ML20215C908
| ML20215C908 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 12/09/1986 |
| From: | Burdick T, Lanksbury R, Mcghee J, Mary Spencer NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20215C906 | List: |
| References | |
| 50-373-OL-86-02, 50-373-OL-86-2, NUDOCS 8612150499 | |
| Download: ML20215C908 (68) | |
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U.S. NUCLEAR REGULATORY COMMISSION REGION III Report No. 50-373/0L-86-02 Docket Nos. 50-373; 50-374 Licenses No. NPF-11; No. NPF-18 Licensee: Comonwealth Edison Company Post Office Box 767 Chicago, IL 60690 Facility Name: LaSalle County Station Examination Administered At: LaSalle County Station Exanination Conducted: October 27-30, 1986
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Approved By:
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. M. Burdic , Chief /MAh i .
Operating Licensing Section Dater '
Examination Summary Examination Administered on October 27-30 1986(_R,eportNo.50-373 Written, oral','iiTsiiiuYator eTaiiiliIatTons,were a,dmin~istereT ~
to'Yi'v/"0L-86-0 e sinliir~
reactor operator (SR0) candidates. In addition, a retake examination, consisting of one section, was administered to one SRO candidate and an SR0 level examination was administered to a candidate for a reactor operator license and an instructor certification.
Results: All candidates passed the written and oral examination and four SR0 candidates passed the simulator examination.
8612150499 861210 PDR ADOCK 05000373 V PDR
DETAILS
- 1. Examiners
- R. D. Lanksbury, Region III M. Spencer, EG&G J. McGhee, EG8G
- Chief Examiner
- 2. Examination Review Meeting Copies of the examination and answer keys were given to the facility personnel for review at the conclusion of the written examination.
Facility personnel provided their consnents to the examiners on October 30, 1986. Their comments as well on this resolution are enclosed as Attachment I to the report.
- 3. Exit Meeting On October 30, 1986, an exit meeting was held. The following personnel were present at this meeting:
CEC 0: G. J. Diederich, Plant Superintendent R. Armitage, Instructor S. Harmon, Instructor NRC: R. D. Lanksbury, Chief Examiner M. Spencer, Examiner S. McGhee, Examiner M. Jordan, Senior Resident I:1spector At the exit meeting the examiners provided the facility with the following candidate generic weaknesses or shortcomings:
- a. Candidate knowledge of local operation of the emergency diesel generator was generally considered deficient.
- b. Candidates were slow to go to and often reluctant to use Alarm Response Procedures during the simulator portion of the examination.
- c. Candidates appeared surprised when asked questions concerning tasks normally performed by the Shift Engineer. Though most always able to respond correctly, it was reconnended that candidates (SRO) be infonted as part of their training that they will be examined at that level.
In addition, the following topics were discussed:
- a. Plant training material supplied to the examiners for examination preparation should be more inclusive. For example, the Reactor
Physics Section should include material taught in the Production Training (Corporate) phase in addition to the site review material provided.
b.. Material provided to the examiners for the LaSalle simulator was
< also considered inadequate. The simulator capabilities and responses were not spelled out well and malfunction gaps made effective use of the simulator difficult.
Attachment:
- LaSalle SR0 Examination Facility Comments and.
Resolution i
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~1 ATTACHMENT LaSALLE SENIOR REACTOR OPERATOR EXAMINATION
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OCTOBER 27, 1986 5.01 Facility Comment:
The facility feels that answers which state " maintaining SRMs onscale",
or " allow for safe Reactor Startup" should be allowed for full credit.
Should not have to state that counts are maintained above Tech Spec Requirements.
Resolution:
The answer key is revised to allow credit for a statement such as
. . . maintained above minimum required" for ". . . maintained above minimum necessary to insure safe startup." The term "on scale" is not by itself an adequate response.
5.04 Facility Comment: .
To be completely accurate for this question, the pump pressure should be given in P.S.I.D. instead of P.S.I.G. Generic comment, no action necessary.
Resolution:
Comment noted as valid, and question answer revised accordingly.
5.05 Facility Comnent:
Although this is included in the LaSalle lesson plan, it should be noted that this information is not a Terminal Performance Objective and LaSalle does not use peaking factors. Generic comment, no action necessary.
Resolution:
Comment noted.
5.08 Facility Connent:
The original question was poorly worded, however, the pen and ink change seemed to clarify the issue. No action necessary.
Resolution:
During the examination the following statement was added to the question:
[ HINT: ECCS systems are designed to be able to provide these three methods of removing core heat].
5.09a Facility Comment:
The answer to part a is an obvious typing mistake and is in error. The moderator temperature coefficient becomes more negative as the fuel temperature increases. Leave the question, change the answer key to part a to read "more negative." Reference attached page 122 of theory lesson plan. (Same' reference as answer key).
Resolution:
Valid comment and answer key revised.
5.11a Facility Consent:
Should accept, for full credit, number 4 or 8. Answer key only specifies number 8. If FLPD is within limits (<1), then since:
= HGR FLPD LHGR Limit the 1% plastic strain limit will not be reached. (i.e. if LHGR is less than the limit, the FLPD will be <1).
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3 Resolution: !
Comment is valid and answer key revised to accept either response.
5.13a Facility Casment:
Should not be required to include all details of the answer key for full credit. The key factors of this answer should identify that deep rods will affect radial flux and are considered " power" rods.
Should not have to mention the coupling of cells, or increased migration length or discuss voiding in the upper region for full credit.
Resolution:
The question required the candidate to discuss why the core responds as it does as well as how it responds. In order to receive full credit, the candidate must include voiding, migration length and cell coupling to answer the question and demonstrate his understanding of the process at work. The keyword answers " power rods increase core power" or " power rods affect radial flux" are not sufficient for full credit. The answer key is revised to indicate the breakdown of the answers which are required for credit.
5.13b Facility Comment:
Should not be required to include all details of the answer key for full credit. The key factors of this answer should identify that shallow rods affect axial flux shape without much affect on power.
The " reverse power" discussion is definitely not appropriate as part of the answer key as this " concept" was already tested under question 5.10. The reverse power affect has very low probability of happening 3
in large commercial reactors and an even lower probability of being observed. Delete reverse power discussion from the answer key.
Resolution:
The question does require a why response to support how the core responds to shallow rod withdrawal and both the positive and negative effects must be discussed to indicate a complete understanding of the processes involved. The discussion of the competing affects explains the " shaping" process and is a necessary portion of the response. The words " power'may decrease" will be deleted from the exam.
5.14 Facility Comment:
General comment for future use of this question. Should possibly use terminology of "overall cycle efficiency" since in the case of loss of feedwater heaters, Rx power would increase causing the generator output to increase. This would cause some confusion on the part of the examinee. No action necessary.
Resolution:
Terminology will be changed to overall cycle efficiency for future reference.
5.15b racility Consent:
Answer key should not specify " transients which cause rapid core flow and power increases." Feedwater controller failure is one of the limiting events that could cause a power increase w/o causing a core flow increase. Accept for full credit statements similar to "some transients are more severe if initiated from less than rated flow."
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o Resolution:
Faciljtycommentonfeedwatercontrollerfailureisnotconsistentwith question because this particular event becomes less severe with core flow less than rated. Key revised to indicate "and power increases" is not required for full credit.
5.16 Facility Comment:
- a. Answer key should accept for full credit statement similar to "orificing creates large pressure drop 9 core entrance, such that pressure drop due to increased 2 phase flow does not affect flow though the bundle."
- b. Drop the word " center" so phase is "...from the higher powered fuel bundles...". High powered bundles are not necessarily center bundles.
i Resolution:
Facility comment "a" above is a satisfactory response to HOW orificing accomplishes its purpose, but does not address WHY orfficing is necessary. Similarity to answer key would allow credit for answering HOW, but no credit for WHY, Comment "b" above is valid comment and word " center" will be removed from the answer key.
5.17 Facility Comment:
The question does not specify to "show work", therefore, full credit should be given for the correct answer w/o showing all the equations.
Resolution:
Writing the equation in the variable format given on the equation sheet is not necessary for credit for that portion of the answer. If candidate 5
provides ccrrect answer full credit will be awarded. If incorrect
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respo'nse is provided indication of how the answer was derived could receive partial credit. If the candidate chose to write the equation with numerical values inserted for variables, it would be considered an acceptable response for partial credit if an incorrect final answer was provided. It should be noted that statement No. 14 of the Rules and Guidelines has the examinee show all work whether stated or not in the individual question.
6.03 Fact 11ty Comment:
Answer key should also accept for full credit, high level in Supp. Pool as 26'9" water level in Supp. Pool. Please reference attached p.11 of Chapter 48 lesson plan on Primary and Secondary Containment to document this level. Also, see answer key to question 7.08 which uses the same level terminology.
Resolution:
Valid comment and answer key will be updated to include 26'9" as setpoint.
6.06b Facility Comment:
The question is misleading for the answer given. The question asks "why is a Rx scram initiated..." (underline emphasis is the facility's).
The correct answer to why we have the scram is to initiate the scram while there is still adequate accumulator pressure available. The answer should not require the statement..." assures the scram accumulator piston will be seated against nitrogen pressure" for full credit. Delete the above statement from the answer key. See attached copy of bases for Tech. Spec.
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I Resolution:
Anser key will be reworded as follows: b. Pressure greater than 1157 psig assures adequate accumulator pressure available [0.5] to insure nonnal scram performance. [0.5]. The other 0.5 points of credit is removed from the question point value and the total points available in section six is adjusted accordingly.
6.11 Facility Comment:
The answer key implies that the governor would only increase D/G load -
this is not true. The governor may increase or decrease load depending on grid frequency. Should accept full credit statements similiar to "makes the machine unstable, load sharing characteristics are hard to control, machine could possibly trip on reverse power or overload."
Resolution:
Although not stated in the question, the implication was that all other operator actions were correct. Were that the case, the answer key is correct per the stated reference. If it is assumed that the operator did not have voltage and frequency on the diesel adjusted properly prior to paralleling to the grid a reverse power trip is definitely possible as soon as the breaker is closed, but whether this is due to the speed droop (bad share) characteristics or the voltage regulator setpoint would be difficult to determine. Statements such as "makes the machine unstable" and " load sharing characteristics are hard to control" are ambiguous and do not adequately describe how the machine will react. The answer key will be revised to credit the assumption that diesel generator frequency is less than grid frequency when the diesel is parallel to the grid and the diesel trips on re-verse power.
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6.14 Facility Comment:
Answee key should also accept " dilution of boron in the S/D cooling loops" as one of the required five answers since question did specify
" Cold Shutdown". Reference SBLC lesson plan page 17 and 18 attached.
Resolution:
Answer key revised to include this comment.
6.15 Facility Comment:
Delete item "g". The question was inadvertantly written in a confusing manner. The wording in the question was "... delta-T greater than 10.1F".
If the delta-T is greater than 10.1, no action, only when the delta-T is less than 10.1 does the Recirc pump trip. Since the words " greater than" were not capitalized or underlined, this could confuse the question being asked. Also, the question was incorrect as asked, since it states l " Main Steam Line - Recirc pump suction Delta-T...". The true setpoint is " Steam dome - Recirc pump suction delta-T." (See attached page from LOP-RR-05). Delete part "g" of this question.
Resolution:
Comment.is noted that some confusion may arise if words " greater than" are not emphasized. However, the question as asked is, representative of knowledge requirements at the senior reactor operator level. In addition, facility training material (Chapter 5, page 40) refers to this interlock exactly as asked in the examination. During the written examination, only one candidate expressed confusion about the terminology used in the question. His question was "Does this nean Pump Suction to Pump Suction delta-T or Steam Dome to Pump Suction delta-T? The response given was " steam dome to pump suction". The answer key will be changed to 8
reflect #3 (Remain in fast speed) as the correct response and the question will remain in the exam.
7.03a Facility Comment:
Answer key should not require the Unit 2 part to specify " Low Water Level Reset Pushbuttons". Full credit should be given for stating
" reset pushbuttons."
Resolution:
Question requires a specific response which is not satisfied by the facility requested change. In order to satisfactorily determine the candidates knowledge level, the answer key response is required.
7.03b Facility Coment:
Answer key has points awarded for " boron dilution" and " temperature reduction" as key words. Full credit should be given for statements concerning " positive reactivity addition" and " power excursions" without stating explicit details of the positive reactivity addition.
Question did not indicate that these were necessary.
Resolution:
In order to satisfy the WHY portion of the question, a statement con-cerning positive reactivity addition must include mention of the agent causing the additon in order to indicate comprehensive knowledge re-quired by the question. An acceptable response would be "To prevent injection of low pressure ECCS systems and the resulting positive reactivity addition." The answer key is revised to indicate that the above is an acceptable response. The supplemental instructions to
" include possible consequences" require the candidate to mention power excursions and/or core damage for credit.
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7.05 Facility Comment:
The question specifies "Per LGP-1-S1," however, the facility feels that the definition of operability from Tech Specs should also be accepted for full credit sirce the procedural requirements are there to meet the intent of the Tech Spec definition. Please see the attached page from Tech Specs.
Resolution:
This question is not designed to test the intent of Tech Specs, but rather, the mechanics of its application. The procedure does clearly provide guidelines which are required. Credit will be given for the Tech Specs definition as a possible answer in lieu of the 3rd answer on the key and the answer key is revised accordingly.
7.07 Facility Conment:
Also, accept the following:
- 1) Steam flow / feed flow mismatch
- 2) Bypass valve closure, if open initially Resolution:
Answer key is revised to include both the extra facility symptoms as acceptable answers.
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7.15 Facility Comnent:
Changes in job titles at CECO have been implemented such that for part c an acceptable answer would be Admin. Superintendent and for part d.
Plant Manager is acceptable. These titles should be accepted for full credit. Reference the attached page from LAP-100-1, Station Organization.
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Resolution:
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The_an.swer key is revised to include these additional answers as acceptable for parts e and d.
8.01 Facility Comment:
Part a8b should also have acceptable answers of "N/A" or "not applicabl'e". This is the same as "any temperature". These responses should be acceptable for' full credit.
Resolution:
Comment is valid and the answer key is revised to accept N/A in lieu of "any temperature".
-8.03 Facility Comrent:
Answer key should also accept for full credit the statement e on tha attached definition of Secondary Containment. See attached page 1-6 from . Tech Specs.
Resolution:
Comment is valid and answer key is revised accordingly.
8.04 Facility Comment:
Since the Unit 2 statement of the question was not capitalized or r
underlined, it is possible the examinee will give Unit 1 limits. For part d 1.07, and for part e 1.08 should be accepted for full credit.
(These are the U1 Safety Limits - reference attached page from U-1 Tech Specs.)
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Resolution:
The candidate is expected to be able to differentiate between Unit I and Unit 2 and to knov which safety limits are for which Unit. The answer key is not changed.
8.05b Facility Comment:
Another acceptable answer to part b is " updates are given any time conditions change. This could be stated several different ways, f.e.,
when classifications are changed or when conditions on the NARS form are changed. NARS is the Nuclear Accident Reporting System used to notify the State and any time anything on the form changes a new form is filled out and the State is notified again. Suggest that the above statements be acceptable for full credit. (Reference the attached page from the GSEP manual).
Resolution:
The statement "When conditions / classification of the event change [0.5]
or at least once per hour [0.5]" will be added to the answer key and point value adjsuted to a total of 1.0 for the b portion of this question.
8.06 Facility Comment:
Examinee may use the term "ALARA" considerations. The ALARA concept is l
a CECO term for "As Low As Reasonably Achievable" when talking about dose.
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Resolution:
"For ALARA considerations" is an acceptable answer and the answer key is revised to reflect this.
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8.07a Facility Comment:
Anot_he,r acceptable answer not stated in the answer key is to " minimize the amount of time spent at the center desk." Reference LAP-1600-2, page 12 attached.
Resolution:
Valid comment and answer key is revised to accept this as one of the two possible answers.
8.07b Facility Comment:
The answer specifies that .5 pts. is awarded for saying the relief individual will "...brief the Unit Operator when he returns." The question does not address returning, the question asks proper procedure for leaving "at the controls." Suggest that the return position be deleted and full credit be granted if examinee answers with "a proper relief by a licensed individual." or similar phrase.
Resolution:
Answer key is revised to reflect this change. The 0.5 credit is removed from the exam and the section point value is adjsuted to reflect this.
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- U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: LASALLE 1&2 REACTOR TYPE: BWR-GES DATE ADMINISTERED: 86/10/27 EXAMINER: SPENCER. M.
CANDIDATE:
INSTRUCTIONS TO CANDIDATE:
Uno separate paper for the answers. Write answers on one side only.
Staple question sheet on top of the answer sheets. Points for each qunstion are indicated in parentheses after the question. The passing drede requires at least 70% in each category and a final grade of at 1ctst 80%. Examination papers will be picked up six (6) hours after tha examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 26 25.00 25.40- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS cW. Co 24.1C N 45v0ft 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 2f i 25.00 25.@0 7. PROCEDURES - NORMAL, ABNORMAL,
! EMERGENCY AND RADIOLOGICAL l CONTROL
.:2p. 5'o .2y. 75 35 00- 45.00 8. ADMINISTRATIVE PROCEDURES, i
CONDITIONS, AND LIMITATIONS
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$$. 0 100.00 Totals i
Final Grade l
All work done on this examination is my own. I have neither given l nor received aid.
Candidate's Signature NAS"B :PY
. i NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
- 1. Ceeating on the examination means an automatic denial of your application cnd could result in more sevrre 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 te examination room to avoid even the appearanc or possibility of cheating.
- 3. Use black ink ordark pencil only..to facilitate legible reproductions
- 4. Print your name in the blank provided on the cover sheet o the examination.
- 5. Fill in the date on the cover sheet of the examination (if necessary).
- 6. Use only the paper provided for answers.
- 7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
- 8. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a anw page, write onlY 2n Enn side of the paper, and write "ast Page" on the last answer sheet.
. Number each answer as to category and number,for example, 1.4, 6.3.
- 10. Skip at least thre lines between each answer.
- 11. Separate answer sheets from pad and place finshed answer sheets face down on your desk or table.
- 12. Use abbreviations only if they are commonly used in facility literature.
- 13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
l 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE l QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
f 16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
l 17. You must sign the statement on the cover sheet that indicates that the l
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.
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- 18. When you complete your examination, you shall:
- a. Assemble 1Riur examination as follows:
(1) Exam questions on top.
(2) Exam aids - figures, tables, etc.
(3) Answer pages including figures which are part of the answer.
- b. Turn in your copy of the examination and all pages used to answer the examination questions.
- c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions,
- d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
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- 5. THEORY OF NUC M AR POWER PLANT OPERATION. FLUIDS. AND PAGE 2
, THERMODYNAMICS QUESTION 5.01 (1.00)
WHY are installed sources not necessary after one fuel cycle?
QUESTION 5.02 (1.50)
How does Xenon concentration affect peripheral rod worth following a ceram from high power and WHY does this occur?
QUESTION 5.03 (1.50)
What are the TWO primary causes of fuel damage during operational transients?
I QUESTION 5.04 (1.50)
Assuming an ideal fluid system with no losses, the speed of a centrifugal pump is decreased from 1800 rpm to 1200 rpm. The information listed below is the 1800 rpm parameters.
1800 rpm parameters Flow = 2000 rpm Pressure =1000psih Power = 150 Hp l
What will the 1200 rpm values be for:
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- a. flow -
- b. pressure?
- c. Power?
l (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
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- 5. THEORY OF NUCr.uAR POWER PLANT OPERATION. FLUIDS. AND
. THERMODYNAMICS QUESTION 5.05 (1.00)
From the below list, choose which selection best describes total peaking factor.
- a. TPF = RPF X APF/LPF
- b. TPF = RPF + APF + LPF
- c. TPF = RPF X LPF/APF
- d. TPF = LPF X APF X RPF QUESTION 5.06 (2.00)
During a startup following a reactor scram, the plant is cooling down et 100 F/hr and Keff = 0.99. With no further rod withdrawal, WHEN will the plant go critical? (Do NOT consider Xenon effects.)
NOTE: Assume alpha TM = -1 X 10E-4 Delta-K/K/F SHOW ALL WORK FOR FULL CREDIT.
QUESTION 5.07 ( .50)
TRUE or FALSE?
As long as MFLCPR is less than 1.0 for every bundle in the core, the Technical Specification requirements for operating and safety limit MCPR are met.
QUESTION 5.08 (1.00)
Adequate core cooling is heat removal from the reactor sufficient to restore and maintain peak cladding fuel temperature below 2200 deg.F.
l What are THREE methods which the operator uses to assure adequate core cooling is maintained?
? HMT: E CCS sysfems; a t e- cdesty'd lo be able. ho (hvade. dese Ortee mdAols of fee m e d gj coge, keaf. 3 l _
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(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
PAGE 4
- 5. THEORY OF NUCTMAR POWER PLANT OPERATION. FLUIDS. AND
. THERMODYNAMICS QUESTION 5.09 (1.00)
For the following conditions, state whether the moderator temperature coefficient becomes MORE NEGATIVE or LESS NEGATIVE:
- a. Fuel temperature increases
- b. Moderator temperature increases
- c. Core age increases QUESTION 5.10 ( .50)
TRUE or FALSE?
It is possible to increase core average power by inserting a control rod.
QUESTION 5.11 (2.00)
Match each of the four lettered items with one of the numbered items.
(A letter-number sequence is sufficient.)
! 3. CPR 7. TOTAL PF
- 4. FLPD 8. LHGR
- a. Parameter by which plastic strain and deformation are limited to less than 1%.
- b. Contains guidelines restricting power ramp rates above the threshold power.
QUESTION 5.12 (1.00) l WHY does the value of the core average delayed neutron fraction (Beta-bar)
DECREASE over core life?
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
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- 5. THEORY'OF NUCH AR POWER PLANT OPERATION. FLUIDS. AND
- THERMODYNAMICS QUESTION 5.13 (3.00)
Describe HOW and WHY the core responds to each of the following rod movements. Include in the discussion the effect on both axial and radial flux distributions. (Assume power is greater than 75%)
- c. The withdrawal of a Deep Control Rod. (From a deep position to another deep position.) [1.5)
- b. The withdrawal of a Shallow Control Rod. [1.5)
QUESTION 5.14 (1.00)
Fill in the blanks below with the word which makes the statement true.
- a. Increasing the amount of condensate depression from 8 deg. to 11 des LaSa11e's thermodynamic will (INCREASE or DECREAGE) cycle efficiency.
- b. Removing a high pressure FW heater from service will (INCREASE or DECREASE) the thermodynamic cycle efficiency.
QUESTION 5.15 (1.50)
- a. HOW does the Technical Specifications MCPR operating limit change (INCREASE or DECREASE) when core flow is less than rated?
- b. WHY is it more possible to violate the MCPR Safety Limit with core flows less than rated and no corresponding adjustment to the MCPR operating limit?
QUESTION 5.16 (2.00)
Explain WHY core orificing is necessary and HOW orificing accomplishes this purpose.
I (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
PAGE 6
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND
. THERMODYNAMICS QUESTION 5.17 (3.00)
The reactor is suboritical with a Keff of .95 a SRM countrate of 200 cps. The control rods are withdrawn and the new countrate is 400 cys.
- a. How much reactivity was added? (1.5)
- b. What would be the STATUS of the reactor (i.e. subcritical, critical, or supercritical) if the same amount of reactivity, determined in a.,
was added again? (1.5) t
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l 1 (***** END OF CATEGORY 05 *****)
PAGE 7
- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION QUESTION 6.01 (1.00)
- a. What is the minimum count rate for operability of a Source Range Monitor which has been verified to have a signal-to-noise ratio of at least 2:17
- b. What signal (in addition to the annunciator) is generated by the Reactor Manual Control system when the SRM downscale trip signal is received?
QUESTION 6.02 (1.00)
RCIC is being used to control reactor level when the level is inadvertantly allowed to increase to +55.5".
- a. How will the RCIC system respond to this level?
- b. After this event, when will RCIC AUTOMATICALLY inject,1f ever?
(Assume no operator action.)
QUESTION 6.03 (2.00)
What will AUTOMATICALLY cause the HPCS pump suction to transfer from the Condensate Storage Tank to the Suppression Pool? (INCLUDE SETPOINTS!)
QUESTION 6.04 (2.00)
Indicate whether each of the following statements concerning the Rod Sequence Control System are TRUE or FALSE.
[ Assume a reactor startup is in progress and Sequence A is being used.]
f a. Any rod in Groups A1, A2, A3, or A4 may be selected and moved first.
- b. When Group Al is the first group moved and is fully withdrawn, a Group A3 rod must be moved next.
- c. After Groups Al and A2 have been fully withdrawn (75% rod density), a Group A3 rod may be selected and fully withdrawn. ,
- d. After all Group A1, A2, A3, and A4 rods have been fully withdrawn (50% rod density), any rod in groups A5 through A10 may be selected and moved.
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(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
PAGE 8
- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION QUESTION 6.05 (1.50)
Indicate whether each of the following evolutions CAN or CANNOT be performed all or in part from the Remote Shutdown Panel,
- c. Reactor cooldown using Safety Relief valves
- b. Vessel water level control using RCIC
- c. Injection of Standby Liquid Control
- d. Initiation and injection of HPCS
- o. Suppression pool cooling using RHR "B" in Suppression Pool Cooling Mode
- f. Reactor Cooldown using RER "A" in the Shutdown Cooling Mode
. QUESTION 6.06 (0.00)^-[,25)
- e. What THREE system parameters will cause the Electric Power Monitoring Assembly (EPMA) to trip the output breakers from the associated Reactor Protection System power supply? (Setpoints NOT required.) [1.0)
- b. Why is a Reactor Scram initiated if the Control Rod Drive charging water pressure is less than or equal to 1157 psig with the Mode Switch in STARTUP or REFUEL? -F1.5] " [f. o]
- c. Why is the CRD low charging water pressure scram required to be active only in STARTUP or REFUEL modes? [0.5]
, QUESTION 6.07 -
(2.00)
?
- a. Describe HOW a fission chamber functions from the time the thermal neutron enters the chamber until the electrical output is generated.
[1.5]
- b. Why is the Uranium Oxide coating of the LPRM detector enriched with U234 (in addition to U235)? [0.5]
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
t__ .._ __ _. , _ _ _ . . _ __. _
PAGE 9
- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION QUESTION 6.08 (1.00)
Choose the answer below which best completes the following statement:
The maximum allowable injection time for Standby Liquid Control (125 cinutes at 1220 psig) is based on ...
- a. overcoming the reactivity addition due to cooldown following a Xenon poison peak.
- b. overcoming the effects of improper mixing and reactivity
" chugging".
- c. providing sufficient time to allow for coast down of Reactor Recirculation pumps.
- d. providing the operators time to complete reactor shutdown by control rod insertion using alternate means.
QUESTION 6.09 (1.00)
List FOUR diesel / diesel generator trips which ARE bypassed during LOCA conditions. (Setpoints are NOT required.)
QUESTION 6.10 (1.00)
- a. What are the conditions which will AUTOMATICALLY (no operator action required) initiate LPCS? [SETPOINTS REQUIRED]
- b. WHY is the LPCS injection valve (F005) NOT permitted to open l
until RPV pressure is less than 500 psig?
QUESTION 6.11 (1.50)
While performing the diesel generator operability surveillance, the operator at the diesel neglects to adjust the speed droop on the governor to the reuired 50% value. BRIEFLY describe HOW the diesel generator will l
react when paralleled to the normal supply if the speed droop remains l ct 0%.
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
l
PAGE 10
.6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION QUESTION 6.12 (1.00)
Choose the phrase below which best completes the following statement:
The LPCS system piping inside the reactor vessel is monitored for integrity by the measurement of the pressure differential between the LPCS piping downstream of its testable check valve and ...
- a. the HPCS piping downstream of its testable check valve
- b. the above core plate pressure tap
- c. the "A" LPCI system piping downstream of its testable check valve
- d. the below core plate pressure tap
- QUESTION 6.13 (2.00)
What are FOUR Unit 1 automatic (no operator action required) initiation signals for the Standby Gas Treatment System? (Include setpoints)
QUESTION 6.14 (2.00)
List FIVE of the positive reactivity effects which Standby Liquid Control cystem is expected to overcome in order to take the reactor from rated power to cold shutdown. -
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
l .
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- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION QUESTION 6.15 (3.00)
Match the conditions below to the result (1, 2, or 3 below) that each condition would be expected to have on fast speed Reactor Recirculation pumps with no operator action.
CONDITION RESULT Suction valve less than 90% open 1. Automatically trip to c.
zero speed
- b. Discharge valve less than 90% open
- 2. Automatically downshift Recirc loop flow mismatch greater to slow speed c.
than 10% Remain in fast speed 3.
- d. 1135 psig RPV pressure
- o. Turbine stop valve closure above 30% power
- f. Turbine Control valve fast closure (low RETS) above 30% power
- g. Main steam line - Recire pump suction Delta-T greater than 10.1 F
- h. Feed water flow less than 30%
(***** END OF CATEGORY 06 *****)
s ,
PAGE 12
- 7. PROCEDURH - NORMAL. ABNORMAL. EMERGENCY AND
. RADIOLOGICAL CONTROL QUESTION 7.01 (1.00)
What FOUR items are required to be logged in the Unit Log when the reactor is declared critical, per LGP-1-1?
QUESTION 7.02 (1.00)
According to LOA-RX-01, Control Room Evacuation, WHERE should the following people report after clearing the Control Room?
[Be as specific as possible.]
- a. Unit 1 NSO
- b. Unit 2 Shift Foreman
- c. Shift Engineer
- d. SCRE QUESTION 7.03 (2.50)
LGA-ATWS-01, ATWS Power Control, directs the operator to prevent automatic initiation of ADS per LOA-NB-11 when injecting boron.
- a. How is automatic actuation of ADS prevented by LOA-NB-11,
' Preventing ADS Auto Actuation? (Both Unit 1 AND Unit 2 methods required for full credit) [1.0]
! b. WHY is it desirable to prevent actuation of ADS when injecting boron? [ Include possible consequences.] [1.5]
QUESTION 7.04 (1.00)
I The LGA General Precautions provide two conditions which allow the ADS SRVs to be CLOSED after automatically openning. What are the TWO conditions?
QUESTION 7.05 (1.50)
LGP-1-S1, Master Startup Checklist, provides guidance to the Operations Staff for determining when a system / component is considered as operable.
What are the THREE conditions which must be met?
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
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- 7. FEQGEDUMR - NORMAL. ABNORMAL. EMERGENCY AND
. B&DJOLOGICAL CONTROL QUESTION 7.06 (1.00)
WHAT are the recommended guidelines for voluntary WHOLE BODY exposure limits under the following emergency conditions, per LRP-1000-1?
- c. To control a fire or to eliminate a serious unplanned radioactive release.
- b. To save a life.
QUESTION 7.07 (2.50)
List FIVE symptoms of a Stuck Open Safety Relief Valve, per LOA-NB-02.
(Do NOT include alarms or annunicators!)
QUESTION 7.08 (1.50)
What are the ENTRY CONDITIONS for Primary Containment Control, LGA-037
[Setpoints Required]
QUESTION 7.09 ( .50)
When using the SRVs for RPV pressure control in accordance with the LGAs, HOW does the operator determine which valve is to be operated next?
QUESTION 7.10 (2.00)
Level Restoration, LGA-04, provides a list of eleven systems which may be used to Rapidly Depressurize the RPV if less than three SRV's are open.
List SIX of these systems.
QUESTION 7.11 (1.00)
WHEN may the MINIMUM STARTUP CHECKLIST (LGP-1-S2) be substituted for the MASTER STARTUP CHECKLIST (LGP-1-S1)? (TWO conditions required for full credit)
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
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PAGE 14
- 7. PROCEDURM - NORMAL. ABNORMAL. EMERGENCY AND
. RADIOLOGICAL CONTROL QUESTION 7.12 (2.50)
What are the ENTRY CONDITIONS for ATWS Power Control LGA-ATWS-017
[Setpoints required.]
QUESTION 7.13 (1.00)
WHY is it desirable to operate an RHR pump below 6000 gym in the Shutdown Cooling Mode until the Recirculation pump in the associated loop has been shutdown?
QUESTION 7.14 (1.50)
DEFINE the following in accordance with LaSalle Radiation Procedure, LRP-1000-1.
a) Radiation Area [0.75]
b) High Radiation Area [0.5]
c) Daily Whole Body Dose [0.25]
QUESTION 7.15 (2.00)
According to the LaSalle Radiation Procedure LRP-1000-1, what JOB POSITION is the minimum level of authority that may grant an individual permission to exceed the following occupational dose limits?
I 50 mrem / day whole body but less than 100 mrem / day a.
l b. 100 mrem / day whole body but less than 1250 mrem /qtr l
- c. 1250 mrem /qtr but less than 5000 mrem /yr
- d. 5000 mrem /yr but less than 7000 mrem /yr QUESTION 7.16 (2.50)
In accordance with the limitations of LGP-1-1, Normal Unit Startup, power operation with one Recirculation Pump is permitted providing seven conditions are met. List FOUR of these conditions.
(***** END OF CATEGORY 07 *****)
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PAGE 15 l
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS QUESTION 8.01 (2.00)
Fill in the blanks on the following table in accordance with the OPERATIONAL CONDITIONS table (TABLE 1.2) from the LaSalle Technical Specifications. ,
MODE SWITCH AVERAGE REACTOR CONDITION POSITION COOLANT TEMPERATURE
- 1. POWER OPERATION RUN a.
- 2. STARTUP STARTUP/ HOT STANDBY b.
- 3. c. SHUTDOWN >200 F
- 4. COLD SHUTDOWN d. </=200 F
- e. f.
- 5. REFUELING QUESTION 8.02 (1.00)
Technical Specifications require surveillance items to be performed within specific time intervals and provide limits for exceeding those times.
i a. What is the maximum extention for a single surveillance item?
I l b. What is the maximum combined time interval for any three consecutive surveillance items?
QUESTION 8.03 (3.00)
LIST the Technical Specification requirements for SECONDARY CONTAINMENT to exist.
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
.. . . _ = -. . _ _ - . . .-
s .
_8. ADMINISTRATIVE PROCEDU m . CONDITIONS. AND LIMITATIONS PAGE 16 f I
l QUESTION 8.04 (3.00)
Fill in the blanks regarding the LaSalle Unit 2 Safety Limits:
THERMAL POWER shall not exceed (a) of RATED THERMAL POWER with the reactor vessel steam dome pressure less than (b) or core flow less than (c) of rated flow.
The Minimum Critical Power. Ratio (MCPR) shall not be less than (d) with two recirculation loop operation and shall not be less than (e) with single loop operation with the reactor vessel steam dome of pressure Greater than (b) and core flow greater than (c) rated flow.
The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed (f) .
(6 0 0.5 ea)
QUESTION 8.05 (2.50)
In accordance with LZP-1110-1,
- n. What THREE " State and Federal" agencies are always notified when a GSEP classified event occurs? [1.5]
- b. How often are updates provided to these agencies? [1.0)
QUESTION 8.06 ( .50)
WHEN may the second person verification for SAFETY RELATED outages be waived per LAP-900-47 i
QUESTION 8.07 ( 3. 00 ) " ' (A .5)
Concerning Conduct of Operations, LAP-1600-2:
- a. The "At the Controls" operator is required to remain within the "At the Controls" area and within line-of-sight of the Unit front panels. State TWO additional requirements he must be able to meet regarding his physical position.
- b. EXPLAIN the procedure which must be followed if the Unit Operator must leave the "At the Controls" area?
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
i
I i . l PAGE 17 1
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS
~
l l
QUESTION 8.08 (2.25)
What THREE requirements must be met per LaSalle Unit 1 Technical j Specifications in order to make Temporary Changes to procedures?
QUESTION 8.09 (1.50)
If a " temporary SYSTEM change" is required to be implemented on a weekend, what THREE conditions must be met to comply with LAP-240-6, Temporary System Changes?
QUESTION 8.10 ( .75)
WHO is responsible for the initial classification of an emergency cvent according to the GSEP?
QUESTION 8.11 (1.50)
What are the SIX general categories of emergency events included in the LaSalle General Stations Emergency Plan, GSEP7 QUESTION 8.12 (1.00)
FILL IN THE BLANK with one of the following TS terms:
"A shall be the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip I functions and channel failure trips."
- a. Channel Calibration l
- b. Channel Check r c. Channel Functional Test
- d. Logic System Functional Test 1
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
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PAGE 18
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS QUESTION 8.13 (2.00)
WHEN are the conditions of " stable and under control" considered to exist in order to place an ECCS system component in manual or pull-to-lock, per LAP-1600-27 QUESTION 8.14 (1.00)
Unit 1 Technical Specification 3.4.4 establishes the following conductivity and chloride limits:
PLANT CONDITION CONDUCTIVITY LIMIT CHLORIDE LIMIT 1 umho/cm 0.2 ppm 1
2 umho/cm 0.1 ppm 2 and 3 Per the TS Basis, WHY is the chloride limit more restrictive at (1.0) the lower steaming rate than when at power?
l l
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(***** END OF CATEGORY 08 *****)
(************* END OF EXAMINATION ***************)
s .
-' EQUATION SHEET
~.
f = ma v = s/c 2 Cycle efficiency =
- }
w = 28 s = v,t + ac E E = aC -
a = (vg - v )/c KE = 1 3mv v g =v + .a A = AN A = A,e" E g
PE = agh w - 6/g A = In 2/tq = 0.693/tg W = v&P' (t, )(e ) .
AE = 931Am
- h* ~
(5~g+e) b 6=sc,AT ,
g . g, ,-rx
', Q = UAAT g , g ,Wx
~Pur = Wg $' .
I=I
~
10
- P=P 10 (EI- TVL = 1.3/u P=P et /T HVL = 0.693/u
~
4
~SUR = 26.06/T T = 1.44 DT SCR = S/(1 - K,gg)
/A
- c)
SUR = 26 g CR x = S/(1 - K,gg )
T = '(1*/o ) + [(6 'o)/A,ggo ] 1(1 ~ *aff)1 " K2(I ~Keff)'2 ~~-
T = 1*/ (o - f)
M = 1/(1 - K,gg) = CR /CR0 g T = (I - o)/ A *ff o M = (1 - K,gg)0/ (l ~ Keff)1 8 " ( aff-1)/K,gg = AKdf/Keff SM = 0 - K aff Weff
~
p= [L*/TK,gg .] + [H/(1 + A,ggT )] ,
1* = 1 x 10 ' seconds
-I P = I$v/(3 x 1010) A aff = A 0.1 seconds I = No Idgg=Id22 l WATER PARAMETERS Idg =Id 2 2
1 gal. = 8.345 lbm R/hr = (0.5 CE)/d g,,t,,,)
I gal. = 3.78 liters R/hr = 6 CE/d (feet) -
1 ft = 7.48 gal. MISCELLANEOUS CONVERSIONS .
10 Density = 62.4 lbm/f t 1 Curia = 3.7 x 10 dps
. Density = 1 gm/cm i kg = 2.21 lba l Heat of varori:acion = 970 Ecu/lbm I hp = 2.54 x 10 BIU/hr 0
Heat of fusica = 144 Btu /lbs 1 Mw = 3.41 x 10 Beu/hr 1 Atm = 14.7 Psi = 29.9 in. Ig. 1 Stu = 778 ft-lbf I ft. H O 2
= 0.433". Ibf /in 1' inch = 2.54 cm F = 9/5 C + 32
- C = 5/9 (*T -32)
s a PAGE 19
'5. THEORY OF NUCM AR POWER PLANT OPERATION. FLUIDS. AND THERMODYNAMICS
-86/10/27-SPENCER, M.
ANSWERS -- LASALLE 1&2 ANSWER 5.01 (1.00)
Intrinsic sources will supply a sufficient neutron population to maintain SRM counts above Technical Specification requirements.
(also ucept : su%'cient to ma, A in sear c. cads al,ua moh%am gegu %c) ' ca.
REFERENCE obove m .'a im u m eeg usned fog safe. sfactug )
LaSalle, Rx Physics Review, Aug. 1985, Rev. 1, P. 62 ANSWER 5.02 (1.50)
Peripheral rod worth will increase. [0.5] High Xenon concentration in the previously high power (central) portion of the core will depress the thermal neutron flux in that portion of the core and " push" the flux into the previously low power regions (peripheral). Thermal flux peaks et the periphery of the core cause higher rod worth. [1.0]
REFERENCE LaSalle, Rx Physics Review, Rev. 1, Aug. 1985, PP. 188, 220 ANSWER 5.03 (1.50)
- 1. Severe overheating of rod cladding due to inadequate cooling. [0.75]
- 2. Rupture of cladding due to strain caused by relative expansion of the fuel, pellet (PCI is acceptable). [0.75]
i REFERENCE I LaSalle, Nuclear Fuel System Description I
THEORY OF NUCTRAR POWER PLANT OPERATION. FLUIDS. AND PAGE 23 5.
. THERMODYNAMICS ANSWERS -- LASALLE 1&2
-86/10/27-SPENCER, M.
ANSWER 5.04 (1.50)
- c. flow = 1,332 gym +/- 10 gym [ 2000 X 0.6666 )
- b. Pressure = 450.00 +/- 10 psta-[ 1000 X 0.666 squared ]
- c. power = 44.5 Hp +/- 0.62 Hp [ 150 X 0.666 cubed ]
( 3 0 0.5ea = 1.5 )
REFERENCE LaSalle Fluid Flow and Heat Transfer, Revi, Aug 1985, pg 46 Grand Gulf, Heat Transfer and Fluid Flow, Vol 2, Chapter 6, page 6-96a.
ANSWER 5.05 (1.00) d REFERENCE I
LaSalle Core Thermal Hydraulice, Rev2, Apr 1986, pg 42 GG, Heat Transfer & Fluid Flow, Vol. 2, Ch. 9, P. 9
. . _ ,m_.'. .- ,, , . - _ . _ _ . _ _ . . . . - . . _ . _ _ _ _ . _ _ _ _
% e PAGE 21
- 5. THEORY OF NUCMAR POWER PLANT OPERATION. FLUIDS. AND
. THERMODYNAMICS ANSWERS -- LASALLE 1&2 -86/10/27-SPENCER, M.
ANSWER 5.06 (2.00)
Delta-K Keff - 1
[0.5]
= --------
K Keff
.99 - 1
= -------
.99
= -0.0101 [0.5]
MUST INSERT 0.0101 Delta-K to reach Keff = 1 K
Delta-K
= Alpha-TM Delta-T K
Delta-K
= Alpha-TM (Cooldown rate X time) [0.5]
K Delta-K Delta-K 0.0101 ------- = (-1 X 10E-4 -------) (-100 F/hr) time K K Delta-K
' O.0101 -------
K time : ------------------------------
Delta-K
(-1 X 10E-4 -------) (-100 F/hr)
K/F time = 1.01 hr [0.5]
l REFERENCE l
LaSalle, Rx Physics Review, Rev. 1, Aug. 1985, P. 120 l
l i
l I
l
s -
- 5. THEORY OF NUC MAR POWER PLANT OPERATION. FLUIDS. AND PAGE 22
. THERMODYNAMICS ANSWERS -- LASALLE 1&2 -86/10/27-SPENCER, M.
ANSWER 5.07 ( .50)
TRUE REFERENCE LaSalle, Core Thermal Hydraulics, Rev. 2, April 1986, P. 74 ANSWER 5.08 (1.00)
- 1. Core submergence [0.33]
- 2. Spray cooling [0.33]
- 3. Steam cooling [0.33]
l REFERENCE LaSalle, BWROG Emergency Procedure Guidelines ANSWER 5.09 (1.00)
- 9. dSE[inegative [0.33]
- b. More negative [0.33]
- c. Less negative [0.33]
REFERENCE LaSalle, Rx Physics Review, Rev. 1, Aug. 1985, PP. 122-123 1
i
( ANSWER 5.10 ( .50)
TRUE
THEORY OF NUCM AR COWER PLANT OPERATION. FLUIDS. AND PAGE 23 5.
. THERMODYNAMICS ANSWERS -- LASALLE 1&2 -86/10/27-SPENCER, M.
REFERENCE LaSalle, Rx Physics Review, Aug. 1985, P. 206 ANSWER 5.11 ( 2.03 P O(1)8 (0.5)
- b. 5 (0.5)
- c. 1 (0.5)
- d. 4 (0.5)
REFERENCE LaSalle Core Thermal Hydraulics, Rev 2, Apr 1986, pg 50, 52, & 82 GE Thermodynamics, Heat Transfer, and Fluid Flow, Chapter 9 ANSWER 5.12 (1.00)
The fuel composition changes over core life as U 235 is consumed and Pu 239 is produced. [0.5] The delayed neutron production of Pu 239 is smaller than that of U 235 [0.5].
REFERENCE LaSalle, Rx Physics Review, Aug. 1985, Rev. 1, P. 96 i
1 l
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PAGE 24
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND
. THERMODYNAMICS
-86/10/27-SPENCER, M.
ANSWERS -- LASALLE 1&2 ANSWER 5.13 (3.00) 0.1s_]
- c. (The core response to the withdrawal of a deep control rod)is to Yaise core power in the upper region of the core, : g;; icily in th: 1-arer: "here the red ::: withdr =n st- Since the void content the of the upper Portion of the core is high at operating conditions, effects of deep con However, the addition of coupling cells to the core and increased migration length for neutrons 4will affect the radial flux and thus the total core power will[ increase. 6).5] [a2s]
[o.1r]
- b. (Thecoreresponsetowithdrawalofashallowcontrolrod)isto
[gggraisethepowerlocallyintheregionwherethecentrolrodis withdrawn 3 The local power increage y ag the core bottom will
-pull the boiling Leundary de~c,=i,ncross.ng+the void content above the withdrawn control rod." The negative effects of voids may or may not be stronger than the positive effects of rod withdrawal [0.()
and perer : y decrease. In addition, radial response _is limited [04) because of the shadowing effect of nearby rods./~he T shallow rod strongly affects the axial power shape and not the overall power.[o.zr]
REFERENCE WNP-2 Reactor Physics VIII Operatng Characteristics, VIII.A.6.C.1) &
2), pg 38 & 39.
LaSalle Rx Physics Review, Rev 1, Aug 1985, pg 176 & 177 ANSWER 5.14 (1.00)
- n. DECREASE [0.5]
- b. DECREASE [0.5]
REFERENCE LaSalle, Thermodynamics, Rev. 1, Aug. 1985, PP. 76-80
PAGE 25
- 5. THEORY OF NUCTRAR POWER PLANT OPERATION. FLUIDS. AND
. THERMODYNAMICS ANSWERS -- LASALLE 1&2 -86/10/27-SPENCER, M.
ANSWER 5.15 (1.50)
- e. Increase [0.5]
- b. Transients which cause rapid core flowhond power increases)become more severe if initiated from low flow / low power conditions. [1.0]
REFERENCE LaSalle, Core Thermal Hydraulics, April 1986, Rev. 2, P. 72 ANSWER 5.16 (2.00)
As the boiling rate increases, two-phase flow resistance increases.
This would tend to divert coolant flow from the higher powered cerler-fuel bundles where it is needed the most (concept 1.0).
Orificing has the effect of providing a large resistance to flow so that any additional resistance caused by two-phase flow is acceptably cmall (concept 1.0).
REFERENCE LaSalle Reactor Vessel & Internals S/D, Chap 2, Feb 1985, pg 30 G. P. Heat Transfer and Thermal Limits, pg 28
, a
- 5. THEORY OF NUCWAR POWER PLANT OPERATION. FLUIDS. AND PAGE 26 THERMODYNAMICS ANSWERS -- LASALLE 1&2 -86/10/27-SPENCER, M.
ANSWER 5.17 (3.00)
- a. CR1 (1-Keff1) = CR2 (1-Keff2) [0.5]
200 (1 .95) = 400 (1-Keff2)
[200 (1 .95)/400] -1 = -Keff2
.975 = -Keff2 [0.25]
delta p = Keff2-1/Keff2 - Keff1-1/Keffi [0.5]
delta p =( .975-1)/.975 - (.95-1)/.95 delta p = ( .0256) - ( .0526) delta p = .027 +/ .002 [0.25] ,
- b. Part b. will be graded independently of part a.
delta p =(Keff3-1)/Keff3 - (Keff2-1)/Keff2 [0.5]
i .027 = (Keff3-1)/Keff3 - (.975-1)/.975
.027 = [1-(1/Keff3)] - ( .0256) 1
.0014 = 1-(1/Keff3)
.9986 = -1/Keff3
.9986 Keff3 = -l Keff3 = 1.0014 +/- .0014 [0.25]
The reactor is supercritical (will accept critical) (0.75)
REFERENCE LaSalle Reactor Physics Review, Rev 1, Aug 1985, pg 96 WNP-2 Reactor PhysicsSection VI Reactor Operations Part A.
Suberitical Reactor (no page numbers available), subsection e.
Determine Criticality.
Also see pp 21 of Section III for reactivity
- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION , PAGE 27
~
ANSWERS -- LASALLE 1&2 -86/10/27-SPENCER, M. ;
ANSWER 6.01 (1.00)
- c. 0.7 (cys) [0.5]
- b. Rod Withdrawal block [0.5]
REFERENCE LaSalle, Source Range Monitoring System Description, Ch. 11, Rev. 1, Nov. 1985, PP. 47 and 51 ANSWER 6.02 (1.00)
- n. The steam supply valve (E51-F045) will close. [0.5]
- b. When level drops to -50", RCIC will reinitiate. [0.5]
REFERENCE LaSalle, Starting and Operating of the RCIC System, LOP-RI-02, Limitations and Actions, Rev. 7 ANSWER 6.03 (2.00)
Low level in the CST [0.5] 3 ft. 1 in. indicated [0.5]
[also accept the following for setpoint value: 5 ft 1 in actual or 715 ft 7 in per Tech Specs]
High level in the Suppression Pool [0.5] 2 in above novmal [0.5]
[also accept the following for setpoint value: 700 ft 1 in per Tech Specsf p n. % (+ 9 w. }
REFERENCE LaSalle, HPCS System Description, Ch. 36, Rev. 1, Nov. 1985, P. 33 LaSalle, Unit 1 Tech Specs Table 3.3.3-2, P. 3/4 3-30
PAGE 28
.6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION.
ANSWERS -- LASALLE 1&2 -86/10/27-SPENCER, M.
ANSWER 6.04 (2.00) 1
- c. TRUE [0.5]
- b. FALSE [0.5]
- c. FALSE [0.5]
- d. TRUE [0.5]
REFERENCE LaSalle, Rod Sequence Control System Description, Ch. 19, Feb. 1985 ANSWER 6.05 (1.50)
- n. Can
- b. Can
- c. Cannot
- d. Cannot
- o. Can l f. Cannot (6 @ 0.25 each)
REFERENCE LaSalle, Remote S/D System, Ch. 74, Rev. 1, Dec. 1985
(
l I .__ -- _. .
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- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION ANSWERS -- LASALLE 1&2 -86/10/27-SPENCER, M.
ANSWER 6.06 (-3. ; - h I)
- a. 1. Undervoltage
- 2. Overvoltage
- 3. Underfrequency
[3 0 0.33 each]
oc/e ll fgesSuAe b.
Pressure greater than 1157 psig assures "-----'*--w fuale. <ceama A cM.
picten ' fill bc :::t d ::: inst nitres 4--"- #"
n pre::"r:
' ' [0.5]. c;;
Prep:r s bilit; -
sec+4=" ^* the pisten 111-and ncia.el musoua yo fvimauvo [1.0]". cui/4.6/e. [ g -t Jau rte.
NoaMal scram per-fog maNce, [0. S*.]
- c. In the STARTUP or REFUEL mode, reactor pressure may be insufficient to assist CRD scram action. [0.5]
REFERENCE
- o. LaSalle, RPS System Description, Ch. 20, Rev. 1, Nov. 1985, P.16
- b. LaSalle, RPS S/B, P. 38, also LaSalle Unit 1 Tech. Specs., P. B 2-13
- c. LaSalls, Unit 1 Tech. Specs., P. B 2-13 ANSWER 6.07 (2.00)
- a. Incident neutron will cause a fission reactionfission with U235 resulting in highly charged fission fragments [0.5]. The fragments will cause ionization of the argon gas [0.5]. The ion pairs migrate to the electrode and the cylinder creating an electrical pulse from the detector [0.5].
- b. Addition of U234 is used to increase detector lifetime. [0.5]
234 1 235 (U + N -> U )
O REFERENCE LaSalle, a. SRM System Description, Ch. 11, Rev. 1, Nov. 1985, PP. 19-25
- b. LPRM System Description, Ch. 13, Rev. 1, Nov. 1985, P. 14 i_ .-_ _ _ _ _ _ _
e <
- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 30 ANSWERS -- LASALLE 1&2 -86/10/27-SPENCER, M.
ANSWER 6.08 (1.00) c.
REFERENCE LaSalle, Standby Liquid Control System Description, Ch. 10, Jun. 1985 P. 18 ANSWER 6.09 (1.00)
Any four of the following 9 0.25 each;
- 1. Engine high water temperature
- 2. Engine start failure
- 3. Engine low lube oil pressure
- 4. Overcurrent with voltage restraint
- 5. Loss of excitation
- 6. Reverse power
- 7. Underfrequency
- 8. Ground overvoltage (neutral ground)
REFERENCE
! LaSalle, Diesel Generator and Auxiliaries System Description, Ch. 17, Rev. 1, Nov. 1985, PP. 25-27 ANSWER 6.10 (1.00) j o. Low RPV level of -129" [0.25]
High drywell pressure of 1.69 psig [0.25]
l
- b. To protect the LPCS piping from overpressurization [0.5]
l l
1
. _ _ . ._ - __,..-. -..~-. -..-. -_.,__ __ -....-. - ..- _ _,-,. - ._-_... . -.
PAGE 31
- 6. FLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION ANSWERS -- LASALLE 1&2 -86/10/27-SPENCER, M.
REFERENCE LaSalle, SDLP-27, Rev 1, Jan 86, P. 17 ANSWER 6.11 (1.50)
(With the speed droop still' set at 0%,)the govenor would try to make the grid frequency equal to its demanded frequency (maintain 60 Hz) [0.5].
until it tripped on This would cause DG load to increase [0.5]
overload. [0.5]
( %o Accept - -A e y uren. w ou.ld lo , stake $e gaid laqueNes qaa.l b }ls ck'modb REFERENCE 60 % [,o y . N is uod cese. f4e gentecedox.
44%ue,4cre("if
+o wo less f auLo s] ~4 usth)I +ke outr<+ basakee +act d LaSalle, Diesel Generator and Auxiliaries System Description, Ch.p<47 Fowed-[o.
Rev. 1, Nov. 1985, P. 35 ANSWER 6.12 (1.00) c.
REFERENCE LaSalle, LPCS System Description, Ch. 38, Mar. 1985, PP. 19-20 and Fig. 38-6 ANSWER 6.13 (2.00)
- 1. High Drywell Pressure [0.25] 1.69 psig [0.25]
- 2. Low RPV water level [0.25] -50 in. [0.25]
- 3. High Radiation - Fuel Fool Exhaust (0.25] 10 mr/hr [0.25]
- 4. High Radiation - Rx Bldg Exhaust (0.25] 10 mr/hr [0.25]
REFERENCE LaSalle, Standby Gas Treatment System Description, Ch. 51, Rev. 1, Nov. 1985, P. 20 l
l l
t
t #
PAGE 32
- 8. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION ANSWERS -- LASALLE 1&2 -86/10/27-SPENCER, M.
ANSWER 6.14 (2.00)
Any FIVE of the following e 0.4 pts. each
- 1. Elimination of steam voids
- 2. Changing water density from hot to cold .
- 3. Reduced Doppler effect
- 4. Reduced neutron leakage from hot to cold
- 5. Decreased control rod worth (as moderator cools)
- 6. Complete decay of rated power Xenon inventory
- 7. 3% shutdown margin
- 8. DikNen of borton in the ShwO& MIId3 I"f 5 REFERENCE LaSalle, SBLC System Description, Ch. 10, Jun. 1985, PP. 17-18
CONTROL. AND INSTRUMENTATION PAGE 33 l
- 6. PLANT SYSTEMS DESIGN ANSWERS -- LASALLE 1&2 -86/10/27-SPENCER, M. i l
ANSWER 6.15 (3.00)
- e. 1. (Automatically trip to zero speed)
- b. 1. (Automatically trip to zero speed)
- c. 3. (Remain in fast speed)
- d. 1. (Automatically trip to zero speed)
- o. 2. (Automatically downshift to slow speed)
- f. 2. (Automatically downshift to slow speed) a.
',e N.. *E..$.?+ 52. 0...'.I
+. +.
H. !!Y...b...a.
. . .., -. ..... - . - n. ,
- h. 2. (Automatically downshift to slow speed)
- 1. 1. (Automatically trip to zero speed)
- j. 3. (Remain in fast speed (FCV runback))
[10 @ 0.3 each]
REFERENCE LaSalle, Recirc'ulation System Description, Ch. 5, June 1985, PP. 46-55 Reactor Vessel Instr. system Description., Ch. 3, April 1985, P. 21
t s PAGE 34 ,
.7. PROCEDUMR - NORMAL. ABNORMAL. EMERGENCY AND l RADIOLOGICAL CONTROL i ANSWERS -- LASALLE 1&2 -86/10/27-SPENCER, M.
ANSWER 7.01 (1.00)
- 1. Time
- 2. Rod position
- 3. Coolant temperature
- 4. Reactor period (Also accept time Rad / Chem notified or peron in Rad / Chem notified of criticality.)
REFERENCE LaSalle, LGP-1-1, Rev. 21, Jun. 1986, P. 14 ANSWER 7.02 (1.00)
- c. Unit 1 Remote S/D Panel [0.25]
- b. Unit 2 Remote S/D Panel [0.25]
- c. S/D Panel SCRE is not at [0.25]
- d. S/D Panel S.E. is not at [0.25]
REFERENCE LaSalle, LOA-RX-01, Rev. 3, Nov. 83, P. 3 f
y, , _ , . . . . , ,---.,-__.,...______.,7__, _
-_-.._-,,-y,-,y -
_ _ , . - - . . , _ - , . , + ,,y- e,,_m, - - -- - . - . --,
I .*
PAGE 35
- 7. PROCEDURRR - NORMAL. ABNORMAL. EMERGENCY ANQ
. BADIOLOGICAL CONTROL
-86/10/27-SPENCER, M.
ANSWERS -- LASALLE 1&2 ANSWER 7.03 (2.50)
- a. Unit 1 - place " ADS MANUAL INHIBIT" switches in the Inhibit position [0.5]
Unit 2 - reset " Low Water Level Reset Pushbuttons" at intervals of less than 105 seconds until leads are lifted to disable the function [0.5]
- b. ADS initiation will result in injection of large volumes of relatively cold, unborated water which adds positive reactivity (boron dilution
[0.5], temperature reduction [0.5]). With the reactor shutdown on soluble boron, the resulting power excursion (could result in core damage)[0.5]. '
( Als o ace 9+ 4e follesi,aj is hea of Md M af8od avel b per d re e<&ef[ed :
REFERENCE To g nage J ad tL, aesa%pagveat W ,et;oaposi4,k<. neachW+gof few $dhihb.v. ) gces LaSalle, LOA-NB-011,3 Rdv. 1, Feb. 1986, P. 2 ANSWER 7.04 (1.00) i
- 1. If an LGA caution requires RPV pressure to remain above 57 psig. [0.5]
- 2. Misoperation in automatic is confirmed by at least two independent indications. [0.5]
REFERENCE LaSalle, LGA-GP, General Precautions, Rev. O, Sep. 1985, P. 2 ANSWER 7.05 (1.50)
' 1. Mechanical / electrical checklists have been performed. [0,5)
- 2. Periodic surveillances have been satisfactorily completed. [0.5)
- 3. Surveillances indicated in LGP-1-S1 or Master Outage Checklist as required prior to startup will be completed prior to startup. [0.5]
7Ae ylm/cageweaf j3 capa.ble of (gko ceceyf k 4.3 dove .
gaa40nia3 :ts spea,a Gaetteawermd aa a.4 eeaenaas aLded issku.nedkhia, ecxAeols , and A apa' e/ee/,eicil pra couee ., en/,9 c< sed water ,s{/ukieu,k n A emenyexeyL any,a e d 4d ue xephes As e s3 cle,,, sJsgd e / ,
efyme
-e -paie s p -y.aea nus ,,a,,,u[,4
t PAGE 36
- 7. PROCEDUMR - NORMAL. ABNORMAL. EMERGENCY AND RADIOLOGICAL CONTROL ANSWERS -- LASALLE 1&2 -86/10/27-SPENCER, M.
REFERENCE LaSalle, LGP l-S1, Master Startup Checklist, Rev. 19, P. 1 ANSWER 7.06 (1.00)
- a. 25 rem
- b. 75 rem REFERENCE LaSalle, LRP-1000-1, Rev. 3, Oct. 1985, P. 35 (26 of 69)
ANSWER 7.07 (2.50)
Any FIVE of the following 9 0.5 each
- 1. Generator load decrease with no change in Rx power.
- 2. Decrease in indicated steam flow.
- 3. ADS /SRV Open indication light.
- 4. Temperature increase in downcomer piping.
- 5. Suppression Pool level increase.
- 6. Suppression Pool (and/or Suppression Chamber) temperature increase.
- 7. Increase in Suppression Chamber background radiation.
- 8. Rumbling noise in Suppression Pool area.
- 9. sk //~ / fed //w ,,rd o M de c/orsRe , f ogc/ M /;d//y REFERENCE fo, 37c LaSalle, LOA-NB-02, Stuck Open Safety Relief Valve, Rev. 6, Nov. 1983, P. 1
t s PAGE 37
- 7. PROCEDUnRR - NORMAL. ABNORMAL. EMRRnRNCY AND RADIOLOGICAL CONTROL ANSWERS -- LASALLE 1&2
-86/10/27-SPENCER, M.
ANSWER 7.08 (1.50)
- 1. Suppression Pool Water Temperature above 100 F [0.3]
- 2. Drywell Temperature above 135 F [0.3]
- 3. Drywell Pressure above 1.69 psig (0.3]
- 4. Suppression Pool Water level above +3 in. [0.3]
(also accept 26 ft 10 in from bottom of pool)
- 5. Suppression Pool Water level below -4.5 in (0.3]
(also accept 26 ft 2.5 in from bottom of Suppression Pool)
REFERENCE LaSalle, LGA-03, Primary Containment Control, Rev. 6, Sep. 1986, P. 1 ANSWER 7.09 ( .50)
Alphabetical sequence.
REFERENCE LaSalle, LGA-ATWS-01, ATWS Power Control, Rev. 2, Jan. 1986, P. 7 l
l l .- __. . ._- . - - _ . _ _ _ _ _ - . . . .- _ - -
e .-
PAGE 38
- 7. PROCEDUDM - NORMAL. ABNORMAL. EMERGENCY AND RADIOLOGICAL CONTROL ANSWERS -- LASALLE 1&2 -88/10/27-SPENCER, M.
ANSWER 7.10 (2.00)
Any SIX of the following 9 0.33 each
- 1. Main Turbine Bypass Valves
- 2. RCIC
- 3. RHR (in Steam Condensing Mode) .
- 5. Turbine Driven Reactor Feed Pumps
- 6. Rad Waste Reboiler
- 7. Offgas preheaters
- 8. Gland Seal Steam Reboiler
- 9. Main Condenser Demerating Steam
- 10. Main Steam Line Drains
- 11. RPV Head Vent.
REFERENCE LaSalle, LGA-04, Level Restoration, Rev. 5, Sep. 1985, P. 21 ANSWER 7.11 (1.00)
All startups following a Hot Shutdown [0.5] or Cold Shutdown that was less than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> [0.5].
REFERENCE LaSalle, LGP-1-S2, Minimum Startup Checklist, Rev. 11, P. 2 i
i i
t .-
PAGE 39
- 7. PROCEDURRA - NORMAL. ABNORMAL. EMERGENCY AND
, RADIOLOGICAL CONTROL ANSWERS -- LASALLE 1&2 -86/10/27-SPENCER, M.
ANSWER 7.12 (2.50)
Any of the following:
- a. RPV level below + 12.5 in,
- b. RPV pressure above 1043 psig.
- c. Drywell pressure abovs 1.69 psig.
[3 0 0.5 each]
- 2. A condition which requires reactor scram, AND either of the following:
- a. The reactor is critical
- b. Reactor power cannot be determined
[2 9 0.5 each]
- 3. As directed. [not required for full credit]
REFERENCE LaSalle, LGA-ATWS-01, ATWS Power Control, Rev. 2, Jan. 1986, P. 1
~
l l ANSWER 7.13 (1.00)
To insure that the RHR pump does not dead head the Recirculation Pump.
REFERENCE LaSalle, LOP-RH-07, Shutdown Cooling System Startup and Operation, Rev. 13, Jul. 1986, P. 8 l
t l
! .J
- 7. PROCEDUMR - NORMAL. ABNORMAL. EMERGENCY AND PAGE 49 M DIOLOGICAL CONTROL ANSWERS -- LASALLE 1&2 -86/10/27-SPENCER, M.
l ANSWER 7.14 (1.50) ;
e) Any area accessible to personnel [0.25] in which there exists radiation at such levels that a major portion of the body could receive in any one hour a dose in excess of 5 millirem [0.25], or in any 5 consecutive days a dose in excess of 100 millirem [0.25].
b) Any area accessible to personnel [0.25] in which there exists radiation at such levels that a major portion of the body could receive in any one hour a dose in excess of 100 millirem (0.25].
c) Whole body dose equilavent received in the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period beginning at the start of the worker's shift.
REFERENCE LaSalle, LRP-1000-1, Rev 3, Oct 85, pg 7&8 (3,4 of 69)
ANSWER 7.15 (2.00) i
- a. The individual's supervisor [0.5]
- b. Radiation-Chemistry Supervisor [0.5]
- c. Administrative and Support Services Assistant Superintendent (0.5]
(riso aceept Adm s stnt.'ve Swpeawtencien)
- d. Station Superintendent [0.5]
(ako oec.ept P%+ Muay )
REFERENCE LaSalle, LRP-1000-1, Rev. 3, Oct. 1985, P. 34 (25 of 69)
PAGE 41
- 7. PROCEDURRR - NORMAL. ABNORMAL. EMERGENCY AND
. RADIOLOGICAL CONTROL ANSWERS -- LASALLE 1&2 -86/10/27-SPENCER, M.
ANSWER 7.16 (2.50)
Any FOUR of the following 9 0.625 each
- 1. Steady State thermal power does not excead 50%.
- 2. Drive flow in operating loop does not exceed 30375 GPM (75%).
- 3. MCPR Safety Limit is increased by 0.01 .
- 4. MAPLHGR is reduced to 85% of the old limit.
- 6. APRM flux noise is not greater than 5% peak-to-Peak and core plate d/p noise is not greater than 1 psi peak-to-peak.
- 7. LOS-RR-SRI, Thermal-Hydraulic Stability Data Acquisition, is performed.
t REFERENCE LGP-1-1, Rev. 21, Jun. 1986, P. 7 1
i
1 PAGE 42
- 8. ADMINISTRATIVE PROCEDU N . CONDITIONS. AND LIMITATIONS ANSWERS -- LASALLE 1&2 -86/10/27-SPENCER, M.
ANSWER 8.01 (2.00)
Each blank worth 0.33 pts,
- c. Any temperature ,,
AIO5 Gyf CQN$t-IE (0,lso onc 9 + N/A " o9
- b. Any temperature
- c. Hot shutdown
- d. Shutdown
- c. Shutdown or refuel (both answers required for full credit)
- f. </=140 F REFERENCE LaSalle, Unit 1 Technical Specifications, PP. 1-9 ANSWER 8.02 (1.00)
- a. 25% of the surveillance interval (or 1.25 times the interval). .
- b. 3.25 times the surveillance interval.
REFERENCE LaSalle, Unit 1 Technical Specifications, P. 3/4 0-2 I
i
t s PAGE 43
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS ANSWERS -- LASALLE 1&2
-86/10/27-SPENCER, M.
ANSWER 8.03 (3.00)
- 1. ALL penetrations required to be closed during accident conditions are either closed or capable of being closed by an operable automatic isolation system.
- 2. All hatches and blowout panels are closed and sealed.
- 4. At least one door in each access is closed.
- 5. Secondary Containment pressure is less than or equal to 0.25 in. of vacuum water gauge.
- b. The seda weeka\sm asscci Y u h k eu $ Se W OR CahdNmeHY pese%4db is CPGAA6LE. CV @ aAY each]
REFERENCE LaSalle, Unit 1 Technical Specifications, PP. 1-6 ANSWER 8.04 (3.00)
- a. 25%
- b. 785 psig
- c. 10%
- d. 1.06 l o. 1.07
- f. 1325 psis REFERENCE Grand Gulf, Technical Specification, page 2-1, pp 2.1.1-4
- LaSalle Unit 1 TS pg 2-1 & 2-2 l
l
1 J
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 44 ANSWERS -- LASALLE 1&2 -88/10/27-SPENCER, M. ;
\
l ANSWER 8.05 (2.50)
- e. Illinois Emergency Services and Disaster Agency (ESDA)
Illinois Department of Nuclear Safety (DNS)
Nuclear Regulatory Commissior; [3 0 0.5 each]
- h. OJhed egdoHoWr/classiNkdiins of 16(e eoeuf gjna[c.
r [0. K.l, 0L
.kr'.
Atleasgonceeachhour. [1. 9] Q o,s j REFERENCE LaSalle, GSEP, LZP-1110-1, Rev. 10, Feb 86, P. 23 (ATTACH. F)
ANSWER 8.06 ( .50)
When significant radiation exposure would result.
latso MetPt : Fort hLAR.n censidex.u.Neas)
REFERENCE LaSalle, LAP-900-4, Equipment Out of Service Procedure, Rev. 24, Jan. 1986, P. 3 ANSWER 8.07 M (2,5)
- a. He must be in position to monitor plant parameters [0.75]
and he must be in position to take immediate action if required [0.75]le (Alto acetpf M in hem of one Mswe : midsm\as amount of ->% e speat gt f g
- b. The Center Desk Operator or another licensed person must stand-irf for the Unit Operator .[0,5] The relief operator will be briefed on plant status [0.5] and will brief the "-it 0; crater ;;han hc -
rcturns-[".5].
( REFERENCE I
LaSalle, LAP-1600-2, Rev. 29, Jul. 1986, P. 12 l
l i
l l
l
/ /
PAGE 45
- 8. ADMINISTRATIVE PROCEDUDE. CONDITIONS. AND LIMITATIONS ANSWERS -- LASALLE 1&2 -86/10/27-SPENCER, M.
ANSWER 8.08 (2.25)
[3 0 0.75 each]
- 1. The intent of the original procedure is not altered.
- 2. The change is approved by two members of the plant management, at least one of whom holds an SROs license on the unit affected.
- 3. The change is documented, reviewed by the Onsite Review and Investigative Function and approved by the Station Superintendent within 14 days of implementation.
REFERENCE LaSalle, Unit 1 Technical Specifications, PP. 6-17 ANSWER 8.09 (1.50)
- 1. Safety evaluation must be completed and reviewed by two SRO's, one of which must have an engineering degree or equivalent.
- 2. Change must be authorized by the Shift Engineer.
- 3. Onsite Review should be conducted promptly following the change (normally the next working day).
REFERENCE LaSalle, LAP-240-6, Temporary System Changes, Rev. 11, May 1986, P. 3 ANSWER 8.10 ( .75)
Shift Engineer (also accept Acting Station Director)
REFERENCE LaSalle, GSEP, LZP-1110-1, Rev. 10, Feb. 1986, P. 3
't J ADMINISTRATIVE PROCEDURKS. CONDITIONS. AND LIMITATIONS PAGE 46 8.
ANSWERS -- LASALLE 1&2 -86/10/27-SPENCER, M.
ANSWER 8.11 (1.50)
- 1. Transportation Accident
- 2. Unusual Event
- 3. Alert
- 4. Site Emergency
- 5. General Emergency
- 6. Recovery REFERENCE LaSalle, GSEP, LZP-1110-1, Rev. 10, Feb. 1986, P. 5 SDLP-22/GSEP, Rev 1, Sep 85, P. 32 ANSWER 8.12 (1.00) e REFERENCE GGNS: TS DEFINITION 1.6 LaSalle Unit 2 TS DEFINITION 1-1 ANSWER 8.13 (2.00)
- 1. If radiation levels and the pressure / temperature in the primary containment are stable. .
- 2. If there is adequate core cooling as indicated by stable reactor coolant system pressures, temperatures, and levels.
- 3. If these parameters are following expected trends.
(only answers 1 and 2 are required for full credit)
- .)
PAGE 47
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS
-86/10/27-SPENCER, M.
ANSWERS -- LASALLE 1&2 REFERENCE LaSalle, LAP-1600-2, Conduct of Operations, Rev 29, Jul 86, pg. 6 ANSWER 8.14 (1.00)
Because the dissolved oxygen content of the reactor coolant is typically higher during low steaming rates (e.g. Startup or Hot (1.0)
Standby)
REFERENCE EIH: U1 TS's, 3.6-6 GGNS: TS 3/4.4.4, Table 3.4.4-1 GGNS: TS 3/4.4.4, Table 3.4.4-1 LaSalle Unit i TS B 3/4 4-2
., a TEST CROSS REFERENCE PAGE 1
' QUESTION VALUE REFERENCE
- ~ ~
C5.01 1.00 HEE 0001024 05.02 1.50 HEE 0001027 C5.03 1.50 HEE 0001030 05.04 1.50 HEE 0001022 05.05 1.00 HEE 0001021 C5.06 2.00 HEE 0001029 05.97 .50 HEE 0001032 05.08 1.00 HEE 0001033 C5.09 1.00 HEE 0001028 05.10 .50 HEE 0001025 05.11 2.00 HEE 0001019 05.12 1.00 HEE 0001023 05.13 3.00 HEE 0001017 05.14 1.00 HEE 0001031 05.15 1.50 HEE 0001026 05.16 2.00 HEE 0001020 05.17 3.00 HEE 0001018 25.00 06.01 1.00 HEE 0001045 C6.02 1.00 HEE 0001044 06.03 2.00 HEE 0001042 06.04 2.00 HEE 0001043 06.05 1.50 HEE 0001035 06.06 3.00 2(HEE 0001034 C6.07 2.00 HEE 0001046 06.08 1.00 HEE 0001048 06.09 1.00 HEE 0001039 06.10 1.00 HEE 0001040 06.11 1.50 HEE 0001038 06.12 1.00 HEE 0001041 06.13 2.00 HEE 0001037 06.14 2.00 HEE 0001047 06.15 3.00 HEE 0001036 s.ccah 07.01 1.00 HEE 0001050 07.02 1.00 HEE 0001052 07.03 2.50 HEE 0001053 07.04 1.00 HEE 0001054 07.05 1.50 HEE 0001051 07.06 1.00 HEE 0001063 07.07 2.50 HEE 0001064 07.08 1.50 HEE 0001058 07.09 .50 HEE 0001056 07.10 2.00 HEE 0001059 07.11 1.00 HEE 0001057 07.12 2.50 HEE 0001055
TEST CROSS REFERENCE PAGE 2
' QUESTION VALUE REFERENCE 07.13 1.00 HEE 0001060 07.14 1. 50-~ HEE 0001065 07.15 2.00 HEE 0001061 07.16 2.50 HEE 0001049 25.00 08.01 2.00 HEE 0001068 08.02 1.00 HEE 0001069 C8.03 3.00 HEE 0001066 C8.04 3.00 HEE 0001077 C8.05 2.50- HEE 0001067 C8.06 .50 HEE 0001072 08.07 -h00A.5 HEE 0001070 08.08 2.25 HEE 0001073 C8.09 1.50 HEE 0001071 C8.10 .75 HEE 0001075 08.11 1.50 HEE 0001074 08.12 1.00 HEE 0001078 08.13 2.00 HEE 0001062 08.14 1.00 HEE 0001076 25.
_ _ _ _ '_'" eW. f He-00 99. o 1
l l
- . - . _ _ - _ . . - . _ . - . - . _ _ . _ . - - _ _ _ _ - . _ _ _ _ _ - _ - _ . -_ .