ML20138R886
| ML20138R886 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 11/04/1985 |
| From: | Munro J, Wilson B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20138R883 | List: |
| References | |
| 50-416-OL-85-01, 50-416-OL-85-1, NUDOCS 8511190199 | |
| Download: ML20138R886 (160) | |
Text
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D ENCLOSURE 1 EXAMINATION REPORT 416/0L-85-01 Facility Licensee: Mississippi Power and Light Company P. O. Box 23054 Jackson, MS 39205 Facility Name: Grand Gulf Nuclear Station Facility Docket No.: 50-416 Written, oral, and simulator examinations were administered at Grand Gulf Nuclear Station near Port Gibson, Mississippi.
Chief Examiner: [/
John F. Munr0" f 2_- /o///!f[
Ddte Signed Approved by:
Brif hu !! [4 [77 A. Wilson, Section Chief Date~ Signed Suninary:
Examinations on August 12-16, 1985 Written, oral and simulator examinations were administered to 13 candidates; six of whom passed.
J 8511190199 851112 PDR ADOCK 05000416 G PDR
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REPORT DETAILS
- 1. Facility Employees Ccntacted:
- J. Cross, Plant Seneral Manager
- G. Lhamon, Operations Training Supervisor M. Shelly, Simulator Supervisor
- K. Beatty, Operations Training Superintendent C. Bottemiller, Training Instructor
- J. Yelverton, Manager of Plant Support
- J. Robertson, Operations Superintendent
- M. Wright, Plant Manager for Operations
- R. Rogers, Technical Support to Plant General Manager
- J. Bailey, Compliance
- Attended Exit Meeting
- 2. Examiners:
- J. Munro, NRC S. Guenther, NRC K. Brockman, NRC
- Chief Examiner
- 3. Examination Review Meeting At the conclusion of the written examinations, the examiners provided G. Lhamon, Operations Training Supervisor, with a copy of the written examination and answer key for review. The following comments were made by the facility reviewers:
- a. SR0 Exam (1) Question 6.06 - The question is unclear since it does not specify that only the referenced annunciator was received. If it was assumed that other annunciators were received, then answer b would be correct. Since students cannot state assumptions on multiple choice questions, full credit sbould be given for answer b or c.
NRC Resolution - The question wording is straightforward and indicates the only conditions that are abnormal with the plant
. . . operating normally at power . . .". No other assumptions are required for the examinee to conclude the correct response. No change to the answer key is warranted.
Enclosure 1 2 (2) Question 6.23 - The answer key for this question references GGNS Lesson Plan OP-N32-2-501. Although the answer given by the key is in one portion of the lesson plan, a more complete answer is presented in section 1.4.a.6.d of the lesson plan. This section lists a Turbine Trip as well as the two answers given in the key.
Therefore, a Turbine Trip should also be considered a correct response. A copy of the lesson plan is attached for your reference (Attachment I).
NRC Resolution - Section 1.4.a.b.d of the referenced facility lesson plan supports the inclusion of a Turbine Trip as a correct response. The answer key has been modified accordingly.
(3) Question 7.02 - This question requested the 3 immediate operator I actions following a failure of rods to insert on a scram with power at 6%. ONEP 05-1-02-I-1 requires the operator to enter EP-10 if power is greater than 5%. Only if power is less than 5% ,
does the referenced ONEP give other immediate actions. Therefore, l it is requested that consideration be given to accepting reason- '
able actions from EP-10 as correct responses, as well as those other actions listed in the ONEP. A copy of ONEP 05-1-02-I-1 is attached for your reference (Attachment II).
NRC Resolution - A typographical error caused the incorrect i presentation of the question as detailed in the comment. The question is invalid as written. Delete from exam.
(4) Question 7.06 - This question asked for three (3) conditions to be met prior to restoring a system to service following an automatic isolation per 0NEP 05-1-02-111-5. The referenced ONEP does not require three conditions, but rather lists only two conditions for verification of system integrity. These two conditions are:
a) verification that the system is intact; and b) verification that the operation of the system will not result in an uncontrolled release to the environment.
The remaining verbage of the sentence states methods by which these two conditions can be verified i.e., visual inspection of accessible areas and/or observation of available Process Radiation Monitoring instrumentation and other available indications for inaccessible areas. Since the caution statement in the referenced ONEP is admittedly poorly worded and examinees were undoubtedly confused by the question, it is requested that this question be removed from the examination. If not removed, generous latitude should be given when considering other reasonable answers. ONEP 05-1-02-III-5 is attached for your references (Attachment III).
Enclosure 1 3 NRC Resolution - A conservative reading of the procedure indicates that three conditions, as per the answer key, should be ve-ified prior to system restoration. Procedural shortcomings with regard to clarity should be evaluated by the facility and appropriate changes effected. No change to the answer key is warranted.
(5) Question 7.09 - 101 03-1-01-1 actually lists 9 conditions in section 6.2.17 to be met prior to transfer to run. Any four (4) of these nine requirements should be given full credit, rather than the 5 listed in the answer key. 10I-03-1-01-1 is attached for your reference (Attachment IV).
NRC Resolution - A review of section 6.2.17 of the referenced procedure and the question wording supports the inclusion of additional administrative conditions as correct responses. The question will be reworded for future use to distinguish between i plant parametric conditions and administrative conditions. The answer key has been changed accordingly.
(6) Question 7.11 - Administrative P: ocedure 01-S-06-2, section 6.6.2.d states that operators r.eed only to memorize the entry conditions for EP-1, EP-3 and EP-10, other Emergency Procedures then are entered from these. This is done to minimize the effort-necessary to recognize an entry into an Emergency Procedure and therefore, minimize the chance of human error. Although operators are to be generally familiar with the Emergency Procedures other than EP-1, EP-3 and EP-10, they may not know the precise entry conditions.
Technical Specification 3.6.3 gives a limit of 120 for suppres-sion pool temperature before requiring depressurization of the reactor.
Because of the similarity between the Technical Specifications and the requirements of EP-5 and due to the proximity of answer b to the limit of Technical Specification 3.6.3, we are concerned that the examinees may have confused the requirements and bence answered b vice d. It is requested that answer b be given con-sideration for full credit. A copy of Administrative Procedure 01-S,-06-2 is attached for your use (Attachment V).
NRC Resolution - The depressurization directed by the action statement of Technical Specification 3.6.3.1 is allowed to be completed within a 12-hour timeframe and is therefore not analogous to the " Rapid Depressurization" required by EP-5. The question is presented in a multiple choice format to determine the !
examinee's understanding of the procedure's entry conditions by l recognition, rather than by memorization. No change to the answer i key is warranted. )
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Enclosure 1 4 (7) Question 7.20 - The answer given in the answer key reflects the system limitation (65 mw) as given in the caution note prior to step 5.2 of the referenced 101 03-1-01-2. However, step 5.2 gives a procedural limitation of 25 mw. Therefore, full credit should be given for an answer of 25 mw, as well as the answer given in the key. 101 03-1-01-2 is attached for your use (Attachment VI).
NRC Resolution - Since the question did not specifically request a systematic limitation, the procedural limitation value of 25 mw will be considered as an acceptable answer to part a. Addi-tionally, an alternative answer of "no available indication" will be required for part b to ensure consistency between the conditions of the question and an answer of 25 mw. The answer key has been changed accordingly.
(8) Question 8.06 - The procedure referenced states "When performing electrical lineup and when applicable, verify ...". (underline ours). Since the question did not reference any applicable or specific breaker, consideration should be given to other reason-able answers and credit given for these other answers as appropriate. A copy of procedure 02-S-01-2 is attached for your reference (AttachmentVII).
NRC Resolution - The question is clearly worded to elicit generic requirements of a governing administrative procedure. No change to the answer key is warranted,
- b. R0 Exam (1) Question 2.06 - The GGNS Lesson Plan referenced (OP-P33-501) was
( in error in stating a sampling subsystem station was located on.
l the refueling floor. It should have stated the sample station was
! on the 185' elevation of the containment. Since the incorrect l reference to the refuel floor in the question could have caused <
f the examinees confusion, consideration should be given to reason-able answers other than those listed in the key.
NRC Resolution - Acknowledged. The incorrect facility reference j material invalidates the question as written. Delete from exam.
! (2) Question.2.07 - The lesson plan referenced gives three limitations imposed on the Standby Service Water (SSW) System by a loss of instrument air:
(a) causes all air operated valves to fail shut (b) loss of make-up water to basins (c) isolation of fill tank
x Enclosure 1 5 It further states that loss of air does not affect overall operation of the system. The hypochlorite and acid additions were not mentioned in the lesson plan since these operations are performed manually, rather than by using the installed systems.
Therefore, reasonable answers other than the loss of chemical addition capability should be considered.
NRC Resolution - The' loss of the ability for both sulfuric acid and hypochloritec injection is detailed by both the facility system description P-41 and the Off-Normal Procedure for " Loss of Instrument Air", 05-1-02-V-9. Alternative answers will only be accepted if consistent with the Automatic Actions of the " Loss of Instrument Air" procedure as it pertains to the SSW system. The answer key has been modified accordingly.
(3) Question 2.20 - As well as the answer. listed in the key, Technical Specification Table 3.3.7.1-1, item 6, footnote (h) notes other conditions rather than just hi-hi activity in the air intake duct (such as two downscales) which will result in an automatic isolation of the Control Room Ventilation System. Responses consistent with this Technical Specification should be given full credit.
NRC Resolution - Acknowledged. The Control Room Ventilation Radiation Monitor signal logic combinations described by Technical Specification Table 3.3.7.1-1 will be considered acceptable alternative answers. The answer key has been modified accord-ingly.
(4) Question 3.17 - Same comments as for Question 6.23 on the SR0 exam.
NRC Resolution - See Resolution (2) per part a., SR0 Exam.
(5) Question 4.01 - Same comments as for Question 8.06 on the SR0 exam.
NRC Resolution - See Resolution (8) per part a., SR0 Exam.
(6) Question 4.03 - Same comments as for Question 7.02 on the SR0 exam.
NRC Resolution - See Resolution (3) per part a., SR0 Exam.
(7) Question 4.06 - Same comments as for Question 7.06 on the SR0 exam.
NRC Resolution - See Resolution (4) per part a., SR0 Exam.
Enclosure 1 6 (8) Question 4.09 - Same comments as for Question 7.09 on the SR0 exam.
NRC Resolution - See Resolution (5) per part a., SR0 Exam.
(9) Question 4.19 - Same comments as for Question 7.20 on the SR0 exam.
NRC Resolution - See Resolution (7) per part a., SR0 Exam.
- 4. Exit Meeting At the conclusion of the site visit, the examiners met with representatives of the plant staff to discuss the results of the examination. Those individuals who clearly passed the oral examination were identified.
There was no generic weaknesses (greater than 75 percent of candidates giving incorrect answers to one examination topic) noted during the oral examination.
The cooperation given to the examiners and the effort to ensure an atmo-sphere 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.
ATasyf2 0
ENCLOSURE 3 U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: GRAND GULF 1 REACTOR TYPE: BWR-GE6 DATE ADMINISTERED: 85/08/12 EXAMINER: MUNRO,J APPLICANT: _________________________
INSTRUCTIONS TO APPLICANT:
Uce 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 passins grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF APPLICANT'S CATEGORY ~
VALUE TOTAL SCORE VALUE CATEGORY g ___________ ________ _______________________________ ___
25 ' " "
- ___I_00 ___ _Ili 2 ___________ ________
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 35? 25%
_ _I __ _ _I ___________ ________ 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 24
___I_00 ___ _ l!! ___________ ________ 3. INSTRUMENTS AND CONTROLS E __* _ EL ___________ ________ 4. PROCEDURES - NORMAL, ABNORMAL,
_23S EMERGENCY AND RADIOLOGICAL CONTROL 174 IGG. efh 100.00 TOTALS FINAL GRADE _________________%
All work done on this examination is my own. I have neither sivGn nor received aid.
5~PL5C5 iT5 5IGU55UR5~~~~~~~~~~~~~~
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 2
--- isEss557sAsics- REAi isAssFEE As5 FEUi5 FE6s QUESTION 1.01 (1.00)
Which of the following statements correctly describes the behavior of the void coefficient of reactivity? -
- a. It becomes more negative as the void fraction decreases.
- b. It becomes less negative as fuel temperature increases.
- c. It becomes less negative as core size increases.
- d. It becomes more negative as moderator temperature increases.
. QUESTION 1.02 ( 1.0( )
What are the units of neutron flux?
- a. neutrons / cm cubed
- b. neutrons / cm / second c.-neutrons / cm squared - second
- d. neutrons / cm squared -
QUESTION 1.03 (1.00)
The rate of change of power in a nuclear reactor is governed by the averase neutron generation time (1-av). How does 1-av change as the core ages?
- a. 1-av INCREASES due to the DECREASE in the effective delayed neutron fraction (B-bar) over ccre life.
- b. 1-av DECREASES due to the DECREASE in the effective delayed neutron fraction (B-bar) over core life,
- c. 1-av INCREASES due to the INCREASE in the effective delayed neutron fraction (B-bar) over core life.
- d. 1-av DECREASES due to the INCREASE in the effective delayed neutron fraction (B-bar) over core life.
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 3
~~~~iE5EE66YEEE5C5~~E5Ei~iEE 5F5R EE6 Flui5 FL5E QUESTION 1.04 (1.00)
Which of the following radiation exposures would inflict the GREATEST biological damage to man?
- a. 1 Rem of GAMMA
- b. 1 Rem of ALPHA
- c. 1 Rem of NEUTRON
- d. NONE of the above; they are all equivalent GUESTION 1.05 (1.00)
Which of the following correctly describes the Maximum Fraction of Limiting Power Density (NFLPD)?
- a. LHGR-actual / LHGR-limit i must be maintained < 1
- b. LHGR-limit / LHGR-actual ; must be maintained > 1
- c. LHGR-limit / LHGR-actual ; must be maintained < 1 .
- d. LHGR-actual / LHGR-limit ; must be maintained > 1 QUESTION 1.06 (1.00)
When does a constant-speed centrifugal pump motor draw the LEAST current?
- a. at ' runout' conditions
- b. at its ' operating point *
- c. while "cavitating'
- d. at ' shutoff head' conditions
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 4
--- iAEEs557sisiCi? sEAi iEEsiFEE En5 FEDi5 FE6s GUESTION 1.07 (1.00)
Which of the following equations is used to perform a BWR reactor heat balance?
NOTE: c=CRD; f=Feedwater; s= Steam; ri=RWCU in; ro=RWCU out
- a. b-rx = (w,+ g) x h,+ w,,x h,+ 5-rad - w,x h,- w, .x h,- w,x h - h pump
- b. b-rx = (wu+ we) xh+w 3 g x h,+ d-rad - w,x h,- wrox h,,- w,x h,- pump
- c. 5-ex = ( w, + w, ) x n,+ w,x n,,+ 6 pump - w,x n,- ws x h,.- w,x h,- h-rad ,
d.' -rx = wgx h,+ w;n , h,;+ -rad - ( w, + wg) x h,- w,,x h,,- w x h,- b pump OUESTION 1.08 (2.00)
- a. DEFINE ' Critical Power'.
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- b. Which one of the following conditions would tend to INCREASE the Critical Power level assuming all other variables remain unchanged?
- 1. Inlet subcooling is DECREASED
- 2. Reactor pressure is DECREASED
- 3. The axial power peak is RAISED
- 4. Coolant flow rate is DECREASED QUESTION 1.09 (1.00)
Which of the following actions will INCREASE your plant's thermodynamic cycle efficiency?
- a. DECREASING power from 100% to 25% .
- b. LOWERING condenser vacuum from 29' to 25'.
- c. REMOVING a high pressure FW heater from service.
- d. DECREASING the amount of condensate depression.
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
- 1. PRINCIPLES'0F NUCLEAR POWER PLANT OPERATION, PAGE 5
~~~~iU5RE66iE5E5C5I~E5Ei~iEEE5555~EE6~FL0i5 FL5E QUESTION 1.10 (1.00)
The change in reactivity associated with a change in Keff from 0.920 to 1.004 is approximately...
- a. 0.080
- b. 0.084
- c. 0.087
- d. 0.091 a
QUESTION 1.11 (1.50)
Using the enclosed Mollier Diagram, LIST the following property values for steam with an enthalpy of 1390 BTU /lbm and an entropy of 1.568 BTU /lba - F.
- a. Pressure 4
- b. Temperature i c. Superheat QUESTION 1.12 (1.00)
FILL IN THE BLANKS i.
The reactor period for any reactor shortly after a scram will be
____________ seconds because of ___ ______________________________.
QUESTION 1.13 (1.00)
I CALCULATE the GUALITY of a 540 degree F vapor-liquid mixture whose specific enthalpy is 1175 BTU /lbm.
1 (xxxxx CATEGORY 01 CONTINUED ON NEXT PAGE *xxx*)
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 6
--- isiss557sAsiCi- siii fiEssFEE Es5 FEUi5 FE5E GUESTION 1.14 (2.00)
The attached figure (15.3-4) illustrates a transient that could cccur at a BWR.
GIVEN (1) A fast closure of BOTH recire. FCVs at 11% per second.
(2) No operator actions are taken.
(3) Valve closure begins at time = 0 seconds.
EXPLAIN the cause of the following recorder indications:
- a. The peak in inlet subcooling at ~11 seconds on graph (a).
- b. The dip in reactor pressure at ~8 seconds on graph (b).
- c. The peak in vessel steam flow from ~12-15 seconds on graph (c).
- d. The reactor scram at ~8 seconds on graph (d).
QUESTION 1.15 (1.00)
Adding latent heat to liquid water at saturated conditions will...
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- a. increase the temperature of the water.
- b. change the water to steam at the same temperature.
- c. change the water to steam at a slightly higher temperature.
- d. decrease its subcooling by making it boil.
GUESTION 1.16 (1.00)
Water is an excellent neutron moderator. What are TWO (2) NUCLEAR FACTORS which make w&ter the moderator of choice for most commercial reactors?
GUESTION 1.17 (1.00)
ANSWER THE FOLLOWING TRUE OR FALSE
- 3. A control rod's worth varies directly with effective core sice.
- b. The slope of the integral rod worth curve is greatest where the differential rod worth is the highest.
(xxxxx CATEGORY 01 CONTINUED ON NEXT PAGE xxxxx)
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 7
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~~~~ TUERE66E 55C5,~55dT TEdU5FER~ UE~ FLUE 6" FLUE QUESTION 1.18 (2.00)
A significant amount of excess reactivity must be loaded into a core at BOL so that 100% power can be attained at the end of a fuel cycle. For each of the following, LIST the approximate value of K-excess which must be loaded to overcome that negative reactivity component at rated-equilibrium conditions.
- a. Moderator temp increase
- b. Void fraction increase
- c. Samarium buildup
- d. Xenon buildup QUESTION 1.19 (1.50)
LIST three (3) factors upon which a reactor's decay heat generation rate is dependent.
QUESTION 1.20 (1.00)
Which of the following is NOT a characteristic of Suberitical Hultiplication?
- a. The subtritical neutron level is directiv proportional to the neutron source strength.
I b. Doubling the indicated count rate by reactivity additions will reduce the margin to criticality by approximately one-half.
- c. For equal reactivity additions, it takes longer for the new equilibrium count rate to be reached, as K-eff approaches unity.
- d. A single notch of rod withdrawal will produce an equivalent equilibrium count rate increase whether Keff is 0.88 or 0.92.
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(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 8 isiEs557sisiCi? sisi iEisifiE Es5 FEUi5 FE5A GUESTION 1.21 (1.00)
The reactor trips from full power, equilibrium xenon conditions. Twenty-four hours later the reactor i s brought critical and power level is main-tained on range 5 of the IRMs for several hours. Which of the following statements is CORRECT concerning control rod motion during this period?
- a. Rods will have to be withdrawn due to xenon build-in.
- b. Rods will have to be rapidly inserted since the critical reactor will cause a hiSh rate of xenon burnout.
- c. Rods will have to be inserted since xenon will closely follow its normal decay rate.
- d. Rods will approximately remain as is as the xenon establishes its equilibrium value for this power level.
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(***** END OF CATEGORY 01 xxxxx)
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- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 9 a -------------------------------------------------------
i QUESTION 2.01 (1.00)
Which of the following is the only normally CLOSED valve in the RCIC steam supply flow path in the at power Standby lineup?
- a. Steam Supply Valve (F045)
- b. Outboard Steam Isolation Valve (F064)
, c. Turbine Trip Throttle Valve
- d. Turbine Governor Valve GUESTION 2.02 (1.00)
Which of the following sequences of components correctly reflects the normal RCIC' water flow path for injection into the Reactor?
What is(are) the automatic isolation signal (s) to the RCIC Vacuum Breaker Isolation Valves (F077,F078)? Setpoints required.
QUESTION 2.04 (1.00) ,
How would a loss of service air affect the operation of the Standby Liquid-Control System (SBLC)? !
- a. The SBLC tank level indication would be inoperable.
- b. The SBLC tank air sparser would be inoperable. '
- c. The SBLC tank level indication and air sparger would be inoperable.
- d. It would have NO impact since the instrument air system supplies all SBLC needs.
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
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- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMG PAGE 10 QUESTION 2.05 (1.00)
The containment flooding mode of RHR is available as a backup when virtually all other means to keep the core covered have failed. Briefly DESCRIBE the flowpath established during the containment flooding mode.
QUESTION 2.06 (1.50) W<b G. . E%
The containment building sampling subsystem station located on the refueling floor pro';Ades a central location for monitoring and grab sampling what three (3) fluid systems?
QUESTION 2.0/ (1.50)
The Instrument Air System provides air for the operation of various valves within the Standby Service Water System (SSW). LIST three (3) limitations imposed on the SSW system by a loss of Instrument Air.
GUESTION 2.08 (1.00)
The plant is operating at power with A, B, and C CCW pumps running and NONE of the pumps selected for STANDBY operation. A LOSp securs and the diesels start and tie in normally. How will the CCW system respond during this transient?
- a. The LSS panel will auto start the 'B' CCW pump on ESF power 20 seconds after the bus is reenergized.
- b. Either the 'A' or 'B' CCW pumps can be started manually on ESF power after the buses are reenergized.
- d. The 'B' CCW pump can be manually started by the operator on ESF power after the bus is reenergized.
(xxxxx CATEGORY 02 CONTINUED ON NEXT PACE xxxxx)
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 11 QUESTION 2.09 (2.00)
Briefly DESCRIBE the operation of the PSW system under each of the following operating modes;
- a. " Local
- Automatic
- b. ' Cascade' Automatic QUESTION 2.10 (1.50)
Answer the following with regard to the SRVs / ADS:
- a. Which ADS interlock / permissive signal is NOT bypassed (i.e. must be present) to allow manual initiation of ADS from the 601 panel? (0.5)
- b. Panels 601 and 631 have red and green SRV indicating lights. An illuminated red light on P601 vindicates that _________________________
while an illuminated red lish", on P631 indicates that _______________.
(BE SPECIFIC.) a c a. u M GUESTION 2.11 (1.50)
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LIST six (6) methods of detectin3 reactor coolant leakage within the drywell.
QUESTION 2.12 (1.50)
LIST three (3) purposes of the A, B, and / or C RHR jockey pumps.
QUESTION 2.13 (1.50)
Fill in the following blanks with the appropriate (if any) .LPCS injection valve (F005) interlocks and setpoints*
Manual opening of F005 with the handswitch is prohibited when _____(a) ___.
.If power is available, F005 will auto open on a LPCS initiation signal of
______ ( b) __________ or _________ ( c) __________. Once open, ______ ( d) ______
signal will auto close the valve. If the auto open signal is manually overridden the valve will reopen automatically if __________(e) _________.
(xxxxx CATEGORY 02 CONTINUED ON NEXT PAGE **xxx)
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2o PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 12 i QUESTION 2.14 ( .50)
Reactor pressure is 900 psig and LPCS is running in response to a valid initiation signal. What is the approximate expected flow indication on the pump discharge flow meter on the 601 panel?
QUESTION 2.15 (1.00)
Reactor Feed Pump (RFP) turbine speed is controlled by either a Notor Speed Changer (MSC) or an Electric Automatic Positioner (EAP). The EAP ...(CHOOSE ONE)
- a. ... will control the RFP turbine's speed only if its speed signal is greater than that from the MSC.
- b. ...is normally used to control feed flow rate over a turbine speed of 0 - 5500 rpm.
- c. ..., unlike the MSC, d,es NOT afford the capability of manual
], speed control by use a local handwheel.
j d. ...will lock in place to prevent a ramp response to a false signal, if it loses its signal from the flow controller. ,
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QUESTION 2.14 (1.50)
- -The Upper Containment Pool terves different functions depending on the
- ' plant's operating mode or cor dition at the time. STATE the various functions of the Upper Containment Pool.
QUESTION 2.17 (1.00)
What provides the motive force t'or the Suppression Pool Cleanup System
- in each of the following lineups?
. a. Normal flowpath
- b. Cmergency flowpath QUESTION 2.18 (2.00)
- LIST FOUR (4) signals which will resv4t in a DIRECT trip of the RWCU recirculation pumps.
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
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- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 13 GUESTION 2.19 (1.50)
The Containment Recirculation Charcoal Filter Trains normally take air from the , but can also draw a suction from _____ (b) and ______(c) _______(a)
________. ______ Air from the filters is normally returned to ____
(d) _______
, but can also be directed to the ______ (e) _______.
QUESTION 2.20 (2.00)
LIST the four (4) signals which will result in an automatic isolation of the Control Room Ventilation System.
(***** END OF CATEGORY 02 ****x)
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- 3. INSTRUMENTS AND CONTROLS PAGE 14 QUESTION 3.01 (1.00) l Assume that APRM *B' currently has 14 operable LPRM inputs and
, is reading 65% power. Which of the following indication (s)
I and/or action (s) will occur as a result of 1 LPRM (of the 14 remaining LPRM inputs to APRM 'B') failing downscale? Assume NO operator action.
l b. LPRM downscale alarm - APRM *B' reading > 65% l
Rod Block - APRM-'B' reading 65%
i d. LPRM downscale alarm - APRM INOP Trip and Alarm -
l Rod Block - 1/2 Scram - APRM *B' reading 65%
- j. QUESTION 3.02 (1.00)
Which of the following axial location sequences correctly describe the axial locations of LPRMs in the core?
- a. BAF 'A'G+9' - 'B'G+27' -
'C'G+45' -
'D*G+d3' -
- b. BAF - "A'G+18' -
'B'G+54' -
'C'G+90' - 'D'G+126'
-c. BAF - "D*G+9' -
'C'G+27' -
'B'G+45' -
'A'G+63'
- d. BAF - "D'9+18' -
'C'G+54' -
'B'G+90' - 'A'G+126' l
l
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
o a
30 INSTRUMENTS AND CONTROLS PAGE 15 00ESTION 3.03 (2.00)
For each of the following situations (i and ii) select the correct Feed-water Control System / plant response from the list (a through e) which follows. An answer may be_used more than once, and NO operator actions are taken.
- a. Reactor water level decreases and stabili=es at a lower level.
- b. Reactor water level decreases and initiates a reactor scram.
- c. Reactor water level increases and stabilizes at a higher level.
- d. Reactor water level increases and initiates a turbine trip and Reactor Scram.
I e. None of the above.
- i. The plant is operating at 90% power in 3-element control when the HPCS system inadvertently initiates and injects.
ii. The plant is operating at 100% power, in 3-element control, when one
~
. Feed Flow Detector FAILS DOWNSCALE.
QUESTION 3.04 (1.00)
The reactor is critical at approximately 10 psig and the 'RX Heatup and Pressurization' phase of 03-1-01-1,RX SU is being performed.
The narrow range P-680 level instruments read the following ' approx-imate" values: ,
NR LT-N004A 37' NR LT-H004B 38' NR LT-N004C 37' The WIDE RANGE P-680 indicators should read which of the following i approximate values?
- a. O inches.
- b. 15 inches
- c. 38 inches
- d. 60+ inches
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
l l 3.- INSTRUMENTS AND CONTROLS PAGE 16 l
QUESTION 3.05 (1.00) l Which of the following is NOT a symptom that you would enpect to zoe as a result of a " Jet Pump Riser Failure'? Assume Recite Flow Control is in " Flux Manual'.
l a. DECREASE in failed Jet Pump flow.
- b. DECREASE in core differential pressure.
- c. DECREASE in reactor (APRM) power.
- d. INCREASE in indicated core flow.
QUESTION 3.06 (1.00)
'The plant is operating normally a' pow?r when you receive a ' Pump A Ssal Staging Flow High/ Low" alarm and note a DECREASE in No.2 Recire Pump seal pressure. . Which of the fol'.owing failures would cause this indication?
- a. Failure of No. I seal
- b. Failure of No. 2 seal .
I' c. Plugging of the No. 1 internal restricting / breakdown orifice
- d. Plugging of the No. 2 internal restricting / breakdown orifice GUESTION 3.07 (2.00)
Briefly explain what condition (s) will generate EACH of the following indications on the Operator Control Module.
- a. Data Fault
- b. Scram Valves
- c. Channel Disagree
- d. Insert Required (xxxxx CA'.EGORY 03 CONTINUED ON NEXT PAGE *****)
- 3. INSTRUMENTS AND CONTROLS PAGE 17 QUESTION 3.08 (1.00)
- a. Fill in the following blank:
Above the HPSPr continous withdrawal of a control rod is automatically limited to _____ notch (es). (0.5)
- b. What is the reason for this limitation? (0.5)
OUESTION 3.09 (3.00)
Consider a Recire Pump Fast to Slow Speed transfer:
- a. After ' tripping CB-5', certain permissives must be met to
'close CB-2' and complete the speed transfer. Indicate the 8 permissives (in 2 groupings) that are left blank on Figure 9 - Transfer Sequence. (2.0)
- b. Briefly explain the reason for ' tripping the FCV to Manual' in the sequence. (1.0) 1
(*x*** CATEGORY 03 CONTINUED ON NEXT PAGE **xxx)
- 3. INSTRUMENTS AND CONTROLS PAGE 18
. QUESTION 3.10 (1.00)
The plant is operating at 100% power with Recire Flow control in
- Flux Manual'. An operator inadvertently INCREASES the ' Pressure Reference Set' on the EHC Turbine Control System by 5 psis.
ASSUME: 1. No further operator action.
- 2. All other EHC control settings are normal.
- 3. Starting Parameters:
TCVs (MSCV & LPSCVs) - 100% Steam Flow Position BSCVs -
0% Steam Flow Position Rx Power - 100% Rated Thermal Power Rx Pressure - 1025 psig NOTES: All valve %s are in % Steam Flow Position.
See Fi3vre 7 (EHC Logic Diagram) for information.
Wh'ich of the following most accurately describes both the INITIAL RESPONSE and FINAL STATUS of the different parameters and components?
Note: c lv ins-<- M &J <tst.n 8. L ~ @ ;-it. t A 4;a g ,, , , ,
a b c d INITIAL RESPONSE
- TCVs IPartial IPartial IPartial INo Chan5e IClose (<100%) IClose (<100%) IClose (<100%) 1
- -BSCVs INo Change IPartial INo Change IPartial i 10 pen (>0%) i 10 pen (>0%)
-Rx Power IIncrease INo Change IIncrease IDecrease
-Rx Pressure IIncrease iNo Change IIncrease IDecrerse I I FINAL STATUS , f l ,
-TCVs I"100% IPartial 10% I"100%
1 IClose (<100%) l I
-BSCVs 10% IPartial 10 pen (as ~10%
i 10 pen (>0%) Inecessary)
- I
-Rx Power l>100% l>100% !"0% l<100%
-Rx Pressure l>1025 psig l>1025 psig I"920 psis l<1025 psig
- Open as necessary for SD Pressure Control
(***** CATEGORY 03 CONTINUED ON NEXT PAGE xxxxx)
- 3. INSTRUMENTS AND CONTROLS PAGE 19 QUESTION 3.11 (1.00)
Which of the following will NOT result in the COMMANDS DISAGREE lamp being lit for RC & IS ?
- a. Depressing the COMMAND DISAGREE '1': SHIFT INHIBIT PB
- b. Depressing the COMMAND DISAGREE '2': SHIFT INHIBIT PB
- c. Depressing the MASTER TEST PB (CLOCK FREQUENCY Section)
- d. Depressing the COMMAND INHIBIT PB's.
QUESTION 3.12 (1.00)
The General Area Radiation Monitors (ARMS) have installed check sources.
These sources...
- a. are normally shielded and are exposed by depressing the green backlit Check Source pushbutton.
- b. are automatically exposed every 17 minutes to test proper module response.
- c. do not affect the ARM's indicated background radiation level in those areas monitcred.
- d. aid in the detection of equipment malfunctions which cause downscale trips.
QUESTION 3.13 (1.00)
How would an SRM detector respond to a pin hole leak which causes a gradual decrease in Argon gas pressure?
- a. Gamma and neutron sensitivity would DECREASE.
- b. Gamma sensitivity would DECREASE but neutron sensitivity would REMAIN UNCHANGED.
- c. GAMMA sensitivity would REMAIN UNCHANGED but neutron sensitivity would DECREASE.
- d. Both gamma and neutron sensitivity would REMAIN UNCHANGED.
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
L
- 3. INSTRUMENTS AND CONTROLS PAGE 20 QUESTION 3.14 (2.00)
The plant is operating at 100% RTP when APRM 'A' fails upscale and results in a reactor half-scram. Utilizing the attached RPS trip logic diascams (Figures $141 A thru C) DESCRIBE in a STEP-BY-STEP fashion (with regard to the opening / closing, ener-Siring /deenergining of ALL applicable contacts and relays) how the APRM vpscale trip results in an actuation of the scram solenoid.
NOTE: IF THE ATTACHED DIAGRAMS CAN NOT BE EASILY READ, ASSIGN THE CONTACTS / RELAYS, ETC NUMBERS AND REFER TO THEM IN YOUR ANSWER.
. QUESTION 3.15 (1.00)
A LOCA signal is received. The DG's start and their output breakers close at time t=0. Which of the following loads is correctly matched with its Load Shed and Sequence System (LSSS) sequencing time?
- a. LPCS pump at time t=5 see
- b. RHR pump C at time t=0 see
- c. SSW pump A at time t=15 see
- d. HPCS pump at time t=0 sec 4
00ESTION 3.16 (3.00)
Ecc Provide the following information with regard to the ATHS AND theaRPT Recirculation Pump trips:
- a. Initiation signals and applicable setpoints
- b. Actions / components actuated
- c. Bypasses (automatic and manual, if any)
GUESTION 3.17 (1.00)
Under what 2 conditions will the EHC System ' Load Reference Control'
- AUTOMATICALLY switch 0FF?
(***** END OF CATEGORY 03 *****)
e i
I i
4
O 40 PROCEDURES - NORMAL, ABNORMAL, ENERGENCY AND PAGE 21
~~~~ - ------------------------
R5655LU5555L C5sTR5L OUESTION 4.01 (1.50)
With the exception of breaker position, what THREE (3) items should an operator check on a breaker during the performance of a system lineup checksheet per Control and Use of Operations Section Direc-tives, 02-S-01-2? Consider Local checks only.
QUESTION 4.02 (1.00)
With a stuck control rod, ONEP-05-1-02-IV-1, 'CRD Malfunctions',
instructs the operator to INCREASE drive water pressure in an attempt to initiate control rod movement. With the reactor at FULL POWER conditions, SELECT the MAXIMUM differential pressure to which the drive water may be raised.
- a. 90 psid
- b. 260 psid
- c. 350 psid
- d. 500 psid -
QUESTION 4.03 (1.50) Dc kh Rm h A reactor SCRAM has occurred, but NOT all of the control rods have inserted to less than the 06 position. Reactor power is indicated as 6% on the APRH's. LIST the three (3) immediate operator action steps that are required per ONEP-05-1-02-I-1, ' Reactor Scram."
NOTE: LIMIT YOUR RESPONSE TO THOSE ACTION STEPS REQUIRED FOR REACTIVITY CONTROL.
QUESTION 4.04 (1.00) 50I-04-1-01-P75-1, ' Standby Diesel Generator
- cautions the operator NOT to operate the diesel generator without air pressure. EXPLAIN the basis for this caution.
(mmmm* CATEGORY 04 CONTINUED ON NEXT PAGE *****)
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 22
~ ~~~~~~~~~~~~~~~~~~~~~~~~
~~~~EI655L6555EL C5EiR5L QUESTION 4.05 (1.00)
Assume that adequate core cooling CANNOT be maintained and
' Alternate Shutdown Cooling' must be established per EP-8.
DESCRIBE the RPV cooling water flowpath that should be estab-lished per EP-8.
NOTE' INCLUDE IN YOUR DESCRIPTION THE SYSTEMS / COMPONENTS WHICH ARE USED.
QUESTION 4.06 (1.50)
Por ONEP-05-1-02-III-5, ' Automatic Isolations', LIST the three (3) conditions which must be met before a system can be restored to .
-service. Assume an automatic isolation HAS OCCURRED and that the cause of the isolation HAS BEEN determined.
QUESTION 4.07 (1.00)
Per EP-2, ' Emergency Cooldown*, which of the following most ,
accurately describes how SRV operation should be used to control pressure, if needed?
NOTE: ASSUME THAT THE INSTRUMENT AIR SYSTEM IS OPERATING PROPERLY
- a. Use numerous SRV's, with short pressure reductions
(
- 50 psis) to equalire Suppression Pool heatup.
- b. Use fewer SRV blowdowns, with increased pressure reduc- ;
tions to minimize SRV cyclic stresses.
- c. Depressuri=e with a sustained SRV opening to maximi =e
-the emergency cooldown rate.
- d. Allow the SRV's to operate by mechanical actuation to ensure design pressure control and heat dispersion.
QUESTION 4.08 (1.00) j I0I-03-1-01-3, ' Plant Shutdown't cautions the operator to reduce reactor pressure to approximately 400 psig, if possible, when it '
is desired to raaintain a HOT SHUTDOWN condition. EXPLAIN the basis for.this caution.
(smsum CATEGORY 04 CONTINUED ON NEXT PAGE *****)
(
l l
I
-- _ , . , _ . _ _ _ _ _ - . _ . _ _ . _ _ _ ,~~ . - _ . _ _
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 23
~ ~~~~~~~~~~~~~~~~~~~~~~~~
~~~~R 5655Ld5 55L 56Ni5dL
.,.0UESTION ,4.09
, (2.00)
Per IOI-03-1-01-1, ' Cold Shutdown to Generator Carrying Minimum Loadr*: LIST four (4) conditions which must be met / satisfied prior to placing the Mode Switch in RUN.
NOTE: INCLUDE SETPOINTS, IF APPLICABLE QUESTION 4.10 (1.50)
LIST the three (3) types of Radiation Work Permits (RWP's) which may be used to control access / account for personnel o >:p o s u r e .
QUESTION 4.11 '(1.00)
Icmediate Actions in ONEP-05-1-02-I-4, ' Loss of Off Site Power',
direct the operator to ensure that certain DC Oil Pumps auto-s.atically start (or start them manually). Which of the following is 140T one of these pumps? ,
- b. Diesel Generator DC Oil Pump
- c. Main Turbine DC Oil Pump
- d. Main' Generator DC Seal Oil Pump
\
\
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
s 1
5 s
b t
i e
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 24
~
~~~~R5656L65553L 55EiR5L'~~~~~~~~~~~~~~~~~~~~~~~
GUESTION 4.12 (1.00)
The Control Room is declared uninhabitable and evacuated. The immediate operator actions for ' Shutdown From the Remote Shutdown Panel', ONEP-05-1-III-1, are completed. RCIC then ISOLATES.
Level subsequently decreases to Level 2. Restoration of level USING RCIC requires which of the following?
ASUME THAT THE THREE CONDITIONS NEEDED FOR RESETTING AN ISOLATION, PER ONEP-05-1-02-III-5, ' AUTOMATIC ISOLATIONS', HAVE BEEN MET.
- a. No Operator Action. RCIC will restart automatically.
- b. Operator Action. Close RCIC TURB TRIP /THROT VLV; Place RCIC TURB FLO CONT in manual at minimum setting; Re-open RCIC TURB TRIP /THROT VLV and establish flow.
- c. Operator Action. Close RCIC TURB TRIP /THROT VLV; reset RCIC TURB TRIP logici RCIC will now restart automatically.
- d. NONE OF THE ABOVE. RCIC cannot be restarted from the Remote Shutdown Panel after isolation. .
GUESTION 4.13 (1.50)
ONEP-05-1-02-III-3, ' Decrease in Recirevlation System Flow Rate',
directs operator actions for an unexpected decrease in reactor coolant system flow rate.
FILL IN THE BLANKS (After the unexpected decrease), if both recirculation loops are still operating, transfer the FCV's to ____(a) ____. Balance loop flows to within ____(b) ____ at less than 70% core flow, or to within
( c)____ at greater than 70% core flow.
(*xxxx CATEGORY 04 CONTINUED ON NEXT PAGE *****)
8
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 25
~ ------------------------
~~~~EE65d[655 EEL E6OTR5L
'0UESTION / 4.14 (1.00)
A plant startup is in progress and condenser vacuum is being established in accordance with IOI-03-1-01-1, ' Cold Shutdown to Generator Carrying Minimum Load'. What is the proper sequence for. component / subsystem startups?
- a. Steam Seal Exhauster, Steam Seal Header, Mechanical Vacuum Pump, Steam Jet Air Ejector.
- b. Steam Seal Header, Steam Seal Exhauster, Hechanical Vacuum Pump, Steam-Jet Air Ejector,
- c. Mechanical Vacuum-Pump, Steam Seal Exhauster, Steam Seal Header, Steam Jet Air Ejector.
- d. Steam Seal Exhauster, Mechanical Vacuum Pump, Steam Seal Header, Steam Jet Air Ejector.
QUESTION 4.15. (1.00)
Per ONEP-05-1-02-V-1, ' Loss of Component Cooling Water', a loss -
of CCW may be either. complete or partial. -In.which of the
'following instances would reduced flow (partial loss) be treated as'a COMPLETE LOSS of CCW?
~
- a. Reactor Recire Pump temperatures above the HI alarm setpoint.
- b. RWCU NRHX Outlet temperature above the HI alarm setpoint.
- d. -CRD Pump Oil temperature above the HI alarm setpoint.
i
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
l e
l
6 l 4o PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 26
~~~~ ~
R5656Ld555EL 55UTRUL'~~~~~~~~~~~~~~~~~~~~~~~
GUESTION 4.16 (1.00)
The unit is operating at 70% RTP; you notice Power start to increase with NO CHANGE in recirculation flow or rod position. You suspect a ' Loss of Feedwater Heating.'
Which of the following is required / appropriate per ONEP-05-1-02-V-5?
- a. A 30% reduction in Recire Flow, monitored by Recire Flow indication.
- b. A 30%~ Power Reduction, using Recire Flow, monitored by APRH's.
- c. Insertion of Shallow Rods, to maintain proper flux shaper prior to reducing Recite Flow.
- d. Insertion of Power Rods, to maintain proper flux shape, prior to reducing Recire Flow.
QUESTION 4.17 (2.00)
EP-3, EP-5, and EP-7 caution the operator to observe certain .
limitations on Suppression Pool Level and Temperature when operating HPCS, LPCS, RHR, and/or RCIC.
- e. COMPLETE THE FOLLOWING* (1.5)
Suppression Pool Level shall not be less than ____(1) ____.
Suppression Pool Temperature shall not exceed ____(2) ____
during HPCS, LPCS, and/or RHR operation; it shall not exceed ____(3) ____ during RCIC operation.
-b. STATE the basis for these temperature / level limitations on the Suppression Pool. (0.5)
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
l s.
l
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 27
- ~~~~~~~~~~~~~~~~~~~~~~~~
~~~~R5656L65iEAL E5NTR6L GUESTION 4.18 (1.00)
You enter an area posted with the following sign: CAUTION OeA a . . . _
namanan ma
~
[KEEPOUT nwe ngso son LIST the MINIMUM and MAXIMUM exposure rates for this area.
QUESTION 4.19 (1.00)-
When raising power per 10I-03-1-01-2, ' Power Operations,' you are cautioned to maintain the Load Demand Limited (LDL) value close.to the Actual Generator Load (AGL) value.
~
- a. STATE how much the LDL value may exceed the AGL value. (0.5)
- b. STATE how you would know if this limit were exceeded (EXCLUDING THE DIGITAL METERS ON 1H13-P680-90). (0.5)
GUESTION 4.20 ( .50)
You are conducting a shutdown of the CRDH system, per SOI-04-1-01-C11-1. You open Drain Valve 107xx to drain the water accumulators. State the local indication (s) which
.should be used to determine that the accumulator is fully drained.
(***** CATEGORY 04 CONTINUED ON NEXT PAGE **xxx)
4.. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 28
~~~~~~~~~~~~~~~~~~~~~~~~
~~~~EI656L655E3L"56N5 EEL QUESTION 4.21 (1.00)
With regard to the Protective Ta93ing System Procedurer 01-S-06-01:
l The minimum level of qualification for an Independent Verifier (per this procedure) shall be ... (CHOOSE ONE)
- a. ... journeyman level
- b. ... NOB for. operations
- c. ...NOA for operations
- d. ... Shift Supervisor l
l' l
I i
(***** END OF CATEGORY 04 *****)
(************* END OF EXAMINATION ***************)
I
f o ra v o $/t
, . mg sa 1,t + 1/2 at' l 2
E = mc- -
l' 2 -x'
<E = 1/2 mv a = (Vf - /3 )/t A = Aa A=Ae3 l PE = m9n vf=v g+ at * = e/t A =
In2/t1/2 = 0.693/ti/2 2 "
y , , .p A* nD 1/2" b
4
[lt173) + (t))]
aE = 931 sm -
m = V,yAo -Ix Q.= m,ah I=leo Q = mCpat 6 = UA4 T I = I n
e'"*
Pwr = Wfah I = I,10-*/U' TVL = 1.3/u P = P l0 sur(t) o HVL = -0.693/u P=Pe* o SUR = 26.06/T SCR = 5/(1 - K,ff)
CR x = S/(1 - K,ffx)
SUR = 26o/t* + (a - o)T CR)(1 - K,ff)) = CR 2 (I - keff2)
T = (t*/o) + [(s - oy Io] M = 1/(1 - K,ff) = CR /CR j 3 T = t/(o - s) M = (1 - K,ffa)/(1 - K,ff))
T = (a - o)/(Io) SDM = ( - K,ff)/K,ff o = (Keff-1)/K ,ff = AKeff /K eff 1" = 10 secono I=0.1secondsj o = [(t*/(T K,ff)] + [i,ff (1 / + IT)]
I j d) =Id P = (IoV)/(3 x 1010) I)d) 2 ,2gd 2 22 I = aN 2 R/hr = (0.5 CE)/d (meters)
R/hr = 6 CE/d2 (f,,g)
Water Parameters Miscellaneous Conversions 1 gal. = 8.345 lem. 1 curie = 3.7 x 10100ps 1 ga; . = 3.78 liters 1 kg = 2.21 lbm 1 ft* = 7.48 gal 1 np = 2.54 x 103 Stu/nr Density = 62.4 1 /ft3 1.w = 3.41 x 100 5tu/hr Density = 1 gm/c. lin = 2.54 cm Heat of vaporization = 970 Stu/lom 'F = 9/5'C + 32 Heat of fusion = 144 Stu/lbm 'C = 5/9 (*F-32) 1 Atm = 14.7 psi = 29.9 in. Hg. 1 BTU = 778 ft-lbf I ft. H O 2
= 0.4335 lbf/in.
e = 2.718
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' 1 VOID REACTIVITY 150. - 1. 2 DOPPLER REACTIVITY l 1 LEVEL INCH REF-SEP-SKIRT !! 3 SCRAM REACTIVITY i 2 VESSEL STEAMFLOW $ 4 TOTAL REACTIVITY 2 3 TURBINE STE AMFLOW z if .
- 4 FEEDWATER FLOW $ 1234 2 _-
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MISSISSIPPI POWER & LIGHT COMPANY
- - GRAND GULF NUCLEAR STATION FAST CLOSURE OF BOTH MAIN UNIT! 2 RECIRCULATION VALVES AT 11% F ECOND FINAL SAFETY Aluc.YSIS REPORT FIGURE 15.3 - 4
Volume, ft'/lb Enthalpy. Stu Ab Entropy. BtrEb a F T p T P ss. Water Evap Steam Water Evap Steam Water Evap Steam
't V
Ie 's by h,g h, sg s,, s, 0.01602 3305 3305 -0.02 1075.5 1075.5 0.0000 2.1873 2.1873 32 32 0.08859 0.09991 0.01602 2948 2948 3.00 1073.8 1076.8 0.0061 2.1706 2.1767 35 15 0.12163 0 01602 2446 2446 8 03 1071.0 1079 0 0 0162 2.1432 2.1594 40 40 0.14744 0 01602 2037.7 2037.8 13.04 1068.1 1081.2 0 0262 2.1164 2.1426 45 45 0.17795 0.01602 1704.8 1704 8 18.05 1065.3 1083.4 0.0361 2.0901 2.1262 50 50 0.2561 0.01603 1207.6 1207.6 28.06 1059.7 1087.7 0.0535 2.0391 2.0946 60 60 0.3629 0.01605 868.3 868.4 38.05 1054.0 1092.1 0.0745 1.9900 2.0645 70 70 0.5068 0.01607 633.3 633.3 48.04 1048.4 1096 4 0.0932 1.9476 2.0359 80 80 0.6981 0.01610 468.1 468.1 58.02 1042.7 1100.8 01115 1.8970 2.0086 $0 90 0.9492 0.01613 350.4 350.4 68.00 1037.1 1105.1 0.1295 1.8530 1.9825 100 100 1.2750 0.01617 265.4 265.4 77.98 1031.4 1109.3 0.1472 1.8105 1.9577 110 110 1.6927 0.01620 203.25 203.26 87.97 1025.6 1113.6 0.1646 1.7693 1.9339 120 120 2.2230 0.01625 157.32 157.33 97.96 1019.8 1117.8 0.1817 1.7295 1.911? 130 130 2.8892 0.01629 122.98 123.00 107.95 1014.0 1122.0 0.1985 1.6910 1.8895 140 140 3.718 0.01634 97.05 97.07 117.95 1008.2 1126.1 0.2150 1.6536 1.8686 150 150 0.01640 77.27 77.29 127.96 1002.2 1130.2 0.2313 1.6174 1.8487 160 160 4.741 5.993 0.01645 62.04 62.06 137.97 996.2 1134.2 0.2473 1.5822 1.8295 170 170 7.511 0.01651 50.21 50.22 148.00 930.2 1138.2 0.2631 1.5460 1.8111 180 ISO 9.340 0.01657 40.94 40.96 158.04 984.1 1142.1 0.2787 1.5143 1.7934 100 190 11.526 0.01664 33.62 33.64 168.09 977.9 1146.0 0.2940 1.4824 1.7764 200 200 210 14.123 0.01671 27.80 27.82 178.15 971.6 1149.7 0.3091 1.4509 1.7600 210 14.696 0.01672 26.78 26.80 180.17 970.3 1150.5 0.3121 1.4447 1.7568 212 212 0.01678 23.13 23.15 188.23 965.2 1153.4 0.3241 1.4201 1.7442 220 220 17.186 0.01685 19.364 19.381 198.33 958.7 1157.1 0.3388 1.3902 1.7290 230 230 20.779 0.01693 16.304 16.321 208.45 952.1 1160.6 0.3533 1.3609 1.7442 240 240 24.968 0.01701 13.802 13.819 218.59 945.4 1164.0 0.3677 1.3323 1.7000 250 250 29.825 0.01709 11.745 11.762 228.76 938.6 1167.4 0.3819 1.3043 1.6862 260 260 35.427 41.856 0 01718 10.042 10.060 238.95 931.7 1170.6 0.3960 1.2769 1.6729 270 270 49.200 0.01726 8.627 8.644 249.17 924.6 1173.8 0.4098 1.2501 1.6599 280 280 57.550 0.01736 7.443 7.460 259.4 917.4 1176.8 0.4236 1.2238 1.6473 290 290 67.005 0.01745 6.448 6.466 269.7 910.0 1179.7 0.4372 1.1979 1.6351 300 300 0.01755 5.609 5.626 280.0 902.5 1182.5 ' O.4506 1.1726 1.6232 310 310 77.67 0.01766 4.896 4.914 290.4 894.8 1185.2 0.4640 1.1477 1.6116 320 320 89.64 0.01787 3.770 3.788 311.3 878.8 1190.1 0.4902 1.0990 1.5892 340 340 117.99 0.01811 2.939 2.957 332.3 862.1 1194.4 0.5161 1.0517 1.5678 360 360 153.01 0.01836 2.317 2.335 353.6 844.5 1198.0 0.5416 1.0057 1.5473 380 380 195.73 0.01864 1.8444 1.8630 375.1 825.9 1201.0 0.5667 0.9607 1.5274 400 400 247.26 0.01894 1.4808 1.4997 396.9 806.2 1203.1 0.5915 0.9165 1.5080 420 420 30S.78 440 0.01926 1.1976 1.2169 419.0 785.4 1204.4 0.6161 0.8729 1.4890 440 381.54 460 466.9 0.0196 0.9746 0.9942 441.5 763.2 1204.8 0.6405 0.8299 1.4704 460 480 0.0200 0.7972 0.8172 464.5 739.6 1204.1 0.6648 0.7871 1.4518 460 566.2 0.6545 0.6749 487.9 714.3 1202.2 0.6890 0.7443 1.4333 500 500 680.9 0.0204 0.5386 0.5596 512.0 687.0 1199.0 0.7133 0.7013 1.4146 520 520 812.5 0.0209 0.4437 0 4651 536 8 657.5 1194.3 0.7378 0.6577 1.3954 540 540 962.8 0.0215 0.3651 0.3871 562.4 625.3 1187.7 0.7625 0.6132 1.3757 560 SEO 1133.4 0.0221 0.3222 589.1 589.9 1179.0 0.7876 0.5673 1.3550 580 550 1326.2 0.0228 0.2994 0.2675 617.1 550.6 1167.7 0.8134 0.5196 1.3330 Goo 600 1543.2 0.0236 0.2433 0.1962 0.2208 646.9 506.3 1153.2 0.8403 0.46S9 1.3092 620 620 1786.9 0.0247 0.1802 679.1 454.6 1133.7 0.8656 0.4134 1.2821 640 640 2059 9 0.0260 0.1543 0.1443 714.9 392.1 1107.0 0.8995 0.3502 1.2498 660 660 2365.7 00277 0.1166 0 1112 758 5 310.1 1068.5 0.9365 0.2720 1.2086 680 660 2708.6 0 0304 0.0808 0.0386 0 0752 822.4' 172.7 995.2 0.9901 0.1490 1.1390 700 700 3094.3 0 0366 0 0.050S 906.0 0 906.0 1.0612 0 1.0612 705.5 705.5 3203 2 0.0508 TABLE A.2 PROPERTIES OF SATURATED STEAM AND SATURATED WATER (TEMPERATURE)
A.3
Volume,it /lb 8 EF.thalpy. Stu/lb Enterpy. Stu/2 a F Energy. Stu/lb
, Pro s. T P w ater Evap Steam C;tir Ev p Steam Weter Ev p Steam Cater Steam
- r 'k 's hg hg h, s, s, o
sg wy u, 32.018 0.01602 3302.4 3302 4 0 00 107b.5 1075 5 0 2.1872 2.1872 0 1021.3 0.0886 0.0446 i 0.10 35.023 0.01602 2945.5 2945 5 3 03 1073 8 1076 8 0 0061 2 1705 2.1766 323 1022.3 0.10 0.15 4!453 0 01602 2004.7 2004 7 13.50 1067.9 1081 4 0 0271 2.1140 2.1411 13.50 1025.7 0.15 0.20 53.160 0 01603 1526.3 1526 3 21.22 1063 5 1084 7 0 0422 2 0738 2.1160 21.22 1028 3 0.20 0.30 64 484 0 01604 1039 7 1039.7 32.54 10b7.1 1089 7 0 0641 2.016S 2.0809 32.54 1032 0 0.30 0.40 72.869 0.01606 792.0 792.1 40.92 1052.4 1093.3 0.0799 1.9762 2.0562 40.92 1034.7 0.40 0.5 79.586 0 01607 641.5 641.5 47.62 1048 6 1096 3 0 0925 1.9446 2.0370 47.62 1036.9 0.5 0.6 55218 0 01609 540 0 540.1 53 25 1045.5 1093.7 0.1028 1.9186 2.0215 53.24 1038.7 0.6 0.7 90 09 0.01610 466.93 466 94 5810 1042 7 1100 8 0.3 1 8966 2.0083 58.10 1040.3 0.7 0.8 94.38 0.01611 411.67 411.69 62.39 1040 3 1102.6 0.1117 1.8775 1.9970 62.39 1041.7 0.8 0.9 98.24 0.01612 368 41 368 43 66.24 1038.1 1104.3 01264 1.8606 1.9870 66.24 1042.9 0.9 j 1.0 101.74 0.01614 333 59 333 60 69.73 1036.1 1105 8 0.1326 1.8455 1.9781 69.73 1044.1 1.0 2.0 126.07 0.01623 173.74 173.76 94.03 1022.1 1116 2 0.1750 1.7450 1.9200 9443 1051.8 2.0 3.0 14147 0 01630 118 71 118.73 109.42 1013.2 1122 6 0.2009 14854 1.8864 109 41 1056.7 3.0
, 4.0 152.96 0.01636 90 63 90 64 120.92 1006 4 1127.3 0.2199 1.6428 1.8626 120.90 1060.2 4.0 5.0 162 24 0.01641 73.515 73.53 130 20 1000.9 1131.1 0.2349 1.6094 1.8443 130.18 1063.1 5.0 6.0 170.05 0.01645 61.967 61.98 138 03 996.2 1134.2 0 2474 1.5820 1.8294 138.01 1065.4 6.0 7.0 176 84 0.01649 53 634 53.65 144.83 992.1 1136 9 0.2581 1.5587 1.8168 144.81 1067.4 7.0 8.0 182.86 0 01653 47.328 47.35 150.87 988.5 1139.3 02676 1.5384 1.8060 15034 1069.2 8.0 9.0 169 27 0 01656 42.385 42 40 156.30 985.1 1141.4 0.2760 1.5204 1.7964 15628 1070.8 9.G 10 193.21 0.01659 38 404 38 42 161.26 982.1 1143.3 0 2836 1.5043 1.7879 161.23 1072.3 to 14.696 212.00 0.01672 26.782 26 80 180.17 970.3 1150.5 0.3121 1.4447 1.7568 180.12 1077.6 14.696 15 213.03 0.01673 26.274 26.29 181.21 969.7 1150.9 0.3137 1.4415 1.7552 181.16 1077.9 15 20 227.96 0.01683 20.070 20 087 196 27 960.1 1156.3 0.3358 1.3962 1.7320 19621 1087.0 to 30 250.34 0 01701 13.7266 13 744 218.9 945.2 1164.1 0 36B2 1.3313 1.6995 2185 1087.9 30 40 267.25 0 01715 10 4794 10 497 236.1 933 6 1169.8 0.3921 1.2844 1.6765 236 0 1092.1 40 50 281.02 0.01727 8.4957 8.514 250.2 923.9 1174.1 0.4112 1.2474 J.6585 250.1 1096.3 50 60 292.71 0.01738 7.1562 7.174 262.2 915.4 1177.6 0.4273 1.2167 1.6440 262.0 1098.0 60 70 302.93 0.01748 6.1875 6 205 272.7 907.8 1180.6 0 4411 1.1905 1.6316 272.5 1100.2 70 80 312 04 0 01757 5 4536 5471 232.1 900.9 1183.1 0.4534 1.1675 1.6208 281.9 1102.1 80 90 320.20 0 01766 4.8777 4.895 290 7 894.6 1185.3 0 4643 1.1470 1.6113 290.4 1103.7 90 100 327.82 0.01774 4.4133 4.431 238.5 888.6 1187.2 0.4743 1.1284 1.6027 298.2 1105.2 100 120 341.27 0 01789 3.7097 3.728 312.6 877.8 1193 4 0 4919 1.0960 1.5879 312.2 1107.6 120 140 353 04 001803 3 2010 3 219 3250 858 0 1193 0 0.5071 1.0681 1.5752 324 5 1109.6 140 160 363 55 0 0;815 2.8155 2 834 336.1 859 0 1195.1 0.5206 1.0435 1.5641 335.5 1111.2 160 180 373 08 0.01827 2.5129 2.531 346.2 850 7 1196.9 05328 1.0215 1.5543 345.6 1112.5 180 200 3L180 0 01839 2.2689 2.287 355.5 842.8 1198.3 0 5438 1.0016 1.5454 3548 1113.7 200 250 400 97 0 01865 1.8245 1.8432 3761 825 0 1201.1 0.5679 0 9585 1.5264 375.3 1115.6 250 300 417 35 0 01859 1.5233 1.5427 394 0 808 9 1202.9 0.5ES2 0 9223 1.5105 392.9 1117.2 300 350 431.73 001913 1.3064 1.3255 409 8 794 2 1204 0 0 6CM 08909 1.4968 4096 11IB ! 350 400 44460 0 0193 1.14162 1.1610 424 2 7804 1204 6 0 6217 0 8630 1.4647 422.7 111E 7 400 450 456 28 0 0195 1.01224 1.0318 437.3 767.5 1204.8 06360 0 8378 1.4738 435.7 1118.9 450 500 467 01 0O!93 0 90787 09276 449 5 755.1 1204.7 0.6490 0 8148 1.4639 447.7 11188 500 550 476 94 0 0199 0 82183 0.8412 460.9 743.3 1204 3 0 6611 07936 1.4547 458.9 1118 6 550 074962 07693 471.7 732.0 1203 7 0.6723 0 7738 1.4461 469.5 IllE 2 600 600 48620 0 0201 700 .503 08 0 0205 0.63505 0 6556 491.6 710.2 1201.8 06928 0.7377 1.4304 4S8.9 1116 9 700 800 51821 0 0209 0.54809 0.5690 509.8 689 6 1199 4 0 7I11 0.7051 1.4163 506 7 1115.2 800 900 L's! 93 0 0212 0 4796S 05009 526 7 659 7 1196 4 07279 06753 1.4032 5232 1113 0 900 1000 544.5B O0216 042436 0 4460 542 6 f 50 4 1192 9 07434 06476 1.3910 53'16 1110 4 1000 1100 555 2/ 00220 0 37863 04006 557.5 631 5 1189 1 0757S O6216 1.3794 553 1 1107.5 1100 1200 67.19 0 0223 0 34013 0 3625 571.9 6130 1164 8 0.7714 05969 1.3633 5569 1104 3 1200 544 6 1180 2 0.7843 05733 1.3577 530.1 1100 9 1300 1300 l:577 42 0 0227 0 30722 0.3299 585 6 14CD 537 07 0 0231 0 278/1 0 3018 599 8 576 5 1175 3 0 7966 05507 1.3474 592.9 1037.1 1400 1500 59620 0 0235 02b372 0 27/2 611.7 550 4 1170 1 0 8035 0!253 1.3373 605 2 1093.1 1500 2000 635 60 0 02L7 016?66 01883 6721 4662 113B 3 0BCS 04256 1.?b81 662 6 10G36 2000 2500 66d 11 0 02cf 010209 01307 731 7 3616 1093 3 09139 03206 1.2345 118 5 1032 9 2500 3000 695 33 0 0343 0 050/3 0 0850 801 8 218 4 1070 3 0 9723 0lE91 1.1619 782 8 973.1 3000 3298 2 70147 00%8 0 0 050d 906 0 0 906 0 10612 0 1.0612 875 9 875 9 37082 TABLE A.3 PROPERTIES OF SATURATED STEAM AND SATURATED WATER (PRESSURE)
A.4
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Lesson:~ Recieculation System - 533-1 Page 36 of 36
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 29
--- isEER657sisics- REsi iEAssFEE As5 FEUi5 FE6s ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J ANSWER 1.01 (1.00) e REFERENCE BFNP REACTIVITY COEFFICIENTS LP,P.3 GGNS OP-NP-513,P.9-10 ANSWER 1.02 (1.00)
REFERENCE BFNP NUCLEAR REACTIONS LP,P. 7 GCNS OP-NP-505,P.4 ANSWER 1.03 (1.00) b .
REFERENCE BFNP NEUTRON SLOWING DOWN AND DIFFUSION LP,P.6-8 GGNS OP-NP-510,P.6,98511,F.4 ANSWER 1.04 (1.00) d REFERENCE BFNP MCD BWR LP,P.4 GGNS OP-RP-502,P.5-7 ANSWER 1.05 (1.00) a REFERENCE BFNP LHGR AND BASES LP,P.8-9 GGNS MCD, THERMAL LIMITS, P.74
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 30
~
~~~~T U5R 667 d EC5,~ 55T fR5 EE5E 5 6~ELU56'EL5s ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J ANSWER 1.06 (1.00) d REFERENCE BFNP PUMPS LP,P.5-6 GCNS OP-HF-514 ANSWER 1.07 (1.00) a, REFERENCE GGNS OP- AD-545, P. 4
-ANSWER 1.08 (2.00)
- o. The assembly power which would cause the onset of transition boiling at some point in the assembly. ,1.0)
(
- b. 2 REFERENCE BFNP TRANSITION BOILING & ATLAS TESTING LP,P.5-6 GEXL CORRELATION & CRITICAL POWER LP,P.3 GGNS.MCD, THERMAL. LIMITS, P.26,32-33 ANSWER 1.09 (1.00) d REFERENCE OFNP RANKINE CYCLE LP,P.5,7-8 GGHS OP-HF-505 ANSWER 1.10 (1.00)
-d 1
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l f 1. ' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 31
~
~~~~TUkR 66E dkf65~~EE Y~TR 5FER d 6~FL6 6~FL6U ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J REFERENCE GCNS OP-NP-511,P.3 ANSWER. 1.11 (1.50)
- o. 1000 psia
- b. 800 F l
- c. 255 F l REFERENCE
! GGNS OP-HF-503,P.22-24 ANSWER 1.12' (1.00)
-80 seconds
! the longest-lived delayed neutron precursor (Br-87) (0.5ea/1.0) l REFERENCE GGNS OP-NP-518,P.6 ANSWER 1.13 (1.00)
(1175 - 536.5) / 657 5 = 0.971 REFERENCE
=GGNS OP-HF-503,P.5 ANSWER 1.14 (2.00)
- s. Due to the pressure spike.
l l
- b. Due to the decrease in reactor power, j c. Due to the cycling of SRVs.
I d. Due to vessel high level (L8).
l i
i r
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,,-,vw,- ,-r -,--,---,,---.-----,.---.w,,---.--,---._-----.~,-en-,,,---- , . ,. ----n .. --- . - - .
- 1. PRINCIPLES OF NUCLEAR POWER PLANT-OPERATION, PAGE 32
~~~~ ~
TU5R566ihdkf65,~UE N T TEd 5k5E~d'_d'ELUE6~FLUU ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J REFERENCE GGNS FSAR FIG. 15.3-4 ANSWER 1.15 (1.00) b REFERENCE GCNS OP-HF-502,P.7 ANSWER 1.16 (1.00)
- 1. It (hydrogen) has a high microscopic scattering cross section .
- 2. It (hydrogen) has a high logarithmic energy decrement per collision.
REFERENCE GGNS OP-NP-502,P.9
~
ANSWER 1.17 (1.00)
- o. True
- b. True REFERENCE l
GGNS OP-NP-512,P.13-14 ANSWER 1 18 (2.00)
- a. 4.77%
- b. 3.6% ,
- c. 1.0% )
- d. 3.0%
REFERENCE GGNS OP-NP-518,P.2,9 (CAF) ***************
I 1
l l
l
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 33
~~~~isEE566Y d55C5~~EEdi~iEE05E5E~IE6~ELUiU~ELUE ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J ANSWER 1.19 (1.50)
- 1. power level
- 2. time at' power
- 3. time since shutdown REFERENCE GGNS OP-NP-518,P.7 ANSWER 1.20 (1.00) d REFERENCE EIH, L-RG-605 (15)
GGNSr OP-NP-515,P.4-7 ANSWER 1 21 (1.00) c
' REFERENCE BFNP XENON & SAMARIUM LP, P.4,12 GGNS LP OP-NP-514, p. 5-10
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 34 ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J ANSWER 2.01 (1.00)
G REFERENCE GGNS LP OP-E51-501 ANSWER 2.02 (1.00) b REFERENCE GGNS LP OP-E51-501 ANSWER 2.03 (1.00)
- Hi DW Pressure [.33, 1.39 Psig C.13
- AND- E.23
- RCIC Stm Pressure Low C.33, 60 psis C.13 (1.0)
REFERENCE -
GGNS LP OP-E51-501, p.9 ANSWER 2.04 (1.00) b REFERENCE BFNP LPt39,P.18 GGNS OP-C41-501,P.5,20 ANSWER 2.05 (1.00)
Flow is established from SSW loop B [0.53 to RHR loop B LPCI injection line [0.53.
REFERENCE GGNS OP-P41-501,P.14 SD-E12,P.40,
. r
- 2. PLANT DESICN INCLUDING SAFETY AND EMERGENCY SfSTEMS PAGE 35 ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J ANSWER 2.06 (1.50) DtIdc Q- <5'b d Ase r 0%
- 1. Reactor water
- 2. RWCU system water
- 1. Loss of ability to make up water to the cooling tower basins.
- 2. Loss of ability to make up water to the SSW fill tank.
- 3. Loss of ability to inject sulfuric acid into the basin.
- 4. Loss of ability to inject hypochlorite into the basin. (3GO.5ea/1.5)
- 8. kaap4 8d et br A* S= a, % %t ,>,c .. Ag e , ,,4 % , , ,, , tw L ,. ,y 2,,t g ., , ,p REFERENCE 8b 7'* " 4 -- tw 8*<* ' *- %
'GGNS OP-P41-501,P.22 SD-P41,P.36-37 Prw . os.t.o2 v-4 sn, .
ANSWER 2.08 (1.00) d REFERENCE GGNS SD-P42,P.3,19 ANSWER 2.09 (2.00)
- a. Each operating pump FCV is automatically controlled to maintain its respective pump's desired flow.
- b. All operating Pump FCVs are automatically controlled to maintain desired
-pressure on the common (36') header.
REFERENCE GGNS OP-P44/47-501,P.3,7
- 2. PLANT DESIGN INCLUDING SAFETY AND EHERGENCY SYSTEMS PAGE 36 l _______________________________________________________
ANSWERS -- GRAND GULF 1 -85/08/12-HUNRO,J ANSWER 2.10 (1.50)
- a. The ECCS pump discharge pressure interlocks (0 5)
- b. 601 - the tail pipe pressure switch has picked up 631 - the (B) solenoid is enersized (0.5ea/1.0)
REFERENCE GGNS OP-E22-2-501,P.7,7-10 ANSWER 2.11 (1.50)
- 1. Drywell cooler temperature
- 2. Drywell cooler condensate flow
- 4. Recite pump seal leak det.
- 5. Vessel head seal leak det.
- 6. Valve packins leak det.
- 7. Drywell pressure
- 8. Drywell air monitorins (6G0.25/1.5)
REFERENCE GGNS OP-E31-501,P.10-14 l
i ANSWER 2.12 (1.50)
- 1. Prevent RHR water hammer
- 4. The C pumF Provides suppression pool level transmitter reference les fill (300.5ea/1.5)
REFERENCE GGNS OP-E12-501,P.10 OP-E32/38-501,P.13-14 l
l I
l l
t
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 37 ANSWERS -- GRAND GULF 1 -85/08/12-MUNR0rd ANSWER 2.13 (1.50)
- c. > 50 psis reactor pressure after a 15 minute time delay
- b. Rx level 1 / -150.3'
- c. High drywell pressure / 1.39 psis
- d. No
- e. the initiation signal is reset (0.3ea/1.5)
REFERENCE GGNS OP-E21-501rP.7 ANSWER 2.14 ( .50)
=ero spa ,
REFERENCE GGNS OP-E21-501rP.12,17 ANSWER 2.15 (1.00)
~
d, RFFERENCE USNRC BWR-4 Systems Manual, pp 3.3 3.3-10 EIH! HNP-x-10018 HNP-x-1286 GGNS LP OP-N21-501 p.10 GGNS SIM. MAL. 121 ANSWER 2.16 (1 50)
- 1. Shielding when the reactor is in operation
- 2. Storage space for the steam dryer and moisture separator assemblies and for fuel transfer during refueling
- 3. Post-LOCA suppression pool makeup water source REFERENCE GGNS SD-M41-irP.12-13
. -. . - _ _ _ . . - = _ -. . - . -- . .. . _.
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 38 ANSWERS -- GRAND GULF 1 -85/08/12-MUNR0rd ANSWER 2.17 (1 00)
- c. Refueling water transfer pumps
BFNP: 1. Inlet inboard isolation valve (69-1) not fully open.
- 2. Inlet outboard isolation valve (69-2) not fully open.
- 3. Reactor return isolation valve (69-12) fully closed.
l 4. Pump flow <= 90 spa for 7 seconds.
- 5. Pump coolin3 water (RBCCW) outlet temp. high (140F) (490.5ea)
GGNS 1. Pump cooling water (CCW) temp high (195F) 2.. Pump suction flow low (<70 spm after 15 see pump run) l 3. F001 closed CO.253 and any of F250, 251, 252, 253 not full open [0.253 ,
- 4. Isolation valve F004 not fully open
- 5. Motor Protection Device Activated (490.5ea)
REFERENCE BFNP LP413,P.10 GGNS SD-G33/G36,P.5 j ANSWER 2.19 (1.50) l a. Containment cooling coolers discharge plenum l b. Drywell
- c. Rx water sampling station vent hood
- d. Containment cooling coolers suction plenum ,
- o. Auxiliary building penthouse exhaust vent l
l REFERENCE i GGNS GD-M41eP.7-8 l
r I
l I
r
2.- PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 39 :
ANSWERS -- GRAND CULF 1 -85/08/12-MUNRO,J ANSWER 2.20 (2.00)
- 1. High-high activity in the air intake duct
- 2. High chlorine gas concentration in the air intake duct
- 3. Low reactor water level (-42*)
4.
O. Highdr[wellpressure A<.,e 4 e e%r sg .ts > (2 psis)vs rrva\c 3.s. r, 4- 1. W REFERENCE-GGNS SD-Z51,P.13 l
TS m eat. 3. 3.7, b n ors l
l l
s 1
e I
l l
I I
i i
I
- - _ _ _ _ _ . . . _ . , _ . , . - . , _ . . _ . . _ _ _ _ _ . _ . ~ _ , . . _ , _ . .___ _ _, . _ _ _ . - . . - _ . .
I e.
i i 3. INSTRUMENTS AND CONTROLS PAGE 40 ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J ANSWER 3.01 (1 00) t I
a REFERENCE GCNS LP OP-C51-4-501 ANSWER 3.02 (1.00) b ;
REFERENCE GGNS LP OP-C51-3-501
! ANSWER 3 03 (2.00)
- 1. e l
- 11. d ,
! REFERENCE BFNPt LP912,P.241 TRANSIENT #20iOI-57,P.53 .
EIHf L-RO-726 ,
! GGNS LP OP-C34-501 l GGNS SIN. MAL. 125 & 69 ANSWER 3.04 (1.00) d REFERENCE BFNP! L/P 93 -
EIHf GPNT, Vol. VI, Chapter 2.3-3, 5, Fig 2.3(3)
L-RG-712, pp 4, 5, 19 GCNS LP OP-B21-501 GGNS LP OP-C34-501
- CAFs .
l
,9-
- 3. INSTRUMENTS AND CONTROLS PAGE 41 ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J ANSWER 3.05 (1.00) d a 1 REFERENCE : ..
BFNPt BF-0I-68,pp'28, 29 EIH: EIH Simulator, Malfunct. ion #36 GGNS SIM. MAL. 11 s
ANSWER 3 06 (1.00) c REFERENCE
- T '
BFNP LP47 P. 28 EIH1 L ; ,9 0 ,7,1 4 , Figure 714-68.HNP-2-2447 GGNS SD B33' 1 p. 5&6 - -
GGNS LP OP-B33-1-501 P. 5 GGNS ARI B33-FAL-L603A
~
ANSWER 3.07 (2.00)
- b. Indication that all pairs of scram valves on all HCus are not in the same state
- c. Indication that the RGDS finds disagreement between the signals received fram the 2.RACS ,
- d. Indication that the withdrawn rod supt be fully inserted before any other. control rod can be moved (0 5 ea)
REFERENCh GGNS LP. P-C11-2-501
\
1 q
a 9
'h *.
l
i
- 3. INSTRUMENTS AND CONTROLS PAGE' 42
[
! ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J l-I ANSWER 3.08 (1.00) r
- o. 2 (0.5)
- b. To prevent an. excessive change in the LHGR. (0.5) l REFERENCE GGNS LP OP-C11-2-501 ANSWER 3.09 (3.00)
- c. Group it -- CB-1 Fully Inserted
-- CB-2 Fully Inserted
-- CB-5 Open
-- CB-5 opposite loop Open
-- CB-2 Open Group 21 -- Pump Speed 20% - 26%
-- Pump Motor Voltage not <75V for 4 Sec.
-- LFMG at Rated Voltage (.15 ea)
- 6. Prevents valve cycling CO.5] when Recirc Pump speed changes CO.53 FEFERENCE CGNS OP-N33-1 ANSWER 3.10 (1.00) a REFERENCE GCNS LP OP-N32-2-501 ANSWER 3.11 (1.00) c REFERENCE GGNS: 30I-04-1-01-C11-1, pp 2, 3
., 1 3.- INSTRUMENTS AND CONTROLS PAGE 43 ANSWERS -- GRAND GULF'l -85/08/12-MUNRO,J
- ANSWER 3.12 (1.00) e d
REFERENCE GGNS OP-D21-501,P.9 SD-D21,P.6
- ANSWER 3.13 (1.00) a REFERENCE
'BFNP LP419,P.5-6 GGNS SD-C51-2,P.16 ANSWER 3.14 (2.00)
.1) APRM 'A' fails)JuPscale -> relay K12A deenersi=es (0.4)
- 2) -> NHS contacts' K12 A in RPS Trip Logic A open (0.4)
- 3) -> Relays K14A & E deenergize -(0.4)
- 4) --> Contacts K14A'& E open (0.4)
- 5) -> Scram-solenoids for RPS A deenergize (0.4)
REFERENCE-BFNP L/P 428 EIH: L-RG-720, Fig 720-1a, -1be -2a, -2b, -3a, -3b
.GPNT, Vol. VI, Chapter 9.3.1-2, 3, 4 GGNS: SD-C71,P.2-7 ANSWER 3.15 (1.00) b 2
REFERENCE GGNS OP-R21-501,P.7,17,21 >
)
f I
i 4
e b
-1 -
i i
_ . - _ _ _ , . _ . -, - ____,._,_.m ._,..,m_
- 3. INSTRUMENTS AND CONTROLS PAGE 44 ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J ANSWER 3.16 (3.00)
ATHS RPT s . --1125 p sig -TCV fast closure
-level 2 (-42') -TSV trip b.-opens CB-2 -opens CB-3
-opens CB-5 -opens CB-4
-initiates Fast to Slow Transfer c.-ATWS Test Switches (HS-M616A/B) -auto if < 30% first sts press
-keylock switches on RPS logic (0.5ea)
REFERENCE
-GGNS SD-833-irP.23-24,21
- !'SWER 3.17 (1.00)
..oad' Reference Control automatically switches off during:
- 1. a load rejection below 12% power (power <12% for > 5 sec.), or -
- 2. a load rejection above 35% power (0.5 ea)
F REFERENCE GGNS LP OP-N32-2-501 9 3 Ts c e ;~c " Tem Per 5 < cl i - ev 1.4. 3. G . A 4O %
4o ' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 45
~~~~ - ----~~~-----------------
RA5i5L55i5At 55sTs5L ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J ANSWER- 4.01 (1.50)
The following checks should be made:
- Breaker charging springs charged. (0.5)
- Charging motor disconnect switch on. (0.5)
- Control power on. (0.5)
REFERENCE GGNS Proc. 02-S-01-2 GGNS LP OP-AD-539 ANSWER 4.02 (1.00)
C REFERENCE EIH: HNP-2-1933, p2 .
GGNS: ONEP-05-1-02-IV-1 ANSWER 4.03 (1.50) D.Id* 4r 850 psis.
- 2. Main Steam Line Low Pressure Alarm cleared.
- 3. All operable APRH's indicating > 5% power.
- 4. All APRM Downscale Alarms cleared.
- 5. Transfer one (1) IRH/APRM Recorder in each Division to APRM and Place IRM recorders in Slow Speed (4 0 0.5 each)
G, A 14, A cc<cOJ A a ,-.s % w e v <n E ..t . s ce- m (. .s . g , J p,,,, g g REFERENCE GGNS: 10I-03-1-01-1, pp 47, 48 ANSWER 4.10 (1.50)
- 1. Work Permit (WP)
- 2. Surveillance Permit (SP)
- 3. General Access Permit (GAP) (0.5 each)
REFERENCE ^
GGHS: 01-5-08-2, pp 17, 18 l ANSWER 4.11 (1.00) b I
l
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 48
~
~~~~R d656L66565L 66 TRUL~~~~~~~~~~~~~~~~~~~~~~~~
. ANSWERS -- GRAND GULF 1 -85/08/12-HUNR0rd REFERENCE GGNS: ONEP-05-1-02-I-4, p3 ANSWER 4.12 (1.00) d REFERENCE GGNS: ONEP-05-1-II-1, p4 ANSWER 4.13 (1.50)
- a. Loop Manual
- b. 10%
- c. 5% (0.5 each)
REFERENCE GGNS: ONEP-05-1-02-III-3, p2 ,
ANSWER 4.14 (1.00) a REFERENCE BFNP: BF-0I-66, pp 5-7 EIH: HNP-1001, pp 19, 20 GGNS: IOI-03-1-01-1, p 33: SOI-04-1-01-N33-1, p4 ANSWER 4.15 (1.00) a REFERENCE GGNS: ONEP-05-1-02-V-1, p 1
,6 .. *
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 49 !
- ------~~----------------
~~~~R A5i5L55fCAL 55sTR5L ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J ANSWER 4.16 (1.00) b REFERENCE EIH: HNP-2-1946 GGNS: DNEP-05-1-02-V-5, p2 ANSWER 4.17 (2.00)
- a. (1) 14.5 feet (2) 212 des F (3) 140 des F (0.5 each)
REFERENCE GGNS: EP-3, _p 6; EP-5, p 2; EP-7, p 1 ,
ANSWER -4.18 (1.00)
GGNS: '2.5 mR/hr - < 100 mR/hr (0.125 credit for 100 mR/hr)
REFERENCE EIH: GET Handbook, p 25 GGNS: 01-S-08-2, p 15 ANSWER 4.19 (1.00)
(od) ">
- 6. k % r..%.Le
EHC. Lock-up preventing LOL increases REFERENCE P r,t< d+ c Tne GGNS: I0I-03-1-01-2, p7
,w-+ n-e e , - -
A
- ,, . a . *~
4 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 50
- ----------~~----~~------
~~~~R dbf6L55fEdt E5sTR6t ANSWERS -- GRAND GULF 1 -85/08/12-MUNR0rd ANSWER' 4.20 ( .50)
- a. Gas Pressure (PI-R131) remains constant
' REFERENCE :
GGNS'SOI-04-1-01-C11-ir p9 i i
ANSWER 4.21 (1.00) b REFERENCE GGNS: Procedure 01-S-06-01 e
t e
-*, -- a~ , , - ,-.-n . ,, . . , , , , ,
- . - - , , - . - . , . - . , . - 4 .,-,,.7-. , - - , . . n . --, - , , ,
Mo s7 Lyt ENCLOSURE 3 U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: GRAND GULF 1 REACTOR TYPE: BWR-GE6
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ~ . _ _
DATE ADMINISTERED: 85/08/12 EXAMINER: MUNR0rJ o _________________________
sPPLICANT: _________________________
INSTRUCTIONS TO APPLICANT:
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 Stade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF
~
CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY
~~'~~~~- ~~~~~~~~ ~~~~~~~~~~~~~~~~~~~~~~'-~~~~~~~~~~
~~~~~~~~ '5~Y ST'
9
_I'6*50_.____ _221'.'" ___________
_ ________ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS 27 PLANT SYSTEMS DESIGNr CONTROL,
_I'_8.50_ _ _ _ _ _ ! _ ! _' 8 ___________ ________ 6.
AND INSTRUMENTATION
________ 7. PROCEDURES - NORMAL, ABNORMAL,
_2 111__ _I_ '.(( ___________
EMERGENCY AND RADIOLOGICAL CONTROL 26.00 ADMINISTRATIVE PROCEDURES,
________ ______ ___________ ________ 8.
CONDITIONS, AND LIMITATIONS teS.s da 157.GG 100.00 TOTALS FINAL GRADE _________________%
All work done on this examination is my own. I have neither givsn nor received aid.
~
5EPEICEUT 5 55GU55UR5~~~~~~~~~~~~~~
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDSr AND PAGE 2 GUESTION 5.01 (1.00)
The rate of change of power in a nuclear reactor is governed by the average neutron generation time (1-av). How does 1-av change as the core ages?
- a. 1-av INCREASES due to the DECREASE in the effective delayed neutron fraction (B-bar) over core life.
- b. 1-av DECREASES due to the DECREASE in the effective delayed neutron fraction (B-bar) over core life.
- c. 1-av INCREASES due to the INCREASE in the effective delayed neutron fraction (B-bar) over core life.
- d. 1-av DECREASES due to the INCREASE in the effective delayed neutron fraction (B-bar) over core life.
4 GUESTION 5.02 (1.00)
Which of the following radiation exposures would inflict the GREATEST ~
biological damage to man?
- c. 1 Rem of NEUTRON
- d. NONE of the above; they are all equivalent GUESTION 5.03 (1.00)
When does a constant-speed centrifugal pump motor draw the LEAST current?
- a. at ' runout' conditions
- b. at its " operating point'
- c. while 'cavitating'
- d. at " shutoff head' conditions
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
'5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 3
-QUESTION 5.04' (1.00)
Which of:the following equations is used to perform a BWR reactor heat' balance?
NOTE: c=CRD;-f=Feedwater; s= Steam; ri=RWCU in; ro=RWCU out
- a. b-rx = (wc + w,)~ x hs+ w,,x h,,+ d-rad - wg x h,- w,;x h,;- w x h,- Q pump b..b-rx = (w,+ w,) x h,+ w,;x
~
h,.+ b-rad - w,x he- w,,x h,- w,x h - b pump e
- c. b-rx = (wc + wp) x h,+ u,x h,,+ f b pump - w,x h,- w.x'h 7 ec
- Wc n h - b-rad d.. $-rx = wg x'h,+ w;x h,.+ b-rad - (w,+ w,) xh-w,,
r 3 x h,,- w, x h,- d pump OUESTION ~ 5.05 (2.00)
- a. DEFINE ' Critical Power'.
- b. Which one of the following conditions would tend to INCREASE the Critical Power level assuming.all other variables remain unchanged?
- 1. Inlet subcooling is DECREASED
- 2. Reactor pressure is DECREASED
- 3. The axial power peak is RAISED 4 '. Coolant flow rate is. DECREASED GUESTION 5.06 (1~.00)
Which of'the following conditions will result in the largest (MOST negative) Doppler / fuel temperature coefficient?
a._1000F fuel temperature with 10% voids
- b. 2000F fuel temperature with 10% voids
- c. 1000F fuel temperature with 30% voids
- d. 2000F fuel temperature with 30% voids (xxxxx CATEGORY 05 CONTINUED ON NEXT PAGE *****)
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 4 GUESTION 5.07 (2.00)
The reactor is supercritical on a 60-second period in the source range at BOL. Assuming no further rod movements and the following initial conditions, CALCULATE the FINAL STABLE values of reactor temperature and pressure. Show assumptions and calculations where necessary.
- a. Reactor temperature 281 F (1.5)
- b. Reactor pressure 35.3 psis (0.5)
QUESTION 5.08 (1.00)
A new Periodic NSS Core Performance Log (P1) is run and several LPRMs are noted to have a Base Crit Code of 2. What does this indicate to the oper-ator AND what should be done to correct the situation?
GUESTION 5.09 (1.50)
- a. BRIEFLY DESCRIBE the Sm-149 transient response following a shutdown" from a 100% equilibrium condition and a subsequent restart to 100%
power after a three-month outage. (1.0)
- b. HOW does the 100% equilibrium Sm-149 concentration compare with the 50% power equilibrium concentration? (<, >, =) (0.5)
GUESTION 5.10 (1.50)
Using the enclosed Mollier Diagram, LIST the following property values for steam with an enthalpy of 1390 BTU /lbm and an entropy of 1.568 BTU /lbm - F.
- a. Fressure
- b. Temperature
- c. Superheat
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 5 GUESTION 5.11 (1.50)
Four variables.are combined to form a dimensionless constant called the Reynolds number which describes the type of flow, either laminar or turbulente.in a system. LIST three (3) of these four variables.
QUESTION 5.12 (1.00)
FILL IN THE BLANKS The reactor period for any reactor shortly after a scram will be
____________ seconds because of __________________________________.
QUESTION 5.13 (1.00)
Which of the following post accident containment hydrogen contributors is dependent on the radiation field intensity inside containment for the acount of hydrogen released?
-a. Zr + H 2 O -> Zr0 + H s
- b. 2H2 O -> 2Hz + Os
- c. 2A1 + 3H 2O -> A1 203 + 3H s
- d. Fe + Hs0 -> Fe0 + Hz QUESTION 5.14 (1.00)
CALCULATE the GUALITY of a 540 degree F vapor-liquid mixture whose specific enthalpy is 1175 BTU /lbm.
(**xxx CATEGORY 05 CONTINUED ON NEXT PAGE *****)
-5. . THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 6 QUESTION 5.15- (1.50)
For each of the following meteorological / evaporative cooling parameter changes, state whether you would expect condenser vacuum to INCREASE.
(lower absolute pressure), DECREASE (higher absolute pressure), or REMAIN UNCHANGED.
- a. Relative humidity increases from 60% to 90%.
- b. A temperature inversion occurs.
- c. Ambient temperature drops from 90 F to 80 F.
. QUESTION 5.16 (1.00)
The THRESHOLD power below which PCI failures do not occur is known to )
DECREASE with fuel burnup. STATE two (2) reasons for this decrease in the PCI threshold.
QUESTION 5.17 (1.00)
Adding latent heat to liquid water at saturated conditions will...
- a. increase the temperature of the water.
- b. change-the water to steam at the same temperature.
- c. change the water to steam at a slightly higher temperature.
- d. decrease its subcooling by making it boil.
(xxx*x CATEGORY 05 CONTINUED ON NEXT PAGE xxxxx) l l
l
- 5. THEORY OF NUCLEAR POWER PLANT CPERATION, FLUIDS, AND PAGE 7 QUESTION. 5.18 (2.00)
The attached figure (15.4-5) illustrates a transient that could occur l at a BWR.
GIVEN: (1) A fast opening of BOTH recire. FCVs at 11% per second.
(2) No operator actions are taken.
(3) Valve cle:::r: begins at time = 0 seconds.
ceau ~8 EXPLAIN the cause of the following recorder indications:
- a. The decrease in core inlet flow after ~10 seconds on graph (a).
-1542
'- b. The peak in vessel pressure at 24KF seconds on graph (b).
- c. The decrease in feedwater flow between '23-30 seconds on graph (c).
- d. The reactor scram at ~1.5 seconds on graph (d).
QUESTION 5.19 (1.00)
Water is an excellent neutron moderator. What are TWO (2) NUCLEAR FACTORS which make water the moderator of choice for most commercial reactors?
QUESTION 5.20 (1.50) i LIST three (3) factors upon which a reactor's decay heat generation rate is dependent.
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 8 QUESTION 5.21 (1.00)
Which of the following is NOT a characteristic of Soberitical Multiplication?
- a. The.suberitical neutron level is directly proportional to the neutron source strength.
- b. Doubling the indicated count rate by reactivity additions will reduce the margin to criticality by approximately one-half.
- c. For equal reactivity additions, it takes longer for the new equilibrium count rate to be reached, as K-eff approaches unity.
- d. A single notch of rod withdrawal will produce an equivalent equilibrium count rate increase whether Keff i s 0.88 or 0.92.
1 d
(***** END OF CATEGORY 05 *****)
i n >-w-..~ .w . > - - , . . - . - - - - - + --- y>, , ----e.n.< +-m m rp , r , w --
60 PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 9 GUESTION 6.01 (1.00)
Assume that APRM 'B' currently has 14 operable LPRM inputs and
< is reading 65% power. Which of the following indication (s) and/or action (s) will occur as a result of 1 LPRM (of the 14 remaini'ng LPRM inputs to APRM 'B') failing downscale? Assume NO operator action.
Rod Block - APRh 'B' reading 65%
Rod Block - 1/2 Scram - APRM 'B' reading 65%
GUESTION 6.02 (1.00) 4 Which of the following axial location sequences correctly describe the axial locations of LPRMs in the core? -
- a. BAF - 'A'G+9' -
- B'9+27' -
'C'G+45' -
'D*G+63'
- b. BAF - 'A'G+18' -
'B'0+54' -
'C'G+90' - 'D'O+126'
- c. BAF - 'D'O+9' -
'C'G+27' "B'9+45' -
'A'G+63'
- d. BAF - 'D'O+18' -
'C'9+54' -
"B'G+90' - 'A'O+126' (xx*** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
, .-. ...---,,mc--.. - , . -_ , - . - -
- , . - + - + ~n- --
- 6. PLANT SYSTEMS DESIGNr CONTROL, AND INSTRUhENTATION PAGE 10 GUESTION 6.03 (2.00)
For each of the following situations (i and 11) select the correct Feed-water Control System / plant response from the list (a through e) which follows. An answer may be used more than or.ce, and N0 operator actions are taken. ,
- a. Reactor water level decreases and stabilizes at a lower level.
- b. Reactor water level decreases and initiates a reactor scram.
- c. Reactor water level increases and stabili:es at a higher level.
- d. Reactor water level increases and initiates a turbine trip and Reactor Scram.
- e. None of the above. >
- i. The plant is operating at 90% power in 3-element control when the HPCS system inadvertently initiates and injects.
ii. The plant is operating at l'00% power, in 3-element control, when one Feed Flow Detector FAILS DOWNSCALE. ,
GUESTION 6.04 (1.00)
The reactor is critical at approximately 10 psis and the 'RX Heatup and Pressurization' phase of 03-1-01-1,RX SU is being performed.
The narrow range P-680 level instruments read the following ' approx-imate' values
- NR LT-N004A 37' NR LT-N004B 38' NR LT-N004C 37' The WIDE RANGE P-680 indicators should read which of the following approximate values?
- a. O inches.
- b. 15 inches
- c. 38 inches
- d. 60+ inches
(*x*** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
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- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 11 QUESTION 6 05 (1.00)
Which of the following is NOT a symptom that you would expect to see as a result of a " Jet Pump Riser Failure *? Assume Recire Flow Control is in ' Flux Manual *.
a.
DECREASE in failed Jet Pump flow.
- b. DECREASE in core differential pressure.
- c. DECREASE in' reactor (APRM) power.
4
- d. INCREASE in indicated core flow.
QUESTION 6.06 (1.00)
The plant is operating normally at power when you receive a ' Pump A Seal Staging Flow High/ Low' alarm and note a DECREASE in No.2 Recire Pump seal pressure. Which of the following failures would cause this indication?
- a. Failure of No. 1 seal ,
- b. Failure of No. 2 seal
- c. Plugging of the No. 1 internal restricting / breakdown orifice
- d. Plugging of the No. 2 internal restricting / breakdown orifice QUESTION 6.07 (1.00)
Which of the following is the only normally CLOSED valve in the e RCIC steam supply flow path in the at power Standby lineup?
- a. Steam Supply Valve (F045)
- b. ~'atboard Steam Isolation Valve (F064)
- c. rbine Trip Throttle Valve
- d. arbine Governor Valve
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
-- - - - - - ---.,7, , + - . . - - , ..-g--.e.~.g .,_,,-e - , , . , - ,. .~._ ,
60 PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 12 GUESTION 6.08 (1.00)
Which of the following sequences of components correctly reflects the normal RCIC water flow Path for injection into the Reactor?
What is(are) the automatic isolation signal (s) to the RCIC Vacuum Breaker Isolation Valves (F077,F078)? Setpoints required.
QUESTION 6.10 (2.00)
Briefly explain what condition (s) will generate EACH of the following indications on the Operator Control Module. -
- a. Data Fault
- b. Scram Valves
- c. Channel Disagree
- d. Insert Required GUESTION 6.11 (1.00)
- a. Fill in the following blank Above the HPSP, continova withdrawal of a control rod is automatically limited to _____ notch (es). (0.5)
- b. What is the reason for this limitation? (0.5)
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 13
~ QUESTION 6.12 (3.00)
Consioer a Recirc Pump Fast to Slow Speed transfer
- a. . After ' tripping CB-5", certain permissives must be met to "close CB-2' and complete the speed transfer. Indicate the 8 permissives (in 2 groupings) that are left blank on Figure 9 - Transfer Sequence. (2.0)
- b. Briefly explain the reason for ' tripping the FCV to Manual' in the sequence. (1.0)
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
40 PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 14 QUESTION 6.13 (1.00)
The plant is operating at 100% power with Recire Flow control in oFlux Manual'. An operator inadvertently INCREASES the ' Pressure Reference Set' on the EHC Turbine Control System by 5 psis.
ASSUME: 1. No further operator action.
- 2. All other EHC control settings are normal.
- 3. Starting Parameters:
TCVs (MSCV & LPSCVs) - 100% Steam Flow Position BSCVs -
0% Steam Flow Position Rx Power - 100% Rated Thermal Power Rx Pressure - 1025 psi 3 NOTES: All valve %s are in % Steam Flow Position.
See Figure 7 (EHC Logic Diagram) for information.
Which of the following most accurately describes both the INITIAL RESPONSE and FINAL STATUS of the different parameters and components?
tuo tt- cyN t (b u e 3 il ,. J 4 .d : < ( ( l u. .: -G . i~ d:a l f -E..I TL w- w .
a b c d .
INITIAL RESPONSE
- TCVs IPartial IPartial IPartial INo Change IClose (<100%) IClose (<100%) IClose (<100%) 1
-BSCVs INo Change IPartial INo Change IPartial i IDpen (>0%) i 10 pen (>0%)
-Rx Power IIncrease INo Change IIncrease ~
IDecrease
-Rx Pressure IIncrease INo Change IIncrease IDecrease FINAL STATUS l g i : f
-TCVs I"100% IPartial 10% I"100%
I IClose (<100%) l I
-BSCVs 10% IPartial 10 pen (as 10%
i 10 pen (>0%) Inecessary)
- I
-Rx Power l>100% l>100% I"0% l<100%
-Rx Pressure l>1025 psis l>1025 psig I"920 psis l<1025 psis
- Open as necessary for SD Pressure Control
(***** CATEGORY 06 CONTINUED ON NEXT PAGE xxxxx) l
6o PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 15 QUESTION 6.14 (1.00)
How would a loss of service air affect the operation of the Standby Liquid Control System (SBLC)?
- a. The SBLC tank level indication would be inoperable.
- b. The SBLC tank air sparger would be inoperable.
- c. The SBLC tank level indication and air sparger would be inoperable.
- d. It would have NO impact since the instrument air system supplies all SBLC needs.
QUESTION 6.15 (1.00)
The containment flooding mode of RHR is available as a backup when virtually all.other means to keep the core covered have failed. Briefly-DESCRIBE the flowpath established durin3 the containment flooding mode.
QUESTION 6.16 (1.00)
The General Area Radiation Monitors (ARMS) have installed check sources.
These sources...
- a. are normally shielded and are exposed by depressing the green backlit Check Source pushbutton.
- b. are automatically exposed every 17 minutes to test proper module response.
- c. do not affect the ARM's indicated background radiation level in those areas monitored.
- d. aid in the detection of equipment malfunctions which cause downscale trips.
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
6o PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 16 QUESTION 6.17 (1.00)
The plant is operating at power with A, B, and C CCW pumps running and NONE of the pumps selected for STANDBY operation. A LOSP occurs and the diesels start and tie in normally. How will the CCW system respond during this transient?
- a. The LSS panel will auto start thp *B" CCW pump on ESF power 20 seconds after the bus is teenergized.
- b. Either the 'A' or 'B' CCW pumps can be started manually on ESF power after the buses are reenergized.
- d. The 'B' CCW pump can be manually started by the operator on ESF power after the bus is teenergized.
QUESTION 6.18 (1.50)
Answer the following with regard to the SRVs / ADS: -
- a. Which ADS interlock / permissive signal is NOT bypassed (i.e. must be present) to allow manual initiation of ADS from the 601 panel? (0.5)
- 6. Panels 601 and 631 have red and green SRV indicating lights. An illuminated red light on P601 vindicates that _________________________
while an illuminated red light on P631 indicates that _______________.
(BE SPECIFIC.) s L am o.e..a GUESTION 6.19 (1.50)
Fill in the following blanks with the appropriate (if any) LPCS injection valve (F005) interlocks and setpoints:
Manual opening of F005 with the handswitch is prohibited when _____(a) ___.
If power is available, F005 will auto open on a LPCS initiation signal of
______ ( b) __________ or _________ ( c) __________. Once open, (d) ______
signal will auto close the valve. If the auto open signal ______is manually overridden the valve will reopen automatically if __________(e) _________.
(xxxxx CATEGORY 06 CONTINUED ON NEXT PAGE xxxxx)
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 17 QUESTION 6.20 ( .50)
Reactor pressure is 900 psis and LPCS is running in response to a valid initiation signal. What is the approximate expected flow indication on the pump dischaP3e flow meter on the 601 panel?
GUESTION 6.21 (1.00)
Reactor Feed Pump (RFP) turbine speed is controlled by either a Motor Speed Changer (MSC) or an Electric Automatic Positioner (EAP). The EAP ...(CHOOSE ONE)
- a. ... will control the RFP turbine's speed only if its speed signal is greater than that from the MSC.
- b. ...is normally used to control feed flow rate over a turbine speed of 0 - 5500 rpm.
- c. ..., unlike the MSC, does NOT afford the capability of manual speed control by use of a local handwheel.
- d. ...will lock in place to prevent a ramp response to a false -
signal, if it loses its signal from the flow controller.
GUESTION 6.22 (2.00)
The plant is operating at 100% RTP when APRM 'A' fails upscale and results in a reactor half-scram. Utilizing the attached RPS trip logic diagrams (Figures #141 A thru C) DESCRIBE in a STEP-BY-STEP fashion (with regard to the opening / closing, ener-gizing/deenergining of ALL applicable contacts and relays) how the APRM upscale trip results in an actuation of the scram solenoid.
NOTE: IF THE ATTACHED DIAGRAMS CAN NOT BE EASILY RCAD, ASSIGN THE CONTACTS / RELAYS, ETC NUMBERS AND REFER TO THEM IN YOUR ANSWER.
GUESTION 6.23 (1.00)
Under what 2 conditions will the EHC System ' Load Reference Control' AUTOMATICALLY switch 0FF?
(***** END OF CATEGORY 06 *****)
. ~ ..
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'7. PROCEDURES - NORMAL,-ABNORMAL, EMERGENCY AND PAGE 18 l
--- Es5iaE55iCat 55siE5E------------------------
QUESTION 7.01 (1.00) l With a stuck control rod, ONEP-05-1-02-IV-1, 'CRD Malfunctions *,
instructs the operator to INCREASE drive waier pressure in an attempt to initiate control rod movement. With the reactor at FULL POWER conditions, SELECT the MAXIMUM differential pressure to which the drive water may be raised.
- a. 90 psid
- b. 260 psid
- c. 350 psid 4
- d. 500 psid QUESTION 7.02 (1.50) DELETED %= h See M d A reactor SCRAM has occurred, but NOT all of the control rods have inserted to less than the 06 position. Reactor power is indicated ,
i as 6% on the APRH's. LIST the three (3) immediate operator action steps that are required per.ONEP-05-1-02-I-1, " Reactor Scram.'
NOTE: LIMIT YOUR RESPONSE TO THOSE ACTION STEPS REQUIRED FOR i REACTIVITY CONTROL.
QUESTION 7.03 (1.00)
SOI-04-1-01-P75-ir "Stanoby Diesel Generator' cautions the operator NOT to operate the diesel generator without air pressure. EXPLAIN l the basis for this caution.
'GUESTION 7.04-
- (1.00)
Assume that adequate core cooling CANNOT be maintained and
' Alternate Shutdown Cooling' must be established per EP-8. l DESCRIBE the RPV cooling water flowpath that should be estab-lished per EP-8.
NOTE: INCLUDE IN YOUR DESCRIPTION THE SYSTEMS / COMPONENTS WHICH ARE USED. ,
i ,
t
(***xx CATEGORY 07 CONTINUED ON NEXT PAGE *****)
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_ , _ . . _ , , , . . . , . . _ . _ _ _ _ , _ _ _ , _ . . . _ _ . _ . ._.___._,.a_.-__.__,_-.,__- . . _ _ , _ _ _ _ _ _ ._
+
(
+
x i 70 P.'0CEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 19
~~~~ ~
~~~~~~~~~~*'~~~~~~~~~~~~~
RI656L65555L 55 TR6L y
s s
4 GUESTION 7.05 (1.50? ',
EXCLUDING specific ~ control room annunciators, LIST three (3) symptoms, per ONEP-05-1-02-I-5 ' Resin Intrusion into the Reactor Vessel', that woul.d be indicative of such an event.
QUESTION 7.06 (1.50)
Per DNEP-05-1-02-III-5, ' Automatic Isolations', LIST the three (3) conditions which must be met before a system can be restored to service. Assume an automatic isolation HAS OCCURRED and that the cause of the,, isolation HAS BEEN determined.'
QUESTION 7.07 (1.00) -
t Per EP-2, ' Emergency Cooldown','which of the following most accurately describes how SRU
operation should be used to control
, pressure, if needed? ,
~
NOTE
- ASSUME'THAT THE INSTRUMENT AIR SYSTEM IS OPERATING PROPERLY
- a. Use numerous SRV's, with short pressure reductions
( ~ 50 psis) to equalize Suppression Pool heatup.
- c. Depressurice with.a sustained SRV opening to maximi =e the emergency cooldown rate.
- d. Allow the SRV's to operate by mechanical actuation to-ensure desisn pressure control and heat dispersion.
QUESTION 7.08 (1.00) ,
IDI-03-1-01-3, ' Plant Shutdown *, cautions the operator to reduce
. reactor pressure to approximately 400 psig, if possible, when it is desired to maintain a HOT SHUTDOWN condition. EXPLAIN the basis for this caution.
r .
(***** CATEGORY 07 CONTINUED ON ' EXT PAGE xxxxx) 4 k
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g *
.- -- +- ,-. . - - - - - - , - - -- --
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- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 20
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~~~~RE655L655 EEL 55NTR5L l
l' QUESTION 7.09 (2.00)
Per I0I-03-1-01-1, ' Cold Shutdown to Generator Carrying Minimum Load,' LIST four (4) conditions which must be met / satisfied prior to placing the Mode Switch in RUN.
! NOTE: INCLUDE SETPOINTS, IF APPLICABLE QUESTION 7.10 (1.50)
LIST the three (3) types of Radiation Work Permits (RWP's) which may be used to control access / account for personnel exposure. ,
QUESTION 7.11 (1.00)
Which of the following fulfills an Entry Condition into EP-5, Rapid RPV Depressurization'?
~
- a. Drywell temperature near the cold reference leg instrument vertical runs has INCREASED to within 93 des F of the RPV saturation temperature.
- b. Suppression Pool temperaure has INCREASED to a value of 122 des F.
- c. Containment pressure has INCREASED to a value of 5.6 psig.
- d. Containment temperature has INCREASED to a value of 201 des F.
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
- 7. PROCEDURES - NORMAL, ABNORMAL. EMERGENCY AND PAGE 21
~ -
~~~~EA655L6555AL 55ETRUL '~~~~~~~~~~~~~~~~~~~~
s, QUESTION 7.12 (1.00) l CAUTION I You enter an area Posted with the following signi gjg ab l MfGM l RADIAD004 AREA *
, KEEP OUT
~iWPRfU Ep IOR' LIST the minimally acceptable PERSONAL and PORTABLE dosimetry required for entry into this area.
QUESTION 7.13 (1.00)
The' Control Room is declared uninhabitable and evacuated. The immediate operator actions for ' Shutdown From the Remote Shutdown Panel', ONEP-05-1-III-1, are completed. RCIC then ISOLATES.
Level subsequently' decreases to Level 2. Restoration of level USING RCIC requires which of the following? i ASUME THAT THE THREE CONDITIONS NEEDED FOR RESETTING AN ISOLATION, PER ONEP-05-1-02-III-5, ' AUTOMATIC ISOLATIONS'r HAVE BEEN HET. I
- a. No Operator Action. RCIC will restart automatically.
~
- b. Operators Action. Close RCIC TURB TRIP /THROT VLV; Place RCIC TURB';FLO CONT in manual at minimum setting; Re-open RCIC TURB TRIP /THROT VLV and establish flow.
- c. Operator Action. Close RCIC TURB TRIP /THROT VLV; reset RCIC TURB TRIP logic; RCIC will now restart automatically.
- d. NONE OF THE ABOVE. RCIC cannot be restarted from the Remote Shutdown Panel after isolation.
(xxxxx CATEGORY 07 CONTINUED ON NEXT PAGE *****)
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- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 22
~~~~kd656L66ECdL 66UTR6L'~~~~~~~~~~~~~~~~~~~~~~~
QUESTION 7.14 (1.50)
ONEP-05-1-02-III-3, " Decrease in. Recirculation System Flow Rate'r directs operator actions for an unexpected decrease in reactor coolant system flow rate.
FILL IN THE BLANKS (After the unexpected decrease), if both recirculation loops are still operating, transfer the FCV's to ____(a) ____. Balance loop flows to within ____(b)____ at less than 70% core flow, or to within
____ ( c) ____ at greater than 70% core flow.
QUESTION 7.15 (1.00)
A plant startup is in progress'and condenser vacuum is being
-established in accordance with I0I-03-1-01-1, " Cold Shutdown to Generator Carrying Minimum Load". What is the proper sequence for. component / subsystem startups? ,
- s. Steam Seal Exhauster, Steam Seal Header, Mechanical Vacuum Pump, Steam det Air Ejector,
- b. Steam Seal Header, Steam Seal Exhausterr Mechanical Vacuum Pump, Steam Jet Air Ejector.
- c. Mechanical Vacuum Pump, Steam Seal Exhauster, Steam Seal Headere Steam Jet Air Ejector.
- d. Steam Seal Exhauster, Mechanical Vacuum Pump, Steam Seal Headerr Steam det Air Ejector.
QUESTION 7.16 (1.00)
Per DNEP-05-1-02-V-ir ' Loss of Component Cooling Water's a loss of CCW may 'tne either complete or partial. In which of the following instances would reduced flow (partial loss) be treated as a COMPLETE LOSS of CCW?
- a. Reactor Recite Pump Lemperatures above the HI alarm setpoint.
b.- RWCU NRHX Outlet temperature above the HI alarm setpoint.
- d. CRD Pump Oil temperature above the HI alarm setpoint.
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 23
- ~~~~~~~~~~~~~~~~~~~~~~~~
~~~~~EE655L555 EEL 55ETE5L GUESTION 7.17 (1.00)
The unit is operating at 70% RTPi vou notice power start to increase with NO CHANGE in recirculation flow or rod position. You suspect a ' Loss of Feedwater Heating.'
Which of the following is required / appropriate per ONEP-05-1-02-V-5?
- a. A 30% reduction in Recire Flow, monitored by Recire Flow indication.
- b. A 30% Power Reduction, using Recirc Flow, monitored by APRM's.
- c. Insertion of Shallow Rods, to maintain proper flux shape, prior.to reducing Recire Flow.
- d. Insertion of Power Rods, to maintain proper flux shape, prior to reducing Recire Flow.
QUESTION 7.18 (1.00) -
ONEP-05-1-02-VI-2, ' Hurricanes, Tornadoes, and Severe Weather *,
provides direction to the operator concerning verification /
vtilization of the Emergency Diesel Generators upon receipt of a severe weather warning. Which of the following most correctly describes these directions?
- a. Verify EDG Operability by performing appropriate Surveillance Tests on D/G's 11, 12, and 13 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
- b. Manually start EDG's 11 and 12. Load these diesels with their ESF Buses and SEPARATE these Buses from Off-Site power.
- c. Manually start EDG's 11 and 12. Pick up approx-imately 50% of the ESF Bus load and MAINTAIN these Buses synchronized with Off-Site power.
- d. Verify EDG Operability by performing appropriate Surveillance Tests on D/G's 11, 12, and 13 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Load D/G 13 with Division III loads and SEPARATE it from Off-Site power.
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
. .- o
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 24
~ ~~~~~~~~~~~~~~~~~~~~~~~~
~~~~R E653L 555IL 56NTR5L QUESTION 7.19 (2.00)
EP-3r EP-5, and EP-7 caution the operator to observe certain limitations on Suppression Pool Level and Temperature when
.cperating HPCS, LPCS, RHR, and/or RCIC.
- a. COMPLETE THE FOLLOWING: (1.5)
Suppression Pool Level shall not be less than ____(1) ____.
Suppression Pool Temperature shall not exceed ____(2) ____
during HPCS, LPCS, and/or RHR operation; it shall not exceed ____ ( 3) ____ during RCIC operation.
- b. STATE the basis for these temperature / level limitations on the Suppression Pool. (0.5)
. QUESTION 7.20 (1.00)
When raising power per I0I-03-1-01-2, " Power Operations,' you ,
are cautioned to maintain the' Load Demand Limited (LDL) value close to the Actual Generator Load (AGL) value.
- a. STATE how much the LDL value may exceed the AGL value. (0.5)
- b. STATE how you would know if this limit were exceeded (EXCLUDING THE DIGITAL METER 5 DN 1H13-P680-9D). (0.5) l l
. QUESTION 7.21 ( .50)
You are conducting a shutdown of the CRDH system, per SOI-04-1-01-C11-1. You open Drain Valve 107xx to drain the water accumulators. State the local-indication (s) which should be used to determine that the accumulator is fully drained.
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 25
~~~~ ~~~~~~~~~~~~~~~~~~~~~~~~
R5655L55555L'55 555L GUESTION 7.22 (1.00)
Upon recovering from a ' Loss of Off Site Power *r DNEP-1-02-I-4 cautions the operator that either the SJAE's be isolated -OR-the condenser vacuum be broken PRIOR to re-energining MCC's -
11B42, 12B42, and 14822. Which of the following is the basis for this caution?
- a. Prevent large reverse flows in the Off Gas system.
- b. Prevent inadvertent initiation of the Mechanical Vacuum Pumps.
- c. Prevent establishing combustible gas mixtures in the charcoal adsorbers.
- d. Prevent electrically tripping the cooling compressors in the Off Gas System.
(***** END OF CATEGORY 07 *****)
A
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 26 1
+
QUESTION 8.01 (1.00)
With' regard to the Surveillance Program Procedure, 01-S-06-12' Answer with True or False
- a. The Shift Supervisor may approve a special test for the purpose of satisfying a Technical Specification requirement.
- b. A system, which has a TS surveillance procedure that makes the system / equipment inoperative, is NOT subject to the restriction (s) of the applicable TS LCO ACTION statement during the performance of this surveillance procedure.
QUESTION 8.02 (1.00)
Which of the following choices will correctly complete the blanks for the MCPR LCO listed below?
The MCPR shall be equal to or ____ (1) ____than ____(2) ____ MCPR(f)
____ ( 3) ____MCPR(p) limits at indicated core flow and THERMAL-POWER as shown in~ Figures 3.2.3-1 and 3.2.3-2. -
NOTE: Figures 3.2.3-1 and 3.2.3-2 are enclosed for reference.
(1) (2) (3)
- a. greater i the smaller of the ; or
- b. less i the larger of the ; or
- c. greater i both i and
- d. less i both i and GUESTION 8.03 (1.00)
What are the minimum number of operable SRM channel (s) required for the following Operational Conditions?
OPERATIONAL CONDITION NUMBER OF SRMs (with IRMs Range 2 or below) ______
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 27 QUESTION 8.04 (1.00)
The APRM Trip Setpoint Formula is (.66H+48%)T. Which of the following choices correctly details the definition of 'T' AND when it is applied?
- a. T = FRTP/MFLPD ; T applied if < 1.0
- b. T = MFLPD/FRTP ; T applied if < 1.0
- c. T = FRTP/MFLPD ; T applied if > 1.0
- d. T = MFLPD/FRTP ; T applied if > 1.0 QUESTION 8.05 (3.00)
Answer the following with regard to the Control of Temporary Alterations Procedure, 01-S-06-3:
- a. The removal or installation of temporary electrical leads or jumpers identified in an approved procedure or Maintenance Work Order are not considered Temporary Alterations previded ,
the procedure or HWO meets TWO requirements. What are these two requirements? (1.0)
- b. The mimimum level of qualification for an Independent Verifier (per this procedure)qshall be ... (CHOOSE ONE) (1.0)
- 1. ... journeyman level W n,-<3 Cr T* ~* r ~n A kw d e
- 2. ... NOB for operations
- 3. ... NOA for operations
- 4. ... Shift Supervisor
- c. Which of the following colors must the wire for Temporary Jumpers be made from? (1.0)
- 1. Purple
- 2. Orange
- 3. Green
- 4. No Color Specified (xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE ***xx) l
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 28 QUESTION 8.06 (1.50)
With the exception of breaker position, what THREE (3) items should an operator check on a breaker during the performance of a system lineup checksheet per Control and Use of Operations Section Direc-tives, 02-S-01-2? Consider Local checks only.
, GUESTION 8.07 (1.00)
The Shift Supervisor may use duplicates of controlled Operations Section Directives provided-that the duplicate is NOT used for a period of time greater than ... (CHOOSE ONE)
- a. .. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
- b. .. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
- c. .. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
- d. .. 7 days GUESTION 8.08 (1.00)
What are the Two (2) provisoes / stipulations that must be met in order to allow 'out of sequence" completion of IOI procedural steps?
(xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE *****)
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 29 GUESTION 8.09 (2.00)
With regard to the Protective Tagging System Procedure', 01-S-06-01:
- a. Fill in the following blanks: (1.0)
"All red' tags will be independently verified for all Safety Related systems and components. Safety Related components in Non-Safety Related systems will be determined through the use of a ____(1) ____ list. Any list component shall be indepently verified, unless ____(1) for ____ALARA reasons, the ALARA committee recommends otherwise. The ____(2) ____ will authorize deletion of independent verification.'
- b. The minimum level of qualification for an Independent Verifier (per this procedure) shall be ... (CHOOSE ONE) (1.0)
- 1. ... journeyman level
~
- 2. ...NGB for operations
- 3. ...NOA for operations 4.- ... Shift Supervisor QUESTION 8.10 ( .50)
Given the following conditions on the unit:
Mode Switch - Shutdown Temperature - 180 des F Pressure - O psig Level - 36 inches
-RHR - SDC Mode RPV Head Bolts are Detensioned State the Operational Condition of the plant as described above.
(***** CATECORY 08 CONTINUED ON NEXT PAGE *****)
i e
i
- - - - - - - - - - . . n- , - , , ----,-e-- - - - - - - --.
v -- - - - - . ,
.-.+---m.- ,e, ..
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 30 QUESTION 8.11 (1.00)
The plant is at 60% power with only one outstanding LCO:
- The LPCS pump is INOP due to an in-progress (1 day) repair.
There is no estimate of repair time.
Ten minutes into the shift, DG 12 fails to start twice during the performance of a scheduled surveillance and is declared INOP. There is no estimate of repair time.
Which of the following actions most correctly detail the allowances and/or limitations imposed by the Technical Specifications in this instance?
NOTE: APPLICABLE TSs ARE ENCLOSED FOR REFERENCE.
- a. Be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- b. Power Operation may continue for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; and then, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- c. Power Operation may continue for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; and then, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- d. Power Operation may continue for 6 days; and then, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the-following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
I l
'8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 31 QUESTION 8.12- (1.00)
Fill in the blank with one of the following TS terms:
"A ________ shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent channels measuring the same parameter.'
- a. Channel Calibration
- b. Channel Check c.- Channel Functional Test
- d. Logic System Functional Test
'0UESTION 8 .1'3 (1.50)
List the three(3) Refuel position interlocks required to be operable per the Technical Specifications for the performance of any Core Alteration. Assume the Reactor Mode Switch is locked in Refuel.
QUESTION 8.14 (1.00)
With regard to the Control of Refueling Operations Procedure, 01-S-06-10:
Answer with True or False:
- a. An SRO will be present on the Containment Refueling Floor at all times when any Core Alterations are in progress.
- b. When irradiated fuel movement not involving Core Alterations is in progress, an SRO will be in charge either in the Containment or the Refueling E:vilding.
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 32
'0UESTION 8.15 (1.00)
The plant is at 60% power with only one outstanding LCO:
Hydrogen Recombiner 'A' is INOP due to an in-progress (1 day) repair. It is anticipated that repairs and return to service will be complete in two(2) weeks.
Tan minutes into the shift an Instrument Technician reports that the Hydrogen Recombiner 'B' 'PWR ADJ" Potentiometer is faulty and will produce only a zero(0) power level signal.
Which of the following actions most correctly detail the allowances and/or limitations imposed by the Technical Specifications in this instance?
NOTE: APPLICABLE TSs ARE ENCLOSED FOR REFERENCE.
- a. Operational Condition 1 may be maintained for approxi-mately 29 days .
- b. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> measures must be initiated to place the unit in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- c. Within i hour measures must be initiated to place the unit in at least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- d. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> measures must be initiated to place the unit in at least STARTUP within the ne::t 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
(***** CATEGORY 08 C0r1TINUED ON NEXT PAGE *****)
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 33 j
QUESTION 8.16 (2.00)
MATCH the following emergency classifications to their appropriate d2finitions.
- a. Unusual Event 1. The occurrence of an event or events which involve actual or likely major failures of the plant functions needed for the protection of the public.
- b. Alert 2. The occurrence of an event or events which indicate a POTENTIAL degradation of the level of safety of the plant.
- c. Site Area Emergency 3. Events are in process or have occurred which involve actual or imminent substan-tial core degradation or melting with the potential for loss of containment integrity and substantial releases of large amounts of radioactive material off-site.
- d. General Emergency 4. The occurrence of an event or events which involve an actual or potential SUBSTANTIAL degradation of the level of safety of the plant.
QUESTION 8.17 (1.00)
In accordance with 10 CFR 55, 'if a licensee has not been actively performing the functions of an operator or senior operator for a period of ___(1) ___ months, or longer, he shalle prior to resuming activities licensed pursuant to this part, demonstrate to the Commission that his knowledge and understanding of facility oper-ation and administration are satisfactory."
FILL:IN THE BLANK WITH ONE OF THE FOLLOWING TIMES:
I
- a. 4
- b. 6
- c. 12
- d. 24
(*x**x CATEGORY 08 CONTINUED ON NEXT PAGE *****)
L- -
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 34 QUESTION 8.18 (1.00)
Unit 1 is operating at 75% rated thermal power. Channel Functional Tests are performed on all of the MSL Radiation Monitoring System
. channels. Channels A and D test UNSAT; Channels B and C test SAT.
Maintenance has no estimate of repair time and will not be able to commence troubleshooting and repair for at least 16 - 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.
Which of the following actions most correctly detail the allowances and/or limitations imposed by the Technical Specifications in this instance?
NOTE: APPLICABLE TSs ARE ENCLOSED FOR REFERENCE
- a. Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- b. Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- c. Place MSL Rad Mon Channel 'A' in the tripped condition within one hour -AND- be in at least HOT SHUTDOWN within
! 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- d. Place MSL Rad Mon Channel 'A' in the tripped condition within one hour -AND- be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
QUESTION 8.19 (2.00)
STATE the four (4) basic PCIOMR rules currently in effect at GGNS (i.e.
exposure < 3.3 Gvd/ST).
QUESTION 8.20 (1.50)
Technical Specifications define SHUTDOWN MARGIN as ...
- ' SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is soberitical or would be suberitical assuming ...
and the reactor is in the shutdown condition;
- List the plant conditions which complete the definition of SHUTDOWN MARGIN.
(***** END OF CATEGORY 08 *****)
(************* END OF EXAMINATION ***************)
't a
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 35 ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J
' ANSWER 5.01 -(1.00) b REFERENCE BFNP NEUTRON SLOWING DOWN AND DIFFUSION LP,P.6-8 GGNS OP-NP-510,P.6,9;511,P.4 ANSWER 5.02 (1.00) d REFERENCE BFNP MCD BWR LP,P.4 GGNS JP-RP-502,P.5-7 ANSWER 5.03 (1.00)
-d REFERENCE BFNP PUNPS LP,P.5-6 GGNS OP-HF-514 ANSWER 5.04 (1.00) a REFERENCE GGNS OP-AD-545,P.10 ANSWER 5.05 (2.00)
- a. The assembly power which would cause the onset of transition boiling at some Point in the assembly. (1.0)
- 6. 2
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 36
~~-----
~~~~TsERs557sAsicS t
{ ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J REFERENCE BFNP TRANSITION BOILING & ATLAS TESTING LP,P.5-6 j GEXL CORRELATION & CRITICAL POWER LP,P.3 GGNS MCD, THERMAL LIMITS, P.26,32-33 j ANSWER 5.06 (1.00) c REFERENCE BFNP REACTIVITY COEFFICIENTS LP,P.3,15
- GGNS OP-NP-513, P.13-14 i
ANSWER 5.07 (2.00) i i
1 l a. Assume: B=0.007 lambda =0.1/sec alpha M=-1E -4 delta k/k/F C.253
- Using
- T = B p / lambda x P C.253 p = B / lambda x T + 1 j p = 0.007 / (0.1)(60) +1 p = 0.007 / 7 = 0.001 delta k/k CO.53 i
(0.001dk/k) / (-1E -4dk/k/F) = 10 F temperature rise and a final temperature of 291 F CO.53 (1.5)
- b. Indicated reactor pressure = ~43.5 psig -(0.5)
- REFERENCE l- BFNP REACTIVITY COEFF. LP l REACTOR POWER &' REACTOR PERIOD LP j GGNS OP-NP-511 ANSWER 5.03 (1.00)
Some ' critical
- LPRM strings' Base distributions are out of tolerance (the i strir.1s' difference distribution exceeds about 15% of the average of the i strint) a CO.53 and a new 00-1 (or 00-2) should be run CO.53.
i REFERENCE
- GGNS MCD, THERMAL LIMITS,P.00 4
i l
1
--~. , e- ,.-..,,,_,---.n ~.n ,~.,.__,n vn n,. . , , . - . , , , , . , - . . . .-,,__...-.--_.;.-L-.-,-.__-...-.-.__..,.----..---..
- 5. - THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 37 ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J ANSWER 5.09 (1.50)
- a. It will increase to a maximum concentration in ~20 days [0.53 and return to the 100% equilibrium value in ~6 days CO.53. (1.0)
- b. They are equal. (0.5)
REFERENCE GGNS OP-NP-514,P.10-11 ANSWER 5.10 (1.50)'
- a. 1000 psia
- b. 800 F
- c. 255 F REFERENCE GGNS'OP-HF-503,P.22-24 ANSWER 5.11 (1.50)
- 1. Viscosity
- 2. Density
- 3. Velocity
.4. Diameter (0.5ea/1.5)
.REFEREhCE GGNS MCD, CORE COOLING MECHANICS,P.24 ANSWER 5.12 (1.00)
-90 seconds the longest-lived delayed neutron precursor (Br-97) (0.5ea/1.0)
y-
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 3" y g--------------------------------------
l ANSWERS -- GRAND GULF 1 -85/08/12-MUNR0rJ REFERENCE GCNS OP-NP-518eP.6 ANSWER 15.13 (1.00) b REFERENCE GCNS OP-PC-505,P.6-10 ANSWER 5.14 (1.00)
(1175 - 536.8) / 657.5 = 0.971 REFERENCE
-GGNS OP-HF-503rP.5 ANSWER 5.15 (1.50)
- a. Decrwase
- b. Decrease -
- c. Increase REFERENCE GGNS OP-HF-506,P.20-21 ANSWER 5.16 (1.00)
- 1. Neutron embrittlement of the cladding.
- 2. Thermally induced pellet growth. -
- 3. Inward motion of the cladding walls (creepdown). (200.5ea/1.0)
REFERENCE GGNS MCD, PCIOMR, P.7
l- i
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 39 4
ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J
- ANSWER 5.17 (1.00) b
. REFERENCE
ANSWER 5.18 (2.00) ,
e...Due to recire. pump trip on low vessel level 2.
b.,Due to the spike,in reactor power.
- c. Due to the vessel level recovery -OR-Due to the feed pump trip on high vessel level 8.
-d. Due to APRM high flux.
REFERENCE GGNS'FSAR FIGURE 15.4.-5 1
- ANSWER 5.19 (1.00)
- 1. It~(hydrogen)-has a high microscopic scattering cross section .
- 2. It (hydrogen) has a hi3h logarithmic energy decrement per collision.
REFERENCE
- GGNS OP-NP-502,P.9 ANSWER 5.20 (1.50) l'. Power level
- 2. time.at power
1 y .
- 5. THEORY OF NUCLEAR POWER.FLANT OPERATION, FLUIDS, AND PAGE 40 ANSWERS -- GRAND GULF 1 -85/08/12-MUNR0rd ANSWER 5.21 (1.00) d -
REFERENCE EIHe'L-RQ-605 (15)
GGNS, OP-NP-515,P.4-7 l
I l
L
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 41 ANSWERS -- GRAND GULF 1 -85/08/12-MUNR0rd ANSWER 6.01- (1.00)
D REFERENCE GGNS LP OP-C51-4-501 ANSWER 6.02 (1.00) b REFERENCE GGNS'LP OP-C51-3-501
. ANSWER 6.03 (2.00)
- 1. e ii. d REFERENCE BFNP::LP412rP.24; TRANSIENT 420;0I-57eP.53 EIH: L-RQ-726 GGNS-LP OP-C34-501 GGNS.SIM. MAL. 125 & 69 ANSWER 6.04 (1.00)
-d REFERENCE BFNP: L/P #3 EIH: GPNT, Vol. VIr Chapter 2.3-3, 5, Fig 2.3(3)
L-RO-712, pp'4r 5, 19
, GGNS LP OP-B21-501 GGHS LP OP-C34-501
- CAF*
l l
I l
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 42 ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J
- ANSWER 6.05 (1.00) d REFERENCE
- BFNP .BF-0I-68, pp 28, 29 EIH: EIH Simulator, Malfunction 436 GGNS SIM. MAL. 11 ANSWER 6.06 (1.00)
C REFERENCE BFNP: LPt7,P. 28 EIH: L-RQ-714, Figure 714-6; HNP-2-2447 GGNS SD 833-1 p. 5&6 GGNS LP OP-833-1-501 p. 5 GGNS ARI B33-FAL-L603A ANSWER 6.07 (1.00)
< 8 REFERENCE GGNS LP OP-E51-501 ANSWER 6.08 (1.00) b o
-ANSWER 6.09 (1.00)
- Hi DW Pressure C.33, 1.39 psis C.1]
- AND- C.23
- RCIC Stm Pressure Low C.33, 60 psis C.1] (1.0) ,
4 I
1 1
i L _,
- 6. PLANT-SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 43 ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J REFERENCE GGNS LP OP-E51-501, p.9 ANSWER 6.10 :(2.00)
- c. -Indication that the RGDS finds disagreement between the signals received from the 2 RACS
'd. Indication that the withdrawn rod must be fully inserted before any other control rod can be moved (0.5.ea)
REFERENCE GGNS'LP OP-C11-2-501 ANSWER 6.11 (1.00)
- a. 2 (0.5)
- b. To prevent an excessive change in the LHGR. -(0.5)
REFERENCE GGNS LP OP-C11-2-501
' ANSWER 6.12 (3.00)
- a. ' Group 1: -- CB-1 Fully Inserted
-- CB-2 Fully Inserted
-- CB-5 Open
-- CB-5 opposite loop Open
-- CB-2 Open Group 21 -- Pump Speed 20% - 26%
-- Pump Motor Voltage not <75V for 4 Sec.
-- LFMG at Rated Voltage (.25 ea)
- 6. Prevents valve cycling [0.53 when recire pump speed changes CO.5]
REFERENCE GGNS OP-N33-1
-6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 44 ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J ANSWER 6.13 (1.00) e REFERENCE GGNS LP OP-N32-2-501 ANSWER 6.14 (1.00) b REFERENCE
'BFNP'LP439,P.18 -
GGNS OP-C41-501,P.5,20 ANSWER 6.15 (1.00)
Flow is established from SSW loop B CO.53 to RHR loop B LPCI injection line [0.53.
REFERENCE GGNS OP-P41-501,P.14 SD-E12,P.40 ANSWER 6.16 (1.00) d REFERENCE GGNS OP-021-501,P.9 SD-D21,P.6 ANSWER 6.17 (1.00) d REFERENCE GGNS SD-P42,P.3,19 l
y - *.e_-- ,..y,___..-.--,-1,%mr;, -w, m c r y --- -s---*~v-----+-w---r-- w e v --w-- - - - - ' - - - = -- '+-ni-w----w- w- - , ~ - - - --- -+
- 6. PLANT SYSTEMS CONTROL, AND INSTRUMENTATION PAGE 45
__________________' DESIGN, ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J ANSWER 6.18 (1.50)
- c. The ECCS pump discharge pressure interlocks (0.5) b.-601 - the tail pipe pressure switch has picked up 631 - the (B) solenoid is energized (0.5ea/1.0)
REFERENCE GGNS OP-E22-2-501,P.7,9-10 ANSWER 6 19 (1.50)
- a. > 50 psig reactor pressure after a 15 minute time delay
- b. Rx level 1 / -150.3'
- c. High drywell pressure / 1.39 psig d.-No
- e. the initiation signal is reset (0.3ea/1.5)
REFERENCE GGNS OP-E21-501,P.7 ANSWER 6.20 ( .50)
=ero spm REFERENCE GGNS OP-E21-501,P.12,17 ANSWER 6.21 (1.00) d REFERENCE USNRC BWR-4 Systems Manuale pp 3.3 3.3-10 EIH: HNP-x-1001; HNP-x-1286 GGNS LP OP-N21-501 p.10 GGNS SIM. MAL. 121
i PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 46
' ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J ANSWER 6.22 (2.00)
- 1) APRM 'A' fails upscale -> relay K12A deenergizes (0.4)
- 2) -> NHS contacts K12A in RPS Trip Logic A open (0.4)
- 3) -> Relays K14A & E deenergize (0.4)
- 4) -> Contacts K14A & E open (0.4)
REFERENCE BFNP: L/P #28 EIH: L-RQ-720,-Fig 720-ta, -ibe -2a, -2b, -3a, -3b GPNT, Vol. VI, Chapter 9.3.1-2, 3, 4 GGNS: SD-C71,P.2-7 ANSWER 6.23 (1.00)
Load Reference Control automatically switches off during:
< 1. a load rejection below 12% power (power <12% for > 5 sec.), or
- 2. a load rejection above 35% power (0.5 ea)
REFERENCE GGNS LP OP-N32-2-501 4 3, ~C.,g. r,.. r, ,. S e h. -. la 2 6 - ** '? %
1 f
l l
l l
l t
L
s.
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 47
- ~~~~~~~~~~~~~~~~~~~~~~~~
~~~~R Ib5bl55iEit 25UTRUL ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J ANSWER 7.01 (1.00) c ,
REFERENCE ,
EIH HNP-2-1933, p 2 GGNS: ONEP-05-1-02-IV-1 ;
(
ANSWER 7.02 (1.50) 4-**rs.sla-swe- (e h'b l
- 1. Place the RPS (Div 1, 2, 3, a 4) CRD Discharge Volume HI Trip Bypass Switches in the BYPASS position.
- 2. Place the RPS (Div 1, 2, 3, a 4) Scram Reset Switches in the RESET position and verify tha) the scram resets.
- 3. Allow the HCU's to recharge, then drive the control rods ,
that are not full-in to posit' ion 00. (0.5 ea) 3 REFERENCE g GGNS: ONEP-05-1-02-I-1, p3 t
ANSWER 7.03 (1.00)
,9 Without air pressure the D/G shutdown features are inhibited.
t
' REFERENCE '
GCNSt O N E P-0 4 01 - P,7 F - 1, p3
- o
. : '. 1 ANSWER 7.04 -(1.00) .
Establish LPCS or LPCI flow .from the Suppression Pool with injection to the RPV (0.5) and open two (2) SRV's to establish return flow to the Suppression Pool. (0.5) .
REFERENCE -
GGNS: 05-S-01-EP-8, pp 1, 2 l
l 4
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 48
~~~~ ~
RA5i5t55i5AL C5 TR6L'~~~~~~~~~~~~~~~~~~~~~~~
ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J ANSWER 7.05 (1 50)
- 1. An-unexplained increase in reactor water conductivity (0.25) accompanied by a very low pH (0.25).
- 2. An unexplained increase in reactor water activity.
- 3. An unexplained decrease in reactor power (due to the increased surface tension of the water).
4.- An unexplained increasae in levels on the MSL RAD monitors'(and possibly the OG RAD monitors).
5.- An unexplained change in condensate or RWCU F/D parameters (particularly flow or d/P). (3 9 0.5.ea) l REFERENCE i GGNS! ONEP-05-1-02-I-5, p2 l
- ANSWER 7.06 (1.50)
- 1. Verification of system integrity by visual inspection of accessible areas and/or
- 2. Verification of system integrity by available indication (s) for inaccessible areas (PRM's, etc) and~
- 3. Verification that operation of the system will not result in uncontrolled release to the environment. (0. 5 ea)
REFERENCE GGNSI ONEP-05-1-02-III-5, p 3 i
I
)
~ ANSWER 7.07 (1.00) b I
k
,,,-,,.--...,,,,--,~,~,.,,-,n.-,-..,--.,.. . _--- ,v - --. . . - - , . - ,--,r-,,.,, --,.-.,,- ,~ ~ w ,e, ,a.--,.,,ms- - m.- - -~,- --w--
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 49
- ----~~-~~~--~~~~~-------
~~~~kd555L55i5dL 55 sis 5L ANSWERS -- GRAND GULF 1 -85/08/12-MUNR0rd REFERENCE GGNS: 05-5-01-EP-2r p6 ANSWER 7.08 (1.00)
To minimize the feecwater-to-reactor differential temperature (cnd feedwater no::7.e thermal stress).
REFERENCE GGNS: IOI-03 .1.-01-3, p2 3
ANSWER 7.09 -(2.00)
-1. Main Steam Line Pressure > 850 psig.
- 2. Main Steam Line Low Pressure Alarm cleared.
- 4. All APRM Downscale Alarms cleared.
- 5. ' Transfer one (1) IRM/APRM Recorder in each Division to APRM-and Place IRM recorders in Slow Speed (4 0 0.5 each) c, pts. sc.cet J tW t .ww <. ve r. n . a ,. % c. ,- w,t 6. 2. t ~1 ef P rouA., g REFERENCE GGNS:-I0I-03-1-01-1, pp 47, 48 ANSWER- 7.10 (1.50)
- 1. Work Permit (WP) 2.' Surveillance Permit (SP)
- 3. General Access Permit (GAP) (0.5 each)
REFERENCE GGNS: 01-S-08-2r pp 17, 18 L
- 7. PROCEDURES - NORMAL,' ABNORMAL, EMERGENCY AND PAGE 50
~~~~ ~ '~~~~~~~~~~~~~~~~~~~~~~~
RAU5ULUUsUdL UU TRUL ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J ,
ANSWER 7.11 (1.00) d REFERENCE GCNS: EP-5, Section 2.1 EP-3, Sections 3.1.5, 3.2.3, 3 3.6, 3.5.3 l
l l
ANSWER 7.12 (1.00)
Personal Dosimeter (Low Range) (0.25) TLD (0.25)
Alarming Dosimeter (or Portable Survey Instrument) (0.5)
( -OR- HP Surveillance )
REFERENCE GGNS: 01-S-08-2, pp 15, 16 ANSWER 7.13 (1.00) d REFERENCE GGNS: ONEP-05-1-II-1, p4 ANSWER 7.14 (1.50)
- a. Loop Manual
- b. 10%
- c. 5% (0.5 each)
REFERENCE GGNS: ONEP-05-1-02-III-3, p2 l
l
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 51
~~~~R d555L55iEAt 55UTR5t ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J ANSWER 7.15 (1.00)-
0 REFERENCE BFNP BF-0I-66, pp 5-7 EIH: HNP-1001, pp 19, 20 GGNS: IOI-03-1-01-1, p 33i.SOI-04-1-01-N33-1, p4 ANSWER 7.16 (1.00) 8 REFERENCE GGNS: ONEP-05-1-02-V-l, p1
. ANSWER 7.17_ (1.00) .
b REFERENCE EIH: HNP-2-1946 GGNS: ONEP-05-1-02-V-5, p2 ANSWER 7.18 (1.00) b REFERENCE
'GGNS: ONEP-05-1-02-VI-2 1
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 52
' '~~~~~~~~~~~~~~~~~~~~~~~
~~~~kd656 LEU 55dL UUU5kUL ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J ANSWER 7.19 (2.00)
- a. (1) 14.5 feet (2) 212 des F (3) 140 des F (0.5 each)
REFERENCE GGNS: EP-3, p 6; EP-5, p 2; EP-7, p 1
-ANSWER 7.20 (1.00) 4 85 *
- c. 65 MW (+0,
-5 MW) p,(, p (0.5)
EHC Lock-up preventing LDL increases i" a h - =
Pe r s m' s.2 .V h .L., <g REFERENCE GGNS: IOI-03-1-01-2, p7 ANSWER 7.21 ( .50)
- a. Gas Pressure (PI-R131) remains constant REFERENCE GGNS: 50I-04-1-01-C11-1, p9
. ANSWER 7.22 (1.00) a REFERENCE GGNS: ONEP-1-02-I-4, p3
, , .,e , , .. ,- _ - - . . - - _ .
- 8. ADMINISTRATIV
E. PROCEDURE
S, CONDITIONS, AND LIMITATIONS PAGE 53 ANSWERS -- GRAND GULF 1 -85/08/12-MUNR0rJ
[ ANSWER 8.01 (1.00)
- e. True
- b. False l REFERENCE
. GGNS Proc. 01-S-06-12 ANSWER 8.02 (1.00) c REFERENCE GGNS TS 3.2.3 ANSWER 8.03 (1.00)
-Operational Condition 2 -- 4 SRMs (.34)
-Operational Condition 3 -- 2.SRMs (.33)
-Operational Condition 4 -- 2 SRMs (.33)
REFERENCE' GGNS TS 3.3.7.6-ANSWER 8.04 (1.00) a REFERENCE GGNS TS 3.2.2
+
d T
I
, . , . . . - + - . - - . - . , . , - - - . . . , , - - - - - - - , , - , i -- - - - - - - . - - , , - ,
9 ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 54 ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J ANSWER. 8.05 (3.00)
- a. The procedure or MWO must' .
-Specify that SS be notified of removal and/or installation. (0.5)
- Specify requirement for independent verification on Safety Related Systems. (0.5)
- 6. 3 (1.0)
- c. 1 (1.0)
REFERENCE' GGNS Proc. 01-S-06-3 GGNS LP OP-AD-524 ANSWER 8.06 (1.50)
The following checks should be made:
- Breaker charging springs charged. (0.5)
- Charging motor disconnect switch on. (0.5)
- Control power on. (0.5)
REFERENCE GGNS Proc. 02-S-01-2 GGNS LP OP-AD-539 ANSWER 8.07 (1.00) e REFERENCE CGNS Proc. 02-S-01-2
-GGNS LP OP-AD-539 1
.8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 55 ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J ANSWER 8.08 (1.00)
The two provisoes are:
- The sequence of major evolutions is not changed. (0.5)
- The intent of the instruction is not changed. (0.5)
REFERENCE GGNS Proc. 02-S-01-2 ANSWER 8.09 (2.00) e .- (1) 'Q-Valve' , (0.5)
(2) Operations Superintendent (0.5)
- b. 2 (1.0)
REFERENCE GCNS Proc. 01-S-06-01 ANSWER 8.10 ( .50)
Refueling / Operational Condition 5 REFERENCE GGNS TSs Table 1.2 ANSWER 8.11 (1.00) a REFERENCE GGNS TS 3.8.1.1 i
i l~
- .- r - -
.- _=
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 56 ANSWERS -- GRAND CULF 1 -85/08/12-MUNRO,J ANSWER 8.12 (1.00) b REFERENCE GGNS TSs Definitions ANSWER 8.13 (1.50)
- 1. ONE-Rod-out.
2.- Refuel Platform position.
3.- Refuel Platform Main Hoist loaded. (0.5 ea.)
REFERENCE GGNS TS 3.9.1 ANSWER 8.14 (1.00) a.. True
- b. True REFERENCE
.GGNS Proc. 01-S-06-10 GGNS LP OP-AD-525 ANSWER 8.15 (1.00) d REFERENCE GCNS TSs 3.0.3 & 3.6.7.1
. . - . ~
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 57 ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J
' ANSWER 8.16 (2.00)
- n. 2
- b. 4
- c. 1
- d. 3 REFERENCE EIH: GET Handbook, pp 57, 58, 60, 61 HNP-x-4420, HNP-x-4520, HNP-x-4620, HNP-x-4720 GGNS Proc. 10-S-01-2,3,4 & 5 ANSWER 8.17 (1.00) a REFERENCE 10 CFR 55.31.e ANSWER 8.18 (1.00) d REFERENCE EIH: U2 TS, 3.3.1, 3.3.2 GGNS TSs 3.3.1 & 3.3.2 4
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 58 ANSWERS -- GRAND GULF 1 -85/08/12-MUNRO,J ANSWER 8.19 (2.00)
- 1. Threshold power is 14.0 Kw/ft.
- 2. Nominal rate is 0.11 Kw/ft/hr with a max. step of 0.2 Kw/ft and a max.
rate over any 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period of 0.12 Kw/ft/hr.
- 3. A 12-hour wait at the final power.
- 4. If power is lowered before ramp or soak completion then continue pre-conditioning from 1 Kw/ft below the previous level.
(0.5 ea)
REFERENCE GGNS MCD, PCIOMR, P.11-13 ANSWER 8.20 (1.50)
- all rods fully inserted except for the single control rod of highest reactivity worth ~which is assumed to be fully withdrawn
- cold (68 deg. F)
- xenon free (0.5 ea)
REFERENCE GGNS TSs Definitions l
l
o o mg s a Vo t
- 1/2 at 2 I
i = mc -
KE = 1/2 mv a = (Vf - 73 )/t A = AN A=Ae'* 3 PE = m9n yf = y, + at *
- e/t x = zn2/t1/2 = 0.693/t1/2 w=v P A= n0 2 t 1/2'ff
- U *1)('d 3 4 [(t1/2)
- IIb))
t.E = 931 am *
. m = V,yAo - T.x Q.= m.&h I"Iec Q = mCpat 6 = UA4 T I
- I n
e~"*
Pwe = Wfah I=I n 10-*/ M TVL = 1.3/u P = P 10 sur(t) HVL = -0.693/u P = P,e*/
SUR = 26.06/T SCR = S/(1 - Kgf)
CR, = S/(1 - Keffx)
SUR = 26e/t= + (a - p)T CRj (1 - K gfj) = CR2 (I ' "eff2)
T = ( **/s ) + [(8 - o V Io 3 M " I/II ~ Edf) = CR /CR, j 7 = g/(o - s)
M = (I - Keffo)/(I - Keff1)
T = (8 - e)/(Io) SDM = ( -Kdf)/Edf a = (Kgf-l)/Keff " #effM eff D *#
["0.1 A= seconds,)
o = [(t*/(T Kdf)] + [a df /(1 + T)] -
Idjj=Id P = (reV)/(3 x 1010) I jd) 2 ,2gd 2 22 I = cN 2 R/hr = (0.5 CE)/d (meters)
R/hr = 6 CE/d2 (f,,g) _
Water Parameters Miscellaneous Conversions 1 gal. = 8.345 lbm. 1 curie = 3.7 x 1010 aps 1 ga . = 3.78 liters ,1 kg = 2.21 lem 1 ft< = 7.48 gal 1 hp = 2.54 x 103 Stu/hr -
Density = 62.4 1 /ft3 1 mw = 3.41 x 100 5tu/hr Density = 1 gm/c lin = 2.54 cm Heat of vaporization = 970 Btu /lom 'F = 9/5'C + 32 Heat of fusion = 144 Stu/lbm .
- C = 5/9 (*F-32) 1 Atm = 14.7 psi = 29.9 in. Hg. 1 BTU = 778 ft-lbf 1 ft. H 2O = 0.4335 lbf/in.
e = 2.718 e " 9NT, , -
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- r '6 's or, 0.0s46 32.018 0.01602 3302.4 3302.4 0.00 1076.5 1075.5 0 2.1872 2.1872 0 1021.3 tages 3.10 35.023 0.01602 2945.5 2945 5 3.03 1073 8 1076.8 0 0061 2.1705 2.17M 3A3 1022.3 e.10 0.15 45.453 0.01602 2004.7 2004.7 13.50 1067.9 1081.4 0 0271 2.1140 2.1411 13.50 1025.7 S.15 0.20 63.160 0 01603 1526.3 1526.3 21.22 1063 5 1084.7 0 0422 2.0778 2.1160 2122 1028.3 a.20 0.30 64 484 0 01604 1039 7 1039.7 32.54 1057.1 1089 7 0 0641 2 016S 2.0809 32.54 1032 0 c.30 0.40 72.869 0.01606 792.0 792.1 40.92 1052.4 10933 0.0799 1.9762 2.0562 40.92 1034.7 0.40 0.5 79.586 0.01607 641.5 641.5 47.62 1048.6 1096.3 0.0925 1.9446 2.0370 47.62 1036.9 0.5 S.6 85.218 0.01609 540.0 640.1 53.25 1045.5 1098.7 0.1028 1.9186 2.0215 53.24 1038.7 0.6 0.7 90 09 0.01610 466.93 466 94 58 10 1042 7 11008 0.3 1.8966 2.0083 58.10 1040.3 0.7 02 94.38 0.01611 411.67 411.69 62.39 1040.3 1102.6 0.1117 1A775 1.9970 6229 1041.7 OA 0.9 98.24 0.01612 368.41 368.43 66.24 1038.1 1104.3 0 1264 1A606 1.9870 E.24 1042.9 as 1.0 101.74 0.01614 333.59 333 60 69.73 1036.1 1105 8 0.1326 13455 1.9781 69.73 1044.1 1A LO 126.07 0.01623 173.74 173.76 94.03 1022.1 1116.2 0.1750 1.7450 1.9200 94A3 10512 2A 3.0 141.47 0.01630 118 71 118.73 109.42 1013.2 1122 6 0.2009 14854 1A864 109.41 1056.7 SA 4.0 152.96 0.01636 90 63 90 64 120.92 1006.4 1127.3 02199 1.6428 1.8626 120.90 1060.2 4A 8.0 162.24 0.01641 73.515 73.53 130 20 1000.9 1131.1 0.2349 1.6094 13443 130.18 1063.1 EA 6.0 170.05 0.01645 61.967 61.98 138 03 996.2 1134.2 0.2474 1.5820 12294 13821 1065.4' E.0 7.0 176.84 0.01649 53.634 53.65 144.83 992.1 1136 9 0.2581 1.5587 1A168 14431 1067.4 7A 8.0 182.86 0.01653 47.328 47.35 150 67 988.5 1139.3 0 2676 1.5384 12060 15034 1069.2 SA 9.0 188.27 0.01656 42385 42.40 156.30 985.1 1141.4 0.2760 1.5204 1.7964 15628 1070.8 94 to 193.21 0.01659 38.404 38 42 161.26 982.1 1143 3 0.2836 1.5043 1.7879 161.23 1072.3 10 14.696 212.00- 0.01672 26.782 26.80 180.17 9703 1150.5 0.3121 1.4447 1.7568 180.12 1077.6 14A96 15 213.03 0.01673 26.274 26.29 181.21 969.7 1150.9 0.3137 1.4415 1.7552 181.16 1077.9 15 20 227.96 0.01683 20.070 20.087 196 27 960.1 1156.3 0.3358 1.3962 1.7320 196.21 10824 20 1 30 250.34 0.01701 13.7266 13.744 218.9 945.2 1164.1 0.3682 1.3313 1.6995 2183 1087.9 30 l 40 267.25 0.01715 10.4794 10.497 236.1 933.6 1169A 0.3921 1.2844 1.6765 2360 1092.1 40 80 281.02 0.01727 8.4967 8.514 250.2 923.9 1174.1 0.4112 1.2474 J.6585 250.1 10953 80 80 292.71 0.01738 7.1562 7.174 262.2 915.4 1177.6 0.4273 1.2167 1.6440 262.0 1098.0 80 70 302.93 0.01748 6.1875 6.205 272.7 907A 1180.6 0 4411 1.1905 1.6316 272.5 1100.2 70 80 312.04 0.01757 5 4536 5.471 232.1 900.9 1183.1 0.4534 1.1675 1.6208 281.9 1102.1 30 < ,
90 320.28 0.01766 4.8777 4.895 290.7 894.6 1185.3 0.4643 1.1470 1.6113 290.4 1103.7 90 '
100 327A2 0.01774 4.4133 4.431 238.5 888.6 1187.2 0.4743 1.1284 1.6027 298.2 1105.2 100 '
120 341.27 0.01789 3.7097 3.728 312.6 877.8 1190 4 04919 1.0960 1.5879 312.2 1107.6 120 140 353 04 0 01803 3.2010 3 219 325.0 568.0 1193 0 0.5071 1.0681 1.5752 324.5 1109.6 140 >
160 363 55 0.01815 2.8155 2234 336.1 859.0 1195.1 0.5205 1.0435 1.5641 335.5 1111.2 160 180 373 08 0.01827 2.5129 2.531 346.2 850.7 1196.9 05328 1.0715 1.5543 345.6 1112.5 130 200 351.80 0 01839 2.2689 2.287 355.5 842.8 1198.3 0.5438 1.0016 1.5454 3543 1133.7 300 1 1
250 400 97 0.01865 1.8245 1.8432 376.1 825 0 1201.1 0.5679 0 9585 1.5264 3753 1115A 250 300 417 35 0 01889 1.5233 1.5427 394.0 808.9 1202.9 0.5882 0 9223 1.5105 392.9 1117.2 300 350 431.73 0.01913 1.3064 1.3255 409.8 794 2 1204 0 0 60 % 0 8909 1.4968 4086 1118.1 350 400 444 60 0.0193 1.14162 1.1610 424.2 780 4 1204.6 0 6217 0.8630 1.4847 422.7 111E 7 400 450 456 28 0 0195 1.01224 1.0318 437.3 767.5 1204.8 06360 0.8378 1.4738 435.7 1118.9 450 500 46701 0 0199 0 90787 0 9276 449.5 755.1 1204.7 0.6490 0.814S 1.4639 447.7 11183 500 553 47694 0 0199 0 82183 0.8418 460.9 743.3 1204 3 06611 0.7936 1.4547 456.9 1118 6 550 600 48520 0 0201 0.74962 0.7698 471.7 732.0 1203 7 0.6723 0.7738 1.4461 469.5 1116.2 600 700 ,503 08 0 0205 063505 0 6556 491.6 710.2 1201.8 06928 07377 1.4304 488.9 1116.9 700 800 518 21 0 0209 054809 05690 509.8 689 6 1199 4 0.7111 0.7051 1.4163 506 7 1115.2 000 900 53! 93 0 0212 04796S 05009 526 7 669 7 1196 4 07279 06753 1.4032 523 2 1113 0 900 1000 544.5B O.0216 042435 0 4460 542.6 f 50 4 1192.9 07434 06476 1.3910 5306 1110 4 1000 1100 SSE 2d 0.0220 0.37B(3 0 4006 557.5 631.5 1189 1 07573 0.6216 1.3794 553.1 1107.5 1100 00223 0 34013 0.3625 571.9 613 0 1184 8 0.7714 05%9 1.3683 5569 1104 3 1200 1200 1300 l 667.19 577 42 0 0227 030722 0.3299 585.6 544.6 1180 2 0.7843 05733 1.3577 530.1 1100 9 1300 1400 527 07 00231 0.27811 0 3018 598 8 5765 1175 3 0.7966 05507 1.3474 592.9 1097.1 1400 1500 596 20 0 0235 025372 0.2772 611.7 558 4 11701 0.8035 0 *.233 1.3373 605 2 1093.1 1500 2000 635 80 0.0257 016706 0.1883 672.1 466.2 1133.3 0 86M 04256 1.7881 662 6 10GS 6 2000 2500 65d11 0 02c;f 010209 01307 731 7 361.6 1093 3 09139 0 3206 1.2345 118 5 1032 9 2500 3000 695 33 0 0343 0 050/3 0 0850 801 8 218 4 1020 3 09728 0.1891 1.1619 7822 973.1 3000 3298.2 70147 00508 0 0 050a 906 0 0 9060 10612 O 1.0612 875.9 875.9 37082 1 ____
TABLE A.3 PROPERTIES OF SATURATED STEAM AND SATURATED WATER (PRESSURE) l A.4
Volume, ft'/lb EMhelpy. SirAb EMeepy. Stellt a F To p water Evap Steam Water Evop C'eem Water Ever Ste:m [F
't % 's At A t
A s at s,e se 0.01602 3305 3305 -0.02 1075.5 1075.5 0.0000 2.1873 2.1873 32
. 82 0.08859 0.01602 2948 2948 3.00 1073.8 1076.8 0.0061 2.1706 2.1767 35 35 0.09991 1079.0 0.0162 2.1432 2.1594 40 40 0.12163 0 01602 2446 2446 8 03 1071.0 2037.7 2037.8 13.04 1068.1 1081.2 0 0262 2.1164 2.1426 45 45 0.14744 0 01602 0.01602 1704.8 1704.8 18.05 10653 1083.4 0.0361 2.0901 2.1262 50 80 0.17796 0.0535 0.01603 1207.6 1207.6 28.06 1059.7 1067.7 2.0391 2.0946 40 80 0.2561 0.01605 868.3 868.4 38.05 1054.0 1092.1 0.0745 1.9900 2.0645 70 TO 0.3629 0.0932 0.01607 6333 633.3 48.04 1048.4 1096.4 1.9426 2.0359 80 80 0.5068 01115 1A970 0.01610 468.1 468.1 58.02 1042.7 1100.8 2.0086 90 to 0.6981 0.01613 350.4 350.4 68.00 1037.1 1105.1 0.1295 1A530 1.9825 100 100 0.9492 0.01617 265.4 265.4 77.98 1031.4 1109.3 0.1472 1A105 1.9577 110 -
110 1.2750 0.01620 203.25 203.26 87.97 1025.6 1113.6 0.1646 1.7693 1.9339 120 180 1A927 0.01625 157.32 157.33 97.96 1019.8 1117A 0.1817 1.7295 1.9112 130 130 2.2230 0.01629 122.98 123.00 107.95 1014.0 1122.0 0.1985 1.6910 1.8895 140 140 2.8892 0.01634 97.05 97.07 117.95 1008.2 1126.1 0.2150 1.6536 1.8686 150 150 3.718 0.01640 77.27 77.29 127.96 1002.2 1130.2 0.2313 1.6174 1A487 160 160 4.741 i
0.01645 62.04 62.06 137.97 996.2 1134.2 0.2473 1.5822 1.8295 170 170 5.993 0.01651 50.21 50.22 148.00 990.2 1138.2 0.2631 1.5480 1A111 130 180 7.511 0.01657 40.94 40.96 158.04 984.1 1142.1 0.2787 1.5145 1.7934 100 190 9.340 0.01664 33.62 33.64 168.09 977.9 1146.0 0.2940 1.4824 1.7764 200 200 11.526 1.4509 1.7600 210 0.01671 27.80 27.82 178.15 971.6 1149.7 0.3091 210 14.123 0.01672 26.78 26.80 180.17 9703 1150.5 03121 1.4447 1.7568 212 212 14.696 0.01678 23.13 23.15 188.23 965.2 1153.4 0.3241 1.4201 1.7442 220 220 17.186 03388 1.3902 1.7290 230 230 20.779 0.01685 19.364 19381 198.33 958.7 1157.1 16.304 16.321 208.45 952.1 1160.6 03533 1.3609 1.7142 340 240 24.968 0.01693 0.01701 13.802 13.819 218.59 945.4 1164.0 0.3677 1.3323 1.7000 250 250 29A25 0.01709 11.745 11.762 228.76 938.6 1167.4 0.3819 1.3043 1.6862 260 260 35.427 1.6729 270 41.856 0.01718 10.042 10.060 238.95 931.7 1170.6 0.3960 1.2769 270 0.4098 1.2501 1.6599 280 49.200 0.01726 8.627 8.644 249.17 924.6 1173.8 280 0.4236 1.2238 1.6473 250 57.550 0.01736 7.443 7.460 259.4 917.4 1176.8 290 1.6351 300 0.01745 6.448 6.466 269.7 910.0 1179.7 0.4372 1.1979 300 67.005 5.609 5.626 280.0 902.5 1182.5 0.4506 1.1726 1.6232 310 310 77.67 0.01755 4.896 4.914 290.4 894.8 1185.2 0.4640 1.1477 1.6116 320 320 89.64 0.01766 3.770 3.788 311.3 878.8 1190.1 0.4902 1.0990 1.5892 340 340 117.99 0.01787 2.939 2.957 332.3 862.1 1194.4 0.5161 1.0517 1.5678 350 360 153.01 0.01811 2.317 2335 353.6 844.5 1198.0 0.5416 1.0057 1.5473 380 380 195.73 0.01836 1.8444 1.8630 375.1 825.9 1201.0 0.5667 0.9607 1.5274 400 400 247.26 0.01864 1.4808 1.4997 396.9 806.2 1203.1 0.5915 0.9165 1.5080 420 t 420 305.78 0.01894 1.2169 419.0 785.4 1204.4 0.6161 0.8729 1.4890 440 440 381.54 0.01926 1.1976 0.9942 441.5 763.2 1204.8 0.6405 0.8299 1.4704 450 440 466.9 0.0196 0.9746 0.0200 0.7972 0.8172 464.5 739.6 1204.1 0.6648 0.7871 1.4516 480 ;
. 450 566.2 0.6545 0.6749 487.9 714 3 1202.2 0.6890 0.7443 1.4333 500 500 680.9 0.0204 520 0.5386 0.55 % 512.0 687.0 1199.0 0.7133 0.7013 1.4146 520 812.5 0.0209 540 0.4437 0 4651 536.8 657.5 1194.3 0.7378 0.6577 1.3954 540 962.8 0.0215 560 0.3651 03871 562.4 625.3 1187.7 0.7625 0.6132 1.3757 5CO 1133.4 0.0221 1
0.3222 589.1 589.9 1179.0 0.7876 0.5673 1.3550 580 580 1326.2 0.0228 0.2994 617.1 550.6 1167.7 0.8134 0.5196 1.3330 8i00 600 1543.2 0.0236 0.2438 0.2675 0.2208 646.9 506.3 1153.2 0.8403 0.46S9 1.3092 620 620 1786.9 C.0247 0.1962 679.1 454.6 1133.7 0.8656 0.4134 1.2821 640 640 2059 9 0.0260 0.1543 0.1802 0.1443 714.9 392.1 1107.0 0.8995 0.3502 1.2498 660 660 2365.7 0 0277 0.1166 0.0808 0.1112 758.5 310.1 1068.5 0.9365 0.2720 1.2086 Geo 640 2708.6 0.0304 0.0386 0.0752 822.4' 172.7 995.2 0.9901 0.1490 1.1390 700 700 30943 0 0366 0.0508 0 0.0508 906.0 0 S06.0 1.0612 0 1.0612 705.5 705.5 3208.2 1
TABLE A.2 PROPERTIES OF SATURATED STEAM AND SATURATED WATER (TEMPERATURE)
A.3
)
t I NEUTRON FLUX I I
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Lesson Recirculaden System'- sa-2 Page 36 of 36
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-- y l'. -:t l W: l
.I i .
6.
' 3 . ,
m ' g< I g
i f't-' t___ J c's r i l:
i :<!
- s. W 7 . ;i >$ '
w ul3
- i -
.w i
e j_ O.O*
m w m o ~*v - *
- TECHNICAL SPECIFICATIONS 3.0.1 OPERABILITY 3.0.2 3.0.3 3.0.4 J 3.2.2 (Figures 3.2.3-1 & 3.2.3-2 Only) -
MCPR 3.3.1 Reactor Protection System Instrumentation 3.3.2 Isolation Actuation Instrumentation 3.5.1 ECCS - Operating 3.6.7.1 Containment Hydrogen Reconbiner System 3.8.1.1 A.C. Souisces - Operating I
o 4
. _ _ . - . , . - . , - - - - - --=A - - - - - - ' ' ' ' ' ' ' ~ ' "'
- 3. .&- .
- i,%
,p
. . sg
(. -.
C 3/4.0 APPLICABILITY
. . LIMITING CONDITION FOR OPERATION 3.0.1 Compliance with the Limiting Conditions for Operation contained in 'the
.' succeeding Specifications is required during the OPERATIONAL CONDITIONS or other
' .[ conditions specified therein; except that upon failure to meet the Limiting
,.:'$.* ; , Conditions for Operation, the associated ACTION requirements shall be met.
? ,#
3.0.2 Noncompliance with a Specification shall exist when the requirements of T- %. . the Limiting Condition for Operation and associated ACTION requirements are
.= not met within the specified time intervals. If the Limiting Condition for
- Operation is restored prior to expiration of the specified time intervals,
, ,. completion of the Action requirements is not required.
-:$ %!- ' 3.0.3 When a Limiting Condition for Operation is not met, except as provided 9
in the associated ACTION requirements, within one hour action shall be initiated
. ;)' $.' , to place the unit in an OPERATIONAL CONDITION in which the Specification does
..{S. r; not apply by placing it, as applicable, in:
.. 1. At least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,
- 2. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and
- 3. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time
((f..
limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual Speci-P .l., fications.
- j This specification is not applicable in OPERATIONAL CONDITION 4 or 5.
a .? 3.0.4 Entry into an OPERATIONAL CONDITION or other specified condition shall
- J not be made unless the conditions for the Limiting Condition for Operation are f,gy met without reliance on provisions contained in the ACTION requirements. This t;s. provision shall not prevent passage through or to OPERATIONAL CONDITIONS as 6 required to comply with ACTION requirements. Exceptions to these requirements are stated in the individual Specifications.
. \k coaun cVLF-UNIT 1-- 3/' 0-1
- *- .M .
I.7 l
i i a 1.6 i- 1 4 ..
y . . p., . . ,.
- ;t '
- . .~.~, .
we. .s4 ,
) ,.u...
-~. _. .
I."- ,
, g' I.t;'4
.. x.
4..... . T 2
1.4 -
- i
.l
- E c- ,
o i
- - 2 . i
__ .1.3 i u -. I 4.
.s*nh : a.--
- W -; -
\
i .
C:'
T
- T.-
r*fcl . .
12
. \ l
.? l l Roted MCPR Operating Limit
= 1.18 1.1 l
l f
. 1.0 0 20 40 60 80 10 0 120 1
Core Flow, % of Roted Core Flow I 1
MCPR g ,
('
l .
l FIGURE 3.2.3-1
- 1 1
P^"^ C'Jt.F-UNIT 1 -- - - - --3/4 2-5 --. - - . . -
9 4
'a . .
- 5. . .
S.,
1.7
.w.. . . . . s. . .
. l.6
.i e-
?- - * -
+
1.5 l'
N l
'. c. I cr 1.4 o- l i i j t
. y .
c- i .
i
. , , i l i ,
i i . I s I ._
' I I
{
i l
\
. t i
'<. <~ -
' $:' I.2 \
. N l.1 1.0 0 20 40 60 80 100 120
. Thermal Power, % of Rated Thermal Power MCPR P
s FIGURE 3.2.3-2 GRAND GULF-UNIT 1 3/4 2-6
..j-,-e-,, -
g 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protection system instrumentation channels shown in ' Table 3.3.1-1 shall be OPERABLE with' the REACTOR PROTECTION SYSTEM RESPONSE TIME as shown in Table 3.3.1-2.
APPLICABILITY:. As,shown in Table 3.3.1-1.
' .g {?
, - ACTION: -.1 -
- a. With the number of CPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip system, place the inoperable channel and/or that trip system in the tripped condition
- Oc. within one hour. The provisions of Specification 3.0.4 are not applicable.
. z.
.;.y b. With the number of OPERABLE channels less than required by the Minimum 3G ~ OPERABLE Channels per Trip System requirement for both trip systems, place at least one trio system ^^ in the tripped condition within one hour and take the ACTION w:uired by Tab,le 3.3.1-1.
SURVEILLANCE REQUIREMENTS
\( . 4. 3.1.1 Each reactor protection system instrumentation channel snall be demonstratea CPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL
- . TEST ana CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at
,Y 9 the frequencies shown in Table 4.3.1.1-1.
. {? 4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of l ;['
all channels shall be performed at least once per 18 months.
~:. -
4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip
~ ~i- functional unit shown in Table 3.3.1-2 shall be demonstrated to be within its t limit at least once per 18 months. Each test shall include at least one chan-nel per trip system such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip system.
- An inoperable channel need not be placed in the tripped condition where this would cause the Trip Function to occur. In these cases, the inoperable channel shall be restored to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or the ACTION required by Table 3.3.1-1 for that Trip Function shall be taken.
- The trip system need not be placed in the tripped condition if this would cause the Trip Function to occur. When a trip system can be placed in the tripped condition without causing the Trip Function to occur, place the
\
trip system with the most inoperable channels in the tripped condition; if both systems have the same number of inoperable channels, place either trip system in the tripped condition.
GRAND GULF-UNIT 1 . . _ _ .
_. , 3/.4_3-1 . _ . . . . _ __
- q . 3.:.. 4.t ? t -
f .
[ ,
I' f ; 7
, TABLE 3.3.1-1 ,
REACTOR PROTECTION SYSTEM INSTRUMENTATION ,', ,
l ;
c, APPLICABLE MINIMUM -
E OPERAT10flAL OPERABLE CHANNELS T FUNCTIONAL UNIT CONDlII0flS PER TRIP SYSTEM;(a) ACTION
- 1. Intermediate Range Monitors: I-
- a. Neutron Flux - liigh 2 3 1 3, 4 3 2 5(b) 3 3
- b. Inoperative 2 3 1 3, 4 3 2 5 3 3
- 2. Average Power Range Monitor (c);
- a. Neutron Flux - liigh, Setdown 2 3 1
. = 3 2 3(b) .
3 w
4 b. Flow Biased Simulated Thermal Power - High 1 3 4
- c. Neutron Flux - liigh 1 3 4
- d. Inoperative 1, 2 3 1 3 3 2 i 5 3 3
- 3. Reactor Vessel Steam Dome Pressure - High 1, 2(d) 2 1 l 4. Reactor Vessel Water Level - Low, i Level 3 1, 2 2 1 i
l S. Reactor Vessel Water Level-High, Level 8 1(g) 2 4 g 6. Main Steam Line Isolation Valve -
Closure II ') 4 4
- 7. Main Steam Line Radiation - High 1, 2 Id) 2 5 II)
- 8. Drywell Pressure - High 1, 2 2 1 n s a
e
., n e.,is . .- t . ; 4c. . , . s . , ...- .. ...,...$v,.
- .:1
's- .f e,[' t
- e. 41 - f i
( .ai : o .'
p.' p
. '4;r&G k l ', i.y.Q{}@:.l,.. p p 7.g.
o
.,3p,
[f ,
-1 . .
s ;.. .
n TABLE 3.3.1-1 (Cont'inued) .-
g, r,'2 ,
= , . r .
REACTOR PROTECTION SYSTEM INSTRUMENTATION .
e !i .
E APPLICABIE MINIMUM . I- ei
- E OPERATIONAL OPERABLE CNANNELS ~
5 FUNCTIONAL UNIT CONDil10NS PER TRIP SYSTEM (a) ACTION
-4
~
- 9. Scram Discharge Volume Water Level - High 1 2 1 S
I93 -
2 3
, I-
- 10. Turbine Stop Valve - Closure I Ih) 4 6
- 11. Turbine Control Valve Fast Closure, Valve Trip System Oil Pressure - Low 1(h) 2 6 a 12. Reactor Mode Switch Shutdown ..
D Position 1, 2 -
2 1
a 3, 4 2 7 0 5 2 3
- 13. Manual Scram 1, 2 2 1 1
I 3, 4 2 8 5 2 9 I
i i
i I
F
]
1
'l
,.L- INSTRUMENTATION
- . TABLE 3.3.1-1 (Continued) {
REACTOR PROTECTION SYSTEM INSTRUMENTATION
.. ACTION AC. TION 1 Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
ACTION 2 -
Verify all insertable control rods to be inserted in the core and 4
lock the reactor mode switch in the SHUTDOWN position within a . -
one hour.
5 ACTION 3 -
Suspend all operations involving CORE ALTERATIONS *, and insert all insertable control rods within one hour.
':h , ACTION 4. -
Be in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
.3
-}- ACTION 5 --~
Be in STARTUP with the main steam line isolation valves closed
- - within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4
- . ACTION 6 -
Initimea a reduction in THERMAL POWER within 15 minutes and reduce ^?3rbine first stage pressure to less than the automatic bypass setpoint within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
ACTION 7 -
Verify all insertable' control rods to be inserted witnin ,
one hour. '
. ACTION 8 -
Lock the reactor mode switch in the SHUTDOWN position within i one hour. -
5 j ACTION 9 -
Suspend all operations involving CORE ALTERATIONS *, and i insert all insertable control rods and lock the reactor mode switch in the SHUTDOWN positicn within one hour.
s
^Except novement of IRM, SRM or special movable detectors, or replacement of LPRM strings provided SRM instrumentation is OPERABLE per Specification 3.9.2.
/
~. $
\
GRAND GULF-UN1r 1 3/4 3-4
~
f . . ,. .
i
'[ _ii _ .' ~ ~ TABLE 3.3.1-1 (Continued)
\g REACTOR PROTECTION SYSTEM INSTRUMENTATION
\:
TABLE NOTATIONS (a) A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required surveillance without placing the trip system in the tripped condi-
- . tion provided at least one OPERABLE channel in the same trip system is monitoring that parameter.
' ~f-l ,
. , (b) The " shorting links" shall be removed from the RPS circuitry prior to and
- A n d y .. during the time any control rod is withdrawn
- per Specification 3.9.2 aj .
and shutdown margin demonstrations performed per Specification 3.10.3.
(c) An APRM channel is inoperable if there are less than 2 LPRM inputs per
.. level or less than 14 LPRM inputs to an APRM channel.
,. 5 (d) ~This function is not required to be OPERABLE when the reactor pressure r s.
vessel head is unbolted or removed per Specification 3.10.1.
y 5 i . ., ,J -
(e) This function shall be automatically bypassed when the reactor mode switch
- e.'-.u - -
.is not in the Runn osition. .
(f) This function is not. required to be OPERABLE when DRYWELL INTEGRITY is not required. .
e
( ..
(g) With any control rod withdrawn. Not applicable to control rods removed
~'(
per Specification 3.9.10.1 or 3.9.10.2.
g..
3 ;__. (h) This function shall be automatically bypassed when turbine first stage pressure is less than 30%** of the value of turbine first stage pressure
~.I @ in psia, at valves wide open (W0) steam flow, equivalent to THERMAL POWER 4, .. less than 40% of RATED THERMAL POWER.
.s, ,.
~Q - . .
i "Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.
- Initial setpoint. Final setpoint to be determined during startup test program.
Any required change to this setpoint shall be submitted to the Commission within 90 days of test completion, e' $
k '
GRAND GULF-UNIT 1 . - . . .
3/4 3-5
e E
.L.:
--, INSTRUMENTATION 1 A .' 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION i 3.3.2 (
The isolation actuation instrumentation channels shown in Table 3.3.2-1 !
$. shal.1 be OPERABLE with their trip setpoints set consistent with the values shown
~j. l in the Trip Setpoint column of Table 3.3.2-2 and with ISOLATION SYSTEM RESPONSE '
TIME as shown in Table 3.3.2-3.
C APPLICABILITY:
W- As shown in Table 3.3.2-1. : -
3 c; . ACTION: .E _f -
~
.- . f,
- a. 'Nthin isolation actuation instrumentation channel trip seipoint
....[$'.b'T
~
f e 'd. ~1ess conservative than the value shown in the Allowable Values column i
of Table 3.3.2-2. declare the channel inoperable until the channel 4
p g is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
~
3l k'. , b. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip system, a g -. --
place the inoperable channel (s) and/or that trip system in the tripped gg condition
- within one hour. The provisions of Specification 3.0.4 i:.,. - .-. .
are not appi: able.
..- - - i
- c. With the nucer of OPERABLE channels less than required by the Minimum
- OPERABLE Channels per Trip System requirement for both trip systems,
\,, 7 - place at least one trip system ** in the tripped condition within one hour and take the ACTION required by Table 3.3.2-1.
~
StiRVEILLANCE REQUIREMENTS
.; O M e.c. 4.3.2.1 Each isolation actuation instrumentation channel shall be demonstrated
$Mf.-- OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and t .:.3 '
CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the
%-}O.
' frequencies shown in Table 4.3.2.1-1.
! 7- 4.3.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of l , $ .'..,
all channels shall be performed at least once per 18 months.
4.3.2.3 The ISOLATION SYSTEM RESPONSE TIME of each isolation trip function shown in Table 3.3:2-3 shall be demonstrated to be within its limit at least once per 18 months. Each test shall include at least one channel per trip system such that all channels are tested at least once every N times 18 months, where N is the total number of redundant channels in a specific isolation trip system.
- An inoperable channel need not be placed in the tripped condition where this would cause the Trip Function to' occur. In thsse cases, ths' inoperable" channel shall be restored to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or the ACTION required by Table 3.3.2-1 for that Trip Function shall be taken.
- The trip system need not be placed in the tripped condition if this would cause the Trip Function to occur. When a trip system can be placed in the tripped condition without causing the Trip Function to occur, place the trip system with the most inoperable channels in the tripped condition; if both b (- systems have the same number of inoperable channels, place either trip system in the tripped condition.
GRAND GULF-UNIT 1 3/4 3-9 .
as . -JW ; ,.. s. . i s Tr i l' ? '* ^ ? 4 'WW >
( .. 3 4. . .b' htqO&$.i,t?]'I .'
(; ,- , f ; "u TEk .
t .
- t+
e .-tpf _-
m-'d - .
- . ,t- / . # -
h.
yl m 1 .
N z
TABLE 3.3.2-1 !l,l c
1501 AT10N AC hilAll0N INSIRll!!! NI ATION .
c Q
- VALVEGROUPS MINIMull APPLICAulE _
1 ~*
TRIP FUNCTION OPERATED BY OP[RA0![ CllANNELS OPERATIONAL .,
f:. '
.E.
- SIGNAL (a} PER TRIP SYSilM (b) CONDITION I ACTION 1
-4 ' 1.
P..lMARY CONTAINMENT ISOLATION
- a. Reactor Vessel Water Level- .
- Low Low, level 2 6A, 7, 8, 10 I 'II'II 2 1, 2, 3 and # ( 20
- b. Reactor Vessel Water Level- g Low Low Level 2 (ECCS - -
Division 3) 6B 4 1, 2, 3 and # 29 ,
- c. Reactor Vessel Water Level-Low Low Low, Level 1 (ECCS -
Division 1 and Division 2) SI "II') 2 1, 2, 3 and # 29
- d. Drywell Pressure - High 6A, 7(c)(d) 2 1, 2, 3 20 R
- e. Drywell Pressure-High -
(ECCS - Division 1 and w Division 2) SI "II") 2 1,2,3 29 g f. Drywell Pressure-High (ECCS - Division 3) 6B 4 1, 2, 3 29
- g. Containment and Drywell Ventilation Exhaust Radiation - High High 7 2I ') 1, 2, 3 and
- 21
- h. Manual Initiation 6A, 7, 8, 10(c)(d) 2 1, 2, 3 and *# 22
- 2. f1AIN STEAM LINE ISOLATION I
. a. Reactor Vessel Water level-Low Low tow, Level 1 I 2 1,2,3 20
- b. Main Steam tine Radiation - Nigh 1, 10 IO 2 1,2,3 23
- c. Main Steam Iine Pressure - Low I 2 1 24
- d. Main Steam Iine Flow - ifi 9h I 8 1,2,3 23
- e. Condenser vacuum - low 1 2 1, 2,'" 3** 23
,sp % .c 1 . ,. .. : %=t W.34 l iM . . . .
As ti .itnaap
'r.' y .- i d, t
, [.;
b I .' } '
.- .3 l
. ;1 O .. . ,
l g TABLE 3.3.2-1 (Continued) ,.
z
- l a ISOL ATION ACillATION INSTHt!MI.N.TATION E VALVE GR0llPS HINIHilH APPLICABLE ,
7 OPERATED BY OPERADIE CilANNELS OPERAIIONAL' E TRIP FUNCTION CONDill0N ACTION SIGNAL (a} PER IRIP SY5itH (b) ,
M' 2. MAIN STEAM LINE IS0lATION (Continued) I
- f. Main Steam Line lunnel f Temperature - liigh 1 2 1,2,3 $' 23'
- g. Main Stecm Line Tunnel ';
a Temp.- ilich 1 2 '
1, 2, 3 i 23
- h. Manual Initiation 1, 10 2 1,2,3 22
- 3. SECONDARY CONTAINMENT ISCLATION -
a Reactor Vessel Water Level-Low Low, Level 2 N.A.(c)(d)(h) 2 1, 2, 3, and'# 25 w b. Drywell Pressure - High N.A.(c)(d)(h) 2 1,2,3 ( 25 D c. F'.el Handling Area N.A.III . 2 1, 2, 3, anf
- 25 w Ventilation Exhaust 4 Radiation - High High ,
- d. Fuel Handling Area '.*;
Pool Sweep Exhaust Ji Radiation - liigh High N.A.ISI 2 1, 2, 3, and *! 25
- e. Manual Initiation 2 1,2,3 6
- 4. REACTOR WATER CLEANUP SYSTEM ISOLATION
- a. A Flow - liigh 8 1 1,2,3 -
.; 27 A Flow Timer
- b. 8 1 1,2,3 } 27
- c. Equipment Area Temperature - 8 1/ room 1, 2, 3 [ 27 High ..
- d. Equipment Area a Temp. - 1 High 8 1/ room 1, 2, 3 27
- e. Reactor Vessel Water E Level - Low Low, level 2 8 2 1, 2, 3 4 2/
> \
P. : ;%td.'J48- '* UM
' d SDE F Wt ' d 4"
~
,, , i s.;..m,u 6 < . .
- / -
/ .
s.c. . .fra.'
' e
.J d.P.
v ..
. e;
{
4
'1 1 4 fI .
C1 8
. l y TABLE 3.3.2-1 (Continued) ;. ! -
2: . .. -
3 ISOLATION ACillAI! Oil IN51ROMEllTAT10N *j C -
VALVE GR0tlPS HitalMHH APPLICA0lE <
7 . OPERATED llY OPLRAlt!I CliANiill5 OPlRATIONAL
E TRIP FilNCTION SIGNAL (a) PIR 1 RIP SYSTEll (tj CONDH10N . ACTION q- :-
- 4. REACTOR WATER CIIANtlP SYSTEH ISOLATION (Continued) ;'
- f. Main Steam line funnel 8 1 1, 2, 3 27 Ambient Iceperature - liigh
- g. Main Steam line Tunnel a Y Temp. - liigh 8 I 1, 2, 3 J 2/
SLCS Initiation III
- h. 8 i 1, 2, 5## 30
- 1. Manual Initiation 8 / 1, 2, 3
.h 26
- 5. REACTOR CORE ISOLATION COOLING SYSIEH 1501Ai10fl '.'
w a. RCIC Steam Line Flow - liigh
- i N
- 1. Pressure 4 1 1, 2, 3 . 27 w 2. Time Delay 4 1 1, 2, 3 27 s .*
N b. RCIC Steam Supply g) s Pressure - low 4, 9 1 1, 2 J ,
21 4
- c. RCIC Turbine Exhaust
- Diaphragm Pressure - liigh 4 2 1, 2, 3 27 t
- d. RCIC Equipment Room Ambient >
lemperat ure - liigh 4 1 1, 2, 3 3 21
- e. RCIC Equipment Room a Temp. :
- liigh 4 l 1, 2, 3 ?? 27 i i
- f. Main Steam line innnel :
Ambient le mper ature - liigh 4 1 1, 2, 3
( 2/ .
9 Main Steam Line Tunnel i-a Temp. - liigh 4 1 1, 2, 3 t 2/
t
- h. liain Steam line Tunnel f.
Iemperature Timer 4 1 1, 2, I i' '
.I T,
.i O O O
- c. <. 3.ra4%t+: - + , e
.<A
,t ' s r .* *C 6eesNd
( '
l 9. f,b S'$' ' a f i
...\
> u. . . -
- j -}.[yi.-
L
- )*
'jl')_,',>*. ; Y
- g 't. A'," f
,sp. L O
p- , .r.?
- 9. f )j.,
. ,sl. d+ :
, 'S;tf<jjg.).
y q; ($.<3:,i
- i
.- Q 9jj. }g:'.;
f i r,
, _ f.
i
- yg- . . ..
- t z.w ,. 4,
- ,+
p l-
- c3 .! ?
g ' TA8tE .3.3.2-1 (Continued) . , Q.' , . :.
s me .
~
ISOLATION ACTUATION INSTRIIMENTATION O
r- VALVE GROUPS MINIMUM APPLICABLE T OPERATED BY OPERABLE CilANNELS OP[RAll0NAL TRIP FUNCTION SIGNAL (a) PER 1 RIP SYS1[M (bj CONDI110N- ACil0N
- 5.
g REACTOR CORE IS0_lATION COOLING SYSTEM 150LATION
- i. RHR Equipment Room Ambient [
Temperature - liigh 4 1/ room 1, 2, 3 -
27
- j. RHR Equipment Room a Temp. - e
.High 4 1/soom 1, 2, 3 27
- k. RHR/RCIC Steam Line Flow -
High 4 1 1, 2, 3 27
- 1. Manual Initiation 4(k) 1 1, 2, 3 -
26-w A m. Drywell Pressure-High 9("I , I 1, 2, 3 27 w (ECCS-Olvision 1 and '
4 Division 2) w
- 6. RHR SYSTEM ISOLATION -
- a. RHR Equipment Room Ambient .
Temperature - High 3 1/ room 1, 2, 3 i 28
- b. RHR Equipment Room a Temp. - liigh 3 1/ room I, 2, 3 , 28
- c. Reactor Vessel Water ,
level - Low, level 3 3 2 1, 2, 3 28
- d. Reactor Vessel (RHR Cut-in Permissive) Pressure - III High 3 2 1, 2, 3 28
- e. Drywell Pressure - High 35 '} 2 1, 2, 3 28 I
- f. Manual Initiation 3 2 1, 2, 3 26 5
l-.
G - - - _ - _ _ _ _ -
?
I
? INSTRUMENTATION r TABLE 3.3.2-1 (Continued)
ISOLATION ACTUATION INSTRUMENTATION
\s {\
ACTION ACTION 20 -
Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTOOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
ACTION 21 -
Close a.
the affected system isolation valve (s) within one hour or:-
In OPERATIONAL CONDITION 1, 2, or 3, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- b. In OPERATIONAL CONDITION *, suspend CORE ALTERATIONS, handling of irradiated fuel in the primary containment and operations with a potential for draining the reactor vessel.
a '
ACTION 22 -
Restore the manual initiation function to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT SHUTDOWN within the next 12 hou'rs and in COLD SHUTOOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
ACTION 23 -
Be in at least STARTUP with the associated isolation valves closed-within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
ACTION 24 -
Be in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
ACTION 25 - Establish SECONDARY CONTAINMENT INTEGRITY with the standby gas treatment system operating within one hour.
ACTION 26 -
Restore the manual initiation function to OPERABLE status within ..Jours or clo,se the affected system isolation valves within ; e next nour ana ceclare tne atrecteo system inoperable.
ACTION 27 - Close the affected system isolation valves within one hour and declare the affectpd system inoperaDie.
( ACTION 28 - Within one hour lock the affected system isolation valves closed,
, .or verify, by remote indication, that the valve is closed and [
\
electrically disarmed, or isolate the penetration (s) and declare the affected system inoperable.
qg -
ACTION 29 -
Close the affected system isolation valves within one hour and y y. , declare the affected system or component inoperable or:
a - a.
7t In OPERATIONAL CONDITION 1, 2 or 3 be in at least HOT SHUTDOWN
?' within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
_ b. In OPERATIONAL CONDITION # suspend CORE ALTERATIONS and opera-tions with a potential for draining the reactor vessel.
ACTION 30 - Declare the affected SLCS pump inoperable.
NOTES '
When handling irradiated fuel in tne primary or secondary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.
The low condenser vacuum MSIV closure may be manually bypassed during reactor l SHUTDOWN or for reactor STARTUP when condenser vacuum is below the trip set- i point to allow openirig of the MSIVs. The manual bypass shall be removed when '
condenser vacuum exceeds the trip setpoint.
. # During CORE ALTERATIONS and operations with a potential for draining the reactor vessel.
- With any control rod withdrawn. Not applicable to control rods removed per Specificatior. 3.9.10.1 or 3.9.10.2.
(a) See Specification 3.6.4, Table 3.6.4-1 for valves in each valve group.
GRAND GULF-UNIT 1 3/4 3-14 . _ _ _ . . . . -
5 -
i
! INSTRUMENTATION
( - TABLE 3.3.2-1 (Continued)
ISOLATION ACTUATION INSTRUMENTATION .
.\
NOTES (Continued)
(b) A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for
.._ required surveillance without placing the trip system in the tripped con-
' dition provided at least one other OPERABLE channel in the same trip system is monitoring that parameter.
(c) Also actuates the standby gas treatment system.
- (d) Also actuates the control room emergency filtration system in the isolation
- 4. - -
mode of operation. ' '
o
{fh2' Q)'"Twoupscale-HiHi,'m'neupscale-HiHiandonedownscale,ortwodownscale
sig'nals'froiIth'e sa e' trip sys't'e'm 'act'uate the trip system and initiate J
- isolation of the associated containment and drywell isolation valves.
, (f) Also trips and isolates the mechanical vacuum pumps.
(g) Deleted.
W- (h) Also actuates secondary containment ventilation isolation dampers and
- (13..
t valves per Table 3.6.6.2-1.
3 Qi (.il.. Closes only RWCU system isolation valves G33-F001, G33-F004, and G33-F251.
'in. -
(j) Actuates the Standby Gas Treatment System and isolates Auxiliary Building
- 2"- penetration of the ventilation systems within the Auxiliary Building,
- , j- ,,
(b) Closes only RCIC outboard valves. A concurrent,RCIC initiation signal is
. -required for isolaw on to occur (
(1) Valves E12-F037A ana E12-F037B are closed by high drywell pressure. All other Group 3 valves are closed by high reactor pressure.
(m) Valve Group 9 requires concurrent drywell high pressure and RCIC Steam
( Supply Pressure-Low signals to isolate.
(n) Valves E12-F042A and E12-F042B are closed by Containment Spray System
- t. initiation signals.
T -
(o) Also isolates valves E61-F009, E61-F010, E61-F056, and E61-F057 from Valve
- h. ._ r _ r- Group 7.
.a t;?-
K. *
.s
.' 4;;
(.
GRAND GULF-UNIT 1 3/4 3-15 - .. ~.-- =
. 3/4.5 EMERGENCY CORE COOLING SYSTEMS
3/4.5.1 ECCS - OPERATING LIMITING CONDITION FOR OPERATION 3.5.1 ECCS divisions 1, 2 and 3 shall be OPERABLE with:
- a. ECCS division 1 consisting of:
- 1. The OPERA 2LE low pressure core spray (LPCS) system with a flow path capable of_taking suction from the suppression pool and
, transferring the water through the spray sparger to the reactor vessel. . .
- 2. The OPERABLE low pressure coolant injection (LPCI) subsystem
" A" of the RHR system with a flow path caDable of taking suction frca the suopression pool and transferring the water to the reactor-vessel.
- b. ECCS division 2 consisting of:
- 1. T"e CDERABLE low pressure coolant injection (LPCI) subsystems s ' -a "C" of the RK8 system. each witn a f !cw path capacie of
. . :t's, t rem tne suepression ecol ana trans rerrina tne
. iter to tre reactor vessel.
e . E'; t :;ERABLE ADS valtes.
( :. E::: ;
-iva ; ::nsist:ng of the OPEF.ABLE high pressure core spray
' 2C51 ;; stem ita a flow path ca;able of tacng stction f rom tne 3 ;;ress'On acci and transf erring the water through the spray sparger to tne reactor vessel.
APPLICABILITY: CFERATICNAL CONDITION 1, 2* # and 3*.
ACTION:
- a. For ECC5 division 1, provided that ECCS divisions 2 and 3 are OPERABLE:
- 1. With the LPCS system inoperable, restore the inoperable LPCS system to CPERABLE status within 7 days.
- 2. With LFCI subsystem "A" inoperable, restore the inoperable LPCI subsystem "A" to GPERABLE status within 7 days.
- 3. With the LPCS system inoperable and LPCI subsystem "A" inoperable, restore at least the inoperable LPCI subsystem "A" or the inoperable LPCS system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
4 Ot.erwise. te in it least HOT SHUICC'..N sithin the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in CCLD SHUTOO'nN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.**
^Ihe ADS is not required to be OPERABLE when reactor steam dome pressure is less than or equal to 135 psig.
- 5ee Special Test Exception 3.10.5.
e g **Whenever two or more RHR subsystems are inoperable, if unable to attain COLD
'( _ SHUTCOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.
GRAND GULF-UNIT 1 3/4 5-1 -
9 V
~. ~ EMERGENCY CORE COOLING SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
('-
.. ACTION: (Continued)
,.y . . . . > . . . . . a . --,. .- c.. . n .- t . - ;-- - '-- - - ' ~ " " ' ' ' ^ '~ ~1
- 1. With either LPCI subsystem "B" or "C" inoperable, restore the inoperable LPCI subsystem "B" or "C" to OPERABLE status within
- i. 7 days.
j h:. -
I 2. With both LPCI subsystems "B" and "C" inoperable, restore at least l t
the inocerable LPCI subsystem "B" or "C" to OPERABLE status l witnin 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. j I 1
} 3. Otherwise be in at least H0T SHUTOOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 7, .
and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ^.
I '
- c. For ECCS division 3 proviced tnat ECC5 divisions 1 and 2 and the
, , RC:C system .re OPERABLE: . .
. .. i . .- k di.ision 3 incceracie, restore tne inoperauie n.is::n to CFEUELE status wit,nin .4 aavs.
j
- 2. C ?e mise. :e in s. .cas*. : T .<::'..N w i tnin u.e ex :. --
ar c in COLD 5.--diGO'..N wi tnin .ne a i w.,my 24 not.rs.
(;
l l
- d. For ECC5 divisions 1 and 2, proviued tnat ECCS aivision 3 is s
- . OPERABLE
I i 1. With LPCI suosystem "A" and eitner LPCI suosystem "i" or 'C" inocerable, restore at least tne inoperaDie LFCI suosystem "A"
, or the incperable LPCI suosystem "2" cr "C" to CPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
t
- 2. With the LPCS system inoperable and either LPCI st.osys . ems "3" or "C" incperable, restore at least the in:cerable i:5 3,3 tem .
or the inoperable LPCI subsystem "B" cr "C" to CFE.'iABLE status i
within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
- 3. Otherwise, be in at least HOT SHUT 00WN within the %xt 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> j and in COLD SHUTOOWN within the following 24 hourc'
^Whenever two or more RhR subsystems are inoperable, if unable tc. ittain C:LD SHUTDOWN as required by this ACTION, maintain reactor coolant tcruerature as low as practical by use of alternate heat remaval methods.
8 GRAND GULF-UNIT 1 , ,
3/4 5-2 .
- g. . . _ . .
r' .
EMERGENCY CORE COOLING SYSTEMS A
LIMITING CONDITION FOR OPERATION (Continued)
ACTION: (Continued)
.. - ~
. " . .e, . F.o r ECCS d i v.i s i o.ns 1. .agd . 2.,,.p rovided. tha.t' E.CCS. di vi s i on' 3 i s ,
. OPERABLE and divisions 1 and 2 are otherwise OPERABLE:
- 1. With one of the above required ADS valves inoperable, restore
'.. the inoperable ADS valve to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce 3i. v
,, , reactor steam dome pressure to 5 135 psig within the next 24
?* -
hours.
- 2. With t'ao or more of the above required ADS valves inoperable, be in at least HOT SHUTOOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce reactor steam dome pressure to 5 135 psig within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
[,[ " f. With an ECCS discharge line " keep filled" pressure alarm instrumentation channel inoperable, perform Surveillance Requirement 4.5.1. 3.1 at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
6 hitn an E;C5 .eacer celta P instrumentation channel inoceraole, restore tr.e inoperacle channel to GPERABLE status with 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or cetermine ECCS heacer delta P locally at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; s otnerwise ceciare the associated ECC5 inoperable.
- h. In tre event an ECC5 systen is actuated ano injects aater into tre Reactor Cooiant System, a Special Report shall be preparea and sub-mitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the useage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.
- Whenever two or more RHR subsystems are inoperable, if unable to attain COLD SHUTCOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.
s GRAND GULF-UNIT 1 3/4 5-3
C '
c.
.a 9-
,' , CONTAINMENT SYSTEMS
\ 3/4.6.7 ATMOSPHERE CONTROL I(f CONTAINMENT HYDROGEN RECOMBINER SYSTEMS LIMITING CONDITION FOR OPERATION s......... 3;.627t1~^Two- independent tohtaitiment'hydf6tjerFretbrf.b'iner'l' y st'eh6l'shaITbs'" ' " " "
APPLICABILITY: . OPERATIONAL CONDITIONS 1 and 2. *
- j. ACTION: P.*, ', ' ' .
- l J ' ., ' , . _' ,
With one containment hydrogen recombiner system inoperable, restore the inoperable system to OPERABLE status within 30 davs or be in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
2 ;
.', SURVEILLANCE RE0VIREhENTS
- 3. ! 4.6.7.1 Each containment nyarogen recombiner system shall be demonstrated OPERABLE: m
- a. R :ea:t en E w e montns cy veritying curing a recomoiner system functionai test tnat the minimum heater sneath temperature increases to greater than or equal to 700 F within 90 minutes. Maintain >700 F
\ for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
- b. At least once per 18 months b'c.
7 - 1. Performing a CHANNEL CAllBRATION of all control room recombiner i.1 instrumentation and control circuits.
I 2. Verifying the integrity of all heater electrical circuits by performing a resistance to ground test within 30 minutes following
, the above required functional test. The resistance to ground for any heater phase shall be greater than or equal to 10,000 ohms.
Verifying duri'nc) a rec'omb'iner' system fuhdt'ional tesi. 'that the 3.
~
heater sheath temperature increases to greater than or equal to 1200 F within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and is maintained between 1150 F and 1300 F for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
- 4. Verifying through a visual examination that there is no evidence of abnormal conditions within the recombiner enclosure; i.e, loose wiring or structural connections, deposits of foreign
, materials, etc.
- c. [ DELETED]
L
(
GRAND GULF-UNIT 1 3/4 6-58
r -
J e
? 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES A.C. SOURCES - OPERATING LIMITING CONDITION FOR OPERATION s ,. ., . '3!8?L1 *As"ii'm'iididiti,' Eh' e ' f61'1Bilhg' I1"sieEtFi'ca}1bGeTs'o'uTc'e'ssT151ibf ^ ' ' '
- OPERABLE:
- a. Two physically independent circuits between the offsite transmission e twork and the onsite Class lE distribution system, and
=- b. Three separate and, independent diesel ge'nerators, each with:
- 1. Separate day fuel tanks containing a minimum of 220 gallons of fuel.
- 2. A separate fuel storage system containing a minimum of:
a) 48,000 gallons of fuel each for diesel generators 11 and 12, and b) 39,000 gallons of fuel for diesel generator 13.
- 3. A separate fuel trans,fer pump.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
ACTION:
k a. With either one offsite circuit or diesel generator 11 or 12 of the above reauired A.C. electrical power sources inoperaole, demonstrate the OPERABILITY of the remaining A.C. sources'by performing Surveil-
-' lance Requirements 4.8.1.1.1.a within one hour and 4.8.1.1.2.a.4,*
. for one diesel generator at a time, within two hours and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter; restore at least two offsite circuits and diesel generators 11 and 12 to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTOOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- b. With one offsite circuit and diesel generator 11 or 12 of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirements 4.8.1.1.1.a within one hour and 4.8.1.1.2.a.4,* for one diesel generator at a time, within two hours and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter; restore at least one of the inoperable A.C.
sources to OPERABLE status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or be in at least HOT SHUT-DOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTOOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Restore at least two offsite circuits and diesel genera-tors 11 and 12 to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from time of initial loss or be in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTOOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
" Specification 4.8.1.1.2.a.4 must be performed for diesel generator 13 only when the HPCS system is OPERABLE.
c o a u n niii r .U u 1I 1._. Vi.tk ... . . . _ . _ _ - . . _ -
m
?
~
ELECTRICAL POWER SYSTEMS
\ LIMITING CONDITION FOR OPERATION (Continued)
ACTION (Cgntinued) , _ .
, ~ c. ~
With ei.ther. diesel generator 11 or.12 of the above required A.C.
electrical power sources inoperable, in addition to ACTION a or b above, as applicable, verify within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> that all required systems,
, subsystems, trains, components and devices that depend on the remaining diesel generator'11 or 12 'as a source of emergency power otherwise, be in at .least HOT. SHUTDOWN within the are alsohou'rs'and OPERABLE;'li COLD SHUTDOWN within the following 24
~ ~ ~
next 1'2' i
- d. With two of the above required offsite circuits inoperable, demonstrate the OPERABILITY of three diesel generators by performing Surveill.ance Requirement 4.8.1.1.2.a.4*, for one diesel generator at a time, within two hours and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter, unless the diesel generators are already operating; restore at least one of the inoperable offsite circuits to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least H0T SHUTOOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
With.only one offsite circuit. restored to OPERABLE status. restore at least too offsite circuits to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> frcm time of initial loss or be in at least H0T SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ano in COLD SHUT,00WN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
\- /
- e. With diesel generators 11 and 12 of the above requireo A.C.
electrical power sources inoperable, demonstrate the OPERABILITY of (
the remaining A.C. sources by performing Surveillance Requirements
. 4.8.1.1.1.a within one hour and 4.8.1.1.2.a.4*, within two hours and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter; restore at least one of the in-
{F[*fd
- +.
.~._
operable diesel generators 11 and 12 to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />
' or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTOOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Restore both diesel genera-tors 11 and 12 to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from time of initial loss or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTOOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- f. With diesel generator 13 of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirements 4.8.1.1.1.a within one hour and 4.8.1.1.2.a.4, for one diesel generator at a time, within two hours and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereaf ter; restore the inoperable diesel generator 13 to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or declare the HPCS system inoperable and take the ACTION required by Specification 3.5.1.
" Specification 4.8.1.1.2.a.4 must be performed for diesel generator 13 only when the HPCS system is OPERABLE.
GRAND GULF-UNIT 1 1/4 A-2 . - -- - - - - - -
&