ML20237C657
| ML20237C657 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 11/18/1987 |
| From: | Coe D, Elin J, Johnston G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20237C651 | List: |
| References | |
| 50-312-OL-87-03, 50-312-OL-87-3, NUDOCS 8712220056 | |
| Download: ML20237C657 (140) | |
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l' N(O. NUCLEAR REGULATORY CCv%SION REGION V 1,
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} (y OPERATOR LICENSING EXAM bATION REPORT
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tenrentc6oi>erato'r(RO) candidates'.WrPteir.9xamitat1'oes and operat-ing ty's pass.,N 5UMM41\\
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Nine R0 ed both the written and
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J operatin's tests.' One RD failyd.+.haDue to the extended shtdo'goperatn4 test and was issu l
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n and.h. sccordance v'th 10 CFRJ5.31 I
(a)(5), those candidates who ps,ed t,he exertrinst % Will he issued itenses 9
W following certific,ation Lu tbc.bcK1f ty licensee of successful com etion NURES 1921 ES-109.B},iW,M.11pi ation and power 'ppep? ions specifie of % e required contepl 1 -
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i REPORT DETAILS 4
q "tTYPE OF EXAMINAll'ONS:
Replacement F
1.
Persons Examined:
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There were 10 candidates for Operator License examinations.
I'1 2.g Examiners at Sitel U
V
- G. Johnston, Region V NRC D. Coe, Region V, NRC r
- Chief Examiner m
- LL
- [
3.
Personnel Presen,t_a1 Exit Interview:
l NRC Personnel
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D. Coe, RV, NRC Facility Personnel l
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P. Turner, Trair,dg' Superintdodent r
O,.
M. Hieronimus, Assistant Operations Superintendent M. Herrell, Assistant TrainWp Superintendent A. Bonino, Licensing
.h G. Wallace, Training InstrucSor 3
K. Molloy, Training Jnstructor y
4.
Written Examination and Fecility Review:
The facility staff was given copies of the written examination and examination key and wasrinstructed to review the examination and provide their epments prior to the examiners departure or to the Region V' office no later then 5 working days after the completion of all operating examinations. After grading of the written examination NRC examiners noted the following:
The candidates did particul ely well on the Reactor Theory portion of the written examination.
.The candidates had difficulty determining how EFIC automatic and manual control is affected following manipulation of MFI, MSI, and AFW trip modules.
They were not aware of the facility radiation exposure limits for untrained visitors.
5.
Operating Examinations:
Duringt.heconductof[theoperatingexamination,theexaminersidentified thefollowing concern.
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The candidates did not understand the basis for the Low Temperature L
Overpressure EMOV setpoint.
It is not related to protecting the Decay Heat System piping or components, as many stated.
Some candidates also could not determine the required pressure control band for various combinations of Reactor Coolant Pumps and Decay Heat Removal pumps.
Some candidates could not correctly describe the ICS response to a failed controlling steam header pressure transmitter.
Candidates were able to locate most required reference material.
However, some candidates could not describe where to find the guidance which was promulgated by plant Standing Orders.
6.
Summary of NRC comments made at exit interview:
No results of the examinations were discussed, but the generic deficiencies noted in paragraphs five above were presented.
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I HANCIl0 SECO REACTOR OPERATOR LICENSE EXAM REVIIM I
5 SECTION I 1.01 The drain lines to the condenser from many sources are designed to withstand water hammer because the potential for water hammer io very high when draining to a vacuum.
Drains from steam traps, steam generators, moisture separators, etc. are frequently less than 212 F.
When the hot water from these drains enter the condenser, it will flash due to the low pressure (vacuum) and will very I
likely result in water hammer.
Therefore, answer "d" should also be acceptable for full credit.
Placing vacuum breakers on the high points of water system can also result in water hammer.
When a pump is shut down, a vacuum breaker will allow the high point to drain, thus leaving on empty section of pipe, assuming.
the vacuum breaker is on the discharge. When the pump is restarted, the resulting surge as the line refills will result in water hammer.
An open vent valve on a high point would have the same effect.
Procedure A.47
" Condensate and Feedwater," contains.a limit and precaution that requires that the condensate pump l
discharge valve be closed prior to starting the. first pump to prevent " hydraulic surges." An open vent line or I
vacuum breaker would aggravate this situation if the l
discharge valve was left open.
In 1978 at Oconee Nuclear Station, this exact event took place resulting in ruptured hydrogen coolers and subsequent release of radioactive water to the environment.
Therefore, answer "c" should also be acceptable for full credit.
1.17a Since the correct answer depends on the assumptions made, as stated in the question, each student's answer should be graded based on his assumptions.
If the assumptions are valid, then the resulting answer (more, less, or same) should be acceptable.
1.23 The phrasing of the question suggests that the answer should be a discussion of how adequate subcooling margin with accurate pressurizer level is possible given that a bubble exists in the head.
It does not necessarily indicate that a discussion of how a bubble is formed is required for an answer.
Therefore, an explanation of the possibility of adequate SCM and PZH level given a head l'
bubble should also be acceptable.
l
_______-_____________a
l PO LICENSE EXAM REVIEW PAGE 2-SECTION II.
2.01 At' Rancho Seco, the battery and the battery charger float on the lines supplying the inverter, with - the battery charger supplying a slightly higher voltage.
When the battery charger is lost, the battery (already connected to the lines) will supply the loads, thus no actual-L transfer takes place.
Therefore, both answers 1 and 3 -
E -
should be acceptable.
See Item.2.01.e.for further j
explanation.
2.01.e The terms "soterce" and " supply" appear to be ambiguous in this question.
The normal power supply for NNI-X, NNI-Y, and ICS is 120. VAC bus SIGB-1.
The sources of 120 VAC for this bus are as follows:
a.
480 VAC MCC S2B2 via regulating transformer and
-switch, l
b.
'4 8 0 ' V A C MCC S3B2 via battery charger and inverter to the static switch, and c.
Battery GB via inverter and static switch.
In order to cause a shift of normal supply (SIGB-1) to the alternate supply (SI-J), all of the above sources must be lost at the same time or a loss of the static switch must occur.
These two conditions were not stated in the question.
Also accept answer 1 since conditions were unclear.
(1 - No power supply shift occurs.
ICS continues powered operation from the same source.)
Reference:
ECN R-0929 2.02.a & b This question appears to be a conceptual type dealing with where the thermal energy is dicipated following a LOCA.
Since this is a conceptual answer, specific components (RB spray pump,'DRR pump, Cooler its) should not be required in a discussion of the concept of thermal energy transfer to the ultimate heat sink.
2.04.a The learning objectives referenced (No. 9, OD 21 I 3205 and No. 13, OD 21 I 3207) are as follows:
9.
State the systems which are physically connected to the system, and describe their interrela-tionships with it.
13.
State and explain the major steps of the i
procedures associated with the system with respect to the following:
- Normal operations
- Infrequent operations
R0 LICENSE EXAM REVIEW-PAGE 3 These objectives do not require memorization of the locked valves (open or closed) associated with a system.
-Neither Licensed Operators nor Licensed Operator candidates are required by the Training Department or Operations Department to memorize locked valves.
The Operators are required to know normal and abnormal-flowpaths (i.e. open or closed).
In view.of the length
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of the ' locked value list, SP.214.0J (approximately 500 valves), and that the answer key requires two responses, any two normally closed valves on the figure should be acceptable for full credit.
2.04.b The line identified by point "A"
is also used for auxiliary spray to the pressurizer during startup and shutdown.
Therefore, " auxiliary" spray should also. be acceptable for full credit.
References-(Attached):
- Procedure A.15, Rev. 32, Enc. 8.1, P. 56 Procedure A.3, Rev. 21, Step 6.6, P. 10
- Procedure B.4, Rev. 40, Step 5.27, P. 12 2.05.c D/G's GEA and GED (Fire Protection Zones 40 and 41) nre also protected by the Fire Protection Water System (Sprinkler).
This answer should also be acceptable for full credit.
Reference:
P&ID M-594, SH.1 - H-7, H-8 2.07 Recent modifications bave been added to the Instrument Air (IA) System.
These include diesel air compressor, extensive bottle air backup, and others.
These modifications have improved the reliability of the air system such that the 30 minute shutdown requirement is no longer necessary.
Shutdowm is now required based on air system pressure and upon imminent loss of control air operated equipment and resulting difficulty in operating the plant.- Therefore, any reasonable answer discussing problems of operation with a loss of air should be acceptable for full credit.
Reference:
Restart Lesson Plan - OD 24 D 1400 2.10 The question does not indicate that a detailed description is required for credit.
Also, the point valve (1.0) suggests minimal details.
The Lesson Plan OD 24 D 1700 does not discuss the effect of loss of air in great detail.
Only a very simplified diagram (TDI shutdowm system) is provided in the student handout.
RO LICENSE EXAM REVIEW PAGE 4 Therefore, a general discussion that the Shutdown Protection System requires air to function (shutdown engine) should also be acceptable for full credit.
Reference:
Restart Lesson Plan - OD 24 D 1700 2.16 The question asked which of the following is not a design difference between control rods and APSR's.
APSR's do have buffer springs in the buffer assembly.
This makes answer "d"
false which qualifies it as a correct answer along with "a",
which is also false.
Since neither "a" nor "d" statements are a design difference, credit should be given for either answer.
SECTION III 3.01.a on loss of ICS power the auxiliary steam control valve setpoint fails "as is."
The valve will continue to control (modulate) pressure at the setpoint that was established prior to loss of ICS power.
Therefore, the correct answer should be modulating or controlling at the setpoint.
This should be accepted for full credit.
Reference:
ECN R-0878, Rev. 1 3.01.b The operation of the valves, as discussed in 3.01.a above, is such that control is available on loss of ICS power (Valve modulates to control pressure).
Since the valve does control during a loss of ICS power and only the setpoint is effected, an acceptable answer should also be " Control is fully available upon regaining ICS power."
Reference:
ECN R-0878, Rev. 1 3.02.a On loss of ICS power from 1004, the reactor trips; when the reactor trips, the CRD breakers open.
Therefore, the l
control rods cannot be moved in manual. "No control available due to the reactor trip" or " react or t rip" should also be considered for full credit.
Reference:
OD 24 D 2500, Pp. 3 and 4 (Attached) l 3.02.b The MFW pumps trip on loss of ICS power.
As discussed in 3.02.a (above), "MFW pumps trip" should also be l
acceptable for full credit.
1 Reference (Attached):
OD 24 D 2500, P. 3 l
l
RO' LICENSE' EXAM REVIEW PAGE 5 3.03.a The TBV's also shut on loss of ICS power.
Therefore.
-" loss of ICS power" should also be acceptable for full credit.
Reference (Attached):
- OD 24 D 1600 Student Handout, (ECN R-0861, Rev.:4)
'3.05 If the pressure channel fails high, the pressure compensation for level control will be inaccurate.
This willicause the valves to control at an incorrect level.
Therefore, " level changes due to a _ change in pressure f
compensation" should also be acceptable for. credit.
Reference (Attached): OD 24 D 3200 -- iia 3.08.c.
The question does not specifically identify the. monitor by number or function (i.e. process or area monitor).
There are areas as well as process instruments that monitor radiation levels _in the Auxiliary Building grade level.
Therefore, "no action or alarm" should be acceptable for full credit.
Reference (Attached':
)
- Procedure A.67, " Area Radiation Monitors," Rev. 7, Enc. P.1 3.08.e.
Same' comment as on 3.08.c (above). Other monitors in the Reactor Building can be used to determine if a leak exists.
Therefore, any answer for. a Reactor Building monitor function should be acceptable for full credit.
3.09.a It is not clear which recorder is first and which is second.
There were originally three trend recorders on
'HICO.
These have been replaced by the two new recorders.
Therefore, a discussion of the purpose of the first recorder should also be acceptable (i.e., selectable trend recording).
3.10.a & b The generator differential current trip must be considered by the Operator as. a possible cause of the trip.
It-is operable under the conditions given in - the question.
The generator differential trip is sensed upstream of the breaker.
It is'possible to get the trip with the.D/G unloaded.
In addition, the relay could fail and cause a trip. A differential current trip is just as likely to occur because of a fault on the generator field as downstream - of the breaker.
The question does not indicate that lower probability causes should not be considered.
Therefore, " generator differential trip" should also be accepted for full credit for both 3.10 a l'
and b.
L_-_..____..
_ - _ ~ _. _ _ _
RO LICENSE EXAM REVIEW PAGE 6 3.12.c:
The question asks for specific conditions throughout the startup that will inhibit rod withdrawal.
The following
)
. additional conditions will inhibit rod withdrawal:
'l.
Asymmetric rod fault greater than 60% power, and -
2.
Safety rods not.out greater than 60% power.
If it is assumed that the plant is started up to 100%,
then these two answers should also be acceptable for full credit.
Reference:
Hancho Seco Correction Bank - CRDCS --13a SECTION IV 4'.03.a The given condition in the question stated that the reactor was manually tripped and that the reactor did trip upon subsequent Operator action.
Therefore,
" manually trip the reactor" should not be required for full credit.
Also, all of the action of Step 2.0 could be applicable because the last given condition states that.the reactor trips.
Therefore, credit should not be lost because Steps 2.1 through 2.4 are listed.
Reference:
E0P's - E.01 4.03.b The instructions to borate are broken down into three parts in E.02, Step 1.5, each of which is a separate action. -Therefore, it should be acceptable to list the--
three boration methods or a combination of the boration methods and the other two answers listed in the key (i.e., deenergize CRDM's, drive rods in) to receive full credit.
Reference:
E.02, Steps 1.5.1, 1.5.2, 1.5.3 4.04.a Shifting to one pump per loop in the event that all pumps are not tripped within two minutes is not an immediate action.
The students are not required to memorize steps in the body of a procedure.
The critical point about the pumps is to maintain some flow through the core if the pumps are not stopped within the required two minutes.
Therefore, it should also be acceptable to state that the RCP's should remain running for full credit.
Reference:
E.03 4.05.c & e Casualty Procedures are to be used when symptoms indicate a problem.
The " response" section of the procedure is not required to be memorized.
A symptom will alert the Operator to the problem, at which point the Operator reads and performs the applicable procedure.
The procedure says to first reduce power prior to tripping
.RO LICENSE EXAM REVIEW PAGE 7 the pump.
Therefore, there is time available to read through the procedure prior to being required to trip the pump.. The wording of indicator explanations 1, 2, and 3 is confusing. (especially 2).
Based on the above, it should be acceptable to answer 1,2, or 3 for."c" and "e"
Reference:
Car.ualty Procedure C.8 t
4.07.a,b,c Casualty Procedure C.10 has been changed. such that the responses are different from previous revisions.
The requirements for tripping the reactor are as follows:
a.
(1) Main feedwater is lost to either OTSG.
(2) 15 inches on startup level or EFIC low range (only one required for credit).
- b. (1) 381 inches on EFIC high range or 95% on operating range (only one required for credit).
(2) Pressurizer level less than 160 inches.
c.
(1) 381 inches on EFIC high range.or 95% on-operating range (only one required for credit).
(2) Pressurizer IcVel less than the enclosure.
4.08.e This Lis an SRO-level question.
The learning objectives do not require RO's to know when a job planning meeting is required.'.
It - is HP' supervision's responsibility to determine when such a-meeting should take place.
Specifically, it_ is the Senior Chem-Rad. Assistant's responsibility.
Therefore, this part of the question should be deleted.
Reference (Attached):
AP 305-4, P. 12 4.09 Each required response should have a reasonable acceptable band assigned to account for inaccuracies j
encountered while reading the Reactivity Balance curves.
4.11 It is not required that. Licensed Operators or candidates memorize. steps in normal operating procedures ("A"-
procedures).
Therefore, any reasonable methods listed should be acceptable for full credit.
I 4
l l
NRC RESOLUTION OF RANCHO SECO COMMENTS l
ON NRC EXAMINATION GIVEN 10/19/87 1.01 Will accept c and d for full credit.
1.17 Part (a) is deleted due to the large number of assumptions possible.
/
1.23 Facility comment does not provide a concise recommendation for modifying the answer key and is, therefore, not accepted. However, an answer which indicates an understanding of the location of SCM and Lpzr instruments in relation to the RV head as the reason this condition may exist will be accepted fo'r full credit.
2.01 If the candidate assumes that " power supply shift" means from l
bus SIGB-1 to bus SI-J via the ABT, then parts (c) and (e) could also be answered with "1."
This will be an acceptable answer for these parts.
2.02 Since the question specified heat transfer components or systems, l
DHR (LPI) pump is deleted from the answer as it is not primarily a
" heat transfer" component.
2.04a The referenced. learning objectives and K and A numbers do not-directly address part (a). Therefore, part (a) is deleted.
2.04b Accepted, based on listed references.
2.05c Licensee proposed additional correct answer is added to answer key such that both answers are required for full credit, based on listed references.
2.07 Licensee provided reference material (Procedure C.23-Loss of Plant Air)
'1 indicate that the 30 minute requirement still exists (at least for the purposes of this examination). However, candidate response which shows an understanding of the extended use of plant air for critical control valves will be accepted.
2.10 The answer key is modified to emphasize that the correct answer shows understanding that air pressure supplies the motive force necessary for fuel shutoff, based on listed reference.
2.16 Accepted.
3.01a Accepted.
3.0lb Deleted due to the technically correct answer (restoring ICS power) necessarily negates the premise of the question and thus could be confusing.
3.02a Answer key clarified to indicate loss of control until CRD breakers are re-shut, based on listed references.
l l
3
l 3.02b
- Answer' key clarified.to indicate loss of control until MFP's 'are reset, based on listed references.-
3.03a Not accepted. The presence of ICS power in order for TBV's to operate.is not considered an " interlock" by facility literature.
3.05 Accepted based on listed reference.
3.08c Accepted based on listed reference.
3.08e Not accepted. The monitor stated in the question is clearly a process monitor and is the only monitor given the listed name in any licensee reference' material provided.
'3.09a Not accepted. The question clearly indicates the trend recorder in question is the one which was ADDED, representing a capability beyond that which existed with the original recorders.
3.10a,b. Accepted.
3.12c The use of the word "startup" is indicated in facility literature as meaning critical operation up to 15% power only (Procedure B.2).
Therefore, the facility suggested answer is neither required for full credit nor will points be deducted if-it is not present.
4.03a Not accepted. The question asks the candidate to_ identify (all) the steps that apply to the given events, thus " manually trip-the reactor" must be included.. However, since it is given that the reactor trips,Jno points.will be deducted for adding parts of step 2.0 to the answer.
4.03b Accepted, based on listed references.
l 4.04a Accepted based on listed reference.
1.
4.05c,e The licensee's learning objective referenced.on the examination L
requires operators to know (be able to state) the RCP tripping i
criteria. However, since confusion over this question was noted l
l during the exam administration, the facility comment is accepted l
l and these parts are deleted.
4.07a,b.c. Accepted based on listed reference.
I 4.08e Question deleted due to the vagueness of the learning objective and K and A reference.
4.09 The grader will allow 10%.
1 4.11 Reasonable alternate answers which can be located in facility
-supplied reference material will be considered on a case basis.
j I
U.
S.
NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
_RANgHQ_@gCQ,____________
REACTOR TYPE:
_PWR-@hW11Z______________
1 l
DATE ADMINISTERED _@ZfigfyL________________
EXAMINER:
COE D.
2 CANDIDATE:
IN@IBUQIlgNS_IQ_Q@NpipBIg1 Une separate paper for the answers.
Write answers on one side only.
Staple question sheet on top of the answer sheets.
Points for each question are indicated in parentheses after the question.
The passing grade requires at least 70% in each category and a final grade of at 1 cast 80%.
Examination papers will be picked up six (6) hours after the examination starts.
- /. OF CATEGORY
% OF CANDIDATE'S CATEGORY
__2069E_ _19166
___SG9BE___
_Y669E__ ______________G01EG9BX_____________
.2E 99__
2E 99 1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
_3E199__ _2E 99. ___________
________ 2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS j
_2E 99__ _25z99
________ 3.
INSTRUMENTS AND CONTROLS
}
_EE 99__
2E 99
________ 4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 199:99__
Totals Final Grade 1
All work done on this examination is my own.
I have neither given nor received aid.
Candidate's Signature i
f
)
NRC RULES AND GUIDELINES FOR LICENCE EXAMINATIONS During the administration of'this examinati'on the following rules apply:
1.
Cheeting on the examination means an automatic denial of your application and could result in more severe penalties.
2.
Restroom trips are to be limited and only one candidate at a time may leave.
You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
3.
Use black ink or dark pencil gnly to facilitate legible reproductions.
a.
Print your name in the blank provided on the cover sheet of the examination.
5.
Fill in the date on the cover sheet of the examination (if necessary).
6.
Use only the paper provided for answers.
7.
Print your name in the upper right-hand corner of the first pe.ge of each section of the answer sheet.
8.
Consecutively number each answer sheet, write "End of Categor y __" as appropriate, start each category on a new page, write gnly gg one side of the paper, and write "Last Page" on the'last answer sheet.
9.
Number each answer as to category and number, for example, 1.4, 6.3.
- 10. Skip at least three lines between each answer.
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
- 12. Use abbre lations only if they are commonly used in facility litetaturg,
- 13. The point value f or each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given.
Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
- 16. if parts of the examination are not clear as to intent, asi. questions of the g,amingr only.
1
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination.
This must be done after the examination has been completed.
- 18. ' When. you compl ete1 your. exami nation, you shall:
.a.
. Assemble your examination as followsx.
[
(1)
Exam. questions on top.
i (2)
Exam aids - figures, tables, etc.
(3)
Answer pages including figures which are part of the ancwer, b.
Turn in your copy of the examination and all pages used to answer the examination-questions, c.
Turn in 'all scrap. paper and the balance of the paper that ycu did not use for answering the questions.
d.
Leave the examination area, as defined by the examiner.
If after leaving, you are found in this-area while the examination is still
'in progress, your license'may be denied or revoked.
PAGE 2
1.__ESJNCIE6ES_gE_Nyg6 EBB _EgyEB_E68NI_g[EBBIlgN 2 ISEBdgpyN8dIC@2_SE8I_IB8NSEEB_8Ng_E691g_E6gy QUESTION 1.01
(.75)
MULTIPLE CHOICE (Choose the best answer.)
Which of the following actions is most likely to cause water hammer ?
a.
Starting a centrifugal pump with the discharge valve shut.
I b.
Feeding a steam generator with water colder than saturation temperature.
c.
Placing vacuum breakers on the high points of water systems, d.
Draining a steam generator when it is below 212 F.
QUESTION 1.02
(.75)
MULTIPLE CHOICE (Choose the best answer.)
Which ONE of the f ollowing statements is CORRECT concerning a
. secondary calorimetric?
a.
If feedwater temper atur e is read erroneously high, then the calculated reactor power will be higher than actual caused from the considerable additional heat energy provi ded.
i b.
The calorimetric equation does not take into consideration the heat added by the reactor coolant pumps, or the heat lost to the containment atmosphere (ambient losses).
c.
Mass flow rate of the secondary system is determined by totaling the average steam flows from each of the steam generators.
d.
The results of the secondary calorimetric may be used as the basis for calibration of the power range nuclear instrumentation.
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
L_____________
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I 1.__EBJNyJELgg_gE_NUGLEBB_EgMEB_EL8NI_gEEBBIJgN PAGE 3
i a
l IdESdgpyN801Cg3_bE81_I68NSEES_8Np_E6Ulp_ELgb j
i 1
L QUESTION 1.03
(.75) l MULTIPLE CHOICE (Choose the best answer.)
l When performing a reactor S/U to full power that commenced five hours after a trip f rom f ull power equilibrium conditions, a 0.5%/ min ramp was used.
How would - the resulting xenon tr ansient vary if instead a 2%/ min ramp was used?
a.
The. Xenon dip for the 2%/ min ramp would occur sooner and the magnitude of the dip would be smaller, b.
The Xenon dip for the 2%/ min ramp would occur later and the magnitude of the dip would be smaller.
c.
The Xenon dip for the 2%/ min ramp would occur sooner and the magnitude of the dip would be larger.
d.
The Xenon dip for the 2%/ min ramp would occur leter and the magnitude of the dip would be larger.
QUESTION 1.04
(.75)
MULTIPLE CHOICE (Choose the best answer.)
Which of the f ollowing' statements concerning the power defect is most correct?
a.
The power defect is the difference between the measured l
power coefficient and the predicted power coefficient.
b.
The power defect increases the rod worth requirements necessary to maintain the desired shutdown margin following a reactor trip.
c.
Decause of the higher boron concentration, the power defect is more negative at beginning of core life, d.
The power defect necessitates the use of a ramped Tavg program to maintain an adequate Reactor Coolant System subcooling margin.
J
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l 1.__EBINCIELEg_CE_N6JCLE88_EgMEB_EL8NI_Q[EB8IlgN PAGE 4
3 ISEBdgp%B8dlC32_SE8I_IB8BSEEB.,8Np_E6Ulp_E69W QUESTION 1.05
(.75)
MULTIPLE CHOICE (Choose the best answer.)
Which of the following chemical agents may be added to control Reactor Coolant System Oxygen concentration while at 100% power?
a.
Hydrazine b.
Hydrogen Peroxide c.
Li thium Hydr o>:i de d.
Hydrogen QUESTION 1.06
(.75)
MULTIPLE CHOICE (Choose the best answer.)
What in the specific volume of saturated steam with a quality of 85% and a temperature of 500F7 a.
0.58 ft3/lbm b.
0.67 ft3/lbm c.
0.93 ft3/lbm d.
1.16 ft3/lbm
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****-)
-I___EBINCIELEg_gE_NUCLEBB_EgyEB_E68NI_g[EB811gN PAGE 5
2 IHEBggpyN9MICS,t_UEGI_IB6NSE[B_8Ng_E69]D_E6gy QUESTION 1.07 (1.00)
MULTIPLE CHOICE (Choose the best answer.)
What is the startup rate if power increases from 3000 cps to 8000 cps in twenty seconds?
a.
O.4 DPM b.
0.7 DPM t.
1.0 DPM d.
1.3 DPM OUESTION 1.08 (1.50)
MULTIPLE CHOICE (Choose the best answers.)
Which THREE of the following statements are TRUE, with regard to centrifugal pump cavitation?
a.
Assuming proper valve lineup, cavitation can be diagnosed from control board indications of pump discharge pressure, pump suction pressure, and pump flow.
b.
The point where cavitati on is most likely to begin is the high-pressure discharge area of the' pump.
c.
If available net positive suction head (NPSH) is positive, cavitation will not occur.
d.
The system fluid velocity has no effect on the probability of pump cavatation, e.
Undesirable consequences of cavitation may include erosion and pitting of the pump impeller, reduced flow rate being maintained by the pump, and damage from excessive vibration.
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1 l
1c__EBINCIE6E@_QE_NUC(E86_EQWEB_C6@NI_QEEB@IlgN FAGE 6
2 L
ISESdQDyN@dlC@i_bE@I_IB@Ng[EE_@ND_CLUlD_E6QW QUESTION 1.09 (1.00)
MULTIPLE CHOICE (Choose the best answer.)
l Which of the following best describes Keff?
a.
The number of neutrons in one generation causing a fractional change of population per neutron generation for a finite reactor.
b.
The number of neutrons in one generation divided by the number of neutrons in the previouc generation or a finite sized reactor, c.
The number of neutrons in one generation divided by the number of neutrons in the previous generation for an infinite reactor.
d.
The number of fast neutrons produced by fissionc of all energies divided by the number of neutrons absorbed in the fuel.
-QUESTION 1.10 (1.50)
Will the following power range instrumention be indicating HIGHER, LOWER or the SAME as actual power, if the instrument has been adjusted to 100% based on a calculated calorimetric?
a.
If the feedwater temperature used in the calorimetric was higher than actual feedwater temperature.
[0.753 b.
If the reactor coolant pump heat input used in the calorimetric is omitted.
[0.75]
OUESTION 1.11 (1.50)
How will an INCREASE in each of the following parameters affect the Departure from Nucleate Boiling Ratio (DNBR) (INCREASE, DECREASE, or NO EFFECT)?
a.
Coolant flow EO.75]
b.
Reactor power E0.75]
(***** CATEGORY 01 CONTINUED ON NEXT PAGE
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1 __fBINg]ELEg_gE_NUg6E88_E9 WEB _E69NI_gEEB91IgN PAGC 7
2 ISEBdQgyN301gg3_Sg61_]BONSEEB,,8Ng_E6Ulg_E69W QUESTION 1.12
(.75)
MULTIPLE CHOICE (Choose the best answer.)
During a reactor trip recovery a 1/M plot is being made, the i ni ti al 1/M data point was 1.0.
One hour after the trip, rod withdrawal was commenced.
Upon stopping rod withdrawal to take 1/M data, you find that the second 1/M point is 1.1.
Which of the f ollowing explains this increase in the 1/M value?
a.
This is not possible, the RO must have made an error when taking count rate data during the interval after stopping the rod withdrawal.
b.
The buildup of Xenon during the 1-hour delay added more negative reactivity than the rod withdrawal had added in positive reactivity.
4 c.
The source-detector geometry is incorrect, cousing more counts to be detected proportionately as reactivity is added.
d.
An inadvertent dilution ic in progress increasing the amount of neutron leakage that the excore nuclear instrumentation can see, l
I l
QUESTION 1.13
(.75)
MULTIPLE CHOICE (Choose the best answer.)
When positive reactivity is added with outward rod motion to a just cri ti cal reactor, a transient positive startup rate (prompt Jump) can be observed.
Which of the following is this prompt jump attributed to?
a.
The higher importance of the longer lived delayed neutron precursors.
j i
l b.
The increase in the prompt neutron fraction.
c.
The decrease in the prompt neutron fraction.
d.
The operating characteristics of the ex-core instruments.
(***** CATEGORY 01 CONTINUED ON NEXT PAGE ****+)
1 l
lt__ESINCIELEE_gE_NgC6E@S_EgBEB_E68MI_ GEES @IlgN1 PAGE O
IHEBMQDIN@MIC@t_HE@I_IB@ySEEB_8MD_E6UlD_E698 OUESTION 1.14 (1.50)
Reactor power is increased from 50% to 100%.
How will DIFFERENTIAL rod worth chance (INCREASE, DECREASE, or REMAIN THE SAME) for each of the f oliowing condi tions?
(Consider each case separately.)
a.
Rod position and baron concentration are held constant, temperature is allowed to decrease.
[0.753 b.
Rod position is constant, baron concentration is diluted to maintain temperature constant.
CO.753 QUESTION 1.15 (1.50)
How will each.of the f ollowing ef f ect the value of differential Baron worth (LESS NEGATIVE, MORE NEGATIVE, or NO EFFECT),
assuming all other conditi ons remain unchanged?
a.
Reactor coolant temperature decreases.
[0.75]
b.
Baron concentration increases.
[0.75]
QUESTION 1.16 (1.50)
How will each of the f ollowing af f ect the value of Shutdown Margin (INCREASE, DECREASE, or NO EFFECT), assuming all other conditions remain unchanged?
a.
Reactor coolant temperature decreases.
[0.753 b.
Xenon concentration increases.
[0.753
(*****
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1
PAGE 9
Iz__EBINCIE6ES_g[_NUC6EOS_EgyES_E68NI_9EESSIlgN 2 ISEBMggyNSMICg3_dESI_IE8BgEgB_8Np_E6Ulp_[69W W O.7[
QUESTION 1.17 For each of the conditions li sted bel ow, will the moderator temperature coefficient becomes MORE NEGATIVE, LESS NEGATIVE, or REMAIN THE SAME7 Assume all other conditions are unchanged.
iu.
c, 3:
DELETED n-
,1 r t i. -
c-b.
Core age increases.
E0.753 QUESTION 1.18 (2.00)
MULTIPLE CHOICE (Choose the best answer.)
Assume the reactor is Xenon free. You then take it to criticality and raise power to 50% at a rate of 5%/ min. (Allowed for this case.)
A trip occurs as power reaches 50%.
How will the Xenon concentration be trending in the following three situations?
Use one of the choices (1-5) below to answer the three questions.
1.
Increasing towards peak Xenon concentration.
l 2.
Increasing towards 100% equilibrium.
l 3.
Decreasing toward a dip.
4.
Decreasing toward zero percent power equilibrium value.
5.
At zero percent equilibrium value.
a.
One hour after the trip.
CO.53 b.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter the trip.
E0.5]
I c.
If after B hours from the trip, the reactor was taken back j
to criticality and power returned to 50% at 1%/ min.
What 1
would be the trend as power reaches 50%7 E1.OJ l
l l
i l
i
?
I i
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
l
1.__ESINCIE6ES_QE_ NUCLE @B_CQWES_EL@NI_QEEB@llgN PAGE 10 2
ISEBdgg1N@dlCQ1_dE@l_18@@@[EB_@NQ_[(glg_E6QW 1
QUESTION 1.19 (1.50)
Answer TRUE or FALSE to each of the following.
a.
A single RCP pump running during hot shutdown draws more motor current amperage than one of four running at power.
CO.53 b.
A single RCP running at cold conditions draws l ess motor amper age than when running at hot conditions.
CO.53 c.
The RCP motor amperage is higher when starting the pump than when running.
CO.5J i
QUESTION 1.20
(.50)
Answer the f ollowing TRUE or FALSE:
Concerning Pressurized Thermal Shock (PTS) events, transients which involve a rapid cooldown of the reactor coolant system are capable of subjecting the reactor pressure vessel to a thermal shock.
QUESTION 1.21
(.50)
Answer the-f ollowi ng TRUE or FALSE:
The buildup of Pu240 over core life (BOL to EOL) causes the doppler temperatur e coef ficient to become LESS NEGATIVE.
QUESTION 1.22 (1.00)
'The RCS is designed to allow and promote natural circulation.
What Reactor Coolant System design feature (s) ensures that natural ci r cul at i on (NC) will occur?
(***** CATEGORY 01 CONTINUED ON NEXT PAGE
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f 1
l L
lt__ESINGIELE@,_QE_GUQ!g88_EQWEB_.E6@NI_QEE6@llgM2 PAGE 11 ISESdQRXU8dlGS1_SE@l_I68MEEE6_@MQ_E6Ulp_E698 l
QUESTION 1.23 (1.00)
How is it possible to have adequate subcooling indication with accurate pressuri:er level indication and still form a steam bubble in the vessel head during natural circulation cooldown.
(*****
END OF CATEGORY 01
- )
a 1
1 2 __P66NI_DE@l@N_lNCLUDIN@_@@EEIy_@ND_EMER@ENCy_@y@lEM@
PAGE 12 OUESTION 2.01 (3.00)
Answer the following questions (a-e) either with "1",
"2",
or "3"
as indicated below based on the modifications made to the NNI and ICS power supplies.
Assume all equipment functions as designed unless otherwise given, that no operator action i s tat:en, and that all i
systems are in a NORMAL configuration prior to EACH given occurence.
"1"
- No power supply shift occurs.
ICS continues powered operation from the SAME source.
"2"
- Power to ICS i s lost.
ICS de-energizes.
"3" - Power supply shift will occur automatically such that ICS may continue powered operation f ron, a DIFFERENT power source.
a.
12Ovac Bus S1D is inadvertently de-energized.
[0.6]
b.
480v/12Ov stepdown transformer associated with inverter S1GB burns up and is de-energized.
[0.63 c.
Battery charger associated with inverter S1GB (H4BGB) shorts and is de-energized.
[0.63 d.
Fire in the NNI-X power supply ABT causes de-energi:ation of ALL 118vac NNI-X power downstream of the ABT.
[0.6]
e.
Inverter S1GB silicon controlled rectifiers (SCRs) fail causing a loss of inverter output.
[0.63 i
(*****
CATEGORY O2 CONTINUED ON NEXT PAGE *****)
2 __E6BNI_DggigN_ INCLUDING _S9Egly_8ND_gdg6GgNCy_gySIEDS PAGC 13 OUESTION 2.02 (3.50)
A design basis LOCA (large break) has occurred.
For each of the following questions, start with the reactor coolant leaving the break and continue to the point where the thermal energy EITHER collects in one place OR dissipates to the environment.
Be sure to identify each system or heat transfer component used to transfer this heat.
Consider each case separately.
I a.
What are the TWO paths that reactor decay heat THERMAL ENERGY will follow during injection phese with the Reactor Building spray actuated?
Neglect ambient losses through the Reactor Building structure and assume only one train of ECCS and supporting systems is operable.
[2.03 b.
What is the path that reactor decay heat THERMAL ENERGY will follow during recirculation phase with Reactor Building spray secured and normal Reactor Building pressure?
Assume no boiling is occurring in the core.
[1.5]
l l
I l
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1 I
i 4
I
(*****
CATEGORY O2 CONTINUCD ON NEXT PAGE *****)
y __
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/
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.)
i 2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENyi' SYGTEMS PAGE 14 b
.)
QUESTION 2.03 (3.00) i For the following question, refer to at t ached 'Ti gures 27 -I-C1, and 27-I-C2.
Note that valve positions shown may not necessarily refY ct normal OR emergency p1 ant conditions.
P-
'l f
3 I
J For each of the following motor operated 54:1vos listed, indicate one j
of the f ollowing :
s I
.\\
s t
s.
l SHbT sid p/1 upon'SFAS
\\
{
"S" if the valve receives an autt,
.ic initiation regardless of i ni :. a. ' position l
1 1:
"O" if the valve receives, an a/.tomitic OPEN cignal upon,.NNS initiation regardless nf initial position
(.
1 I
I
,a "N" if the valve receives NO automatic signal upon SFhj initiation.
i )
On Figure 27-I-C1 L'
\\
)
- 1) SFV-23604 f
f{
- 2) SFV-23508
~ />
- 3) SFV-25003/25004 V
- 4) SFV-23810
- 5) HV-23801
- 6) HV-23802 t'
(?
/
g
' (j On Figure 27-I-C2 f
- 7) SFV-26039/26040 p'
)
- 8) HV-26046 F
- 9) HV-26106/26105
- 10) HV-2OOO1/20002 EO.3 pointc eachJ 6,
\\
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=
OUESTION 2.04
':. 5 0
,',0 l
l c
1 v
y For the f oll owing questions, refer to 49tached Fi guren 2 7 -2.-C1 and g, ' ~
f
\\
27-I-C2.
1.
,\\;
l
(-
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'u 4 -,
-- - t _ m m.>
,e 4 sc
.. s.
+-,.%--
r4m.. -
C b. C., u... ' ~
[
j l
s '
l b.
For each of the points "A"
and " fs " as labelet on these twoj!
Figuren, what is the purpoue for thin flowpath?
ri o]
l
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l 1
\\
t l
\\
. t l
s t
s
(
t
(
).
\\,
\\
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s-(*****
CATEGORY O2 CONTINUED ON NEXT,PAGE
- 5.*)
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y
\\" wwer ne follow 2ng qu,<sti'ans rbsrding the Fire Protection System.
A r
f,y U
r
/'
)
i D. Hj w i s t'oc Fi re Pp/.estion hntSr System pressurized when ther e is
(
,j NO sys em /actuand?
[ 0. 4 3.
,i
,b. [ W is the Fire f,f btect i on Water Of utem pressuri z ed when there IS z
i J Cyste",'U$, nand?
E 1 ; '3 2 ts i
l, c.1 4 hat, entomati c fi re prcito-ti on. systs'h pr atects n + Oh Erner g en c y
./
Dscsel Ge 1e'ra tor s (GEA and'6EB)? /.'.;f. 51 5
~
7 c i
\\
f(
.s d.
What automatic fire, protection Gystem protects the TP Emergency Diesel Generators (GEA2 /and GEB2)? ' EO.33 c
7 g'
t.
I j'
/
QUESTION 2.06 (2.00) l y
The Component Cooling Water System suppliet five eeparate components
-1 on each Tiesttor Cool ant Picap end *1ot or.
l t,'
\\
,/
l W). s t kre,f our (4) of those,cive components?
y s
j s
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/;'
I;UESTION 2.07 (1.00)
/
i Y,2 The plant air supply f rom the normal, compressors i t, lost due to a lube
{
oi? fire at the compressors.
Whst is the basis for the regt!.i c ement tnat a plant shutdown must l
4 coamente i,* mediately if the 14RMAL plant air supply cannot bc o
r.3/ t or ed iY 30 mi nutes? [
I
(
)
[
' I' \\
\\
l
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t
' 5{.
L Y
(*****
CATEGORY O2 CONTINUED ON NEXT PAGE +++++)
2:__EL9NI_DEpigN_INCLUDINQ_g8EgIy_SUD_EdEBgENCy_Sy@IEUS PAGE 16 QUESTION' 2.08 (2.00)
Match the components listed in Column A with the correct location where they penetrate tne RCS.
Answers may be used more than once.
[0.5 each]
y, Column A Column B
- a. in - PZR ypray 1.
P-210B Suction cb.
Normal Makeup Line 2.
P-210D Suction Leb!nwc Line 3.
P-210A Discharge o.;
Decay Heat Removii Lin./
4.
P-210D Discharge l
1 5.
P-210C Discharge 6.
A Hot Leg 7.
B Hot Leg DUESTION 2.09 (1.00) j; The HPI and LF:I systems provide 100% capacity per train when operating in'their emergency modes. TPe Reactor Building Spray system trains i
have. only o 50% capacity in the emergency mode.
1 Why i t;.50?. 't.:apaci ty per train for the Reactor Building Spray system an acceptable condition?
\\
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j7 QUESTION 2.10 (1.00) s Regarding the Transamerica Delaval Inc. (TDI) Emergency Diesels:
Why will the loss of air pressure in the TDI Emergency Diesel Generator start-air syctem will defeat thL entire Shutdown Protection System?
[' f tp r.
$i s
p; l 1
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1--
l 1
2.__E69NI_ DES 19N_lNCLUDJNg_@@Egly_8MD_gMESQgNCy_@y@IgMS PAGE 17 J
i QUESTION 2.11
(.75)
]
MULTIPLE CHOICE (Choose the best answer.)
With respect to the 125 volt DC system:
On a sustained loss of all AC power sources, HOW long will the station batteries BC and BD carry their respective loads?
a.
0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> b.
2.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> c.
4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> d.
6.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> QUESTION 2.12
(.75)
MULTIPLE CHOICE (Choose the best answer.)
Which one of the following statements is accurate concerning the OTSG's?
a.
Primary and secondary side blowdown (during plant heatup) is accomplished by means of drain connections near the lower tubesheet, b.
The startup range instruments will provide indication flooding of the aspirating ports.
c.
The auxiliary feedwater header penetrates near the top of the OTSG shell and sprays the feedwater on the upper cylindrical baffle.
d.
The Orifice plates, located in the lower downtomer section are adjusted to their minimum opening to retard level oscillations.
(*****
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2 __EL8NI_DESIGy_ INCLUDING _SOEEIy_6ND_ EMERGENCY _SygIEMg PAGE 1G 3
1 l
QUESTION 2.13
(.75)
MULTIPLE CHOICE (Choose the best answer.)
Which one of the following correctly describes the trip system of the main turbine?
a.
When the auto-stop (turbi ne control ) oil pressure decreases, the interface trip valve will open allowing the EHC Control oil to dump to drain.
b.
When the EHC Dil pressure decreases, the interface trip valve will open, allowing the auto-stop (turbine control) oil to dump to drain, c.
The interf ace trip valve is solenoid actuated and when open, will dump both auto-stop (turbine control) oil and EHC control oil to drain.
d.
A full turbine trip ret;uires the servo valves for all f our sets of turbine valves (throttle, governor, reheat and interceptor) to open.
QUESTION 2.14
(.75)
MULTIPLE CHOICE (Choose the best answer.)
Which one of the following statements is CORRECT regarding the design of the internal vent valves?
a.
The vent valves are designed to open in the event of a HOT leg break when the pressure differential reaches at least 15 psi.
b.
The vent valves are designed to open in the event of a COLD leg break when the pressure differential reaches at least 15 psi.
c.
In the event of a HOT leg break, the valves should begin to open with a delta p of about 0.3 psid and be f ully open at l
1.5 psid.
d.
In the event of COLD leg break, the valves should begin to open with a delta P of about 0.3 puid and be fully open at 1.5 psid.
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2.__PLgNI_DEglgN_JNCLUDJNQ_@@EgII_gND_EMESgENCy_gySIEMS PAGE 19 I
DUESTION 2.15
(.75)
MULTIPLE CHOICE (Choose the best answer.)
Which one of the f ollowing statements about the TDI Emergency l
Diesels is true?
a.
The air start reservoirs f or either diesel are designed to allow at least 10 starts each without recharging.
b.
The fuel oil day tank for each diesel must be refilled once per day during. operation at rated power.
c.
During startup, oil is supplied to the main drive end bearing by an auxiliary gear-driven pump.
d.
System design does not provide for cross-connecting starting air supplies between the two diesels.
QUESTION 2.16
(.75)
MULTIPLE CHOICE (Choose the best answer.)
Which one of the following is NOT a design difference between the safety / regulating control rods and the APSRc?
a.
APSR drives have ball valves and bypass parts.
b.
APSR couplings have larger diameters and shorter keys to prevent coupling an APSR drive to a safety or regulating roti or vice-versa.
c.
On APSR drives a small button on the lower portion of the segment arm prevents the lead screw from being disengaged when power is lost.
d.
APSRs do not have buffer springs in the buffer assembly.
(***** END OF CATEGORY O2
- +)
3 __INSIguMggIg_ ego _CONIBO65 PAGE 20 QUESTION 3.Oi
.;7 /,0 l
Answer the f ollowing questions regarding the aunillary steam pressure control valve.
a.
Following a loss of ICS electrical power, you are required to adjust auxillary steam header pressure.
What position will the control valve be in following the loss of ICS power?
E0.5J
-+g_,.
- .1-t
.- +. 2 7 I ~
DELETED s
unu 4-r m-w n, ms %
m, 4,
1g is_ : g _,; ; ;r y-t.,
- g,,; -
7,5:
m 4.- s r rr n _ ;
_.; 7 c.
What effect, if any, will a loss of instrument air to the auxillary steam pressure controller (PC-36014A) have on auxillary steam pressure?
E0.5]
OUESTION 3.02 (2.00)'
The reactor is operating at 100% power when'all electrical power to the ICS is lost.
What effect will this have on your ability to l
CONTROL the following components, assuming no other power losses or malfunctions have occured?
a.
Control rods CO.53 b.
Main Feed Pumps CO.5]
c.
Turbine. Bypass Valves
[0.53 d.
Atmospheric Dump Valves
[0.53 QUESTION 3.03 (2.00)
You are maintaining the plant in Hot Standby at 527 F and 10-6 an.p t. in the intermediate range.
Turbine bypass valves are in " normal" controlling steam pressure at 855 psig.
Suddenly, the TBV's go shut and RCS temperature begins to rise.
a.
What are THREE (3) conditions which provide interlock functionc to the TBV*s and which could have caused this event?
Setpoints are not required.
E1.53 b.
Are the TBV interlock functions active during TBV automatic j
control, TBV manual control, or both?
[0.53
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l
PAGE 21 3_ _. ___I _NS_ T RU ME_N_T S_ A_ N_D__C_O_N_T_R_O_L_S QUESTION' 3.04 (1.50)
You are operating at 100% power, steady state, when a stsam line break occurs in the P-318 steam supply line.
Both OTSG's depressurize at the same rate to 550 psig, a.
Assuming you took no action to this point with the EFIC or feedwater systems, in what position (open, shut, or modulated) do you
-expect the following valves to be for EACH OTSG?
b, You isolate the break to the D OTSG, and A OTSG pressure recovers to 750 psig while B OTSG continues to decrease.
Assuming you took no action to alter the automatic response of the EFIC system, in what position (open, shut, or modulated) do expect the following valves to be for EACH OTSG?
E1.03 Auxillary Feedwater control valves, Auxillary Feedwater block valver..
QUESTION 3.05 (2.00)
You are operating at 80% power when the B OTSG steam pressure EFIC instrument channel B malfunctions and fails high.
Other than AFW initiation and Vector (Feed Only Good Generator) functions, which THREE (3) EFIC valve CONTROL or LOGIC functions will be affected by the loss of this instrument?
l
(*****
CATEGORY 03 CONTINUED ON NEXT PAGE *****)
j
3___INSIgutggIg_ggD_CONIggLE PAGE 22 DUESTION 3,06 (3.50)
Answer the f ollowi ng questions either:
T for ALWAYS TRUE F for ALWAYS FALSE 7 for may bo EITHER TRUE or FALSE depending on f actors not given in the statement Assume all plant equipment FUNCTIONS EXACTLY AS DESIGNED unless otherwise SPECIFICALLY GIVEN in the statement.
Treat each part independently.
a.
An OTSG overfill condition exists due to a stuck open AFW control j
valve.
When OTSG 1evel reaches the high level.setpoint, EFIC will J
isolate AFW flow from the affected OTSG.
[0.73 b.
Plating an EFIC Channel in maintenance bypass will still allow a manual trip or an ESFAS trip of either AFW trip module.
CO.73 c.
When the trip busses of the MFI Trip Module are reset following MFI actuation and after the MFI initiation condition has cleared, the Main Feedwater control and startup valves will go to 50%.open.
[0.7]
d.
An MSI trip exists on A OTSG.
You may restore manual control of j
A OTSG main feedwater valves by taking A OTSG MFI Trip Modules to MANUAL.
[0.73 e.
Following an AFW actuation you have taken manual control of A OTSG AFW control valves at the Bailey TT20 Hand / Auto stations and have placed the AFW Trip Module in manual to restore level.
If a subsequent A OTSG low level setpoint were reached, the A OTSG AFW control val ves will automatically revert'to AUTOMATIC control.
[0.7]
i 1
QUESTION 3.07 (1.00)
EFIC channel A is in maintenance bypass.
How will EFIC respond if RPS channel B is placed in Bypass?
l
)
)
(*****
CATEGORY 03 CONTINUED ON NEXT PAGE *****)
___________o
l 2.__IUSISUMEUIS_9ND_CgyIBg6S PAGE l
QUESTION 3.08 (2.50)
Indentify the automatic action, if any, associated with an 01 arm condition on each of the following monitors.
Component numbers a r t-not required.
Include automatic State of California notifications.
a.
Retention Basin Monitors R15017A/B EO.53 b.
RHUT Monitor CO.5]
c.
Auxillary Bldg. Grade Level Monitor CO.53
- d. Reactor Bldg. - Stack Monitor EO.53 e.
Reactor Bldg. Atmospheric Leak Detection Monitor
[0.5J QUESTION 3.09 (1.00)
In addition to replacing the existing trend recorder, a second trend recorder is being installed on HICO.
For what primary purpose is the SECOND trend recorder necessary?
a.
[O.53 b.
Will the indications on this second recorder be available during a loss of NNI power? EO.53 QUESTION 3.10 (2.00)
Following an SFAS, all Emergency Diesel Generators are running
- unloaded, a.
If GEA2 were to trip due to a vali d emergency automatic trip condition, which TWO (2) automatic trip functions would you suspect?
[1.03
- b. Answer question (a.) above assuming Diesel Generator GEA had tripped instead. E1.O]
(***** CATEGORY 03 CONTINUED ON NEXT PAGE
- )
PAGE 24 ss__INg16UDgNI@_8ND_CgNIBg6@
QUESTION 3.11 (3.50)
You are operating at 50% power, steady state, with three Reactor Coolant Pumps (RCPs) running.
RCP "D"
is shutdown.
All.other plant equipment is operable.
The RCP "A"
power monitor output feils to
- zero, a.
For each of the following systems, what automatic actions result from this failure?
If there is NO response, state "NONE."
- 1) RPS
[0.5]
- 2) ICS CO.53
- 3) CFIC CO.53 b.
Reactor Coolant loop flow instruments provide ACTUAL indication of loop / core flow.
Which of the above three systems in part (a.),
if any, are using actual flow inputs f ollowing this malf unction?
[1.O]
- c. To what TWO (2) systems do the RCP power monitors provide valve control interlocks or permissives?
[1.03 QUESTION 3.12 (2.50)
Answer the f ollowing questions regarding the Nuclear Instrument System assuming NI-7 has failed high and RPS Channel "C"
is bypassed.
Answer each part independently, a.
The plant is operating at 87% power.
An I and C technician working in the "C" RPS cabinet asks if he can de-energi:e NI-3.
Will this action affect the Source Range instruments (YES or NO)?
[0.53 b.
Your are conducting a plant shutdown from 30% power.
What specific conditions will UN-bypass the High SUR rod withdrawal inhibit circuit?
Specify channel or component numbers and setpoints.
[1.03
- c. You are perf orming a plant startup from a Source Range level of 10 to the fourth cps.
What specific conditions throughout the startup will inhibit rod withdrawal?
Specify channel or component numbers anti setpoints.
E1.OJ
(***** END OF CATEGORY 03 *****)
i
l 4.__EB9EEDU3ES_;_UgBM962_gB$g6M8L _EME6GENgY_gND PAGE 25 3
j i
68919699198L_QgNIggL i
)
QUESTION 4.01 (2.50)
Answer each of the following questions TRUE or FALSE based on the requirements of the Emergency Operating Procedures (EOP's). Answer each part i ndependentl y.
[0.5 points each]
a.
You are executing the " Loss of Subcooling Margin" CE.03]
procedure when you determine that a steam generator tube rupture (SGTR) is occurring.
You must IMMEDIATELY EXIT E.03 and TRANSFER to
" Steam Generator Tube Rupture" CE.063.
b.
You are executing the " Inadequate Core Cooling" EE.07] procedure when you determine that a CGTR is occurring.
You must IMMEDIATELY EXIT E.07 and TRANSFER to SGTR CE.06].
- c. You are executing the SGTR CE.063 procedure when you determine that a loss of subcooling margin (SCM) has occurred.
You must IMMEDIATELY EXIT E.06 and TRANSFER to " Loss of SCM" EE.033.
d.
You are executing the " Excessive Heat Transfer" [E.053 procedure and have just shut the Turbine Bypass valves' f rom A OTSG when you determine that the e.:cessive heat transfer transient has terminated.
You must COMPLETE the isolation of A OTSG and CONTINUE with E.05 until the procedure transfers you to a different procedure.
e.
You are executi ng " Loss of Heat Transfer" [E.04] when you determine that a loss of offsite power has occurred.
You must STOP PERFORMING E.04 and TRANSFER to " Vital System Status Verification"
[E.02].
QUESTION 4.02 (2.00)
(All rOS on be kHam)
The reactor is initially subtritical at normal operating pressure and temperature.
What four (4) dif f er ent transient symptomt, would require entry into the Emergency Operating Procedures?
(*****
CATEGORY 04 CONTINUED ON NEXT PAGE *****)
P(40E 26 di__BBgCEQUBES_;_Nggd86,_9BNgBMSL _EMEBgENCY_ONg 3
689196991C06_C9NIBg6 DUESTION 4.03 (3.00)
The following sequence of events occurs:
Two (2) Reactor Protective Channels trip but the reactor does NOT trip Ian Anti cipated Transient Without Scram),
THEN the reactor does NOT trip by manual reactor trip pushbutton, 1
THEN the reactor trips upon subsequent operator action, BUT two (2) rods remain out of the core, a.
Which IMMEDIATE ACTION steps specified in the Emergency Operating Procedure / are APPLICABLE TO THIS SEQUENCE OF EVENTS?
[1.O]
A 5.01 b.
What three (3) methods are specified in the EOP's to reduce reactor power FOLLOWING failure of the manual reactor trip pushbutton?
E1.53 c.
Is the plant designed to achieve greater than 1.0% shutdown margin following a reactor trip with two (2) rods stuck out?
(YES or NO)
E0.5]
OUESTION 4.04 (1.50)
The execution of the " Loss of Subcooling Margin" procedure E.03 requires ALL Reactor Coolant Pumps (RCP's) to be tripped within two (2) minutes of entering E.03.
a.
If RCP's are NOT tripped within two (2) minutes of entering E.03, what action must be taken with the RCP's? EO.62 b.
If RCP's are NOT tripped when subcooling margin was loct, WHY could core damage result?
EO.92 l
1
(****-*
CATEGORY 04 CONTINUED ON NEXT PAGE *****)
4:__E8pgEpuBEg_;_NgBd86,_6pNgBd862_EDEggENgy_8Np PAGE 27 889196991986_G9NIgg6
-QUESTION 4.05 C,@e>7 l,I Answer the f ollowing based on "RCP Mal f uncti ons" procedure EC.6].
Treat each condition (a - e) separately, and apply it to only ONE RCP.
For each of the conditions (a - e), indicate ONE of the following:
"1" if a Reactor Coolant Pump (RCP) trip is required by procedure, "2"
if NO RCP trip is required, but the parameter is used by ITSELF to determine if a RCP trip is required, or "3"
if the parameter is NOT used by ITSELF to determine if a RCP trip is required.
a.
Lube oil reservoir level' Hi-Lo Alarm EO.5]
b.
Stator temperature is 160 C E0.53 11 :_g ge" e n _ r E DELETED
-,na mm,3
<-m>
r,.
rr # 4 m m -4
- mme,
,mei
., c d.
Seal injection flow is ZERO
[0.52 m
i m m,-
sn a 4,,+m--m 4 am
-m,,
4
+.x.--
4 4 e :- -
rn QUESTION 4.06 (1.50)
You are implementing "Small Reactor Coolant Leak" procedure EC.33.
What are the THREE (3) conditions which require you to manually trip the reactor?
QUESTION 4.07 (3.50)
Answer the following questions based on the Loss of Steam Generator j
Feed or Steam Generator Overfeed procedure EC.103.
Clearly specify the parameter (s) and value(s).
a.
You are experiencing a Loss of Steam Generator Feed transient while at 50% reactor power.
Under what TWO (2) conditions must you manually trip the reactor?
fl.53
- b. You are experiencing a Steam Generator Overfeed transient while at 90% reactor power.
Under what THREE (3) conditions must you manually trip the reactor?
[1.23
- c. You are experiencing a Steam Generator Overfeed transient while at 10*/. r e ac t or power.
Under what TWO (2) conditions must you
]
manually trip the reactor?
E0.03
(***** CATEGORY 04 CONTINUED ON NEXT PAGE
- )
i J
4; PROCEDURES - NORMAL _ABNgRMAL1_gMgRGENCY_AND PAGE 28 1
E8 Dig 6gg1CeL_CQNIBQL 1
I 1
QUESTION 4.08 M 2, f Answer the f ollowing questions regarding f acility radiation control procedures.
{
a.
What is your weekly whole body exposure administrative limit and whose authorization is required to exceed it?
[0.53 b.'What is your quarterly whole body exposure administrative limit and whose authorization is required to exceed it?
CO.5]
c.
What is the maximum whole body exposure that an untrained visitor
.may receive?. CO.5]
d.
How often should posted radiation signs showing current exposure information be' updated?
[1.03
.. J,,
+%
-1,,
_. 3 3,s pm,-cmm_rme g7g us m o 4
.,m t m
.s-s
-,,3 s
e n.._- _
, a_
c.,_
u QUESTION 4.09 (3.50)
You have just completed a rapid plant shutdown to subtritical conditions.
The Shift Supervisor asks you to determine the ACTUAL shutdown. reactivity of the core prior to commencing plant cooldown.
A copy of " Reactivity Balance Calculations" procedure B.6 is provided.
Interpolate the graphs as necessary, use Enclosure 5.4 as an answer sheet, and show all work including interpolations.
Given the f ollowing current ' plant status, what is the ACTUAL shutdown reactivity of.the core?
Date:
10/20/87 Times 1200 Cycl e:
7.
Time since shutdown:
3.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> Previous power history:
58 days at 80% power Time in cycle:
700 ppmB RCS temperature:
527 degrees F RCS pressure:
2125 psig Control rod positions:
Group 1 Saf ety Rods withdrawn in preparation for-the cooldown, Group 2-7 rods fully inserted. APSR's are 25% WD.
(*****
CATEGORY 04 CONTINUED ON NEXT PAGE *****)
l 4 __EBOgEDyGES_;_NOGU@61_@pNOBd@61_EdESGENgy_@ND PAGE 29 BBD1969EIG06_GQNIgg6 l
l DUESTION 4.10 (1.00) l One of the initial conditions required by the " Reactor Coolant System l
Heatup to Hot Shutdown" procedure EB.2] prior to exceeding 200 degrcos F is to verify the reactor protection system channels are in Shutdown Bypass and the high flux trips are set to less than or equal to 5%
power.
l l
a.
Why was 5% power chosen as the limit for the high flux trip?
CO.53 b.
Is the 5% high flux l i mi t during Shutdown Bypass conditions a Technical Specification requirement?
(YES or NO). [0.5]
OUESTION 4.11 (1.uu?
You ar e executing the " Plant Heatup and Startup" procedure [B.2] from an RCS temperature of 150 degrees F to 250 degrees F.
What TWO (2) methods will this procedure use to ensure maximum tube-to-shell OTSG differential temperatures are not exceeded?
l J
I i
(***** END OF CATEGORY 04 *****)
(************* END Ur EXAMINATION ***************)
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EFFECTI'.'E LATI 08-31-57 Rev. 27 WP3873P
.f-* s.
D-0052P B.6 REACTIVITY BALANCE PROCEDURE 1.0 PURPOSE 1.1 To provide a method for estimating critical rod positions or critical boron concentrations prior to startup.
1.2 To provide a method for determining available shutdown margin during power operation and for determining actual shutdown margin with the reactor shutdown.
1.3 To provide a method for estimating the effect of withdrawing safety rod groups on subcritical multiplication.
1
2.0 REFERENCES
2.1 Technical Specification 3.5.2 2.2 Physics Test Manual 61-10000732-19 (present revision) 2.3 Memo R. G. Rosenstein to RG Mc Andrew, RCR 001/NFM-87-004, Subj:
Physics Data, Jan. 26, 198-7 2.4 OP 8.6 Revision 26 (Superceded by this revision) 3.0 LIMITS AND PRECAUTIONS 3.1 If criticality is not achieved within 0.8% delta K/K of the predicted critical rod position or boron concentration, then the data used in this procedure needs to be reviewed by Plant Nuclear Engineer and the shutdown margin needs to be re-evaluated by Plant Nuclear Engineer and an SRO.
3.2 To ensure accurate results, the plant conditions for which this procedure is being used need to correspond to the reference conditions on the enclosures, or appropriate corrections made, per this procedure.
For estimated critical condition calculations, group 8 rods are assumed 3.3 to be at 25% WD.
After the APSR pull near EOC, the group 8 rods are assumed to be 100% WD for these calculations.
An SRO shall either perform the calculations in this procedure or check 3.4 them.
Rev. 27 8.6-1
4.0 PROCEDURE i
This procedure is divided into'the following sections.
Select the section that applies to the evaluation you want to perform and. proceed to that section.
L Estimated Critical Position, 4.1 Estimated Critical Boron Concentration, 4.2 At Power Shutdown Margin Verification, 4.3 Suberitical Shutdown Margin, Verification, 4.4 Subcritical Multiplication, Enclosure 5.15 4.1 Estimated Critical Rod Pos,ition (ECPJ NOTE:.1 will be used for this calculation. The fuel excess reactivity is obtained based on burnup.
The reactivity worths associated with boron, xenon, temperature, and samarium are found and summed with the fuel excess reactivity.
Then the' control rod group 5-7 reactivity position is found which when summed with all the above giver a total core reactivity of 0% delta K/K.
The corresponding rod position is the ECP.
The upper and lower rod position limits are then determined.
The actual critical position ir recorded during startup.
4.1
.1 Complete Enclosure 5.1.
4.2 Estimated Critical Boron Concentration'n IEC81 NOTE:.2 will be used for this calculation.
The fuel excess reactivity is obtained based on burnup.
The reactivity worths associated with xenon, temperature, desired control rod insertion, and samarium are summed air-fuel excess ~ reactivity. The boron concentration is then found which when summed with all the above gives 0% delta K/K.
Then the upper and lower rod positions are de te rmined.
The actual critical boron concentration / critical rod position is recorded later.
4.2
.1 Complete Enclosure 5.2.
4.3 At power Shutdown Margin Verification 4.3
.1 Verification of Shutdown Ma~rgin with Reactor At Power Rev. 27 8.6-2
PROCEDURE (Continued) h
,f~
r
' NOTE:
Ifthereactoriscriticaland411controlro$sare operable, then the minimum shutdown margin exists provided the shutdown margin ins 6rtion limits are not violated per Tech Spec 3.5.2.
/
If the reactor is operating with a known inoperable rod, then shutdown margin is adjusted accordi.ngly.
See Actual
-Shutdown Margin Calculation, Step 4. 3.2.
4.3
.1.1 No Known Inoperable Rod; Complete Enclosure 5.3.
4.3
.2 Actual' Shutdown Margin Calculation At Power With Known Inoperable Rod NOTE:
This is required only when a control rod has been declared inoperable to demonstrate compliance with Tech Spec Shutdowr.
Margin requirements while' operating with an inoperable rod..4 will be used.
The 532*F, all rods in, stuck rod out, 1% delta K/K shutdown boron concentration is determined based on burnup.
The reactivity worth associated with xenon, samarium',-and the inoperable control rod are summed and the sum converted to-equivalent boron.
This equivalent boron.is then subtracted from the all rods in-stuck rod out, 1% delta K/K shutdown boron ~ concentration to determine the required' shutdown boron for.the actual
. ("h; plant conditions.
The minimum shutdown margin criterion is L. '
satisfied if the actual RCS boron is greater than the value just determined.
. 4.3
.2.1 Complete Enclosure 5.4.
4.4 Subcritical Shutdown Margin Verification 4.4
.1 Verification of Shutdown Margin With Reactor Subcritical NOTE:
If the reactor is subcritical and all control rods are operable, then the minimum shutdown margin exists provided the RCS boron concentration is equal to or greater than the minimum in Enclosure 5.12, or, if control rod group 1 is withdrawn, Enclosure 5.13.
4.4
.1.1 Complete Enclosure 5.4.
4.4
.1.2 If an inoperable rod exists or xenon or samarium worth is
'j being taken into account, then Shutdown Margin is adjusted -
accordingly.
See Actual Shutdown Margin Calculation, step 4.4.2.
,e Rev. 27 8.6-3
PROCEDURE -(Continued) 4.4
.2 Actual Shutdown Margin Calculation, Suberitical With Inoperable Rcc NOTE:
This is required only when a control rod has been declared inoperable or it is desired to take credit for existing xenon reactivity..4 will be used,.
The 532 F, all rods in, stuck rod out, 1% delta K/K shutdown boron concentration is determined based on burnup.
The reactivity worth associated with the xenon, samarium, and the inoperable control rod are summed and the sum converted to equivalent boron. This equivalent boron is then subtracted from the all rods in-stuck rod out, 1% delta K/K shutdown boron concentration to determine the required shutdown borce for the actual plant conditions. The minimum shutdown margin criterion is satisfied if the actual RCS boron is greater than the value just determined.
4.4
.2.1 Xenon Credit Desired.
Complete Enclosure 5.4.
5.0 ENCLOSURES i
5.1 Estimated Critical Rod Position 5.2 Estimated Critical Boron Concentration 5.3 Shutdown Margin Verification At Power', All Rods Operable 5.4 Required Shutdown Boron Concentration or Actual Shutdown Margin Calculation 5.5 Suberitical Shutdown Margin Verification; No Xenon Credit 5.6 Core Excess Reactivity 5.7 Differential Boron Worth 5.8 Xenon Worth After Shutdown 5.9 Temperature Reactivity. Coefficient 5.10 Samarium Worth After Shutdown 5.11 Control Rod Worths, HZP 5.12 Shutdown Boron Concentration, All Rods In-Stuck Rod Out 5.13 Shutdown Boron Concentration, All Ros In-Stuck Rod Out, Control Rod Group 1 Out 5.14 Inoperable Rod Penalty Rev. 27 B.6-4
ENCLOSURES (Continued) fs 5.15 Subcritical Multiplication Worksheet 5.16 Control' Rod Group Worth 5.17 Shutdown Reactivity For Suberitical Multiplication Calculation-1 S
pe
(..
Rev. 27 B.6-5
EWCLOSURE 5.1 ESTIMATED CRITICAL ROD POSITION Calculation Effective for:
Date Time with the following reactor conditions:
RCS temp
'F Burnup EFPD Boron ppm Amount of time Shutdown Hrs.
Reactor Power Prior to Shutdown 1.
Core Excess Reactivity Obtain the 532aF fuel excess reactivity at the currect cycle burnup from Enclosure 5.6.
Always positive.
% OK/K 2.
Boron Worth NOTE:
Use the 532*F differential Boron worth, a.
Obtain the differential Boron worth for the present RCS temperature and burnup from.7.
% SK/K/ ppm.
Always negative.
b.
Obtain the Boron worth by multiplying 2a by the present Boron' concentration.
Always negative.
% OK/'
% SK/k/ ppm x ppm =
3.
Xenon Worth a.
Obtain the current Xenon worth by running the off-line Xe-worth calculation program or by calling up Bailey computer group 32 (Reactivity Balance) or Enclosure 5.8.
Circle mothed used.
Always negative.
% OK/..
is not 532aF).
4 Temperature Correc tion (only if Tave a.
Obtain the 532aF temperature coefficient from.9.
% OK/K/oF Always negative, b.
Subtract 532*F from present RCS temp.
- F
-532 0F
'F 4.
c.
Obtain temperature effect by multiplying temperature difference obtained in (b) by temperature coefficient obtained in (a).
OF x
%$K/K/*F
% AK/K i
Product can be + or ENCLOSURE 5.1 PAGE 1 0F Rev. 27 l
I B.6-6
ENCLOSURE 5.1 (Continued)
ESTIMATED CRITICAL ROD POSITION 7 -.
5.
Samarium Worth a.
Obtain Samarium worth from Enclosure 5.10.
Always negative.
% SK/K
^
6.
Total Inserted Rod Worth Required a.
Algebraically sum reactivity values obtained in Steps 1-5 and reverse the sign.
% OK/K b.
If the value obtained in step 6 is positive, THEN boron concentration must be reduced to obtain criticality at the desired rod position.
Go to Enclosure 5.2.
7.
. Estimated Critical Rod Position Rod Index a.
Obtain the Gp 5-7 position at Gp 5 which the inserted worth is Gp 6 equal to the result of 6 from Gp 7.11.
Use 5.11A for Xenon free and 5.118 for near peak Xenon b.
Determine the rod configurations corresponding
~1, to the 20.8% OK/K limit Calculation valid for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> if Xenon free; 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> if not Xenon free.
Calculation valid until:
NOTE:
TF the value in step 6 is less than 0 8% OK/K, THEN, enter 300% for the upper limit rod index.
Value in item 6 - 0.8% OK/K
%SK/K Value in item 6 & 0.8% AK/K :
%$K/K Lower Limit Upper Limit Rod Index,
Rod Index Gp 1-4 100%_ Gp 5 Gp 1-4 100%
Gp 5 GP 8 25%
Gp 6 Gp 8
_25%
Gp 6 Gp 7 Gp 7 ENCLOSURE 5.1 PAGE 2 0F 3 Rev. 27
{
B.6-7
ENCLOSURE 5.1 (Continued)
ESTIMATED CRITICAL ROD POSITION 8.
Record the actual critical rod configuration at 10E-0 Amps.
Gp 5 Gp 6 Gp 7 Rod Index 9.
Forward a copy of coinpleted Enclosure to the Plant Nuclear Engineer.
Calculation Performed By Date Calculation Checked By Date ENCLOSURE 5.1 PAGE 3 GF '
Rov. 2/
B.6-8
ENCLOSURE 5.2 ESTIMATED CRITICAL BORON CONCENTRATION 9
Calculation Effective for:
Date Time with the following reactor conditions:.RCS temp
Reactor power Prior to Shutdown 1.
Core Excess Reactivity obta'in the 532*F fuel excess reactivity at the current cycle burnup from Enclosure 5.6.
Always positive.
% OK/K 2.
Xenon Worth a.
Obtain the current Xenon worth by running the off-line Xe-worth calculation program or by calling. up Bailey computer group 32 (Reactivity Blalance) or Enclosure 5.8.
Circle method used.
Always negative.
% OK/K 3.
Temperature Correction (only if Tave is not 532*F) a.
Obtain the 532"F temperature coefficient from.9.
%aK/K/0F
/~',
b.
Subtract 532*F from present temp.
'F
-532
'F
(,,/
"F c.
Obtain temperature effect by multiplying temperature coefficient obtained in (a).
- F x
%SK/K/*F
% OK/K 4.
Control Rod Worth a.
Record the desired critical control rod configuration below:
Gp 5
% WD 4
Gp 6
% WO Gp 7
% WD Rod Index
% WD b.
Obtain inserted worth of Gp 5-7 from Enclosure 5.11.
Always negative.
% OK/K l
l ENCLOSURE 5.2 PAGE 1 0F 3 Rev. 27 B.6-9
ENCLOSURE 5.2 (Continued)
ESTIMATED CRITICAL BORON CONCENTRATION
.5.
Samarium Worth a.
Obtain Samarium worth from Enclosure 5.10.
Always negative.
% OK/K 6.
Total Boron Worth Required Algebraically sum reactivity values obtained in a.
Steps 1-5 and reverse the sign.
% OK/K 7.
Estimated Critical Boron Concentration a.
Obtain the differential boron worth from.7.
%SK/K/ ppm Always negative.
b.
Obtain the boron concentration required to provide the total worth obtained in Step 6 by dividing 6a by 7a
% oK/K +
%SK/K/ ppm =.
ppm This value is the Estimated Critical Boron.
c.
Determine the rod positions corresponding to the t.8% SK/K limits.
Value in step 4b - 0.8% OK/K =
% OK/K Value in step Ab + 0.8% OK/K =
?. OK/K Lower Limit Upper Limit Rod Index
% WD Rod Index
% WD Gp 5
% WD Gp 5
% WD Gp 6
% WD Cp 6
% WD Gp 7
% WD Gp 7
% WD 8.
Actual Critical Rod Configuration at 10 E - 8 Amps.
Gp 5
% WD Gp 6
% WD Gp 7
% WD Rod Index
% WD Calculation valid for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> if xenon free; 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> if not xenon free.
Calculation valid until ENCLOSURE 5.2 PAGE 2 CF 3 Rev. 27 B.6-10
./
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'i ENCLOSURE 5.2 (Contin'ued)
F-
' 'N w
ESTIMATED CRITICAL BORON CONCENTRATION I
O
/s 9.
Forward a copy of completed Enclosure To the Plant Nuclear Enginedr.
Ip.
l
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6 ga Date
$ 'l.
Calculation Performed By
),
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Calculation Checked.By'
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- ENCLOSURE 5.3 SHUTDOWN t%RGIN VERIFICATION AT POWER ALL RODS OPERABLE p
h j'
(
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1 C4 ' F"etent. control rod index is'
% WD.
M#
2.0- Lowcet control red-index allowed at present cycle burnup by TS 3.5.2
- j Figures 3.5.2-1 through 6 for shutdown margin is
% WD.
-\\
.i 3.0 If Step 1.0 is gecator than or equal.to Stop 2.0, then adequate shutdown r'
margin exists.
. i C, a 1.0 > 2.0-7 Yes / No v
i.
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4.0 If. control rod index is less than that allcwed by TS 3.5.2 for shutdown-
,u,
t margin,,then inform the Shift Supervisor immedd.ately.
xs
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5.0 Forward a, copy of the completed Enclosure l o ths Plant Nuclear Enginee.
t
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Performed Gy Date Time
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B.6-12
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A ENCLOSURE 5.4 l
t 3
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' AT POWER /SUBCRITICAL SHUTDOWN MARGIN CALCULATION 1
f
(
Calculation Effective for: Date Time with the following
)-
reactor conditions:
Power
%FP
_RCS temp
'F Burnup EFPD Gp 1 rod position Inoperable rod:
Gp #
Rod #
n.
9.. '
1.
Obtain reference shutdown boron concentration from:
ppm Fnclosure 5.12 (No Xe) Gp 1-7 @ 0% WD, Gp 8 0 25% WD, or\\C3 closure 5.13 (No Xe) Gp 10100% WD, Gp B 25% WD, f or' the applicable conditions of burnup and RCS temperature.
For shutdown margins at power, use
'.12 for a-temperature of 532*F.
Circle one used.
1 2.
Adjus,t for non-referm corditions as follows:
'1 a.
Obtain appropriate Xenon worth from Bailey computer group 32 (Reactivity Balance), or
~
Enclosure $3.8, or off-line Xe-worth calculation o
F program.
Always negative. Use,0 for conservatism if desired.
Circ'le' method used.
% AK/R V.
b.
Obtain Samarium wirth from Enclosure 5.10.
g"
% AK/K Always negative.
yl c.
If operating with an inoperable rod or shutdown q
ywith a stuck rod, obtain Inoperable Rod Penalty i
(from Enclosure 5.14.
Always positive.
% AK/K f
The sum in step d below will always be R,13 :
negative unless on Inoperable Rod Penalty l
l greater than the sum of Xonon and Samarium worth is being used.
' i d.
Algebraically sum (a)-& (b) + (c) =
% AK/K 4
e.
Obtain boron concentration adjustment by dividing (d) by differential. boron worth from Enclosure 5.7.
1
% l\\K/K +
% AK/K 4
Always Negative ppm ppm l
1
)
ENCLOSURE 5.4 PAGE 1 0F 2 Rev. 27 E.6-13 f i 1-I, f, ' () j.
f i
a 1
4
ENCLOSURE 5.4 (Continued)
REQUIRED SHUTDOWN BORON CONCENTRATION OR ACTUAL SHUT 00WN MARGIN CALCULATION 3.
Algebraically subtract (2e) from'(1) to obtain reauired boron concentration for 1% AK/K shutdown margin (assumes worst case stuck rod is out),
ppm a.
RCS boron concentration must be maintained greater than the step 3 value to satisfy Tech Spec shutdown margin requirements for operating with an inoperable rod.
4.
If desired, calculate actual shutdown margin as follows:
a.
Subtract (3) from the actual boron concentration; the result should be positive unless a 1% AK/K shutdown margin has not been established, ppm minus ppm =
. ppm Actual (3)
NOTE:
Shutdown margin is expressed in -% OK/K.
If the actual boron concentration is greater than the required boron concentration of (3), as it should be, the shutdown margin will be more negative than -1.0% AK/K.
b.
Calculate actual shutdown margin by multiplying (4a) times differential boron worth from Enclosure 5.7 and subtracting + 1.0% OK/K.
(
ppm X
% AK/K) - 1.0% SK/K =
% OK/K ppm Always Negative If credit for Xenon is taken, calculation good until:
Time Date Do not exceed time corresponding to Xenon worth used in Step 2.
5.
Forward a copy of the completed procedure to the Plant Nuclear Enginec'.
Calculation Performed By Date Calculation Checked By Date 1
ENCLOSURE 5.4 PAGE 2 0F 2 Rev. 27 B.6-14
ENCLOSURE 5.5 A
SUSCRITICAL SHUTDOWN MARGIN VERIFICATION NO XENON CREDIT DESIRED 1.0 The present RCS boron concentration is ppm.
i L
2.0 If control rod. group 1 is in, then the boron required for present cycle burnup is ppm from Enclosure 5.12.
OR If control rod group 1 is withdrawn, then the boron required for present cycle burnup is opm from Enclosure 5.13.
3.0 If Step 1.0 is greater than or equal to Step 2.0, then adequate shutdown margin exists.
1.0 1 2.0 7 Yes / No (Circle ono) 4.0 If shutdown margin is less than 1% delta K/K, then inform the Shift
^
Supervisor immedi.ately.
'.., e 5.
Forward a copy of the completed procedure to the Plant Nuclear Engineer
/
/
Date Time Performed By
/
/
Date Time Checked By-ENCLOSURE 5.5 PAGE 1 Of 1 Rev. 27 B.6-15 I
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ENCL 5.8 s
XENON WORTH AFTER SHUTDOWN, 532
- F in
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POb ER BEFOR E SHUTDOWN l'
l 100%
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5 10 15 20 25 30 35 40 45 50 55 Sb TIME SINCE TRIP, HOURS This is the Xenon Reactivity due to Iodine Decay After Shutdown from the indicated Power Level.
Equilibrium xenon worth changes over cycle life are neglected in this graph.
This assumption is consistent with the accuracy of the graph.
Rev. 27 Enclocure 5.3 E.6 - 13 Page 1 of 1
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ENCL 5.10 SAMARIUM WORTH AFTER TRIP 532
- F ASSUMES EQUILIBRIUM SAMARIUM EXISTS BEFORE SHUTDOWN m
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This curve also includes plutonium buildup after trip.
Rev. 27.10 B.6 - 20 Page 1 of 1
,0 ENCL
. 5. iia
.s, CONTROL ROD WORTHS, 532* F NO XENON o
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% WD Rev. 27.11 B.6 - 21 Page 1 of 2
ENCL 5.11b s
CONTROL ROD WORTHS, 532* F i
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80 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 ROD INDEX,
% WD Rev. 27.11 B.6 - 22 Page 2 of 2
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'NOI1V81N3DN00 NOB 09 Rev. 27.13 B.6 - 24 Page 1 of 1
ENCL 5.14 l;
INOPERABLE ROD PENALTY NO XENON ROD GROUPS i-7 AT O' 40 f
THE WORTH OF ANY INOPERABLE..JD IS EQUAL TO OR LESS THAN THE VALUE IN THE FIGURE O,
N l
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'O 50 100 150 200 250 300 350
- BURNUP, EFPD Rev. 27.14 B.6 - 25 Page 1 of 1
y
.t' ENCLOSURE 5.15
-SUBCRITICAL MULTIPLICATION WORKSHEET RCS Boron Concentration RCS Temperature Cycle Burnup NOTE:
Avoid calculations of multiplication where an RCS temperature change is. expected.
This calculation does not take into account the effect of coolant density change on
~
the NI detectors.
1.
Case Descriptions f
INITIAL a.
Control Rod Group 1 position
% WD b.
Control Rod Groups 2 + 3 + 4 position
% WD c.
.Countrate NI 1 cps d.
Countrate NI 2 cps 2.
Shutdown Roactivity
~
a.
Calculate Net Shutdown Reactivity from Enclosure 5.17 for current conditions.
Always negative.
R1=
% OK/K b.
Calculate-Net Shutdown Reactivity from Enclosure 5.17 for safety groups 1-4 withdrawn.
Always negative.
R2=
% SK/K 3.
Suberitical Multiplication a.
Calculate the expected suberitical multiplication in changing from initial to final conditions.
R1 (100 + R )
2 M=
R2 (100 + RI)
( 100-
)
M=
( 100-
)
ENCLOSURE 5.15 PAGE 1 0F ?
Rov. 27 l
B.6-26 l
o ENCLOSURE 5.15 (Continued)-
SUBCRITICAL MULTIPLICATION WORKSHEET 4.
Expected Countrate.
a,
' Multiply the initial countrate by M from step 3a to obtain the expected countrate.
X
=
cps NI 1 Countrate 3a X
=
cps NI 2 Countrate=
3a 5.
Initial NI Countrate plus 20%
a.
Multiply ic by 1.2 X 1.2 =
cps ic b.
Multiply Id by 1.2 X 1.2 =
cps Id 6.
Observed Countrate a.
Record the observed NI Countrate after reaching the final conditions.
Observed NI 1 Countrate =
cps Observed NI 2 Countrate :
cps Prepared by Date Reviewod by '
Oate ENCLOSURE 5.15 PAGE 2 OF 2 Rev. 27 B. 6-- 2 7
ENCLOSURE 5.16
\\
CONTROL ROD GROUP WORTH RCS TEMPERATURE = 532*F SAFETY GROUP WORTH, %dK/K 1
- 1.5 2+3+4
- 3.7 1+2+3+4
- 5.2 5+6+7
- 3.5 a
i ENCLOSURE 5.16 PAGE 1 0F 1 Rev. 27 8.6-28
ENCLOSURE 5.17 I
-~
SHUTDOWN REACTIVITY FOR SUBCRrlICAL MULTIPLICATION CALC,LATION U
NOTE:
This calculation does not calculate actual shutdown reactivity of the core.
It is for use only in conjunction with Enclosure 5.15 for estimating changes in NI countrate during approach to criticality.
4 1.
Present Boron Concentration Obtain the present RCS Boron concentration from Chemistry Dept.
a.
ppm b.
Obtain the differential Boron Worth for the present cycle burnup.
from Enclosure 5.7.
Always negative.
%AK/K/ ppm Obtain the Reactivity Worth of the present RCS Boron Concentration c.
by multiplying la by 16.
Always negative.
X
=
%SK/K la ib 2.
Inserted Control Rod Worth
'.s Obtain the worth of the safety banks inserted a.
in the core from Enclosure 5.16.
Group worth is zero if withdrawn.
Always negative or zero.
%dK/K b.
Obtain the worth of the regulating rods inserted in the core from Encisoure 5.16.
Always negative.
%dK/M l-3.
Xenon Worth Obtain the current Xenon' Worth from the off line a.
1 XENON _ Program, or the plant computer, or Enclosure 5.8.
Circle method used.
Always negative or
%aK/K~
zero.
4.
Fuel Worth Obtain the fuel excess reactivity for the present a.
burnup from Enclosure 5.6.
Always positive.
%aK/K ENCLOSURE 5.17 PAGE 1 0F 2 Rev. 27 B.6-29
(Continued)
ENCLOSURE.5.17 SHUTDOWN REACTIVITY FOR SUBCRITICAL MULTIPLICATION CALCULATION t
5.
Net shutdown Reactivity, R.
1 1
%oK/K
]
Algebraically sum Ic + 2a + 2b + 3a + Aa = R = _
a.
~
t Always negative.
4 Date Prepared by Date Reviewed by e
9 ae ENCLOSURE 5.17 PAGE 2 OF 2, END Rev. 27 B.6-30
l
.i '
EQUATION SHEET f = ma v = s/t Cycle efficiency = N Jork out) 2 w = mg a = v,t +
at E = mC a = (vf - v )/t 9
A = AN KE.=
my vf = v, + at A = A,e PE = mgh w = e/t A = In 2/tg = 0.693/tg W = vaP (t )(t )
i x
~
AE = 931Am h
i (tg+t) l l
Q = [nC AT 7
7,-Ex p
Q = UAAT 7
7 ux Pwr = W En g
I=I 10 *
~
o 8
(t)
P=P 10 TVL = 1.3/u ti P=P e HVL = 0.693/u o
SUR = 26.06/T T = 1,44 DT SCR = S/(1 - Keff) fA 0\\
- 1f SUR = 26 CR = S/(1 - K ff )
g, 1( ~ eff}1
- 2(
eff)2 T = '(1*/p ) + [(g_p)/x ff}
~
p T = 1*/ (p - T)
M " I/(1 - Kegf) = CR /CR y
0
~
~#
E eff M = (1 - K
/
eff)0 (1 - K,ff)3 0 " ( eff~l)/K
= AK,gg/Keff eff SDM = (1 - K,ff)/K,gg
[ 1*/TK,'f f J + [B/(1 + A,ff )]
t* = 1 x 10 seconds
~
p=
T P = I4V/(3 x 10 0)
A,gf = 0.1 seconds-1 I = No Id3y=Id22 WATER PARAMETERS Id =1d g
2 1 gal. = 8.345 lbm R/hr = (0.5 CE)/d (meters)
I gal. = 3.78 liters R/hr = 6 CE/d (feet) 1 ft = 7.48 gal.
MISCELLANEOUS CONVERSIONS 3
10 Density = 62.4 lbm/ft 1 Curie = 3.7 x 10 dps 3
Density = 1 gm/cm 1 kg = 2.21 lbm 3
Heat of vaporization = 970 Etu/lbm 1 1,p = 2.54 x 10 BTU /hr 0
Heat of fusica = 144 Btu /lbm 1 Mw '= 3.41 x 10 Btu /hr 1 Atm = 14,7 psi = 29.9 in. I'g.
1 Btu = 778 ft-lbf 1 ft. H O = 0,4335 lbf/in 1 inch = 2.54 cm 2
F = 9/5 C + 32
- C = 5/9 ( F - 32) 1
lu__EBINCIELES_g[_NyCLEQB_EgWEB_[($N1_g[ES$11gN PAGE 30 1
ISEB0991N9dlC@z_MEGI_lE@NSEEE_@ND_E(glD_E(gW ANSWERS -- RANCHO SECO
-87/10/20-COE,D.
ANSWER 1.01
(.75) or C. cr b[.
b.
(0.75)
REFERENCE HTFF,Section III, Part b 193CO6K104 ANSWER 1.02
(.75) d.
(0.75)
REFERENCE HTFF section II, Part B, pgs 129 - 162 193OO7K106 193OO7K10S i
ANSWER 1.03
(.75) c.
(0.75)
REFERENCE Reactor Theory RT16.2 - 16.6 192OO6K110 ANSWER 1.04
(.75) b.
(0.75)
REFERENCE Reactor Theory RT-13.4 - 13.5, Figurec 13.5, 13.6a, 13.6b, 13.0 192OO4K108 l
4
)
i r
lt__ESINCIELgg_gE_gyckgeg_EgWgs_EleNI_gEEBellgN1 PAGE 31 ISEBdgDENed1Cg1_egeI_IgeNSEEB_eND_ELylp_E(gW ANSWERS -- RANCHO SECO
-87/10/20-COE,D.
ANSWER 1.05
(.75)
.d.
(0.75)
REFERENCE 192OOOK121 ANSWER 1.06 L.75) a..
CO.75]
REFERENCE HTFF,Section II, Part A, pgs 64 - 98 193OO3K125 ANSWER 1.07 (1.00) d.
(1.00)
REFERENCE Reactor Theory RT-10, Figure-10.4 192OOK109 ANSWER 1.08 (1.50) a.
(0.5) c.
(0.5) e.
- (0. 5 )
REFERENCE HTFF,Section III, Part B, pgs 319 - 336 192OO4K101 l
1 l
Iz__EBluCle6ES_QE_NgC6E@S_EgWEB_[6@NI_g[EB@IlgN PAGE 32 1
IbESugE18@dlCS1_bESI_IBQNg[E8_@NQ_E691p_E69W ANSWERS -- RANCHO SECO
-87/10/20-COE,D.
ANSWER 1.09 (1.00) b.
(1.0)
REFERENCE Reactor Theory RT-6.3 l
- 192OO2K108 l
ANSWER 1.10 (1.50)
I a..
-Lower (0.75) b.
Higher (0.75) j REFERENCE HTFF section II, Part B, pgs 129 - 162 193OO7K108
'I i
i ANSWER 1.11 (1.50) a.
Increase (0.75) b.
Decrease (0.75)
REFERENCE I
HTFF,Section II, Part C, pgn 243 - 272 193OO8K105 ANSWER 1.12
(.75) b.
(0.75)
REFERENCE Reactor Theory RT-8.5 - 8.11 192OOSK103 192OO8K106 i
j l
1.__ESINQlELES_QE_NgC6E88_EgMEB_EL8NI_QEEB811gN1 FAGE 33 IHEBdggyB901C@1_HE@l_I68N@[EB_@dQ_E6WlE_ELQN ANSWERS -
RANCHO SECO.
-87/10/20-CDE,D.
ANSWER 1.13
(.75) b.
(0.75)
REFERENCE I
RANCHO SECO EXAM BANK REACTOR THEORY RT-F3.
192OO3K108 ANSWER 1.14 (1.50) a.
Decrease.(Lower temperature increases the density thus reducing the. number of neutrons available for capture by the r od s..)
(0.75) b.
Increase. (Decrease boron concentration increasen the..u.ober of neutrons availabl e f or interaction with the rodc.) (0.75)
REFERENCE Reactor Theory RT-14.2 192OO5K107 ANSWER 1.15 (1.50) a.
More negative (0.75) b.
Less negative (0. 75)-
1 REFERENCE Reactor Theory RT-12.4 192OO3K109 1
ANSWER 1.16-(1.50) l a.
Decrease (0.75) 1 b.
Increase (0.75) i I
It__ESINCIE6EQ_QE_NYCLE@S_EgWEB_[6@NI_QCEB@llgM3 PAGE T4 ISEBUQQyy@dlCS _ME@l_IB@NSEE6_@NQ_[6plQ_ELgd z
ANSWERS -- RANCHO SECO
-87/10/20-COE,D.
REFERENCE Reactor Theory RT-7.3 - 7.4 192OO4K109 192OO4K110 e7I ANSWER 1.17 h.s.,
E' pcLETED c--
c- -
b.
Mor e negati ve (0.75)
REFERENCE Reactor Theory RT-12.2 - 12.3 192OO4K106 ANSWER 1.10 (2.00) a.
(1) Increasing towardo peak Xenon concentration.
(0.5) b.
(4) Decreasing toward ero percent power equilibrium value.
(0,5) c.
(3) Decreasing toward a dip.
(1.0)
REFERENCE Reactor Theory RT16.2 -16.6, Figures 16.4,5,6 192OO6K110 ANSWER 1.19 (1.50) a.
TRUE ( A cingle pump running has a higher flow than when all 4 are operating due to reduced discharge pressure so more worI~ is done and more amperage drawn.) (0.5) j b.
FALSE (At col d conditions, fluid density is higher, so more mass i s moved so more worl: is done and more amperage J
drawn. ) (0. 5) 4 c.
TRUE (It must accelerate more mass, which requires more work, and amps.) (0.5)
REFERENCE HTFF section III, Part B, pgs 319 - 336 191004K107 191004K109 191004K112 i
is__ESINCIC6E@_QE_NgC6EQB_EQWEB_C66NI_QEEB@IlOBt PAGE 35 IUES50DYNQUICS _ME@I_lB@NEEE6_@ND_E6MID_ELQW t
ANSWERS -- RANCHO SECO
-07/10/20-COE,D.
ANSWER 1.20
(.50)
True (0.5)
REFERENCE HTFF,Section II, Part C,
pgs 243 - 272 193010K106 ANSWER 1.21
(.50)
Fal se (0.5)
REFERENCE Reactor Theory RT-13.3 192OO4K107 ANSWER 1.22 (1.00)
The heat sink must be at a higher elevation than the heat source.
(1.0) 1 REFERENCE HTFF,Section III, Part B, pgs 351 - 360 OO2OOOK513 OO2OOOK516 ANSWER 1.23 (1.00)
(During nat. circ. cooldown, bypass flow i s minimal.
Therefore, the water in the upper head is decoupled from the cooldown of the RCS.) The water (& metal) in the upper head mcy stay warm enough to flash to steam when the RCS is depressurized, even though the bulk temperatures in the RCS are cool.
(1.0)
L'on win l ateert chu uss,'m a4 locahen a f <>s-h-wh (*se A 4 AW'M MM )
_l REFERENEL
/g.
g gj (Q/
d HTFF,Section III, Part D, pgs 351 - 360 j
193OOBK120 l
1 1
i i
2___PL@dl_QE@l@@_1NCLUDIN@_@@[EI!_6ND_EMEB@@NCY_@Y@l[ME PAGE 36 ANDWERS -- RANCHO SECO
-07/10/20-COE,D.
ANSWER 2.01 (3.00) a.
1
.(SID'was the "old" manual power suppl y f or NNI)
EO.6]
b.
1 (Provides alternate power to the NORMAL NNI/ICS cupply bus S1GB-1, which is maintai ned)
[0.6]
f c.
3erl(power source switches uninterrupted to battery GB)
E0.6]
d.
2 (trips Si and S2 per ECN R-0826)
[0.6]
Oe. 3 ort (Static switch auto transf ers to Xf rmr X31GB1) EO.6]
skl[Y, />
REFERENCE Learn Obj 2,5 OD 24 D 1600 SFAsttAmes cwa-s p y M
~!
Id f
9 7 an F( 082 EC STM Chap 33 pp 33-34 4 glq g g {
016000A202 ANSWER 2.02 (3.50) a.
(vopor) to RB atmosphere -> RD emergency cooling units -> NECW ->
NSRW -> Nuclear Service Spray Pond CO.25 each]
(spill also coessidet-hosphere -> RD spray -> RD sump {o[fy / gh3 ceded &% g ff b.
,, g 8 NSRW -> Nucl ear Service Spray Pond EO. g each]
Oo S REFERENCE Learn Obj. 1,2 OD 21 I 3201 Learn Obj. 9,10 DD 21 I 3205-Sys Trng Man Chapt 27,28,30,31 OOOO26K302 OO603OK403 l
_g _ _.
?
.((
q L___ELONI_DEglGN_lNC6gDlNG _g@ Eely _6ND_EMEgggNCX_SYSlEt!g -
-PAGE 37 ANSWERS -- RANCHO SECO
-87/10/20-COE,D.
't
-v
'i I
\\' 1 ANSWER-2.03 (3.00)'
t
.1) S --
cn 2).5 i3) O
- 4) O ~(throttled)
.5) N.(L.O.)
6)'N gy
- 7) O
~
'Os,
- 8) N
- e
- 9) N
- 10) N EO.3 occh3
(
b, p(
REFERENCE t-Learn Obj 2 DD 21 I 3201
(
Learn Obj 8 OD 21 I 3204 Sys.Trng Man Chapt 27 pp 6-21 OO603OU404 t,
\\
/, lf s
, \\p 1
T ANSWER 2.04
- 6. lor /, O c
^"i EO 2T' DRETED C~
^5'
- . 25 g
b.
A - for baron dilution post LOCA [0.53
/i B -'for piggyback recirc mode post LOCA CO,33 p.
\\
REFERENCE Learn Obj 9 OD 21 I 3205 Learn Obj 13 OD 21 I 3207 Sys Trng Man Chapt 27 pp 6 -- 21
!(.
f, Dwg M-522 sheet 2 OO6000K406 OO6020K403
'N-s r-
\\ }b i:
ry 2 t._
P {= A N T_,,,D E_ @ l @ N_ _ { N C L U D I N G S A F E T Y A N D E M E R G E N C Y S Y S T E M S PAhE 38 y
L
.ANSyERG -- RANCHO SECO-p
-87/10/20-COE,D.
lg
,s S
\\
e
\\
h
- t'
( 2. 50 ) ('s ' )s ANSWER 2.05 a.
from Plant, Service Water System (through a 1" line CO.53 w
(
)
b.
1 Motor driven IO.53 and nM 1 Di esel dri ven 't.4 re pump EO.53 i
?
[ ?..
.L D A r T [0. 2 5 ) $ Al b.f) g R,..e & f g c & (A / r & $ & [0,z 5'}
)
c.
Cardou (C02) d.
(dry pipe */ sprinkler system C?.5j J.
REFERENCE T
Learn Otaj 9. OD 21 I 5500
<h i
Learn Obj 9 OD 24 D 1700 086000K401/K403iK406
\\)'h
/
'g g
w.m 4
/
N c
t v x
(
/ '
ANSWER 2.06 (2.00)5 L-f, l'
Upper Radial Bearinghe* f h ib b bC'
^
okftYler )
f'y. b 2f C001ing 3achat (o,~ (sce>C In+tarrd k
3.
Motor air cooler
'I s
,f 4.
Upper Motor Bea-ing lube oil cool er RC pump cooler (or ful eyec.feh dec/fr ) (ar.y 4 at 0.5 stach)
!O.
\\
REFERENCE v
S.7) '( pg 41.
i.
t i OdTfoO:uO2 C
- S l' 2. y 2-/ca
{'
i.
I y
~'t b
I ANSWER 2.07 v (2 v0)
- t
(
p'
, '.. i Only 30 linnutes of air pressure is availiaale if the entire pl ant
.ai r system 1,9 in use. (discount,s accumulat(fors andseackup nitregen
- g uupply.)
5 (1.Q'
.a OR **LOLSof afr~ 4 s C s N C o l t/tt JtE 11 Cs'H!'f.5f C
't' l'f3N 6
1
~f8 f REFERENCE
.)
SD (13a pg B5 (Am e%
t ckj la n f % $ s'e n h.f phe 3 o r~
fIf'!'" It 'I C,
f4 *f~
l
^
t 3
.s
.s
\\
6 n;
i li l.>
1 i
..?
'i e
.1.1 r'
'; [ : /
r!
21,__[(@yI_Qgg[@N_lyg6gD1N@_@@EEIY_@dQ_Edg6GENGY_"Y@Igdg PAGE 39 ANSWERS -- RANCHD $ECO
-07/10/20-CDE,D.
(
1 ANSWER 2.08 (2.00)
\\
a) 4 ( P +2 i OD Di s's.h '>
(.y5) b)
3 (P-210A:Disch)
( 0,, 5 )
c) 1 (P-210D Suct.i on) 40i S) d)
7 (Bt Hot Leg) 10.5)
- 4;,*.***********.
i' PEFERENCE i'
F and D's M520 sheet 2
- OO2OOOK106 OO2OOOK108 OO2OOOK109
}
~
- p. 7, (1.00)
ANFWER 2.07 The I.i.'S i s redundant to. the Reactor Building Cooling units.
REFERENCE SD 2.9 pg 12.
0260QOK301 9
ANSWER 2.10 (1.00)
St art-ai r is neece.1 to supply moti.ve f orce to r! rr:
t'
.-tm.
shutoif fuel to the engine, u-t ' _.'
-c'.
" +^
REFERENCE OD 24 D 1700 0640DOK603 064000A306 ANSWER 2.11
(.75) b.
(0.753.-
REFERENCE SD 43, pg 17, DD 21 1.6500 063OO ')A101 063OOOA403
)
__,._____._J
-_l Er__EL9NI_DEEl@d_lNC69 DING _@@ Eely _@ND_,EDEBGEUCf_@X@lEdC PAGE 40 ANSWERS -- RANCHO SECO
-87/10/20-COE,D.
ANSWER 2.12
(.75) d.
(0.75)
REFERENCE SD 14, pg 32, OD 21 1 0300 035010K107 035010K503 035010A101 ANSWER 2.13
(.75) a.
[.753 REFERENCE SD 39 pg 23.
045010K423 ANSWER 2.14
(.75) d.
[.753 REFERENCE OO2OOOK108 ANSWER 2.15
(.75) d.
[0.753 l-l l
l ANSWER 2.16
(.75) l-OR d.
a.
[0.753 REFERENCE SD 37a pgs 1 - 18.
OO1000K103 i
PAGE 41 5 __INgIByMENIg_AND_CONISO60 t
ANSWERS -- RANCHO SECO
-87/10/20-COE,D.
-ANSWER 3.01
'i.5:' /,0 a.
~
lMed(oe w+nH:9) to,.s-3
~
'r:1?
- ._c'
.Jc c J,. s m. m. -
- c.. _,,' A,.
- ;.. ;., - OCLEMO L.
9,
_ m. v..-
c.
(control valve f ails open) Au::111ary steam pressuro goes to ma::imum (possibly relief setpoint)
E0.53 REFERENCE ECN R-0878 Rev 1
', OD 24 D 1600 Learn Obj.
4 039000A402 (2.1 importance, but due to licensee ECN and training emphasis and protection from overcooling on loss of ICS power, is
' included) 000065A208 (c 47/ cgg Qg,7s org gg9}
ANSWER 3.02 (2.00) v rods may
~N+ OT' be controll ed in manual r EO.53 4
a.
Control
- '. 20: ec.M I a /oS E MY /* /
b.
MFPs ~
- ~= " " "' r'tr.
E - _ c ;- n _' :.
- c. TBVs may be switched to alternate power E0.25] and controlled in (tSu(p,5}
manual only EO.253
- d. ADVs continue to control in auto or manual (or "no effect")
[0.51 REFERENCE l
Learn Obj 6 OD 24 D 2500 Learn Obj 1,3 OD 24 D 1600 ECN R-OB23, R-0861 041020K401/K603 (No good KAs for ICS as a system)
ANSWER 3.03 (2.00) a.
1.
Main condensor low vacuum EO.5]
2.
Main condensor high differential vacuum EO.5]
3.
Main condensor circulating water pumps not running (breaker position)
[0.53 b.
Both EO.53 REFERENCE STM Chap 32 pg 74 OD 24 C 0701 ECN R-0826 Rev i pg 5 of 15 Answer b.
based on phonecon 9/15/87 D.
Coe and T.
Hunter 04102OK401/402 (importance 2. 9/2. 3 but included based on recent
A _ilNSIB!LMENIS_8lJD_CgNIggbS' PAGE 42 ANSWERS -- RANCHO SECO
-87/10/20-COE,D.
'i
' modifications)
ANSWER 3.04 (1.50)
'a. ALL MFW valves are shut' E0.53
[0.253-
~B OTSG AFW control ^ valves are shut EO.253 block valves shut
[0.253 REFERENCE Learn Obj 3, Table 2,'OD 24-D 3200 (i ni ti ati on stpts)
Learn Obj.6c, OD 24 D 3200, Encl. 4 pg 5 (Vector logic)
OOOO40A102/A110/A111
' ANSWER 3.05 (2.00)
- 1) ADV CONTROL for B OTSG (all 3 ADV's will open)
[0.673
E0.663 3)- B OTSG rate of level increase CONTROL E0.663 m pre.uur<. es,-pena + ton he a orsa /evet.aj eut [s,sQ p
REFERENCE gy ff,.cq goo Learn Obj'Sc, 6b, 8,
and 9 DD 24 D 3200 l
CD 24 D 3200.II and IId (rate centrol)
/
IIc.3 and Fig. 22 (MFI)
IIe (ADV) 059000K102 061000K102/A101
- 016000K106 o
i
It__INSIByMENIS_AND_CONISOb@
PAGC 43 ANSWERS -- RANCHO SECO
-87/10/20-COE,D.
ANSWER 3.06 (3.50) a.
F (overfill condition only trips MAIN FW valves)
[0.7]
b.
T E0.73 c.
?
(Depends on state of ICS control signal)
[0.7]
d.
F (an MSI trip seals in the MFI trip modules)
EO.73 c.
?
(only if the Control Enable Trip Module has been reset)
[0.73 REFERENCE a.
Learn Obj 5c, OD 24 D 3200 para II.c.3 059000A306/A412 b.
Learn Obj 10a (cl osest ), OD 24 D 3200 II.c.1 061000K402 (auto start, no KA for manual start) c.
Learn Obj 10c, OD 24 D 3200 II 059000A410/411 d.
Learn Obj 10b OD 24 D 3200 II.c.3 059000K419/A411 e.
Learn Obj 10b,c OD 24 D 3200 III.a.2 061000K411 ANSWER 3,07 (1.00)
The bypass for the EFIC channel Will be automatically removed.
E1.03 REFERENCE Learn Obj. 12 OD 24 D 3200 III.a.3.2 No KA relating RPS to AFW(EFIC) 012OOOA203 (incorrect RPE channel bypcssing?)
ANSWER 3.08
-( 2. 50 )
a.
Diverts release to preselected retention basin EO.5]
- b. Stops discharge pump and closes discharge valve EO.52 c.
State of CA notification U. C osit AJoN F (0,5)
- d. Stops Reactor Bldg. exhaust and supply fans and shuts equalizing block valve EO.5]
e.
none EO.53 REFERENCE Learn Obj 1,
OD 24 D 4400 pp 2, 4-6 Learn Obj 8, OD 21 I 6104 073OOOK401
3.__INSIByt][NI@_@ND_ CON 16CL@
PAGE 44 2
ANSWERS -- RANCHO SECO
-87/10/20-CCE,D.
ANSWER 3.09 (1.00) a.
To achieve cold shutdown conditions. [0.5]
b.
yes
[0.53 REFERENCE Learn Obj 1,2 OD 24 D 1600 pp 7-G also see Rancho RO Exam Bank question NNI-15 016000A402 ANSWER 3.10 (2.00) a.
Lube oil Low Pressure
[0.5]
or Engine Over d
[0.53 d^
nW %& c0 feu & [o S) r 6.
Ground fault
[0.5]
or Engine
[0.53 4.n q h or W 'Qverspeeg[i c u rt-C h f {0rS 1
(
tv r An*+
(generator di H current trip highly unlikely due to unloaded condition)
REFERENCE Learn Obj 7 OD 24 D 1700 pg 19 STM Chap 45 pg 119 also see Rancho RO Exam Bank quest EDG-1 064000K402
)
1
's __INgIBUdENIS_BND,CgNIBOL@
PAGE 45 ANSWERS -- RANCHO SECO
-87/10/20-COE,D.
i l
ANSWER 3.11 (3.50)
]
a.
- 1) NONE (No reactor trip if < 55/ power)
LO.5]
- 2) Runback to 45% reactor power CO.53
- 3) NONE (requires ALL RCPs to trip bef ore responding)
[0.5]
b.
RPS (flux / delta flux / flow)
CO.5]
ICS (BTU limit calc and FW ratioing)
EO.5]
c RCP seaJc (or MU and Purif.)
CO.53 CCW
[0.53 REFERENCE 1
Learn Obj 9 OD 21 I 5000, 5004 pg 4
)
Learn Obj 9 OD 21 I 4900 STM Chap 2 pg 92, Chap 32.II.C ECN R-0825 DBR pp 7-9 Learn Obj 4 OD 23 K 0500, 0502 pg 5 Learn Obj 8 OD 21 I 1000, 1003 pg 5 STM Chap 7 pg 37
{
Learn Obj 8 OD 21 I 2500, 207 pp4-9 GTM Chap 5 pp 93-97 OO3OOOK303/K304/K305/K103/K112 ANSWER 3.12 (2.50) a.
NO CO.5]
b.
NI-5 AND NI-6 [0,5] both < 10% power
[0.5]
c.
Either SR (NI-1 or 2) > 2 dpm
[0.5] OR Ei+" ~ IR (NI % -se 4) > 3 dpm CO.5]
REFERENCE Learn Obj B OD 21 I 4600 STM Chap 34b pp 90-92 Tech Spec 3.5 also see Rancho RO exam Bank question NI-6 015000k103/K402/K604 l
l l
4:__ESQQEDyEE@_ _UQBd@6t_@@NQBd@kz_EDEBgEdgy_AND PAGE 46 EBD19600lGB6_G9 NIB 96 ANSWERS -- RANCHO SECO
-87/10/20-COE,D.
' ANSWER 4.01 (2.50) a.
False (SCM is. higher priority than SGTR) b.
False (Although listed after SGTR, ICC is the more severe case of Loss of SCM and must be treated accordingl y) c.
True d.
False (hold at that point, stabilire the plant, check for other problems) e.
False (E.02 should be conducted in parallel with E.04)
LO.5 each]
REFERENCE Learn. Obj. 2a,b,c,e OD 24 B 4300 OD 21 J 4900 pg 6 OOOO40G012 OOOOO9G012 0000746012 OOOO38G012 OOOO40K304 OOOO38K306 OOOOO9K321 ANSWER 4.02 (2.00)
Loss of SCM Loss of heat transfer Excessive heat transfer SGTR
[0.5 each]
REFERENCE Learn. Obj. 1b, 6 (cl osest )
OD 24 B 4300 EOP E.01 pg 2 OOOO40G012 OOOOO9G012 000038G012 OOOO74G012 l
)
4.
' PROCEDURES - NORMAL _ABNgRMAL _EME;RGEN_gY_AND PAGE 47 1
1 889196991CeL_C9t!IB96 ANSWERS - RANCHO SECO
~B7/10/20-COE,D.
1 ANSWER.
4.03 (3.00) a.
(E.01) (2.1) Manually trip the reactor
[0.53 (2.1.1) If.the reactor has NOT tripped, then immediately go to E.02, Step 1 CO.53 b.
1.
Deenergine the CRDM's CO.53 2.
Drive rods in
[0.53
- 3. Borate (8w6T -/o HP.r)
CO.5]
- k. OordYC ($A f g ggy}
[g,g]
g" gppg*fg,(gAf g gp,y) [0,f]
]
c.
No
[0.53 b^7 3; /<IN)
REFERENCE E.01,.E.02 No Learn. Obj. found OOOO2VG012
-OOOO29K310/311/312 OOOO29A111 ANSWER
-4.04 (1.50)
RCPb Are fe[+ t~unnfn 3 c.' C c-m 24
[O.63 a.
E4 r*
b.
rt CP'_C^
E' 33 morereactorcoolant'liquidmassgculd be pumped out the break than if the pumps were stopped
.I.G.+5 causing insufficient liquid in the vessel to cool the core (ICC).
J.Dc37",$f REFERENCE E.03 OD 21 J 5100 pp. 2,3 Rancho Seco ATOG (B8<W 76-1127470-00 DRAFT;' pp. 89-93 Learn. Obj. 10a OD 24 B 4300 OOOOO9K323
ds__EB99E99 BEE _I_N98d96t 6Ed960661_EdEB@ENQy_@NQ PAGE 48 B09196991906_99 NIB 96 ANSWERS -- RANCHO SECO
-87/10/20-COE,D.
ANSWER 4.05 (2.50) a.
1 b.
1 c-2 DELETED
~
d.
3 T
pE LETE D EO.5 each]
REFERENCE C.8 Learn. Obj. 2 OD 21 J 1100 OOOO15K303 ANSWER 4.06 (1.50) 1.
Pressuri:er level 160 inches and decreasing Co.5]
2.
MUT level less than or equal to 18 inches EO.5J 3.
RCS subcooling less than VSM E0.5]
REFERENCE C.3 Learn. Obj. 2 OD 21 J 0901 OOOOO9G010 ANSWER 4.07 (3.50) 4 uru 3-+
t, getu mvgn ;
^
r e, n,
g e,.?
7 y
,m er e,
qit n+--
, -- L7,
m, L:.g i
. r ; - _ ; ;, 3.
, _7 g s
.s
- t,., - ;,, m g..-
77v r e, s
uw
-,s 1
r ut es s en M7 p7,_
r_' 7 r-. '
- 4i,__ gTy i.
- r,
- r.,
7, eg,
rj
,_g g_.
- g T.
M'J T : : :-:
O i.. ;.?, c ;
P.12 f
?
~
e-.r, r i it a s r es, q 3.
ne 3_
.; ; t o g,- - r-
-- _ a - <
m_, -
g _- r$7,g e r. _-
gr.y g n33 re -
7_
- r., r.
,3
. _7 h, Q hFg) las't do BN$4s'- OTS $- {Oo 75h doc'c~~;~;
C r'. ^ 2
- 2) jf" crn SLA l4Jt l{or-W/C-f.esu-R 4 475 REFERENCE 6, ;} ggj y EFl C. h t) h h*f h 95*f on C.10
/
Learn. Obj. 2,3 OD 21 J 1300 OOOO54G010 YD OL e z a <. / 6o" fo, r3 c, 1) 3a1 2 ro.53 Enc I,qL wqt le 9C% en n
s a-aw
4a__PBQGEDUBE@_ _Ng6d66t_8999Bd@62_EdEEgENgy_AND PAGE 49 BOD 1969 GIG 66_G9N1BQ6 ANSWERS -- RANCHO SECO
-87/10/20-COE,D.
ANSWER 4.08 t-2. ^ ^ ' 2 f f
a.
100 mrem /wk ist line supervicor CO.53 b.
1000 mrem /qtr Chairman of PRC CO.5]
c.
300 mrem CO.5]
d.
7 days EO.53 AND if conditions change EO.53 (based on survey results) 1 ;- c:
U.3: DEL.F TE b e
REFERENCE Learn. Obj. 16, 17 OD 21 F 0406 Quest RAD-10, quest bank AP305-1 pp 5, 7 AP305-4 pg 11 AP305-7 pg 3 194000K103/104 ANSWER 4.09 (3.50) 1.
798 CO.53 2a.
-3.25 [0.53 b.
O
[0.23 c.
O
[0.23 d.
-3.25 CO 13 e.
+379 CO.53 3.
+419 E0.53 4a.
+281 CO.5J b.
-3.4 C0,53 Red,$m fc_
k w
- 3 a esvo REFERENCE Learn Obj. 4 OD 21 J 0401 B.6 OO4000A402
ds__EBOGEQQBE@_ _NgBd@61_@@NOBd@61_EdEB@ Edgy _@ND PAGE 50 680196991C06_C9 BIS 96 ANSWERS -- RANCHO SECD
-87/10/20-COE,D.
ANSWER 4.10 (1.00) a.
Limits reactor power to that which can be handled by natural circulation (f or physics testing)
EO.53.
b.
YES EO.53 REFERENCE B.2 pg. 10 Learn. Obj. 3 OD 24 D 3400 Rancho Seco RO Exam Banh question B-29 (modified)
Tech Spec 2.3.1 and Table 2.3-1 012OOOG010 ANSWER 4.11 (1.00) 1.
(f rom 150 F to 230 F) OTSG 1evels maintained at 85% of operating range.
[0.53 2.
(from 180 F to 250 F) OTSG's are kept under a vacuum using the main condensor to promote condensation on the OTSG shell.
[0.53 3.
Sc o f av s+*~
% m M PtN per A'+7 % % 4.05)
REFEl<ENCE B.2.3.1 pg 2, B.2.4.2 pg10 g/
gN A.6.4.2 pg 7 OD 21 J 0000 Rancho Seco RO Exam Bank question B-15 (modified) 035010G010 No Learn. Obj. found
5.4 AT POWER /SUBCRITICAL SHUTDOWN MARGIN CALCULATION Calculation Effective for:
Date 20/ 7Fime /20'O with the following reactor conditions:
Power O
%FP' f(CS temp fE7 *F Burnup /2 O EFPD Gp i rod position /oD9544/A Inoperable rod:
Gp #
Rod #
79N ppm 65]
1.
Obtain reference shutdown boron concentration from:
h.12 (No Xe) Gp 1-7 @ 0% WD, Gp 8 @ 25% WD, or Enclosure 5.13 (No Xe) Gp 1 @ 100% WD, Gp 8 25% WD,
$ g~ -7f 0 +X _
for the applicable conditions of burnup and RCS g
531-M -7Po+//l5 temperature.
For shutdown margins at power, use p 799.12 for a temperature of 532*F.
Circle one used.
2.
Adjust for non-reference conditions as follows:
a.
Obtain appropriate Xenon worth from Bailey computer group 32 (Reactivity Balance), or.8, or off-line Xe-worth calculation program.
Always negative.
Use'0 for conservatism t
if desired.
Circle method used.
~~2'
% OK/K
.)
O' b.
Obtain Samarium worth from Enclosure 5.10.
%SX/KhI).
Always negative.
O c.
If operating with an inoperable rod or shutdown with a stuck rod, obtain Inoperable Rod Penalty-from Enclosure 5.14.
Always positive.
% 6K/K h2 NOTE:
The sum in step d below will always be negative unless an Inoperable Rod Penalty greater than the sum of Xenon and Samarium worth is being used.
-3, tr % nK/K[/]
d.
Algebraically sum (a) & (b) + (c) =
e.
Obtain boron concentration adjustment by dividing (d) by differential boron worth from Enclosure 5.7.
- 3,2f% OK/K +
' M OE7
% SK/K E37' ppm {[)
+
Always Negative ppm ENCLOSURE 5.4 PAGE 1 Cr 2 b'
Rev. 27 B.6-13
ENCLOSURE 5.4 (Continued)
REQUIRED SHUTDOWN BORON CONCENTRATION OR ACTUAL SHUTDOWN MARGIN CALCULATION 69
(
3.
Algebraically subtract (2e) from (1) to obtain required 79$
j 4
boron concentration for 1% OK/K shutdown margin 3 79
-i (assumes worst case stuck rod is out).
4/9 ppm a.
RCS boron concentration must be maintained greater than the step 3 value to satisfy Tech Spec shutdown margin requirements for operating with an inoperable rod.
' 4 If desired, calculate actual shutdown margin as f
follows:
Subtract (3) from the actual boron concentration; a.
the result should be positive unless a 1% OK/K shutdown margin has riot been established.
7 00 ppm minus MIT ppm = 2 & / ppm h4f Actual (3)
NOTE:
Shutdown margin is expressed in -% OK/K.
If the actual boron concentration is greater than the required boron concentration of (3), as it should be, the shutdown margin will be more negative than -1.0% OK/K.
b.
Calcult.te actual shutdown margin by multiplying (4a) times differential boron worth from Enclosure 5.7 and subtracting + 1.0% OK/K.
( 281 ppm X,00857 % OK/K) - 1.0% OK/K = 3.h%OK/K hh]
~
ppm Always Negative If credit fo Xenon is taken, calculation good until: Time MOD Date
/0/2b A7.
Do not exceed time corresponding to Xenon worth used in Step 2. '
5.
Forward a copy of the completed procedure to the Plant Nuclear Engineer.
Date Calculation Performed By Date Calculation Checked By ENCLOSURE 5.4 PAGE 2 0F 2 Rev. 27 8.6-14
7.--,._.
i
-g.
ATTACHMENT 3 SCENARIO EVENTS L /vr scenario No. 3 kroup:
1 si.oiation racin ty:
Examiners:
b o 7/A 6 &
Candidates:
)
i Initial Conditions:
2 ere (3e e Y f['/
i EVENT i
NUMBER TIME MALF.~ N0.
DESCRIPTION
/
l l
l P B 4 / o, 8 d.,h - r 1l l
l4 K L A C A (L % :D j
i 3
l l
l M/h. E L & sio W V
~
l l
l G, /~
j
'l l
l lars L6 Lh n A i
- aco -96 l
1 j
i J
ES-302-4 ATTACHMENT 4 OPERATOR ACTIONS Scenario No.
Event No.
/
! of Page QX CW h[eaf,Tucrcb 46 Q W Brief
Description:
i Time Position Candidate Actions / Behavior I
0 fot<J YewG CVGQ9Cpatu-<h
/
17c H <-v la v(, a 8'
Yvea / i, m
. E'
(
Ik % w --d.L I, L v,,L t
,Mb%
g engaggange --
M 8, 9 m
ms O
D e
l ES-302-4 l
l ATTACHMENT 4 OPERATOR ACTIONS Scenario No.
Event No.
Page l of Brief
Description:
( /) [b
/ k h,
,/
Time Position Candidate Actions / Behavior l
W(4')
l l c,w revan p--,
a
~
A os n
T < X h 4'e 'eec4v-n C a'? E, i
r i
M NM e
O
_______m._---------
./
ES-302-4 ATTACHMENT 4-OPERATOR ACTIONS i
Scenario No.
Event No. 3 Page 3 of
[ h ' [ A !C [ c u e [ v % g, g Brief
Description:
/
, 3 Time Position Candidate Actions / Behavior i
l D
0cr N
JW5e n
e n,
n n L c,c d r L,,i/ d a
v 84 N
1 h
1
./
ES-302-4 ATTACHMENT 4 OPERATOR ACTIONS Scenario No. _3 Event No.
Page of Brief
Description:
OGC Tu [e !a!q
)
u e 3'c -7Cfjg-Time Position Candidate Actions / Behavior E
u,,
(jc 7k% g D
iha~<
e.
as?
u T/n /~~4.
I un s
/, ~, A cs, /m. A e
.f y
$000 Y
G 51 b U*^
hC Y
Detsu a ex dua<La L
G k Ecp
, o e Ju~s y
i
/
leda a d L G S,.
~.
M.
O e
ATTACHMENT 3 SCENARIO EVENTS Simulation Facility:
) on/
A Scenario No.
Group:
/
Examiners: ' b o /A S in--
Candidates:
Initial Conditions:
8 p/tvr r>voccccb 6/O
{
J EVENT NUMBER TIME MALF. NO.
DESCRIPTION
/
l l
lDcp s / C,l _
l l
l i., m L,
2l l
l MLcH bue/XXK/s /d 3
l l
l %d we IJI /ww cLJn l
-l l C/s 349o4 6'
l l
l Lom docca %-
~~
l l
l D ; 4 2 4 4 /o s / ~, K fl l
l AIDU sL k w y
ES-302-4 ATTACHMENT 4 OPERATOR ACTIONS Scenario No.
Event No.
Page of
/
~
2cP see /G, d~~
Brief
Description:
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