ML20135B551

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Exam Rept 50-397/OL-85-01 of Exams Administered on 850529-30.Exam Results:Three Senior Reactor Operator Candidates Passed Exams
ML20135B551
Person / Time
Site: Columbia Energy Northwest icon.png
Issue date: 08/26/1985
From: Pate R
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:
Shared Package
ML20135B548 List:
References
50-397-OL-85-01, 50-397-OL-85-1, NUDOCS 8509110098
Download: ML20135B551 (39)
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U. S. NUCLEAR REGULATORY COMMISSION REGION V Examination Report No. 50-397/0L-85-01

! Facility: Washington Nuclear Plant No. 2 Docket No. 50-397 Examinations administered at Washington Nuclear Plant No. 2. Richland, 4 Washington from May 29 to ay 30, 198 > 7 ,

l Chief Examiner: _ ~ ( [ [ _' f/ 2 .8 I '

'; R.J. Pate, Chief Date Bigned Rea t r Safety Br deh (Acting)

Approved:

- ( _ -- ,

/ W.J. Pate, Chief

[Date Cbifaned 95

Operations Section 1

l Summary:

4 Examinations on November 6-8, 1984 Written examinations were administered to four SRO candidates. An operating examination (oral) was adminintered to one SRO candidate. Three SRO l candidates panned the examinations.

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REPORT DETAILS

1. Persons Examined Examinations were administered to four senior operator candidates.
2. Examiner:

R.J. Pate

3. Examination Review Meeting An exam review meeting was held immediately after the written exam was administered, on May 29. 1985. The following utility representatives were in attendance:

John Wyrick Mark Westergren Tim Messersmith Sam McKay Additionally, the following NRC representative was present Robert Pate The responses to the comments provided by the utility representatives are included as enclosure (1). Additional comments were provided by letter from C.C. Sorensen to J.B. Martin, dated June 10. 1985. The responses to these comments are included as enclosure (2). Where applicable the examination keys have been changed.

4. Exit Meeting An exit meeting was held with the facility on May 30. 1985. The attendees were NRC:

Robert Pate - Chief. Reactor Safety Branch Utility:

Jack Shannon - Deputy. Managing Director Jerry Martin - Assistant Manager Director for Operations Chris Powers - Plant Manager. WNP-2 John Wyrick - Nuclear Technical Training Rick Stickney - Manager Technical Training l Jack Baker - Assistant Plant Manager. WNP-2 i s

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The examiner reported that there were no candidates that were a clear pass on i the Operating Examination (Oral). The criteria used for determining whether a j candidate passed the oral examination was discussed.  ;

a l The current status of the plant simulator was discussed. The facility staff stated that the repair of the simulator should be complete prior to the next NRC exam.

The small number of candidates (one) taking the oral examination made it possible to identify any weaknesses in the facility training program.

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The examiner noted that many of the comments on the examination review

resulted from incomplete or out-of-date reference material. The examiner accepted the facility cosaments when based on additional information. However, the effort of making the comments and NRC having to respond to the comment could be saved with better reference material.

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.i Enclosure (1)

I SRO EXAM REVIEW COMMENTS AND RES01.UTIONS

) Comments on the following questions were accepted and the master answer key suitably modified:

SRO EXAM t

Section 5: No comments accepted without examiner comments.

f Section 6: 6.01 ; 6.02; 6.07.

Section 7: 7.03; 7.06; 7.09; 7.11.

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Section 8: 8.06; 8.07.

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! Comments on the following questions were not accepted as explained below

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j SRO EXAM ,

Section 5 Question 5.01c: +

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1 Facility comment: The discharge head will change only by'a little l with increasing temperature. Should accept little l or no change.'

I i Response Will accept little change, but not no change.

Question 5.02b: -

Facility Comment Question does not require the calculations be shown.

Response: The calculation in the key does not have to be shown

A correct answer will be accepted for full credit, but

! no partial credit will be.given without the calculations i being shown.

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Question 5.04 Facility Comment: Reviewer provided suggestions on assignment of partial credit.

Response Distributions of partial credit is at the discretion of the examiner.

Question 5.05 Facility Comment: Reviewer request that full credit be given if candidate says how the power changes locally and overall without describing the effects discussed in the answer key.

Resolution: Full credit will be given if the candidate describes what is happening in the core. No change to key.

Section 6 .

Question 6.04 Facility Comment: Manual should not be one of the required responses.

The low FW flow in initiated at 30% flow.

Responses Manual will be required for full credit. The 30% flow will be included in the answer key.

Section 8 Question 8.02 Facility Comment: Memorization of tech. specs. should not be required.

Response Overtime limits have a direct affect on the operator and senior operator and the limita should be known.

No change to answer key.

u.-___-_-___________--__-___-_

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1-Enclosure (1) 1 Response to Facility Comments Provided in Sorensen letter.

Question 7.09 l

! The answer key has been changed to accept other items listed as 4 indications of Jack of containment integrity as requested by the

! facility.

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s i U. G. NUCLEAR RCCULATORY COMMIGGION SENIOR REACTOR OPERATOR LICCNGC EXAMINATION f I rACILITY: _WNE-Q___________________ f

] RCACTOR TYPC: _QWB; QED _____ _____ _____ _

i DATE ADMINISTERCO:_DDigDfg2________________

f CXAMINCR: _@ujb_EgbL_______________

.i APPLICANT: __________________.______

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INQIQVCIIQNG_IQ_6EELIC6HIa Use separuta paper for the answers. Write answers on one side only. [

f Utmple question sheet on top of the answer sheets. Points for each

questiun are indicated in parentheses after the question. The passing

] grade requires at least 70% in each category and a final grade of at s I lunst 00%. Cmamination papers will be picked up six (6) hours after l

) the esamination starts. ,

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1 l  % OF '

i CATCCOHY  % OF APPLICANT'G CATCCORY

_ VALUG. _JDIot. ___GGDBC_._ .McLUG_ ..._______.___GeIEGD9%_____________

j _CD400_. .C3.QQ _c....____. ________ 5. THCORY OF NUCLEAR POWCR PLANT OPERATION, FLUIDG, AND

!' . THCAMODYNAMICU i

j _CO*QQ_. _CD*QQ ....._____. ________ G. PLANT HYGTCMG DCGIGN, CONTHOL,

AND INGTRUMENTATION J

{ _CE*QQ__ _CQ.QQ ....______. ________.7. I'ROCCOURCG - NOAMAL, ADNOMMAL, i '

LMCAGENCY AND RADIOLOGICAL

, CONTOOL i I

_23.Q0__ _Ca.QQ .__.__.____ ___.__.. U. ADMINIGinATIVE Ph0GEDURCG, CON 0!TIONG, AND LIMITATT.ONG i h i 100.Q0_. IQQ.QQ ___________ ___ .... TOTALU ,

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, FINAL GRAOC _________________%  !

i I l All work'done un this wammitiatiun is my own. I have neither [

!' Given nur received mid. . i i

____.._______.____...._____________ E

! APPLICAN T # D UIGNA TURE  !

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D , ._ _II IC Q B Y_0C_ NWGL G OB_ E DWG B_ EL oSI_ D EGBoIIDU4_ E L UIDD._o!fD Pact c D

IUGBdQD1dedIGD QUCGTION 5.01 (3.00)

Duscribe llOW and Wily centrifugal pump dischargo hand in affacted for each of the following (considor each condition suparatelyl' .

a. Guctinn pressure increases. 11.0)
b. The discharge valve is throttled closed. (1.0)
c. The t uenpo r a t ur e of the fluid being pumped increatus. (1.0)

CNPGH is not lost.3 QUCGTION 5.02 (3.00)

The reactor is suboritient with a Keff of .'JS a GAM countrate of 000 e;p u . The control rudu are withdrawn und the now countrate is 400 cps.

.4 . Blow much "eautivity was added? (D.01

b. What would be thw statum of thw reactor if the u manes amount of rwactAvity, duturmined in a., was addud again? (1.0)

QUCGTION 5.03 (3.00)

C< plain what happutin in the cure and why, when r oc: t reula t ion finw i ra ducreamud, whilu at power and with no contrul rod muveement.

QUCGTION 5.04 (3.50)

a. The and of cure life ur und of c yc l e, tu unsually deifino un what? 10.51 3 b. Describe the three methuda of entendinu nporations beynnd and of cycle. 13.01 1

(***** Caf f_CODY 05 CON TINUCD ON NCX T PAGC * * * * * )

3 __ItEQ01_QE. NUCLE 88_EQWER_EbeNI QEEBeI1QN._ELWIDS._eNQ PAGE D ILEB50Q1Ned1C2 QUCGTION 5.05 (3.00) l Describe the core rumpanse, both lucal and overall f or each o f the following. (Assume power level is greater inan 75%.) ,

m. The withdrawal of a Deep Control Rod. (From a deep position to another deep position.) C1.53 '
b. The withdrawal of a Ghm110w Control Rod. C1.53 QUCGTION 5.06 (3.00)

Deluw is 11 stud some of the data for a recirculation pumps at vlow [

und fast speed. Using the data provided, determine the values for the four 14) missing parameters.

Glow Upeed Fast Opeed 15 HZ 60 HZ 450 ApH --m-- APM

--b- CPM 47,D00 CPM

--c-- ft af head 005 (t af head 150 HP --d-- HP  ;

(Ghow all work.) (4 W 0.75 en) (3.01 -

v QUCGTICil 5.07 (D.00)

A reactor has.just sc r anened from extended full power operation.

Ten (103 hours0.00119 days <br />0.0286 hours <br />1.703042e-4 weeks <br />3.91915e-5 months <br /> later couldown in complete, and the 001 is measured at that time to be 1% dk/k. Describe the changes, if any, to the COM  :

fur the next DO hours. (Include in your answer any adverse condi-tions,)

i QUCGTION 5.00 (D.00) 9 Captain Wily cure urificing is necessary and HOW ori ficing accomplishee y this purpose. 12.0) l

! l***** CATEGORY 05 CONTINUCD ON NEXT PAGE *****)

5.. IUCOBY_QE_NUCLEoB_EQWER_ELONI_QECBoIIQN._ELUIQG._oNQ PAGC 4 IUCBdQQYNod1GQ QUCGTICN 5.09 (2.50)

With regard tu the MAPuHCR t h e r m.41 limit-

m. Driefly, M4A T is the reason, ur bases for having a MAPLHGn thermal limit? (1.0)
b. M11Cl4 TWO o f the following four parameters affect the MAPLHGR LIMIT 7 (0.5)
1. Moderator Temperature U. lypu ut fuel
3. Fuel expusure 4, neactor pressure
c. IF an P-1 isselected on the Procesu Computer, the program pruvidos, .unu n g other things, MAPRAI. WilAT is the reintion-ship between MAPRAT and MAPLilGR7 (1.0) 9 I

(***** CND Uf" CAT EGunY 05 * * * * * )

G.__ELodI_QYQICdQ_DCGIGN. CQNIBQL._6MQ INGIGudCUI6IIQM PAGE 5 QUCGTION 6.01 (3.001 With regard to the Doactor Vussul Luvul Instrumentation, duar ribo tho five (5) types of levels used. (Includa rangu, refurunce point for zero and, YCO or NO if any automatic functions are initiated.) (3.0)

QUCUTION G.02 (3.50)

The Luakage Octection Gysta.'m can bw divided into two general groups:

abnormal lunkago WITHIN the primary containment and abnormal leakage l'UTCIDC thu primary con t a i nnwn t . The Lankagu Detection Gyatum provides INDICATION, ALRAM, and ICOLATION GICNALG for what system components?

(Includu both loutdo und Outside Primary Containment.) \ 2 QUCG TION G.OJ ( *.) J J 6 l With r eegard tu the I n t e r media t o Mange Munitoring Gystem (IRM), answer tho f u llowi rig quwetionu.

a. 'Jha t i. o 11.4 pur pum. of the range owitches? (0.51
b. Thu INM is ruadittu 24 un range 5.
1. What would be the reading if range G wa's solucted? CO.33]
2. What would be thu reading if range 7 was aulacted? Co.333
3. Uuuld thuru be any alarm +a resulting from the soluction of rango 77 CO.343 (1.0)
c. What three cunditions will renuit in an IhM inoperative nod Dluck? C3 G 0.5 ea] (1.51

.o GUCGTION 6.04 (3.00)

Thrs rucirculation pump will be downshifted from fast speed tu ult w I

  • upued for any of five reasons.

al What are the five (5) reasons? . (1.5) b) What aru the sequence of evwnta that the.rectreuintion system undurguus (include incomplatu transfer). (1.51

(***** CATCC04Y OG CONTINUCD ON NCXT PAGC *****)

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S. PL,6NI_91EIEUS_QCSJgN _QQNIBQL_6N9_INQIBydgNTel190 PAGE 6 QUCCTION G.05 (3.00)

, a. When the DEH is in Auto Turbine Control:

[ 1. The ASC programs are in control of what three items when in speed control?. CO.G3 d) 8h U. What two items are the ASC programs in' control o f when in load control? CO.43 4' 6.l LA*

J .b. Driefly describe the four modes of turbine operation used at WNP-2. C2.03 (3.0)

CUEGTION G.0G (3.00) 5 iWhen the HPCS system logic is initiated, what components receive

< signals und what type of signals are thaty? (i.e., open, close, l

g, start or stop) (3.0)

QUCGTION G.07 (3.50)

Concerning the Reactor Water Cleanup (RWCU) system:

n. List two conditions that will cause the blowdown flow control valve (FCV-DD) to auto close. (1.0) 9 b. Chauld.the blowdown restricting orifice bypnos valve (V-Gil be open with reactor pressure greater than 125 psig? Explain your answer. (1.0)
c. What conditions (list four) will close the inboard isolation w* eve (V-117 (1.0)

Valv e d - How would a loss of service air affect system operation? (0.5)

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G.__ELeNI_GYSIEMS_DEQIGN._CQNIBQL._68Q_INGIBudENIoIIQN PAGE 7 QUESTION G.08 (3.00)

Assume the RCIC system receives an initiation signal, all system components function properly, except the items listed below.

Cach failure is present prior to the initiation signal being received.

Describe the RCIC systems response for each of the following and justify your answer. Consider each item seperately,

a. The turbine exhaust valve (RCIC-V-GO) is stuck shut. (1.0)
b. The Ramp Generator portion of-the RGSC (Ramp Generator Gignal Converter) has failed producing a large signal. (1.0)
c. The D/P cell, for the RCIC flow control element, has a perferated diaphram. (1.0)

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(***** END OF CATEGORY OG *****)

2 ___pnoCcgungg_ _NgBMDb _6DNgBMek2_EMEBgENCY ANg pAGE O BADIOLOGIC66_ggNIgg6 QUCGTION 7.01 (3.00)

Using WNP-2 Figures 7 & D answer the fullowing questions concerning thermal stresses:

a. When considering thermal stresses, the turbine rotor is the most critical element because of it's large diameter.

Why then is the ist stage turbine temperature used vise turbine rotor temperature to determine the turbine starting procedure? (1.0)

b. The turbine 1st stage temperature is 175 F. How long would it take to bring the turbine generator to 100% power? (Consider turbine limits only.) (1.0)
c. With the unit at 10% rated load, WHAT is the recommended load changing rate (%/ min) to increase power to 95%. (Assume 10,000 cycle fatique index.) (1.0)

QUESTION 7.02 (1.50)

During Core Alterations procedure 2.1.2, Nuutron Monitoring System requires at least a GRMs in specific locations. WHAT are the specified locations?

QUCGTION 7.03 (2.00)

With the unit operating at 35% power:

One recirculation pump is secured and it's discharge valve closed

, by procedure. After 4 minutes the operator attempts to open the discharge valve. The-discharge valve fails to open. WHAT two (2) steps are taken to prevent excessive couldown of this recirculation loop?

3 QUCGTION 7.04 (2.00) e In what way could changing RWCU system flow advarsly affect the recir6ulation pumps (per procedure 2.2.1, Reactor Recirculgtion System)?- -

(***** CATCCORY 07 CONTINUED ON NEXT PACE *****)

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L.__RBQCEQQBES_=_NQBdeL._eBNQBtleL._EMEBGENGLeNQ PAGE 9 i BeQIQLQGICeL_GQNIBQL I

QUESTION 7.05 (1.00)

I' l .How are the control rods verified inserted after a reactor scram (4 ways required for full credit) C4 9 0.25 ea3 (1.0) f .

f-QUCCTION 7.06- (1.00)

After a reactor scram with the loss of condensate booster pumps, WitY should the RWCU be. lined up-to return to the RPV before restarting a. booster pump, per procedure 3.3.1, reactor scram recovery?

_ QUESTION 7.07 (3.001 During power. operation the operator receives m' rod drift alarm.

The rod is selected and is found to be still drifting out. No automatic scram setpoints have been reached. What are the j- irrvnediate operator (control room) actions? (Actions involving uperati o ns /coaanunicatio n at the HCU are not required.)

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QUESTION 7.08 (2.00)

On a complete loss of CRD drive flow when are you required to scram the reactor and how is it accomplished? '

l QUECTION' 7.09 (2.50) '

There are eight (8) " Indications" listed in procedure, 4.3.1.1, j, " Primary-Containment Not Operable", that indicate a~ loss of Primary i

Containment _-integrity. WHAT are five (5)-of these " Indications"7 4

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QUESTION 7.10 (4.00)

In reference to the Loss of Condenser Vacuum procedure:

, a. ' List 4 of the automatic actions associated with a decreasing condenser vacuum. Include setpoints. (2.0)

b. What are the four immediate operator actions assuming vacuum has not yet decreased to the point where the automatic actions of part "a" have occurred? (2.0)

QUCGTION 7.11 (3.00)

During power operation at 100%:

a. A loss of GM-7 occurs, WHAT are 4 of the 5 imediate operator action step required? CAn action step may consist of more than one action item.3 (2.0)
b. One of the indications of the loss of SM-7 is a scram from the loss of RPG-MG-1. HOW does the loss of RPS-MG-1 cause the reactor to scram? CGive the chain of events.3 (1.0) 2 o

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

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Q.__eQdINISIBeIIME_EBQCEQUBEG._CQNDIIIQUS._eUQ_LIMIIGIIQNS PACE 11 QUESTION O.01 (3.00)

According to UNP-2 Technical Specifications, WHAT is the minimum staffing requirements for WNP-2 when in condition 3 Emode 33 AND indicate what type of license is required for each staffing position, if any?

QUCCTION O.02 (3.00)

What are the guidelines, per Tech Spec, used when assigning overtime for unit staff members who perform safety-related functions at WNP-27 QUCGTION O.03 (2.50)

It is discovered today, May 14 day shift (08-1 GOO), that a monthly surveillance item due on Tuesday May 7 mid shift (12-0000) was not performed. This item has been performed on time for the past six months. Has the specified time interval for this surveillance item been violated? (Yes or No) EXPLAIN your answer.

QUESTION O.04 (2.50)

n. Describe how a Gafety Limit and a Limiting Cafety System Setting relate to each other. (1.5)
b. What does the term " Limiting Condition for Operation" mean? (1.0)

QUESTION 8.05 (3.00)

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Answer the following questions with regard to the issuance of a Radiation Work Permit (RWP).

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a. What are the radiological limits that require the use of a RWP7 (1.5) 2 b.* What are the responsibilities of the Shift Manager, when

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. reviewing a RWP fnr approval? . (1.5) i

(***** CATEGORY 00 CONTINUED ON NEXT PAGE *****)

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B.__eQdINISIBeIIME_EBQCEQUBES._CQNQIIIQNS._eNQ_LINIIeIIQNS PAGE 12 QUESTION 8.06 (3.00)

With regard to equipment clearance and tagging procedures answer the following:

a. If the person the clearance :L s issued to cannot be contacted for m' clearance release, how is the clearance release accomplished? (1.0)
b. How is an addition and/or deletion of tags _made to a previously authorized clearance order prior to the clearance order being acceptanced by the authorized individual? (1.0) c What is meant by the term " Redundant Verification" 7 (1.0)

QUESTION O.07 (3.00)

What is the guideline or instruction given for each of the following per the standing operating orders, 1.3.1 attachment I7

a. Overriding the automatic action of an ECCS system. (0.75)
b. Placing a controller in the manual made from the automatic mode. (0.75)
c. A safety-related motor operated valve has been manually seated.(0.75)
d. The instructions for aligning more than 2 valves or circuit breakers. (0.75)

QUESTION O.00 (3.00)

According to WNP-2 Technical Specifications 4.1.D.6 which states each affected control rod shall be demonstrated to be coupled to its drive mechanism...:

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a. When is the coupling to be verified? (1.5)

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b. How is the coupling verified? (1-51 e

(***** CATEGORY 00 CONTINUCD ON.NCXT PAGE *****)

Q __GDdINISIBeIIMC_EBQCERURCS._CQNQIIIQNS._eND_LIMIIGIIQNS PAGE 13 QUESTION 8.09 (2.00)

In reference to the Emergency Plan Implementing Procedures:

a. Who has the sole responsibility for timely classification and declaration uF any emergency situation? (0.4)
b. What is the normal line of succession for the position responsible For classification and declaration of an emergency situation? (1.6) h'

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

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

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EQUATION SHEET .

f a ca y a s/t Cycle efficiency = (Network out)/(Energy in) 2 -

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

E'= mc  ;

2 KE = 1/2 mv a = (Vf - V,)/t A = AN A=Aeg

, PE = mgn Vf = V, + at w = e/t A = an2/t1/2 = 0.693/t1/2 y ,yg - t 1/2 eff=[(tm)(t)] g

[(t 1/2I

  • I*bI3

, AE = 931 am -

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I'=Ie"g Q = mCpat (i = UA At. I = I,e i Pwr = Wfah I = I,10**/ M TVL = 1.3/u 5

P = P 10 ""(*I HVL = -0.693/u p = p et /T o -

SUR = 26.06/T SCR = S/(1 - K,ff)

CR x = S/(1 - K,ffx)

SUR = 26a/t* + (s - o)T CR j (1 - K ,ffj) = CR 2 I ~ eff2)'

T = (t*/s) + [(a - o)/ o] M = 1/(1 - K,ff) = CR j/CR, T = 1/(o - s) M = (1 - K,ffa)/(1 - K,ffj) 7=(s-o)/(Ao) SDM = (1 - K,ff)/K ,ff
o = (K ,ff-1)/K ,ff = AK,ff/K,ff 1* = 10-5 seconds l

_ T = 0.1 seconds-I o = [(t*/(T K,ff)] + [s,ff (1 / + AT)]

Idjj=Id P = (I+V)/(3 x 1010) Id 2 ,2 7d 2

jj 22 I = oN 2 R/hr = (0.5 CE)/d (meters)

R/hr = 6 CE/d2 (feet)

~

Water Parameters Miscellaneous Conversions 1 gal. = 8.345 lem. 1 curie = 3.7 x 1010*dps

, 1 gal. = 3.78 liters 1 kg = 2.21 lbe .

l 1 ft* = 7.48 gal. 1 hp = 2.54 x 103 Stu/hr Density = 62.4 lbg/ft3 1 av = 3.41 x 106 8tu/hr Density = 1 ge/c# lin = 2.54 cm Heat of vaporization = 970 Stu/lem *F = 9/5'C + 32 Heat of fusion = 144 Stu/lbe 'C = 5/9 (*F-32) 1 Ata = 14.7 psi = 29.9 in. Hg. 1 BTU = 778 ft-lbf 1 ft. H 2O = 0.4335 lbf/in.2

~

+ - - + , - - - - . . . - - . - .

2. m & .. _ . . .

O e

e 9

6 FIGURE 3 fe --- APPLY INITIAL LOAD = 8% M10%-ele--20*& -el

, *I 1

4h 4

X . .

I T INCREASE 3 LOAD \ ,

<ExameLE v \ b'14

, Js\ N \

aO\\4 5% LOAD 3 N .T

\

[

i q

%\ k j A. E 533g; TO SYNC SPEE R XN N m I 1 l - .

IEXAMPLE H 2)J 0 100 200 300 400 500 DEGREES F INITIAL HP TURSINE FIRST-STAGE METAL TEMPERATURE I

W

  • " - - - * = - e,----.~ . _ , , _ _ , , _ ,

'"O*

h e-*W=** 4 4 g .g M g, g _4 _

e e

i FIGURE 7. START UP RECOMMENDATIONS 700-1200 MW NUCLEAR STEAM SYSTEM UNITS I

81101021 A i

APRit 1944 I

L l

. ~ - . . . . , , ,

- - - - ~ --- ..

. t FIGURE 1 400 P.

E I 300 g ass a y b y o 2 g iso *F f

= ' /

moc 4 I

g 1ss " #

/

r o a 100 r E /

o o 20 40 so so 100 PERCENT RATED LOAD FIGURE 2 400-1 / b.'

i '

//AP'f

i l,00 V/I6 Dh 20, f

/Isew f/W "

l 5 "* f/Y 5 100

{

140*F w -

$ v o

o 1 2 3 4 5 TIME TO CHANGE LOAD HOURS i

e 4

l i

.I FIGURE I. LOAD CHANGING RECOMMENDATIONS 700-1200 MW NUCLEAR STEAM SYSTEM UNITS

, 8110102 2A APRIL 1984 i

L_

/VJ43 E

5. Ti tCQRY_. O!" NUCLEAR POWCR PLANT OPERATIQL_FLUlg_G,_AND PAGE 14 IUEBUQDYNAMICQ ANCWERO -- WNP-2 -05/05/29-MORGAN, T.

ANSWER 5.01 (3.00)

a. Head increases CO.53 pump is still putting the same amount or work into the fluid, therefore the same delta pressure increase across the. pump, so as suction pressure increasen so will the discharge head. CO.53 (1.0)
b. Ilead increases CO.53 as system resistance to Flow increases pump head increases. CO.53 (1.0)
c. Head decreases CO.53 as temperature increases system resistance to Flow decreases (lower viscosity), therefore pump head decreases. CO.53 (1.01 REFERCNCE C. E. Thermodynamics llent TransFor and Fluid Flow Chapter 7, Fluid Statics, Dynamics, and Delivery, Pg 7-104 through 7-124.

ANCUCR 5.02 (3.00)

a. CR1 (1-Keff1) = CR2 (1-Keff2) CO.753 200 (1 .95) = 400 (1-Keff2)
  • 200 (1 .95)/400 -1 = -Keffa

. 975 = - !;e f f 2 CO.253 delta p = Koff2-1/Keff2 - Keff1-1/Keff1 CO.753 delta p = .975-1/.975 - .95-1/.95 delta p = ( .0256) - ( .0526) delta p a .027 CO.253 (2.0)

6. Part b. will be graded independently of part a.

delta p= Keff3-1/Keff3 - Keff2-1/KefF2 CO.753 f/ f; c

.027 = Keff3-1/Keff3 - .975-1/.975 c/!/

.027 = 1-1/Keff3 - ( .025G) Mi7

- .0014 = 1-1/Keff3 .c r 3

.990G = -1/Keff3 T7EI20

.990G Keff3 = -1 CO.253 Keff3 = 1.0014 super critical (will accept critical) (1.0)

REFERENCC WNP-2 Reactor PhysicsSection VI Reactor Operations Part A.

Suberitical Reactor (no page numbers available), subsection e.

Determine Criticality.

h/Sa S

  • nr- ly' 2 I e f $ e c su G kr tvac.fy, s

I f

5. THEORY OF NUCLEAR __POWCR PLANT OPERATION,_ S UID92_AN9 PAGE 15 THERMODYNAMICG ANGWCRC -- WNP-2 -05/05/29-MORGAN, T.

ANGWER 5.03 (3.00) i When recirculation flow decreases the boiling boundary moves down j so there are more voids in the core. CO.53 This adds negative

} reactivity and power. level starts to decrease. E0,53 As power level drops,the fuel cools down which cools the water and the boiling boundary starts to move up again. CO.53 Power will cont-inue to decrease and boiling boundary will continue to move up increasing reactivity until the reactivity balance is zero and the reactor returns to steady state at a new lower power level. CO.53 The boiling boundary does not return to the same level as before the power decrease. It remains somewhat lower than before recirc-ulation flow was decreased. The reason is that although the neg-ative reactivity comes From only one source (boiling boundary moving down) the positive reactivity comes from two sources. The first

, is the boiling boundary moving back up due to the decreased heating caused by the power level decrease. The other is the positive reactivity from the doppler coefficient resulting from the fuel

. cooling down. El.03 (The final results is an increase in void reac-f Livity to off set the doppler reactivity decrease.) (3.0) t t

l'.CrERENCC UNP-C III Neutron Life Cycle, Criticality, and Reactivity-

$28& r7YYeof3 ff 5 4 ~ U, i

l O

9 e O i

i i

L ,

3.__IUEQBY_QE_UUCLE6B_EQWEB_EL6MI_QEEBoIIQN._ELUIQQ._6NQ PAGE 1G IUCBdQQYNoblCS ANGWERG -- WNP-2 -05/05/29-MORGAN, T.

ANSWER S.04 (3.50)

a. Defined as the last. day, the core can produce rated power at rated conditions with all the control rods removed from the core. (0.5)
b. Coastdown -

a made of operation in which the reactor power level may drift downward with exposure after and of cycle, this is a reduction in negative coefficient effects with the reduced reactor power compensating for the fissionable istope depletion (power coefficient). C1.03 Final Feedwater Temperature R.Juction (FFWTR) - a made of operation where the feed temperature is reduced by reducing the extraction steam to the feedwater heaters, which increases the core inlet subcooling. The increased subcooling reduces the core void fraction, which results in a core reactivity increase. C1.03 Increased Core Flow - a mode of operation where core flow rates greater than the referenced 100% rate value is used. The core reactivity is increased by reducing the core void content, which results from recirculatng more water through the core. C1.03 (3.5)

REFERENCE Reference, WNP-2 Reactor Physics VII Core Aging, VII, E.1.a-b-c, pg 0, 9, 10, 11 & 12.

3 I

e e

L___THCORY__ OF NUCLCAR POWR PLANT _QPgBATIQL_FlyIgg _AND PAGE 17 TIICRMODYNAMIGQ ANSWCRC -- WNP-2 -05/05/29-MORGAN, T.

ANGWCR 5.05 (3.00)

n. The core response to the withdrawal of a deep control rod is to raise core power in the upper region of the care, uspecially in the areas where the rod was withdrawn. Since the void content of the upper portion of the core is high at operating conditions, the effects of deep control rod withdrawal, although axially dampened, will be substantial radially. C1.53
b. The care response to withdrawal of a shallow control rod is to raise the power locally in the region where the control rod is withdrawn. The local power increase at the core bottom will pull the boiling boundary down, increasing the void content above the withdrawn control rod. The negativu effects of voids may be l

stronger than the positive effects of the rod and power may decrease. The shallow rod strungly aFfects the axial power shape and not the overall core power. C1.53 (3.0) j REFCRCNCC t.*NP-C Reactor Ph/nics VIII Operatng Charactoristics, VIII.A.6.C.1) &

2), pg 30 & 39.

1 ANGUCR 5.0G (3.00)

a. G0/15 = x/450 c. (1000/450) squared = 005/x (4)(450) = x (4) squared x = 005 x = 1000 rpm x = 50.3 ft of head
b. 1000/450 = 47,2OO/x d. (1000/450) cubed = x/150 i 4x = 47,200 (4) cubed x 150 = x x = 11,000 gpm .x = 9,600 hp i (4-@ 0.75 ea) '

(3.0)

REFERENCE i,

WNP-2 Requalification Program, Reactor Recirculation System and G.E.

Thermodynamics Heat Transfer and Fluid Flow.

O t

e e

i 9

(

l i

L

1 l

_S __IUggRY Og_UggLg$B_EQWER_ELANT_gPgBATION 2 _FLgIQQ _AND PAGE 10 l IDGBd9916edIG9 l

. 1 ANOWCRG -- WNP-2 -05/05/29-MORGAN, T ANGWER 5.07 (2.OO)

The xenon peak following a shutdown (scram) can have important effects on reactor operations throughout core life. If the reactor was shut down by 1% ok/k as measured at the time of peak xenon, then the GDM will decrease as xenon decays. Gince xenon (peak) is greater than the 1% dk/k a reactor restart would r.c c u r . (2.0)

REFERCNCE Wr1P-2 Reactor Physics Sec. V Fission Product Poisons, part 2 Xenon behavior after Reactor Chutduwn (nu page numbers)

ANOWER 5.00 (2.001 As the boilinu rate increases, two-phase flow recistance increases.

this would tend to divert coolant flow From the higher powered center ruel bundlet, where it is needed the most C1.03. Orificing has the e f f.cc l of providing a large resistance to flow so that any additional eu!istans.u caused by two-phase flow is acceptably small C1.03. (2.01 REFERLNCE W:4P 2 C.E. Thermud y nanic s , Chapter 9 DWR Thurmal Limits, Pg 9-58 ANOWER 5.09 (2.50)

a. Minimize fuel dmnage during a DDA LOCA by limiting the peak clad temperature (to < 2200 F) -OR- limiting bundle stored energy. (1.0)
b. 2 and 3. (0,5)
c. MAPRAT = /APLHCR/LIMLHGR -or- a

/APLHCR/MAPLHGR limit (1.0)

-or- =

(ffAPLHGR ) actual / (MAPLHCR) LCO max REFERENCE WNP-2 G.C. Thermodynamics, Chapter 9, DWR Thermal Limits pg 9-60 t

9_ 7 s

,9-74

-+-w~,, a - . + - -

G.__ELoNI_GYSIEdS_QGGIGN._CQUIBQL._68Q_INGIBudENIoIIQN PAGE 19

. ANSWERG -- WNP-2 -05/05/29-MORCAN, T.

ANGWCR G.01 (3.00)

Narrow Range, 0-GO. inches above instrument zeru CO.33, bottom edge of steam dryer skirt (527.5 inches), yes CO.33 Wide Rango 150 inches below to GO inches above instrument CO.33 zero, (527.5 inches), yes. CO 33

-n 8 0 Fuel Zone,.140 inches ba'

- 300 be/as.) C27 $.

t o -he- i nc he s - CO.33, (M inches), no CO.3J Upset range, O tu 100 inches above instrument zuro CO.33, (527.5) no. CO.33 s l Chutdown, O to 400 inches above instrument zero CO.33, (527.5) no. CO.33 (3.0)

GCFCRCNCC WNP-C Cystem and Procedure Vol. 1, Nuclear Dailer Instrumentation, Data Gheut 1, pg 55, SG, & 57 <

ANGWCR G.OC (3.50)

Inside

} 1. Containment Area Temperatures l C. RPV Head Seal Leakoff

3. Valve Gtem Leakage RRC-V-GO A&D i 4. Drywell Equipment and Floor Drain Gumps (4 0 0.5 eal CD.03 Outside 3

1, Reactor Duilding Equipment and Floor Drain Gumps

2. Area Leak Detectors - Various Reactor Oldg, Ste un Tunnel and Turbine Bldg area temperatures
3. System Isolation - RHR, RWCU, RCIC and MSIV's (3 G 0.5 ea) C1.53 (3.5) i, .

REFERENCE Gystem and Procedure Vol. I, Leak Detection Gystems, pg 2.

' s sl 6

p!s. cms OG o u YshMse oG .

g N Meacf M*l gag gR c (velve leeLape) z q,,,eg RLC Ruc Rw c u. msov.

G. EL6NI_SIGIEdG_QCGIGN._CQNIBQL._eNQ_INGIBudENI6IIQN PAGE 20

. ANGWERS -- WNP-2 -85/05/20-MORGAN, T.

1 i

ANOWCR G.03 (3.00)

.! a. The different position of the IRM range switches provide the i . attenuation necessary for a particular IRM channel to cover six decades of reactor power indication. (0.5)

b. 1. 24 CO.333 3 ne CO d43 or /cs [4) *!e k C" ^#

(1.0) g

c. 1. Detector liigh Voltage Low CO.53 l

C. Module Unplu0ged CO.53

3. IRM Mode Owltch not in operation CO.53 (1.5)

REFERENCE 4

WNP Oystems and Procedures Vol. II, IRM pg 14, & 23, and figure 10, ANSWER G . 0'4 (3.00) '

a. 1. fianual initiated downshift because of low reactor power operations.

! C. Low total feedwater Flow,with a 15 see time delay.

G30'7h

3. Low' reactor vessel water level (level 3).
4. Main steam line/ pump suction line differential temperature l high . (49.1#F )
5. Turbine trip r governor valve fast closure I ith power b.

greater than ) .[> l'f 2 F5') M"" M MA.Yb. J em3 - (1.5)

}

Any of the downshift signals will trip RPT-3A(D), allowing the pump to start coasting down . As the pump coasts down, the LFMG comes up to rated speed and voltage C O .(/3 Dreakers 2A(B) For ,

l each pump will close when pump speed is between 20-2G% CO.%3 and the other close permissives are met. (Any time the luw speed transfer is initiated, the recirculatoin loop flow controller automatically shifts to the manual mode CA-93 the low speed

', transfer sequence is activated, an incomplete).transfer When sequence t imear starts CO.33. If the pump is not between 20-2G% speed, 5

or breaker 2A(B) is not closed after 40 seconds, the incomplete transfer relay trips breaker 1A(D) C O .p . ES 0.0 ee] (1.51 REFCRENCE WNP-2*, ,Gyntem and Procedures, Vol. I, Reactor Recirculatiorg System, Fast to Glow Speed Transfer Gequence, pg 34.

  • Vol. V Feedwater Gystem, Gec VII Interlock / Control Actions "M" pg 23 e

f

G.__ELoNI_SYSIEdQ_QEGIGN._CQNIRQL._eMQ_INGIBudENIGIIQM PACC 21

, ANCWERG -- WNP-2 -05/05/29-MORGAN, T.

1 ANCWCR G.05 (3.00)

a. 1. Setting speed demand Getting acceleration values I

Cenerating speed holds

2. Getting Loading rates Generating load hold signals C5 9 0.2 ea3 C1.03
b. Mode 1 - Ruactor Gtart - the DEH has little control other than maintaining the setpoint for BPV operation, or maintain plant pressure if the startup is placed in a hold.

Mode 2 - Turbine Start - the turbine is latched rolled off the turning guar, accelerated to synchronous speed, and turbine control is transferred from the throttle valves to the governor valves.

i Mode 3 - Turbine Load Control - starts when the main generator output breakers close. Luad is pick-up until OpV just closed.

Mode 4 - Turbine Follow Reactor Manual - the 10% bias to the load control signal. A 3% closing bias is applied to the bypass valves.

The turbine is slaved to the reactor. [4 9 0.5 ua3 C2.03 (3.0)

REFERENCE WNP-2, Cystem and procedure, Vol V DEH L.P., pg 22, 23, and 57-G2.

ANGWER 6.0G (3.00) i 1. HPCS pump -

start signal.

2. CGT Suction Valvo (HPCS-V-1) - open signal, i f Suppression pool I

suction valve (HPCS-V-15) is not fully open.

3 The injection valve (HPCS-V-4) -

open signal.

4. The CGT flow test valve (HPCG-V-10 & 11) - close signal,
5. The suppression pool test flow valve - close signal.
6. Division 3 diesel generator - start signal

, C6 9 0.5 ea3 (0.0)

REFERENCE WNP-2 System & Procedure HPCS LP, pg 7. .

I i

[

l l - - - - -

G.__ELoNI_SYSICd2_QEQIGN _CQNIBQL._eUQ_INSIBudEUIGIIQM PACE 22  !

ANSWERG -- WNP-2 -95/05/29-MORGAN, T.

ANOWCR G.07 (3.50)

a. 1. Low pressure in piping upstream of the FCV. (0.5)
2. High pressure in piping downstream of the FCV. (0.5)
b. No, the orifice bypass valvu is opened when the influent pressure is low. This would increase the flow rate. The purposu of the orifice is to limit fluw rates to the condenser or radwaste when fr/.....

the influent pressure is high. I" 5 - o k ,

  • f t M,<r'M{ <' u >< ' ** M I'* 1.0)

(Also acceptable: 1. Becausu it violates plant operating procedures and 2. Decause it will " starve" the RHX and increase F/D influent temperature)

c. 1. Low Reactor Water-Level 2
2. RWCUG Inlut und Outlet High Flow Differential
3. High Ambient Temperature in RWCU Equipment Room
4. High dT across Equipment Room ventilation ducts 5' R L>c44 Fa P e ** J y T"Y ( 4 required (7 0.25 each) (1,0) 6 3.I'*/ tc/c t r e A n H. p pr y ,

i

d. Cervice air is used in the Filter-demineralizer backwash operation of the precoat evolution. C4 *-/c/ *'d b E L e<>734 (0.5)

REFERENCE WNPD Cystems and Procedures Training RWCU L.P., pg. 6,~1,14,16, & 22 44 ANGWCR G 00 (3.00)

a. RCIC will,not initiate, CO.253 the RCIC steam stop valve (RCIC-V-45) will not open if the exhaust vnlva (RCIC-V-GO) is not full open E0.753. (1.0)

~

b. The turbine will trip on overspued, CO.253 the ramp generator is usually the low signal which controls the turbine on quick starts with this signal high the turbine i +. up in speed before sufficient oil pressure is available to the governor valve to close it CO.753. [w,'I/#"'8'*Orh) (1.0)

, c. RCIC will inj at maximum rate und the min flow valve will not respond, CO.253 the flow signal is at minimum due to the zero d/p sensed therefore demanding Max. flow from the RCIC 1 system CO,753. (1.0)

REFERENCE

  • WNP-2 Gystem & Procedures Vol III RCIC L.P. pg 11, 12 & 13 I

i l

I

! l 2.__EDQCEQUECQ_=_NQBdQL._oQNQBdQL._EdCOGENCY_QNQ PAGC 23 SQQ1QLQGICOL_CQNIBQL

.. ANGWCRS -- WNP-2 -05/05/29-MORGAN, T.

I i i i

f ANOWT:n 7.01 (3.00) i

n. Rotor temp is not directly measured. 1st stage metal temp is j used as a good approximation. (1.0) i b. 3.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> (3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> 55 minutes) C3.0-4.0 hrs acceptable 3 (1.0)
c. 10% = 90 F ist stage temp l 95% = 205 F ist stage temp 205 r - 90 F= 105 F 195 F from figure 2 (oF fig 0) 1.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> power change 95% - 10% = 05%

05% / 70 min = 1.09 or 1.1% / min F st cred'k de #~ 3 sr i w ,jto- f- s L,w <q coo rk.

i nCrCnCnCu V'o II upe t e reda t ~~ /nr w. el. a .s ku

} WNP 2 Training Handout,3 DCil, pg 53, fig 7 & O d i

i ANOWER 7.02 (1.50) l 1

One ORM in the core quadrant in which the core alteration is taking place, CO.753 and another in an adjacent quadrant. CO.753 (1.5)

REFERCNCC WNP 2 Plant Procedure Manual, 2.1.2 Neutron Monitoring Gystem pg 2

- ANOWER 7.03 (2.00) i

1. Glowl r reduce seal purge injection Flow (maintaining seal temp per p ocedure ) . A/J.) 2ccc rf shp pg., y c h f& pu my s (1.01
2. Maintain or establish RWCU flow from the loop. (1.0) i 1

l REFERENCC j WNP- 2 ppm 2.2.1 Reactor Recirculation Gystem pg G ANGWER 7.04 (2.00) 3 l Any oberation that changes the RWCU non-regen hx heat load ,would

!. affect'RCCW temperature and may cause a undesirable thermal.

4' transient to the recire pump seals. (2.01 i

l t

I

! 2.__EBQGEDuBES - NQBtieL._8CNQBtieL._Et1ERGENGL.eNQ PAGE 24 i BeDIQLQGICeL_CQNIBQL '

i

. l

~ ANSWERS -- WNP-2 -05/05/29-MORGAN, T.

REFCRCNCC WNP-2 PPM 2.2.1 Reactor Recirc Sys pg 4 j ANSWER '7.05 (1.00) ,

1. RGCG display e
2. GDG display 4e#/"' # N

/ g CRD position printout ((*~ra F )

3.

4. Full core display, E4 0 0.25 ea3 (1.01 REFERENCE' WNP-2 PPM 3.3.1 Scram Recovery pg 3 ANSWER- 7.06 (1.00)

The riow is returned to the RPV to repressurize the feed water piping.4o pr~ e.e s.a f~ e <c o.s t a v e & r s., a l S /ew s t me./lc o & lramm+) (1.0)

RtrCRCNCC WNP-2 PPit 3.3.1 Reactor Gerasw Rucovery, pg 2 ANGWCR c7.07 (3.00)

1. Depress continuous in button
2. If Rod motion stops then reduce power to ~65%.with recirc flow
3. Insert deep rods to maintain power below flow-biased APRM scram
4. If rod motion continues er.annally scram the reactor before heat Flux exceeds 105% or an APRM scran set point
5. Refer to T.S. 3.1.3.1 C4 0 0.75 ea3 (3.0) 5- REFERENCE Wr4P-D ' PPM 4.1.1.1 Rod Drift.pg 2 & 3

{1 f_

i 1

i L

I I

'L..

. _ _' : --_ --__-__:_--__r._ _ r- ~ ~ ~ ? mm h:._ - _ . _ _ _ _ _ _ - . _ ~ - - . - - _ _-. - - _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - _ - - . _ _ - -

l I

2.__EBQCEQUBES_=_NQBdoL._oRNQBboL._EdCBGENC1_oNQ PAGE 25 BOQIQLQGICoL_CQNIBQL ANSWERG -- WNP-2 -GS/05/29-MORGAN, T.

t t

ANOWCR 7.00 3 l .tf udo N et CKD fsmf b(2 00)rN arHW

'. 3 Whun a second accumulator trduble alarm is received C1.03 scram the

, reactor by placing the made switch in shutdown. C1.03 (2.0)

RCFERCNCC l WNP-2 PPM 4.1.1.2 Complete Loss of CRD Drive Flow, pg 2 3

ANSWER 7.09 (2.50)

1. One or more ur the suppression chambers drywell vacuum breakers j are not operable and closed.

4

2. One or more of the Reactor Duilding suppression chamber vacuum breakers are not operable and closed.

' 3. The dr ywell or suppression chamber oxygen concentration greater than 3.5% by volume during operational condition 1 after 24 l hours it om the time the reactor power exceeded 15% and not within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> oF reducing reactor power to less than 15%

l p c al in. i. nar y to a scheduled reactor shutdown as 1ndicated on containment oxygen recorders CMG-02R-1 and CMG-02R-2 at P.s n e l K1' ar.d K2,

] 4. The prin.ary containment air lock or associated doors are not operable.

3. One ur more of the primary containment equipment hatches are not closed and sealed.

G. Oou or more of the sealing mechanisms associated with the l primary containment penatration is found to be inoperable.

7. Priniary containment leak rates are tuo high an defined in Technical Gpecifications 3.G.1.2, 3.G.2.1.e or 3.G.3.

O. Primary containn.ent automatic iso 1 tion valve r;ot operable.

J. &r ! C o** 5 l* $ I /"

REFERENCE l* C' o Y "* ='9bD" **' * ** l~ '" 7 } peer o i. Tsel.5 re

?

, e e

i

  • i o

i I

I e

l

2.__EBQCEDUBEG_=_NQBd6L._6DNQBdaL._EdEBGENC1_eNQ PAGE as BeDIQLQGICoL_CONIBQL ANSWERS -- WNP-2 -05 / 05 / 29--MORGAN , T.

ANSWCR 7.10 (4.00)

a. 1. Main turbine trip G vacuum (19" Hg
2. MGIV Closure G vacitum (10"Hg
3. EPU Closure G (7" Hg.
4. RFPT Trip G 0" Hg.
5. Low Vacuum alarm G 25" Hg.

(4 required O O.5 each) (2.0)

b. 1. Rapidly reduce reactor power with recirculation flow
2. Verify normal gland steam seal header pressure of 200 psig
3. Place standby sut of air ejectors in servico
4. If reactor power (5%, start mechanical vacuum pump (4 required G 0.5 each) (2.01 REFERENCC WNP -2 PPM AP 4.6.5.1 Loss of Condenser Vac.uum pg 1 &2 ANGWCR 7.11 (3.00)
a. 1. Transfer RPG bus A to the Alternate A power supply and reset scram and isolation signals.
2. If all c.uo li ng (CRD and RCCW) was lost (and cannot be restcred within 60 secondu) to the Reactor Recirculation Pumps (ARC-P-1A and 1D) o Placa the Reactor Mode Gwitch in Ghutduwn.

o Trip both Reactor Recirculation Pumpo, a Refer to PPM 3.3.1, " Reactor Scram"

3. If CRD cooling is still available to the Reactor Recirculation Puir.p s TitCN shift both Reactor Recirculation Pumps to the 15 11z MG sets and balance recirculation loop flows if necessary.
4. If the reactor has scrarrmed THEN carry out PPM 3.3.1, " Reactor Scram".
5. Notify plant personnel and the Dittmar Luad Dispatcher.

. G .4kT waJ (2.0)

, b. 1. The loss of RPS bus 1 deenergizes NGGGG ((Y #* I

2. NGGGG deenergizing trips CW pump 'C' O. The loss of CW pump 'C' causes a losg of condenser vacuum

}1 4, The loss of condensor vacuum fs",cfamp N -h -_ t ur h ; f ; ./.

_ j .2 2 -

' OW Ng A penswd . C4 0 0.25 em3 (1.01 REVCRCNCC j WNP-2 PPM AP4.7.1.0 Loss of Power to GM-7 pg 1, 2& 3 l

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. ANSWCRS -- WNP-2 -05/05/29-MORGAN, T, ANGWER O.01 (3.00) 1 - Shift Manager - SRO 1 - Control Room Cupervisor - SRO 2 -

Reactor Operators -

RO 2 -- Equipment Operators - None 1 - Shift Technical Advisor - None 4

(number CO.23 title CO.23 license CO.23) (3.0)

REFERCNCE

! WNP -2 Tech Gpecs Tablo G.2.2-1 pg G-G J

, ANSWER O 02 (3.00) f 1. An individual should nut be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> straight, CO.53 excluding shift turnovers. (0.5) i 2. An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />

in any 24-hour period CO.53, nor morn than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48-hour period CO.53, nor more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any 7-day period i CO.53, all excluding shift turnover time. (1.5)
3. A break of at least O hours should be allowed between work periods

, CO.53, including shift turnover time. (0.5)

} 4 Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff an a shift. (0.5) 1 REFERENCE WNP-2 Tech Upecs Cec G.2.2.f pg G-2 i

ANGWCR O.03 (2.50)

, No, CO.53 A maximum allowable extension is not to exceed 25% of the surveillance interval or 7.75 days C1.03 ANO a total maximum combined

., interval for any three consecutive tests not to exceed 3. 25 t inwes the specified surveillance interval. C1.03 (2.5) li RCFERCNCC '

L WNP-2 Tarb Spec 4.0.2 Gurveillance Requirements pg D/4 0-2

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, ANSWERS -- WNp-2 -05/05/29-MORGAN, T.

ANSWCR O.04 (2.50)

.. Safety Limits are limits upon important process variables below which the reasonable maintenance of the cladding and primary systems are assured. LSGG's are settings on instrumentation which initiate automatic protective action at a level such that the Safety Limit will not be exceeded. (1.5)

b. The LCO's specify the minimum acceptable levels of system per-formance necessary to aseture safe startup and operation. When these conditions are met, abnormal situations can be safely controlled. (1.0)

RCf'CRCNCC WNP-2 Tech Opue GL & LGSG Dases pg B2-1 & D/4 0-1 ANSWCR O.05 (3.00)

a. o Direct Radiation >/= 2.5 mrem /hr CO.53 o Airborne Radioactivity >/= 25% oF MpC CO.53 o Contamination >/= 1000 dpm/100 cm2 Data Gamma or

>/= 100 dpm/100 cm2 alpha CO 53 (1.5)

b. Tu determine the e f f ec: t s of the task on the overall plant and the effect of other plant parameters on the task to be performed. (1.5)

RCFCRCNCC WNP-2 Ilealth physics Program Discription IIPD 3.1.0 hadiation work permit pg 2 ANOWCR O.06 (3.00)

a. The worker's supervisor or Department Manager can initiate the release after satisfying himself that equipment and personnel

> safety will not be jeopardized. As a last rusart, the Shift Manager has the authority to initiate a clearance release. (1.0)

b. The change needs to be initiated by the Shift Manager to signify his approval. (1.0) c . . Rettundent Verification is that requirement directing a spcond knowledgeable individual to make an independent verification of correct equipment s t at us ' (for safety related and Fire protection equipment} (1.O)

1 Q.__oQdlulSIBoIIMC_EBQCEQUBES._CQNQlIIQUG._eND_LidII6IlQNS PAGE 29

, ANSWCRS -- WNP-2 -85/05/29aMORGAN, T.

A REFCRCNCC WNP- 2 Equipment Clearance and Tagging Procedure 1.3.0 pg 2, 5 and change t 84-1058

- ANGWCR C.07 (3.00)

a. Operators are not to override the automatic actiuns of ECCG and _,

other safety Features, unless continued operat on of result in unsafe plant conditions.Or

-b. An operator may place'a controller in go, / e 4mo esuhwif'[1 j

Tr,om[t[o 0.75) f Gre M to manua automatic made whenever, in the operntor's judgement, continued N D"( y automatic operation is undesirable. (0.75)

c. When a ~ safety related motor operated valve has.been manually seated or back seated, the valve shall be declared inoperable until matur operation can be demonstrated. (0.75)
d. Instructions for aligning more than 2 valves or circuit breakers should be written on a Component Gtatus Change Order and carried by the operator performing the change unless the operation is pur f or mad using the procedure or chucklist. (0.75)

RCFERENCC WNP-O Otanding Orders / Night Orders, 1.3.1 Att 1 pg 2-5 ANGWCR 'O.00 ( 3 .~ 0 0 1- -

y

a. 1. Priur to reactor criticality after completing Core Alterations that could have aFFected the control rod drive coupled integrity,
2. Anytime the control rod in withdrawn to the " Full Out" pcsition in subsequent operation,
3. Following maintance on or modification to'the. control rod or control rod drive system which could have affected the control rod drive coupling integrity.

E3 0 0.5 ea3 (1.5)

b. Each affected control rod shall be demonstrated to be coupled to its drive mechanism by observinD any indicated response nF the nuclear instrumentation while withdrawing the control rod to the

) Fully withdrawn position C1.03 and then verifying that the control rod drive does not,go to the overtravel position CO.53. (1.5)

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WNP C , Tech Sphe 4.1.3.G Reactivity Control Gystems, Gurveil7ance requirements pg 3/4 1-12

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/.NOWERG,-- WNP-2'

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,- -85/05/29-MORGAN, T.

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Ib.. < !OCR.O.'09 (e.00)

a. Plant' Emergency Director (0.4)

! b'. Pla'nt Manager , / (0.4)

Assiatant Plant: Manager (0.4)-

Operations Manager- (0.4)

Chift Manager (0.4)

REFERENCE EPIP 13 .1' .1, pg 1 & 2; . 13.1.~2 pg 1 1

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