAGGAR, >
l
l
i RNSWER 4.14 (2.00) .0 REM / Quarter (0.5) ! . Accumulated dose does not exceed 5(N-18)
O 7 2. NRC-4 documentation complete i
' 3. Written rurainsion from Tech.-P.S. Superintendent and the
'
sponsoring department superintenden (1.5)
i (
REFERENCE l PMP 6010. RAD.001 PP. 35, 36 PLA. GEN. 12, i t
I
.
!
,
s
.
.
t
b
i
,
..
l !
l l
. _ . . - . . - _ - . . . . , _. .. . _. _ , . ___ _ . , _ . , ,
- - . - - - - --y.-+--*--.------w m = e-e v. w-. - - - . - - , - - , - - - , y-r,----w-- - - - -- ---r+y --
y*-,
.- -. . . _ - - .. - . - , _- - .=- . _ _ _ .
I
+
!
..' % *
n b E... .,R C O PaV
'
. U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR IICENSE EXAMINATION I
.
!
, . FACILITY: .D.C. COOK ci j .1
' REACTOR TYPE: WESTINGHOUSE
~'
DATE ADMINISTERED:' JULY 16.1985
,
EXAMINER: T.D.REIDINGER APPLICANT:
'
INSTRUCTIONS TO APPLICANT:
i Use separate paper for the answers. Write ' answers on one side only. Staple question l sheet on top of the answer sheet 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%.
% Of
', Category % of Applicant's Category Value Total Score .Value Categorv j Theory of Nuclear Power Plant
Operation, Fluids, and Thermodynamics
' Plant Systems Design, Control,
! and Instrumentation 1 Procedures - Normal, Abnormal,
,
Emergency, and Radiological Control
.; Administrative Procedures, Conditions,
and Limitations
,
i TOTALS i
'
,
l Tinal Grade %
!
All work done on this exam is my own, I have neither given nor received afd.
J
'
i ,
'
I
<
Applicant's Signature i
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. . -. . - _ _ _ - _ _ - _ _ _ _ _.- _ _-. _ _ __ _ ___
.
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Section 5 - Theery of Nuclear Pcwer Plant Operations, Fluids and Thermodynamics
!
l Which of the folicwing is NOT a characteristic of subcritical (1.0)
j multiplication?
4 If the reactor is shutdown long enough, the source range instruments will lose their ability to determine the subcritical multiplication level even though the core may still be at FOL.
- Doubling the indicated count rate by reactivity additions will reduce the margin to critical by approximately one hal For equal reactivity additiens, it takes longer for the equilibrium subcritical multiplication level to be reached as Keff approaches unity.
- If ten steps of rod withdrawal increases the subcritical
, multiplication level by 10 cps, then twenty steps of rod l withdrawal will increase the subtritical multiplication
- level by approximately 20 cp i Which of the following will NOT change over core life? (1.0) The acceptable AFD target band.
1 The minimum acceptable shutdown margi t The control rod reactivity worth.
, _The power defect reactivity wort .3 Which of the following den:enstrates the effects of the (1.0)
delayed neutron fraction changing over core life?
l A icwer boron concentratio A higher rod bit A higher startup rate for equal reactivity addition A larger (more negative) moderator temperature coefficient.
J
- . .
.-- - . - - - - - - - . - . . -.
- ._ _ . _ _ _ _ _ _ _
.
. An estimated critical position has been calculated for a reactor (1.0)
startup that-is. to be performed 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> after a trip following a 60-day full power run. Which of the following actions will contribute to a higher actual rod positions than the calculated ECP?
a. Feeding the steam generators to increase level by 15%.
b. Delaying the startup six (6) hours longer than anticipated, c. Increasing the steam dump pressure setpoint by 100 psig, Increasing the pressurizer level using the d ute mode boron concentration contro /l v/7 / Which of the following will contribute to a smaller (less (1.0)
negative) doppler-c -fficicnt over core life?
RFW a. Fuel densificatio b. Clad cree c. Crud buildup on the fuel cladding, d. Fission gases released to the gap between the fuel and claddin .6 Which of the following is a true statement concerning the (1.0)
moderator temperature coefficient?
a. The MTC tends to drive the neutron flux toward the top of the core over core life.
b. The MTC increases (more negative) as the boron concentration increases, c. The MTC effects the axial neutron flux distribution n; ore than the radial neutron flux distribution, d. The MTC will not change with a change in temperature if the boron concentration is maintained constan ,
, y, ., - - - . - - , , , , , - ,, ,
. - .- . .. - -_ - . - . . - . .
.
. Which of the following statements describes the behavior of (1.0)
Xenon and Samarium? After a reactor trip occurs, xenon concentration initially increases and samariun. initially decreases.
l After a reactor trip occurs xenon will eventually decay to a xenon free condition, but a samarium free conditicn will not occur until after the next refueling outage, The xenon and samarium peak concentration following a t' rip ,
occurs at a time independent of the previous power leve Xenon concentrations may increase or decrease when taking the plant from Mode 3 to full power but. samarium will always
,
decrease during this transient after the core's equilibrium samarium has been reache .8 Which of the following radioactive isotopes found in the reactor (1.0)
coolant would NOT indicate a leak through the fuel cladding? Xe-133 Co-60 Kr-85 The largest contribution of hydrogen released to containment (1.0)
due to an accident involving inadequate reactor vessel void formation is from? recoolingand)
- 067- Q56- a zirconium - steam reaction an aluminum - steam reaction the release of dissolved hydrogen in the coolant from the hydrogen overpressure on the volume control tank radiclysis of the coolant
'
5.10 Void fonnation can occur in the RCS during certain accident (1.25)
conditions while on natural circulation resulting in a phenomenon called reflux boiling phenomeno Explain how reflux boiling cools the reactor core? (Bespecific.)
5.11 DC Cook Unit 1 is at EC power when the reactor tripped due (1.0)
to a feed regulating valve failing open. While reviewing the post-trip log, the STA notes the first out annunciator ,
was negative flux rate. Additional alarms indicated S/G's ;
levels rose higher and tripped the turbine. The STA also '
noted that power dropped from 20*/ to 16% before the tri Describe why it is possible to have cold water accident as this and yet have a power drop. /fQ/5 /NMNdh
J
. .- _ _ _ .
,.
'
5.12 What is the highest value of the Quadrant Power tilt ratio for (0.5)
which no corrective action is required?
5.13 Describe how the change in enthalpy across the steam generators (1.25)
1 changes as reactor power increases from 70% to 100% power?
5.14 At equilibrium operation at 70% power, a feedwater heater ,
malfunction causes a lower feedwater temperature. Explain how this transient affects the following parameters, including reasons for any changes and identification of how equilibrium values differ from the initial conditions. Assume all controls
, are in aut T average (0.75) , Reactor Power (0.75) VCT Level .(0.75)
5.15 Explain how Xenon oscillation could be caused by an inadvertent (1.0)
boron dilution. Be explicit as to what happens in the upper and lower core sections. (Assume Rods in Auto.) How does the operator suppress a Xenon oscillation? (0.5) What would most likely happen if the operator took no action (0.5)
other than to re-establish bcron concentration?
5.16 List four (4) significant factors which must be considered in (1.0)
,
calculating shut down margin at powe .17 Refer to figure 13.1.
r (0.5)
What is theatprimary a (+1.1.%) reason 501 power vice afor the dog)
(-24.45; house at 50% to be limited to(Pointa.)
power? What is the bases for the limit of 92% maximum power level? (0.5)
,
(Point b.) Are we limited to operating inside the target band 5% at 917 (0,5)
Defend your answe .18 What are three (3) reasons for establishing regulating (1.5)
group insertion limits? When is the violation of the rod power dependent insertion (0,5)
limits acceptable? What is the four (4) pump zero-power rod insertion limit? (0.5)
1
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5.19 Explain why a severely damaged core could experience a (.75)
re-criticality even though all control rods are inserted and the proper boron concentration is maintaine .20 A positive reactivity addition occurs in the core after a trip (.75)
from power because of the increase in concentration of a certain fissile isotope. What is the name of this fissile isotope and why does its concentration increase after a trip?
5.21 If the moisture content of steam from the steam generator is (1.25)
excessively high, will the power level calculated using the heat balance be erroneously high or low? Explai END OF CATEGORY 5
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Section 6 Questions - Plant Systems Design, Control and Instrumentation The reactor is in hot shutdown and primary pressure is (1.5)
1850 psig. You are the SR0 and Mr. Goodwrench (C&I)
wished to conduct maintenance on the turbine first stage pressure instrument. Would you authorize him to do the maintenance? Defend your answe .2 According to DC Cook Technical Specificationc, the Power (1.0)
Range Nuclear Instrumentation is only required in power operations. In spite of this fact, no more than one channel can ever be removed from service during shut down
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modes. Explain wh .3 What automatic start signal (s) for the Motor Driven Auxiliary (1.0)
Feedwater Pumps (MDAFP) is/are lost when the switch is in the after-trip or neutral position? Which of the following components provides the most significant (1.0)
reduction in cesium from the reactor coolant when placed in service? The reactor coolant filter upstream of the VCT
! The cation bed demineralizer The mixed bed demineralizer The deborating demineralizer Recently DC Cock in mode 1 experienced a reactor trip and safety
, injection at 68% power due to the loss of a control rcom Instrument distribution IV inverter.
l Explain the exact cause of-the reactor tri (1.0)
i l Explain why the safety injection occurred (1.5)
after the reactor tri .6 Which of the following will actuate the Unit 1 (1.0)
Rod Control Urgent Failure annunciator light? Internal failure of the power cabinet because of failure of the slave cycle Internal failure of the power cabinet due to the removal of a printed card l Internal failure of the logic cabinet from phase failure detector Normal alarm occurring during the dropped rod recovery procedure
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. You add water to raise the level in the RCP Number 2 seal (1.0)
standpipe. Assume no other changes in the system. Would you expect any changes in No. I seal leakoff flow? Explain your answe .8 Power reduction normal operation procedure, OHP 4021.001.003 (1.0)
step 6.12 states, " Transfer the feedwater controls from auto to manual which normally is done between 15% to 20% of power."
What is the primary reason for this step in not using automatic control of the main feed regulating valves? Be specifi .9 DC Cook has experienced a station blackout and Mr. Goodwrench (.75)
was authorized to prepare the CD emergency diesel generator and output breakers for local operation. What general guidelines would you as SR0 give to the C&I technicians concerning the specific leads and wires which must be lifted to disable the diesel from control room?
6.10 Why is speed droop a desirable characteristic for parallel (1.0)
operation but undesirable for safeguards actuation?
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6.11 Mr. Goodwrench has the one (1) CD diesel in test mode and (1.0)
loaded 2800 KW and a spurious SI signal is received from the results of a faulty test procedure. How does the diesel
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generator and its electric load respond in this condition?
6.12 How does the Cardox System fail on loss of power? (1.0)
6.13 Would the failure of a turbine first stage pressure channel (1.0)
result in an inadvertent actuation of the steam dump? Explai .14 List at least three (3) plant service system whose failure (1.5)
would prevent the opening of any of the steam dump valve .15 Rods are in automatic at 210 steps and Mr. Goodwrench just (1.0)
reported to you that he has experienced a dropped rod (no scram) and rods haven't moved out to compensat Explain why rods didn't move out to compensat .16 While performing a cool down with RCS terperature at 450*F and . (1.25)
pressure at 1100 psig, Mr. Goodwrench (CEI) requests that he be allowed to calibrate two (2) 0-60 psig containment pressure transmitters simultaneously. He states that this is acceptable
, since the containment spray logic will not be made up. Would
- you allow this work to occur? Defend your answer, t
6.17 Indicate whether the following statement is true or false. On (1.0)
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Unit 1, the automatic reactor trip signal will cause both the undervoltage coil to de-energize and the shunt coil to operat .18 On the ground detections system the brightest test light -(0.5)
indicates the phase with the ground.
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i 6.19 Give three (3) plant operating or accident situations that (1.5)
j require the Breaker Synchronizing Selector switch for each 4160 volt normal bus to be in the auto positio *
6.20 For each of the Emergency Diesel Generator HEA lockout trips listed below indicate whether the trip functions only during i normal testing (surveillance) or will function during either ;
abnormal testing (SI, blackout etc.) or normal testing, low lube oil pressure (.50)
- generator phase differential (.50) overspeed (.50) overcurrent (.50) >
6.21 List the radiation monitoring systems that respond to a (1.0)
high radiation signal and initiate automatic actions to prevent the release or spread of radiation at DC Cook Unit '
Be specific, list five (5) out of the ten (10) system List all the automatic actions which are initiated in the (0,5)
event a high radiation signal is detected for five radiation j monitoring system END OF CATEGORY 6 i
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Section 7 - Procedures - Normal, Abnormal, Emergency and Radiological Control The techriical specifications require that the Unit 1 Boron (1.0)
injection tank be operable. With the Boron injection tank inoperable, in certain modes, actions trust be taken in one hour to restore the tank to an operable status or further action must be taken. Which one (1) of the following describes the applicable modes and the further action must be taken?
Applicable Mode _s Further Ac,t_io,n , 2, 3, and 4 Be in cold shut down and borated to a shut down margin equivalent to 1% delta K/K at 200*F within the next 20 hour2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> , 2 and 3 Be in hot shut down and borated to a shut down margin equivalent to 1.6% delta K/K at 200 F within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, , 2, 3, and 4 Be in hot standby and borated to a shut down margin equivalent to 1.0% delta K/K at 200 F within the next 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> , 2, and 3 Be in hot standby and borated to a shut down margin equivalent to 1.0% at 200*F within the next six hour .2 List five (5) areas at DC Cook which are formally designated (1.25)
as " Extreme High Radiation Areas." If you are using the manipulator crane during a refueling (1.0)
and an interlock f ails, whose approval must be obtained in order to bypass the interlock? The plant is at 100% power and Mr. Goodwrench reports to you (1.5)
that the train N Batterycharger has been removed from service for maintenance for four (4) days. What are' your actions as i SRO per applicable Technical Specifications? Procedure OHP 4021.017.002 (RHR Operation) cautions the operator (1.0)
not to throttle component cooling to the RHR heat exchanger for-temperature control. Explain this precautio .6 DC Cook, Unit 2 is in a critical, hot standby condition at end (1.5)
of life and peak Xenon. A Xenon follow test is in progres Mr. Goodwrench (C&I) technician requests your permission to
-
remove the steam flow channels from service to blow down the level pots in preparation for refueling shut down. As the Shift Supervisor would you permit this evolution? Explai .7 A primary to secondary leak is detected during full power (1.0)
operation at noon on day 1. Prior to this time, no primary to secondary leakage has been detected. What is the latest
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time at which the unit must be in hot standby if the initial leak rate is .2 gpm and is increasi by .003 m every eight (8) hours? Defend your answe S ho Inthe"LossokallACpower"OHP 4023.001.016 there is a (1.0)
caution statement that says: The non-faulted steam generators are to be depressurized as quickly as possible. You may not decrease RCS pressure below 390 psig or core exit thermocouples below 446 Two (2) reasons exist for these limitations; l provide one of these reason .9 The H.P. Technician reports that he has detected a small piece (1.0)
of metal on the floor in the auxiliary building which is reading 6 rem ganma/hr at one foot. How far from this source should you erect a Danger - High Radiation barricade?
7.10 What are your actions in the event a Safety Limit is violated (1.5)
per Technical Specifications? List three (3) action : 7.11 Recently DC Cook has written two (2) LER's concerning (1.25)
noncompliance with the Technical Specification that the flow
,
rate of reactor coolant to reactor vessel shall be _3000 gpm whenever a reduction in reactor coolant system boron concentration is made. What is the basis for this Technical Specification for mode 6?
7.12 Mr. Goodwrench has nctified you that he has received a alert (1.25)
level alarm followed by a high level alarm on the failed fuel element monitor. What are your immediate actions per the emergency procedure?
7.13 List all the amergency procedures which uses the Emergency (1.0)
Procedures "Immediate Actions and Diagnostics OHP-4023-081-002" for assisting it, diagnosing and identifying transient condition .14 The plant is at 1001 power and Mr. Goodwrench (RO) informs you (1.25)
that the plant has symptoms or indications of decreasing generator load with constant reactor power. What invrediate actions per abnormal procedures would you reconmend in
.
dealing with this problem? (List 4)
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7.15 DC Cook abnomal precedures list five (5) specific categories by which emergency boration is require ' List the five (5) specific categories for emergency boratio (1.0) List two (2) conditions by which you would terminate emergency (1.0)
boratio .
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7.16 Mr. Goodwrench has inforn.ed you that he has received a high (1.25)
activity alarm on monitor R17-B (CCW system). Plant is at ,
61% pwr and there has been no change in equipment status since last shift. What are your actions as a SR0 per procedure to identify and isolate if the count rate continues to increase?
7.17 What are the operator's emergency actions when receiving a (1.25)
100 psig control air pressure low alarm and air pressure is dropping steadily?
7.18 Which one of the following is a symptom of inadequate core (1.0)
cooling per procedure OHP - 4023.001.015. (Choose One)
a. Three (3) core exit themocouples monitored by plant computer exhibit readings at or above 1200 b. Computer not available, two '
(2) out of four (4) hot leg RTD's are pegged hig c. Five (5) incore thermocuples are off scale above 1200'F and emergency core cooling is not being delivered to the Reactor Coolant Syste '
d. Five (5) incore thermocouples are off scale about 700 F and auxiliary feedwater is not being delivered to the intact steam generator, ,?c4 3 e 4*P% 49 b % e_ R c_ L
% .19 a. What is your plant's operating philosophy concerning the use (1.0)
of er.ergency procedures ~for a safety injection initiated due to an event which occurs below the low pressure SI setpoint?
b. Whose approval is required to with hold or reset any ECS (1.0)
equipment for the conditions stated in part a?
7.20 In the procedure OHP 4022.008.002 " Initiation of ECC (1,0)
Recirculation Phase", a statement is made that after a
,
loss of coolant accident, resetting of the safety
injection signal in order to change over to recirculation phase must be delayed as long as pcssible. Explain the reason for the delay.
END OF CATEGORY 7
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Section 8 - Administrative Procedures, Ccoditions and Limitations
' If a system has been returned to service by the clearance permit (1.0)
procedures, a valve lineup need not be done on system during a startup system lineup. True or false? You have experienced a delay of 24.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> since commencing (1.25)
reactor startup operations. What direction do you as the SR0 provide to Mr. Goodwrench (the R0) concerning the status of the completed system lineups and surveillance tests? Which of the four (4) emergency classifications is it required (.50)
to sound the Nuclear Emergency Alarm and announce directions over the P.A. per the intnediate DC Cook action guidelines? List two (2) occasions in which the Shift Supervisor will be (1.5)
responsible for the initial notifications to individuals or agencies instead of the Plant Manager as normally specified in the initial off-site notification precedur .5 Radiation safety does not take precedence over repair or damage (1.0)
control functions during natural emergency condition True or false? Which of the folicwing is a 10 CFR 20 occupational dose limit (1.0)
that does not require form NRC-4 to be kept on record? rem /qtr - whole body
, mrem /qtr - whole body I rem /yr - whole body mrem /qtr - hands and forearms Recently an operations meno was issued that if any roof vents (1.0)
or doors are opened while fuel is being moved in the auxiliary building, then the fuel handling operations must be stopped while the doors are in the open position. What is the basis for this precaution? Mr. Goodwrench (IEC) technician r.iissed a weekly hot shutdown (1.0)
surveillance and the plannelp]pnt to enter the hot standby mod What conditions have to be met prior to entering the next higher mode of operations? (List two (2)). For change sheets which are not considered to be "on the spot"; (1.0)
changes to review and approval must be completed prior to placing the change in effect. List the signatures required to complete the chang (Listfour(4)).
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8.10 When included in the sequence of a procedure, the activities (1.0)
,
required by hold points shall be completed prior to continuing
- work beyond that point. True or false?
8.11 Choose the following statement (s) which is/are correc (1.0)
a. Acting department head may act as the pennit holder in cases when the permit holder cannot be contacted for unexpected reason b. A red tag and a striped tag clearance permit can be issued in emergency conditions for the same component at the same time, c. All of the isolation points for a clearance requests can be divided among the multiple clearance permit ! d. Permit holder can leave the area but not the site while
- performance type testing (no safety or equipment damage concerns) is in progres .1 Mr. Goodwrench (I&C) technician wishes to remove one emergency (1.5)
diesel from service for maintenance. As a SRO, what are the
- requirements which must be satisfied prior to removing one (1)
diesel from service at 100% power?
8.13 The reactor operator assigned to the control room has suddenly (0.5)
taken ill. The unit supervisor may relieve the 0 for limited
- 8/)& //
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l time not to exceed two (2) hcur Yes or no?
Defend your answe (.5)
8.14 Standing orders can be used to change any procedure.or (1.0)
instruction used to control or operate the plant. True/ False 8.15 Instructions and procedures shall be cancelled in a prescribed manner whenever cancellation is necessary.
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a. List the three (3) steps to cancel a procedure or instructio (1.5)
b. Who can approve the cancellation? (List two (2)). (.5)
8.16 What identifies the procedures required to be present during (.50)
the conduct of the activity, operations or evaluation?
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8.17 During a reactor startup, with power starting to indicate on the power range instrument, a intent >ediate range channel fails low.
4 Which of the following is the correct action to be taken (1.0)
in accordance with technical specifications? With the number of operable channels one less than the total number of channels, startup and power operation may proceed provided the inoperable channel is placed in the tripped condition in one hour.
- With the number of operable channels one less than required by the minimum channel operable requirements and with thermal power level less than or equal to 5% of rated, restore the inoperable channel to operate status prior to increasing thermal power above 5% of. rated thermal powe . With the number of operable channels less than
, required by the minimum channels operable requirement, be in at least hot standby within 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> . With the number of operable channels one less than the required minimum channels operable requirement, plant operation may continue until the next required channel functional test provided the inoperable
'
channel is placed in the tripped condition within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> ,
- While trying to satisfy the action statement in part a, (1.0)
the reactor pcwer increased and mode 1.was entere Which of the following is the correct action to be taken? Reenter mode 2 within I hour and complete the required action previously initiate . Complete the required action and then continue the power accessio . Within one hour comence a reactor shutdown and be in hot standby within 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> . Continue the power accession and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> notify the NRC of the technical specification violation.
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8.18 The Overpressure Protection System which includes two (2) PORV i with lift settings of 435 psig was installed to comply with
'
10 CFR 50, Appendix Under what condition (s) must this system be operable? (1.0) Three (3) conditions must be met for the PORV to be considered (1.0)
operable. State two (2) condition .19 For each location below indicate the reactor coolant leakage criteria per the Technical Specification that would appl Unknown location (0.5) Through pressurizer code safety valves to the PRT (0.5)
4 Valve QRV 301 body is leaking (0.5)
" Valve 51-170L2 (low head safety injection isolation valve- (0.5)
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coldleg)
8.20 Refer to figure 2.1-1 from Technical Specifications.
, What is the basis for the four (4) curves marked A? (.75)
j What is the basis for the four (4) curves marked B? (1.0)
EhD OF CATER 0GY 8
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} MASTER CORY Section 5 Answers - Theory of Nuclear Power Plant Operations, Fluids and Thermodynamics 5.1 (d)
Reference: Rx Theory 001-000-K5.09 (3.7); K5.17 (4.2)
5.2 (b)
Reference: Rx Theory 001-0100-K.5.20 (3.6); A.4.04 (4.1); K5.35 (3.6)
5.3 (c)
Reference: Rx Theory 001-000-K5.17(4.2)
5.4 (c)
Reference: Rx Theery 001-000-K5.51(~'.7)
5.5 (b)
Reference: Rx Theory 001-000-K5.51(3.7)
5.6 (c)
Reference: Rx Theory 001-000-K5.45(3.6)
5.7 (d)
Reference: Rx Theory 001-000-K5.34(3.5)
5.8 (c)
Reference: Rx Theory 000-076-K3.05 (3.6) K3.01 (3.1)
5.9 (a)
Reference: TMI - Report to Comissioners & Public Volume II, Part 2, i Page 527 000-074-K1.02 (4.8) A2.06 (4.6)
5.10 Reflux boiling is the phenomenon of two (2) phase flow in the RCS where steam entering the s/g U-tubes is condensed and then back flows into the reactor via the hot leg where it picks up additional heat and the process repeats itsel Reference: 000-017-Kl.01(4.6)
000-009-Kl.01 (4.7)
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. g @c9eRnrDA6 b 5.11 g6tfrTI64Jttf'fdffhifrA was rapidly reduced by the cold water slug which went into the reactor cold leg temperature drops due
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to high level -in S/G caused by the feed regulatory valve failur Reference: PGS-10 059-000-K4.02 (35)
5.12 1.02 Reference: Technical Specifications 000-001-Kl.16 (3.4)
000-001-Kl.18 (3.8)
000-001-K1.11 (3.3)
5.13 Although the saturation tenperature of the steam exiting the steam generators decreases slightly as power increases, the temperatures of the feedwater increases with power because of the improved
' effectiveness of the low and high pressure feedwater heaters. Since-feedwater temperature increases, the enthalpy change across the steam generator decrcases with powe Q s/g = M (h2-hy )
Reference: RX theory 059-000-K3.03 (3.7)
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5.14 T-average initially decreases due to increased heat removal through S/G's to heat FW. Rod control system will eventually react to the delta - T signal (if significant enough) to raise T-average back to its initial level, Reactor power will initially increase due to MTC. The rod control system may initially insert rods due to an unnecessary mismatch between nuclear and turbine power. The power level will reach equilibrium at a value slightly higher than the initial power level to overcome the added effect of feedwater heatin The decrease in T-average will cause the RCS fluid to contract, lowering the pressurizer level. The CVCS will react to lowering changes which will cause VCT level to decrease. When T-average subsequently increases, charging will reduce and the VCT level will return to its initial leve Reference: 004-000-K1.02 (3.8) .
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5.15 Dilution will cause rods to insert pushing flux towards bottom core. Top half of core: Flux decreases, Xenon increases as burnout decreases. As iodine decreases, Xenon will start to decay to lower flux equilibrium level and neutron flux will start to increase. As flux increases, Xenon will again build and push flux towards bottom or core; causing cycle to repeat. Bottom half of core: Increased flux.will cause Xenon to burn out further increasing flux. As increased flux burns out Xenon, flux will migrate towards top of
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core reducing burnout and causing Xenon to increase. Xenon continues to increase until iodine from previous high power level is a decayed. Xenon will decrease causing flux to increase and repeat cycle, Observe history of Delta-I swing and initiates insertion before reaching positive Delta-I peak. Maintain inserted position until Delta-I drops below target band and then withdraw rods.
, Nothing as the cscillation would probably be convergent
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(self-damping) and the oscillation would gradually sto Reference: RX theory 001-000-K-5.08(4.4)
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001-000-K-5.53 (3.7)
001-010-K5.35 (3.6)
'5.16 xenon worth , XfMu ( f t. J.-e e 7 T ' f/ W baron worth , 6 r.' u ' (.'Cr. uw /;Cf/70' ' - stuck rod worth , /g'i 7 N '[' inserted rod worth / jT i/. h'Us' s '/2 core excess reactivity
' K5.06 (3.8)
Reference: OPH 4021.001.012 M
,3 < . . , . . . r . . /. , - ,w>y cA= m e- _ __
n ,A :'
- 5,17 a- cgp r ni ce4 i: ,. , <&,r s.j '~ n~ , es ? - ti' G i, <: ..i . . w w ' ,'.- s ' . - ~. . - ( ,', - - f d'-
mo 7 . w. 5 .
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p,. ni n,, CAF /A gy'j-(c Kfnii l'- l (nif,j'.,l f,fbD $cl
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- s iu ,m, e r- rc - ve,;{ r 2 s M H w Reference: T/S 001-000-K5.54 (3.6); C01-000-K5.55 (3.4)
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a 5.18 . hot channel factors shutdown margin acceptable power distribution limits effects of rod ejection accident
, When it is necessary to rapidly reduce power to avoid or minimize a situation harmful to plant personnel or equipmen steps - Bank C . - v/.,, r/
'f1 1Rf r- {% 0 Ct , .f 9-Reference: Technical Specifications 3.1.3.5 Page 3/41-23 001-010-K5.01(3.4)
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5.19 Core geometry becomes altered and boron can be removed from the core as temperature falls or as coolant boils off, t
- - Reference
- Mitigation of core damage Page 12 j 002-000-K5.1,4 (4.2)
. 5.20 Plutonium 239, which builds up after a trip due to the Jecay of Neptunium 239.
).
Reference: Rx Theory 001-000-K5.32 (3.0)
5.21 Erroneously high. The energy content per mass of steam will
actually be lower than charts assuned in the heat balance calculation which assumes that the steam leaving the steam generator is nearly 100% quality (no moisture content).
i Reference: Steam Tables 035-010-K5.01 (3.9)
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END CATEGORY 5 - ANSWERS ;
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Section f hnswers - Plant Systems Design, Centrol, and Instrumentation No. Low pressurizer DC Cook pressure 7 trip is 1815/1900 psi If the turbine first stage pressure instrument is placed in test for maintenance purposes, it will de-energize interlock circuit P-7 this will unlock the lcw pressurizer pressure trip functions and .
will result in a reactor protection actuatio '
Reference: NS-11-46 013-000-KI.16(3.4)
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012-000-K4.01 (4.0) If more than one (1) PR NI channel is de-energized, P-10 will block source range high voltage causing a loss of source range. Source range is required in shut down mode Reference: T/S 012-000-K6.04 (3.6)
015-020-K3.6 (3.9) The switch in the neutral vice auto position would defeat two (2) of the four (4) required engineered safety features auto start signa These are the S/G water level Lo-lo and loss of main feedwater pump Reference: AS-11 061-000-K3.01(4.6)
061-000-K4.02 (4.6)
! (b)
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Reference: NS-6-47 000-076-K3.01(3.1)
004-020-K3.5 (3.9)
' The reactor trip occurred due to indication-of low RCS flow with reactor power greater than the P-8 setpoint.
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' The safety injection occurred due to an indication of low steamline pressureconcurrentwithhighsteamflowcausedbytheoperationol A Reference: LER 64-008 and drwg 1-98361 013-000-Kl.01(4.4);AZ.04(4.2);
063-000-K302(3.7)
> (b)
Reference: 10-029 - Annunciator Response 001 050 K4.01 (3.8);
, A2.01(3.9) The No. 2 seal standpipe maintains the back pressure on No. 1 seal j by raising the back pressure more flow will be directed out of the No. I seal leak off.07 OG BO 41M r4c. W SM A g/7*us 4.AN' und- c/'<fe 4 Oww<. t o#A4Mi ytS suovy?
Reference: NS-6 003-000-Kl.03(3.6) #/ SdM 6%fC'f-
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- The reason is that at such a low power level turbine impulse f
pressure Levec is 9,too p>low 1 to provide-reliable
+t=cets auto JEYMc F/otv /NUM/MM ? ON<eiMs5/&evel
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control . SN 47 4w Reference: PG5-10-48, OHP 4021 001 003 035-010-K4.01 CO-Kl.04 .9 g)Adirection that only all wires marked with yellow labels and red lettering, or yellow labels and black lettering are be lifted or jumpere ^
p) u.y orvt , wn '
e-Reference: OHP 4023.032 003 064-050-K21 (3.6)
6.10 Droop is desirable for parallel operation because it allows load sharing between power sources. With a flat speed droop, any frequency changes result in very large swings in load. For safeguards actuation, a flat droop is required so that loads sequencing cn are operated at the proper speed and voltage; if not flow might not meet FSAR requirement Reference: AS-11 064 000 K4.01 (4.1) K4.10 (4.0)
6.11 The diesel operating in test mode with a SI signal will return the diesel to standby operatio The emergency loads with automatically energize by offsite powe Reference: Technical Specifications 3/48-4 064-050-K36(4.1)
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6.12 The master pilot valve (at tank) which is energized shut will fail open. The local valve is energized cpen will fail shut. It must be opened to get a discharge.
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Reference: AS-18-20 086 000 K35 (3.7) K4.06 (3.3)
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i 6.13 No. A failure of a turbine first stage pressure channel should not I
result in an inadvertent actuation of the dumps unless simultaneous l failure of both channels occur since one channel feeds the loss of Icad arming circuit and the other channel feeds the load rejection error signal circuitr Reference: RGS 12 041-020-K35(3.6)
6.14 Steam dump valve air supply sys l
i air solenoid DC supply circ water system air ejectors (aux stm) (condenser) 041-020-K35 (3.6)
Reference: PGS 12-52
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6.15 Automatic rod withdrawal is blocked by a possible urgent failure. /# ,4*e0 vgfjg/ l
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Ls.ne fuar sinye pn ve, ec.s, Mo r, er+r,the ~ w> e sw> swi cc.ac, .,ir . t Reference: NS-4-74 001-000-K6.02 (3.3)
6.16 No!@Containn'ent sprays require logic 2/4 for containment high pressure signal to actuate, also required in modes 1, 2, and 3 per
- Technical Specifications. But bypass push buttons will isolate via the spray and gate logicGoodwrench logics.$M which will prevent shouldspray actuation be directed not when to taketesting two (2)
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channel to test pgt echnicalT Specification which require three (3)
channels operableCF Steamline isolating (MSIVs close) would occur
_2.9 psig thus isolating stm to sta dumps. #NY Mm w e/t., Su/F/cnwd Reference: Ns-11-VI-1, P 11-00, DWL OP-1-98512-3 026-020-K36 (4.1)
6.17 True
j Reference: NS-11-20 001-000-K6.27 (4.1); 012-000-K6.03 (3.5)
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6.18 False Reference: PGS 15-57 063-000-A2.01(3.2)
, 6.19 Paralleling normal to reserve power during S/D Reserve to normal during startup j Return to normal lineup-t 'W > '
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. ., ..: i . :e , 9 following
. i. w .; a , blackout. .. .. , , a J f.;. '.' ^. :"' ' / 4 1 '
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Reference: PGS 14-24 062-000-K4.05(3.2)
6.20 normalce: re,r eitherta res.r r S'/ WQc w r i eithern P V f St/6&Awf" normalte Mf,r Reference: AS-10-60 064-000-K4.02(4.2)
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6.21 Control room area monitor R-1
' Containnent area monitor R-2 Spent fuel area monitor R-5
, Containment air particulate monitor R-11 [e7f-/*O
, Containment radiogas monitor R-12 C G4 ' -/-v) Component cooling water monitor R-17A/B Waste disposal system liquid monitor R-18 S/g blowdown liquid nonitor R-19 S/g blowdownTh m sYonitor R-24 10. Unit vent radiogas monitor R-26 (Ep 7Q ~ ev -; y ri tew:c ccivm/x'x?e Ar <frwis**('Myrspj Reference: AS-21-37 072-000-K1.01 (3.5); 072-000-K2.02 (3.9) See Reference AS-21-37
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f-l -shoFU> cnutt END OF SECTION 6 ANSWERS Aw' W tca re icec ec ,*rcre cpm cw1w kh"
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Section 7 Answers - Procedures - Normal, Abnormal Emergency and Radiological Control (d)
Reference: Technical Specifications 3/41-16 006-050-K36 (4.3) . Power containment inside crane wall 10% pwr Regenerative Heat Exchanger Area _10T pwr Ice condenser, lower plenum __20% pwr Incore Instrument Room Reactor Cavity Pit CVCS/ Boric Acid Evaporator Feed Ion Exchanger Room Drum Storage Facility Reactor Coolant Filter / Seal Water Filter Room Seal Water Supply Filter Room 10. Spent Resin Storage Tank Room 11. Spent Fuel Pit Demineralizer Room 12. Deborating demineralizer 13. VCT Rooms 14 Fuel Transfer Tube Rocms Reference: PMP 6010 RAD.001 068-000-K4.01(4.1)'
.- The shift supervisor shall obtain permission from the D. C. Cook Senior Reactor Operator in chare of core alterations (SR0CA) in containment prior to going to bypass on the cran Reference: NS-16-52 034-000-K35(3.5) . Declare the turbine driven auxiliary feedwater pur:p inoperable and follow action statement. . ,
4" # 9tuf Y to 07kt 7% % M 7P ," cl' Yyfs' Since the charger will be out for four (4) da s, be in hot shut down .
within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. 04' gra> rc4'e 70 p Srp$ cm Mw 7d@
061-000-K36 (4.1) $'?$ekf'
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Reference: T/S
, Sufficient CCW flow to the RHR heat exchanger must be maintained to
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prevent the flashing of CCW to steam as a result of heat input from RHR flow through the hi.at exchanger, also create excessive water hame Reference: OHP 4021 .017.002 005-000-K4.02(3.5);K9(3.6) No. Removing these channels from service is contrary to Technical Specifications since they provide reactor protection. These signals include bi steam flow isolation and low s/g leve.1 with flow mismatch t,rg' **~, ", ,.,,, 4._, 7,,e , , w . ,,y O '( ' >/* om"ra r - m 4f**W**'l'*e #A l
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Reference: NS-11-44 035-010-K9(3.5);012-000-K4.06(3.5)
4 u w ,.x .: + y u . f t. t > H *N*""W NY 0)k d f r/3' i sf C 1;f t. .' ( .ta ll? Ms t ' "
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! A leak rate increasing by .009 gpm (13 gpd) per day will not exceed i a Technical Specification limit until the flow rate exceeds 500
'
gp The initial flow rate is 288 gpd and increases at a rate of 13 gp gpd = .35 gpm S/g limit = 1 gpm 1 gpm = 1440 gpd
.2 gpm = 288 day 1 = 288 day 2 = 288 + 13 = 301 i day 3 = 301 + 13 = 314 day 4 = 314 + 13 = 327 day 5 = 327 + 13 = 340 day 6 = 340 + 13 = 353
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day 7 = 353 + 13 = 366 day 8 = 366 + 13 = 379 day 9 = 379 + 13 = 392 day 16 = 484 + 13 = 497 approximately 500 gpd
Reference: Technical Specificatiors 3.4. K3.06 (4.1); K3.05 (4.0); K36 (4.0)
.,,m ,- + Prevent injection of accumulator nitrogen into the RCS; prevent reactor core returning to criticality due to moderator temperature effect Reference: OHP 4023.001.016 000-055-K3.02 (4.6)
l 7.09 High Radiation Area - Whole body dose 100 mrem /hr t
! (6 REM)(1 FT)2 D
100 M Rem ,
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(6 REM)(1 FT)
- 60 = dj
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7.7 ft = d
Reference: Radiation Protection Manual 073-000-K5.01 (3.0)
7.10 The facility shall be placed in at least Hot Standby within one'(1)
hour, The safety limit violation shall be reported to the Commission and
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to the Chairman of the NSDRC within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> *
. s) tty,wom.w y a e c Mn%
, Safety limit violation report shall be prepared.'t9 chD, 6 GN - h e' * I f
/ L) ncts. n.s ska:- 1 rL -rre m Safety limit violation report shall be submitted within 14 day Reference: T/S!7 012-000-K4.02 (4.3); K27 (4.7)
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- - , e - .11 Provides adequate mixing, prevents stratificatian and ensures that reactivity changes will be gradual during boron concentration. The reactivity change rate associated with boron reductions will be within the capability for operator recognition and contro Reference: Technical Specifications 3/ K5.20(3.7)
lineup to the mixed 7.121.h0nce flowalert level alarm derrineralizer andhad been reached increase flow to 120- vegpm@tiotify Chc Dept. Supervisor and/or Chem. Dept. foreman, Rad Protection Dep and Nuclear Section. (7,2 3 j 2.hiHigh Alarm - Notify Operations Supervisor, Technical Supervisor arid R6d Protection Superviso ( M5)
3 jConsulttheEmergencyPlanEC-1 /, * ' # C)
Reference: OHP 4023.002.001 000 076 K3.06 (3.8)
7.13 Epric= Actuatier of SI
. 4/ loss of reactor coolant
~ [/r loss of secondary coolant & S
[.CSteamgeneratortuberupture Reference: OHP 4023-001.004 000-040-K3.04 (4.7)
7.14 Reduce load M m ... ;:.0 q : t r _ Check proper operation of air , ejectors [C.fM /M dM Check proper operation cf circulating water
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4 Check proper operation cf. flow Check proper cperatiori of condenser water boxes vent system
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Reference: CHP 4022.053.001 000-051-K27 (3.9) - - -
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7.15 excessive control rod insertion failure of a control rod to drop following a reactor trip Uncontrolled reactor coolant cooldown following reactor trip Unexplained or uncontrolled reactivity increase Shutdown margin _1.6% Delta K/K (1) Control rods are above the insertion limit and rod insertion limit low alarm is cleared. (2) Sufficient RCS boron concentration is achieved to ensure adequate shutdown margin for anticipated temperature and Xenon condition Reference OHP 4022-005-002 000 024 K3.01 (4.4)
7.16 Record time /date of alarm and record and trend count rat / 08A/ Isolate CCW flow to all components which are not in service or which may be shut down without adversely effecting plant operations in progres . If counts increase, separate the CCW System into two (2) train Start the standby CCW pump . Open standby heat exchanger outlet valv Separate pump discharge headers Separate the East and West trains with the miscellanecus header on the last train Isolate each component on the leaking header one at a time and watch for a leveling off of counts / surge tank leve Reference: 0HP 4022-016-003 000-026-K3.03 (4.2)
7.17 If pressure in 100 psi control air header decreases to 80 psig, take the following actions: trip the reactor /N 2. - Initiate a cooldown of the RCS if control air is available .)/Af.)
I If all control air is lost, maintain s/g level with auxiliary feed water numps while heat is being removed through.the s/g safety ' pg/
valves l Maintain pressurizer level within Technical Specifications by intermittent charging pump operations
- ,J/)f]-
l Reference: OHP 4023.00.1.006 000-065 K3.08 (3.9); A2.06 (4.2)
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Feference: 0HP 4023.001.015; 000 074 K3.11 (4.4); A1.11 (3.7) A1.12 (4.4)
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7.19 a.-- Safety injections occurring below the low pressure SI setpoint must be assessed by licensed operators as the implied initial conditions for emergency procedures are no longer valid and the use of the procedures could cause an unsafe condition. It is the licensed I
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operator's. duty and responsibility to maintain thereactor in a safe conditio The Unit Supervisor should approve any with holding or resetting of ECS equipment as required to n.aintain the reactor in a safe conditio Reference: 0S0.061 .Page 2-2'No. 27 (4.5)
} 7.20' - This is to allow automatic restarting of the safeguards equipment in
- case a blackout occurs during this period.
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Reference: OHP 4022.008.002 013-000-K1.01(4.4);K4.064.3)
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4 ERD.0F CATEGORY 7 ANSWERS a
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Section 8 Answers - Adniinistrative Procedures, Conditions, and Limitations
, True Reference: OHP 4021.001.001 P 2-1 #11 (4.0) If the delay extends more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the status of system
- lineups andi. surveillance tests shall be reverified and signed off prior to proceeding with the heatu Reference: Precautions OHP 4021.001001 P 2-1 #10 (4.0) (1) site area
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(2) general Reference: PMP 2080 EPP015 P 2-1 #4 (3.3) In the SS judgement, the Plant Manager will be unable to complete the notifications in a timely manner or the SS is delegated this responsibility by the Plant Manage P 2-3 #34 (4.7) False - takes precedence Reference: PHP 2080 EPP.001 P 2-1 #13 (3.7) (b)
Reference: 10 CFR 20.101 Page 2-1 No. 12 (3.9) The basis is that if the d'oor or any roof vents were open during a
, fuel handling accident, the fuel handling area ventilation would not be capable of maintaining the required negative pressure. This would cause an uncontrolled and unmonitored release to occur.
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Reference: April 4, 1985 Aux Bld Integrity Memo P 2-1 #12 (3.9)
8.8- The grace period is not likely to expire before the surveillance can be completed in the mode being entere Permissicn to use the grace period is obtained froni the Operation production Supervisor or Operation Superintenden Use of grace period is logged as an open item in the Applicable Control Room log until the surveillance is performe Reference: MPI-4030 Page 2-3 No. 36 (3.9)
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. (1) Management Staff (2)SR0 (3) Q/A Supervisor (4) PN SRC (5)PlantManager Reference: PMI-2010 Page 2-1 No. 1 (4.1)
8.10 True Reference: PMI-2010 Attachment 2 P 2-1 59 (3.9)
8.11 (a)
(d)
Reference: PMI-2110 Page 2-1 No. 11 (4.0)
8.12 When an emergency diesel generator is to be removed from service the ,,_
other diesel generator for that unit-.is_to be , ~'
will then be left running [uritil ~the clearAce% proven operable n the diesel _and
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generator being removed from service has been hung and verifie Reference: PM1-2110 Attachment 1; Page 2-1 No. 11 (4.0)
8.13 No! The unit supervisor cannot be relied on for relief of the R0 because having the SR0 perform the functions of a reactor operator even for a limited time, would result in loss of the oversight function of the supervisor which might decrease the probability of correctly detecting abncrmal events early enough to mitigate potential adverse consequence Reference OHI-4011 P5 P2-2 No. 19 (4.7)
8.14 False Reference: PMI-2260 P2-2 No. 24 (4.0)
8.15 )__Cncenier sheet ..ght upper hand cu.ic, inse,t "can:cl." -
f (k Under nape of revision on cover sheet, insert the word Cancelled and surmary of why the instruction or procedure is being cancelle (3) ext . c. is ion oiod un cuve. sheel shuuid reficci the c 2
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-ceacelled 4 place of the revitica number, i Originating Department Head Quality Assurance Supervisor Plant Manager Reference: PMI-2010 generic P2-1 No. 2 (4.3)
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8.16 Procedures identified by a double asterisk imediately before the procedure number are required to be present during the conduct of the activit Reference PMI-2010 P1 P2-KSA-K9 (3.9)
8.17 l Reference T/S 3/43-7 015-020-K36 (3.9)
8.18 Whenever RCS temperature is less than the minimum pressurization
M- M S ' ' l l ' > ' ' ' 'W "
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temp ratu r : p b n %f -' .~> .Wrk s~ rat ?
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es. c y si, n & n 2...rn t u
- . backup emergency air supply is charged PORV low pressure setpoint is enabled The PORV isolation valve is opene a/ z; a - nim cq_,;-/cylg,g,..,,9,,,a,,,.,,.,,,, ,,,,,,,,,,,,3,;,,,,,.
Refererice:' 'OY4d21., BOT.004 P6.21 010-000-K4.03 (4.1) <unc c e Uc4,;,( - ,),
8.19 gpm s gpm 3. .A'gpm /6 ? %>
l I gpm i
Reference: Technical Specifications 3.4. P4-16 002-020-K4.01 (3.8); K36 (4.1)
S.20 Curves marked A - the coolant average enthalpy at the core exit is equal to the saturated water enthalpy. *: na; e* /- 4 :<J / > 9. f in ^'-
,-tyysne,7c p yyi. t r v //'"ir Curves marked B - quality is equal to 15% at core exit.
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Reference: T/S, Rx theory, W_ 002-000-K5.09(4.2)
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END OF CATEGORY b ANSWERS
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