ML20196D616

Forwards Comments Compiled by Training Dept Re Reactor Operator & Senior Reactor Operator Written Exams Conducted at Plant on 880919

ML20196D616
Person / Time
Site:	Vogtle
Issue date:	09/21/1988
From:	Bockhold G GEORGIA POWER CO.
To:	Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
Shared Package
ML20196D574	List: ML20196D570 ML20196D602 ML20196D616
References
NUDOCS 8812090145
Download: ML20196D616 (98)

Text

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GoW a Pw Co"mi ENCLOSURE P um oun s P y,t Om:n &;. '(N WV+ 5 t< Q Ge : ;1? 4 T

'c e;' ore 4 4 72 % P114 A:4 M 4 :+ (

• SEP 221988 L '

voce proieci (;tgugiti[ygnir September 21, 1988 Plant Vogtle - Units 1 & 2 Cornents - 9/19/tu R0/SRO Written Examination Log: N07-01530 Security Code: NC Hr. John Hur?o Nuclear Regulr ary Commission 10' Hacietta .,. eet. NW.

Atlanta. M 30323

Dear Mr.11unro:

Attached are the concents compiled by the Training Department regarding the R0 and SRO written examinations conducted at Vogtle Electric Generating Plant on Septem'er 19, 1988. Your consideration of there r,omentr in grading of the examinations is appreciated.

Very truly yours, l ,

'C b G. Bockhold. Jr.

General Manager POR:kss xc: W. G. Hairston C. K. McCoy ssi;0:0145 591125 P Dr< ADOCL 050004:4 V FDC

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GUEST 10N 6.17 (1.00) i State the locations where the TWO Safety Grade Cold Shutdown Letdown Flowpaths discharge.

6.17 (1.00) go.5 each) .

g, PRT l

2 Excess Letdown Heatexchanger inlet

, pgfEftENCC l

VEGP Training Tt :tt . CVCS, Chapter 5.a, p. 0 (3.6/3.4) 004010A307 ..(KA's)

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eum QUESTION g (1.00) l State tha2 M ations eshare the TWO SAFCTY GRADE Co\d Shutdow.4 Lutdown i Flowpaths discharge.

i A 64.ER 2.16 (1.00) t L

1. PRT. '

2. Excess Letdown Heat Exchanger flowpath (OR upstream of the excess 1stcown heat exchanger).

REFERENCC VEGP Training Text, CVCS. Chapter 5.a. p. 6 3.6/3.4 l

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QUESTION NUttBER: 2.16 (2)/6.17 FACILITY COMMENT: Accept these alternate answers for excess letdown:

1) Seni return fl.vpath to the charging pump header OR

2) Reactor coolant drain tank, since these are specific locations that excess letdown flowpath discharges to.

REFERENCES:

Training Text, CVCS, Chapter 5.a. Pg. 8 1X4DB-111 IX4DB-ll4 l

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CHEMICAL AND VOLUME CONTROL SYSTEM intermediate leg and passes through the excess letdown flow piping to the excess letdown isolation valves (HV8154 and HV8153). It then flows tnrough the tube of the excess letdown heat exchanger giving up much of its heat to the ACCW System. Excess letdown is reduced .

from normal loop temperature to approximately 190_ F. It then passes first through a check valve and then through the excess letdown hand control valve (HV123) which provides a means of flow modulation or isolation. The flow continues to the excess letdown

.three-way flow direction valve (HV9143) where it can be directed to the RCDT or the RCP seal water-return header.*It'is normally.

directed to -the seal water return' header. The excess letdown.then continues,through.the normal seal-water. return fIow' path to-the charging. pump suction header. The^three-way divert valve (HV9143) provides an' alternate"flow ' path'to 'the reactor coolant drain tnnk for use.during-operations such as a plant heatup.

Safety Grade Cold Shutdown Letdown Flow The safety grade cold shutdown letdown flowpath provide letdown from the reactor vessel head vent through two trains of solenoid isolation valves (HV9095A and HV9096Al Train A and HV90958 and HV9096BI Train B). From these valves, the flow can be directed to the PRT through isolation valves (HV0442A and/or HV04428), or to the excess letdown flowpath, upstreae of the excess letdown heat exchanger, through a motor-operated isolation valve (HV9099).

i h. Chemical Control, Purification and Makeup (Figure 5a-6)

The water chemistry, chemical shim and makeup requirements of the RCS are such that the following functions must be provided:

1) Means of addition and removal of pH control chemicals for startup and normal operation.

2) Control of oxygen concentration during normal and shutdown operation of the plant.

l l 3) Neans of purification to remove corrosion and fission products.

y 4) Neans of addition and removal of soluble chemical neutron absorber (boron) and makeup water at concentrations and rates l compatible with all phases of plant operation including emergency situations.

1. pH Control The chemical control slement employed f or pH control is lithium l

hydroxide (LiOH). This chemical is chosen f or its compatibility with the materials and water chemistry of borated water / stainless steet/zirconius systemsg in addition, Li-7 is produced in the core region due to irradiation of the dissolved boron in the coolant.

The L10H is introduced to the Reactor Coolant System via the charging flow. A chemical mixing tank is provided to introduce the 5a-8 Revision 2

QUESTION 5.03 (2.50)

Compare the calculated Estimated Critical Position (ECP) for a startup to be performed 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after a trip from 100% power, to ACTUAL control rod position if the following events / conditions occurred.

Consider each independently. (State your answer as Actual position is HIGHER THAN the ECP, the Actual posicion is LOWER THAN the ECP, or the Actual position is the SAME AS the ECP)

a. One reactor coolant pump is stopped two minutes prior to criticality.

b. The S/U is conducted 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> earlier than planned.

c.

The steam dump pressure setpoint is reduced to 950 psig.

d. Condenser pressure is reduced from 7 psia to 2 psia.,

e. All S/G 1evels are allowed to decrease by 5% as the rod withdrawal occurs during the startup.

ANSWER 5.03 (2.50)

a. SAME AS

b. HIGHER THAN

c. LOWER THAN

d. SAME AS

e. SAME AS REFERENCE VEGP Lesson Plan, LO-LP-33510-01 l Learning Objective #14 1

001010A201 001000K534 001000K104 . . (KA's) 0

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QUESTION NUMBER: 5.03b FACILITY COMMENT: Answer assumes Xc peaks 8-9 hours following a reactor trip.

Answer could change based on assumption for peak Xe. (i.e.,

Xe peaks at 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> - SAME AS). Recommend allowing credit for answers based on initial peaking assumptions.

REFERENCES:

LO-LP-33430-02, Pg 7

i Ill. LESSON OUTLINE: NOTES '

Zenon-135 Behavior Following Shutdown D.

3

1. d approx. equal to 0 a ., 0 ,**N g ,d = 0, burnout ceases

b. d "*1 = 0, production from fission

/. lies c

2. Only production is by continuing decay of 133 I

3. Only removal is by l3 Ie decaying to 135 Cs I "I ~

Ia N, g

a. Initially produced at a faster rate than it decays

1) Lost a major removal process

2) only lost SI of production

3) 'Ie concentration initially increases ..

b. If shutdown lasts long enough i

1) Rate of removal approaches and then exceeds rate of production

2) When gg rates are equal this is tre Time T @ '(6-10 peak Io concentration bre) approx.

-5150 pcs a) )g N g= kg, Ng , (2500 p m) b) Time .te reach peak varies from 6 to 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, taking longer for higher ,

power levels - rule of thumb is i appros. equal to_the square root of

.equil. power.

s, c) I35 h wrth large et of p

3) After peak, 135Is decay dominates and d

concentration decreases to aero. (0 , ,

5. 135 1e worth after shutdown @bM'S'M Refer to Reactor

6. Feaking of I3I Ie concentration - important Engineering Manual consideration for reactor startup for most recent curve

a. least be enough t(d,,,,,availabletoover-135 come -p'distroduced by xe - W ortant at EOL l 7 l

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J QUESTION 5.16 (1.00)

Answer the following questions plant. in reference to subcooling margin of the a.

. What is the subcooling margin of the plant exists if the following conditions That = 587 F Tave = 572 F Ppar = 2235 psig Psg =

1033 psig Tcold = 557F

b. If power decrease? is raised from 50% to 100%, why does the subcooling margin ANSWER 5.16 0 (1.'0) c.

Tsat for 2250 psia (2235.psig)

= 652.67 F (0.50)

Subccoling margin = Tsat - Thot =

l 652.67 - 587 = 65.67 F + or -

0.05 F (0.50) b.

Subcooling margin decreases because That will increase as power increases (0.50) 6 AEFERENCE VEGP Lesson Plan, LO-LP-04110-03 i

001000K556 ..(KA's) l l ._

QUESTION NUMBER: 5.16a FACILITY COMMENT: Margin of tolerance is too small (0.05'F). T is given as 587. Answer requires 5 significant figures. HEecommend expanding margin to allow for rounding off in c:1culation.

Suggest 22*F.

REFERENCES:

N/A

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, A QUESTION 5.18 (1.50)

Following the taking of critical data, L ctartup rate of 0.15 DPN at 0.1% of full powerin the intermediate range, a stable l reactor is at no-load Tave, is established. The I plant is belew the point of adding heat.and the Baron concentration is 800 ppm. The

a. What will the reactor power be after 2 minutes? '

b.

At (No what power level calculation does the point of adding hest begin?

required.)

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g wER 5.18 (1.50)

a. P = Po10edp (SUR) (t) (0.5)

=

(0.1 ) 10ex p (0.15) (23

=

0.199526% (0.5) i, Round off to 0.2 x 15 acceptable.

l b.

1E-6 to 3E-6 ampd or 1% to 2% of full power. (0.5)

REFERENCE VEGP Lesson Plans, LO-LP-33230-01, P. 6 Lesson Objectives, LO-LP-33230-01, #3 19.7009K114 ..(KA's) l

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QUESTION NUMBER: 5.18(b)

FACILITY COMMENT: LicensecandidagesaretaughtthattheP0AHisequivalentto approx. 5 x 10- amps which is also the point where power becomes visible in the power range (-1-27.). This is' consistent with actual indications seen on the simulator during training. Recommend wider tolerance on acceptable answer.

(Cannot find' reference to values listed in. key.)- i

REFERENCES:

N/A  ;

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i

9 CUESTION 6.04 (2.00)

List FOUR permissives that must be satisfied before the Diesel Generator output breaker will shut following a start by the Safeguards Sequencer.

ANSWER 6.04 (2.00) l' (any 4 0 0.5 each)

1. DG ready to load (>440 RPM, 90% voltage)

2. Preferred incoming breakers open.

I

3. No bus faults as indicated by no bus lockouts on preferred bus. '

4. DG close permissive from sequencer.

5. No lockouts on Diesel engine or generator.

REFERENCE VEGP Lesson Plans, LO-LP-01401-01, p. 8 (3.5/4.0)

• 064000K411 ..(KA's) a b

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r QUESTION NUMBER: 6.04 FACILITY COMMENT: Answer gives partial list of requirements for DG breaker closure. Recommend accepting additional requirements as correct answers.

- Breaker "Local-Control Rocc" selector switch in the control room position (See TRS-LR).

- Breaker control switch in "AUT0" (See CS-R).

REFERENCES:

1X3D-BA-D02D

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QUESTION 6.07 (1.50)

How must each of the following parameters change to achieve an increase in the Overtemperature Delta Temperature Setpoint (OTdelta T)?

Uae INCREASE, DECREASE, REMAINS THE SAME.

a. Tave

b. Reacter Pressure

c. Delta Flux Penalty ANSWER 6.07 (1.50)

a. Decreases

b. Increase (0.5 each)

c. Decreases REFERENCE VEGP Training Text, Chapter 25 Logics sheet 5.

012000A101 ..(KA's)

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QUESTION NUMBER: 6.07c FACILITY COMMENT: Delta flux penalty [f (Delta I)] can only cause the OT Delta T setpoint to decrease. Changes in the parameter in either direction beyond the allowed range will decrease the setpoint.

Recommend delete part c.

REFERDiCE: TS Table 2.2-1 l

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TABLE 2.2-1 (Continued) 8 TABLE NOTATIONS N NOTE 1: OVERTEMERATURE AT AT **

3,, (1 f 13 5IIO o -

2

, , , D (1 f ,5) 1 - T *] + K 3 (P - P') - },N)[]

w idhere: A1 =

Measured AT by RID Manifold Instrumentation; 1+ 5 =

Lead-lag compensater on measured AT; ti, 12 =

Time constants utilized in lead-lag compensator for AT, t i 18 s.

1 2 1 3 s; I

=

3, 3 Lag compensator on measured AT;

,, ,_ v3 =

• Time constants utilized in the lag compensator for AT,13 = 0 s; AT, =

~

Indicated AT at RATED THElWHL POWER; s.

X < 1.10; K2 = 0.012/*F; I * **5 =

3, 5 dynamic compensation;The function generated by the lead-lag compensator for T

14. rs =

Time is constants utilized in the lead-lag compensator for T,,g, 5 4 5; 4 1 281 s, 1 =

Average temperature. *F; 1

=

g, g tag compensator on measured T,,9;

s. =

Time constant utilized in the measured I,, lag compensator, to = 0 s;

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TABLE 2.2-1 (Continued)

TABLE NOTATIONS (Continued)

NOTE 1: (Continued) 1 j I* $

i SS8.5*F (Nominal T,,g at RATED THERMAL POWER);

K3 = 0.00056/psig; i

=

P Pressurizer pressure, psig; P' =

2235 psig (Nominal RCS operating pressure); '

S =

Laplace transform variable, s ';

and f,(al) is a function of the indicated difference between top and bottom detectors of the power-range neutron ton chambers; with gains to be selected based on measured instrument response during plant startup tests such that: (

- (1) For gt'9b between -33.5% and + 6.5%, f (AI) = 0; where q ad % t are Percent RATED THERMAL I

• POW R in the top and bottom halves of the core respectively, and q + g is total THERMAL . , ,

POWER in percent of RATED THER m t POWER; (2) For each percent that W magnitiade of qg g exc d " 33. 2 , the AT Trip Setpoint shall be autamatically reduced by 1.27% of its value at RATED TENGALl PERER; and.*

(3) For each percent that the magnitude of qg gexceeds+6.2,theATTrip[Setpointshall i

be automatically reduced by 0.83E of its value at RATED TEment POWR. -

1 l NOTE 2:

The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2.5%.

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a QUEDTION 8.02 (1.00)

While operating in Mode 1, control power is lost to a pressurizer .

power operated relie+ valve, which one statement below is CORRECT 7

a. Technical Specifications requires no action provided another PORV is operable and all pressuriger code safety valves are operable.

b. Technical Specifications requires the power supply to be removed from the associated block valve after verifying it to to open, if the PORY is not operable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

c. Technical Specifications requires the associated block valve '

to be shut and its power removed if the PORV is not made operable within i hour and continuous operation is desirable.

d. Technical Specifications requires action restore two PORVs to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in het standby within next 6 h,ours and hot shutdown withik fallowing 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ANSWER 8.02 (1.00) d l

stEFERENCE VEGP T.S. 3.4.4 o'0000A203 ..(KA's) i i

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QUESTION NUMBER: 8.02 FACILITY COMMENT: For the given conditions, the following Tech Spec actions would be required

- Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV to OPERABLE status or close the associated block valve and remover power from the block valve, AND

- Restore at least a total of two PORVs to OPERABLE status within the a^ollowing 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

IAW TS 3.4.4b, the correct answer would be "c" and "d".

Accept "c" since it is the most correct, or since no question is totally correct, delete question.

REFERENCES:

TS 3.4.4b 4

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s REACTOR COOLANT SYSTEM 3/4.4.4 RELIEF VALVES LIMITING CONDITION FOR OPERAT_ ION 3.4.4 All power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With one or more PORV(s) inoperable, because of excessive seat leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV(s) to OPERABLE status or close the associated block valve (s); otherwise, be in at least HOT STANDBY following within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,

b. With one or more PORV(s) inoperable dug to causes other than exces-sive seat, leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV(s) to OPERABLE status or close the associated block valve (s) and remove' power from '

the block valve, and

1. With only one PORY OPERA 8LE, restore at least a total of two PORVs to 0PERA8LE status within the following 72 hourt or be in H0T STAN06Y within the rext 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, or

2. With no PORVs OPERABLE, restore at least one PORV to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in HOT STANOBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

c. With one or more block valve (s) inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (1) restore the block valve (s) to OPERA 8LE status or close the~ block valve (s) and remo e power from the ' clock valve (s) or close the PORV and remove power from its associated solenoid valve; and (2) apply ACTION b above, as appropriate, for the isolated PORV(s).

d. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.4.4.1 Each PORY shall be demonstrated OPERABLE at least once per 18 months by:

a. Operating the valve through or,e complete cycle of full travel, and

b. Performing a CHANNEL CALIBRATION.

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i V0GTLE - UNIT 1 3/4 4-10 I

QUESTION 8.07 (1.00)

In accordance with VEGP 10003-C, "Manning the Shift", during WHAT modes must a SRO be present in the Control Room?

ANSWER 8.09 (1.00)

Modes 1 to 4 Co.53 for either unit Co.5? ,

REFERENCE VEGP Operations Administrative Procedure, 10003-C, p. 1 194001A103 ..(KA's)

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QUESTION NUMBER: 8.09 FACILITY COMMENT: The question only asked for the modes an SRu was equired to be present in the control room and did not elicit the response "in either unit". Recom:nend full credit should be given for stating modes required.

REFERENCES:

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QUESTION 8.12 (1.50)

Liot SIX typoo of cctivitioc which cro record 0d in tho Shift Supervisor's log except for the time when the activities occurred.

ANSWER 8.12 (1.50)

(any 6 0 0.25 each)

1. The name and position of each operator on shift.

2. Major equipment status changes. ,

I

3. Major system and equipment testing.

4. Personnel injuries.

5. Entering and exiting a Technical Specification action statement.

6. Significant svents, such as reacter trips or unexpected power changes.

7. Implementing the Emergency Plan.

9. Significant security incidents.

. 9. Mode changes. ,

)

REFERENCE VEGP Operations Administrative Procedure, 10001-C, sec. 2.2.1, p. 1 194001A106 ..(KA's)

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QUESTION NUMBER:. 8.12

~ FACILITY COMMENT: Answer gives tieneric list of activities that are recorded in the SS's log. Recomunend credit for s'jecific examples of those events that are required to be recorded.

REFERENCES Ngg i

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T - J QUESTION 9.14 (2.50)

List FIVE duties of the Emergency Director which are NON-DELEGABLE according to Emergency Response Procedure,.91102-C, Duties of the Emergency Director? ,

a' ANSWER 8.14 (2.50) i (any 5 e 0.5 each)

1. Classifying and declaring the emergency. (including downgrading and terminatt,,7g.)

2. Recommending protective actiorm to off-site authorities and content of messages.

3. Authorizing personnel radiation exposures in excess of 10CFR20 limits.

4. Deciding ?,o evacuate non-essential personnel form site et Alert Classification. '

5. Deciding to request assistance from federal support groups.

6. Deciding to notify off-site authorities responsible for emergency measures.

QEFERENCE ,

I i VEOP Emergency Response Procedure, 91102-C, sec. 2.3, p. 2 19400;f4116 ..(KA's) l I  !

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QUESTION NUMBER: 8, .4 FACILITY COMMENT: Recommend full credit to be given for downgrading and terminating the emergency. These duties are separate and independent of declaration and classification.

REFERENCES:

91001-C Pg 1, Step 2.2 t

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• Vogtle Electric Generating Plant 01-C NUCLEAR OPERATIONS g g ,,,, g ,,

OMMON Unit Georgia Power "* " 1 o f 12

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R EPl_ACEMENT EFERGNNCY CLASSIFICATION AND IMPLEMENTING INSTRUCTIONS MINUIL SET 1.0 PURPOSE 1.1 The purpose of this procedure is to provide instructions in the classification of off-normal events into one of four emergency classification levels. This procedure also provides initial implementing instructions for each emergency classification.

2.0 RESPONSIBILITIES 2.1 The On-Shift Operatione Supervisor (OSOS) is -

responsible for initial classification of events. The OSOS shall assume the responsibilities of the Emergency Director until relieved. The OSOS then becomes responsible for recognizing changes in plant conditions and advising the Emergency Director concerning classification of events.

2.2 The Emergency Director has the following non-delegable responsibilities relative to emergency classification.

2.2.1 Classifying and declaring the emergency.

2.2.2 Declaring changes in the amergency classification, including downgrading and terminating.

2.3 The Technical Support Center (TSC) and the Emergency Operations Facility (EOF) Managers are responsible for:

2.3.1 Providing recommendations on emergency classifications to the Emergency Director.

3.0 PREREQUISITES ,

An off-normal event has occurred, or is in progress.

o %vA W:5' E

QUESTION 9.17 (1.00)

Whct to tho Clocrcnco Palicy uccd whcn piccing a clocecnco en o FAIL OPEN Air-Operated Valve (What must be done to Tag the valve closed)?

ANSWER 8.17 (1.00)

(0.5 wech)

1. HOLD tag the handswitch in the CLOSED position.

2. HOLD tag the handwhael in' the CLOSED position.

REFERENCE  ;

VEGP-00304-C, p. 8, Equipment Clearance and Tagging 194001K102 ..(KA's)

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QUESTION NUMBER: 8.17 FACILITY COMMENT: Answer only addresses a FAIL OPEN A0V with a handwheel.

Recommend credit also be given for clearance policy towards FAIL OPEN A0V without a handwheel as follows.

- HOLD tag the handswitch in the closed position.

- Mechanically or hydraulically gas the valve in the closed position and HOLD tag the gagging device.

REFERENCES:

00304-C, Pg 8, 9. Steps 4.1.4.2. and 4.1.4.3.

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noCEME No t.EvisiCN ogogyo VEGP 00304-C ,La l3'I

. 8 of 42

, m-3.8 PLANT' PERSONNEL It is the responsibility of all plant personnel to adhe're to the requirements of this procedure.

3.9 INDEPENDENT VERIFIER -

i The Operations Department Individual who is responsible for verifying the position of a safety-related component as described on the CLEARANCE SHEET in accordance with the provisions of Procedure 00308-C.

"Independent Verification Policy".

4.0 INSTRUCTIONS 4.1 CLEARANCE PHILOSOPHY 4.1.1 When clearing power supplies to solenoids remove fuses, where practical, instead of links. Fuses shall be bagged and HOLb tagged as part of the CLEARANCE. No more than two fuses per bag.

4.1.2 The scope of each CLEARANCE should be just enough to adequately clear the equipment. This is to reduce interference with other CLEARANCES, and use of PARTIAL RELEASES.

! 4.1.3 When HOLD tagging a Motor Operated Valve (MOV) as a i

fluid boundary, the handswitch shall be HOLD tagged in i

che position in which the valve handwheel will be HOLD tagsed. The breaker shall be opened or off, as applicable, and the handwheel shall be HOLD tagged.

4.1.4 When using Air Operated Valves (A0V) as boundary valves perform the followias:

4.1.4.1 For a FAIL CLOSE A0V

a. HOLD tag the handsvitch in the closed position

b. HOLD tag the air supply valve closed and check that the air line to the valve is depressurized.

c. If the valve has a handwheel, HOLD tag it in the closed position.

4.1.4.2 , For a FAIL OPEN ACV with handwheel

a. HOLD tag the han'dswitch in the closed Josition

b. ~ HOLD tag the handwheel in the closed position v ..,

o nGcEQVat No A.EvisiC N imaggso 3 VEGF 00304-C A4- llT~ 9 of 42 {

4.1.4.3 F.or a FAIL OPEN . AOF without handwheel

a. *ROLD tag the handswitch_in the closed position

b. Mechanically or hydraulically (as appropriate) gag

. - the the gagging valve in the closed position and HOLD tag device.

4.1.5 When restoring Air operated Valves used as boundary valves perform the following 4.1.5.1 For a FAIL CLOSE A0V

a. If the valve has a handwheel, remove the handwheel HOLD tag and restore to the operate or open position.

b. Remove the HOLD tag from and open air supply valve

c. Remove HOLD tag from handswitch and operate as required.

4.1.5.2 For a FAIL OPEN ACV with handwheel

a. Remove HOLD tag from and open air supply valve

b. Remove HOLD ag from handwheel and place handwheel in :he open position, i . 4.1.5.3 For a FAIL OPEN Act.' uithout handwheel

a. Remove HOLD tag from and open air supply valve, t

b. Remove HOLD cog from gagging device and remove gag.

I 4.1.6 The handswitch pesition on HOLD TAGS shall be in the same position as the controlled component, e.g. A handswitch tagged c'esed means the valve is closed. If a differenca exists due to unforeseen circumstances an information cas shall be attached to the handswitch and the CLEARANCE stating conditions. An example may bei i

The motor of a MOV is burned out and manual operation is necessary.

I 4.1.7 HOLD TAUS should be placed in a manner that they will be easily visible to anyone preparing to operate the equipment.

1 13 4%

l

f M I

J QUESTION 1.02 (2.00)

State your how each of the following will answerAssume to INCREASE, affect shutdown margin. Limit separately. EOL. DCPREASE. or NO CHANGE. Consider each case a.

Boronand power concentration no rod motion. is decreased 20 com whileg maintaininconstant

b. Bank main ad'hhighk'isincreasedfrom125 steps ng,\ constant power and boron concentration.to 200 steps wntle

c. Reac

%p. all rods insert. 6 d.

While ehutdown, the RCS is cooled down by 40 degrees .

ANSWER 1.02 (2.00)

a. DECREASE

b. NO CHANGE

c. INCREASE

d. DECREASE REFERENCE VOGTLE LO-LP-33510, Objs. 1, 5 & 6.

3.2/3.4, 3.5/3.7 3.8/3.9 190002K114 192002K113 19:002K110 ..(KA's) 4

QUESTION NUMBER: 1.02c FACILITY COMMENT: Per attached reference, the defined SDM assumes a stuck rod.

Under this assumption SDM prior to and immediately following a reactor trip is the same. Recommend changing answer to be ,

NO CHANGE.  ;

i

REFERENCES:

LO-LP-33510-1, Obj. 1 L

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

3 Ill. LESSON OUTLINE: NOTES I. INTRODUCTION The shutdown margin (SDM) calculation is used to determine the amount of reactivity by which the reactor is or would be 0.Mi,s subcritical.from any given conditions. It is necessary for g.g.J3Q-g l the ';eactor operator to know how to calculate a SDM to operate the reactor safety and to ensure compliance with -

Technical Spa ifications. /,d 7F US'd W

• I' [

les G CM- ' Lule> hh b O II. PRESENTATION A. Shutdown Margin (SDM)

1. Define in Tech Specs:

0h% \

2. Shutdown margia shall be the instantaaews amount' of reactivity by which the reactor is ruberitical or would be subcritical free its preerst condition assuming all red cluster assemblies (shutdows

'and centrol) are fully insett M eacept for the single rod cluster assembly of highest reactivity

' worth which is ase m i to be fully withdreva.

J. Purpose of SDN , .

a. Ensures reactor can be safely made suberitical p N, 3 free operating condition.

b. Reactivity transients during postulated accidents can be acceptably controlled.

c. Reactor remains subcritical when shutdown.

B. Tech Spec Limits (T.S. 3.1.1.1/2)

1. H0DE 1 and 2 (Power ops, startup) 44 ;bI 1

a. SDM skal be greater than or equal to 1.3I delta k/h.

b. Limit based os possible steam line rupture NTC aore negative at i

at EG., so-load Tave. EOL. therefore cooldown will add

2. MODE 3 and,4 (Est standby. Hot Shutdown) acre + reactivity.

a. SDN shall be greater than or equal to teff*

delt Lil. -(14 l .*J 8 SW M ( u e . f. W J l l. M'N

b. >

' Limit more restrictiv*/ 4 f (*k b t#~t.. Qs<., d n 4J Tiq l 4tn a su ,w

$ $l< $fb, tL f'l uUam- [

A5 i[ Mif thy.

e CUESTION 1.10 (1.00)

If tho UNIT 2 rC00 tor werO ts opercto ccntinuouoly ct caprcxtoctolv 100%

power for the entire first fuel cycle, describe how the axial flux peak would move or behave over core lif e as f ollows:

a. State why the axial flux peak is initially located below the core centerline.

b. What causes the axial flux osak to move upward over core life 7 a

ANSWER 1.10 (1.00)

a. The lower core inlet temperature CO.253 in combination with the

negative MTC CO.253 result in more power being generated in the bottom half of the core.

b. Fuel depletion in lower regions of the core.

REFERENCE V0GTLE LO-LP-33520-01, Obj. 3.

2.9/3.1 192005K112 ..(KA's)

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QUESTION NUMBER: 1.10a l

l FACILITY COMMENT: Recommend elecepting answers that discuss or describe operational characteristics of a negative MTC (i.e., better moderation, higher power, etc.). See attached.

REFERENCES:

LO-LP-33520, Obj. 3 L i

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LO11 LO-LP-33520-DO Ill. LESSON OUTLINE: NOTES

8. Axial Power Distribution Objective 3

1. Psax at 40% core height BOL due to

a. T entering the bottom of the cold Reactor, therefore

b. Better moderation

2. EOL

a. Puel in high flux areas begins to burn LO-TP-33520-00-003 o

out

b. Peak flattens and moves up

3. Control rod position effects

a. Would signif f cantly disrupt the smooth Objective 3 flux distribution

b. Cause local power peaks

c. Orsrationally attempt to run at power with all rods out

d. Try to compensate for power changes using dissolved boron

4. Xenon distribution

a. Can cause xenon oscillations  !

ihd t. c.J Ju d.Jls

b. Talk aoout Tenon osc1Tta @ a= 1M C. Radial power distribution Objective 1

1. A core of uniform fuel enrichment would have Objective 3 a pronounced central radial power peak

2. To flatten the radial neutron flux fuel is LO-TP-33520-00-004 radially zoned

a. Use three enrichments

1) 2.1 v/o U-235

2) 2.6 w/o U-235

3) 3.1 v/o U-235

b. High enrichment is used in outer region

c. The two lower enrichments are mixed See Vogtle Text Vol.

checker - board fashion 2. Ch. 4. Section C.

Page 4-82 5

)

QUESTION 4.03 (1.50)

In cecordcnco with tho Stcnding Ordcro, ctoto SIX opcectieno peccccuros thct coy bo pcrformed without having the procedures ir hand.

ANSWER 4.03 (1.50)

CAF i t

REFERENCE VOGTLE Standing Order, # CAF.

V0GTLE 00054-C, "Rules for Performing Procedures," step 4.2.4.1.

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QUESTION NUMBER: 4.03 FACILITY COMMENT: Standing order referenced in question has not been issued or written. Recomend deleting question.

REFERENCES:

Standing Order Book l

- __~ _ . ,

TABLE OF CONTENTS DATE ISSUED 08-05-88

- V0GTLE ELECTRIC GENERATING PLANT OPERATIONS STANDING ORDERS UNIT ONE ORDER NO. DATE ISSUED STANDING ORDER TITLE 1-87-63 12-03-87 REQUIREMENT FOR IST RETESTING 1-87-71 12-29-87 ACCUMULATOR #3 INCREASING LEVEL 1-88-10 02-19-88 CRITICAL COMPONENT 1-88-21 05-09-88 POWER LEVEL MONITORING PROGRAM 1-88-22 06-11-88 FIRE PROTECTION LCO PROGRAM 1-88-23 07-08-88 CONTROL 8 LOG. CHILLED WATER OPERATIONS 1-88-25 07-28-88 MANUAL OPERATION OF MOTOR OPERATED VALVES f

,e,* * ***'

7 ..,

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Table of Contents DATE OF ISSUED 8-4-88 -

VOGTLE ELECTRIC GENERATING PLANT OPERATIONS STANDING ORDERS UNIT 2 ORDER NO DATE ISSUED STANDING ORDER TITLE 2-87-03 08-24-87 UNIT 2 ACTION REQUIRED FOR UNIT 1 PCB RECLOSURE

! 2-87-14 11-20-87 UNIT II EMERGENCY HORN 2-87-16 12-14-87 FIRE PROTECTION P!V INTERFACE 2-87-17 12-14-87 CONTROL OF RWST OUTLET VALVE 2-88-01 01-14-88 TRACKING OF VALVE /8REAKER MISPOSITIONING 2-88-08 04-14-88 UNIT 2 ELECTRICAL BUS AND SUPPORT SYSTEM OUTAGES ,

2-88-09 04-14-88 VALID ANNUNICATOR 10ENTIFICATION PLAN 2-88-10 04-15-88 POTENTIAL CONTAMINATION IN THE SPENT FUEL POOL 2-98-16 05-06-88 PREREQUISITE CONDITIONS REVIEW 2-88-18 05-19-88 VALIO METER 10ENTIFICATION PLAN 2-88-19 05-20-88 ROV!NG WATCH SYSTEM QUALIFICATION 2-88-20 05-24-88 ROVING WATCH PROGRAM 2-88-21 06-02-88 MAINTENANCE WORK ORDER REQUESTS 2-88-22 06-06-88 RCS LEVEL MONITORING 2-8b-23 07-13-88 UNIT II FIRE PROTECTION RESPONSE 2-88-24 07-27-88 ADMINISTRATION OF OPERATING PROCEDURES OURING HOT FUNCTIONAL TESTING 2-88-25 08-04-88 FIELD SHIFT SUPERVISOR DUTIES MNO N011VW80(Ni

__ ____ ~

119 QUESTION NUMBER: +rO9/5.13 STATEMENT: Answer lists generic causes of water hemmer. Allow credit for specific examples of operations that could cause water hammer.

REFERENCES:

N/A

r QUESTION NUMBER: 6.11n STATDfENT: Coincidence with causing a full faed water isolation, the SIS

& P-14 signals also result in a turbine trip. Some students may include the main turbine on the list of components tripped. Recommend that this be included as an optional response.

REFERENCES:

1X5DN203-1 i

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QUESTION NUMBER: 6.13 STATEMENT: Scme students may include the S/G blowdown isolations and S/G sample valves in their answers since they are actuated from a MDAFW actuation signal. Recommend these be included as optional responses.

REFERENCES:

Figure 13d-3

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QUESTION NUMBER: 7.09 a STATEMENT: Answer a. should be changed tc greater than 24*F.

REFERENCES:

19001-C. Attachment D. Rev. 5.. Pg 16

s e l

PRoCEcumt No LEvlSioN pActNo VEGP 19001-C $ 16 of 17 Sheet 1 of 1 ATTACHMENT.D VERIFICATION OF NATURAL CIRCULATION The following conditions support or indicate natural circulation flows e RC8'aubcooling monitor indication - GREATER THAN 24'F.',

e SG pressures - STABLE OR LOWERING.

e RCS hot leg temperatures - STABLE OR LOWERING.

e Core exit thermcouples - STABLE OR LOWERING.

e RCS cold leg temperatures - AT SATURATION TEMPERATURE FOR SG PRESSURE.

i END OF ATTACHMENT D J

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QUESTION NUMBER: 7.15 STATEMENT: Answer should be expanded to include:

- At the end of each calendar year.

- Not used for 6 days.

REFERENCES:

00930-C Steps 5.2.12.2 and 5.2.12.6. Pg 14 L

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Nxstt e: No. navisios n os s o-93 ,

< t 5.2.11.7 Anindividuaika)naybeaddedasnecessarytocompleta the work, to an existing specific RWF by the job supervisor after verif ing the individual's remaining exposure limit and wit the concurrence of NF 4

j supervision.

5.2.12 Termination Of, Radiation Work Fermits 5.2.12.1 Specific RWFs shall be terminated or suspended whenever the ob is completed or cancelled, or whenever sign ficant changes in radiological conditions occur.

5.3.12.2 General RWPs shall be terminated *whenever significant changes in radiological conditions occur or at'the end

• I of each' calendar year. General RWFs shall be reviewed' i by :Health Physics supervision each calendar quarter.

s 5.2.12.3 HP should ensure that terminated RWFs are clearly

) identified es terminated and inform the Shift '

Supervisor of RWFs to be terminated.

! 5.2.12.4

! HP should also ensure that terminated RWFs are removed from the job location and/or control points and *

[

returned to the Health FWfsics office j 5.2.12.5 Specific RWFs that are not used for 3 working days

! should be auspended. The RWF can be reactivated upon j

RF notification by the work party supervisor or Work i

Flanning Group.

5.2.12.6 Any RNF that has not been used for 6 days shall be-j terminated unless RF has been apprised of the need for i sus j

WorgendingtheRWFbytheworkpartysupervisororthe Planning Group. ,

5.3 ColrrAMINATION CONTROLS l

4 5.3.1 Use of Protective Clothing (PC) 1 5.3.1.1 Fersonnel shall wear and use protective cicthing in a manner consistent with the training each has received.

Dressing / undressing guidelines are given in Table 1.

i 1 5.3.1.2 HP shall ensure that PCs are commensurate with the l 1evels and state of the contamination expected for the

( area entered. Suggested protection levels are shown in l Table 2.

I 5.3.1.3 Outer screet clothing and jewelry should be removed l prior to donning PCs. Violations of this policy may result in the confiscation of contaminated clothing and personal effects.

I mu.s

r---_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

1 l

l I

QUESTION NUMBER: 8.18 l t l

l STATEMENT: Question does not ask what the O ENTRY" margin is, but what l limitations are placed on an individual when he/she has reached this margin. Recommend full credit fort "the individual should NOT be allowed to enter a RCA".

i I

1

REFERENCES:

Exam key ,

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i QUESTION NUMBERI 8.20  !

i STATINENT: Part c. In accordance with 13105 a sample must be taken within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after an indicated level change of 7% (67  !

gals) or more which corresponds to a 1% volume change as l required by surveillance requirements 4.5.1.b. Recommend ,

that ansvar c. be modified to meet the abeve condition. '

REFERDicES : 13105-1, Rev. 4 TS 3/4.5.1 b

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juvisios nca no VI( P 13105-1 >

4 l 4 of 30

.- l 4.2.1.6 I When'the desired _ Accumulator level is reached CLOSE the fill valve opened in Step 4.2.1.5. Independent verification required.

4.2.1.7 REPEAT Steps 4.2.1.5 and 4.2.1.6 if required to adjust levels in additional Accumulators.

4.2.1.8 CLOSE l-HV-8871. Independent verification required 4.2.1.9 CLOSE 1 HV-8888. Independent verification required i 4.2.1.10 STOP the running SI Pump and ENSURE its control switch is in AUTO: Independect verification required. ';

SI Puap A 1-HS-0998A '

SI Pump 5 1-HS-0999A a

NOTE

'The,horon concentration of ant ,gg  !

Accumulator -imust be verified "

, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after'an indicateda lay 31 ' rise of 7!' (6.7 gallons) or, d ""**

. mot e per Surveillanc,e , ,

Requirement 4.3.'i;7b(

4.2.1.11 NOTIFY Chemistry to sample the Accuriulator as renuired and REPORT the boron concentration of each one sampled. i 4.2.1.12 If requirati, AbJU,1T Accumulator pressure (s) per Sub-subcection 4.2.3 of this procedure. -

4.2.1.13 u

-i LOG Accumula tor boron concentration sample results in [

the Unit Ceutrol Log vnen reported by Chemistry.

4.2.2 Pressurizing The Accumulators 4.2.2.1 OPEN Nitrogen Supply Header Isolation 1+2402-U4-016.  !

CAUTION t

Accumulator Nitro ea supp y/ vent (

I valves (1-HV-8875 -H) wil nold 1 a pressure on the Accurulator side only. Therefore, when i  :

1-HV-8380 is opened, all  ;

Accumulators are pressurized 1 t

simultaneously.

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3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS LIMITIt:G CONDITION FOR OPERATION 3.5.1 Each Reactor Coolant System (RCS) accumulator shall be OPERABLE with:

a. The isolation valve open,

b. A contained borated water volume of between 6616 (36% of instrument span) and 6854 (64% of instrument span)' gallons (LI-0950, LI-0951, o LI-0952, LI-0953, LI-0954, LI-0955, LI-0956, LI-0957),

c. A boron concantration of between 1900 and 2100 ppm, and

d. A nitrogen cover pressure of between 617 and 678 psig. (PI-0960A&B, PI-0961A&B, PI-0962A&B, PI-0963A&B, PI-0964A&B, PI-0965A&B, PI-0966A&B, PI-0967A&B)

APPLICABILITY: MODES 1, 2, and 3*.

ACTION:

a. With one accumut' tor inoperable, except as a result of a closed isolation valve restore the inoperable accumulator to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b. With one accumulator inoperable due to the isolation valve being closed, either imediately open the isolation valve or be in at least HOT STANDBY within i hours and reduce pressuMzer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

1 SURVEILLANCE REQUIREMENTS 4.5.1.1 Each accumulator shall be demonstrated OPERABLE:

a. At least or.ce pe- 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by:

l 1) Verifying c.ontained borated water volume and nitrogen l ,

cover prea the tanks, and I

2) Verifying $nat each accumulator isolation valve is open.

(HV-8808A,B,C,0) l

• Pressurizer pressure above 1000 psig.

l V0GTLE - UNIT 1 3/4 5-1

t

=

u EMERGENCY CORE COOLING SYSTEMS r

SURVEILLANCE REQUIREMENTS (Continued)

b. i At least once per 31 days and within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after each solutions volume increase of greater than or equal to 1% of tank volume c' (67 gallons) by verifying the boron concentration of the accumulator-solution;'and ' I f

c. At least once per 31 days when the RCS pressure is above 1000 psig by verifying that the circuit breaker supplying power to the isolation valve operator is open.

4.5.1.2 Each accumulator water level and pressure channel shall be demon-strated OPERABLE at least once per 18 months by the performance of . CHANNEL CALIBRATION.

e eh k

V0GTLE - UNIT 1 3/4 5-2

QUESTI0h NUMBER: 1.22c STATEMENT: Steam flow is not used in Vogtle calorimetric calculation.

Recotamend delete c. from question and reassign point values co a., b., and d.

REFERENCES:

14030 L

P0oCEDURE No. LEWSloN PAGE No.

VEGP 14030-1 10 4 of 13 Sheet 1 of 7 DATA SHEET 1 LIANUAL CALORIMETRIC NOTE The plant computer should be used when available for obtaining the following datas when not available, indications in ( ) should bs used.

1.0 RECORD the following data:

1.1 Time Date 1.2 Average Steam Pressure (See Note 1)

SG 001 F0400A (PI-514) PSIG PO401A (PI-515) PSIG P0402A (PI-516) PStG SG 002 PO420A (PI-524) PSIG P0421A (PI-525) PSIG P0422A ( P'I-526 ) PSIG SG 003 P0440A (PI-534) PSIG PO441A (PI-535) PSIG P0442A (PI-536) PSIG SG 004 P0460A (PI-544) PSIG PO461A (PI-545) PSIG PO462A (PI-546) PSIG Total = PSIG Average Steam Pressure = PSIG NOTE 1: If a data point is not acceptable (e.g., value abnormally high or low, indicator bad, total loss of indication), enter "Bad" for the value. Calculate the "Total" based on the remaining values and obtain the "Average" by dividing the "Total" by the number of remaining values.

,u,

ProCEDU AE No. REVIS!oN . PAGE Wo.

VEGP 14030-1 10 5 of 13 Sheet 2 of 7

~

SHEET 1 1.3 Average FW ' arature (See Note 1)

SG 001 TL _ 15208)

• F SG 002 T04> II-15209) 'F SG 003 T0458A (TI-15210) __,

'F SG 004 T0478A (TI-15211) 'F '

Total = 'F Average FW Inlet Temperature = ____ 'F 1.4 Total FW Flow (See Note 2)

SG 001 F0403A (FI-510A) MPPH ,.

F0404A (FI-511A) MPPH {

SG 002 F0423A (FI-520A) MPPH F0424A (FI-521A) MPPH SG 003 F0443A (FI-530A) MPPH F0444A (FI-531A) MPPH SG 004 F0463A (FI-540A) MPPH F0464A (FI-541A) MPPH Sum = ,MPPH Total FW Flow = Sum /2 = MPPH NOTE 1: If a data point is not acceptable (e.6., value abnormally high or low, indicator bad, total loss of indication), enter "Bad" for the value. Calculate the "Total" based on the remaining values and obtain the "Average" by dividing the "Total" by the number of remaining valuee.

NOTE 2: If a data point is not acceptable (e.g., value abnormally high or low, indicator bad, total loss of indication), record the indicated value of its redundant data point for that loop in its place.

m,

7_ 9 PQoC EDUEE No. EEVISloN PAGE Wo.

VEGP 14030-1 10 6 of 13 r

Sheet 3 of 7 DATA SHEET 1 1.5 Blowdown Flow (Enter zero if secured) (See Note 3)

SG 001 F0407A (FI-ll71B) GPM SG 002 F042'7A (FI-1172B) GPM SG 003 F0447A (FI-ll733) GPM SG 004 F0467A (FI-1174B) GPM Total = GPM 1.6 Indicated Power (A Drawer)

N41 7 N42 ,%

N43  %

N44  %

1.7 PROTEUS Calorimetric Power (Enter NA if not available)

Ulll8 MWT l

t I

NOTE 3: If blowdown indication is lost, unter 0 gpm for the l applicable loop. This is the default value used in U1118.

. 184%

7 PROCEDURE NO. WEVISiOPS WQGE NO.

VEGP 14030-1 10 7 of 13 Sheet 4 of 7 DATA SHEET 1 2.0 Determine the following steam table data:

2.1 Ave rage Steam Enthalpy = BTU /lba (Use Average Steam Pressure +14.7 and Saturated Steam' Tables, quality factor of 1) 2.2 Average F'd Enthalpy = BTU /lba (Use Average FW Tempersttre and Average Steam Pressure

+14.7 and Subcooled Tables) 2.3 Average Blowdown Enthalpy = BTU /lba (Use Average Steam Pressure +14.7 and Saturated Liquid Enthalpy at this pressure, quality factor of 0) 2.4 Average Feedwater Specific Volume = fc 3/lba (Use Average W Temperature and 1000 psia and Subcooled Tables from PTDB.)

3.0 Determine the temperature corrected FW Flows

,a 2 Correction Factor = -5 p 0 917 3.1 1 + 0.98 x 10 - 440 ,

p

- ~. 2.4

. .- 2 1 + 0.98 x 10"$ 0.01917

=<

- 440 ,

= (5 decimal places)

P 3.2 Corrected FW Flow = x P

=

x_ _

= neFu l 1

l 444 %

~ u. g PROCEDCE NO. .

KEVISION PAGE NO.

VEGP 14030-1 10 8 of 13

~

Sheet 5 of 7 DATA SHEET 1 4.0 Determine the total reactor thermal powers n ,

4.1 9" (Step.3.2h* [ Btu /lba / Step 2.1}~[ Step 2.2%

MPPH / \ ( 5tu/lba /-

~

(Step [ Step 2.1) ~ [ Step 2.3$

• 495.12 MPPtf spa 1.5\*

~

/ ,\3tu/lba / (Btu /lba/, 10 6 gpa 55.29* MBTU/Hr

= ( )x -

( )x -

x d21'.1.2, D

10

- 55.29

= -

- 55.29

=

MBTU/Hr

• Net RCS heat input - losses determined during post core load hot functional testing.

l l

r I

i l

QUESTION NUMBER: 2.02(2)

STATEMENT: SI pumps should be as follows:

850 gpm (425 each) 0 1150 psig. The pressure given in the key (1750 psig) is the design maximuni pump pressure. Pump head at design flow is approx. 1150 psig (2680 ft.).

REFERENCES:

LO-LP-13201-01 l

- ~ .

LO-LP-13201-01 Ill. LESSON OUTLINE: . NOTES

c. Design Pressure--1750 psig

d. Design Temp--300'F

e. Design Flow--425 spat

f. DesignHead--7680'ftl LO-TP-13201-00-004

g. Maximum Flow--650 gpm

h. Design head at max flow--1650 ft

i. Design shutoff head--1625 psig ,

j. Required NPSH at any flow--less than 15 ft

k. Motor has two motor coolers (30 gpm each NSCW)

1. Pump bearings have oil cooler (10 gpm NSCW)

.. and will run only 40 seconds without NSCW flow. Pump has eleven impellors and a mechanical seal.

m. Each pump is 100% capacity for ESF analysis. '

n. Handswitches on QMCB-A and respective shutdown panel.

1) Give alarm on SSMP when in Pull-to-Lock.

2. RWST--covered in LO-LP-13001

3. HV-8806 RWST to SIP Suction Isolation Valve

a. QMCB/ Lockout-A

b. 480 V MCC 1 BBD

c. Out of position alarm (SSMP. Group I:on. Light)

4. HV-8923A SIP-A Suction Isolation Valve

a. QMCB-A

b. 480 V MCC 1 ABD

c. Out of position alarm (SSMP, Cp. Mon.)

\

6

F QUESTION NUMBER: 2.03 STATEMENT:

h RMWST is alternate makeup C'4ST is normal maksup Change answer to reflect these sources as stated above.

REFERENCES:

LO-LP-25102-02-c, Obj. 9

--147-19-1 r-Pe-lh-4ter 4 . 4. 4 -

,- e r

co . w . ts , a 1. a -c Ill. LESSDN OUTLINE: NOTES i

5) Temperature a) SFP -

• ~ ~ " ~ ~

b) ,SFP HX outlets

6) Flow out of backflushable filter

• 1.... . . 4h '. . M .fr* uAv .. . . . -

I A

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v v M 9:_vh v

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c;r n o m v y HJ .

i t

v. t. < ~< .

v i n ,. . A M ^ ds v

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t A 'A -

bA u

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O M4 h Si

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e.

~ - s. ui a t . . z - c LESS.ON OUTLINE:

111.

NOTES euk M 1 a n - .t 'cn .

o.. -

u

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a, _~

vn c-jdesdirs % u b .

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O. -

QUESTION NUMBER: 2.04 STATEMENT: Also accept as correct answer "to limit AFW flow to a depressurized SG".

REFERENCES:

FSAR 10.4.9.1.1

a VEGP-FSAR-10 effects of hydraulic instability (water hammer). The AFW is also designed to permit testing of the pumps during normal plant operation.

'The AFW eedwater auxiliary- is['provided flowwith-flowylimiting~

to i a depres'surizedorifices"to111mit'the  : steam generator f and

~ ~

to ensure that"the"required"minimum-flow'is directed to the

remaining three intact steam generators.

10.4.9.1.2 Power Generation Design Basis The AEW supplies feedwater to the steam generators at a sufficient flowrate to support normal low-power transients auch  !

as startup, cooldown, and hot standby.

10.4.9.1.3 Codes and Standards 1

Codes and standards applicable to the AFW are listed in ,

table 3.2.2-1. The AFW is designed and constructed in accordance with Safety Class 3 requirements up to the motor-operated valves in each branch line located upstream of

the stop-check valves, which serve as the containment isolation valves; the remaining valves and piping to the main feedwater system are Safety Class 2.

10.4.9.2 System Description 10.4.9.2.1 General Description 7 a

j The system consists of two motor-driven pumps, one steam turbine-driven pump, and associated piping, valves, instru- .

ments, and controls as shown on figures 10.4.9-1 and 10.4.1-1 (sheet 3). Table 10.4.9-1 provides data for the various active

components in the AFW. Table 10.4.9-2 presents minimum system flow requirements for various modes of operation. A plan view of the AFW pumphouse is provided in figure 1.2.2-27.

Each motor-driven auxiliary feedwater pump is sized to supply

, the feedwater flow required for removal of 100 percent of the  !

j decay heat from the reactor. The turbine-driven pump is sized ,

to supply up to twice the capacity of a motor-driven pump. The system capacity is sufficient to remove decay heat and to provide adequate feedwater for ccaldown of the reactor coolant system (RCS) at 50'F/h. Following transients or accidents, cooldown may begin w.4 thin 1 to 2 h at zero load hot standby. ,

l For a transient or accident condition, the minimum flow is  ;

I j 10.4.9-2'

~ - - - - ,,m _e., , , _ . _ _ _ _~,-c,__,., . _ _ _ _ , . , - , . - _ _ , _ _ , _ . , , - , _ _ _ _ , _ _ __

,,.____,_m ,

-g4 -

QUESTION NUMBER: 2.05b STATEMENT: . Accept additional corract answers as stateo in AFW SOP.

- AFW piping hot to the touch.

- AFW pump starts but fails to develop adequate discharge pressure.

REFERENCES:

13610-1 Pg 11, Step 4.4.4.1

~

PRoCEDU"E NO. REVISloN PAGE No.

VEGP 13610-1 7 11 of 23 4.4.3.6 OPEN the MS SPLY 70 AUX FW TD PMP 1 valves 1-HV-3009 and 1-HV-3019. Independent Verification Required. .

4.4.4 Response To High Auxiliary Feedwater Pump Discharge Temperatures NOTE Backleakage of feedwater into the Auxiliary Feedwater System can disable the AFW pumps due to steam or vapor binding.

If backleakage through a check valve continues after attempting to reseat it then the affected pump should be inspected frequently for steam binding and ov.trheating, and operated as necessary to prevent it from overheating.

.4.4.4.1 Check'valv'e backleakage may be Nantified by'sny of thei following symptoms: '

i

a. ~ Auxiliary Feedwater Piping. hot 'to the touch,'

b. Auxiliary Feadwaterl Pump ~ discharge temperature.

greater than 200*F as shown on the ERF'computeri terminal,3 AW PUMP TEMP ELEMENT COMPUTER POINT TDAFW TE-5166 T9887 MDAFW B TE-5211 T9888 MDATW A TE-5210 T9889

c. Auxiliary. Feedwater Vincreases following ' pump' pipingshutdown; temperature. rapidly

d. Auxiliary ~feedwater pump starts but'failsLtor

~

develop adequate discharge pressure.,

4.4.4.2 If an Auxiliary Feedwater Pump is vapor bound COOL the pump as follows:

a. CLOSE the discharge valves to the Steam Generators for the affected pump, i

i I

m -#

QUESTION NUMBER: 3.04a STATDIENT: The upper limit for a thermocouple is 2300 (page 11).

However, the upper limit of indication for a thermocouple (asked for in questien) is 2200'F (page 13). Accept either answer.

REFERENCES:

LO-LP-36102-00-c, Pg 11 6 13

LO-LP-36102-00-C lil. LESSON OUTLINE: NOTES CONDITION IV: SIGNIPICANT CORE DAMAGE SUCH THAT DETECTOR INSERTION IS HINDERED A. Assumptions

1. Core shutdown - all rods may or may not be fully inserted.

2. Core damage so severe as to allow only partial or ~

no insertion of incore detectors.

a. Most likely to occur in upper, central regions

b. Amount of insertion available will give operator a general idea of amount of core damage.

B. Using MIDS to Locate Damage

1. Insert detector until stopped by damage and reading detector position to build 3-D map.

2. Picoammeter not needed for this core condition.

III. CORE EXIT THERMOCOUPLES A. General System Description

1. 50 chromel-alumel thermocouples used to measure LO-TP-36102-00-C-014 exit temperature LO-TP-36102-00-C-015

2. Located on upper core plate at various locations. LO-TP-36102-00-C-016

3. Clad in stainless steel shesithed with aluminum oxide insulation.

4. Thermocouple characteristics

a. T/C output linear with respect to temperature i

b. Accuracy

1) 12'F from 0 to 530'F

2) 13/8'T from 530 to 700'F Upper limit - +2300'F 3) 11

i LO-LP-36102-00-C  !

lll. EESSON OUTLINE: NOTES

b. g updated monthly and is based on full flow ,

conditions. '

i NOTE: Keep in mind that under accident conditions, F calculations generated by computer using mIIknlYactorsarainvalid.,

Under post-accident conditions, the real parameter of concern is basically the thermocouple reading -

itself.

C. Thermocouple Indications on Plant Computer

1. Computer programmed for inadequate cooling

• Presently Plant evaluations

• Vogtle SPDS system uses all T/C for 5 T/C with range to 2200*F '

~

a. monitoring ICC.

b. Remaining TC range to 1200*F.

2. Computer corrects all thermocouple readings for reference temp. variations.

a. 210*F variations

b. Accuracy 25% between 750'F and 2200'F D. Post Accident Monitoring

1. Computer inoperable

~

a. Expanded readings can be taken at the in-core display panel in control room

b. Reference temperature readings may be taken locally at Remote Processing Units (RPU) A3 and B3.

i 1) In case of off normal containment conditions.

l l

i 2) Tables will be provided for conversions LO-TP-36102-00-C-017

3) Readings taken with millivolt potentio-l a ter 13

~

p QUESTION NUMBER: 3.09(2)

-O STATEMENT: P-6 setpoint is 10 amps. Change answer to reflect correct setpoint.

REFERENCES:

LO-LP-28102-01

e LO-LP-28102-01 Ill. LESSON OUTLINE: NOTES b) Will block automatic reinitiation of SI once reset, until P-4 signal is removed.

5) Arms Steam Dumps if C-9 is present (TR A)

6) Shif t's steam dumps f rom load rejection to plant trip controller (TR B)

2. P-6,'SourceRange.'BlockPermissive),

s. '5etpoint s . 1 1 10-10 ICAon.any'1/2IRNIS[

detectors (NR-35,.NR-36)

b. Resets at 5 x 10 ~1I

c. Function

1) Allows operator to manually block SR High Flux trip and de-energize SR High Voltage (both TR A and TR B switches, QMCB)

2) Loss of P-6 (either train) will automa-tically unblock trip

d. Permissive Status Light on BPLP (QMCB-C) comes on when P-6 is present

3. P-7, at power trip permissive

a. Setpoints - either of the following

1) greater than 10% turbine power (p-13) or h coincidence

2) greater than 10% power on 2/4 PR NIS (P-10)

b. Functions unblocks at power trips

1) Pressurizer Low Pressure

2) Pressurizer Hi Level

3) RCP Undervoltage

4) RCP Underfrequency

5) Two loop loss of flow trip

c. Permissive Status Light on BPLP goes out when P-7 is present 5

i

?

QUESTION NUMBER: 3.21 STATEMENT: The high vibration function ja actually an alarm function.

only and applies to Units 1 & 2. The change that was implemented on Unit 2 was to place a'30 see time delay on the  :

vibration alarm on fan startup.- By substitsting alarm for trip in questions 3.21 (a) & (c) they remain valid. Since the alarm applies to Units 1 & 2 recommend that 3.21 (b) be t deleted and point values be adjusted on.3.21 (a) & (c).

REFERENCES:

1X3D-BD-K03A  :

2X3D-BD-K03A  ;

IX4DB133-1 2X4DB133-1  ;

RQ-LP-60301-00-c 1

5

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c. Haste Monitor Tank Capacity Unit 2 will have 2 additional 20,000 gal tanks designated as common,

1. both units have 2 10,000 gal tanks

2. pinch points of waste water system

3. Add 2 20,000 gal tanks

a. 2 new pumps b, d lvl unit 2 southwest corner

c. cost > 800k 8

4. increase surge capacity

5. being built now SW corner D lvl unit 2

d. NSCW System - Unit 2 Changes

1. NSCW Tower Fan Vibration'Roset)

Unit 2 will'have~a 30'sec time delay'on the vibration trip for the NSCW tower-fans. "

a.~30 see time'delayi

b. local reset .

c. reset at breaker 1

2. NSCH Aux Containment Cooler Isolation Valves Unit 2 aux ennt cooler isolation valves are interlocked so that the outlet valves must be opened first. When opening the outlet valves they will stroke open for 3 sec, wait for 60 sco, and then finish stroking open. This allows the header to backfill before allowing flow through the system.

a. outlet opens first

b. open 3 sec, wait 60 sec, finish stroke

c. then open inlets

d. reduces waterhammer

1. outlet header lower pressure

2. slow backfill during 60 see wait

3. after test may install on unit 1 HMW. 6/20/88 PAGE 7 2

QUESTION NLHBER: 4.02 STATEMENT Although RQ-H0-60301-00-001 specifically states rectangular for Unit 1, the general concept is square vs circular shapes. '-

Accept either rectangular or square for Unit 1.  !

REFERENCES:

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QUESTION NUMBER: 4.06(2) i l

STATEMENT: The 4500 mrem /yr administrative Ifmit is not achievable unless increases in radiation exposure have been authorized (the quarterly limit is 1000 mrem). Recommend deleting part 2 from the answer key and re-distributing points within the question.

REFERENCES:

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l PL.OCECULE NO REVl510N PAGE NO VEGP 00920-C 4 8 of 10 TABLE I RADIATION EXPOSURE ADMINISTRATIVE LIMITS J

(REF: Frocedure 009ZO-G)

VISITOR (4.19) 100 mrea/qtr Whole Body j PREGNANT 400 ares /yr Whole Body i NO RCA ENTRY RCZ VISITOR (> 19) 300 area /qtr Whole Body E)3REMITY = Hands.

RCA VISITOR (> 19) 1200 mrea/yr Whole Body Forearms (Elbow to PARALLEL BADGED WORKER 7500 area /qtr Skin Wrist). Lower Less NRC 18750 area /qtr Extremity (Knee to Ankle) 10 MPC/7 days Inhalation Ankles. Teet.

130 MPC/otr Inhalation P RENATAL 500 mrea/qtr Whole Body WHOLE BODY = Head (NOT PREGNANT) 2000 mrea/yr Whole Body trunk. Upper Leg y 7500 area /qtr Skin to knee. Upper Arm s 18750 mrea/qtr Extremity to Elbow. Active d' 10 MPC/7 days Inhalation Organs. Gonads.

130 MPC/qtr Inhalation e

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THESE LIMITS CANNOT BE EXTENDED - THE INDIVIDUAL MUST BE ON RADIATION WORKER LIMITS TO RECEIVE MORE EXPOSURE.

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RADIATION WORKER 1000 area /q:r Whole Body MARGIN = (Quarter i

4500 area /yr~ Whole Body WB Limit) - (WB 7500 mres/qtr Skin Current Exposure) 18750 ares /qtr Extramity 2

' 10 MPC/7 days Inhalation "NO ENTRY" MARGIN 130 MPC/qtr Inhalation is 200 area -

...... ......... - ~ ..... _.. - .. .......- Manual RVP S1gn IT PRIOR SITE EXPOSURE IS DOCUMENTED. THESE WHOLE BODY In Required - HP LIMITS MAY BE EXTENDED TO 3000 MRD(/QTR OR 5000 MREM /YR Lab Supr nuet i

USING TORM 00920-1 OR FORM 00920-2. Authorize Entry.

ALL OTHERS NO OCCUPATIONAL EXPOSURE INCREASE WHOLE BODY (WB) ADMINISTRATIVE LIMITS

Individual must be on Rad Worker Limite (1000 ares /qtr. etc.)

UP TO 1500 area /qtr & 4500 ares /yr Auth: HP Toreaan/ Worker's Forenan UP TO 2000 mrea/qtr & 4500 area /yr Auth: HP Lab Supv/ Worker's Supv UP TO 2500 area /qtr & 4500 ares /yr Auth: HP Supt / Worker's Supt

To exceed 2500 mren/qtr or 4500 ares /yr. TLD aust be read and evaluated

UP TO 3000 area /qtr & 4500 aren/yr Auth

HP Supt /Ceneral Manager l 4500 area /yr & 4999 ares /yr Auth: HP Supt / General Manager m.,

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QUESTION NUMBER: 4.09d STATEMENT: Answer should be 6.

REFERENCES:

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REVI$loN PAGE No PAccEDURE No VEGP 12004-C 10 13 of 28 INITIALS 4.1.32 'At approximately 30% (348'NWe') turbine power the' station electrical loads should be transferred from the Reserve Auxiliary Transformers to the Unit Auxiliary Transformers,

a. TRANSFER the 13.8kV busses per 13420, "13.8kV AC Electrical Distribution System",

b. TkANSFER the 4160V AC bus 1NA05 per 13425, "4160V AC Non IE Electrical Distribution System". ,

c. TRANSFER MFPT Speed Control to AUTO.

(1) ENSURE MF?T Master Speed Controller SIC-509A is in AUTO, (2) ADJUST MFPT A(B) Speed '

1 Controller SIC-509B(509C) to have the same output ,

value as the Master Speed l Controller SIC-509A,  !

PLACE MTPT A(B) S  !

(3)

ControllerSIC-50B(509C) heed i

in AUTO. j 4.1.33 Prior to reaching 50% power (580 MWe),

PERFORM the following:

a. ENSURE both circulating water pgs running and if necessary START the second circulating water pump per 13724, I "Circulating Water System".

b. START a second Condensate Pump and PLACE the third in standby,

c. START the second Main Feed Pump per 13615, "Condensate And Feedwater

! System",

d. RAISE Turbine-Generator gas pressure to between 70 psig and 75 ps 13810. "Generator Gas System,1,3 per w

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QUESTION NUMBER: 4.22b STATEMENT: Clarification given during exam stating flowrate (gpm) was not required. Recommend deleting point value from flowrate and r9 assigning.

REFERENCES:

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ENCLOSURE 4 SIMULATION FACILITY FIDEllTY REPORT Facility Licensee: Georgia Power Company Facility Licensee Docket No.: 50-424 and 50-425 Facility Licensee No.: NFP-68 Operating Tests administered at: Vogtle Electric Generating Plant Training Simulator Operating Tests Given On: September 20-22, 1988 During the conduct of the simulator portion of the operating tests identified above, the following apparent performance and/or human factors discrepancies were observed and discussed with the facility's staff instructors.

1. During the conduct of a faulted #3 steam generator (main steam line rupture inside containment) event. #1 steam generator's response in both pressure and level decrease was significantly more rapid than that of #2 and #4 steam generators.

2. Indication (meter) of NI reactor power was inconsistent with the actual power.

3. At stcady state operation, turbine load indicatinn varied significantly from load demand indication.

4. The simulator exhibitted a loss of its "ringback" function.

5. During the conduct of a failure of P1 455 (pressurizer pressure transmitter) high, the associated PORY failed open and was overridden to remain open, however, the R0 took the switch to "close" and it closed.

6. Several spurious Train A SSMP alarms were exhibited.