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Exam Rept 50-244/87-22OL During Week of 871005.Exam Results: Six Senior Reactor Operators & Four Reactor Operators Passed All Exams & One Reactor Operator Failed Written Exam
	ML20147C915
Person / Time
Site:	Ginna
Issue date:	11/24/1987
From:	Keller R, Temps R; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
Shared Package
ML20147C864	List: IR 05000244/1987022; ML20147C860;
References
	50-244-87-22OL, NUDOCS 8803030192
	Download: ML20147C915 (200)
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{{#Wiki_filter:, . - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . U.S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REFORT NO.

50-244/87-22 (OL) FACILITY DOCKET NO.

50-244 FACILITY LICENSE NO.

OPR-18 LICENSEE: Rochester Gas and Electric Company 89 East Avenue Rochester, New York 14649-0001 FACILITY: Ginna Nuclear Power Plant EXAMINATION DATES: October 5-8, 1987 CHIEF EXAMINEP: hdMN 8'/4 Yh7 h_R.R. Temps,OperationsEngineer Oa'te APPROVED BY: MM h /////f7 /R.N/)(eller, C/nef, PWR Section ' 'Da't e OpeHitions Branch Division of Reactor Safety SUMMARY: Written examinations and operating tests were administered to six senior reactor operator (SRO) and five reactor operator (RO) candidates. Six SR0s and four R0s passed these examinations. One R0 failed the written examination.

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7_ _ _ . _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ DETAILS TYPE OF EXAMINATIONS: Replacement EXAMINATION RESULTS: I RO l SRO l l Pass / Fail l Pass / Fail l l l l

1 I I l Written l 4/1

6/0 I I i i I I I I I l Operating l 5/0 l 6/0 l l l l l

1 I I l Overall l 4/1 l 6/0 l l l

I I I I I 1.

CHIEF EXAMINER AT SITE: R. Temps 2.

OTHER EXAMINERS: D. Silk I. Kingsley (Sonalysts) T. Guilfoil (Sonalysts) 3.

The following is a summary of generic strengths or deficiencies noted on operating tests.

This information is being provided to aid the licensee in upgrading license and requalification training programs. No licensee response is required.

STRENGTHS a.

The candidat s were trained together as teams for the simulator examination. These same teams were then examined on the simulator by the NRC. The benefits of this team approach to training were apparent in the manner by which the candidates conducted them-selves during the simulator scenarios. Malfunctions and casualties were handled in a calm and collected manr;er with team members usually correcting each other when mistakes or misdiagnoses were made.

OFFICIAL RECORD COPY OL EXAM GINNA - 0004.0.0 11/06/87 .

. _ - _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ ._ -3-DEFICIENCIES a.

Most candidates did not follow procedure 0-5.2 properly during routine power changes on the simulator. Specifically, boration/ dilution calculations were not performed and load changes using the turbine were commenced before control rods responded to the dilution /boration in effect, which is contrary to step 5.3 of procedure 0-5.2.

b.

When entering the controlled area, most candidates had to be reminded by their examiner to check the radiation / contamination status boards prior to entry into the auxiliary building, c.

Most SRO candidates were deficient in knowledge of fuel handling operations, procedures and casualties. This weakness is attributed to the licensee's training department for failing to provide adequate training in this area.

d.

Student lesson plan SLPRRT08CH states that procedure 0-1.2.2 may be used to calculate Shutdown Margin (SDM). None of the SR0 can- ' didates were able to use this procedure to calculate a SDM. This

deficiency is attributed to the licensee as procedure 0-1.2.2 in its present form contains insufficient information for i determination of SDM. It should be noted that most of the candidates arrived at the conclusion that 0-1.2.2 was deficient for purposes of determining SOM.

! 4.

The following is a summary of generic deficiencies noted from the , grading of the written examinations.

This information is being provided to aid the licensee in upgrading license and requalification i i training programs. No licensee response is required, j DEFICIENCIES NOTED ON R0 EXAMINATION: (by question number) i 2.02 Effect of loss of control air on various control valves.

i i 2.08c Full explanation of the effects of a pressurizer level I control channel failing low on various systems, l specifically that isolation of letdown occurs.

' I 3.10c The fact that loss of 125 VDC Bus B will result in a - reactor trip.

l DEFICIENCIES NOTED FROM SR0 EXAMINATION: (by question number) ,

i l , l 5.15 The effect that changes in various parameters have on - , Shutdown Margin.

l t i , OFFICIAL RECORD COPY OL EXAM GINNA - 0005.0.0 11/03/87

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_ _ _ _ _ _ - _ _ - _ _ - .__- __ ____ - _ __. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. __ . _ _ - _ _ - _ _ _ _. ___ _ _-_ i-4- ~l . 5.17 Factors effecting axial and radial flux distribution.

-l 6.11 Effect of first_ stage impulse ~ pressure instrument failure on the EHC system and other control systems.

I 7.11 Control of contair, ment depressurization valves.

j

8.02 Administrative requirements for reactor post 1 trip review.

8.11 Administration of temporary procedure changes.

i 8.12 Determination of component operability of a generic nature l using information from procedure A-52.4.

t 5.

Simulation Facility Fidelity Report: i a. The Ginna facility recently acquired a plant reference simulator

for use in operator training. This was the first time this simulator was used to administer NRC operating examinations.

, , Although there were a fair number of outstanding trouble reports ' i written against the simulator, most were of a minor nature and did ' not impact on operation of the simulator. The NRC was able to make , full and accurate evaluatio.is of the candidates during the , simulator portion of the operating examinations. The outstanding i trouble reports ~are being pursued for resolution by the utility in

association with the simulator vendor.

, i b. It is apparent that the facility has placed great emphasis on ensuring

fidelity of the simulator with respect to both the control room's actual physical layout and the modeling of actual plant systems' ' responses, c. On two separate occasions, major simulator problems occurred. Both ! were of a non-recurring nature. The first problem appeared to be a i breakdown of the model which occurred at the end of a scenario.

In this scenario, a steam break resulted in a large and rapid decrease in , reactor coolant pressure to the point where the accumulators injected.

, However, earlier in the scenario, one of the accumulators had drained '

due to a leak and contained only nitrogen. Upon injection of the accumulators, the nitrogen in the empty accumulator was introduced into the reactor coolant.

The simulator model was unable to handle this complication as evidenced by the fact that large fluctuations in reactor coolant pressure were noted to occur (0 to 3000 psig spikes) as well as level fluctuations in the faulted SG even though it was blown d ry. The second event involved the simulator "freezing up " at the end of a scenario. As noted, both events happened at the end of their scenarios and did not interfere with the evaluations of the candidates, . i i OFFICIAL RECORD COPY OL EXAM GINNA - 0006,0.0 , 11/06/87 !

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l-5-6 ',

Personnel Present at Exit Interview: . NRC Personnel

R. Temps, Operations Engineer ' L. Briggs, Senior Operations Engineer E. Yachimiak, Operations Engineer N. Perry, Resident Inspector ' Facility Personnel R. Marchionda, Ginna Training Manager T. Meyer, Superintendent-Ginna Support Services T. Schuler, Operations Manager J. Wayland, Training Coordinator-Simulator T. White, Senior Reactor Operator i 7.

Summary of NRC comments made at exit interview: The chief examiner reviewed the number and type of examinations ad- . ministered during the week and gave an overview of the observations and generic strengths and weaknesses observed during conduct of the examinations. The chief examiner also briefly reviewed some ' problems noted with facility training material and procedures. These had been previously discussed with the training department, details of which can be found in section 10 of this report.

It was also noted to the facility that.me cleanliness of the areas in the auxiliary building accessed by the txaminers during the plant walkthroughs appeared to be quite good.

8.

Summary of facility comments made at exit interview: ' The facility stated that the examinations were conducted in a profes-

sional manner and that they had no complaints with respect to the conduct and content of the operating portion of the examinations. The facility did voice some concern over the written examinations.

Specifically, they felt that a number of the RO examination questions were not specific to the Ginna plant and that the SRO examination required too much memorization of specific knowledges; that is, that there is a difference in knowledge level required between what the NRC expects and what the training department expects.

i , 0FFICIAL RECORD COPY OL EXAM GINNA - 0007.0.0 11/06/87 , n, ,,, - - -, n,- n,-r--,c...,.,-,.. -, - - - -,, - -,,,- --,,,,-v------,-- w.-,-.. ,- r

-6-9. Examination Review: Following administration of the written examinations, an examination review session was held. Facility comments on both examinations were quite numerous and were discussed on a line item basis. Some changes to the examination answer keys were made at the review session; other changes were made based on comments submitted by the utility.

Not all facility comments result d in changes to the examination grad-ing keys. Utility comments can be found in attachment (3) and NRC resolution to the comments along with additional NRC instituted changes to the answer keys are in attachment (4).

10. Training Material and Procedure Discrepancies: The chief examiner met with Ginna training department representatives on October 8, 1987, to discuss NRC concerns with the training materials sent for examination preparation and to point out discrepancies noted in some of the plant's procedures. Additional discrepancies were found in procedures subsequent to this meeting and are identified in section 10.b of this report. Most of these items have been brought to the attention of the Ginna senior resident inspector for information and tracking purposes.

A. Deficiencies Noted at Meetin.gl 1) System Description manuals are not detailed enough for proper for-mulation and development of examination questions.

2) Questioning in certain areas was limited due to insufficient material being sent; areas specifically lacking were fuel handling procedures and testing and surveillance procedures that the Opera-tions department are responsible for. Other areas lacking in ma-terial were discussed and guidance has been given to the training department regarding information to be sent for future examina-tions.

3) Material not updated to reflect procedures which have been deleted. One example given was malfunction GEN-7 from the sim-ulator documentation book which referenced procedure E-4. Pro-cedure E-4 no longer exists.

! 4) Standard Technical Specifications statement 3.0.3 was found to have been improperly incorporated into step 3.1 of procedure , l A-52.4. The senior resident inspector confirmed this finding with Ginna's Technical Manager.

5) Procedure 0-1.2.2, as currently written, cannot be used for deter-mination of SDM.

,

0FFICIAL RECCRD COPY OL EXAM GINNA - 0011.0.0 11/06/87 _ _

-7-B. Discrepancies Noted Subsequent to Meeting: 1) Procedure 0-1.2.2 was found to have the following inconsistencies: a) Step 5.7.5 states to use the value from step 5.7.4 to determine differential boron worth from figure 4. This is inconsistent with figure 4 which requires the value for core burnup rather than baron concentration, which is the value given in step 5.7.4.

b) Step 5.11.3 states to fill in the value for differential boron worth obtained from step 5.7.2. The actual value is found in step 5.7.5, not 5.7.2.

2) Question 4.01 on the R0 written examination was deleted because a precaution statement on which the question was based was deleted from procedure 0-2.1. However, the same precaution statement can be found in procedure 0-1.2. Also, the lesson plans for procedures 0-1.2 and 0-2.1 still contain reference to the deleted precaution.

3) The facility supplied system drawings were compiled as of 7/25/87.

However, not all of the drawings supplied are up-to-date. For example, drawing 33013-1248 Revision 1 was supplied but this drawing is currently on revision 3 which was issued on May 28, 1987. The facility needs to ensure that all materials supplied to the NRC are current.

Attachments: 1.

Written Examination and Answer Key (RO) 2.

Written Examination and Answer Key (SRO) 3.

Facility Comments on Written Examinations after Facility Review 4.

NRC Resolution of Facility Comments and Additional Changes to the Answer Keys as a Result of Examination Grading 0FFICIAL RECORD COPY OL EXAM GINNA - 0012.0.0 11/06/87

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. ,wi , - " ~ ~U..S.1 NUCLEAR REGULATORY COMMfSSf0N REACTOR OPERATOR LICENSE EXAMINATION FACILITY: GlNN@____________________ _ REACTOR. TYPE: PWS-WEg2_________________.

'DATE.ADMINSTERED: 8Z/19/99_________________ , .

EXAMINER: YAgHIMIAK _E.____________ i PV d [ h._b_ [_(_k __( [__ __ CANDIDATE __ IgS1BuC11geS_1g;CeNg1gezet Use. separate paper.for-the answers.

Write answers on one side only.

Staple = question sheet on top of~the answer sheets.

Points for each-question are indicated in parentheses after the question.

The passing , grade requires at least 70% in each category and a final grade of at ! least 80%. Examination papers will be picked up six (6) hours after ' the examination starts.

-

I % OF

CATEGORY- % OF CANDIDATE'S CATEGORY __Y8LUE_._1gl@L ___SCQBE___ _yeLUE__ ______________CG1EGQBy_____________ 2N.D C25E99~_ 2Et99 ________ 1.

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

___________ ________ 2.

PLANT DESIGN INCLUDING SAFETY

_2E199__ _3Eigg AND EMERGENCY SYSTEMS

? t. GC' p _'25199_['_3Eigg ___________ ________ 3.

INSTRUMENTS AND CONTROLS

! - - . ??.en[- } 25. 00 45,00 4.

PROCEDURES - NORMAL, ABNORMAL, i U EMERGENCY AND RADIOLOGICAL CONTROL ' 92.Co _199Eg__l) ________% Totals ' ___________ Final Grade i i t ? t l All work done on this examination is my own.

I have neither given

nor received aid, i s a b e b $

ch enda .* +o Osn* d&%). '

l ! t . - ...

NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply: 1.

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

2.

Restroom trips are to be limited and only one candidate at a time may leave.

You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.

3.

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

4.

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

5.

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

6.

Use only the paper provided for answers.

7.

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

8.

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

9.

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

10. Skip at least three lines between each answer.

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

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

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

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

' 15. Pa'tial credit may be given.

Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.

16.

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

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

This muut be done after the ex ami nat i on has been completed.

18. When you complete your examination, you shall a.

Assemble your examination as follows: (1) Exam questions on top.

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

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

b.

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

c.

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 ex amination is still in progress, your license may be deni ed or revoked.

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'l.- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION Page

1 IdE6dODXN@dlCS _dE@l_16@NSEE6,@NQ_ELulD_ELQW t , - QUESTION 1.01 (2.00) .WHAT'are the FOUR (4) limits-that must be observed by an operator to ennun l that Technical Specification core thermal li nits are met during normal-operations? - QUESTION.

1.02 (2.00) L WHAT are FOUR (4) conditions that indicate to an operator that natural ' circulation flow exists? List applicable parameters and trends.

' QUESTION 1.03 (3.00) , L j For the f ollowing condi tions/ actions given below, determine HOW (INCREASES.! [ DECREASES, or NO CHANGE) and WHY RCS temperature changes in terms of' core [ reactivity.

Assume NO operator action is taken other than that indicated.

' Consider each part separately.

Compare FINAL equilibrium RCS temperature.

a.

An unsaturated mixed bed demineralizer is.placed into service.

Reactor , power is at 90% with rods in manual.

. b.

Control bank D cods are pulled cut from 100 tc 120 stens.

Reacte-v.

' , is at 10% anc steam dumps are in service.

i . c.

A turbine ~ control valve fails open while at 60% reactor power causing , turbine load to increase to 400 MWe, Rods are in automatic.

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OUESTION 1.04 (2.00)

Answer the following questions TRUE or FALSE:

' L a.

If a centrifugal pump is in service and the operator INCREASES sts '

flowrate, the pump's available Net Positive Suction Head (NPSH) will ' INCREASE.

b.

If a condensate pump was operating at "RUNOUT" conditions, motor current , would indicate LESS than normal values.

c. When starting a condenser circulating water pump, motor starting curreni is REDUCED by CLOSING the pump's discharge valve.

l d.

If - a centrifugal pump's speed is DOUBLED, its flow will DOUBLE.

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1 __ESINCIE6ES_QE_NUC6E@@_Eg8E6_E(@NI_ GEE 6@IlgN t IHE6MgpVN@MlCS _HE@I_I6@NSEE6_@NQ_E(Ul@_E(gW t OUESTION 1.05 (2.40) For EACH of the following conditions, HOW does the Moderator Temperature , ' Coefficient (MTC) change (MORE NEGATIVE, LESS NEGATIVE, NO CHANGE).

a.

Reactor power INCREASES from 20% to 50% at EOL.

b.

Baron concentration INCREASES from 1200 ppm to 1500 ppm at BOL, HZP.

c.

Core age INCREASES from MOL to EOL at HZP. 300 ppm boron concentration, d.

Control bank D rods are INSERTED from 220 steps to 120 steps, baron concentration at 1200 ppm.

-GUf ST-f ON - 1.06 -H-trOO+ -- - Which ONE of the f oll owing core aging factors ALONE would cause the-f'uel temperature coefficient to become LESS negative? - ./ - A.

fuel densification , ~ ~ ~ ~ , ~', / * ( l- ' , B.

clad creep ./ C.

Pu-240 buildup ' , Ds-ff[sion product gas buildup in fuel / gap clad - . QUESTION 1.07 (2.25) If reactor power is INCREASED from 50% to 100% at BOL, HOW will Differential Rod Worth change (INCREASE, DECREASE, or NO CHANGE) for 'the following situations? JUSTIFY EACH ANSWER.

Consider each case separately, a.

Rod position and baron concentration are held constant, RCS temperatore is allowed to DECREASE.

b.

Boron concentration is held constant, rods are WITHDRAWN from 110 steps on bank D to maintain RCS temperature constant.

c.

Rod position is held constant, baron concentration is DECREASED to maintain RCS temperature constant.

(***** CATEGORY 1 CONTINUED ON NEyT PAGE 88ed4) ]

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1:__EBINCIE6ES_QE_ NUCLE @B_EQWEB_E6@NI_ GEE 6@IlgN1 IHE6DggYN@DICS _HE@I_IB@NSEEB_@NQ_E6Ulg_E69W

OUESTION 1.08 (0.95) During a reactor startup, the FIRST reactivity addition caused count rate to increase from 100 cps to 200 cps.

The SECOND reactivity addition caused count rate to increase from 200 cps to 400 cps.

WHICH ONE of the following statements is CORRECT? A.

The FIRST reactivity addition was larger.

B.

The SECOND reactivity addition was larger.

C.

The FIRST and SECOND reactivity additions were of equal magnitudes.

D.

Not enough information was given to determine the relationship between reactivity values.

GUESTION 1.09 (1.00) WHAT are the TWO parameters / conditions which can be used to determine the rate of decay heat generation? QUESTION 1.10 (2.40) - a.

Given the attached graph (Attachment C), explain WHY xenon concentration changes between points A-B, and B-C.

(1.20) b.

HOW long does it take for xenon concentration to reach equilibrium after the given step power change? (0.60) c.

At WHAT power l evel is equilibrium xenon concentration ONE-HALF of the 1 0 0 7. reactor power equilibrium xenon concentration? "( O. 60 ) QUESTION 1.11 (1.50) Given the partially filled out RGLE c al or i met r i c (Attachment B).

complete the required spaces and CALCULATE Core Thermal Power in MWt.

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IHEBMggYN@OICS _HE@I_IB@NgFEB_@Ng_E(U1g_E6gW t QUESTION 1.12 (1.80) HOW will each of the following affect the results of a secondary calorimetric power calculation? Limit your answer to CALCULATED LOWER THAN ACTUAL, CALCULATED HIGHER THAN ACTUAL, or CALCULATED SAME AS ACTUAL.

Assume the plant is at 100% and consider each case separately, a.

Measured feedwater temperature is 10 degees HIGHER than actual feedeAter temperature.

b.

Measured steam generator pressure is 30 psig LOWER than actual steam generator pressure, c.

Measured feedwater flow i c, l.OES lbm/hr HIGHER than actual feedwater flow.

QUESTION 1.13 (2.70) Complete the attached (Attachment A) Cri ti cal Rod Position Calculation, form 0 - 1. 2. 2., and CALCULATE how much makeup water (i n gallons) must be addeC to the RCS if the reactor is to be taken critical 10 hours after a reactor trip? Assume the plant was at 75% power for three ( 3 ', days prior to the trip and that core burnup is ISO (MWD /MTU).

. (***** END OF CATEGORY

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_ 0 __E68NI_QEg1QN_lNCLUDIN@_@@EEIY_@ND_EDEBQENCY Page

SYSIEd5 OUESTION 2.01 (2.50) a.

WHAT is the function of the reactor coolant pump's flywheel? (1.00) b.

WHY is there a minimum required differential presssure of 220 psid across the RCP No. 1 seal? (0.50) c.

WHAT are THREE (3) reactor coolant pump components that would be affected by a loss of Component Cooling Water? (1.00) QUESTION 2.02 (2.50) HOW (OPEN, CLOSED, AS IS, REMAINS OPERABLE) do the following valves respond on a complete loss of instrument air.

Assume the plant is at 100% power with all systems in automatic.

a.

CV-56, steam generator "A" atmospheric relief 'alve b. FCV-1108, reactor makeup to charging pump suction flow control val ve c.

FCV-135. letdown line pressure control valve d.

LCV-Il28, RWST to charging pump suction level control val ve

e.

FMV-487A. turbine driven auxiliary feedwater pump flow control valve OUESTION 2.03 (3.00) a.

WHAT are TWO (2) reasons for using letdown when in colid plant, cold (1.00) shutdown conditions? . b. WHAT are THPEE (3) flowpaths that charging flow can be directed to AFTER the regenerative heat exchanger 7 (1.00) c. WHY is low pressure letdown control valve (PCV-135) necessary for normal, at power, letdown operations? (0.50) d. Electrical power is lost to the heat tracing lines from the boric acid tanks to the suction of the charging pumps.

WHY is this a concern? (0.50) . i (*e*

CATEGORY 2 CONTINUED ON NEXT PAGE
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2.__PL@NI_DESl@N_lNCLUDISG_@@EEIY_@ND_EDER@ENCY Page

SYSIEUS QUESTION 2.04 (3.00) d.

WHAT are the NORMAL, EMERGENCY, and LEAST desirable sources of water for the Auxiliary Feedwater System (AFW)? (1.50) h.

WMAT are the TWO (2) AFW pump turbine overspeed trips and HOW is each reset? (1.00) c.

HOW do the AFW turbine steam supply valves (MS-3505A, MS-3504A) fail on a loss of AC power (OPEN, CLOSED, AS IS, REMAINS OPERABLE)? (0.50) GUESTION 2.05 (3.00) Given the below conditions following a Loss of Cool an t Accident. WHAT SIX (6) safeguards pumps should be running after two minutes into the LOCA? containment pressure: 18.0 psia RCS pressure: 850.0 psia with: Loss of all offsite electrical power Diesel Generator "B" out-of-service s QUESTION 2.06 (2.00) a.

WHAT is the design basis for the containment spray system? b.

HOW is NaOH added to spray pump flow so that injection phase pH is maintained within design limits? Include a description of system components and flowpaths in your answer.

. QUESTION 2.07 (3.00) For EACH of the following Radiation Monitoring System detectors, list FOUR (4) automatic actions (not including annunciators) which occur upon a HIGH alarm, a.

R-ll or R-12, containment particulate or gas monitors b.

R-13 or R-14, auxiliary building Particulate or gas monitors with auxiliary building filters "IN" (***** CATEGORY 2 CONTINUED ON NEXT PAGE

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, 2c__ELONI_QggigN_lNCLUQ1N@_g8EEIy_8NQ_EdEBGENCY Pege 10 SYSIEUS QUESTION 2.08 (2.00) a. List THREE (3) locations WHERE halon is used? (1.00) b.

WHY is Fire Water Storage Tank pressure maintained above 100 psig? (0.50) c.

WHAT TWO (2) conditions will cause the Motor Driven Fire Pump to automatically trip if running? (0.50) QUESTION 2.09 (2.00) a.

WHAT are SIX (6) points of possible fluid in-leakage into the Component Cooling Water (CCW) system? Assume non-accident conditions.

(1.50) b.

WHAT TWO (2) control room indications can be used by the operator to detect in-leakage into the CCW system? (0.50) GUESTION 2.10 (2.00) Draw AND l isb el a one-line diagram showing HOW 4160 VAC bus 12A supplies power to 125 VDC bus lA.

Include ALL major components except breakers.

s . (***** END OF CATEGORY

4***4)

3 __INSIBUMENIS_9ND_CONIBOLS Page 11

GUESTION 3.01 (2.00) For each of the phrases below, STATE whether the phrase describes the Source Pange (SR). Intermediate Range (IR), AND/OR Power Range (PR) nuclear intrumentation.

More than one nuclear instrument may apply for each.

a.

Uses circuitry to eliminate effects of gamma radiation on instrument.

b.

Uses gaseous baron in detectors for neutron sensitivity.

c.

Detectors operate in the Ionization Region of the gas filled detector characteristic curve, d.

Provides Reactor Protection System trips which must be blocked at some time during a reactor startup to full power.

QUESTION 3.02 (2.00) WHAT are the FOUR (4) Reactor Protection System (RPS) trips that protect the core against DNB during rapid transients while at 100% power? (2.00) OUESTION 3.03 (3.00) . For each of the following situations determine WHICH rod c on t r ol interlocks would prevent rod motion.

For each situation, state in which DIRECT!GN rod motton is prevented and WHAT operator actions must be taken to reestablish rod motion.

Assume rod control is in AUTO.

Consider each case separately.

STATE ALL APPLICABLE SETPOINTS.

a.

A Power Range nuclear instrument fails HIGH during a power locrease with reactor power at 6 0 */.. . b.

Turbine first stage i mpul se pressure transmitter (PT-485) fails LOW during a plant startup with reactor power at 25%. c.

Loop 1 T(hot) f ails HIGH with reactor power at 100%. (**e** CATEGORY 3 CONTINUED ON NEXT PAGE

- - - e)

A

3a__INSIByMENIS_BNQ_GQNIBQLS Page 12 (a. 95) OUESTION 3.04 '2.75) a.

What TWO '2) conditiond must exist before Si can be reset by an operator after it has been automatically actuated? ' S S 0) 41. 00i

b.

What is the reason for resetting SI after actuation? (0.50) c.

WHAT TWO (2) SI signals should be blocked and WHAT conditions (setpoints and coincidences) must be present for the blocks to be effective during an RCS cooldown and depressurization.

(1.25) -OtfES T 1 ON -3.-05-OW ).e Match the proper Steam Dump Control (s) or Control l er (s) in Column B to e ch statement in Column A.

More than one response in Column B may apply.

(example: 'b.

1,2,3 OR c.

5) COLUMN A COLUMN a Operates Steam Dumpe based on a

Turbine p controller temperature deviation (Tavg-Tref) 2.

S .m pressure controller b.

Used to reset the C-7 arming signal o.

Load rejection c on t r ol l er c.

Contains no trip open control features

Steam Dump Control Mode Selector Switch d.

Uses auctioneered high Tav 5.

Steam Dump I nt er l oc k e.

Compares the actual rameter to Selector Switch that of a pre-set alue for operation f.

Used to all a cooldown val ves to open

below P- . g.

Uc d to turn the system off i. Uses quick open biscables for rapid transients QUESTION 3.06 (2.50) WHAT are FIVE (5) CONTROL systems that would be affected by a loss of 120 VAC instrument bus IC? Assume the plant is at l O O Y. power.

(***** CATEGORY 3 CONTINUED ON NEyT PAGE e

- s)

,

3:__IN@l6UMENI@_gNg_CgNJggL@ Page 13 OUESTION 3.07 (2.50) a.

Explain HCW a reactor trip is avoided when testing the nuclear instrumentation when the reactor is at 5% power AND 50% power.

Include Source Range AND Intermediate Range testing with appropriate interlocks, bypass switches, and logics.

(1.50) b.

A power range nuclear instrumentation channel fails HIGH.

HOW do the ROD STOP and PO'c!ER MISMATCH PYPASS swi tches af f ect Rod Control System operation when individually placed to the failed channel? (1.00) QUESTION 3.08 (3.00) a.

WHAT input signal is used to adjust programmed level for the pressurizer level control system? (0.50) b.

WHAT is the normal programmed pressurizer level at no load AND full lead? (0.50) c.

If pressurizer level control channe) LT-427 fails LOW during 100% power operation while in automatic c o n t r e,1, WHAT reactor protection signal will cause the reactor to trip? Explain HOW LT-427 f a i l i n g uC'.d causes this to occur.

Include setpoints and coincidences.

Assume NO operator action is taken and that LT-427 is the controlling channel.

(2.00) OUESTION 3.09 (2.50) HOW (OPEN, CLOSE, NO CHANGE) does Feedwater Regulating Valve (FCV-466) respond INITIALLY to each of the following conditions.

Assume reactor power is at 100% with FCV-466 in AUTO.

Consider each part separately.

' a.

controlling S/G level transmitter fails LOW b.

safety injection actuation c.

controlling S/G steam flow transmitter falls HIGH d.

controlling S/G pressure transmitter fails LOW e.

controlling S/G feed flow transmitter fails HIGH i (ne**e CATEGORY 3 CONTINUED ON NEXT PAGE eee*e) J

3 __ INS 16UDENI@_@yp_CQNI6Q6@ Pcge 14

QUESTION 3.10 (1.75) For each of the items listed below, explain HOW a loss of 125 VDC Bus 18 would affect their operation including any automatic actions that would take place.

a.

Inverter !B (0.50) b.

Turbine Driven Auxiliary Feedwater Pump / Turbine (0.50) c.

Reactor Protection System (0.75) d (***** END OF CATEGORY

- - - - )

4:__PBQGEQUSES_;_NQSd@(t_@BNQBd@(t_EdESGENQY Pago 15 GNQ_B@Qlg(QGlq@(_QQNISQL . l' ) j-euesT-I ow- .O:

2.0c:

v ^ A precaution in RGLE procedure 0-2.1, "Normal Shutdown to down," states that auxiliary feedwater flow shall be ' to a maximum of 150 gpm whenever the steam generator s below the lo-lo level setpoint.

When regaining level f ow that setpoint, auxiliary feedwater flow should alsg b mited to 150 gpm until 307. level has been achieved, J4HA he basis for this precaution? p ) OUESTION 4.02 (3.00) Answer the f ollowing in accordance with Emergency Operating Procedure (EOP) E-2, "Faulted Steam Generator Isolation."

a.

HOW is a faulted steam generator identified? (1.00) b.

List FIVE (S) operator actions that should be performed to isolate a faulted steam generator.

(2.00) GUESTION 4.03 (2.70) For the f ollowing Immedi ate Acti on steps of procedure E-0, "Reactor Trip or Safety injection," state HOW the actions are completed by the operator, a.

verify reactor trip (2 indications) b.

veridy containment isolation (3 indications) c.

verify FW isolation (2 indications)

d.

verify service water system operation (2 indica + - is ' OUESTION 4.04 (2.00) Answer the following questions as they apply to EOP FR-S.1, "Response to Reactor Restart /ATWS."

a.

WHAT action should the operator take if the rod drive MG sets will not trip and the reactor remains critical? b.

WHAT action should the operator take if the turbine has not automatically tripped, will not manually trip and will not runback? (**e:* CATEGORY 4 CONTINUED ON NEXT PAGE *oee4) i

_. -_ _ _ _ _ _. l 41__EO9EE998EE_2_U90U9b1_@@NQBd@L1_EjE69ENgY Page 16 l BNQ_B@QJQLQQlg@L_ggNI6QL QUESTION 4.05 (2.50) You have been selected to assist in work which must be' performed inside the containment in a radiation field of 850 meem/hr gamma. 100 mead /hr thermal neutron, and 200 mrad /hr beta.

Assume that you are 29 years old and have a life time exposure through the last quarter of 53 REM on your NRC Forrn 4.

Additionally, you have accumulated 0.5 REM this quarter.

a.

Assuming you wear the appropriate protective clothing, HOW long may you work in this area without exceeding your lOCFR2O limits? Show ALL work.

(2.00) high radiation b.

During a declared emergency, you volunteer to enter a

area to perform work necessary to prevent further effluent release, in accordance with the Station Procedures, WHAT is your maximum allowed whole body exposure? (0.50) CUESTION 4.06 (1.75) WHAT are FIVE (5) symptoms of a steam generator tube rupture other than decreasing pressurizer l evel and pressure? DUESTION 4.07 11.25) Describe the TWO (2) methods by which RCS overpressure pratettion can be ' maintained while in Cold Shutdown with the RHR system in service.

Include any applicable setpoints.

I IZ i'N I u

OLIEST1ON 4.08-t37001 - a.

If Di esel Generators (D/G) "A" and "B" jacket water temperatures alarm HIGH due to a loss of service water, WHAT actions would an operator take LOCALLY to supply the alternato source of cooling water to these D/Gs? (1.00) b.

HOW would an operator trip D/G "A" if the local selector Gwitch was in the emergency position AND an $1 signal was present? +17 (.KJt ( 0. 50), c. When D/G 1A is paralleled to 400 VAC Bus 14, WHAT TWO (2) D/G control s r egul ate load AND power factor? (1.00) i (e**** CATEGORY 4 CONTINUED ON NEXT PAGE

- - - - )

P l ' r . - - _ _, _. _ _ _ _ _ _ _. _. _. _

4:__P_6QCEQ6B(S_:_NQBd@(t_@BNQBd@(t_EDEGGENCY Pcge 17 GNQ_8@Q1Q(Q@lC@(_CQNIRQL (?oo) QUESTION 4.09 t 2dW For the following questions, assume EOP FR-C.1, "Response to Inadequate Core Cooling," is in effect.

a.

RWST High/ Low level alarm actuates.

WHAT action is the operator required to take? (0.50) b.

WHEN can a faulted or ruptured S/G be used for RCS cooldown? (0.50s c.

WHY should the RHR pumps NOT be allowed to run for longer than 30 minutes without Component Cooling Water (CCW) flow avai l abl e to the RHR heat nxchangers? (0.50) d.

Both RCPs were tripped prior to entering FR-C.I.

-WHEN ond WHY would they be required to be restarted? tii444-; (U.Sc) OUESTION 4.10 (1.50) Prioritize the below list of Critical Safety Function Status Tr ees from HIGHEST to LOWEST.

1.

Integrity - RED 2.

Insenter7 COANCC 3.

Core Cooling - RED

Subcriticality - YELLOW 5.

Containment - ORANGE 6.

Heat Sink - YELLOW OUESTION 4.11 (2.80) . a.

State FOUR (4) symptoms for entry into procedure AP-CVCS.2, "Immediate Boration."

(1.60) b.

Poric Acid (BA) ficw is NOT as expectad, WHAT are THREE (3) valves an operator can open to establish a BA flewpath tf Emergency Borate Valve MOV-350 will not open? (1.70) (***** END OF CATEGORY

- - - - )

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

- - - - )

' ' a

'

;v > ;

.. ~ Cycle efficiency = (Net wrk: .. - ' v " 5/t.

f = ma cut)/(Energy in)

s = V t + 1/2 at w = mg o

-At E = mc A" KE = 1/2 mv 4 = (Vf - V )/t o x = tn2/t1/2 = 0.693/t1/2 at w = e/t l/2*ff " [(t )(t )] yf= U l

2 nD [(t /2) + (t I3 W = v 6P-4, b ~D d = 931 m di = V Ao g,te e - * -- = . . Q = mCpat .yx I=Ie 6 = UALT 10.x/TVL

I=I pwr = 'd ah

f TVL = 1.3/u sur(t) HyL = -0.693/u P = P 10 P = P e*I o SCR = S/(1 - Keff) SUR = 26.06/T CR = S/(I - Keffx) x-6eff2) ' CR (1 - Keff3) = CR (

SUR = 26s/L* + (s - p)T j _ 'M = 1/(1 - Keff) = CR /CR j

7 = ( t=/o) + ((s - o Y iol M * (1 ~ Keff0)/(1 - KOff1} T= U(o-6) SCM = (1 - Keff)/Keff T = (8 - o)/(Io) (* = 10 seconds

-l o = (Kafs-II/Ka f f * AK

- eff 'eff _ = 0.1 seconds A eff (I + M a = ((t=/(T Keff)] + (I ' / =Id Idj 2 =2 2 j 1d p = (tov)/(3 x 1010) I)d g

'

R/hr = (0.5 CE)/d (ceters) R/hr = 6 CE/d2 (feet) l t = o ti 14iscellaneous Conversions ' 'Jater Pareeters curie = 3.7 x 1010eps I gal. = 8.345 lem.

kg = 2.21 lbm 1 gal. = 3.78 liters Ihp=2.54x10}Stu/hr = 7.48 gal.

1 m = 3.41 x 100 Stu/hr J 1 ft

Oensity = 62.4 lbm/ft lin - 2.54 cm , Density = 1 ga/cm3 l

F = 9/5'C + 32 ric1t ci vaporization = HG 3:u/1tsm
C = 5/9 ( *F-32)

. Heat of fusion = ida R u/lem 1 Biu = 778 ft-lbf 1 Ata = 14.7 psi = 29.9 in. nc.

1 1t. ti;0 s 0.0335 Itf/in.

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

.. _ _ _ _ _ _ _ _ _ _ _ _ _ _, . ATTACHMENT A O-1.2.2:1 0-1.2.2 CRITICAL _FOD POSITION CAirULATION l _' 1.0 PURPOSE: 1.1 This procedure provides the method for calculating critical Rod Position.

2.0 REFERENCES: 2.1 Design data, current cycle.

3.0 INITIAL CONDITIONS: 3.1 Approximate time reactor is to be made critical is known.

4.O PRECAUTIONS: The critical rod position calculation sheet should be 4.1 filled just prior to returning critical.

A time delay in using the calculation will cause errors due to the transient effect of Xenon.

. 5.0 INSTRUCTIONS: Calculate estimated critical rod position using steps 5.1 5.3 through 5.9.

Use attached curves to assist in calculations.

5.2 CAo010h /0!S

TIME: 0900 NAME: 5.3 DATE: ' ( 1), prior to shutdown, The core is critical (K eff = at a given % of power, Xenon concentration, Baron concentration, and rod position.

Tr.e algebraic sum of all changes in reactivity, after shutdown, will determine the amount of positive or ne gative reactivity e that has been added to the core duas to the shutdown.

sum is time d e pe nt.e rit due to Xenon, The algebraic the time after shutdown must be considered therefore, to determine the Xenon worth.

Ti.e algebraic sum is used to determine the criticat rod position.

A positive algebraic sum is the an>unt of excess

- - - - - - - - - _ _ _ _ _ _ _ _ _. , , ATTACHMENT A O-1.8 282 5.3 (con't) reactivity in the core that must be compensated for by leaving control rods in the core.

The bank and height of control rods left in the core is the critical rod position.

NOTE: A negative algebraic sum indicates the core would not become critical with all rods out.

5.4 POWER DEFECT: '75 g pr.

5.4.1 Power prior to shutdown.

Boron concentration at the above power level.

5.4.2 . SAMPLE DATE: /0

TIME: /200 /# 00 Boron ppm /./ Using power and boron values above, determine reactiv- , 5.4.3 ity increase from power defect curve or power defect table.

+ ocm Attached Figure 1 or Figure 1A.

NOTE: This will always be positive.

5.5 ROD WORTH 5.5.1 Determine rod position at steady state power level (value 5.4.1) //

DATE: / 0/ Y/ P7 TIME: R4OO t ' ROD POSITION: O , BANK L35 - STEPS Determine worth of rods in core prior to shutdown using , 5.5.2 Integral Rod Worth Curve attached, Figure 2A for BOC or Figure 2B for EOC.

This will always be a positive value.

+ pcm , REACTIVITY CHANGE DUE TO XENON 5.6 Determine hours shutdown using shutdown time of power 5.6.1 level used in 5.4.1 to time of estimated criticality.

' IO hours , i t-

. _ _ _. _ _ _ -ATTACHMENT A O-1.2.2:3 5.6.2 To_ determine reactivity change due to Xenon, a steady state power level must be used.

A steady state condition will be defined as at least 40 hours.

If-power prior to shutdown was not steady state, it will be necessary - to obtain power level from the Reactor Engineer or designated alternate.

5.6.3 Using the above values 5.6.1 and 5.4.1 (if steady state) determine reactivity change from attached curve Figure 3, 3A, 3B, 3C, etc.

ocm NOTE: This value may be , 0, or + depending on the time after shutdown.

/ ) 5.6.4 Time and date of estimated criticality.

/0 S/ 87 / DATE ' / TIME /OCO NOTE: Criticality must occur within one hour of expected criticality time during the first 80 hours after shutdown.

If not, a new , critical rod position must be calculated and the rest of the procedure marked N/A.

,

5.7 BORON 5.7.1 Boron prior to shutdown value 5.4.2.

/200 Boron ppm 5.7.2 Boron in core now:

SAMPLE: C900 / OU Boron ppm DATE: /D B7 TIME:

5.7.3 Change in boron concentration.

. /4O Boron ppm (Subtract 5.7.2 from 5.7.1) 5. '7. 4 Obtain average boron concentration (5.7.1) + (5.7.2) =

Differential boron worth determined from differential 5.7.5 boron worth curve (attached figure 4) using value from step 5.7.4.

DCm ' ppm i ,

j.

I l ATTACHMENT A O-1.2.2:4 5.7.6 Change' in boron (5.7.3) x Differential boron worth i ncm'

(5.7.5).

' NOTE: If boron has.been added to the system the resultant reactivity will' carry a negative value.

If boron has been decreased in the system the resultant reactivity will carry a positive value.

! 5.8 SAMARIUM Use power prior to shutdown (from step 5.4.1) and. hours 5.8.1 shutdown (from step 5.6.1) to determine Samarium worth from attached Figure 5.

b b ocm Determine the Samarium value by taking the algebraic 5.8.2 difference between the worth determined in step 5.8.1 and the equilibrium valve of (-) 615 pcm.

O oc.

(- 615) (5.8.1) = - NOTE: This value will always be negative.

5.9 . TOTAL REACTIVITY CHANGE (carry sign for each value) . + pcm 5.9.1 Power defect value 5.4.3 + pcm 5.9.2 Rod worth value 5.5.2 pcm 5.9.3 Xenon value 5.6.3 ocm 5.9.4 Boron value 5.7.6 C) ocm -

5.9.5 Samarium value 5.8.2 " pcm l 5.9.6 Total 5.9 through 5.9.5 5.10 ESTIMATED CRITICAL ROD POSITION '? ' The algebraic sum of the reactivity changea determines the rod position that will cause the reactor to attain criticality.

When the sum is positive, the reactor will go critical with the rods still inserted some number of steps.

When the sum is zero, the reactor will go critical with all rods withdrawn.

When the sum is negative, the reactor will not go critical, ' and some further steps must be taken to increase the sum 2 0.

The rod position at which the reactor goes critical is the "Critical Rod Position".

This value is found using the integral rod worth curve (attached Figure 2A or 28).

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

_ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ .. ATTACHMENT A O-1.2.2 5 i i: 5.10.1 Estimated Critical, Rod Position.

If the value at l J step 5.9.6 is negative, mark this step N/A.

BANK

STEPS NOTE: If the estimated critical rod position is acceptable, the following steps (5.11, 5.11.1, 5.11.2, 5.11.3, 5.11.4 may be marked N/A).

5.11 Desired Critical Rod Position.

g BANK , /OO STEPS

5.11.1 Reactivity associated with desired critical position.

DCm 5.11.2 Reactivity associated with algebraic difference of estimated (step 5.9.6) & desired (step 5.11.1) critical rod position.

Dcm (Step 5.11.1) (Step 5.9.6) = - Differential boron worth determined from differential 5.11.3 boron worth curve (attached Figure 4) using value from step 5.7.2.

' DCm ppm i 5.11.4 Amount of boron change required to achieve desired ' critical rod position (5.11.2)/(5.11.3).

Indicate boron change with proper sign (- for dilution, + for boration).

. opm (Step 5.11.2)/(Step 5.11.3) = Verification of calculation by Reactor Engineer.

5.12 COMPLETED BY: DATE COMPLETED: SHIFT SUPERVISOR:

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TO

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33 l EORAT.

James F.' Sweet ". ) j .. ~ ' . .I

Carl H. Peck .

i - I O K . . , \\ - 3."

. ,, -. . - - - - - - - ---. --- . .- -.. - .

( U1 0-6.3 18 ATTACHMENT B Rochester Gas artl Electric Corpotution , l G1MM UFATICN IC&E CAIGtlMM11 TIC 3 Attach: mat III ConttVlled Copy #4 Refer to O-6.3 for Instructions [nTE 10/5/1987 TIME 0800 rulforeim BY CANDIDATE

I , Shift Supervircr:

I 1.1T'!1MN1131 CATGtIMM11tIQi:

Avg. Steam Genetutor Pressuru (A&B) 735.0 ruig + 14.7 = paia Of 420.0 Avg. Feat #ater Tceperaturu = 113 at steam ptre,suru (psia) f at steam pressure (psia) aid asy. fecd trep. =

h(ure aceprescal water table)

IDh Avg. Flew Pate (h), 6000.0 _nnmr.

S/G 'Iherml Ibe'er . Q = 1.9276 x 10-5 (M) (tq-hf) . ) Q = 1.9276 x 10-5 ( )(

s (!!et Heat Ircses)

AVEPAGE - 0.3% Blcutic'en Correction Famr - 0. 5 *. (QBD f rum O-6. 3.1) 's (100% = 1520 fG't) __M.<t CDRE 'IHEle'E IGOR __ II. DF:LTA-T 'Ilm'AL OE0;: H A IrOP F TI-407B F - TI-405B ~ O OF TI-408B P ~ TI-406B - 'Iha IDM flcs' reasuremnt dull only te u:nl wtwn t ecdsater RIin were tnd.

1U11 :

Obtain hg using asy. fewd temp only when steam table is tud.

If any chanyya atu r.nic to bicuiun ficv arnther caloriretric rhauld Le [erfomxl.

If the caloriretric is above 1520 rut, anl af ter unit puur rohrtion tus tscen perfomd, within appmxirately tw hcum, [nrfom anotJv'r caloriretric to verify the adtyncy of the Icad tutuction.

If a dirwJrceent exists, tutvcen the pwr twqe durv>!1s anl the caloriretric, adjust tiv1 q1iru of the IIIS FAer rany? dunmls, to aqne.

If the Reactar is at p%ier, with ore or rotu f'. tin In d.'ater pqn in rauvice P D II:: in a 24 h m r [ericd.

a Caloriretric will te lerfotnd at 1(unt orce RELDRD CNIID7RY 3.3.6 PEVIUul) BY: ,

- =. = - ._. - .. _ .____ -. . ._.- - - _. -. - ATTACHMENT B M 3811 < ' aW +d + %

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_ _ _ _ _ _ _ _ _ _ _ -_ _ _ __ l Page 18 Iz__ESINQ1ELgg_QE_NyQLg65_EQWES_g69NI_QEgGGIlGNg IUUSbQQyNgdiggi_bggI_I69NSEg6_ONQ_ELylp_ELQW ! , ANSWER 1.01 (2.00) y f. [ 2 citg4 - control banks rods together with no individual r od A-44H" f r om bank demand l position proper rod sequencing and overlap C4 X O.50? - rod insertion limits are not violated delta 1 is within required band - REFERCNCE 'Ta r k. : verb 3 lO.A*I RGLE Enabling Objective RHT12C-1.7 t RGLE Lesson Plan RHT12C page O K/A 173007 K1.05 3.1 ' 193OO9K105 ..(KA's) ANSWER 1.02 (2.00) ! - RCS subcooling CO.25] (areater than requirements of ES-1.1 figure)

based on core exit TCs (0.253 - S/G pressure (0.253 stable C0.125] or decreasing 00.1253 - RCS hot leg temperature (0.253 stable CO.1253 or decreastng CO.12D] E - core exic TCs CO.253 stable CO.1253 or decreasing CO.125] - RCS cold leg temperature (0.253 at saturation for S/G pressure (0.25] C4X O.503 REFERENCE RGLE Enabling Objective RHT11C-1.5,1.6 RGLE ECP ES-1.1 Attachment A K/A 193008 Kl.22 4.2 193008K122 ..(KA's) . . t i ! i I " (***n* CATEGOHV 1 CONTINUED ON NE*1 PAGE

- - - - )

! . I . __ _

_ _ _ _ _ _ _ _ _ _ 12__ESINClg(gS_g[_NUC6E@S_EghgS_[6@@!,g[gB@Ilgdt Page 19 ISESdQQYN@ digs _U[@l_IB@NS{[G_Q$Q_{(U}Q_{(gW t ANSWER 1.03 (3.00) a.

INCREASES CO.2S3 because boron will be removed [0.253 adding posi t i ve reactivity to the core CO.253 that will have to be offset by an RCS temperature increase (because rods are in manual) CO.2S3 b.

NO CHANGC CO.253 because the initial temporary increase in RCS temperature will cause S/G pressure to increase CO.253 causing the steam, dumps to open enabling them to remove more heat CO.253 causing RCS temperature to decrease to its initial value (0.253 c.

INCREASES CO.253 because the rod control system will respond to the increase in P (i mpul se) (0.253 causing rods to withdraP ladding positive C O. g 3 -caust nq ' hew. 7 tg -t ne rwse -E% 251-c++sul t+rtq-ttt-amt w S r eac ti vi ty W4 ~ ~E;C impe aw outscc-( e,, p e,%h CI /*' #j l ' ' ' C O 163 r REFERENCE

l

"I b l FN b '>d edom #5 M*' s;9.:.,ea e.s n u..,,La kw..aw ) a RG7.E Enabling Objective RRTOSC-S.S (c. ', c] RGLE Roactor Control - Student Handbook - pages 22,23,26,27 r/A 192000 K1.20 3.8 K/A 192006 K1.21 3.6 192OOBK121 192OOSK120 ..(KA's) (p3L, (d.i) /) 10 hD) CC VC).b{ C QL'F d - r ) 2 E A' /(.(~ld d(awd (dd.u e ANSWER 1.04 (2.00) g g., 2.g (Mulk Q ad , a.

VALSE L)1 s o tTC c'v 3 5) . ow l it J 'c 'g ~-9 ' b.

FALSE [4 X O.SOJ Y % J EG bil; brY Y'I c.

TRUE d.

TRLIE REFERENCE RG7.E Enabling Objectives RHTO7C2-2.4,2.S,2.9.2.9 RGtE Flutd Flow Text Chapter 6 pages 29,30,40 REG procedure T-BA, "Startup and Shutdown "A" and "H" Circulating Water Pump" V/A 191004 K l. OS 2. 3 K/A 191004 K1.07 2.9 K/A 191004 K1.12 2.5 K/A 191004 Kl.15 2.6 191004K115 191004K112 191004K107 191004KloS ..(KA's) . <*eoee CATEGO4v 1 CONT INUE D ON NEXT PAGE e****)

Pcgo 20 it__E61NQJg({Q_QE_NQQLg66_EQWE6_E(@NI QEg6911QNt ISE60duyNGUlQgt dg@I,169N@EEG,@NQ,E(ylQ E(QW ANSWER 1.05 (2.40) a.

MORE NEGATIVE b.

LESS NEGATIVE C4 X O.60] c.

MORE NEGATIVE d.

MORE NEGATIVE REFERENCE RGLE Enabling Objective RRTOSC-2.6 Student Handbook pages 15-19 RGLE Inherent Reactivities - - K/A 1o2004 K1.06 3.1 192OO4K106 ..(KA's) ,p # ANSWER 1.06 ~ ,_.. a-cr: 56i' REFERENCE RGLE Enabling Objective RRTOSC -3. 6 RGLE Lesson Plan RRTOSC page 45 K/A 192004 K1.07 3.0 192OO4K107 ..(KA's) l l l ANSWER 1.07 (2.25) a.

DECDEASE CO.251 moderator density increases (0.2S] neutrons travel less distance and are affected by control rods less (0.233 l b.

DECREASE CO.253 flux decreases CO.25] less neutrons area available to ! interact with control rods [0.233 ~ c.

INCREASE Co.251 boron neutron absorbtion decreases [0.253 more neutrons can interact with control rods CO.2S] , REFERENCE i

! RGLE Enabling Objectives RRTO7C-2.7,2.9 RGLE Reactor Control - Student Handbook - p a g ta s 6.17 K/A 192005 K1.OS 2.8 K/A 192003 Kl.06 2.6 K/A 192005 K1.07 2.5 192OOSK107 192OOSK106 192OOSK105 ..(KA's) (e**ee CATEGORY I CONTINUED ON NEx? I' AGE

- - - - )

( l-i

_ . __ Pcge 21 lt__E61NQlC([$_QE_NQQ(EOS_CQWE6_E(QNI_QEg6811gNt ISESO99yN@dlG$t_dgQl.16QNSEqG_QNQ,E(Q1Q_E(Qg ANSWER 1.08 (0.95) A CO.953 REFERENCE RGLE Enabling Objective RRT-04C-3.7 RGLE Reactor Kinetics - Student Handbook - page 29 K/A 192008 K1.04 3.8 192OO8K104 ..(KA's) 0,' ANSWER 1.09 (1.00) A:-

xLg 0>. - power history CO.503 OP2 qqpg \\k'l10:0&a - ,)p(4 E t' - time after shutdown (0.50] REFERENCE RGLE requalification Exam 3 Question 1.05 K/A 192000 Kl.27 3.1 192008K127 ..(KA*s) ANSWER 1.10 (2.40) A-8 ! Nonon decrease 9 due to Durnout (0.601 d.

ti.?e

time = B-C ! xenon increases due to an increase in todine production from the higher fission rate CO.603

b.

' 50 hours [0.601 c.

20% +/- 5% reactor power CO.60)

!

REFERENCE RGLE Enabling Objective RRTO6C-1,3.1.4 RGLE Poisons Reactivity - Student Handbook - page 4.7 K/A 192006 K1.11 3.I 192OO6K111 ..(KA's) , ANSWER 1.11 (1.50) 1200.7 DTU/lbm (0.50] Hg = 397.310 DTU/lbm (0.50) Hf = /VOC /IICW U l')) 92.I'/. TO.5 W r Core Thermal Power = t4e*6e CATE GORY 1 CONTINUED ON NExT PAGE *eeee> , t ...

_ _ _ _ _ _ _ _ _ - - i Iz_ EBINQJE(gS_QE_NQG(g85_ggWEB_E(9NI_QEgB911gN Pcgg 22 i ISESDQQyN9DJgh,UE@I_IB9N@E[6.9NQ,E(Q19_E(QW REFERENCE RGLE Enabling Objective RHT10C-2.S AGLE Procedure 0-6.3 Attachment III K/A 193003 K1.23 3.3 K/A 193007 K1.00 3.1 193007K100 193003K123 ..(KA's) ANSWER 1.12 (1.80) 4.

CALCULATED LOWER THAN ACTUAL b.

CALCULATED HIGHER THAN ACTUAL C3 x 0.601 c.

CALCULATED HIGHER THAN ACTUAL REFERENCE RGLE Enabling Objectives RHTOSC-1.0,1.9 Westinghouse Thermal Hydraulic Pricioles and Applications to the PWR 11 pages 7-67,7-68 K/A 193003 Kl.23 3.3 K/A 193003 K1.03 2.5 19300SK103 193003K123 ..(KA's) ANSWER 1.13 (2.70) ,y a psy n3 Y 'O C p u s line 3.4.3 +1050 pcm (0.25] Inne U.S.2 +325 pcm (0.233 line 5.6.3-2250 pcm (0.403 line S.7.5-9.94 pcm/ ppm (0.2S] line S.7.6 +1392 pcm (0.40] _ line 5.11.1 =11^0 pcr C O. 25 3-4 ( Q3 t ha t L E' d.

1 i ne S.11. 2 --550 p 1 E C. Z '.M - pG8 ge m CD M3 ' 1ine S.I1.4 M vpdor4tM-- - ( t C. n, dpW u. Y S,; . ~ . "*t96 4 g a l 1 on s 4 0.--25 3 ' makcup at er needed for dilution = ~ U [

' REFERENCE t.,2 7,w.u $ n h 'n 7 - - RGLE Enabling Objective RRTOBC-2.1.2.3 RGLE procedure 0-1.2.2 M '11'11I.ke, ptOTi.,C) , RGLE ECC Calculations - Student Handbook - pages 1,4 ' K/A 192000 Kl.07 3.S 192008K107 ..(KA'si s i (***** END UF CATEGORY

- see)

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

I ?z_.PL99I_Dggigy_lyCLyplyg_S9g[Iy_999,gDg69ghCy Pege 23 [ S!sIgng

l . > ! . ANSWER 2.01 (2.50) a. Olves a long coast down time on a loss of power (0.503 which ensures adequate heat removal (while decay heat generation is high)CO.SO3 b. ensures No. I seal faces do not touch CO.SO3 c.

upper oil cooler lower oil cooler [3 X O.333 thermal barrier heat exchanger + CCP b4LdA s:}O REFERENCE \\ RGLE Enabling Objectives R1301C-1.1,2.3.S.2 RGLE Leuson Plan R13OlC pages 9.11.14,15 K/A 003000 Kl.12 3.0 K/A 003000 KS.02 2.8 i: / A 003000 A3.03 3.2 ' OO3OOOA303 OO3OOOKSO2 OO3OOOK112 ..(KA's) ANSWER 2.02 (2.50) a.

REMAINS OPERABLE u. CLOSED c.

OPEN (5 X O.SO) d.

CLOSED e.

OPEN REFERENCE RGLE PLID's 33013-1231,1237.1264.1265.1266 K/A 078000 K3.02 3.4 076000K302 ..(KA's) . ANSWER 2.03 (3.00) a.

RCS purtfication (0.50] pressure control C0.503 . b.

RCS cold leg

f)C4 ed C (^CW; L'i fy func) C)t2, & b '(' ( tC r st i r . i c. prevents coolant flashi h at the letdown orifices [O.303 d. baron precipitation (which may block BA flow) [O.503 (***s* CATEGORY 2 CONTINUED ON NEXT PAGE

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- _ _ _ __ 3t.Q(@NI_DESJgN_lNQ(yplNg_gGEEly_QN9,gDg6@ENCZ Pcgo 24 l 915I500

1 ! REFERENCE RGLE Enabling Objectives R1601C-1.2,2.1.2.6 and R1801C-2.1 RGLE Lesson Plans R1601C pages 7,9,203 R1801C page D , K/A 004000 K1.17 3.4 l K/A 004010 K1.01 3.4 K/A 004010 K4.02 3.2 K/A 004010 K4.03 3.1 OO4010K403 OO4010K4C2 OOA010K101 OO4000K11/ ..(KA's) l l ANSWER 2.04 (3.00) l l a, condensate storage tank j service water system C3 X O.503 ! city firewater system l el e c t r-i ca l--overspeed-EO.-254-++s et s-nit t ex w th s Hv-{@r253 - b. mechanical overspeed f %254 must be manually reset (0. 253 [0,S c] c.

REMA)N OPERABLE t0.503 Lo 5 32 REFERENCE RGLE Enabling Objective 9 R4201C-3.,12.,16.

RGLE Lesson Plan R4201C pagon S,6 REG procedure ER-AFW.1 aage 2 K/A 061000 K4.01 3.9 K/A 061000 K4.07 3.1 - K/A 061000 A2.03 3.1 061000A203 061000K407 061000K401 ..tKA's) ANSWER 2.09 (3.00) - SI 1 A 4 0. 5G4 and 1C fOv503 . - RHR 1A LO.SG) q b~ c d O' 'J - J SW IA (or IC) EO.500- - MDAFWP 1A (0.S0} - -NkP--f C; 50 ) - REFERENCE RGLE Enabling Objectives R2701C-2.3.2.4 RGLE Lesson Plan R2701C page 19 K/A 006000 K2.01 3.6 K/A 006000 A3.02 4.1 0060004302 OO6000K201 ..(KA's) (***** CATEGORY 2 CONTINUED ON NEXT PAGE 4e8**1

l_

7t__P(OUI _Q(gigy,10C(UglyG,QQ((I!_GUQ_E[ERGEUCY Page 25 SYSIE05 ANSWER 2.06 (2.00) a.

remove heat CO.233 to maintain containment preGsure below design limits CO.25] AND reduce containment iodine inventories CO.2S) so that of f-si te radi ation exposures are within 10 CFR 100 guidelines (0.253 h. 4-1 i. qui d-) ot -oduc. tor CO. 2SJ-4 s-used-to-di-ve*4-*1pm*y-puep-d behega f i c.. [4 452 back ts-th=i p o g, ', suu t a w.

EG. ;) v4a ac line (v.;53 2/M ? $$CWI CO.%?2 G3 via 3 &:; y'Ts' y 5.

r REFEREN_CE pr414) { tL'y ^t4) '% M.,) L iL%j [o. "Ln} x c, x,g3z,4 kp fjf . RGE.E Enabling Objectives R2401C-1,,2.

_ b dj' L / 'O ' ' ' / , ' RGLE Lesson P1an R2401C pages 3.4,O G4 " k/A 026000 K4.04 3.7 . -. - 3 "' '" LC' K/A 026000 GO.07 3.5 026000G007 026000K404 ..(KA's) , ANSWER 2,07 (3.00) ,

.o h n..n, ;.O.

k e U h D, N '1 Y J ' i.4 Containment purge supply untts trap . containment purge exhaust units trip containment depressuri.7ation valves close C4 K O.375) {^. a g

"f c ont ai nwn t air test valves close J

k " fl~2 'd "y containment RMS sample valves close f other various valves and damper 5 close s b.

aux 111ary butIding exhaust fan (IF) trips intermediate tJulIding ehhaust fan lA or 18 trips C4 i O.375) i n t er rr e d i a t e butlding eshaust fan 1C trips r adi oac t i ve gas r el ease vent valve ROV-14 cicses REFERENCE RGF.E Enabling Oujective R3901C-3.2

RGLE Lesso, P1an R3901C pages 11,12 K/A 072000 K2.02 3.S K/A 072000 K4.01 3.3 K/A 072000 V4.03 2.2 072OOOK403 072OOOK401 072000K202 ..(KA'i) r t***** CATEGOR4 2 CllNTINOED (~1 N NExi PA:T

- e**i

_ _ _ _ _ _ _ _ _ - _ _ _ _ _ 7t._P(@UI_QgglGG_lNQ(QQ1NQ,$QEgly_QUQ,gDEQQgNQY PCQe 26 i UYSIEd5 ' ANSWER 2.00 (2.00) b.0cphont.theo n a.

relay room computer room C3 X O.333 technical support center b.

to provide adequate pressure CO.25) to the highest norrle CO.25] c. undervoltage E6-2&L , SI Cer25-1 L2 y o, g,g] 994Av4444eVI l REFERENCE.

RGLE Enabling Objectiven R5901C-2.0.2.9.2.10 i RGLE Lesson Plan R5901C pages 14.17 r K/A 086000 K4.02 3.0

K/A 086000 K4.05 3.0 ' K/A 086000 GO.07 3.0 - 086000 GOO 7 OB6000K405

086000K402

..(KA's) f ANSWER 2.09 (2.00) a.

- RCP therm) barrier excess letdown heat oxchangter - ' CVCS non-regenerative heat exchanger - CCW surge tank make-up valve. AOV-823 C6 X O.251 - a , sample heat exchangers - - RHR beat exchangers i - RHR pumps b.

surge tank level (0.253 radiation monitor (PMS-R-17) (0.25) [ f REFERENCE i

RGLE Enabling Objectives R2801C-1.2,4.1.4.2.5.1 RGLE Lesson Plan R2901C page 11.13 x/A 000000 K3.01 3.4 K/A 008000 A2.02 3.2 OOGOOOA202 008000K301 ..(KA's) toee** CATEGORY 2 CONTINUED ON NEXT PAGE eeeoa)

2i__P6@NT_QE@lgN_lNQ(yDlN@_gAFETY_AND_EMERGENGY Page 27 S_Y_S_T_E_M_S_ ANSWER 2.10 (2.00) 4160 VAC bus 12A

station servive xfmr CO.30] No.

14 CO.203 ""' ~~~~~

-480-V AG-{ O. 30 bus 14 6 t'o.So]

l___:

MCC [0.301 IC CO.203 -__

l l-----l ' ' , . !

battery charger [0.303 1A [0.20] _____ l l 125 VDC bus 1A

REFERENCE RGhE Enabling Objectives RO701C-1.2,4.1; RO9010-2.1 RGLE drawing's 33013-623,652,756 K/A 062000 K1.03 3.5 062OOOK103 ..(KA's) . .. , (****4 END OF CATEGORY

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3:__INSIBUDENIS_@ND_CQNIggLS Page 28 ANSWER 3.01 (2.00) a.

SR.0-0,35 3 IP C O. 2 5 7.

@,~-y b.

SR C O. 2 5 3-(0.6(J] c.

I R Hb-250- PR +O c-264-CO. 6 o') d.

SR -eor-254 IR EG,-254-PR -00r-25-1 (C>.603 REFERENCE RGLE Enabling Objectives R3301C-2.2,2.3,3.1 RGLE Lesson Plan R3301C pages 7,8,10,13,14,15.18,21,23 K/A 012000 K6.04 3.3 K/A 015000 K5.01 2.9 K/A 015000 K5.02 2.7 015000K502 015000K501 012OOOK604 ..(KA's) ANSI'ER 3.02 (2.00) o'rST RCP bus underfrequency RCP bus undervoltage RCP loop low flow [4 X O.50] RCP breaker open / c> Cu P2 V.

g,pw g REFERENCE RGLE Enabling Objective R3501C-3.1 RGLE Lesson Plan R3501C page 20 K/A 012000 K4.02 3.9 o,QU, 012OOOK402

..(KA's) c p, > > cT LTT a-ANSWER 3.03 (3.00) ()d {,M ) '

0964 a.

outward motion i. prevented [0.25] due t o t he -Pf4 r od stop (0.25] at 103% CO.25]' defeat failed channel CO.25] (p2) b. outward motion is prevented LO.25] due to Pimpulse CO.253 less than 12.8% CO.253 take manual control o' rods [0.25] c. all motion is prevented CO.251 due to Tavg - (Tavg) avg CO.253 > or equal to +/- 4 degrees F CO.253 defeat the failed channel [O.25] C. k h D i2.

CCLt intob OL la [' # to ou m o n t ca.3C c Vn*+ Co. 'K3 ' ) Y 'A' v C ' dry n4 CC M (4**** CATEGORY 3 CONTINUED ON NEXT PAGE * ****)

3:__INjlBUdENIS_@NQ_CQNIBQLg Page 29 . REFERENCE RGLE Enabling Objectives RIC01C, RIC05C, and RIClOC-1.2.1.3 RGLE Lesson Plans RIC01C pages 5,6; RIC05C pages 7,8; RIC10C pages 5,7 K/A 001050 K4.01 3.4 OO1050K401 ..(KA's) C2 M ANSWER 3.04-42r76+ a.

r-ea+:-tW r-i-p (P d' E O. 50 2 - (60 second) time delay has timed out [0.50] b.

te return control of safeguards equipment to the operator CO.50] c.

Iow pressurizer pressure C ^ '?S 3 Cp.3 03 low SG pressure C ^. 25 3 = [b.Ce>] both car bc bi cc' ed CO.252 when 2/3 RCS pressure channels [0.231

< 1992 psig CO.25] . . ('o,po} REFERENCE RGLE Enabling Objectives R2701C-3.1,3.2,3,3 RGLE Lesson Plan R2701C page 17 K/A 013000 K4.01 3.9 K/A 013000 K4.12 3.7 013OOOK412 013OOOK401 ..(KA's) F ANSWER 3.05 (3.00) ' b.

L c.

c.

1,3 L12 X O.25J 1,, 2 e.

4., - 5 ,q'

' / h.

1,3 / REFERENCE RGLE Enabling Objectives R4501C-2,1,2.2.2.3,2.4 RGLE Lesson Plan R4501C pages 5,12,13,16,17 K/A 041020 K4.09 3.0 K/A 041020 K4.17 3.7 K/A 041000 K4.18 3.4 041020K418 041020K417 041020K409 ..(KA's) (***** CATEGORY 3 CONTINUED ON NEXT PAGE 4****)

3a__INSI6UdgNIS_AND_ CONI 6QLS Page 30 AMSWER 3.06 (2.50) - steam generator level control pressurizer pressure control pressurizer level control C5 X O.50] - rod control - steam dump (RCS temperature) control REFERENCE RGLE Enabling Objectives RICl2C-1,1,1.2 RGLE Lesson Plan RICl2C page 6 O.

b Ig '"b Ah).hJ'6g#g gY f, /. ('o, gg K/A 000057 EA2.19 4.0 ., ( OOOO57A219 ..(KA's) .- ._ -p,. ipj trJ1) . or byt&WS. b) Co.?o]

7 4D I,"

t ' / ' -g) 'M ANSWER 3.07 (2.50) H igd #b ' a.

- each channel has a level trip bypass switch CO.50] to prevent _a-trip p ' 122'~~ trip logic at low power levels (l ess than 10% or 25%) due to the scheme (0.25] p_ - - at high power l evel s (>25%), the t r i pyarClil oc k ed CO.25] by the P -- 6 CO.253 or P-10 permi ssi ves Jor25']

'emoves the overpower rod stop b. - the "Rod Stop Bypass fSw'-h t

function for the set'ec t ed c h an n el CO.25] to allow rod motion CO.25] - the "Pow smatch Bypass Switches" remove CO.25] the selected ,ghar1Tnil from the rod control system averaging circutt (0.25] M ' /U N -'k k dM## - REFERENCE b, TO U) l'7 -;).O - N~ ybA& l d* '.> l LlU CD. 3:>J . g x RGLE Enab1ing Objectives 3.6,3.7,4.1 ('M ) . RGLE Lesson iext RGE-33 pages 5,9,13,18,23 ',7 79 d,,, g .gg L,D K/A 000033 EA1.O2 3.O r - K/A 000033 EA2.09 3.4 ,.y ld C o,Jo]

,g , OOOO33A209 000033A102 ..(KA's) ]lge.d U/q > M h ANSWER 3.08 (3.00) L M4 J Ab) A blf] I ' ( dt/0 TH(/J) a.

Tavg +auct hi gi : CO.50] b.

no load -- 19.5% CO.25] full load -- 49% CO.25] c. High PZR level trip CO.50] 87% CO.25] with 2/3 channels CO.25] due to letdown isolation CO.50] with charging pump flow CO.25] -a t-- -mi n i mu m-so a w4 r N_25 L.

CA2 Gddl) L 'V.}$ b , (***** CATEGORY 3 CONTINUED ON NEXT PAGE

- - - 4)

,

. 3:__1N@I690ENI@_66D_ CONI 60L@ Page 31

REFERENCE RGLE Enabling Objectives R1901C-1.2,3.1,4.1 RGLE Lesson Plan R1901C pages 14,13 RGLE Simulator Documentation No.

6.3.4.11.3 K/A 011000 K4.04 3.0 K/A 011000 K4.05 3.7 K/A 011000 A2.11 3.4 K/A 000028 EA2.02 3.4 K/A 000028 EA2.12 3.1 011000K405 011000K404 011000A211 0000284212 OOOO28A202 ..(KA's) ANSWER 3.09 (2.50) a.

OPEN b.

CLOSE c.

OPEN [5 X O.50] d.

CLOSE e.

CLOSE REFERENCE RGLE Enabling Objectives R4401C-3.2,3.3 RG&E Lesson Plan R4401C pages 5,6,7.8 K/A 035000 K4.01 3.6 035000K401 ..(KA's) ANSWER 3.10 (1.75) a.

inverter output voltage will drop 10.25] causing auto static transfer to alternate power supply CO.25] b.

steam supply valves (MOV-3504A/3505A) fail as is (c l osed ) LO.25] feedwater pump discharge valve (MOV-3996) fails as is (open) [0.25) EO.20] u auniiig c.

ewiv-t - i ;+ ccE ;- 520 Opens (0.75 32A up..ia ._ y g o n .

+ r eac-t or-t o ip 4 0v 2Gb ] g r,c } .,gj .g,

, ,

J REFERENCE (, , ), g d Co.'f o] ,;aMbij G 6cdfC / 16) Cc.usd' RGLE Enabling Objectives RO901C-2.4, R3501C-2.1, R4201C-9.0,12.0 RG&E Lesson Plans RO901C pages 4,5; R3501C page 14; R4201C pages 5,6 RGtE Lesson Text RGE-9 page 4 RGLE drawing 33013-756 K/A 000050 EA2.03 3.5 OOOO58A203 ..(KA's) (***** END OF CATEGORY

- - 4*)

f

_ 4:__EB9GEggBES_;_NgBd861_9BNg609L _EUE6GENQY Page 32

GND_B9 dig 6gGIg86_ggNIBg6 / _ANGWER-M.04 ( 2 rOO) y' cduce the pcccibil.ty CO.25] of watw-hamma [0.13] buuuuwe uf wivam EO.dO] -ecte-ing the (nedring rn sn1 REFERENCE RGLE Enabling Objectives ROPO3C-1.1 RGLE Lesson Plan ROPO3C page 4 RGLE Requali f ication Exam No.

3, question 4.07 K/A 000054 EK3.03 3.8 OOOO54K303 ..(KA's) ANSWER 4.02 (3.00) a.

either S/G pressure decreasing in an uncontrolled manner CO.50] either S/G completel y depressurized CO.503 b.

- close or verify closed MSIV and bypass valves - isolate MFW flow - isolate AFW flow - close steam supply valves to TDAFW pump - verify S/G ARVs closed C5 X O.40] - verify S/G blowdown system isolated - isolate sample system - isclate steam support heating - isolate steam to upstream traps REFERENCE RGLE EOP E-2 pages 3,4 K/A 000040 EK3.04 4.5 K/A 000040 EA2.01 4.2 ' OOOO40A201 OOOO40K304 ..(KA's) .. (4**** CATEGORY 4 CONTINUED ON NEXT PAGE

- - - 4)

_ _ __ ___ __ - 4:__PBggEgyBEg_ _UgBD@61_@@NQBU@b1_EDERgENgY Page 33 ONp_B8 pig 69@lC@6_CQNIBg6 ANSWER 4.03 (2.70) a.

- one train of reactor trip breaxers open CO.30] - neutron flux decreasing CO.30] b.

- annunciator A-26 actuated CO.3OJ - all CI status lights on MCB bright CO.303 - train A and B XY rel ay lights deenergized CO.303 c.

- MFW pumps tripped CO.303 - MFW pump discharge valves closed CO.303 d.

- verify SW pump running CO.303 - verify SW header pressure greater than 40 psig CO.30] REFERENCE RG6E Enabling Objective REPOOC-1.4 REG EOP E-O page 4-9 K/A 000007 GO.10 4.2 OOOOO7G010 ..(KA's) 'b eb ddd , D, 'gg) ANSWER 4.04 (2.00) , aw - f(ki) ~ a. manually insert the control rods C1.00] DS ' b.

shut the main steam isolation valves CO.50] and bypass valves CO.50J REFERENCE RGLE Enabling Objectives RGLE EOP FR-S.1 page 3 K/A 000029 GO.10 4.5 000029G010 ..(KA's) . (***** CATEGORY 4 CONT INtJED ON NEXT PAGE

- - - - )

I

4-- 88ggggUBES_ _NQ80961_8BNOBMG61_EDESGENgy Page 34 BNQ_86D]QLgGigG6_GQNIBQL ANSWER 4.05 (2.50) 55 REM CO.253 a.

5(N-18) = Total lifetime to date = 53 + O.5 = 53.5 rem hb. 25'] Total lifetime available = 55 - 53.5 = 1.5 rem -[0.253-Total this quarter available = 3 - O.5 = 2.5 rem [O.253 gamma 850 mrem /hr x 10F = 850 mrem /hr [O.253 th neut 100 mrad /hr x 50F = 500 mrem /hr [0.253 beta negligible due to protective clothing [0.253

1350 mrem /hr total dose rate total = 1.5 rem 50.253/(1.35 rem /hr) ' 67 min [O.253 b.

25 REM [C.503 REFERENCE RG&E Enabling Objectives RRCO3C-1.6; RAD 621-2.5,2.6 RGLE Lesson Plan RRCO3C page 6 RGLE procedure A-1, "Radiation Control Manual" page A-1:13 10 CFR 20 K/A 194001 K1.03 2.8 194001K103 ..(KA's) ANSUER 4.06 (1.75) _ og w,& '8 ejeckorfDuj 7,Wnw %.f . k;d - condenser ai radiation abnormal - S/G blowdown radiation abnormal - S/G level increasing in an uncontrolled manner [5 X O.351 - S/G steamline radiation abnormal ' - S/G chemistry samples indicate radiation abnormal - 0 'A ; REFERENCE _ y> M V ofuk;t' thlCAL+LV c ! RGLE Terminal Objective REPO3C RGLE Lesson Plan REPO3C page 3 EOP E-3 page 2 K/A 000038 EA2.02 4.5 OOOO38A202 ..(<A's) ! ! (***** CATEGORY 4 CONTINUED ON NEXT PAGE

4**d)

l , ,

s Page 35 3:__E69gggyBg@_ _UO8D@61_@@NQBD96t_gDERGgNgY BNQ_6@ pig 6QGJg@6_ggNI6QL ANSWER 4.07 (1.25) M If b-2 [O.203 RCS PORVs 00.30] set tc 'ZM psig (0.25] ' - an open RCS vent path CO.503 . .) g .,gg ( ' , - - REFERENCE LA do RGLE Terminal Objective ROPOOC REG Procedure 0-1.1E page 0-1.1E:1 K/A 002000 K4.10 4.2 OO2OOOK410 ..(KA's) (/.6Ch ANSWER 4.08-(-3 TOM-L'o 3 5 ) ( o. 3 5) ~ a. connect a length of fire hose FO.2G) from fire hydrant 444-Mril&l to the D/GW the- -Icce service water semplv v a ' v r> t V-ft66W M1-7G+-e n d--o&* " N" LO ' Co 303 f.ir mwat.ec--supp l y ea1s.

W--46MF '. L G. 231 b. place the START /STOP switch to STOP [0.50]

~

~'

O ' c.

governor controls load CO.50] , .,s voltage control regulates power factor CO.503 ""RN1[A3D h&} ' W 3 9-REFERENCE RGLE Enabling Objectives RO801C-6.2,7.7 REG procedure ER-D/G.2:2 RGLE Lesson Text RGE-8 page 9,12 K/A 064000 K1.02 3.1 M/A 064000 K4.02 3.9 K/c 064000 A3.05 2.G (2.00) - ANSWER 4.09 (2.00) a.

realign the SI system for Cold Leg Recirculation [0.50] jgt-J,, q b.

if an intact S/G is not available (0.50] , MWM b, m'_ c.

limits damage to the RHR pump seals EA.-ba(0.'A6) .Dt .O ' ^~ d.

to establish core cooling (flow) [0.503-i L eor e-e*+ t -T EH Or254-emge t*a ter --th an--l W.++-F. :C.DE' f

l l (***** CATEGORY 4 CONTINUED ON NExT PAGE

- - - - )

_ - 'St__88QQEQQ8gQ_ _UQ8d@(t_@@NQBd@(t_QUg8GENQY Pago 36 GNQ_8@Q1Q(QGiQ@(_QQNI6QL REFERENCE RG&E Enabling Objective RFRCIC-1.2 RG&E Lesson Plan RFRCIC pages 2,3,4,6 K/A 000074 EK3.11 4.0 OOOO74K311 ..(KA's) ANSWER 4.10 (1.50) 3.

1.

5.

[ 6 )( o.30] 2.

nx c.252 4.

6.

REFERENCE RG&E Enabling Objective RFROOC-1.1 'RG&E Lesson Plan RFROOC pace 9 K/A 194001 A1.02 4.1 194001A102 ..(KA's) ANSWER 4.11 (2.80) a.

- control bank at or approaching the low insertion limit - control bank at the low low insertion limit - an uncontrolled cooldown of the RCS following a RX trip C4 X O.40] - an unexplained or uncontrolled reactivity increase - one or more control rods not known to be fully inserted following a R) trip b. (LCV-1128)gur from RWST (V-350 BAST to charging pumps C3 X O.40] (V-358)fWST9;sypgS >& gin,g pumps to char (ft(/-IICf) Wn f REFERENCE RGLE Enabling Objectives RAPO6C-1.1,1.2,2.1 REG procedure AP-CVCS.2 pages 3,6 K/A 000024 EK3.01 4.1 K/A'OOOO24 EK3.02 4.2 OOOO24K302 OOOO24K301 ..(KA's) (***** END OF CATEGORY 4 *****) (********** END OF EXAMINATION

- - - - )

i' ', v;. : - . .. .~ .--., - - - - -... - - - - . - - .n ,. . - ' . f Coryl 'o f. ~ /\\ &acLAe. ( )

W W

'"'N - e c.cro 6 +. h.,by . - , U. ' S.

NOCl. EARL REGill.A10RY COMiIISSIOli.. ~ ' SENIUR REACTOR OPEknlOR lilCENSE EX AMINAT ION F ACllil 1 / ..G lilN().. _ _ __ _.

,, REACTOR TYPE: .f HR --W EC R...... _,, _.. _ _ DAT E ADf1INIS1ERED:,l87/J O/05. - ,. _..

EXAMINER: ,.M I N G St.E L, I. . CAND1DATE: _ _,, _ _ _ _ _ _ _.. _ , _.. _,,, _ _ _. J tGIBUCIIOtg_,1O_G8tLDIDATE,: Use separate paper ior i he ensegers.

Nrite answers on one side on1v.

-Staple question sheet on top of the ar.swer sheets.

Points.for-each .questi on ar e i ndi c.ated i n ' parentheset after t ite quest ion.

lhe pessino grade requires at least 70% in each category arid a final grade or'at least 80%. Exarni nat ion papers ni11 be picked up six (6) hours e f t er the examination starts.

. Ot

, CATEGORY .%.OF CANDIlnIl'5 CIVI t-! suR, ~ M.6l.d@-._10101.

_,_q U1.4 _ ,tjeUA .. _. ..,_,,GOl E Gptik _ 22. so ..;; . =- 2% 00 . h.. I t thuHi UF NUCl E AR POWER Pl niai ' OPERAI luf t.

Fl.l!! OS, ni lt'> l i tbhl iUlh l!A111 Lb 13.0 0 l _....._. _.20.= $'.0 , 6.

PLAWI SVS IEl15 uESION. CONTRO: s AllD INST RUMEtlI A'l IDH 14*Go . F HUCEutIRES - Nt1HfiAl. nbNi.*Ri ini.. - i.erf 25.90 _.

.. .a :a.

EMERGEllCY AND RADIO (OGIGAt_ ' -COrilhul.

14 50 . _ _ ~ N. 25. Cf2 , _., __.

.. _ .. _ G.

Ahl11N1STRATIVE P H O C E D U Rii u. " CONDITIONS. AftD LIhllATIONS , 94. oo - ! I'~.'."""" .___.._ __...._..% Iot415 l . Final Grade , I AlI work done on thln eraminot2on is ray own, I have neither given ' nor received sid.

. . - . .. ... - -. .-

can n aat e's sionatore , Imb,9ery vab.t. el. 9,g A,_ - [ 1* e4 44., e f pr.n%. L

. NRC RllLES AND UUIDELINES FOR LICENSE EXAl11 NATIONS Du.-i ng the administration of this examination the following rules apply: -1.

Cheating on t he exaroi nat i on means an automatic denial of your eppli cat i oni and could result in more severe penalties.

2.

Restroom trips are to be liinited 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 cheatino.

3.

Use black ink or dark penti1 on1y to feci1itate 1egible reproductions.

4.

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

5.

Fi11 in the dat e on the cover sheel of fhe examtnstion (1+ nec ut-s or v ). 6.

Use only the paper provided for answer s.

7.

Print your n arie in i he opper rloht-hend corner of the first pov o4 e rm h section of the entiwec sheet.

Es. Consecutavely nonibor rech a%wer sheet, wi11o "End o+ Lotennrv " - appropriate, start escl. coteaur ois n new peye, write onl y un one unde of the paper. and e:r> t o- "t a s P.. q m " 04, the 1.. v t a n s.or sheet.

9.

Nuinber eact enca*et at t o cat ence s enJ nimber. i ci-example, 1.4 6.3.

J0. Skip et l ea s t tirer lanos bot i e. r ah m uu er, f 11. Separat e an s wer theet-4: em p m; a n o p i ec.e 42n)s.htd en v ue.

sheets d ec(- z down on your desk or table.

12. Use abbt evi ations on1' if *sev .w e conwo::1 s used in 4 sci 1)t' I ii et el pi _ 13. The point velvr for each question is indicated in parentheses atter the question and cen teowd * a nolde for t he dept h of a n s wei reuoared.

14. Show all calculations, methoos, or ausump ti ons used t o oL. L o a t. oli ensoer to mathematical problems whether Indicrted in t t ee oues t i er. x not. 15. Partial credit may be gi ven.

Therefore. (4NSWER ALL PARIS UF T HE.

C!UES1 ION AND DO NOT LEAVE Ain ANbWER E: LANK.

16. If parts of the examination ar e not cleer as to intent, ask quent2cos or the enminor on1y.

17. You must sion the statement on the cover sheet that indicetes i hei tta> work is your own and you have not received or been Q1ven assastence lei completino tho en eaa nsi i nn.

l hi a.

most be done e f i er the ex ana ris t i co ime been comp)eted.

i 18. When you complete your ex a nii n a t. i on, you s tia11 : a.

Assemble your examination as follows: (1) Exam questions on top.

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

(3) Answer pages including floures which are pert of the ensamr.

b.

Turn in your copy of the examination and all paces used to answer the exam) nat ion quections.

c.

Turn in all scrap paper and the balance of the paper ttne t you did not use f or answering the questi ons.

d.

Leave the ey aminat J on erce, os defincd by t he ex anii ner. 1+ a f t et l eavi no. You are found in th)s eres wh11 o the en eo.i nat i on 2e st)11 i o pr oqr ests, voor 1) cense uwv be den)ed or tevoked.

( .

_IMEOBLOE_NVCl EOR POWhiR_ F%,Grir.OPERATIOT4 r ELMID5, AND PAGE

JHER[10D,YN9(IICS QUESTION 5.01 (2.00) One RCP trips at 30% r ated thermal power without a reacter, protection system actuation or a change in turbine load.

Indicate whether the following parameters will INCREnSE, DECREASE, or ret 1AIN lHE Snlm.

As sure no operator action.

a.

Flow in the operating RCS loop b.

Reactor vessel Delta-P (0.5 esch) c.

Core Delta-1 d.

The operatino 1000 s t. o wi yereretor e t e -. m fIor OllESTION 5. O.3 Q.00> The r eac tor is sober)Lical be 4. i *.* p c in. l oe courit r ate t u lib cus, u t t> r a posi t i ve> r eac t i vi t " ) r c er t i on.

tho c n' in t r c i r:- s iicr ees ras to M fi.

4. -

in, reactiv)tv was added tu i h e-c or r> a.

1670 pcm 'b.

l'/SO pcm c.

1824 pcm d.

1892 pcm OtJESTION 5.03 (1.Oni During a reactor startup, the +2rst react 3vity eddi t t ori caused cou.d ret.

to increase from 100 cps to 200 cps.

lhe second reactivity addition ceused count rate to increase 4 tom '.'00 c p E to 4uO cpe.

Wl:1cli DNE of the f ol l oe)1 ti.. ctatements is CORRECT? a.

The firrt reactivtv eddi t t ore wss l er oer. b.

The second reactivity addition was larger, c.

The first and second reactivity addi t i ons were equal. d.

There is not enough dats 01ven to determine the relatiorishap bet wet n r eac ti vi t y ve l o rn. (*44*4 C 61 E Gf iP Y O'r CO' f li4UED CN NLi/,1 Fabt 44444>

b.

THEpBLOF,.NLICLE88,.E'OMEB..PL8NLOPER8T I 091_ ELU 1 DS,,,8ND PAGE

ISEBMODYN8MICS QUESTION 5.04 (1.50) Would the actual critical rod position be HIGHER THAN. LOWER THAU. or THE SAME As the calculated Estimated Critical Rod Posit 2an (ECP) for a startup 15 hours a+ter a trip if the following events or cotidi t i ons occurred.

Consider each independently and assume all other plant conditions are unchanged.

(0.5 esch) a.

One Reactor Coolant Pump is stopped one minute prior to critical 11v.

b.

The steam dump pressure setpoint is increased to a value just below the code safetius setpoints.

c.

The startup is delsved 2 more hours.

QUESTION 5.05 ( 1.S0) The plant is operat.inq at 100 percent ptwer with RCS leve at S/3.u. thcos ees F and a steam pt essure of 730 psto.

Uhat... u n t.

1N>E be chanced to in er rp. to maintain these conditions.nih 10 percer:t of the tubes 2n exh v i e on oenerator pluoced? Assume PCS flow rate is constant and no tubes pluow! inittal1y, a.

574.5 degreen F b.

576.5 degreer F c.

578.5 ocoroes F d.

580.5 degreet F QUESTION 5.06 ( 1. '-50 ) In order to matntain e 200 deoreet F tiibcoo12 no mer o2 ri in t he h. b wiin i reducing RCS pressure to 1600 pulg, steam generator pressuru muut be reduced to appr o>t i ma t el y a.

405 psig b.

325 psio c.

245 psio d.

1('. psio . (444it L A T EGOf,Y M LUlll ll4UL D ()l4 lie > I P6bE 448A4

h_. TtLEQBY. 0F_.[MCLE9E'._ POWER Pt Oty _pPERAT ION,. FLUI DS, AND PAbE

IHER!10DYN9MICS . QUESTION 5.07 (2.00) Will the Departure f rom Nuclear Boiling Ratio (DNBR) INCREASE, DECREASE. or REMAIN THE SAME if the following plant parameters INCREASE durino power operation? Consider each perameter independent 1y.

(O.5 ecch) a.

Reactor Coolant System (RCS) Pressure b.

RCS Temperature c.

RCS Flow d.

Reactor Power QUESTICN 5.08 (1.50i The reactor is :ir oduc i r m Iv0 percerit r ouet wtn n 1 st at s on blact..ot ve t or =. . Naturel circulot2on 14 e e_ t c h i i e md er,a u or c: delie-I goes t. o zio l. A de, heat i <: 2 portrot poer. .sh a t >s 'ho c o> e m a s s-flow rete; a.

2.7 percrrnt b.

3.2 percent c.

3.7 percent d.

4.1 p r> r c en t DUESTION 5.09 (2.no) Answor the iolIc uino u v e:_t 2.;r.t i RUi. or Fe1se: w. 5 e.. c i.1 a.

If the flowrate through on operating RHR pump INCREASES. the pumo's (6t Posit)ve Suction Heed will DECREASE.

b.

If a condensate pump is operatino at RUNOUT conditions, motor current will indicate LESS t h a n n r. r re a l runnino current.

c.

When starting a condenser circulating water pump, motor startino current i s REDUCED t,y CLOSING the pump's discheroe valve ptior to starting.

d.

If a centrifuge 1 pump't speed is DOUbi_ED. Its flow will DuU ut E.. (***41 CnTEGliRi W CONTINUED ON NE)i F4bE 44444>

- Us. _ItjEQBY_QE_UUCLEOB_EQUES_ELOULDEEROIlOUn FL_ UIDSdlJD Pi40E '; IUEBUQQYUQUICS QUESTION 5.10 (1.00) Within minutes followinq a reactor trip f rom f ull power. SUR atta2 ris an equilibrium value of

p-mv i ma t el y -1/3 dpm.

This SUR is the result of: a.

the shortest-lived group of cielayed neutron precursors, b.

the longest-lived group of delayed neutron precursor s.

c.

spontaneous fission of U-235.

. d.

spontaneous fission of Pu-239.

DUESTION 5.11 (1.50) What will r eact or pove-t r-oilcr m i r a i t e.c c 14 the t eac t or is placed cu c .75 dpm SUR at an inittsi power of 10 E-10 arnos ? a.

3.16 " 10 E o anip e b.

5x 10 E -9 emps c.

10 E-O a n.p =. d.

5< 10 E-8 ao.ps-DLfESTION 5.12 t.* Which ONE of the f011owino c_r e ..a l tio +actore alone wuoid cuuse ihe 4. iel temperature coefficirot to become 1esu neqattve c a.

Fuel dens) f i c at) on b.

Clad creep c.

Pu-240 buildup d.

Fission oreduct oss bui1dop 2o fuel /c1ad oep i l

(*94** CATEGDFn 05 CliNTINUED ON NEAT PAbE 44414) , I

h_._IMEOBLQE_UUC; E AB,,POljES, EIANLORERGTIOW FLLH DS,, Al@ PAUE o IUE6dQDyUGd.ICS QUESTION 5.13 (1.00) How does total power coefficient very at end of core life as power is increased f rom 0 to 100 percent? a.

Increases continuously b.

Increases, then decreases c.

Decreases continuous 1v d.

Decreases, then increases OUESTION 5.14 (1.00) Folicwing a e ecct or irin t r :.ce, 1<O pa cent puou. A u ion c onu:iit r oi a v., o11 peak in approx i mat el y houts.

a.

5to 7 b.

8 to 10 .c.

11 10 15 d.

14 to 16 OUESTION M.15 ( ?. CM How dor s-ove)lable Shu t do.n > f i-o o i ri chance ilNGRhASES. DE CPti;6 E i. n E11hilld THE 9AME) for the f ol l oon no con._ii t 2 co s with no operstor a c t i oit-: ( t o. i t< w s stated)> Assume the plant ac i n; t.3.41 1s at 50 percent pouer eiid ell c oo t t systems are in (1ANUAL.

(0.5 each) a.

RCS boron concentration decteuses IS ppm.

b.

Control rods are manual ly inserted 20 steps.

c.

A steam leak occurs in a main steem 12ne equivalent io 1 percent rer li., power (disregard effects of Xenon).

d.

Fuel deptetion and + 1 ssi ori produc t butIdup.

(*4*** CATEGORY 05 COMIINUED O4 l iE ( ) PAoE 44414)

Et_._ISEO8Y_DE_NUGLE68_POWE8,_E60NLDEEB01 ION, _ELtHDS._ AtJD PAGE

IUEBOODYNatlicS QUESTION 5.16 (1.00) The plant is operating at 50 percent power with all control systems in manual.

Assuming the following initial parameter values, what must final RCS boron concentration be in order to lower lave by 4 degtees F? Initial RCS Baron = 600 ppm Total Power Coefficient-20 pcm/ percent = Moderator Temperator Ccefficient = -15 pcm/ degree F Differential Boron Worth = -10 pcm/ ppm a.

594 ppm b.

596 ppm c.

606 ppm d.

608 pp r, OUEST10N 5.17 (1.50) Indicate whether (ech of tho f ol i cei nc-heve e me.j or 2nfluence cn F Ab f Al.

FLUX D I S T R I Bi tT I Ot t (RFDi, A: l At FLO) OfSTRIIRITIDH 4 AF D). or RiiD I (,L Al4D A<16! FLUX D I Sl F< I BUT I Di l, to.', 04ch> a.

Fuel a s s.'mb l y orlds b.

Control r ect incertion c.

Reflector (44444 EllD OF LallbulW On 4s44 )

r b __ELANL SYSIEUS_ DESJ GU, _CONIBN _ AND_,1 NSIBUMENT AJ 1011 PAGE

OUESTION 6.01 (1.00) Which DNE of the fol1owino 1 cads is NOT automatical1y sequenced crito a cafeguards bus following a Safety injection Si gnal ? a.

Containment fan 1D b.

Auxiliary feedwater pump 1A c.

Containment spray pump 1A d.

Safety injection pump 1C OUESTION 6.02 (1.On) Selett t f ir-I ris t r urnen t Boa thet. (<pon deener o) r 1 nci. m 11 remove t #it capability of coerato*2 to caenipula+e t1CD control l er a lei flanual. a.

A i ns tr ustient bue 6.

B instrument bue c.

C 2 nst ru neret bi s d.

D instrunieni but DUESTION 6.03 (1.00> Which ONE of t he to11c m ro. t nu n t op c.

will F RE'/d N1 flanual F'..d itt.hde mej s.

P-1.

overpower b.

P-2, 3ow pcwer

c.

P-3.

dropped rod d.

Tavo deviation (44444 C A l f.bi)P Y ti6 L' tin i 114 ti; ON Ncxi F nL+_

e 4 4 4.'

6....._PL At.lT,_ SYSTE!1S_ DES IGth _GONT ROL,. AND.. IblST RUMENT AT 10N PAbE

OUESTION 6.04 (2.00) Refer to FIGURE 6-1 (SI System) to answer the followino questions: a.

List the TWO conditions which must be satisfied to cause MOV-826A and 826B to automaticall y CLOSE.

(1.0) b.

State whether the followinq volves are OPEN or SHUT durino the Haah Head In.)ettion Phase in the event of a LOCA.

L1.O> 1.

MOV-878A 2.

MOV-878E: 3.

MOV-8968 4.

MOV-897 OUESTION 6.05 t2.uO) F or each of the f ol 1 ovi no p, oc e.a > sd)et>cn moni t or i ng clierit r_1 i,, !stu i i. i automatic actions (e'acludano alarms) w11: occur when the essociatrd H1 oli Radioacti vi tv al ar m actucteE.

( T her e er e FOUr< cctionc requared + ve + et h.' a.

R-18.

11auid w+ste dispose) to circulating wet er system (1.

/ b.

R-21, r e t e n t l oc. t ent: monitor (J.O.

OUESTION 6.06 (2.SO) a.

L i t,t the THREE slalials i iis t 211 cause en cutomstic ster t o& Doti.

MDAFPs.

(1.'.> b.

Why is each MDAFP eut oil.al i c el l y limited to 230 opm discheroc fl o6s.

sv 0., c.

With both MDAFPs takino suction from the Service Water bvstem, whert le p urnp rec i r cul ati on f l ow dir ec t ed'- ('.'.r (t**4* CATEGOR( 06 col 4T INUL D Ull NE Al PAbh 444*4)

. b.__ELONTESIEtjE_DESIGth.COUIEh_ OlyD_IUST8U(1EUlOJ100 FwOL 1O QUESTION 6.07 (2.50) State how the f ollowing components respond (F AIL OPEN. FAIL CLD5ED. hEllAlfi FUNCTIONAL. DIVERT TO ETC.) vih en instrument air pressure is lost with .... the plant at 100 percent power.

iO.0 each) c.

Steam oenerator atmospheric relief valves b.

PCV-135. Letdown line pressnre control valve c.

FCV-1108 Reactor makeup to cheroing pump suction flow control vc-lve d.

AOV-754A. CCW to RCP A thermal barrier heat exchanger i sol e t i on se). e.

LCV-112A. UCT level control val vn . DUESTION 6.09 Q.Un) The pl ar.t is opere:tino c<t 30 per tent p o.m r .n t t i eJ1 control s. v r t es,s 2 t! eutomatic.

State HOM corit r o) r.,d s uz)J i n ) t i a '! 1 y respond ( I N 5'- R I. WITHDRAW. or NO C HONhE ) tu r -c h 04 i tic f ol i c in iio oc cier t er stes. A s ia m I.

reactor doou not trip and no t ur bi ne rulibac t: cccurs.

LO.S each/ a.

PCS boren c oncen t r o t. 2 0n &- rr .: H 10 p u oi. b.

FR chaunes N 4 '- drifto LOW.

c.

PT-485 ( t tir b) ne f it et tetene pressure) drifts HIGH.

d.

Channel 1 1-hot inrtontgneaue1-(6112 HIGH.

(44*4A 1:A1 EGur > o e-l.0111 I N.JE D ON NEA1 Fii(.E 44444

6.__E! AU.TESIEttS_ DESIGth,CONTROLt._ AND_, I NSTRUt1ENT AT 1019 PAGE

1 QUESTION 6.09 (2.00) Complete the f ollowing statements t'egardino the interlocks associated with the Fuel Transfer System.

The answers are among the choices provided below.

320 700 960 1200 2250 2600 1.

The slack cable lim tino c trcuit pr even t s the hoist from nio v i r us dows iet d with a load of _ __.

pounds or less.

(0.5) 2.

'T he gr i pper 2 riterl oc k carcult or everits t he or apper air solenoid salve f r om operati na at a load cf ___ pounds or more.

(0.5 3.

The overloed circtilt pr everit s u p w or-d hoi st mvvestivtil whcie the loed exceede _, ._ pmincz.

(O.i,) 4.

The hoist trip c i rciai t prevents upward moven,en t of the hotet wh er s f l i rs oripper is disenow.1ed with a load of pounds or more.

(u.5-Ot TEST ION 6.10 ( 1. 5t:0 a.

Explain the purpose of the e t eon, pr essure input s t oreal in the Ste.'n " Generator Wster Level Control System.

(O.S+ b.

How will INDICATED s t m i, fi o L:.ncare to nrlUAL steem flow <t Wi partLtn* - p o w er-if. durino c pow r t rict ease f rom 20 per cent to luu percen. pot io. the steam pressure sinrial s t i c i.s at its 50 per cerit power value ' (1.

> (4**44 CATEGORY 06 CDN1ll:UED ON NE)1 PAbE 444444

ht__ PLONT_SYSTEtjS_,DESIFth. CON 180L, OND_INSIfiVL1ENI61, ION PAGE

OUESTION 6.I1 (2.00) Describe the effect(s) ori the f ollowino plant systems if the turbine first stage pressure transmitter (PT-485) fails LOW during full power operation with all control systems alioned for automatic operation.

a.

Rod Control System (2 effects) (0.G ecch) 6.

Steam Dump System c.

Steam Generator Water L evel Control System d.

Electro-Hydraulic Control System DUESTION 6.12 (2.50) a.

E >< p } a i ri HON a r c ~ c* cir ir2e is .<o:ded olien i,eti a n.. i h e-n.. 1._ u instrumentation when the reector is at S powe, AND 50*/. powr-r. loc l iitie Source Rence AND Int ermedi-t e Reriot t es. L a na with eppe out 3 at e i n t ei l oc h t.. bypass switches. and l ea n. 'J.' b.

A power r anoe riitt l eet i rist r omerit e t s ori creenrrl fells HIGH.

HOW do the ROD STOP and POWER MISM,iTLH B iF C.56 uvi t che s affect Rod Contr ol System operetien whi n I r.d1.2d ul1< p1ated to the f i. 2 l e d trainel' ii.0 - QUESTION 6.13 (3.008 a.

WHAT i npiit stonel is used to ed iu z t pr our somhed level f ut t i.e pressurizer levcl con t r cil w; t t -- ' i-b.

WHAl 25 the nor inel u oer-moed pi ei sur 2 :er ] t. s e l ei no Iusd i i lu roii load? w.', c.

If pressurtner level control criennel L 1 --4 2 7 f all e LOW durano 100, g 41-operation while in.utomatic control, WHAl reactor protection st onal will cause the reector to trip? Explain HOW LT-427 failir,a LOW ceuser this to occur.

Include setpoi nt s and coi nci denc es.. Assume NO operator action i s t aken end t hat LT-427 it the controllino chennc1.

(,. u-(**4t* EliD OF CATEGORY 'o 44t+4) ..

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ . . Z,_ PROCEDURES - NORMALt_GBNOFR10L.. Et1ERGEtJCY AND FriOE

80D I QLOG I.C8L,_CO!HEQL DUESTION 7.01 (2.50) a.

Select the Critical Safety Function (CSF) condi tion which takes precedence in requiring operator response? (1.0) 1.

RCS Inventory - Orange Path 2.

Core Cooling - Red path 3.

Containment Integrity - Red Path 4.

Subtriticality - Orange Path 5.

PCS Inteor)ty - Red Path b.

List the TWO Emergerity Frocedures (by title or number) wl1ch et e Nul i cnmed i alul y eaited upon di sc over y of e red poth air e C BF- / ti.Or OUESTION 7.02 ( 1. 'i B Refer to Floure I-1.

Mi n a niunt Subcool0d Tenoureture.

Select t he or o. ip of 2 ridi crt i c.r w et -i t t, e.= u h w r,.. i er s u nj( af i s u it,. i Circulation lii accordance ' s i 1. h E5-0.1.

Reacl o.

Ir10 ker u oi is u. Attechnem t - A.

A ns.u i.e rior o.a l c or. t e i s sioen 1.

t azidii l or.w. RCS S/G COKE EX]1 PRESSURE (psig) FRESSUFES 1-1401 ~l - t U L I. IttER;1CCOU! i E S a.

950 Decr easing Consta % Coastont 523 decr ee r, I-b.

1150 Irice ee s i n.s Cor Ei cnt Incre:4s i r.ci 'i ' i '. d..y ee-F - c.

1300 Decreasirio Decreasing Decreasino 575 deot et f-d.

1650 Lecr ess a rio Dect cana rio Corst en t 6v0 dcgreet F-e.

1950 Constant Decreasino Constant 625 decreec. F (4**4i CATEGUR) O) LON1 I Hut.b f ilJ NLA F '4bF 44444 . _ -

L PBOCEDUBES_ _UO6b6L,_9B(40EMGl_,..EUEBGENCY_8t4D PAGE

80D.IOLO91 COL _.cgyIBpt QUESTION 7.03 (1.00) The plant i s in Col d Shutdown with an RCS heatup in progress.

A grid voltage fluctuation resul ts in B RCP trippino at 0400.

Select the earlJtat time that B RCP :nay be rest art ed to insure against motor overheetanu? Assume all other startino prerequisites are satisfied.

a.

Immediately (After 2 mi nutes wi th lift pump oper ati ng. ) b.

0415 c.

0430 d.

0445 e.

0500 DUESTION 7.04 < J. ',W The t ru.edi at e ac t i on s. c4 E -v, Reac t o-it 3p c>r Safety Inicction. 2nclude t. i s - stepn which are taken to PREVEiil excessive GCB coolcuwn.

List these TNu steps.

QUESTION 7.05 ( ). 'iO ) E-0.

Reactor Trip or 9eiet3 In s ec t 2 on. 4oldout paoe 1 i t t i-100 conditioin that tcoether reouiro the operetor to trip both FCPs.

Whet are t hnw TWO conditions? OUESTION 7.06 (1.On) What TWO conditions constitute ' edver s e con t a? nnient " es used in Eneercent Procedures? Include values of parameters.

(44444 f.f. l F Boh 07 TON)II)UEb DN t hi i 1 F e staE 44444>

._ _________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ L__E80CEDUBg;L,.UO80GL, _ODNOBMh_.EMERGEl4Cy_AND PAGE IS BODIOLOGIGOL,_COUIBOL QUESTION .7.07 (1.50) During a loss af all AC power. ECA-0.0 has the operator cooldown end depressuri:e the RCS using steam generator atmospheric reliof valves, a.

Why must the RCS be depressurized? (0,5) , b.

Why shouldn't the RCS be depressuriced below 170 psto? (0.5) c.

With RCS pressure at 750 psig and decreasino durino the depressurization, how should the operator respond i f pressuriner l evel as l ost or vessel head voida no occur t? (0,5) OUESTION 7.08 (1.50) Thr pl erit-is in a rrA v e l )rio s h.it oc.k n i2th c.u e al ter a t 4 on _ a is viv.o ees.

' - . You are the sontor r c> a c t or operetor moni tor i nq fut] niovosien t io tno sor i t fuel pit when a spent 4 $el e s t onit; 1 Iw unl u t c tied pi em-it or el v .n i o.,1 ]. e., feet into the noont tim) 01t I 191 THPFF .e c t 3 o rin ruovat ej t o L.

t.4,,. .. attordance with F. 6 - 8, F ur-1 aid i. m.- Lomponer.t Mo v eri.en t 3o t h e S u ci, .J c . Pat.

QUESTION '.09 C.'u) / a.

Pr ovi de the followino ohole body exposure 11m3ts +or an oct.up a t i on e l radiation wcrker who is 20 vesr s ol e has rio previous document ed - rediat2on exoncuro, arid h e. t c comol et ed NRC-4 f or m on fa1e.

1.

Normal administrative quartorly linit 2.

Manimum admini s tr at i ve ct _t e r t er i y 1Im1t 3.

Normal administrative vaarly limit ( 0. '5 e a c h ) 4.

Maximum 10 CFR 20 yearly limit b.

In accordanco with A-1.

Radiation Coritrcl Manual. 1ist an indivacuct h, ti tl e who may extend the norms 1 edmi ni s tr ati ve over ter l y limit to the maximum administrative quarterly limit.

M 'i> t r i (**148 CATEG0fo O L ou l l h. je i. Du No.. i 6%e 44444 < l . _ _ - ,

,. - - - - - - - - - - - - - - - 7i_..fROCEDUBES _NO8M N.,9BNO8M N _EUERGENCY AND PAGE

BGD.IOLOatCAL_COUTROL . QUESTION 7.10 (2.50) Following a valid reactor trip sional, the operator notices the reattor trip breakers are still closed and after unsuccessfully attemptino to manuativ t' r i p the reactor, he beci ns Etnergency Boration, a.

State the THREE steps required to 2nitiate Emeroency Borstion in accordance with FR-S.I.

Response to Reector Restart /ATWS. (1.5) b.

The reactor is subsequently tripped but 1HREE control rods ' f ail to insert.

How much boric aci d shoul d be sdded pri or to terminating Emergency Doretion? (1.0) DUESTION 7.11 (2.00) Prne erior e n-11. E c.n t o f o4 con:. ) n...e n t W.1 ,s t m.1 i. k. + _, t..r ic,. ,i).

. places requirements on thn creration of the containmont depres9imication velves. 7 9.' w id M1), a.

At phet por. ) 11 s e c o-il n; n. < t 1 0, it ut e-n o. u i i hc ol e b e-( t e r e d '.- iv',, v b.

At what neoat i ve cont e) s i". a ' ' pr ensure mu a. the velvos be opened? ( 0. * i.' c.

Whet i tt t h c me n 2,u i.y cona i) - t s e t2me pe,1od 611 ows.d +o-797v end h<1 io be open 1ri e calendar ye r ~ t<>.5) d.

At whet cun ul et i ve t i me per i od f o,- 7970 end '/971 bei no open nivs t the Operations Maney?r be not s iie P (0.5) OUE9 TION 7.12 -'-) i ' Provide the fol1owino overatano 1imits as defined in F-2.

FCS F#e sutivs... Limitations, and Sotpoints, t0.5 each) a.

Maximum heatuo rete b. Maximum cooldown rato u i t. h PCPn runnino c. Maximum cooldown tete with na t or -1 catculat3on d.

Maximum delta-T between the pres aurt: er end the RC5 1 Mos e.

Mes2 mum bor on cone.en t r e t 2 c r.

dafforence between the pr et sur 2 r e, m ed it2 RCS (6t4s4 c, a _: s $, ti' C' urn t u Lli 014 Ne n P,6L 44ea4, . i

, 2r_ E80CliD.UBES - _ NOB!th.,8pNpMV4 3.EVERGENGy AND PAGE

BSDIOLOGIGOL GOH.TBOL QUESTION 7.13 (2.00) The reactor is et 40 percent power with bank "D" contr ol rods et 1 6 u s l e k t.

during the initial load increese f ollowino a refuelino shutdown.

a.

What is the mexitnum load increase rate? ( 0. 5 ) 6.

What is the maximum bank "D" rod withdrawal rate? (0.5) . c.

Describe the power and rod position requirements which must be satisfied in order to remove all power rate and rod rate restrictions.

(1.0) DUESTION 7,14 r1.Gn> lhe plent 2 c over uting at / u pe, t erit pe et if a t t.

all control systends i ri automatic.

L19t TliREE s o u r c. a r-cf l e ;J

INT O ti.e CCW a v s t es, 2.

cmoi C. with AP-CCW.1. Leakaoe Ini o F Cid Looo.

(*44'44 END DP C al EG0r: p 07 4444.43 a

_ _.. _ _ _ _ _ _ __ _ _ _ _ _ _ - ___ _ - _ _ _ _ b.. qD01NISIGOT.IW.,PPOCEDURES._,COfjD I T IGNS,..AND._t,Iij!] AT IONS PAGE

3 OUESTION 8.01 (1.00) In the absence of the Fire Protection and Safety Coordinator, the Shatt Supervisor may assume his responsibt?.1 ties in accordance with A-905 Open F1ame. Weldino, and Gri ndi ng Permi t.

Comolet e the f ol1owinq sentences regarding administrative weldino requirements.

a.

Prior to commencing welding, the work aree must be inspected to ensure all moveable combustible material bel ow ar.d wi thi n a mi niit. uni radi us o f ______ f eet around the weldi no si t e has been removed.

(0.") 6. After welding has been completed, a firewatch must remain on station 4or a minimum of _ _,, addit 2onal oii n u t e c. (0.5) , QUESTION 8.O2 (l.50) ei. tenc n it re g., i t, , 3,,g , ,,,, ego,, vg to t, t u t., o.

.,, g

,,,, b.

Who is r enputisi b l e - For ccmoi.el a tio 8. h u Po=L le j p Re v i ew F or i.i ; i<a.

. c.

Whote orrmittior. It required t c. r +. E t s e t the tesctor' us.

. i QUESTIDN G.03 C~.. SO.i . a.

Upon essighir o en Event C1 eu) 4 a cet 1 on. What Is t he mer a mu.n t i me p.fr 1 od a l l owed for noti +1 cation of the following (0.5 each) 1.

tFC - 2.

New York Stato 3.

Wayne snd hora oe Countie b.

Select the lowest level E <eint Classification for which Pr ui ec t > < *. (n. t,. Recommendation <a should be evcluoted and refortec.

(1.0) 1.

Unusual Event 2.

General Emeroency < 3.

Site Emeroency 4.

Alert i l (*48** C(il E GUP, Ub i.(>N I INUE D ON Ntx1 Pn66 44*44) , ' . - _ _.. _.

_.

_ - - _. _

_8DUlt415IB8IIVE P80.CEDUBEh CONDITIONS, _ AtJD_ LIMIT AT IONS PAGE

. OUESTION 8.04 (2.25) a.

In accordance with Technical Specifications, the Overpte=sure Fratection System (OPS) must be operable whenever one or more RCS cold legs la lose than or equal to __. (1).. (0.D) deorees F or when the _ (2) ( 0. ". > system is in operation.

b.

List the TWO acceptable methods of providing an OPS in accordance m li.

Technical Specifications.

(1.25) OUESTION 8.05 (2.7D) a.

Using Fioure 8-1 Control hnd i nwr i lon L 2 n.i t s Ver sus Cor e Power. E t al e the Rod Insertion Ia mi t (2 n e. t ep r. ) for opor at i on at 25 percent power.

(0.70) b.

St ate T HF EF. batt.

f c>r Rod ]i_srtton L;m2Lt.

t 1. ', c.

What m e.) ot opa ato octaen it r eoult ed 1t control r cdu are inbet ted below the Rod Insertivi L2elt ( 0. N OUESTION 8.06 ( 2. O <. a a.

What ere the Technical Spec 2ficel lon 11 mitt for RLb succific. atti 3-(3.U) b.

What is the basis for the lechnical bpec2fication limits for RLS spetifit acti vi t y? (0.5) c.

Whet plent E,yttem t. h vin g e as r coui r ed in response tu e tuei el eme:i t failure eith RCS = p it t i ( 1. ext 1vity tae l ou leconics] Spec)ficettun 1a.o1: (v.r., as**** C ATEGOR's 08 CON 1INUED O'4 NEni FAbE 4**44) - _ _- .

< O x _.6Dd] U I ST6011 yE_ E80CEDURES,,GOND I T I QNS.._ AND_ L i t11 T(31 I ONS PA6h .co DUESTION B.07 (1.GO) Rsfer to Figure B-2, Target Axial Offset.

Given each of the following i ndi c at i ons, determine whether axial fim difference is being maintained within the target band AND explain briefly why or why not.

(0.5 each) POWER AFD AFD AFD AFD LEVEL CHANNEL 1 CHANNEL 2 CHANNEL 3 CHANNEL 4 a.

51% +3 OOC +5 +4 b.

70% -8-7 -6 OOL c.

857 o-6 -6 s OUESTION 8.08 (1.00) In accordante with SI-.. nds er s t W e -4 i h et E,r.er o~n t y Pl ie n. whel e s_ t i o n < =., must be taken regardino the emergency diesel generators in the event oi steadv Hur t 2 c ens-f- or c e nind, DUESTION 8.09 (2.00) Describe FCUP requi remt ni s that n.u t t be setisfied to ens ur e. Corit a t rio.eiit Inteor i ty is estabilchc-d it, accordance with lechnice) Spec 1 4 1 cat i co ne. OllEST ION 8.30 (1.5v) Provide the MINIMUt1 whole bodv radletion dose rates at wh) ct, tiie f oll om o, designations are applied, a.

Radiation area b.

High radiation area (0.5 each) c.

Locked high red 13 tion area (***** CATEGubi 06 CON 11NUEI' ON tr i T PMI *4444>

- 8 __eDMINisIB911yg_EBgCgogggs. gOsplIIOt4.S,.8ND. LIM 1101]Ot4S PAGE

1 OUESTION 8.11 (1.50) What are the requirtmentt, for i n p l einen t i ng a terr.por ary chance to operatino procedures listed in Technical Specifications? OUESTION 8.12 (2.00) The p1 ant is operating at 60 percent power when it is discovered that tiie CCW supply line to B Containment Spray Pump is crimped shut, a.

What action (s) must t> e teken in accordance with Technical Specificalitur prior to beginning repairs on the CCW supply 1ino? (1.O) b.

Prior to c ompletino repair s en the CCW <wpplv line. In dieer1 v. neret.o.

f ails a routirse operability test.

State whether A Containment Spray Put.g. ii DFEFAPLE. aid th ( cor (Mi t i 41 (1.Os GUESTION 8.13 ( 3. '~ 0 ) Technical Specifications require at, invectioation to det ermine t he onor ce of a significant increene in RCS ledace.

a.

List fHREE indicationi provi ac d in l eclin a c ci duec.l f i c at l osis whi c h ar e used to defloe when a "caontiacant increase" in RCS leakace exists.

t1.Dt 6.

The plant is operat no at 7 5*,. p ower and the la;eut l ed rate d e:. L e sh>ws: 11.1 GPM - Total RCS 1(M.coe rete 1. 7 GPt1 - Leakaoe int a the Pr eu eim a r er Pelief l eni 1.2 GPM - L e c t e o r-l iit t, t he heett er (oolent U,can Tens 0. 5 GPl1 - Total pr a mar y to seconderv leM.aco 5. 2 GPt1 - Leskaoe past RCP sealo 0.5 GPM - Leakaoe past pressurt:er PORV (FCV-430) What RCS leakaoe limits. if any, have t>een exceeded? Pefer to attachtd Technical Specifications.

(2.0) (4s44* END GF LAiEGUFi O t, 4446 4/ (************* END OF EX AMINATlf N

444t********44)

.

EQUATION SHEET f=ma v = s/t Cycle efficiency = (Met work out)/(Energy la)

w = ag .s = V,t + 1/2 at

E=m , KE = 1/2 av ,, (yy, y )fg g, 13 3, g,- A t

g o PE = agn Vf = V, + at w = e/t A = in2/t.1/2 = 0.693/t1/2 w = v aP.

l/2'

' C(*1"}(b}1

A= ((g /2) * I*b)3 , l AE = 931 m

m = Y,yAo-Ex , Q = mCpat Q = UAaT I = I,e'"* I = I,10"*/ M Pwr = W sh f TVt. = 1.3/u sur(t) P = P,10 HYL = -0.693/u ? = P e /T t o SUR = 26.06/T SCR = S/(1 - K,ff) , CR = S/(1 - K,ffx) x SUR = 26s/t* + (8 - p)T CR;(1 - K,ffj) = G (I ' Ieff2)

T = ( t*/s ) + [(a - o V Io ] M = 1/(1 - K,ff) = CR;/CR, T = 1/(o - s) M = (1 - K,ff,)/(1 - K,ff;) T = (s - o)/(Is) SDM = ( - K,ff)/K,ff o = (X,ff-1)/K,ff = d,ff/K,ff t' = 10 seconds I = 0.1 seconds ' o = C(t=/(T K,ff)] + ti,ff (1 + IT)) / Ib ",Id ll

P = (tov)/(3 x 1010)

Id gd jj

2 I = eN R/hr = (0.5 CE)/d (meters) R/hr = 6 CE/d2 (feet) Water Parameters Miscellaneous Conversions 1 gal. = 8.345 lem.

1 curia = 3.7 x 1010eps l 1 ga;. = 3.78 liters i kg = 2.21 lbm 1 f t- = 7.48 gal.

I np = 2.54 x 103 Stu/nr Oensity = 62.4 lbrp/ft3 1 mw = 3.41 x 106 5tu/hr Density = 1 gm/c r3 lin = 2.54 cm Heat of vaoorization = 970 5tu/ltm

F = 9/5'C + 32 Hest of fusion = 144 Stu/1cm

'C = 5/9 ('F-32) 1 Atm = 14.7 psi = 29.9 in, ng.

1 BTU = 778 ft-lbf I ft. H O = 0.4335 itf/in.

.

Table 1.

Saturated Steam: Temperature Table Abs Press.

Speofsc Volume Enthalpy Entropy Temp tb per Sat.

Sal.

Sat.

Temp Fahr SqIn.

Liquid Evap Vapor Liquid Evap Vapor Liquid Evap Vapor Fahr I p v, vig vg hl h fg h se sig sg I g 37 0 0 08859 0 016022 3304 7 3304 7 0 0179 1075.5 1075.5 0 0000 2.1873 2 1873 32 a 14 0 0 09600 0 016021 3061.9 3061.9 1.996 1074 4 1076.4 0 0041 2.1762 2.1802 34 8 36 0 0 10395 0 016020 28390 2839 0 4 008 1013.2 1077.2 0 0081 2.1651 2 1732 36 8 38 8 011249 0 016019 2634l 26342 6 018 1072.1 1078.1 0 0122 2.1541 2.1663 38.8 40 0 1.12163 0 016019 2445 8 2445 8 8 027 10710 1079 0 0 0162 2.1432 2 1594 40 8 42 0 0 13143 0 016019 2272 4 2272.4 10 035 1069 8 107E9 0 0202 2.1325 2.1527 42 8 44 0 0 14192 0016019 2112 8 2112 8 12 041 1068 7 10803 0 0242 2.1217 2.1459 44 8 46 0 0 15314 0 016020 1965 7 19657 14 047 10676 1081 6 0 0282 2.1111 2.1393 as e 48 0 0 16514 0 016021 1830 0 1830 0 16 051 1066 4 1082.5 0.0321 2.1006 2.1327 48 8 50 0 0 17796 0 016023 17048 1704 8 18 054 1065.3 1083.4 0 0361 2.0901 2.1262 58 I 52 0 0 19165 0 016024 1589 2 1589 2 20 057 10642 1084 2 0 0400 2.0798 2.1197 52 3 54 0 0 20625 0 016026 14824 1482 4 22 058 10631 10851 0 0439 2.0695 2.1134 54 I SE O O22183 0 016028 1333 6 1383 6 24 059 1061.9 1086 0 0 0478 2.0593 2.1070 56 8 55 0 0 23843 0 016031 1292.2 1292.2 26 060 1060 8 1086 9 0 0516 2.0491 2 1008 58 8 60 0 0 256tl 0016033 1207.6 1207 6 28 060 1059 7 10873 0 0555 2 0391 2.0946 E8 8 E2 0 0 27494 0 016036 1129 2 1129 2 30 059 1058.5 1088 6 0 0593 2.0291 2.0885 E2 s E4 0 0 29497 0 016039 1056.5 1056.5 32.058 10574 1089 5 0 0632 2.0192 2.0824 64 8 EE O O31626 0 016043 989 0 989.1 34 056 1056.3 1090 4 0 0670 2 0094 2 0764 EE e EB D 033889 0 016046 926 5 926.5 36 054 1055 2 1091 2 0 0708 1.99 % 2 0704 El 8 JD D 036292 0 016050 868 3 868.4 38 052 10540 1092.1 0 0745 1.9900 2.0645 70 8 17 0 0 38844 0 016054. 814 3 814.3 40 049 1052.9 1093.0 0 0783 1.9804 2 0587 72.8 74 0 0 41550 0 016058 764I 764.I 42 046 1051 8 1093 8 0 0821 1.9708 2 0529 14 I " JE O O44420 0 016063 717.4 717.4 44 043 10507 1094 7 0 0858 1.% 14 2.0472 7E I 78 0 0 47461 0 016067 673 8 673.9 46.040 1049.5 1095 6 0 0895 1.9520 2.0415 78 8 IO s 0 50683 0 016072 633 3 633.3 48 037 1048 4 1096 4 0 0932 1.9426 2 0359 88 8 87 0 0 54093 0 016077 595 5 595.5 50 033 1047.3 1097.3 0 0969 1.9334 2 0303 12 8 e4 0 0 57702 0 016082 560 3 560 3 52 029 10461 1098 2 0 1006 1.9242 2 0248 84 8 86 g 061518 0 016087 227.5 527.5 54 026 1045 0 1099 0 0 1043 1.9151 20193 85 g - ae 0 0 65551 0 016093 496 8 4%8 56 022 1043 9 1099 9 01079 1.9060 2.0139 fS I 90 0 0 69813 0 016099 4681 468I 58 018 1042 7 11008 0.1115 1.8970 2 0086 98 I 97 0 0 74313 0016105 441.3 441.3 60 014 1041.6 1101 6 01152 1.8881 2 0033 32 e 14 0 0 79062 0 016111 416 3 416 3 62 010 1040 5 1101 5 0 1188 1 8792 1 9980 94 8 SE D 0 84072 0 016117 392 8 392 9 64 006 1039 3 11033 0 1224 1.8704 I.9928 SE s 18 0 0 89356 0 016123 370 9 370 9 66 003 19382 1104 2 0 1260 18617 1.9876 SI 8

i Abs Press.

Specific Volume [nthalpy [ntropy Ternp Lb per Sat.

Sal.

Temp fahr SqIn.

Liquid Ivap Vapor Liquid Evap Vapor Liquid Evap Yapor Fahr h, h ig h s, sig s t . g g V4 I p v, _ vfr ~ O94924 0 016130 350 4 350 4 67999 10371 11051 01295 I8530 1.9825 les e 102 O 100789 0016137 331.1 331 1 69 995 1035.9 1105 9 0.1331 18444 1.9775 182 I 100 O ' 184 O I06965 0 016144 313.1 313 1 71 992 1034 8 1106 8 0 1366 1.8358 1.9725 184 e Its O 1.1347 0 016151 29616 29618 13 99 1033 6 !!DT6 0 1402 18273 1.9675 les e ICE O 12030 0 016158 28028 28030 7598 1032.5 1108 5 0.1437 1.8188 1.9626 les e 1100 1 2750 0 016165 26537 26539 7T98 1031.4 1109 3 0.1472 1.8105 1.9577 lie s 112I 13505 0 016113 25137 25I 38 79 98 1030 2 1110 2 0 1507 18021 1.9528 112 0 1140 14299 0 016180 238 2I 238 22 81 97 10291 1111 0 0.1542 13938 1.9480 114 e ils a 15133 0 016188 225 84 225 85 83 97 10?T9 1111.9 0.1577 13856 1.9433 Ils e Ils a 16009 0 016196 214 20 214 21 85 97 1026 8 11123 0.1611 13774 1.9386 lit a 12I O 1.6927 0 016204 203 25 203.26 8797 1025 6 1113 6 0.1646 13693 1.9339 12eI 122 I 13891 0 016213 19294 192.95 89.96 1024 5 11!4 4 0.1680 1.7613 1.9293 122e 114 a !8901 0 016221 183 23 183.24 91.96 1023 3 1115 3 0.1715 1.7533 1.9247 124 8 12E r 1 9959 0 016229 174 08 174 09 93 96 1022 2 1136.1 0 1749 13453 1.9202 12E 8 17f 8 21068 0 016238 165 45 165.47 95.96 10210 1117.0 0 1783 13374 1.9157 128 I r10 0 2 2230 0 016247 15T32 157.33 97.96 1019 8 Ill78 0.1817 13295 1.9112 13e e 32 g 2 3445 0 016256 149 64 149 66 99 95 1018 7 1118 6 0.1851 13217 1.9068 132 8 1.90 2 4717 0 016265 14740 142.41 101.95 10175 1119 5 0 1884 13140 1.9024 134 0 135 0 2 6047 0 016274 135 55 135 57 103 95 1016 4 1120 3 0 1918 13063 I.8980 136 0 138 0 2 7438 0 016284 129 09 129.11 105.95 1015 2 !!21.1 0 1951 1.6986 1.8937 13e s 14D D 28892 0 016293 12238 12300 107 95 10140 1122 0 0.1985 1.6910 1.8895 148 8 1420 3 0411 0 016303 11T2I 11722 109 95 1012.9 1122 8 0 2018 16534 1.8852 142 8 1440 3 1997 0 016312 III14 11136 111.95 10113 1123 6 0 2051 1.6759 1.8810 144 I tes O 3 3653 0 016327 106 58 106 59 113 95 1010 5 1124 5 0 2084 1.6684 1.8769 145 I see n 3 5381 0 016332 10168 10130 115 95 1009.3 1125 3 0 2117 1.6610 1.8727 les s 150 D 3 7184 0 01634'3 9705 9707 117.95 1008 2 1826 i 0 2150 1.6536 1.8686 15e 8 1520 3 9065 0 016353 9266 9268 119 95 1007 0 1126 9 0 2183 1.6463 1.8546 152.5 - 154 O 4 1025 0 016363 8850 8852 121 95 1005 8 11273 02216 1.6390 1.8606 154e 15E D 4 3068 0 016374 84 56 8457 123.95 1004 6 1128 6 0 2248 1.6318 1.8566 156 0 158 0 4 5197 0 016384 80 82 80 83 125.96 1003.4 1129 4 0 2281 1.6245 1.8526 15s I isa s 4 7414 0 0!6395 1727 7729 12796 1002 2 1130 2 0 2313 1.6174 1.8487 188 8 152 0 4 9722 0 016406 73 90 7392 129 96 1001.0 1131 0 02345 1.6103 1.8448 IE2 s 164 0 5 2124 0 016417 70 10 70 72 13196 9998 1131.8 0 2377 1.6032 1.8409 164 s IEE D 5 4623 0 016428 6767 6768 13397 998 6 1I32 6 0 2409 1.5961 1.8371 186 8 1Es D 5 7223 0 016440 64 78 6480 13597 997 4 1133 4 0.2441 1.5892 1.8333 168 8 110 0 5 9926 0 016451 62 04 62 06 13797 996 2 1134 2 0 2473 1.5822 1.8295 lie s 172 0 6 2736 0 016463 5943 5945 13998 995 0 1135 0 02505 1 5753 1 8258 172 8 174 0 6 5656 0 016474 56 95 5697 14198 993 8 1135 8 0 2537 1.5684 1.8221 114 8 lis D 68590 0 016486 54 59 54 61 14399 992 6 1136 6 0 2568 1.5616 I8184 IIs 8 178 0 71840 0 016498 52 35 5236 I4599 991.4 1137 4 0 2600 1.5548 1.8147 lie s

__ Entha!py Entropy Temp lb per Sat.

Sat.

Sal.

Sat.

Sal.

Temp Abs Press Specific Volume Iabr SqIn.

Liquid Evap Vapor Liquid Evan Vapor Liquid Evap vapor fant I I p vi vig vg hg h eg h

sl8_5J g 153 0 7 5110 0 016510 50 21 5022 148 00 990 2 1138 2 0 2631 1 5480 1.8111 las s 182 3 7 850 0 016522 48172 18.189 150 01 989 0 1139 0 0 2662 1 5413 1.8075 182 8 . 134 8 8 203 0 016534 46 232 46.249 152 01 987 8 1139 8 02694 15346 I8040 184 0 , 195 0 8 568 0 016547 44 383 44 400 154 02 986 5 1140 5 0 2725 1 5279 1 8004 las t las a 8947 0 016559 42 621 42.638 156 03 985 3 1141 3 0 2756 1.5213 13969 188 0 l I 190 0 9 340 0 016572 40 941 40 957 158 04 9841 11421 0 2187 1.5148 11934 1900 192 0 9 747 0 016585 39 337 39.354 160 05 9828 1142 9 0 2818 1.5082 11900 IS2 G 154 8 10168 0 016598 37 808 37 824 162 05 981 6 11433 02848 1.5017 13865 1948 ISE 0 10 605 0016611 36 348 36 364 164 06 980 4 1144 4 0 2879 1.4952 13831 ISE D 19s a 11.058 0 0!6624 34 954 34.970 16608 979.1 1145 2 0 2910 1.4888 17798 ISEO , 2000 11.526 0 016637 33 622 33 639 168 09 977.9 11460 0 2940 1.4824 1 7764 200 D 2c4 O 12 512 0 016665 31 135 31 151 172 11 975 4 !!475 0 3001 1.4697 17698 204 O 2ct o 13 568 0 016691 28 862 28878 176.14 972 8 1149 0 0 3061 1 4571 1 7632 2c8 0 212 0 14 696 0 016719 26 782 26 799 18017 970 3 1150 5 0 3121 14447 17568 2120 flE D 15 901 0016747 24 878 24 894 18420 9678 11520 0 3181 1.4323 13505 215 0 223 0 17186 0 016775 H131 23.1 6 188 23 965 2 1153 4 0 3241 1.4201 1 7442 220 8 224 8 18 556 0016805 21 529 21 h 5 19227 962 6 1154.9 0 3300 1.4081 13380 2240 228 8 20 015 0016834 20 056 2 d13 196 31 960 0 1156.3 0 3359 1.3961 13320 228 0 232 8 21 567 0016864 18 701 18318 20035 957 4 1157 8 0 3417 13842 17260 232 0 235 8 23 216 0 016895 17.454 1T471 20440 954 8 1159 2 0 3476 13725 13201 23Eo 240 0 24 968 0 016926 16.304 16 321 208 45 952 1 1160 6 0 3533 1.3609 13142 240 0 2440 26826 0 016958 15 243 15 260 21250 949 5 1162 0 0 3591 1.3494 13085 2440

24s a 28 796 0 016990 14 264 14 281 216 56 946 8 1163 4 03649 1.3379 13028 248 0 2520 30 883 0 017022 13 358 13 375 22062 9441 1164 7 0 3706 1.3266 16972 252 0 255 8 33091 0 017055 12.520 12.538 224 69 941.4 1166.1 0 3763 1.3154 16917 255 0 2500 35 427 0017089 11 745 11362 22836 938 6 1167.4 0 3819 13043 16852 250 8 254 3 37894 0 017123 11 025 11.042 23283 935 9 11687 0 3876 1.2933 16808 2E4 e i 2Es e 40 500 0 017157 10 358 10 375 236 91 9331 1170 0 0 3932 1.2823 1 6755 2E8 0 [ 272 D 43.249 0 017193 9 738 9 755 240 99 930 3 1171 3 0 3987 12715 16702 272 0 275 0 46147 0 017228 9 162 9180 24508 9275 1172.5 04043 1 2607 1.6650 215 0 285 O 49 200 0 017264 8 627 8 644 24917 924 6 1I73 8 04098 12501 I6599 230 0 224 a 52 414 0 01730 8 1280 81453 253 3 9217 1175 0 0 4154 1 2395 16548 2:40 288 0 55 795 001734 76634 76807 2574 918 8 1176 2 0 4208 12290 16438 2ts a 297 0 59 350 0 01738 72301 72475 261 5 915 9 1117 4 0 4263 1 2186 I6449 292 0 215 0 63 084 0 01741 68259 6 8433 265 6 913 0 1178 6 04317 1 2092 1 6400 295 0

, Abs Press.

Speofic Volume [nthalpy [ntropy lemp lb per Sal.

Sat.

Sal.

Temp fahr SqIn.

Liquid [vap Vapor liquid Evap Vapor liquid Esp Vapor Fahr I p vg vtg vg hr h it h s, sig

I g t 3sa 3 67 005 001745 6 4483 64658 2697 910 0 II79 7 0 4372 11979 16351 300 8 364 3 71119 0 01749 6 0955 6 1130 273 8 9070 1180 9 0 4426 1.1877 16303 304 0 388 8 75 433 001753 5 7655 57830 278 0 9040 1182 0 0 4479 1.1776 16256 30s O 3128 79 953 0 01757 5 4566 5 4742 2821 901 0 1183 1 0 4533 1 1676 I6209 312 0 315 3 84 688 0 01761 5 1673 5 1849 286 3 897.9 1184.1 0 4586 1.1576 16162 31E o 328 8 89 643 0 01766 48%l 4 9138 290 4 894 8 1185 2 04640 1.1477 I6tl6 320 0 3248 94 826 0 01770 4 6418 4 6595 294 6 8916 1186 2 0 4692 1.1378 16071 224 0 328 a 100 245 0 01774 44030 4 4208 298 7 888 5 1187 2 0 4745 I1280 16025 328 8 332 0 105 907 0 01779 4 I188 4 1966 302 9 885 3 1188 2 0 4798 11183 15981 332 0 335 8 111 820 0 01783 3 9681 3 9859 307.1 882.1 1189 1 0 4850 1.1086 1.5936 3360 340 t 117 992 0 01787 3 7699 3 7878 311 3 8788 11901 04902 10990 15892 340 0 3448 124 430 0 01792 3 5834 36013 315 5 875 5 1191 0 0 4954 10894 15849 3440 348 8 131 142 0 01797 3 4018 3 4258 319 7 872 2 1191 1 0 5006 10799 I5806 348 0 3528 138 138 001801 3 2423 3 2603 3239 868 9 11923 0 5058 10705 15763 352 0 355 B 145 424 001806 3 0863 31044 3281 865 5 1193 6 0 5110 10611 1.5721 35E o 36a s 153 010 0 01811 29392 29573 332 3 862 1 1194 4 0 5161 1 0517 1 5678 350 0 3 54 s 160 903 001816 78002 28184 336 5 858 6 1195 2 0 5212 I0424 1.5637 354 0 358 3 169 113 0 01821 2 6691 2 6873 340 8 855I 1l95 9 0 5263 1 0332 1 5595 3Es o 312 3 177 645 0OlB26 25451 25633 345 0 851 6 11961 0 5314 10240 15554 372 e 375 3 186 517 0 01831 2 4279 2 4462 349 3 848I 1192 4 0 5365 10148 15513 31E a 3ss e 195 729 0 01836 2 3170 23353 353 6 844 5 1198 0 0 5416 1.0057 1.5473 3:00 184 8 205 294 0 01842 2 2120 2 2304 357 9 8408 1198 7 0 5466 0 9966 1.5432 3:40 318 0 215 220 0 01847 2.1126 2.1311 362 2 8372 1199 3 0 5516 0 9876 1 5392 383 8 312 8 225516 001853 2 0184 2 0369 366 5 833 4 1199.9 05567 09786 15352 392 0 396 0 236 193 0 01858 1.929! 19477 370 8 829 7 1200 4 0 5617 09696 I 531.3 35E O , 4C3 8 247259 0 01864 I8444 18630 375.1 825 9 1201.0 0 5667 09607 15274 4c3 0 454 a 258 725 0 01870 17640 17827 3794 822 0 12015 0 5717 0 9518 1.5234 404 0 4tt s 270 600 0 01875 16877 1 7064 3838 818 2 1201.9 0 5766 0 9429 1.5195 4cto 4120 282 894 0 01881 1 6152 1 6340 388 1 814 2 1202 4 25816 0 9341 1.5157 412 0 415 D 295 617 00iB87 1 5463 1.5651 392 5 810 2 1202 8 0 5866 09253 1.5118 4 50 420 0 308 780 0 01894 14808 14997 396 9 806 2 I2031 0 5915 0 9165 15080 4200 424 8 322 391 0 01900 1.4184 I4374 401.3 802 2 1203 5 ' 0 5964 0 9077 15042 424 0 423 0 336 463 0 01906 13591 I3782 405 7 798 0 1203 7 0 6014 0 8990 15004 428 0 4320 351 00 0 01913 130266 132179 4101 793 9 1204 0 0 6063 0 8903 1.4966 432 0 4350 366 03 0 01919 I24887 126806 414 6 789 7 1204 2 0 6112 0 8816 14928 436 0 448 0 381 54 0 01926 I19761 121687 419 0 785 4 1204 4 06161 0 8729 I4890 443 0 444O 397 55 0 01933 II4874 I16806 4?)5 781 I 1204 6 0 6210 08643 14853 4440 4470 414 09 0 01940 110212 112152 4280 776 7 1204 7 0 6259 0 8557 14815 4484 772 3 1204 8 0 6308 0 8411 14778 4530 4 52s 431 14 0 01947 1 05764 10771,1 4?,J s - -~~. .-u e .nu, ,fn 7c 7 R f 204 8 0 6356 0 8385 14741 4w a

~ 1 . ~ Ab5 Press Specific Volume [nthalpy [ntropy Temp lb per Sal.

Sat.

Sal.

Sat.

Sal.

Temp 1ahr SqIn.

Liquid Evap Vapor Liquid Evap Vap0f liquid Evap Vapor Fahr i p vg vig vg hg h rg h

Sig sg I r s50 0 466 87 001961 0 97463 0 99424 441 3 763.2 1204.8 0 6405 0 8299 1.4704 468 e 414 8 485 56 0 01969 0 93588 0 95557 4461 758 6 12043 0 6454 0 8213 1.4667 (se s 458 o 504 83 001976 0 89885 0 91862 4507 754 0 1204 6 0 6502 08127 1.4629 458 8 417 0 524 67 001984 0 86345 0 88329 455 2 749 3 1204 5 a6551 0 8042 1.4592 4723 415 0 545 11 001992 0 82958 0 84950 459 9 744 5 1204 3 0 6599 0 7956 1.4555 475 g 423 0 566 15 0 07000 0 79716 0 81717 464 5 739 6 12051 0 6648 03871 1.4518 488 8 484 0 58781 0 02009 076613 0 78622 4691 7343 1203 8 0 66 % 0 7785 1.4481 484 0 4II D 610 10 0 02017 0 73641 0 75658 473 8 729 7 12035 0 6745 03700 1.4444 488 e 417 O 633 03 0 02026 0 70794 0 72820 478 5 7246 1203.1 0 6793 0 7614 1.4407 4923 415 3 656 61 0 02034 068065 0 70100 4832 719.5 12023 0 6842 03528 1.4370 496.8 500 0 680 86 0 02043 0 65448 067492 487.9 714 3 1202.2 0.6890 01443 1.4333 See e ' 564 0 705 78 0 02053 0 62938 0 64991 4923 709 0 12013 0 6939 0 7357 1.4296 584.8 SCI S 73140 0 02062 0 60530 0 62592 497.5 7033 1201.1 0 6987 01271 1.4258 5000 512 0 75732 0 02072 0 58218 0 60289 5023 698 2 1200.5 01036 0 7185 1.4221 512.0 515 0 784 76 0 02081 0 55997 0 58079 507.1 692.7 1199.8 0 7085 03099 1.4183 515.0 , 570 0 812 53 0 02091 0 53864 0 55956 512.0 687.0 2199.0 01133 0 7013 1.4146 529.8 574 8 841 04 0 02102 0 51814 0 53916 516 9 6813 1198.2 03182 0 6926 1.4108 524I $78 0 87031 0 02112 0 49843 051955 521 8 675 5 1197.3 03231 0 6839 I.4070 528 8 537 8 900 34 0 02123 0 41947 050070 526 8 669 6 11 % 4 03280 0 6752 1.4032 532 8 535 0 931.17 0 02134 0 46123 0 48257 5313 663 6 !!95 4 03329 0.6665 13993 535.8 540 g 96239 0 02146 644367 0 46513 536 8 657.5 !!94 3 0 7378 06577 13954 548 8 5440 995 22 0 02157 0 42677 0 44834 541 8 6513 1193.1 0 7427 0 6489 13915 5440 - Sag e 1028 49 0 02169 041048 0 43217 5469 645 0 !!91.9 03476 0 6400 13876 548 8 5570 1062 59 0 02182 039479 0 41660 552 0 6385 1190 6 01525 0 6311 13837 552s 555 0 1097 55 0 02194 0 37966 0 40160 557.2 632 0 1189.2 03575 0 6222 13797 555 8 550 D !!33 38 0 02207 0 36507 0 38714 562.4 6253 !!873 0 7625 0 6132 13757 568 5 554 0 1870 10 0 02221 0 35099 0 37320 5676 6185 11861 0 7674 0 6041 13716 554 8 SEE D 120732 002235 0 33741 0 35975 572 9 611.5 1184 5 0 7725 0 5950 13675 568 8 577 0 1246 26 0 02249 0 32429 0 34678 578 3 604 5 11823 0 7775 0 5859 13634 5728 516 0 1285 74 0 02264 0 31162 0 33426 583 7 597.2 1180.9 07825 0 5766 13592 575 8 5:00 1326 17 0 02279 029937 0 32216 5891 589 9 1179 0 03876 0 5673 13550 508 8 53: o 1367 7 0 02295 0 28753 0 31048 594 6 582.4 1176 9 07927 0 5580 13507 584 3 5ee 0 1410 0 0 02311 0 27608 0 29919 6001 5743 1174 8 0 1978 0.5485 13464 588 8 511 0 1453 3 0 02328 0 16499 0 28827 6057 566 8 1172 6 0 8030 0 5390 13420 592 8 515 0 14978 0 02345 0 25425 0 27770 611 4 558 8 1170.2 0 8082 0 5293 13375 5960 , -

. Abs I'vess.

Specific Volume EnthalpY Entropy Temp ~ pr Sal.

Sal.

Saf.

Sat.

Sal' Temp fahr Sq in liquid Evap Vapor liquid Evap Vapor liquid Evap Vapor Fahr h ig h

sig

i I p v, vfg vg h l a g 500 O 15432 0 02364 0 24384 0 26747 617I 550 6 1167.7 0.5134 0.51 % 13330 500 0 504 0 15897 0 02382 0 23374 0 25757 622.9 542 2 11651 0.8187 0.5097 1.3284 544 I Eat 0 16373 0 02402 0 22394 0 24796 628 8 533 6 1162.4 0 8240 0.4997 13238 EOS 8 El2 O 16861 0 02422 0 21442 0 23865 634 8 5241 1159 5 0 8294 0 4896 13190 512 I E1E E 17359 0 02444 020516 0 22960 640 8 515 6 1156 4 0.8348 0.4794 13141 515 8 520 0 1786 9 0 02466 01%I5 022081 646 9 506 3 1153 2 0.8403 0.4689 13092 528 8 524 0 1839 0 0 02459 0 18737 0 21226 653 1 4?6.6 1149 8 0.8458 04583 13041 524 I E2s 0 18924 0 02514 0 17880 020394 6595 486 7 11461 0.8514 0.4474 1.2988 528 s 512 0 19470 0 02539 0.17044 0 19583 665 9 476.4 !!42.2 0 8571 0.4364 1.2934 532.8 53E 0 20023 0 02566 0.16226 0 18792 672.4 465 7 1138.1 0 8628 0.4251 1.2879 53E 8 E40 0 20599 0 02595 0.15427 0 18021 6791 454.6 11337 0 8686 04134 1.2821 548 8 544 0 2118 3 0 02625 0.14644 0 17269 6859 443.1 1129 0 0.8746 0 4015 1.2761 544 8 548 0 2178 1 0 02657 0 13876 0 16534 692 9 433.1 1124 0 0 8806 0 3893 1.2699 54s s $52 0 2239 2 0 02691 0 13124 015816 700 0 418 7 1118 7 0 8868 03767 1.2634 552 I E55 0 23011 0 02728 0.12387 0 15115 707.4 4053 1113.1 0 8931 03637 1.2567 555 8 650 0 2365 7 0 02768 0.11663 0.14431 714.9 392.1 !!07.0 0.8995 03502 1.2498 560 0 554 0 24311 0 02811 0.10947 013757 722.9 3773 1100.6 0.9064 03361 1.2425 554 s Ess o 2498 1 0 02858 0.10229 0.13087 731 5 362.1 1093 5 0 9137 03210 1.2347 Ess a E12 0 2566 6 0 02911 0 09514 0 12424 7402 345 7 1085.9 0 9212 0 3054 1.2266 572.8 576 0 26368 0 02970 0 08799 0 11769 749 2 328 5 1077.6 0.9287 02892 1.2179 57E 8 5to 0 2708 6 0 0303'1 0 08080 0 till7 758 5 310.1 1068 5 0 9365 02720 1.2086 See 8 EI4 0 27821 0 03114 0 07349 0 10463 768 2 2902 1058 4 0.9447 0.2537 1.1384 584 0 - Ell a 28574 0 03204 0 06595 0 09799 778 8 268 2 1047.0 0.9535 0.2337 1.1872 les s E32 0 2934 5 0 03313 0 05797 0 09110 790 5 2431 1033.6 0 9634 0.2110 1.1744 532.8 595 8 30134 0 03455 0 04916 0 08371 804 4 212.8 1017.2 0.9749 0.1841 1.1591 595 s 700 0 30343 0 03662 0 03857 0 07519 822.4 1723 995 2 0.9901 0.1490 1.1390 70s 8 702 0 3135 5 0 03824 0 03173 0 06997 835 0 144 7 9793 1.0006 0 1246 1.1252 782.8 104 0 3177.2 0 04108 0 02192 0 06300 854 2 102 0 956 2 1.0169 0 0876 1.1046 784 e 105 0 31983 0 04427 0013G1 0 05730 873 0 61.4 934.4 1.0329 0 0527 1.0856 785 8 705 47* 32082 0 05078 0 00000 0 05078 906 0

906.0 1.0612 0 0000 1.0612 185 41*

- . _ . -. _ . . Table 2: Saturated Steam: Pressure Table Specific Volume Enthalpy Entropy Abs Press.

Temp Sat.

Sat.

Abs Press.

ib/Sq In.

Fahr Liquid Evap Vapor Liquid Evap Vapor liquid Evap Vapor Lb/Sq In v h l gg g s, s gg s p h h g p t vg v,g g sse855 32.018 0 016022 33024 3302 4 0 0003 1075 5 1075 5 0 0000 2 1872 2 1872 sess55 s 25 59 323 0 016032 1235 5 1735 5 27382 10601 1087 4 0 0542 2 0425 2 0967 s 25 s Se 79 586 0 016071 641.5 641.5 47623 1048 6 10 % 3 0 0925 19446 2 0370 s 5e 1s 10114 0 016136 333 59 33360 6973 10161 1105 8 01326 I8455 I9781 Is Ss 16224 0 016407 13515 73532 130 20 1000 9 1131.1 0 2349 16094 18443 Se is e 193 21 0016592 38404 38420 16126 982I !!43 3 0 2836 1.5043 17879 le a 14BSE 212 00 0 016719 26 782 26 799 18017 970 3 1850 5 0 3121 I4447 11568 14 ESE 15 a 213 03 0 016726 26274 26 290 181.21 %97 1150 9 0.3137 1.4415 1.1552 15 3 28 8 227.96 0016834 20 070 20 087 19627 %01 1156.3 03358 1.3%2 13320 2s e 3e s 25034 0 017009 131266 13 7436 218 9 945 2 1164.1 0 3632 1.3313 16995 38 s as s 26725 0 017151 10 4794 10 4965 236 1 933 6 1169 8 0 3921 12844 16765 as e 5s 3 281 02 0017274 8 4967 8 5140 2502 923 9 1174 1 0 4112 1.2474 16586 Se e Es a 29211 0 017383 7.1562 7.1736 262 2 915 4 11116 0 4273 12167 I6440 ses 7s 3 302 93 0 017482 61875 6 2050 272 7 907 8 1180 6 0 4411 1.1905 16316 78 s se s 312 04 0 017573 5 4536 5 4711 2821 900 9 I!83 I O4534 1.1675 16708 se s se s 320 28 0017659 4 8779 4 8953 2907 894 6 1185 3 0 4643 1.1470 1.6113 98 s iss e 32782 0 017740 4 4133 4 4310 2985 888 6 1187.2 04743 1.1284 1.6027 Ise e Ils a 33419 0 01782 4 0306 4 0484 3058 883I !!88.9 0 4834 1.1115 1.5950 118 e 128 8 341 27 0 01789 3 7097 3 7275 312 6 877 8 1190 4 0 4919 1.0960 1.5879 12s a 133 3 347.33 0 017 % 3 4364 3.4544 319 0 872 8 11913 0 499: 1.0815 1.5813 13s 8 les s 353 04 0 01803 3 2010 3.2190 325 0 8680 1893 0 0 5071 10681 1.5752 14es 15s s 35843 0.01809 2.9958 3 0139 330 6 863 4 1194.1 0 5141 1 0554 1.5695 15s s 15s s 363 55 001815 ~ 2.8155 2 8336 3361 8590 11951 0 5206 10435 1.5641 Ils a 11s s 36842 0 01821 2.6556 2 6738 341.2 854 8 11 % 0 0 5269 1 0322 f.5591 178 s * 13s 3 373 08 001827 2 5129 2.5312 346 2 8507 1196 9 05328 I0215 15543 ist s 1se s 377.53 0 01833 2.3847 2 4030 350 9 8463 1197.6 0 5384 1.0113 1.5498 Its s 2ss a 381.80 0 01839 2.2689 2.2873 355 5 842 8 1198 3 0.5438 10016 1.5454 288 8 218 s 38591 0 01844 2 16373 2.18217 359 9 8391 11990 0 5490 0 9923 1.5413 213 s 228 a 389 88 001850 206779 2 08629 364 2 835 4 1899 6 0 5540 0 9834 15374 278 e 23s 3 39310 0 01855 197991 1.99846 368 3 831 8 12001 0 5588 0 9748 1.5336 23s e 24s 8 397.39 001860 189909 1.91169 3723 828 4 1200 6 0 5634 0 9665 15299 248 8 253 s 400 97 0 01865 1 82452 1 84317 376.1 825 0 1201.1 0 5679 0 9585 1.5264 25s s 26s e 40444 0 01870 115548 117418 319 9 821 6 1701 5 05722 09508 1.5230 26s 8 21s 3 40780 0 01875 169137 I71013 3836 818 3 1201.9 0 5764 0 9433 15197 21s 8 2ss 8 All 07 0 01880 163169 165049 387 1 815I 1202.3 0 5805 0 9361 15166 288 e 29s e 414 25 001885 157597 1.59482 390 6 812 0 1202 6 0 5844 0 9291 1.5135 29s s 300 e 41735 001889 I52384 1.54274 394 0 808 9 1207 9 05882 0 9223 1.5105 3sse 35s a 43173 0 01912 1 3 % 42 1.32554 4098 794 2 1204 0 0 6059 0 8909 I 4968 35s e sCO s 444 60 0 01934 1 14162 1 16095 474 / 780 4 1204 6 nm? ""

Specilit Volume Enthalpy Entropy Abs Press.

Temp Sat.

Sat.

Sal.

Abs Press.

lb/Sg in.

Fahr Liquid Evap Vapor liquid Evap Vapor-Liquid Evap Vapor LblSq In.

A h

S

W h l fg g

ig g p I 'rg g P I V 4500 456 28 0 01954 101224 103179 4373 7615 1204 8 06360 0 8378 1.4738 450 0 Sco 0 46701 001975 0 90187 0 92162 449 5 7551 12047 06490 0 8148 1.4639 SCO 3 5500 476 94 0 01994 0 82183 0 84177 460 9 743 3 1204 3 0 6611 0 7936 I.4547 550 0 Ec0 0 486 20 0 02013 0 74962 0 76975 4713 732 0 1203 7 0 6723 0 7738 14461 sco 0 550 0 494 89 0 02032 968811 0 70843 481 9 720 9 1202.8 06828 01552 14381 E50 0 720 0 50308 0020$0 0 63505 0 65556 491 6 710 2 1201 8 0 6928 07377 1.4304 100 0 1500 510 84 0 02069 058880 0 60949 500 9 699 8 1200 7 0 7022 0 7210 14232 150 0 0000 518 21 0 02087 054809 056896 509 8 689 6 1l99 4 0 7111 0 7051 I4163 Sta 0 050 0 525 24 0 02105 051197 0 53302 518 4 679 5 1198 0 0 7197 0 6899 1 4096 050 0 SCO O 53195 0 02123 0 41968 050091 526 7 669 7 1196 4 0 7219 06753 14032 3c0 0 $50 0 538 39 0 02141 0 45064 047205 5347 660 0 1194 7 0 7358 0 6612 1 3970 350 0 10000 544 58 0 02159 0 42436 0 44596 542 6 650 4 1192 9 07434 0 6476 13910 Ic 00 10500 550 53 0 02177 040041 0 42224 550 1 640 9 1191 0 07507 06344 1.3851 19500 11000 556 28 0 02195 0 3sb63 0 40058 557.5 631 5 II891 0 7578 06216 33794 11000 11500 561 82 0 02214 0 35859 0 38073 564 8 6222 1187 0 0 7647 06091 13738 11500 1200 0 567 19 0 02232 0 34013 0 36245 5719 6130 11848 0 7714 0 5969 1.3683 1200 0 12500 57238 0 02250 0 32306 0 34556 5788 603 8 1182 6 0 1780 0 5850 13530 12500 13c0 0 57742 0 02269 0 30722 0 32991 585 6 594 6 1180 2 0 7843 0 5733 13577 1300 0 13500 582 32 0 02288 029250 0 31537 5923 585 4 II77 8 07906 0 5620 1.3525 1350 0 14t00 58707 0 02307 0 27811 0 30178 5938 576 5 1175 3 0 7966 0 5507 1 3474 1400 0 1450 0 591 70 0 02327 0 26584 0 28911 605 3 5674 1172 8 0 8026 0 5397 1 3423 1450 0 15000 596 20 0 02346 0 25372 0 21719 6117 5584 18701 0 8085 0 5288 13373 15000 15500 600 59 0 02366 0 24235 0 26601 618 0 549 4 1167 4 0 8142 0 5182 13324 15500 1ECO O 604 87 0 02387 0 23159 0 25545 624 2 540 3 1164 5 0 8199 05076 13274 Isc0 0 16500 609 05 0 02407 0 22143 0 24551 630 4 531.3 1161 6 0 8254 0 4971 1.3225 1550 0 17c0 0 613 13 0 02428 021178 023607 636 5 522 2 1158 6 0 8309 0 4867 1 3176 17000 1150 0 61712 0 02450 0 20263 0 22713 642 5 513 1 1155 6 0 8363 0 4765 1 3128 17500 1000 0 62102 0 02472 0 19390 0 21861 648 5 503 8 1152 3 0 8417 04662 13079 10:0 0 18500 624 83 0 02495,018558 0 2l052 654 5 494 6 1149 0 0 8470 0 4561 1.3030 19500 1500 0 628 56 0 02517 0 17761 0 20278 660 4 485 2 1145 6 08522 0 4459 12981 15C0 0 15500 63222 0 02541 0 16999 0 19540 666 3 475 8 1142 0 0 8574 0 4358 12931 1350 0 - ! 20000 635 80 0 02565 0 16266 0 18831 672.1 466 2 1138 3 0 8625 0 4256 12881 2230 0 i 21c0 0 64216 0 02615 014885 0 17501 683 8 446 7 IJ30 5 0 8727 0 4053 12780 2I000 22s0 0 649 45 0 02669 0 13603 0 16272 695 5 426 7 1I22 2 08823 03848 1.2676 22000 23000 655 89 0 02727 0 12406 0 15133 707 2 406 0 1I13 2 08929 03640 12569 2200 0 24000 66211 0 02790 011287 014076 719 0 384 8 11037 0 9031 0.3430 1.2460 24000 25C00 668 II O02859 0 10209 013068 731 7 361 6 1093 3 0 9139 0 3206 1.2345 25000 2500 0 673 91 0 02938 0 09172 0 12110 744 5 3376 1082 0 0 9247 0 2977 1 2225 21s0 0 27C0 0 679 53 0 03029 0 08165 0 11194 757 3 312 3 10697 0 9356 0 2741 1 2097 2100 0 2300 0 684 96 0 03134 0 07171 0 10305 770 7 285 1 1055 8 0 9468 0 2491 1.1958 20000 25C0 0 690 22 0 03262 0 06158 0 09420 7851 254 7 1039 8 0 9588 0 2215 1.1803 29000 3DCS O 69533 0 03428 0 05073 0 08500 801 8 218 4 1020 3 0 9728 0 1891 1 1619 3ces0 3100 0 700 28 0 03681 0 03711 0 07452 824 0 169 3 993 3 0 9914 0 1460 1.1373 3100 0 31c0 0 70508 0 04472 0 01191 0 05663 875 5 56 1 9316 1 0351 0 0482 1 0832 3200 0 33912-70541 0 05018 0 00000 0 05078 906.0

906 0 10612 0 0000 10612 32cs 2- .-

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- _ _ _ - _ _. . .. I E0P2 i TITLE:

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REACTOR TRIP RESPONSE I i

I I PAGE 12 0F 15 I ( ii > i iii I Ii l ' - 2400 '! I

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. _. _ _ _ _ _ _ _ _ _ _.

. _ -. _ _ _ _. __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ . t GINfM SENIOR REACTOR OPERATOR LICENSE EXAMINATION KEY

5.

THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PA0E

THERMODYNAMICS ANSWERS -- GINNA-87/10/05-KINGSLEY, I.

i ANSWER 5.01 (2.00) > a.

INCREASE b.

DECREASE c.

INCREASE d.

INCREASE (0.5 each) REFERENCE LP POCO 2C, P. 43-47 003000K502 003000K504 ...(KA'S) ANSWER 5.02 (2.00) a REFERENCE WESTINGHOUSE Fundamentals of Nuclear Reactor Physics, Chapter 8 - 51 LP RRT04C, P.

LP RRT04C, LO. 3.7 001010K508 192002K112 ...(KA'S) AMdWER 5.03 (1.00) a , i REFERENCE Westinghouse Reactor Theory and Core Physics, Chapter I - 4,28 LP RRT04C, P. 58 LP RRT04C, LO. 3.8 004000K508 ...(KA'S) l ' t

i ' ! .-...- ,,. _ _ - _ _ _ -.

5.

THEORY OF NUCtr4R POWER PLANT OPERATION, FLUIDS. AND PAGE

THERMODYNAMICS ANSWERS -- GINNA-87/10/05-KINGSLEY, I.

(

ANSWER 5.04 (1.50) a. THE SAME b. HIGRER TRAN (0.5 each) c. LOWER TRAN REFERENCE LP RRT080, P.

16-32 001010A207 ...(KA'S) ANSWER 5.05 (1.50) , d REFERENCE WESTINGHOUSE Thermal-Hydraulic Principles, Chapter 12 - 8 Steam tables LP RHTO6C, LO. 1.1 002020K508 002020K501 ...(KA'S) - ANSWER 5.06 (1.50) c

REFERENCE Steam tables LP RHT04C, LO. 1.4 002000A104 ...(KA*S) I a . i --, - - - --

5.

THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS AND PAGE

THERMODYNAMICS i ANSWERS -- GINNA-87/10/05-KINGSLEY, I.

l ANSWER 5.07 (2.00) a.

INCREASE b.

DECREASE c.

INCREASE d.

DECREASE (O.5 each) REFERENCE LP RHT100, P. 12 AND 13 LP RHT100, LO. 2.3 193008K105 ...(KA'S) ANSWER 5.08 (1.50) d REFERENCE WESTINGHOUSE Thermal-Hydraulic Principles, Chapter 2 - 39 LP RHT06C, LO. 1.1 000055K102 ...(KA'S) ANSWER 5.09 (4,04) /. S a. TRUE b. FALSE IMM / REFERENCE RG&E FLUID FLOW TEXT, CHAPTER 6, PAGES 29, 30, 40 RG&E PROCEDURE T-8A, STARTUP AND SHUTDOWN A AND B CIRC WATER PUMP LP RHT07C, LO, 2.4, 2.5, 2.8 2.9 191004K105 191004K112 191004K115 191004K107 ...(KA'S)

5.

THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS. AND PAGE

THERMODYHAMICS ANSWERS -- GINNA-87/10/05-KINGSLEY, I.

ANSWER 5.10 (1.00) b REFERENCE LP RRT04C, P. 78 LP RRT04C, LO. 3.13 192008K123 ...(KA'S) ANSWER 5.11 (1.50) a REFERENCE LP RRT04C, P. 25 LP RRT04C, LO. 2.6 192003K109 ...(KA'S) !

- fl!ET.

b.12 (1.00) REFERENCE LP RRTOSC, P. 45 LP RRT05C LO. 3.6 192004K107 ...(KA'S) a imonz.n o.la (1.00)

REFERENCE LP RROSC, P.

4, LO 5.4 LP RR05C, P. 52 , LP RROSC, STUDENT HO, IR-5.24b

LP RR05C, LO. 5.4 001000K549 ...(KA'S)

_ _ - L THEORY OF NUCr.rAR POWER PLANT OPERATION. FLUIDS. AND PAGE

THERMODYNAMICS ANSWERS -- GINNA-87/10/05-KINGSLEY, I, , ANSWER 5.14 (1.00) b REFERENCE LP RRT06C, STUDENT HO, P. 6.7b LP RRT060, LO. 1.4.C 001010K526 015000A103 015000K504 ...(KA'S) ANSWER 5.15 (2.00) a. SDM DECREASES i ' b. SDH REMAINS THE SAME (0.5 each) c. SDM REMAINS THE SAME

d. SDH INCREASES REFERENCE LP RRT080, P. 14 AND 15 , LP RRT080, LO. 2.3 ' 192002K114 ...(KA'S) l AhCWER 5.16 (1.00)

c REFERENCE LP RRTO5C, P. 58

LP RRT05C. LO. 5.5

004000A404 ...(KA'S)

, l t i.

, , ,_ __ _ --. - __ _. - ,

___ _. - _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ __ . _ _ _. _ ____ . ____ _ _ _ _.

_ _ _ _ _ _ _ 5.

TIMORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE

l THERMODYNAMICS ANSWERS -- GINNA-87/10/05-KINGSLEY, I.

l i ANSWER 5.17 (1.50)

a. AFD I b. RFD AND AFD (0.5 each) c. RFD AND AFD I REFERENCE ! LP RRT090, P. 17 AND 20 ' LP RRT090, LO. 1.7 193009K107 001000K525 ...(KA'S) .

4 .

_.. - _ _ _ _ _ _ _ _- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___ __ __ _ -___ __ __ -_ __-__ 8.

PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE

ANSWERS -- GINNA-87/10/05-KINGSLEY, I.

ANSWER 6.01 (1.00) e REFERENCE LT RGE-8 P. 17 064000K411 ...(KA'S) ANSWER 8.02 (1.00) e REFERENCE LP RIC12C, P. 5 t LP RIC120, LO. 1.2 000057A106 ...(KA'8) ANSWER 6.03 (1.00) a REFERENCE LP R3001C, P. 23 LP R3001C. LO. 3.5 001000K407 ...(KA'S)

6.

PLANT SYSTEMS DESIGN. CONTROL AND INSTRUMENTATION PAGE

ANSWERS -- GINNA-87/10/05-KINGSLEY, I.

ANSWER 6.04 (2.00) < a. 1. Boric acid storage tank level less than 10 percent, and (0,5) 2. either HOV-825A or 825B has left the closed position. (0.5) b. 1. SHUT 2. OPEN g (0.25 each) 4. OPEN REFERENCE LP R2601C P. P-10 LP R2601C, LO. 2.4, 2.5 006020A302 006000K402 ...(KA'S) ANSWER 6.05 (2.00) a. 1. Overboard valve (RCV-018) will close 2. Laundry waste pucp will trip (0.25 each) 3. Monitor tank pump will trip 4. Waste condensate pumps will trip b.

1. Discharge valve closes 2. Recirculation valve opens (0.25 each) 3. Neutralizing tank pump trips 4. Retention tank discharge pump trips (if in auto) REFERENCE LP R3901C P. 13 AND 14 LP R3901C, LO. 3.2 073000K401 ...(KA*S)

6.

PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE

ANSWERS -- GINNA-87/10/05-KINGSLEY, I.

ANSWER 6.06 (2.50) ^ a.

1. Low-low level in either steam generator 2. Both MFP circuit, breakers open (0.5 each) 3. SIS b. To prevent pump runout (0,5) c.

Condensate storage tanks (0.5) REFERENCE LP R4201C LP R42010, LO. 5,9,10 061000K408 061000K404 061000K402 ...(KA'S) ANSWER 6.07 (2.50) a.

REMAIN FUNCTIONAL (N2 backup) b. FAIL OPEN c. FAIL CLOSED (0.5 each) d. FAIL OPEN e.

DIVERT TO VCT REFERENCE AP IA.1, P.

P AND ids 078000K302 ...(KA'S) - - - -. .. . - - ... -. --. - - _ _ .-.

6.

PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE

ANSWERS -- GINNA-87/10/05-KINGSLEY, I.

ANSWER 6.08 (2.00) a.

INSERT b. WITHDRAW or No Cd4#G 6 (0.5 each) d. NO CHANGE (Tavg deviation rod block) REFERENCE LP RIC010, P. 5 LP RIC010, LO. 1.3 LP RIC05C, P. 7 LP RIC010, LO. 1.2 LP RIC100, P. 6 LP RIC100, LO. 1.2 FIGURE RGE-CR-5 001000K403 ...(KA'S) ANSWER 6.09 (2.00) JAG '- ' 2. 96 (0.5 each) 3.

250 1200 . REFERENCE RGE TEXT CH. 37 P. 6,7 034000K401 034000K402 034000K403 ...(KA'S) ~. _ _- _ _. - __ _ _ .

8.

PLANT SYSTEMS DESIGN. CONTROL AND INSTRUMENTATION PAGE

ANSWERS -- GINNA-87/10/05-KINGSLEY, I.

ANSWER 6.10 (1.50) a. Density compensation (0.5) b. Indicated steam flow will be higher than actual. (1.0) REFERENCE LP RIC050, P. 6 LP RIC05C, LO. 1.2 LP R4401C, P.6 LP R44010, LO. 3.2 191002K102 191002K103 ...(KA'S) ANSWER 6.11 (2.00) a. Rods step in. (0.25) Automatic rod withdrawal is blocked. (0.25) b. Maximum steam dump demand (but does not <> pen valves unless armed). (0.5) c.

Steam generator level reference decreases (to 39 percent)(or FRV begins to close). (0,5) d. Load limiter runback inoperable. (0.5) REFERENCE LP RIC050, P. 7 AND 8 l LP RIC05C, LO. 1.2 j 016000A201 ...(KA'S) l l l

6.

PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE

ANSWERS -- GINNA-87/10/05-KINGSLEY, I.

>/- 6 M ANSWER 6.12 f - each channel has a level trip bypass switch [0.50] to prevent a trip a.

at low power levels (less than 10% or 25%) due to the 1/2 trip logic l scheme [0.25] - at high power levels (>25%), the trips are blo ked [0.25] by theI a b i l [0.25] vt/ml 4 pot-J P-6 [0.25] or P-10 permissives b. - the "Rod Stop Bypass Switch" removes the overpo er rod stop function for the selected channel [0.25] to allow rod motion [0.25] - the "Power Mismatch Bypass Switches" remove [0.25] the selected channel from the rod control system averaging circuit (0.25] REFERENCE RG&E Enabling Objectives 3.6,3.7,4.1 RG&E Lesson Text RGE-33 pages 5,9,13,18,23 K/A 000033 EA1.02 3.0 K/A 000033 EA2.09 3.4 000033A209 000033A102 ...(KA'S) ANSWER 6.13 (3.00) Tavg (0 0+ ig) [0.50] a.

b. no load -- 19.5% [0.25] full load -- 49% [0.25] c. High PZR level trip [0.50] 87% [0.25] with 2/3 channels [0.25] duetogetdownisolation[0.50]withchargingpumpflow[0.25] at =1 - t -".o speed [0.25] mwnum REFERENCE , RG&E Enabling Objectives R1901C-1.2,3.1,4.1 ' RG&E Lesson Plan R1901C pages 14,13 RG&E Simulator Documentation No. 6.3.4.11.3 K/A 011000 K4.04 3.0 K/A 011000 K4.05 3.7 K/A 011000 A2.11 3.4 K/A 000028 EA2.02 3.4 K/A 000028 EA2.12 3.1 000033A209 000033A102 ...(KA'S) i l l !

f 7.

PROCEDURES - NORMAL ABNORMAL. EMERGENCY AND PAGE

RADIOLOGICAL CONTROL ANSWERS -- GINNA-87/10/05-KINGSLEY, I.

ANSWER 7.01 (2.50) a. 2.

(1.0) b. 1. ECA-0.0 (Loss of ALL AC Power) (0.75) { 2. ES-1.3 (Transfer to Cold Leg Recirculation) (0.75) 0 ( Aeactor Tn'd o r i q1r ty.Inje c fidn) ( g,7 y) LP REP 500, P.

LP REP 500, LO, 1.11, 1.13 LP RFR000, P.

194001A102 ...(KA'S) ANSWER 7.02 (1.50) e.

REFERENCE ES-0.1, P.

12, 13 LP REP 000, LO, 1.7 002000A402 ...(KA'S) ANSWER 7.03 (1.00) k b, C, $, & 6 REFERENCE P-2, P. 4 003000G010 ...(KA'S) ANSWER 7.04 (1.50) 1. Verify turbine trip (0.75) 2. Verify feedwater isolation (0.75) N.sf.m b k [C 7 S) 3, (f +)

7.

PROCEDURES - NORMAL, ABNORMAL EMERGENCY AND PAGE

RADIOLOGICAL CONTROL ANSWERS -- GINNA-87/10/05-KINGSLEY, I.

REFERENCE LP REP 000, P. 4 LP REP 000, LO, 1.4 000007K301 ...(KA'S) ANSWER 7.05 (1.50) 1. At least one safety injection pump running, (0.75) and 2. RCS pressure less than 165 psig greater than the highest steam generator pressure.

(0.75) REFERENCE LP REP 000, LO, 1.5 E-0, FOLDOUT 000007G007 003000G010 ...(KA'S) ANSWER 7.06 (1.00) 1. 4 psig in containment (0.5) 2. 100,000 R/HR in containment (or 1,000,000 R integrated dose) (0.5) REFERENCE LP REP 500, P. 4 LP REP 500, LO, 1.1 194001A102 ...(KA'S) ANSWER 7.07 (1.50)

[ N* a. To minimize RCS inventory loss via the RCP seals).

(0,5) b. To prevent injection of accumulator nitrogen into the RCS.

(0.5) c. Continue depressurizing.

(0.5) REFERENCE ECA-0.0, P.11 LP REC 000, LO, 1.1 0000550007 ...(KA'S)

7.

PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE

RADIOLOGICAL CONTROL ' ANSWERS -- GINNA-87/10/05-KINGSLEY, I.

ANSWER 7.08 (1.50) 1. Notify personnel to evacuate the area (to the East end of the Auxiliary Building).

2. Notify control room, health physics and shift supervisor.

3. Consider altering ventilation lineup to increase filter flow and negative pressure in Auxiliary Building.

4. Ensure doors are closed and openings are covered leading to affected area.

5. Refer to SC-205, Local Radiation Emergency.

(any 3 at 0.5 each) REFERENCE RF-8, ATTACHMENT 2 000068G010 000036G010 ...(KA'S) ANSWER 7.09 (2.50) a.

1. 2.0 Rem 2. 2.5 Rem (0.5 each) 3. 5.0 Rem 4.

10.0 Rem (5(N-18)] b. A plant superintendent (0.5) REFERENCE A-1, P.

10 AND 11 LP RAD 62T, LO, 2.0,.;. 0 194001K103 ...(KA'S) l l .. _ _ . . . _ __._ . -

7.

PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE

RADIOLOGICAL CONTROL ANSWERS -- GINNA-87/10/05-KINGSLEY, I.

ANSWER 7.10 (2.50) a.

1. Start 1 boric acid pump.

2. Open MOV-350.

(0.5 each) 3. Verify flow.

b. 525 gallons (3 X 175 gal / rod) (1.0) REFERENCE FR-S.1, P. 4 AP-CVCS.2, P.3 LP RFRS10, LO, 1.2 000029G010 000024G010 ...(KA'S) ANSWER 7.11 (2.00) a.

+0.5 psig b.

-0.3 psig ' c. 90 hours f /5 /rs d. 70 hours l'// Sr3 REFERENCE O-11, P. 1 AND 2 029000G001 029000G002 029000G011 ...(KA'S) l ! l . -. - . _ -.. _ _ . - _ _ _ _ _ _ _

7.

PROCEDURES - NORMAL. ABNORMA1,. EMERGENCY AND PAGE

RADIOLOGICAL CONTROL ANSWERS -- GINNA-87/10/05-KINGSLEY, I.

ANSWER 7.12 (2.50) a. 50 degrees F/hr b. 50 degrees F/hr c. 25 decrees F/hr (0.5 each) d. 200 degrees F e.

50 ppm REFERENCE P-2, P. 1 AND 2 002000G010 ...(KA'S) ANSWER 7.13 (B:00) /. # a.

3 percent /hr (0.5) b. 3 steps /hr (0.5) Theplantmustoperat)It100 percent (0.25) with bank "D" withdrawn at ' c.

, least 210 steps (O.

for at least 72 u ative hours (0.25) within a 7 day operatin riod. (0.25) , - REFERENCE ObbGiO 062000G010 ...(KA'S) \\ ANSWER 7.14 (1.50) 1. RCP thermal barrier 2. NRHX (or excess letdown heat exchanger) 3. Surge tank makeup (any 3 at 0.5 each) 4. Sample heat exchanger REFERENCE AP-CCW 1, SYMPTOMS LP RAP 01C, LO, 1.3 000026G011 ...(KA'S) l ' l

8.

ADMINISTRATIVE G OCEDURES. CQHDITIONS. AND LIMITATIONS PAGE

ANSWERS -- GINNA-87/10/05-KINGSLEY, I.

\\ ANSWER 8.01 ( MKI[, 5 a.

(0,5) /@ b. 30 .5) REFERENCE A-905, P. 2 AND 3 194001K116 ...(KA'S) ANSWER 8.02 (1.50) a. After every unplanned reactor trip.

(0.5) b. STA (0.5) c. Dre::t1 := %.:a:: (0. V 5 WMY Sf 5- ' ' ' d G. 2) REFERENCE LP RAD 53T, P. 2 194001A103 194001A109 ...(KA'S) ANSWER 8.03 (2.50) a.

1. I hour 2. 15 minutes (0.5 each) , 3. 15 minutes b. 3.

(1.0) REFERENCE SC-201, P.1 SC-240, P.2 194001A116 ...(KA'S)

8.

ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE

ANSWERS -- GINNA-87/10/05-KINGSLEY, I.

ANSWER 8.04 (2.25) a.

1. 330 (0.5) /,(dbd-MYfOvMfdWJf'f b*flf 2. RRR (0.5) f . b. 1. Two pressurizer PORVs (0.5) with a lift setpoint of less than or equal to 435 psig (0.25), or 2. RCS vent (0.5) REFERENCE TECHNICAL SPECIFICATION 3.15 LP RAD 52T, LO, 1.1 005000G005 010000G005 ...(KA*S) ANSWER 8.05 (2.75) a.

114 to 119 steps on bank C (0.75) b.

1. Ensures adequate SDH (0.5) 2. Limits ejected rod worth (0.5) 3. Ensures acceptable nuclear peaking factors (0.5) c. Immediate boration (0.5) REFERENCE T.S., P. 3.10-11 T.S., P. 3.10-20 AP-CVCS.2 LP RTS100, LO, 1.4 LP RAP 06C, LO, 1.1 001000G006 001000G014 194001A108 ...(KA'S) . . . _ _. . , . _ _ _ -,

8.

ADMINISTRATIVE PROCEDURES CONDITIONS. AND LIMITATIONS PAGE

ANSWERS -- GINNA-87/10/05-KINGSLEY, I.

ANSWER 8.06 (2.00) Less than or equal to.2 uci/gm dose equivalent I-131 (0.5) and less a.

than ar equal to 84/E(bar) uci/gm of total specific activity (0.5).

b. Activity above limits would indicate > 1 percent fuel defect (or, limits ensure acceptable public radiation exposure in event of SGTR) (0.5) Increase CVCS letdown flow (to 60 gpm) (0.5) c.

REFERENCE T.S., P. 3.1-22 AND 23 AP-RCS.3, P. 3 LP RTS01C, LO, 1.1, 1.4 LP RAP 17C, LO, 1.1 000076G003 000076G004 002000G005 002000G006 ...(KA'S) ANSWER 8.07 (1.50) a. No (0.25), average of channels exceeds limits. (0.25) b. Yes (0.25), average of channels within limits. (0.25) c. Yes (0.25), average of channels within limits. (0.25) REFERENCE TECHNICAL SPECIFICATIONS. 3.10.2.8 LP RTS100, LO, 1.2.A 001000G011 194001A108 ...(KA'S) ANSWER 8.08 (1.00) (e':.}k Start onegemergency diesel generator (0.5) andtielgtosafeguardbus.7 a[/or18. (0.5) REFERENCE LP RSC57T, P.5 194001A116 ...(KA'S) .. - -.

8.

ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE

ANSWERS -- GINNA-87/10/05-KINGSLEY, I.

ANSWER 8.09 (2.00) a. All non-automatic containment isolation valves which are not required to be open during accident conditions are closed (and blind flanges are i installed where required).

(0.5) b. The equipment door is properly closed and sealed.

(0.5) c. At least one door in each personnel air lock is properly closed and sealed.

(0.5) d. All automatic containment isolation valves are operable (0.25), secured in the closed position, or isolated by closed manual valves or flanges (0.25) (as permitted by LCOs).

The containment leakage is within Tech Spec. limits.

(0.5) e.

(any four at 0.5 each) REFERENCE TECHNICAL SPECIFICATIONS, P. 1-4 LP RTS000, LO, 1.1.F 103000G005 ...(KA'S) ANSWER 8.10 (1.50)

a.

(>) 1.2 mrem /hr i b.

(>) 100 mrem /hr (0.5 each) c.

(>) 1000 mrem /hr REFERENCE A-1, P. 5 AND 6 194001K103 ...(KA'S) ,

_ _ _ _ _ _ _ _ _ _ _ _ _ 8.

ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE

ANSWERS -- GINNA-87/10/05-KINGSLEY, I.

ANSWER 8.11 (1.50) 1.

Intent of procedure cannot be changed.

(0.5) 2. Must be approved by two members of plant management staff (0.25), at least one of whom is the shift foreman who holds a senior operator license.

(0.25) 3. Change must be documented (0.1), reviewed by PORC (0.1), and approved by the station superintendent (0.1) within 10 days of implementation. (0.2) REFERENCE LP RTS600, P.

LP.RTS60C, LO. 1.1.B 194001A103 ...(KA'S) ANSWER 8.12 (2.00) a. Test A containment spray pump to determine operability.

(1.0) b.

CAF.

NOTE TO REVIEllERS: STS 3.0.3 CURREllTLY UllDER REVIEW BY FACILIT F0P. IMPLEliEl1TATION IN PROCEDURE A-52.4 ggf g[

' sm $ pd pfw

-

REFERENCE .

- 4 (/# . (7 p,)g );) '.h[ TECHNICAL SPECIFICATIONS, P. 3.3-3 A-52.4, P.2 (,y

5 026 BOG 011 ...(KA k [ ##' C b M G^ h.2 f

i - ' ' _. - - _ _ -.. _ _ _.. _ _. -

8.

ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE

ANSWERS -- GINNA-87/10/05-KINGSLEY, I.

ANSWER 8.13 (3.50) a.

1. The equilibrium containment particulate radioactivity increases by 20

percent over the normal reading, or 2. The equilibrium containment radioactive gas increases by 20 percent over the normal reading, or 3. The containment dewpoint increases greater than 3 degrees F above the normal reading, or 4. The reactor makeup rate increases by 0.25 gpm over the normal rate, or 5. The time between containment sump pump actuations decreases by 10 percent from the normal interval.

(any three at 0.5 each) b.

1. Primary to secondary leakage (S/G U-tube) (1.0) 2. Unidentified leakage (1.0) REFERENCE TS 3.1.5.1, 3.1.5.2 002000G005 002000G006 002000G011 ...(KA'S) f . i I

A TT AC N h\\6 N T e ] .... : ... . ....,,7 - ~ [ . .7 , ,. ..- (

l ROCHESTER GAS AND ELECTRIC CORPORATION e 89 EAST AVENUE, ROCHESTER, N.Y. 14649-0001 ^ ' < l I ,1t t *~ on i ' .=r a coor tie 546 2700 October 12, 1987 Robert Chief, Operations Branch Division of Reactor Safety US Nuclear Regulatory Commission Region I King of Prussia, PA 19406

Dear Mr. Gallo:

Enclosed are the facility comments on the Operator Licensing examination administered by your staff at the R.E. Ginna Nuclear Power Plant the week of October 5, 1987.

If you have any questions, please contact Jeffrey Wayland at (315) 524-4446 extension 209.

Stanle M. Spector Superintendent Ginna Production L_ - >

-. - , \\ l ! R. E. GINNA RO EXAM COMMENTS ,- Exam Date: 10/05/87 .- ,

~ -. < t

1.01

.Recommisndation: Accept.the items observed to i mainta!.n the core safety. limits as an alternate answer.for full credit.

-(Thermal power, reactor

coolant system pressure, and coolant temperature). . The parase "core' thermal. limits" has no precise meaning per Technical Specifications.

The candidate must then determine what information

is being sought in order to respond.

In addition to the intended meaning of "Hot Channel Factors",

the Technical Specification Core Safety Limits i j could also be used.

The referenced lesson plan does not refer to core thermal limits.

g , , 1.03 b.

Recommendation: Both modes of steam dump , , operation are possible with the conditions , stated.

Both no change and increase should be

accepted for full credit if consistent with the accompanying explanation.

The status of the steam dump system is essential in formulating a response.

If steam dumps are controlling average RCS temperature (AUTO-AUTO

MODE), then the response would be as indicated

by the key.

If the steam dumps are controlling on steam pressure (AUTO-MIRUAL, then the increase in RCS temperature will cause an increase in steam generator temperature and this steam generator pressure.

Steam dumps will open to reduce steam pressure to the setpoint of the controller.

This will reduce steam generator , ' temperature to the original value and tend to ! lower RCS temperature.

But RCS temperature will D 1 return to the original value.

The heat Q transfer rate from the primary to the secondary has increased, and with temperature of the steam back at the initial value, the average RCS temperature must increase: > t ~ ' . f Q = UA(Tavg - Tatm) < ! ! , i ! - -.. . - . .--.

, .

i 1.05 c.

Recommendation: Allow more negative and less negative as acceptable responses as assumptions cannot be determined with no explanation requested.

The question is asking for the change in MTC as boron changes from MOL to EOL, with the EOL value of 300 ppm being given.

The candidate could intarpret this to mean that boron is constant from MOL to EOL at 300 ppm.

This would eliminate the boron change effect and cause fission product buildup to be the dominate factor.

This would cause MTC to become less neaative.

This interpretation is probable because part b of the question specifies a range of boron, while part c states a single value.

1.06 See SRO exam comments for Question 5.12.

1.07 b.

Recommendation: Accept alternate answers (no change, increase) if properly supported by assumptions on flux changes.

Withdrawing control rods will tend to place the tip of the rod into an area of decreased flux and then reduce the rod worth.

The question states that power is being increased to 100%, and thus flux toward the top of the core will be increasing, along with the average flux.

If the candidate construes the power increase and rod withdrawal to cause an upward shift in flux that would override the increase in rod height and increase in average flux factors, then no changes or increase would be correct answers.

1.09 Recommendation: Accept time at power and power level as full credit responses in addition to power history and time after shutdown.

The answer assumes the decay heat will be evaluated with the reactor shutdown.

Although decay heat only becomes an operational concern after shutdown, the initial value of decay heat ', at the time of a trip would be a function of power level and time at that power.

L 1.13 Recommendation: Allow full credit for dilution estimates where data point fell beyond the table provided.

. The value for group C at 100 steps (line 5.11.1) is incorrect on the answer key.

The correct value from figure 2A of attachment A is 1675 125 pcm.

This in turn causes a correction on line 5.11.2 of the key to -1158 pcm.

To add 1158 pcm, with a boron worth of -9.94 pcm/ ppm, a dilution of 116.5 ppm must occur.

Starting at 1060 ppm, the final ppm will be 943.5.

The boron change table provided with attachment A is good for values down to 1000 ppm.

The candidates could n_qt, then, solve for the amount of dilution ceyond the 1964 gallons required to get to 1000 ppm.

2.04 b.

Recommendation: Accept "mechanical overspeed must be manually reset" for full credit.

Concern: TDAFWP has only 1 overspeed trip per tech manual Ref: Worthington Tech Manual 2.05 Recommendation: Either accept 5 answers for full credit or accept CNMT Recirc Fans A or D for 6th pump.

Some candidates asked whether fans were acceptable and were told no.

Strongest recommendation is for accepting 5 answers full credit.

2.06 b.

Recommendation: Full credit should be given for the following: portion of discharge flow diverted through eductor which draws NaOH from tank eductor outlet returns to spray pump suction.

No credit should be lost for not including words "liquid jet" or for not including size of piping "2" line.

See LP R2401C, Pg. 7.

Concern: Answer not correct.

i , - - - - - -

[1 .I ' I' N / ' , ' . i T '~ i I , , .

ROCHEST AN I hip AN Y - GINAS NUCLEAR STATION ' ' ' ROCHESTER, NEW YORK l t@ cQ O c,ae3 - , D ref , i

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' INSTRUCTIONS FOR 465 H. P.

NON-CONDENSING STEAhi TURBINE TURBINE SERIAL NUhtBER 26635 ' , WELLSVILLE WORKS ORDER NUhiBER U - 15336 i .l ' [ . i '> .- l l sir u.

oj $ $ .2, WORTHINGTON CORPOR ATION i

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. _ _ _ _ _ _ -

26. OVERSPEED GOVERNOR (See Figure 7) ~@ Gene ral All Worthington turbines are equipped with an emergency device to protect i I them from injurious overspeeding. This overspeed governor acts entirely indenpeno-nt of the normal speed governor. The device is set to shut down the turbine at a p.e-determined speed which is, usually,10% to 15% above the l maximum operattag speed. The tripping speed of your turbine is given on the , Turbine Data Sheet, vhich is found at the front of this Instruction Book.

The overspeed governor is located either in an overspeed governor cup, secured to the end of the turbine shaft, or in the turbine shaft proper. In either ~ case the device will be found at or near the governor end of the turbine s haf t.

The overspeed governor cup, when used, is screwed onto the end of the } turbine shaft and secured by a lockwasher, (See Figure 2). The assembly of l the overspeed governor parts is shown on Figure 7.

! Ope ration The accompanying sketch is intended to clarify the method of ope ration of a typical overspeed governor and its attendant linkage. (NOTE: Do not use this sketch when ordering replacements or spare parts; use part numbers from ~ Figures 3 and 7 for this purpose).

l Essentially the overspeed governor consists of weight (1) and spring (2).

l Bushing (4) serves to guide weight (1) and adjusting screw (3) is, as its name implies, used to adjust the pre-compression of spring (2) upon which the tripping.

speed of the turbine depends. The center of gravity of weight (1) is not concentric with the centerline of the turbine shaft so that as the speed of rotation of the shaft

i

, . t '.l ' - ! ! . increases the centrifugal force of weight (1) increases until, at the tripping spa ed, it overcomes the restraining iorce of spring (2). Weight (1) then flies j . out, practically instantaneously, and strikes trip finger (7) displacing it from l - its normal position in the direction indicated by the arrow. The motion of trip finger (7) is transmitted through connection link (8) to trip lever (9) which pivots ' in the direction shown by the arrow and unlatches trip valve (10). Trip valve (10) l is then moved up to its tripped position by spring force with the result that high .l t i pressure oil is directed to the closing side of the steam inlet valve servo-motor , l and the opposite side opened to drain. In addition, the trip oil line is opened to i drain and those devices, such as a trip throttle valve, non-return valve, etc.

! which are maintained in their normal operating position by oil pressure from this i ! line, are moved to their tripped position.

. ~. ,Te s tin g On all newly installed turbines and after work has been done on the overspeed I governor or its attendant linkage of existing turbines, the overspeed governor must be tested at least three times to assure proper operation of this important safety device. This test should be conducted as soon as it is possible to bring the turbine up to speed and must be conducted with the turbine uncoupled from . l the driven machine. To execute the test proceed as follows: I l I ( (a). Start and warm up the turbine as outlined under Starting Instructions.

(b). After the turbine speed is brought up to point where the speeu ' governor takes over control and all phases of operation are satio- - i factory, open the throttle valve wide and then close it at least a l i . quarter of a turn to preclude the possibility of it sticking and thus ' 1. making it difficult if not impossible to close by hand after it is heated.

' 19.1 I

< - , ,

. 2.07 a.

Recommendation: Accept R-ll/12 sample pump as an additional correct answer.

Concern: With alarm on R-11/12 Containment Vent Isolation occurs which closes valves to and from R-11/12 package.

This results in tripping R-ll/12 sample pump.

See LP R3901C, Pg. 10.

2.08 a.

Recommendation: Accept telephone room as an additional correct answer.

Concern: Telephone room also uses halon, see system description Lesson Text RGE-59, Pg. 10.

c.

Recommendation: Accept additional answer of , , overCurrent.

, I Concern! Motor Driven Fire Pump will also trip on overcurrent.

, Ref.

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[

h, b R z' !!?v42 M c0 e: Og-0 c c.

g _.o o o Ogg g . w in V O c'bs8 o ~' @ g <l % W9 % i S e 'S $ a W ~~ ~hA/ %; L

)hn05 *)O AG'll

f =

J r +g 9g

- -- 3

, % ic C'f % i

hWNI h N

I [y Ibtf ,_; O :i4 tri p / IC I7 N 7-p* ca ps $p I h :: h l ' E

- D l4 h l

Dl s

j

< m, > --e V s, ~ " w i t , - $ d )N v Y' h l .i w a i o , e

@ m-%A

- g ~ = i , </J........, u i .. ,... . .

_ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - 2.10 Recommendation: Give 0.5 pts for Bus 14 answ(=r.

Concern: Voltage of buses not asked for in the question (worth.3 pts) 3.01 Recommendation: Delete IR as answer for part a and give full credit for SR.

Concern: Compensation for gamma in the source range is done in circuitry.

Although there is circuitry involved in producing compensating voltage, typically IR compensation is taught as being accomplished in the detector.

See RGE-33 NIS System Description, Pg. 11.

3.02 Recommendation: Should include the following as additional correct answers: Low PRZR Pressure, and OT/._\\T Trip.

Ref.

RGE-35 Attachment concern: Enabling objective referenced concerns rod stops.

Answer does not include all trips which provide DNB protection.

,I' \\ ' 3.04 a.

Recommendation: Change answer sto "sequence finished" (time delay timed out) for full credit.

only one condition exists.

Concern: Ginna has no P-4.

Rx trip is not an input to SI reset.

See Logic diagram 33013-sheet 6.

c.

Re::ommendation: Give full credit for "2/3 PRZR Press < 1992 psig" Concern: Credit should not be given for phrase "both can be blocked".

Response not elicited by question.

__ . '~$, D 2,. - - _., _ - . _;.e __.

TRIP SETPOINT LOGIC BYPASS PURPOSE

4

~ Manual 1/2 None Operator Judgement -

Source Range 10 cps 1/2 P-6 Shutdown Reactivity High Flux P-10 Change Start-up Accident Intermediate Current Equivalent to 1/2 P-10 Start-up Accident Range High 25% Power Flux Power Range High 24% 1/2 P-10 Start-up Accident Low Setpoint) Power Range High 108% 2/4 None Over Power Flux (High Sepoint) Single Loop 91% 2/3) P-8 DNB Low Flow or or ) 1/2 RCP Breaker Open 1/1) Loops Two Loop 91% 2/3) P-7 DNB Low Flow or or ) 2/2 RCP Breaker Open 1/1) Loops g gRCP Low 70% Normal Voltage 1/2 Under-P-7 Anticipatory Loss Noltage Voltage on of flow (DNB)

- 2/2 buses RCP UF on llA/11B 57.7Hz 1/2 + 1/2 P-7 DNB + credits d.

OTNT 120% (68.4 ) 2/4 None DNB penalties b OP,WT 107.7% - penalties 2/4 None Escessive KW/FT .,

(61.4 F)

'

Pressurizer Low 1873 psig 2 /4 P-7 LOCA (DNB) Pressure Limits Range of OT8T (lead-lag compensated) Pressurizer High 2377 psig 2/3 None RCS Integrity Pressure Limits Range of OTET Pressurizer High 87% 2/3 None Prevent water L2 vel relief through , Safety valves (Integrity) RGE-1

3.05 Recommendation: Delete question.

Column B items 1 & 3 not RG&E nomenclature.

Correct answers: a.

Auto-Auto b.

Steam Dump Mode Sel Sw (SDMSS) c.

Man / Auto or Man / Man d.

Auto / Auto uses averacte Tavg e.

Auto-Auto on turb trip f.

Man / Auto or Man / Man g.

Auto / Auto (Auto-Auto is Tavg control, Man-Auto is Pressure control, Man-Man is Valve Position control) Concern: Generic Westinghouse terms used in this question, not plant specific.

See System Description RGE-45.

3.07 a.

Recommendation: Accept the following answer: - Source range will have been defeated by P-6 for both power levels stated.

P-6, 1/2 IR > , 10-10 ICA (mantal block) - At 5% power lovel trip bypass switch must be used to prevent trip - At 50% power IR trip will have been defeated.

This is done manually when power > P-10, 2/4 - PR > 8% - Level trip bypass always used when performing , P.T.

Concern: Question confusing.

Above (P6) 10-10 ICA the source range should be defeated and can be tested with no problem.

Above P-10 intermediate range trip should be blocked ' (before 20% power) and IR can be tested.

There , are level trip bypass switches for each channel.

System Description RGE-33, Pg. 10, RGE-35, ' Pg. 21, 22 i i .)

_ _.. _ _ _. _ ,,,

. _ _. _ _ _ _ _ _ __ _ __ _ - _ _ _ _ _ _ _ _ _ -. .' 3.08 a.

Recommendation Change ansv;er to average Tavg or Tavg Concern: Ginna uses Tavg not auctioneered Tavg.

See System Description RGE-20, Pg. 7 3.10 c.

Recommendation: Accept full credit for: Reactor trip breaker B Concern: Bypass breakers not normally installed.

System Description RGE-35, Pg. 25 ,

[ , r l ,

. _.

4.01 Recommendation: Delete question.

Concern Precaution in question was deleted 1-16-87.

See attached info.

. . I ! l ! l ' i l l l l l i .., - -. -. - - -,......, _ - _... -. _., ..- - -..-..-,_

ROCHESTER GAS ANO ELECTraC COUCRATION GINNA STATION PROCEDURE CHANGE NOTICE '~ ,//,

aoctov tNo

>~ s . afv No iNitea tion D At t PCN No a b NEW PROCEDURE /C INITI ATOR

<P RESPONS!BLE MAN AGER PERM ANENT CH ANGE ._ b TEMPORARY - REVIEW ONLY

PL ANT STAFF b TEYPOR ARY. PERV ANENT DUTY ENGrNEER

PERIODi0 REv?EW SHIFT SUPERv!SOR

E XPIR ATION O ATE
PORC REVIEW BY DATE NO b

< FOR TEVPORARY PCN'S ONLvi OPERATORS < STAFF ACKNOWLEDGEYENT YES %g auv atLM pp m dis uaA. ulaw a attu. Lus d. Aus4 s ' ' R . h d u L t:I o i m. u i s u a a d' Q& L ' O-W LM ! ^ RE ASON:

Os O_ w a > Doe V J s/v-

_ G '> m'?. ) PORO RECOMVENDATION.

b APPROVAL DISAPPROVAL PORO COMM EN75 OR MODIFIC ATIONS.

_ / PORC REVIEW D ATE _ EFFECTIVE DATE ~

k lb ' cat [Q0Rv i P$'$ REVIEW . ^ I

.. >; ...v.. ~..., 0-1.1:19 yc q._ a Cross-tie to turbine pump discharge.

, Valve 4360 Closed . s ' }

the discharge valves from all three auxiliary + J2.4 Close feedv,ater pumps.

MOV 3996 Closed MOV 4007 Closed MOV 4008 Closed AOV-4297 Closed AOV-4298 Closed Vent lA and 1B auxiliary feedvater pumps.

5.42.5 Start a motor driven auxiliary feedwater pump.

5.42.6 Aux. FWP On 5.42.7 As the discharge valve associated with above pu=p cycles open and begins to throttle closed, close discharge valve.

Valve # Closed of the flow control valves 4480 and 4481 feed 5.42.8 By use Maintain levels equal and at both steam Generators.

30 - 39% level.

, PRrcAUTION: Aux il ia ry -f-etdw a ter "i nw__ sh all b.a.

1-i m i t e d - t o-a-m a x imum-o f-150- ( gpm )-- .w.h e n e v e-r-e t-ee m. g+aesa t o r 1 a.v el._.is.

b+1-ow -t+. c le lo-lev el cetrnint n' M4, --When-regaining-stoc= generetor hvel f m below 15t, .h e-Auxi-1-ia=y feedwater-f ace chalt.-be--l-imited to._a . max-imuw-of-1M-gpe-unti-i--304--level-has- =b.an =echleved. -In whe event -of-a* ~ 5. 2. Ac;ue;ien involy-ino--water-lauels.

f :r b a i cw--the-f e e d ring '5t narrow- .eaege ' i... aEU!dmit-'snEy!r1T alisuva a&ninunrfiew ete af-2ac-,qADn-4 ng

h wuda;ien. the 15^ spr 1 Leit.

r-ed-net--be +=a 'ad Complying-with.

, these receristion. elli ra h a ttm 70:;itili;y of ;. cater h::::r-occurring, r used by ellevir; -tets t: :nter---the-f: d-rig.

Continuous flow rata =ust he less than 230 gpm per pu=p.

Cold water introduced to the stea= generators has a tendency to give the appearance , . h .. _ $ -

0. 1:20 l m ... .s.. .....e of a decroacing lovol, cvoid cdding wator too fact.

Suddon changcc in ' . ' the reactor coolant te=perature (on the

. order of 10 F) must be avoided.

, - , !

-

5.42.9 Alternate the use of motor driven auxiliary feedwater pumpi during single pump operation.

Change pumps every 2 hours or more frequently if necessary.

Observe pu=p for overheating.

NOTE: Secure the RHRS from service and align for safety injection less than or equal to 350 F.

C 5.43 Verify stopped.

lA RER Pump 5.44 Verify stopped.

1B RER Pump 5.44.1 Perfor= PT-2.4.1, if necessary, at this time.

5.45 Close the loop isolation valves: RHR to Loop A hot leg MOV-700 Closed Breaker OFF Fuses IN RER to Loop A hot leg MOV-701 Closed Breaker OFF Fuses IN

RER from Loop B cold leg MOV-720 Closed Breaker OFF Fuses IN RER from Loop B cold leg MOV-721 Closed Breaker OFF Fuses IN 5.45.1 Charging Pu=p in pull-stop may be removed from the pull-stop position at this time.

(Refer to step 4.2) , o y ! f. h .. l t

[. >

_ _ _ _ _ _ , g 3.gg' p g? . QOCHESTEQ GAS AQ3D ELECTQtC COQPOQATiON . ' GINNA STATION PROCEDURE CHANGE NOTICE ' **"' % , W noeswu so 0--d-l Y ,ws ns,ososts _l2-ll / fb Y' ' E l esv no nu so / s b NEW PROCEDURE [ E / [b INIDATOR <#___ RESPONStBLE M ANAGER PERMANENT CHANGE

TEMPORARY REVIEW ONLY PLANT ST AFF b TEMPORARY PERMANENT

OyTY ENG:NEER !

PERIODIC REVIEW SHIFT SUPERvlSOR I

EXPIRATION OATE PORC REVIEW BY D ATE

OPERATORS / ST AFF ACKNOWLEDGEMENT YES NO t* FOR TEVPOR ARY PCN'S ONLY1 ham %& 4.

& . v l -}-: = h /. _ .V ?

.,r __ f I

,e-pt ' ' * _ l _, _e a o * ,. <_ cn 1.

v_, - a.. o p ~ m nwn i s w m,. zu m ~ l f,,: ) . .) .. RE ASON: J [A-b '! w

b A > Oco % a s/c-h9 ta?.) 5 0' ' > , PORC RECOMMENDATION.

~ APPROV AL DtS APPROV AL

PORC COMMENTS OR M00lFIC ATIONS- .. s N-eaume e e l-/4-f7 N6-/7 " E,,EC1,vE o ATE , ORC REv;iw oATE , ' . catacoay u PS'S REVIEW ' C e O m.

.

., - . =. - - _ . i . j , ! , 4.04 Recommendation: Delete part a.

If not, then ~ accepts trip Bus 13 & Bus 15 as alternate answer.

j , concern: This question assumes candidate memorizes RNO in precise order.

It also~ implies that MG sets can be directly, tripped from CB, Tripping bus-13/15 is what should result in MG ' set trip.. System Description RGE-30, Pg.

6.

l 4.05 a.

Recommendation: If candidate used lifetime dose

as limiting, give 0.5 pts irregardless of whether L he wrote down available quarterly dose.

Concern:- Answer gives credit - for calculating , , amount available this quarter.

This is an easy !

calculation which may be done without being written down.

, i ! 4.05 a.

Recommendation: As long as the answer is ! < ' consistent with the assumed QF, full credit

should be given.

I ' concern: According to 10CFR20.4, Neutron Flux Dose Equivalent table, it takes a 40 times i stronger flux for a thermal neutron to impart ! ' the same dose equivalent as a 10 MeV neutron.

! This implies the quality factor for fast neutrons i ' is 40 times the quality f actor for thermal ' , neutrons.

In direct conflict with this,10CFR20.4 , item (3) implies the QF for all naturons is 10.

l i i It is obvious that an exact number for QF of a

thermal neutron is not clearly available.

l i ! 4.06 Recommendation: Accept the following as i additional answers: I ! t i 1.

Steam flow - feed flow mismatch - ref. Westinghouse ERGS [ ! Low Press version rev 1 (E-3, Pg. 4) '

2.

Charging pump speed increase same reference ! 3.

S/G level deviation - alternate answer for i keyed answer of "S/G level increasing in an j uncontrolled manner".

I . Concern Other symptoms exist.

! l I l l ! ! L , l I __ _ -. _, _, _. - _ _ _ _ _ _ _.. _ _ _ . -.. .=._, _._. _ _._._._._,_, _. _,. _. - - _ _ _ _ _. _. _ _ -. _, _ _ _ _, _ -..

Sh ^ ~.-C . .g .. - . Since the primary system pressure (nominally 2235 psig) is initially much ,6>, greater than the steam generator pressures (nominally 1000 psig) reactor coolant flows from the primary into the secondary side of the af fected steam generator.

In response to this loss of reactor coolant, pressurizer level decreases at a rate which is dependent upon the size and number of failed A tubes, as shown in Figure 1.

RCS pressure, Figure 2, also decreases as the steam bubble in the pressurizer expands.

Normally, charging flow will automatically increase and pressurizer heaters will energize in an effort to stabilize pressure and level.

However, if leakage exceeds the capacity of the [I-[ Chemical and Volume Control System (CVCS), reactor coolant inventory will continue to decrease and eventually lead to an automatic reactor trip signa >. If turbine load is not reduced, reactor trip will most likely occur on overtemperature AT.

For the expected case, however, turbine load will be decreased either automatically or manually so tht reactor trip will occur on low pressurizer pressure.

Normal letdown flow would isolate and pressurizer heaters would turn off on low pressurizer level.

On the secondary side, leakage of contaminated primary coolant will increase r, the activity of the secondary coolant resulting in high radiation indications from the air ejector radiation monitor and blowdown line radiation monitors.

Although these alarms may lag indications of a loss of reactor coolant, cepending on the transport time to the radiation monitors, they have sounded nearly simultaneously with pressurizer low level indications during past tube failure events and generally provide the Earliest diagnosis of a steam generator tube rupture. As primary coolant accumulates in the affected steam _ _ _ generator, normal feedwater flow is automatically reduced to compensate for_

high steam generator level.

Consecuently, a mismatch between steam flow from em _ - - .i and feedwater flow to the affected steam generator may be observed.

This _ % _ _ potentially provides early confirmation of a tube f ailure event and also identifies the affected steam generators.

however, such a mismatch may not be noticeable for smaller tube failures because of the relatively large normal p feedwater/ steam flow rates.

The water level in the affected steam generators % _ may not be significantly greater than that of the intact steam generators prior to reactor trip as the normal feecwater control system automatically compensates for changes in steam flow rate and steam generator level due to primary-to-secondary leakage.

E-3 LP-Rev.1 0202V:Ib

t

. .- - 4.07 Recommendation: Accept the following additional answers: 1.

PRZR PORVs set at 410 psig - Ginna actual setpoint see attached 0-1.1, Rev 87 2.

600 psig letdown line relief (V-203) - provides RHR overpressure protection and if RHR inservice would provide overpressure protection for RCS (System Description RGE-16, Pg. 6) . Concern: Question does not ask for Tech Spec items or values specifically.

' ,

F d I < !

. ' t ROCHESTER GAS AND ELECTRIC CORPORATION

i GINNA STATION CONTROLLED COPY NUMBER , k_) PROCEDURE NO.

0-1.1 REV. NO.

PLANT HEATUP FROM COLD SHUTDOWN TO HOT SHUTDOWN-TECHNICAL REVIEW E~/ O PORC REVIEW DATE 6'& b44 ' Q d PLANT SUfERINTENDENT ,t f-/.'.3-/7 EFFECTIVE DATE s QA Y NON-QA CATEGORY 1.0 START: REVIEWED BY: DATE TIME THIS PROCEDURE CONTAINS

PAGES COMPLETED: DATE TIME: et

l j 0-1.115 _ .. _. -.---- 4.3 A reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures s 330 F

unless; 1) the pressurizer water volume is less than ,h J 324 cubic feet (38% level) or 2) the secondary water temperature of each steam generator is less than 50 F

above each of the RCS cold leg temperatures.

(TS-3.1.1.1.K Limit) (Admin. limit less than 0 F)

4.4 Except during diesel generator load and safeguard sequence testing or when the vessel head is removed or the steam generator manway is open, no more than one safety injection pump shall be operable whenever the overpressure protection system is required to be operable.

4.5 Whenever a maximum of one safety injection pump may be operable by 4.4 at least two of the safety injection pumps shall be demonstrated inoperable at least once per twelve hours by verifying that the control switches are in the pull-stop position per 0-7.

4.6 Both pressurizer PORV's must be operable with a lift setting of 5 410 psig and aligned per 0-7 whenever the temperature of one or more of the RCS cold legs

is s 330 F, or the residual heat removal system is in operation except one PORV may be inoperable for seven days.

If these conditions cannot be met, the' Primary System must be depressurized and vented through a 1 1.1 square inch vent (s) within next p B hours.

Maintain RCS in a vented condition until both PORV's have been restored to operable status.

Venting cf system is accomplished by using step 5.39 of 0-2.2.

4.7 The No. 1 seal bypass valve should noj; be opened unless either the pump bearing temperature (seal inlet temperature) or the No. 1 seal leakoff temperature approaches its alarm level.

The No. 1 seal bypass valve should then be opened only if all of the following conditions are met: 1.

Reactor coolant system is greater than 100 psig but less than 1000 psig.

2.

No. 1 seal leakoff valve is open.

3.

No. 1 seal leakoff flowrate is less than one GPM.

4.

Seal injection water flow rate to each pu=p is greater than six GPM.

F .

_ _ _ _ _ _ _ _ _. 4.08 a.

Recommendation: Accept the following as correct answer: connect fire hose to D/G cooling in accordance with appropriate ER procedure.

Concern: This is an aux operator duty.

Control room operator should know that fire hose connection is available and what set of procedures to use (i.e., ER-D/G.2) ) l

4.08 b.

Recomendation: Accept "mechanical overspeed trip" as alternate answer.

Concern: There is another means for tripping , ! D/G - see attachment from ALCO manual.

l i L. _.,,,

, - - ALCO operating instructions 21-1h ,, , ( OPERATING DIFFICULTIES 2.

The emergency stop pushbutton (24) \\ will trip the breaker and stop the ege Investigateimmediately and determine the immediately, causes of any abnurmal conditions that de- = velop in the engine.

3.

Emergency mechanical stopping cf%e engine can be accomplished by pullint @e Most operating difficulties are caused by emergency shutdown (overspeed trip) hi mdle improper operation or adjustments. The located at the free end of the engine.

m majority of operating difficulties, however, can be detected by a careful operator and ~ ~

remedied before they develop into serious NOTE: If the engine is to be shutdown trouble, for a long period, it is recommended that the lubricating oil be circulated through the - system at least once per week. Dpen the In attempting to determine the causes of indicator cocks, engage the barring device, faulty engine operation, the operator should and bar the engine over slowly, about 2-1/2 first determine what work had been done on revolutions, while the lube oil priming pump the engine just prior to the difficulty. Causes is operated.

of operating difficulties can often be traced , back to improper adjustments or maintenance

procedures.

PROTECTIVE DEVICES Themost obvious reasons for poor engine ' operation are often overlooked. Abnormal DESCRIPTION pressures and temperatures are the most -- common result of operating difficulties.

llowever, before assuming, that an abnormal All protective devices required for safe - temperature or pressure indicates a serious operation of the engines are also incorporat-difficulty, first check to see that all valves ed in these control panels. The proteMive are in their proper operating positions and devices are listed in the "Data Sheet" publ{c-that pressure and temperature sensing de-ation along with the tripping values, vices are working properly.

e f Gauges are provided at the engine gitige Eeveral engine operating difficulties and panelas shown on Fig.1. Location of tAer-their possible causes are indicated in the mometers are shown on the piping diagrams public ation entitled "Trouble Shooting of the various systems.

Charts" (MI-17078).

OPERATION STOPPING ENGINE The system is located on the engine con-trolpanel, Fig.1. It consists of an indicat-1.

The engine may be stopped remotely or ing light and for most faults the engineTill locally by depressing stop button on engine shut down or the breaker will trip. The hairn gauge and control panel (24, Fig.1), will sound at each alarm condition and nj(ny be silencedbyoperation of the alarm silge , NOTE: After stopping engine, the re-pushbutton (23). An alarm reset pushbutton mote shutdown reset pushbutton must be de-(22) is provided to reset system after fault v' pressed before attempting to restart, is cleared.

July 1968

4.09 d.

Recommendation: Delte part d.

Concern: This question requires memorization of a subsequent action step of an FR procedure.

K/A manual provides no specific guidance for depth of knowledge.

t 4.10 Recommendation: Consider above information when grading.

CSFSTs are prioritized by function and color.

Memorization of which functions have which color paths is not required.

Concern: There is no orange path for inventory.

See RGE CSFSTs.

4.11 b.

Recommendation: Accept the following as alternate answers: 1.

RWST to charging pump suction AOV 2.

RWST to charging pump suction manual valve 3.

BAST to charging pump suction manual valve . -____-~_-_______ __ __ __

/ R. E. GINNA SRO EXAM COMMENTS Exam Date: 10/05/87 l l .. - --

-_ _ _ -. _ _ _ _ . _. _ 5.09 d.

Recommend deleting this question.

Information given is insufficient to determine the correct answer.

When the pump is operating in a system, the statement in 5.09d is FALSE.

Under ideal conditions, looking only at the pump curve, the statement is TRUE.

5.11 The-question was changed ' from " 10 E-10 ... amps?" to "... 1 E-10 amps?" by the examiner during the exam.

5.12 and 1.06 Recommend deleting this question.

The referenced objective and LP clearly states the lifetime effects on Doneler oniv Power coefficient for the four cases given in the anewer. (L.O. 3.6 P 44,45 of RRT05C attached) The answer is correct for Doppler only Power coefficient.

It is not correct for Fuel Temp Coefficient.

The factors that affect FTC are temp, fission products and Pu-240 buildup (LP RRT05C, P 39-41) ! Clad creep does not change the FTC valve at any ' given temperature.

What is changes is the j temperature change per percent power.

ST/SP ! is a Doppler only Power Coefficient term, not an

FTC term.

l FTC = AE Clad creep coes not change ST ' , l this value for any given temperature.

i l DOPC = FTC x AT - This term changes APower with clad creep and it is only a DOPC term.

Also, the Westinghouse manual discusses the

temperature affect on FTC and the Pu effect on l FTC.

It discusses densification, clad creep, l gas conductivity and Pu buildup with respect to l DOPC.

I i l l l

" _ _ _ _ _ _, _ . _ - - - - - - ~ - m /, Q6 k Sa /2- ' ,. 1 g._ ' ]3,. .._.

_ -- . j e., . EO # CONTENT INSTR / STUDENT ACTVTY/COMMNTS . 2) Example: While operating at power, the RO borates.

' CB +, -ApC, Prx'* B Tg +, + aog (immediate effect) . . Orx < OSG

Tavg +' + A0iavg (delayed effect) Result Reactor reaches steady state with . . and og = 0 O

OSG rx O

(-) (+) (+) of = co + AoCB + 00f * A0iavg f. Variation in af NOTE: Once

again just as 1) T effect on =g with C and B , DBW, and T ' avg and MTC you Recall =g = aog must get the students to ATg think of a > change in reactivity { with respect " to a given , change in fuel , temperature.

LESSON PLAN NUMBER RRT05C PAGE 39 Or

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .

% ._ EO # CONTENT-INSTR / STUDENT ACTVTY/COMMNTS ge., .-..'C' ... i.e., the more reactivity added (greater fraction of neutrons affected) by a given change:in Tg, results in a greater magnitude of g.

Plot of lo9 vs. log E for Show TP 3.6 - a different temperatures.

IR-5.18 Explain plot for doppler peaks at three different temperaures where: T1<T2<T3 AT -2 = AT -3

2

i

there is a large l For AT -2, doppler broadening change in [ and self-shielding'than for AT -3'

NOTE: The increase in absorption I capability becomes smaller and I71 smaller at higher and higher Tg.

Also, the resonance peaks as they broaden will overlap at the higher fuel temperatures j and thus the o insertion per i degree rise is smaller at a [ high fuel temperature than at a low fuel temperature.

Thus, for a given change in T, i - ' et a higher temperature , eff I loog lI AP l AK ! +, I +, ll +, ! ATg ATg l ATg l , l fl + ! NOTE: This is not a large change.

{ . > l ' LESSON PLAN NUMBER RRT05C PAGE 40 OF

, l

I [ i _ _ _ _ _. _ . . - ..

.- . , . EO f CONTENT INSTR / STUDENT (3 ACTVTY/COMMNTS . L1 ..)_. 6 b 21 Core age effect on =2 , As the core ages: Fis_si_on__ product _ poisons build - up more (-) o Pu-240 b.uilds_up_more (-) o but - no sel'f-shielding U-238 burns out --> less (-) o - In cycle 16, clad _ creep and fuel pellet swell effects have already occurred for much of the core yellding a very small change in effective fuel temperature over core life.

Much of the fuel is reused and Pu-240 is already present.

Since Pu-240 has a very large

e,(10 b), it is a very significant poison, especially at lower temperatures, as a () result of increased heat ~ t.ransfer_ efficiency,_and-the . continued buildup of Pu-240 causes PTC to become more ) over core life.

__ ' LESSON PLAN NUMBER RRT05C PAGE 41 OF

__ __ . -_- . ! EO # CONTENT INSTR / STUDENT ACTVTY/COMMNTS

, 2) Doppler-only Power Defect Show TP

IR-5.20 l Doppler-only power defect: total ao summed over a power: ) (oo) x AP ' (Ap) = , -(AP) Doppler-only Doppler-only - , power defect power coefficient

r Power Defect units = pcm . , . Values from 0 pcm at 0% to , -1325 pcm at 1001, BOL ' Values from 0 pcm at 0% to

1-1625 pcm at 1001, EOL , Core Lifetime Effects on 3.6 - ~~ poppleI-only_f_owar_C_oef ficient O; and Defect -

Four lifetime effects: Iu-240_ production, fue.1 densification, fuel to clad _ gap , g a s_e o tis t h e r mafc~o rid uc tilyity, l clad ~ creep.

'

Partial refueling: fuel ! - assembly reuse up to 3 cycles, !

largest a Doppler effect

! character early cycles when ) i mostly unused fuel assemblies

inrtalled, j . ( I

! ! ! f , i . . .. f LESSON PLAN NUMBER RRT05C PAGE 14 OT

! ! h ! . ' .

- - - -

__ _ _ _ _ _

- . _ _ _ ._.

._ . _ - _ _ _ _ _ _

, , EO f CONTENT INSTR / STUDENT ACTVTY/COMMNTS (~)1

% Fuel to_ clad _ gap gaseous the'rcal conductivity-heavy, s' low' Xe, Kr fis'sion product ~~ ' gases obstruct Helium, gap thermal conductivity +, fuel temp +: Dgppler-only power coe_fficient_nore negative with burnu ~p Fuel densification: fuel - Tisilit' size + kith burnup, fuel ~ to clad gap +, gap thermal conductivity +, fuel temp t: Doppler-o_nly power coefficient m5ie negativ,_e with burnup , Pu-240 production: 100,000 - Tirn~Pi:-T40 resonance ~ absorption peak at 1 ev outweighs U-238 inventory drop.

Total core resonance absorption cross-section + with burnup: Doppler-only_ power coefficient more negativeiwith , (]) burnup.

I' C1ad__ creep; Irradiation - effects, differential pressures cause plastic clad deformation (shrink).

Fuel to clad gap size +, gap thermal I conductivity +: ~ ppler-only Do g wer coefficient more_positiye EL1h__burnup.(less effect with later cycles) NOTE: Net lifetime effects: Pu-240 l buildup is more dominant in l later cycles causing a more negative DOPC.

l T LESSON PLAN NUMBER RRTOSC PAGE 45 OF _59

_ - _ _ _ - _ _ , . GINNA NUCLEAR STATION LICENSED OPERATOR TRAINING PROGRAM REACTOR THEORY COURSE 'b) LESSON 5.0 V INHERENT REACTIVITIES I e I I [ 5 i P-ss I

' ' /, l < s('s 01 < r2 < rs l < log ca l

AT -2 " 47 -3

2 i o - ! log I For AT -2 there is a larger change in doppler broadening and

i.e., the increase in absorption self-shielding than f or 4 -3, d smaller at a higher and higher p,

capability becomes smaller an y T.

Also, as resorance peaks broaden they overlap, further f reducing the overa.'.1 effect.

Thus, for a given change in T, at a ' higher initial Tg.

ef f ll llAof ll AP 4K 4, 4, lafl +, t.7, a T, l l12, l l Core Ace Effect In cycle 16, cla6 creep and fuel pellet swell effects have already occurrcJ for much of the core yielding a very small change in th.e effective fuel temperature over core life (680'F - 600'F).

Much of, the fuel is reused and Pu-240 5 *ar*ns**, it I b ) is a present.

Ei.Re Pu-240 has a very large na (10 very significant poison, especially at lower temperatures,-.as_a.

result of increased heat __ transfer _ efficiency,_and'the continue ~ [ buildup of Pu-240 causes FTC to become more negative with core (age.

_ _.

.______ __ _ _ _ _ x___

. i , FTC VARIATIONS T, EFFECT -- IR-5.18 .

, _ _ _ _. _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -. _ _ _ _ _ _ - - _. - - - _ _ _ _ _ _ - - - __ _ - - - - - - - - _ _ _ _ _. _ _

GINNA NUCLEAR STATION LICENSED OPERATOR TRAINING PROGRAM REACTOR THEORY COURSE hl LESSON 5.0 INHERENT REACTIVITIES .

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u n = u,

,, .. reist :tn tr:t::n :? ?:'M ' .

I ! e.

ME DOPPLER ONLY POWER COEFFICIENT i VS. POWER LEVEL AT BOL AND EOL--IR-5.19 i - _, - _ = - - ,, - _...,, _ _ _ , _, _ _, _. _.,, _ _,. _ _ _ _. _., _.. _ _ _. _ _ _ - = _..., ,, -... -. - _ - - - - --_.--

' q7 g4 - c -- - - , . (Vest ko m L L eev /hevd temperature increases, the amount of reactivity inserted per degree , of change in effective fuel temperature decreases.

[ h i g?i O_o_ppler-Only Power Coefficient and Power Defect ss2 p;.- d The Doppler-only power coefficient is a combination of the m . effective fuel temperature as a function of power and of fuel -

f temperatures (Doppler) coef ficient.

[ , d2 pcm) d_ T, 'F 4-d pcm _2 'F dP % Power) dP (% Power) dT , , ,, Doppler Only Doppler or Fuel Variation in Power Coefficient Temp. Coefficient Effective Fuel Temperature with Reactor Power . As shown in figure FND-RF-224, as reactor power increases, the ef fective f uel temperature increases and thus the Doppler-only power coefficient is defined a function of reactor power.

The Doppler-only power coefficient is always negative and its magnitude decreases as power level and ef f ective f uel temperature increase.

Typical first cycle Doppler-only power coefficients are between-15 pcm/% power and -11 pcm/% power at BOL.

The Doppler-only power defect is the net negative reactivity added to the core due to the Doppler ef fect as power is increased.

The defect is zero pcm at zero percent of full power and approximately -1200 pcm at 100 percent of full power.

A typical Doppler-only power defect for BOL is given in Figure FND-RF-235.

Variations in Doppler Power Coefficient Over Core Life ggc Over core lif e, the Doppler power coef ficient and Doppler-only ,y gower def ect change due to f our phenomena: the depletion of U-238 I and buildup of Pu-240, the change in the helium _ gap thermal , _ f conducitivity, fuel densification and clad creep.

.

2-40 l 0465C i

, d i e6 l < - -s C Y u i . '.. e , ie, , - p E ' !

,- a-10 - . N , '

, . i ~ ' z W b h - ? C - >

1- ' f E-12 - t:: t - ' ' t/

ex 's.

, '

i

-

$ l '1 l g !i . .; , , t, ll t a n 7.t-14 < F - - L <.

.' - - i - e- , 'd , g:

&C ' l,t ~ e-5l-t --

,

! l l l-16 0:

20

60

80 0 I d r. - < hh POWER LEVEL (% FULL Po*ER) ,p {i l , ' '

}i

C h ,' FIGURE FND-RF-235: DOPPLER ONLY POWER COEFFICIENT AT BOL, i th,

.? CYCLE 1 (REV 1) - -,; . . J ' w . Fast fissioning of U-238 and conversion of U-238 to Pu-239 i decreases the quantity of U-238.

The effect of the decrease in - 1 l; , ! U-238 is nearly offset by the production of Pu-240.

As shown in . ' 1 ', ; figure FND-RF-234, Pu-240 has a single resonant peak of Ii L approximately 100,000 barns at i ev.

, i 2 41 W '

.

.I - 0465C ' '

'l

i i . i i . d3 nq 1250 - .. ,, - [.fI .

'

31 .~h,l J 'f'. y '759 i5l

1000 - ?- i " .

...., - ' E

$ \\ w '

i

. $ 750 - k ! > d

E soo - tw 250 -

0

40

80 40 0 power LEVEL (% FULL PO*ER) i FIGURE FND-RF-224: EFFECTIVE FUEL TEMPERATURE AS A FUNCTION OF POWER (REV 1) The _EOL Doppler temperature coef ficient becomes more negative at lower ef f ective f uel temperatures (< 800'F) due to the buildup at the resonance of Pu-2 1 The absorption cross section, o,, , energy is so large that ef fectively every neutron of that energy is absorbed.

At low temperatures, an increase in temperature will not 2-42 i 0465C 't' 0' l \\ _.

,e 'l . ] - , e ' + ecrease the o, at the resonance energy significantly, All

neutrons at that energy will still be absorbed.

Therefore, even --- -. th h the Pu-240 is too evenly distributed to have a self-shielding rect.

a temocratura "c ene FM1 result in an increased absorption of off-resonance neutrons and no chance in the absorption y t of on-resonance neut_ronL Thus, the Doppler temperature coefficient _ r _ ,is more negative at low temperatures at EOL.

y.

, ,ec

m

" w::

g.

gl .., , kkb i @: " .s00 - bii.

y.; - n: ,

a ,

E f N ', l< t s " .;.... ! [ I t000 -

0 }.; i t '* BA f ' o , a: , k[h f j

O j,

I-1500 -

,j p il ,1 .! 00*Pkte OtttCT U , , ' ! 'L !! 2000 - l I

O to

60

400 i

l power LEVEL (% FULL power)

, 'i fe , f!GURE IND-Rf-236: DOPPLER ONLY POWER DEFECT AT BOL, CYCLE 1 ^! , (REV 1) ll 'l 2-43 W - i - l 0465C l ' --

, - - l r - - - - - -.,. _ , _ , ,e ./ /J C F I ' ' ' absorbed since no self-shielding effects are evident for Pu-240.

' The depletion of U-238 is more significant than the buildup of hu-240andtheDopplertemperature,coefficientbecomesslightlyless . [ cegative at EOL.

. ) L i i

' - Figure SNP-RF-il shows the BOL and _EOL. Doppler _temperaturej tsdhe three other factors that effect the Doppler '

coefficient influence the ef f ective f uel temperature as a f unction I of power.

t- . Initially the fuel rods are pressurized with helium gas.

! 'I Fission produces gases such as xenon and krypton which pollute the i , ){ - + helium gas, cause a reduction in the gap thermal conductivity.

The g result is a greater increase in f uel temperature for a given power }D ' increase at EOL. This is a minor effect.

, 't

Fuel densification is a decrease in the dimensions of the fuel . , { pellets.

The decrease in pellet diameter causes an increase in the

{

. fuel-to-gap distance and a greater resistance to heat transfer.

', fThus,thefueltemperaturewillrisemoreforagivenpower nerease p at EOL.

Hence, feel densification increases the magnitude of the n Doppler power coefficient.

However, the densification is very small pj g i b his has only a minor effect on Doppler power coefficient.

4r r? sa.In.,/, $:' ~ g-Clad creep is the predominant factor affecting the ef f ective l jj E fuel temperature at a given power level.

Clad creep is the l M ID a phenomena in which the f uel cladding shrinks and come in to contact ! IF C.

t c- - h} j' with the fuel.

The rate of clad creep is a function of the f irradiation and the dif f erential pressure across the clad.

Figure k i '/ ZNP-RF-26 illustrates how clad creep causes the clad to contact the .l fuel.

Al ' r d The result of clad creep is a decrease in the gap size, an , , increase in the gap thermal conductivity, and a marked decrease _in - ef f ective f uel temperature over core lif e.

This H the predominant . _~ effect.

The net effect of the helium polution, fuel densification ' 2-45

eex c i

. h ' and clad creep is to decrease ef f ective f uel temperature over core .' g.

life.

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.'4 y

ue 37o.330cr . (f - i,4 _ _. - _ _ _._.

._ g swg gJ . _ _. _ _.__. _ _... p ~g ,

_ g..e [v v / ( f -i.6 . _. EOL ~' _ _._

_ J-t

_, k... gb \\ / __ . . _.. _.. _ _ _ _. _.

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V BOL.

8 0.326 - 2.2 -- - - - - - - - - - -.- - } gg _ _ _ _ _ _ _ _ _ _. _ _. _ _ _ _ l R . . - l . -[- E O L - - - -. - -. . 2.4 . _.

._a .- .. -.... .... _. _ _ - -.. - -. . - _.

. _.

- 2.6 - _- o.324 ' . _ _.. _. _ ._ .. _ ._.

.

* ' *

i s n s s 800 600 TOO 900 900 1000 1100 1200 EFFECTIVE FUEL TEMPERATURE T ('F) g FIGURE SNP-RF-11: DOPPLER TEMPERATURE COEFFICIENT CURVE I (AT BOL AND E0L, CYCLE 1) (REV 2) p _The decrease in the rate of ef f ective f uel temperature change the pfedominant factor affecting e wi t,h____pg _ Doppler power coefficient.

Recall that:, l l l

c ) i l \\ l l 2 46 ,,; ) ' ( 0465C l

_- e.

(h) (h) = L Doppler Power Doppler Temp.

Effective Fuel ! Coefficient Coefficient Temperature Change (pcm/% power) (pcm/*F) With Power (*F/% power) ,_ l so - ' t < i . .

' b ! CL AD I.D.

I 26 - e.

I'.

! ..'g j- - ,...

as -

I L 1! I .3 PELLET ,0.

T ill .c ? .; w _ .> T: I b ) i

1 l l

1 I I I I I I

l i I I I l l l l l '

I-rf: CYCLE 2 CYCLE 3 2{ .p. CYCLE i

9 F . ,-

.! ' EXPOSURE END OF !" ,. n .e LIFE l; 7,5 ..: -

' hills F GURE ZNP-RF-26-SH-4: "CLAD CREEP" EFFECTS IN A PWR FUEL ROD

,e - , E@k,. ki ' (RE". 1)

%

. i' : j$ in Figure SNP-RF-42, at EOL the Doppler Therefore, Os illustrated Y ' ' power coefficient is always less negative than that at BOL.

[ ' '%. ? q .g{ '! i b~ l '/; -Variation in DoDDier Postr Defect with Core Life . g: -- ) h' i t l if, ) Since the magnitude of t'.e Doppier power coefficient is I b l ,,-

f. censistently less at E 01., the P.igni'ude -i the Doppler-only power

f: \\

"4 i \\

. , ,.c 2-4'l j i W .l ' l-- 0465C - .,f . - - ~ ~ ~

r . - p T

Neutrons are born in fuel as fast neutrons, but have little I ef f ect until they are slowed down in the moderator.

If a neutron passes through the fuel at one of the resonant peak energies it will , be absorbed.

As fuel temperature increases more off-resonance energy neutrons are absorbed in the fuel.

Although the probability ,th of an on-resonance energy neutron being absorbed decreases at higher [ re temperatures, the ef f ect of self-shielding in a lumped f uel causes j;glf . f the on-resonance neutron to still be absorbed.

Therefore, an increase in fuel temperature reduces the resonance escape [,$ h Br probabi',hy, p, and introduces negative reactivity.

[[3[::., E al !,.s L. ; R: - ~ F. - t@ z Two ef f ective f uel temperatures are used to calculate the f uel

4 temperature (Doppler) coefficient.

One is for only one fuel rod and ) ::pg Cl

+.e is less than the rod average terrperature.

The core ef f ective f uel - N . it temperature is the second temperature.

It is greater than the core _ " average fuel temperature.

& ;+ 00ppler-only power coef ficient is the product of the ef fective " ' fuel temperature change as a function of power and the fuel C_ L~ _ d temperature (Doppler) coefficient.

The effective fuel temperature is principally changed by varying the power level of the reactor.

As reactor power increases, C the effective fuel temperature increases.

The Doppler-only power a coef ficient is always negative and its magnitude decreases as p:>wer level and effective fuel temperature increase.

! r,- - At EOL the (Doppler-only power coefficient)is less negative primarily because clad creep reduces the change in fuel temperatu_r_e [ with_ power.

. , prG . Q f ' L ' ' )j g l?/3 2-52 0465C _ _ _ _ _ _ _ _

. 5.13 Recommend deleting this question.

The KA referenced ntates: Knowledge of: Definitions and effects of factors affecting power defect: moderator temperature defect fue:. temperature defect moderator void defect redistribution individual contribution effects (the summation of all defects) The facility objective referenced states: State a typical valve for BOL and EOL Power coefficient and Power Defect and explain the variation in Power coefficient and Power Defect due to: a.

Concentration of Boron b.

Effective fuel temperature c.

Tavg Neither the KA or the facility objective indicate that the level of knowledge required to answer the question is required.

The curve referenced in the key is the EOL curve.

The BOL curve is also in the handout and if the question was written from that curve the answer would be different.

Our analysis has indicated that a typical valve for the Power Coefficient a!!d the effect of Boron changes, fuel temp changes and moderator temp changes are what is necessary to safely operate the plant.

It is our contention that to memorization of Core Design curves is not necessary to safely operate the plant.

Also, the wording of the answers is confusing.

When discussing negative coefficients, increase could mean that the value gets more negative, or it could mean that the value is being incremented in a positivo direction.

. _

GENERAL COMMENT 5.02, 5.03, 5.05, 5.08 Each of these questions requires a significant amount of calculation.

If the intent is to determine if the candidate understands the principles involved, then multiple choice is a poor format for these questions since the work needed to arrive at the answer is not required.

The examiner cannot tell if.the proper method was used or if the candidate just "guessed" correctly.

In addition, due to the high point values associated with these questions, a candidate that knows the principles involved but makes a math mistake could fail the exam and be denied a license.

Recommend not using multiple choice questions when calculations are involved.

We suggest referring to the scrap paper to give partial credit.

6.02 Recommend deletion.

, KA referenced: Ability to operate and monitor the following: Manual control of components for which automatic control is lost Facility objective referenced: Describe the operators actions for the failure of an instrument bus , i Neither the KA or the objective requires the operator to memorize the knowledge required to answer this question.

Our analysis has shown [ and our objectives require the operator to know l the major plant responses to the loss of an I instrument bus and the actions required to l recover from the loss.

l It is not reasonable to expect an operator to memorize all equipment powered by each instrument bus.

For a loss of any instrument bus the operator has three main concerns.

A turbine , l runback will be initie bed and auto rod control ! will be lost.

He must stop the runback and drive rods in manual to stabilize.

The remaining , concern is S/G level control.

Some portion of auto control is lost with each instrument bus l l i ,_ _ . - - - -

loss.

The operator is instructed to take manual ' control of feedwater for the affected S/G.

This will work in all cases except !.nstrument bus C where manual is also lost.

In this case a Rx trip is imminent and the operators respond according to the EOP's.

See LP RICl2C for verification.

6.04 b.3 Answer should be changed to OPEN.

This is RWST suction path for SI pumps.

Answer changed by examiner at facility review.

6.07 (b-e) Delete this question.

It is not necessary for the operator to have the fail position of control board valves memorized.

The fail position of each of these valves is clearly marked on the control board.

6.08 b.

Recommend accepting NO CHANGE as an alternate answer.

LP R3001C P 20. states that a rate of ch t. to between turbine 1st stage pressure and average nuclear power is required to cause rod motion.

0F error It also states that at least a il.5 signal must be generated before rod motion starts.

If the PR channel drifts low slowly enough, a very small power mismaten will be generated, this will generate a small error signal, and thus no rod motion will occur.

6.09 Recommend deletion.

The referenced KA states that a knowledge of the fuel transfer system interlocks is necessary.

, l It does not detail the extent of this knowledge (i.e., memorization of the interlocks vs. being aware that they exist and being able to determine their values from available procedures) Our contention is that memorization is not required to safely operate the fuel transfer system.

l

' . , 6.13 a.

Change answer to Average Tavg Aucutioneered Tavg is not used at Ginna

Key changed by examiner at facility review at minimum speed" to c.

Change answer from " ... ... at maximum speed".

" An indicated low level will cause charging pump speed to increase to attempt to raise level.

Refer to LP RICO3C Pg. 4 ! l l , I i ._ ,. - - _ . ___ . -

7.01 a.1 Comment - There is not RCS inventory orange path

Comment accepted by examiner at facility review 7.01 b.

Add alternate answer: E-0 Rx-trip or safety injectica When E-0 is entered, the foldout page must be opened.

This requires monitoring of the Red Paths.

These Red Paths are not to be entered though, until the immediate actions of E-0 have been completed.

The reason for this is the same as that for when the CSFSTs are to be monitored.

This reason is delineated in the Westinghouse Background document for E-0 attached.

Alternate answer was accepted by the examiner at the facility review

- . .___ __-. _, - _ _. - - -. _ _. -.

. - _. - -. ), 0 { N,

, STEP DESCRIPTION TABLE'FOR E-0 STEP

/Rb y ' STEP: Initiate Monitoring of Critical Safety Function Status Trees . PURPOSE: To initiate monitoring of the status trees' 'E?) . BASIS: At this point in E-0, no transition to an Optimal Recovery Guideline has been made-and SI termination criteria have not been met.

The operator will remain in E-0 until either a transition to a recovery guideline is nade or SI can be (p;i' cq N terminated. Monitoring the Critical Safety Function Status Trees will ensure that the plant remains in a safe condition while the operator remains in E-0.

The basis for-the placement of this instruction in Step 27 follows.

The operator is trained to monitor the Critical Safety Function Status Trees when a transition out of E-0 is made (see Users Guide in the Executive Volume). Since a transition out of E-0 is expected, the Critical Safety , Function Status Trees are monitored soon af ter the reactor trip or safety injection. However, if the operator does not make a transition out of E-0 due to lack of appropriate synptoms, Step 27 gives explicit instruction to monitor the Status Trees while remaining in E-0.

Placement of this instruction after the verification of autosctic actions and the diagnostic sequence is due to '.., various reasons. Verification of automatic actions ensures that plant equipment is operating properly.

These steps are performed prior to monitoring the Status Trees since the proper operation of the safeguards equipment is the first means of preventing or correcting any challenges to the Critical Safety Functions. The diagnostic sequence can be performed fairly quickly and any transition to another Optimal Recovery Guideline would require tnat the Critical Safety Function Status Trees be monitored.

Hence, the step to explicitly monitor the Status Trees in E-0 follows these actions.

In addition, any extreme challenges to the Critical Safety Functions due to equipment failure are addressed by explicit transitions out of the immediate action steps in E-0.

. ACTIONS: - ' Initiate monitoring of Critical Safety Function Status Trees INSTRUMENTATION: . N/A - . P l E-0

LP-Rev. 1 0072V

7.02 Recommend deleting this question.

No correct answer exists.

In answer (e),

thermocouples are 625 F.

This is not a normal expected value following the referenced

transient.

For CETs to reach 625 F they must have increased.

The requirement for Natural Circulation is that CETs are constant or decreasing.

,

( ,, ,. .- ,, - - , --

7.03 Recommend deletion.

The operator 1s not required to memorize - this requirement.

Whenever a RCP is to be started, procedure S-2.1 is to be used.

The starting duty requirement is a precaution of this procedure and the.very first instruction in the procedure is to check to see if the starting duty requirement is met.

Ref.

S-2.1, RCP Operation . - - -. . - - - --

a, _ ROCHESTER GAS AND ELECTRIC CORPORATION GINNA STATION . ' UNIT #1 GINNA STATION , OCMPLETED . CONTROLLED COPY NUMBER ' - ~ DATT:- TIME:- . PROCEDURE NO.

S-2.1 REV. NO.

_ REACTOR COOLANT PUMP OPERATION _ f TECHNICAL REVIEW

, - J PORC REVIEW DATE d'Of"[9 ) PLANT SUPERINJENDENT / ' ' . < / v , 5-09-17 } EFFECTIVE DATE - . . QA % NON-QA CATEGORY 1.0 . REVIEWED BY: , .. i THIS PROCEDURE CONTAINS

PAGES

. .

i ! - S-2.1 1 ' . S-2.1 . REACTOR COOLANT PUMP OPERATION .- - v ' 1. O __ PURPOSE: - - -. ~.,. l.1 This instruction describes the procedures for starting normal operation and stopping a reactor coolant pump.

2.O REFERENCES: 2.1 Reactor Coolant System Flow Diagram.

2.2 CVCS Flow Diagram, Sheet #1.

2.3 RCP Instruction Manual - 5710 - 71A.

2.4 Westinghouse Nuclear Service Division Technical Bulletin NSD-TB-75-18.

3.0 INITIAL "ONDITIONS: 3.1 Plant is in any coadition f rom cold shutdown to hot shutdown.

3.2 Reactor coolant pump (RCP) is stopped.

3.3 Seal water is being supplied in accordance with Operating Instruction S-3.2, Charging and Volume ~ Control and a positive differential pressure is established across the labyrinth seal (normally 30-40-inches of water column).

For exceptions to this condition, refer to 4.1 below.

NOTE: Prior to operation of the RCP, the non-regenerative heat exchanger temperature controller TC-130 must be adjusted to the

1270F setpoint, and the seal water heat exchanger outlet temperature at TI-120 adjusted to produce a seal water supply temperature less than 1300F.

3.4 Minimum RCP running pressure (as indicated on PR-420) , has been established in accordance with Figure S-2.2.

. .me

, S-2.1:2,,, . . ' \\ NOTE: If the residual heat removal loop is in . service, observe the maximum system pressure --Nb} fp limitation and maintain reactor coolant pressure at 360 psig as indicated on PR-420

by setting of the low pressure letdown . controller PC-135.

3.5 - Vapor seal head tank (standpipe) has sufficient water, refer to 4.4 below.

3.6 Component cooling flow ~of at least 25 gpm is established to the thermal barrier cooling coils as measured by FI-610 and 614.

3.7 Component cooling flow is established to the motor thrust bearing oil coolers such that coolant flow readings on (A pump /B pump) FI-609/613 minus the total flow from FI-610/614 (25 gpm) and FI-611/615 (5 gpm) equal a flow of at least 150 gpm through the thrust bearing motor cooler.

This should maintain bearing temperatures at 1400 - 1600F.

- 3.8 RCP #1 seal leakoff flow t 0.25 gpm as read on FI-175 and FI-176.

, 4.O PRECAUTIONS: 4.1 Do not stdrt or operate an RCP unless seal injection q/: water is being supplied.

NOTE: The RCP can be operated without seal ir.jection water provided that: 4.1.1 Reactor coolant temperature is below 1500F.

4.1.2 RCP seal leakage rate is 5 gpm or less, and at least 25 gpm of component cooling water is flowing through , the thermal barrier cooling coil FI-610 and FI-614.

. 4.2 A pressure differential of at least 220 psid must be

maintained across the No. 1 seal.

This condition is met if the reactor coolant pressure is above 360 psig, as shown by the loop A hot leg pressure recorder PR-420.

This minimum pressure is based on 15 psig in . the volume control tank.

The presence of this delta P is evidenced by the extinguished No. 1 seal low delta P alarm (PT-173 and PT-174) light on the ' annunciator panel.

- '

.

- - - - - -,

___ S-2.133 ! , . . . 4.3 Never start an RCP unless its oil lift pump has been . delivering oil to the upper thrust shoes for at least two minutes.

Observe the oil lift pump indicating lights for motor operation and oil pressure.

' 4.4 Prior to initial operation or af ter the system has been - depressurized, the RCP's vapor seal head tank must contain sufficient water to ensure that the vapor seal does not operate dry.

Verify the low level alarms LA-490B and 491B are cleared.

If not, provide more water by opening valves 550A and 550B, whichever applies, and operate valve 508, 4.5 The RCP's are not designed for "jogging" operations.

l Too frequent starting of high inertia or heavily I loaded drives may damage the motor windings.

To prevent such damage the following maximum starting duty should be observed: 4.5.1 Restarting the motor: After any period of running or after any attempt to start where the motor has failed to achieve full speed before it is stopped, a restart should not be made until the motor has been allowed to cool by standing idle for a period of not less than 30 minutes.

4.5.2 Consecutive starting limits: Within any two hour period, the number of starts should be limited to a 'g maximum of three with a minimum idle period of 30 minutes prior to each restart.

When three starts or '.) attempted starts have been made within a two hour period, then a fourth start should not be made until the motor has been allowed to cool by standing idle for at least one hour.

NOTE: Only one reactor coolant pump is to be started at any one time.

Starts should not average more than six (6) per day throughout the life of the RCP motor.

4.6 Tne RCP should not be operated continuously until the Reactor Coolant System has been thoroughly vented.

' See Operating Procedure 0-2.3.2.

4.7 The nominal component cooling flow to the RCP motor bearing oil coolers must be 155 gpm (each RCP).

The flow is divided, with 150 gpm going to the upper bearing oil cooler and 5 gpm to the lower bearing oil - cooler.

-

___ _______-__ - - ____ _ _ - _ _

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. S-2.184 . . . ,, . , . . - ' 4.8 If component cooling water flow to the RCP motor is , lost, the RCP must be stopped within two minutes or before either the upper or lower bearing water - temperature has increased to 2000F (high temperature G, alarm is set at 1850F).

Ref er to procedures, AP-CCW.2 or AP-CCW.3, Loss of Component Cooling during Power

, Operation /While Plant is Shutdown.

4.9 Opening the No. 1 seal bypass valve has occasionally resulted in lifting and jamming either the No. 1 or No. 2 seal ring.

The purpose of the bypass line is to allow for additional seal injection water flow through the pump bearing for cooling purposes.

The No. 1 seal bypass valve should be opened only if additional cooling of the pump bearing is needed.

The No. I seal bypass valve should not be opened unless either the pump bearing temperature (seal inlet temperature) or the No. l' seal leakoff temperature approaches its alarm level.

The No. 1 seal bypass valve should then be opened only if all of the following conditions are met: 4.9.1 Reactor coolant system is greater tnan 100 PSIG but less than 1000 PSIG.

4.9.2 No. 1 seal leakoff valve is open.

4.9.3 No. 1 seal leakoff flowrate is less than one GPM.

4.9.4 Seal injection water flow rate to each pump is greater than six GPM.

- 4.10 Never let the volume control tank pressure fall below 15 psig when an RCP is operating.

4.11 The pump should not be operated beyond the low and high oil level alarm in the upper and lower oil pot.

4.12 Adhere to additional precautions in the RCP Instruction Book No. 5710-71A, Ref.

2.3.

. 4.13 During filling and venting, if there is any chance that the Reactor Coolant System will be equal to or less , than Volume Control Tank Pressure, keep the #1 seal ! [ outlet valves AOV-270A, and AOV-270B closed; however, do not run the pumps with these lines closed.

' 4.14 Maintain a positive differential across the thermal l j barrier labyrinth at all times during filling and ' venting.

- - i , l ,

S-2.1 5 ' ' . , . .. ., ,

4.15 A reactor coolant pump shall not be started with one . ' - or more of the RCS cold leg temperatures < 3300F f.'h unless 1) the pressurizer water volume is less than C/ 324 cubic feet (38% level) or 2) the secondary water temperature of each steam generator is less than 500F - .. above each of the RCS cold leg temperatures.

. 5.0 INSTRUCTIONS: 5.1 Starting.

5.1.1 Check that maximum starting duty, 4.5 above, is not limiting.

5.1.2 Start the RCP oil lift pump and verify that the oil pressure indicating light is on.

Let the oil lift pump run for about two minutes before attempting to start the RCP.

NOTE: A pressure interlock prevents starting of the RCP unless a minimum oil pressure is available to the upper thrust shoes of the motor thrust bearing.

5.1.3 Start the RCP (pumps must be started sequentially; any order of starting permissible).

5.1.4 Monitor RCP running current, bearing temperatures, m.

7") labyrinth seal differential pressure, seal leakoff flows and component cooling water flows and tempera- ~ ture.

5.1.5 Manually stop the oil lift pump after the RCP has run for at least 50 seconds.

S.2 Normal Ooeration.

5.2.1 Monitor component cooling water inlet temperature to RCP.

. ,. NOTE: The inlet temperature of the component cooling water to the RCP should not exceed j,. 1000F dur'ing normal plant operation.

The temperature is permitted to go to 1200F for a maximum period of 8 hours, 15 times per year.

,

5.2.2 Maximum motor bearing temperature is expected to be , 1670F (high temperature alarm is 1850F).

, . I e .. <*

S-2.126 , . .. -

' , ., 5.2.3 WhIn componsnt cooling wator flow to any RCP motor is . ' - lost,- ref er to Emergency Procedure AP-CCW.2 or AP-CCW.3, Loss of Component Cooling.

/ %_ i ) 5.2.4 Maintain a normal labyrinth seal differential pressure of about 30-40 inches water column (low positive - , differential pressure alarm is set at about plus 15 inches water column).

NOTE: A loss of RCP seal injection water may be tolerated, refer to 4.1 above.

5.3 Stopping.

5.3.1 Stop the RCP.

Oil lift is not needed.

NOTE: When plant conditions warrant, it is permissible to stop the RCP by merely tripping the RCP.

However, such operation should be minimized to minimize restarting.

5.3.2 When an RCP is stopped, either component cooling water through the thermal barrier or seal water to the RCP must be continued until reactor coolant temperature is reduced below 1500F.

5.3.3 Cooling water to the motor oil coolers should be maintained for at least one half. hour after stopping the motor.

- ,, . l l .>. t l ' s

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7.04 Add alternate answer: Step 12 - Check if Main Steam Lines should be Isolated . The purpose of this step is to ensure steamlines '

are isolated, if required,.to minimize the ' consequences of and/or terminate the mass and i energy release associated with a high energy line break.

Ref.

Westinghouse Background document for E-0 , i h e ! j i

d r i

, ' .

P ! ! ! .

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. . . ..- - STEP. DESCRIPTION TABLE FOR E-0 STEP

~~ STEP: Check If Main Steamlines Should Be Isolated . PURPOSE: To ensure main steamlines are isolated when required BASIS: ' Main steamlines are isolated, when certain setpoints are reached, to either minimize the consequences of and/or terminate the mass and energy releases T;- associated with a high energy secondary line break.

ACTIONS: ' o Determine if main steamlines should be isolated o Determine if main steamline isolation and bypass valves are closed o Close valves INSTRUMENTATION: o Plant specific instrumentation to determine if main steamline isolation is 1: required , l o Main steamline isolation and bypass valves position indication 'e CONTROL /EOUIPMENT: Switches for main steamline isolation and bypass valves l KNOWLEDGE: N/A h PLANT-SPECIFIC INFORMATION: Setpoints or means to determine if main steamline isolation is required h n

- ?

., A; E-0

LP-Rev 1 0072V

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _. ___ __ ___ ____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ ___ 7.07 a.

Add alternate answer: To inject accumulators An additional reason for the RCS depressurization is to allow the accumulators to inject to replenish lost RCS inventory.

Ref. Westinghouse Background document for ECA-0.0 t ! ! I i ! i . ! . l l , c _ . _ .. - .. -. - -.. - - - . - - -

- _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ ' ~ 7. 6 7 a . Th decrease in pressurizer level will also be accompanied by a decrease in D RCS p ssure. Without the beneHt of charging pumps and pressurizer heaters, f coolant through the seals will deplete the inventory of hb water ~ the loss in the pres urizer causing pressure to trend downward in concert with the level. This \\tr(ndwillcontinueuntilthepressurizerisenpty,atwhichtime /

flashing will occb either in the head of the reactor ve sel or in the RCS hot leg piping. At this .ointtherateofpressuredecja ill be reduced due to thelargervolumeofho\\wateravailableforflasingineitherofthese \\ locations.

If ac power is not restored, the ressurization will continue g.. kh' until eventually the entire RCS s satura at approximately the setpoint pressure of the steam generator sesfety v ives.

The time history of the pressure decay, like the pressurizer' evel transient, will be controlled by the amount of leakage from the R sea s.. OncetheentireRCSsaturafes,coolingviatesteamgeneratorsafetyvalves will maintain RCS press te and, therefore, RCP 1 akage at essentually constant values.

Seal leakage will continue to deplete the S inventory ultimately ' draining the upp head and causing steam ' voids to form in the steam generator O') \\ C U-tubes.

Sign icant voiding in the U-tubes will stop natural circulation through the RCS coolant loops, and reflux boiling will be Quired between the j core and the steam generators to remove aecay heat.

If ac po er is still not resto d, this situation will continue until enough inventory 5 ost to pr ent the removal of decay heat and an inadequate core cooling cogdition may

occur.

Plant Resconse With Operator Controlled Cooldown p.

gj{';) The scenario described above is predicated on the assumption that subsequent ' to the loss of ac power, heat is removed only through the steam generator , safety valves. Without the ability to replenish water lost through the RCP seals, this situation will eventually result in saturation of the RCS and a h stabilization of temperature and pressure at values slightly above the v conditions in the steam generator.

If the operator does nothing to change - l

ECA-0.0

LP-Rev. 1 I 0084V l l . _ _ . _,

_ - _ - - _ _ _ _ _ _ _ _ _. _ _ - _ _. . _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ __- . . this situation, the possibility of core damage will be greater tnan would g, exist if actions were taken to reduce RCS pressure and temperature below these 'S, ' conditions. This benefit derives from the reduction in RCP seal leakage that accompanies the reduction in RCS pressure and temperature.

Reducing seal leakage will extend the time necessary to uncover the reactor core and, y therefore, increase the time available to restore ac power before a potential inadequate core cooling situation can develop.

In addition to reducing the amount of water lost from the RCS, reducing RCS pressure and temperature will also reduce the differential pressure and the temperature to which the RCP g s seals are exposed and thereby reduce the rate and potential magnitude of seal (.^ degradation.

Finally, decreasing RCS pressure via secondary cooling can allow yectionofthewaterinthepassivelowpressureaccumulatorstoreplenish some of the lost RCS inventory. Thus, there are advantages to having an operator take timely action to cool the RCS in the event of a complete loss of ac power, and the analyses described in subsequent sections have considered this option.

In those analyses, cooling below the safety vabe setpoint conditions is assumed to be accomplished by coordinated manual or local control of the steam generator PORVs and the turbine-driven AFW pump.

There are several restrictions that must be observed should an operator take action to manually cool a plant without having normal shutdown systems availabic.

One restriction is related to the potential for returning the reactor core to a critical condition because of the effects of negative moderator feedback. Without ac power the systems normally used to borate the RCS are unavailable, and, unless pressure can be reduced sufficiently to allow accumulator injection, the only sources of negative reactivity to maintain the ' core suberitical are the control / shutdown rods.

There may be situations when _ it is not possible to depressuri:e below the accumulator injection pressure l without also returning the core to a critical low power state.

Such situa-tions would arise because of the combined effects of low system boron .. . ECA-0.0

LP-Rev. 1 0084V ,

- _ 7.11 c.

& d.

Allow a tolerance of i 15% in the answer The operator is not required to have the exact numbers memorized.

He should have a feel for their approximate value and then be able to look the exact values up in the procedure.

Accepted by examiner at the facility review 7.13 Recommend deletion.

This is a precaution in an operating procedure.

As such, it is read before the actions of the procedure are started.

If the procedure is in effect when shift turnover takes place, it is the oncoming operators responsibility to review the portions of the procedure which are complete (this includes the precautions).

For this reason we feel this knowledge is not required to be memorized.

Ref.

O-9 step 5.1 (Pg. 2) .

. - -- .- 8.01 a.

Recommend deletion or at least a , 10 ft tolerance.

, , The shift supervisor performs the Fire Protection and Safety Coordinator function very infrequently.

On the-rare occasion when'he would be expected to perform this function, he , is expect 2d to review the procedure.

This would refamiliarize him with the process and the . in f orma '.lon needed for him to perform this ' function.

i According to A-503, step 3.1, a procedure is required to be used for maintenance, repair, or inspection activities for which a procedure exists.

In accordance with A-503, the shift supervisor , must une the procedure to perform this function.

Therefore, memorization of the requirements is

not necessary.

A tolerance was agreed to by the examiner at the facility review.

The value of the tolerance , ! waa not agreed upon.

, ! 8.01 b.

Recommend deletion.

This is the responsibility of the firewatch, not i the Firs and Safety Coordinator or the Shift

i Supe ~ir-- ! , Tits shift Supervisor may only neume the ! re-sponsibility of the Fire end S tfet,y Coerdinator , 4~ for issuance and approval of r-Laits.

The Shift , Supervisor may not be used as a firewatch since

,

he would not be allowed to leave the area.

, According to A-52.1, steps 3.4.4 and 3.4.5, the ' Shift Supervisor must be free to go where he

feels h6 is mv.ded, and is required to be in the l control room during emergencies.

l f -

, ! !

! .

, - I . i

i.

> > -w -..~, .._,y~ .- --,---.---.,,..,w--... --_ _. _ - - - -, _, _ _ _ _ _. _.,, _ _ _ _ _. _ - -

. -

4 8.02 b.

. Recommend deletion.

, This is the STA's responsibility, not the Shift Supervisors.- The Duty Engineer, Operations Manager, and Superintendent would ensure the review was done by the ST7.. . 8.02 c.

Modify answer to read: Superintendent (0.3) Operations Manager (0.2) According to procedure A-25.4, both the Superintendent's and the operations Manager's approvals are required.

Ref.

A-25.4, steps 3.6.2 and 3.8.1

Accepted by examiner at facility review

< 8.04 b.1 Accept alternate answer: Overpressurization system operable Implicit in system operability is that both PORVs are operable with a lift setpoint of $ 435 psig.

Ref.

A-52.4, step.3.1

Accepted by examiner at the facility review 8.08 Accept alternate answer:

, Start D/G(s) and load onto safeguards ' bus (es) '

According to A>503, step 3.1, a procedure is to d be used for operational activities, if ona i exists.

Therefore, the operator, once realizing he had steady hurricane force winds would enter SC-2 which would give him the specifics of how many D/Gs to start and which buses should be loaded onto the D/G.

He should have an overview of the procedure (i.e., know that a diesel or

diesels should be started and that safeguards buses must be loaded) i <

-_.

... - m - ... . .- - - - .. . -

,- - ~

8.12 b.

Recommend deletion.

KA references: - Knowledge of limiting conditions for operations and safety limits Ability to recognize indications for system - operating parameters which are entry-level conditions for technical specifications Having the "ability to recognize the indications" does not necessarily mean the definition of operable must be memorized.

Our operators are trained that if there is a questions as to whether a piece of equipment is inoperable, the expanded definition in A-52.4 should be consulted.

We recommend that they do not memorize this definition as this may lead to misinterpretation if it is not accurately recalled.

It is the operators responsibility to be able to recognize when the definition applies and be able to correctly interpret it.

We believe this satisfies the KA requirements referenced.

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

__ _ _ __ __ _ _ _ _. _. _ - - _ - _ _ _ _. _ _. ' , i ..{ ATTACHMENT 4 . l ! NRC Resolution of Facilitv Commonts and Additional' Changes to Answer Key as a Risoit of Examination Grading RO Examination Response: . I 1.01 s Comment noted.

Answer key unchanged.

While "Core Thermal Limits" is not [ specifically defined by Technical Specifications, it is a term that is used

in the facility Lesson Plan to classify "Hot Channel. actors."

t "

1.03b.

. Comment noted.

Answer key modified as follows to inclu1e the condition when steam dumps are in AUTO-MANUA'_ mode ! ' INCREASES CO.253 because rod withdrawal adds positive reactivity CO.253 which is offset by MTC CO.253 causing steam dumps to open CO.253 1.05c.

Disagree with comment.

Question clearly states to consider EACH of the conditions given.

Also, question is clear in stating that boron is at 300 ppm.

RGLE reference Inherent Reactivities - Student Handbook - MTC Variations, iigure IR-5.16a, shows that MTC becomes MORE negative due to core life distribution changes.

Answer key unchanged.

1.06 '- Question deleted.

See SRO question 5.12.

Reduce section value by 1.00 ' point.

1.07b.

Comment noted, however, question graded according to answer key.

Question was written to elicit a resporAe at the RO level of knowledge.

Facility comment assumes a level c4 knowledge well beyond what most RO's would be j expected to know and what is taught.

1.09 , i Accept alternate answers.

Add "power level and time at power" to answer l l key.

t

i l L ~

_ - - - _ _ - - - -. _ __ _ _ , !? , . 1.13 Comment accepted.

Full credit given for determination of dilution by _ , estimation.or by use of tables provided.

) 2.04b.. [ ! Comment accepted.

Answer key modified to delete "electrical overspeer] _ resets automatically" and points redistributed to the one correct-answer.: 2.05

Comment noted.

Answer key modified to eliminate "TDAFWP" as an answer ' leaving five pumps required for full credit.

Points redistributed to the , remaining five answers.

' 2.06b.

Comment accepted.

Answer key modified accordingly.

2.07a.

Comment accepted.

Added "R-11/12 sample pump" to answer. key.

, I 2.08a.

, Comment accepted.

Added "telephone room" to answer key.

[ ' ! ! i 2.Clc.

Comment accepted.

Added "overcurrent" to answer key.

2.10

, Comment accepted.

Deleted reference to voltage and points redistributed.

I r ] 3.Ola.

Comment accepted.

Answer key modified to delete "!R" as an answer.

Points redistributed to the remaining responses.

,

._ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ - _ 3.02 Comment accepted.- "OT delta T" and "low pressurizer pressure" trips added to-answer key.

3.04a.

Comment accepted.

"P-4" deleted on answer key.

Section value reduced by ! , 0. 50 poi r ts.

3.04c.

Comment accepted.

"Both can be blocked" deleted from answer key.

Points redistributed.

3.05 Comment accepted.

Question deleted.

Section value reduced by 3.00 points.

3.07a.

Comment noted.

Answer key modified as follows: SR reactor trip blocked by P-6 CO.303 or bypassed by using the level trip , bypass switch CO.303 due to 1/2 trip logic CO.15] IR rcactor trip blocked by P-lO CO.303 or bypassed by using the level trip bypass switch CO 303 to prevent trip due to 1/2 trip logic CO.153 l ' 3.OBa.

' Comment accepted.

"(avg Tavg)" added to answer key.

I a 3.L9c.

See Sh3 6.13c.

i 3. toc.

, Comment noted. Answer key modified as follows: "loss of (DC) control power to train "B" equipment [0.303 causing a reactor trip CO.453" i i

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - 4.01 Comment accepted.

Section value reduced by 2.00 points.

4.04 Comment accepted.

Answer key modified to include "trip bus 13 and bus 15" as alternate answer.

4.05a.

Both comments noted; however, a full credit answer must recognize that lif etime dose is more limiting than quarterly dose.

Addi ti on al l y, RG&E lesson plan RRCO3C page 6 uses a quality factor of 5 for thermal neutrons.

If the quality factor for thermal neutrons is unclear, then the facility training material shoulo indicate this f act or resolve the issue.

4.06 Comment accepted for parts 1 and 2.

Part 3 not accepted.

S/G 1evel deviation does not adequately describe the behavior of S/G 1evel during a S/G tube rupture.

S/G 1evel increasing is acceptable.

Answer key modified to add "steam flow / feed flow mismatch" and "charging pump speed increase" as alternate answers.

4.07 Comment noted.

Answer key modified as follows: "2 PZR PORVS cet at 410 psig" and "600 psig letdown line relief" as alternate answers.

4.08a.

j Comment accepted.

Answer key modified as follows: "connect a length of fire hose CO.353 from a fire hydrant CO.353 to the D/G ! service water connection CO.303' 4.08b.

Comment accepted.

Added "mechanical overspeed trip" as alternate answer.

l . ~ _. -

, - _ _- _ - -- _. - - - - _ _ _ _ - - _ - - - _ _ - - - - _ _ - 4.09d.

Comment accepted.

"If core exit TCs are greater than 1200 F" deleted from answer key.

Section value reduced by 0.50 points.

4.10 Comment accepted.

Inventory, answer "2", deleted from the answer key.

Points redistributed accordingly.

4.11b.

Comment noted.

Answer key modified to delete actual valve numbers f rom the required answer.

i.

SRO Examination Resonse 5.09d.

Comment accepted. Question deleted. Section value reduced by 0.50 points.

5.11 Comment noted.

5.12 , ' Comments noted. Although the ref erences provided for the question were incorrect, the question is valid. Further review determined that there are two correct answers for the question as written. However, as only one response was asked for, question deleted from examination due to possible

l conf usi on among the candidates. Section value reduced by 1.00 point.

i 5.13 ! Comment accepted. Questi on deleted on the basis that wording of the answer selections may have been confusing to the candidates. Section value reduced by 1.00 point.

5.02,5.03 i ! 5.05,5.08 Comments noted.

! ! l t

. _. .- - _ ____ ___ _ _ _ ___ _ _ _ _ _. _ _. -. _ _ _ _ _ _ _ _ _ _ _ _ __ . _ _ _ _ _ _ _ 6.02 Disagree.with comment. The question did not ask for specific component responses. The fact that the plant will trip due to the loss of the vital bus is just as important as the operators knowing that the loss of the vital bus-will-also cause loss of manual control of systems at the main control board which could complicate post-trip recovery actions.

6.04b.3 Comment noted.

6.07b-e.

- Comment not accepted. The question is based upon a high value KLA objective and asks for component responses to an abnormal condition. The question is appropriate for the examination level.

The operator should be able to determine a valve's failure mode based upon its application in a system.

i 6.08b.

, . ,

Comment accepted.

NO CHANGE "'added as alternate answer to the key.

" ! , 6.09 i Comment accepted. Question deleted. Section value reduced by 2.00 points.

' 6.13a.

I Comment noted.

! 6.13c.

I Comment accepted.

at maximum speed.." incorporated in answer key.

l " ..

7.Ola.1 l Comment noted.

t 7.Olb.

I ! f Comment noted,

, mm.--v~ -,-yy, w - rn,-- n

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. __ __ _ . _ _ _ _ - _ _ _ _ 17.02 Comment not accepted. The question tests a candidate's ability to apply natural circulation criteria to a set of conditions. The question did not imply that CET temperatures were changing yet it clearly indicated that.

the other three parameters were changing.

! 7.03 . Comment noted. It is important that operators be aware that time limits associated with restarting an RCP exist. We agree that while a specific value need not be memorized awareness of a time requirement is important.

The answer key was modified to accept any answer except a.

, 7.04 Comment accepted.

Check if MS line should be isolated added to answer " " key.

7.07a.

Comment accepted. " To inject ac cumul ators added as alternate answer " to the key.

7.11c,d.

Comment noted.

7.13 Comments noted. Deleted part c of the question. Section value reduced by 1.00 point.

8.Ola.

Comment accepted. Tolerance of +- 10 feet added to answer key.

8.Olb.

Comment accepted. Question deleted. Section value reduced by 0.50 points.

_ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ 8.02b.

Comment not accepted. The SS is responsible - f or ensuring his crew members . . perform their assigned responsibilities.

8.02c.

Comment noted.

8.04b.1 Comment noted.

8.08 Comment accepted.

Start D/G(s) and load onto safeguard bus (es) " added " as alternate answer to tne key.

3.12b.

Conment not accepted. The question is testing for general knowledge and app.'ication of an important statement in procedure A-52.4.

Additional Comments Due to Examination Gradino: 1.01 Answer key modified as follows due to inconsistency between the f acili' y training material and plant operating documents (technical soecifications): "control banks rods together with no individual rod > +/- 12 steps from bank demand position" 1.03c.

The answer key was modified to ensure all correct responses were given appropriate credit as follows: INCREASES CO.253 because the rod control system will respond to the increase in F(impulse) CO.203 and the Tref /Tavg mismatch due to the increased steam flow CO.20] causing rods to wi thdraw CO.103 adding positive reactivity CO.103 increasing temperature per program CO.15]

, . .. . __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ I.

2.01c.

The answer key was modified to ensure all correct responses were given appropriate credit by adding "RCP bearings" as a fourth answer.

2.03b.

The answer key was modified to ensure all correct responses were given appropriate credit by adding "manual charging", "alternate charging" and "auxiliary spray" as alternate answers.

2.07a.

The answer key was modified to ensure all correct responses were given appropriate credit by adding "containment ventilation isolation" as an answer.

3.03a.

The answer key was modified to ensure all correct responses were given appropriate credit by adding "overpower (P-1) rod stop" and "< l.71 degrees F (f rom OTWT) " to the answer.

3.03c.

Answer key was modi fied to ensure all correct responses were given appropriate credit by adding "out motion is prevented [0.253 due to OPWT or OTWT < 1,71 degrees F [0.25]." 4.08b.

Value of question was only 0.50 not 1.00 points.

Section value reduced by 0.50 points.

4.09c.

Answer key modified by requiring "and other overheating concerns" as an additional answer.

4.11b.

Answer key was modified to ensure all correct responses were given appropriate credit by adding "(FCV-110A) normal boration path."

, _. _.......... }}

v t e Inspection Report - Ginna - 1987022
80 \| 81 \| 82 \| 83 \| 84 \| 85 \| 86 \| 87 \| 88 \| 89 \| 90 \| 91 \| 92 \| 93 \| 94 \| 95 \| 96 \| 97 \| 98 \| 99 00 \| 01 \| 02 \| 03 \| 04 \| 05 \| 06 \| 07 \| 08 \| 09 \| 10 \| 11 \| 12 \| 13 \| 14 \| 15 \| 16 \| 17 \| 18 \| 19 \| 20 \| 21 \| 22 \| 23 \| 24 \| 25 \|
1987 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 7(0) 8(0) 9(0) 10(0) 11(0) 12(0) 13(0) 14(0) 15(0) 16(0) 17(0) 18(0) 19(0) 20(0) 21(0) 22(0) 23(0) 24(0) 25(0) 26(0) 27(0) 28(0) 29(0) 99(0) Security Operator Exams Emergency Planning Other

IR 05000244/1987022

Text

=t.tu y;n mcy

Dear Mr. Gallo:

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