ML20214L814

Exam Rept 50-416/OL-86-02 on 860908-11.Exam Results:Four of Five Reactor Operator Candidates Passsed & All Six Senior Reactor Operator Candidates Passed.Master Copy of Exams Encl

ML20214L814
Person / Time
Site:	Grand Gulf
Issue date:	10/22/1986
From:	Brockman K, Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20214L754	List: AECM-86-0297, Forwards Comments Re Review of Reactor Operator & Senior Reactor Operator Exams Administered on 860908.Marked-up Tech Specs & Procedures Encl ML20214L751 ML20214L814
References
50-416-OL-86-02, 50-416-OL-86-2, NUDOCS 8612030194
Download: ML20214L814 (137)

Text

m ENCLOSURE 1 EXAMINATION REPORT 416/0L-86-02 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 and operating examinations were administered at Grand Gulf Nuclear Station near Port Gi son, Mississippi.

Chief Examiner: zw / e / /2o[u KpE.BrocRmad7 / Date Signed Approved by: _n

/0/4 [$

John F. Munro, Actin 7 Section Chief .Date' Signed Summary:

Examinations on September 8-11, 1986 Oral examinations were administered to 11 candidates, all of whom passed; simulator examinations were administered to 11 candidates, all of whom passed; written examinations were administered to 11 candidates, 10 of whom passed.

Based on the results described above, 4 of 5 R0s passed and 6 of 6 SR0s passed.

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PDR ADOCK 05000416 L V PDR

REPORT DETAILS

1. Facility Employees Contacted:

J. Cross, Site Director R. Hutchinson, General Manager K. Beatty, Training Superintendent M. Shelley, Operations Training Supervisor

2. Examiners:

• K. E. Brockman, RII M. Spencer, EG&G J. Sherman, EG&G R. Miller, Sonalyst

• Chief Examiner

3. Examination Review Meeting At the conclusion of the written examinations, the examiners provided M. Shelly with a copy of the written examination and answer key for review.

The comments made by the facility reviewers are included as Enclosure 3 to this report, and the NRC Resolutions to these comments are listed below.

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a. SRO Exam 4

(1) Question 5.02 NRC Resolution: Comment not accepted. Cosmic ray interactions are negligible compared to other neutron sources, as per GE, BWR Reactor Theory, Chapter 2. No change to exam / answer key.

i (2) Question 5.03 NRC Resolution: Facility comment is acknowledged. Answer key has been changed to the following:

K SDM = (1 Keff)/ eff K

0.3 = (1 Keff)/ eff 0.3 Keff = 1 Keff K

1.3 eff - 1

,d .

2 K

eff = 1 = 0.77 [0.75]

1.3 CR2 = CR3 (1 Keff)1 (1 Keff)2

= 30 (1-0.77)

(1-0.98)

= 345 cps [0.75]

The equation [Keff = 1-SDM] is only accurate when eff is very nearly one. Therefore, it is not appropriate for this problem, and no credit is given for its use.

(3) Question 5.11 NRC Resolution: Comment acknowledged. No change required to this examination. The utility is encouraged to _ update the materials they utilize to be consistent with this comment. The confusion which may occur during training due to using non-specific training aids, is potentially more damaging than that which would occur on the licensing examination.

(4) Question 5.15 NRC Resolution: Comment accepted. Answer key is changed to show "E" as the correct answer for part 4.

(5) Question 5.20 NRC Resolution: Comment accepted. +,..swer key is changed to allow

+/-5% on part a, +/-10% on parts b and c, for full credit.

Correct answers are: a. 1333, b. 444, and c. 44.4. Note should be made that the methodology for calculations is worth 0.4 points, and the math worth 0.1 points.

(6) Question 6.01 NRC Resolution: Comment accepted. The startup level control valve is physically positioned in series (not parallel) with the

1. 5 and #6 feedwater heaters. Answer key is changed to show "B" as the only correct choice.

(7) Question 6.07 NRC Resolution: Comment accepted. Updated material justifies answer key being changed to show APRM downscale setpoint of 4%,

flow signal upscale high at 108%. Point values are changed such that each part is worth 0.20 points. Total point value unchanged.

.3 (8) Question 6.12 NRC Resolution: Comment accepted. Answer key is changed to (b) Minimum Flow Valve [0.25] @ 1250 gpm [0.25]. This question will be reworded in the future to avoid misunderstanding.

(9) Question 6.13 NRC Resolution: Comment accepted. Add " Diesel Generator Output Breaker in lower or Control Power Failure" to answer key as another correct answer.

(10) Question 6.15 NRC Resolution: Comment accepted. Part "B" is deleted from exam along with 0.5 point value. Question, section and total point values adjusted accordingly.

(11) Question 7.11 NRC Resolution: Comment is not valid because, per Examination Standards ES-402, part 7 of the SRO written examination shall "contain questions regarding fuel, fuel handling, and core loading, including procedures and limitations concerning core loading and alteration, fuel transfer and storage, and detection and prevention of criticality". Questions concerning fuel handling and refue'ing operations will continue to be valid examination topics. Answer key remains unchanged.

(12) Question 8.01 NRC Resolution: Comment acknowledged. No change to this examination required.

-(13) Question 8.13 NRC Resolution: Comment accepted. The answer key is changed to the list of 14 items requiring one hour notification shown in 01-S-06-5, Attachment III. Point value remains unchanged.

b. R0 Exam (1) Question 1.09 NRC Resolution: Comment not accepted. The Xenon peak occurs approximately 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> following a scram from 100% power. In addition, the burnout of Xenon when the reactor is at low power is insignificant in this case because of the large amount of Xenon contributed by iodine decay shortly after the scram.

4 (2) Question 2.02 NRC Resolution: Comment accepted. Answer key modified to reflect TBCW Makeup Valve will f ail closed (FC).

(3) Question 2.05 NRC Resolution: Comment accepted. Answer key niodified _to correct typographical error - 3.0 psig changed to 3.0 psid.

(4) Question 2.08 NRC Resolution: Comment partially accepted. Answer key to Part A was modified to include " generator differential current" and

" generator ground overcurrent, or generator lockout". Answer key to Part B was not modified since Technical Specifications are not specific as to when these trips are. active.

(5) Question 2.09 NRC Resolution: Comment accepted. Answer key modified to allow alternate wording of correct response.

(6) Question 2.10 NRC Resolution: Comment accepted. Answer _ key modified to accept

'yes'; a method of initiating the HPCS SSW pump from the Control Room must be given.

(7) Question 2.16 NRC Resolution: Comment accepted. Part B was deleted since the only identification methodology used would require valve numbering memorization.

(8) Question 3.03 NRC Resolution: Comment accepted. Answer key modified. The Rod Pattern Controller will provide additional rod movement restric-tions beyond those originally addressed.

(9) Question 3.06 NRC Resolution: Comment accepted. Answer key modified to add the CRD system as an acceptable response.

5 (10) Question 3.09 NRC Resolution: Comment not accepted. As a minimum, the operator should be aware of what is the cause of a " Channel Disagree" lamp in order to diagnose whether he has taken an improper action or the system has failed to perform properly. However, the answer key was modified to include pressing of the Display lamp test pushbutton and to accept the definition of a Channel Disagree per SOI-04-1-01-C11-2, pg. 16.

(11) Question 3.10 NRC Resolution: Comment partially accepted. Maintaining the Reactor Vessel Temperature above 70 degrees Fahrenheit is of the utmost importance in ensuring the Reactor Vessel integrity.

However, the answer key for Part B was modified to also accept an action to prevent further cooldewn as CAUTIONED in the immediate operator actions of ARI 04-1-02-H13-P680-3A-D2.

(12) Question 3.13 NRC Resolution: Comment accepted. Although the answer key was not modified to include all the possible answers on various ranges, any answer on other than range 8 was examined for its validity. (Note that 4% reactor power cannot be read below range 8 of the IRMs; refer to Lesson Plan OP-C51-2-501 Figure 2.)

(13) Question 4.03 NRC Resolution: Comment acknowledged. No modification to present examination required.

(14) Question A. 08 NRC Resolution: Comment accepted. Answer key mcdified to allow additional responses, as per the wording of the plant procedure.

4. Exit Meeting At the conclusion of the site visit the examiners met with representatives of the plant staff to discuss the examination.

There were no generic weaknesses noted during the oral examination.

The cooperation given to the examiners and the effort to ensure an atmosp-here in the control room conducive to oral examinations was also noted and appreciated.

The licensee did not identify as prcprietary any of the material provided to or reviewed by the examiners.

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U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: - GRAND GU.LF 1 REACTOR TYPE: BWR-GE6 DATE ADMINISTERED: _g6/09/08 EXAMINER: SPENCER. M.

CANDIDATE:

INSTRUCTIONS TO CANDIDATE:

Uae separate paper for the answers. Write answers on one side only.

Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.

% OF

% OF CANDIDATE'S CATEGORY CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 70.

25.00 23754 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS 27.X5 h6- }4 6. PLANT SYSTEMS DESIGN. CONTROL, AND INSTRUMENTATION

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25.50 24.'86 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 27.75 26.14 8. ADMINISTRATIVE PROCEDURES,

/ CONDITIONS, AND LIMITATIONS 9

10 Nm>~ Totals Final Grade

.All work done on this examination is my own. I have neither given

-nor received aid. -

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)hf Candidate's SignatEre'

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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:

1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.

2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.

3. Use black ink or dark pencil oniz to facilitate legible reproductions.

4. Print your name in the blank provided on the cover sheet of the examination.

5. Fill in the date on the cover sheet of the examination (if necessary).

6. Use only the paper provided for answers.

7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.

8. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write only on one side of the paper, and write "Last Page" on the last answer sheet.

9. Number each answer as to category and number, for example, 1.4, 6.3.

10. Skip at least three lines between each answer.

11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.

12. Use abbreviations only if they are commonly used in facility literature.

13. The point value for each question is indicated in parentheses after the l question and can be used as a guide for the depth of answer required.

14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.

15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.

16. If parts of the examination are not clear as to intent, ask questions of the examiner only.

l 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.

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' - 18 '. When you complete your examination, you shall:

a; . Assemble your' examination as follows:

(1)... Exam questions on top.

(2) Exam aids ' figures, tables, etc.

.(3) ' Answer pages including figures which are part of.the answer.

-b. Turn in your copy of the examination and all pages used to answer the examination questions.

c. TurnLin all scrap paper and the balance of the paper that you did not use for answering.the questions.

d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.

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'5. tTHEORY OF-NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 2

- ,THERMQDYNAMICS s

' QUESTION' 5.01- (1.00)

Which of the following actions will INCREASE Grand Gulf'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.

i QUESTION 5.02 (1.00) 4 List TWO commonly occurring natural sources of neutrons in a BWR?

QUESTION 5.03 (1.50)

The reactor is shutdown with a shutdown margin of 30% Delta K, and an indication of 30 cps on the SRM instrumentation. Control rods are withdrawn to a-Keff of 0.98. What is the new indicated SRM reading?

QUESTION 5.04 (1.00)

The heffchangesovercorelifebecause

a. As }( gfg decreases, Beta must also decrease.

b. There is an increased percentage of fissioning from Pu-239, which has a smaller delayed neutron fraction than U-238 and U-235.

c. As U-235 is fissioned and effectively used up, there are fewer fast fissions which result in fewer delayed neutrons.

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d. The microscopic cross sections for the isotopes producing delayed neutrons progressively become smaller with core age.

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(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)

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5. TSEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS. AND PAGE 3 7, THERMODYNAMICS QUESTION 5.05 (1.50)

How will moderator temperature coefficient of reactivity' change with the below plant parameter changes? (MORE NEGATIVE, LESS NEGATIVE or NO CHANGE)

a. Moderator temperature increase

b. Core void fraction increase

c. Fuel temperature increase QUESTION 5.06 (1.50)

How will the void coefficient of reactivity change with the below plant parameter changes? ( more negative, less negative, or no change )

a. Void fraction increases

b. Fuel temperature increase

c. Core age increase QUESTION 5.07 (1.00)

Choose the best word in parentheses to complete these statements.

a. The insertion of a deep rod will have the greatest affect upon (radial, axial) flux.

b. The withdrawal of a shallow rod will have the greatest affect upon (radial, axial) flux.

(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)

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5; THEORY! OF NUCLEAR' POWER PLANT OPERATION. FLUIDS, AND PAGE 4

,. . , THERMODYNAMICS.

QUESTION 5.08' (1.00)

How will control ~ rod' worth change with the below plant parameter changes?

(MORE NEGATIVE, LESS NEGATIVE, or NO CHANGE)

a. Moderator temperature increase

<-b. Fuel temperature increase

c. . Control rod density increase-d..~ Voids fraction increase QUESTION 5.09 (1.00)

State the TWO purposes of loading burnable poisons in a BWR core.

QUESTION 5.10 (1.00)

~The~following are common methods used to continue plant operation beyond-the end-of-cycle. Match them with the statement on the right best describing them.

>- =a. Derating 1. 'Results in decreased plant efficiency but allows

b. Coastdown maintenance of full turbine. load,

c. Feedwater temperature 2. Continued operation at a lower, reduction but constant power level.

d. Excess core flow 3. Reaching 100% power on a less than 100% flow control line.

4. Load is allowed to drop while operating uith all rods out.

(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)

~5. THEORY OF NUCLEAR POWER PLANT' OPERATION. FLUIDS.:AND PAGE 5

,. ,7, THERMODYNAMICS QUESTION 5.11 (2.00)

Using Figure One'at the-end of this exam, identify the belon listed lines on'the Figure by alphabetic letter.

1. Rod Block intercept line-

2. Pump-constant speed line

3. Minimum Pump speed line-(28%)

4. Natural' Circulation Line

-QUESTION 5.12 (1.50)

List THREE factors determining the type-(ie. CAPTURE or FISSION) of_ neutron interaction occurring in the core.

QUESTION 5.13 ( .50)

'Which of the following statements best describes the effects

resin beds have on conductivity.asithe beds become depleted.

a. . Increase slightly then decrease

b. Decrease slightly then increase

c. Increase drastically then stabilize

d. Decrease drastically then stabilize 2

QUESTION 5. lgj / .50) nab Hydrogenisen;w4_[rc/6inconcentrationsfrom 1c e to in air.

! Choose from below the best se:ection to complete the above.

a. 1.7% 91.3%

b. 4.1% 74.2%

c. 9.5% 57.8%

I d. 11.6% 77.3%

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-( ***** ' CATEGORY 05 CONTINUED ON NEXT PAGE ** *** )

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5. THEORY OF NUCLEAR POWER PLANT OPERATION,-FLUIDS. AND PAGE-- 6

, 3 , THERMODYNAMICS:

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-QUESTION 5.15' (1.00)

Using: Figure-Two at the end of this exam,-identify the below by

-olphabeticalfletter.

1. Partial film boiling 2: Log Q/A

-3. Log (Ts - Tsat)

4. CHF QUESTION 5.16 (1.50)

-List the effect on critical power that each of the below changes'would-produce. (INCREASE --DECREASE - REMAIN.'THE SAME)-

a. Increase subcooling-

b. Decrease mass flow-

c. Increase system pressure ( 1000 to 1050 psig ) ,

1 QUESTION 5.17 (1.00)

From the below list, choose which selection best describes total peaking factor,

a. TPF = RPF X APF/LPF

b. TPF = RPF + APF + LPF

c. TPF = RPF X LPF/APF

d. TPF = LPF X'APF X RPF

(***** CATEGORY 05 CONTINUED ON NEXT FAGE *****)

5. THEORY OF NUCLEAR POWER PLANT.0PERATION. FLUIDS. AND PAGE 7

,, THERMODYNAMICS QUESTION 5.18 (3.00)

For each of the following limiting parameters, state: (1) item measured, (2) limiting condition, (3) cause of failure, and (4) failure mechanism.

a. LHGR

b. APLHGR

c. CPR QUESTION 5.19 (1.00)

Briefly describe the affect core age increase has on moderator temperature coefficient.

QUESTION 5.20 (1.50)

Assuming an ideal fluid system with no losses, the speed of a centrifugal pump is decreased from 1800 rpm to 1200 rpm. The-information listed below is the 1800 rpm parameters.

1800 rpm parameters Flow = 2000 gpm Pressure = 1000 psig Power = 150 Hp What will the 1200 rpm values be for:

a. flow

b. pressure

c. Power i

M/STS0:?Y

(***** END OF CATEGORY 05 *****)

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6. PLANT SYSTEMS DESIGN. CONTPOL. AND INSTRUMENTATION -PAGE '8 QUESTION 6.01- (1.00)

The location Sf the Startup Level Control Valve-(C34-F513) is: (choose.

best selection 1from below)

a. in-parallel with both reactor feed water ptimps

b. in series with.FW Heater 5 & 6

c. in parallel with FW Heaters 5 & 6

d. in parallel with FW Heaters 5 & 6 and both feed water pumps QUESTION 6.02 (1.50)

a. What'is-the initiating event or action for the "Setpoint Setdown Mode? (0.5) i

b. The setdown setpoint (18 inches) remains for (select one of the below) (1.0)

' 1. 2 minutes

2. Until auto reset at 54" level

-3. Until manually reset

4. .10 sec QUESTION 6.03 (2.00)

List FOUR of the positive reactivity effects the Standby Liquid Control System is designed to overcome.

(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)

6. -PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION. _PAGE 19

. QUESTION' 6.04' .(1.00)

The relief-valves on the'SLC system discharge.to: (select one from the below list) a- between F001A(B) and the storage tank

b. directly to the storage tank

c. between the pump and-the motor operated pump suction valve

[F001A(B)]

d. to the inlet of the test tank

e. to the drain drums.

-QUESTION 6.05 (1.50)

. Describe how a fission chamber detects neutron flux. ( neutron to measured output-)

-QUESTION (1.00) 6.06 Why is U-234 used in the coating of the LPRM detectors (inaddition tv U-235)?

QUESTION 6.07 (2.00)

_ ~ Excluding INOP, list the initiating signals for an APRM rod withdrawal Block.~ Include applicable setpoints.

QUESTION 6.08 ( .50)

Where do the scram discharge volume vent and drain valves discharge to?

QUESTION. 6.09 (1.00)

List TWO affects removal of the " shorting links" will have on the RPS neutron monitoring system.

(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)

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6. PLANT SYSTEMS DESIGN. CONTROL, AND INSTRUMENTATION PAGE 10 QUESTION 6.10 (1.00)

List the requirements and setpoints for the RHR min. flow valves (F064) to AUTO close.

QUESTION 6.11 (1.50)

List the initiating events / requirements for auto initiated containment spray.

QUESTION 6.12 (2.00)

Answer the below questions about the Low Pressure Core Spray System.

a. What is the design flow rate?

b. What provides minimum flow and at what value?

c. The discharge relief valve (PSV-F018) discharges into what system?

d. List the auto initiation signals and setpoints.

QUESTION 6.13 (2.00)

List FOUR conditions capable of actuating the "HPCS NOT READY FOR AUTO START" annunciator.

QUESTION 6.14 (1.00)

Per the Grand Gulf Technical Specifications, state two reasons for maintaining the ECCS discharge piping full.

QUESTION 6.15 (1.50)

a. List which ADS initiation requirement (s) is/are NOT bypassed when ADS is manually initiated.

b. Which ADS valves are equipei with the " low-low" set function?

c. How many SRV's can be operated from the Remote Shutdown Panel?

(***** -

CATEGORY 06 CONTINUED ON NEXT PAGE *****)

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6. PLANT SYSTEMS DESIGN. CONTROL. AND' INSTRUMENTATION PAGE 11 QUESTION -6.16 (1.25)

List the MSIV's minimum and maximum operating time AND the bases for each.

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QUESTION. 6.17 -(2.00)

Matchethe type of fire protection systems in Column A with the correct description in. Column B.

COLUMN A' C6LUMN B

l. Automatic Wet Sprinkler a. -Piping is dry and system initiated

_ by solenoid valve by sensing heat-

2. Automatic Dry Pipe rate-of rise.

3. Automatic Preaction b. Piping filled w'ith pressurized water and which initiates if fused spray

4. , Automatic Deluge. nozzle reaches melting point.

c. Piping requires air pressure to keep deluge valve closed provided fused nouzles are sealed.

d. Piping requires a heat rate-of-rise detector signal and melted sprinkler head fusable link to complete flow path.

QUESTION 6.18 (2.00)

List the electrical buses each of the below diesel generators supply power to.

a. D/G #11

b. D/G-#12 -

c. D/G #21

'd. D/G #22

(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)

6. PLANT SYSTEMS DESIGN. CONTROL AND INSTRUMENTATION PAGE 12 QUESTION 6.19 (2.00)

List all power sources available to:

c. 4160V Bus 16AB

b. 4160V Bus 17AC MSE3 CWY

(***** END OF CATEGORY 06 *****)

7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 13

.,,BADIOLOGICAL CONTROL QUESTION 7.01 (1.00)

SOI-04-1-01-N64-1, "RWCU System", cautions that, whenever pos-sible, the RWCU system should be operated at MAXIMUM permissible temperature and flow rate to both Feedwater lines during LOW Feedwater Flow conditions. STATE the reason for this procedural caution.

QUESTION 7.02 (1.50)

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 3% on the APRM's. LIST the three immediate operator action steps required per ONEP-05-1-02-I-1, " Reactor Scram."

NOTE: LIMIT YOUR RESPONSE TO THOSE ACTION STEPS REQUIRED FOR REACTIVITY CONTROL.

QUESTION 7.03 (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 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 logic; RCIC will now restart automatically,

d. NONE OF THE ABOVE. RCIC cannot be restarted from the Remote Shutdown Panel after isolation.

(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)

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17. PROCEDURES NORMAL. ABNORMAL. EMERGENCY AND PAGE 14

' RADIOLOGICAL CONTROL

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QUESTION '7.04 (1.00)

-Per EP-2, " Emergency.Cooldown", which one of the following most <

occurately describes ~how SRV operation should be used to control pressure,'if needed?

NOTE: ASSUME THE INSTRUMENT AIR SYSTEM IS OPERATING PROPERLY

a. Use numerous'SRV's, with short pressure' reductions

( ~ 50 psig) to equalize Suppression Pool heatup.

b. Use fewer SRV blowdowns,. with increased pressure reduc-tions to minimize SRV cyclic stresses.

c. Depressurize with a sustained SRV opening to maximize the emergency cooldown rate.

( d. Allow-the SRV's.to operate by' mechanical actuation to L

ensure design pressure control and heat dispersion.

QUESTION 7.05 (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-57

a. A 30% reduction in Recirc Flow, monitored by Recirc Flow indication. ,

b. A 30% Power Reduction, using Recire Flow, monitored by APRM's.

c. Insertion of Shallow Rods, to maintain proper flux shape, prior to' reducing Recirc Flow.

d. Insertion of Power Rods, to maintain proper flux shape, prior to reducing Recire Flow.

QUESTION 7.06 (2.50)

LIST five (5) Entry Conditions for EP-3, " Containment Control".

(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)

.7.  : PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND. PAGE 15

,.., RADIOLOGICAL CONTROL-QUESTION <7.07" -(2.50)

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.

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. (1.0).

' QUESTION 7.08 (1.00)

.Upon recovering from a " Loss of Off Site Power", ONEP-1-02-I-4 cautions.the operator that either the SJAE.'s be isolated -OR-the condenser vacuum be broken PRIOR to re-energising MCC's 11B42, 12B42, and 14B22.

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 l charcoal adsorbers.

d. Prevent electrically tripping the cooling compressors in the Off Gas System.

QUESTION 7.09 (1.00)

' A SINGLE MSIV CLOSES and you. determine a high flow condition was reached in the other steam lines. Given that a Group I Isol-ation DID NOT OCCUR - STATE your Immediate Actions, l

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(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)

I

. i

7. -PROCEDURES' " NORMAL. ABNOEMAL. EMERGENCY A D ri '-

PAGE' 16 1,

,, RADIOLOGICAL-CONTROL- >

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,~

r QUESTION 7.10 .(1.50) '- .

Reguarding:" Emergency Exposure Guidelines":  :-

/ ,

a. List the whole body. dose limit to prevent immediat'r[ serious damage togplantgpersonnel. rf f d,

40 Ok -

' I b.

Li"st the whole body dose limit to save a life. - 4

~

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c. When is prior approval by GGNS General' Manager not required?

~ ,

QUESTION 7.11 (2.00) ,

Procedure 09-S-02-100 list several General Rules.for handling fuel. ~One of these rules is, ." Fuel assemblies shall be stored in such a' manner that water will drain freely from the assemblies in the event of, flooding and subsequent draining of the fuel storage area."

List-FOUR other General Criticality Rules.

1 ,

QUESTION 7.12 ( .50)

The ADS System must be declared INOP if the ADS Air System pressure isjat

.what value?

QUESTION 7.13 (1.00)

Listtheentryconditiontemperaturesto05-S-01 kepi 5 for'the.drywell and '

containment.

(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)

'7. PROCEDURES - NORMAL. ABNORMAL, EMERGENCY AND PAGE~ 17

,, RADIOLOGICAL CONTROL

-QUESTION 7.14 (2.00)

Match each of the terms below with its-definition.

TERM

1. Site Area Emergency

2. General Emergency

3. Unusual Event

4. Alert DEFINITION

a. The occurence of an event'or events which involve actual or likely major failures of plant functions needed for protection of the public.

There exists a SIGNIFICANT actual or potential release of radioactive material and some radiation exposure to the near-site public.

-b. The occurrence of an event or events which involve an actual or potential SUBSTANTIAL degradation of the level of safety of the plant.

c. Events are in process or have occurred which involve actual or, imminent-substantial-core degradation or melting with potential for lona of containment integrity and subsequent releases of large amounts of radioactive material Off-Site,

d. The occurrence of an event or events which indicate a POTENTIAL.

degradation of the level of safety of the plant. The situation may be one in which time is available to take precautionary and

-construc ive steps to prevent a more serious event or to mitigate any consequences that may occur.

QUESTION 7.15 -(2.00)

List EIGHT of the immediate Operator Action steps per the ONEP - Reactor Scram #05-1-02-I-1. Note - some action steps may contain more than one action.

QUESTION 7.16 (2.00)

List the FOUR Control Room Operator Immediate action steps required prior to-leaving the Control Room per 05-1-02-II-1.

(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)

7. PROCEDURES -~ NORMAL, ABNORMAL, EMERGENCY'AND' PAGE 18

,, RADIOLOGICAL CONTROL y QUESTION 7.17 (1.00)

-What is the reason for manually isolating the inservice Hydrogen Pressure Regulator following a turbine trip? '

QUESTION 7.18 (1.00)

List the operator actions required by a stuck open SRV.

/

VAS"R OPY

(***** END OF CATEGORY 07 *****)

c- 3

8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS - PAGE 19 QUESTION 8.01- (1.50)

List the.THREE LCO's associated with the two SRM channels during Refueling-Operations.

QUESTION '8.02 (2.75)-

E

' List the Technical Specification requirements for secondary containment integrity to exist. '

. QUESTION 8.03 (2.00)

The Technical Specification limits on idle loop startup are based to protect what items?

(i.e. idle. loop to vessel temperature 50 Deg)

(i.e. idle loop to operating loop temperature 50 Deg)

QUESTION 8.04 (1.00) 4 The APRM Trip Setpoint Formula is (.66W+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 =1MFLPD/FRTP ; T applied if > 1.0 i

t-9 r

(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)

L

~ 8. " ADMINISTRATIVE PROCEDURES. CONDITIONS, AND' LIMITATIONS PAGE 20, QUESTION 8.05 (1.00)

FILL ~IN'THE BLANK.with one of the following TS terms:

"A shall:be the injection of a simulated signal into the channel as close to the sensor as practicable 4

to verify OPERABILITY including alarm and/or trip

' functions and channel failure. trips."

a. Channel Calibration

-b. Channel Check- '

c. Channel Functional' Test 3 d. Logic System Functional Test i

t QUESTION 8.06 (2.00)

ADEQUATE CORE COOLING must be assured prior to. securing an ECCS system that has automatically initiated. LIST four plant conditions'(per the " Conduct of Operations" procedure) which will assure Adequate Core Cooling exists.

QUESTION 8.07 (1.00)

LIST the action (s) which the Shift Supervisor shall perform prior to the intentional removal of any safety-related systems or components from service. (Per 01-S-06-2)

QUESTION- 8.08 (1.50)

With the Mode Switch locked in the Refuel position:

i " CORE ALTERATIONS shall not be performed using equipment assoc-lated with a Refuel position interlock unless at least the full associated Refuel position interlocks are OPERABLE for such equipment."

LIST these three Refuel Position Interlocks.

(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)

m .x.

8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 21 QUESTION 8.09 (1.00)

Unit 1 Technical Specification 3.4.4 establishes the following conductivity and chloride limits:

PLANT CONDITION CONDUCTIVITY LIMIT CHLORIDE LIMIT 1 1 umho/cm 0.2 ppm 2 and 3 2 umho/cm 0.1 ppm Per the TS Basis, WHY is the chloride limit more restrictive at the lower steaming rate than when at power? (1.0)

QUESTION 8.10 (3.00)

Fill in the blanks regarding the Grand Gulf Safety Limits: (3.0)

THERMAL POWER shall not exceed (a) of RATED THERMAL POWER with the reactor vessel steam dome pressure less than (b) or core flow less than (c) of rated flow.

The Minimum Critical Power Ratio (MCPR) shall not be less than (d) with the reactor vesse l steam dome pressure greater than (b) and core flow greater than (c) of rated flow.

The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed (e) .

The reactor vessel water level shall be above (f) .

(6 @ 0.5 ea) l l

l

(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)

8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 22 QUESTION 8.11 (3.00)

Indicate whether the following statements are TRUE or FALSE. (3.0)

m. STA's are required to be licensed Senior Operators.

b. A Shift Supervisor may concurrently fill the position of the STA while on shift.

c. The Fire Brigade must include at least one of the following: SS, STA or COF.

d. All core alterations must be directly supervised by a licensed Senior Operator (or Senior Operator limited to fuel handling).

o. An Operator license is required for an operator to perform a core alteration.

f. During Operational Condition 4 or 5, an individual with a valid Operator license may be designated to assume the control room command function during an absence of the Shift Supervisor from the control room.

QUESTION 8.12 (2.00)

Answer the following regarding allowed working hours and overtime per the Grand Gulf TECHNICAL SPECIFICATIONS.

.a. An individual should not work more than hours straight, excluding shift turncver. (0.5)

b. There should be at least a(n) hour break between all work periods. (0.5)

c. An individual should not work more than in any seven day period. (0.5)

d. Any deviation from the above guidelines shall be authorized by .

(0.5)

QUESTION 8.13 (3.00)

List six types of events requiring one hour notification to the NRC.

(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)

,8. ' ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 23 QUESTION 8.14- (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 hours.

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. hours 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 />.

(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)

8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 24 QUESTION 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.

Ten 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?

NOfE: APPLICABLE TSs ARE ENCLOSED FOR REFERENCE.

a. Operational Condition 1 may be maintained for approxi-mately 29 days .

b. Within on hour, measures must be in 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 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 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 next 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 CONTINUED ON NEXT PAGE *****)

8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 25 QUESTION 8.16 (1.00)

All Fuel is removed.from the core; however, Fual Loading is scheduled to commence. TWO (2) Control Rods are removed from the core under the allowances of the Technical Specifications.

Which one of the following actions most accurately details the allowances and/or limitations imposed by the Technical Specifications in this instance?

NOTE: APPLICABLE TECHNICAL SPECIFICATIONS ARE ENCLOSED FOR REFERENCE

a. Fuel Loading may not commence until all Control Rods are inserted.

b. Fuel Loading may commence and continue as long as the Shut-down Margin requirements of TS 3.1.1 are satisfied.

c. Fuel Loading may commence - however the four fuel assemblies surrounding the removed Control Rods may not be loaded.

d. Fuel Loading may commence AFTER one of the Control Rods is inserted. The four fuel assemblies surrounding the removed Control Rod may not be loaded.

kASIEMC?Y

(***** END OF CATEGORY 08 *****)

(************* END OF EXAMINATION ***************)

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EQUATION SHEET f = ma v = s/t w.: ,.ye+

is,t 2 Cycle efficiency = "I '

E = aC -

a = (vg - y )/t KE = lsev vg = v, + at A = AN A = A,e" E PE = mgh to = 8/t A = in 2/tg = 0.693/tq W = v4P-

. t g(eff) = (t,,)(es)

AE = 931Am .

( + )

I) = faC AT I . I .o EX

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Pwr = W'f [a -X/ M I=I o 10 P=P 10 (t)- M = 1.3/u y.p t o

e /T HVL a 0.693/u

'SUR = 26.06/T ~

T = 1.44 DT SCR = S/(1 - K,gg)

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• AT SUR = 26 g CR =S/k1-K,ggx) x T = '(t*/o ) + [(g_;o)/x o] R1 (3 - Keff}1 " CR 2(1 ~~Keff)'2 T,= 1*/ (o - D M = 1/(1 - K,gg) = CR g/CR0 T = (3 - p)/ A,gg o g , (y ~ geff)0 Il ~ eff)1 P"I eff ~) eff

• A eff eff SDM = (1 - K,gg)/K,gg

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p= [1*/TKygg .] + [B/(1 + A,fgT )] 1* = 1 x 10 seconds P = I4V/(3 x 10 0) g aff =A 0.1 seconds" E = No Idgy=1d22 WATER PARAMETERS Id =Id2 g

1 gal. = 8.345 lbm R/hr = (0.5 CE)/d (meters)

I gal. = 3.78 liters R/hr = 6 CE/d (feet) -

1 ft = 7.48 gal.

MISCELLANEOUS CONVERSIONS ,

Density = 62.4 lbm/ft 3 1 Curie = 3.7 x 10 dps 10 Density = 1 gm/cm 1 kg = 2.21 lbm Heat of vsfori:ation = 970 Etu/lbm i hp = 2.54 x 103 BTU /hr Heat of fusica = 144 Btu /lbm 0 1 Hw = 3.41 x 10 Btu /hr i 1 Atm = 14.7 Psi = 29.9 in. Eg. I Btu = 778 ft-lbf 1 ft. H 2O = 0.4333 lbf/in 1' inch = 2.54 cm l F = 9/5 C + 32 "C = 5/9 ( F - 32)

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 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 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 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 l requirements, the ACTION may be taken in accordance with the specified time

( 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-(

fications.

(E - This specification is not applicable in OPERATIONAL CONDITION 4 or 5.

se :

c .. 3.0.4 Entry into an OPERATIONAL CONDITION or other specified condition shall

. . . not be made unless the conditions for the Limiting Condition for Operation are met without reliance on provisions contained in the ACTION requirements. This provision shall not prevent passage through or to OPERATIONAL CONDITIONS as required to comply with ACTION requirements. Exceptions to these requirements are stated in the individual Specifications.

k GRAND GULF-UNIT 1 3/4 0-1

p 4 %yw

' f, h).1 .,

p APPLICABILITY SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL CONDITIONS or.other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement.

4.0.2 Each Surveillance Requirement shall be performed within the specified time interval with:

a. A maximum allowable extension not to exceed 25% of the surveillance interval, but

b. The combined time interval for any 3 consecutive surveillance intervals shall not exceed 3.25 times the specified surveillance interval.

4.0.3 Failure to perform a Surveillance Requirement within the specified time interval shall constitute a failure to meet the OPERABILITY requirements for a Limiting Condition for Operation. Exceptions to these requirements are stated in the individual Specificatons. Surveillance requirements do not have to be performed on inoperable equipment.

4.0.4 Entry into an OPERATIONAL CONDITION or other specified applicable condition shall not be made unless the Surveillance Requirement (s) associated with the Limiting Condition for Operation have been performed within the applicable surveillance interval or as otherwise specified.

4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2, & 3 components shall be applicable as follows:

a. Inservice inspection of ASME Code Class 1, 2, and 3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves j shall be performed in accordance with Section XI of the ASME Boiler i and Pressure Vessel Code and applicable Addenda as required by 10 CFR

, 50, Section 50.55a(g), except where specific written relief has been

', granted by the Commission pursuant to 10 CFR 50, Section 50.55a(g)

(6) (1).

I

b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications:'

ASME Boiler and Pressure Vessel Required frequencies Code and applicable Addenda for performing inservice terminology for inservice inspection and testing inspection and testing activities activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days GRAND GULF s 1 3/4 0-2

4 APPLICABILITY SURVEILLANCE REQUIREMENTS (Continued)

c. The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection and testing activities.

d .' Performance of the above inservice inspection and testing activities shall be in' addition to other specified Surveillance Requirements.

e. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.

==

1 i

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GRAND GULF-UNIT 1 3/4 0-3

E h .

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 l

- 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.

,,, 5 ACTION:

a. With the number of OPERABLE 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

• within one hour. The provisions of Specification 3.0.4 are not applicable.

b. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for both trip systems, place at least one trip system ** in the tripped condition within one hour and take the ACTION required by Table 3.3.1-1.

SURVEILLANCE REQUIREMENTS

( 4.3.1.1 Each reactor protection system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at

_ the frequencies shown in Table 4.3.1.1-1.

4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of I 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 functional unit shown in Table 3.3.1-2 shall be demonstrated to be within its 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

, TABLE 3'.3.1-1 -

REACTOR PROTECTION SYSTEM INSTRUMENTATION ,

o APPLICABLE MINIMUM I E OPERATIONAL OPERABLE CHANNELS

FUNCTIONAL UNIT CONDITIONS PER TRIP SYSTEM (a) ACTION

1. Intermediate Range Monitors:

'~

a. Neutron Flux - High 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 - High, Setdown 2 3 1 R*

3 3 2 w

5(b) 3 3 a b. Flow Biased Simulated Thermal ~'

Power - High 1 3 4

c. Neutron Flux - High 1 3 -4 i

d. Inoperative 1, 2 3 1 3 3 2

. 5 3 3

3. Reactor Vessel Steam Dome 8" '

Pressure - High 1,2(d) 2 1 i 4. Reactor Vessel Water Level - Low, i level 3 1, 2 2 1

5. Reactor Vessel Water Level-High, '

Level 8 1(e) 2 4

6. Main Steam Line Isolation Valve -

1 Closure 1(*) 4 4 I 7. Main Steam Line Radiation - High 1, 2(d) 2 5

8. Drywell Pressure - High 1, 2(I) 2 1

~'

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I g o-Q - - " y, d 1, y ,

, TABLE 3.3.1-1 (Continued) , ,

l REACTOR PROTECTION SYSTEM INSTRUMENTATION .

E G APPLICABLE MINIMUM E OPERATIONAL OPERABLE CHANNELS FUNCTIONAL UNIT CONDITIONS PER TRIP SYSTEM (a) ACTION

{

, 9. Scram Discharge Volume Water Level - High 2 1 1I93 S 2 3

10. Turbine Stop Valve - Closure 1(h) 4 6

11. Turbine Control Valve Fast Closure, Valve Trip System Oil Pressure - Low I Ih) 2 6

!w 12. Reactor Mode Switch Shutdown 1 Position 1, 2 2 1

,w 3, 4 _

2 7

,a 5 2 3 a

13. Manual Scram 1, 2 2 1 3, 4 2 8

5 2 9 l

) y 1 -

1 .

i 3

INSTRUMENTATION TABLE 3.3.1-1 (Continued)

{~

REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION AC. TION 1 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 />.

ACTION 2 -

Verify all insertable control rods to be inserted in the core and lock the reactor mode switch in the SHUTDOWN position within one hour.

ACTION 3 -

Suspend all operations involving CORE ALTERATIONS *, and insert all insertable control rods within one hour.

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 />.

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 />.

ACTION 6 -

Initiate a reduction in THERMAL POWER within 15 minutes and reduce"*forbine 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 within /

one hour. (,

ACTION 8 -

Lock the reactor mode switch in the SHUTDOWN position within one hour.

ACTION 9 -

Suspend all operations involving CORE ALTERATIONS *, and insert all insertable control rods and lock the reactor mode switch in the SHUTDOWN position within one hour.

"Except movement of IRM, SRM or special movable detectors, or replacement of LPRM strings provided SRM instrumentation is OPERABLE per Specification 3.9.2.

f k

GRAND GULF-UNIT 1 3/4 3-4

. - . . ._ ~ . - _ .

b e TABLE 3.3.1-1 (Continued) r-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.

(b) The " shorting links" shall be removed from the RPS circuitry prior to and

..s during the time any control rod is withdrawn *'per Specification 3.9.2

- - -- - 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.

(d) This function is not required to be OPERABLE when the reactor pressure f;' vessel head is unbolted or removed per Specification 3.10.1.

(e) This function shall be automatically bypassed when the reactor mode switch

- is not in the Run position.

(f) This function is not required to be OPERABLE when DRYWELL INTEGRITY is not required.

(g) With-any control rod withdrawn. Not applicable to control rods removed l ( per Specification 3.9.10.1 or 3.9.10.2.

(h) This function shall be automatically bypassed when turbine first stage pressure is less than 30%** of the value of turbine first stage' pressure

" in psia, at valves wide open (VWO) steam flow, equivalent to THERMAL POWER i ,

less than 40% of RATED THERMAL POWER.

"Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.

i ** 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.

GRAND GULF-UNIT 1 3/4 3-5

• .-s TABLE 3.3.1-2 .

c2 m -

3 REACTOR PROTECTION SYSTEM RESPONSE TIMES ,

o .

G; RESPONSE-TIME d: FUNCTIONAL UNIT ,

(Seconds) z I

1. Intermediate Range Monitors: .

a. Neutron Flux - High NA

b. Inoperative NA

2. Average Power Range Monitor *:  !

a. Neutron Flux - High, Setdown t#4

b. Flow Biased Simulated Thermal Power - High 5 0.09**

c. Neutron Flux - High < 0.09

d. Inoperative RA u, 3. Reactor Vessel Steam Dome Pressure - High 5 0.3s

]; 4. Reactor Vessel Water Level - Low, Level 3 5 1.05 u, S. Reactor Vessel Water Level - High, level 8 1 1.05 J, 6. Main Steam Line Isolation Valve - Closure 1 0.06

7. Main Steam Line Radiation - High NA ,

8. Drywell Pressure - High NA

9. Scram Discharge Volume Water Level - High NA

10. Turbine Stop Valve - Closure -< 0.10

11. Turbine Control Valve Fast Closure, Valve Irip System g.

Oil Pressure - Low < 0.10

12. Reactor Mode Switch Shutdown Position HA

13. Manual Scram NA ur .

• Neutron detectors are exempt from response time testing. Response time shall be measured from the detector output or frs*i the input of the first electronic component in the channel.

• • Not including simulated thermai power time constant. ,

1. Measured from start of turbine control valve fast closure.

-... . . . . . . . . Ec. : .... .. ,. . ,

~-

/ /

/ 3 s ,

c) sz IA8tf 4.3.1.1-1

• RE ACIUR PR0l[Cil0N SYSl[M INSIRUMENIAl10N SURVElll ANCE' R[QUIRIMENT'S O

r- UPERATIONAL CilANNE L I CilANN[L IUNCIl0NAL CilANNE L CON 0lil0NS FOR WillCH E fDNCIIONAL UNil .Cill[K ((51 CAllBRAll0N(#} SURVElllANCE REQUIRED G 1. Intermediate Range Monitors: .

" 5/U,5,II'I 5/U, W

4. Neutron flux - liiuh R 2 5 W R 3,'4,

5 .

b. Inoperative NA W

' NA 2,'3, 4, 5 l 2. Average Power Range Monitor: IO j i

a. Neutron Ilum - High, S/U,5,(b) S/U, W 5A 2>

Setdown 5 W SA 3,} 5 ,

i  !

b. Flow Biased Simulated .

l Thermal Power - High 5. D(h) y g(d)(e). $g, g( 0 g. ,

w

~

.2

c. Neutron flux - High 5 W W Id) , 5A 1i .

Y N d. Inoperative NA W NA 1,;2, 3, 5

3. Reactor Vessel Steam Dome $

R I9I

Pressure - liigh 5 M 1,$2IN

4. Reactor Vessel Water Level - '

low, Level 3 5 M R I9 1,l2 .

5. Reactor Vessel Water level -  ?

1 liiuh, Level 8 5 M R I9I 1 p l

6. Main Steam Iine isolation f Valve - Closure NA H R 1' i  :

1 /. Main Steam line Radiation - gj) liigh 5 H R 1,.2

8. Drywell Pressure - liiuh 5 M k(U} 3. /(

O I

I o T A8tf 4.3.1.1-1 (Continued) '

y

< g RE ACIOR PROILCil0N SYSIEM INSIRUNINI All0N SURVE lil ANCE WlQUIREMENIS CilANNil OPERATIONAL

@ fuNCil0HAL CalANN[L CON 0lil0NS FOR WillCH q CllANNEL Litf CK IL51 CALIBRAll0N SURVEILLANCE REQUIRED FUNCTIONAL UNIT f _

3 9. Scram Olscharge Volume Water H Level - liigh 5 M R IGI 1, 2, S III Turbine Stop Valve - Closure 5 M R IU 1 ,

10.

11. Iurbine Control Valve Fast i Closure Valve Trip System Dil 5 M R 1 Pressure - Low

12. Reactor Mode Switch 1,2,3,4,5 +

shutdown Position NA R NA M NA 1,2,3,4,5

13. Manual Scram NA i

U i (a) Neutron detectors may be excluded from CalANNit CAllBRAfl0N.

i

'I (b) The IRM and SRM channels shall be determined to overlap for at least 1/2 decade during each CD startup af ter entering OPERAll0NAL CON 0llI0N 2 and the IRM and APRM channels shall be deter-mined to overlap for at least 1/2 decade during each controlled shutdown, if not performed within the previous 7 days.

(c) [UttEI!D)

(d) !his calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by.a heat b4 ance during OPERAll0NAl CON 0lll0N 1 when lil[RMAL POWER > 25% of RATED IM RMAL F0WER. Adjust the APRM channel if the absolute difference is greater than 2% of RAIE0 litt NMAL POW [R. Any APRM chan arl gaisi adjustmesit made in compliance with Specification 3.2.2 shall not be included in determining the absolute dillertnce. ,

f0I ICCli Sp0C /h)1l-(e) Ihis calibration shall consist ut tue adjustment of the APRM flow biased channel to conform to , '

calitrated flow signal.  ; see IsPs 101 l_

(t) 1he !?RMs shall be calibrated at least once per 1000 M /l using the itP system.

(9) Callt. ste trip unit at least once per 31 days.

(h) Verify measured drive flow to be less than or equal to established drive flow at the existing flow con-trol v.lve position.

i (i)

Ihis ca.libration shall consist of veritying the f> 1 1 second simulated thermal power time constant.

(j) Not applicable when the reactor piessure vessel head is unholted or removed per Specification 3.10.1.

(k) Not appeicable when ORYWLil IHilGRilY is not required.

(1) Applica' ale with any control rod withdrawn. Not applicable to control rods removed per Spatilita-

. Lion 3.ta.10.1 or 3.9.10.2.

O O

6 INSTRUMENTATION 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 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 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.

APPLICABILITY: As shown in Table 3.3.2-1.

ACTION: _

a,. .Wi.th an isolation actuation instrumentation channel trip setpoint

. . . .less conservative than the value shown in the Allowable Values column of Table 3.3.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

b. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip system, place the inoperable channel (s) and/or that trip system in the tripped condition

• within one hour. The provisions of Specification 3.0.4 are not appli<able.

c. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for both trip systems,

,, place at least one trip system ** in the tripped condition within one hour and take the ACTION required by Table 3.3.2-1.

.k SURVEILLANCE REQUIREMENTS 4.3.2.1 Each isolation actuation instrumentation channel shall be demonstrated

~ OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and 1

CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.2.1-1.

4.3.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of 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 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 3.3.2-1 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

( systems have the same number of inoperable channels, place either trip system in the tripped condition.

GRAND GULF-UNIT 1 3/4 3-9

9 o

5 TABLE 3.3.2-1 .

5 ISOLATION ACIUA110N INSIRUMENIATION C

g, VALVE GROUPS MINIMUM APPLICABLE

• OPERATED BY OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION SIGNAL'(a) PER TRIP SYSTEM (b) CON 0lil0N ACil0N E.

-4 1. P..lMARY CONIAINMENT ISOLATION

a. Reactor Vessel Water Luel-Low Low, Level 2 6A, 7, 8, 10(c)(d) 2 1, 2, 3 and # 20

b. Reactor Vessel Water Level- g low Low level 2 (ECCS - .-

Division 3) 68 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 y e. Drywell Pressure-High A (ECCS - Division 1 and w Division 2) SI "II"I 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 *I 1, 2, 3 and

• 21

h. Manual Initiation 6A, 7, 8, 10 ICIId) 2 1, 2, 3 and *# 22

2. MAIN STEAM LINE ISOLATION

a. Reactor Vessel Water Level-Low low Low, Level 1 1 2 1, 2, 3 20

b. Main Steam Line Radiation - High 1, 10 III 2 1,2,3 23 c, Main Steam Lirie Pressure - Low 1 2 1 24

d. Main Steam Line Flow - liigh I 8 1,2,3 23

e. Condenser Vacuum - low 1 2 1, 2,** 3"* 23

. l .

f F i .

O TABLE 3.3.2-1 (Continued)

C

. i ISOLATION ACTUATION INSTRlMENTATION O

r- VALVE GROUPS HINIMUM APPLICABLE T OPERATED BY OPERABLE CHANNELS OPERATIONAL E TRIP FUNCTION SIGNAL (a) PER TRIP SYSTEM (b) CONDITION ACTION

2. MAIN STEAM 1INE 1501ATION (Continued)

f. Main Steam Line Tunnel lemperature - High 1 2 1,2,3 23

g. Main Steam Line Tunnel a Temp.- High 1 2 '

1, 2, 3 23

h. Manual Initiation 1, 10 2 1,2,3 22

3. SECONDARY CONTAINMENT ISOLATION -

a Reactor Vessel Water Level-Low Low, Level 2 N.A.(c)(d)(h) 2 1, 2, 3, and # 25 m b. Drywell Pressure - High N.A.(c)(d)(h) 2 1,2,3 25 1 c. Fuel Handling Area N.A.II) 2 1, 2, 3, and

• 25 w Ventilation Exhaust 4 Radiation - High High

" d. Fuel Handling Area ,

Pool Sweep Exhaust Radiation - liigh High N.A.(j) 2 1, 2, 3, and'* 25

e. hanual Initiation 2 1, 2, 3 26 N.

N.A.A.((c)(d)(h) c)(d)(h) 2

• 25

4. REACTOR WATER CLEANUP SYSTEM 150'LAT10N

a. A Flow - High 8 1 1, 2, 3 27

b. A Flow Timer 8 1 1, 2, 3 27 '

c. Equipment Area Temperature - 8 1/ room 1, 2, 3 27 High

d. Equipment Area a Temp. -

High 8 1/ room 1, 2, 3 21

e. Reactor Vessel Water tevel - Low Low, Level 2 8 2 1,2,3 2/

_ _J

- . . . . . m c)

$3 TABLE 3.3.2-1 (Continued) .

z 150LAIlON ACIUAIIGN INSTRUMENIATION r; VALVE GROUPS MINIMini APPLICABLE a OPERATED BY OPERAutE CilANNELS OPERA 110HAL 55 TRIP FUNCil0N SIGNAL (a) PIR 1 RIP SYSIEM (b) CONDlil0N ACTION Z 4. *

,, REACIOR WAIER CL EANUP SYSTEM 1501 ATION (Continued)

f. Main Steam Line Iunnel 8 1 1, 2, 3 27 Ambicnt Temperature - liigh

g. Main Steam Line Tunnel a  !

Temp. - Nigh 8 1 1, 2, 3 21 II)

h. SLCS Initiation 8 1 1, 2, 5## 30

i. Manual Initiation 8 2 1, 2, 3 26

5. REACTOR CORC ISOLATION COOLING SYSTEM 150lAll0N o, a. RCIC Steam Line Flow - liigh 31 1. Pressure 4 1 1, 2, 3 27 ca 2. Time Delay 4 1 1, 2, 3 27

$$ b. RCIC Steam Supply Pressure - Low 4, 9I "I 1 1, 2, 3 21

c. RCIC Iurbine Exhaust Diaphragm Pressure - liigh 4 2 1,2,3 27

d. RCIC Equipment Room Ambient lemperature - liigh 4 1 1,2,3 21

e. RCIC Equipment Room a Temp.

- Nigh 4 1 1, 2, 3 21

f. Main Steam line Tunnel Anb ien t leniperature - liigh 4 1 1, 2, 3 27 9 Main Steain Line lunnel a Temp. - liigh 4 1 1,2,3 2/

h. Main Steam Line Tunnel temperature Timer 4 1 1, 2, I .' /

Q ._

(' <

U (h . , ' . '.. .:<

f .H ,

I-l-

o -

g TABLE.3.3.2-1 (Continued) z '

C ISOLATION ACTUATION INSIRUMENTATION ,

c- APPLICABLE r- VALVE GROUPS MINIMUM 7 OPERAIED BY OPERABLE CHANNELS OPERATIONAL g TRIP FUNCTION SIGNAL (a) PER TRIP SYSTEM (b) CONDITION AC110N

5. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION

i. RHR Equipment Room Ambient Temperature - High 4 1/ room 1, 2, 3 27 j .- RHR Equipment Room a Temp. - e High 4 1/ room 1, 2, 3 27

k. RHR/RCIC Steam Line Flow -

High 4 1 1, 2, 3 27

1. Manual Initiation 4 III 1 1,2,3 26-w D m. Drywell Pressure-High 9("I 1 1,2,3 27 w (ECCS-Division 1 and 4 Division 2) w

6. RHR SYSTEM ISOLATION

a. RHR Equipment Room Ambient --

Temperature - 1;igh 1, 2, 3 28 3 _-. , _ 1/rCom ,

b. RHR Equipment Room a s Temp. - High ~'s 3 1/ room 1, 2,' 3 .?2. -

c. Reactor Vessel Water Level - Low, Level 3 3 2 1, 2, 3 2tf ~-..

~

d. Reactor Vessel (RHR Cut-in ,

Permissive) Pressure - g) '

tu.ji- - 3 -2. 1, 2, 3 28- _

e. Dryweil Pressure - High 3

0) g 1, 2, 3 28

f. P.anual Initiation 3 2 1, 2, 3 26 s

-N-,

+

INS M NTATION TABLE 3.3.2-1 (Continued)

ISOLATTDN ACTUATION INSTRUMENTATION ACTION \

ACTION 20 -

Be in at least HOT SHUT 00WN 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 21- -

Close the affected system isolation valve (s) within one hour or:,

a. 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.

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 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 />. I 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 close the affected system isolation valves within tne next hour and declare the affected system inoperable.

ACTION 27 -

Close the affected system isolation valves within one hour and declare the affected system inoperable.

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.

ACTION 29 -

Close the affected system isolation valves within one hour and declare the affected system or component inoperable or:

a. In OPERATIONAL CONDITION 1, 2 or 3 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 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 />.

!' b. In OPERATIONAL CONDITION # suspend CORE ALTERATIONS and opera-tions with a potential for draining the reactor vessel.

j ACTION 30 -

Declare the affected SLCS pump inoperable.

! NOTES

• When handling irradiated fuel in the primary or secondary containment and

! during CORE ALTERATIONS and operations with a potential for draining the l reactor vessel.

i ** The inw condenser vacuum MSIV closure may be manually bypassed during reactor 4

SHUTOOWN or for reactor STARTUP when condenser vacuum is below the trip set-point to allow opening of the MSIVs. The manual bypass shall be removed when

, condenser vacuum exceeds the trip setpoint.

l # During CORE ALTERATIONS and operations with a potential for draining the l reactor vessel.

i ## With any control rod withdrawn. Not applicable to control rods removed per Specification 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.

l l

! GRAND GULF-UNIT 1 3/4 3-14 i

k

9 9

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 mode of operation.

(e) Two upscale-Hi Hi, one upscale-Hi Hi and one downscale, or two downscale

. _ _ _ signals from the same trip system actuate the trip system and initiate isolation of the associated containment and drywell isolation valves.

(f) Also trips and isolates the mechanical vacuum pumps.

(g) Deleted.

(h) Also actuates secondary containment ventilation isolation dampers and valves per Table 3.6.6.2-1.

(i) Closes only RWCU system isolation valves G33-F001, G33-F004, and G33-F251.

(j) Actuates the Standby Gas Treatment System and isolates Auxiliary Building l penetration of the ventilation systems within the Auxiliary Building. l (k) Closes only RCIC outboard valves. A concurrent RCIC initiation signal is

-required for isolation to occur.

(1) Valves E12-F037A and 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-F0428 are closed by Containment Spray System initiation signals.

(o) Also isolates valves E61-F009, E61-F010, E61-F056, and E61-F057 from Valve Group 7.

GRAND GULF-UNIT 1 3/4 3-15

9 f.. .

CONTAINMENT SYSTEMS L

3/4.6.7 ATMOSPHERE CONTROL (g CONTAINMENT HYOR0 GEN RECOMBINER SYSTEMS LIMITING CONDITION FOR OPERATION 3.6.7.1 Two independent containment hydrogen recombiner systems shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. _ _

ACTION:  ;... . . , . _ ., ; ... .

With one containment hydrogen recombiner system inoperable, restore the inoperable system to OPERABLE status within 30 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 />.

SURVEILLANCE REQUIREMENTS 4.6.7.1 Each containment hydrogen recombiner system shall be demonstrated OPERABLE:

a. At least onc per 6 months by verifying during a recombiner system functional test that the minimum heater sheath 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 by:

1. Performing a CHANNEL CALIBRATION of all control room recombiner instrumentation and control circuits.

i. 2. Verifying the integrity of all heater electrical circuits by b performing a resistance to ground test within 30 minutes following lt the above required functional test. The resistance to ground for any heater phase shall be greater than or equal to 10,000 ohms.

3. Verifying during a recombiner system functional test that the 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. [ DELETE 0]

l ,

GRAND GULF-UNIT 1 3/4 6-58

CONTAINMENT SYSTEMS.

f' i CONTAINMENT AND DRYWELL HYDROGEN IGNITION SYSTEM LIMITING CONDTION FOR OPERATION 3.6.7.2 The. containment and drywell hydrogen ignition ~ system consisting of the following:

a. ' At least two igniter assemblies in each enclosed area specified in Table 3.6.7.2-2, 2 ..

~ ~

b. All igniter assemb' lies adjacent to any inoperable' igniter assembly in each open area specified in Table 3.6.7.2-2, and

c. Two independent containment and drywell hydrogen ignition subsystems each consisting of two circuits (as listed in Table 3.6.7.2-1) with no more than two igniter assemblies inoperable per circuit.

shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2 ACTION:

a. With less than two igniter assemblies OPERABLE in any enclosed area

' specified in Table 3.6.7.2-2, restore at least two igniter assemblies to g OPERABLE status within 7 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 />.

b. With any adjacent igniter assemblies within an open area as specified in

_. Table 3.6.7.2-2 inoperable, restore the igniter assemblies in that open area so that all igniter assemblies adjacent to an inoperable igniter assembly are OPERABLE within 7 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 />.

i I c. With one containment and drywell hydrogen ignition subsystem inoperable,

! restore the inoperable subsystem to OPERABLE status within 7 days 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 />.

l SURVEILLANCE REQUIREMENTS __

1 l 4.6.7.2 The containment and drywell hydrogen ignition system shall be demonstrated OPERABLE:

a. At least once per 92 days by energizing the supply breakers and:

1. Verifying a visible glow from the glow plug tip of each normally accessible igniter assembly specified in Table 3.6.7.2-2, l

(

GRAND GULF-UNIT 1 3/4 6-59

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

C

2. Verifying that each circuit of each containment and drywell hydrogen

. ignition subsystem is conducting sufficient current to energize the minimum required number of igniter assemblies specified in -

Table 4.6.7.2-1. -

b. At every COLD SHUTDOWN, but no more frequently than once per 92 days, by energizing the supply breakers and verifying a visible glow from the glow plug tip of each normally inaccessible igniter assembly specified in Table 3.6.7.2-2.

c. At least once per 18 months by:

1. Verifying the cleanliness of each glow plug by a visual inspection.

2. Energizing each glow plug and verifying a surface temperature of at least 1700*F.

l l

l l

i GRAND GULF-UNIT 1 3/4 6-60

-,-rv - - - - *m--r -- - - -- w- , .g - -- -ee w --,--- - -w -- .--.v-mm

S' i

REFUELING OPERATIONS MULTIPLE CONTROL ROD REMOVAL LIMITING CONDITION FOR OPERATION 3.9.10.2 Any number of control rods and/or control rod drive mechanisms may be removed from the core and/or reactor pressure vessel provided that at least the following requirements are satisfied until all control rods and control

< - rod-drive mechanisms are reinstalled and all control rods are inserted in the Core,

a. The reactor mode switch is OPERABLE and locked in the Shutdown position or in the Refuel position per Specification 3.9.1, except that the Refuel position "one-rod-out" interlock may be bypassed, as required, for those control rods and/or control rod drive mechanisms to be removed, after the fuel assemblies have been removed as specified below.

b. The source range monitors (SRM) are OPERABLE per Specification 3.9.2.

c. The SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied.

d. All other control rods are either inserted or have the surrounding four fuel assemblies removed from the core cell.

e. The four fuel assemblies surrounding each control rod or control rod .

drive mechanism to be removed from the core and/or reactor vessel are removed from the core cell.

.~'~ f. All fuel loading operations shall be suspended unless all control rods are inserted in the core.

APPLICABILITY: OPERATIONAL CONDITION 5.

ACTION:

With the requirements of the above specification not satisfied, suspend removal of control rods and/or control rod drive mechanisms from the core and/or reactor pressure vessel and initiate action to satisfy the above requirements.

i GRAND GULF-UNIT 1 3/4 9-16

5. THEORY'OF NUCLEAR POWER' PLANT OPERATION. FLUIDS. AND PAGE 26 j ,T H E R M 0 D Y N A!j E S ANSWERS - . GRAND GULF 1 -86/09/08-SPENCER, M.

ANSWER. 5.01 (1.00) d REFERENCE Grand Gulf, Heat Transfer and Fluid Flow.

ANSWER ^ 5.02 (1.00)

/G// ctzM T~ For A?ay 2

- 1. Photoneutron source j of Oog 7pg

2. Spontaneous fission

'3. -Alpha-neutron reaction [any 2 @ 0.50 each]

REFERENCE GG, GE Reactor Theory, Ch. 2, PP. 2-4 ANSWER 5.03 (1.50) eff = 1 - SDM SDM = ( l- Eef(-)/ Re.g

= 1 - 0.30 03= ( l - r,g)/ r,pg.

= :0.7 [0.5]

0.3 kau = 1- kes; 5  : CR (1-Ke*f ) /.7 (,f4 : J 1 1

=

=

30(1 - .7) 9 -[0.5]

R e (+ =

• = 0.77 (0,75]

N GR t  : ct, ( l _ teg ),

CR = 5/(1 - e f f )

2 2

=

(.1 - K.+4 )2.

9/(1 - 0.9

= 450 cps [0.5] = 3 o ( l- 0,77)

REFERENCE ( l- O,4B)

GG, GE Reactor Theory, Ch. 3, P. 3-10b

34-5' c.fr ( d,W) ,

7- -

.5. THEORY'OF NUCLEAR' POWER' PLANT OPERATION. FLUIDS. AND PAGE 27.

, sIEERMODYNAMICS

ANSWERS - . GRAND GULF 1 -86/09/08-SPENCER, M.

ANSWER 5.04 (1.00) b.

REFERENCE GG,~ Reactor Theory, Ch. 3,.P. 3-73 ANSWER 5.05 (1.50)

a. MORE NEGATIVE

b. 'NO' CHANGE Ec. MORE NEGATIVE -[3 @ 0.5 each]

REFERENCE GG,. Reactor Th'eory, CH. 4, P. 4-66 ANSWER 5.06 (1.50)

a. more-negative

b. more negative

c. less negative [3 @ 0.5 each].

, REFERENCE cGG, Reactor Theory, Ch. 4, P. 4-70

' ANSWER 5.07 (1.00)

a. radial

b. axial l -REFERENCE

! GG, Reactor,. Theory, Ch. 5, P. 5-46 l~

l I

L ._

c-1 THEORY OF NUCLEAR POWER PLANT OPERATION FLUIDS. AND PAGE 28

, ,THERMODYNAMLQS ANSWERS -- GRAND GULF 1 -86/09/08-SPENCER, M.

ANSWER 5.08 (1.00)

a. more negative

b. no change-

c. less negative

d. less negative [4 @ 0.25 each)

REFERENCE GG, Reactor Theory, Ch. 5, P. 5-41 ANSWER 5.09 (1.00) h 1. To control excess reactivity [0.50]

2. Shape thermal flux (avoid local power peaking) [0.50]

REFERENCE GG, Reactor Theory, Ch. 6, P. 6-34 ANSWER 5.10 (1.00)

a. 2

b. 4

c. 1

d. 3 [4 @ 0.50 each)

REFERENCE GG, Reactor Theory, Ch. 7, P. 7-52 VASIB CPY

~

5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. A@ . PAGE 29 j lrHERM0 DYNAMICS-ANSWERS -- GRAND GULF 1 -86/09/08-SPENCER, M.

- ANSWER' 5.11 (2.00)

1. E

.2. F

3. B

4. A [4 @ 0.5 each)

~

a REFERENCE' GG, Reactor Theory,.Ch. 7, P. 7-32 4

ANSWER 5.12 (1.50)

1. Material cross section 20f5 If '# 8 $ C##SS #8C Ub" Gnos o As d:fkuur -f' Ac rots

2. Energy of neutron 11 .3. Mass of target nucleus

4. Energy of the-target nucleus [any 3 @ 0.50 each]

REFERENCE 1

GG, Basic Atomic & Nuclear Phy., Ch. 5, P. 5-41 ANSWER 5,13 ( .50) b.

l. REFERENCE GG, Chemistry, Ch. 5, P. 17 ANSWER 5.14 ( .50)

.b i- ' REFERENCE GG, Chemistry, Ch. 7, P. 7-22 i

J

l 5 '. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND' PAGE- 30

_g.  ; THERMODYNAMICS .

-ANSWERS ---GRAND GULF-1 -8s/09/08-SPENCER, M.

ANSWER' 5.15 (1.00)'

1. C

-: 2 . 'G

3. H E- [4 0 0.25 each]

4. (fh

. REFERENCE GG, Heat. Transfer & Fluid Flow, Vol. 2, Ch. 8, P. 9 ANSWER 5.16 (1.50)

a. INCREASE b '. DECREASE

c. DECREASE [3 @ 0.5 each]

REFERENCE GG, Heat Transfer and Fluid Flow, Vol. 2, Ch. 9,-PP. 26 -23 ANSWER 5.17 (1.00) ,

d' REFERENCE GG, Heat Transfer & Fluid Flow, Vol. 2, Ch. 9, P. 9 i

f f

L -

5 .-

THEORY'OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 31 !

, jTHERMODYNAMICE I ANSWERS - GRAND GULF 1 -86/09/08-SPENCER, M.

t

' ANSWER 5.18 (3,00)

n. ,LHGR b. APLHGR c-. CPR

1. local fuel - average fuel total fuel-pin power pin power in- bundle power in mode mode power

2. 1% plastic. clad temp. of boiling strain on 2200 Deg F transition cladding  ;

3. fuel pellet decay heat and loss of nucleate expansion store heat boiling around following LOCA cladding

4. fuel'eladding gross cladding fuel cladding cracking due failure due to due to lack of to high stress lack of cooling cooling

[12 @ 0.25 each]

REFERENCE GG,' Heat Transfer & Fluid Flow, Vol. 2, Ch. 9, P. 33 ANSWER 5.19 (1.00)

As the core ages, control rods are withdrawn for fuel burnup, and the result of this action is an increase in core sine, a decrease in the negative effects of leakage, a decrease in the number of fuel nuclei, and an increase in moderator-to-fuel ratio to cause a positive trend of.the total moderator temperature coefficient. (1.0)

REFERENCE GG, Reactor Theory, Ch. 4, P. 11

l

. '5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND- PAGE 32

' THERMODYNAMICS

~

o.

-ANSWERS -- GRAND GULF.1 -86/09/08-SPENCER, M.

ANSWER' 5.20 1(1.50)

a. flow = 1,332 gpm [.2000 X 0.6666 ] 16

b. pressure = l .f++

00 - [-1000 X 0.666 squared ] 4 JO/o .i 44 + - s I /07o

c. power = .44-: Hp [ 150 X 0.666 cubed ] ( 3 @ 0.5 ea = 1.5 )

(o.4 tk h4 0.\ W kl REFERENCE

. Grand Gulf, Heat Transfer and Fluid Flow, Vol 2, Chapter 6, page 6-96a.

1 l

1.

t 4

4

'6. " PLANT SYSTEMS DESIGN.' CONTROL. AND INSTRUMENTATION PAGE 33

• ANdWERS -- GRAND GULF 1 -86/09/08-SPENCER, M.-

ANSWER 6.01 (1.00) g 8.

REFERENCE GG, Lesson Plan - OP-C34-501, P. 17 ANSWER 6.02 (1.50)

a. Reactor water level of + 11.4" (0.5)'

b. 3 (1.0)

REFERENCE GG, Lesson Plan - C34-501, P. 12 ANSWER 6.03 (2.00)

1. 3% S/D Margin

2. Decreased control rod worth (as moderator cools)

3. Reduced neutron leakage from boiling to cooldown

4. Reduced Doppler effect

5. Changing water density from hotito cold

6. Elimination of voids

7. Complete decay of rated power xenon inventory

[any 4 @ 0.5 each]

REFERENCE GG, Lesson Plan - C41, P. 11 ANSWER 6.04 (1.00) c MAS"R C:PY

J6. ' PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 34

• ANSWERS -- GRAND GULF 1 -86/09/08-SPENCER, M.

REFERENCE GG, Lesson Plan -'C41, P. 20 ANSWER- 6.05 (1.50)

The thermal neutron is unchanged. The Uranium Oxide is fissioned by a neutron-and the fission fragments being highly charged cause ionization of the' argon gas. [1.0] The ionization of the argon causes an electrical discharge between the cathode and anode and appears as a " pulse" from the-detector. [0.5]

REFERENCE GG, Lesson Plan - C51-1, P. 5 ANSWER 6.06 (1.00)

The sensitivity of the detector would decrease more rapidly if only U-235 at normal power levels. [1.0)

-- OR --

Addition of U-234 provides a method of replacing or regenerating U-235 by neutron capture. [1.0]

REFERENCE GG, Lesson Plan --C-51-3, P. 5 ANSWER 6.07 (2.00) 1.

0 20 4*l*0to 0.z0 Down scale [0f Gfr] le.ss thanfA' [0,3&] power in RUN [0,26']

O.20 0.2o 0,20

2. High Neutron Flux [Q 25'] greateg than 12% [0,,35']Z'in STARTUP [0,,.263; greater than .66W + 42% [0,F5]'*_tn RUN. [0.J&]

go 2 .rly n 4I ogscale (, O. Z o ] 44- l C 8 */o C O,2.0,3 GG, Lesson Plan - C-51-4, P. 17 ANSWER 6.08 ( .50)

Suppression Pool REFERENCE GG, Lesson Plan - C-71, P. 4 ,

p

6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION- PAGE 35-

~'

iAN.!WERS -- GRAND-GULF 1 --86/09/08-SPENCER, M.

ANSWER 6.09 (1.00)

1. Allow SRM scram functions

2. l Removes all neutron monitoring coincidence [2 @ B.50'each]

REFERENCE oGG , Lesson Plan - C-71, P. 14 ANSWER 6.-10 (1.00)

1. Flow great-r than 1000 gpm ( + / - 100 gpm )

2. RHR pump t. aker closed - [:2 @ 0.5 each] <

REFERENCE oGG, Lesson Plan - E-12, P. 12 ANSWER 6.11 (1.50)

1. Containment pressure [0.25] greater than or equal to 7.84 psig [0.25 WITH

2. Drywell pressure [0.25] greater than or equal to 1.39 psig.[0.25]

3. 10.85 minute [0.25] after a LOCA [0.25]

REFERENCE GG, Lesson Plan - E-12, P. 17 '

ANSWER 6.12 (2.00)

a. .7115 gpm' [0.25] ( + / - 10 % )

~

b. -P rY$1n ori$ $ b'. ] 9 .8.[9.25]- ( + / - 50 gpm )

3-

c. Suppression pool [0.25]

I d .- Reactor level [0.25] 1 - 150.3 inches [0.25]

! OR Hi Drywell Pressure [0.25] 1.39 psig [0.25]

l l

• 3 myw .=,w we,-e- .w,--es-.i--- .e -c-.,.--m.g 33.m,-wy--m -w,. 4 ---g.ery.e g * - - - - - - - - --mtyye g-m --

py-ey-em---y-- , ,-y,- ,r -p,,wv=------= ----

~

6. ;-PLANT-SYSTEMS DESIGN. CONTROL.'AND INSTRUMENTATION PAGE 36

-ANSWERS -- GRAND GULF 1 -86/09/08-SPENCER, M.

. REFERENCE GG,, Lesson Plan - E-21, PP. 4, 6,.and 17 ANSWER' o6.13 (2.00)

1. Voltage regulator not-in " Auto"

2. Engine lock-out not reset

3. Generator lock-out not reset

.4 . Unit Mode Sel. Switch not in " Auto"

5. HPCS motor breaker control power failure

6. Undervoltage circuit control power failure [any 4 @ 0.5 each]

} <}

gen. ker ouhf A breake* in l o w., ce ca,,1,oj yo u,. -f a,*Ivre ,

GG, Lesson Plan E-22-1, P. 16 ANSWER 6.14 (1.00)

1. Prevent water hammer

2. Start cooling at the earliest moment -[2 0 0.5 each]

REFERENCE GG, Tech. Spec., P. 3/4 5-1 1,00 ANSWER 6.15 p.Ae7

a. ECCS discharge pressure interlocks [0.5]

h. F051B [0.25] and F051A [0.25)) Del.t.4 &v ,., ex4 m

c. Six [0.5]

REFERENCE GG, Lesson Plan - E-22-2, P. 7, 8 f

F 16- .

= PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE :37 INRfERS.--GRANDGULF1 -86/09/08-SPENCER, M.

LANSWER- 6.~ 16 (1.25)

Greater than~or equal to 3 sec. [0.25]

-Less_than or equal to 5 sec. [0.25]

Bases Fast enough to limit the release of radioactivity to the environment [0.25]

in the event of a guillotine break of one steam line outside primary containment ~[C.25].

Slow enough to-prevent pressure transients from exceeding design limits (0.25].

REFERENCE GG, Lesson Plan - B13/21, P. 17 ANSWER 6.17- (2.00)

1. b

2. c

3. d f

4. a [4 @ 0.5 each]

I REFERENCE GG,-Lesson Plan --P-64, PP. 11, 12 l

' ANSWER' 6.18 (2.00)

a. ESF. Bus 15AA l b. Bus 16AB l

l

-c. Bus 25AA

d. Bus 26AB (4 @ 0.5 ea = 2.0) l- REFERENCE l GG, Lesson Plan - P-75, P. 4 l

i

n.

6. PLANT SYSTEMS DESIGN. CONTROL.~AND INSTRUMENTATION PAGE 38

-- " ANSWERS 1'--- GRAND GULF.1. -86/09/08-SPENCER, M.

.. ANSWER .6.19 (2.00)

a. 1. D/G #12

2. ESF X-former #11

-3. ESF X-former #12

'4. ESF X-former #21

b. 1. D/G #13

2. ESF X-former #11 1 -3. ESF X-former #12

4. ESF X-former #21 [8 @ 0.25 each]

; REFERENCE GG,

. Lesson Plan - R21,.P. 4 k

e i

l 3

i' 1

7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 39 '

EADIOLOGICAL CONTROL ANSWERS - GRAND GULF'l -86/09/08-SPENCER, M.

ANSWER 7.01 (1.00)

MASTER COPY To limit Feedwater Nossle temperature transients.

REFERENCE.

GGNS: SOI-04-1-01-G33-1, page 2 ANSWER 7.02 (1.50)

1. Place the RPS (Div 1, 2, 3, & 4) CRD Discharge Volume HI Trip Bypass Switches in the BYPASS position.

2. Place the RPS (Div 1, 2, 3, & 4) Scram Reset Switches in the RESET position and verify that the scram resets.

3. Allow the HCU's to recharge, then drive the control rods not-full-in to position 00. (0.5 each)

REFERENCE GGNS: ONEP-05-1-02-I-1, p3 ANSWER 7.03 (1.00) 4 d

REFERENCE GGNS: ONEP-05-1-II-1, p4 ANSWER 7.04 (1.00) b REFERENCE GGNS: 05-5-01-EP-2, p6 f

l l

l VAsiiR COPY

~ --

r--

c

7 '. PROCEDURES -' NORMAL. ABNORMAL. EMERGENCY-AND PAGE 40'

.R&DIOLOGICAL CONTROL ANSWERS -- GRAND GULF.1 -86/09/08-SPENCER, M.-

ANSWER 7.05~ (1.00)

b.

REFERENCE EIH: HNP-2-1946

'-GGNS: ONEP-05-1-02-V-5, p 2 i-

?

ANSWER 7.06 (2.50)

1. Drywell Pressure > 1.23 psig (+-0 psig)

2. Drywell Temperature > 135 deg F. (+-0.deg )

3. Supression Pool Temperature > 95 deg F (+-0 deg-)

4. Suppression Pool Level > 18.81 ft (+-0 ft )

~5. Suppression Pool ~ Level < 18.34 ft (+-0 ft - )

6. Containment Temperature > 90 deg F (+-0 deg ) (5 @ 0.5 each)

REFERENCE

_GGNS: EP-3, p 1-(2.50)

L ' ANSWER 7.07

a. (1) 14.5 ' feet i

(2)-212 deg F t

! (3) 140 deg F (0.5 each) l b. To. ensure adequate NPSH for the respective ECCS Pumps (1.0)

REFERENCE.

GGNS: EP-3, p 6; EP-5, p 2; EP-7, p1 2

a

.e<,e --

,,,--,-+-,ew--...=ve,1.,~1i.,4-,g-rer,*,+,ee+-r~++,---r--,-- -,,..weiw--- ~2,-=w-,+-- em = n w-- -rp -+ -w- v-4f w ef

W 7 ~. PROCEDURES - NORMAL.' ABNORMAL EMERGENCY AND PAGE 41~

. '., RADIOLOGICAL COHIBQL

ANSWERS -- GRAND GULF ~1 -86/09/08-SPENCER, M.

ANSWER 7.08 (1.00)-

a REFERENCE GGNS: ONEP-1-02-I-4, p 5 ANSWER 7.09 (1.00)

1. Place the Mode Switch ~in Shutdown (0.25 for Scram.the Reactor).

- 2. Close all-Group I Isolation Valves (0.5 each).

' REFERENCE GGNS: ONEP 05-1-02-I3, Section 4.3.1, page 2 ANSWER 7.10 (1.50)

a. 25 REM.

b. 75 REM

c. When situation is life threatening and time does not permit prior approval. [3 @ 0.5 each]

REFERENCE o GG, 01-5-08-2, P. IS,.PP. 6-3.5 .

r l

i' I

l l

7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 42 RADIOLOGICAL CONTROL ANSWERS -- GRAND GULF 1 -86/09/08-SPENCER, M.

ANSWER 7.11 (2.00)

1. No more than one fuel bundle should be suspended above the fuel storage array and this at a height no greater than 6 feet for a prolonged period of time.

2. Damaged or dropped fuel, either new or irradiated, shall not be moved until all factors have been studied and the situation fully analyzed to preclude further damage or contamination.

3. A fuel array of up to three fuel bundles outside of a normal storage area or outside a normal shipping container should be maintained with an edge-to-edge spacing of 12 inches or more from all other fuel.

4. A fuel array of four or more fuel bundles outside of the normal fuel storage areas or properly designed fuel shipping container is prohibited.

5. No more than two fuel bundles should be allowed in or around a fuel prep machine at any time. This fuel should be separated from the main body of stored fuel by at least 12 inches.

6. Load handling over the fuel storage areas is to be limited to one fuel assembly or weight equivalent per crane unless the fuel storage area is empty. An exception to this requirement is a properly designed fuel shipping container suspended over the fuel shipping container storage area.

[any 4 @ 0.5 each]

i i REFERENCE

'GG, 09-S-02-100, P. 3, PP 63

! ANSWER 7.12 ( .50)

Less than 150 psig ( + / - 5 psig )

REFERENCE GG, 04-1-01-P53-1, P. 9

~

-7._. PROCEDURES - NORMAL'. ABNORMAL. EMERGENCY AND -

PAGE _43-

.,, - RADIOLOGICAL CONTROL ANSWERS -- GRAND GULF 1 -86/09/08-SPENCER, M.

ANSWER 17.13 (1.00)'

Drywell'= above 330 Deg. F Containment : above 185 Deg. F [2 0 0.5 cach]

REFERENCE

~GG, 05-S-01'-EP5, P._1, PP. 2.1 ANSWER 7.14 (2.00)

1. a

2. c

3. d

4. b [4 @ 0.5 each]

REFERENCE GG, 10-S-01-4, 10-S-01-2, 10-S-01-3 and 10-S-01-05

7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 44

. RADIOLOGICAL CONTROL s- .

ANSWERS -- GRAND GULF 1 -86/09/08-SPENCER, M.

ANSWER 7.15 (2.00)

1. If any of the monitored parameters in Section 2.0 exceed the setpoint given without an automatic scram, manually scram the Reactor.

2. Arm and depress all four MANUAL SCRAM pushbuttons, RPS Div. 1, 2, 3-,

4 on Panel 1H13-P680.

3. Place the Mode Switch to SHUTDOWN

4. Verify all control rods are fully inserted by observing the full-in green light for each rod is displayed. Also, OD-7 may be used, indicating Position 00. Also verify the scram discharge volume vent and drain valves closed.

a. If any control rods have position indications greater than 06 and reactor power greater than 5% on APRM's, enter EP-10, Reactivity Control.

b. If any control rods are not inserted, and EP-10 was not entered, perform the following:

1) Place the RPS Div. 1, 2, 3, and 4 CRD DISCH VOL HI TRIP BYP switches to the BYPASS position.

2) Place the RPS (Div. 1, 2, 3, and 4) SCRAM RESET switches to reset; verify scram reset.

3) Allow HCU's to recharge, then drive the control rods that are not full-in to Position 00.

5. Select the six SRM's and the eight IRM's and DRIVE IN, and switch IRM/APRM LVL recorders to IRM's.

6. Verify power decreasing as indicated on the nuclear instrumentation.

7. If two reactor Feed Pumps are operating and Reactor water level is being maintained, trip one of the pumps.

8. Verify that the Recirculation Pumps have shifted to low speed, or shift them manually.

9. Verify Reactor water level stabilizes near or above + 18".

10. Verify that the Turbine and Generator trips on reverse power (15 seconds time delay, 5 seconds time delay if the turbine has already tripped), or trip manually. Refer to ONEP 05-1-02-I-2, Turbine and Generator Trips.

11. If an MSIV closure occurs, enter EP-1, Level Control.

12. Verify reactor pressure is maintained with either the bypass valves or relief valves.

r1

7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 45 3,:

.

• RADIOLOGICAL ~ CONTROL ANSWERS ~--' GRAND GULF 1. --86/09/08-SPENCER, M.

[any 8 @ 0.25 each]

REFERENCE-GG, 05-1-02-I-1 ANSWER 7.16 (2.00)

.1 . Place the Reactor Mode Switch in Shutdown.

L 2. Verify all control rods are fully inserted.

> 3. Reset the scram if time permits.

4. Manually initiate the RCIC System if time permits. [4 @ 0.5 each]

REFERENCE

.GG, 05-1-02-II-1 ANSWER 7.17 (1.00)

. -The hydrogen temperature will decrease during coastdown. The hydrogen

. pressure regulator will maintain pressure at 60 psig. Upon restart of 4

turbine generator the hydrogen will expand and overpressurize the generator casing.

l REFERENCE GG, 05-01-02-I-2, P. 8 c.

t i

ANSWER 7.18 (1.00)

' 1. Attempt to close. [0.25]

l 2. If any stuck open SRV cannot be CLOSED within approx. 2 min [0.25]

OR if any SRV is stuck open and suppression pool average temperature reaches 105 Deg. F [0.25] then scram the reactor. [0.25]

REFERENCE GG, 05-S-01-EP-3 l

b I

1.

'8. ' ADMINISTRATIVE' PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 46

. a t ANSWERS -- GRAND GULF 1

-86/09/08-SPENCER, M.

MASTER COPY ANSWER 8.01 (1.50)

1. Continuous visual indication in the control room

2. One of the required SRM detectors located in the quadrant where CORE ALTERATIONS are being performed and the other required SRM detector located in an adjacent quadrant

3. Unless adequate shutdown margin has been demonstrated, the shorting

-links shall.be removed from the RPS circuitry prior to and during the time any control rod is withdrawn. [3 @ 0.5 each]

REFERENCE GG, T.S., Page 3/4 9-3, PP. 3.9.2 ANSWER 8.02 (2.75)

Secondary Containment Integrity shall exist when:

a. All Auxiliary Building and Enclosure Building penetrations required to be closed during accident conditions are either: [0.25]

-1. Capable of being closed by an OPERABLE secondary containment automatic isolation system, or [0.25]

2. Closed by at least one manual valve, blind flange, rupture disc or deactivated automatic valve or damper, as applicable, secured in its closed position, except as provided in Table 3.6.6.2-1 of Specification 3.6.6.2. [0.25]

b. All Auxiliary Building and Enclosure Building equipment hatches and blowout panels are closed and sealed. [0.5]

c. The standby gas treatment system is OPERABLE pursuant to

! Specification 3.6.6.3. [0.5]

d. The door in each access to the Auxiliary Building and Enclosure Building is closed, except for normal entry and exit. [0.5]

e. The sealing mechanism associated with each Auxiliary Building and Enclosure Building penetration, e.g., welds, bellows or 0-rings, is OPERABLE. [0.5]

' REFERENCE GG, Technical Specification, P. 1-7, PP. 1.38 NASER CPY

8. ~ ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE~ 47

?5 ' ANSWERS - GRAND GULF ^1 -86/09/08-SPENCER, M.

ANSWER ~ 8.03- (2.00)

Idle loop to. operating loop undue. stress on. vessel nozzles and bottom head region. [1'.0]

Idle loop'to vessel thermal shock to recirculation pump and recirculation nozzles [1.0]

REFERENCE GG, Technical Specifications, P B 3/4 4-1 ANSWER 8.04 (1.00) a REFERENCE GGNS: TS 3.2.2, page 3/4 2-3 ANSWER 8.05 (1.00) c

-REFERENCE GGNS: TS DEFINITION 1.6 ANSWER 8.06 (2.00)

'1. Reactor Water Level maintained > TAF.

2. Core being sprayed.by HPCS or LPCS.

3. -Reflooding flow of 1 LPCI pump' injecting into.the core with reac-tor water level high enough to produce 2 phase flow through the core.

4. Sufficient steam flow through the core. [4 @ 0.5 ea = 2.0]

REFERENCE GGNS: Procedure 01-S-06-2, page 5

- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS;'AND LIMITATIONS PAGE 48 N.INWERS--GRANDGULF1- -86/09/08-SPENCER, M.

' ANSWER.

8.07 (1.00)

The Shift Supervisor shall perform a verification (by test or

-inspection) of the operability of redundant. safety-related systems.or components.

)

REFERENCE GGdS: Procedure 01-S-06-2, item i, page 15

. ANSWER ;8.08 (1.50) ,

1) One-rod-out-2)- Refuel Platform Position

3) ' Refuel Platform Main Holst Fuel-loaded (Oi.5 each)'

REFERENCE' GGNS: TS 3/4.9.1

ANSWER 8.09 (1.00)

! 'Decause the dissolved oxygen content of the reactor coolant is-t typically higher during low steaming rates (e.g. Startup or Hot Standby)- (1.0) l

[

REFERENCE l

-EIH: U1 TS's, 3.6-6 GGNS: .TS 3/4.4.4, Table 3.4.4-1 l

l-l i

i l

t _

8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE- 49

' ANSWERS -- GRAND GULF 1 -86/09/08-SPENCER, M.

J T

ANSWER 8.10 (3.00)

c. 25%

b. 785 psig

c. 10%

d. 1.06

o. 1325 psig

f. the top of the active irradiated fuel (6 at 0.5 each)

REFERENCE Grand Gulf, Technical Specification, page 2-1, pp 2.1.1-4 ANSWER 8.11 (3.00)

a. False

b. False

c. False i

d. True

e. False

f. True (6 at 0.5 each)

REFERENCE Grand Gulf, Technical Specifications, section 6.2.2

'8.

ADMINISTRATIVE PROCEDURES.' CONDITIONS. AND LIMITATIONS PAGE 50

. . . .EANSWERS'--' GRAND. GULF-1 -86/09/08-SPENCER, M.

' ANSWER 8.12. (2.00)

e. 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />

b. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />  ;

c. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />

'd.

Accept any of the following; GGNS General Manager, or his designee, or.a higher levels of management, in accordance with established

procedures (4 at 0.5 each)

REFERENCE Grand Gulf, Technical Specifications, page 6-2.

i l

l'

r a

8. ADMINISTRATIVE PROCEDURES. CONDITIONS, AND LIMITATIONS PAGE 51

? NSWERS -- GRAND GULF 1 -86/09/08-SPENCER, M.

ANSWER 8.13 (3.00)

1. Failure of PS or other system subject to LSSSs to initiate required protective nction when monitored parameter reaches setpoint or failure to c plete required protective function.

2. Operation of un t or systems when any parameter or operation subject to LCO less cono rvative than established in T.S.

3. Abnormal degradati n discovered in fuel cladding, coolant pressure boundary, or primar containment.

4. Reactivity anomalics reater than or equal to 1 % of delta k/k.

5. Shutdown margin less co.servation than specified in T.S.

6. Short-term reactivity inc cases that correspond to reactor period less than five seconds or,

6. If suberitical, unplanned r activity insertion greater than 0.5 %

delta k/k; or any unplanned riticality.

7. Failure / malfunction of one or ore comonents which prevents or could prevent fulfillment of functio l requirements of system (s) used for accidents analyzed in SAR.

8. Personnel error or procedureal in dequacy which prevents or could prevent fulfillment of functional equirements of systems required for accidents analyzed in SAR.

9. Conditions arising from natural or ma -made events that require unit shutdown, operation of safety systems, or other protective measures required by T.S.

10. Errors discovered in transient / accident - alyses or in methods used for analyses as described in SAR or T.S. ses that have/could permit operation less conservative than assumed i analyses.

l1. Performance of structures, systems, or compo ents requiring remedial action or corrective measures to prevent oper tion less conservative than assumed in accident analyses in SAR or T. bases.

12. Discovery of conditions not specifically conside ed in SAR or T.S.

that require action or corrective measures to pre ent an unsafe condition.

13. Offsite releases of radioactive materials (liquid a gaseous) which exceed T.S. limits.

See a4+ah L p y 1

p.grJ ATTACHli3NT XVII Page 1 of 2 j ,' J id#! GUIS NUCLEAR STATION ADMINISTRATIVE PROCEDURE rs q./

? l 01-S-06-5 [ Revision 14l

. k gp r)f h/ev Nf.V i bP l Attachment IIIlPage 1 of 6 l Page of _

[, d NRC NOTIFICATION REQUIREMENTS I. Immediate Notification (within one hour of occurrence) ,

I 9

1. The declaration of any of the emergency classes specified in the ,*

licensee's approved emergency plan. 50.72(a)(1)(i) j

a. The notification.will normally be made and documented in accordance with 10-S-01-6.

2. The initiation of any nuclear plant shutdown required by the Plant's Technical Specifications. 50.72(b)(1)(1)(A) (See_4ttachment V Page 1 for

' examples)~

3. Any deviation from the Plant's Technical specifications authorized pursuant to 10 CFR 50.54(x). 50.72(b)(1)(1)(B)

4. Any event or condition during operation that results in the condition of ..

the plant, including its principal safety barriers, being seriously ..

degraded; or resul.ts in the plant being: ,50.72(b)(1)(li) (See Attachment c V Page 4 for examples) .-

a. In an unanalyzed condition that significantly ccmpromises plant '

safety. 50.72(b)(1)(ii)(A)  ; -

b. In a condition that is- outside the design; basis of the plant; or 50.72(b)(1)(ii)(B) .s. ,

c. In a condition not covered by the plant's operating and emergency procedures. 50.72(a)(1)(ii)(c)

5. Any natural phenomenon or other external condition that poses an actual threat to the safety of the plant or significantly hampers site personnel in the performance of duties necessary for the ;sfe operation of the plant. 50.72(b)(1)(iii) (See Attachment V Page 6 for examples.)

6. Any event that results or should have resulted in Emergency Core Cooling System (ECCS) discharge into the reactor coolant system as a result of valid signal. 50.72(b)(1)(iv) (See Attachment V Page 8 for examples)

7. Any event that results in a major loss of emergency assessment capability, offsite response capability, or communications capability (e.g. ,

significant portion of Control Room indication, emergency notification system, or offsite notification system). 50.72(b)(1)(v) (See Attachment V Page 18 for discussion).

8. Any event that poses an actual threat to the safety of the plant or significantly hampers site personnel in the performance of duties necessary for the safe operation of the Plant including fires, toxic gas releases or radioactive releases. 50.72(b)(1)(vi) (See Attachment V page 17 for examples)

J 01-S-06-5 ATT III

ATTACllMENT XVII Page 2 of 2

\.M i GULF NUCLEAR STATION ADMINISTRATIVE PROCEDURE hi l 01-S-06-5 l Revision 14l l Attachment IIIlPage 2 of 6 l Q'

m. ,

Page _ of NRC NOTIFICATION REQUIREMENTS

/

9. Any Licensed Material event which causes.or threatens to cause exposure

, of the whole body of any individual to 25 rems or more of radiation; .

exposure of the skin of the whole body of any individual of 150 rems or more; or exposure of the extremities of any individual to 375 rems or more of radiation. 10 CFR 20.,403 (a)(1)

10. Any Licensed Material event which caused or threatens to cause loss of one working week or more of the operation of any facilities affected.

-10.CFR 20.403 (a)(3) - - -

11. Any Licensed Material event which caused or threatens to cause damage to property in excess of $200,000. 10 CFR 20.403(a)(4)

12. Immediately after determining the loss or theft of licensed material in such quantities and under such circumstances that it appears that a -

substantial hazard may result to persons in unrestricted areas.

10 CFR.20.402(a)(1) = ,. _ _ , .

II. The NRC regional office (Region'II) shall immediately (l' hour) be notified by -

telephone and telegraph, mailgram or facsimile of the following occurrences:

(10 CFR 20.205(b)(2) & (c)(2))

1. The detection of removable radioactive contamination in excess of 0.01 microcuries (22,000 DPM) per 100 square centimeters of packages surface on the external surfaces of the package, or

2. If radiation levels are found on the external surface of the package in excess of 200 millirem per hour, or at three feet from the external surface f of the package in excess of 10 millirem per hour.

i l NOTE l i I i

l The final delivering carrier must also be notified immediately of this event.l (I

III. Four-hour reports (within four hours of occurrence).

1. Any event, found while the reactor is shutdown, that had it been found while the reactor was in operation, would have resulted in the plant, including its principal safety barriers, being seriously degraded or being in an unanalyzed condition that significantly compromises plant safety.

50.72(b)(2)(1) (See Attachment V Page 4 of examples)

2. Any event or condition that results in manual or automatic actuation of any engineered safety feature (ESF), including the Reactor Protection System (RPS). However, actuation of an ESF, including the RPS, that results from, and is part of, the preplanned sequence during testing or reactor operation need not be reported. 50.72(b)(2)(ii) (See Attach = cat V Page 8 for examples) 01-S-06-5 A'IT III

-8! ' ADMINISTRATIVE' PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 52 i ANSWERS.-- GRAND GULF 1 -86/09/08-SPENCER, M.

[ Exceeding T.S. limitsifor radioactive materials in tanks.

(any 6 @ 0.5 ea)

REFERENCE

-Grand Gulf, I.;;;en Plan, OF TS 500 A Lmio.h+rets'v< fre e sluvs 0l- S

• 0 C,"5. s AW*b"~' 5

' ANSWER 8.14 (1.00) d

. REFERENCE GGNS TSs 3. 3.1' & 3. 3. 2 '

ANSWER 8.15 (1.00) d REFERENCE ,,,

GGNS TSs 3.0.3 & 3.6.7.1" ANSWER 8.16- (l'00)

-a REFERENCE j~GGNS: TS 3/4.9.10.2 1

i I

1

TEST CROSS REFERENCE PAGE 1 QUEETIOli VALUE REFERENCE 05.01 1.00 KZS0001695 05.02 1.00 KZS0001696 05.03 1.50 KZS0001697 05.04 1.00 KZS0001698 05.05 1.50 KZS0001699 05.06 1.50 '.KZS0001700 05.07 1.00 KZS0001701 05.08 1.00 KZS0001702 05.09 1.00 KZS0001703 05.10 1.00 KZS0001704 -

05.11 2.00 KZS0001705 05.12 1.50 KZS0001706 05.13 .50 KZS0001707 05.14 .50 KZS0001708 05.15 1.00 KZS0001709 05.16 1.50 KZS0001710 05.17 1.00 KZS0001711 05.18 3.00 KZS0001712 05.19 1.00 KZS0001713 05.20 1.50 KZS0001763 25.00 06.01 1.00 KZS0001714 06.02 1.50 KZS0001715 06.03 2.00 KZS0001716 06.04 1.00 KZS0001717 06.05 1.50 KZS0001718 06.06 1.00 KZS0001719 06.07 2.00 KZS0001720 06.08 .50 KZS0001721 06.09 1.00 KZS0001722 06.10 1.00 KZS0001723 06.11 1.50 KZS0001724 06.12 2.00 KZS0001725 06.13 2.00 KZS0001720 06.14 1.00 KZS0001727 06.15 1.50 KZS0001728 06.16 1.25 KZS0001729 06.17 2.00 KZS0001730 06.18 2.00 KZS0001731 06.19 2.00 KZS0001732 27.75 07.01 1.00 KZS0001694 07.02 1.50 KZS0001733 07.03 1.00 KZS0001734 07.04 1.00 KZS0001735 07.05 1.00 KZS0001736

=

TEST CROSS REFERENCE PAGE 2 QUESTIQN VALUE. REFERENCE .

___--?_- _-_-_. _______.--

07.'06 2.50 KZS0001737 07.07 2.50 KZS0001738 07.08 1.00 KZS0001739 07.09 1.00 KZS0001740 07.10 1.50 KZS0001741 07.11 2.00 KZS0001742 07.12 .50 KZS0001743 07.13 1.00 KZS0001744 07.14 2.00 KZS0001745 07.15 2.00 KZS0001746 07.16 2.00 KZS0001747 07.17 1.00 KZS0001748 07.16 1.00 KZS0001749 25.50-08.01 1.50 KZS0001750

• 9.02 2.75 KZS0001751 e8.03 2.00 KZS0001752 08.04 1.00 KZS0001753 08.05 1.00 KZS0001754 08.06 2.00 KZS0001755 08.07 1.00 KZS0001756 08.08 1.50 KZS0001757 08.09 1.00 KZS0001758 08.10 3.00 KZS0001759 08.11 3.00 KZS0001760 08.12 2.00 KZS0001761 08.13 3.00 KZS0001762 08.14 1.00 KZS0001764 08.15 1.00 KZS0001765 08.16 1.00 KZS0001766 27.75 M_mm__

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10. When you conplete yeu- exaninationr /cv shall*

a. Assenble your exaninct:or as #ollowc:

(1, E:: a n quest:ans on top.

(2) E:: c m cids - figures, tzbles, etc.

(3) Answer p:43es including ":gures w h i c :' cre part of the answer,

b. Turn in your copy of the ce: e n i n a t i o n and all pages used to answer the exanination questions,

c. 'vrn in all scrap paper end the balance of the paper that you did c: o t use for answerina the e

cuestions,

d. Leave the enanination ::r e a r es defined by the e :: a n i n e r . If after leavina, w +/ o u are found in this area while the e: am tnation is still in pro 3ress, yOUT licen2e nay be denied or revoked.

f

4

1. PRIMCIPLES OF NUC! E AP 00WER PLANT OPERATION, FAGE 2

~~~~ ~

THE5506Y5I5555I~C j~Yk555E55~550 E[U5D'EEUW GUESTION 1.01 (2.00)

What are four (4) nechanisms that may cause the reactivity of the core to decrease during reactor operation? S e c c i f :. c elements or icottoes e 4 are NOT required. '2,00)

GUEbT10N 1.02 (2.50)

Cor each of the r ollowin;3 events , STATE WHICH coeff icient of reactivity would cet '~IRST to chen3e reertavity and whether POSITI','E or NEGATIVE reactivit,y is sdded due to that coefficient,

a. Control od drop et power (0.50)

b. SRV o~ enirm < ~

at ,o o w e r (0.5))

c. Loss of shutdeun cooling when shutdown (0.G3)

d. Ma:n turbine t r '. o s while at 3 0 '. Power (0.50)

e. Loss of one f e edu t:t e r heate" (ext cetion s t e e re . sc.sted) (0.50)

GUESTION 1.03 (' 50)

Etcte whs'her control rod worth INCREA23 OECREASESr ;r is u N A r E C T E D ss pl:nt conditions ch cq u ' - ' ' - am cold to hot : t 1 *. ,o o w e * . JUST7Y your answer. (1,50) ev.n_ raiT .7 0 'M t . . n,

<c.on, b '.3 e 2 . '

h lo5 2nd bb[ tbh nunOOT C[ C O f'i k r c 1 ICC1 IU {'1 red IO 00 W i t h d *' ' W ' te l' C h 1 U V e C r i t i c i . t y .

>j n i. c a*

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e.,,,i,,.,

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E -

V.s. H .h A,

., A, ,! O, Y I

.h,4 (m![ [r T .)

+

'( O lv; 1 [.

. ma- (T1< _ .a u . .- C_ r4 i _ . ..

.P'

• A tt e )tlt-O't cV r;. , -pAN et - [ *, A tts*

t '. ' s ;

s-*3O -

. j [i u p [p . 3 e

s $$ sd.  ; .* C, i.s m -.

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[ T, 9 l 'V ") p 3

s4 v, ., . v ,1 - m us ,

g CR2 = o ep: . .:6 -- 300 cp-i rom =.

o c, co' g .

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l 1 0 0 '. roc deici;y, I

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%u roEw,0 . - en m on m' o-l l

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4

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4mpn LUTD e t_ n- w GUESTION 1.05 (1.50) ,

1

, T. n u r_ re o c. . ,ea. r. .

For a constant reactor period, it takes the sa.e 3.iount of tiee to chenee extcr pcwer f r c ': l' to 5 ?. as it doer to change it from t A, */

. n

+n

, r n */. e

<. 7TDi/i ?T d.yi J W 'F . 'i . (14 ,ay': A )

Qt_lCCT.TGhi

_s u 1 . 0 2,. (, _7 . A,, 0 ,'s Your reactor has ,, u s t scrammed "non ertended full oever coeration, 4 .

Ten (10) h o u r ". la:ter coo' dow is completer and the reactor :. s neasured .

at tha+ time to be shutdown b v, 1 ". dk/'.. Describe t h r.- chances, if a n "> 7 to 4 the Teettivity by wh:ch the reactor is snu down cve- the next 20 h o u r s..

Include a r.y adverse conditions !at may occur n c. result. (2.00) 4 I

4 rssU r_ c.', T ". n. ", 1. . n. 7.- (*.

.. . C o. ',

'~

4

3. O ,.atn W4Y tiie uel Temperature Coefficient : Doppler Coefficient) ry p.rre.

. o s.,_- t,u,-. , s. e

. r o'e t.4 or.- '. , C x, A. S r ,4. . / .T N r '.. '. G_ r_'

4

. o. y o o. r _v,,e"_ A T_ tJ ; "A 0 N Whether i t beco es MCOE or LESS n e oaativ2).

(*. n.on' u )

O vi l t ev2v T 0 F. .

..nou s ( 1. c; sO. 3 .

The P vfa c t C r .

^

..en *0 C91TICA'ITY f r O iD 2 C0ld CJ9ditlOn 2nd

., 00 00 cecoic 00CITIVE perio_ _ m t t a i r: e r' .

2. om control rocs nuclear i ri s t e u ro.r. t .r. t : a n i how can the cperCtOr t.

• l l W h t- n the lecting r:nac F 1. Sc 2n r e r.c h e d ?

(pod PCCit100 1hd r9CirCV11t10n 3ra held C;f ' tant.) , (0.50)

b. In Wh:Ch Of the ID. LOW 1h3 intcPVal Wa: the hectin: r w s re: ~ e dntered' '1.00)

(1) Intervel -

r< actor power i n r. o . ed by 2 +'acter '

6 in '.43.3 seco',d , -

r (2) Intervel 2 - reactcr power i n e r s c u : c.' by e factor :

,e 7,.; ii a C t'., ,

_-_-, C O .;a ,_

(3T IntOTVO. 'd O?c101 power i n O T- 6 "d Oy i' E C t O! C c

.; . n . .sc. n . ,u O C C O r.st -

(***)4 C A I E C, O R O '. C O N T I f U E .. [' Ni_~ X - sACL r*et* '

. . . ~ . . . . . . _. . - . . ~ - -

f

-; 2 t .

d 4 f 4 .

4 1: ,

1-j' 1 1. . PRINCIPLES OF NUCLEAR.*GWER PLANT OPERATION, PAGE -4  !

' ~ ~

~~~'T55EEd6YU552C57~55dT T5I55E55~5s6~f[UfD EEU~  !

. ______________________________________.._____ q I

4 1

GUESTION 1.09 (1.00) i I The reactor trips f' r om full power, equilibrium .:enon conditions. Four (4) hourc leter the "ocetor is brought critical and power level is main--  ;

tairied on range 5 of the IRMs for several hours. Which of the'following statements is CORRECT concerning' control rod motion during this period?

, (1.00) l 4 .

c. Rods u 'I have to be withdrawn due to xenon build-in. ,

f b. Rods will have to'be rapidly inserted since the critical reactor

] will cause a high rate af xenon burnout.

c. Roos will have to be inserted since :enon will. closely follow

{ its norn.a1 decay' rate.

1

d. Rods will apr>tonimately remain as is as the nenon establishes.

I

ts equilibrium value for this power level.  !

i 4

1 l GUESTION 1.10 (2.25) a Assume the reactor is operating at 100% power and one recirculation  ;

pump trips. Indiccte how each listed indiceted parameter w;uld FIRST 4 CHANGE-(Increase or Decrease) and E:<CLAIN why the change' occurs.

I-l a'. reactor power (0.75)  !

l ~b.. indicated reactor water level (0.75) j .

i

e. feeducter flow (0.75) l

, i f

(v m t entec 0RY 01 CON'IEED ON NEXT On0E anu '

I v

h 4

5 i

t

.+-ervv-+_ s

_ . -. .- . _ ~ .- -- ..

t i

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION > PAGE 5 THERMDDYNAMICS, HEAT TRANSFER AND CL'JID FLOW 44 e, n

aUw o c ee t r n.a A' 2sI 4. A . s e nni s v/

l

'4 Match ene of the following numbered items with each of the l four lettered statements. A letter-number sequence is

!' sufficient. (2.00)

, ruc_ .y .._ c.rmr.rAer,-

.. r.

a .

i 2. APLHCP. 6. MCPR

3. CPR 7. TOTAL c' F -

c1. F t_ c. m_ u. i or .

a. P a r .n e t e r by which plastic strain 2nd deformation are limited to lesc then 1%.

b. Ratio of bencle power required to produce onset of transition boiling somewhere in the bundle to actuci bundle power.

L

i. c. Paraneter by which p .;.? a k. clad temperature is Ti c i n t a i n e d less then 2200 degrees F du-ins postulsted design basis i accident.

i

d. Contains guidelines restricting ,souer ramp ratec above the t

th"c: hold power.

OUESTION 1.12 (1.00) fi i Which of the following are the incica+ ions of a CINT::1FUG AL punp in c RUNDUT condition') (1\ Jk&

b. High power consumption, law 3 charge head. nd low eleu.

l c. Nigh power contugtton, high discharge head, and low flo; .

i

! d. 'lich pou n r ensumption, .au discharge heac. and high flow, i

.l i

i i

i 1

- 'i -

j I

l r

1

, 1 4

.---m,y..,. -,,-,ry..,_,---,,-.w,,..m..-,,-w--y. -+,--~.-..,,,_e-,.c.,cr--,-.,.---.y-..,,e.,- ...-----%------- - - - - - , . , - * - , - %w. , - . - -- ~ . . - -

-. . - .- - = . - - . ..

.. .. 9 ,

t

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 6

~

~~~~T55RE66~5555CE~iI5ET TRd55EEE~556~EL6ED~EEEE i

+

QUEGTION 1.13 (1,00)

) During a REACTOR STARTUPr the reactor operator is sequentially

{ withdrawing control rods te m a k.e the reactor critical. Which of the following statements correctly describes the SRM's response.

(1.00) 1 1

l a. It u:ll teke a shortcr length o f' time for the SRM count rate tc

slowly increase to a new stable ccunt rate.

I,- b. It W: 11 take the same encuit of time for the SRM i count rate to slowly increase to a new stable count 1

rate.

2 No change will be observed in the SRM count rate until the reactor is criticel.

i 7

d. It will take 3 longer length of time for the SRM count rate

to slowly increase to a new steble count rate.

4 i QUESTION 1.14 (3.00) i Attached FIGURE 6 2 illustrates a transient that could occur at j Grend Gulf One.

! GI'J EN : (1) The reactor is init: ally operat:rg at 100% Rated Thermel ra avgp

, (2) R e a t-t o r ceed Pumps trip at time 0 j (3) No operctor action: are taken

~XPLAIN the cause ci the following:

I

a. Why reacter wEte- level be3 i ns INCREASING c#ter j ~30 seconds. (0.50) i i D. Why reactor pour is DECRE ASING be tweer. time i ~0 to '? seconds. (0.50) i

c. Why TOTA _ cteen flow STADILIZEE c'ter N0 seconds, i

i f0.50) 9

. b ICaCEoI [JIeCbuIe C t W C kf I o Geconds and s t a '-' i l l 2 0 s after '30 S e C O n 'I S , (0.50)

{

P. Why the r G- L C l O T CCRAMMED at *7 1+ 0 c o r .

'O 50) i I

:*r m CATEGODS 0 C0"TINUEC 0 T ' AGE u t r .* '

t i

t J

.I f

I

- ... , ~ ~ ._ ,- - . - , . . _ _ . _ . ~ _ _ , . - - , . ~ . _ . - . . . - , . _ _ . _ _ _ _ ,

7

. - , . - = .- _- - . -- . -- .-. . .. . ... . - = - . . .

a.

e 4 ,. . 4.

E 4

i

f. L. : P r(IN CIP L E'3 0F' NUCLEAR c'0WER PL ANT OPERATION, PAGE 7

~~~iHEEEOD 5555C5~~HE5T~TREU5E55~dUE~fEU53~EL 5E i

s

! G'JE3 TION 1.15 (1.50) f List the effect on critical power that esch of the below chanaes would produce. (INCREASE - DECREASE - REMAIN THE SAME) a.. Increase subcocling

b. Decrease mass flow j

c. Increate systen pressure ( 1000 to 1050 psis )

f, OUESTION 1.16 (1.00) 4 i Which of the following actions will INCREASE Grand Gulf Unit One's thermodynanic cycle eff:ciency?

! a. DECTEASING power # rom 100% to 25% ,

i b. LORERING condenser vtcuum from 29' to 25',

4.

c. REMOVING a high pressure FW heater # rom service.

. d. DECREASING the cmount of-condensate d e p r e t. s t o n ,

i QUESTION 1.17 (1.00) ,

Which of the following e q u a t :. o n s is ned to perforn a ?WR reactor heet bclence?

NOTE: c :- C R D ) f-Ceedwater; s2Stean; ai=RWCU in; 'o=RWCU out

a. G-rx =--(w + w i n h + w '

h + 0-rcd -w , h - w / h - w >

h - Q pump r

b. G: 2 (w + u ) h + w , h 4 0-tad - w " ,

- w h - w ,

h - 0 pump I c. G-r, = (w + w ) h + w , 5

• Q p u c.p - w m - w '

t. - - n - G-rad I d. 0-r: = u -

h + w e h + n-rah - (w + w ) >

h - w , S - w >

h - G pump 5

( x *. *

• r EO OF CATEG0;Y 0; umi t

y ..~ ~. - .- - - - - -- . .

1. , =

r r-i 2 .- PLANT DE3IGN' INCLUDING SAFETY.AND EMERGENCY SYSTEMS DAGE 8

'0UESTION '2.01 (1.50)

LISTthree (3) conditions which will START Both-(A and B)

Ctandby Service Water Pumps and INCLUDE SETPOINTS.

, (1.50) f GUESTION 2.02 (3.00) i

. Consider an Off-Normal Event in which Instrument Air System pressere is lost.

a. How will the following velves Fail? (CLOSED 7 OPENr AS IS)

(e.e0s n a >

i ,

1. CRD FCV

' 2. RFP Minimum Flcu Valve i i '3. Feedwater Startup Flow Control Velve

4. Drywell Chillers Tem,oerature Control Valves l 5. TBCW Mcke-up Valve

'. b. EXFlaIN why High Radiation levels in the Off-Cas Building nay I occur. (0.50) i 4 f l ..00ESTION 2.03 (1.00)

} The-plant is operrtins at power with 'A' and ' C CCW pumps running and the*B' CCW pump selected for STANDBY operction. A LOP occurs and the

f. diesels start and tie in normally. Which one of the "ollowins most l accurately deceribec how the CCW cyctem w:11 respond during th: E i transient' j a. The LSS panel will auto start the 'S' CCW pump on ESF power efter tha bus is reene 3.ced,

c. Both .he*A' and 'C' CCW pv=po can be started manually on E3F power efte* the busec are reenergized.

c. The 'B' CCW pump will net auto starte but can be manually stc ted by the opercior on ES~ power afte: the bus is neenergized.  ;

d. The '3' CCW Punp ull; euto a art on a low CCW gressure s;gnal after the ESF but is reenerg1 ed.

(**AAY CATECOR, Q1 ' N N1 I?J U E D ON < E '( 7 -AGE 3 414*)

t-E i.

j . c. . .

i.<

l t-

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 9 i

1

} 'c i

! ~9UESTION' 2.04 (. .50) i FILL IN THE BLANK

DC.11.12 the only powerEscurce svPPlyins ESF Sus 15AA. When j '

paralleline the Normal' Power Supply back to ESF Bus 15AAr the

, synchroscope should be~ turning slowl"> in the ________ direction. -

4 0UESTION 2.05 (2.00) l LIST the DESIGN "ALUES for the fcilowing containment paraneters; t

!. c. . Maximum Dryuell Internal o ressure (0.50)

I i 'b. Maximum EXTERNAL-to-INTERNAL Differential Pressure

! for the DRYWELL (0.50) .

! c. Manmimun Containment Shell Internal Pressure (0.50)

d. Manisum EXTERNAL-to-INTERNAL Differential Oressure i- ror the CONTAINMENT (0.50)

I f

I j QUESTION 2.06 (2.00) l- The Standby Gas Treatment System (S'G7S) J is in operation with an auto initistion signcl present. Euplain the e f f e c t '. s ) on SSGTS T" sin 'A' for the # allowing canditions.

(2.00)

! a. The SBGTS Mode Select Switch is in AUT0; the l l operator places SOGTS Div I MAN INIT RESET SW to

RESET t r, d then returns to NORMAL, then he places .

i ohe ENCL BLDG RECIRC CAN *A' and SSGTS ~ILTER TRAIN j 'A' switchec to stop.

1 j 5. If the SBGTS Mode Select Switch is in STANDEY; '

l and the operater performs the same switch l manipulations sc in part 'a'.

i i

i i

i (**rx* CATEGORv 02 CONTINUED ON NEX' PAGE *T***' ,

a

.h -

)

i 1,.

t i '

i 4

i i  !

1 3

i i  ;

! _-., n - - - - - . __ -

r

5- _

'8

• 9 ..

t t

r 2.- ' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS FACE 10-

'GUESTION 2.07 (1.00)

I Recirculation Hydraulic Power Unit (HPU) 'A' is operating as ,

follows:

(1) SOTH sub1 cops are in the READY mode with subloop 2 selected as the LEAD loop.

(2)- An AUTO transfer to subloop 1 occurs due to a subloop 2

• T A N L O W ' CONDITION.

i . (3) The operstar refills the tank and clears the low level i' condition.

! If subloop 1 trips, what is the final status of HPU 'A' l.

and the flow control valve. (1.00) i GUESTION 2.08 (1.50) >

c. LIST three (3) trip functlant which are net bypassed when the DIVISION I EMERGENCY DIESEL GENERATOR lu

cperatins in the AUTO mode. (0.75)

I j b. LIST three (3) additional trip functions which are 1

Dettve when the DIVISION I EMERGENCY DIESEL GENERATOR ic ,

operat.2ng in the TEST mode. (i.e. surveillance in progress) (0.75) i j QUESTION 2.C9 (3.50) c, Whct c -' e the three (0) differentici temperature interlocks + hat must ce natisfied in the recirculstion pump starting sequenect prior to sta-t:nc e recirculation pump (INCLUDE SETPOINTS)? (1.50)

b. Li t the two (2) inputs into the ATWE recirev1ztion pump trip cirevit (INCLUDE CE~oCIFS). (i.00)

c. How is en ATWS trip daf"erent from I heirculction Punp Trip (RPT)' /Antwer :. n terms cf pump power supplieu chly. . Assume high power ope'atien when the events occur.) (1.00)

( k i

• bk $ '

- , - - . - - - . ~ _ . . _ _ _ ..,.

. ~ . . .. . . . , _ . . _ - . - . - . . . - - . . _ _ _ - . . - -- . . . . . . . . . - . .

t.

i 4

l- .

p b i i

~2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 11 1

QUESTION 2.10 (1.50) a .Can HPCS SSW (div III) be initiated manuclly' If 0c how?

, .b List the auto initiation-signals for SSN "B' ( D I '.' I I ) .

i t-(SETPOINTS NOT REQUIRED)

II QUESTION 2,11 (1.50)

! Stcte the two (2) tri ?s of the RCIC' turbine which cennot be

{ reset in the control room (Include any SETo0INTS).

2 EXPLAIN how to rese the RCIC turbine in either of these conditions.

1 (1.50)

, QUEETION 2.12 (1.00)

, LIST F O !!" (4) conditions that automa;1cally cauce a DIRECT tri?

of the Reector Water Cleanup Punes. (Setpoints NOT required)

QUESTION 2.13 (1.00) L 2

! The plant is oper; ting et 90% power when the operator j notices an INCREASE in the 62 seal pressure an RECIRC PUMP 'A'. The 12 seal pressure for RECIRC PUMP 'A' h2 NOT l incesased. These observations are followed shortly by a l DUMP 'A' SEAL 0TAGING FLOW HIGH/ LOW alarm.

[ List two malfunctions which give this alarm for the 1 s+ed

! conditions.

3 4

4 s .

I i

3 E

4 e

f l'

i 6

1 1

. - .- . ~ _ . _ _ _ ._. _. _ _ . _ . _ _ . _ . _ _ ____ ._ , . . _ . . .

p l

i c: .. '

1 1

.i

) .2.- PLANT DESIGN. INCLUDING'5AFETY ANS EMERGENCY SYSTEMS PAGE 12 i

f I

1 DUESTION' 2.14 (2.00)

Answer.the follau:.n3 with regard to tne RHR system and its various

} mcdes of operation; i

a. Match the following actions, events, or interlock s in Column A l with itc' initiation SETPOINT in Column E'. (1.50) j Column A Column B.

I 1. . Shutdown cooling isolates 125 psis

2. Allows manual operation of the 135 psis LPCI INBOARD INJECTION valve 350 psis

3. Input to ADS 50 psig

!; 400 pcis l

b. Where does the Shutdown Cooling Mode of RHR takes its ,

t suction? Assume normal operation. (0.50) 1 I GUESTION 2.15 (2.00) 1 1 l For eich at the following HPCS initial conditions, indicate the final l petition of the velvec follow n3 an automatic HPCS initiation

• a, CST suction valvo open, Suppression c' col suction valve shut (0.50)

b. . CST cuetion valve shutt Suppression Poc1 suction valve shut (0.50)

c. . CST suction valve shut, Suppression Pool suction valve open (0.50)

d. HPCS test return velvet open (0.50)

?

i-DUE9 TION  ?.16 r: 1.50) ,

3. List which ADS initiation r eru tr ement( s ) is/are NOT bypassed when ADS is manually init:2ted,

b. Which ADS valvea tra ' quipped w th the " low-Icu' set function?

c. How many SRV's can be operated fi or, the Renate Shutdown Panel?

( .K r a *

• END OF CATCGORY 02

• xn * )

---,--n ._ _ - . , . _

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l 3. INSTRUMENTS'AND. CONTROLS oACE 13 1

l 1

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QUESTION 3.01 (1.50) <

i j Indicate et what' reactor water level each of the following

~

j ' action's is directly initiated. If more than one level  ;

! applies, indicate all of the applicable levels,

! (1.50)

I

a. Direct reactor scram l b. Ctandby Gas. Treatment System star ts  !

c. RCIC starts j d.- HPCS injection valvc closes j e. Recirculation pumps transfer to the LFMG

f. Recirevlation flow control valves run back (RFP :n tripped)

QUESTION 3.02 (1.00)

WHEN Total Steam Flow decreases below 45 % r a level program inches is _____ b> ( _____

signc! set at ____ (a) ____

4 .

i (added to/ subtracted " rom) the actual uensed level far the Feed

Water-Level Control System. i

! (1.50)

, t 1

j i QUESTION 3.03 (2.50) 1-

3. A reactor plant startup is in progress. For each of the following

, conditions, stsLe whether or not the RCIC will allow the attempted j .

control rod movement, (NPTE: 'A' SEQUENCE IS SELECTED). (2.50) i, 1. The first control aos withdrawal attenpt is a control rod in j Group 2 (100% rod density).

4 1

{ 2. Croup 1 it "ully withdtcan. Grev 2 is still fully inserted.

A withdrawal attempt is made o r, a control rod in Gravp 3.

• l

! 3. Groupt 1 cnd 2 are 'ully withdtcun. end a control rod in Group 3 1 is :,e l e c t e d . A, attempt is nade to fully withdraw the rod, i

4. Groups 1-4 are fully w 4 ther wn. The nert con cl rod selected [

i ir in Group 7 and :! ne-notch withdrawL1 is a t t e m p t .'d .

i S. Recetor power it N: ene an attempt "

made to withdesu +1e celected control rad three . ctches.  ;

(r**** CATEGORi 03 CONTINUED fh KEXT oAGE *****)

i 1

5 d

)

1 1

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1 -

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1 3 . . .

1.

QUESTION -3.04 (2.50)

For EACH of the"followins conditions, state whetherfa scramr j half-scramr rod blocRr or no action is directly generated. j j '. For conditions that produce.more than one-action, state the  ;

, more'severeJaction (i.e. half-scram ~1s more' severe than a  ;

tod block.3. (2.50) a.. Loss'of one'RPS MG set, Mode Switch in STARTUP/ MOT  ;

' STAND 9Y -

b. Turbine trip at 30% power

c. Inboard Main Steam Isolation Valves (MSIV's)-feil clocad  !

on Main Steam Lines 'A' AND 'C'r Mode Switch in RUN l

s

d. APRM P downscaler Mode. switch in RUN ,

e. Scram discharge volurie level is at 40 sallons, [

Mode switch in STARTUP t

. QUESTION- 3.05 (3~.00)  ;

j. s. Which cf the following provides the signal for a

' SCRAM due to a Turbine Control Valve (TCV) Fast Closure.

(0.50) l'. TCV position limit switches

2. 9 ate.of TCV position change a

'3 . Power to the TCV fast seting solenoids

4. ETS oil pressure at - '

TCV  :

b. When'is c TCV Fast Closut initiated (irclude setpoints, bypasse2r parameters sensed)' (1.50)

c. How will the TCV fail (OPENr EMUT, or AS IS) en: (1.00) 1, Loss o" electrical signal to the servo talve coils I

2. Lett of Low Pres:ure Contrel Fluid OUESTION 3.06 'i.00) i i

. List fcur (4) systens that have ccipenents that can be operated  ;

or controlled from the Reacte Shutr'own Eystem control '

panels. '1,00)

(**vrr CATEc0tY 03 CONTINUED nN NEYT PACE v r.v r * )

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3 .- INSTRUMENTS AND CONTROLS PAGE 15 1

1. 9UESTION 3.07 11.00):

a

Which of.the following INDICATIDNS would you expect to see '

! as a result'of a ' Jet Pump Riser Failure'? '(1.00) t .

l e.- A DECREASE in failed det Pump flowi'an IRCREASE in f indicated Core Flowr and a DECREASE in Core

Differential' Pressure.

b. .A DECREASE in: core differential pressure > an INCREASE in Reactor Powerr. cnd an INCREASE in  !

Indicat'ed Core Flow.- l

t I c. A DECREASE .irc reactor .-( APRM) powerr a DECREASt in

[~ the Failed Jet, Pump Flow, and a DECREASE in Core "

l ~ Differential Pressure.

i d. .An.' INCREASE in indicated core-flow, an INCREASE in

[ Failed det Pump Flour.and an INCREASE in Reacter c'Ower.

f', . NOTE ASSUME RECIRC FLOW CONTROL IS-IN FLUX MANUAL- i s

h

+

b l

i L

(r**** CATEGORY 03 CONTINUED ON NEXT PACE r****)

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~ 3. 1NSTRUMENTS AND CONTROLS PAGE 16 GUESTION 3.08 (2.00)

Select.the correct Feeduater Control G y s t e n: ./ Plant Response from the list ( i throi>sh 4 below ) Por th? followins malfunctions:

(NOTE: NO OPERATOR ACTION IS TAKEN) t 1. Reactor water level decreases and stabilizes at a lower level. .

2. Reactor water level decrescos and initiates a reactor s t r a n. ,

3. Reactor water level incr-3ses and ctabilizes at a h'. sher level.

4 Reactor water level incroaces and initiate a turbine trip,

a. The plant is operating at 70% powerr in 2-element control, when ONE (1) S t e e r. Flow Signal to the reed Water Level Contral n/ stem FAILS to ZERO. (1.00)

b. The plant is operating at 1001 power, :n 3-elenent control, when *he Level Program Modifier signal fails to its MAXIMUM NORMAL OPERATING selve,

('.00)

QUESTION- 3.07 (1.50)

List TREE f3) cperator actions which will result in the Catmands Disagree Lamp being LIT. (1.50) i i

t

($113A ((qTEGO'sY /) 3 CONTINUED ON '2 EXT ~ AGE ro*#)

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_ _ . . . . . - - . . . . . . . - . - - - . . . - . . - - . - . . - - . . - . - - - - . . . - . - . _ - . . ~ - -

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t 3 '. ~ INSTRUMENTS.AND' CONTROLS- 'PAGE-- 17

! GUESTION - 3 .- 10 -(2.50) 4' i

The reactor. Plant is inLCOLD SHUTDOWN uith RHR.in the-j; . Shutdown Cooling Modes when t.. e ' REACTOR TEMPERATURE LOW

• f

, alarm-is receised. Answer the.followins..

s ?s . LIST the possible causes for the alarm. (INCLUDE

} SETPOINTS) (1.00)

.b. .What is the IMMEDIATE cperator action? ~(0.50)  ;

c. Is this'elarm an indication that a Limiting ~

Condition for Operation (LCO) has been entered?-  !

(0.50) i

d. .If an LCO is enteredr how long does the operator j have to restore Reactor Vessel Temperature to  !

within Limits? (0.50)'

E QUESTION 3'.11 (2.50) I i

Regarding the RPS System; i

a. STATE whether the solenoids associated with the fc;10 wing valveu are Energized or Deenersized, when the operstar-inserts a i MANUAL SCRAM. (1.00) i l'. Beet-up Scram Valvec >

2. Scram Discharge Volume Vent and Drain Valves  !

b. LIST the reactor SCRAM functions which become e"fective ,

unen the operator takes the Mode Sw:tch from RUN tc STARTUP/

HOT STANDE'Y. (INCLUDE SETFOINTS)  !

-( 1. 5 0 )

[

t GUESTION 3.12 (2.00) I

?

EXPLAIN the EEOUEMCE of events for the Feeduster Control  !

System 'Setpoint Setdown Mode' feature from actuation [

following s r e a c t 'a r SCRAM to a reset condition. Ensu*e [

thct your e :p ! metion addrettas the following:

2.00)

- all applicable setpoint.(s)  ;

- specific effect(s)  !

- any timer settingc l

- reset metnod'.;)  !

I l

1**r** CATECORY 03 C0fM LNUC0 DN NE"T  ?.G E ***u) t

. ..- ~ - . - - - - - -

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i 3 ,: INSTRUMENTS AND CONTROLS PAGE 18 j - ____________________________

I DUESTION 3.13 (3.00) 4 l- With regard to-the Nuclear Instrument.ation Sy stem , answer

'the following:

1 (NOTE: FIG. 137 IRM RANGE SCALE ic l PROVIDED FOR REFERENCE)

a. An operator switches.an IRM which is reading 39 on -

Range 6 to range 5. (1 00) ,

i fi) What will this IRM indicate on

! Range 5 ?

4

! ( 2 .4 . LIST all, if any, alarn(s) a n ci / o r automatic i cetion(s) which w i'1 : ' occur.

b. WHICH AGAF value (P1 printout) is scre conservative;

! (1) 0.?9 (0.50)

- (2) 1.01

c. LI51 ALLr if anyr alara(s0 a r.d / o r autcmatic actionto) i which will occur if the APRM Channel Mode Switch fer .

APRM 'A' is placed in " Power-Flow'. (1.00) ,

d. Four percent (4%) Reactor Power should indicate 3PPro"imatly______ on IRM rance . (0.50) 4 4

l j (r**** END 0 CA'!EGORY 02 **tri 1

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4. . PROCEDURES - NDRMAL, ABNORMAL, EMERGENCY AND PAGE #f9

___________..____________________________________ f e RADIOLOGICAL CONTROL

____________________ ,a 4J

. QUESTION 4.01 (1.50)

Reactor Recirculation' pump 'A' was secured ONE hour aso. ithe .

Reactor-is in HOT SHUTDOWN) The operator attempts to start the recire pumpi he observes cotor anps increasing and the the pepp breaker trips open. (1.50) i

a. WHEN can the operator attempt ~another start?

b. HOW LONG must the operetor wait between additioncI ,

. start attempts?

c. Recire pump 'A' starts, runs for ten (10) minutes-and trips.

Can the operator attempt an immediate restart? .

,0UESTION 4.02 (2.00) ,'

All control room. key lock switches are required to have the /

key in the hand switches in accordance with PROCEDbRE 't 1 02-S-01-9. LIST four (4) exceptions to this restirqnent. *

'QUESTICh 4.03 (i.00) i i During the activity of shutting the plant downrin accordance with,0P-IP-503rit is required to INSERT the SRM detectors.

State' 01.00) a .' When are the detectors inserted?

b, What count rate is required te be naintained while the detectors are being :nserted?

GUESTION 4.04 (3.00)

List the entry concitions for the "CCNTAINhENT CONTROL ,

ENERGENCY PROCEDURE'. Include the initiatins levels /setpoints. (3.00)

\

(xr m CATEGJRY 04 CONTINUED ON NEXT PAGE n a tr -

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4. -PROCEDURES - NORMAL 7 ABNORMAL, EMERGENCY 6ND PAGE 20

~~~~R ED55L5Gicst-c5 sis 5E-~~--------------~~~~-~~

QUESTION 4.0S (1.50)

In.accordence with Off Normal Procedure CONTAINMENT HYDROGEN CONTROL, list the THREE (3) conditions which .

-require the starting of the hydrogen ignitors.

(1.50)

GUEST 'ON 4.06 (2.50)

The reactor is in Optrational Ccndition 2 (STARTUP) and reactor power is 3%. A Drywell entry is required to investigate a pessible leak, Ctate the fallowing; (2.50)

a. The required status of the TIP system prior to entry.

<a ~

b. The OEPARTMENT responsible for completing the Orywell Entry Check List.

c. The responsibilities of the STANDBY MAN during the Drywell entry.

'a

• 't GUESTION 4.07 (1.00)

When pstforming n whole body friskethe frisker probe should be held __ ( a) __ inch (es) from the surface being surveyed, sad the speed at which the probe is moved should be __ (b) _

inch (es) per second. (1.00)

, >+

GUESTION 4.09 (2.50)

-LIST five (5) conditions when a r- tm P (Radiation Work Permit) will be used. :2.50) i 1

(***** CATEGORY 04 CONTINUED ON NEXT OACE *****:

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'4. . PROCEDURES - NORMAL,. ABNORMAL, EMERGENCY AND PAGE 21

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~~~~E5Df0LUUfd5L UUNTEUL DUESTION 4.09 (1.00)

Select the Emergency Procedure (s) which the operator would enter after a reactor SCRAM with the following conditions existing: (1.00)

-Reactor water level tu zero (0.0) inches on both 1821-R623A and 1921-R623B (Level Indicators on 1H13-P601) i -Drywell pressure is 1.5 psig l

-Suppression Pool 1cvel is 18.5 feet

-MSIV's-closed on steam tunnel high t er:pe r a t u r e -

-IRM's inserted and indications decreasing or- Range 3

3. EP-1 Level Control, and ER-10, Reactivity Control

b. EP-3, Containment Control, and EP-10, React.ivity Control e, EP-1, Level Control, and EP-3, Containment Control

d. 'EP-3, Containment Control

e. EP-1, Level Control GUEST 10N 4.10 (2.00)

In the procedure for SHUTDOWN FROM THE REMOTE SHUTDOWN PANEL (05-1-02-II-1). whet are the four (4) i n.n e d i a t e actions you are to perform pr or to eniting tne control roon? (Assume you hcVe cdequete time.) (2.00)

QUESTICN 4.I' ( * .00)

The' occurrence rf en event or events which involve an actuel or potential SUSSIANTIAL degradation of the level of safety of the plant is classified as: (1.00) a .- Unusual Event

b. Alert Emergency

c. Site Area Emergency

d. General Enersency

(****x CATEGCRY 04 CONTINUED ON NEXT PAGE **n*)

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! . 4. PROCEDURES NORMAL, ABNORMAL,~ EMERGENCY' AND -PAGE 22

RADIOLOGICAL CONTROL t

____________________ f j ;-

! QUESTION 4.12' (2.50) 1 l During a REACTOR HEATUP AND PRESSURIZATION. .the operator. ,

l performs specific actions at various reactor temperatures

! and pressures. Match the Action taken to the appropriate l . Pressure or Temperature. (2.50) l

[

! ACTION PLANT STATUS '

i

s. Realign the head vents (1) 500 to 650 psig l j b. Place RCIC in stendby (2) 190 deg. F  ;

c. Perforn RCIC low-flow (3) 950 psis 3

verification (4) 60 psis l

~

d.; Place SJAE and GS Gen (5) 150 psig in service (6) 90 pcis

-e. Mak e Orywell entry (7) 212 deg. F l and vet. v -recire pumpt

' ' 00ESTION 4.13 ( .50) ,

t j TRUE/ FALSE.

l ONLY the Cuntrol Room Operator or the Assistant Control Room i j' Operator may mcks comments in the CONTROL ROOM OPERATORS j LOG 900M. f I f

! i

[

2 0UESTION 4.14 (1.00) i L In accordance with Drocedure 09-S-02-100 (Criticality Rules),

j during normal opetations the irradiated FUEL ASSEMBLIES i

shall be stored in the (a) . i

( During OUTAGES, the irr3dicted~FEEE~E555MEE155 53y~~~

l be stored in the._(b) _____ ________ ________ . r i

(1.00) 1 QUESTION 4.15 (1.50)

~

i As a Licensed 0peratorr you are re'.leving the Control Roon '

Operator at shift turnover. LIST can (6' 2*ect of

infornation which should be discussed before y o i. relieve the

[ ~ Control Room Operetor on duty.  ;

l'  :

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4. PROCEDURES.- NORMAL'r ABNORMALr EMERGENCY AND PACE 23

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~~~~E5656666565E~C6NTR6L

-QUESTION 4.16 (1.00)

'Per EP-2, ' Emergency Cooldown*, which one of the following most accurately describes how SRV cperatien should be used to control pressure, if needed?

NOTE: ASSUME THAT THE INSTRUMENT AIR SYSTEM IS OPERATING PROPERLY

a. Use numerous SRV's, with chcrt pressure reductions

( ~ 50 psig) to equalize Suppression Pool heatup.

b. Use fewer.SPV blowdowns, with increased pressure reduc-tions'to ninimize SRV cyclic stresses.

c. Depressurine with a sustained SRV opening to manit: ire the' emergency cocidewn rate.

d. Allow the SRV's to operate by mechanical actuction to ensure design pressur e control and heat dispersion.

(***** END OF CATEGORY 04 *****)-

(*****x******* END OF EMAMINATION ***************)

O - _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

MASTER

+ , ,

.1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE- 24 IUEBUOpyOODICDa_bE6J_TB8NSEEB_ BUD _ELUID_ELOW ANSWERS -- GRAND GULF 1 -86/09/OG-BROCKMAN, K.

ANSWER 1.01 (2.00)

1. Fuel burnup

2. Fission product poison buildup

3. Resonant absorber buildup

4. Moderator temperature increase

5. Fuel temperature increase

6. Void fraction increace (recirc flow decrease)

7. Control rod density increase (4 of 7 required O O.5 each)

REFERENCE BWR ACADEMIC SERIES ON REACTOR THEORY ANSWER 1.02 (2.50)

a. Doppler or fuel temperature (.25), negative (.25)

b. Void (.25), negative (.25)

c. Moderator temperature or Fuel temperature (.25), negative (.25)

d. Void (.25), positive (.25)

e. Moderator temperatur e (.25), positive (.25)

REFERENCE DWR ACADEMIC SERIES ON REACTOR THEORY ANSWER 1.03 (1.50)

Rod worth increases (0.5). Increasing coolant temperature decreases the moderator density allowing the neutrons to travel further before being absorbed. This increases the probability of the neutrons interacting with a control rod (1.0).

REFERENCE BWR ACADEMIC SERIES ON REACTOR THEORY

1.- ' PRINCIPLES OF NUCLEAR POWER PLANT QPERATION, PAGE 25~

ISEBdgDyOOd1CG,_UEBI_JBGUSEE6_6ND_ELUID_E60W ANSWERS -- GRAND GULF 1 -86/09/08-BROCKMAN, K.

ANSWER 1.04 (2.00)

. Drawing a straight line between the last two *'s predicts 33-36 control rods must be withdrawn.

(0.1 for each point plotted, 0.7 for line, and.O.6 for prediction) 5 10 15 20 25 30 35 40 45 50 55 1.O*----l----l---- ----l---- ----

---

l----!----l----!---- ----1.0 0.9- -0.9 O.8- * -0.6 O.7- -0.7 1/M O.6- -0.6 O.5- -0.5 O.4- -0.4 O.3- * -0.3 O.2- * -0.2 O.1- * -0.1

+ -

0.0 ---- ----l----l----t---- ----i----i----l---- ----l----l----0.0 0 5 10 15 20 25 30 35 40 45 50 55 Control Rods REFERENCE NEDE - 12255, pg 265-6 ANSWER 1.05 (1.50)

TRUE (0.5)

Using the equation Power = Power (initial) x in time / period and solving for time results in the equation:

time = Period a In Power / Power (i ni t i al )

From this it can be seen that since 5/1 yields the same value as 50/10, and since all other' factors in the equation are equal, the time is equal (1.0 FOR EXPLAINATION) l

fit __E81NGLE6ES_QE_NQQLE68_EQWE8_E66NI_QEE8611QNr PAGE 26

_IMEBMQD188 mig @t_SEGI_18QNSEE8,6MD_ELQID_E(QW LANSWERS -- GRAND GULF 1 -86/09/08-BROCKMAN, K.

. REFERENCE ~

'BWR. ACADEMIC SERIES REACTOR THEORY ANSWER 1.06 (2.00)

If the reactor was shut down by 1% dk/k as measured at the time of.

peak xenon, then SDM will. decrease as xenon decays. Since xenon (peak) -is greater than the 1% dk/k, an inadvertent-criticality could result. (2.0)

REFERENCE BWR ACADEMIC SERIES of REACTOR THEORY TECH SPEC BASES 4

ANSWER 1.07 (1.00)

a. As voids increase,,the density of the moderator decreases so-the neutron slowing down time and length become longer,

,thus resonant absoption increases due to. neutrons to spending more timef i n the resonant energy spectrum, (0.5) and thus, the' fuel temperature coefficient becomes more negative (0.5).

REFERENCE

DWR ACADEMIC SERIES-of REACTOR THEORY ANSWER 1.08 (1.50)

.a. Operator- can notice'that period-has become l'onger and that power change on IRMs, SRMs is leveling off

'(turning around due to. power overshoot). (0.5)

b. ('2 ) From P = Poe(t/T) --> T = t/in P/Po, in Interval 2 the period has lengthened from 80 seconds. The other intervals have 80 second periods (1.0)

-REFERENCE

-BWR ACADEMIC SERIES OF REACTOR THEORY

l

. .- n.

.1z -- PRINCIPLES OF NUCLgBRlEgWER_ELONI_gEgRBTIgN a .PAGE 27 LIUEBUggyd6MICS,_Ug3I_IB6USEgB_Oyp_Ebulp_E6gy ANSWERS -- GRAND GULF'1 -86/09/08-BROCKMAN, K.

ANSWER 1.09 .1.00)

(

.a REFERENCE

.DWR ACADEMIC SERIES OF REACTOR THEORY ANSWER' 1.10 (2.25)

a. Decrease (0.37) due to increased void content in the core as flow decreases (0.38).

b. Iricrease (C.25) due to increased voiding in the cure (0.25) and recirc pump no longer taking a suction on the annulus (0.25).

4 T. c. Decrease (0.25) due to steam flow decrease (0.25) and l evel

. increase (0.25).

l

. REFERENCE-GGNS SI'ULATOR M MALFUNCTION #12-DWR ACADEMEIC SERIES REACTOR. THEORY -

4 ANSWER 1.11 (2.00)

a. 8

b. 3

c. 2

,d.- 5

-(0.5 each)

REFERENCE BWR ACADEMIC SERIES HEAT TRANSFER AND FLUID FLOW 1 J

l .

i-(. -ANSWER 1.12 (1.00)

[. _d .

l

[ REFERENCE l DWR Academic Series Heat Trans and Fluid Flow pg. 6-109 I

i i

-~ , - - - - . . - - - - . . . - , . - - , - - - - - . - - - - - - . . - - - - - . . - . . . .

1-' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION 2 PAGE 28

.IHE8MODYN8DICS 2 _HEBI_IB8NSEE8_8Np_ELUlp_E60W ANSWERS -- GRAND GULF 1 -06/09/08-BROCKMAN, K.

ANSWER 1.13 (1.00) d.

REFERENCE BWR ACADEMIC SERIES REACTOR THEORY ANSWER 1.14 (3.00)

a. RCIC and HPCS begin injecting.

b. Feed flow is rapidly dropping to zero; therefore, core inlet subcooling is decreasing; adding negative reactivity.

c. Bypass Steam Control Valves are fully open to control reactor pressure, and the main turbine has tripped ,

d. Reactor power decreases due to a reactor scram, after the scram decay heat is removed to the condenser via the BPSCV's that control reactor pressure at the EHC pressure-set.

e. Low reactor water level.

(0.5 EACH)

REFERENCE LP OP-DT-509 BWR ACADEMIC SERIES REACTOR THEORY ANSWER 1.15 (1.50)

,a. INCREASE

b. DECREASE

c. DECREASE C3 0 0.5 each3 i- t REFERENCE GG, Heat Transfer and Fluid Flow, V o l. . 2, Ch. 9, PP. 26, 20 f

ANSWER 1.16 (1.00) d REFERENCE

.BWR ACADEMIC SERIES HEAT TRANS, RANKINE CYLLE i

i-

)' . ~

t k: .

+ . . . .

I j ~ li_; PRINCIPLES OF NUGLEAf3_PgWER_ PLANT _gPgRATIgh PAGE- 29  !

I THERMODYNAMICS,-HEAT TRANSFER AND FLUID FLOW  !

ANSWERS -- GRAND GULF 1 -86/09/08-BROCKMAN, K.

}

e" .!

\-

r' i 5

4

j. ~ ANSWER '1.17 (1.00) l

i.  !

1 i

REFERENCF  !

, GGNS OP-AD-545,P.10 i j -,

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' 2." -PLANT DESIGN _INCLQQ1NQ_SAFEIY_AND_EMERGENGY_SYSTEijg PAGE 30 ANSWERS - ' GRAND GULF'l-' -86/09/08-BROCKMAN, K.

l ANSWER' '2.01 ( l '. 50 )

1. RWL. LOW (.25) -150"-(.25)-

2. .HI'D/W pressure (.25), 1. 39 ' pnig (. 25)

3. . LOSS OF OFFSITE POWER _after ESF is reenergized (.5)

REFERENCE GGNS.L.P. OP-P41-501 ANSWER 2.02 (3.00) a.- 1. FAI

2. FO.

.3. FC

'4 FC

5. FC (0.5 each)

b. Valve stem sealing _ air to Off-Gas' system is lost (0.5)

REFERENCE GGNS: ONEP 05-1-02-V-9; OP-C11-1A-501, p 13 ,

ANSWER 2.03 (1.00)

C

. REFERENCE-

.GGNS: SD P42, pp'3, 19; OP-P42-501; 04-1-01-R21-1; Prints E1226 L E1116 ,

ANSWER 2.04 ( .50).

Counterclockwise (Slow)

REFERENCE GGNS: 04-1-01-P75-1 i-4

- - _ . , , - . . . . . ~ . ., . _ ..._.....,__.m.-

he-

,2,- P(QNI_DE@l@N_ LNG (UDid@_g8EETV ANQ_EMER@ENGY_SYETEMS PAGE 31 ANSWERS -- GRAND-GULF 1 -86/09/08-BROCKMAN, M.

ANSWER 2.05 (2.00) i

a. 30 psig l

l

b. 21 Psid .

l

c. 15 psig

d. 3.0 psid (0.5 each) l

' REFERENCE GGNS: SD-M41-1; TECH SPECS AN3WER- 2.06 (2.00)

a. .SBGTS train "A" will continue to run. (1.0)

b. SBGTS train "A" will stcp (0.5), but will auto restart on SBGTS train "B" low flow or recirc fan "B" low flow or ENCL BLDG pressure high. (0.5)

RFFERFNCE GGNS SD T-48 PG 3,5; OP-T4B-501 FG 8,13; 04-1-01-T4D-1 ANSWER 2.07 (1.00)

Both subloops are in the MAINTENANCE MODE (0.5)

Valve' Motion Inhibit Interlock ar tivates or flow control valve "A" is locked up. (0.5)

REFERENCE GGNS SD-B33-1, D-33-2

t 2, PLANT-DESIGN INCLUDING EAFETY AND EMERGENCY SYSTEMS PAGE 32 ANSWERS -- GRAND GULF 1 -86/09/08-Bh0CKMAN, K.

AN5WER 2.08 (1.50)

a. engine overspeed low lube oil pressure generator lockout

'Ealso accept: generator ground overcurrent and generator phase differential current if generator lockout is not listed as an answer]

(0.25 each)

b. high crankcase pressure low turbo charger oil pressure high vibration high bearing temperature high jacket water t e.n p high lube oil tcmperature (any 3, 0.25 each)

ANSWER 2.09 (3.50)

a. - vessel bottom head drain and steam dome temperature  :

differential (0.3) less than 100 deg F (0.2)

- recirculation loop and steam dome temperature differential (0.3) less than 50 deg F (0.2)

- less than 50 deg F temperature differential (0.2) between the two recirculation loop suction lines (0.3)

b. - reactor vessel level (0.25) < 1evel 2 or -41.6" (0.25)

- reactor pressure (0.25) / 1125 psig (0.25)

c. - The ATWS trips both the high and low speed power supplies to the recirculation pumps (0.5). [M 4rl5P breakers C6- 2. ad c6-5]

- The RPT trips only the high speed power supply and causes a high to low speed transfer to occur (0.5).

REFERENCE GGNS L.P. OP-933-1-501

2c__ PLANT DESIGrJ_INGLQQLNG_QAFETY_AND_EMERGENQY_ SYSTEMS PAGE 33 ANSWERS -- GRAND GULF 1 -96/09/08-BROCKMAN, K.

ANSWER 2.10 (1.50)

a. Yes.

'May start.with SSW pump handswitch or May cause auto start of SSW pump by manually starting'HPCS pump or May cause auto start of SSW pump by manually starting HPCS diesel.

CFor any credit, must answer yes and give one method that the operator may initiate the SSW pump 3 (0.25)

b Low reactor water level High drywell pressure Loss of offsite power B or C RHR pump running D/G # 12 running (0.25 each)

REFERENCE GGNS OP-P41-501 ANSWER 2.11 (1.50)

TRIPS: mechanical overspeed (.25), 110% (.25) local-manual trip (.50)

RESET locally by pulling the mechanical trip rod to the reset position (expect answer to reflect necessity of resetting mechanical trip rod locally) (.5)

REFERENCE GGNS LESSON PLAN OP-E51-SO1 ANSWER 2.12 (1.00)

1. Pump cooling water temperature high

2. Pump suction low flow

3. FOO1 closed and any of F250, F251, F252, F253 not fully open

4. FOO4 isolation valve not fully open

5. Motor protection device activation (any 4 at 0.25 each)

= 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 34

~ '

i . ANSWERS -- GRAND GULF 1 -86/09/08-BROCKMAN, K.

REF RENCE GGNS SOI-04-1-01-C11 pp 2,3 ANSWER 2.13 (1.00)

I a Failure of recirc pump A #1 seal b Plugging of recirc pump A #2 seal internal restriction / breakdown orifice (0.5 EACH)

REFERENCE GGNS SD-B33-1-501 pg 5 ANSWER 2.14 (2.00)

. a. 1. 135 psig

2. 50 psig

3. 125 psig (0.5 each)

b. Recirculation loop B (0.5)

REFERENCE GGNS LESSON PLAN OP-E12-501 ANSWER 2.15 (2.00)

a. CST suction valve stays open, Suppression Pool suction valve stays shut

, b. CST suction valve opens, Suppression Pool suction valve stays shut

} c. CST suction valve stays shut, Suppression Pool suction val ve stays open 1 d. HPCS test return valves close (0.5 EACH)

I

t. .- - - . . . - . - . - - . - - - , . . . - . - - . -... - -- -.- - .- - - .- - - . - - - - - - - - - -

2- PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 35 ANSWERS - GRAND GULF 1 -86/09/OG-BROCKMAN, K.

ANSWER 2.16 (1.50)

a. ECCS discharge pressure interlocks EO.53

"' ' ; T C

._a; . .m _-1,, ,

.252 - Tcle.hd

c. Si :t EO.5]

REFERENCE GG, Lesson Plan - E-22-2, P. 7, 8 l

l l

l l

l

4 e'  %, .

3. INSTRUMENTS AND CONTROLS PAGE 36

. ANSWERS -- GRAND GULF 1 -86/09/08-BRCCKMAN, K.

2 ANSWER 3.01 (1.50)

-a.. ' level 8 or +53.5", l evel 3 or +11.4" (0.125 ea)

6. level 2 or -41.6"

c. level 2 or ~41.6"

d. level 8 or +53.5"

e. level.3 or +11.4"

f. level 4 or +32.7" (b.- f. O.25 each)

REFERENCE GGNS TEC SPECS ANSWER 3.02 (1.00)

a. G inches

, b. added to (0.5 each)

REFERENCE' i

GGNG L.P.-OP-C34-501 Feedwater Control f ' ANSWER 3.03 (2.50)

a. 1. yes

2. no

3. no

4. yes

5. no (0.5 each)

REFERENCE GGNS L.P. OP-C11-2-501 GGNS POM 04-1-01-C11-2 P

I

m

• ' e  %, .

- 3. INSTRUMENTS AND CONTROLS PAGE 37 ANSWERS -- GRAND GULF 1 -86/09/08-DROCKMAN, K.

' ANSWER 3.04 (2.50)

a. half-scram

b. no action

c. no action

.d. rod binde

e. scram (0.5 EACH)

REFERENCE ONEP 05-1-02-III-2 C

ANSWER 3.05 (3.00)

a. 4 or ETS oil pressure at the TCV (0.5)

b. The TCV Fast Closure is initiated by actuation of the Load Reject and Turbine Trip Controller (0.5). Load Reject is defined as at least a 35 % GEN load drop at ,

greater than 10% per sec. (0.25) The fast closure is bypassed if reactor power is less than 40 % (0.5) as sensed by first stage turbine pressure (0.25).

c. 1. shut (0.5)

2. shut (0.5)

REFERENCE GGNS TECH SPECS

-GGNS L.P. OP-N32-2 PG 15, 16 ANSWER 3.06 (1.00)

- SRV's

- RCIC

- RHR

- SSW

- CRD (0.25 each)

REFERENCE GGNS ONEP 05-1-02-I1-1 C

• e'  %.

3 n__ INSTRUMENTS AND CONTROLS PAGE 38 ANSWERS -- GRAND GULF 1 -86/09/08-BROCKMAN, K.

ANSWER 3.07 (1.00) c.

REFERENCE GGNS: SIM MAL 11 ANSWER 3.08 (2.00)

a. 1

b. 1 REFERENCE GGNS: OP-C34-501 ANSWER 3.09 (1.50)

1. Depressing the COMMANDS DISAGREE "1" SHIFT INHIBIT pushbutton.

2. Depressing the COMMANDS DISAGREE "2" SHIFT INHIBIT pushbutton.

' 3. Depressing the COMMAND INHIBIT pushbuttons.

4. Depressing the DISPLAY test pushbutton.

(0.5 each)

[Also accept when RGDS finds disagreement between 2 RACS signals 3 REFERENCE 1 GGNS SOI-04-1-01-C11 7/'PG 2,3;16.

C

• i' . \, .

3- INSTRUMENTS AND CONTROLS PAGE 39 ANSWERS -- GRAND GULF 1 -86/09/08-BROCKMAN, K.

NSWER 3.10 (2.50)

a. vessel head flan'ge temp less than or equal to 75 degrees F.

Bottom head temp less than or equal to.75 degees F.

Shell flange temp less than or equal to 75 degees F.

Bottom head drain temp less than or equal to 75 degrees F. (4 at 0.25 each)

b. Notify the SHIFT SUPERVISOR /SUPERINTENDANT.

(Also accept an action to prevent further.cooldown as CAUTIONED in the immediate operator actions of ARI 04-1-02-1H13-P680-3A-D2) l (0.5)

c. NO (0.5)

d. one-half hour (0.5) l REFERENCE GGNS ARI 04-1-02-1H13-P680-3A-D2 GGNS TECH SPEC 3/4.4.6 ANSWER 3.11 (2.S0)

a. 1. ENERGIZED

2. DEENERGIZED (0.5 each)

b. APRM High-High ( 15'/. )

IRM High-High (120/125 scale)

IRM INOP (Low Volt, Out of Oper, Module Unplug)

(.25 each function; .25 each setpoint)

REFERENCE GGNS: OP-C71-501, pp 17, 18; OP-C51-4-501, p 25; TS's ANSWER 3.12 (2.00)

- Auto initiated at + 11.4" ( 0. 5)

- Level signal increased to +54" (0.3) for 10 seconds (0.2)

- After 10 seconds (0. 2 ) , +D4" replaced by +18" signal (0. 3)

- No Reset until operator actuation of "Setpoint Setdown Reset" (0. 5)

_ . _ _ . _ _ _ . . _ _ _ _ . _ . _ _ _ _ - _ . _ _ _ _ _ _ . . _ . _ _ . . . . . _ . . . _ _ . _ . ._--._...._ . . _ _ . ._.m____._..__._..

!r-1:

i.

l, f

• ;b s, .

3- -INSTRUMENTS'AND CONTROLS PAGE 40 p .

,; LANSWERS '- GRAND GULF 1 -

-86/09/08-BROCKMAN, K.  ;

O- P i e i t

! . REFERENCE

! 'GGNS: OP-C34-501  :

i i

l ANSWER 3.13 (3.00)

i. a. . '( 1 ) 39 on range 5 (.5)

(2) IRM hi rod block (.25) i

. IRM hi hi 1/2 scram (.25) .

b. 0.99 is more conservative-(.5)

! c. APRM INOP (.33)

Rod block (.33) l 1/2 scram (.34)

d. reads 100 (O.25) 'on range 8 (0.25)

REFERENCE-f GGNS'OP-C51-1, 2, 4 -501 LESSON PLAN f 4 .

i I

I A

P h

3 h

L k

l r

n

_WTwh- _ _ _ _mW_ _ ___, __ _ _ __ _

e A ' t, .

4, PROCEDURES - NORMAL 3_ ABNORMAL 2_EMERQENCy_AND PAGE 41 BODI96991CO6_990IB96 ANSWERS -- GRAND GULF 1 -86/09/08-DROCKMAN, K.

ANSWER 4.01 (1.50)

a. may attempt immediate restart, no wait required.

b. 45 minutes

c. no (0.5'EACH)

ANSWER 4.02 (2.00) a Custody of the SLC pumps key shall be maintained by the Shift Supervisor or the Shift Guperintendent. Keys may be obtained by the operators for emergency or test ccnditions.

b Custody of the mode switch key shall be maintained by the Shift Supervisor or the Shift Superintendent when the mode switch is required to be locked in a specific position.

c The keys for D SRV solenoids owhpanel P631 shall be maintained in a box attached to the P631 panel. The keys shall be maintained in the box except when operation of any SRV is required, d If keys are required to be removed from hand switches when necessary to comply with a license commitment.

e. (also accept) as directed by shift supervisor when noted in log or on clearance (ANY 4 AT O.5 EACH)

REFERENCE GGNS FOM O2-S-01-9 ANSWER 4.03 (1.00)

a. IRM channel s on range 3

b. between 10E02 and 10E05 REFERENCE GGNS POM 03-1-01-3 FLANT SHUTDOWN

e- b i, ,

4. PROCEDURES - NORMALx_AijNQRMAks_EMERGENGY_QND PAGE 42 R8DlQLQGIG6(_QQNIBQL ANSWERS -- GRAND GULF 1 -86/09/08-BROCKMAN, K.

ANSWER 4.04 (3.00) a High drywell pressure (0.25) above 1.23 psig (0.25) b High drywell temperature (0.25) above 135 F(0.25)

'C High suppression pool temperature (0.25) above 95 F (0. 25) d High suppression pool level (0.25) above 18.81 feet (0.25) e Low suppression pool level (0.25) below 18.34 feet (0.25) f High containment temperature (0.25) above 90 F(0.25)

REFERENCE GGNS 05-S-01-EP-3 ANSWER 4.05 (1.50)

a. Before the containment hydrogen concentration reaches 3.5%

b. Containment pressure reaches 7.84 psig

c. RPV water level drops to TAF REFERENCE GGNS ONEP 05-1-02-III-11

c. le_ 1, ,

4, PAGE 43 PROCEDURES - NORMALt_QBNQRMBL t_EMER'dENCY_6ND E6DLQ60tiLC6(_CQNIGOL ANSWERS -- GRAND GULF 1 -86/09/08-DROCKMAN, K. -

ANSWER 4.06 (2.50)

a. The TIP'S will be fully extracted or inside tne vessel (.5) and tagged out (.5)

b. The Health Physics Department (.5)

c. The STANDBY MAN will be stationed outside the Drywell entrance (.3) he will maintai n communications with the control room and the personnel in the Drywell ( . 4) he will have av&ilable any standby equipment which may be required (.3)

REFERENCE GGNS AP 01506-7 ANSWER 4.07 (1.00)

a. 1/2 inch

b. one to two (also accept either)

REFERENCE GGSN ADMIN PROC 01-S-08-2 i

l l

a b 8, .

Sc__E8QQEDgGEQ_ _NQEM86t_GQNQBdQLt_EMESQEUQY_QND PAGE 44 BSD106QGIGGL_GQNISQL ANSWERS -- GRr.ND. GULF 1 -86/09/08-BROCKMAN, K.

ANSWER 4.08 (2.50)

a. Entrance into or work in a hi contami nati on area.

b. Entrance into or work in a posted neutron radiation area

c. When performing work which will open a potentially highly contaminated system,

d. Jobs requiring greater then 0.1 MAN / REM

e. When directed by Health Physics

f. Entrance into or work in a hi radiation area

g. Entrance into or work in a 6e{y hi radiation area

h. Entrance into or work in a potentially airborne radi a ti on area

i. Entrance into or work in an airborne radiation area (eny five at 0.5 each)

REFERENCE GGNS ADMIN FROC 01-S-OO-2 ANSWER 4.09 (1.00)

c. (1.0)

REFERENCE GGNS 05-S-01-EP-1,EP-3 1

• %',s 4, PROCEDURES - NORMALt_ABNQRMAL1_E tjEl3GENCY _AUD FAGE 45 R6DIOLOGICOL_CQUIBQL ANSWERS -- GRAND GULF 1 -86/09/08-BROCKMAN, K.

ANSWER 4.10 (2.00)

a. P1 ace the mode switch in the shutdown position

b. Verify all control rods are fully inserted

c. reset the scram

d. manually initiate the RCIC system (0.5 EACH's REFERENCE GGNS ONEP 05-1-02-Il-1 ANSWER 4.11 (1.00)

b. (1.0)

REFERENCE GGNS EXAM DANK GGNS EPP 10-S-01-3 ANSNER 4.12 (2.50)

a. 2

b. 4

c. 5

d. 1

e. 3 (0.5 EACH)

REFERENCE GGNS EXAM DANK PROC. 01--S-02-2 ANSWER 4.13 ( .50) l Fa1se. (O.5) (Comments al1 owed by others 1f entey iu fol1 owed ta y signature) u

F

= **t.e d___P8QQEQUBES_n_NO8dQ6t_GEUQBdQ6t_EUEEGENGY_AND PAGE 46 86010600LC66_CONIGQL ANSWERS -- GRAND GULF 1 -86/09/08-DROCKMAN, K.

REFERENCE GONS ADMIN PROC 02-S-01-4 '

ANSWER 4.14 (1.00)

a. SPENT FUEL POOL

b. CONTAINMENT FUEL POOL (.5 EACH)

REFERENCE GGNS PROC 09-S-02-100 ANSWER 4.15 (1.50)

a. status of safety-related equipment

b. running equipment and train alignment

c. inop equipment and LCO'u including surveillance requirements

d. reasons for alarms that are lit

e. tagged equipment

f. tests, evolutions, and surveillances in progress at shift change

g. unusual events which have occurred in the past 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

h. abnormal system valve lineups (ANY SIX AT O.25 EACH)

REFERENCE GGNS PROC 02-S-01-4 ANSWER 4.16 (1.00) b REFERENCE GGNS: 05-5-01-EP-2, p 6 u _ _ _ _ _ __ _ _.___ . .