ML20207T483
| ML20207T483 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 02/24/1987 |
| From: | Huenefeld J, Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20207T477 | List: |
| References | |
| 50-302-OL-86-02, 50-302-OL-86-2, NUDOCS 8703240062 | |
| Download: ML20207T483 (73) | |
Text
{{#Wiki_filter:. _ _ . _ _ __ .. - _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _____ _ _ m . l-ENCLOSURE 1 EXAMINATION REPORT 302/0L-86-02 Facility Licensee: Florida Power Corporation Facility Name: Crystal River Nuclear Plant
~ Facility Docket No.: 50-302 Written and oral examinations were administered at Crystal River Training Center near Crystal River, Florida. Oral examinations were administered at Crystal River
- Power Plant near Red Level, Florida Chief Examiner: ~
b J. C. Huenef d# f Date Signed - Approved by: vh AN!AS.t # kav!P7 g ohnf. Munro M ectJ4n Chief Date Signed-Summary: Examinations on December 17-19, 1986 Written examinations were administered to four SR0 upgrade candidates, all of whom passed. Oral examinations were administered to four SR0 upgrade candidates, all of whom passed. Based on the results described'above, four of four SR0s passed. 8703240062 870303 PDR ADOCK 05000302 V PDR
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REPORT DETAILS
' Facility Employees Contacted:
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Johnie Smith Paul McKee Larry Kelley 2.- Examiners: James Huenefeld, Chief Examiner 3 .- Examination Review Meeting At the conclusion of the . written examinations, the examiner provided i Johnie Smith, with a copy of the written examination and answer key for l review. The comments made by the facility reviewers are included as l Enclosure 3 to this report, and the NRC. Resolutions to these comments are.
-listed below.
Question 5.05 - While the utility did not comment, it was discovered on review prior to grading that the answer is " b", vice " a". The answer key was changed accordingly. Question 5.09 - Disagree. The utility comment does not contend that
'the question is flawed, or that the answer is incorrect, but rather that the reference has not been approved. There is no requirement that material used for operator training be approved by the NRC for such use. The generic reference was used for this answer only because the AT0G documentation provided by the facility for examination' preparation was not sufficient. No change required.
Question 5.10 - See 5.09 Question 5.22 - Agree. Prior to future use, the question will be reviewed to assure that adequate information to identify the plant condition is included. The answer key was expanded slightly to ensure that all correct answers were included. Specifically, the following changes (underlined) were made to the answer key: increasing PZR level decreasing or stable RCS pressure relatively stable or decreasing RCS temperature Question 6.02 - Agree with the comment, but not the recommendation. The stipulation that setpoints are not required was added to the answer key.
m 2 b I' '* Question 6210.- Disagree. - The question as phrased... " behavior. of a
- Geiger-Mueller detector"... is correct, and elicits only one~ answer.
In addition, the information provided with the facility comment does not support thel contention that "all instruments in use at CR-3' are designed to prevent this occurrence (saturation)." No change required.~ Question' 6.12 Disagree. . The question asks for the five-diesel start permissives. They are listed in the training material as "permissives". While many other conditions, some of which are shown on the submitted logic, must exist for auto start of- the. diesel, the question clearly asks for the "permissives" and does not require recitation of- all conditions. No change required. Question 6.13 - Agreed. The answer key was changed.
. Question 6.17 - No utility coment. _ During the examination a typo-graphical error was discovered and corrected by announcement to the candidates.
Question 7.03 - Agreed. The answer key was changed.
*. Question 7.04 - Agreed. The referenced material states "...the operator should make every attempt to regain feedwater to at least one steam generator. This includes ... service water." However, the AT0G prepared for Crystal River at Appendix C, page C-33, supplied with the utility comment specifically states "... HPI cooling is preferable to the use of non-condensate grade feedwater." Prior to utilizing ~ the referenced training material (RDT 3-4) .the training staff should correct this error. Prior to the next examination at Crystal River, the- training staff should assure that the document they referenced here, " Crystal River U-3, Abnormal Transient Operating Guidelines (ATOG)" is supplied to the examiners for preparation of the examina-tion. The examination and answer key were changed to delete this question.
Question 7.13 - Agreed. ROT 4-15 on the EFIC system supports the utility comment and recommendation as well as other references. The answer key was changed accordingly. Note, however, that RDT 3-3 pg. 8 is in error and should be corrected prior to future use. Question 7.15 - Agreed. The answer key will be changed accordingly. Note that OP 204 pg. 5 states that pressure control selection should be performed per OP 501 section 7.2. OP 501 revision 09 has no section 7.2. This should be corrected. Question 8.03 - Agreed. 10 CFR 20.402 reporting requirements will be added to the answer key. Question 8.06 - Disagree. Operations must supply the leader. The leader must know how many people and from where they are supplied. Both are required to assure there are a sufficient number of qualified
1 . . ' s f-
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d members present. Also, the requirement is that the people come from
. maintenance; they are not. required to come from a subset of maintenance.
(i.e. , . the question ' states department). . Note that ,NUREG 1122,. at PWG 19, gives this an importance -rating of 4.2_ for SR0s. 'No change required.
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. Question 8.10 - Agreed. The answer _ key was changed accordingly.
*- Question 8.19 - Agree with the comment, but disagree .with the recom-mendation. It is our. policy to accept the wording of the latest
. procedure revision. The answer key was changed accordingly.
Generally, philosophical questions submitted by the utility are not addressed. However, in -this case, the general coment raises a question regarding references for theory questions. The documents cited in the utility comment (other than training material utilized at other B&W plants) are all official B&W documents transmitted to the NRC and the utility. ~While-it is true that there are areas where the designs differ, thus invalidating specific state-ments, it is also true that the theoretical principles being elucidated are generic. It is only with respect to these principles that selected category 5 questions reference these document.s. If, of course, a particular question's answer is_ incorrect due to design differences, or incorrect for any other reason, the answer will be changed. This material ~ will be requested for ' future examinations. Theory questions that refer only to training material utilized at other B&W plants will be minimized; every attempt will be made to find the plant specific training material which teaches the concept.
- 4. Exit Meeting At the conclusion of the site visit, the examiner met with the above representatives of the plant staff and the resident inspector to discuss the results of the examinations. The following general observations were made:
- a. The candidates could not discuss specific controls for shutdown reactivity, and did not know about the technical specification requirement to maintain Keff _ 0.99 while in the shutdown modes.
- b. The candidates were not familiar with the phenomenon of saturated repressurization.
- c. The candidates were uncertain whether or not to apply the emergency cooldown limits during an OTSG tube rupture while. in natural circula-tion.
- d. There was considerable confusion in integrating the several APs/EPs needed to handle a given event. This seemed to be in spite of obvious familiarity on the part of the candidates.
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4 The cooperation given to the examiners and the effort to ensure an atmosphere in the control room conducive to oral examinations was also noted and appreciated. The licensee did not identify as proprietary any of the material provided to-or reviewed by the examiners. I l l
r g y, D M A STER ob U.S. NUCLEAR ~ REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION Facility: CRYSTAL RIVER Reactor Type: PWR-8&W177 Date Administered: 86/12/17 Examiner: HUENEFELD, J. Candidate: ANSWER KEY INSTRUCTIONS TO CANDIDATE: Use separate paper for the answers. Write answers on one side only. Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts. Category % of Candidate's % of Value Total Score Cat. Value category 34 ~ 27.d4 5. Theory of Nuclear Power Plant Operation, Fluids and Thermodynamics J 31 25 / 6. Plant System Design, Control and Instrumentation
.M. E' 3 2( 23.M 7. Procedures - Normal, Abnormal, Emergency, and Radiological Control 7
29 23.M 8. Administrative l Procedures, Conditions, and Limitations XI , TOTALS I 4 i l Final Grade ,% l l All work done on this examination is my own; I have neither given nor l received aid. l l Candidate's Signature l
n, y o o NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
- 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
- 2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
- 3. Use black ink or dark pencil only'to facilitate legible reproductions.
- 4. Print your name in the blank provided on the cover sheet of the examination.
- 5. Fill in the date on the cover sheet of the examination (if necessary).
- 6. Use only the paper provided for answers.
- 7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
- 8. Consecutively number each answer sheet, write "End of Category " as appropriate, start each category on a new page, write only one sTde of the paper, and write "Last Page" on the Tast answer sheet.
- 9. Number each answer as to category and number, for example,1.4, 6.3.
- 10. Skip at least three lines between each answer.
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
- 12. Use abbreviations only if they are commenly used in facility literature.
- 13. The point value for each question is indicated in parenthesis 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
- QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
- 16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given 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 your examination as follows:
(1)Examquestionsontop. (2) Exam aids - figures, tables, etc. (3) Answer pages including figures which are a part of the answer.
- b. Turn in your copy of the examination and all pages used to answer the examination questions.
- c. Turn in all scrap paper and the balance of 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 NUCLEAR POWER PLANT OPERATION, FLUIDS.
AND THERMODYNAMICS -PAGE 1 QUESTION 5.01' (2.00) DESCRIBE the response of neutron power for each of the following two (2) situations: (2.0) 1. The reactor is critical at 10**-8 amps and control rods are inserted by 1% RI. 2. The reactor-is at 100% FP with the ICS in " Track," and control rods are inserted by 1% Rod Index. (Note: a detailed description of ICS ops is NOT required.) i , QUESTION 5.02 (2.00) MAKE a sketch of'a graph of neutron countrate versus time (linear scale) assuming a short duration rod pull is made with the reactor ! critical below the point of adding heat. NEGLECT feedback effects. ! EMPHASIZE on your graph the effects of prompt and delayed neutrons. (2.0) QUESTION 5.03 (0,5) Inserted rod worth with increasing temperature during a plant heatup. (0.5)
- a. increases
- b. decreases
- c. remains constant QUESTION 5.04 (0,5)
The negative moderator coefficient of reactivity in magnitude as rods are withdrawn. (0.5)
. increases
. decreases remains constant
(***** CATEGORY 05CONTINUEDONNEXTPAGE*****) d
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS,- PAGE 2
.AND THERMODYNAMICS
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QUESTION 5.05 (0.5) The negative moderator coefficient of reactivity in magnitude as boron is increased. (0.5)
- a. increases
- b. decreases
- c. remains constant QUESTION 5.06 (1.50)
AFW can have a proportionately larger cooling effect on the RCS for the same flow rate than MFW. . STATE three (3) reasons for this.
-(Assume forced convection in the RCS.) (1.5)
QUESTION 5.07 (1.00) WHICH one-(1) of the following would NOT be an acceptable means for elimination of a void that has fonned in an idle loop during a single loop natural circulation cooldown? ' (1.0) a.) venting through the hot leg vents
- b. bumping an RCP
- c. establishing full HPI flow until the void collapses
- d. allowing the void to condense through ambient heat losses QUESTION 5.08 (2.00)
The reactor is shut down by 6% delta k/k with a source neutron countrate indication of 50 cps. Rods are withdrawn to raise the source range indication to 285 cps. WHAT is the value of reactivity when counts are 285 cps? (SHOWallwork) (2.0) QUESTION 5.09 (1.00) STATE the two (2) most positive indications in a subcooled RCS that natural circulation has been lost. - (1.0) (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, PAGE 3 AND THERMODYNAMICS QUESTION 5.10 (1.50)
With the RCS saturated, WHAT is the best indication of a loss of natural circulation flow or interruption of boiler condenser cooling (reflux boiling)? (1.5) QUESTION 5.11 (1.00) During an overcooling event, the RCS will remain subcooled unless: (SELECTone.) (1.0)
. RCPs are lost.
. emergency feedwater actuates.
. the overcooling is not rapidly terminated.
. the pressurizer empties.
QUESTION 5.12 (1.50) During a plant heatup, OP-202, a vacuum is drawn in the Steam Generators. This action promotes even warming of the Steam Generator shell. EXPLAIN why this occurs. (1.5) QUESTION 5.13 (1.00) The water occurs during hammer WHAT rangeproblem during heatup of feedwater at Crystal temperatures? River typically)(1.0) (SELECTone. i a. below 180 deg F
- b. 210 to 230 deg F
- c. 230 to 300 deg F
- d. above 300 deg F i
f (***** CATEGORY 05CONTINUEDONNEXTPAGE*****) l l
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, PAGE 4 AND THERM 00YNAMICS
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QUESTION 5.14 (1.00) There are three (3) regions of heat transfer in the steam generator.
- a. NAME the region that provides the greatest heat flux. (0.5)
- b. NAME the region that expands the largest amount due to a power level increase. (0,5)
QUESTION 5.15 (2.00)
- a. Approximately HOW MUCH of the RCS flow will bypass the core as it flows through the vessel? (SELECT one.) (1.0)
- a. 1%
- b. 6%
- c. 15%
- d. 19%
- b. Even though the above amount of flow bypasses the core, it still picks up heat energy. DESCRIBE the mechanism by which this occurs. (1.0)
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f QUESTION 5.16 (3.00)
- a. WHAT special complication is associated with conducting a natural circulation cooldown with only one (1) OTSG and with C OTSG full of water? (1.5)
- b. WHAT special complication it associated with conducting a forced cooldown on one (1) OTSG and with the other OTSG dry and depressurized? (1.5)
QUESTION 5.17 (2.00) IS the boiler-condenser or reflux boiling method of RCS circulation ! susceptible to blockage by buildup of non-condensables? EXPLAIN. (2.0) (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, PAGE 5 AND THERMODYNAMICS QUESTION 5.18 (0.50)
ANSWER TRUE or FALSE. After a reactor trip early in life, emptying a full makeup tank of deborated water into the reactor coolant system will result in a reactivity addition of less than 1.0% delta k/k. (0.5) QUESTION 5.19 (1.00) WHAT does the term " gray" mean in gray APSRs? (1.0) QUESTION 5.20 (0.50). ANSWER TRUE or FALSE. A one (1) decade per minute (DPM) startup rate implies that reactor power will change by one-half (1/2) of one decade (a factor of 5) in thirty (30) seconds. (0.5) QUESTION 5.21 (2.00) Reactor power drops to about 7.5% instantly after a reactor trip. Given that the delayed neutron fraction is less than 1%, WHY doesn't reactor power drop instantly to less than 1%? Address the importance of delayed neutrons and subcritical reactivity in your answer. (2.0) QUESTION 5.22 (1.50) STATE three (3) RCS symptoms that, when taken separately, are indeterminate but when observed coincidently, indicate that pressurizer level is no longer a good indicator of RCS inventory. (1.5) (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, PAGE 6 AND THERMODYNAMICS QUESTION 5.23 (1.50)
During power operations HOW can you determine, at a glance, that the amount of positive reactivity that will be inserted by doppler and moderator temperature during a reactor trip will NOT exceed the amount of negative reactivity that will be inserted by control rods? (1.5) QUESTION 5.24 (3.00) WHICH contains the greater enthalpy, a cubic foot of saturated water at 70 degrees, or a cubic foot of saturated water vapor at 70 degrees? (Use the steam tables provided to prove your answer.) (3.0) (***** END OF CATEGORY 05 *****)
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- 6. PLANT SYSTEMS DESIGN, CONTROL, AND PAGE 7 ,
INSTRUMENTATION QUESTION 6.01 (3.00) WHY is it-NOT necessary for the ICS to program OTSG water level versus power in order to achieve equilibrium conditions? (3.0) QUESTION 6.02 (1.50) STATE six (6) of the seven (7) RCP start permissives as stated in OP 302. Setpoints are NOT required. (1.5) QUESTION 6.03 (0.50) ANSWER TRUE or FALSE. The self-powered neutron detectors are susceptible to thermionic effects causing them to read very high when exposed to very high temperatures. (0.5) QUESTION 6.04 (0.50) ANSWER TRUE or FALSE. The incore instrumentation is compensated for gamma radiation. (0.5) QUESTION 6.05 (0.50) ANSWER TRUE or FALSE. The SPDS displays the average of the five (5) highest of the 52 incore thermocouples. (0.5) (***** CATEGORY 06 CONTINUE 0 ON NEXT PAGE *****)
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- 6. PLANT-SYSTEMS DESIGN, CONTROL, AND PAGE 8 INSTRUMENTATION QUESTION 6.06 (0.50) .,
' ANSWER TRUE or FALSE..
Essentially complete gamma compensation is achieved in the Intermediate Range nuclear instrumentation by surrounding the detectors by four (4) inches of lead. ' (0.5) QUESTION 6.07 (0.50) ANSWER TRUE or FALSE. Voiding in the downcomer. region of the reactor vessel can cause a significant increase in nuclear instrument indication. (0.5) QUESTION 6.08 (1.50)
-Given a loss of forced RCS flow, the T(c) indication may decrease even though natural circulation of the RC does NOT exist. WHY? (1.5)
QUESTION 6.09 (2.00)
- a. WHAT effect does elevated reactor building temperature have on the indicated level of a level indicator that has a wet reference leg? EXPLAIN. (One or two sentences should be sufficient.) (1.5)
- b. IS this effect observable on control room instrumentation?
(YESorNO) (0.5) (***** CATEGORY 06CONTINUEDONNEXTPAGE*****)
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- 6. PLANT SYSTEMS DESIGN, CONTROL, AND PAGE 9 INSTRUMENTATION 1
QUESTION 6.10 (1.00) WHICH one (1) of the following phases best describes the behavior of a geiger-mueller detector when exposed to extremely high INCREASING radiation fields? The meter will: (SELECTone.) (1.0)
- a. saturate at full scale and stay pegged.
- b. saturate at some maximum value and remain at that value.
- c. saturate at some maximum value and then slowly drop off,
- d. saturate at some maximum value and drop abruptly to zero.
QUESTION 6.11 (2.00) STATE four (4) of the six (6) loads serviced by the Decay Heat closed Cycle Cooling system. (Multiple pumps from a single system are to be counted as a single load. Pumps and the motors that drive them are to be considered as a single load.) (2.0) QUESTION 6.12 (3.00)
- a. STATE the three (3) types of emergency diesel auto-start signals. (1.5)
- b. STATE the five (5) permissives that must be satisfied in order for an auto-start to occur on signal. (1.5)
QUESTION 6.13 (1.00) WHERE do the DC system and the SW system surge tanks vent and relieve to? (1.0) QUESTION 6.14 (2.50)
- a. WHY is it more important to check for leakage from the SW system into the RB than into other locations? (1.0)
- b. STATE three (3) different SW supplied components that are instrumented to monitor for leakage from 'the SW system to the RB. (1.5)
(***** CATEGORY 06CONTINUEDONNEXTPAGE*****)
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- 6. PLANT $YSTEMS DESIGN, CONTROL, AND PAGE 10 INSTRUMENTATION QUESTION 6.15 (1.50)
WHAT is the safety significance of maintaining the SW system pressurized? (1.5) QUESTION 6.16 (1.00) WHAT are the two (2) sources of water to the Industrial Cooler (CI) system? (1.0) QUESTION 6.17 (1.50) EXPLAIN WHY the Nuclear Instrumentation tends to become less conservative as power increases from)( FP to 100% FP. (1.5) if$ QUESTION 6.18 (1.50) At WHAT Intermediate Range (IR) level is the Source Range instrument deenergized? DOES it take one or both of the IR instruments to cause this to happen? (1.5)' QUESTION 6.19 (1.50) WHY is it important that the NNI X or Y +/- 24 VDC system be totally deenergized by the power supply monitor should a degraded voltage condition develop? (1.5) QUESTION 6.20 (1.50) HOW is the RCS flow signal to the ICS derived? (1.5) i (***** CATEGORY 06CONTINUEDONNEXTPAGE*****) 1
O O PLANT SYSTEMS DESIGN, CONTROL, AND PAGE 11 6. INSTRUMENTATION _ QUESTION 6.21 (1.50) GIVE a general, one (1) or two (2) sentence, description of the major difference between a fixed water spray system, a wet pipe sprinkler system, and a pre-action sprinkler system. Also, GIVE cne (1) example application of each type. (1.5) QUESTION 6.22 (1.00) WHAT is the minimum volume of fuel oil that a day tank must have in order for the associated DG to be considered OPERABLE? (1.0) (SELECT one.)
. 100 gallons
. 200 gallons
. 300 gallons
. 400 gallons
(***** END OF CAiu. 06*****)
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- 7. PROCEDURES - NORMAL, A8 NORMAL, EMERGENCY PAGE.12 AND RADIOLOGICAL CONTROL
-QUESTION 7.01 (1.50)
WHY is it necessary that all of the hydrazine be removed from the RCS prior to placing a makeup desineralizer in service? (1.5)- QUESTION'7.02 (0.50) ANSWER TRUE or FALSE. During a plant heatup (OP-202), pressurizer level is to be controlled such that there is an out-surge from the surge line nozzle. (0.5) QUESTION 7.03 (1.50) According to ATOG there are three (3) specific instances during a loss of forced RCS flow when EFW should be " turned on full." STATE one (1) of these three-(3). (1.5) 4 (0.50) ANSWER TRUE or FAL Upon loss of all feedwater, t of service water as a source of emergency feedwater is preferable to e ishing HPI cooling. (0.5) Or9ffMk. f W NO6ht.y vi p ~[ l (***** CATEGORY 07CONTINUEDONNEXTPAGE*****)
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- 7. PROCEDURES - NORMAL. ABNORMAL, EMERGENCY PAGE 13 AND RADIOLOGICAL CONTROL QUESTION 7.05 (1.00)
WHICH one (1) of the following statements best describes why two (2) HPI pumps should be used for HPI/PORV cooling? (SELECTone.) (1.0) (a.) One HPI pump is NOT adequate for core heat removal. (b.) One HPI pump CANNOT keep up with the leak out of the PORV, (c.) The time to match core decay heat is much shorter for two pumps than for one ' ump. (d.) Two pumps result in balanced heat removal. , QUESTION 7.06 (0.50) ANSWER TRUE or FALSE. Following inadequate core cooling AND hydrogen production, only the PZR vent may be used for RCS depressurization. (0.5) QUESTION 7.07 (1.00)
- a. WHAT is the maximum allowable flow for one (1) HPI pump? (0.5)
- b. At WHAT RCS pressure will this runout condition occur? (0.5)
QUESTION 7.08 (1.00) According to ATOG, under WHICH one (1) of the following conditions should a reactor coolant pump NOT be started or bumped? (SELECT one.) (1,0)
. RCS subcooled with natural circulation existing.
. RCS subcooled with no natural circulation.
. RCS saturated with natural circulation existing (HPI on).
. RCS saturated with no natural circulation existing (HPI on).
1 (***** CATEGORY 07CONTINUEDONNEXTPAGE*****) l l
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY PAGE 14 AND RADIOLOGICAL CONTROL QUESTION 7.09 (2.00)
Within 24 hours after a LOCA, action must be taken to preclude the possibility of boron precipitation in the reactor vessel. GIVE a one (1) sentence description of each of the two (2) methods that are to be used at Crystal River. (2.0) QUESTION 7.10 (0.50) WHAT parameter are you procedurally directed to observe to independently verify that generator output voltage is, in fact, 22 kV7 (0.5) QUESTION 7.11 (1.00) WHICH one of the following conditions does NOT constitute grounds for an immediate manual reactor trip: 'SELECTone) (1.0) (a.) Two (2) main steam isolation valves on different loops have been inadvertently shut. (b.) Reactor' power is 10% and main feedwater is lost. (c.) Pressurizer level is 295 inches and decreasing. (d.) All four safety groups drop into the core while at 100% power. QUESTION 7.12 (1.50) Subcooling margin is lost as a result of a small break LOCA. STATE the three (3) major actions, automatic or manual, that must be initiated. (1.5) (***** CATEGORY 07CONTINUEDONNEXTPAGE*****)
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- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY PAGE 15 AND RADIOLOGICAL CONTROL QUESTION 7.13 (1.50)
Various procedures (APs/EPs) require that you verify proper OTSG level following actuation of the EFIC system. LIST the correct EFIC level for each of the following conditions. (1.5)
- a. RCPs running... reactor tripped
- b. no RCPs running... adequate subcooling margin
- c. no RCPs running... inadequate subcooling margin OR less than two (2) HPI pumps QUESTION 7.14 (1.50)
LIST the Imediate Actions for AP-555 (Continuous Control Rod Withdrawal) . (1.5) QUESTION 7.15 (1.00) While 1A must operating at 95% be removed from power, service. you WHAT(astwo SS00)(are
- 2) actionsinformed are you that RCP required to ensure have been taken PRIOR to securing the pump per the limits and precautions of OP-204 (Power Operation)? (1.0)
QUESTION 7.16 (2.00) During full power operation you discover that the PORV block valve is open and will not respond to the control switch on the Main Control Board. Per Technical Specifications WHAT actions must be taken? (2.0) QUESTION 7.17 (3.00) WHAT are the nine (9) Immediate Actions for AP-380 (Engineered SafeguardsSystemActuation)? (3.0) (***** CATEGORY 07CONTINUEDONNEXTPAGE*****) l
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- 7. PROCEDURES - NORMAL. ABNORMAL, EMERGENCY PAGE 16 '
AND RADIOLOGICAL CONTROL _ s QUESTION 7.18 (2.00)
- a. During an approach to criticality, you note that RCS Tave has decreased to 521 deg F. -WHAT actions are required per Technical Specifications? (1.0)
- b. STATE two (2) of the four (4) bases for the minimum temperature for criticality. (1.0)
QUESTION 7.19 (3.00) LIST the nine (9) "Imediate" actions (remedial actions not s required) of the Reactor Protection System Actuation s procedure (AP-580). (3.0) QUESTION 7.20 (2.50)
- a. DEFINE shutdown margin. INCLUDE any assumptions about axial power shaping rod or control rod positioning. (1.5)
- b. Duringaplantheatup, modes 5,4,and3,thereactormubtbe shutdown by >/= 1% delta k/k. When group 1 rods are withdrawn for the heatup, does their inserted worth count as part of' that 1% delta k/k? WHY? (1.0) t
(***** END OF CATEGORY 07 *****)
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- 8. $DMINISTRATIVE PROCEDURES, CONDITIONS, . PAGE 17 AND LIMITATIONS
~_
QUESTION 8.01 (0.50) ANSWER TRUE or FALSE. Compliance with Normal Operating Procedures is required by the Code of Federal Regulations. 3 .. (0.5)- QUESTION 8.02 (0.50) AN,SWER TRUE or FALSE. The Cott of Federal Regulations allows a licensed senior operator to take reasonable action that departs from a Technical Specification in an emergency. (0.5) ( QUESTION 8.03 (2.50) STATE'five (5) reportable events that according to the Code of Federal Regulations must be reported to the NRC either immediately or within one (1) hour. (2.5) t QUESTION 8.04 (1.50) WHAT are the Nuclear Instrumentation requirements in Modes 3-5 if the CRDM breakers are closed and the rods are capable of being withdrawn (i.e., how many power range, intermediate range, and source range channels are required)? (1.5) QUESTION 8.05 (1.00) Assume that a MVP breaker fails to shut. An initial investigation is undertaken via work request to determine the cause of failure. No cause for the failure is evident. WHAT are the minimum testing requirements for reestablishing the operability of this breaker? (1.0)
~
- s .,
(***** MTEGORY 08 CONTINUED ON NEXT PAGE *****)
)A s 4 i 0 , - - - - ,s - ~ , ^4 .Ls -- - - , , - , , . _ _ . , , , . - - - , . ,,-,--- - ,, - m.--- . - - - , -ng,y.,,--,-,wa- - , , . , , , . - - - - , ~ ,-- -----s,-
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~ .
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, PAGE 19 AND LIMITATIONS
~.__
QUESTION 8.10 (1.00) The authorit WHAT two (2)y to startup and return to power operation rests.with positions? (1.0) QUESTION 8.11 (2.00) AI-500 requires the Nuclear Shift Supervisor to perform five (5) functions in the event of a reactor trip or plant shutdown. LIST the five (5) functions. (2.0) QUESTION 8.12 (2.00) According to the emergency plan implementing procedure, there are seven (7) Emergency Teams that may be activated by the Emergency Coordinator during an emergency. LIST four (4) of these Emergency Teams. (2.0) QUESTION 8.13 (1.50) WHAT are the three (3) categories of decisions that the Emergency Coordinator cannot delegate the responsibility for? (1.5) QUESTION 8.14 (0.50) If generating complex personnel are to evacuate the site, HOW many suitable evacuation routes are available from the site? (0.5) QUESTION 8.15 (1.50) In any case where a " General Emergency" has been declared, WHAT is the minimum protective action recommendation that should be made regarding the welfare of the surrounding population? (1.5) (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
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- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, PAGE 20 AND LIMITATIONS QUESTION 8.16 (1.00)
The on-shift Nuclear Shift Supervisor has the authority of the Emergency Coordinator until relieved by the designated Emergency Coordinator. WHO is the designated Emergency Coordinator? (1.0) QUESTION 8.17 (1.00) According to EM-201, Duties of an Individual Who Discovers an Emergency, WHAT is the first action that an individual who discovers an emergency condition locally out in the plant should take? (1.0) QUESTION 8.18 (2.00) STATE the five (5) conditions required for containment integrity to exist. (2.0) QUESTION 8.19 (1.50) CP-115 (In Plant Equipment Clearance and Switching Orders) requires PRC approval prior to issuance of a clearance which cannot meet the double valve protection guidelines.
- a. WHAT are the double valve protection guidelines? (1.0)
- b. WHEN may the clearance Authority approve clearances which do not meet these guidelines without PRC approval? (0.5)
QUESTION 8.20 (1.00) All steam systems at CR-3 have been administratively established as RCAs. GIVE two (2) other examples of systems that are normally not contaminated, yet have been administratively established as RCAs. (1.0) (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
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- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, PAGE 21 AND LIMITATIONS
' QUESTION 8.21 (1.00)
Using one (1) or two (2) sentences, STATE the main difference between an RWP and an SRWP. (1.0) QUESTION 8.22 (1.00) l WHAT minimum radiation field, in mr/hr, constitutes a "High Radiation Area"? (1.0) QUESTION 8.23 (1.00) WHAT significant personnel safety hazard, other than radiological, is associated with using tools in the spent fuel pool area, and WHAT must be done to eliminate it? (1.0) QUESTION 8.24 (1.00) WHAT restrictions apply to raising an irradiated fuel asssembly
'with the new fuel elevator? (1.0)
I !~ (***** END OF CATEGORY 08 *****) (************* END OF EXAMINATION *************)
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, PAGE 22 AND THERMODYNAMICS ANSWERS -- CRYSTAL RIVER -86/12/15-HUENEFELD, J.
ANSWER 5.01 (2.00)
- 1. The reactor will go, and remain, subcritical and power will decrease to the subcritical multiplication level corresponding to the amount of negative reactivity inserted. [+1.0]
- 2. Reactor power will decrease and level off at a new critical power level, with the amount of the decrease being determined by the power deficit. [+1.0]
REFERENCE
- 1. Crystal River: Nuclear Energy Training, Module 3, " Reactor Operation," pp. 13.5-2 and 13.5-4.
1
~
\. __ _ _ - _ _ _ _ _ - _ _
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, PAGE 23 AND THERMODYNAMICS ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J.
ANSWER 5.02 (2.00) Countrate [+1.0] Prompt response to g Constant period dominated rod withdrawal k by delayed neutrons
, }
[+1.0] _ Time REFERENCE
- 1. Oconee: Fundamentals of Nuclear Reactor Engineering, Duke Power Co., pp. 101 through 104.
- 2. Crystal River: Nuclear Energy Training, Module 3, Unit 6 (general).
ANSWER 5.03 (0.5) (a.) [+0.5] REFERENCE
- 1. Crystal River: Nuclear Energy Training, Module 3, Figure 16-6.
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, PAGE 24 AND THERMODYNAMICS ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J. !
ANSWER 5.04 (0.5) (b.) [+0.5] REFERENCE
- 1. Crystal River: Nuclear Energy Training, Module 3, Figure 16-3.
ANSWER 5.05 (0.5) V (A. ) [+0.5] see u hed d*## x' I REFERENCE l 1. Crystal River: Nuclear Energy Training, Module 3, Figure 16-3. ANSWER 5.06 (1.50)
- 1. AFW enters near the top of the OTSG. This results in an effective increase in heat transfer area.
- 2. AFW is not pre-heated, and is therefore colder than MFW.
- 3. - AFW has a steam pressure reduction effect that MFW does not have because it is injected into the steam space of the OTSG.
[+0.5] each REFE81ENCE
- 1. B&W Technical Document, Emergency Procedures Technical Bases.
(c' et&sll W %
- 4. t:,'C
}Y
. s,u4c ,< M. v w.r. .ci n fldg fnd+:.4 % <+~u h s b N
so .__ .. _ _ . . . _ . _ _ _ .__ _._ _ . . _ ._ . .
._ _ _ _ . _. . _ . ___i._.
so _ ..._. ._. . ._g ___ ..
~: . _ _ _ . . . . . . . . . ...
... __. g .,
~_ mg'f .: ... m . i~ :'-% ,,oo .l g - -
%%c= % '
1 2 -- - - - - - -^ -
~ .= -- - - -
- N=3x _
x -.
.\
,o. o_
.o g...
l:; _: ::--* :._ -_. . _ _ .g.,
.. _.w
. c,,o.i.
= *ho . ._ . . . . _ . _ . > .,
l ... . . . . __ ._.._.. . . _. g ..
= .__
yo 3 _. 8 =so .
._ _ . __ . _ . _ _ . . . ._. i L :- -
. . !:: .1
*. ~..:: ..
-.o o ioo soo soo .oo soo ooo sootRafoR TEWPERATU8tt (*F)
FIGURE SNP-RF.68: MODERATOR TEMPERATURE COEFFICIENT CURVE (AT 80L AND EOL, CYCLE 1) ROOS IN (REV. 1) . increasing the moderator temperatura increases the probability of neutron leakage into the control rod and loss to the fission chain reaction. Therefore for a given rooderator temperature change, more negative reactivity is inserted when control rods are in the core. Consider the equation for eT 2 2 dl dL 1 df 1 2 f dT th)
+ - 8 +
"T f dT p dT (dT
\_
3_23 W 0469C
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- 5. THECRY OF NUCLEAR POWER PLANT-OPERATION, FLUIDS, PAGE 25 AND THERMODYNAMICS l
ANSWERS -- CRYSTAL Rt2ER -86/12/17-HUENEFELD, J.
' ANSWER 5.07 (1.00)
(c.) [+1.0] REFERENCE
- 1. B&W Technical Document, Emergency Procedures Technical Bases.
ANSWER 5.08 (2.00) Given p = -6% CR(1) = 50 cps CR(2) = 285 cps 1 1
----- = K = ---- = 0.943 OK if estimated p = k - 1 [+0.5]
1-p 1 1.06 CR 1-K 1-K 1 2 50 2
--- = ------ = --- = ---------
[+0.5] CR 1-K 285 1 - 0.943 . 2 1 50(0.057)
--------- = 1 - K therefore K = 0.99 [+0.5]
285 2 2 i l K-1 0.99 - 1 p = ----- = -------- = -0.01 [+0.5] K 0.99 . or 1% shutdown l REFERENCE
- 1. Basic Reactor Theory, Formula Sheet.
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=
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, PAGE 26 I
AND THERMODYNAMICS ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J. l
. ANSWER 5.09 (1.00)
- 1. A divergence developing between the incore T/C and Thot.
- 2. A "decoupling" between Tcold and SG pressure.
[+0.5] each REFERENCE
- 1. B&W Technical Document, Emergency Procedures Technical Bases,
- p. II.B.9.
ANSWER 5.10 (1.50) A trend of incore T/C temperature versus RCS pressu e increasing away from the SG Tsat along the saturation curve. [+1.5] REFERENCE
- 1. B&W Technical Document, Emergency Procedures Technical Bases,
- p. B-9.
ANSWER 5.11 (1.00) (d.) [+1.0] REFERENCE
- 1. B&W Technical Document, Emergency Procedures Technical Bases,
- p. II.B-9.
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, PAGE 27 AND THERMODYNAMICS ANSWERS -- CRYSTAL RM.ER -86/12/17-HUENEFELD, J.
ANSWER 5.12 (1.50) The warming of the shell is promoted because when a vacuum is drawn, the water in the OTSG flashes to steam at the saturation temperature. The steam helps to evenly warm the OTSG shell. [+1.5] REFERENCE
- 1. Crystal River: Basic Thermodynamics, OP-202, p. 23.
ANSWER 5.13 (1.00) (b.) [+1.0] REFERENCE
- 1. Crystal River: OP-202.
ANSWER 5.14 (1.00)
- a. nucleate boiling region ;+0.5;
- b. nucleate boiling region ,+0.5 REFERENCE
- 1. Crystal River: ROT-3-2, pp. 30 to 37.
-r -- - - - -
7- r - v' y - - - - - - - - -- -- ---
----m *--
e O O
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, PAGE 28 AND THERMODYNAMICS ANSWERS -- CRYSTAL RI_V_ER E -86/12/17-HUENEFELD, J.
ANSWER 5.15 (2.00) i
- a. (b.) [+1.0]
- b. By conduction [+0.5] from the vessel and other components that are heated by conduction and gamma heating [+0.T]. ,
1 REFERENCE
- 1. Crystal River: ROT-3-2, pp. 47 through 48.
ANSWER 5.16 (3.00)
- a. The idle OTSG will lag significantly behind the operating OTSG; a loss of subcooling margin in the idle loop could result if RCS pressure is reduced too fast. [+1.5]
- b. There is very little coupling between the tubes and the shell of the idle OTSG. Because the tubes are already under extra tension due to the depressurized secondary side, it is important that cooldown rate be limited to minimize tube-to-shell delta T. [+1.5]
REFERENCE
- 1. Crystal River: ROT-3-3, p. 11.
ANSWER 5.17 (2.00) . Yes [+0.5] . The potential exist that non-condensable gases could ! collect on the heat transfer surface of the primary side OTSG tubes. If this were to occur, non-condensable gases of sufficient quantity could shut off the boiler-condenser process due to the increase in resistance to heat transfer. [+1.5] I REFERENCE
- 1. Crystal River: ROT-3-5, p. 6.
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, PAGE 29 AND THERMODYNAMICS ANSWERS -- CRYSTAL RIVER -86/1'2/17-HUENEFELD, J.
ANSWER 5.18 (0.50) True [+0.5] REFERENCE
- 1. Crystal River: ROT-3-8, Section 10, p. 2.
ANSWER 5.19 (1.00) The " gray" APSRs are made of inconel and they are not as strong as the previously used black APSRs. [+1.0] REFERENCE
- 1. Crystal River: ROT-2-10, p. 10-8.
ANSWER 5.20 (0.50) False [+0.5] f REFERENCE 1.- Crystal River: Basic Reactor Theory
- 2. Crystal River: NUS, Module 3.
l l
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, PAGE 30 AND THERMODYNAMICS ANSWERS--CRYSTALRIV.ER -86/12/17-HUENEFELD, J.
ANSWER 5.21 (2.00) The delayed neutrons behave as a source. Keff after a reactor trip will be well above 0.9; therefore, significant subcritical multiplication of the delayed neutrons take place. [+2.0] REFERENCE
- 1. Crystal River: Basic Reactor Theory.
- 2. Crystal River: ROT-1-33, p. 20. ,
. ANSWER 5.22 (1.50)-
- 1. increasing PMvep'c. ,ske
- 2. . decreasing'RCS pressure
- 3. relatively stable,RCS temperature
[+0.5] each """*"'ij REFERENCE
- 1. Crystal River: ROT-3-11, p. 23.
ANSWER 5.23 (1.50) l By observing that all control rods are operable and aligned with their groups within the limits of the Technical Specification rod withdrawal l curves. [+1.5] REFERENCE
- 1. Oconee: OP-0C-SPS-RT-RC.
- 2. - Crystal River: Basic Reactor Theory.
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, PAGE 31 l AND THERMODYNAMICS ANSWERS -- CRYSTAL RIV.ER -86/12/17-HUENEFELD, J.
l ANSWER 5.24 (3.00) At 70 degrees v(f) = 0.016050 ft**3/lbm rho = 62.305 lbm/ft**3 [+0.5] m(f) - rho y = 62.305 lbm [+0.5] H(f) = (38.052 BTU /lbm)(62.305) = 2370.841 BTU [+0.5] v(g) = 868.4 ft**3/lbm rho = 0.00115 lbm/ft**3 [+0.5] m(g) = ev = 0.00115 lbm [+0.5] H(g) = (1092.1)(0.00115) = 1.26 BTU [+0.5] The 70 degrees liquid water contains the most ethalpy per ft**3. REFERENCE
- 1. Oconee: Thermodyr=mics, Fluid Flow and Heat Transfer for Nuclear Power Plants.
i
. O Q
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND PAGE 32 INSTRUMENTATION.
ANSWERS -- CRYSTAL RH.ER -86/12/17-HUENEFELD, J.
' ANSWER 6.01 .(3.00)
There are two frames of reference from which the candidate may
- answer this question. The first frame of reference is to discuss-the heat transfer across the steam generator "mm'the primary to the secondary. The second frame of reference is to discuss the primary heat balance and the secondary heat balance. Either frame of-reference is equally correct. Representative full credit answers from each frame of reference are given below:
FRAME OF REFERENCE 1: The ICS controls selected parameters to ensure a balance between the heat being generated in the reactor, and the heat being transferred at the steam generator. There-are four variables that affect this ener produced (i.e., generated megawatts)gy , the balance: primary tothe total energy secondary heat transfer area (i.e., OTSG level), the primary side temperature (i.e., Tave), and the secondary side saturation pressure (i.e., steam header pressure). By fixing any three of these four variables, the remaining variable must by necessity assume a specific value in order for equilibrium to exist. If level is not at this specific .value, then an upset condition exists. The ICS'was designed to fix the following three variables: generated megawatts, Tave, and steam header pressure. The ICS will control upsets in these three variables so as to drive them to their selected setpoints. Feedwater flowrate is adjusted in concert to ensure an equilibrium steam header pressure is achieved and that any upset condition is eliminated. OTSG level is, by virtue of this control of feedwater, indirectly driven to the correct level. Any drifting from the correct level will cause an upset in steam header pressure along with a resultant correcting adjustment in feedwater flow. I
s .
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND PAGE 33 INSTRUMENTATION ANSWERS -- CRYSTAL RIEER -86/12/17-HUENEFELD, J. l l
FRAME OF REFERENCE 2: In order for equilibrium heat transfer to exist, the primary heat balance must equal the secondary heat balance. In the primary, the only variables to be controlled are Thot and Tcold. The ICS does not control them specifically, but rather controls their mid-point by fixing a constant Tave. In the secondary, the variables of concern are: feedwater flow, inlet enthalpy, and outlet enthalpy (i.e., steam header pressure). The inlet enthalpy, because it coincides with the liquid phase, is very nearly a constant. The two remaining variables of control are directly controlled by the ICS. OTSG level is not one of the variables affecting a primary heat balance, nor a second cry heat balance. REFERENCE
- 1. Crystal River: ROT-3-2, pp. 31 and 32.
ANSWER 6.02 (1.50)
- 1. reactor power ( 30% (administrative)
- 2. oil lift pressure > 200 psig
- 3. NSCCCW return flow > 260 gpm/ pump
- 4. upper and lower oil reservoirs above low alarm
- 5. seal injection flow > 3 gpm/ pump
- 6. controlled bleedoff valves MUV-258, 259, 260, and 261 open
- 7. Tc > 500 degrees F to start the fourth RCP m e ;ees e., c ,w t. e e. 0 b ,- G el c ,d it.,
Any six (6) [+0.25], tw6 +1.5 maximum REFERENCE
- 1. Crystal River: ROT-3-5, p. 25 and OP-302.
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- 6. PLANT SYSTEMS DESIGN, CONTROL, AND PAGE 34 INSTRUMENTATION l
ANSWERS -- CRYSTAL RTV_ER 86/12/17-HUENEFELD, J. ANSWER 6.03 (0.50) True [+0.5]
' REFERENCE
- 1. Crystal River: ROT-3-9, pp. 3 and 4.
ANSWER 6.04 (0.50) True [+0.5] REFERENCE
- 1. Crystal River: ROT-3-9, p. 7.
ANSWER 6.05 (0.50) False [+0.5] REFERENCE
- 1. Crystal River: ROT-3-9, p. 15.
ANSWER 6.06 (0.50) False [+0.5] REFERENCE
- 1. Crystal River: ROT-3-10.
~.
. - , . - , -- ..w-,, ,. , m --r-< ---e - ig.-,-- --ca. -i+-- _ , - - - - - m.m---w -- - -,,.. 9-- ,,-.----w- - -- - ------. mew +y.-v-- ----
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- 6. PLANT SYSTEMS DESIGN, CONTROL, AND PAGE 35 l INSTRUMENTATION ANSWERS -- CRYSTAL RtEER -86/12/17-HUENEFELD, J.
I ANSWER 6.07 (0.50) True [+0.5] REFERENCE
- 1. Crystal River: ROT-3-10, p. 16. j i
ANSWER 6.08 (1.50) RCP seal injection can flow down the cold leg stratifying in the low point, at the location of the Tc RTD. [+1.5] REFERENCE
- 1. Crystal River: ROT-3-11, p. 20.
ANSWER 6.09 (2.00)
- a. The high ambient temperature decreases the density of the wet reference leg. This causes the detector to sense a larger relative pressure of the fluid being measured, therefore, increasing the indicated level. [+1.5]
- b. Yes [+0.5]
REFERENCE
- 1. Crystal River: ROT-3-11, pp. 23 and 24.
-:------------- g_ s-,-,- - , - - , ,,,_-- ,--,,n-e,- -. ,,,. -g-- - .,, ,-- , -- - , ,
t o o
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND PAGE 36 INSTRUMENTATION ANSWERS -- CRYSTAL RLEER -86/12/17-HUENEFELD, J.
i ANSWER 6.10 (1.00) (c.) [+1.0] I REFERENCE
- 1. Crystal River: ROT-3-12.
ANSWER 6.11 (2.00)
- 1. decay heat removal heat exchangers
- 2. decay heat services seawater pump motors
- 3. DC pump motor handling units
- 4. . decay heat pumps and motors
- 5. reactor building spray pumps and motors
- 6. makeup and purification (MU) pumps and motors IA and IC
- Any four (4) [+0.5] each, +2.0 maximum.
REFERENCE
^
- 1. Crystal River: ROT-4-2, p. 5.
l l l 1 1
. - _ _ _ , _ , , _ . _ _ . , , _ , _ - . . . . , _ _ , . _ . , . . . . _ _ _ _ _ . , . . _ . - , _ . . . _ , - - . . . . , , . _ . _ _ . . . - . . . . . . _ . . - _ _ . - . _ . . . , . . . - . . . , ~ , _ . . . _ . . . - .
. O O 6.-~ PLANT SYSTEMS DESIGN, CONTROL, AND PAGE 37-INSTRUMENTATION ANSWERS -- CRYSTAL RIVER -
86/12/17-HUENEFELD, J.
. ANSWER 6.12 (3.00)
- a. 1. ESactuation(HPI)
- 2. ES bus degraded voltage
- 3. ES bus undervoltage (loss of voltage)
[+0.5] each
- b. 1. diesel start mode select switch (43) on main control board (one for each diesel) selected to auto
- 2. air shutoff valves EGV-35(A), EGV-39(B) open
- 3. control at engine - normal switch (on the engine gauge panel) selected to normal
~
- 4. 86 lockout relay (generator control cabinet) reset
- 5. SDR seal-in reset (accomplished by pressing the Reset pushbutton CPB4) located on the engine gauge panel
[+0.3] each REFERENCE
- 1. Crystal River: ROT-4-6, p. 55.
ANSWER 6.13 (1.00) the waste gas system [+1.0] c.- tkr wk,.\ka. u ok. l.1ft n . fw.
- 1. Crystal River: ROT-4-2, p. 47.
-- - -- ,- n ---,,-e
, - - - ~v v --
w- +,- -- mn+ - e,--,- -------,w .~ ,-
. O O
- s ,
-6. PLANT SYSTEMS DESIGN, CONTROL, AND PAGE 38 INSTRUMENTATION j ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J.
ANSWER 6.14 (2.50)
- a. Because the SW system is a source of unborated water that could potentially dilute water in the RB sump. [+1.0]
- b. 1. RB AHUs
- 2. letdown coolers
- 3. RCDT
- 4. CRD coolers
- 5. RCP coolers Any three (3) [+0.5] each, +1.5 maximum.
~ REFERENCE
- 1. Crystal River: ROT-4-2, p. 33.
ANSWER 6.15 (1.50) To ensure that a minimum pressure (60 psig) exists for all components inside the RB thereby eliminating the SW system as a flowpath for release from the RB. -[+1.5] REFERENCE
- 1. Crystal River: ROT-4-2, p. 9.
I ANSWER 6.16 (1.00)
- 1. demineralized water [+0.5
- 2. condensate system [+0.5))
REFERENCE
- 1. Crystal River: ROT-4-2, p. 12.
^
. O O 4 y
- 6. PLANT SYSTEMS DESIGN, CONTROL,'AND PAGE 39 INSTRUMENTATION
, ANSWERS -- CRYSTAL RLEER -86/12/17-HUENEFELD, J. ANSWER 6.17 .(1.50) As power is increased, the inlet water temperature decreases causing the water between the thermal shield and the reactor vessel wall to become more dense and therefore to shield more neutrons from the detectors. [+1.5] 1 REFERENCE
- 1. Crystal River: ROT-4-10, p. 34.
ANSWER 6.18 (1.50) 1 x 10**-9 amps [+0.5]. It takes both IR instruments greater than setpoint to cause the SR to deenergize [+1.0]. ) i REFERENCE
- 1. Crystal River: ROT-4-10, Figure 11.
ANSWER 6.19 (1.50) If the NNI system were allowed to operate with a degraded voltage, instrument readings would be affected. Actions taken by either the operator or by automatic system response would be in response to inaccurate ' readings.. [+1.5] REFERENCE
- 1. Crystal River: ROT-4-9, p. 15.
a j .. e
,_-.,--.,,__,m,.,-_m.-_r _- ..,,_ _ . . .. ._,,, _. -..,,,.,-....._..,,~. - .
_,m_ ..-_ _ - - , __ , . ,
. O O
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND PAGE 40 INSTRUMENTATION ANSWERS -- CRYSTAL RHER -
86/12/17-HUENEFELD,'J. ANSWER 6.20 (1.50) A simulated flow signal is derived from signal generators which produce a signal equivalent to full flow from one pump anytime the breaker to that pump is closed. [+1.5] REFERENCE
- 1. Crystal River: ROT-4-9, p. 11.
ANSWER 6.21 (1.50)
- 1. Fixed Water Spray System - A fixed system designed for specific discharge patterns generally over one specific fire hazard.
- 2. Wet Pipe Sprinkler System - A general area system that is fully pressurized. Water flow is actuated when individual sprinkler heads get hot enough to soften fusible solder links.
- 3. Pre-action Sprinkler System - Similar to the wet pipe system, except the system is dry. A main flow control valve must be actuated separately by a heat or smoke detector.
[+0.5] each For examples see ROT-4-7, Tables 1, 2, and 3 on pp. 16, 20, and 24, respectively. REFERENCE
- 1. Crystal River: ROT-4-7.
w.-- - , _
CRYSTAL RIVER O O PAGE 40a TABLE 1 - FIXED WATER SPRAY SYSTEMS I P:otected Area l Detector l Monitor l Actuation l Remarks l Type l System ! Req'ments 1 1 I l l Aux. Bldg. Charcoal l heat l Fire Srvce 1 2 detec. l Filter Banks (5) l l Panel l [ l l I l Control Complex Filter l heat l Fire Srvce 1 2 detec. l Banks (2) l l Panel l l l l l l Hydrogen Seal Oil l heat l Fire Srvce l 1 detec. l Unit l l Panel l l l l l l Turbine Lube Oil l heat l Fire Srvce l 1 detec. [ Storage Tank l l Panel l l l l l l Feedwater Pump l heat l Fire Srvce l 1 detec. l Consoles l l Panel l l l 1 1 I Unit Auxiliary l heat l Fire Srvce l 1 detec. I Also Trips Transformer l l Panel l l Water Wall I i I l Start-Up Transformer l heat l Fire Srvce l 1 detec. l Also Trips
, l l Panel l l Water Wall i I I I
., Unit Transformers l heat l Fire Srvce l 1 detec. l Any 1 Detec g (3-1,3-2,3-3) l l Panel l l Trips Other l l l l 2 Units &
l l l l Water Wall l I I I TSC Air Cleanup Units l smoke l PYR-A-LARMl 1 detec. l Alarms on l l l l Fire Service i l l l Panel 69 i l N ROT-4-7 16 Rev. O
; CRYSTAL RIVER PAGE 4Cb TABLE 2 - WET PIPE SPRINKLER SYSTEMS Protected Area-- l Detector Type l Monitor l Actuation l Remarks l l System l Req'ments I l
Fire Pump House l press. sw.li Fire Srvce l1 l i sprinkler l l l Panel l head l l l l 95' El. Turbine Bldg l press. sw.ll Fire Srvce l1 sprinkler l2 headers l l Panel l head l l south / north 119' El. Turbine Bldg l press. sw.[I Fire Srvce l1 sprinkler l2l headers i l l Panel l head l south / north l 1 95' El. Aux. Building l press, sw.llPyrotronic l1 sprinkler1 lSee note 5 l l Module 5 l head l l l l 119' El. Aux. Building l press. sw.llPyrotronic l1 sprinkler lSee note 6 l l Module 5 l head l 1 1 I I 95' El. Interm. Bldg l press, sw.lPyrotronic l1 sprinkler [
} l Module 5 l head [
l 119' El. Interm. Bldg l press. sw.li Pyrotronic l1 sprinkler l I i 1 l Module 5 l head l I I ~ 95' El. Control Comp. l heat lPYR-A-LARM l I l ./ (Ch,em/ Rad) l l 1 detec. lSee note 1 l l 1 124' El. Control Comp.l press. sw.lI Pyrotronic l1 sprinkler lSee I i note 4 l l Module 5 l head l I office Building l press, sw.lI Fire Srvce l1 sprinkler lI I Records Storage Vault l l Panel l head l 1 1 Technical Support l smoke I l Center "' lPYR-A-LARM l 1 detec. lSee note 2 l l pe l l l l Environmental Whse l press. sw.l1 Local only l1 l sprinkler lSee note 3 l lat present l head l l Flammable Liquids I l press. sw.l1 Local only l1 sprinkler lSee note 3 I Storage Area ,l lat present i head l ; I I I I I I I I l s , i I ) ! I ROT-4-7 20 t Rev. O !
Q O PAGE 40c lER LER SYSTEM TXBLE 3 - PRE-ACTIOJ{ SPRINK i Actuation l Remarks _ _ _ l Rea'ments li actcd Area-l Detector Type i l Monitor System I l l IFire Srvce i 11head sprinkler l actionlafter pre-heat Panel l l alarm
' Diocol Gen. Il (20) l l 1
- rol Rooms l l i I I -
I B e
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- - - - - -a ---, _ _-__,_._ _ -,,-.- - ,,,--.,,w-.,._____.____,,-,__,,w-.--..,,,,,.,_.., -
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. . O .O :
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND ~PAGE 41- ..
INSTRUMENTATION s ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J. Q.) ANSWER 6.22 (1.00) (d.) [+1.0] REFERENCE ,
, .c
' 1._ Crystal River: Technical Specifications, p. 3/4 8-1. 'q L2. Crystal River: ROT-4-6.
k l i L w i i 4 i l l i-i. i) % , s l l 3 5 .
-r- - , _ . _ , _ _ . - - . - . _ - _ _ . , _ , - - . _ _ _ _ _ _ , . - _ . . . . _ . . . . . . - , . _ _ . - . , . _ . _ - - _ _ , . _ . _ - - _ _ . . . . _ . . , _ _ _ . . . _ _ _ . -
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7.kPROCEDURES-NORMAL,' ABNORMAL, EMERGENCY - PAGE 42 3s 1 4 6 RADIOLOGICAL CONTROL , j ah _ _ g'
' ANSWERS -- CRYSTAL RIVER -86/ 2 ?d17-HUENEFELD,' J. '
- y
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~ ANSWER ( 7'.01 s.1 =
(1.bO) n _ Q* - g. 3ys-s, - 4 Because'jlh'e hybazineNINs A g' eater ahfinity fo[: tile ). -Ndhangeresin r
~
than do other foris like' chlorine. Sustsined use of the' demineralizer?with hydrazine in the RCS' could lead to a release of
. undesirable ions from the,demineralizer. [+1.5]. i C '
h= , et , ,
;s 4 ,i
+ qREFERENCE Y U .
t ,/.- 1x S. >
- 1. Crystal River: OP-202, p. 4. 'Q .
U e , I s, .
' 2;.l, 3 hD. i ,f, '
i o id - ), W
,g ANSWER 7.02 4,(0.50)7
,,3 S; , t True [+0. 5]
r
,'qf' i -'
f, !ke ,s l' ' ' / REFERENCE h' -,' s ,
,,N
- 4. Cryhal River: OP-202, p. 29. 4/
s
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y , , i (
* , ANSWER 7.03 5 (1.50) EN i
,e , ,
.ej ,
s . 1.- (EFW must be turned on, full) if natural circulation stops and-i .
% the 4 team genera,thr level is, beleg the setpoint. (It can be throgled when natural circulatier starts.)
2.. (EFW st be turned on full) if its acthatioil was delayed. (It ekn be' throttled when natural circulatidarstarts.) ,( d 4 3. (EFW must be. turned on full) if it is in,jecting into only one generator. (It can be thpttled when natural circulation Qj pp ,[ n starts.) g ic , 4 y, ', , [ .5] -9 Mete 0,i y l c. ce c h9$ . m . e 6 v = y ., e d b. h.lt c. e $ f- '
\ <
s REFERENCE {,
. v
- 1. Crystal River: ROT-3-3, p. 7.
>i,
,,,-.._---.m--.-, - - - . - _ - - - _ , - . , - _ - , . -
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- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY PAGE 43 AND RADIOLOGICAL GONTROL ANSWERS -- CRYSTAL RIVER ~ 86/12/17-HUENEFELD, J.
ANSWER 7.04 (0,5d True [+0.5] j D,*,('P s$ f<< k w Io'bl y<N*n .f REFERE
-1. Crystal River: ROT-3-4, p. 2.
4
- . ANSWER 7.05 (1.00)
(c.) [+1.0] 4 REFERENCE
- 1. Crystal River: ROT-3-4, p. 4.
ANSWER 7.06 (0.50) True [+0.5] REFERENCE
- 1. Crystal River: AP-530, p. 10.
ANSWER 7.07 (1.00)
- a. 545 gpm +0 -45 gpm ;+0.5;
- b. 600 psi +/- 50 psi +0. 5, REFERENCE
- 1. Crystal River: ROT-3-4, p. 14. ~.
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. O O
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY PAGE 44 AND RADIOLOGICAL CONTROL ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J.
ANSWER 7.08 (1.00) (c.) [+1.0] REFERENCE
- 1. Crystal River: ROT-3-5, p. 23.
ANSWER 7.09 (2.00)
- 1. Gravity drain through the DH drop line to the RB sump. [+1.0]
- 2. Feed via the auxiliary spray flow path to the pressurizer.
[+1.0] REFERENCE l
- 1. Crystal River: ROT-3-6, p. 7.
1 ANSWER 7.10 (0.50) l exciter current [+0.5] REFERENCE
- 1. Crystal River: OP-203, p. 20.
. O O 9 ,
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY PAGE 45 AND RADIOLOGICAL CONTROL ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J.
ANSWER 7.11 (1.00) (b.) [+1.0] REFERENCE
- 1. Crystal River: OP-203, p. 3.
- 2. Crystal River: OP-502, p. 4.
ANSWER 7.12 (1.50) j
- 1. initiate full HPI
- 2. stop all reactor coolant pumps
- 3. feed up OTSGs to 95% on the OR
[+0.5] each REFERENCE
- 1. B&W Technical Bases Document.
ANSWER 7.13 (1.50)
- a. }V T H
- b. 65% EFIC Hi range
- c. 95% EFIC H1 range
[+0.5] each REFERENCE
- 1. Crystal River: ROT-3-3, Rev. 4, Objective 11.
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. O O I
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY PAGE 46 AND RADIOLOGICAL CONTROL ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J.
ANSWER 7.14 (1.50)
- 1. IF asymmetric rod condition exist THEN go to AP-542
- 2. Select "J0G" on " SELECTOR" switch
- 3. Sto supply) p rod withdrawal (transfer rod (s) to alternate power REFERENCE
- 1. Crystal River: ROT-5-27.
- 2. Crystal River: AP-555, Rev. O.
ANSWER 7.15 (1.00)
- 1. power reduced to (60%
- 2. T(c selected to the unaffected leg 3, $ sic)efA tetd a t t. M % % M c- % JE' c P REFERENCE I
- 1. Crystal River: ROT-5-2, Rev.1, Objective 3C.
- 2. Crystal River: OP-204, Rev. 43, p. 5.
ANSWER 7.16 (2.00) JLestior_e e t!Le block vaWye to o erabilityMthin one hour [+0.7] or close the block valve and remove power from it [+0.7], on_clo_se_the PORV and_ remgve power from the solenoid [+0.6]. REFERENCE
- 1. Crystal River: STS - 3.4.3.2.
im
O O y ';
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY PAGE 47 AND RADIOLOGICAL CONTROL ANSWERS--CRYSTALRbR -86/12/17-HUENEFELD, J.
ANSWER 7.17 (3.00)
- 1. verify a valid actuation
- 2. depress "HPI actuation" pushbuttons
- 3. ensure HPI trains start [+0.4]
- 4. ensure BWST suction valves open [+0.4]
- 5. ensure HPI valves open [+0.4]
- 6. ensure LPI trains start
- 7. ensure EDGs start
- 8. ensure diverse containment isolation actuation
- 9. verify adequate subcooling margin
[+0.3] each with exceptions noted REFERENCE
- 1. Crystal River: AP-380, Rev. 6.
- 2. Crystal River: ROT-5-22.
ANSWER 7.18 (2.00)
- a. Restore Tave to within its limit within 15 minutes or be in HOT STANDBY within the next 15 minutes. [+1.0]
- b. 1. ensures that HTC is within its analyzed range
- 2. ensures that the protective instrumentation is within its normal operatir.g range
- 3. ensures that the pressurizer is capable of being operable j with a steam bubble
- 4. ensures that the pressure vessel is above its minimum RT(NOT)
Any two (2) [+0.5] each, +1.0 maximum REFERENCE
- 1. Crystal River: STS 3/4.1.1.4.
. O O
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY PAGE 48 AND RADIOLOGICAL CONTROL ANSWERS--CRYSTALRIVIR -86/12/17-HUENEFELD, J.
ANSWER 7.19 (3.00)
- 1. ensure GRP 1-7 rods fully inserted l
- 2. ensure flux decreasing [+0.4]
- 3. ensure main turbine TVs and GVs closed
- 4. ensure main block valves closed [+0.4]
- 5. ensure low load block valves closed [+0.4]
- 6. maintain PZR level >/= 50 inches
- 7. ensure steam header pressure at 1010 psig
- 8. ensure output breakers open
- 9. close the block orifice bypass valve
[+0.3] each with exceptions noted REFERENCE
- 1. Crystal River: AP-580, Rev. 6, pp. I through 3.
- 2. Crystal River: ROT-62-82, Generic Objectives.
ANSWER 7.20 (2.50)
- a. Shutdown margin shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming: 1) no change in axial power shaping rod position, and 2) all control rod assemblies (safety and regulating) are fully inserted except for the single rod assembly of highest reactivity worth which is assumed to be fully withdrawn. [+1.5]
- b. No. Even if group 1 rods are to be withdrawn, the shutdown value must be >/= 1% delta k/k (i.e., Keff must be (0.99).
[+1.0] REFERENCE
- 1. Crystal River: Technical Specifications, pp. 1-3 and 1-9.
- 2. Crystal River: Plant Heatup, OP-202, p. 3,7.
O' O r .
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, PAGE 49 AND LIMITATIONS _
-ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J.
ANSWER 8.01 (0.50) True [+0.5]
-REFERENCE 1.- 10'CFR 50, Appendix B, V.
ANSWER 8.02 (0.50) True [+0.5] REFERENCE
- 1. 10 CFR 50.54, (x).
l l l'
. . O O ~ Q
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, PAGE 50 AND LIMITATIONS ANSWERS--CRYSTALRiiTER -86/12/17-HUENEFELD, J.
ANSWER 8.03 (2.50)
- 1. any emergency of the classes in the Approved Emergency Plan
- 2. receipt of a package with > 0.01 uc/100 cm**2 on surface
- 3. 1ech Spec safety limit exceeded
- 4. aJtomatic safety system does not function as required
- 5. limiting control setting exceeded
- 6. 'imiting conditions for operation exceeded
- 7. results of a trace investigation of radioactive shipment
- 8. theft or diversion attempt of licensed special nuclear material
- 9. event'that threatens effectiveness of physical security system
- 10. plant shutdown required by Tech Specs
- 11. departure from Tech Specs authorized to protect public health and safety
- 12. degradation of plant and/or principal safety barriers
- 13. natural phenomenon or external condition threatens safety or hampers personnel duties for safe operation
- 14. any event that causes (or should have caused) ECCS injection
- 15. major loss of emergency assessment capability, offsite response capability, or communications capability
- 16. plant event that threatens safety or hampers personnel duties for safe operation; includes fires, toxic gas release and f.f, ! .*
-4 4 Iocr'E 20 403 du. ^ ^ ^ " ^ ^%.
Any five (5) f+0.5] e 5 maximu REFERENCE
- 1. Crystal River: ROT-3-15, pp. I and 2.
1 ;L
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, PAGE 51 AND LIMITATIONS ANSWERS --~ CRYSTAL RIVER -86/12/17-HUENEFELD, J.
ANSWER 8.04 (1.50) no power range channels two IR channels two SR channels [+0.5] each REFERENCE
- 1. Crystal River: ROT-4-10, p. 40.
ANS'.iER 8.05 (1.00) The breaker must successfully pass two (2) consecutive retests. [+1.0] REFERENCE
- 1. Crystal River: AI-500, p Ba.
ANSWER 8.06 (1.50) Operations - three individuals, one is leader Maintenance - two individuals [+1.5] REFERENCE
- 1. Crystal River: OSIM, p. V-13.
e i ._ _ _ _ _ _ _ __ _
. ~
l the reg- (3His A reuston on tne orag m al j unoer a bu.21t u s or t au.;; un tua pat 6 me .....4.....m- - . . - . . . - . .- _ PSAR containing those original pages C shall notif y the NRC Operations significantly compromises plant l **' Center via the Emergency Notification safet); 3erate a that are still apphcable plus new re-t to the placement pages shall be filed within System of; (B) In a condition that is outside the l of this 24 months of either July 22.1980, or f*
' (i) The declaration of any of the design basis of the plant: or
! as pro- the date of issuance of the operating g Emergency Classes specified in the 11 (C)In a condition not covered by the 3 (4) of heense, whichever is later, and shall censee's approved Emergency Plan;
- plant's operating and emergency pro-lanalysis bring the PSAR up to date as of a .
or cedures.
' (ii) Of those non-Emergency events ( i) Any natural phenomenon or hitted as maximum of 6 months prior to the y operat- date of filing the revision. I specified in paragraph (b) of this sec- other external condition that poses an nforma- tion. actual threat to the safety of the nu-(ii) Nat less than 15 days before N i
gontains 150 Mite) becomes effective, the Direc- ,1 (2) If the Emergency Notification clear power plant or significantly ham-
- d. This tor of ?he Othee of Nuclear Reactor
System is inoperative, the licensee pers site personnel in the performance i
changes Regulaiton shall notify by letter the shall inake the required notifications of duties necessary for the safe oper-non and licensecs of those nuclear power plants i via commercial telephone service, atson of the plant. mission ther dedicated telephone system, or (iv) Any event that results or should the li-initially subject to the NRC's system- l r7 any other method wruch win ensure have resulted in Emergency Core atic evaluation program that they need not comply with the provisions of
- ng & sum MS Marge inm of th - 1 fo*t eN Oper ons Ce ter the reactor coolant system as a result this section while the program is being (3) The licensee shall notify the ate, the l of a valid signal.
lupdated conducted at their plant. The Director 1 NRC immediately after notification of (v) Any event that results in a major 'ude the of the Of fice of Nuclear Reactor Reg- . the appropriate State or local agencies I ss of emergency assessment capabil-5 the fa. ulation will notify by letter the beens- and not later than one hour af ter the ity, offsite response capability, or com-ce of each nuclear power plant bemg i time the licensee declares one of the munications capabihty ( e.g., signifi-ld in the ns per, evaluated when the systematic evalua.
; Emergency Classes.
cant portion of control room indica- ! in sup. tion program has been completed. (4) When making a report under tion. Emergency Notification System, hdments Withm 24 months after receipt of this - paragraph (a)(3) of this section, the li- or offsite notification system). bis that notification the licensee shall file a censee shall identify: L eviewed complete PSAR which is up to date as i I (i) The Emergency Class declared; or (vD Any event that poses an actual threat to the safety of the nuclear ' lyses of of a maximum of 6 months prior to (ii) Either paragraph (b)(1). "One. power plant or significantly hampers j by or on the date of filing the revision. Ilour Report." or paragraph (bx2), Smission (4) Subsequent revisions shall be "Four-Ilour Report," as the paragraph site personnel in the performance of ' rmation filed no less freqpently than annually
- of this section requiring notification of duties necessary for the safe operation I the Non-Emergency Event. of the nuclear power plant including 3 within and shall reflect all changes up to a maximum of 6 months prior to the l (b) Non-cmergency events-gQ fires, toxic gas releases, or radioactive
.ated in- date of filing. 8 oitf'reporti,1f not reported as a dec. releases. aration 'of an Emergency Class under (2) Four hour reporfs. If not report- , on a re-paragraph (a) of this section, the 11- ed under paragraphs (a) or (b)(1) of (5) Each replacement page shall in. 11 be ac- clude both a change indicator for the dentifies area changed. e.g., a bold line verticaj. censee shall notify the NRC as soon as this section, the heensee shall notify h follow- ly drawn in the margin adjacent to the practical and in all cases within one the NRC as soon as practical and in all
, signed portion actually changed, and a page hour of the occurrence of any of the cases, within four hours of the occur-following: rence of any of the following:
ks of the change identification (date of change (tXA) The initiation of any nuclear (D Any event, found while the reac-lbe filed or change number or both). plant shutdown required by the tor is shut down, that, had it been . Reactor ,. found while the reactor was in oper- ! g gulatory (33 FR 9704. July 4.1968, as amended at 41 plant's Technical Specif; cations. ~ ' ! 20555. FR 16446. Apr.19.1976; 41 FR 18303. May ABT Ani deviationi.frodi th'e plant's ation, would have resulted in the nu- ' 3.1976. 45 FR 30615. May 9.19801 Technical Specifications authorized clear power plant, including its princi-hde (i) a pal safety barriers, being seriously de- ' zed of fi- kursuarlt to i 50.5Kx) of this part,,, (ID Any event or condition"during graded or being in an unanalyzed con-
~
- I"" * ""' '*""" "4"
) the in- menth br "Pnanng nuckar pomu re operation that results in the condition dition that significantly compromises i changes
"""" of the nuclear powerplant, including plant safety.
hbmittal. its principal safety barriers, being seri- (iD Any event or condition that re-
- non and tai General requirements.' (1) Each ously degraded; or results in the nucle- sults in manual or automatic actuation hmission nuclear power reactor heensee licensed ar power plant being; of any Engineered Safety Feature nmission
- ( ESP), including the Reactor Protec-changes ,other requirements for immediate notifi tion System (RPS). Ilowever, actu-(ation of ration of the NRC by licensed op. rating nu 'These Emergency Classes are addressed ation of an ESF, including the RPS, ricar s m er reactors are contamed else- in Appendix E of this part.
instons of ' Commercial telephone number of the that results from and is part of the lutted to whrre m ihn chapter. m particular. n 20 205 20 403. 50 36. and 73 71. NRC Operauons Center is (202) 951-0550. preplanned sequence during testing or 456 457 l l
,. - - - , - .--- - ~ - - - - - _ ,--. . . - - - - --- - - - - - - -
i 20.403 10 CFR Ch.1 (1-lhdities) klar Rsgulutary Csmmissian (34 FR 7500. May 9.1969, as amended at 38 (d) Reports made by licensees in ro"8 s FR 1271. Jan.11.1973; 48 FR 33859. July 26. sponse to the requirements of this se' (1) Estimates of each individue 19831 gesure as required by paragraph tion must be made as follows: p secuon; O 20.403 Notifications of incidents. (1) Licensees that have an installed Emergency Notificatlan System shaa. fu) Levels of radiation and c (a) Immediate nott/feation. Each 11- make the reporta required by part. M ons of ramacme mah censee shall immediately report any graphs (a) and (b) of this section le 8'Id* events involving bypt.pduct, source, or the NRC Operations Center in accoc6- till) The cause of the ext special nuclear material possessed by ance with I 50.72 of this chapter ; kvels or concentrations; and the licensee that may have caused or (2) All other licensees shall make the fiv) Corrective steps take I threatens to cauw reports required by paragraphs (a) and , planned to prevent a recurrence (1) Exposure of the whole body of (b) of this section by telephone and by tb) Any report filed with the
' any individual to 25 rems or more of telegram, mallgram, or facsimile to the alssion pursuant to paragraph radiation; exposure of the skin of the Administrator of the appropriate NRC mis section shall include for car whole body of any individual of 150 Regional Office listed in Appendix D ' Mdual exposed the name social a rems or more or radiation; or exposure of this part. a IF number, and date of birth.
of the feet, ankles, hands or forearms estimate of the individual s ext of any Individual to 375 rems or more (27 FR 5905. June 22.1962. as amended at, The report shall be prepared s 28 FR 6323. July 3.1963; 42 FR 43965. SeyL of radiation; or 1,1977; 43 FR 2719. Jan.19,1978; 48 FR this Information is stated in a se (2) The release of radioactive materi- 33859, July 26.19831 part of the report. al in concentrations which, if averaged < icHI) In addition to any notif over a period of 24 hours. would $ 20.404 [ Reserved] g required by 1 20.403 of this par exceed 5,000 times the limita specified Icensee shall make a report in for such materials in Appendix B, 8 20.405 Reports of overeuposures and eaa, of levels of radiation or release-Table II of this part; or cessive levels and concentrations. Goactive material in excess of (3) A loss of one working week or (a)(1) In addition to any notificatlas specified by 40 CFR Part 190,
' more of the operation of any facilities required by 1 20.403 of this part, each renmental Radiation Pro af fected; or licensee shall make a report in wTitirty Standards for Nuclear Power (4) Damage to property in excess of concerning any one of the foliostag attons,' or in excess of license
$200,000. types of incidents within 30 days oiits tions related to compliance s 4 (b) Ticenty four hour nottficaffon, occurrence:
' CTR Part 190.
Each Ilcensee shall within 24 hours of (1) Each exposure of an individual te . (2) Each report submitted I discovery of the event, report any radiation in excess of the applicable paragraph (ex1) of this sectioi event involving licensed material pos- limits in il 20.101 or 20,104(a) of this describe:
, sessed by the licensee that may have part, or the license; (1) The extent of exposure of i caused or threatens to cause: (11) Each exposure of an individual sais to radiation or to radioact; (1) Exposure of the whole body of to radioactive material in excess of the terial;
, any individual to 5 rems or more of ra- applicable limits in il 20.103(axlk diation: exposure of the skin of the (11) Levels of radiation and <
20.101(a)(2), or 120.104(b) of this park trstions of radioac tive matei whole body of any individual to 30 or in the license; ,olved; rems or more of radiation; or exposure (111) Level
- f radiation or concentis' of the feet, ankles, hands, or forearms tions of rt. active material in a re- ( 111 ) The cause of the ex to 75 rems or more of radiation; or levels, or concentrations; and stricted ares in excess of any other ap (iv ) Corrective steps tak (2) The release of radioactive materi- plicable limit in the license; al in concentrations which, if averaged planned to assure against a (iv) Any incident for which notifles" rence including the schedt over a period of 24 hours, would tion is required by 1 20.403 of this achieving conformance with 4 exceed 500 times the limits specified part; or for such materials in Appendix B, (v) Levels of radiation or concentrs. Part 190 and with associated Table II of this part; or conditions tions of radioactive material (whether (d) For holders of an opera (3) A loss of one day or more of the or not involving excessive exposure of operation of any facilities affected; or any individual) in an unrestricted ares " P (4) Damage to property in excess of in excess of ten times nn) applicable dm d in p ra
$ 2,000. limit set forth in this part or in the ik 0 (c) Any report filed with the Com- cense. ordan e ith the p edu mission pursuant to this section shall (2) Each report required under pars. stribed in i 50.73 (b ), (c), 'd ). '
be prepared so that names of individ- graph (a)(1) of this section must de- (g) of this chapter and must unas who have received exposure to ra- scribe the extent of exposure of Indl. elude the information reque diation will be stated in a separate viduals to radiation or to radioactive paragraphs (a) and (c) of this part of the report. mater ial, including: o na o a" n[ ents, 3 p ,, , no 256
+4 g
.a - e v
m sM y 48 Y 3 4. --
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- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, PAGE 52 AND LIMITATIONS ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J.
ANSWER 8.07 (0.50) False [+0.5] REFERENCE
- 1. Crystal River: OSIM, p. V-9.
ANSWER 8.08 (1.00)
- 1. The PORV block valve (RCV-11) is operable (capable of closing).
- 2. The operator on the switch does not leave the PORY (RCV-10) switch unattended (dedicated operator) until he assures that the PORV (RCV-10) is closed and there is no flow thru through
- the PORY (RCV-10).
[+0.5] each i REFERENCE
- 1. Crystal River: OSIM, p. V-14 ANSWER 8.09 (1.00)
- 1. corrective maintenance is to be performed around the clock
- 2. a dedicated operator is to be assigned to monitor the defective annunciators
[+0.5] each REFERENCE
~~
- 1. Crystal River: OSIM, p. V-21.
O
' i,
' O i
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, PAGE 53 l AND LIMITATIONS j ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J. l 1
l ANSWER 8.10 (1.00)
- 1. Nuclear Operations Superintendent
< 2. Nuclear Plant Manager o, % c.,( 08 [+0.5] each REFERENCE
- 1. Crystal River: AI-500, p. 3.
ANSWER 8.11 (2.00)
- 1. ensure the plant is under control with existing procedures and requirements
- 2. call the man-on-call, NUC Ops Superintendent, SOTA, Resident NRC Rep, and the NRC
- 3. determine subsequent actions (i.e., cooldown) start back up
- 4. fill out the RX trip / shutdown report and assign shutdown number
- 5. ensure the information is entered in the N0 logs and the SS00 logs
[+0.4] each REFERENCE
- 1. Crystal River: AI-500, Rev. 55.
- 2. Crystal River: Section 2.4, p. 17.
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- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, PAGE 54 AND LIMITATIONS ANSWERS -- CRYSTAL RilTER -86/12/17-HUENEFELD, J.
ANSWER 8.12 (2.00)
- 1. Emergency Medical Team !
- 2. Radiation Emergency Team '
- 3. Plant Fire Brigade
- 4. Environmental Survey Team
- 5. Sampling Team
- 6. Emergency Repair Team
- 7. Dose Assessment Team Any four (4) [+0.5] each, +2.0 maximum.
REFERENCE
- 1. Crystal River: EM-202, p. 12.
ANSWER 8.13 (1.50)
- 1. Emergency Classification
- 2. Notifications
- 3. Protective Action recommendations
[+0.5] each REFERENCE
- 1. Crystal River: EM-202, p. 6.
ANSWER 8.14 (0.50) only one [+0.5] REFERENCE
- 1. Crystal River: EM-202, p. 5.
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- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, PAGE 55 AND LIMITATIONS _
ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J. ANSWER 8.15 (1.50) Evacuate all people within a 2-mile radius and shelter all people 5 miles in the potentially affected sectors. [+1.5] REFERENCE
- 1. Crystal River: EM-202, p. 5.
ANSWER 8.16 (1.00) Director, Nuclear Plant Operations or his designated alternate, the Man-On-Call. [+1.0] REFERENCE' l
- 1. Crystal River: EM-202, p. 4.
l ANSWER 8.17 (1.00) contact and notify the control room [+1.0] REFERENCE
- 1. Crystal River: EM-201, p. 3.
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+
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,(
8. ADMINISTRATIVE PROCEDURES. CONDITIONS,~ AND LIMITATIONS - PAGE 56 ANSWERS -- CRYSTAL RIVER - 86/12/17-HUENEFELD, J. ANSWER 8.18 (2.00) 1. all penetrations required to be closed off-during an accident are able to be closed off by an operable automatic- closure system and OR closedauto de-energized by avalve manual valve, blind flange, or closed
- 2. containment air locks are operable
- 3. equipment hatch is closed and sealed 4.
ontainment leakage rate is within limits 5. all penetration seal mechanisms (i.e., 0-rings, gaskets) are operable [+0.4] each REFERENCE 1. Crystal River: Technical Specifications, p.1-2. ANSWER 8.19 (1.50) a=. a. [+1.0]500 psig and/or 200 deg F and M 0.5 in. diameter opening b. with m li.icn jusu T;cet'r. "0 approval of the MOC [+0.5] REFERENCE 1. Crystal River: ROT-5-40, Rev. O, Objectives 13 and 14.
- 2. Crystal River: CP-115, Rev. 56.
e
, - - _. .__., _ .g _ _. .._ _ _ _ -
--__e . . . , . . . , ,,m y._- ,,,._.-,--,..,-,,,,,.,y,.,yy _
m -,- - -._--__y_. _ _ . . . - - _ .--- -. , . _ , , , , . . _.,_ ,_.. _,
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- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, PAGE 57 AND LIMITATIONS.
ANSWERS -- CRYSTAL RIVER - 86/12/17-HUENEFELD, J. ANSWER 8.20 (1.00)
- 1. portions of the condensate system
- 2. turbine building sump
- 3. nitrogen system Any two (2) [+0.5] each, +1.0 maximum REFERENCE
- 1. Crystal River: RSP-101, p. 22.
- ANSWER 8.21 (1.00)
An SRWP authorizes groups of individuals to conduct routine tasks, and RWP is more detailed and specific. REFERENCE
- 1. Crystal River: RSP-101, p. 8 i
ANSWER 8.22 (1.00) 100 mr/hr [+1.0] REFERENCE
- 1. Crystal River: RSP-101.
4 h ** 4
- - - - - .---._,w.- -..,--m.-. - - - ,-.---,,,,--,.y ,-,.ve,-----~y--.,.,v,,--,-.,vy .w-._.,_--,,.y_,. ,m,ww w , re -v--evmmvw--wr e=ww*+--
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- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, PAGE 58 AND LIMITATIONS ANSWERS--CRYSTALRIER -86/12/17-HUENEFELD, J.
ANSWER 8.23 (1.00) Electrical . The crane power bus must be deenergized to prevent accicantal contact and electrocution. [+1.0] REFERENCE
- 1. Crystal River: FP-601, p. 5.
l l ANSWER 8.24 (1.00) l l It must NOT be done. (An irradiated fuel assembi must never be I raised, no matter what the circumstances.) [+1.0 l REFERENCE I
- 1. Crystal River: FP-601, p. 4.
[, - O O efe-r@ EQUATION SHEET Where al'" "2 (density)1(velocity)i(area)i (density)2(velocity)2(area)2
-_____________________________=___..._______ _
KE=]2 PE = agh PEg + KE +P V l 1 7 = PE 2+KE 2+P Y22 where V = spectfic volume P = Pressure O" p(Tout-Tj ,) Q = UA (T,y,-Tsta) Q = ni(hi-h2) P = Po10(SUR)(t) P = Po et /T SUR = 26.06 T = (B-p)t T p delta K = (K,ff-1) -CR 1 (1-K,ffg) = CR 2 (1'Keff2) CR=S/(1-K,ff) (1-K,ffg) (1-K,ff) x'100% SDM = M = (1-Keff2I K eff decay constant = in (2) " 0.693 A 1=Ae-(decayconstant)x(t) g t t 1/2 1/2 Water Parameters Miscellaneous Conversions 1 gallon = 8.345 lbs 1 Curie = 3.7 x 10 10 dps 1 gallon = 3.78 liters 1 kg = 2.21 lbs 1 ft 3= 7.48 gallons I hp = 2.54 x 103 Btu /hr 3 6 Density =62.4lbg/ft 1 MW = 3.41 x 10 Btu /hr Density = 1 gm/cm 1 Btu = 778 ft-lbf Heat of Vaporization = 970 Btu /lbm Degrees F = (1.8 x Degrees C) + 32 Heat of Fusion = 144 Btu /lba 1 inch = 2.54 centimeters 2 1 Atm = 14.7 psia = 29.9 in Hg g= 32.174 ft-lbm/lbf-sec
ENCLOSURES 3 e :- % l l CO R POR A?80N December 23, 1986 TRA86 0233 i Mr. John Munro Chief, Operator Licensing Section Region _ II U.S. Naclear Regulatory Commission suite 2900 101 Marietta Street,'NW Atlanta, GA 30323 Subjects Crystal River Unit 3 12/17/86 NRC-Issued License Exas Dear Mr. Munro As pe r the cur rent practice f or examination revie vs af ter NRC-issued operator examinations, please find enclosed our review and comments on the December 17, 1986, Senior Re a c t o r Operator i Examinatione given at Cryotel River Unit 3. We are including our f consents and recommended action for each question under review. l l If you desire any f urther information, please contact Mr. Johnia
$aith . Nuclear Operations Training Supervisor, at (904) 793-0504 Ext. 107. Thank you f or your a t tention in this very important matter.
Very truly yours,
$La c-. KdL C. Kelley Manager, Nuclear Operations Training JCS/LCK/lb Attachment f(?o 22 6 % 27 NUCLEAR OPERAT NS TRAINING DEPARTMENT: 8200 West Seven Rivers Drive e Crystal River, Florida 32629 A Florida PmQtens Company
e , The following-comments sumarize items noted by the CR-3 staff during a review of the Senior Reactor Operators examination given on December 17, 1986, and are submitted for your consideration in grading this examination. CATEGORY 5_ l QUESTION # 5.09 COMENT: This question requests that the, operator " State the two 12), most positive indications" for a loss of natural circulation. At present, there is no approved reference which defines the two most positive indications". The reference listed (B&W Technical Ecument) has not been approved by the NRC for operator training and therefore cannot be used as a stand alone reference. RECOMENDATION: AP-530 lists the indications which the operator should use to verify natural circulation (see attachment 1).
- Accept any two of the indications listed.
QUESTION # 5.10 COMENT: See comment and recommendation for 5.09 above. QUESTION #5.22 COMENT: This question request the candidate to supply a list of symptoms for a specific plant condition without adequate information to identify that condition. While it is granted that due to the recent emphasis on this condition, the candidates on this exam most likely identified the correct symptoms, it would be unwise to assume that the same will always hold true in the future. RECOMENDATION: No action is requested on this exam, however, it is requested that this question be reworded prior to future use. CATEGORT 6, QUESTION i 6.02 COMEliT: This question states "Setpoints are NOT required." The answer key lists setpoints. RECOMENDATION: Delete setpoints from answer key.
QUESTION #6.10 COMENT: To answer this question, the candidata must assume either a bare GM detector -(as they would be with. no special features). or the instruments actually in use at CA-3. The question does not specify which is correct. While the answer specified in the key is correct for the first case, all instrumnts in use at CR-3 are designed to prevent this occurrence, see attachment 2.
- RECOMENDATION
- Accept either answer A or C as correct.
QUESTION # 5.12 COMENT: This question requests five diesel start permissives. While it is true that the conditions listed in the key must exist for auto start of the diesel, there are two additional conditions required by the diesel start sliw it, see attachment 3. RECOMENDATION: Add <6 psig lube oli and <250 rpm to the list of answers and ac:eot any five. QUESTION #6.13 COMENT: The DC and SW system surge tanks have a dual " relief" system. Durin operation, bleeder valves (shown as relief valves on attachment 4)g , will relieve excess pressure to the waste gas system. Operation of these valves is not considered abnormal. The safety relief on the tank relieves to the ventilation system through filters. RECOMEN0ATION: Change answer key to read: Vents to WG system Relieves to ventilation system. CATEGORYJ, QUE5 TION # 7.03 COMENT: Question states " State one (1) of these three". Answer key lists three responscs with a value of .6 each, (question is worth 1.5 points). RECOMEN0ATION: Accept any one response for full credit. l QUESTION # 7.04 COMENT: At CR-3, the use of service water to feed the OTSG's during a loss of feedwater is not a desired or workable alternative due to plant design, see attachment 5. RECOMENDATION: Delete question from exam.
\
QUESTION #7.13 COMMENT: Response "A" should be 24" Vice 30", see attachment 6. RECOMMENDATION: Change answer key as indicated. QUESTION #7.15 COMMENT: Question asks for two actions which are required prior to securing a RCP per 0P-204. There are actually three actions, see attachment 7.
- 1. Reduce to <60% power.
- 2. Assure that TC is selected to the unaffected loop.
3 Select pressure control from the loop with one RCP. RECOMMENDATION: Allow full credit for any two of the above. CATEGORY 8
-QUESTION #8.03 COMMENT: Question asks for five events wnten are reportable 1mmediately or within one hour. Answer key does not address requirements of 10 CFR 20.403 which contains imediate notifications, see attachment 8.
RECOMMENDATION: Add 10 CFR 20.403 requirements ~to answer key. QUESTION #8.06 COMMENT: We do not feel the operator should be held responsible to list this i type of information from memory. This 'is not an' item which the i operator can control or one which will affect the ability of the Fire
- Briilade to respond as long as there are a sufficient number of qua'ifled members present.
Also, Butiding service is the subset of Maintenance from which members i ere drawn. i- RECOMMENDATION: Delete question from use on future examinations. Accept either Building Services or Maintenance as correct. QUESTION #8.10 COMMENT: Second part of answer should read: Nuclear Plant Manager or " Man On Call", see attachment 9. RECOMENDATION: Revise key as indicated. 4 1
- - + + - - ** wy,-pyems *ep e , s-i,wpyp pewm,y m ywt enny-.--ge.-y y-
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8<, is 44 o ny - ~'s o r r u s t iir o ' p . i 6' jg - s _ 4.4 an operater may fill out the clearance order with a Licensed } operator ' review sad www.srance.. to indienee eeneussence. the I,1 ceased Operator adst sign outside the ' Operator Recording Order' block. The respor.sibilities include researchlag all cisarasses and lasuring chan the number and type of togs issued, and their intended leestica, is such that the persea g classification involved is the clearance cas perform the activity without the risk of persossel or equipment safety. l This, must be done by first guestioalag the person ng classifi-cation requestin, the elaaranee/ fag to determine the scope of the latended clearance / fag and them to still e whatever is necessary to accurately record the Clearance. I 4.5 The plaat Review Committee (PAC) sust review and approve any Clearance, prior to issuance, whlen asets any coadition speci. fled belove
======L =^=-rand am,
- a. The clearames is to he issued for aa or ahmasmal evale* fan (i.e., repair of RCf=11 la other than Mode $ er 6, and other emergemey. repairs).
- b. The clearance to ne issued sammet meet the danbla yalva
- agagasting guidelines of 1500 peig andlar 2 200F, and l.5'
~
greater than er equal to f/2 Lash diameter opealag. 15 IICIpT!0m The Clearance Aetherity any, with approval of the l Raa-ca-call, approve Clearances which de met meet these guidelines. If the guidelines casset be set, l5 4
- and approval is gives, it shall be annotated la l
,Iten 5 by writing or stampias la red lak 'Does Wet l Meet Deeble vale saideline' initialed and dated by {
l ! the clearance authority. l 1 ] cp-115 Rev. 58 Pa** ' l
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'QUE$fl0N#8.19
-COMMENT: The double valve protection guidelines have been rewritten. Also, the requirements in the exception to this rule are reworded. Both are i recent changes and may not have been reviewed by all candidates as of the exam date, see attachment 10.
RECOMMENDATION: Revise answer key to accept either wording. GENERAL During review of this examination. It was noted that several questions, particularly in category 5, referenced documents which are either not readily available or approved for use. These documents include the B&W Technical Document. Emergency Procedures Technical Bases and the training material utilized at other B&W plants. While it is
. true that most of this material is common to each plant, there are those areas where l designs differ. We feel that questions used on the examinations should come from the c.aterial supplied by Florida Power Corporation. If additional material is needed to crite the examination, a request should be made to Florida Power Corporation.
l e I 9
NC REV 05 Date 10/O3/86 AP-530 FOLLOW-UP (dont'd) ACTIONS DETAILS
- 5. E adequate subcooling margin Establish HPI:
h agt exist,
.T.Hfd i
- 4. Establish HPI a. Open:
1 o MUV-73 o MUV-58 ' 1
- b. Ensure required CTSG b. Start 2 HPI pumps !
levels
- c. Refer to AP-3'80, c. Open:
Engineered Safeguards Actuation. o MUV-23 o MUV-25 .
, o MUV-24 o MUV-26 l Required CTSG Levels W/0 RCP CONDITIONE LEVEL i
- o Adequate Subcooling 65%
Marcin o Less Than Adequate 95% refer , I subceolina Marcin to AP-380 o L2 HPI Pumps 95% Available
- 6. E verified, natural circulation is Verify natural circulation by observing:
$$f.HgotoFollow-upstep o Te =TSAT of OTSG.
- 17. .
,. o AT (incores - Tcold) develops 1 and stabilizes j l
o The average of the $ highest 1 inceres follows TH within l 10'T. l o Tg, Te, and Incores lower when OTSG pressure is lowered. L , AP-530 Page 5 of 13 NC l I
u -- . _ _ _ _ _ - - - t FI4RIDA PONER CORPORATION CRYSTAL RIVER UNIT 3 NUCLEAR OPERATIONS TRAINING DEPARTMENT PROGRAM: REPIACEMENT OPERATOR TRAINING COURSE MITIGATION CORE DAMAGE 1 LESSON: RADIATION MONITORING IA880N Not ROT 12 P LEss0N LENGTNs 1 BOUR RE7ISION: 1 DATE: 9/3/86
. .-.d .A?em-Nudlear I tructor Approved sys_ XMS, Baviewed prior to use (/ Nuclear Training ~
Date Instructor 8 gnature supervisor Reviewed and
)
concurred with
' ruoleat C p-ations ~
Training support Supervisor e
[-- l monitoring, it is important to know that a very high input rate may paralyze the counters as the input rate j increases, the observed count-rate reaches a flat maximum, stays fairly constant for a while, and then drops. h [g) some G-M detectors have a built-in holding circuit. The function of the holding circuit is to take the detector response, once it reaches about 3/4 of full scale, to a full scale value and hold it at this value until the strength of the radiation field is reduced. Whether a G-M detector has a holding circuit, or not, saturation in high radiation fields does not damage the detector. The detector will recover once the high radiation field is reduced. u 2.1 asnavIon gt &cINTILLATICN DETECTORS IN HIGN RADIATION 912L28 If the s,cintillation detector is used as a pulse-type instrument, similar to the a-M detector, the resolving time, , is on the order of 1 second. A random coincidence describes an event in which two unrel,ated
-rays (i.e., not belonging to the same disintegration) are detected within the resolving time of the detector. This effect could result in detector .
saturation similar to that of the G-M detector. However, due to the extremely short resolving time of . the saintillation detector, saturation is not considered k, very likely in any radiation fields that could exist in Page 5
286 o Pulses formed in the counter probe are fed to the c' rate meter circuit. The meter reading then gives the average pulse rd in the G-M tube. With these circuits, the device needs no zero con ~ or warm-up periodu However, some counters "sa_turate" in a ve radiation fi_ eld.[That is, the pulse rate becomes so high that the co' Mutt faiIs to function,Jroperly. As a result, the meter reada' sero rather than off-scale.) The condition of saturation can also ruin the woMe of the rapid loss of some types of quenching The G-M survey meter may be used to detect beta gamme . The tube thickness is about 30 mg/cm2. This is much too ' thick to allow alpha penetration, but the device detects beta of E 5 6 MeV. The sliding' shield (~1500 mg/cm 2) rejects all beta from norm sources. For low-energy beta, an end-window type G-M survey me l can be used. In an end-window tube, one end of the cylinder has ' l very thin covering, which can be used as a window, Mica , a few. , thick, is quite often used as the window substance. These devi spond to beta of E > ~30 kev (22)and alpha of E.>~3 MAW (24)gg , window is very fregile, care must be taken in using these devices. End-window counters are used extensively to detect 14 C, which gives off a low-energy beta. We must be always aware of the highly directional beta response for both the end-window and e,ther 0-M survey mete'rs. The G=M survey meter only detects beta if the open area in the sliding shlald is facing the source. '. o b. Enercy Denendence. The G-M survey meter is not a precise instrument exposure-rate measurements. The response of this device is not di-rectly proportional to the energy absorbed in the sensitive volume.(33 , The energy absorbed per unit mass in any medium is a function of th, I photon energy fluence (t) and the mass energy-absorption coeffioten, (en/p).UThis is true if the secondaries do not have too high an ,, e nergy. The count rate for a G-M survey meter depends upon the efficiency of the counter (counts per incident photon). This, in turn is a function of the cathode material. Regardless of the cathode a stance used, the count rate for a given exposure rate is not constant photon energy varies. Thus, the response of these instruments in to 0 9***# * * " O e e ,g,, og eeme ees* -e44e 8***M88 8 8
286 o Pulses formed in the counter probe are fed to the c" rate meter circuit. The meter reading then gives the average pulse ti in the G-M tube. With these circuits, the device needs no zero con'
- or warm-up period However, some counters " saturate" in a ve radiation field. (That is, the pulse rate becomes so high that the co' d6cuit fails to functionyperly. As a result, the meter reads" sero rather than off-s_cale. The condition of saturation can a*so ruin thTtubETeEbe of the rapid loss of some types of quenching
'Ihe G-M survey meter may be used to detect beta j gamma. The tube thickness is about 30 mg/cm2 This is much too " l
' thick to allow alpha penetration, but the device detects beta of E 56 MeV. The sliding shield (~1500 mg/cm 2) rejects a!! beta from norm
~
sources. For low-energy beta, an end-window type G-M survey me can be used. In an end-window tube, one and of the cylinder has', very thin covering, which can be used as a window, Mica, a few , i thick, is quite often used as the window substance. These devi spond to beta of E > ~30 kev (22)and alpha of E >~3 MeV.(24)3g,,, window is very fragile, care must be taken in using these devices.' l i End-window counters are used extensively to detect I 14 C, which gives off a low-energy beta We must be always aware-of the highly directional beta response for both the end-window and other G-M survey me,te'rs. The G-M survey meter only detects beta if the open area in the sliding shield is facing the source. I
.b. Eneray Decondenes j
The G-M survey meter is not a precise instrument exposure-rate measurements. The response of this device is not di-rectly proportional to the energy absorbed in the sensitive volume.I33 , The energy absorbed per unit mass in any medium is a function of th, t photon energy fluence (t) and the mass energy-absorption coefficieqt (pen /p).UIThis is true if the secondaries do not have too high an , energy. The count rate for a G.M survey meter depends upon the efficiency of the counter (counts per incident photon). This, in turn is a function of the cathode material. Regardless of the cathode sub-stance used, the count rate for a given exposure rate is not constant photon energy varies. Thus, the response of these instruments in to
. . . o \
l FLORIDA POWER CORPORATION l CRYSTAL RIVER UNIT 3 NUCLEAR OPERATIONS TRAINING DEPARTNENT l PROGRAM: REPLACENENT OPERATOR TRAINING COURSE: SYstEN TECENOLOGY 1 LESSON: DIESEL GENERATOR LESSON NO: A07-4-6 LEs$0N LENGTE: 4 NOURS . AIVISION: 3 OATE: 10/9/86 l l l , . I i Rawlauq( Igist in East PREPARED BY: 2M_1
. Date Instructor signature Nuclear Instructor j REVIEWED and CONCURRED with:
#4M] Lif i
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+ i A07-4-6 Fafe 65
NOTE: A locally started engine vill respond to a remote (control room) STOP cosaand, with the '43' switch in ' MANUAL', the affected diesel will not be
- able to respond to an Es actuation, or loss of power to the Es bus. With the '43' switch in ' MANUAL' and the RS1 switch in
'WORMAL', only low speed (300 RPN) remoto (control room) starting of the diesel is permitted.
There are three ' auto start' signals:
- 1. ES actuation (NPI)
- 2. IS bus degraded voltage -
i .
- 3. ES bus undervoltage (loss of voltage) ,
Permissives required for ' auto start'
- 1. DIEstL START N00E SELECT switch (43) on main control board (one for each diesel), selected to AUTO. .
- 2. Air shutoff valve (EGV-35 (A) E0V-39 (5) open.
- 3. CONTROL AT ENGINE - NORNat, switch (on the er.gine gauge panell selected to NORMAL.
I i
- 4. 86 lockout relay (generator control cabinet) reset.
. $. 50R seal-in reset (accomplished by Pressing the RESET push button (Ps4)) located on the engine gauge panel.
1 i < i i ROT-4-6 Page 11
~ , . . - . . . - - - - . . - , - . - . , . . . . . - . - - - - - .
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FLORIDA POWER CORPORATION CRYSTAL A!YER UNIT 3 - NUCLEAR OPERATIONS TRAINING DEPARTNENT PA00RAN: REPLACINENT OPERATOR TRAINING
-COURSE: SYSTEN TECNNOLOGY LESSON: PRINAAT SUPPORT STsTINs REVIEW LESSON NO: ROT-4-2 LESSON LENGTE: 4 500RS ARV!sION: 1 DATE: to//J/f4 taviamed Ezing h Bag PREPAR G ST: MMO4 ' 7 _' -
i Nuclear Instructor Date Instructor Signature
- AEVIEVID and -
/ ,f CONCURRED with: /4+tuis
(/NuclearTraining 1 - - -
,s.,ervisor APPROVID ST: 6
" Nuc14ar operations
, , Tralelag support Supervisor h
cw <*/a/a 3k . l l l
CEApTER 1 cooling water flow through the HX while keeping total DC flow constant. These velves are controlled with a manual confroller in the control Room or from ES 4160V switchgear rooms A and 8. The bypass type regulator is used instead of throttling flow to the II to maintain total DC Systes head constant. The bypass valves fall shut to ensure any valve fault will result la full cooling flow to its NX. NOTE: All other DC coolers are regulated by assual valves on the outlet of the coolers. l 1.4.4 surse Tank fact.-1A and. -1M A $000 galles pressurised surge tank is installed la each systes. I the tanks are pressurised with altrogen to prevent pump l cavitaties. Upon reaching a pressure of 13, pels, bleeder valves
; vent to the relief valve header through charcoal' filters. Nitroges is added to the surge tank if pressure falls below 5 pois. The
'l tank may also be vented to the Weste Oas systes if the system is i contaminated. The surge tanks are located la the Beat Eschanger i Rooms of the A8 on Ef. 95' (Figure 10). I 1.4.5 2131m i a. DC - Valded carbon steel, i'
- b. Sea Water F14 aged carbon steel with PVC or polyurethane i
.; lining. l
; R07 4 2 page 5
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- C.K & W..it.COX we TECHICAL 00CNElli I. - The BWST Inventory should be carefully monitored when this vuode of depressurleatten is utillsed for entended periods of ties.
[ d. Total toes of feedvetor (NN and EFW) will require a solid f water cooldown without the benefit of secondary heat re-moval.- At tempt e should be made to obtain feedwate r frois any avelleblo source, however IF1 eesting to preferable te (,PW4544**bf'adHeesdessatar.giede feedveter't For cases of thle type wh(eh require eseessive seeldown times, SWIT inventory should be carefulty monitored. NP1 fross the EW8T will flov (nto the affected Ofs0 and will not be returned
- to the reseter bul(dias susp. Therefore, backup water suppuen for the BW87 may be nosassary and should be prepared. Offette releases may be very signif! cant.
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I DAT8: 10 1-83 Appendia c, rese C-31 'I ,
, ,, , FIARIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 NUCLEAR OPERATIONS TRAINING. DEPARTMENT PROGRAM: REPIACEMENT OPERATOR TRAINING COURSE SYSTEMS TECHNOIA0Y i LESSON: ENERGENCY TIEDNATER AND EFIC LEss0N Not ROT-4-15 LESSON LENGTN 8 NOURS REVISION: 1 ISSUE DATE: 13-3-86 Frepared Byt 7D # 9/3-% _ , loleer Instructor Reviewed end Concurred Withs a%J,JJ U.4 =Unk ) Nuclear Trein16 ~2 Academio iallet Approved By: N mis i yt,earTraining Supervisor i Reviewed Prior it 313,l,ture Data Instructor signa ' 4
/
. l blinking. Tho bypoco tunt b3 icplcocntcd b0fere oither OTs0 pressure reaches 400 7820. This bypass feature bypasses the EFW initiate and the Main steam and Feedwater Isolation functions.
The other bypass that must be implemented is the EFW initiation based on lose of all RcP's. The NI/RPS provides a bypass permissive signal when reactor power i decreasse below 54. The actual bypass function is l l performed in the EFIC cabinets by placing the "RCP l 1 BYPASS / RESET toggle switch located on the " Initiate a modules in all 4 EFIC cabinets acaentarily in the typass i * (Up) position prior to turning off the last Rcs. The ! bypass may be verified by observing that7hr #acy l SHUTDOWN" light on the sabinet alara panels is blinking. l 4.3 Abnormal operations I The EFW / EFIC systems were designed to handle the . ! following abnormal eventes , I. a) 14ee of main feedwater b) loss of main feedwater with loss of offsite power.
, c) Lees of main feedwater with loss of offeite and onsite Ac power.
l d) Main feedwater line break e) Main steam line break /EFW line break f) Small break IACA 1 j , 4.3.1. Loss of Nein Feedwatera Upon less of all main feedwater l 4
- both EFW pumps are automatically started by the EFIC l 1
- system. A minimum flow of 740 GPM is suffielent to mitigate the effecte of a less of main feedwater event. l 4
This a11e6we e sin;1e failuss.en.am ERM.tsnin.to occur mad . .. still meet design requiremente. After EFW initiation, the I l steam generator level will be automatically controlled at i i 39 i
ch;ut 24". Tho cnly cp0rotar cchicne aro to etnfira that EFW flow has been initiated and that a' level has been established in both OTsG's. 4.3.2 fees of Main Feedwater with Lgga of offsite Power = Upon loss of offsite power (which causes loss of all Main Feedwater pumps and RC pumps), the EFW system will be - used to establish natural circulation in the RCS. The turbine driven pump will be started when the system is initiated and ASV-5 and AsV-304 are opened . The motor driven pump (EFP-1) start is delayed until 5 seconds , af ter the diesel generator breaker closes or, if a NFI actuation is present, it is delayed until the block loading sequence is completed and then started with a 5 second time delay. The total time, including the 10 seconds required for the diesel start, from signal to pump start is 30 seconds. While the motor driven EFW pump is operating and powered gggs the diesel generator, the load on the generator is monitored and it it exceeds 3000 KwAan alarm is actuated that informs the operater that the diesel generator is above it's 30 minute rated load. When this happens, a 30 minute timer is aise started. If at the end of 38 minutes the load on the diesel has not been reduced below 2000 RW, another alarm will be received that the diesel generator is at 35 minutes of it's 30 minute rating. If af ter 30 minutes the diesel load has not been reduced below 3000 MN, the motor driven EFW pump will be tripped. 30 1.
7 Rev. 44 11/15/04
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1 N'?\Ap"**,3.h,$! U s .'.'}l,. OPERAf!NG PROCEDURE 09-2M Ff,0RIDA POWER CORPORAf!0N A CRYSTAI,R!YER UNIT 3 POWER ofERAf!0N i '. 1 TIIS FROCEDURE ADORESSES SAFETT REIATED CONp0NINTS i ) i l l } \ I APPROYED ST: Responsible Sectica Superlatendeat/ ! 4 Supervleer ! _A hYW_ ! , Date #4///fd i i *
- IWftRfAffAf!0W CONTACT
- Nuclear Operations Superlatendent t
I e 1
._.---..- -- -.- ,-- _ - - - - - - . ~ ~ ~ ~ -
.a. . .
4.1.1.3 The naminua rate of power increase below 20t FP shall not escoed 10% per hour. Above 20% ft, Transmission systes limits shall apply. 4.1.2 fahalance and Quadrant power filt Limits. 4.1.2.1 Malatala power labalance per Nuclear teactor specialist's Osidence. 4.1.2.2 If the Operator fails to move er naves the asial power shaping rode (A,79A's) la the wrong direction, the action could result la a reactor trip. 4.1.2.3 ApSA's shall not be inserted or withdraws more thaa sa lacre-mest et $%, after which power imbalance shall be verified. 4.1.2.4 7e alaisise p=vst tilte, e:Latala..sc6.pesha.wih..s. woup-... at the same level. 4.1.2.5 prior to perforalag a heat balance, iseure quadrant power tilt is < 1 2%. 4.1.2.6 !asure Quadreat power filt is within the lialts of s.t.s. l4
- 3.2.4. l l 1 .
l 4.2 nearTom cooLast stattu Lim 1ft l
- 4.2.1 Thate shall he at least three (3) ACp's la operation.
( 4.3.3 Reduce reseter power te ( 60% when lattiatlag four (4) to three (3) RCp opera'ies. t Assure prior to tripplag the RCp that the '
, fg transaltter is' selected for the unaffacted cold leg. For RC pressure contzel, seleet the loop :ith : e fit att &c.ths ses ---
(resetet protective systen) per Op-901, Rosetor pen-Nuclear Instruncataties, festion 7.2. Op-204 seg. 5 '44 page 5 .m u_.___________
Rev. 34 08/29/86 sffective cate 74 la _ Document Section NFORMC. ATION R. Nuclear ONL _t COMPLIANCE PROC 100Rt CF-111 FLORIDA POWER CORPORATION CRTsTAL RIVER UNIT 3 DOCUMENTING. REPORTING, AND REVIEWING NONCONIORMING CPERATION5 REPORTS ( . } Turs FRocsness accassses n0s-sArm Ass.atun Conreams APPROYtB RY: Respanalble Section Superintendant Supervisor mW
..t. 4thc r g- -
INftRPRETATION CONTACT: Nuclear safety & Reliability Supt, esen. e.ed.eppe t Smumpe# # # * .* _ - - - - _ - _ , _m-, _-_m.. . , _ , . __.-. - _,,.. _ . _ _ _ _ , . . , _ . _ _ _ , _ __ _mm, _ _ _ -_ . _ _ , ...m.
s o. . l NCOR IMMEDIATE NOTIFICATION REQUIREMENT! ENCLOSUSE 4 (Continued) (Page 1 of 10) C. 10 CFR 20.402 Reports of Theft or loss of Licensed Material
! mediate Notification: CR-3 $nall report to the NRC Operations Center via the Emergency Notification System (alternate means -
comercial telephone (202) 9510550) feediately after CR-3 determines that a loss or theft ."of licensed material has occurred in such quantities and under such circumstances that it appears to CR-3 that a substantial hazard may result to persons in unrestricted areas. U. 10 CfM 20.403 Notifications of Incidenta
- a. ! mediate Notification: CH-3 must imediately report to OHMS (see EM-206 for phone nebers) and shall immediately report to the NRC Operations Center via tne Emergency Notification System (alternate means - comercial telephone (202) 951-05$0) any events involving by-product source, of special nuclear material possessed by CR-3 that may have caused or threatens to cause:
. (1) Expasurg of_ tbo wac14.Mdy. Af. ao. individual..to 25. rems or more of radiation; exposure of the skin of the whole' body of any individual of 150 rems or more of radiations or exposure of the
' feet, ankles, hands, or forearms of an individual to 375 rems or more of radiation; or (2) The release of radioactive material in concentrations which, if averaged over a period of 24 hoiers, would eaceed 5000 times the 11mits specified in 10 CFR 20, Appendix s, Table !! or (3) A loss of one working week or =re of the operation of any facilities affected: or Cp-111 .... say,
- 34. ..
pace 31 .
. b, .
NCOR IMEDIATE NOTIFICATION REQUIA(MENT 5 (NCLO5URE 4 (Continued) (Page 8 of 10) (4) Damage to property in excess of $200,000.
- b. Twenty-Four Hour Notification: <A-3 must tunedtately report to DMR$
(see (M-206 for phone numbers) and shall within 24 hours of discovery of the event, report to the NRC Operations Center via the Emergency Notification System (alternate means - anamercial tripphane (202) g51-0550) any event involving Ilcansed material possessed by CR-3 that may have caused or threatens to causes (1) Exposure of the whole body of any individual to S rems or more of - ! radiation; exposure of the skin of the whole body of any I individual to 30 rems or more of radiation; exposure of the feet, ankles, hands, or forearms to 75 rems or more of radiationi or (2) The release of radioactive material in concentrations which, if i averaged ever a period of 24 hours, would exceed 500 times the Itatts spectfled for such materials in 10 CFA 20. Appendix I, , Table !!! or .
,(3) A loss of one day or more of the operation of any facilities affected er -
2 (4) Damage to property in excess of $2000. i E. 10 CFR 50.36 Technical Specifications i All reports performed under this requirement shall be made to the MC Operations Center via the Emergency %tification $ystem (alternate means - connerical telephone (202) 951-0550).
- a. Safety Limit (ST5 2 0): If any safety limit is exceeded, CH 3 shall l
- nottfy the NRC Operations Center as required by sections A.b.i . (10 j
Cp-111 Rev. S4 page 32 i
9
*>*
- 2.1.3 The personnel assigned to shift cperati!ns perfora three general functions:
- 1. Continuous normal operation of the plant and its associated equipment, including normal planned power changes, startups, and shutdowns.
- 3. Maintala the plant la a safe condition during abnormal conditions.
- 3. protect the health and safety of the public, plant personnel, and plaat equipment during and folloviaq an energency situation.
A Senior Reactor Operator.(SRO) shall be la the control Center during Modes 1 through 4. All major plant operations are conducted from the control Center with the Nuclear Shift Supervisor la consand. , The operatlag shift, consistent with the provisions of 709AN (plant Operating Quality Assurance Manual), effects the I,oad 81spatcher er other authetised e orders. 3.1.4 Authority to shut deva the resetor rests with:
- a. Nuclear Operator
- b. Chief Nuclear Operator
- s. Assistaat Nuclear Shift Supervisor
- d. Nuclear shift supervisor
- e. But!4ar Operatless Superlatendeak l f. Nuclear plant Manager or 'Naa On-Cali" Authority to start up and retura to power operation rests with
- a. Neelear Operations Superlatendeng i b. Nuclear plant Manager or 'Nas ca-Call' I
1 1 I AI-900 Rev. * '8 5 pese 3 I
. w.. /
9 2.1.3 The personnel assigned to shift operations perform three general functions:
- 1. Continuous normal operation of the plant and its associated equipment, including normal planned power changes, startups, and shutdowns.
- 2. Maintala the plant in a safe condition during abnormal conditions.
- 3. protect the health and safety of the public, plant personnel, and plaat equipeent during and following an energency situation.
A Senior Reactor Operator.(SRO) shall be in the control center
, during Modes 1 through 4. All major plant operations are conducted from the control Center with the Nuclear Shift Supervisor la cosaand. , The operatlag shift, consistent with the provisions of p00AM (plant Operating Quality Assurance haual), effects the I,oad Dispatcher or other authorized orders.
2.1.4 Authority to shut down the reactor rests with:
- a. Nuclear Operator ,
- b. Chief Nuclear Operator
- c. Assistant Nuclear Shif t Supervisor
- d. Nuclear Shift Supervisor
- o. Nuclear operations Superlatendent j f. Nucisar plant Manager or *Maa On-Call' Authority to start up and return to power operation rasts withi
- 4. Nuclear Operations Superintendeng
- b. Nuclear plant Manager or 'Nea Ca-Call' AI 500 Rev." ) $ page 3 e
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