Text

Exam Repts 50-387/87-05OL & 50-388/87-05OL on 870316-19. Exam Results:Three Out of Five Senior Reactor Operators & Two Out of Four Reactor Operators Issued Licenses.Exam & Answer Key Encl
	ML20213G559
Person / Time
Site:	Susquehanna
Issue date:	05/07/1987
From:	Collins S, Keller R, Kolonauski L; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
Shared Package
ML20213G509	List: IR 05000387/1987005; ML20213G507;
References
	50-387-87-05OL, 50-387-87-5OL, 50-388-87-05OL, 50-388-87-5OL, NUDOCS 8705180345
	Download: ML20213G559 (98)
	v • d • e

-

y

$

U.S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NOS. 87-05(0L) and 87-05(0L)

FACILITY DOCKET NO /388~

LICENSEE: Pennsylvania Power & Light C North Ninth Street Allentown, PA 18101 ,

FACILITY: Susquehanna Steam Electric Station (Units 1 and 2)

EXAMINATION DATES: March 16-19, 1 7 CHIEF EXAMINER: Yt/F)

Lynn Kolonauski actor Engineer (Examiner) Date REVIEWED BY: )

Robert M. Keller, Chief, Projects Section 1C d([77 Date APPROVED BY: M(tMtfd S'amuel J. Collins, Deputy Director, DRP 5 Y/87 Date SUMMARY: Operator licensing examinations were administered to five (5) Senior Reactor Operator candidates and four (4) Reactor Operator candidates during the week of March 16, 1987. Two candidates failed the R0 written examination. Two candidates failed the SRO operating examination. The remaining candidates passed their respective written and operating examination Overall, three (3) SR0 and two (2) R0 licenses were issue ~

hDR ADOCK 05000387 PDR

.

t

. .-

.,

..-

4 2-

' REPORT DETAILS

.

TYPE OF EXAMS: Initial Replacement _X_ Requalification EXAM RESULTS:

l R0 i SR0 l l Pass / Fail l Pass / Fail i I I I l .1 l l l Written Exam l 2/2 -l 5/0 l l l 1 l 1 I I I l Oral Exam l 4/0 l 3/2 l l l 1 l l I I l Simulator Examl 4/0 l 3/2 I I I I I I I I I l0verall l 2/2 l 3/2 I I l l l CHIEF EXAMINER AT SITE: Lynn Kolonauski, NRC OTHER EXAMINERS: David Lange, NRC Allen Howe, NRC Brian Hajek, NRC Consultant Summary of generic strengths or deficiencies noted on oral exams:

While the examiners noted individual strengths and weaknesses, no generic strengths or weaknesses were identifie . Summary of generic strengths or deficiencies noted from grading of written exams:

The SR0 candidates did very well in sections 5, 6, and 7. However, the SRO l

candidates indicated a weakness in answering the questions concerning the ,

,

SSES Technical Specification i The RO candidates were strongest in sections 2 and 3. However, no generic strengths or weaknesses were demonstrated by the R0 candidates as a group.

I l

l i

i

!

l i

. . .

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

_

.

. 3 Comments on availability of, and candidate familiarization with plant reference material available in the control room:

All main control room reference material necessary to conduct the opera-ting examinations was readily available in the control room. While selec-ted SRO candidates were proficient in locating and using this material, the remaining SR0 candidates seemed somewhat unfamiliar with its location and use. These comments are documented in the individual operating exam-ination report . Personnel Present at Exit Interview:

NRC Personnel Lynn Kolonauski, Reactor Engineer Examiner Tracy Lumb, Reactor Engineer Examiner (Trainee)

James Stair, Resident Inspector Facility Personnel Thomas R. Markowski, Day Shift Operations Supervisor Howard J. Palmer Jr. , Supervisor of Operations Robert G. Byram, Plant Superintendent Kenneth Roush, Supervisor of Nuclear Instruction Arthur Fitch, Operations Training Supervisor William G. DiDomenico, Simulator Instructor William Lowthert, Plant Training Manager 7. Summmary of NRC comments made at exit interview:

While the examiners noted individual strengths and weaknesses while administering the oral exams, no generic strengths or weaknesses were identified for the licensing clas The personnel of the SSES Training and Operations departments were very cooperative throughout the examination period. The examiners acknowledged the cooperation and assistance of the simulator instructor In conducting the plant walkthrough examinations, the examiners experi-enced no access delays and noted that the plant appeared clea The training material provided for preparation of the examinations was well written and generally complet Despite the lack of malfunction flexibility provided by the SSES simula-tor, the machine was still acceptable for NRC operator examination pur-poses. The examiners recognized the age of the simulator and the facilty plans to improve the simulator's range of malfunctions within the next two year I J

.-

a Summary of facility comments and commitments made at exit interview:

The facility acknowledged the NRC comments and thanked the examiners for their cooperation and professionalism. The facility felt that the written exams were challenging and fai . CHANGES MADE TO WRITTEN EXAMS DURING EXAMINATION REVIEW:

All comments about the written examinations were resolved during the exam review conducted with the facility. The examiners left the site with no unresolved comments. The following list represents significant changes made to the written examination R0 EXAM Answer N Change Justification 2.02 Added throttling of These are additional methods HX outlet valve or as identified by the facility RHRSW valv in RHR SDC procedur .05 Accepted " false" with The Suppression Pool may also explanation, be used if the CST is unavailabl .06 Added " rejecting This is an additional unfiltered water to condition identified in the Radwaste". RWCU procedur .07 Also accepted valves ON-159 Attachments A and B listed in ON procedur list additional valve responses to the NS4 signa references).

2.10 Added requirement for Additional information candidates to include required for full credi LPCI injection logi .04 Added "When Cont In Additional correct answer PB is depressed" as identified during revie acceptable answe _

.

>

,

-*

R0 Exam (Continued)

-Answer N Change Justification 3.05 Corrected runback to Corrected typo; took 30%.-Also noted that possible question SSES does not normally interpretation into increase recirc flow consideratio until 60% powe .08 Corrected for number Original answer incorrec of key switche .08 Changed "30%" to "24%" Original answer incorrect of rated first stage due to lesson plan error, pressur .02 Corrected from "None" Original answer incorrec to "E0-100-101".

4.10 Accepted " increase". Clarification of procedure provided by facilit SRO EXAM Answer N Change Justification 5.09 " Increases" Original answer incorrec .02 Removed "Rx pressure This interlock has been below 600 psig" from jumpered out at SSE answe .04 Accepted also- These are additional correct power level changes, answers not contained in the CRDH system parameter original question referenc change .04 Accepted also- This is an additional cause local SRI switch of a rod drif actuatio .05 "e" added to ke Each of these signals is a "a" added to key, possible RPS initiation signal for events 4 and .06 Deleted from exam; Question is inaccurate points redistribute because fuel zone range is calibrated without recirc flo ,

.

SR0 Exam (Continued)

Answer N Change Justification 6.06 " accurate" is also Upon ADS. actuation, acceptabl pressure indication will be at the lower end of rang .07 Rod block added to A rod block will occur answe due to the APRM flow uni .09 Added Circ Water Inadvertantly omitted pump tri .in original answe .03 Accepted additional The Discussion Section of alternate answers as the ON contains a broader marked on key, explanation for the basis of each ste .10 Accepted also-Loss Alternate correct answer of NPSHA for LP ECC identified in E0P Base .02 Question changed It was necessary to be during examinatio more specifi Answer changed to Answer changed to match Site Emergenc modified questio .06 Answer modified Original answer inaccurat as'shown on ke With this specific TS problem, there are several possible and equally correct interpretation Attachments: Written Examination and Answer Key (SR0) Written Examination and Answer Key (RO)

-

ESTER hf7RCl7171C17 l

.

U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: S__----

-_USQUEH A_N__N_A__1 &_2_--- _ __ _ __

REACTOR TYPE: _gWR-gE4______________ __

DATE ADMINISTERED: 87/03/16 EXAMINER: _KgLgNAugKI1_ ________

CANDIDATE: _ _

_ ______

IN@IgyCIJgNg_Ig_C@NQ1991El Use separate paper for the answer Write answers on one side onl Staple question sheet on top of the answer sheet Points for each question are indicated in parentheses after the question. The passing trade requires at least 70% in each category and a final grade of at Icast 80%. Examination papers will be picked up six (6) hours after the examination start % OF CATEGORY % OF CANDIDATE'S CATEGORY

__YelyE_ _19106 ___gCggE___ _ygLyE__ ______________C9IEgggY______ ______

2St99-_ _23t99 ___________ ________ THEORY OF NUCLEAR POWER PLANT

'

TH RMOD NAM C I

_243 99__ _24 99 ___________ ____ _ PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION

_2@g99__ _2@ 99 ___________ ________ PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL

_22199-_ _22199 ________-__ ________ ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 199t99-_ ___-_______ _ _ _ _ _-

x Totals Final Grade All work done on this examination is my ow I have neither given nor received aid.

, _ - - - - - - - - - _ - _ _ _ _ - - _ - - - - = _ _ - - - = = _ -

Candidate's Signature

!

l

,

b

.'

. 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 penaltie . Restroom trips are to be limited and only one candidate at a time may leav You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheatin . Use black ink or dark pencil only to facilitate legible reproduction . Print your name in the blank provided on the cover sheet of the examinatio . Fill in the date on the cover sheet of the examination (if necessary). Use only the paper provided for answer . Print your name in the upper right-hand corner of the first page of each section of the answer shee Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write gnly gn gne side of the paper, and write "Last Page" on the last answer shee . Number each answer as to category and number, for example, 1.4, . Skip at least three lines between each answe I 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or tabl . Use abbreviations only if they are commonly used in facility literatur . 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 require . Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or no . Partial credit may be give Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLAN . If parts of the examination are not clear as to intent, ask questions of the examiner onl . You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been complete .

.

10. When you complete your examination, you shall: Assemble your examination as follows:

(1) Exam questions on to (2) Exam aids - figures, tables, et (3) Answer pages including figures which are part of the answe Turn in your copy of the examination and all pages used to answer the examination question Turn in all scrap paper and the balance of the paper that you did not use for answering the question Leave the examination area, as defined by the examine If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.

l l

f I

l

>

I l

l E---____________-__-_____-___--______-_-----________.___-_ _-

..

5:__IMEggy_g[_@gghg86_EgggB_fb@NI_QEg@@I]QNx _Ebgippt_@NQ PAGE 2 IMEBDggyN9digg QUESTIDN 5.01 (2.00)

o. According to the SSES Technical - Specification, " shutdown margin (1.5)

is the amount of reactivity by which the reactor is subcritical or would be subcritical assuming all control rods'are fully

"

inserted encept for . . .

List the three (3) additional conditions for this definitio Give the minimum acceptable value for shutdown margin as stated (0.5)

in SSES TS 3/4. (Include units.)

QUESTION 5.02 (2.00)

The purpose of the end-of-cycle recirculation pump trip (EOC-RPT) (2.0)

(as described in the SSES Tech Spec Bases) is to recover the loss of thermal margin that occurs at ED List and briefly explain two (2) of the reasons that cause a high reactor pressure transient to be more severe at EO QUESTION 5.03 (1.50)

SSES Unit 1 Procedure DP-164-OO1 (" Reactor Recirculation"), contains (1.5)

o prerequisite that limits the temperature differential between the Reactor Vessel Steam Dome and Bottom Head Drai If the Bottom Head Drain temperature recorder reads 395 degrees F, what is the maximum allowable reading on the narrow range RPV pressure instrument that would meet this requirement?

Show all work completed in arriving at your answe (seses CATEGORY 05 CONTINUED ON NEXT PAGE 88888)

.

5 __IHEg6Y_gE_NyCLE@g_EgMEB_E68NI_gEEg611 ggt _E(ylpgi_ONg PAGE 3 IDEggggyNODICS

.

QUESTION S.04 (3.00)

The HPCI system will automatically swap from the CST suction to the suppression pool suction on low CST level. Failure of this cutomatic transfer could result in inadequate Net Positive Suction Head Available (NPSHA) for the HPCI pum The following information applies to the HPCI system:

HPCI pump suction elevation = 650.25 feet Transfer level elevation = 673.75 feet Suctica transfer point = 3.6 ft above bottom of CST Net Pasitive Suction Required = 15 feet Head loss = 7.1 feet CST diameter = 40 feet CST water temperature = 80 deg F c. Calculate the NPSHA for the HPCI pump if the level in the CST (2.0)

is at the suction transfer poin Given: NPSHA = u (P-Psat)

+ z h1 If the level in the CST is at the suction transfer point and the (1.0)

suction failed to transfer, how long could the pump operate before cavitation could be expected (i . e. , before the CST would be emptied)7 Assume HPCI at rated flow condition QUESTION S.05 (1.00)

SSES Tech Spec 3.1.4.1 states: (1.0)

"The Rod Worth Minimizer (RWM) shall be operable in operational conditions 1 and 24 when thermal power is less than or equal to 20% of rated thermal power."

Why is the RWM not required to be operable if reactor power is greater than 20%7 QUESTION S.06 (1.00)

CSES procedure EO-100-113, " Level / Power Control", requires a (1.0)

reduction in RPV water level in order to reduce reactor power during an ATW Explain how lowering reactor water level will reduce reactor powe (***** CATEGORY OS CONTINUED ON NEXT PAGE *****)

.

01__IMEggy_g[_NgC(g@B_EgMEB_P(@NI_gPEB@llg@t_E(ylQgt_@Ng PAGE 4 IUE6dgQyN@dlCS

.

QUCSTION 5.07 (2.00)

A startup is about to commence at SSE The Reactor Engineer has (2.0)

calculated the estimated critical position for this particular startu State the EFFECT on the ECP for each of the following situation That is, state whether the reactor will go critical with LESS rod withdrawal, MORE rod withdrawal, or that there will be NO CHANGE in the EC Consider each case separatel c. RWCU isolates (significant decay heat). l Moderator temperature drop Reactor head vent is inadvertantly close Shutdown Cooling isolates (significant decay heat).

QUESTION 5.00 (2.50)

c. Step 6.30 of SSES GO-100-02 (" Plant Startup and Heatup") directs (1.0)

the operator to continue with control rod withdrawal until a stable positive period of 100 seconds is achieve Assume that this period is maintained and that an operator wants to verify the accuracy of the period mete He decides to time l the indicated power increase from 20 on IRM range 3 to 100 on IRM !

range How long should this take if the period meter is accurate?

l Assume that the reactor scrams from full power. Explain the (1.5)

response of indicated reactor power level as indicated by nuclear i instrumentation versus actual core thermal output for approximately I thirty (30) minutes after the scram.

! l l

l l

(*ss** CATEGORY OS CONTINUED ON NEXT PAGE *****)

.

5:__IBEggy_QE_NUGLE86_EgMEB_EL8NI_QEEBOIlgN t_ELUID@t_8ND PAGE 5 IME8599yN@dlGS

.

QUESTION 5.09 (3.00)

c. Define the term Critical Power (CP). (1.0) State how Critical Power would change for each of the following (2.0)

events ( i . e. , INCREASE, DECREASE, or NO CHANGE).

Assume that the reactor is at full powe Consider each event separatel . Loss of a feedwater heater string 2. Main Turbine Trip (Consider for the time immediately prior to the reactor scram.)

3. Recirc Flow Control system fails to maximum demand 4. Feedwater Control system fails to maximum demand QUESTION 5.10 (1.50)

'

o. There is an orifice located downstream of each pump in the (0.75)

Control Rod Hydraulic System (CRDH). What condition are these orifices designed to prevent? What is the adverse effect to the pump if this condition occurs? (0.75)

,

QUESTION 5.11 (1.50)

i Tech Spec Table 3.4.4-1 specifies the Reactor Coolant chemistry (1.5)

limit Explain why the most restrictive chloride limits are l given for startup conditions.

L

,

(48888 CATEGORY 05 CONTINUED ON NEXT PAGE *****)

I

.

5 __IBEggy_gE_Ngg6E86_EgMEB_E6@NT DEEB@llgN1 _E691pyt_@Ng PAGE 6 IUEggggyN@digg

.

DUESTION 5.12 (3.00)

Figure 1 contains charts of several key reactor parameters following a Feedwater Controller Failure to Maximum Deman For the areas marked, give the cause of each parameter change as ctated belo A- State why reactor power rises at '20 seconds then immediately (0.75)

decrease B- State why feedwater flow drops sharply at '20 second (0.75)

C- State why core flow drops at '20 second (0.75)

D- Explain the variations in steam flow from '20 - 30 second (0.75)

>

f (***** END OF CATEGORY OS *stes)

, .

6.__PL9NI_SYgTEMS_ DESIGN t_CQNTRgL t_AND_INSTRUMENTATIgN PAGE 7

.

QUESTION 6.01 (3.00)

c.-SSES DN-024-OOl lists sixteen (16) Emergency Diesel Generator (1.0)

(EDG) trips for a non-emergency start. Of these, list those EDG trips which are NOT bypassed for an emergency star What action (s) outside of the control room would an operator (1.0)

take in order to manually shutdown an EDG f ollowing an emergency start? Why is it undesirable to stop the EDG by using the Emergency (1.0)

Stop pushbutton (located on local control panel OCS21A) in any other case than an emergency?

QUESTION 6.02 (3.00)

Several RPG scram signals are listed below. For each signal, state (3.0)

whether or not the scram signal can be bypassed (either manually or cutomatically).

IF the signal can be bypassed MANUALLY, give the Reactor Mode Switch

~ position (s) for which the bypass can be initiated and give the operator cction(s) taken to initiate the bypas ,

IF the signal can be bypassed AUTOMATICALLY, give the Reactor Mode Switch position and applicable plant conditions (if any) that initiate the automatic bypas Scram Discharge Volume Hi Hi Level Scram Drywell Hi Pressure Scram c. MSIV Closure Scram

!

l i ,

i I

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

+

.

6 __PL@N1_gyglEdg_gEgigN2_CgglBgL 1,9Ng_1NglBUDEN1811gN PAGE 8

.

QUESTION 6.03 (3.00)

c. From the following list, choose the RCIC interlocks which (1.25)

are defeated when RCIC is controlled from the Remote Shutdown Pane . RCIC Turbine Overspeed Trip RCIC Turbine Trip on High Exhaust Pressure RCIC Turbine Trip on High RPV Water level Automatic Suction Transfer from CST to Sup Pool on Low CST level RCIC auto initiation on RPV Level 2 (-30")

,

b. Using attached Figure 2, "RCIC", trace the flowpath for Pressure (1.25)

Control using RCIC while cooling the plant down from the Remote Shutdown Panel, c. Answer TRUE or FALSE: If the CST level switch fails low, it is (0.5)

possible to switch the RCIC pump suction back to the CST by closing F031.

'

,

QUESTION 6.04 (2.50)

c. While at power, the ROD DRIFT alarm at 1C651 annunciates. Name (1.0)

three (3) other indications or alarms in the main control room you would look for to verify the control rod drift b. List three (3) malfunctions within the Control Rod Drive (1.5)

Hydraulic System or in systems that interface with the CRDH system that could be causing the control rods to drif s (***88 CATEGORY 06 CONTINUED ON NEXT PAGE *****)

,

.

bz__PL9NI_SYSIEdg_QE@lGNx _CQNIBQL1 _8NQ_INSIBQDENI@llgN PAGE 9

.

QUESTION 6.05 (2.50)

Assume that a reactor scram occurs for each of the following event For each event, match the condition that sends the scram signal to RP (Your answers should read 1-5 with a letter for each answe Conditions may be used more than once.)

-EVENTS-1. Loss of Stator Cooling at 100% power (0.5)

2. Main Steam Line Low Pressure in RUN Mode (0.5)

3. Complete loss of Feedwater at 100% power (0.5)

4. Closure of a single MSIV while at 100% power (0.5) Turbine trip (without bypass) at 20% power (0.5)

-CONDITIONS- High Neutron Flux RPV Low Water Level b. Low Reactor Pressure High Reactor Pressure Turbine Stop Valve Closure MSIV Closure

!

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

l l

-

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

.

6:__PLONI_gYgIEDS_QEglGNt_CQNIBQLt _8NQ_INSIBudENI@IlgN PAGE 10

.

QUESTION 6.06 (2.50)

c. Assume that Drywell temperature increases to 200 deg F during an (1.5)

accident at SSES. For the following RPV water level ranges, state whether indicated level will be HIGH, LOW, or ACCURATE as compared to the actual water level during the acciden Wl(kd ~LK@gittl57

!. I.m i I wu =- y 4. Wide Range Narrow Range 5. Shutdown Range Extended Range Upset Range b. SSES Unit 2 is rapidly depressurized during an ADS blowdown. For the following RPV instrumentation, choose the phrase that correctly completes the sentenc . The narrow range water LEVEL indication will indicate (0.5)

(A LOWER THAN ACTUAL, A HIGHER THAN ACTUAL, or AN ACCURATE)

RPV water level readin . The narrow range PRESSURE indication will indicate (0.5)

(A LOWER THAN ACTUAL, A HIGHER THAN ACTUAL, or AN ACCURATE)

-

. reactor pressure readin QUESTION 6.07 (2.50)

For the following events, choose the auto action or actions, if any, (2.5)

that will be initiated by the APRM Trip Unit AUTO ACTIONS - ROD BLOCK / HALF SCRAM / SCRAM c. At 100% power, APRM C has thirteen (13) LPRM inputs b. At 100% power Flow Unit A fails downscale l

c. APRM B fails downscale at 75% power Flow Unit A oypassed with joystick at 100% power o. APRM E fails upscale in Startup Mode

,

!

i

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

. - _

.

6:__PLBNI_gy@lEgg_DEgigNt_CQNIBQL1 _BNQ_lNgIBUDENIBIlgN PAGE 11

.

QUESTION 6.00 (1.50)

a. List the automatic start signals for the ESW pumps A, B, C, and D. (1.0)

Include any applicable time delay b. Assume that a Loss of Offsite Power (LOOP) occurs and the ESW (0.5)

pumps are running. If a LOCA then occurs, will the ESW pumps continue to run?

DUESTION 6.09 (2.50)

a. Give the conditions for which the Recirculation Flow Control (1.0)

System will enforce the #2 Speed Limiter. (That is, limit demand signal to 45% of rated speed.) Include setpoints, if applicabl b. SSES Unit 1 is operating at 40% power. The Recirc pumps are at (1.0)

minimum speed in individual speed control. If the Manual / Auto Transfer Station is inadvertantly placed in AUTO, briefly explain what happens to the speed of Recirc pump c. Answer TRUE or FALSE: If a scoop tube lockout occurs at full (O.5)

power, the #1 and #2 Recirc speed limiters will still be able to runback Recire if their respective logics are satisfie QUESTION 6.10 (1.00)

SSES Unit 1 OP 149-005, "RHR Operation in the Suppression Pool (1.0)

Cooling Mode", recommends the use of RHR pumps 1P202C and D for use in Suppression Pool Cooling if no LPCI initiation signal is presen Why are these pumps preferred?

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

e 2r__PggggpuBEg_;_NgBD@61_9BNgBD66t_EDEggENgY AND PAGE 12

.

889196991906_99NI696 QUESTION 7.01 (2.50)

s. Assume that a reactor scram from full power is imminent. As the (1.5)

Control Room SRO, you could take certain actions to reduce the impact of the scram on the plant. Name three (3) of the actions that you would tak b. The bases of EO-100-01, " Scram", list two (2) purposes served (1.0)

by putting the made switch to SHUTDOWN after a scram occur State the two reason QUESTION 7.02 (1.75)

Assune that the control room must be evacuated due to dense smoke caused by a fire. If the control room is evacuated prior to scramming the plant manually, EO-100-OO9 instructs the operator to open the following breakers: CD2A on RPS distribution panel 1Y2OIA and CBBB on RPS distribution panel 1Y201 a. State 2 purposes (i e. , plant equipment response) accomplished (1.0)

by opening these breaker b. Besides opening the breakers listed above, the operator must (0.75)

perform one other local immediate action. State the actio (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)

}

.

2 __P899gggBEg_;_Nggd@61_9pNQ858L _EDEggENCy_ONQ t PAGE 13 60919LQQJCOL_CgNIBgL

.

QUESTION 7.03 (3.00)

a. SSES Unit 1 is in Operating Condition 5 with the vessel head (1.5)

remove Both loops of the shutdown cooling mode of RHR are then lost. Section 3.3.1 of SSES ON-149-OO1 (Loss of Shutdown Cooling) suggests the following actions. Fill in each blank with the REASON for each suggested ste .3.1 a. MAINTAIN water level 90 to 100 inches on shutdown indication to _ ._ _ _ _ _ _ ____ ____________ OR-3.3.1 b. OPERATE Reactor Water Cleanup System at its maximum flow rate in accordance with OP-161-OO1 to

.

3.3.1 c. START Reactor Recirculation System with at least one pump in minumum speed in accordance with OP-164-OO1 to _______________________ _

_.

NOTE: DO NOT consider part "a" when answering parts "b" and "c".

b. Section 3.3.2 a. of DN-149-OO1, " Loss of Shutdown Cooling", (0.75)

lists systems available to ADD water to the RPV. List three (3)

of these system c. Section 3.3.2 b. of ON-149-OO1, " Loss of Shutdown Cooling", (0.75)

lists systems available to REMOVE water from the RPV. List three (3) of these system (***** CATESORY 07 CONTINUED ON NEXT PAGE *****)

. . - . .

..

21__PggCgggggg_;_Ng858(t_@@NggDg(g_gDERgENCY_ANg PAGE 14 80919L991 COL _CgNIBgl

&

DUESTION 7.04 (2.00)

For each of the following plant conditions or events, (2.0)

1. . state the EDP or EOPs (Emergency Operating Procedures) entered (by title or procedure number), and give the specific EDP entry condition met. Include the setpoint, if applicable, Suppression Pool Temperature = 105 deg F , Drywell Pressure = 2 psig Main Steam Line Rad Monitors read 7.5 times normal full power backgroun Reactor Pressure = 1100 psig HVAC Zone III exhaust reads 3.0 mrem /h QUESTION 7.05 (3.00)

The following questions concern GO-100-OO2, " Plant Startup and Heatup".

a. A CAUTION statement located in Section 6.2 of this procedure (1.0)

'

states: DPENING MSIVs WITH CONDENSER VACUUM ESTABLISHED AND REACTOR _ HEAD VENT OPEN MAY RESULT IN ___(1) ___ OR ___(2) .

Fill in each blank.

, According to this procedure, under what conditions is the reactor (1.0)

j considered critical?

l I Why is it important to monitor turbine first stage pressure (1.0)

during shell warming?

QUESTION 7.06 (2.25)

Tables SC-2 and SC-3 list Secondary Containment Maximum Operating values for area temperature and radiation levels. (See Figure 3.)

l For Table SC-2, briefly explain the difference between (1.5)

L " Maximum Normal" and " Maximum Safe" area temperatures.

l For Table SC-3, define " Maximum Normal" area radiation leve (0.75)

!

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

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

.. ._ . .

.-

Zt__P89gEQUBES_ _Ng8586t_8BNg8596t_EDE8@ENGY_@NQ PAGE 15 809196991GOL_GQNIBQL

.

QUESTION 7.07 (1.50)

e. OP-143-OO1, "SJAE and Mechanical Vacuum Pump", states that the (1.0)

mechanical vacuum pump should not be operated when reactor power is greater than 5%. Give two (2) reasons for this precautio Section 3.6 of OP-143-OO1, " Main Condenser Vacuum Breaker (0.5)

Dperation", states that if steam seals are lost and cannot be restored immediately, Main Condenser vacuum must be broke Briefly explain wh QUESTION 7.08 (2.50)

e. ON-135-OO1, " Loss of Fuel Pool Cooling", provides alternate (1.0)

'

methods of cooling spent fuel assemblies. Assume that the fuel pool -

requires makeup because of a leak. List three (3) sources of makeup to the fuel poo .b. Assume that a Loss of Fuel Pool Cooling has occurred as a result (0.5)

of PCV-11036 failing closed. All attempts to regain cooling (including the use of RHR in Fuel Pool Cooling Assist Mode) have failed. The only means of heat removal available is to allow the ,

fuel pool to boi Complete the following sentence with the limits specified in DN-135-OO1: Boiling should not occur bef ore _(1) _ hours after loss of cooling and the fuel pool level must be maintained

_ ( 2) _ feet above the irradiated fuel bundle c. Assume that you are following section 3.5.2 of DN-135-OO1, which (1.0)

contains the steps in providing fuel pool boiling. Briefly explain i WHY you must ensure that Zone III exhaust ventilation must be aligned to exhaust the area directly over the fuel poo QUESTION 7.09 (1.50) ,

i Assume-that Startup Bus 20 was lost due to a fault from Startup (1.5)

. Transformer T-20. In accordance with DN-OO3-OO2, " Loss of Startup Bus:20", what are the three (3) conditions necessary for tie breaker OA10502 to close to f eed startup bus 2O?

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

- . -- -- - .. -. _. . . . ..

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

9 , PROCEDURES - NORMALt _ABNgRMAL1 _gMERggNCy_ANg PAGE 16 R991gLgg1C8L_CgNIBgL

,

DUESTION -7.10 (3.00)

c. Using the attached graphs of the Heat Capacity Temperature Limit (1.5)

(HCTL) and the Heat Capacity Level Limit (HCLL) (Figures 4 and 5),

determine the suppression pool water level for which a controlled RPV depressurization would be required i n accordance with the Primary Containment Control Procedure (ED-100-103).

Given: Reactor pressure = 500 psig Suppression Pool temperature = 170 deg F b. Step SP/L-B of this procedure states that HPCI and RCIC should (1.0)

not be used if Suppression Pool water level decreases below 18.5 feet. Give two (2) reasons for this ste ; c. Why is rapid RPV depressurization required if Suppression Pool (0.5)

level reaches 12 feet?

QUESTION 7.11 (2.00)

OP-153-OO1, " Standby Liquid Control System", directs the operator to inject SBLC if required by EO-100-102 (RPV Control).

a. What is the limiting condition given by EO-100-102 at which (1.0)

I SBLC MUST be injected?

'

b. Under what two (2) conditions (as stated in EO-100-102) may (1.0)

SBLC injection be terminated?

l l

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

' y w w e- r y y-,y w----swm -M- y7e-+y----y,-*---6-m -e * +=--3+t- --eem --Y-4me-+ -+-ywe ee---------e---w

. . - _ _ . - - . ..

-

.

Dz__0DglNigIB8IlVE_P89CEDUBEg1_CQNQlIlgNgi_8NQ_LIMlI@IlgNg PAGE 17

.

QUESTION O.01 (3.00)

a. State the four (4) Safety Limits as given in Section 2.1 of the (2.0)

SSES Technical Specifications. For each, include any applicable setpoints or conditions and the applicable plant operational condition (s). What IMMEDIATE action must be taken in accordance with the SSES (1.0)

Tech Specs if a Safety Limit is exceeded while the unit is at power?

QUESTION 8.02 (3.00)

Classify each of the following events as an UNUSUAL EVENT, an ALERT, a SITE AREA EMERGENCY, or a GENERAL EMERGENCY. A copy of the SSES Emergency Classification Guide (EP-IP-OO1) is attached as Figure i . Building Vent Monitoring System indicates a total site 14TCI)

release rate for I-131 equal to 1.5 E3fcGi/ mi MSL "C " t5 2. A h linea break)mh(dt of SSES occurs at Mm CmtainM Unit 2. An RO reports 1,1<ti)

RPV water level at -180" on the Fuel Zone Range. He then reports that "SIV "C" has failed to clos V OLA60dVd -

ut nis/tr-chawjdk y( gyp

for 5 minn QUESTION B.03 (2.00)

Answer the following TRUE / FALSE questions concerning the SSES Emergency Plan.

If the Emergency Plan is initiated, the Shift Supervisor (SS) (0.5)

!

assumes the role of the EMERGENCY DIRECTOR until relieved by the PLANT SUPERINTENDEN The TSC is the central location for offsite emergency managemen (0.5) With regard to in plant radiation levels, The RESTORATION (0.5)

,

'

ORGANIZATION may be implemented as long as in-plant radiation levels are decreasing.

I It is the responsibility of the RECOVERY MANAGER to make the (0.5)

final determination regarding the establishment of the RESTORATION ORGANIZATION.

!

,

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

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

. -

0 __0951NiglB81]yE_PBgCEQUBE@t_C9NpillgNg2_@NQ_Lidl1911gNg PAGE 18

,

QUESTION 8.04 (2.00)

a.~ List the minimum shift complement for Operations personnel as (1.5)

stated in DI-AD-015, " Minimum Shift Manning". Include position and type of NRC license held, if an Fill in the blank: The Day Shift Supervisor may REDUCE the (0.5)

Minimum Shift Complement from that given in DI-AD-015, as long as _ _

_ _

__________________ .

QUESTION 8.05 (1.50)-

MATCH the following tags with their designated meaning as given in (1.5)

SSES AD-DA-103, " Protective Permit and Tag System".

-TAG- -MEANING- Yellow The tagged device is not to be operated until properly cleared by the permit holde Red The tagged device is in a controlled status-and is not to be operated unless the restrictions on the tag are me Gtriped The tagged device is not to be operated except by request of the permit holder ,

AND on orders of the S.D. representative.

]

l l

l

<

I

!

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

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

- - _ -

~

9__b9 MINI @l80llVE_P6gCEpuSE@t_CgNQlligN@t_@NQ_LIMII8IlgN@ PAGE 19

.

QUESTION B.06 (3.00)

( cftcf sDY'

SSES Unit 1 is operating at 90% power. APRM "B" is bypasse APRM "F" then fails upscale and is removed from service by the I & C Departmen What actions are required by the SSES Tech Specs for the above (1.0)

situation? Assume now that in addition to the above failures, APRM "A" is (2.0)

bypassed. Also, the I&C Supervisor informs you that the Channel Functional Test on APRM "C" i s due immediatel l What actions are required by Tech Specs in order to allow this

, surveillance to be completed? j

NOTE: USE THE ATTACHED TECHNICAL SPECIFICATIONS TO ANSWER THE ADOVE QUESTIO DE SURE TO REFERENCE ALL APPLICABLE TS THAT YOU USC IN DEVELOPING YOUR ANSWE ******************************************************************

QUESTION 8.07 (3.00)

SSES Unit 1 is at 100% power. The "B" Main Steam Line Radiation Monitor fails downstal Can reactor operations continue? What Tech Specs apply? (1.0) What Tech Specs would be applicable if Main Steam Line Monit r (2.0)

"C" then failed downscale?

- - - - ****

NOTE: USE THE ATTACHED TECHNICAL SPECIFICATIONS TO ANSWER HE ABOVE QUESTION. BE SURE TO REFERENCE ALL APPLICABLE TS THAT YOU USE IN DEVELOPING YOUR ANSWE ********************************************************* ********

LK1 (130 03/I'l87 Chomyd -h) ;

\

\1. - C A M S S WY Vti\\( V V L hl (*"YLW if MS wkod n KA' aFPM

\(- Wo - W h3 Norf (***** CATEGORY OO CONTINUED ON NEXT PAGE *****)

~

Dz__A901Niglg8IIVE_PBgCEpuBE@t_CgNQlligNgt_8NQ_ Lid 11811gN@ PAGE 20

.

QUESTION 8.00 (3.00)

c. SSES Unit 1 is at 50% power. You learn from the previous Shift (1.0)

Supervisor that during individual control rod scram time testing, control rod 19-30 was f ound to be immovable in the fully withdrawn position. The testing is scheduled to continue during your shif What actions are you required to take in accordance with the Tech Specs if control rod 27-38 is also found to be immovable in the f ull y wi thdrawn posi tion?

(NOTE- A core map is attached as Figure 7.) If a single control rod was found to be immovable with normal (2.0)

Control Rod Drive pressure during Startup, could the startup continue? Give any applicable Tech Spec requirement ******************************************************************

NOTE: USE THE ATTACHED TECHNICAL SPECIFICATIONS TO ANSWER THE ABOVE DUESTION. BE SURE TO REFERENCE ALL APPLICABLE TS THAT YOU USE IN DEVELOPING YOUR ANSWE ******************************************************************

3(iL{ V1 QUESTION 8.09 (2.50) 2, g l330 cbydM9W bas &

While at power, 125 VDC battcry ch= g=c ID61/ and 250 VDC bettcry ID652 become inoperabl kk# Give all applicable Technical Specification (1.0) Give all applicable Tech Specs, if, in addition to the above (1.5)

malfunctions, 480 V swing bus 19229 became inoperabl ******************************************************************

NOTE: USE THE ATTACHED TECHNICAL SPECIFICATIONS TO ANSWER THE ABOVE QUESTION. BE SURE TO REFERENCE ALL APPLICABLE TS THAT YOU USE IN DEVELOPING YOUR ANSWE ******************************************************************

,

.

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

I

'

Bz __0DMINigl8@IIVE_P89CEQQ8ES 1 _CgNgillgNg1_@NQ_LIMlI@IlgNg PAGE 21

..

QUESTION 8.10 (2.00)

SSES Unit 1 is in cold shutdown. A unit startup is scheduled to (2.0)

commence in three (3) days. The Maintenance Supervisor reports the failure of the outboard blower for the MSIV Leakage Control Syste He estimates that it will take ten (10) days to return the blower to servic In view of the above malfunction, is it possible to maintain the present startup schedule without violating the SSES Tech Specs?

Explain your answer referencing all applicable Tech Spec requirements.

l ******************************************************************

NOTE: USE THE ATTACHED TECHNICAL SPECIFICATIONS TO ANSWER THE ADOVE QUESTIO DE SURE TO REFERENCE ALL APPLICABLE TS THAT YOU USE IN DEVELOPING YOUR ANSWE ******************************************************************

!

l QUESTION B.11 (2.00)

l For each radiation worker described below, determine if any SSES Administative or Federal (10 CFR 20) radiation exposure limits have been exceeded. Show all work and state which particular limits, if any, have been exceede year old male, with an NRC Form 4 (1.0)

lifetime exposure = 100 rem He has received 2000 mrem this current quarter.

l year old male, with an NRC Form 4 (1.0)

l lifetime exposure = 50 rem He is a new employee and has no documentation of his radiation exposure for the current quarter. He then receives 400 mrem during his first month at SSES, which is in the same calendar quarte (***** END OF CATEGORY 08 *****)

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

-,

. 97-.

,

,.-

' p . . .

- THEORY OF NUCLEAR POWER PLANT OPERATION t _FLUIpS 1_ANQ PAGE 22 IUER5ggyNAdlCS

-.

ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-KOLUNAUSKI, L.

.

ANSWER 5.01 (2.00) . The single rod of greatest reactivity worth is withdraw (0.5) .

The reactor is cold; 68 deg F. y The reactor is Xenon fre (0.5)

-02 l O.38*/ delta k/k .

(0.5)

(with the highest worth rod analytically determined, or O.28% ,

delta k/k with the highest worth rod determined by test.)

REFERENCE SSES SCO23 A-6, Reactor Control Specific Objectives 10, 11, 12 pages 7-10 Tech Spec 3/4.1.1 Bases

.

292002, Neutron Life Cycle (Group II Rx Theory)

K1.10 Define SDM K1.14 Predict change in SDM with core param changes 2.9 i

ANSWER 5.~ O2 (2.00)

(2 required, 1.0 each)

1. At EOC, there is a slower scram reactivity insertion rate (0.5)

because the' control rods are further 6ithdrawn than at BOC (0.5).

2. The void coefficient is more negative at EOC (0.5), so more positive reactivity is added at EOC than BOC for the same change in core void content (0.5).

3. Beta, the delayed neutron fraction, is smaller at EOC (0.5).

This results in a shorter reactor period at EOC than at BOC for the same positive reactivity addition (0.5).

REFERENCE SSES Transient Analysis; SCOO7C, Pressure Increase Transients Specific Objective pages 19-20 292003, Reactor Kinetics (Reactor Theory Group II)

K1.06 Explain effects of delayed neutron fraction on Rx powe .7 295001, Partial loss of forced core flow circulation (AE Group II)

AK1.03 704 of op implic of the thermal limit concep .1

.

5:__IUgD8y_DE_NUg6E86_EgMEB_ELONI_DEEBATIONt _FLUIQS t_ANQ PAGE 23 IUE8dgQyNOMIGS

.-

ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-KOLONAUSK1, ANSWER 5.03 (1.50)

Maximum delta T = 145 deg F (0.25)

X - 395 deg F = 145 deg F (0.5)

X = 540 deg F Psat for 540 deg F -> 962.79 psia (0.5)

962.79 - 14.7 = 948.09 psig (0.25)

REFERENCE Saturated Steam Tables SSES Procedure OP-164-OO1 SSES SCO23 D-1 Rev O, Thermo Fundamental; Specific Objective 6 SSES SCO23 D-3.Rev O, Steam Tables; Specific Objective 1 293003, Steam (Group II Thermodynamics)

K1.23 Use saturated steam tables ANSWER 5.04 (3.00)

e. NASHA = L) (P-Psat) +z h1 (2.0)

= [O.016ft3][(15 - 0.5)1bf3[144in2] + (673.75 - 650.25)ft - 7.1ft Ibm in2 ft2

= 33.4 + 23.5 -7.1 feet

= 49.8 feet CST vol ume = Tr r2 h = TC (20 f t ) 2 (3.6 ft) = 4524 ft3 (1.0)

HPCI flow = 5000 gpm Time = 4524 ft3 * 7.48 gal /ft3 * 1 min /5000 gal

= 6.77 min REFERENCE i SSES Unit SCO23 E-4, Fluid Mechanics- Pumps; Objective 8 l Saturated Steam Tables 293006, Fluid Statics (Group III Thermodynamics)

K1.10 Define NPSH 2.8

291004, Pumps (Components)

K1.06 Need for NPSH; effects of loss of suction 3.3

,

i

'

a THEORY OF NUCLEAR POWER PLANT OPERATIONt _ FLUIDSt _AND PAGE 24 IBERMODYNAMICS

.

ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-KOLONAUSKI, ANSWER 5.05 (1.00) N The purpose of the RWM is to limit rod war . (1.0)

When thermal power is greater than 20%, here is no rod with high Gnough worth which, if dropped at the design rate of the velocity limiter, could result in cladding damage (a peak enthalpy of 280 cal / gram).

REFERENCE SSES SYO17 K-6 RWM, Objectives 1, 5, pages 2-3 Mitigation of Core Damage, SCOO6 L SSES TS Bases 29 314, Inadvertant Reactivity Addition (Group I EA Evol)

SG #3 KN of LCOs and safety limits ANSWER 5.06 (1.00)

>

If RPV level is decreased, the natural circulation driving head (1.0)

is decrease This reduction in core flow will cause an increase in voiding and decreased reactor porte REFERENCE SSES SCOOO6 K SSES EUP Specific Objectives 2, 11 295037 ATWS (EA Evol Group I)

EA2.02 AB to relate RPV water level to ATWS 4.2

. ANSWER 5.07 (2.00)

a. more rod withdrawal (O.5)

b. less rod withdrawal (0.5)

c. no change in ECP (0.5)

d. more rod withdrawal (0.5)

REFERENCE SSES Licensed Operator Science, SCO23 A-7 Objective 3 292005 Control Rods (Group II Reactor Theory)

K1.09 Explain direction of change in CRW for a change in mod temp. etc. .

5:__IUEg8y_gE_NUCLEOB_EgWE8_ELONI_9EEBOIlgNz _ELUlygz_8NQ PAGE 25 IUE8dggyN@dlC@

,

ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-KOLONAUSKI, ANSWER 5.08 (2.50)

a. P = Po e**(t/T) --> t = T In (P/Po) (1.0)

t = 100 sec * In(100/20)

= 100 sec * 1.61

= 161 sec = 2.7 min The nuclear instrumentation will show a continual drop in the (0.75)

fission rate on a -80 second perio The fission rate will continue to decrease until it drops into the source range (where subcritical multiplication effects will cause the fission rate to stabilize).

Core thermal output will drop to the decay heat level, which is (0.75)

about 67. of the original power leve REFERENCE SSES SCO23 A-9 Licensed Operator Science pages 1-4

" Characteristics of an Operating Plant" Objectives 6, 7, 10 SSES Transient Analysis SCOO7 B, Section 2.6, Objective 3 292008 Reactor Operational Physics (Group I Reactor Theory)

K1.08 Describn power, period response once criticality reached K1.27 Reactor power response to control rod insertion K1.29 Define decay heat ANSWER 5.09 (3.00)

a. Critical Power is the bundle power needed to produce the critical (1.0)

quality or the bundle power needed to cause OTB to occur in the bundl . (inlet subcooling ^) CP increases (0.5) (pressure ^) CP decreases (0.5) (core flow ^) CP derrrrr;; inenaeu (0.5) (inlet subcooling ^) CP increases (0.5)

REFERENCE SSES SCO23 G-3 Specific Objectives 3.3, 3.4, Core Thermal Limits (Group 1 Thermo)

K1.17 Define CP K1.22 KN of effect of subcooling on CP K1.23 KN of effect of core flow of CP K1.24 KN of effect of pressure of CP ..

D __INEggy_gE_NgCLE@B_PQMEB_EL@NI_QPEB8IlgNt_ELylg@t_@NQ PAGE 26 IUEBDgQyN@gICS

.

ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-KOLONAUSKI, ANSWER 5.10 (1.50)

c. Pump runout (excessive system flow demands when the scram (O.75)

accumulators are being charged).

b. The increase in flow rate will cause the pump motor to draw (O.75)

a larger current amount which can result in overheating of the moto REFERENCE SSES SCO23 E-4 Pumps page 22, Objective 5 291004 Pumps (Components)

Kl.12 Pump runout- corrective reasures, et .8 ANSWER 5.11 (1.50)

The limit on chloride concentration is to prevent stress corrosion (0.5)

cracking of the stainless stee During startup, there is not much deaeration taking place and the (0.5)

dissolved oxygen content in the coolant may be hig With a higher oxygen concentration, the chloride limit must be (0.5)

lower to ensure that stress corrosion cracking is prevented.

t REFERENCE SSES System Lesson Plan, RWCU SYO17 L-1 page 18, Objective 6 204000 RWCU (Group II Plant Systems)

,

SG #6 KN of TS Bases for LCOs 3.4 l

l

,

I i

,

i

!

i

.

_ _

.

5- IMEQBy_QE_NyCLE98_EQWE8_ELONI_gEEg8IlgN1 _E(ylppg_@NQ PAGE 27 IUE8dggyN901CS

..

' ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-KOLONAUSKI, ANSWER 5.12 (3.00)

A- Rx power increases due to increased subcooling (+p) then (0.75)

decreases sharply due to scram from turbine trip (hi level).

B- FW flow decreases due to FW pump trip caused by high RPV water (0.75)

level (+54").

C- Core flow decreases due to EOC-RPT initiatio (0.75)

D-(Steam flow drops to zero upon turbine trip)then oscillates as (0.75)

the BPVs (and the SRVs) open to reduce reactor pressur REFERENCE SSES SCOO7 D pages 9-12, Objective 4 259002 Rx Water Level Control System (Group I Systems)

KN of effect of loss of FWCS on following parameters:

K3.02 FW system 3.7 / K3.06 Main Turbine 2.0 /

K3.07 RFC Kl.lO KN of relation between RPS and-FWCS 3.9

.

6:__P(@NI_gYgIEgg_gEgigy3_CgglBg61_@Ng_ lng 16UDENI@IlgN PAGE 28

ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-KOLONAUSKI, ANSWER 6.01 (3.00) . High Generator differential current (0.33) Low Engine lube oil pressure (0.33) Engine overspeed (0.33) By taking the mode selector switch to local and depressing the (1.0)

emergency stop pushbutto Automatic cooldown is bypassed because the unit shutdown relay (1.0)

is opened, which prevents the cooldown circuit from energizin (Shid5 ) (>*d Mp )

REFERENCE SSES SYO17 G-1, Objective 5 DN-024-OO1 264000 Emergency Diesel Generators (Group I Systems)

K1.04 KN of E D/G relation to cooling water system K6.03 KN of loss of lube oil effect on E D/G K4.02 KN of E D/G trips t ANSWER 6.02 (3.00)

a. Manually bypassed (0.5) in the Shutdown or Refuel Modes (0.5) by placing the SDV high water level bypass switch in the BYPASS positio (manual, keylocked, on benchboard IC651) (0.5). Signal NOT automatically or manually bypassed (0.5). Automatically bypassed (0.5) when mode switch not in RUN (0.5]n_ r_ LR 3/(([f7

+ ._ _ ; . m L c i m. 600 gu y 40.5). '

CA"~ ;

REFERENCE **P"* " ~

SSES SYO17 L-5 Objective 6 i

l 212000 RPS (Group I Systems)

A4.04 AB to bypass SDV hi level scram signal 3.9

,

K4.12 KN of RPS interlocks which provide bypassing of scram signals (auto or man)

!

l I

!

I

.

6 __P(8NI_SygIEd@_pESJgN _CQNIBgL i _ANQ_lN@lBUDEN1811gN 1 PAGE 29

ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-KOLONAUSKI, .

ANSWER .6.03 (3.00)

' (1.25) See attached Figur (1.25) TRUE (0.5)

REFERENCE SSES SYO17 C-5, Objective 5d 217000 RCIC (Group I System)

K1.01 KN of RCIC relation with CST Control Room Abandonment (Group I EA Evol)

AA2.03 AB to operate to control Rx pressure ANSWER 6.04 (2.50)

c. SCRAM DISCHARGE VOLUME NOT DRAINED on panel IC651 (3 reqd, 0.33 each)

ROD OUT BLOCK on panel 1C651 SCRAM PILOT VALVE AIR HEADER LOW PRESS ALM on panel 1C601 Full Core Display Red Drift lights illuminated LK Wet- atto - C KW yeraW%r5 b. 1. Scram valve leakage YoWtr Ltd CHAM *f8 (0.5)

i Cooling water pressure high (OrNow) (0.5) Low Instrument Air Supply pressure (0.5)

acctyf atro Seltet Wd lnWf (Wl/IWN.M adlydtdl0 cal lj

REFERENCE SSES ON-155-OO6 Rev 2 SYO17 K-2 295019 Loss of Instrument Air (Group II AE Evol)

l AK2.01 KN of effect of air loss on CRD system 3.9 j 201001 CRDH (Group II System)

l A2.12 AB to predict impact of high cooling water flow on CRDH 2.9 l A2.11 AB to predict impact of valve openings 2.7 I

!

__ ___

.

6 __PL@NI_gYSIEgg_pEQ1@Nt _CQNIBQLt_@NQ_lNglByDENI@ll@N PAGE 30

ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-KOLONAUSKI, ANSWER 6.05 (2.50)

1. (Turbine Trip) (0.5 each) (MSIV closure) . . (hi steam flow-MSIV closure) (or a.) (Or t.) (or a.)

REFERENCE SYOl7 L-5 Objectives 5,6 SSES Transient Analysis, SCOO7C- Objective High Reactor Pressure (Group I EA Evol)

EK2.01 KN of relation between hi ex press and RPS estutMcn w10 "<(" fl0W " 'I' 8 I ANSWER 6.06 (2.50) g g _. gag du b/ Wee- If k(cire, diu opediN> o, c. 4. la" accurate ( each)

2. accurate high 3. high high . higher than actual g 3fg[37 (0.5) lower than actual '

RF ts0y'O (0 5)

\0wb~&o(-W (acewadt)

REFERENCE SYOl7 J-2 Objective 9 / SCOO6 G Objectivea 4,5 295028 High DW Temperature (Group II EA Evol)

EA2.03 AB to interpret RPV level ANSWER 6.07 (2.50)

c. Half Scram, Rod Block (0.25 each)

b. Hali Scram (RPS A)3 Vodblock (0.5)

c. Rod B1ock \c h a. h f(c W W t (O.5)

d. None d[ 3/IIIII (0.5)

O. Rod Block, Half Scram (0.25 each)

REFERENCE SYO17 I-4 Objective 2 215005 APRM (Group I Systems)

KN of interrelations between APRM and:

- __ .__

.

6:__ PLAN 1_gYglEdg_QEglGNt _CgNIQQLg _8Np_lNgIQUDENI@llgN PAGE 31

'

ANSWFRS -- SUSQUEHANNA 1&2 -87/03/16-KOLONAUSKI, K1.01 RPS (4.0)

K1.03 RUM (3.5)

ANSWER 6.08 (1.50)

c. Pump A and B start 40 seconds after respective DG start signa (0.5)

Pump C starts 44 seconds after C DG start signa (0.25)

Pump D starts 48 seconds after D DG start signa (0.25) YES (0.5)

REFERENCE SSES SYO17 M-1 Objective 4 264000 Emergency Generators (Group 1 Systems)

AB to predict impact of following: A2.09 Loss of AC A2.10 LOCA ANSWER 6.09 (2.50)

(0.2cxh)

a. Condensate pump discharge pressure <100 psig OR (0.2 )

Individual FW pump flow is <20% OR (O. 5)

1 or 2 FW Heater Hi H1 Level AND (O 25)

RPV level < 30" (Low level alarm point) LK.' 3fgg[py ( .25)

OR M Circ.Why pcketivL tYt The Recirc pump A speed will increase to the lower limit of (1.0)

the Master Controller (which is ed% speed).

47 FALSE (0.5)

REFERENCE SYO17 L-9 Objectives 2,3 202002 Recirc Flow Control (Group 1 Systems)

K4.02 KN of RFC interlocks for Recirc pump speed control K3.OS KN of effect of RFC malfunction on Recirc pump speed 3.3 l

.

$1__P(@@I_@Y@l@D@_Q@@l@Nt_Q9MI69(t_@@p_ INSTRUMENTATION PAGE 32

ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-KOLONAUSKI, ANSWER 6.10 (1.00)

These pumps would trip if a LOCA signal was received for Unit 2, (1.0)

leaving RHR pumps 1P202A and B available f or LPCI initiation on Unit REFERENCE OP-149-OO5, Rev 3 219000 Suppression Pool Cooling Mode (Group II Systems)

K1.04 KN of relation between LPCI pumps and Sup Pool Cooling 3.9 i

!

\

>

a Z.__PgggEggBEg_;_NggdOL t_8gNQBM8(t_EdEggENGy_8NQ PAGE 33 809196991GOL_ggNIBgL

,

ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-KOLONAUSKI, ANSWER 7.01 (2.50) Reduce recirc pump speeds to minimu (3 required, 0.5 each)

Shift Unit Auxiliary Busses to Startup Bu Start Main L-O Suction and Turning Gear Oil Pump . It seals in the scram signal by sending an additional scram (0.5)

signal to RP . It changes the plant mode so that the MSIVs remain open if (0.5)

MSL pressure drops to 861 psig. [This maintains the steam supplies to equipment important for plant cooldown and shutdown (the FW pumps, the SJAEs, Offgas, and turbine seals) and keeps the Main Condenser available as a heat sink.]

REFERENCE EOP 100-101 Lesson Plan, page 2 295006 Scram (Group I EA Evol)

AA1.04 AB to operate / monitor Recirc related to scram AA1.01 AB to monitor RPS- Scram AK2.04 KN of turbine trip logic related to scram ANSWER 7.02 (1.75) . The reactor is scramme (0.5) The inboard and outboard MSIVs and MSL drains are isolate (0.5)

l Manually close RFP Discharge Isolation Valves HV-10603A- (0.75)

l l REFERENCE EO-100-OO9 i

295016 Control Room Abandonment (Group I EA Evol)

AK2.02 KN of local control stations 4.1 l

l

-

..

Z __PggCEDUggg_;_gggd@6t_@BNggd66t_EdEggENCy_8NQ PAGE 34

,

R_ AD_I O_ L_O_ G_ _I C_ A_ L_ _C_O_N__T R_ OL_

__ _

ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-KOLONAUSKI, Wt ANSWER 7.03 (3.00) b~ NIM N W" . a - establish natural circulation. M pKttd hhtd $tyRh6CodY(16.5)

b - ensure maximum heat removal . 3 pd CtrC Glsc gg (0.5)

c prevent thermal stratificatio # yiar(WA)r cwt +WtmMaved ) "' ' - ~ 0. 5 ) f ~ 4- ( RWCU return to vessel Condensate (3 required, 0.25 each)

CRD Core Spray Condensate Transfer LPCI c. RWCU suction from vessel (3 required, 0.25 each)

RWCU letdown Skimmer Surge Tank Letdown Fuel Pool Cooling SRV Blowdown REFERENCE SSES ON 149-001 295021 Loss of SDC (RHR) (Group II EA Evol)

KN of reasons for following responses: AK3.01 Raise Rx Water Level AK3.04 Max RWCU flow AK3.OB Alt Heat Removal KN of oper implic of SDC concepts: AKl.02 Thermal Stratification AK1.03 Adeq Core Cooling AK1.04 Natural Circulation 3.7 l

LE3AQfi ANSWER 7.04 (2.00) gey,get a3 g, ye , e MWw, M - WO -W l c. Primary Containment Control - Sup Pool Temp > 90 deg F (0.4)

i b. RPV Control and PC Control - DW pressure > 1.72 psig (0.4)

c. RPV Control - 7 X NFPB- MSIV isolation signal (0.4)

d. RPV Control - >1037 psig (0.4)

I a. Secondary Containment Control -

Zone III > 2.5 mrem /hr (0.4)

REFERENCE PPOO2 Specific Objective 1, page 3 l

223001 Primary Containment (Group I System)

SG #15 AB to recognize EDP entry conditions Secondary Containment (Group I System)

SG #15 AB to recognize EDP entry conditions 4.1

,

.

!

,

. _ _ . . . _ _ _ _ _,. . _ _ _

__r - _ . . - , . ,

_ . _ _ _ _ _ _ , _ . _ _ _ . , _ . . _ _ .

..

Z __P8QQEQUBES_;_Ng8BOLx_9BBgBMBLx_EBE8GENgy_@ND PAGE 35 BODI96991GOL_ggNIBg6

.

ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-KOLONAUSKI, ANSWER 7.05 (3.00) . placing the drywell under vacuum, or (0.5) may cause long path recirculation to leak into vesse (0,5) The reactor is considered critical when a slightly positive stable ( 1. 0 )

period is achieved with increasing neutron flux and no rod motio It is possible to inadvertantly "unbypass" the turbine stop and (1.0)

control valve closure scrams if first stage pressure is allowed to increase to Jd*/. of rate REFERENCE G0-100-002 292000 Reactor Operational Physics (Group I Rx Theory)

Kl.07 Define criticality as related to a Rx startup. Main Turbine (Group II Systems)

K1.04 KN of cause/effect relations between RPS and Mn Turb ANSWER 7.06 (2.25)

g o.ty " Max Normal" = the Tech Spec isolation setpoints of the Leakage (0.75)

Detection Syste " Max Safe" = maximum environmental temperature at which (0.75)

continued operability of affected equipment is no longer assure " Max Normal" = ten times the Area Rad Monitor alarm setpoin (0.75)

i

'

REFERENCE EO-LOO-104 Bases / PPOO2 Objective 3 290001 Secondary Containment (Group I System)

SG #10 AB to explain all system limits 3.4 i

<

'

.

'Za__Pggggggggg_;_NgBd@61_9BNggd@61_EDEBgENgy_8Ng PAGE 36

,

R___ A D _I O_ L O_ G_

_ __ _I C_ A L _ C___ _O N _T R_ O L

_

ANSWERS - SUSQUEHANNA 1&2 -87/03/16-KOLONAUSKI, f

, ANSWER 7.07 (1.50) . Hydrogen / oxygen explosion hazard (0.5) Unprocessed radiolytic gas discharges (0,5) To prevent drawing cold air across the turbine seal (0.5)

(to prevent bending of the shaft)

REFERENCE SSES OP-143-OO1, Rev 7 271000 Offgas (Group II Systems)

K1.06 KN of cause/effect relation between OG and Mn Steam Plant Wide Generics K1.15 KN of safety procedures related to Hydrogen ANSWER 7.08 (2.50)

c. Demineralized Water Storage -

(3 required, 0.33 each)

Refueling Water Storage Tank RHR Service Water Emergency Service Water Fire Protection System (1) - 25 hours2.893519e-4 days 0.00694 hours 4.133598e-5 weeks 9.5125e-6 months , (2) - 22 feet (0.25 each) Because the airborne radioactivity will increase with boilin (1.0)

REFERENCE ON-135-OO1, SYO17 L-2 Objective 5 l

233000 Fuel Pool Cooling (Group III System)

K3.06 KN of effect on area rad levels if FPC lost K4.01 KN of FPC design feature which maintain adequate level SG #5 KN of limiting conditions and safety limits 3.4

.

b

!

i

!

l

!

l

'

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

.

-22__PggCEDyBEg_;_NgBN@L2_9pN9BDOL 1_EgggggNCy_@NQ PAGE 37

,

RA__D_I

_ __ DL_OG I C AL___ CON __ TROL ___

_ _

ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-KOLONAUSKI, ANSWER 7.09 (1.50)

~ All tie bus lockouts reset (0.5)

2. Control switches in normal after trip position (0.5)

3. Unit Auxiliary Busses 12A and 12B not already being fed (0.5)

by SU Bus-2 REFERENC ON-OO3-OO2 262001 AC Elec Dist System (Group I System)

A3.03 AB to monitor automatic actions of AC Elec Dist-Load shedding A1.05 AB to monitor changes in AC Dist- Breaker lineups 3.5

,

j ANSWER 7.10 (3.00)

I- a.-From the HCTLll urve, if Rx pressure = 500 psig, g4 y/$/37 (0.5)

l SP temp -> deg Fg3

-

Delta T = 1 -170 = pc deg F [3 (0.5)

,

From the HCLL curve, if delta T = p6 deg F, G (0.5)

! ADS should be initiated prior to reaching JWi feet.

l- . If suction is taken from the Suppression Pool, the HPCI and (0.5)

! RCIC pumps will cavitat . Both system discharge spargers will uncover, and'will blow (0.5)

steam directly into the Suppression Chambe Containment pressure suppression function is lost with suppression (0.5)

pool level below 12 feet. (Containment failure could result if a primary system break occurs.) g g REFERENCE 5tpct atse- toJs of- NWtg - Cf /LVCl Fuction ,69 ECCS SSES EO-100-103 Bases ,

g ,g Ly 295030 Low Suppression Pool Water Level (EA Evol Group I) b"NA"I-EK1'.03 KN of Heat Capacity EK3.07 KN of NPSH for ECCS pumps EK2.08 KN of SRV discharge submergence .

i :

l $

l

!

l t-

-

Z1__Pggggguggg_;_Nggd@bt_@'_NQBD66t_EDE8GENCY @NQ PAGE 38

, RA_D_IO_L_O_G__ICA_L_C_O_N__TR_OL_

__ __ _

ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-KOLONAUSKI, ANSWER 7.11 (2.00)

a. Before suppression pool temperature reaches 110 degrees F if not (1.0)

shutdown with rod . If all control rods are inserted to Minimum Subcritical Banked (0.5)

Withdrawal Position (LE O2 with one control rod stuck out).

T~ ( dkk YCd,> in OR, SLC tank reaches 100 gallon (0.5)

REFERENCE OP-153-OO1 EO-100-102 211000 Standby Liquid Control (Group I System)

'A1.01 AB to monitor tank level changes K5.03 KN of oper implic of SDM C5 KN of LCOs and saf ety limits 4.4

!

>

l l

'

[

)

r l

,

_. _ _ _ . - , _

_ . _ _ _

&

.

~9 __89MINigIBOIlyE_P8QCEDUBE@t_CQNpillgN@t_8NQ_LIMlI@IlgN@ PAGE 39

ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-KOLONAUSKI, &gh -

SL = ANSWER B.01 (3.00) Oycm = . THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with (0.5)

the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flo OP CON 1 and 2 2. The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than (0.5)

1.06 with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flo OP CON 1, 2 3. The reactor coolant system pressure, as measured in the reactor (0.5)

vessel steam dome, shall not exceed 1325 psi OP CON 1, 2, 3, 4 4. The reactor vessel water level shall be above the top of the (0.5)

active irradiated fue OP CON 3, 4, 5

, 0. L Be in at least hot shutdown within two (2) hours (and comply with (1.0)

-

the requirements of TS 6.7.1.- Safety-Limit Violation).

REFERENCE SSES Tech Specs Su.Mem G.~1. High Reactor Pressure (Group I EA Evol)

EK1.05 KN of exceeding safety limits ANSWER B.02 (3.00)

t. S 1. ALERT Il-rCT 2. % AL EMERGENCY f.1 rC1 sire (K M (6/ti REFERENCE SSES EP-IP-OO1 294001 Plant Wide Generic A1.16 AB to take actions in the facility E Plan _ .__

.,

-

_ _ _ .

...

s___8DglNigIB@IlVE_PBQCEDUBE@t_CgNQlligN@z_@ND_LldlI@IlgNg PAGE 40

' ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-KOLONAUSKI, ANSWER- 8.03 (2.00)

a. TRU (0.5)

b. FALSE (EOF) (0.5)

c. TRUE (0.5)

d. FALSE (VP Nuclear) (0.5)

REFERENCE SSES E Plan 294001 PW Generic A1.16 AB to take actions as required by the E Plan ANSWER 8.04 (2.00)

a. Shift Supervisor SRO -1 Nuclear Plant Operator -5 (1.5)

Unit Supervisor SRO -2 Radwaste Operator -2 Assistant Unit Supervisor SRO (or RO) -1 Plant Control Operator RO -4 Auxiliary Systems Operator -1 b. the-Tech Specs limits (given in Table 6.2.2-1) for Minimum Shift (0.5)

Complement.are me REFERENCE SSES Tech Specs, DI-AD-015 294001 Plant Wide Generics A1.03 AB to use procedures related to shift staffing ANSWER 8.05 (1.50)

c. 2 (0.5 each) REFERENCE AD-QA-103 294001 Plant Wide Generics K1.02 Kn of tagging procedure . , . . . . - _ , _ . _ . ..

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

..,

9:___0951N1@l80IlyE_P89CEQUBE@z_CgNQlIlgNS z_8Ng_L1811011gN@ PAGE 41

L ANSWERS -- SUSQUEHANNA-1&2 -87/03/16-KOLONAUSKI, . ANSWER' 8.06 (3.00)

e. Per TS Table 3.3.1-1 (RPS Instrumentation), each RPS trip system (1.0)

requires two (2) operable APRM This requirement-is not me Take acti on] TS 3.3.1. a3 > [ place the inop channel or trip system in the tripped condition within one (1) hou '~'C "C" 1... U u.: ir ippeh con Itzon.)

This surveillance will requir removing APRM "C" from service, (1.0)

so now RPS "A" will have 1 s than the required minimum number of operable channels per t p system because APRM "A" is bypasse TS 3.3.1 b applie Action 4 of TS Table 3. .1- must be taken-> Ebe in STARTUP withi six (6) hour __

i BUT- Because APRM "D" is opera e, note "a" of TS Table 3.1.1-1 (1.0)

allows you to place RPS Chan 1 "B" in the inop condition without without tripping RPS Channe "B" for up to two (2) hours in order to perform the requ' e surveillances.

i

!

gg MCMMON d.tfCrildrA N SuW4iMLL CAhhei N 215005 APRM (Group I System) d Oht , kCoA44 $I if N Y SG#11 AB to recognize TS entry conditions ggg g 4 %4 y hgym A.o.rcum wih acca Lt 3/31/97 (2.o; k Wr "c" ir A +nyp A cuhhce pu Tr 3. 3. t. &

i Wi* AfRH 'A bpmcl , a scuolley awsor a cw m Afrm "c" k ( M n- M Itrwm uAM occu * , k(AAA,4 jfV f "G" clou ook hbt h bt planJ M M b7y<d CMClih(/n ir-Ihear, (t is for% S c4 m surettmu cn "cums (W inr as A )

C Aho , AWtn "C" cw w Actad iny wM prhmi$ & surmiluu .

l H A surnibtt was not pr 6mCl , A h4 W - 6dA r r ne uJodd nd occ E 3 5 1 b kn yFlb - % W. 5. 3.1 - i Achcms l %d 4 w(g ,(du 4 (IAP A 6 heuro wim m rnu r<rMee uhtn 1 I

.

,

9___0951NigIB811VE_PBgCEpuSESt_CgNQlIlgN@t_@NQ_LidlIGIlgNg PAGE 42

ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-KOLONAUSKI, ANSWER 8.07 (3.00) YE From Table 3.3.1-1 (RPS Instrumentation), (0.5)

6- 2 MSL Rad Mon are required per trip syste I TS 3.3.1.at [With less than the minimum number of required operable channels per trip system (RPS B), place that trip system in the tripped condition within one hou From Table 3.3.2-1 (Isolation Instrumentation), (0.5)

1e, 3b- 2 MSL Rad Man are required per trip syste [ TS 3.3.2.bt [With less than the minimum number of required operable channels per trip system, place that trip system in the tripped condition within one hour.]

b.(TR 3._3.1.b4 place RPS B in the tripped condition within one hour (0.5)

and take action given in TS Table 3.3.1- #6-[5d[ ion $3 [Be in Startup with the MSIVs closed in six hours, (6 33)

or at least in Hot Shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .]

}TE 3.3.2.ch [ place at least one trip system in the tripped (0.5)

condition within one hour and take the-action given by TS Table 3.3.2- .33

1e-l Action 20:)EDe in Hot Shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and in cerS)

Cold Shutdown within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .]

0.33

3b-1 Action 21] EBe in at least Startup with the associated g>.95 )

isolation valves closed within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least Hot Shutdoun within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in Cold Shutdown within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .]

REFERENCE SSES TS 223002 PCIS (Group I Systems)

SG#11 AB to recognize TS entry conditions RPS (Group I System)

SG#11 AB to recognize TS entry conditions ,

8___8QulNi@lB9IlVE_PggCEQUBE@t_CQNQ111QN@t_8NQ_LidlI@IlgNS' PAGE- 43

. ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-KOLONAUSKI, ANSWER 8.08 (3.00) . Declare the rod inoperable in accordance with]TS 3.1.3.1.a/ (0.33)

Within one hour, verify:] action a.1 4)-- =hi 5 in NA, (0.33)

and disarm the associated DCV Fol l ow acti on[ 3.1. 3.1. a. 2} -- f i x rod within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in (0.33)

Hot Shutdown within next 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months UL NO utr$), you may main in Op Con 2 as long as TS 3.1.3.1.a. (2.0)

AND a.2 are met 0 ), but you cannot go to the RUN mode (i e. , 3r34/j 7 change Op Con) because of TS 3.0.4 (QE) . g- g CN h4A(,

REFERENCE Or?.- $ 5 - Sfark h m ( SSES TS ( (WI' cadA M T ej to WW )

M \ CdM 8 T 3 ~5.l. 3 .1. A . 1. f>

201001 CRDH (Group II Systems) 3, A SGW11 AB to recognize TS entry conditions II 3 ' 3 * 2 yd$

ad ni ANSWER B.09 (2.50)

a. TS 3.8.3.1.b.1.a) 1) 1D612 lost DC Division I (1.0)

c) 1) 1D652 lost DE. Division I l TS 3.8.3.1 Action b} one DC dist system not energized reenergize within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or HOT SD within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and COLD SD within the next 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months b.(TS3.8.3.1Actioneh480VACswingbusinop; (1.5)

declare LPCI loop inop-lTS3.5.1. (TS 3.8.3.1 Action a- 1 AC load group los No fully applicable TS Action statement--> TS 3. O. 3 )

REFERENCE SSES TS l 263000 Dnsite AC Power Distribution (Group II Systems)

SG#11 AB to recognize TS entry conditions 3.9

,

!

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

. .

8.__ egg 1NiglggllVE_PggCEQUBE@t_CgNgillgNgt_ggp_LigilgIlgNg PAGE 44

ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-KOLONAUSKI, ANSWER B.10 (2.00)

NO (0.5). A reactor startup in three days would represent a violation of the SSES Tech Spec TSl3.6.1.4] requires that two MSIV LCS be operable in Op Con 1-2- (0.75)

The action stateme t for TS 3.6.1.4 allows 30 days of continued operation, but Tech Spec l3.0./ does not allow entry into an Operational (0.75)

Condition while r61ying on an action statemen REFERENCE SSES TS 223001 Primary Containment (Group I Systems)

SG #11 AB to recognize TS entry conditions ANSWER B.11 (2.00)

e. NONE. [5 (N-18) = 135 rem; not exceeded.] (1.0)

[2000 mrem /qtr < 2500 mrem /qtr, < 3000 mrem /qtr3 YE [5 (N-18) = 85 rem; not exceede (1.0)

[400 < 1250 mrem /qtr by 10CFR203 SSES Admin limit of LE 300 mrem /qtr if current quarter exposure not documente REFERENCE SSES AD-OO-735 External Dosimetry Program 10 CFR 20 294001 PW Generic K1.03 KN of 10 CFR 20 and related rad con requirements 3.8

,

, - - . . - , - - - - . , - , , - _ . , ,,-----,.,-,.~-r--. .n..- , . ~ , - - - - . . . - - , , - - - - , - , . , ,

,. _

hi~rgd n ?e l7 f }.

.

~

U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION

,

FACILITY: SUSQUEHANNA 1&2 REACTOR TYPE: ~BWR-GE4 DATE ADMINISTERED: 87/03/16 EXAMINER: 3. K .' HAJEK CANDIDATE: M4 STER INSTRUCTIONS TO CANDIDATE:

Use separate paper for the answers. Write answers on one side onl Staple-question sheet on top of the answer sheet 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 start % OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 25.00 25.00 PRINCIPLES OF NUCLEAR-POWER PLANT OPERATION, THERMODYNAMICS,

'

HEAT' TRANSFER AND FLUID FLOW 25.00 25.00 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 25.00 25.00 INSTRUMENTS AND CONTROLS 25.00 25.00 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 100.00 % Totals Final Grade All work done on this examination is my ow I have neither given nor received ai Candidate's Signature

~

.

. - . . . _ - . . -

, . . ,

NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS

.

'uring the administration of this examination the_following rules apply: Cheating on.the examination means an automatic denial of your application and could result in more severe penaltie . Restroom. trips are to be Jimited and only one candidate at a time may leav You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.

I Use black ink or' dark pencil onIV to facilitate legible reproduction . Print your name in the blank provided on the cover sheet of the examinatio . Fill in the date on the cover sheet of the examination (if necessary). Use only the paper provided for answer . Print your name in the upper right-hand corner of the first page of each section of the answer shee . Consecutively number each answer sheet, write "End of Category __" as-appropriate, start each category on a new page, write oniv on one side of the paper, and write "Last Page" on the last answer shee . Number each answer as to category and number, for example, 1.4, , 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 tabl

l 12. Use abbreviations only if they are commonly used in facility literatur i 13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.

!

'

14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or no t l 15. Partial credit may be give Therefore, ANSWER ALL PARTS OF THE l

QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.

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

17. You must sign the statement on the cover sheet that indicates that the I

work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.

!

.

I r'

-

i

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

... ,

. 18.'When you complete your examination, you shall: Assemble your examination as follows:

(1) Exam questions on to (2) Exam aids - figures, tables, et (3) Answer pages including figures which are part of the answe Turn in your copy of the examination and all pages used to answer the examination question ' Turn in all scrap paper and the balance of the paper that you did not use for answering the questions, Leave the examination area, as defined-by the examine If after

1eaving, you are found in this area while the examination is still in progress, your license may be denied or revoke .

. _ _ .__ _ _ __-

1. PRINCIPLUS OF NUCLEAR POWER PLANT OPERATION, PAGE 2

~

THERMODYNAMICS, HEAT TRANSF2R AND FLUID FLOW

.

QUESTION 1.01 (2.25)

With the reactor operating at full power, state which reactivity coefficient will cause the initial or greatest power change for each of the following malfunction A fault occurs in the EHC system that causes reactor pressure to oscillat (0.75) The recirculation system master flow controller fails such that recirc-flow demand decreases slowl (0.75) Extraction steam is blocked to a high pressure feedwater heate (0.75)

QUESTION 1.02 (2.00)

In the case of a loss of the shutdown cooling system, the Off Normal procedure, ON-149-001, Loss of RHR (Shutdown Cooling) System, recommends several operator actions that should be take How will adhering to each of the following recommendations assure effective core cooling 7 Restart at least one reactor recirculation pump OR operate Reactor Water Cleanup at its maximum rate. (1.0) Maintain vessel water level between 90 and 100 inches on shutdown indication instrumentatio (1.0)

QUESTION 1.03 (2.25)

Explain why changes in the following three plant parameters or operating conditions cause reactivity effects that result in more severe pressure transients at the end of core life than at the beginning, thus requiring the EOC RPT protection functio Void coefficient of reactivity (0.75) Delayed neutron fraction (0.75) Control rod density (0.75)

.

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

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

-1 PRINCIPLES OF NUCLEAR' POWER PLANT-OPERATION, PAGE '3

- THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

..

~

QUESTION 1.04 (2.00)

During an ATHS,. Emergency Operating Procedure EO-100-113 recommends lowering reactor level. Briefly explain HOW

,

and WHY this affects J Reactor power.- ( 1. 0 ) ' SLC boron' concentration in the active fuel region. (1.0)

'

QUESTION. 1.05 (1.50)

3 u With the reactor operating at 100 percent power, and the

,

S a"p<.Fy)* 'enthalpy of main steam at 1205 BTU /lbm, what tail pipe

' 's temperature would you expect to see in order to verify p>dg that an SRV had cracked open and was continuously .

g #j**t' leaking? . Assume a.ta11 pipe pressure of 75 psia. ShowJs W41Me 5 all your work, and state all your assumption (1.5)

I

! QUESTION 1.06 (2.00)

Concerning a potential rod drop accident,

{ Why is a positive reactivity excursion due to a dropped rod more severe during a startup than when at power? Give two reason (1.0)

Give two observations or methods that are required

'

i in the' process of withdrawing control rods to minimize the chance and severity of a rod drop acciden (1.0)

QUESTION 1.07 (2.00)

i Emergency operating Procedure, EO-100-112, Rapid l Depressurization, provides for overriding ECCS pump initiations if more than one control rod is out further than notch 02. What are the two reasons and the associated adverse consequences for this provision that

could result if ECCS injection was permitted? (2.0)

i

-

L .

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

.

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

. , PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 4 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

.

QUESTION 1.08 (2.50)

For the changes in reactor operating conditions listed below, indicate whether NPSHA ( Available NPSH) for the reactor recirculation pumps INCREASES, DECREASES, or REMAINS THE SAM Control rods are withdrawn from full in until reactor power reaches mid scale on Range 3 of the IRM (0,5) Reactor power is maintained at mid scale on Range 6 of the IRMs, and the vessel heats up until the head vents need to be close (0.5) Reactor power 10 maintained at mid scale on Range 6 of the IRMs, and reactor pressure increases from 100 psig to 920 psi (0.5) Reactor power is increased from 10 percent to 50 percent by withdrawing control rods. Recirculation pumps are operated at minimum spee (0.5) Reactor power is increased from 70 percent to 100 percent by changing recirc pump speed. Control rods are maintained at 100 percent rod patter (0,5)

QUESTION 1.09 (2.50) O b#

s.A+

,#' 8 IE* A failure occurs in th Feedwater/ Level Control System

$M gy *"" which causes the level setpoint to instantaneously s, ' change from 37 inches to 31 inches while operating in

[,,, ,a three element contro Explain the initial changes in s# j-j eh' reactor power that result from this faul Include

.b gg'* ed considerations of changes in flow and the specific gd g # reactivity coefficients involve (2.5)

u

{8'

o

.

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

.

-

..' 1 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW PAGE 5 ,

l

'

!

QUESTION 1.10 (3.00)

State whether each of the following statements related to adequate core cooling is either TRUE or FALSE. If a statement is FALSE, tell why it is false, Adequate core cooling is assured if the core is covered with liqui (0.75) Adequate core cooling is assured if the core is covered with a two phase mixture, the larger percentage of which.is stea (0.75) If all level indication has been lost, operation of HPCI is sufficient to assure adequate core cooling.(0.75) If the reactor water level reaches the ADS initiation setpoint, and the RPV temperature is increasing, Blowdown Cooling must be initiated to assure adequate core coolin (0.75)

QUESTION 1.11 (3.00)

With regard to Xenon-135 in the core, Explain why the xenon level initially increases after a power decreases, and initially decreases after a power increas (2.0) Why are the notch worths of control rods drastically altered by Xenon-135 for a startup ten hours after a reactor scram? (1,0)

~

.

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

. ' . PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 6 e

QUESTION 2.01 (3.00)

For each of the following ADS manual or automatic initiation conditions, indicate whether ADS will initiate, when, and why or.why no Manual initiation attempted, one Core Spray pump runnin (0.6) ADS 102 see timer timed ou No low pressure pumps runnin (0.6) ADS A/B Logic Control Switch in Inhibit positio Then, all automatic initiation requirements become satisfie (0.6) ADS A/B Logic Control Switch in Inhibit positio Manual initiation attempte (0.6) Level 1 and 3 signals receive Hi D/W pressure signal not received. Required low pressure pump (s)

running, and three minutes elapsed since receipt of low level signal (0.6)

QUESTION 2.02 (2.50)

When operating RHR in the Shutdown Cooling mode, Why is it required that a flow rate of at least 4000 gpm be established within lo seconds after s arting the pump, and what adverse condition could occur if this was not accomplished? (1.0) How is the rate of flow to the vessel controlled? (0.5) Why must the flow rate through the heat exchanger be limited to less than 10,000 gpm? (0.5) How is the cooldown rate controlled? (0.5)

)

~

.

(***** CATEGORY 02 CONTINUED CN NEXT PAGE *****)

,,- 2 , PLANT DESIGN' INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 7 c,

QUESTION 2.03 (2.00)

For the Core Spray System, explain how a leak is detected in the piping between the inside of the RPV and the core shroud. Include a description of how and why the monitored parameter will change if a break occurs in this pipin (2.0)

QUESTION 2.04 (2.00)

J O' 714f What are the normal and alternate supplies for 4160 KV 3' ESS Bus IA, AND in what order will power be supplied to it SQ. 9 (2.0)

( $'g, Bus 1A if the normal source is lost? QUESTION 2.05 (2.50)

State whether each of the following statements regarding the Reactor Core Isolation Cooling (RCIC) System is either TRUE or FALS If the statement is FALSE, explain wh The turbine exhaust is directed to the barometric condenser where it is condensed and pumped to the Suppression Poo (0.5) Water from the Suppression Pool should only be injected into the vessel during automatic initiation or during an emergency conditio (0.5) The RCIC System is capable of remote manual startup, operation, and shutdown from the Remote Shutdown Pane (0.5) If 110 pounds of pressure cannot be maintained in the RCIC discharge piping, it might be necessary to isolate the HPCI system discharge piping keeptill syste (0.5) If the RPV water level reaches +54 inches and the F045 valve (Steam Admission valve) shuts, it cannot be reopened until the RPV water level reaches -30 inche (0.5)

I t .

(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)

!

i

r

., 2 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 8

.

QUESTION 2.06 (3.00)

The Reactor Water Cleanup System (RWCU) is used for letdown during a reactor startup, During'a startup, to what two locations can letdown be routed, AND under what conditions is each location used? (1.0) How and why does the letdown rate limit the plant heatup rate? (2.0)

QUESTION 2.07 (2.50)

For the Primary Containment Isolation System (PCIS),

O ta /S

_po List five of the 1FystSES that will isolate from an 1_",7pyp N4S si al if a L A should occur when the reactor

"JO is ope ing at 40 percent powe (1.5) RHR also receives isolation signals from N4S, but should not isolate under the conditions listed in l Part Why is this, AND under what conditions will the RHR isolation be activ (1.0)

(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)

!

l

'

.

- _ _ _ _-. _

e 22 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAG 3 9

.

QUESTION 2.08 (1.00)

For each of the following conditions, match the correct statement for Recirculation System pump restart limitations, Starting a recirc pump from AMBIENT condition (0.5) Starting a recire pump from HOT condition (0.5) A maximun of ONE start is allowed for the recirc pump. After a 45 minute period has elapsed, another restart attempt can be mad . A maximum of ONE start is allowed for the recire pump. After a 25 minute period has elapsed, another restart attempt can be mad . A maximum of TWO starta is allowed for the recirc pump. After a 25 minute period has elapsed, another restart attempt can be mad . A maximum of TWO starts is allowed for the recirc pump. After a 45 minute period has elapsed, another restart attempt can be mad QUESTION 2.09 (3.00)

What are the problems associated with operating HPCI under each of the following conditions 7 Using the CST as suction if auto transfer to the suppression pool fails on low CST leve (0.75) Suppression Pool level is above 23 ft. 9 in., and suction is being taken from the CS (0.75) Suppression Pool level is below 18.5 ft., and suction is being taken from the CS (0.75) HPCI turbine speed is below 2150 rp (0.75)

b b* N

+ aL%,

kJ ,

.

..

,

(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)

_ _ _ _ - _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - - _ - _ _ - _ _ _ _ _ _ _ - _ _ _ - _ _ - _ _-_ --_--_ _ ._ - ___- ___-.

. , 22. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS -PAGE 10 i

.-

LQUESTION 2.10 '(1.00)

au f.-w Manual start of AmHdt RHR pumps can be accomplished by depressing only one of the Manual Initiation pushbuttons. Why then is it necessary to depress both pushbuttons when manually initiating RER? (1.0)

' QUESTION.R2.11 (2.50)

For an EMERGENCY START of the Emergency Diesel Generators, which of the following will cause an engine trip? (2.5) Low turbocharger oil pressure Low' lube oil pressure Main bearing high temperature High jacket water temperature l Overspeed i Generator overvoltage Generator overexcitation , High Generator Differential Carrent ) Reverse power Incomplete start sequence i

.

(***** END OF CATEGORY 02 *****)

- - - - - - - . - - _

- . ,

e , 3 INSTRUMENTS AND CONTROLS PAGE 11 o

QUESTION 3.01 (2.00)

What methods, are available to insert control rods if they fail to insert on a scram signal, and if normal insertion actions fall? Consider the actions recommended in E0P-100-102, RPV Contro Give fou (2.0)

QUESTION 3.02 (3.00)

For each of the following malfunctions or changes in plant parameters, state whether the Narrow Range Reactor Level instrumentation will indicate ACTUAL, HIGHER THAN ACTUAL, OR LOWER THAN ACTUAL LEVE A leak in the reference leg occurs that allows a continuous flow of water through the le (0.5) Temperature in the Drywell during a LOCA increases and remains at 100 degrees F above norma (0.5) A steam leak occurs outside the Drywell, and one Main Steam line fails to isolat The reactor water level is maintained by HPC (0.5) A leak occurs across the instrument equalizing valv (0.5) A rapid vessel depressurization occurs that causes an elevated Drywell temperatur (0.5) An instrument technician has mistakenly calibrated the Narrow Range Level instrumentation instead of the Wide Range Level instrumentation while the reactor is operating at 75 percent powe (0.5)

<

QUESTION 3.03 (1.00)

Calibration of the APRMs is checked after an LPRM is bypassed. Since count and averaging circuits automatically adjust the output for the bypassed LPRM, why is it necessary to check the calibration after the LPRM is bypassed? (1.0)

!

-

.

t

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

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

o , INSTRUMENTS AND CONTROLS PAGE 12 O

QUESTION 3.04 (3.00) Under what two conditions will the Rod Position and Information System detect a control rod drift? (1.0) How would you determine which control rod is drifting? (1.0) Give two conditions that could cause a rod to drif j, (1.0)

yl$'p { 5 )

pdt 4a:;, 4 ,s ,'/,,g,p b

,.2 ewy reA QUESTION 3,05 (3.00)

.

,

~k3 O W. lp*

The plant is operating at 60 percent power, and the-bdak feedwater flow signal to the recirculation flow control system fails to zero. Consider each of the following

questions independentl Se sure to consider the effects that will occur in BOTH UNIT How will the speed of BOTH recire pumps be affected? Why? (1.0) What will happen to the speed of SOTH recirc pumps if the signal failure clears in about 30 seconds before any operator action is taken? Why? (1.0) How will the speed of BOTH recirc pumps be affected if the speed controller output on A MG set fails just prior to the loss of the feedwater signal?

Why? (1.0)

QUESTION 3.06 (2.00)

Concerning the IRM system, What four signals will cause a reactor scram? (1.0)

o dm1*e, Under what conditions is this scram function 4 bypassed? (1.0)

,

)

.

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

. . . . . .

_ _ _ _ _ _

o , INSTRUMENTS AND CONTROLS PAGE 13 o

QUESTION 3.07 (1.00)

Reactor Feedpump turbine speed is controlled with two controllers, the motor speed changer (MSC) and the electric automatic positioner (EAP) or Auto Speed Controller, Which controller will be in control? (0.5) Where must the MSC be positioned for the EAP to have control over the full range of turbine speed? (0.5)

QUESTION 3.08 (3.00)

List WHEN and HOW (automatically or manually) each of the following Reactor Cerams are bypassed, MSIV Closure (0.75) Scram Discharge Volumn:Hi Hi (0.75) Turbine Control Valve Fast Closure (0.75) Mode Switch in shutdown (0.75)

QUESTION 3.09 (3.00)

Using the attached diagram of the EHC Pressure Control Unit (Figure 5), for reference, Explain the purpose of the zero and three psi biase (1.0) Explain what will happen to control valve position 9ah g,g should a failure occur that causes the output of 1Y g ~~

t h e f g oliiir n'i n g i e gli1~a t'6 r t o F A I L L O W .

-

(1.0)

s_ _ .

g.D,.,pv 3,J Explain what will happen to control valve position should a failure occur that causes the output of-QA yh t sdD-k

-

.

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

=,- 3. INSTRUMENTS AND CONTROLS PAGE 14 s

QUESTION 3.10 (3.00)

With the plant operating at 75 percent power, consider each of the following channel / component operating situations for Rod Block Monitor Channel APRM A has been bypassed due to maintenance on the amplifier that has caused the unit to be taken out of Operate. APRM C has just failed upscale and caused a half scram. Relative to operation of the RSM, and to assure the safest completion of any necessary rod movement, would it be preferable to have APRM A or C bypassed, and to take the trip with the other APRM7 EXPLAI (1.0) An LPRM input to the selection matrix has failed hig The LPRM is also an input to APRM D. How will bypassing or not bypassing of this LPRM in the APRM cabinet affect operation of the Rod Block Monitor? EXPLAI (2.0)

QUESTION 3.11 (1.00)

~

What are four AUTOMATIC ACTIONS (not alarms) that occur on receipt of a Refuel Floor High Exhaust Hi Hi Radiation signal? (1.0)

.

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

o PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 15 RADIOLOGICAL CONTROL

.

QUESTION 4.01 (2.00)

State whether each of the following statements is either TRUE or FALSE according to AD-QA-300, Conduct of Operation If the statement is FALSE, state why it is FALS Only licensed operators are permitted to manipulate controls that directly affect reactivity or power level of the reacto (0,5) In the Off-Normal Procedures, the Operator Actions are written in a logical sequence to provide guidance, and are to be performed as necessary in the sequence as directed in the procedur (0.5) ECCS actuation may be inhibited or overridden if specifically directed to do so in a Symptom-oriented Emergency Operating Procedur (0,5) When a nonradioactive oyotem to or becomeo contaminated, further use of the system may proceed no soon no shift Supervision provides for continuous radiation monitoring of the syste (0,5)

QUESTION 4.02 (3.00)

For each of the following conditions, indicate whether or not EMERGENCY CPERATING PROCEDURE entry is require If entry is required, state which procedure (s) to ente If entry la not required, state "None." Consider each sub part no a separate ite Aonume no additional abnormal conditions are present for each individual ite RPV level la 10 inche (0.25) Reactor power la 12 percent, Startup mod (0.25) Reactor power 10 93 percent seven minutes after a load rejec ggpv (0.25) Power operationo, Orcup I-loolation occur (0.25) Supprecolon Pool level is 23.3 fee (0.25) Drywell preocure la 2.5 poi (0.25) CRD-HCU North Area Monitor indicating 130 MR/H (0.25) Supprecolon Pool temperature is 95 degrees (0.25) Reactor shutdown, RPV pressure is 1090 psi (0.25) Drywell temperature 10 160 degreco (0.25) Zone III HVAC Exhaust Radiation Level is 4.5 MR/HR.(0.25) Suppression Pool level is 24.7 fee (0.25)

.

'

(***** CATEGORY C4 CONTINUED ON NEXT PAGE *****)

_ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _-

. , , PROCEDURES - NORMAL, ABNORMAL, EME9CENCY AND PAGI RADIOLOGICAL CONTROL
l l
QUESTION 4.03 (3.00) l While taking hourly instrument readings, you note that the condenser is slowly losing vacuu f According to ON-143-002, Loss of Main Condenser I Vacuum, which of the following possible symptoms j should you check to either confirm that vacuum is decreasing, or to identify the cause of the vacuum
'
loss? (2.0)
, Off gas isolation valve closure Generator output decreasing Circ water system pump trip and alarm Reactor pressure decreasing Recombiner isolated Low RBCCH discharge pressure SJAE inlet stea?. pressure decreasing Cooling Tower basin level low Steam seal regulator malfunction 2 Reactor water level high Explain why reducing reactor po.ar can slow the decrease in the rate of loss of condenser vacuu (1.0) l i l QUESTION 4.04 (3.00)
According to AD-QA-300, Conduct of Operations, .
Attachment E, Immediate Operator Action List, what are I the immediate operator actions to be taken if the Main Steam Line radiation levels are increasing, but the trip I point has not yet been reached? Include all alternative actions until the reactor is in a safe isolated conditio (3.0)
l t
i
, 1 i
'
.
)
i i
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
,
'
l
.
.-. . ._ - - - . _ - . _ . - - - - . - - - _ _ - - . _ - . . . - . - - - . - . - .
_ _ _ _ _ - _ _
. , PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 17 RADIOLOGICAL CONTROL
QUESTION 4.05 (2.50)
According to the Precautions in OP-100-002, Remote Shutdown - Normal Plant Operating Lineup, What adverse operating and monitoring conditions may result if a transfer switch at the Remote Shutdown panel is placed in the EMERG position while the plant is operating? (2.0) What may occur if a handswitch position at the Remote Shutdown Panel is changed while its associated transfer switch is in the NORM position? (0.5)
QUESTION 4.06 (3.00)
According to the Precautions and Cautions in GO-100-002, Plant Startup and Heatup, Where should the " Pressure Set" and " Bypass Jack" be positioned before condenser vacuum reaches 7" HgV (22.2" HgA)? (1.0) What two adverse effects could occur if the MSIVs were opened with condenser vacuum established and the head vent open? (1.0) How is SRM/IRM overlap established? (0.5) HPCI and RCIC must be operable prior to exceeding what RPV pressure? (0.5)
)
_. - ,
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
,
...,- PROCEDURES -' NORMAL. ABNORMAL. EMERGENCY AND PAGE 18 RADIOLOGICAL CONTROL
< ,
,
QUESTION 4.07 (3.00)
According to ON-118-001, Loss of Instrument Air, how do each of the following components respond (FAIL OPEN or FAIL CLOSED) to low Instrument Air header pressure? Outboard MSIVs (0.5)
i' Feedwater heater normal drain valves (0.5) Service Water outlet throttle valves from RSCCW and TBCCW heat exchangers (0.5) Emergency Service Water supply and return valves for RBCCW and TBCCW heat exchangers (0,5) Drain and vent valves for the Scram Discharge Volume (0.5) Condensor reject and makeup control valves (0.5)
QUESTION 4.08 (1.50)
According to AD-00-735, External Dosimetry Program, What is the station administrative quarterly whole body exposure limit for a Radiation Worker with an up to date NRC Form 4 on file? (0.5) If you are working in a radiation field of 200 mR/hr, and your Self Reading Dosimeter has a full scale range of 500 mR, when do you need to return to Health Physics to have it rezeroed? (1.0)
QUESTION 4.09 (2.00)
'
According to ON-155-001, Stuck Control Rod, What are two indications of a stuck rod that could be observed while attempting to move rods? (1.0) Which of the following conditions could be causing the rod to stick? (1.0) Drive water pressure at 220 paid Scram Discharge Volume Not Drained Alarm Drive water flow at 2 gpm while attempting to insert the rod one notch HCU valve alignments
>
-+
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
,
g # PROCEDURES - NORMAL,' ABNORMAL, EMERGENCY AND PAGE 19
~

RADIOLOGICAL CONTROL
'
.
!
l
' QUESTION 4.20 .(2.00)
'
Describe how each of the following parameters will change if a Safety Relief valve inadvertently opens and remains open, according to ON-183-001, Stuck Open Safety / Relief Valv Indicated feedwa.ter flow relative to indicated steam flow- (0.5) Generator Load (0.5) Reactor power (0,5) Suppression' Pool level (0.5)
l l
l l
l
!
i I-
.
(***** END OF CATEGORY 04 *****)
(************* END OF EXAMINATION ***************)
<
- ., , , . , . - - - _ , . - - - . , ~
-
,
- - . - - . . . . - - e._- , ,,. , -, ,,. ~,
'sy@l?/ A-W
- . .
O J
Owe E>E
>=.Jat EeM e>w U G
.
l
,
.

a M e w*
Se h
U
>
N w .
Z
. o

a> . =3 m *M N $ D
, - .

w n o b E ihO M De
- -
W&
. E z
- =
= =3 c.m u e
-
I i E
'
'
up - M
'
M W M l w c z >= W as -- cz; b MM L
-- c 'l3 .
b
___ u _
_ > _
@a - b _ . , ca ss NM
_
M-b-
+~
\
+' 4 U r +1
+ 8 N N =
= N N
+d d +
- = = L-
-
han W9 g E M EWW w-=
--M .
- = = ===
,_ M NO MOM

- ==
-s m
~
=w
=
_
G l
,
- - - - - . . . . . . . _ .
,
, . . . _ . . . _ . . . _ _ . _ . . . . _ _ . _ _ _ . _ . . . . _ . _ . . _ _ _ . . , _ _ _ .__ ..._. - - _ . _ _ __
, PRINCIPLES OF NUCLEAR POWER PLAp?T OPERATION, PAGE 20 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW e ANSWERS -- SUSQUEEANNA 1&2 -87/03/16-3. K. HAJEK ANSWER 1.01 (2.25) void Void Moderator temperature REFERENCE SCO23 A-7, ppg. 12 - 1 Specific Objectives 3, 13, 14, 1 K&A 292008, K1.08 (4.1/4.1), KI.20 (3.3/3.4), K1.22 (3.5/3.6)
ANSWER 1.02 (2.00) Prevents thermal stratification. [RWcu bgi4ul( Alco amer ++ h N Promotes natural circulatio % % m-e.Ly Hx- w h-h, gg 4 g REFERENCE SCOO6 K (Mitigation of Core Damage - Natural Circulation), pg. Specific Unit Objective K&As 293008, K1.37 (3.2/3.4)
ANSWER 1.03 (2.25) The void coef. is greater at EOC than at BOL, and will add more reactivity should the voids collaps Beta is smaller at EOL, and the reactor will respond faster to positive reactivity insertion Control rods are further out of the core and must initially travel through a low flux region which will reduce the initial negative reactivity insertion rat REFERENCE SCO23 G-3 (Reactor Core Thermo:Th. Limits), ppg. 18 -
1 Specific Objective K&A 293009 K1.24 (2.7/3.2), 202001 K1.28 (3.9/4.1),
K4.13 (3.7/4.0), K5.05 (3.5/3.6).
..
,
- - -
.-.3-
,- ,
1 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 21 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW.
ANSWERS ---SUSQUEHANNA 1&2 -87/03/16-B. K. HAJEK
' ANSWER 1.04 (2.00) Power is reduced [0.5) due to a decrease in natural circulation which causes more voiding in the core. [0,5) SLC mixing is retarded [0.5) due to less natural circulation in the core and increased voiding which reduces the boron concentration. [0.5)
REFERENCE SC006 K (Mitigation of Core Damage - Natural Circulation), pg. Specific Unit Objective K&As 293008, K1.34 (2.9/3.1), K1.37 (3.2/3.4), 292008, K1.20 (3.3/3.4).
ANSWER 1.05 (1.50)
This is a throttling process, which is adiabati Therefore h1 = h [0.5)
The Superheat Stean Tables [0.5) can be used to find a temperature of 350 F, [0.5]
or The Moller Diagram (0.5) can be used to find the same temperature. [0.5)
No interpolation is necessary for either metho REFERENCE SCO23 D-8, ppg. 2 - CE Steam Tables Specific Objective K&As 218000, A1.01 (3.4/3.6), A3.02 (3.6/3.7), A4.06 (3.5/3.6), SG 15 (4.2,4.4), 239002, K4.06 (3.5/3.7),
K5.04 (3.3/3.5), et .'
.
,- - -
, -- - , - -
j.; ' PRINCIPLES OF NUCLEAR POWER' PLANT' OPERATION, PAGE 22 THERMODYNAMICS,-HEAT TRANSFER'AND FLUID FLOW I*l ~ANSWEFP~- SUSQUEHANNA.2&2 -87/03/16-B. K. HAJEK
.
ANSWER' .1.06 (2.00)
-a, 1.- Rods'are worth more at startu'p ~ There is no void feedback at startup
- . Assure proper rod sequencing ~by use of. pull sheets, RWM, or.RSC . Monitor core response during rod pul . Perform coupling check at position 4 [0.5] for.any two or other resonable response REFERENCE SCO23 A-6-(Rx Th-Rx Control), pg. 1 Specific Objectives.24, 25, 2 SC006 L (MCD).
Specific Unit objective 2, K&A.201003 K3.01 (3.2/3.4), K4.02 (3.8/3.9), A2.02 (3.7/3.8).
!
i
'
l ANSWER 1.07 (2.00)
.
1 .- To prevent a. cold water. injection [0.5) that could
'
-result in a pccitive reactivity insertion causing a power increase. [0.5]
\
~ To prevent boron dilu. tion (0.5] and preclude a power increase or fluctuation.- [0.5]
! REFERENCE EO-100-112, Bases, pg.3.
, PP002 Specific Unit Objective 14, 16, 19.
i
'SC006 L (MCD).
Specific Unit Objective 2, 3.
!- K&A 295037 EK1.03 (4.2/4.4), EK1.06 (4.0/4.2), EK2.04 (4.4/4.5).
l
'
i is
!
-
.
l l
. -
. -- . . - _ _
.- . . . - ...~ _
_ . . . - - - - - . , . - . - . - . _ . - . . . - . . - _ - -
, PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 23 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-B. K. HAJEK-ANSWER 1.08 (2.50) Remains the same, Decreases [due to decreasing water density) Decreases [due to decreasing water density] Increases [due to increased FW subcooling) Decreases [due to flow losses exceeding the increase in FW subcooling)
REFERENCE SCO23 E-4, ppg. 35 - 4 Specific Objectives 8, K&As 202001, K4.02 (3.1/3.2), K6.07 (3.3/3.3), A2.11 (3.7/3.9), A2.12 (3.6/3.8), 291004 K1.06 (3.3/3.3)
ANSWER 1.09 (2.50)
t*t When the level setpoint is lowered, feed flow will decrease. [0.5] This will reduce subcooling, and cause warmer water to enter the core. [0.5] Power will decrease [0.5) due to the moderator temperature coefficient [0.5], and due to increased voiding (void coef). [0.5)
REFERENCE SC007B (Plant Reactivity Evolutions), Specific Objective K&A 259002 A1.03 (3.8/3.8)
ANSWER 1.10 (3.00) True True Fals Low pressure ECCS is required to assure sufficient volum False. Wait until reaching TAF because as long as water covers the core, adequate cooling is availabl REFERENCE SC006 L (Review of Systems and Procedures used for Mitigating Core Damage), ppg. 2, 3, 1 Specific Objective .
. _ _ _ _
. , .
1.,-PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE. 24
.TFERM0 DYNAMICS, HEAT TRANSFER'AND FLUID FLOW L 6, ' ANSWERS -- SUSQUEHANNA'I&2 -87/03/16-B. K. HAJEK K&A 295031 EK1.01 (4.6/4.7), EA2.04 (4.6/4.8)
ANSWER 1.1 (3.00) Xenon is produced primarily from the decay of Iodine-135. [0.5) It is removed primarily from burnup. [0.5]. On a power decrease, it is initially still.being produced at the higher power production rate, while the burnup rate has decreased. [0.5) The opposite is true for a power
. increase. [0.5)- Xenon will have built up or peaked in regions of the core that previously had high flux, and will b .
much lower in regions that previously had low flu [ 0 .' 5 ] This will result in a shift in flux during the startup condition to the previously low flux regions, and a corresponding increase in the reactivity of the control rods in those location (0.5]
REFERENCE SCO23 A-8 ~(Rx Th/Xe and Sm), ppg. 5- Specific objectives 1, 2, 4, 5, K&As 292006 K1.02 (3.1/3.1), K1.03 (2.9/2.9), K1.04 (2.9/2.9), K1.06 (2.7/2.7), K1.07 (3.2/3.2), K1.08 (2.8/3.2).
'
.
_ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ . . . _
ll
, PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 25 l ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-B. K. HAJEK e
ANSWER 2.01 (3.00) ADS will not start. [0.3) Both Core Spray pumps must be running (or at least one LPCI pump]. [0.3) Nothing happens [0.3] until pump requirement is met. [0.3) ADS does not initiate [0.3) because Inhibit prevents satisfying the [drywell and level)
permissives. [0.3] ADS initiates provided the pump permissive is satisfied. [0.3] Manual start overrides the other permissives. [0.3) ADS initiates [0.3] after seven minutes to allow initiation for a leak outside the D/W. [0.3)
REFERENCE SYO17 C-4 (ADS), pg. 7 and Table Specific Unit Objectives 1, 3, 6, K&As 218000 K4.02 (3.8/4.0), K4.03 (3.8/4.0), K5.01 (3.8/3.8), K6.01 (3.9/4.1), K6.02 (4.1/4.1).
ANSWER 2.02 (2.50) To keep the min flow valve (FQj7) from opening
[0.5) and draining the vessel to the suppression poo3. [0.5] By throttling the F017 valve [ located upstream of the F015 injection valve).
10.5 ) To preclude vibrations in the heat exchange (0/5) By throttling the heat exchanger bypass valve ,
(OS9 (F048) , odl4k Mn, rs, et,HetSW . Q4 % f "3 gar Valve numbers OR descriptions are acceptable. Both are not required, buts deduction will be made if wrong numbers are given or incorrect descriptions are give REFERENCE YO17 C-1, Attachment D and Figures 3 and '
pecific Unit Objectives 3, 4, 5, 8, 9, 1 K&As 205000 K1.02 (3.6/3.6), K3.02 (3.2/3.3), K4.05
~ 08-i#9 001 dg 17

'
!
~
.
-, - ,
w
, . 2a bLANT DESIGN INCLUDING SAFETY AND EMERGE!!CY SYSTEMS PAGE 26 ANSWERS -- SUSQUEEANNA 1&2 -87/03/16-B. K. HAJEK
.
(3.6/3.7).
ANSWER 2.03 (2.00)
A differential pressure switch measures the dP between the bottom of the core plate (SLC above core plate tap)
[0.5] and the inside of the Core Spray sparger pipe just inside the RPV shroud. [0.5)
-
-If a break occurs, the dP will increase-[0.5] because the pressure drop across the steam separator will be included. [0.5] - - . _ .
'
[ Normal dP is -3.5 psi Alarm at. Ave psidJ2
.
, p;f,,,m y g g g o,$-
"d ,~( _
REFERENCE SYO17 C-2 (Core Spray), ppg. 15 - 1 Specific Unit Objective K&As 209001 K4.04 (3.0/3.2)
ANSWER 2.04 (2.00)
Normal source is ESS101 Transformer OX201 from S/U Bus 10 OA103. [0.5)
First alternate (0.25] is ESS 201 Transformer OX203 from S/U Bus 20 OA104. [0.5)
Second alternate (0.25] is D/G [0.5)
REFERENCE SYO17 G-5 (AC Dist), pg.7 and Figures 4 and 1 Specific Objectives 2, 6, K&As 262001 K1.01 (3.8/4.3), K1.03 (3.4/3.8), K4.06 (3.6/3.9), K6.02 (3.6/3.9).
i e
, PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 27 ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-B. K. HAJEK
.
ANSWER 2.05 (2.50) False. [0.2] The turbine exhaust goes directly to the S OR Gland seal exhaust, valve steam leakoff, and exhaust line drain pot water is . . .
[0.3) True .
bbC. ok $ ( di ll pd t v h'~4ApWh
.
N udrM CST elen h [1Sc e. ' Tru True, M * M. M M .3%f. klIOh False. [0.2] It won't open automatically until level reaches -30 inches, but it can be opened manually any time after the high level condition clears. [0.3)
REFERENCE SYO17 C-5 (RCIC) ppg. 1 & 6, 24, 2, 8, 1 Specific Unit Objectives 3, 5, 6, 7, 1 K&As 217000 K1.03 3.6/3.6), K1.08 (3.3/3.4), KK4.01 (2.8/2.8), K4.02 (3.3/3.3), K4.06 (3.5/3.5), K4.07 (3.6/3.6).
ANSWER 2.06 (3.00) To the condenser [0.25] if vacuum is established Nedds ad dh
[0,25], and to Radwaste [0.25] if vacuum has not *mNO" been established. [0.25] hM M b E'd % ^ M m m .sp Q h. y d '?thssf n o.T p k "YG'E * Since some water is not being returned to the O t^ -
vessel, it is not passing through the regenerative heat exchanger. [0.5] This limits the capacity of ' #~ I the regen Hx and the Non Regen HX must maintain system below 130 degrees F on its own. [0.5]
Otherwise, the RWCU will isolate. [0.5] A slower heatup rate will produce less excess water, allowing more to be returned to the vessel. [0.5]
~# ^ "
S 0 7 L-1 (RWCU), pg. 2g #
'
?O Specific Unit Objectives 2,~3, K&As 204000 K1.01 (3.1/3.3), K1.06 (2.8/2.8), K1.07 (2.6/2.7), K1.09 (3.2/3.3), K3.02 (3.1/3.1), K4.03 (2.9/2.9), K4.04 (3.5/3.6), K4.07 (2.9/2.9), A2.12 j (2.7/2.8), SG7 (3.4/3.3), SG10 (3.2/3.2).
.
. , PLANT DESIGN INCLUDING' SAFETY AND EMERGENCY SYSTEMS PAGE 28 ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-B. K. EAJEK
.
LANSWER 2.07 (2.50) MSIVs A)So c~uetb RWCU System VAlvti IU4#
TIP System in OWDM Containment Atmosphere Control System 7%g Drywell Floor Drain / Equipment Drain System Reactor Building Chilled. Water to Rx Recirc pumps HVAC Isolation 5+ypALssim foot %/[ Actuation)
~1 sad *m Ma .ar6 each for any five 03 RHR will not isolate because it is needed for the LPCI injection mode. [0.5) If it is operating in Shutdown Cooling (different operating mode than in Part a), it will isolate. [0.5)
REFERENCE MA*O SYO17 E-3 (PCIS), . pp . T4 57 _ oM-1 W O*** pol 1,id5 -
Specific objective 5/ - - -
K&As 223002 K1.01 (3.8/3.9), K1.02 (3.3/3.5), KI.03 (3.0/3.2), K1.08 (3.4/3.5), K4.04 (3.2/3.6), SG4 (3.6/3.7)
z.0 8 a tl W5)
b (o.5)
N GYol7 - L-8 (Rsd 6pefc Ob'fcSA T jdfAo
.
, . . . . -. .c., , , -
e 2- PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS- PAGE 29 ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-B. K. HAJEK
.
ANSWER 2.09 (3.00) Continued operation of HPCI will quickly empty the (o,pgj CS (Low CST level alarm at 45" and swap to SP at 3' 7.5"] The combination of HPCI steam discharge and high (gy gg pool level could result in exceeding the design loads of the Suppression Pool.- The HPCI steam exhaust spargers become uncovere ( 0. 76) Below 2150 rpm, oil pressure will be too low to maintain the turbine stop valve in its open g ,35)
position, [and to prevent governor valves from operating too close to their seats.]
REFERENCE SYO17-C6 (HPCI), ppg. 17 f Specific Unit Objectives 5, 6, 8, 9, 1 SC006 Specific Objectives 2, K&A 206000 K4.09 (3.8/3.9), K4.10 (3.7/3.8), K4.19 (3.7/3.8), K5.05 (3.3/3.3), K6.04 (3.5/3.7), K6.05 (3.5/3.7).
^
ANSWER 2.10 (1.00)
(Q BecauseAonly one ESW Division will start on depressing only one pushbutton, [ leading to RHR pump motor damage ] [o.73 (~0 6 iil n d b d. A LICII^'cc.Wm G
a % b% biAdm mios 6N,*4 g g, 4 .,-
REFERENCE G -* * t &n SYO17 C-1 (RHR), ~ cog. 32 - 3 N N "' D Specific Unit Obje'ctives 5, 8, 1 W Fol7 A/S AJd bM
K&As 203000 K1.16 (3.1/3.2), K4.04 (2.6/2.7), A4.05 g&) 'a g Q
'
(4.3/4.1), SG7 (4.2/4.3) ;gS l g de ley. .c ,
8 ~' '$-
M pc.M d D
!
i
-

l
.
l b, e , h l
L-
,
-
, , PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 30
' ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-B. K. HAJEK
>

ANSWER 2.11 (2.50)
b, e, and h (will cause trips during both emergency and non emergency start All others will cause trips only during non emergency starts.]
-
'
0.25 for each response - not listing a response implies it is not an emergency tri REFERENCE SYO17 G-1 (DGs), ppg. 29 - 3 Specific Unit Objactive K&As 264000 K4.01 (3.5/3.7), K4.02 (4.0/4.2)
,-
'
.
I o
. * , - - , , - - - . - , , - . - , - , . , , - - , - - - - . , - , , . - , , , c 2-s--..- -- --- ,--
3 INSTRUMENTS AND CONTROLS PAGE 33 ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-B. K. HAJEK s
ANSWER 3.01 (2.00) TrytodrivearodwithincreasedCRDflow.(esivgdadd4Iad5k 3 Remove air from the top of the scram valves to L'tk ' # D dJ allow them to fail to'the scram positio (This is done by closing the Inst Air filter valves and opening the Scram Air Header Drain valve.]i.e.,IulskertA t e M De-energize the scram solenoid valve .
[This is done by opening breakers CB2A in RPS Trip] 14 drE4 da 9'~'
Sys A and CB8B in RPS B.] J n+d*++ H5 8 4 x Alternately scram rods and empty the SD N Lb22- Pull fuses to de-energ4% the scram pilot solenoid This is recommended in SSES-EPG, but considered to 'Gd, k n am Yodo be less reliable than Item 3 above due to hot short CAF to determine if credit should be Jif%H5/V4%
g give . Scram rods individually. h SgI a Sc % fend sw'd[" Cl '
D 2 14 REFERENCE SYO17 K-2 Specific Objectives 5, 8, SC006 Specific Objectives 2, PP002 Specific Unit Objectives 14, 1 EO-100-102 Flow Chart and Bases K&A 295037 EK3.07 (4.2/4.3)
-
!
.
,
, -3.. INSTRUMENTS AND CONTROLS PAGE 32 ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-B. K. HAJEK
$
ANSWE .02 (3.00) HIGHER THAN ACTUAL [because reference leg temperature will not cool to ambient, density will be less . . .] HIGHER THAN ACTUAL (due to lower density . . . ] HIGHER THAN ACTUAL [due to colder than normal water in the vessel and a higher variable leg pressure) HIGHER THAN ACTUAL (due to d/p decreasing across the valve to zero) HIGHER THAN ACTUAL [due to reference leg flashing]
L w'.Ll M4,Lph oscill fcAto sq $w ACTUAL (since both are calibrated at normal 4 temperature and pressure] o REFERENCE SYO17 J-2 (RPV Inst), ppg. 8- 1 Specific Objective SC006 G (MCD-RPV Level Inst), ppg. 8 - 1 Specific Unit Objective K&As 259002 K2.09 (2.9/3.0), 295012 AK1.02 (3.1/3.2),
216000 K4.05 (3.9/4.1), KS.01 (3.1/3.2), K5.06 (3.4/3.6), K5.07 (3.6/3.8), K5.10 (3.1/3.3)
ANSWER 3.03 (1.00)
Bypassing a low reading LPRM could cause the APRM to read hig Likewise, bypassing a high reading LPRM
could cause the APRM to read Io R E F E R E N C E T 'l N SYO17, ppg. 2 Specific Unit Objective K&As 215005 K3.05 (3.8/3.8), K6.03 (3.1/3.3)
.
+
-_A
.,- 3 INSTRUMENTS AND' CONTROLS 'PAGE .33
~ ANSWERS -- SUSQUEHANNA'I&2 -87/03/16-B. K. HAJEK
.w ANSWER 3.04 (3.00)
- . When any rod other than the selected rod is not at an even numbered notch 3 n. On % ,JJfuelr.d4d p*cq u C195,$ When the selected rod is not at an even notch and no rod motion is requested, or if it [n, A+g doesn't get to the next notch before the timer times ou 'E 3. 10b Orvdinat/R4 In fMWb 55 kp*M. . Depress the DISPLAY ROD DRIFT pushbutton [0,5] and check for the red light on the Full Core Displa [0.5) . Excessive CRD seal leakage Excessive cooling water pressure Air in the system Scram valve leakage Directional control valve malfunction Collet mechanism failure to latch Depressing Cont Insert Rod pushbutton 0.5 each for any two REFERENCE SYO17 K-7 (RMCS), ppg. 3, 6, Figure Specific Objectives 3, ON-155-006 Rod Drift AR-104-001 pg. 50 Rod Drift K&A 201002 K4.03 (3.6/3.6)
.
'
.
_m - _ - . - _ - - - . - , ~ . - , - . _ _ _ - - _ _ - - .
-, INSTRUMENTS AND CONTROLS PAGE 34 ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-B. K. HAJEK e
hl0 nrm wCMus Atc.vt Jyr ANSWER 3.05 (3.00)
30 M > 6 0 'lo Both pumps will run back to 456 percent (0.33]
because the feedwater flow signal to the recirc system is less than 20 percent (or - the flow limiter-is not bypassed at less than 20 percent feedwater flow). [0.67] On Unit # 1, the condition will clear N($4bbypuNd'
automatically, and the pumps will return to their l > gd) .
previous speed. [0.5] (
' nafe c m.hss-On Unit # 2, nothing will happen because the ,
limiter must be reset by operator action after the 4Y' f3 condition clears. [0.5] ) Pump B will run back as in Part a. [0.25] Pump A speed will remain the same (0.25] because of a scoop tube lockup caused by the loss of speed signal. [0.5]
REFERENCE
, SYO17 L-9 (Rx Recirc Control), ppg. 3 - Specific Unit Objectives 3, 4, 6, 1 K&As 202002 K1.08 (3.1/3.2), K3.05 (3.2/3.3), K4.02 (3.0/3.0), K6.04 (3.5/3.5)
ANSWER 3.06 (2.00) . 120/125 of scale [0.25] HV low [0.25] Module unplugged [0.25] Function switch not in operate [0.25] Mode switch in RUN [0.4]
AND [0.2]
Companion APRM on scale. [0.4]
REFERENC SYO17-I2, ppg. 17 - 1 Specific Objective K&A KI.01 (3.9/3.9), K4.02 (4.0/4.0), K6.06 (3.2/3.4)
i
-
.
-
, INSTZUMENTS AND CONTROLS PAGE 35 ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-B. K. HAJEK e
.
ANSWER 3,07 (1.00) The controller with the lower of the two electrical output signal At the high speed sto REFERENCE SYO17 D-3 (Rx FN-Sys), ppg. 11 - 1 System-Unit Objective K&As 259001 K5.03 (2.8/2.8), A1.04 (2.8/2.7)
ANSWER 3.08 (3.00) Automatically (0.25] bypassed when the mode switch is not in RUN. [0.5)
M Manually bypassed [0.25] with tese keylocked switched when mode switch is in Shutdown or Refuel. [0.5) -
Tku- wq Automatically bypassed [0.25] when turbine first stage pressure is less than X percent of q
its rated valuel. [0.5) - hos Ibs n 2.Y pned pvv44; Automatically bypassed [0.25] after a 10 seconds delay. [0.5)
Dofe: PoirdSnoY M "
REFERENCE .
p SYO17 L-5 (RPS), ppg. 31 - 3 Specific Objective hMcT,4FCd y n hmA*W K&A 212000 K4.12 (3.9/4.1) , ggf,*m
, p t40f
-
.
_
3 INSTRUMENTS AND CONTROLS .PAGE 36
. ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-B. K. HAJEK e
ANSWER 3.09 (3.00) Since the regulators are identical and redundant, the biases assure that only one regulator will normally be controllin When the A output falls at least 3 psi, the 3, backup, regulator will take control and limit the $t ed4 , '
closure of the control valve g
There is no backup for a failure that causes an h96' 'k increase in output signal, and-the control valves will ope REFERENCE SYO17 A-8 (EHC Pressure Control & Logic), ppg. 7-8, and Figure Specific Objective 1, 2, K&As 241000 K1.01 (3.8/3.9), K1.08 (3.6/3.7), K3.01 (4.1/4.1), K3.08 (3.7/3.7), K4.01 (3.8/3.8)
ANSWER 3.10 (3.00) It would be better to bypass APRM C [0.5] because with C bypassed, APRM E would automatically become the reference APRM. [0.5] 'If the LPRM is not bypassed, the high reading will be averaged with the other LPRM signals. [0.75]
However, if the LPRM is bypassed, its output would drop to zero, and would trip the dow scale trip unit. [0.75] This would remove it from the count circuit, and result in a RBM input based on the readings of the properly operating LPRMs. [0,5)
REFERENCE SYO17-K5, ppg. 5 - Specific Objectives 4, KSAs 215002 K1.01 (2.9/3.0), K1.02 (3.2/3.1), K4.01 (3.4/3.5), A2.03 (3.1/3.3).
.
e
-
6 - 3r ~ INSTRUMENTS AND CONTROLS PAGE 37
.
.. ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-B. K. HAJEK ANSWE .11 (1.00) RB Zone III Recirc Fan A/B start . SEGT Train A/B start . Control Room Emergency Outnide Air Supply Fan A/B start . Isolation of RB Zone III damper . RB Zone III Recirc Dampers ope .25 each for any fou REFERENCE Sy017 B2 (Process. Rad Mon Sys), pg. Specific Objective K&As 288000 K1.03 (3.7/3.7), K1.02 (3.4/3.4),.K1.05
~(3.3/3.6), K4.01 (3.7/3.9), K4.02 (3.7/3.8)
l
,
l
.
t
.
I e
g . ' PROCEDURES ~-' NORMAL, ABNORMAL, EMERGENCY AWD PAGE'.38 RADIOLOGICAL CONTROL
'
ANSWERS -- SUSQUEHANNA 2&2 -87/03/16-B. K. HAJEK ANSWER 4.01 (2.00) False. [ 0.' 2 5 ] - An exception is when an individual is receiving license training. [0.25] Tru ' Tru Fals [0.25] Operation shall be restricted until the cause of the contamination is identified and corrected and the system has been decontaminated or'if the system'Is required, a safety evaluation
.of the system as a radioactive system must be performed.-[0.25]
REFERENCE AD-QA-300, Conduct of Operations, ppg. 28, 23, 30, 3 PWG A1.02'(4.2/4.2),' EOP SGs 12 (Varies 3.5 and above),
Abnormal Procedures SGs 12 (Varles 3.4 and above)
ANSWER 4.0 (3.00) h 101 59w - g g g g-l) None ' R P V M 'r$ ES d,*44 e cry,3 dagWs None , 103 (0.125 each] ei %d% ced.CMv4 edm M . cud re-t . 103 REFERENCE PPOO2 (EOPs)
Specific Unit objective 1, 1 EOPs 570, 571, 572, 58 KAAs 295024-38, System Generic K/As II (4.3/4.5).
.
=
.
i
, 4.. PROCEDURES - NORMAL, AB4ORMAL, EMERGENCY AND' PAGE 39 i RADIOLOGICAL CONTROL o ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-B. K. HAJEK ,
l i
ANSWER 4.03 (3.00) YES: 1, 2, 3, 5, 7, 8, i NO: 4, 6, 1 j 0.2 for each correct answer. If candidate only lists the YES answers,'NO will be assumed for the others, Reducing power assists in maintaining vacuum by reducing the input of the non-condensible gasses (0.5] and lowering the condenser heat load. [0.5)
REFERENCE ON-143-001, Loss of Main Condenser Vacuum K&As 295002-AA2.02 (3.2), G5 (3.2).
ANSWER 4.04 (3.00) Reduce Recirc pump speed to minimum [0.5] , Insert CRAM Array rods (0.4] l OR [0.2)
Insert in-sequence rod [0.4) If power decrease did not stop the radiation increase, [0.5)
Manually scram the reactor [0.4)
AND [0.2)
Isolate the MSIV [0.4]
REFERENCE AD-QA-300, Conduct of Operations, Attachment K&As 272000 A2.01 (3.7/4.1), SGI (3.6/3.9), SG8 (3.5/3.5), SG14 (3.4/3.4), SG15 (3.7/4.2)'.
l l
[
!
l
,
o
- - _ - , -
o . . PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 40 RADIOLOGICAL CON ~ROL

l
ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-B. K. HAJEK ANSWER 4.05 (2.50) Changing the transfer switch to EMERG may 1) change the position of associated valves, (0.5)
or 2) change the operation of associated equipment, (0.5]
or 3) cause a loss of Control Room indication for the components controlled from the Remote Shutdown panel. [0.5]
4) Also, it may result in operation of equipment that cannot be monitored at the Remote Shutdown panel. [0.5) Nothin RdFERENCE OP-100-001, Remote Shutdown - Normal Plant Operating Lineup K&As 295016 AK2.01 (4.4/4.5), AK2.02 (4.0/4.1), AK 3.03 (3.5/3.7).
ANSWER 4.06 (3.00) Pressure set should be above reactor pressure (0.5]
Bypass Jack at zero [0.5) A vacuum could be pulled on the Drywell. [0.5]
Long path recire could leak into the vessel. [0.5)
b g'T~. S ,
C With 0.5 decade of overlapp 1""i" 202; LG; .-:J.::~) (0.f)
7 psi REFERENCE GO-100-002, Plant Startup and Heatup, ppg. 7, 8, 17, 2 K&As 241000 K1.06 (3.8/3.9), 239001 K1.17 (3.1/3.2),
215004 K4.06 (3.2f3.2), A2.05 (3.3/3.5), 206000 SG1 j (4.3/4.4), SG10 td.7/4.4).
Mok N Memel M
%C,b ' O 3 3.es A h g w c_. L
, % $ady 4 eadAQscahML4Q 4% A 4 7.Jt 3m 2. J .Re. b kaf sum m suJe eM 4 6
--
tod<Asex4.7 % + h m 444.f4 $
nw-MM'"
aw k 4 m % - <se.k sera n.u w o a W p
d A 14*b y h ao N " k to m 2 q }#
.
- _ - _ _ - _ - _ - _ _ - . . _ _ - _ - - _ _ _ _ _ _ _ _ __ __ - _-
..
. .
lj 4.-PROCEDURES - NORMAL,' ABNORMAL, EMERGENCY AND PAGE .41 L
RADIOLOGICAL CONTROL-o ANSWERS -- SUSQUEHANNA 1&2 -87/03/16-B. K. HAJEK
-
L
/
ANSWER 4.07 '(3.00) Fall closed Fall closed Fall open Fall closed Fall closed Fall closed REFERENCE ON-118-001, Loss of Instrument Air, ppg. 5-7, Paragraphs D, B, L, M, K&As 295019-AK2.01 (3.8/3.9), AK2.02 (2.9/3.0), AK2.03 (3.2/3.3),.AK2.05 (3.4/3.4), AK2.07 (3.2/3.2)
=
ANSWER 4.08 (1.50) mrem [n.te . Espom syIdelids ud erkJ-) When it reaches 75 percent of scale (1.0)
OR when it reaches 375 mR [1.0)
OR in about I hr 52.5 min after starting the job. (1.0)
REFERENCE AD-00-735, External Dosimetry Program, pp , 1 K&A 294001 K1.03 (3.3/3.8)
ANSWER 4.09 (2.00) . Rod position indication does not change (0.5] b' No indicated change in NI response (0.5){p 6 4;) q w 3 M C , 3, and (0.25 for each correct evaluation. p) is poe .]
(l.0 )
REFERENCE ON-155-001, Stuck Control Rod, p .
K&As-201003 A2.01 (3.4/3.6) ,
-
.
_ _ _ . . _ . . . . . _ _ _ _ _ _ _ . _ . _
42 PROCEDURFS - NORMAL, ABNORMAL, EMERGENCY AND PAGE 42 4 ,
RADIOLOGICAL CONTROL O ANSWERS --'SUSQUEHANNA 1&2 -87/03/16-B. K. HAJEK ANSWER 4.10 (2.00)
, Feed flow greater than steam flow Decreases Reactor power remains the same(Lg +> @d is ** bcM.Ae. '% prwC ' # '
F Oscillates
} k i: 6 obtv M d not M gJ,gly,", " g $g4g gg]
REFERENCE ON-183-001I, Stuck Open Safety / Relief Valve, p .
K&As 2390d2 K4.06 (3.5/3.7), A1.06 (3.7/3.8), A2.03 (4.1/4.2).
i Y t- & rytAk 6%Yts 'Mj 4 SD & Y 44 .44 u c. AM rise.

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

. , TEST CROSS REFERENCE PAGE 1 QUESTION VALUE REFERENCE QUESTION VALUE REFERENCE o ________ ______ __________ ________ ______ __________
01.01 2.25 BRH0000320 03.01 2.00 BRH0000342 01.02 2.00 BRH0000321 03.02 3.00 BRH0000343 01.03 2.25 BRH0000322 03.03 1.00 BRH0000344 01.04 2.00 BRH0000323 03.04 3.00 BRH0000345 01.05 1.50 BRH0000324 03.05 3.00 BRH0000346 01.06 2.00 BRH0000325 03.06 2.00 BRH0000347 01.07 2.00 BRH0000326 03.07 1.00 BRH0000348 01.08 2.50 BRH0000327 03.08 3.00 BRH0000349 01.09 2.50 BRH0000328 03.09 3.00 BRH0000350 01.10 3.00 BRH0000329 03.10 3.00 BRH0000351 01.11 3.00 BRH0000330 03.11 1.00 BRH0000352
______ ______
25.00 25.00 02.01 3.00 BRH0000331/ 04.01 2.00 BRH0000353 02.02 2.50 BRH0000332 04.02 3.00 BRH0000354 02.03 2.00 BRH0000333 04.03 3.00 BRH0000355 02.04 2.00 BRH0000334 04.04 3.00 BRH0000356 02.05 2.50 BRH0000335 04.05 2.50 BRH0000357 02.06 3.00 BRH0000336 04.06 3.00 BRH0000358 02.07 2.50 BRH0000337- 04.07 3.00 BRH0000359 02.08 1.00 BRH0000338 04.08 1.50 BRH0000360 02.09 3.00 BRH0000339 04.09 2.00 BRH0000361 02.10 1.00 BRH0000340 04.10 2.00 BRH0000362 02.11 2.50 BRH0000341 ------

25.00 25.00 ------
______
100.00 l
.
( ^ s

v t e Inspection Report - Susquehanna - 1987005
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 \|
1987 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 7(0) 15(0) 19(0) 20(0) Security Operator Exams Emergency Planning Other

IR 05000387/1987005

Text

Navigation menu

Search