Text

Requalification Exam Rept 50-416/OL-86-03 on 861215-18. Exam Results:Four of Six Reactor Operators & Four of Six Senior Reactor Operators Passed.Requalification Program Marginally Satisfactory.Exam Questions & Key Encl
	ML20207T716
Person / Time
Site:	Grand Gulf
Issue date:	03/09/1987
From:	Brockman K, Munro J; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	;
Shared Package
ML20207T700	List: AECM-86-0415, Forwards Comments on 861215 Requalification Exams,Per Request.W/Five Oversize Drawings; B211035, Marked-up Schematic Diagram:B21 Automatic Depressurization Sys Relay Logics Unit 1; ML20207T699; ML20207T716; ML20266F053; ML20266F057; ML20266F062; ML20266F066;
References
	50-416-OL-86-03, 50-416-OL-86-3, NUDOCS 8703240271
	Download: ML20207T716 (174)
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3 ENCLOSURE 1-EXAMINATION REPORT .50-416/0L-86-03 o

Facility Licensee: System Energy Resources, Inc.

Facility Name: Grand Gulf Nuclear Station Facility Docket.No.: 50-416 Written and operating examinations were administered at the Grand Gulf Nuclear Station near Vicksbur , Mississippi.

Chief Examiner: m (M mI* / 5 /YM4 87 Date K E. Brockinpf -

Approved by:

%HWF F. Munro, S(crion Chief MM b Date Signed Summary:

Requalification Examinations on December 15-18, 1986 .

Operating examinations were administered to twelve candidates, twelve of whom passed. Twelve candidates were administered written examinations, eight of whom passed.

Based on the results described above, four of six R0s and four of six SR0s passed the examination. The overall evaluation of the Grand Gulf Requalification Program, based upon these results, is MARGINALLY SATISFACTORY. ..

Twenty-two facility comments were submitted; the cause of'five of them (23%) was due to either inadequate reference material, lack of reference material, or no learning objective provided by the facility.

B703240271 870312 PDR ADOCK 05000416 V PDR

REPORT DETAILS

1. Facility Employees Contacted:

K. Beatty

M. Shelley

Attended Exit Meeting

2. Examiners: "

K Brockman L. Lawyer S. Guenther

Chief Examiner

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

The'NRC Resolutions to facility comments are listed below.

a. R0 Exam

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(1) Question 1.11 NRC Resolution:

Concur. Based upon the potential confusion concerning the meter indication, the question is deleted.

(2) Question 2.06 NRC Resolution:

7 Concur. Ihe examiner piacea [ cat] cneck at Facility in tne answer key. The facility has provided the correct response. Answer key revised to reflect facility response.

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(3) Question 2.09 NRC Resolution:

Do not concur. The question has the candidate " Explain the consequence"; thus, no assumptions need to be made. The candidate must explain all the consequences of water buildup in the tail-pipe.

Without a subsequent blowdown, the standing water is not a problem.

No change to answer key required.

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Enclosure 1 2 (4) Question 2.10 NRC Resolution:

Do not concur. The question is explicit in stating the operator took two immediate actions end asks for the HPCS response. The candidate must describe the overall system response to these actions. There is no basis, technically or process-wise, to address each action individually; credit will be awarded based on how the overall system response is described.

(5) Question 2.12 NRC Resolution:

Do not concur. The 501 for this was not provided by the facility.

Furthermore, this question was systematic in nature, not procedural. No change to answer key warranted.

(6) Question 2.13 NRC Resolution:

Concur. Answers which state a need to compensate for minor pressure changes will be accepted. No change to answer key warranted.

(7) Question 3.01 NRC Resolution Concur. The graphs did not clearly discriminate between the choices. Either response B or D is accepted. Answer key modified.

(8) Question 3.03 NRC Resolution Do not concur. Indicated core flow is derived from one source, total jet pump flow. An assumption that the question is talking about recirculation flow is unwarranted. No change to answer key required.

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Enclosure 1 3

.(9) Question 3.05 NRC Resolution Concur. The question is NOT confusing. The output of the IPC is in milliamps not in either psi or percent steam flow. .These two measurements, psi /% steam flow, are used as representative values.

The operators.should be aware of the process for converting either of these two signals to a usable fora;. The answer key has been revised to accept the facility calculations. Part b is not double jeopardy; a mistake made in part"a" would be treated as an " Error Carried Forward". It is noted that the facility reference

. material for the EHC system had numerous errors throughout the description and operation sections. The facility training staff acknowledged this deficiency to the examiner. It. is reconmended that a complete revision of this lesson plan be conducted. Answer key revised.-

(10) Question 3.06 NRC Resolution Do not concur. During the examination, the proctor clarified this question to all examinees. No other assumptions were needed to answer this question accurately. No change to the answer key required.

(11) Question 3.08 NRC Resolution Comment Acknowledged. Answer key revised to accept "B" solenoid energizing only. The requal lesson material does not cover the complete operation of the SRV logic; therefore, the System Description was used to develop the question. The System Description does not describe the logic coincidence provided in the facility comment. The question has significant K/A values and stanas as-1s, except for the celetion in part 'e'.

(12) Question 3.10 NRC Resolution ,

Concur. . Facility reference material did not indicate a difference in pump response; however, the difference does occur and is

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appropriate. Answer key revised.

(13)Questiond.07 NRC Resolution Do not concur. The examinee need only recognize normal / abnormal meter readings as given. However, since no polarity was indicated for each voltmeter, part a.2. is deleted. Answer key revised.

Enclosure l' 4

b. SRO Exam (1) Question 5.10 See R0 exam question 1.11' (2) Question 5.11 ,

NRC Resolution: ,

Do not concur. The question is based upon Requal Objective 13, and all required information is contained within this lesson plan.

The complit ted calculation is not required, only a brief overview addressing key parameters. No change to the answer key necessitated.

(3) Question.6,01 See R0 exam question 3.01 (4) Question 6.02 See R0 exam question 3.03 (5) Question 6.04 See R0 exam question 3.05 (6) Question 6.05 See R0 exam question 3.06 (7) Question 6.06 See R0 exam question 2,09 (8) Question 6.07 See R0 exam question 2.10 (9) Question 6.09 i See R0 exam question 2.12 (10) Question 6.12 See R0 exam question 3.08. Also, Concur for part b.1.

Answer key revised to reflect same as R0 exam.

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Enclosure 1 -5 (11) Question 7.06-NRC Resolution:

Do not concur. The question is a verbatim extraction from 0P-RH-502, objective #3. Per the Code of Federal Regulations, the SR0 is responsible for radiation protection limits and procedures

- therefore, the item is applicable, irrelevant of learning objectives or other job analysis data. .

(12) Question 7.07 See R0 exam question 4.07 (13) Question 8.07 NRC Resolution:

Do not concur. This item requires operator action within 15 minutes. It is expected that candidates should be familiar enough with these actions to recognize the proper requirement. No change in the answer key is warnnted.

(14) Question 8,09 NRC Resolution:

Do not concur. LC0 3.8.1.1, action c, is most restrictive and is entered because of the inoperability of an Emergency Diesel Generator and redundant equipment; it is, therefore, the controlling requirement. No change to the answer key required.

(15) Question 8.11 NRC Resolution:

Do not concur, It is a well-established NRC po!!cy te raquira knowledge of all action statements of one hour or less. Only recognition of correct action statement was required. No change to the answer key warranted.

[ (16) Question 8.12 NRC Resolution:

Concur. Due to an error in question development the question is deleted.

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Enclosure 1 6

(17) Question 8.13 NRC Resolution:

- Comnent acknowledged. The intent of the question was to determine if the candidate is aware that an emergency condition, in of itself.. does not waive independent verification requirements.

No change to the answer key required.

(18) Question 8.15 NRC Resolution:-

Concur. Answers were transposed during development. Answer key revised.

(19) Question 8.16 a & b NRC Resolution:

Concur. This procedure appears to contradict lesson AP 525, Sections 3.b.2, 5.a.3, and 5.b.1. However, the. facility comments are procedure accep)ted, . Forbased 'upon part b., the controlled concur. Answer key document (i.e.,upon revised based the submitted documentation.

4. fritMeeting At the conclusion of the site visit, the examiners met with- +

representatives of the plant staff to discuss the results of the examination.

There were two generic weaknesses noted during the operating examination. The areas of below normal performance were: (1) Use of Emergency Operating Procedures - conversion to the new flow path format has resulted in a need to continue " practicing" their utilization to ensure effective and accurate use; (2) Emergency Plan Classification -

cerplete verification between the F0Ps and the classification

[ guidelines is needed to eliminate some inconsistencies (IFI 50-416/0L-63-01).

Use of Procedures (Normal and Abnormal) and crew communications were noted as having improved significantly from previous examinations.

, The cooperation given to the examiners was noted and appreciated.

l The licensee did not identify as proprietary any of the material provided to j- or reviewed by the examiners.

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'U. S. NUCLEAR REGULATORY COMMISSION SENICR REACTOR OPERATOR REQUALIFICATION EXAMINATION-FACILITY: _QRANQ_QU,lg_1__________L, s

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DATE ADMINISTERED: 86/12/J$________________

. EXAMINER: _gAgTh_g______*__________

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CANDIDATE: I !'4 8 ___~~j_D _ ( 9 _T _$_ $ __ ,___

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Read the attached instruction hage r.abef ully. 'llis examination replaces t) the current cycle f acil,i ty' administered raqual,ification examination.

Retraining requirements f or f ailure of ,, this exarrdnation age the same as for failure of a requalification examinatiori prepared ,and ' administered by 6 your training staff. Pointsi for1 each quee?;1on Qthe' indicated in parentheses after the question. The passing grade requires at least 70*f in each category and a final grade of at least 80%. Examination (. papers will be picked up four (4) hours after the examination starts. -

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% OF CATEGORY % OF CANDIDATE'S CATEGORY ;. J VALUE TOTAL SCORE VALUE _____g CATEGORY 3.

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_19299__ 2Dz99 _ _ _ _ _ _ - . . _ _ _ ______.._6.' ' PLANT'SYBTEMS DESIGN, CONTROL, 3

s SAND It!STRUMENTATION

_1E199__ _2Ez99 ___________ ___..__ _ 7. ' PROCEDURES - NDftMAL, AP, NORMAL,

\ EMERGENCY AND RADIOLOGICAL \

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! q All work done on this examination is my own. I havs naither given nor received aid. ~

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FRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS Duribg 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.

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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.

' 36 Use black ink or dark pencil only to facilitate legible reproductions.

47 yPrint your name in the blank provided on the covar '

sheet of the

+xaminatioi.) ? i

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Fill in thh'.date on the' cover shnet of the examination (if necessary).

Y h.- Use only the paper provided for answers.

. .h 7fl Priht your name in the upper right-hand corner of'the first page of each ys pq sec,t.i on of the answer sheet.

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/ C. Consecttively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write only gn gne side

) of the paper, and write "Last Page" on the last answer sheet.

4 L 9. Number each answer as to category and number, f or. 'ex ampl e, 1.4, 6.3.

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10. Skip at.least thtee 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 commcnly used in facility litetatute.

13. The point value for each question is indicLted in parentheses after the

question and can be used as a guide'for the depth of answer required.

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

15. Partia1Meredit may be given. Therefore, ANSWER ALL PARTS OF THE

% QUESTION' AND DO NOT LEAVE ANY ANSWE.R BLAtN.

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

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

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. 10. M en you complete your examination, you shall:

a. Assem'b le your examin'ation as f ollows:

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./ (1) Exam , questi ons on top.

(2) Exam' aids - figures,'t' ables, etc.

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

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

c'.' Turn in all scrap paper and the balance-of the paper that you did not use for answering the. questions.

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

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5,.__INE98Y_9E_NWGLE0B_Pgge_!seNI_QEgB911QNf_E6Q1pgg_9Ng PAGE 2 IMEBU99YN0DIGE i:

I QUESTIDN 5.01 (1.00)

STATE f or which condition the reactivity coef ficient contr'ibution would be MORE NEGATIVE. EXPLAIN your choice.

Doppler coefficient with a 25% Void Fraction in the core,

-OR- .

Doppler coefficient with a 75% Void Fraction in the core. .

QUESTION .5.02 (1.00)

Which-one of the f ollowing correctly describes the f ormation and removal processes at equilibrum atom density for Xe 1357

a. = (FP + Iodine burnout) -

(Xe decay + Xe135 burnout)

b. = (FP +-Iodine decay) -

(Xe decay + Xe136 decay)

c. = (FP + Cesuim decay) - (Xe decay + Xe135 burnout)

d. = (FP + Iodine decay) -

(Xe decay + Xe135 burnout)

QUESTION 5.03 (1.00)

The distance from the onset of bulk boiling to the point of transition

-boiling is described by which one of the following terms?

a. Critical Boiling Length

b. Boiling Fraction

c. Film Boiling Region

~d. Boiling Volume 4

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

THE98Y_9E_N996E98_E9 WEB _EL9NI_9EE8911pN i _E691p]z_9NQ PAGE 3 Dz__INEB590YN051gs QUESTION 5.04 ,

(2.00)

If steam goes through a throttling process, indicate whether the following '

parameters will INCREASE, DECREASE, or REMAIN THE SAME. )

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a. Enthalpy

b. Pressure

c. Entropy

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c. Temperature r

QUESTION 5.05 (1.00)

While operating at power a RWCU filter domineralizer ruptures causing resin intrusion into the Rx vessel. State the effects (increase, decrease, remain the same) for each of the following:

a. Reactor water PH >

b.. Reactor water conductivity

c. Steam line nitrogen-16 activity

d. Reactor water activity

- QUESTION 5.06 (1.00)

Consider the equation below and answer the following: .

S CR = count rate of neutrons CR = ------

S = source strength 1-K K = Keff

a. Which term (s) determine (s) the total neutron production RATE?

b. CHOOSE ONE. With a Keff<1, succeeding generations of neutrons will

[ increase / decrease 3 in population, at a [ decreasing / increasing] rate.

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

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QUESTION 5.07- (1.50)

A surious HPCS initiation signal at 15% power produces a more significant thermodynamic result than a spurious initiation signal at 85% power.

a. Identify this statement as TRUE/ FALSE.

b. Explain your answer in part a.

QUESTION 5.00 -(1.00)

During operatipn the Rx Recirculation loop temperatures are indicating 525 deg F while the dome pressure is at 985 psig. State the VALUE of the temperature difference between the dome and loops AND briefly EXPLAIN the reason fcr this difference.

QUESTION 5.09 (2.00)

Refer to figure #619 " Plant Response to Control Rod Withdrawl in Power Range".

a. For points labeled 1-5 (choose one for each)

1. fuel temperature at Point 2 is (Higher / Lower) than at Point 1.

2. the heat transfer rate at Point 2 is (Higher / Lower) than Point 1.

3. the most significant negative reactivity contributor stopping the power rise,at Point 4 is the (void / fuel) coefficient.

b. Briefly- explain why Reactor pressure levels of f at Point 5.

QUESTION 5.10 (1.00) j During full power. cycle operation you note that the control room LPRM indications are reading at 100% of meter scale. Explain why this would be an abnormal reading?

s QUESTION 5.11 (1.00)

State what changes might need to be made to the ENC threshold power values if the GGNS-1 core were forced to operate for more than 1000 MWD /MTU j during cycle-2. Include in your answer, limiting VALUES and FACTORS which determine threshold power level adjustments.

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

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Dz__IBED$Y_9E_N996E98_E9NE8_E69NI_QEE89119Nt_E6Q1ppz_9NQ PAGE. 5 IMEBMQQyN9519S e

. QUESTION 5.12 (2.00)

J Attached figure #630 illustrates a transient that could occur at a BWR. l GIVEN: (1) feedwaterncontroller failure (130%)

(2) bypass valves available (3) and of cycle one (4) SRVs higher than nominal setpoints (5) no operator action Explain the cause of the'following recorder indicat, ions:

a. Vessel steamflow increase from 0 to *12 secs.

b. Neutron flux drop at *11.7 secs.

c..Vassel water level rise from *14 to *17 secs.

d. Bypass valve flow at >16 secs.

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QUESTION 5.13. (2.50) -

For EACH of the following sets of conditions EXPLAIN which one would result in the greatest reactivity change due to control rod insertion.

a. An area of high relative flux vs. Iow relative flux. .

. b. Edge of the core vs. middle of the core. .

c. Rod A (inserted) vs.. rod B inserted beside rod A. (all other rods out) 4 9

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

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QUESTION 6.01 (1.00)

Refer to figure 603, which one of the graphs A-D correctly deplicts the Recirculation Flow Control System, Master Controller output signal, based upon the demand signal shown?

QUESTION 6.02 ( .50)

Answer TRUE or FALSE With both recirculation pumps running or both the recirculation pumps off, indicated core flow is a summation of loop flows.

QUESTION 6.03 (2.00)

For each of the following abnormal conditions, state whether or not the LPCS would inject rated flow into the reactor vessel upon a valid initiation signal and Rx pressu; e decreasing below pump shutof f head.

Note: all other system components function properly,

a. FOO1, Pump suction valve from the Suppression-Pool is closed.

b. FE-NOO3, Flow transmitter, has a failed 'HIGH" High Pressure Tap.

. c. 25% of the suction strainer area has become clogged by debris.

d. Instrument air to the Testable Check Valve has been ccmpletly interrupted.

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

l 6t__ELONIigygIgBE_QEgipN _CQNIBQ(g_@NQ_JNgIBUDENI@IlgN i PAGE 7-a ; .

QUESTICN 6.04 (2.50)

The following conditions exist relative to the E-H Control Systems IPC output at 956.0 psi Load Demand "off" Recirc Flux Control " Manual" 110% Failure "off"

' Speed Demand 1812 rpm

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Biases "A" 8: "B" set at 3%

MHC Start DEV meter 100%

Refer to figure #598 " Valve Lift Control Circuit" and answer the f ollowings a.. Given the above conditions, what is the MAXIMUM value the operator can increase.the load demand signal (once turned on,in MWE) without receiving an "EHC IN/L LRL" light on P690-9C7 F

b. A failure occurs that results in the " feedback" signal-from the E/H converter, decreasing it to a zero signal. For each of the following provide the initial value as a result of this f ailure:

1. MIN _1 output (i n MWE)

2. MIN 2 output (in MWE)

3. Summer #1 output (in .% . steam flow)

QUESTION 6.05 (1.50)

For EACH of the following ADS /SRVs state the MODE that the valve will operate in, if any, in response to EACH of - the conditions listed below.

Notes' Position referred to is in the control room, all other swiches are normal, address each valve for each condition.

1. An ADS /SRV in the '"OFF" position.

2. An AND/SRV in the " AUTO" position.

a) An ADS initiation signal it eceived.

b) Rx pressure sensors signal valve opening.

c) Rx pressure exceeds the safety mode setting.

QUESTION 6.06 (1.00)

a. Explain the basis for the vacuum breaker installed on the SRV discharge line to the Suppression Pool.

b. EXPLAIN the consequence, should the SRV vacuum breaker fail to open when required.

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

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QUESTION 6. O . (2.00)

The HPC Diesel. Generator' Mode Select Switch'was inadventently left in the "rauem rosition (all other controls normal). A valid LOCA signal is received. HPCS Bus 17AC becomes deenergized and the operator takes the following mitigating actions:

---Placed the HPCS Pump Control Switch on 1H13-P601 to TRIP.

---Places the HPCS D/G Mode Select Switch to Automatic.

EXPLAIN the resultant' responses of BOTH the HPCS D/G AND the HPCS Pump to these actions. Limit your discussion to initiating and start /stop signals.

o QUESTION 6.08 (1.00)

-The HPCS, Diesel Generator controls are in the following alignment:

' Engine lockout - reset

-Generator lockout - reset Voltage Regulator in " Manual" Unit Mode Selector in " Automatic" All other controls / power supplies are in their normal alignment. ,

a. Explain the effect of this alignment on the AUTOMATIC start capability of the HPCS D/G.

b. The operator repositions the Unit Mode Selector to the " Manual" position

and attempts a MANUAL. start. Would the HPCS D/G start '(yes/no)?

QUESTION ~ 6.09 (1.50)

With regard to the RHR Steam Condensing Mode answer the followings

a. The system is operating in accordance with SOI 04-1-01-E12-1, EXPLAIN the control method by which overpressurization of the RCIC suction line ,

is prevented. { include signal conditioning equipment, controller (s) and/or valve (s)}.

b. State the system alignment which prevents a total loss of NPSH, in the event that the RHR Heat Exchanger supply valves (F065A&B) to RCIC fail closed.

c. Why might if be neccessary, due to operating in the Steam Condensing

_ Mode, to align one RHR pump and one set of Heat Exchangers in the Suppression Pool Cooling Mode?

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

61__EL9NI_SigIEDS_ DESIGNi _CQhIBOL_8ND_INSIBUDENI8IIgN 1 PAGE 9 QUESTION 6.~10 (1.00)

During RCIC. system automatic operation a turbine trip signal was received and has been reset. The operator is preparing to reopen the Trip and Throttle valve. EXPLAIN the operation of the Ramp Generator / Signal Converter as the operator reestablishes RCIC operation by re-opening the Trip and Throttle valve.

Note Flow controller in AUTO, no other steam supply valves have operated.

QUESTION 6.11 (2.00)

The reactor is at '45% power as indicated by the APRM's. The Reactor Operator is withdrawing rod 28-53 from position 12 to position 48 as required by the sequence A rod pull sheet. As the rod passes through position 18, the rod block annunciator alarms and the rod settles to position 20. A withdraw block is indicated on the Operatcr Control Module's Rod Motion Section.

a. Explain what caused the rod block.

b. Explain why the system enforces a rod block under these conditions.

QUESTION 6.12 (2.00)

Answer the following questions in regard to the RPV pressure instrumentation. ,

a. PT-(PIS)-N068A has failed low; A Rx overpressure condition subsequently occurs (greater than the relief setpoint for an SRV. Which solenoid (s)

(A,B) on the SRV would energize AND which pressure transmitter (s) will cause the A,B solenoid (s) to energize?

b. At rated Rx pressure, should a failure occur on the ABOVE CORE PLATE pressure tap, which causes a depressurization of the input to the CRD system, EXPLAIN the responses, if any, of:

1. CRD drive water dif f erential pressure INDICATION

2. CRD drive water Pressure Control Valve (FOO3)

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

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PAGE' 10 RAD 1Q(QQ1Q@L.,_QQNT8Q( .l q2

-' QUESTION '7.01. -(1.00)

In accordance:with 04-1-01-M51-1 "Dr y ..u t 1 Cooling System", should CRD cavity temperatures rise abnormally, at which one of the following . ,

temperatures could neutron. monitoring' cables and/or equipment damage begin l

'to occur? ,

a. 125.deg F i

b. 145 deg:F . ,

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.c. 165.deg F

d. 185 deg F .

1 QUESTION 7.02 (1.00) l I

Refer to figure #635. lWhich one of the following is correct con'cerning i step LP-30 of EP-14 Level / Power. Control? ;

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a. Once RPV pressure d ops below the MAFP in' table LP-T-1, adequate core cooling-is assured.

b. Once RPV pressure drops-below the MAFP sufficient steam flow through the core does not exist to, provide adequate core cooling.

c. If there are no SRVs open and' pressure remains above the table',

sufficient steam. flow through the core exist to provide adequate core cooling..

d..If at least 2 SRVs are open, and pressure remains above the MAFP, injection must be restablished in order to adequately cool the core

-and increase RPV water level.

QUESTION 7.03 (1.00)

Select which one of the below is a Saf ety Limits a.. Thermal power shall not exceed 25%'of rated at less than 785 psig and core flow less than 10%.

l t

..b~.-MCPR shall not be less than 1.06 with Rx vessel pressure greater than

-785.psig or core flow greater than 25%. ,

I At greater than 25% power MCPR will be less than 1.06 unless Rx vessel c.

pressure is less than'785 psig or core flow less than 10%.

d. MCPR shall not be less than 1.06 with vessel pressure greater than 785 psig and core flow greater ,than 10%.

~

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

f a w -~ - - . - , , , ..o a.-- .c. ~,,n -----,,,,.,n n., , , , , - - _ , , , _ . , . , , l. .- , , , , , ,

. Zz__PNQgggggg_ _NQBde61_8BN985061_EMEB@ENGL9ND PAGE 11

~'

M2106991996_G9 NIB 96 _

~

QUESTION .7.04 ~(2.00)

~The Conduct of. Operations procedure has specified four conditions any one

.of which meet the requirements o'f " Adequate Core Cooling". State these conditions.

QUESTION 7.05 (1.50)

According to.01-S-06-2 " Conduct of Operations", the manipulation of controls at GGNS by anyone who is not a licensed RO or a SRO is permitted only when specific conditions exist. State two conditions which must exist to allow another individual (non-licensed) to manipulate the controls.

\

! QUESTION 7.06 (1.30) 1 In regard to RPP-OB-S-01-24 " Radiation Work Permit", answer the followings

a. .Under emergency conditions what may be substituted for an RWP7

b. Refer to figure _#628. Complete the missing information for blocks 1 -

4, Waiver Guidelines.

QUESTION 7.07 (1.00)

SOI 04-1-01-L11'-1 " Plant DC Systems", has the operator check f or grounds on DC bus 11DA in the following manners (1) Verify approximately 62v DC on V1 and V2.

(2) Depress the MID POINT OFFSET pushbutton and interpret voltmater readings:

V1 V2

1. 62 62

2. O 125

a. For each of the above conditions 1- & 2, state what type of ground exist e.g. no ground, positive bus and/or negative bus ground.

b. For the above circumstances, had the voltmeter reading been between the given values, what condition would this indicate?

l l

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

-Zs__P8QQEQQ8ES_ _NQ85@6t_@BNQ85@6t_EME8GENCY_@NQ PAGE 12

- - - Be9196991G9L_GONTROl; QUESTION 7.08 (1.00)

Hotwell level' control is in Manual, the condensate makeup bypass valve is being used to control hotwell level. It is.not possible to determine, by flow measurement, if adequate reject flow exist for CRD pump suction.

According to 04-1-01-N19-1'" Condensate System", how can it be determined that adequate reject flow does exist?

QUESTION 7.09 (1.50)

Refer to figure #636.

~

a. Under what condition (Suppression Pool temperature) does the operator leave step SP/T-177

b. After satisfying step SP/T-17 how does the operator continue in the l procedure?-i.e. to what procedure / step (s) does the operator procede, )

and what conditional items apply. ,l QUESTION 7.10 (1.50) 1

~

a. List the radiation dose standards for the following as stated in 10 CFR 20 for a restricted area without a completed NRC form 4.

REM /calander Qtr.

-1. Skin of whole body

2. Hands and forearms, feet and ankles

3. Whole body, head and trunk, active blood forming organs, lens of eyes or gonads

b. What 3 conditions must be met prior to exceeding the whole body limit stated above?

QUESTION 7.11 (1.50)

Refer to figure #638.

Explain why RCIC injection is terminated at step SP/L-19 of EP-3 AND what adverse consequences could result from continued injection of RCIC.

9

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

Zd__P89CEQUBEE_:_N98596t_@pNQBM@61_EME8@Eggy_9Np PAGE 13

-899196991CB6_C9 NIB 96

.l l

1 QUESTION -7.12 (1.00) l l

The CRD Hydraulic. System Operating Procedure requires the CRD Flow Control Valve to_.be closed before starting a CRD pump., What is the possible af f ect of starting a pump with this valva open?

QUESTION 7.13 (1.50)

The Power Operations procedure requires all turbine bypass valves to be

' fully closed when withdrawing control rods with reactor power above the Low Power Setpoint. What is the basis for this requirement?

QUESTION 7.14 (1.00)

Per procedure 03-1-01-3, " Plant Shutdown", when the plant is shutdown to a " hot shutdown" cor.dition the moderator temperature should be reduced to about 400 deg. F. What is the reason for reducing moderator temperature to this point?

l i-l (***** END OF CATEGORY 07 *****)

l

.._..r.m - _ _ . . . , _ ._ _ , . - _ . . . - _ _ _ _ , . _ , . . . _ - . , -_

l O___09 MINI @IB911yE_BBgCEQUBE@t_CQNpillgN@t_@NQ_LIMII@IlgN@ PAGE 14 ;

1 QUESTION. ,8.01 (1.00)

Which one of the following is NOT a mandatory requirement for an temporary Item Control Area (ICA)? ll

a. The area must have restricted / limited access.

b. The area must be controlled by a licensed SRO.

c. The area must be clearly posted as an ICA for SNM.

d. The area must be posted, if necessary, as a radiation area per 1 applicable HP directives. I 1

1 QUESTION 8.02 (1.00) j Which one of the following individuals DOES NOT approve SNM Movement Plans for core alterations 7

a. Refueling SRO

b. Technical Superintendent ,

c. Rx Engineering Superintendent

d. GGNS General Manager c.- Manager, Plant Operations QUESTION 8.03 (1.00)

Which one of the following DOES NOT consitute an unreviewed safety question involving 10 CFR 50.59 when evaluating changes, test and experiments to the facility or facility procedures?

a. if the, margin of safety as defined in the basis for any technical speci f i cati on is reduced.

b. if a possibility for an accident or malfunction of a different type than any evaluated previously in the SAR may be created.

c. if the probability of occurance or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the SAR may be increased.

d. ,if'the suspension of a Limiting Condition for Operation, identified in Technical Specifications, is completed pursuant to Section 3/4.10 "Special Test Exceptions" during applicable operational conditions.

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

O

m2 m 0 __eDg1N1SIBBIlME_PBQCEDUBES _CQND111gNS t t_@ND_ LIMIT 6TIONS PAGE 15

- - 3, QUESTION- 8.'04 (1.00)'

Refer'to attached Tech Specs 3/4.5.1, TSPS 078 L 082 and Table 3.3.3-1.

Unit 1 is in Operational Condition 1, at 75% RTP, Tech Spec 3/4.5.1 action d.1 is being applied to unit operation.

Which one of the below sets of circumstances is the cause for implementing this action statement?

a. LPCI pump 'A' start TD relay is inoperativegTwo channels of HPCS low CST' level logic are incapable of being tripped 3 ADS timer trip system c 'B' is inoperative.

b. LPCI pump 'A' motor has an overcurrent lockout conditiongOne LPCI pump

'B' ADS high pressure permissive is inoperativeg DG 12 is inoperative.

'is inoperative 3 LPCI pump

c. ADS timer l trip system 'A' 'B' motor has an overcurrent lockout condition;One channel HPCS CST low level logic has been tripped.

d. ADS. trip system 'A'~is inoperable;RCIC Trip & Throttle valve has failed closedgLPCI pump 'B' has an overcurrent lockout-condition.

-QUESTION 8.'05 (1.00)

Whose signature is required to authorize batch release of waste fluids?

Assume a release less than .25 mpc.

a. Shift Chemist

b. H.P.. Supervisor

c. Shift Supervisor or Superintendent

d. Radwaste Supervisor s

QUESTION 8.06 (1.00)

A Notification of an Unusual Event would be initiated if an injured worker was either overexposed to radiation and/or contaminated and also

a. was-sent home.

b. required hospitalization.

c. exceeded the weekly exposure limit.

d. had internal contamination.

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

8 __8pd1NISI68IIVE_PBgCEQUBE@t_CgNpillgNSz _9Np_LidIIBIlgNg PAGE 16 QUESTION 8.07 (1.00)

Concerning Emergency Action Levels, a Site Emergency declaration is made if ALL AC power is lost for

a. greater than 15 minutes.

b. less than 15 minutes.

c. greater than 10 minutes.

d. No EAL, ESF restores AC power.

QUESTION 8.08 (1.00)

Refer to attached Tech Spec LCOs.

Condition 3 activites have continued for 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months above a Rx dome pressure of 135 psig, with the f ollowinc deficiencies on) .

One MSIV is inoperable RCIC is inope'rable Which one of the following actions most accurately outlines the allowances and/or limitations imposed by Tech Specs in this instance 7

a. Condition 3 must be maintained;depressurization and/or entry into Condition 4 acceptable.

b. St'artup activities may continue; Condition 2 may be entered but not exceeded.

c. Startup and power operation into Condition 1 may be acomplished provided that the affected MSL is isolated.

d. Startup activities may continue into Condition 2 provided subsequent restoration of RCIC to operable status within 14 days.

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

F .

ADMINISTRATIVE PROCEDURES.. CONDITIONS. AND_ LIMITATIONS PAGE 17 QUESTION 8.09 (1.00)

Diesel Generator 11 which supplies bus 15AA is' inoperative. LPCI pump 'B' is also inoperable. The Tech Specs for ECCS (TSPS 80) and AC power sources are attached. Which statement below is correct concerning continued operation in Condition 37 a.-The action statements for both LPCI and DG are applied independently.

b. Since neither system is required in Condition 4, the unit must be taken to Condition 4 within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

c. LCO 3.0.3 applies.

d. LCO 3.8.1.1 action c applies.

QUESTION -8.10 (1.00)

Turbine overspeed protection is required by Tech Specs because an overspeed condition could cause

a. excessive overfrequency condition on electrical power systems, supplying safety related equipment,

b. turbine components to become missiles which may damage safety related equipment.

c. the Rx thermal power to exceed the limits in the Units license,

d. the Rx to go prompt critical.

QUESTION 8.11 (1.00)

When a fire suppression Spray / Sprinkler system is declared inoperable for the portion that protects the Diesel Generator Building while the DGs-are required to be operable, the required action is to (choose one)

a. Commence a Unit shutdown within one hour.

b. Establish a bi-hourly fire watch patrol for the affected area.

c. Establish a continous fire watch with backup fire suppression equipment around redundent equipment within one hour.

d. Log ambient temperature readings for the affected area hourly.

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

# Ez_16EdlEIEI6SIISE_EOSEEENOEEE_ESdSil19 doi _6EE_EldlIOIl90E PAGE - 119 QUESTION- 8.12 (1.00)

H l f

SY During Unit 1 startup with the Rx at 2%' power (condition 2) , . one abnormality exist: An APR channel is f ound to be inoperable.

Which one of the following tatements is correct? (Tech Specs attached)

a. Operation in Condition 1 i not allowed until the inoperable channel is repaired and declared op rable.

b. Be in at least hot shutdown ithin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c. The channel need not be place in the tripped condition; restore to cperable within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be n hot shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

d. Place the channel and/or trip s tem in the tripped condition within 1 hourgOperation in Condition 1 s allowed.

QUESTION 8.13 (1.00)

a. For operations personnel to be considered eligible to independently

. verify an action in accordance with 01-S-06-29 " Independent Verification Program", the person must be quali f i ed as a mi ni mum _________. (f ill-in Title)

b. TRUE/ FALSE. An emergency condition eliminates all requirement for Independent Verification.

QUESTION B.14 ( .50) ,

TRUE/ FALSE. Roping off, posting and activating a flashing light (in an extra high radiation area >1000 mrem) will replace the requirement that locked doors be provided.

l l

l QUESTION 8.15 (1.00)

Refer to attached Power Distribution Tech Spec LCO's.

For each of the following identify which LCO would apply to the given condition.

.a. Bundle 43-60 has an FLPD of 1.136.

b. Bundle 35-56 has an APRAT of 1.013.

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

.e 1

Ozf_6gM181EISBIlyg_BBggggUBgE1_ggBgIIlggg1_Gyg_(101IGIIgyg PASE 19 QUESTION 8.16 (1.00)

a. Who (by position) controls authorization to the fuel handling area in the Auxiliary Building'during fuel movement operations?

b. Where is the official copy of the SNM transfer form kept during refueling operations?

QUESTION 8.17 (1.00)

What_are the Two (2) provisoes / stipulations that must be met in order to allow "out of sequence" completion of IDI procedural steps?

QUESTION 8.18 (1.50)

Tech Specs defines Shutdown Margin as the amount of reactivity by which the Rx is subtritical or would be subtritical assuming...

State the three assumptions and all conditional requirements which complete the definition of Shutdown Margin.

I l

(***** END OF CATEGORY OB *****)

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

l 1

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2 g = mc ht gg = 1/2 av 2 , ,gy, , f,)fg 4 , 3; g , 3,,

PE = agn yf = y, + at n = e/t a = an2/t1/2 = 0.693/tijp y , , .p A* nD 2 t

l/2'ff ' [(t l){t b))

4 [(t1/2)+(t)) 3 aE = 931 am ,,y go ,tx av I = I,e o . ph Q = mCpat 6 = UA T I = I ,e"*

Pwr = Wfah I = 1,10**/D L TVL = 1.3/v 8

P = P 10 "'I*) HVL = -0.693/v t

P = P,e /T SUR = 26.06/T SCR = 5/(1 - K,ff)

CR, = S/(1 - Kdfx)

SUR = 26e/t= + (s - e)T CR j (1 - K,ffj) = CR2 II ~ "eff2)

T = (t=/s) + [(s - sy Ie] -

M = 1/(1 - K,ff) = CR)/CR, T = s/(e - s) M = (1 - K,ff,)/(1 - K,ffj)

T = (s - e)/(Is) SDM = ( -Kgf)/Kdf a = (X,ff-1)/K,ff = 4Keff /K df t= =.10 secones I = 0.1 seconds"I -

e = [(t=/(T Kgf)] + [s df (1 /

+ IT))

Ijj=1d d

Id j 2 ,2 gd 2

P = (seV)/(3 x 1010) 22 1 = eN R/hr = (0.5 CE)/d2 (,,g,73)

R/hr = 6 CE/d2 (f,,g) .

Water Parameters Miscellaneous Conversions 1 gal. = 8.345 lem. I curie = 3.7 x 1010gp, 1 ga;. = 3.78 liters I kg = 2.21 lbm 1 ft' = 7.48 gal. 1 hp = 2.54 x 10 3 8tu/nr .

Density = 62.4 lbg/ft3 1 av = 3.41 x 106 Stu/hr Density = 1 gm/cW lin = 2.54 cm l Heat of vaporization = 970 5tu/lom 'F = 9/5'C + 32 Heat of fusion = 144 8tu/10m . 'C = 5/9 ( *F-32) 1 Atm = 14.7 psi = 29.9 in. Hg. 1 BTU = 778 ft-Ibf I ft. H O 2

= 0.4335 lbf/in.

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$80 0.9492 0.01613 265.4 265.4 77.98 1031.4 1109.3 0.1472 1A105 1.9577 ut 110 1J750 0.01617 203.25 203.26 87.97 1025.6 1113.6 0.1646 1.7693 1.9339 130 130 1A927 OA1420 157.32 157.33 97.M 10193 1117A 0.1817 1.7295 13112 130 330 2.2230 DA1625 0.1985 1.6910 1A895 140 0.01629 122.98 123.00 107.95 1014.0 1122.0 140 2A892 0.2150 1.6536 1A686 190 3.718 0.01634 97A5 97A7 117.95 L 2 1126.1 150 127.M IU' 1130.2 0.2313 14174 1A4s7 160 4.741 OA1640 77J7 77.29 360 62.04 42.06 137.97 996.2 1134.2 0.2473 1.5822 1A295 170 370 SA93 OA1645 1A111 50.21 50.22 148A0 990.2 1138.2 02631 1.5480 ISO 180 7.511 OA1651 190 OA1857 40.94 40.M 158A4 984.1 11/2.1 0J787 1.5148 1.7934 390 9.340 977.9 !! 44.0 02940 1.4824 1.7764 300 300 11.526 OA1664 33.62 3344 168 09 0.01671 27A0 27.82 178.15 971.6 li49.7 0.3091 1.4509 1.7600 210 310 14 123 26.80 180.17 970 3 1150.5 CJ121 1A447 1.7568 212 232 14.696 0.01672 26.78 23.3 188.23 M 5.2 1153.4 0.3211 1A201 1.7442 32s 820 17.186 0A1478 23.13 19.381 198J3 958.7 1157.1 03308 SJ902 1.7290 330 330 20.779 0.01685 19.364 16.221 208.45 952.1 1160.6 0.3533 1.3600 1.7142 See 340 24.968 0.01693 16.304 13 319 218.59 945.4 1164A 0.3677 1.3323 1.7000 350 350 29A25 0A1701 13302 11.762 228.76 938.6 1167A 0J819 1.3043 1.4062 350 260 35.427 OA1709 11.745 270 0.01718 10.042 10.060 A38.95 931.7 1170.6 0J960 1.2760 1.6729 370 41.856 1173A 0.4098 12501 1.6600 See 49.200 CA1726 8.627 8.644 249.17 924.6 300 1176.8 0.4236 1.2238 14473 age 57.550 OAl?36 7.443 7A60 259A 917.4 390 910.0 1179.7 04372 1.1979 1.6351 300 i

300 67.005 OA1745 6.448 6.4M 269.7 5A26 200 4 902.5 1K2.5 0.4506 1.1726 1A232 318 310 77.67 041755 5.409 320 4.914 290.4 994A 1185.2 0.4440 1.1477 1A116 320 39.64 OAl?H 4A96 340 3.770 3.788 311J 878 3 1190.1 SA902 1A990 1.5002 340 117.99 0A1787 0.5161 1A617. .1.5678 300 1.939 2.957 337.3 862.1 1194.4 360 153.01 OA1811 1A067 1.5473 300 2.317 2.335 353.6 844.5 1198.0 0.5416 380 195.73 OAISM 825.9 1201.0 0.5067 OA607 1.5274 400 0.01864 13444 1A630 375.1 400 247.26 806.2 1203.1 0.5915 0.9165 1.5080 420 305.78 0.01894 1.4808 1.4997 396.9 420 785A 1204.4 0.6161 OA729 1.4890 44C 440 381.54 0.01926 1.1976 1.2169 419.0 l 03746 0.9942 441.5 763.2 1204.8 0.6405 02299 1.4704 46C 460 466.9 0.0196 0.6648 0.7871 1A515 est 0.0200 0.7972 03172 464.5 739.6 1204.1 480 M6.2 7143 1202.2 0.E890 0.7443 1.4333 Sac 0.6545 0.6749 487.9 500 6803 0.0204 687.0 11994 0.7133 0.7013 1A146 52C 520 812.5 0.0209 0.5386 0.5596 512A 536 8 657.5 1194.3 0.7378 0.6677 1.3954 54C S40 962.8 0.0215 0.4437 0 4651 S6C M 2.4 625.3 1187.7 0.7625 0.6132 1.3757 560 1133.4 0.0221 0.3651 0.3871 Sec 589.1 589.9 1179.0 0.7876 0.5473 1.3550 S40 1326.2 0.0228 0.2994 0.3222 550.6 1167.7 OA134 0.51M IJ330 tes 0.0236 0.2438 0.2675 617.1 600 1543.2 506.3 1153.2 0A403 0.4689 1J092 Gac 0.0247 0.1962 0.2208 644.9 Get 420 1786.9 679.1 4544 1133.7 OA664 OA134 1.2821 640 2059 9 0.0260 0.1543 0.1802 1.2458 $$t 714.9 392.1 1107.0 03995 0.3502 640 2365.7 0 0277 0.1166 0.1443 1.2004 Get 0.1112 758 5 310.1 1068.5 CAM 5 0.2720 640 2708.6 0.0304 0.0008 0.0752 822.4' 172.7 995.2 0.9901 0.1490 1.1390 fot 700 30943 0 0366 0.0386 0 1A612 70!

0 0.0508 906.0 0 906.0 1A612 705.5 3208.2 0 0508 TABLE A.2 PROPERTIES OF SATURATED STEAM AND SATURATED WATER (TEMPERATURE)

A.3

9eleme.1P/m tath*8py.Stettb takopt. DWe ai tasey, gyln

. P'*** WF tretee toep tesem Ceter temp Steam test toop geoom 9 pease gleem 4 P,g,4-peas em 8, e, r, e, 4, A s 8, aj, ab ag 3302A 9 00 8075.5 3076 5 9 f.1872 3.3473 g geg13 gag,g e.e886 32.018 0.01402 3302A 3945.5 3 03 1973 8 19769 00061 2.1705 9.1796 333 8e22.3 Ele e.le 36.0t3 901402 3945.5 0 0271 3.1140 f.1411 41.453 3.01602 3004.7 7004 7 1340 1067.9 1081 4 1340 e.16 15n 3 31.27 3DH 5 1084 7 0 0472 20728 3.1160 3122 132,57 103 3 e.1,5 g,,

e.20 63.160 te1603 1626.3 0 0641 2 0165 3.0809 0.30 64 444 0.01604 1039 7 1039.7 32.54 1057.1 1089 7 3234 1032 0 eJ0 340 72369 0.01606 792.0 792.1 40 92 10524 1093.3 0.0799 1.9762 2.0M2 40.tr 10H 7 0.40 e3 79.586 0 01607 641.5 6413 47.62 1048 6 1096 3 0 09n 1.9446 3.0370 4742 tou t 3,5 0 01609 640 0 6401 H 25 1045 5 1098 7 01028 1 9186 2.0215 51.24 1038 7 06 eA 85.718 58 30 3042 7. .J1po4, 03 ,..tSu6,,3.0083 ,,14,10,,1060L3 eJ'

.- . - e.y - 90 09.9 001610 en 93 .466 94 e2 94 38 0.01631 411.67 411 A9 62 39 1040 3 1102 6 01117 137M 1.9970 8229 1041.7 ' ' g's M24 0 01612 368 41 36843 66 24 1038 1 1104.3 0 1764 13406 1.9870 K.14 1042.9 et e.9 1.0 101.74 0 01614 333 59 313 60 69.73 1036.1 11058 0.1326 13455 1.9781 3923 1064,1 g3 RA 126 07 0.01623 173.74 173 M 94.03 1022.1 1116.2 0 1750 1.7450 1.9200 94A3 1051A 3A 3.0 14147 0 01630 lit 71 11873 109.42 10132 1122 6 02009 11854 14864 109 41 1056.7 S.0 4.0 152.96 0 01636 9063 90 64 120.92 1006 4 1127.3 0.2199 1A428 13626 120.90 1060.2 4.0 8.0 162 24 0.01641 73.515 73 53 130 20 1000.9 1131.1 0.2349 1.6094 13443 130.14 1063.1 E.0 6.0 170 05 0.01645 61.967 61.98 1M 03 996.2 1134.2 0.2474 15820 12294 138Al 1065 i 6.0 7A 17684 0 01649 53 6M 53.65 144 83 992.1 1136 9 02581 1.5587 1A168 144Al 1067A FA 3.0 182 86 0 01653 47.328 47J5 150 87 988 5 1139.3 02676 1.5384 12060 15024 1069.2 S.0 9.0 182 27 001656 42385 42 40 1%.30 985.1 1141.4 0.2760 13204 1.7964 1H28 1070.8 9.L 10 193.21 0.01659 38.404 38 42 161.26 982.1 1143.3 02836 1.5043 1.7879 16123 1072A 10 14.696 212 30 0 01672 26 782 26 to 180.17 970.3 1150.5 0.3121 1A447 1.7568 180.12 1077A 14.896 15 213 03 0.01673 26 274 26.29 181.21 969.7 1150.9 0.3137 1A415 1.7552 181.16 1977.9 15 0.01683 20 070 20 0b7 196 27 960.1 11 % 3 0.3358 1.3962 1.7320 196.21 1082A 30 30 227.96 30 250.34 0 01701 13.7266 13 744 218 9 945.2 1164.1 0.3682 1.3313 1.8995 1183 1087.9 30 40 26725 0 01715 10 4794 10 497 236.1 . 9336 11693 0.3921 1.2844 1A765 2MA 1092.1 40 0.01727 84967 8 514 250.2

923.9 1174.1 0 4112 1.2474 )4585 250.1 10953 80 80 261 02 to 292.71 0.01738 7.1 % 2 7.174 262.2 915.4 1177.6 0.4273 1.2167 1A440 362A 1098A 00 70 302.93 0.01748 61875 6205 272.7 907A 1180 6 0 4411 1.1905 1A316 272.5 11002 70 to 312.04 0.01757 5 4536 5 471 232.1 ' 900.9 1183 1 0.4534 1.1675 1A208 281.9 1102.3 80 90 320.28 0 01766 4A777 4.895 293 7 894 6 1185.3 0 4643 1.1470 1 A113 290.4 1103.7 90 100 32722 0 01774 4.4133 4.431 2M.5 388 6 1187.2 0.4743 1.1284 1A027 2982 1105.2 100 120 341.27 0.01789 3 7097 3.728 312.6 8772 1190 4 0 4919 1.0960 1.5879 312.2 1107A 120 340 353 04 0 01803 3 2010 3 219 325 0 368.0 1893 0 0 5071 1.0681 1.5752 324 5 1109.6 140 363 55 0 0;815 2A155 2334 3361 859.0 1195.1 0.5206 1.0435 1.5641 335A 1111.2 Mo 360 ISO 373 08 001827 2.5129 2.531 346 2 850 7 1196.9 05328 1.0215 1.5543 3454 11123 180 2.287 355.5 842A 1198.3 0 5438 1.0016 1.5454 3542 ,2113.7 300 300 35130- 0 01829 2.2689 250 40097 0 01865 12245 12432 376.1 825 0 1201.1 0 5679 0 9545 1.5264 375A ~ 3115A 380 300 417 35 0 038E9 1.5233 1.5427 394 0 806 9 1202 9 05882 fl9223 13105 3922 1117.2 300 350 411 73 0 01913 1.3064 1.3255 409.8 7942 1204 0 0 60 % 0 89>) 1A968 408 6 1818 1 330 1.1610 424.2 780 4 1204 6 0 6217 0 8630 1.4847 422.7 111E 7 400 400 444 60 00193 1.14162 450 4M 28 0 0195 1.01224 1.0318 437.3 767.5 1204.8 06360 0A374 1A738 4352 1118.9 4E0 0 90787 09276 449 5 755.1 1204 7 0 6490 0.8148 1A639 447.7 1118 3 900 500 4E 7 01 00193 %0 850 47693 00199 0 82183 0 8418 460.9 743.3 12043 O M11 0.7936 1A547 456.9 11186 600 485 20 0 0201 0 74962 0.7698 471.7 732.0 1203 7 0.6723 0 7738 1A461 469 5 111f.2 800 0.63505 0 6556 491.6 710.2 1201.8 0 692R 07377 14304 488.9 1116.9 700 703 . 5".t3 08 0 0205 833 514 21 0 0209 054809 0 M90 509.8 689 6 1199 4 0 7111 07051 1.4163 5067 1115.2 000 04796S 0 5009 526 7 669 7 1196 4 0 7279 06753 1.4032 523 2 1113.0 900 900 53195 0 0212 2000 544.H 0.0216 042435 0 4460 542.6 f 50 4 1192 9 0.7434 06476 1.3910 53* & 11104 1000 04006 557.5 631 5 1189 I O M78 06216 1.3794 553.1 1107.5 1100 1100 5%2d 0.0720 0 378f 3 1200 00223 0 34013 03625 571.9 613.0 1184 8 07714 0.5969 1.3683 566 9 1104.3 3200 g $57.19 544A 11to 2 0.7843 05733 1.3577 580.1 1100 9 3300 1300 E77.42 0 0227 0 30722 0.3299 585 6 5988 5765 1175 3 0 7966 0 5507 13474 592.9 1037.1 1400 1400 557 07 0 0731 0 278/1 03018 1500 5 % 20 0 0235 02b372 0.2712 611.7 5504 31701 0A055 0 5283 1.3373 6052 1093.1 1500 2000 635 80 00h? 0 16760 0 l883 472 1 466.2 1138.3 08CS 0 42 % 1.7881 . M24 10G5 6 2000 2500 0 02c,E 01307 731 7 361 6 1093 3 09139 0 3206 1.2M5 718.5 1032.9 2500 66211 O1020's 218 4 1070 3 0 9728 01891 1.1619 782 2 973.1 3000 3000 655 33 0 0343 0 050/3 0 0850 801 8 0 906 0 1.0612 0 1.0612 875.9 BM.9 3708."

32982 70147 00%8 0 0 0502 906 0 _

TABLE A.3 PROPERTIES OF SATURATED STEAM AND SATURATED WATER (PRESSURE)

A.4 l

2 3 S

(aKIK) i TIME

. STOP ROD PULL M -

START ROD PULL VOID FR ACTION 3 REACTOR $

i PRESSURE p

3-4 5 -

3 .*

REACTOR POWEPR LEVEL S s

P O TIME e

REACTOR PERIOD N g TIME s __

TURSINE 3

STEAM FLOW i

- nw I

Figure -619 Plant Response to Control Rod Withdrawal in Power Range

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TIME (SEC)

FIGURE 630 FEEDWATER CONTROLLER FAILURE, MAXIMUM" DEMAND WITH TURBINE BYPASS NTROLP - OP-DT-520 REV. 1

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RAISE /L067.R A L+

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, ' Lesson: . Ra ci a' tic n' Wo rr Pe rmi t 7 Course: Radioactive Material Control Procedures w

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AREA TO BE MAX. LENGTH MIN. HP- ESCORT ENTERED OF WAIVER SUPERVISION REQUIRED COMMENTS APPROVAL HIGH RAD. AREA, 1 WK. HP. SUPV. HP OR NEUTRON AREA RW III (jW[()1ULD MREM / DAYNOT EXCEED v

HIGH CONT. AREA, 1 WK. RAD. CON. HP ONL_Y ALLOWED IF POT. AIRBORNE SUPV. INDIVIDUAL HAS AREA PREVIOUS EXPERIENCE WORKING IN SIMILAR AREAS AIRBORNE AREA 2 100% MPC s ,

AIRBORNE AREA MUST BE RESPIRATOR 125%, < 100% , RAD. CON. ~HP QUALIFIED IF RESPI-MPC SUPV. RATORS REQUIRED TO

} ENTER AREA VERY HIGH RAD. 1 DAY FP- ONLY ALLOWED IF -

AREA INDIVIDUAL HAS PREVIOUS EXPERIENCE

- WORKING IN SIMILAR AREAS -

OTHER RWP AREAS 1 WK. RAD. CON. HP OR SHOULD NOT NOT LISTED SUPV. RW III 100 MREM / DAY ABOVE i

c. .

FIGURE 628 TRAINING REQUIREMENT WAIVER GUIDELINES

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j .2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in Figure 3.2.1-1.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than cr equal to 25% of RATED THERMAL POWER.

ACTION:

With an APLHGR exceeding the limits of Figu're 3.2.1-1, initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or-reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the limits determined from Figure 3.2.1-1:

a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least.,15% of RATED THERMAL POWER, and

c. Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITINC CONTROL R00 PATTERN for APLHGR.

d. The provisions of Specification 4.0.4 are not applicable.

S i

GRAND GULF-UNIT 1 3/4 2-1

I i

o 14 l c I I I I I I FUEL TYPE l

5 CURVE p A 8CR2tO l B 8CR 160 l F C 8CRO71 T 13 -

E 12.6 12.6 12.6 Z O A A g R 12 V _ 12.4 -

112.4 / .

% 12. 12.1 j

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O 5,000 10,000 15,000 20,000 25,000 30,000 35,000 40,000 AVERAGE PLANAR EXPOSURE (mwd /t) l l

FIGURE 3.2.1-1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR)

VERSUS AVERAGE PLANAR EXPOSURE )

INITIAL CORE FUEL TYPES 8CR210, 8CR160 AND 8CR071 1

i ,

POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall not exceed 13.4 kw/ft.

' APPLICABILITY: OPERATIONAL CONDITION 1, when' THERMAL POWER is greater than or equal to 25% of RATED THERMAL-POWER. -

ACTION:

With the LHGR of any fuel rod exceeding the limit, initiate corrective action within 15 minutes and restore the LHGR to within the limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER'to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.4 LHGR's shall be determined to be equal or less than the limit:

a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER,

c. Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating on a LIMITING CONTROL R0D PATTERN for LHGR, and

d. The provisions of Specification 4.0.4 are not applicable.

l l

t i

i GRAND GULF-UNIT 1 3/4 2-7

i 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protection system instrumentation channels shown in Table 3.3.1-1 shall be OPERABLE with the REACTOR PROTECTION SYSTEM RESPONSE TIME as shown in Table 3.3.1-2.

APPLICABILITY: As shown in Table 3.3.1-1.

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 cha'nnels 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 denonstrated 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 ind simulated automatic operation of 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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

- - - _ _ _ _ m___ _ _ _ _ _

TABLE 3.3.1-1 R_EACTOR PROTECTION SYSTEM INSTRUMENTATION c> APPLICABLE MINIMUM E OPERATIONAL OPERABLE CHANNELS T 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 f 2. Average Power Range Monitor (c);

a. Neutron Flux - High, Setdown 2 3 1

' R

3(D) 3 2 5 3 3 w

4 b. Flow Biased Simulated Ther:aal Power - High 1 3 4

c. Neutron Flux - High 1 3 4

d. Inoperative 1, 2 3 1 3 3 2 j 5 3 3 1

3. Reactor Vessel Steam Dome i Pressure - High 1, 2(d) 2 1

, 4. Reactor Vessel Water I.evel - Low, Level 3 1, 2 2 1 i

i 5. Reactor Vessel Water Level-High, 1

i Level 8 1(') 2 4

6. Main Steam Line Isolation Valve -

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

^

8. Drywell Pressure - High )

1, 2 2 .1

c ,

l.

I

- GRAND GULT NUCLEAR STATION ADMINISTRATIVE PRCC;;;.72 l 01-5-15-2 iRevision 3l l Attachment III IPage 1 of I l TECHNICAL SPECIFICATION POSITION STATEMENT TSPS No.: 082

.Part 1 Originator: Moulder 1/15/86 Name / Date Technical Specification 3.5.1 and 3.5.2 -

References:

IE 85-94 Technical Specification Change Required ( M NO YES ( ) If Yes CR0 Position If the min flow valve on any ECCS system will not perform .

its intended function then the associated ECCS function is inop.

Position Affect / Comments: Example: If E12-F064A will not function automatically then the LPT1 and CTMT spray mode of RHR 'A' will be declared inop. However, S. pool cooling and SDC will not be declared inon if the min flow valve can be closed and adequate flow est. in these modes for cumo protection. The min flow valve will be deacti-varad closed and the nume breaker racked out until the min flow valve in renmired innlema clans are made to olace the system in SP cooling -

ne w ttnd specific attention is plac34 on system performance until adequate flow is established. /

)

Part 2 Compliance Superintendent Y 71 A L _

/

/ /[/ F[

~ ' 'Signatugf Date

.e Part 3 Review and Approval: Approve Disapprove (1) _ //l5[8G CK) ( )

Technical pehalandent / Date (2) b (g) ( )

s (3) ('l#Md//L/doan/

b

/ 9 RC/ et~ g5 Dateh/-/44

/' Date g ( )

- /-/ 7 -8 [ C>Q ( )

(4) g*Nanager Approval / Date

0ne of the following is required: GGNS General Manager Manager, Plant Operttions 01-S-15-2 ATT III

- - + - ~

'i s

TABLE 3.3.1-1 (Continued)

E REACTOR PROTECTION SYSTEM INSTRUMENTATION E

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

9. Scram Discharge Volume Water 2 1-Level - High -

1I9j' ,

S 2 3

10. Turbine Stop ialve - Closure I Ih)- 4 6

11. Turbine Control Valve Fast Closure, Valve Trip System Oil Pressure'- Low 1(h)- 2 -6 5

/

w 12. Reactor Mode Switch Shutdown 1 Position '

1, 2 , 2 1 )

w 3, 4 2 7' O - 5 2 3

13. Manual Scram 1, 2 2 1-3, 4 2 8 5 2 9 9

e a

D 9

9

INSTRUMENTATION TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION

~ ACTION 1 -

Be in at least HOT SHUTOOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 5 -

Be in STARTUP with the main steam line isolation valves closed within6 hours'of.;inatleastHOTSHUTOOWNwithin12 hours.

ACTION 6 -

Initiate a reduction in THERMAL POWER within 15 minutes and reduce turbine first stage pressure to less than the automatic bypass setpoint within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

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 CORF ALTERATIONS *, and insert all insertable control rods o;c 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.

%ND GULF-UNIT 1 3/4 3-4

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION

-TABLE NOTATIONS ,

(a) A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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 dJring 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 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 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 (VW0) steam flow, equivalent to THERMAL POWER less than 40% of RATED THERMAL POWER.

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

- Initial setpoint. Final setpoint to be determined during startup test program.

Any required change to this setpoint shall be submitted to the Commission within 90 days of test completion.

GRAND GULF-UNIT 1 3/4 3-5

/

, - - -. , . , , . , - , - . , . ~ , , . . _ , . - - . . _ . - -

INSTRUMENTATION 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3 The emergency core cooling system (ECCS) actuation instrumentation channels shown in Table 3.3.3-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.3-2 and with EMERGENCY CORE COOLING SYSTEM RESPONSE TIME as shown in Table 3.3.3-3.

APPLICABILITY: As shown in Table 3.3.3-1.

ACTION:

a. With an ECCS actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.3-2, 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 one or more ECCS actuation instrumentation channels inoperable, take the ACTION required by Table 3.3.3-1.

c. With either ADS trip system "A" or "B" in' operable, restore the inoperable trip system to OPERABLE status within:

1. 7 days, provided that the HPCS and RCIC systems are OPERABLE.

2. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to less than or equal to 135 psig within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.3.3.1 Each ECCS actuation 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 4.3.3.1-1.

4.3.3.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months. .

4.3.3.3 The ECCS RESPONSE TIME of each ECCS trip function shown in Table 3.3.3-3 shall be demonstrated to be within the 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 ECCS trip system.

1 GRAND GULF-UNIT 1 3/4 3-27

1 Q ,

TABLE 3.3.3-1 2,

5 ,

^

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION o

c MINIMUM OPERA 8LE APPLICABLE E CHANNELS PER OPERATIONAL d- TRIP FUNCTION TRIP FUNCTION (a) CONDITIONS ACTION z

U A. DIVISION I 1 RIP SYSTEM

~ 1. RHR-A (LPCI MODE) & LPCS SYSTEM

a. Reactor Vessel Water Level - Low Low Low, Level 1 2(b)' 1, 2, 3, 4*,'5" 30

b. Drywell Pressure - High . 2(b) 1,2,3 30

c. LPCI Pump A Start Time Delay Relay 1 1, 2, 3, 4 * , 5* 31

d. Manual Initiation 1/ system (b) 1. 2, 3, 4*, 5* 32

2. AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A"#

a. Reactor Vessel Water Level - Low Low Low, Level 1 2 1,2,3 30

b. Drywell Pressure - High 2 1, 2,.3 30

c. ADS Timer 1 1,2,3 31

d. Reactor Vessel Water Level - Low, Level 3 (Permissive) 1 1,2,3 31 -

w e. LPCS Pump Discharge Pressure-High (Permissive) 2 1,2,3 31 N f. iPCI Pump A Discharge Pressure-High (Permissive) 2 1,2,3 31

g. Manual Initiation 2/ system 1, 2, 3 32 T

N B. DIVISION 2 TRIP SYSTEM U

T. RHR B & C (LPCI MODE)

a. Reactor Vessel Water Level - Low, Low Low, Level 1 2 1, 2, 3,'4*, 5* 30

b. Drywell Pressure - High 2 1,2,3 30

c. LPCI Pump B Start Time Delay Relay I 1, 2, 3, 4*, 5* 31

d. Manual Initiation 1/ system Ib) 1, 2, 3, 4* , 5* 32

2. AUTOMATIC DEPRESSURIZAll0N SYSTEM TRIP SYSTEM "B"#

a. Reactor Vessel Water Level - Low Low Low, level 1 2 1, 2, 3 -30

b. Drywell Pressure - High 2 1,2,3 30

c. ADS Timer 1 1,2,3 31

d. Reactor Vessel Water Level - Low, Level 3 (Permissive) 1 1, 2, 3 '31

e. LPCI Pump 8 and C Discharge Pressure - High (Permissive) 2/ pump 1, 2, 3 31 Manual Initiatiorr

f. 2/ system 1, 2, 3 32 s

t I

O TABLE 3.3.3-1 (Continued)

E o EM:RM NCY CORE COOLING SYSTEM ACTUATION INSTRUNENTATION c3

% MININJM OPERA 8LE APPLICABLE

CHANNELS PER OPERATIONAL E: TRIP FUNCTION TRIP FUNCTION I ") CONDITIONS ACTION

?!

H C. DIVISION 3 TRIP SYSTEM

1. HPCS SYSTEM

a. Reactor Vessel Water Level - Low, Low, Level 2 4 1, 2, 3, 4*, 5* '33

b. Drywell Pressure - High## 4 g) 1, 2, 3 33

c. Reactor Vessel Water Level-High, level 8 2 1, 2, 3, 4*, 5* 31 Condensate Storage Tank Level-Low ) 4*, 5* 34

d. 1, 2, 3,

e. Suppression Pool Water Level-High 2(d) 2 1,2,3,4*,5* 34

f. Manual Initiation ## 1 1, 2, 3, 4*, 5* 32 D. LOSS OF POWER

1. Division 1 and 2

a. 4.16 kV Bus Undervoltage 4 1, 2, 3, 4**, 5** 30 td (Loss of Voltage)

b. 4.16 kV Bus Undervoltage '4 1, 2, 3, 4**, 5** 30 c.a (BOP Load Shed)

E c. 4.16 kV Pus Undervoltage 4 1, 2, 3, 4**, 5** 30

(Degraded Voltage)

2. Division 3

a. 4.16 kV Bus Undervoltage 4 1, 2, 3, 4**, 5** 30 (Loss of Voltage)

(a) A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months during periods of required surveillance without placing the trip system in the tripped condition provided at least one other OPERABLE channel in the same trip system is monitoring that parameter.

(b) As o actuates the associated division diesel generator.

(c) Provides signal to close HPCS pump discharge valve only.

(d) Provides signal to HPCS pump suction valves only. .

Applicable when the system is required to be OPERABLE per Specification 3.5.2 or 3.5.3.

- Required when applicable E5F equipment is required to be OPERABLE.

Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 135 psig.

1. Prior to STARTUP following the first refueling outage, the injection function of Drywell Pressure -

liigh and Manual Initiation are not required to be OPERABLE with indicated reactor vessel water level on the wide range instrument greater than level 8 setpoint coincident with the reactor piessine less than 600 psig.

4 INSTRUMENTATION TABLE 3.3.3-1 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION ACTION ACTION 30 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement:

a. With one channel inoperable, place the inoperable channel in the tripped condition within one hour

or declare the associated system (s) inoperable.

b. With more than one channel inoperatie, declare the associated system (s) inoperable.

ACTION 31 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, declare the associated ADS trip system or ECCS inoperable.

ACTION 32 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or declare the associated ADS trip system or ECCS inoperable.

ACTION 33 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel (s) in the tripped condition within one hour

or declare the HPCS system inoperable.

ACTION 34 - With the number of OPER,ABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within one hour

or declare the HPCS system inoperable.

^The provisions of Specification 3.0.4 are not applicable.

GRAND GULF-UNIT 1 3/4 3-30

REACTOR COOLANT SYSTEM 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.4.7 Two main steam line isolation valves (MSIVs) per main steam line shall be OPERABLE with closing times greater than or equal to 3 and less than or equal to 5 seconds.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

a. With one or more MSIVs inoperable:

1. Maintain at least one MSIV OPERABLE in each affected main steam line that is open and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , either:

a) Restore the inoperable valve (s) to OPERABLE status, or b) Isolate the affected main steam line by use of a deactivated MSIV in the closed position.

2. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.4.7 Each of the above required MSIVs shall be demonstrated OPERABLE by

-- verifying full closure between 3 and 5 seconds

when tested pursuant to

Specification 4.0.5. The provisions of Specification 4.0.4 are not applicable for entry into OPERATIONAL CONDITIONS 2 or 3 provided the surveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reaching a reactor steam pressure of 600 psig and prior to entry into OPERATIONAL CONDITION 1.

i t

l

The 3 seconds is the time measured from start of valve motion to complete l valve closure. The 5 seconds is the time measured from initiation of the actuating signal to complete valve closure.

(

l l GRAND GULF-UNIT 1 3/4 4-24 L

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS - OPERATING ,

LIMITING CONDITION FOR OPERATION For Tech Spec 3.5~.1 3.5.1 ECCS divisions 1, 2.and 3 shall be OPERABLE with: See TSPS 5 08A

a. ECCS division 1 consisting of:

1. The OPERABLE low pressure core spray (LPCS) system with a flow path capable of taking suction from the suppression pool and transferring the water through the spray sparger to the reactor vessel.

2. The OPERABLE low pressure coolant injection (LPCI) subsystem "A" of the RHR system with a flow path capable of taking suction from the suppression pool and transferring the water to the reactor vessel.

3. Eight OPERABLE A05 valves.

b. ECCS division 2 consisting of:

1. The OPERABLE low pressure coolant injection (LPCI) subsystems "B" and "C" of the RHR system, each with a flow path capable of taking suction from the suppression pool and transferring the water to the reactor vessel.

For Tech Spec 3.5.l. c 2. Eig,t OPERA,tE A05 ,al,e,.

See TSPS #079 :- TCCS division 3 consisting of the OPERA 8LE high pressure core spray (HPCS) systefr. with a flow path capable of taking suction from the suppression pool and transferring the water through the spray sparger tc the reactor vessel.

OPERATIONAL CONDITION 1, 2* # and 3*.

APPLICABILIT[

ACTION:

a. For ECCS division I, provided that ECCS divisions 2 and 3 are OPERABLE:

1. With the LPCS system inoperable, restore the' inoperable LPCS system to OPERABLE status within 7 days.

2. With LPCI subsystem "A" inoperable, restore the inoperable LPCI subsystem "A" to OPERABLE status within 7 days.

3. With the LPCS system inoperable and LPCI subsystem "A" inoperable.

restore at least the inope_rable LPCI subsystem "A" or the inoperable LPCS system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

4. Otherwise, be in at least HOT SHUT 00WN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUT 00WN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .**

"The A05 is not required to be OPERABLE when reactor steam dome pressure is less than or equal to 135 psig.

See Special Test Exception 3.10.5.

- Whenever two or more RHR subsystems are inoperable, if unable to attain COLD SHUT 00WH as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.

GRAND GULF-UNIT 1 3/4 5-1

ENERGENCY CORE COOLING SYSTEMS LINITING CONDITION FOR OPERATION (Continued)

. ACTION: (Continued)

b. For ECCS division 2, provided that ECCS divisions 1 and 3 are OPERABLE:

1. With either LPCI subsystem "8" or "C" inoperable, restore the inoperable LPCI subsystem "B" or "C" to OPERA 8LE status within 4

7 days.

2. With both LPCI subsystems "B" and "C" inoparable, restore at least the inoperable LPCI subsystem "8" or "C" to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

3. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months *.

c. For ECCS division 3, pro'ided v that ECCS divisions 1 and 2 and the RCIC system are OPERA 8LE:

1. With ECCS division 3 inoperable, restore the inoperable division to OPERA 8LE status within 14 days.

2. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

d. For ECCS divisions 1 and 2, provided that ECCS division 3 is OPERABLE:

1. With LPCI subsystem "A" and either LPCI subsystem "B" or "C" inoperable, restore at least the inoperable LPCI subsystem "A" or the inoperable LPCI subsystem "B" or "C" to OPERABLE status i

within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

2. With the LPCS system inoperable and either LDCI subsystems "B" or "C" inoperable, restore at least the inoperable LPCS system or the inoperable LPCI subsystem "B" or "C" to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

3. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months *.

l "Whenever two or more RHR subsystems are inoperable, if unable to attain COLD i

SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.

l .

l I

GRAND GULF-UNIT 1 3/4 5-2 l

EMERGENCY CORE COOLING SYSTEMS

-LIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued)

e. For ECCS divisions 1 and 2, provided that ECCS division 3 is '

OPERABLE and divisions 1 and 2 are otherwise OPERABLE:

1. With one of the above required ADS valves inoperable, restore the inoperable ADS valve to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to 1 135 psig within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2. With two or more of the above required ADS valves inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to 1 135 psig within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

f. With an ECCS discharge line " keep filled" pressure alarm instrumentation channel inoperable, perform Surveillance Requirement 4.5.1.a.1 at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

g. With an ECCS header delta P instrumentation channel inoperable, restore the inoperable channel to OPERABLE status with 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or determine ECCS header delta P locally at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; otherwise declare the associated ECCS inoperable.

h. In the event an ECCS system is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and sub-mitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the useage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

"Whenever two or more RHR subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.

GRAND GULF-UNIT 1 3/4 5-3

PLANT SYSTEMS 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3 The reactor core isolation cooling (RCIC) system shall be OPERABLE with an OPERABLE flow path capable of automatically taking suction from the suppression pool and transferring the water to the reactor pressure vessel.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3 with reactor steam dome pressure greater than 135 psig.

ACTION:

With the RCIC system inoperable, operation may continue provided the HPCS system is OPERABLE; restore the RCIC system to OPERABLE status within 14 days, otherwise be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to less than or equal to 135 psig within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.7.3 The RCIC system shall be demonstrated OPERABLE:

a. At least once per 31 days by:

1. Verifying by venting at the high point vents that the system piping from the pump discharge valve to the system isolation valve is filled with water,

2. Verifying that each valve, manual, power operated or automatic in the flow path that is not locked, sealed or otherwise secured in position, is in its correct position.

3. Verifying that the pump flow controller is in the correct position.

b. At least once per 92 days by verifying that the RCIC pump develops a flow of greater than or equal to 800 gpm in the test flow path with a system head corresponding to reactor vessel operating pressure when steam is being supplied to the turbine at 1025 + 20, -80 psig.*

c. At least once per 18 months by:

1. Performing a system functional test which includes simulated automatic actuation and restart and verifying that each auto-matic valve in the flow path actuates to its correct position, but may exclude actual injection of coolant into tae re h tor vessel.

a The provisions of Specification 4.0.4 are not applicable proviJed the surveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is adequate to perform the test.

GRAND GULF-UNIT 1 3/4 7-7

3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES A.C. SOURCES - OPERATING LINITING CONDITION FOR OPERATION 3.8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERA 8Lf.:

a. Two physically independent circuits between the offsite transmission network and the onsite Class 1E distribution system, and

b. Three separate and independent diesel generators, each with:

1. Separate day fuel tanks containing a minimum of 220 gallons of fuel.

2. A separate fuel storage system containing a minimum of:

a) 57,200 gallons of fuel each for diesel generators 11 and l

12, and b) 39,000 gallons of fuel for diosal generator 13.

3. A separate fuel transfer pump.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a. With either one offsite circuit or diesel generator 11 or 12 of the above required A.C. electrical power sources inoperable, demonstrate the OPERA 8ILITY of the remaining A.C. sources by prforming Surveil-lance Requirements 4.8.1.1.1.a within one hour anJ 4.8.1.1.2.a.4,*

for one diesel generator at a time, within two hours and at least

! once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; restore at least two offsite circuits and diesel generators 11 and 12 to OPERA 8LE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT SHUTDOWN within the next li hours and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l b. With ont offsite circuit and diesel generator 11 or 12 of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirements 4.8.1.1.1.a within one hour and 4.8.1.1.2.a.4,* for one diesel generator at a time, within two hours and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; restore at least one of the inoperable A.C.

sources to OPERA 8LE status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or be in at least HOT SHUT-DOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Restore at least two offsite circuits and diesel genera-

tors 11 and 12 to OPERA 8LE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from time of initial loss or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

" Specification 4.8.1.1.2.a.4 must be performed for diesel generator 13 only when the HPCS system is OPERA 8LE.

GRAND GULF-UNIT 1 3/4 8-1 Amendment No. 5 l

ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

For Tech Spec 3.8.l.l c.

ACTION (Continued)

See TSPS f OBO

c. With either diesel generator 11 or 12 of the above required A.C.

electrical power sources inoperable, in addition to ACTION a or b above, as applicable, verify within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months that all required systems, subsystems, trains, components and devices that depenil on the remaining diesel generator 11 or 12 as a source of emergency power are also 0PERABLE; otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . (

d. With two of the above required offsite circuits inoperable, demonstrate the OPERABILITY of three diesel generators by performing Surveillance Requirement 4.8.1.1.2.a.4*, for one diesel generator at a time, within two hours and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter, unless the diesel generators are already operating; restore at least one of the inoperable offsite circuits to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least-HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With only one offsite circuit restored to OPERABLE status, restore at least two offsite circuits to 0PERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from time of initial loss or be in at least NOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 'and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

s. With diesel generators 11 and 12 of the above required A.C.

electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirements 4.8.1.1.1.a within one hour and 4.8.1.1.2.a.4*, within two hours and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; restore at least one of the in-operable diesel generators 11 and 12 to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ,

or be in at least HOT SHUTDOWN within the next-12 hours and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Restore both diesel genera-tors 11 and 12 to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from time of initial loss or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

f. With diesel generator 13 of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirements 4.8.1.1.1.a within one hour and 4.8.1.1.2.a.4, for one diesel generator at a time, within two hours and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; restore the inoperable diesel generator 13 to OPERA 8LE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or declare the HPCS system inoperable and take the ACTION required by Specification 3.5.1.

" Specification 4.8.1.1.2.a.4 must be performed for diesel generator 13 only when the HPCS system is OPERA 8LE.

GRAND GULF-UNIT 1 3/4 8-2

ADMINISTRATIVE FROCEDURE

. GRAND GULF NUCLEAR STATION .

6 01-S-15-2 Rev. 4 At t achment III Pese 1 of 1

TECHNICAL SPECIFICATION POSITION STATIMENT f

Oe#i Part 1 Originator: b A VA, e ussel.l 8//W@f >

Name / Date Technical Specification 3.f. /

Raforencee: 1)/A

.. Poeition J..C.O. 3.$.1.c 'okouracs 7wer HA*$ se "c4 Pants ok , _

ra u esoc on suc Trav FMo s no $w//srs1,w Ave. **. Mis is isvirt/M?n 7D #960FAJ 7##7' A'MS stosr De c'A/sAlf eA s4V70mM7/CMlv '

T4/*G A SvCTieAs fep e9 W fe/Afo:59a v JDe4 .

I 1 ;

1

. Poeition Affect /Commmente: TP 7wr fisc7M the NA*S ass kraes?:cste w

-7AAAa f f.P M O 7D 7W J2.wArsusu Adt. _ 7 Mr socrsou _5Naos.h W As.cowen in Passa.4 M ic %Tro n ?wr scLMDSin rDcc Mn't l yus' tY<.<Dink tunicw t%LMn ywr khni?c 74.dvicet #4s i

3r5A) CQltRECTEb. JF 7UE' MA'S M77w Js A,4mitw Ad2%DWAJ AkD RE ALIGNem 24cr JD 71u f.f7' B M ff M t M *. D n 7 h A su-1 A 77 x M7)lC 7XM46 Pre ms 2MN cs'M'A 'styrz? ina- #fd,.S _Me sr*

! 3r DerCLortGun /*1/of twrit. 7nwr @ M W s:t/ iS cM 2dCTFD ,

Part 2 Compliance Superintendent c- / / T"$

" ' Si'gnaturg# ~ / Date Part 3 Review and Approval: Approve Disapprove (1) _

(8 '

( )

Tect!&ic u dMagshdent / Date

{

(2) I ///5[ W ( )

S / ate

._. (3) $ l lI h[ / ( ( )

(4) /[ (M * ~~

( ) '

Manager, Approval / 'Dete

0ne of the following is required: GGNS Ceneral Manager Manager, Plant Operatione

GRAND GULF NUCLEAR STATION ADMIN 2STRATIVE PROCEDURE 01-S-15-2 i Rev. (e '

Attachment I111 Page 1 of J3 l l TECHNICAL SPECIFICATION POSITION STATEMENT Part 1 Originator: ed s.e t/ 10/3/95' Nhme / Date Technical Specification 3. 6. f . I , dems sn msna e nsEerences: 7Ecx 5;ct 8bs3 sec- 3/4. W. i , FM2 be c r. Er. t, 3

!* 'V MtC. 6c*bitic LcITER tw@ P 4tc" d 'Siv , frf Gs.02 / S3, Dec Position .$EE A7TACMD LI$r of ALs- "d e n w p r o s g s ir m s S A sy s12 nq ys*,~ s , ctves/c vDt,75, srt o M eiers ,rwe r p rPrn o $y r y er' p r m.m,+ ,r c ja. ssc_

_ iEN:124 5't' 11 og r 2 45 A ses.ece ese asespry A me #, 77/nr5 a,,wesv ic AM MP k. 5 o e av s e twf Armasw"6 bes92 6frcanTWL tv ,sf wep, 9 02 Mo? s*srY D BicAust Ak-nero _sinrfw'rM~ e wooLb TWEN Mffy Aiss-Acnon,s s rww ar c.

I Position Af fect/ Consents: 72<t /,s r td,t.t //GA>/o# e tssy,at e., fa 7&r A01/H4mv cf TN6 PivviRED A c i7s ti MD wae s4cr f.s A cAsc,c di37 TO 2inDvCG" ?NE BD/Soff /D Nets it*<*/Mer a f 23F Ms.rsfC s 73'M,5.

A f) //7 Part 2 Compliance Superintendent

/ / 8 Y ' T5 SLgnacp Date Part 3 Review and Approval: Approve Disapprov e l '

~ '

(1) /2.[1c/85 Q<)' ( )

( Te 'nical Sup4nnt_pudent / Date (2) h. l. et48 o/1/ff (X) ( )

to - / Date a>l{Shh?. 026U WwuoM l

5 et' g I

/ Date

(

( ( >

(4) '- / } ".5V ~hb ( ( )

' *Mandier, Approval / Date

0ne of the following is required: GGNS General Manager Manager, Plant Operations

TSPS No.: 080 Pegs 2 of 3 When Div I is declared INOP the following items shall be verified operable per LCO 3.8.1.1, ACTION c.

1. LPCI mode of RHR-B System

2. LPCI mode of RHR-C System

3. CTMT Spray mode of RHR-B System 4 Suppression Pool Cooling mode of RHR-B System

5. Shutdown Cooling mode of RHR-B System (required only in Mode 3)

6. Suppression Pool Makeup System, Div II

7. CTMT and Drywell Isolation Valves as listed in Tech. Spec. Table 3.6.4-1 section 1, Div II

8. Post LOCA Vacuum Breaker, Div II

9. Secondary CTMT Isolation Valves as listed in Tech. Spec.

Table 3.6.6.2-1, Div II

10. Standby Gas Treatment B System

11. Hydrogen Recombiner System, Div II

12. CTMT and Drywell Hydrogen Ignition System, Div II

13. Control Room Emergency Filtration System Div II

14. MSIV Leakage Control System Div II

15. FW Leakage Control System, Div II l 16. RHR-B Room HVAC l

17. RHR-C Room HVAC

18. ESF SWGR Room HVAC, Div II l

19. Standby Liquid Control System, Div II

20. Drywell Purge Compressor, Div II

21. Battery Charger, Div II (either 1B4 or IBS)

-- - , ,. ,,- , , , , - - - . - - , - . - - . _ _ . , , , , , 7- _ ,. - . - . , , - - , - . , , , , . , , - - - - - . , . - -

TSPS No.: 080 Pzgs 3 of 3 When Div II D/G is declared INOP the following items shall be verified OPERABLE per LCO 3.8.1.1, ACTION c.

1. LPCS System

2. LPCI mode of RHR-A System

3. CTMT Spray mode of RER-A System

4. Suppression Pool Cooling mode of RHR-A System

5. Shutdown Cooling mode of RRR-A System (required only in Mode 3)

6. Suppression Pool Makeup System, Div I

~

7. CTMT and Drywell Isolation Valves as listed in Tech. Spec.

Table 3.6.4-1, section 1, Div I

8. Post LOCA Vacuum Breaker, Div I

9. Secondary CTMT Isolation Valves as listed in Tech. Spec. Table 3.6.6.2-1, Div I

10. Standby Gas Treatment - A System

11. Hydrogen Recombiner System, Div I

12. CTMT and Drywell Hydrogen Ignition System, Div I

13. Control Room Emergency Filtration System, Div I

14. MSIV Leakage Control System, Div I

15. FW Leakage Control System, Div I

16. LPCS Room HVAC

17. RHR-A Room HVAC

18. ESF SWGR Room KVAC, Div I

19. Standby Liquid Control System, Div I

20. Drywell Purge Compressor, Div I

21. Battery Charger, Div I (either 1A4 or 1A5)

'~5___IBEQBY_QE_NyC(E@B_EQWEB_E6@NI_QEEB@I1QNt_ELylQS t_@NQ PAGE 20 IBE85QQYN@dlCS ANSWERS -- GRAND GULF 1 -86/12/08-CASTO, C I /

ANSWER 5.01 (1.00) 757. Void Fraction in the core.(0.5) This is because of the increased reasonance capture which would occur (due to the longer slowing down length).(0.5)

REFERENCE EIH: L-RQ-604 GGNS: Reactor Physics L/P, pp 1.7 - 9, 10, 13 BSEP: 02-OG-A, pp 39 -49;GE RxTh sec 4.,OBJ 6.3 BFNP: Reactivity Coefficient LP, pp 4, 5; RQ 85/03/01 ANSWER 5.02 (1.00) d.

REFERENCE GGNS OP RT 802 p. 18 292006 K1.03 2.9/2.9 K1.04 2.9/2.9 ANSWER 5.03 (1.00) a.

REFERENCE GGNS HF 508-008 p. 19 ANSWER 5.04 (2.00)

a. RTS

b. Decrease

c. Increase

d. Decrease REFERENCE GGNS OP HF 303 pp.3-18 293004 K1.16 2.1/2.3

k 4 UNITED STATES

_#j S "00 ,% NUCLEAR REQULATORY COMMISSION

.- >t t .r, .g REG lON ll 7 a 101 MARIETTA STREET, N.W., SulTE 2000

a o ATLANTA, GEORGIA 30323 s.,

/

4 4 4 1

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3 THEQ8Y_QE_NQQ(E@R_PQWg8_P6 ANT _QEg8@llQNt _E(ylQ@g_@NQ PAGI 21 ISE859QYNed1GE ANSWERS -- GRAND GULF 1 -86/12/08-CASTO, C ANSWER 5.05 (1.00)

a. decrease

b. increase

c. increase

d. increase REFERENCE GGNS 65-1-02-I-5 p. 1 204000 K3

K3.01 3.2/3.6 K3.06 2.6/2.7 I

4 ANSWER 5.06 (1.00)

I

a. Keff CO.33

b. increase, decreasing CO.25 ea.]

REFERENCE 4 GGNS OPRT 802 p. 10 OBJ 5 292008 K1.05 4.3/4.3 292003 K1.01 2.9/3.0 ANSWER 5.07 (1.50) i

a. True CO.53 j b. At low power, the cold water sprayed into the core exit region has a more significant effect on moderator temperature and void content. C1.03 REFERENCE GGNS OP E22-1-501 209002 A2.01 3.8/3.8 i

1 i

ANSWER 5.00 (1.00) i 19.33 deg. F +-2 deg CO.53

! Due to the additional subcooling in the downcomer region.CO.53 V

REFERENCE GGNS OP-833-1-501 p. 10 293003 K1.23 2.8/3.1

~ . .. _. . _ _ . _ . . - . _ . . . . - - - . . . . . _ - - , . . . _ _ _ _ - . . _ . , _ - . - _ . - - _ - . , . . _ . - _ . - , _ . .m.., --

i UNITED STATES

/>s Katg*r NUCLEAR REOULATORY CO'*CISSION

.. .o$

3 f, o REGION il I,

-a 101 MARIETT A STREET, N.W., SulTE 2900 a ATLANTA, GEORGIA 30323

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Et__IHEQBY_QE_MWGLEBB_EQWEB_ELONI_QEg@@IlQN t _E(ylQS t_@NQ PAGE 22 ISE8599YNed1GE ANSWERS -- GRAND GULF 1 -86/12/08-CASTO, C ANSWER 5.09 (2.00)

J

a. 1. higher

2. higher CO.5 ea.]

l 3. void

b. EHC controls steam flow to maintain a constant Rx pressure. CO.53

, REFERENCE GGNS GE Rx Theory Chap. 7 pp 7-16 OBJ 5-5.

292008 K1.19 3.1/3.2 I

j ANSWER 5.10 (-1TOO T~ j /

C* / ' 2- Yr /

Due to a significantly di f or(ent method of computing the TIP and LPRM i

l calibrations, cycle 2 read ngs should generally be about 20*/. lower than actual.

l REFERENCE j GGNS OP IP 523 OBJ 2 p. 12 l

l ANSWER 5.11 (1.00)

(

! The applicable threshold power must be decreased CO.253 by a factor which j is calculated by taking the ratio of the current power level CO.253 to rated power CO.253 *(10 KW/ft), however this new threshold power is not l

allowed to fall <0KW/ft. CO.253.

REFERENCE GGNS OP IP 525 OBJ 13 l 293009 K1.36 2.9/3.4 l K1.37 2.6/3.3 t

i i

ANSWER 5.12 (2.00) i j a. incrose steam production due to pressure / power rise l b. Rx trip on LO

{ c. swell due to SRV openings

d. decay heat production i

l REFERENCE l GGNS OP DT 520 OBJ 1 l 259002 K3 l

l !

1

jk3 E8%g' U%ITED STATES

,,#. ,o4 NUCLEAR REOULATORY COMMISSION 3 .f, o REGION il 5

e I 101 MARIETTA 5TREET, N.W., SulTE 2000 o ATLANTA, GEORGIA 30323

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5. THEORY OF NUCLEAR POWER PLANT OPERATIONg _FLyIQHt_@NQ PAGE 23 IMEROQQyN@diqS ANSWERS -- GRAND GULF 1 -86/12/08-CASTO, C K1.02 3.2/3.3 K3.01 3.8/3.8 ANSWER 5.13 (2.50)

a. High relative flux -

causes a greater reactivity change due to CRW being proportional to flux tip/ flux avg. therefore, the higher the relative

, flux the greater the change. (0.333 for area /O.5 for Exp.)

\

b. larger effect for the middle - due to absorption of neutrons which have j n high probability of causing fission. Whereas control rods at the edge j absorp neutrons which have a high probability of leakage.

,I i c. #1 has higher worth. When inserted #1 depresses the flux around itself, ,

this increases the flux in other regions, when #2 is inserted the flux l has been depressed therefore its worth is lower (than its worth in an unrodded core). ,

REFERENCE North Anna ROP- pp. 6.11,6.12,6.19 Obj. B K/A OO1-OOO-K5.02 (2.9/3.4)

TPT CNTO Core Control 6-13 GGNS GERxTh sec. 5 OBJ 3.4 l

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>2 E8tg# UNITED STATES

#o, NUCLEAR REOULATORY CO'. MIS $12N

[ ,, 'n REGION il

$ $ 101 MARIETTA STREET. N.W., SulTE 2000

4 ATLANTA, GEORGIA 30323 a

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6 t__E60NI_SYSIEdS_QE@l@Nt _QQNIBQ6t_QNQ_lNSIBudENI@IlQN PAGE 20 ANSWERS -- GRAND GULF 1 -86/12/08-CASTO, C ANSWER 6.01 (1.00)

d. o /- h ,

I et p t REFERENCE GGNS OP B33-2-501 p. 7, OBJ 38 202002 K4.08 3.3/3.4 5.03 2.4/2.4 ANSWER 6.02 ( .50)

TRUE REFERENCE GGNS OP-821-501 216000 K1.10 3.2/3.4 1.23 3.3/3.4 ANSWER 6.03 (2.00)

a. would not inject

b. would inject

c. would inject

d. would inject REFERENCE GGNS OP-E21-501 OBJ 3,78 209001 K1.02 3.4/3.4 1.06 2.0/2.1 4.05 2.6/2.6 A2.02 3.2/3.2 ANSWER 6.04 (2.50) 373

m. 86r2 MWE [ 1. 0 3

b. 1. 17ILMWE / 7 3

7 c' r /~ 7 ' #

2.'O-MWE 1 '/ , L

3. LT4-GF /3 5 CO.5 ea. close approximations acceptable]

REFERENCE GGNS OP-N32-2-501 241000 K3.06 4.1/4.1

l r .

46'"8% UNITED STATES

,4! Io., NUCLEAR RECULATORY COMMISSION

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8 101 MARIETTA STREET, N.W., SulTE 2000 ATL ANTA, GEORotA 30323 t

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t___PkedI_EXEIEnt_EEE19N _QQUIRQ(t_@NQ_1NgISQQ(NI@IlQN g PAGE 25 ANSWER 3 -- GRAND GULF 1 -C6/12/OS-CASTO, C 3.00 3.7/3.7 4.11 2.6/2.6 ANSWER 6.05 (1.50) 1.a. ADS mode 2.a. ADS mode

b. Valve does not operate b. Relief mode

c. Safety mode c. Safety mod CO.25 ea.3 el ' I fey ej suft / o er ~ ~ S*' ^~' ,tt ye cie REFERENCE GGNS OP-E22-2-501 pp 18 OBJ 6B 239002 K4.OB 3.6/3.7 219000 K4.01 3.7/3.9 4.03 3.8/4.0 ANSWER 6.06 (1.00)

a. Prevents drawing water from the Suppression Pool CO.253 when the line cools after discharge CO.253.

b. Water in the discharge line will cause a pressure transient CO.253 in the event of a subsequent blowdown CO.253.

REFERENCE GONS OP-E22-2-501 pp. 7 239002 K4.03 3.1/3.3 5.06 2.7/3.0 ANSWER 6.07 (2.00)

The HPCS D/0 would start CO.53 due to LOCA and undervoltage signals 0.25ea.

HPCS would not auto start CO.53 the pump control switch was taken to the trip position [0.253 which breaks the auto start signal CO.253.

REFERENCE GONS S0! 04-1-01-P81-1 pp. 34 209002 K1.04 3.0/3.0 A1.07 2.5/2.0 A1.00 3.1/3.3

UNefE0 STATES

.j ys** 450,,'o,,

NUCLEAR REULATORY COMMISSION f, o RE060N il f $ 101 MAR 8Eff A 87REET.N.W SulTE 2900

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6t__PLONI_gygl[Ng_Q{gigN _QQUISQLt_6NQ_INSISQUENIGIlQN t PAGE 23 ANSWER 3 -- GRAND GULF 1 -C3/12/08-CASTO, C ANSWER 6.08 (1.00) i

a. Would not auto start (0.253 due to voltage regulator switch in manual CO253.

b. Would start Ccaf3 CO.53.

REFERENCE GGNS OP-922-1-501 Table 3 OBJ 7E 209002 K1.04 3.8/3.8 .

l 6.01 3.6/3.6 l l ANSWER 6.09 (1.50) 4 I

! a. PIC R605 maintains RCIC suction pressure CO.253. F065 is modulated

{ by R605 to maintain RCIC suction pressure below the high limit setpoint CO.253.

, b. The CST is not valved out, but is controlled by check valves CO.51

c. The RCIC pump turbine exhaust is directed to the Suppression Pool 50.53. ;

i REFERENCE ~- -

GGNS SD E12 pp 37, E51 pp 12, SD'E12 pp. 3,37 (!!A) OBJs 6.a.2,99.

l j ANSWER 6.10 (1.00) e 4

The ramp generator is at maximum, since it does not reset unless F045 is closed CO.503, the signal converter would control the governor valve CO.53 (flow controller) .

i I l REFERENCE GGNS OP E51-501 p. B OBJ 3A 217000 A1.05 3.7/3.7 4.02 3.9/3.9 i

2.12 3.0/3.0

! 3.02 3.6/3.5 r

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< y,p ** "%,,g ,

NUCLEAA RECULATORY COMMISSION

c, g RE010N il

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. h__!NGUI_EYEIEGE DEE19&. GQUIBA 060_INEIBWUENIBIlgd PAGE 27 ANSWER 3 -- GRAND GULF 1 -0$/12/03-CA3TO, C !

l t

i ANSWER 6.11 (2.00) f

a. The operator attempted to withdraw the rod greater than four positions between the LPSP and HPSP.

b. To alert the operator of possible power peaking due to withdrawl of the same rod (or gang of rods) when above the LPSP and below the HPSP.

REFERENCE GGNS OP C11-501 !

201005 K4.03 3.5/3.5 1 l

5.11 3.3/3.3 ANSWER 6.12 (2).00).b d,/~

l d# l_. 2 ?

a.AV R -A-by-RIE-N068 E-- !

B- by PIS N06B B and F [0.2 ea.3

b. 1. di ff erenti al to 0 4* I

2. no /af f ect on valve operation (no response)

[0.5 es.3 l REFERENCE l GGNS SD B21 !!! B pp 19 SD B21 p 17 SD C11-1A p 5. .

216000 K1.00 3.7/3.9 3.0B 3.9/4.1 A2.03 3.0/3.1 !

I K3.27 2.9/3.1 t

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/'g>* U49'g NUCLEAR REOULATORY cot 4400SSION r ,.

e R...oN i, 101 MARIETTA STRGET.N.W.,8UITE 2900 7, !!

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59019kQ91G94_GQUI596

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ANSWERS -- GRAND GULF 1 -96/12/08-CASTO, C i

ANSWER 7.01 (1.00) j d.

REFERENCE 3

GGNS S0! 04-1-01M51-1 j 223001 A2.10 3.6/3.8

)

ANSWER 7.02 (1.00) l i

b.

REFERENCE i GGNS OP-Ep/SPDB-510 pp. 13 OBJ 3 i

! i f ANSWER 7.03 (1.00) f I

d.

1 I REFERENCE

) GGNS TS 3.3 i i l ANSWER 7.04 (2.00)

! 1. RWL maintained above TAF l

2. Core is being sprayed by either HPCB of LPCB (

3. sufficient steam flow through the core ,

4. reflooding flow of one LPCI pump is injecting into the core with j j

i RWL high enough to produce two phase flow up through the core.

i

{ REFERENCE 3

00ND 01-B-06-2 sec 5.10 d !

1 r I ANSWER 7.05 (1.50) !

1. The individual is in training to qualify for an operator license [

2. Under the direction of a Itcensed operator RO or SRO present at the l controle j i

) !

l

1 I I

unetso eit.tss fif* "'%,,%, ..

NUCLEAR RCULATORY COAMMSSION y c, - nassoNil 101 MAA4ETT A ST AGE 7, N.W., SulTE 2000 3

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ZJ._P80GEQW6EE_ _NQBU%a. 696965%a._EUEBGENGY 999 PAGE 29 !

j 8601h001G%_GQUISh l ANSWERS -- GRAND GULF 1 -86/12/09-CASTO, C l j REFERENCE 1 1

GONS 01-S-06-2 6.4.5 ,

294001 K1.05 3.2/3.7 '

I

. ANSWER 7.06 (1.50) i II l a. an SRO or Health Physics personnel. [

b. 1- 100 [

2- not allowed 3- 1 day 4- CHEM / RAD SUPT.

f

! REFERENCE ,

! GGNS RPP-09-S-01-24 OBJ 2,3 l l INPO GPG B2-000 {

,7 I r <. ;

{ ANSWER 7.07 (1700) l i

! a.1. no ground CO. 5 a.3 j 2. posttive._ bus. ground-('t- >W M-

, b. A high resistance ground exist Co.503 l l i i MEFERENCE !

}

GGNS 801 04-1-01-L11-1 i 263000 K1.04 2.6/2.9 l j A2.01 2.9/3.2 l A4.04 3.0/3.2 l' SWG 15 3.4/3.9 s

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i then adequate reject flow does not entst.

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a. when temperature is < 95 deg. Co.53

b. Procede to SP/T-1 and -2 CO.53 and remain untti all entry conditions have cleared. CO.53 REFERENCE GGNS OP EP/SPDS 504 OBJ 3 ANSWER 7.10 (1.50)

a. 1. 7.5

2. 19.75

3. 1.25

b. Dose does not exceed 3 rem /qtr accumulated occupational dose does not exceed 5(N-18) completed NRC form 4 Co.25ea.3 REFERENCE GONS ID 2796 LP OP-PS 601 ANSWER 7.11 (1.50)

Due to suppression pool temperature being above 140 deg. F CO.53, RCIC lube oil will not be cooled CO.53 thersjorekRCIC is secured to prevent damage to the turbine C0.53. Ud vec.pi/ / /$ /r ( . p y , . . t . ., ,( .. . Jo

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(/ , '. l (, % , / . = + o Ea /, \' s /V REFERENCE GONS OP-EP/SPDS 504 OBJ 3 217000 K6.03 3.5/3.5 A1.00 3.5/3.6 A2.19 3.5/3.6 ANSWEH 7.12 (l.00)

It is possible to cause a rod deift as a result of the cooling water pressure transtant.

REFERENCE GONG CR0 OP (COB)

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ANSWERS -- GRAND GULF 1 -86/12/08-CASTO, C ANSWER 7.13 (1.50)

The Rx power input for rod control is determined from first stage pressure (0.53. With the bypass valves open, the RCS senses Rx power as less then actual and the potential exist for a non-conservative rod withdrawl [1.03 REFERENCE GGNS 03-1-01-2 and TS 3.4.1.4 (EQB)

ANSWER 7.14 (1.00) minimize thermal stresses on the feedwater nozzles.

REFERENCE OGNS 03-1-01-3 (EQB)

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REFERENCE GGNS 09-S-02-300 6.7.2 ANSWER 8.02 (1.00) a.

REFERENCE GGNS 09-S-02-300 pp. 4 ANSWER B.03 (1.00) d.

REFERENCE 10 CFR 50.59 ANSWER B.04 (1.00) c.

REFERENCE GGNS TS 3/4.5.1 ANSWER 8.05 (1.00) c.

REFERENCE GGNS OP AP 552-004 ANSWER B.06 (1.00) b.

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REFERENCE GGNS OP EM 501-014 ANSWER 8.08 (1.00) a.

REFERENCE GGNS 3.7.3/3.4.7 ANSWER 8.09 (1.00) d.

REFERENCE GGNS 3.8.1.1 ANSWER B.10 (1.00) b.

REFERENCE GGNS TS bases 3/4.3.9 ANSWER 8.11 (1.00) 4 C.

REFERENCE ~

GGNS 3.7.6.2 i -_

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a. NOB

b. False REFERENCE GGNS 01-S-06-29 sec. 6.4.2 ANSWER 8.14 ( .50)

TRUE REFERENCE GGNS TS 6.12 l

l ANSWER 8.15 (1.00)

a. 3/4.2.1 3.2.1. g d'7

b. 3/4.2.4 3.2.4 i, p REFERENCE GGNS 3/4.2.1 3/4.2.4 ANSWER B.16 (1.00)

a. R67ueling SRD /4 wl d'yst k e 4,p[Ato3' N$ wM

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REFERENCE l GGNS OP IP 505 OBJs 2a 2c I

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ANSWER B.17 (1.00) i The two provisoes are

- The sequence of major evolutions is not changed. (0.5)

- The intent of the instruction is not changed. (0.5)

REFERENCE GGNS: Procedure 02-S-01-2 ANSWER 8.18 (1.50)

All control rods fully inserted CO.253 except for the single control rod of highest worth which is assumed to be fully withdrawn [0.253 The Rx is in the shutdown condition CO.253 cold CO.253 68 deg F [0.253 Xe free CO.253 REFERENCE GGNS def. 1.39 e

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1 U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR REQUALIFICATION EXAMINATION FACILITY: _QRANg_ GULF _1____________

REACTOR TYPE: _@WR-GE6_________________

DATE ADMINISTERED: _g6f12f_lj________________

EXAMINER: _gAgTgi_C________________

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CANDIDATE: ____M_A_3_T_%_3_7_______

INSIBUCIlgNg_Ig_C@NQ1981E:

Read the attached instruction page carefully. This examination replaces the current cycle facility administered requalification examination.

Retraining requirements for failure of this examination are the same as for failure of a requalification examination prepared and administered by your training staff. 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 four (4) hours after the examination starts.

% OF CATEGORY % OF CANDIDATE'S CATEGORY

__YBLUE_ _Igl@b ___gCgBE___ _y@ lye __ ______________C@IE@QBY_____________

_laz99__ _22z99 ___________ ________

l. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

_19199__ _29199 ___________ ________ 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS

_19z99__ _2Dz99 ___________ ________ 3. INSTRUMENTS AND CONTROLS 19:99___ 29199_ ___________ ________ 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL

_Z2199__ ___________

Totals Final Grade i

All work done on this examination is my own. I have neither given nor received aid.

l Candidate's Signature i

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

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

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

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

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

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

6. Use only the paper provided for answers.

7. Print your name in the upper right-hand corner of the first page of gach 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 gnly gn gny 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 litgtatutg.

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

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

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 gxamingt only.

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

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10. 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 pageu including figures which are part of the answer.

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

c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questione.

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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QUESTION 1.01 (1.00)

Which of the following post accident containment hydrogen contributors is dependent on the radiation field intensity inside containment for the amount of hydrogen released?

a. Zr + H O -> ZrO + H 2 2

b. 2H O -> 2H +0 2 2 2

c. 2A1 + 3H O -> Al O + 3H 2 23 2

d. Fe + H O -> FeO + H 2 2 QUESTION 1.02 (1.00)

Which one of the following correctly describes the formation and removal processes at equilibrum atom density for Xe 1357

a. = (FP + Iodine burnout) -

(Xe decay + Xe135 burnout)

b. = (FP + Iodine decay) -

(Xe decay + Xe136 decay)

c. = (FP + Cesuim decay) -

(Xe decay + Xe135 burnout)

d. = (FP + Iodine decay) -

(Xe decay + Xe135 burnout)

QUESTION 1.03 (1.00)

The distance from the onset of bulk boiling to the point of transition boiling is described by which one of the following terms?

a. Critical Boiling Length

b. Boiling Fraction

c. Film Boiling Region

d. Boiling Volume

(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)

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A variable speed motor drives a centrifugal pump operating at 1600 rpm to deliver 40 gpm of water at 50 ft. head pressure. The work input is 65 hp.

It is desired to increase the flow capacity to 50 gpm.

Calculates (show work)

a. The new driving speed of the motor.

b. The new head in feet.

c. The new work input.

QUESTION 1.05 (2.00)

If steam goes through a throttling process, indicate whether the following parameters will INCREASE, DECREASE, or REMAIN THE SAME.

a. Enthalpy

b. Pressure

c. Entropy

h. Temperature QUESTION 1.06 (1.00)

While operating at power a RWCU filter domineralizer ruptures causing resin intrusion into the Rx vessel. State the effects (increase, decrease, remain the same) for each of the followings

a. Reactor water PH

b. Reactor water conductivity

c. Steam line nitrogen-16 activity

d. Reactor water activity

(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)

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Consider the equation below and answer the following:

S CR = count rate of neutrons CR = ------

S = source strength 1-K K = Keff

a. Which term (s) determiness) the total neutron production RATE 7

b. CHOOSE ONE. With a Keff<1, succeeding generations of neutrons will

[ increase / decrease 3 in population, at a [ decreasing / increasing 3 rate.

QUESTION 1.08 (1.50)

A surious HPCS initiation signal at 15% power produces a more significant thermodynamic resul t than a spurious initiation signal at 85% power.

a. Identify this statement as TRUE/ FALSE.

b. Explain your answer in part a.

QUESTION 1.09 (1.00)

During operation the Rx Recirculation loop temperatures are indicating 525 deg F while the dome pressure is at 985 psig. State the VALUE of the temperature difference between the dome and loops AND briefly i EXPLAIN the reason for this difference. l QUESTION 1.10 (2.00)

Refer to figure #619 " Plant Response to Control Rod Withdrawl in Power Range".

a. For points labeled 1 -5 (choose one for each)

1. fuel temperature at Point 2 is (Higher / Lower) than at Point 1.

2. the heat transfer rate at Point 2 is (Higher / Lower) than Point 1.

3. the most significant negative reactivity contributor stopping the power rise at Point 4 is the (void / fuel) coefficient.

b. Briefly explain why Reactor pressure levels off at Point 5.

(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)

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n 1___EBINglELES_gE_NUQ6E88_EgMEB_E68NI_gEEB811gN1 PAGE 5 IHEBdggyN8digSz_HE81_168NSEEB_8Ng_E6U1g_E699 OUESTION 1.11 (1.00) ,7 During full power cycle 2 op ation you note that the control room LPRM indications are reading at 10 ' of meter scale. Explain why this would be an abnormal reading?

QUESTION 1.12 (2.00)

Attached figure #630 illustrates a transient that could occur at a BWR.

GIVEN: (1) fecdwater controller failure (130%)

(2) bypass valves available (3; end of cycle one (4) SRVs higher than nominal setpoints (5) no operator action E: plain the cause of the f ollowing recorder indications:

a. Vessel steamflow increase from O to *12 secs.

b. Neutron flux drop at *11.7 secs.

c. Vessel water level rise at *17 secs.

d. Bypass valve flow at >16 secs.

QUESTION 1.13 (1.00)

a. Explain why decreasing the H20/U ratio (moderator / fuel) results in a reduction of the resonance escape probablity, I

b. Explain why decreasing the H20/U ratio increases the value of 'f' in the six-factor formula.

QUESTION 1.14 (1.00)

STATE for which condition the reactivity coef ficient contribution would be MORE NEGATIVE. EXPLAIN your choice.

Doppler coefficient with a 25% Void Fraction in the core,

-OR-Doppler coefficient with a 75% Void Fraction in the core.

(***** END OF CATEGORY 01 *****)

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The plant is operating at power with A, B, and C CCW pumps running and NONE of the pumps selected for STANDBY operation. A LOSP occurs and the diesels start and tie in normally. How will the CCW system respond during this transient?

a. The LSS panel will auto start the "B" CCW pump on ESF power 20 seconds after the bus is reenergized.

b. Either the "A" or "B" CCW pumps can be started manually on ESF power after the buses are reenergized.

c. SSW will automatically tie in to the main CCW supply header on decreasing header pressure.

d. The "B" CCW pump must be manually started by the operator on ESF power.

after the bus is reenergized.

QUESTION 2.02 (1.00)

SBLC System A is in a normal STANDBY lineup with one systematic deviation - the TEST TANK OUTLET VALVE (F031) is OPEN.

Which of the following most accurately describes the effects on the STORAGE TANK OUTLET VALVE (FOO1) and SBLC PUMP A of placing the SBLC Keylock Control Switch for Pump A to START.

a. Valve FOO1 Opens - SBLC Pump A Starts after the valve reaches its Full Open position.

b. Valve FOO1 Opens - SBLC Pump A starts concurrently with the valve opening.

c. Valve FOO1 does Not Open - SBLC Pump A Starts

d. Valve FOO1 does Not Open - SBLC Pump A does Not Start

(***** CATEGORY O2 CONTINUED Of4 NEXT PAGE *****)

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For each of the following abnormal conditions, state whether or not the LPCS would inject rated flow into the reactor vessel upon a valid initiation signal and Rx pressure decreasing below pump shutoff head.

Note: all other system components function properly,

a. Foot, Pump suction valve from the Suppression Pool is closed.

b. FE-NOO3, Flow transmitter, has a f ailed 'HIGH' High Prwssure Tap.

c. 25% of the suction strainer area has become clogged by debris.

d. Instrument air to the Testable Check Valve has been completly interrupted.

QUESTION 2.04 (1.50)

LIST two criteria which define an ESF Bus " Bus undervoltage" and DIFFERENTIATE between this condition and an ESF Bus " Loss of Power".

QUESTION 2.05 (1.50)

What are the three (3) methods employed by the Combustible Gas Control System that will reduce the percent of H2 in the containment in an accident situation AND how does each reduce the H2 concentration?

QUESTION 2.06 (1.50)

A failure of the RCIC Trip and Throttle valve control power occurs, List three means, via alarms and indications, the operator is made aware of this condition.

l QUESTION 2.07 (1.00)

The RHR Test Return line valves (F024 A/B F021) are interlocked with their control swi tch and automatic ini tiation signals. State which automatic feature can AND cannot be overridden by the control switches for these valves.

(***** CATEGORY O2 CONTINUED ON NEXT PAGE *t***)

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Refer to figure #583 " Hydraulic Power Unit", STATE two (2) reasons for venting the pilot line when isolating an HPU.

QUESTION 2.09 (1.00)

a. Explain the basis for the vacuum breaker installed on the SRV discharge line to the Suppression Pool.

b. EXPLAIN the consequence, should the SRV vacuum breaker fail to open when required.

QUESTION 2.10 (2.00)

The HPCS Diesel Generator Mode Select Switch was inadventently lef t in the Manual position (all other controls normal). A valid LOCA signal is received. HPCS Bus 17AC becomes deenergized and the operator takes the following mitigating actions:

---Placed the HPCS Pump Control Switch on 1H13-P601 to TRIP.

---Places the HPCS D/G Mode Select Switch to Automatic.

CXPLAIN the resultant responses of BOTH the HPCS D/G AND the HPCS Pump to these actions. Limit your discussion to initiating and start /stop signals.

QUESTION 2.11 (1.00)

The HPCS Diesel Generator controls are in the f ollowing alignments Engine lockout -

reset Generator lockout - reset Voltage Regulator in " Manual" Unit Mode Selector in " Automatic" All other controls / power supplies are in their normal alignment.

a. Explain the effect of this alignment on the AUTOMATIC start capability of the HPCS D/G.

b. The operator repositions the Unit Mode Selector to the " Manual" position and attempts a MANUAL start. Would the HPCS D/G start (yes/no)?

(***** CATEGORY O2 CONTINUED ON NEXT PAGE *****)

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With regard to the RHR Steam Condensing Mode answer the following:

a. The system is operating in accordance with SDI 04-1-01-E12-1, EXPLAIN the control method by which overpressurization of the RCIC suction line is prevented. (include signal conditioning equipment, control l er (s) and/or valve (s)).

b. State the system alignment which prevents a total loss of NPSH, in the event that the RHR Heat Exchanger supply valves (F065A&B) to RCIC fail closed.

4

c. Why might if be neccessary, due to operating in the Steam Condensing

. Mode, to align one RHR pump and one set of Heat Exchangers in the Suppression Pool Cooling Mode?

QUESTION 2.13 (1.00)

During startup under Cold Conditions the operator adjust the Control Rod Drive Pressure Control Valve to maintain a +260 psid between CRD and reactor pressure. Explain how this pressure differential is maintained as reactor pressure increases during the ensuing startup.

QUESTION 2.14 (1.00)

The Recirculation Pump (s) discharge / suction valves take approximately 2 minutes to close. Explain why this time frame is necessary.

(***** END OF CATEGORY O2 *****)

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. 3 __INglBUDENig_AND_CQNIB06g PAGE 10 QUESTION 3.01 (1.00)

Refer to figure 603, which one of the graphs A-D correctly deplicts the Recirculation Flow Control System, Master Controller output signal, based upon.the demand signal shown?

QUESTION 3.02 (1.00)

NOTE: 'A'.feedwater control is selected, Rx power at 100%, all controls in normal / automatic.

A failure in a Rx Vessel Level Narrow Range Instrument has occurred and resulted in the following related trips / indications:

"A" NR level indication reads at minimum "B" & "C" NR level indications read at maximum High level alarm (level 7) Low level alarm (level 4) are in

" Water level signal failure" alarm is in Two channels of level 8 have tripped One channel of level 3 has tripped Feedwater flow is at zero State which NR level transmitter has failed AND state in which direction itihas failed (HIGH/ LOW).

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' 'DijESTION 3.03 ( .50) d Answer TRUE or FALSE With both recirculation pumps running or both the recirculation pumps off, indicated core flow is a summation of loop flows.

s QUESTION 3.04 (1.00)

,e Answer TRUE or FALSE e a. If the command words from the two channels of the RACS disagree, a DATA FAULT., signal is generated.

b. If the command word from the RGDS system and the Acknowledge word from the Transponder card' disagree, a DATA FAULT signal is generated.

(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)

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3 __INgIBUDENIg_9ND_CQNIB9Lg PAGE 12 QUESTION 3.07 (2.00)

The RCIS system provides a means of substituting data from a good channel of rod-position for data from a bad channel of rod position.. In order to perform tiis substitution, certain conditions must be met. State

.these five (5) conditions.

QUESTION 3.08 (2.50)

Answer the following questions in regard to the RPV pressure instrumentation.

a. PT-(PIS)-N06BA has failed low; A Rx overpressure condition subsequently occurs (greater-than the relief setpoint for an SRV). Which solenoid (s)

(A,B) on the SRV would energize AND which pressure transmitter (s) will cause the A,B solenoid (s) to energize?

b. What installed design feature prevents an excessive inventory loss in the event of a pressure (or level) transmitter line break?

c. At rated Rx pressure, should a tailure occur on the ABOVE CORE PLATE pressure tap that causes a depressurization of the input to the CRD system, EXPLAIN the responses, if any, of

! 1. CRD drive water differential pressure INDICATION

2. CRD drive water Pressure Control Valve (FOO3)

QUESTION 3.09 (2.00) j Answer the following in regard to the Feedwater Control Systems

a. State the one condition which will cause an EAP " Lock-up" to occur.

b. While operating at 100% power, all controls in auto, the following events occur:

1. The 125V DC to the Turbine High Water Level trip circuit 'B' has deenergized.

2. The "C" NR Reactor Water Level transmitter fails high.

EXPLAIN the effect on the High Water Level Turbine Trip coincidence circuitry.

c. While operating at 100% power (all controls in automatic, MSC at high speed stop) the operator inadvertenly attempts to operate the Hydraulic Jack on a reactor feedwater pump, STATE the response of the feed pump turbine speed. (increase, decrease, remains the same)

(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)

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3___INSIBydENIg_@ND_CQNIBQL@ PAGE 13 QUESTION 3.10 (2.50)

Answer the following with regard to the Recirculation Pump Speed Control Sequences:

a. While transferring from fast to slow speed breaker 5A opens, but 5B remains closed. Explain the response of a pump to this occurance.

b. STATE TWO methods by which the " Incomplete Sequence Relay" can be reset after its operation.

c. List four (4) signals which will automatically transfer the Recirc Pumps from fast to slow speed. Include setpoints.

QUESTION 3.11 (1.00)

During RCIC system automatic operation a turbine trip signal was received and has been reset. The operator is preparing to reopen the Trip and Throttle valve. EXPLAIN the operation of the Ramp Generator / Signal Converter as the operator reestablishes RCIC operation by re-opening the Trip and Throttle valve.

Note Flow controller in AUTO, no other steam supply valves have operated.

(***** END OF CATEGORY 03 *****)

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QUESTION 4.01 (1.00)

In accordance with 04-1-01-M51-1 "Drywell Cooling System", should CRD cavity temperatures rise abnormally, at which one of the following temperatures could neutron monitoring cables and/or equipment damage begin to occur?

a. 125 deg F

b. 145 deg F

c. 165 dog F

d. 185 deg F QUESTION 4.02 (1.00)

Select which one of the below is a Safety Limits

a. Thermal power shall not exceed 25% of rated at less than 785 psig and core flow less than 10%.

b. MCPR shall not be less than 1.06 with Rx vessel pressure greater than 785 psig or core flow greater than 25%.

c. At greater than 25% power-MCPR will be less than 1.06 unless Rx vessel pressure is less than 785 psig or core flow less than 10%.

d. MCPR shall not be less than 1.06 with vessel pressure greater than 785 psig and core flow greater than 10%.

QUESTION 4.03 ( .50)

TRUE/FASLE. According to SDI 04-1-01-M71-1 " Containment and Drywell Instrumentation and Control", subsequent to an M71 system MANUAL isolation, resetting of the isolation signal logic is required prior to opening the individual valves by their handswitches.

QUESTION 4.04 (2.00)

The Conduct of Operations procedure has specified four conditions any one of which meet the requirements of " Adequate Core Cooling". State these conditions.

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(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****) )

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l as__EBDGEDWBEH_:_NDBMBLi_8BNQ8M@La_gMEBgg_gY_@ND N .PAGE 15 8901DLQ91G96_GQUIBDL-QUESTION 4.05 (1.50)

According to 01-S-06-2 " Conduct of Operations", the manipulation of controls at GGNS by anyone who is not a licensed RO or a SRO is permitted only when specific conditions exist. State two conditions which must exist to' allow another-individual (non-licensed) to manipulate the controls.

QUESTION 4.06- (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 (3) immediate operator action steps that are required per ONEP-05-1-02-I-1, " Reactor Scram."

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

QUESTION 4.07 (1.00)

SOI 04-1-01-L11-1 " Plant DC Systems", has the operator check for grounds on DC bus 11DA in the following manners (1) Verify approximately 62v DC on V1 and V2.

(2) Depress the MID POINT OFFSET pushbutton and interpret voltmeter readings:

V1 V2

1. 62 62

2. O 125 P

a. For each of the above conditions 1 & 2, state what type of ground exist e.g. no ground, positive bus and/or negative bus ground.

I

b. For the above circumstances, had the voltmeter reading been between the given values, what condition would this indicate? .

, QUESTION 4.08 (1.00)

(

, Hotwell level control is in Manual, the condensate makeup bypass valve is being used to control hotwell level. It is not possible to determine, by

( flow measurement, if adequate reject flow exist for CRD pump suction.

l According to 04-1-01-N19-1 " Condensate System", how can it be determined

! that adequate reject flow does exist? i

(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)

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4 __P8QQ[Qy@[]_;_UQQd@(t_@@@QBd@bt_EdgB@gNgy_AND PAGE 16 5891969GlC@(_CQUIgg(

QUESTION 4.09 ( .50)

SDI 04-1-01-P41-1 " Standby Service Water System", cautions the operator not to have open FOO6 (pump recirc valve) at the same time FO14 (SSW inlet to RHR Htx) and F068 (SSW outlet from RHR Htx) are open. What is the basis for this caution?

QUESTION 4.10 (1.50)

Refer to figure #636.

a. Under what condition (Suppression Pool temperature) does the operator leave step SP/T-177

b. After satisfying step SP/T-17 how does the operator continue in the procedure? i.e. to what procedure / step (s) does the operator procede, and what conditional items apply.

QUESTION 4.11 (1.50)

a. List the radiation dose standards for the following as stated in 10 CFR 20 for a restricted area without a completed NRC form 4.

REM /calander Qtr.

1. Skin of whole body

2. Hands and forearms, feet and ankles

3. Whole body, head and trunk, active blood forming organs, lens of eyes or gonads

b. What 3 conditions must be met prior to exceeding the whole body limit stated above?

QUESTION 4.12 (1.50)

Refer to figure #638.

Explain why RCIC injection is terminated at step SP/L-19 of EP-3 AND what adverse consequences could result from continued injection of RCIC.

l

(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)

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The CRD Hydraulic System Operating Procedure requires the CRD Flow Control Valve to be closed before starting a CRD pump. What is the possible affect of starting a pump with this valve open?

QUESTION 4.14 (1.50)

The Power Operations procedure requires all turbine bypass valves to be fully closed when withdrawing control rods with reactor power above the Low Power Setpoint. What is the basis for this requirement?

QUESTION 4.15 (1.00)

Per procedure 03-1-01-3, " Plant Shutdown", when the plant is shutdown to a " hot shutdown" condition the moderator temperature should be reduced to about 400 deg. F. What is the reason for reducing moderator temperature to this point?

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(***** END OF CATEGORY 04 *****)

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

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an2/t1/2 = 0.693/t1/2 g , , .p A= nD 2 1/2' 1" d)

, [(g1 /2) * (t e)3

E = 931 am

. m = V,yAo -Ex Q = m.ah I

I ,e Q = aCpat 6 = UA4 T I= I,e'"*

Pwr = Wfah I = I,10**/D L

! TVL = 1.3/v sur(t) HVL = -0.693/v P

p,p = P',10t/T l o

SUR = 26.06/T SCR =_5/(1 x II,ff) .

I CA, = 5/(1 - K,ff ,)

SUR = 26e/s= + (s - e)T s CR j (1 - K,ff)) = CR 2 II * "eff2)

T = ( **/e ) + ((s - s y Ie]

M " 1/(I - Keff) = CR /CR, j T = s/(e - s) M " (I - Keffo)/II~Keff1)

T = (s - e)/(ie)

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["0.1 A= seconds,j -

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Ijj=1d d

Id 2 ,2 2 gd P = (seV)/(3 x 1010) jj 22 2 E = eN R/hr = (0.5 CE)/d (meters)

I R/hr = 6 CE/d2 (f,,g) ,

Water Parameters Miscellaneous Conversions 1 gal. = 8.345 lem. I curie = 3.7 x 1010 dps 1 ga;. = 3.78 liters I kg = 2.21 lba 1 ft* = 7.48 gal. I hp = 2.54 x 103 8tu/hr .

Density = 62.4 lbg/f t3 1 av = 3.41 x 106 8tu/hr Density = 1 gm/cW Iin = 2.54 cm Heat of vaporization = 970 Stu/lem 'F = 9/5'C + 32 l Heat of fusion = 144 Stu/lbm ,

'C = 5/9 (*F-32) 1 Atm = 14.7 psi = 29.9 in. Hg. 1 BTU = 778 ft-Ibf I ft. H O 2

= 0.4335 lbf/in.

e = 2.718 l

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868 4 38 05 1054.0 1092.1 0.0745 1.9e00 24645 70 i 70 0.3629 0.01405 868 3 633.3 48.04 1048.4 1096.4 0.0932 1.94M 3A358 to 90 0.9048 0A1407 633.3 l en! M.02 1042.7 1800A 0 1115 12970 2.0004 to to 0.0001 0.01410 468.3 0.9492 OA1613 350.4 360.4 68 00 1037.1 1105.1 0.1295 13S30 1,9e25 les

, 300 77.98 1031.4 1109.3 0.1472 1A105 1.9577 11e 134 1J750 041617 265.4 2M.4 j

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! 140 24092 0.01629 122.98 123.00 97A7 117.95 2006.2 11N.1 0.2150 1.6536 1.0646 19e

' 190 3.718 OA16M 97.05 ges 77J9 127.M 1002.2 1130.2 0.2313 1.6174 1A487 l 340 4.741 SA1640 77J7 l

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=0 9.M0 SAiM7 0.2940 1.4824 1.7794 809 11.SN 0A1464 33.42 33.64 1M 09 9773 1144 4 see 178.15 971.6 1849.7 SJO91 1.4509 1.7600 210 l 310 14.123 048671 27A0 27A2 26.80 190.17 970J 1150.5 0.3121 1A447 1.7968 212 212 14.696 0.01672 26.78 age 23.15 188J3 966J 1153.4 OJ241 1A201 1.7442 330 17.1M 0.01678 23.13 age i

19.381 198.33 958.7 1157.1 0.3308 1.3808 1.7390 its 20.779 0A8685 19.364 16.304 16.321 208.45 M2.1 1840.6 0.3533 1.3000 ' 3.7142 340 I Sao 24.968 0.01693 1.3333 1.7000 350 13.002 13.819 218 H M S.4 1164A 0.3677 350 29A25 OA1701 1867.4 OJS19 1.3043 1.0062 age 0.01700 11.745 11.M2 228.76 938 6 200 35.427 1170.6 SJ000 1.270s 1.6729 370 0.01718 10.042 10.060 A38.9S 931.7 i 370 41.856 SAOBS 12501 1.0009 MS 0A1724 8.427 S.644 249.17 924.4 1173A See 49.200 1.3238 1.6473 age 87.550 DA17M 7A43 7A40 290A 917.4 11MA OA236 300 910.0 1179.7 0A372 1.1979 1.6381 300 300 47.005 OA1745 6.448 6.446 299.7 S.626 290.0 902.5 1182.5 OAS06 1.1726 1A232 310 330 77.47 OA1755 S.409 age 4.934 290.4 894 2 110S.2 0.4640 1.1477 1.4116 320 M.44 0.017M 4.896 See 3.788 311J SM.8 1190.1 0.4902 1A000 1.9032 See 117.99 0.01787 3.770 300 2.939 2.957 332.3 862.1 11M.4 0.9161 1A617.*.1.S$78 No 153.01 OA1811 353A S44.5 11M.0 0.9416 1A067 13473 300 '

380 195.73 0 41836 2J17 2.335 375.1 825.9 1201.0 0.5467 04607 1.5274 400 I 400 247.26 0.01864 1.8444 1A630 430 396.9 806.2 1203.1 0.5915 0.9165 1.9000 420 308.78 0.01894 1.4808 1A997 440 1.2140 419.0 785.4 1204.4 0.6161 02729 1.4000 440 381.54 0.01926 1.1976 400 0.9942 441.5 MS.2 1204.8 0.M05 0A299 1.4704 440 466.9 0.0196 0.9746 1.4516 400 CA172 464.5 739.6 1204.1 OA648 0.7871 480 566.2 0.0200 0.7972 487.9 714.3 1202.2 0.8000 0.7443 1.4333 900 SCO G80.9 0.0204 0.4545 0.6749 Sto 687.0 1199.0 0.7133 0.7013 1.4146 520 812.5 0.0209 0.5306 0.55M S12.0 3.3954 See 0 4451 534.8 657.5 1194.3 0.7378 0.4677 S40 962.8 0.0215 0 4437 0.6132 1.3757 580 0.3471 562.4 625.3 1187.7 0.7625 S40 1833.4 0.0221 CJ651 0.5473 IJ550 See 0.2994 0.3222 509.1 S09.9 1179.0 0.7876 SSO 1326.2 0.0228 550.6 1367,7 OAIM 0.llM 1.3330 900 0.0236 0.2438 0.2675 617.1 600 1543.2 506.3 1153.2 0A403 0.4089 1.3092 830 CA247 0.1962 0.2200 646.9 Me 620 1796.9 454.6 1133.7 OAGM SA1M 1.2821 2059 9 0.0260 0.1543 0.1002 679.1 000 See 392.1 1807.0 SAD 95 0.3502 1.2498 Geo 2365.7 0 0277 0.1146 0.1443 714.9 000 :

754.5 310.1 1066.5 0.936S 0.2730 1.2006 640 27006 0 0304 0.0008 0.1112 0.07S2 822.4 172.7 995.2 0.9901 0.1490 1.3390 700 700 3094 3 0 0366 0.0386 0 1.0612 705 0 0.0$08 906.0 0 906.0 1.0612 705.G 3208.2 0.0508 TABLE A.2 PROPERTIES OF SATURATED STEAM AND SATURATED WATER (TEMPERATURE)

A.3 i

T.

hiuste. gepa, Salltelpy. Dem4 GeheSy Seiset a Sang 1., . _ _F _ _ _ _ _ 9_ - ._ _ -

pelo

{. pelo w % % N % %% % % % g M6 mAm asuN =0re atteA me0 wFSA was e e.mte s.le7: e seni. .S M ,

l als mar 3 sSel.S EMS.S 3 03 len.S tom.s ee0u s.lsol .lpu sat aars.3 aN 49.463 3084.7 fn04 7 13 50 1967.9 3081 4 00271 3.1140 3.1411 1330 gggg.y ggg ,

! e.15 an p.16e aman ism.3 ISM 3 : .22 leu 6 1984 7 00422 a07M 2.n40 au2 MM3 an.

! SJO 64 484 0.01604 1039 7 1039 7 32.b4 1057.1 1089 7 00H1 3.0168 2A800 31.b4 1032 0 m30 '

geo a0l4% 792.0 792.1 40 92 1052.4 1093.3 0.0799 1.9762 2.0662 40.9;t 1034.7. 3.4 l 72.869 I e.S 79.S86 0 01607 641.5 641.6 47.62 1048 6 10M 3 0 0925 1A446 2.0370 47A2 10369 g ,

S3 25 1045 5 1058 7 01928 1.9186 2.0215 5324 ION 7 g4 0

9.6 85.718 0 01609 M0O S40.1

! .. . - . a gy . 90094 & Ol610 466 93 .. 466 94 . H 40 3042 7. 34(10A 03 .. 18966,, sag 3 . 9&40,.1984 3 ' 0J' .

4A 94.38 0 01611 411.67 411.69 62.39 1040 3 1802 6 0.1117 1A7M 1.9970 6229 1041.7 as

! E9 98.24 0.01612 He el MS 43 46 24 30M.1 1804.3 OlM4 14606 1.9870 8624 1942.9 E9

! 3.0 101.74 001614 333 69 333 40 40.73 10M.1 8105 8 0.1326 1A4SS 1A731 08.73 1944.1 1A SA 1M 07 0.01623 173.74 173.M M.03 1922.1 1116.2 0.1 H 0 1.7450 1.9200 94AB 1051A SA

! 3.0 141.47 0 01630 118 71 118 73 109 42 10132 1122 6 0.2009 LASS 4 late 109.41 1056.7 S.e i 4.0 152.96 0 01636 90 H 90 64 120.92 1006 4 1127J 02199 1A428 1A626 130.90 1960.2 4.0 S.O 162 24 0A1641 7s.515 73.53 130 20 1000.9 1131.1 0.2349 1A094 13443 130.18 1963.1 SA ,

j 1 S.0 170 05 0.OlMS 61.M7 Si tt 19803 996.2 11M.2 0.2474 19820 1 AIM IMAI 106S 4' SA i 7A 176 84 0.01649 S3 434 53.45 144 83 992.1 11M 9 0.2581 1.5587 1A148 144Al 1067A 7A

! 8.0 IS2 86 0 01653 47.328 47.35 150 87 908.5 1139.3 0 M76 14384 1A060 190A4 80092 SA l 9.0 ISE27 0 01456 42.385 42 40 156.30 985.1 1841.4 0.2740 1.5234 1.7964 156JS 1070A 9.4 30 193.2) 0.01659 38404 M 42 161.26 982.1 11433 02836 1.5043 8.7879 161m 1072J 30 l

14.496 212.00 0 01672 M M2 M SO 183.17 970.3 1850.5 0.3121 1A447 1.7568 180.12 1077.6 84ms j

) 15 213 03 0.01673 26 274 M.29 181.21 969.7 1150.9 0.3137 1A415 1.7552 181.16 1977.9 il

, 30 227 16 0 01683 20 070 to ob7 1M 27 963.1 1156.3 0.3358 IJ962 1.7820 19621 Im2A 30 l 30 250 6 0.01701 13.7266 13 744 218 9 945.2 1164.1 0.3682 1.3313 1.0805 314A Im7A 30 40 267.25 0 01715 10 4794 10 497 FM.1 . 933.6 1149A OJ921 1.2844 14765 3364 8002.1 de 50 28102 0.0172/ 84967 8 514 250.2

923.9 1174.1 0 4112 12474 ) ASS 6 29& 1 30063 Se 80 292 71 0.01738 7.1562 7.174 262.2 915.4 1877.6 0.4273 1.2167 1A440 303 4 1088 4 80 70 302.93 0.01748 61875 4205 272.7 907A 1180 6 0 4411 1.1905 1A316 272.5 1100.2 79 ,

0.4534 1.1675 1A208 301A 1802.1 j S0 312.04 0.01757 54536 5 471 242.1

900.9 1883 1 SS t 90 320.26 0 01766 4.8777 4A95 293 7 994 6 1185.3 0 4643 1.1470 1All3 390.4 1103.7 SS 800 327A2 0.01774 4.4133 4.431 298.5 808 6 1187.2 04743 1.1384 BA027 3982 31062 Sep I 120 341.27 0 01789 3 7097 3 728 312 6 877A 1193 4 0 4919 1.0900 1.5879 3122 Il07A 330 340 353 04 0 01833 3 2010 3 219 325.0 868.0 1893 0 0.S071 1.0681 1.5712 3245 1109.6 340

, 360 363 55 0 0'815 2A155 2334 336.1 859 0 1195.1 0.5206 1.0435 3.5641 835.5 1111.2 380 180 373 08 0 01827 2.5129 2.531 346.2 850 7 11M.9 0 5328 1.0715 1.SS43 34&A til2A 380 300 35180 0 01839 2.2689 2.287 MS.S M2A 1198.3 05438 1.0016 1.M54 354A 2113.7 MO l 375J 3115A l 250 40097 0 01865 1.8245 1A432 376.1 825 0 1201.1 0.S679 0 9585 1AM4 MO ,

300 417 35 0 01889 1.5235 1.S427 394 0 806 9 1202 9 0.5882 0.9223 1.5105 392A 1117J 300 350 di173 001913 1.3064 1.3255 409.8 794 2 12M 0 0 6055 00909 1A968 4084 IIle 1 350 400 444 60 0 0193 1.14162 1.1610 424.2 780 4 1204 6 0 6217 0 8430 1.4847 422.7 1112 7 400 450 456 28 0 0195 1.01224 .l.0318 437.3 M 7.5 1204.8 06360 02378 1A738 435.7 1118.9 400 S00 467 01 00199 0 90787 09276 449.5 FS$.1 1204 7 06490 08148 1A639 447.7 1118A 900 953 476 94 00199 0 82183 0 Salt 440.9 743.3 1204 3 0 6611 0 79M 1.4547 4M.9 1118 4 860 400 48523 0 0201 0 74M2 07698 471.7 732.0 1203 7 0.6723 0 77&3 1.4461 409 5 1136.2 000 '

703 .503 08 0.0205 0.63505 0 4556 491.6 710.2 1201.8 0 4928 0 7377 1.4304 408.9 1116 9 700 S33 514 21 0 0209 0 54809 0.5690 S09.8 689 6 1199 4 0 7111 0.7051 1.4163 506 7 till.2 000 900 131 93 0 0212 0 4796R 05009 S26 7 669 7 llM 4 0 7279 0 6753 1A032 6232 1113 0 900 1000 S44.55 0.0216 0 42435 0 4463 542.6 f 50 4 1192 9 07434 06476 1.3910 54".4 1830.4 1000 :

1100 550 24 00720 0 374(3 0 4006 557.5 631 S 1889 1 0 7573 06216 1.3794 553.1 1107.5 1300 !

3200 $57.19 0 0223 0 34013 0.362b 571.9 613.0 lite t 0 7714 0 6969 1.3683 SM 9 8104.3 1300 1500 57742 0 0227 030722 0.3299 $85 6 S44.6 1 ISO 2 0.7843 05733 1.M77 500.1 3100 9 1300 t 1400 517 07 0 0731 0 27871 0 3010 $$8 8 576 5 l175 3 0 7966 0 5507 l.3474 502.9 1017.1 3400 3500 5%20 00235 02h372 0.2772 611.7 550a 1170 1 0 8085 0?283 1.3373 4052 1093.1 1900 '

2003 635SO 0 OT.,7 0167% 01883 672 1 462 1138.3 05625 0 4256 1.7881 462 6 1068 6 3000 2500 65d11 0 02c,f 0 1020's 0 1307 731 7 361 6 1093 3 C 9139 03206 1.2345 FIS.S 8032.9 2900 <

3000 69533 0 0343 0 050/3 0 08h0 801 8 218 4 10703 0 9728 01891 1.1619 782A 973.1 3000 1

0 906 0 10612 0 1.0612 875.9 875.9 37083 ,

3298 2 101 47 00%8 0 0 050d 906 0 TABLE A.3 PROPERTIES OF SATURATED STEAM AND SATURATED WATER (PRESSURE)

A.4

t S (AKtK t S TIME

. STOP ROD PULL 8

START ROD PULL VO10 FRACTION Sf t nw REACTOR g PRESSURE 3-i $ _

5 ,,

REACTOR POWEPR LEVEL E l t IO Tiug 8 s

+

REACTOR PERIOD g TIME s

3 STEA F W

- nME Figure 619 Plant Response to Control Rod Withdrawal in Power Range 1

9 I

I

8

{

OP-DT-520/Revo 1 Page 13 of 13 150.

r q 100. i

1. FEEDWATER FLOW 50.

o. ... <

~ 1 s 100.

y 11 2

50. 2. NEUTRON FLUX 5

6 o. ..

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4. VESSEL PRES RISE (PSI)

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, 25. T l / 6. BYPASS VALVE FLOW

25. .. . I I !

25.

/ { 7. RELIEF VALVE FLOW

.ts, .r. l

o. 4 8. 12. 16.

TIME (SEC)

FIGURE 630 FEEDWATER CONTROLLER FAILURE, MAXIMUM DEMAND WITH TURBINE BYPASS l

! NTROLP - OP-DT-520 REV. 1 L

OP-B33-2-501/Rev. 2 Page 37 of 40 E h Flow

rCV CCNTROL .

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M 0pa N TO REI e- -e

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O NTROLP - OP/B33-2/501 REV 2 2

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l OP-N32-2-501/Rev. 1 Page 73 of 77 8!A$ A SYPA$$ C0fifR0'. VALyt CONTROLLER =

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ANSWERS -- GRAND GULF 1 -86/12/OS-CASTO, C

/nAS7 .

ANSWER 1.01 (1.00) b REFERENCE GGNS: OP-PC-505, pp 6 - 10 ANSWER 1.02 (1.00) i, d.

REFERENCE ,

GGNS OP RT 902 p. is l 292006 K1.03 2.9/2.9

) K1.04 2.9/2.9 i

I f ANSWER 1.03 (1.00) a.

REFERENCE GGNS HF 508-008 p. 19 ANSWER 1.04 (1.50)

a. V2 rpm 2 50 rpm 2 rpm 2 = 2000 rpm i

V1 rpm 1 40 1600

b. h2 (rpm 2)^2 h2 (2000)^2

} -~ = -------- -- = -------- h2 = 78.125'

! h1 (rpm t ) ^2 50 (1600)^2 I

I

c. W2 (rpm 2)^3 W2 (2000)^3 i

-- = -------- -- = ------- W3 = 126.95 hp l W1 (rpmt)^3 65 (1600)^3 l REFERENCE i l GONS OP CP 903 p.5 OBJ 2

{ 293006 K1.08 2.5/2.6 i

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ANSWERS -- GRAND GULF 1 -86/12/00-CASTO, C A

ANSWER 1.05 (2.00)

,a. RTS

b. Decrease i

c. Increase

d. Decrease REFERENCE C3'48 OP HF 303 pp.3-18 ',.

193,004 K1.16 2.1/2.3 .

ANSWER 1.06 (1.00)

/ ,a. de>krease

b. increase '

c. increase

d. nncrease i REFERENCE GGNB 65-1-02-I-5 p. 1 , i 204000 K3 K3.01 3.2/3.6 K3.06 2.6/2.7 l

r ; ANE k'ER 1.07 (1.00)

a. Velf t O'. 5 3

' b; increase, decreasing, CO.25 ea.)

RET'sHENCE s' GGNS OPRT Ob2 p. 10 OBJ 5 292000 K1.05 4.3/4.3 292003 K1.01 2.,7/3.0 8

ANSWEH 1.08 (1.50)

a. True to.53

b. At low power, the cold water sprayed into the core exit region has a more significant effect on moderator temperature and void content. t1.03 1

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1. ~ PRINCIPLES OF' NUCLEAR POWER PLANT OPERATION t PAGE 20 i lHgggDDYNgg!C@t_ HEAT TRANPFER AND FLUID FLOW ANSWERS -- GRAND GULF 1 -86/12/08-CASTO, C i

l i

REFERENCE. 'l '

GGNS OP E22-1-501 .

is 209002 A2.01 3.8/3.8

'r, s i5

I ANSWER 1.09 (1.00) ,

19.33 deg. F +-2 deg CO.53 Due te the additional subcooling in the downcomer region.EO.53 REFERENCE '

GGNS OP-B33-1-501 p. 10

.293003 K1.23 2.8/3.1 i \

ANSWER 1.10 (2.00)

),; a. 1. higher-

2. higher EO.5 ea.]

3. void b..EHC controls steam flow to maintain a constant Rx pressure. [0.53 a

REFERENCE GGNS GE Rx Theory Chap. 7 pp 7-16 OBJ 5-5.

292008 K1.19 3.1/3.2-I' ANSWER 1.11 (r7UUT

(,

L Due to a significan y diffe, rent &__ C C- ll2/ f 7 method of computing the TIP and LPRM f calibrations, cycle readings should generally be about 20% lower than i, actual.

l REFERENCE

[ GGNS OP IP 523 OBJ 2 p. 12 4

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. 1 1 __E81NCIE(E@_QE_NQCLE@B_EQME8_E(@NI_QEEB@IlgN t PAGE 21 IHE65QQYN@dlCS t_HE@l_IB@NSEE8_@NQ_E6Q1Q_E(QW 1

ANSWERS -- GRAND GULF 1 -86/12/OB-CASTO, C ANSWER 1.12 (2.00)

a. increse steam production due to pressure / power rise

b. Rx trip on L8

c. swell due to SRV openings

d. decay heat production REFERENCE GGNS OP DT 520 OBJ 1 259002 K3 K1.02 3.2/3.3 K3.01 3.8/3.8 ANSWER 1.13 (1.00)

a. decreasing the number of H2O molecules results in a longer slowing down length therefore neutrons travel through more fuel at epithermal energy range decreasing the resonance escape probability.

b. A fewer number of H2O molecules in the core results in a higher probablity of neutrons being absorbed by fuel atoms.

REFERENCE TPT CNTO Rx Core Control 3-9 GGNS GE RxTh sec. 1 OBJ 2 ANSWER 1.14 (1.00) 75*/. Void Fraction in the core.(0.5) This is because of the increased j reasonance capture which would occur (due to the longer slowing down length).(0.5)

REFERENCE EIH: L-RQ-604 GGNS: Reactor Physics L/P, pp 1.7 - 9, 10, 13 BSEP: 02-OG-A, pp 39 -49;GE RxTh sec 4.,OBJ 6.3 BFNP: Reactivity Coefficient LP, pp 4, 5; RQ 85/03/01 i

J

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2.__E60dI_pESigN_ lng 6UplNg_@@EEIy_8Np_EDE8@gNgy_@y@ Igd! PAGE 22 ANSWERS -- GRAND GULF 1 -86/12/08-CASTO, C ANSWER 2.01 (1.00) d _h l l'l REFERENCE GGNS SD-P42,P.3,19 ANSWER 2.02 (1.00) c REFERENCE GGNS: OP-C41-501; Print E1169-05 ANSWER 2.03 (2.00)

a. would not inject

b. would inject

c. would inject

d. would inj ect REFERENCE GGNS OP-E21-501 OBJ 3,7B 209001 K1.02 3.4/3.4 1.06 2.0/2.1 4.05 2.6/2.6 A2.02 3.2/3.2 ANSWER 2.04 (1.50)

BUV - 90% Bus UV for 9.0 seconds 80% Bus UV for O.5 seconds w/ LOCA signal 70% Bus UV for 0.5 seconds (2 D O.5 each)

LOP - All 3 lines supplying the ESF Bus are deenergized (O.5)

REFERENCE GGNS: OP-R27-501, p 7, 8; 04-1-01-R21-1, p5

7 t 0 UNITED STATES 9

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1. Containment purge system, by dilution.

2.. Hydrogen Recombiner, by the reaction 2H + 0 ---- 544 O.

2 2 2

3. Hydrogen Ignitors, by burning hydrogen at low concentrations. .

[0.25 ea.3 REFERENCE GGNS ID 356 OP-E61-501 ANSWER 2.06 (1.50)

Loss of " position indication light" Ecaf3 EO.53 Alarm "RCIC OOSVC" EO.53

, Indreatton crr' val ve:TessWconteol p5wer EO.53 vput fi n u L es 5 (o v i ll e r.L ' ^ c, k4-~ s LrfllT REFERENCE 6( If7'!9 7 GGNS OP-E51-501 Table 6 OBJ 7 217000 K6.01 3.4/3.5 A3.06 3.5/3.4 A4.10 3.6/3.5 4

ANSWER 2.07 (1.00)

LPCI can be and Cont. Spray cannot REFERENCE GGNS OP-E12-1 OBJ 3D 219000 A4.14 3.7/3.5 l

i ANSWER 2.08 (1.00)

1. Prevents reverse flow from operating the subloop.

2. Allows positive transfer to the other subloop.

REFERENCE GGNS OP-1333-2-501 OBJ 3A

< SWG #7 3.6/3.6 l

[

l 9 3KECoq' UNITED STATES

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2.__P(@NI_ DESIGN _lNCLUDING_S@EEIy_@ND_Ed@RG@NCY_@YSIEd@ PAGE 24 ANSWERS -- GRAND GULF 1 -86/12/08-CASTO, C ANSWER 2.09 (1.00)

a. Prevents drawing water from the Suppression Pool [O.253 when the line cools after discharge CO.253.

b. Water in the discharge line will cause a pressure transient [O.25]

in the event of a subsequent blowdown EO.253.

REFERENCE GGNS OP-E22-2-501 pp. 7 239002 K4.03 3.1/3.3 5.06 2.7/3.0 ANSWER 2.10 (2.00)

The HPCS D/G would start [0.53 due to LOCA and undervoltage signals O.25ea.

HPCS would not auto start CO.53 the pump control switch was taken to the trip position [O.253 which breaks the auto start signal [0.253.

REFERENCE GGNS SDI 04-1-01-P81-1 pp. 34 209002 K1.04 3.8/3.8 A1.07 2.5/2.8 A1.08 3.1/3.3 ANSWER 2.11 (1.00)

a. Would not auto start CO.253 due to voltage regulator switch in manual CO253.

b. Would start [caf3 EO.53.

REFERENCE GGNS OP-922-1-501 Table 3 OBJ 7E 209002 K1.04 3.8/3.8 6.01 3.6/3.6

UNITED STATES

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2___P(QNI_ DESIGN _lNC6UDING_@@FEIY_8ND_EdERGENCY_SYSIEd@ PAGE 25 ANSWERS -- GRAND GULF 1 -86/12/08-CASTO, C ANSWER 2.12 (1.50)

a. PIC R605 maintains RCIC suction pressure [0.253. F065 is modulated by R605 to maintain RCIC suction pressure below the high limit setpoint

[0.253.

b. The CST is not valved out, but is controlled by check valves [O.53.

c. The RCIC pump turbine exhaust is directed to the Suppression Pool CO.53.

REFERENCE GGNS SD E12 pp 37, E51 pp 12, SD E12 pp. 3,37 (IIA) OBJs 6.a.2,98.

ANSWER 2.13 (1.00)

The FCV opens up as reactor pressure increases maintaining a constant flow, therefore, a constant pressure to the PCV.

REFERENCE GGNS SD C11-1A 201001 K4.08 3.1/3.0 A1.01 3.1/2.9 ANSWER 2.14 (1.00)

To ensure valve operation does not intefere with Post-LOCA Recirc Pump coast-down.

REFERENCE GGNS OP B33-1-501 pp 11 202001 K1.01 3.6/3.7

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3 1__1 dele 9dEd1E_6dE_E96169hg PAGE- 26

~ ANSWERS -- GRAND GULF 1 -86/12/08-CASTO, C ANSWER 3.01 (1.00) d.

REFERENCE GGNS OP B33-2-501 p. 7, OBJ 3B 202002 K4.08 3.3/3.4 5.03 2.4/2.4 ANSWER 3.02 (1.00)

A-low to.5 ea.]

REFERENCE GGNS OP-B21-501 259002 K5.01 3.1/3.2 ANSWER 3.03 ( .50)

TRUE REFERENCE GGNS OP-B21-501 216000 K1.10 3.2/3.4 1.23 3.3/3.4 ANSWER 3.04 (1.00)

a. False

b. True REFERENCE GGNS OP C11-2-501 201005 A2.10 2.8/3.0

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3 __INgIByMENig_AND_CQNIBQL@ PAGE 27 ANSWERS -- GRAND GULF 1 ' 86/12/OB-CASTO, C ANSWER 3.05 (2.50)

L,' ) , $ ' 3 2 7

a. EN!r.-T MWE C 1. 03

b. 1. 175 MWE / ~J 3 , ~7 (c / 2, f

2. O MWE 3 <r ,1

3. 17% SF #3,y CO.5 ea. close apprdimations acceptable]

REFERENCE GGNS OP-N32-2-501 241000 K3.06 4.1/4.1 3.08 3.7/3.7 4.11 2.6/2.6 ANSWER 3.06 (2.00)

a. 1.a. ADS mode 2.a. ADS mode

b. Valve does not operate b. Rel i ef mode

c. Safety mode c. Safety mode g, CO.25 ea.3

b. "A" sol enoid valve. CO.53 Y1 ( j e.k ! If(([C tt ( Lt "V7 g, g e i3 ( '"y REFERENCE 't*

GGNS OP-E22-2-501 pp 18 OBJ 6B 239002 K4.08 3.6/3.7 218000 K4.01 3.7/3.9 4.03 3.8/4.0 ANSWER 3.07 (2.00)

CO.4 ea.]

1. Rod must not be moving

2. Rod's good channel of data must be selected for display l 3. Raw data must not be selected l 4. No other rod in that gang has substituted position already active

5. Rod must be selected REFERENCE GGNS OP C11-2-501 201005 K6.02 3.2/3.3 A2.02 2.8/3.2 l

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3 __1NglggdEylg_ANQ_CQN16QL@ PAGE 2D-ANSWERS -- GRAND GULF 1 -86/12/OB-CASTO, C

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ANSWER 3.08 6

a. M B W PIS-N0681 ct ll v) Y ]

B- by PIS NO68 B and F [0.2 ea.]

b. a restricting orifice [caf high flow valve] [O.53

c. 1. the indication would read at maximum

2. no affect on valve operation (no response)

[O.5 ea.]

REFERENCE

( GGNS SD B21 III B pp 19 SD B21 p 17 SD C11-1A p 5.

216000 K1.08 3.7/3.9 3.08 3.9/4.1 A2.'03 3.0/3.1 l K3.27 2.9/3.1 l

ANSWER 3.09 (2.00)

a. The output of the RFP EAP controller drops below 5 volts [O.53 l b.

A trip is generated in the B & C channels completing the coincidence '

l for a turbine trip. [1.03

c. increase [caf for interlocks 3 CO.53 )

REFERENCE GGNS OP C34-501 OBJ 7A, 6.c.4.

259002 K4.06 3.1/3.2 3.06 2.8/2.8 l

t i

ANSWER 3.10 (2.50)

/ tso [

/

a.{Thepump CO.53 ' A ' / %will w j" not Lvtransfer to s'n the~LFMG, it will coast down to zero rpm.

\

?L L r ow % > f S f mA b.-Depress the "stop" or "stop. lock" pushbuttons on the CB-5 handswitch.

[O.25 ea.]

c. LL3, 11.4" .

[1.5 = .2 parameter, .1 setpoint]

FW' Flow <22% for 15 sec.

i Steam dome temp / pump suction delta T 47.4 deg F EOC RPT logic 1/2 stop valves (40 psi) 2/2 control valves REFERENCE GGNS OP B33-1-501 202001 K4.16 3.3/3.6 2.26 2.9/3.1 I

l l

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3 __IN]IBQUENIS_@ND_CQNIBQL@ PAGE 29 ANSWERS -- GRAND GULF 1 -86/12/08-CASTO, C ANSWER 3.11 (1.00)

The ramp generator is at maximum, since it does not reset unless F045 is closed [0.503, the signal converter would control the governor valve CO.53 (flow controller)

REFERENCE GGNS OP E51-501 p. 8 OBJ 3A 217000 A1.05 3.7/3.7 4.02 3.9/3.9 2.12 3.0/3.0 3.02 3.6/3.5

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REFERENCE GGNS SOI 04-1-01M51-1 223001 A2.10 3.6/3.8 ANSWER 4.02 (1.00) d.

REFERENCE GGNS TS 3.3 ANSWER 4.03 ( .50)

False REFERENCE GGNS SOI 04-1-01-M71-1 223002 K4.06 3.4/3.5 A4.03 3.6/3.5 l

ANSWER 4.04 (2.00)

1. RWL maintained above TAF

2. Core is being sprayed by either HPCS of LPCS

3. sufficient steam flow through the core l 4. reflooding flow of one LPCI pump is injecting into the core with I

RWL high enough to produce two phase flow up through the core.

l REFERENCE l GGNS 01-S-06-2 sec 5.10 i

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4t__P8QgEDU6g@_;_NQBd@(t_@@NQBd@(t_Edg6@gNQY_@NQ PAGE 31 B00196QGlC@(_CQNIBQL ANSWERS -- GRAND GULF 1 -86/12/08-CASTO, C ANSWER 4.05 (1.50)

1. The individual is in training to qualify for an operator license

2. Under the direction of a licensed operator RO or SRO present at the controls REFERENCE GGNS 01-S-06-2 6.4.5 294001 K1.05 3.2/3.7 ANSWER 4.06 (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 that are not full-in to position 00. (0.5 each)

REFERENCE GGNS: ONEP-05-1-02-I-1, p3 7r ANSWER 4.07 (-iWOT a.1. no ground [O.25 ea

2. Ansktiv: :.,us p nd-<{ d U ,3 3 '

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b. A high resistance ground exist CO.503 REFERENCE GGNS SOI 04-1-01-L11-1 263000 K1.04 2.6/2.9 A2.01 2.8/3.2 A4.04 3.0/3.2 SWG 15 3.4/3.8

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4 t__P@ggggy@g@_;_MQ80@(t_@lNQBd@(t_gd[8@gNgy_AND PAGE 32 6091969GlC@(_CQNIBQ6 ANSWERS -- GRAND GULF 1 -86/12/08-CASTO, C ANSWER 4.08 (1.00)

Check the CRD 0 concentration if it is greater than hotwell O concentration 2 2 then adequate reject flow does not exist.

(if it'is less adequate flow does exist)

REFERENCE GGNS 04-1-01-N19-1 201001 K5.04 2.4/2.4 K1.01 3.1/3.1 A2.OB 2.8/2.8 ANSWER 4.09 ( .50)

This alignment may cause pump runout.

REFERENCE GGNS SOI 04-1-01-P41-1 203000 K1.16 3.1/3.2 ANSWER 4.10 (1.50)

a. when temperature is < 95 deg. CO.53

b. Procede to SP/T-1 and -2 [O.53 and remain until all entry conditions have cleared. [0.53 REFERENCE GGNS OP EP/SPDS 504 OBJ 3 ANSWER 4.11 (1.50)

a. 1. 7.5

2. 18.75

3. 1.25

b. Dose does not exceed 3 rem /qtr accumulated occupational dose does not exceed 5(N-18) completed NRC form 4 CO.25ea.3

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REFGRENCE GGNS ID 2796 LP OP-PB 601 l' s ANSWER 4.12 (1.50) i Due to suppression pool temperature being above 140 deg. F....[O.53,,RCIC lube oil will ^ ntst be cool ed [0.53 theref ore RCIC i s secured to prevent l-damage to the turbine [0.53. ,;g yegg g f- ,g g g ,,,3 gu G REFERENCE b .pg w W cc <- 2. t - s 7 '

GGNS OP-EP/SPDS 504 OBJ 3 III 2170C0 K6.03 3.5/3.5 -)

A1.08 3.5/3.6 < ,

A7.19 3.5/3.6

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>l ANSWER 4.13 (1.00) i It is possible to cause a rod drif't'as a' resp'It pf the cooling water y .. pressure transient. ,

REFEPhiNCE- '

GGNS CRD GP .iEOP) 'i

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! i ANSWER 4.14 (1.50) --

This Rx power input for rod control is determined from first stage pressure CO.53. With the bypass valves open, the RCS senses Rx power as less then actual and the potential exist,for a nonpconservative rod withdrawl [1.03 s

REFERENCE -

GGilS 03-1-01-2 and TS 3.4.1.4 (EGB).

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ANSWER 4.15 (1.00) r minimize thermal stresset On the feedwater nozzles. -

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