Text

Exam Rept 50-334/OL-86-16 on 860722-31.Exam Results:Seven Reactor Operators,Two Senior Reactor Operator (SRO) Retake Candidates & One SRO Candidate Passed All Portions of Exam
	ML20210R276
Person / Time
Site:	Beaver Valley
Issue date:	09/15/1986
From:	Barber G, Keller R, Kister H; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
Shared Package
ML20210R261	List: ML20210R258; ML20210R276; ML20210R294;
References
	50-334-OL-86-16, NUDOCS 8610070222
	Download: ML20210R276 (94)
	v • d • e

U. S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO. 50-334/86-16(0L)

FACILITY DOCKET NO. 50-334 FACILITY LICENSE NO. DPR-66 LICENSEE: Duquesne Light Company Post Office Box 4 Shippingport, Pennsylvania 15077 FACILITY: Beaver Valley Unit 1 EXAMINATION DATES: July 22-31, 1985 CHIEF EXAMINER:

N er,

//A. - 9 M ctor Engineer (Examiner) D&te REVIEWED BY:

R. M. Keller,' Chi'ef, Projects Section 1C Th/(Af Date APPROVED BY: /

Harry B. Kistir, Chief Daye /

Projects Branch No. 1

SUMMARY

Oral, written and simulator examinations were administered to twelve reactor operator and three senior reactor operator and one senior reactor operator retake candidate. In addition, just the oral and simulator portion was administered to one senior reactor operator retake candidate. Seven reactor operators passed all' portions of their examinations and will be issued licenses. Both senior reactor operator retake candidates passed all required portions of their exams and will be issued licenses. Of the three remaining senior reactor operator candidates, one passed all portions of the exam and will be issued a license. The specific details of the candidates that failed all or portions of their exam can be found in the examination results table on the following page.

8610070222 860923 PDR ADOCK 05000334 V PDR

2 REPORT DETAILS TYPE OF EXAMS: Initial Replacement X Requalification EXAM RESULTS:

l RO l SR0 l i Pass / Fail i Pass / Fail l I I I I I I i

[ Written Exam l 10 / 2 l 4/1 I l l l I I I I I 10ral Exam l 11 / 1 1 5/0 l l l l l l l l l l Simulator Examl 9/3 1 3/2 l I .I I I l l l l

, l0verall l 7/5 1 3/2 I I I I I I I I I I. CHIEF EXAMINEk AT SITE: G. S. Barber, NRC II. OTHER EXAMINERS: R. M. Keller, NRC N. F. Dudley, NRC B. S. Norris, NRC D. M. Silk, NRC B. Gruel, PNL L. Defferding, PNL i

1 1

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III. Generic observations and weaknesses noted during the operating exams:

1. Candidates became preoccupied with hanging caution tags and 00S stickers while instrument malfunctions and casualties were still in progress. Attention to these administrative requirements is commendable, however, timing in the cases noted was improper.

2. Nonpermanent marker pen was used to list diesel trips, AFW start signals and other oral examination answers on panels, components and switchgear throughout the plant. If these operator aids are needed, they should be replaced with permanent tags or labels.

3. The control room key cabinet was improperly locked. It could be opened by turning the combination lock approximately one quarter turn. Three instances were observed where Maintenance and Radiological Controls personnel took keys without an operations supervisor's approval. The follow-up on this issue was turned over to the senior resident inspector and will be discussed in Inspection Report 86-18,

4. Some R0s could not do a manual RCS subcooling calculation.

5. An examiner observed three plant personnel in the Spent Fuel Pool Area without proper anti-contamination clothing.

6. Some R0s could not explain the functions and principles of operation of the incore instrumentation system.

7. An SR0 could not verify the accuracy of a manual calculation for an unplanned release. The unplanned release was due to a steam gener-ator tube rupture with a stuck open atmospheric dump valve.

8. Candidates frequently reported proper SIS, CIA and CIB valve and pump alignment prior to checking or verifying the actual indication.

IV. Simulator Deficiencies noted during the Operating Examinations:

1. During a scenario, the RO identified that annunciators were alarming and resetting without a horn. When he attempted to acknowledge and test the alarm, the simulator froze. In a subsequent scenario, the manual tap changers failed to operate and the Building Services Panel Alarms could not be acknowledged. The simulator instructors attributed these deficiencies to an electrical storm that occurred the previous night and to the lack of surge suppressors for the simulator's power supply.

2. There were instances during scenarios where the turbine driven AFW pump did not start even with the proper valve alignment.

3. Candidates were distracted by erroneous electrical spikes in the megawatt, steam flow and feed flow recorders.

4

4. Instructors stated that the BOL snapshots were not as accurate as the MOL snapshots.

5. The operators normally have the source range fuses removed when operating in Mode 1. There is no administrative or procedural basis for this action.

6. Changing the lineup of the Baron Recovery System sometimes resulted in isolating CCR to the Reactor Coolant Pumps. The normal flow (48 gpm) is very close to the trip setpoint (50 gpm).

7. AFW pump flow was 350 gpm with the normal supply tank empty and suction valves from river water closed.

V. Generic weaknesses noted from grading of written exams:

A. R0 candidates were unable to adequately explain the following:

1. The effect of starting a RCP or bypassing a string of feedwater heaters on actual critical rod position during reactor startup.

2. The effect of interstitial fission gas absorption on fuel centerline temperature.

3. What determines differential pressure across the #1 seal of a RCP. ,

4. The purpose of opening the turbine driven AFW pumps recircula-tion valve.

5. The automatic actions associated with the Containment Purge Exhaust Monitor during Refueling.

6. Immediate action substeps for E-0, Reactor Trip /SI. For example, the actions required to verify AFW flow.

7. FRGs may be entered from other than red path conditions. For example, FR-H.1 is required to be entered from E-0 when flow is less than 350 gpm.

8. Conditions that cause high activity in the RCS per A0P-43, High Reactor Coolant Activity.

B. SR0 candidates were unable to adequately explain the following:

1. The design features that protect the CCR system in the event of a leak in the thermal barrier.

2. The advantages and disadvantages of using alternate dilute.

3. Reasons for closing the MSIVs on a steam line rupture.

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4. System parameters checked that verify that inadequate core cooling no longer exists.

5. When pressurizer venting should take priority over containment hydrogen limits.

6. Given the RCS leakage T.S., determine if leakage limits are exceeded for a given set of plant conditions.

VI. Training / Reference Material:

1. The training material was improved from the previous examination.

However, there were several instances where the material was incorrect or inadequate. (See Attachment 4)

VII. Personnel Present at the Exit Interview:

NRC Personnel G. S. Barber, Reactor Engineer (Examiner)

8. S. Norris, Reactor Engineer (Examiner)

D. M. Silk, Reactor Engineer (Examiner)

W. Troskoski, Senior Resident Inspector T. J. Kenny, Senior Resident Inspector A. J. Lodewyk, Reactor Engineer Facility Personnel T. D. Jones, General Manager - Nuclear Operations W. S. Lacey, Plant Manager, BV-1 J. D. Sieber, Senior Manager, BV-1 L. G. Schad, Coordinator, Simulator Training A. J. Lindgren, Simulator Supervisor T. E. Kuhar, Nuclear Operations Instructor A. Nowinowski, Westinghouse Training P. A. Russell, Nuclear Operations Instructor VIII. Summary of NRC Comments made at exit interview:

The chief examiner reviewed the number and type of examinations admin-istered during the previous two weeks and presented generic weaknesses observed during the simulator and oral examinations.

IX. Examination Review:

An examination review was conducted. Facility comments were discussed on a line item basis. All items were considered during grading but not all items resulted in a change to the master exam. Attachment 3 details the Facility's Comments on the written exam. Attachment 4 details the significant changes to the examinations.

6 Attachments:

1. Written Examination and Answer Key (RO)

2. Written Examination and Answer Key (SRO)

3. Facility Comments on the Written Examination

4. NRC Response to Facility Comments

kTrnahmen /

AAs TER U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: BEAVER VALLEY 1&2 REACTOR TYPE: PWR-WEC3 DATE ADMINISTERED: 86/07/22 EXAMINER: SILK, D.

APPLICANT: _ b kct__k_g ___________

INSTRUCTIONS TO APPLICANT:

Uno separate paper for the answers. Write answers on one side only.

Staple question sheet on top of the answer sheets. Points for each quastion are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at loost 80%. Examination papers will be picked up six (6) hours after the examination starts.

% OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY

_3_133__ _ ___________ ________ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS 00

_I'_5 00I____ I_I__

_'5 ___________ ________ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 00

_'5 00I_I____ I I__

_'_5 ___________ ________ 7. PROCEDURES - NORMAL, ABNORMAu, EMERGENCY AND RADIOLOGICAL CONTROL 50 ADMINISTRATIVE PROCEDURES,

_'I_I__0__ _1'S _1__ 00 ,__________ ________ 8.

CONDITIONS, AND LIMITATIONS 100.00 100.00 TOTALS FINAL GRADE _________________%

All work done on this examination is my own. I have neither givon not received aid.

~~~~~~~~~~~~~~

dPPL5CEUTI5~55GU55URE L

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 2 UULbl1UN 's.UI (J.UU) c.. Would the -positioning of a -neutron -source TOO - CLOSE -to -the -

neutron detector, being used for constructing a 1/H plot, result in OVERPREDICTING (not conservative) or UNDERPREDICTING (conservative) the reactivity addition needed to reach criticality? EXPLAIN. (1.0)

6. How does the initial source range level (cps) af#ect critical rod position? EXPLAIN. (1.0)

c. How does the Positive reactivity insertion rate affect the source range count level at which criticality is achieved? EXPLAIN. (1.0)

GUESTION 5.02 (3.50)

c. Explain both HOW AND WHY the following factors affect differential baron worth (more negative, less negative or no change).

1. Baron concentration increase (0.75)

2. Moderator temperature decrease (0.75)

3. Fission product buildup (0.75)

4. Core burnup from MOL to EOL with constant rod position (0.751

b. Why does the critical boron concentration drop rapidly from 0 to 150 MWD /MTU of burnup as seen in Figure 1? (0.5)

QUESTION 5.03 (3.00)

a. How does DNBR change (increase, decrease. no change) as the following are increased? (Consider each s e o a r a t e l y 'r . (1.0)

1. Tavs i 2. RCS pressure

3. RCS flow

4. Reactor power (Constant Tavs)

6. What adverse fuel assembly condition could result if actual heat fluv exceeds the critical heat f l u:, in a PWR core? Explain. (1.0-

c. From rigure 2, what parameter is being limited on Section A of the figure and what is the significance of it? (1.0)

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

So THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 3 QUESTION 5 04 (2.00)

Af ter operation at -100% -power for several weeks near the end -of cycle e---

power is reduced to 75% vsing reds only.

a. Explain HOW and WHY Xenon concentration will change over the next 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months . (1.5)

6. what rod motion would be required to maintain the plant at 75% power over the same 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months assuming no change in boron concentration 0 Include applicable time frames. (0.5)

GUESTION 5.05 (1.50)

a. Does Beta bar effective increase, decrease, or remain the same from BOL to EOL? Explain your answer. (1.0)

b. For two equivalent positive reactivity additions to a critical reactor, will the SUR be the same. larger, or smaller at EOL as compared to BOL? No explanation is necessary. (0.5)

GUESTION 5.06 (2.50)

a. To increase the discharge head of a variable speed centrifugal hydro-pump from 1200 to 1800 psia by what factors should the speed and power inputs be increased? Show vovr calculations. (1.5)

b. What pressure is needed at the suction of a feed pump to provide 215 feet of NPSH if the water is at 384 Fo (1.0)

OUESTION 5.07 (2.00)

When would a rod be worth more - if it were droppec while at oower or if it were stuck out while all other rods were inserted? EXPLAIN.

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

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 4 QUESTION 5 08 (Z.00)

- -- - -With -ell systems -in-manual and no operator action, what effect ( inc rea se, ---

decrease, no change) will decreasin3 the circulating water temperature have on the following?

a. Condenser vacuum

5. Condensate temperature

c. Steam generator pressure

d. Electrical output

e. Reactor power GUESTION 5 09 (2.50)

a. What effect does increasing moderator temperature have on control roo worth? Explain. (1.0)

b. What is the effect of a dropped rod on long term reactor power and Tave? Explain. Assume all systems in manual and no reactor trip occurs. (1.5)

GUESTION 5.10 (3.00)

The reactor is at 70% power and Tave is 568 F. A governor valve failure reises load 15%.

a. From a reactivity standpoint. explain how anc why reactor power responds. Assume rods in manual. (1.5)

6. Using Figure 3, caleviate the new Tave if rods are in automatic.

Show all work and state all assumptions. (1.Si

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

i

60 PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 5 QUESTION 6 01 (2.00)

a. nny As vor Auss ai cumvennebana vu1La3r more noLAceaoAr avrAn3 a startup than at 100% power? (0.5)

b. At 100% power the N44-8 Power Range Detector fails high. With rods in manual, give five annunciators associated with the NIS that alarm.

(1.5)

QUESTION 6.02 (2.50)

Delcic a . lJhat is the major advantage of drawing an RCS activity sample from pga the letdown line instead of directly from the RCS? (0.4)

b. Besides the Condenser Air Ejector Radiation Monitor, list four rad-iation monitor alarms that may be indicative of a primary to see-ondary leak? (1.6)

c. What automatically happens when a high-high alarm from the Condenser Air Ejector Radiation Monitor occurs? (0.5)

OUESTION 6.03 (2.20)

3. How would an operator determine the location of a 10 GPM leak from the component cooling water system by using the indications available to him in the main control room? (0.7)

b. What three design features of the component cooling water system min-imize the effects of a rupture of the RCP thermal barrier? (1.5)

GUESTION 6.04 (2 00)

a. What is the reason for maintaining a minimum pressure of 15 psig in the volume control tank (0,5)

b. Normal operations has the '1C' charging Pump breakers 1E15 and 1F15 disconnected from the bus. What prevents tying both emergency busses together? (0.5)

c. When is the Alternate Dilute mode used and what disadvantage accomp-antes its use? (1.0)

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

60 PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 6 Delete 6.os-QUESTION 6.05 (2.00)

The plant is operating at JOI power when the contro[11ng rirst stage i m Eu _1_s e_ p r e s_s,u r.e _ t r_ a n s m i t t e r P T 4 4 6 f a i l s H IG H . Explain the effects of this failure and the sequence of events (control and protection) that lead to a reactor trip. Assume BOL, no operator action and initial plant conditions are in a normal system line-up for 30% power.

(Setpoints are not required.) (2.0)

GUESTION 6.06 (2.70)

a. What is used to control RCS pressure during cold solid plant operat-lons? (0.4)

6. What three plant conditions provide inputs to the interlocks assoc 1-ated with RHR suction valve MOV-RH-701? Setpoints are required.(1.5)

c. Prior to entering a water solid operating mode, describe how over-pressure protection is enabled? (0.4)

d. If the air supply system for PORV's PCV-RC-455C & D fails, describe how the overpressure protection system functions? (0.4)

GUESTION 6.07 (2.50)

a. Why is the operability of the steam generator code safety valves important during power operation? (0.5)

b. Give two reasons (NOT CONDITIONS) why the MSIV's are required to close during a steam line rupture. (1.0)

c. Which mode (HSB, HZP. HFP) and time in cycle (BOL, MOL, EOL) will have the most severe effect on a main steam line break accident.

E:rolain each separatelv. (1.0)

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

l

- ~ -

60 PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 7 i ______________________________________________________

)

QUESTION 6.08 (2.30)

a. urtsinating at une 91ovv ist sus, provloe a sKeten wnlen snows now power is supplied to one 120 VAC Vital Bus and one 125 VDC Bus. The

- - '~

s1Cetch shbv1d TncTUde al1 ~ rio r m a 17 a l t e r n a t e a n'd~ e m e r g e n c y supp1Tes.

Label all electrical components, busses and transformers. (Breakers are not required.) (1.5)

b. Why do the two 480 V Emergency Motor Control Centers MCC1-E13 and E14 feeder breakers remain closed during a loss of offsite power?

(0.4)

c. What signal is needed to allow sequencial loading following a loss of offsite power? (0.4)

DUESTION 6.09 (3.20)

a. What two simultaneous conditions will cause the quench spray flow cut-back valves (MOV-10S-103A,B) to close? (0.8)

b. What is the purpose of the orifice that is parallel to quench spray flow cut-back valves? (0.8)

c. In the recireviation spray coolers, what is the reason for the recir-culation water pressure bein3 3reater than the river water pressure?

(0.8)

d. CIB has been reset and the spray pumps have been secured. If a CIB signal recurs, will the quench spray pumps restart automatically?

EXPLAIN. (0.8)

GUESTION 6.10 (3.60)

3. The safety injection accumulators are required to be maintained with-in certain pressure limits. What pr oblems e::Is t if the pressure is significantly above and below its limits) (0.8)

, b. What conditions are required before automatically transferring from the injection phase to the recirculation phase? Include logic and coincidences. (0.4)

c. What is the sequence of the automatic valve realignments that occur to transfer from the injection phase to the recirculation phase?

(2.0)

d. If the low head safety injection pumps fall during recirculation, what can be done to provide suction to the high head safety inject-ion pumps? (0.4)

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

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 8

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~~~~Rd6 UL55iCdt C6NTRUL QUESI1UN /.01 (1./U1

-- ---c. - - A --Oecay Tenk--Discher se-is-4n-progress-when Gaseous Wa ste Cas Monitor 4-RM-1GW-108B, becomes inoperable. Briefly explain what has to be done to continue the release. (1.0)

b. Durin3 a liquid release, a liquid waste effluent high-hiSh activity alarm comes in. The problem is idertified and corrected. When pump-ing to the cooling tower is re-established, hi3h a,ctivity is still present. Is this to be expected? Explain. (0.7)

GUESTION 7.02 (2.50)

a. What are the two entry conditions to FR-H.1, ' Response to Loss of Secondary Heat Sink'? (0.5)

6. What two conditions, caused by a loss of secondary heat sink, calls for tripping the RCP's and initiating feed and bleed? (0.8)

c. In the response to inadequate core cooling, what three system para-meters are checked to verify adequate core cooling has been re-covered? (1.2)

DUESTION 7.03 (2.50)

Answer the following concerning E-0, Reactor Taip or Safety Injection *

a. The Main Turbine has not tripped and you attempt a manual trip as required, with no response. What additional action are you required to take in order to shutdown the turbine? (0.6)

o. List three plant conditions that require SI initiation? Include setpoints. (0.6)

c. What four parameters are checked to determine if SI flow should be terminated? (0.9)

d. Following an SI reset, what condition must be met before an automatic reinitiation of SI will occur? (0.4) l (anax* CATEGORY 07 CONTINUED ON NEXT PAGE *****)

1

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 9

~~~~ - ~~~~~~~~~~~~~~~~~~~~~~~~

RI6EUL66fCAt C UTR6L UUESTIUN 7 09 LZ.3UJ

- Answee -the -f e l l ow ing w onc e r n i ng - the--EOP s i cule s__o f--usage .___ _ _ _ _ _______ __ __

a. How does the operator know if the sequencial performance of subtasks within a procedure is required? (0.5)

b. When does monitoring of the STATUS TREE's begin? List two circum-stances. (1.0)

c. The ST.A reports the following:

1. Heat Sink -

Orange Path

2. Containment - Red Path

3. Core Cooling - Orange Path

4. Suberiticality - Yellow Path List the order in which the above conditions should be addressed?

(1.0)

QUESTION 7.05 (2.00)

In order to maintain the plant at 100% power, work must be performed inside the containment in a radiation field of 850 MREH/HR samma and 300 MREM /HR thermal and fast neutron. The maintenance man selected is 28 years old and has a lifetime exposure through last quarter of 48 REM on his NRC Form 45 additionally, he has accumulated 1.0 REM so far this quarter.

a. How long may the man work in this area without exceeding his 10 CFR limit? Show all work. (1.2)

b. During a declared emergency, this individual volunteers to enter a high radiation area and perform work necessarv to prevent further effluent release. In accordance with the Station Procedures, what is his man-imum allowed whole body exposure and whose authorization is needed?

(0.8)

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

70 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 10

~~~~ ~ ------------------------

RA656L6U5 CAL 66 TR6t uutdilun /.vo (4.sh)

___ _ Answer--the f ollowing Ausstions in-r egard_to- a m eac tor - star top. - - ---- .-----

a. An id e RCP in a non-isolated loop, with the RCS at 250 F, shall not Odde be started unless what specification is met? (0.5) 7,064

b. What should be done if criticality is not achieved by 500 pcm past the ECP? (0.5)

c. The Reactor Coolant System lowest operating temperature (Tave) is not allowed to go below 541 F during a reactor startup. What are the four bases for this limit? (1.5)

QUESTION 7.07 (2.00)

Answer the following questions in regards to Operating Manual 1.6.4 0,

' Response to Voids in Reactor Vessel.'

a. What symptoms would be indicative of a void in the reactor vessel?

(1.0)

b. When should venting the reactor vessel take priority over containment hydrogen limits? (0.5)

c. When should venting the pressurizer take priority over containment hydrogen limits? (0.5)

GUESTION 7.08 (3.30)

a. What are the two reasons for stopping all RCp's in the case of a small break LOCA? (0.6)

b. What are two criteria for determining if RCp's should be stopped if HHSI pumps are running? (0.8)

c. In accordance with E-3, Steam Generator Tube Rupture, list four ways that a ruptured steani generator can be identified. (1.0)

d. What is the defAnttion of adverse containment conditions? (0.9)

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

I

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 11

~~~~R d6 6E6656AE"_6UNTR6E~~~~~~~~~~~~~~~~~~~~~~~~

UUE5IlDN 7.OY (3.bO)

--- --- - o,-Wh il e --mov i ng -f ue l , -a -h i g h-h i gh-a l e r s --free --a wonta i nmen t-pur se -exhaus t monitor occurs. What automatic actions follow besides an evacuation alarm? (1.6)

b. What is the basis for the requirement that two RHR loops be operable when water level is less than 23 feet above the vessel flanse? (0.6)

c. Under what conditions is it permissible to stop RHR flow during refueling? (0.8)

GUESTION 7.10 (3.00)

a. per A0p-43, High Reactor Coolant Activity, what three plant condit-ions can cause high RCS activity due to the release of irradiated corrosion products? (1.0)

b. In the event of high RCS activity, what is the reason for securing the followinst containment sump pumps, primary drain pumps and their containment isolation valves, containment vacuum pumps, and contain-ment isolation valves for reactor plant sample systems? (0.5)

c. An increased concentration of what two gases sampled from the VCT sas sample siace would be indicative of failed fuel? (0.5)

d. Under what conditions can power operations continue if the specific activity of the primary coolant is greater than its Technical Specif-cation linit? (1.0)

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

k. . .,

c

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 12 QUESTION 8.01 (1.50) ine concentration or Ine ooric acto solv 11on in sne nerveting water Storage Tank (RWST) shall be verified once per 7 days in accordance -

-- ~ ~ -~ Ui t h~T e3 n i c a l S p e c iTic a t i o n 3 . 5 . 5 . The~cTe m is t~s a mipl e'd TWeNST~ -

on the following schedule. (All samples taken at 1200 hours0.0139 days 0.333 hours 0.00198 weeks 4.566e-4 months .)

April 1 ---

April 8 --- April 16 --- April 24 --- April 31

a. EXPLAIN why or why not surveillance time interval requirements were exceeced on April 16. (0.75)

b. EXPLAIN why or why not surveillance time interval requirements were exceeded on April 24. (0.75)

QUESTION 8.02 (1.00)

What restrictions are placed on the manning and composition of the Fire Brigade? (1.0)

GUESTION 8.03 (2.50)

The RCS is heating up at 50 F per hour with the RCS presently at 325 F.

Maintenance reports that Chargins Pump 1B repairs will not be completed for one hour but that Charging Pump 1A is operable. Technical Specificat-ions Action Statement allows 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to repair an inoperable pump in Mode 3. What action, if any, should be taken? (2 5)

OUESTION 8.04 (2.50)

What action (s) (BOTH operational AND administrative) must be taken if the RCS-PRESSURE-Safety Limit is exceeded in accordance with Technical Speci-fications? Consider ALL Modes AND include applicable time limits in your answer.

QUESTION 8.05 (2.50)

Discuss the relationship between Limiting Conditions for Operations, Limiting Safety System Settings, and Safety Limits in terms of preventing release of radioactivity to the environment.

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

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 13 QUESTION 8.06 (2.00)

Ine plant is operating at av4 Aoaa anu you are une un 2nii w an s , 5 ourmi -

visor. Explain what actions, if any, would need to be taken in re3ards to shif t staf firig ~f or~ the ToITowing conditionsFCONSIDER-EACWCASE SEP - ~ ~ ~

~

ARATELY.

a. Your on-shift BOP operator is seriously injured in the plant and you send him to the hospital for treatment. (1 25)

b. The on-coming STA calls and says he won't be in. (0.75)

GUESTION 8.07 (2.00) .

The plant is operating at 75% power and the latest leak rate data shows' l 13.2 GPM - Corrected RCS leakage rate ,

1.5 GPM - Leakage into the Pressort:er Relief Tank 1.2 GPM - Leakage into the Primary Drains Transfer Tank 3.4 GPM - Leakage through SI-23, RCS Loop 1A, cold les isolation (Previous leakage rate was 1.6 GPM) 0.8 GPH - Total primary to secondary leakage 4 2 GPM - Leakage past RCP seals What RCS leakage limits, if any, have been exceeded? Refer to attached Technical Specifications.

OUESTION 8.08 (2.50)

Per Technical Specifications, when does containment integrity exist? (2.5)

QUESTION 8.00 (2.00)

3. If an emergency condition develops, who assumes the role of the Emergency Director? (0.5)

b. When is an immediate deviation from Technical Specifications just-ified? (1.5) l

(***** CATEGORY 08 CONTINUE 0 ON NEXT PAGE *****)

1 l

l I

L

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 14 QUESTION 8.10 (3.50)

a. wnat is tne MAN 1NUM numDer or operaole excore enannels incicating

~ ~ -~

AFD outside the target band before AFD is considered outside ~ ~ --

i t s t a r g e tT n d~~b~y T e EFn i c a l S p~e di f i d a ti~o n s ? - ' 00 ~.TT--- ~ ~ ~

b. Assume the plant is operating at full power and the Axial Flux Difference (AFD) has been outside the target band for the last 5 minutes. What are the TWO actions specified which you may choose between to meet the Tecnnical Specification requirements? Include time limitations. (1.0)

c. Assume that it is 0310 on 05/13/85 and the plant is presently at 45% power. Considering the AFD penalty history below, at what date and time may power be increased above 50%? EXPLAIN.

(Show all work.) Assume no deviation outside the band after 0310 on 05/13/85.

TIME WENT GUT TIME BACK DATE OF BAND IN BAND POWER 05/12/85 0310 0318 85%

05/12/85 1557 1637 65%

05/13/85 0148 0310 45% (2.0, OUESTION 8.11 (3.00)

a. What are the responsibilities of the Nuclear Station Opercting Fore-man (NSOF) at shift change? (1 0)

b. If clearance is needed to do maintenance on a piece of non-ESF equip-ment, how is permission granted to do the work and who gives the per-mission? (1.0)

c. Can a SRO soley avtnorize the installation of a jumper for non-Technical Specification related equipment? Explain. (1.0)

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

(xxxxxxxxxxxxx END OF EXAMINATION x*xusxxxsxxxxxx)

g:

. . f a na v o s/t Cycle efficiency o (Metworv out)/(Energy in) 2 -

o = og s = V,t + 1/2 at 2

E = ac KE = 1/2 av 2

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CR, = S/(1 - K,ff,)

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T = (1*/s) + [(s - o)/ o] M = 1/(1 - K,ff) = CR j/CR, l T = s/(o - 8) M = (1 - K,ff,)/(1 - K,ffj)

T = (s - o)/(lo) SDM = (1 - K ,ff)/K ,ff a = (K ,ff-1)/K ,ff = 4K ,ff/K,ff t* = 10-5 3,,,,q, T = 0.1 seconds-I o = ((t*/(T K,ff)] + [i,ff (1 / + ST)]

l Idjj=Id P = (t+V)/(3 x 1010) Idjj 2 =2Id22 2

2 I = eN '

R/hr = (0.5 CE)/d (meters) l R/hr = 6 CE/d2 (feet)

Water Parameters Miscellaneous Conversions 1 gal. = 8.345 lba.- 1 curie = 3.7 x 1010 dps I ga:. = 3.78 liters 1 kg = 2.21 lbm 1 ft* = 7.48 gal. I hp'= 2.54 x 10 3 Stu/hr Density = 62.4 lbm/ft3 1 av = 3.41 x 106 Stu/hr Density = 1 gm/cm3 .

lin = 2.54 cm Heat of vaporization = 970 Stu/lbm *F = 9/5'C + 32 Heat of fusion = 144 Stu/lbe *C = 5/9 (*F-32) 1 Atm = 14.7 psi = 29.9 in. Hg. 1 BTU = 778 ft-lbf TLRLJMitaSLDS6R3fMJh 2

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0.8 1.0 1.2 0.4 0.6 O 0.2

FRACTION OF RATED THERMAt. POWER

.l 3 REACTOR CORE SAFETY t.lMIT

THREE t. OOPS IN OPERATION

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REACTOR COOLANT SYSTEM _

2 OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION' 3.4.6.2 Reactor Coolant System leakage s' hall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGE,

c. 1 GPM total primary-te-secondary leakage through all staam generators not isolated from the Reactor Coolant System and 500 gallons per day through any one steam generator not isolated from the Reactor Coolant System,

d. 10 GPM IDENTIFIED LEAKAG,E from the Reactor Coolant' Systen, and

.I

e. 28 GPM CONTROLLED LEAKAGE at a Reactor Coolant System

~

pressure of 2230 +20 psig.

APPLICABILITY: MODES 1, 2, 3 and 4.

/hih., .

p ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, be in at least $0T STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ,

b. With any Reactor Coolant System 1eakage greater than any one of tha above limits, excluding PRESSURE SOUNDARY LEAKAGE, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at leasc HOT STANDBY within the .next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REOUIREMENTS 4.4.6.2 Reactor Coolant Systen leakages shall be demonstrated to be within each of the above Ifmits by:

a. Monitoring the containment atmosphere particulate and. gaseous :

radioactivity monitor at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

?.'r i.A/

.,/

BEAVER VALLEY - UNIT 1 3/4 4-13

~ __

A

g. REACTOR COOLANT SYSTEM I

2 PRESSURE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.4.6.3 Reactor cool' ant system pressure isolation valves shall be ooerational.

APPLICA9ILITY Modes 1,2, 3 and 4 Action:

1. All pressure isolation valves listed in Table 4.4-3 shall be functional as a pressure isolation device, except as specified in 2. Valve leakage shall not exceed the amounts indicated.

2. In the event that integrity of any pressure isolation valve specified in Table 4.4-3 cannot be demonstrated, reactor operation may continue, provided that at least tyto valves in each high pressure line having a non-functional valve are in,and re.ain in, the mode corresponding to the isolated condition. a)

3. 'If Specification 1 and 2 cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4 The provision of specification 4.0.4 is not applicable for entry into Mode 3 or 4 l

0

(*) Motor operated valves shall be placed in the closed ,,asition md pcwcr supplies deenergized.

I i . ,g'-

(l BEAVER VALLEY - UNIT 1 3/4 4-14a Order dated April 20,1981 l

1 L - - _ _ . ,_ ____ _ _

a

' REACTOR COOLANT SYSTEMS SURVEILLANCE REQUIREMENT (8) on each valve listed in Table 4.4.6.3.1 pe i test ShedpriortoenteringMode1afterevery l 4.4-$di sha$ak$$e accomp time the plant is placed in the cold shutdown condition for refueling, after each time the plant is placed in a cold shutdown condition for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months if testing has not been accomplished in the proceeding 9 months and prior to returning the valve to service after maintenance, repair or replacement work is performed.

4.4.6.3.2 Whenever integrity of a pressure isolation valve listed in Table 4.4-3 cannot be demonstrated the integrity of the remaining valve in each high pressure line having a leaking valve shall be determined and recorded daily. In addition, the position of the other closed valve located in the high pressure piping shall be recorded daily.

h* N :o (a) To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved procedures and supported by computations

. showing that the method is capable of demonstrating valve compliance with the leakage criteria.

BEAVER VALLEY - UNIT 1 3/4 4-14b '9/ M // M d/M fII/79// M I AMENDMENT NO. 101

- - - - . - - - - - - ~,,,.n.-.. . . . - , -

A TABLE 4.4 3 7<-

REACTOR COOLANT SYSTEM _ PRESSURE ISOLATION VALVES Pkximum(a) (b)

System Valve No. Allowable Leakace Loop 1, cold leg 'SI-23 < 5.0 GPM SI-12 ][ 5.0 GPM Loop 2, cold leg SI-24 < 5.0 GPM SI-11 ][ 5.0 GPM Loop 3, cold leg SI-25 < 5.0 GPM SI-10 ][5.0GPM

(.

NL .

l'I 1. Leakage rates less than or equal to 1.0 gpm are considered acceptable.

l'

2. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

3. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm

are considered unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

4 Leakage rates greater than 5'.0 gpm are considered unacceptable.

(b) Minimum test differential pressure shall not be less that 150 psid.

( g,j }

3/4 4-14c Order dated April 20,1981

y MASE .

5. THEORY OF D'JCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 15 ANSWERS -- BEAVER VALLEY 182 -86/07/22-SILK, D.

ANSWER 5.01 (3.00)

a. Overpredicts (not conservative) (0.5). The source neutron flux dominates the detector reading until the flux level from core multiplication is higher than if the source was further from the detector (0.5).

b. It doesn't (0.5). The critical rod position reflects the positive reactivity necessary to bring the reactor critical and is independent of source magnitude (0.5).

c. The faster the rate, the lower the source range counts at criticality (0.5) due to the reduced time for subcritical multiplication (0.5).

REFERENCE BVPS Reactor Theory Hanual Chapter 5, pp 36,47,49 3.1 001 000 K 5.18 4.3 010 K 5.16 3.5 ANSWER 5.02 (3.50)

a. 1. Delta baron worth becomes less negative (0.25) due to increased competition for neutrons by more boron atoms (0.5).

2. Delta boron worth becomes more negative (0.25) because more neutrons are thermalized due to denser moderator and since boron is a 1/v absorber, the probability of absorption increases (0.5).

3. Delta boron worth becomes less negative (0.25) due to increased competition for neutrons by the poison atoms (0.5).

4. Delta boron worth.becomes more negative (0.25) due to reduced boron concentration from MOL to EOL (0.5).

b. Negative reactivity caused by the buildup of Xe and Sm (0.5).

REFERENCE BVPS Reactor Theory Manual Chapter 8, p 34, 45, 37 3.1 001 000 K 5.20 3.2 K 5.28 3.8 K 5.30 3.1 4

i l

I

I l

1 t, .

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 16 !

ANSWERS -- BEAVER VALLEY 182 -86/07/22-SILK, D.

ANSWER 5.03 (3.00)

e. 1. Decrease (0.25)

2. Increase (0.25)

3. Increase (0.25)

4. Decrease (0.25)

b. Clad failure (melting, burnout) probability is greatly increased because film boiling will reduce the heat being transferred from the fuel (1.0).

c. Coolant outlet temperature is limited (to below saturation temperature)

(0.5). If coolant becomes saturated then there will be no change in RCS hotles temperature and thus no indication of core power (0.5).

REFERENCE BVPS Thermodynamics Manual, Chapter 7, pgs. 14-17, 19 3.4 003 000 K 5.01 3.9 3 2 002 000 K 5.09 4.2 K 5.01 3.4 ANSWER 5.04 (2.00)

a. After the power decrease, the production of xenon from fission (0.25) and from the decay of iodine (0.25) is greater than the removal by decay of xenon (0.25) and burnout by flux (0.25). After five hours, the removal rate is greater than the production (0.25) and positive reactivity is being added entil equilibrium at about 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months (0.25).

b. Rods will need to be withdrawn for about 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months (0.25) and then inserted for the next 35 hours4.050926e-4 days 0.00972 hours 5.787037e-5 weeks 1.33175e-5 months (0.25).

REFERENCE BVPS Rx Theory Manual chapter 7 P3s. 13-16 i 3.1 001 000 K 5.13 4.0 K 5.32 3.5 i

1 l

l i

- I l

l

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 17 ANSWERS -- BEAVER VALLEY 182 -86/07/22-SILK, D.

ANSWER 5.05 (1.50)

a. Decreases (0.5) Pu 239 concentration increases and Pu 239 has a smaller beta. (0.5)

b. Larger SUR (0.5)

REFERENCE BVPS Reactor Theory Manual, Ch 5, ps 14-17 EO 3 001/000 K5.47 2.9/3.4 ps 3.1-3 ANSWER 5.06 (2.50)

a. The speed needs to increase to 1.225 times the original speed to raise the dische 1pe head f rom 1200 to 1800 PSIA.

2 1/2 H1/H2 = (N1/N2) = 1200/1800, H2/N1 = (1800/1200) = 1.225 (0.75)

The horsepower needs to increase to 1.837 times the original horse-power to raise the discharse head from 1200 to 1800 PSIA.

3 3 P2/P1 = (N2/N1) = (1.225) = 1.837 (0.75)

b. NPSH = (Psue.t - Psat)/ density (0.3)

Psat = 205.29 PSIA (0.2) 1/ density = 0.018416 ftxx3/lbm (0.2) 215 ft Ibf/lbm = (P - 205.29 lbf/inux2) (144 inum2/ft**2)

(0.010416 ftxx3/lbm)

P = 286.3 PSIA or 271.6 PSIG (0.3)

REFERENCE BVPS Thermo Manual chapter 4 pss. 14, 21d, 33 Appendix ps. A-9 2.6 3.6 1

s

! So THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 18

_____.c_______

ANSWERS -- BEAVER VALLEY 1&2 -86/07/22-SILK, D.

~

{

1 ANSWER 5.07 (2.00)

The stuck rod would be worth more (0.5). Reactivity worth is proportional to the relative flux squared (0.5). For a dropped rod, the flux is depressed adjacent to it (0.5) whereas if the same rod was stuck out, while the others were inserted, it would be exposed to a much higher flux than the flux in the rest of the core (0.5).

REFERENCE BVPS Rx Theory Manual chapter 8 pgs. 14-16

, 3.1 000 003 EK 1.03 3.8 005 EK 1.05 4.1 ANSWER 5.08 (2.00)

a. Increase

b. Decrease

, c. Decrease

d. Increase

e. Increase (0.4 each)

REFERENCE BVPS Thermo Manual chapter 6 pg. 20 3.5 039 000 A 1.05 3.2 3.2 002 000 K 5.11 4.2 ANSWER 5.09 (2.50)

a. As moderator temperature increases, the migration and thermalization I

len3ths of neutrons in the core increases, therefore more neutrons will migrate to the control rods (0.5) thus increasing their worth (0.5).

b. Reactor power would remain constant (0.5). The negative reactivity inserted by the dropped rod would be countered by positive reactivity inserted by MTC (0.5) since Tave would be lower (0.5).

REFERENCE j BVPS Rx Theory Manual chapter 6 pgs. 16, 20

, chapter 9 pg. 3 a ___________________________________________________________________________

', 3.1 001 000 K 5 10 4.1 i1 l-(

~

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 19 CNSWERS -- BEAVER VALLEY 182 -86/07/22-SILK, D.

K 5.29 3.9 -

000 003 EK 1.16 3.2 ANSWER 5.10 (3.00)

a. Reactor power will increase (0.5) to match secondary power from pos-itive reactivity inserted by MTC (0.5) which will be countered by negative reactivity from power defect as power increases (0.5).

b. 0=UA (Tave - Tstm) 70% = U A (568 - 521.3) = 70/85 U A (Tave - 518.7) (0.5)

{521.3 and 518.7 are from the steam tables for their corresponding pressures in Figure 3} (0.5) lave = 570.7 F (0.5)

REFERENCE BVPS Thermo Manual chapter 7 pgs. 1-4 3 5 039 000 K 5.08 3.6 A 2.05 3.3 l

l i.

l' l

l l

_t _

s.

6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 20 ANSWERS -- BEAVER VALLEY 182 -86/07/22-SILK, D.

ANSWER 6.01 (2.00) -

a. The ratio of gamma to neutron flux is greater (0.5)

6. PR comparitor deviation NIS PR high,setpoint rod stop block rod w/D HIS PR high setpoint neutron flux high NIS PR neutron flux rate high PR low setpoint flux deviation or auto defeat Computer alarm rod deviation / SEQ NIS PR Tilts (5 of 6, 0.3 each)

REFERENCE BVPS OM 2.1 rg. 16 2.2 pg. 7 NS-8 figures 8, 9 3.9 015 000 K 3.01 4.3 K 6.02 2.9 LP-SGS-2.1 3,43 ANSWER 6.02 (2.50)

' ped ba Allow for the decay of N-16 (0.4)

b. Steam generator blowdown sample monitor AFWP turbine exhaust monitor Main steam safety valve effluent monitor Steam generator blowdown tank discharge monitor (0.4 each)

c. The condenser air ejector discharge will be diverted to the containment (0.5).

REFERENCE bVPS OM 43.1 pgs. 9, 15, 21 AOP-42 pgs. 1, 2 Appendix pg. A-6 2.7

- 3.3 000 037 EK 2.02 2.4 3.3 000 037 EK 3.10 3.7 3.9 073 000 K 1.01 3.9 2336 RCS 6 2353 MSS 10 l

l

6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 21 ANSWERS -- BEAVER VALLEY 112 -86/07/22-SILK, D.

ANSWER 6.03 (2.20) pgered bpfb4' alcre W 5-P d% (.q)

a. Slau indi :ter 7:r:1101 t; thw l e e!: uill indic t; ; 1:wcr then ..wi--

e:1 +1;u :nd :tn:rrelly high c o r r e n er- t trererete ec. OP ef'erted ccerraent tr:peretw .ee usuld be much lo.ei then ncr;;l ;f th: 1:;k erc d;un;t eee ef thw su=yuneui w. auch h;3 her th n nw. m.1 ;f the leak us- er tre : cf the ;;p;nent '0.7).

b. High flow will cause RCP thermal barrier CCR outlet valves to close Pressure buildup will seat check valve Piping between valves is designed for 2485 psis (0.5 each)

REFERENCE AOP-20 P3+ 1 BVPS OH 15.1 pg. 16 3.10 008 000 K 3.01 3.5 3.3 000 009 EK 3.15 3.2 LP-SOS-6.3 1 ANSWER 6.04 (2.00)

a. To ensure required adequate back pressure in the RCP seals (0.5)

b. Key interlock will only allow one breaker to be racked in at a time (0.5)

c. For load follow and permits the dilution of water to follow the initial xenon transient (0.5) but using it adds large amounts of non-hydrogenated water to the RCS (0.5).

REFERENCE BVPS OM 7.1 ps. 38 7.2 pgs 1, 3 3.1 004 000 K 1.06 3.1 K 2.03 3.5 K 5.01 3.3 3.2 002 000 K 1.06 4.0 LP-SOS-7.1 3,5 i

I l

6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 22 CNSWERS -- BEAVER VALLEY 182 -86/07/22-SILK, D.

i ANSWER b 6.05 (2.00) -

Control rods will automatically move outward (0.4) due.to temperature error and power mismatch error durin3 the transient (0.4). With a small MTC, reactor power will rise (0.4) causing a (C-2) overpower rod stop (0.4) and power overshoot results in an OTdT or OPdT trip despite Doppler feedback (0.4).

REFERENCE BVPS NS-10 pss 3 to 12 NS-8 P3 39 3.1 001 000 A 1.02 3.4 3.5 045 010 K 4.21 3.2 3.9 012 000 K 4.02 4.3 2352 MSS 3 ANSWER 6.06 (2.70) pc y .cp f a+ f i

a. Letdown pressure control valve ("CU C;; 142) (0.4)

6. Will not open at RCS pressure > 430 psis Will auto close at RCS pressure > 630 psis Will not open if pressuri=er vapor temperature > 475 F (0.5 each)

c. Manually placing two keylock switches in their automatic position (Enables the PORVs' low pressure setpoint) (0.4)

d. The backup supply are two nitrogen filled accumulators (0.4)

REFERENCE

. BVPS OM 10.1 pgs. 2, 15 10.2 P3s. 6, 7

l 6.1 pgs. 52, 53 3.2 006 000 K 4.08 3.5 3.4 005 000 K 4.01 32 ; K 4.07 3.5

3.3 010 000 K 4.03 4.1 j 3.8 078 000 K 3.02 3.6 LP-SOS-10.1 4 RCS PZR Pressure relief system 7 i

e t

'f I

I l.

l l

6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 23 ANSWERS -- BEAVER VALLEY 182 -86/07/22-SILK, D.

ANSWER 6.07 (2.50) -

a. Ensures that secondary system pressure will be limited to within its design pressure during the most severe transient (0.5).

b. 1. Minimize positive reactivity effects of RCS cooldown associated with the blowdown (0.5)

2. Limit pressure rise within containment during a steam break in containment (0.5)

c. Hot Zero Power (.25) because of the greatest mass in the SG results in the largest RCS cooldown (.25)

EOL (.25) because MTC is at its maximum negative value (.25)

REFERENCE T/S B 3/4 7-1 T/S B 3/4 7-3 FSAR 14.1-35 to 38 3.5 000 040 EK 3.01 4.5 EK 2.01 2.5 i EK 1.05 4.4

, 3.5 039 000 K 4.05 3.7 Objectives PGS-10-17 ANSWER 6.08 (2.30)

a. See sketch (1.5)

b. To preserve power to the Diesel Generator Auto Loading Sequence Circuits (0.4)

c. Permissive signal from associated undervoltage devices (0.4)

REFERENCE BVPS OM 37.1 pgs 78, 79 BV exam Bank Question 6-4 a 3.7 062 000 K 4.09 2.9 K 4.03 3.1 LP-SOS-36.1 2,7

i 4160v Sus lAE I) wD yyv/qny (.l5) .

mm 480 v 3us 8N (. I 7) i I MCC- MCC- ( gg)

E13 l

[15) E9 l 1

1) I)

]) l l l

\

Inverter I Static Line Battery #1 Voltage Charger Regulator

1 I

[,gg)

[ 15) i

(.1 T) )

\ n

! n -

[ 3sttory Il1, -

il l """=

(.lS) .

l 125v DC SWBD #1 Vital 3us #1 [.lf)

(15) f

60 PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 24 ANSWERS -- BEAVER VALLEY 112 -86/07/22-SILK, D.

ANSWER 6.09 (3.20) -

a. Associated quench spray pump running (0.4)

RWST low low level (0.4)

b. Reduces quench spray flow to minimize negative pressure when con-tainment returns to subatmospheric pressure following a LOCA (0.8)

c. Only out leakage can occur and dilution of borated water by river water in containment is not possible which ensures necessary shut-down margin (0.8)

d. M: '0.5). CIO init :t F>rhb'>tte^= ="*+ ha daara**=d h"#cre the W il eviomaT,1cally r e , L., t ;0.5;.

REFERENCE ye'> fo .1) ci 6 myx/ wi // ad ^dc $W BVPS OM 13.1 P3s. 2, 8, 12, 19 queach srmy (aaPS r 3.6 103 000 K 1.08 3.8 026 000 K 4.04 4.1

026 000 K 1.02 4.1

! 026 020 K 4.03 4.3 LP-SOS-13.1 3,4,5 4

t l

t i

i i

60 PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 25 l

ANSWERS -- BEAVER VALLEY 112 -86/07/22-SILK, D.

4 ANSWER 6.10 (3.60) -

s. Higher pressure would tend to increase the amount of accumulator water carried out the break (0.4). Lower pressure results in less rapid delivery of accumulator water to the reactor tending to delay core recovery (0.4).

b. SI signal (0.2) and 2/4 low level on RWST (0.2)

c. 1. Containment sump to LHSIP's suction valves (SI-860A,B) open

2. LHSIP's miniflow isolation valves (SI-855A,B,C,D,) close

3. LHSIP's discharge valves to HHSIP's suction (SI-863A,B) open

4. HHSIP's suction from RWST (CH-115B,0) closed (0.3 for valves

5. LHSIP's suction from RWST (SI-862A,B) closed 0.5 for order)

d. Manually align an outside recirculation spray pump (0.4)

REFERENCE BVPS OM 11.1 pgs. 3, 7; Fig 11-12 i BV Exam Bank Question 6-11 a,b BVPS NS-13 pgs. 9-13 3.2 006 000 K 6.02 3.9 K 4.06 4.2 3.4 005 000 K 3.05 3.8 LP-SOS-11.1 3,6 l

s i

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 26

~~~~ - ----------------~~~-----

RA656L55 feat 56ETR5L ANSWERS -- BEAVER VALLEY 1&2 -86/07/22-SILK, D.

ANSWER 7.01 (1.70)

a. Decay Tanks sampled and analyzed (0.5)

Independent verification of release rate calculation and valve line-up (0.5)

b. Activity should decrease as soon as water in the pipes from the dis-charge of the pumps is purged W44 . -If .:t; t;.... i,; i = rel asa ;h:;1d be rt:pp d 'O.0). ('O,7)

REFERENCE TS P3 3/4 3-63 BVPS OM 19.4 pg. 9 17.4 pg. 67 3.11 068 000 Sys sen 4 3.3 1 071 000 Sys sen 5 4.0 ANSWER 7.02 (2.50)

+

a. While performing E-0 if total AFW flow < 350 spm (0.2) all Heat Sink Red Path - SG narrow range level in 10:;t c r. ej SG < 5% (.15)

- Total feedflow to SG's < 350 GPM ~(.15)

b. Wide range level in two SG's < 10% (0.4)

Pressurizer pressure greater than 2335 psis (0.4)

c. RVLIS full ranSe indication [>61%) (0.4)

At least two RCS hot les temperatures $ 350 Fj (0.4)

Five hottest exit TC's (< 1200 F) (0.4)

REFERENCE BVPS EDP FR-C.1 P3+ 11 FR-H.1 pgs. 1, 2 3.4 000 074 EK 3.11 4.4 3.5 000 054 EK 3.04 4.6 a

LP-SOS-53A-FR-H EO 2 LP-SGS-53A-FR-S E0 2 i

h 9

J

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 27

~~~~ -

RA656L665 CAL 66NTR6L'~~~~~~~~~~~~~~~~~~~~~~~

ANSWERS -- BEAVER VALLEY 1&2 -86/07/22-SILK, D.

ANSWER 7.03 (2.50)

a. (1) Close main steam trip valves '0.0) dualwck S furf;ne (2) Close non-return valves (+rM cyg%g g f

b. (1) Pressuri: er pressure < 1845 PSIG (2) Containment pressure > 1.5 PSIG.

(3) Steamline pressure < 510 PSIG (0.2 each)

c. (1) RCS subcooling criteria met (0.23)

(2) Feed flow to intact SG's(> 350 GPM)or Narrow RanSe level in at least one intact SG (0.23)

(3) RCS pressure - table or incr easing) (0.22)

(4) PZR level (> 5% (0.22)

d. Reactor trip breakers must be closed (0.4)

REFERENCE BVPS E0P E-0 pgs. 3, 5, 14, 17

' 4.6 3.1 000 007 EK-3.01 LP-SOS-53A-E-0 E0 1,3 ANSWER 7.04 (2.50) i

a. Letters denote sequencial importance, bullets do not (0.5)

b. As directed in E-0 (0.5)

When transferring out of E-0 (0.5)

! c. 2,3,1,4 (1.0)

REFERENCE E0P Ex Vol pss. 3,6,8 SWPWGKA 22 4.3 LP-SGS-53A-Intro E0 lb, 2a e

4

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 28

~~~~ ---~~-------------------

RA656LU55CAE~C6NTR5t ANSWERS -- BEAVER VALLEY 1&2 -86/07/22-SILK, D.

ANSWER 7.05 (2.00)

c. 5(N-18) = 50 REM (0.2)

Total lifetime to date = 48 + 1 = 49 REM (0.2)

Total lifetime available = 50 - 49 = 1 REM (0.2)

Total this quarter available = 3-1 = 2 REM (0.2)

Lifetime is more restrictive than quarterly limit 0.85 REM /HR 3amma + 0.30 REM /HR neutron =1.15 REM /HR dose rate 1.0 REM /1.15 REM /HR = 0.87 HRS = 52 MIN (0.4)

b. 25 REM whole body one time exposure (0.4)

Emergency Director (0.4)

REFERENCE 10 CFR 20.4; 101 BVPS RCH ps. 9 System wide and plant wide generic KRA (SWPGKaA) 10 3.9 ANSWER 7.06 (2.50)

Dek/ch9 The a tual Pressurizer water level is less than 60% or The secondary water temperature of each SG is less than 25 F above each of the in-service RCS cold less temperature (0.5 for either)

b. Return bank to 500 pcm below ECP and recalculate ECP (0.5)

c. Ensures that:

1. The moderator temperature coefficient is within its analyzed temperature ranse

2. The protective instrumentation is within its normal operating range

3. The pressurizer is capable of being in an operable status with a steam bubble

4. The reactor vessel is above its minimum NDTT temperature (0.3 each)

REFERENCE BVPS OM 50.4 ps. 30& 10 TS ps. B 3/4 1-2 TS 3.4.1.6, BVPS OM 6.4 pg. 3 3 4 000 000 Sys Gen 5 3.9

' 4.2 3.1 001 010 A 2.07 i LP-SGS-6.3 5 c.

)/ i

-a ..

7. PROCEDURES - NORMAL, ABHORMAL, EMERGENCY AND PAGE 29

~~~~

R bibLbb5bkb~bbhikbL'~~~~~~~~~~~~~~~~~~~~~~~

q ____________________

ANSWERS -- BEAVER VALLEY 1&2 -86/07/22-SILK, D.

2336 RCS 8 -

ANSWER 7.07 (2.00)

a. Abnormal pressurizer pressure and level responses to charging and spraying (0.5)

Indication of departure from subcooled conditions (0.5)

b. If the potential for interruption of core cooling with hydrogen in the vessel exists (0.5)

c. If the pressurizer bubble is interferring with the ability to maintain pressure control (0.5)

REFERENCE BVPS OM 6.4 pgs. 121, 126 3.6 028 000 K 5.01 3.9 3.3 000 009 EA 2.01 4.8 EA 2.38 4.3 LP-SGS-6.9 5 ANSWER 7.08 (3.30)

a. 1. Prevent excessive inventory loss (0.3)

2. Preclude core uncovery from RCP's tripping at a later time (0.3)

6. 1. Highest RCS SG D/P < 145 PSI (0.4)

2. No CCR Flow to RCP's (0.4)

c. 1. Unexpected increase in S/G narrow range level.

2. High S/G sample radiation.

3. High S/G steamline radiation.

4. High S/G blowdown line radiation. (0.25 each)

d. Containment pressure > 5 psis or containment radiation > 100000 R/hr or integrated containment radiation > 1000000 R (0.9) l REFERENCE BVPS Exec Vol E-0 step 23 pg. 33 BVPS EOP E-3 pgs. 2, 3

+

E-0 Attachment 6 3.3 000 038 EK 3.06 4.5 1

l l

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 30

~~~~~~~~~~~~~~~~~~~~~~~~

~~~~R5656E66 65E~C6ATR6L ANSWERS -- BEAVER VALLEY 1&2 -86/07/22-SILK, D.

LP-SOS-53A-E-1 EO 3 -

LP-SOS-53A-E-3 E0 1 ANSWER 7.09 (3.00)

a. Containment purge exhaust and supply dampers close Containment purse to exhaust fan damper closes Main filter bank bypass dampers close Main filter bank inlet dampers open (0.4 each)

b. Ensure that a single failure will not result in a loss of heat removal capability (0.6)

c. Maybestopped(for one hour per eight hour period)for core alterations in the vicinity of the hot legs (0.8)

REFERENCE AOP-22 Ps* 1 TS pgs. B 3/4 9-2, 3/4 9-8 3.4 000 025 Sys sen 5 3.9 3.11 034 000 K 3.01 2.9 LP-SOS-10.1 9 ANSWER 7.10 (3.00)

a. Plant heatup, plant cooldown, abnormal pressure / temperature transients ocarcJ shedi (0.33 each)

b. Preclude potential high airborne and increased radiation levels in the auxiliary building (0,5)

=

c. Xenon and iodine (0.25 each) or Eryffon l

d. C c st;cn

.u__- -

--"_enati~;r

____ u,33

'vp to 40 liovi a l Fr:vided th:t Op:r tisii under-

,m+ ...___s 4 n .f m,

+"- Jait'= Lutel yearly-CP e r O t i i n j tile 1 1.O) OfMCO M Q (ceT[7nge p( M h&/ fIld U(h y dCf REFERENCE no Y hte(f %c lumi{ cn +hc' Tec t s yc G rLe AOP-43 pgs.

TS pg. 3/4 4-18 1, 2 (7 $ f, pre g,q. ,) y AN idi. < d h i' i/* -' & "

. __________________________________________a4xlCW_12C______________________

3.11 000 076 EK 3.01 3.1 EK 3.05 3.6 i

Sys sen 5 3.6 2336 RCS 5,8

{

f

s

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 31 ANSWERS -- REAVER VALLEY 112 -86/07/22-SILK, D.

ANSWER 8.01 (1.50) -

a. Interval requirement not exceeded CO.253. Eight days does not exceed 1.25 times the specified interval CO.53.

b. Interval requirement exceeded CO.253. The last 3 consecutive intervals exceed 3.25 times the specified interval CO.53.

REFERENCE TS P3 3/4 0-2 SWPWGKRA 5 3.9 LP-SOS-11.1 7 ANSWER 8.02 (1.00)

The Fire Brigade shall not include three members of the minimum shift crew necessary for the safe shutdown of the Unit or any personnel required for other essential functions during a fire emergency (1.0).

1- REFERENCE l SAP pg. 8 TS pg. 6-1 SWPWGKA 19 4.2 ANSWER 8.03 (2.50)

Technical Specifications require that all LCO's be satisfied prior to (The Mode entry ' ntothe 3 (.

an heatup oper ational must bemode.) { 0.7C b Sing discontinued (

)you and are Taveabout held to atenter less than 350 F until Charging Pump 1B is proven operable (4 ,61 I

REFERENCE TS pg. 3/4 0-1; TS pg. 3/4 5-3 3.4 005 000 Sys sen 5 4.0 i ___________________________________________________________________________

, LP-SGS-7.1 9 i

6 4

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 32 ANSWERS -- BEAVER VALLEY 1&2 -86/07/22-SILK, D.

ANSWER 8.04 (2.50) -

Gn7 60 Modes 1&2-- Be in HSB with pressure within limits in one hour.

61) G L)

(.75)

(,ts) G egs ( p s, Modes 3,4,5-- Reduce pressure to within limit in 5 minutes. (. 5) 6w C 3> <$)

All Modes-- Notify the NRC, Manager of Nuclear Operation, and 0 C (within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ). (1.0)

REFERENCE TS pgs. 2-1, 6-12 SWPWGKRA 5 3.9 2336 RCS 8 ANSWER 8.05 (2.50)

LCO's indicate lowest performance level of equipment required for safe operation of the facility (0.5). If proper automatic action occurs prior to reaching Limiting Safety System Settings, then Safety Limits will not be exceeded (1.0). If Safety Limits are not exceeded then fuel and RCS inte3rity will be maintained (1.0).

REFERENCE TS pgs. B 2-2,3 10 CFR 50.36 e i _ _ _ - _ - _ - _ _ - _ _ _ - _ _ _ - - - _ _ _ _ _ _ _ _ - _ _ - - _ - _ _ - - _ _ _ _ - _ _ - - - - _ _ _ - - - _ _ _ - _ _ - _ _ - - _ - _ _ _ _ -

SWPWGKAA 5 3.9

{

ANSWER 8.06 (2.00)

)

2. You may operate for up to two hours with one less than_ minimum 3

complement (0.75) provided that immediate action is taken to bring the complement up to minimum (0.5).

L b. The on-shift STA will have to wait for a relief to come in (0.75).

REFERENCE i; TS pg. 6-4 1 1 _ _ _ - _ _ - _ _ - - - _ _ _ _ _ _ _ _ _ - - - _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ - -

SWPWGKA 23 3.5 i.

+

i s i

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 33 ANSWERS -- BEAVER VALLEY 112 -86/07/22-SILK, D.

2 j ANSWER 8.07 (2 00) _

RCS Pressure Isolation Valve Limits exceeded. (1.0)

UNIDENTIFIED Leakage limits exceeded. (1.0)

I REFERENCE i TS 3.4.6.21 TS 3.4.6.3 i

3.2 002 020 Sys sen 5 4.1 2336 RCS 8 i

ANSWER 8.08 (2.50) i All penetrations required to be closed during accident conditions are

, either:

I Capable of being closed by an operable containment auto-isolation

' valve system (0.5), or Closed by manual valves, or blind flanges (0.5)

All equipment hatches are closed and sealed (0.5)

Both doors in each personnel air lock are properly closed unless being l

used at which time at least one air lock door shall be closed (0.5) and air lock leakage is within limits (0.5)

The containment leakage rates are within limit (0.5)

,! REFERENCE TS 1-2 I

SWPWG K&A 5 3.9 s

LP-SGS-12.1 5 f,

4 l

I 4

l I,

l

a

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 34 ANSWERS -- BEAVER VALLEY 182 -86/07/22-SILK, D.

ANSWER 8.09 (2.00) -

O. Shift Supervisor (0.5)

b. In an emergency when this action is immediately needed to protect the public health and safety and no action consistent with the license conditions and Tech Specs that can provide adequate or equivalent protection is immediately apparent. (1.5)

REFERENCE SAP P3+ 49 EPP pg. 5-3 SWPWGKRA 36 4.7 ANSWER 8.10 (3.50)

a. 2 (0.5)

b. Within 15 minutes (0.2)

1. Restore the indicated AFD to within the target band (0.4), or

2. Reduce the thermal power to <90% of rated thermal power. (0.4)

c. Accumulated penalty over the past 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is 89 minutes. (1.0)

The penalty will be reduced to 60 minutes at 1618 minutes on 05/13/85 and then power may be increased. (1.0) 85% 0318-0310 = 8 (0.25) 65% 1637-1557 = 40 (0.25) 45% 0310-0148 -

82/2 = 41 (0.5) j 89 min. total penalty 05/13/85, from 1557; 81 min left 21 min -> 1618 05/13/85 (1.0) l I

REFERENCE TS 3.2.1; TS pg. B 3/4 2-2 39 015 020 Sys sen 5 3.9 LP-SGS-2.1 3,7 l

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 35 ANSWERS -- BEAVER VALLEY 112 -86/07/22-SILK, D.

ANSWER 8.11 (3.00) - l

c. Review plant status by inspection of control room instrumentation Review entries in loss Conduct a briefing with the off-goins NSOF using Shift Relief Turn-over Checklist (0.33 each)

b. The Nuclear Shift Supervisor (0.33) expresses permission via the Equipment Clearance permit (0.33) and Maintenance Work Request (0.33)

c. No (0.5). Technical evaluation has to be done with OSC concurrence prior to installation (0.5).

REFERENCE SAP pgs. 12, 27, 23 SWPWGKRA 14 4.0 l

t l'

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SENSITIVEINF0 MAIL mi 0N U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: BEAVER VALLEY 1&2 REACTOR TYPE: PWR-WEC3 DATE ADMINISTERED: 86/07/22 EXAMINER: BARBER, S. l APPLICANT: ___

k__________________ ,

IN5TRUCTIONS TO AFPLICANT:

Use separate paper for the answers. Write answers on one side only.

Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% i r, each category and a final grade of at leest 80%. Examination papers will be picked up six (6) hours after the e"aninetion stetts.

/. OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY g-

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,5,,v.

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, i

HEAT TRANSFER AND FLUID FLOW

-e -e o o PLANT DESIGN INCLUDING SAFETf

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___1_n ___________ ________

2.

AND EMERGENCY SYSTEMS 25.00 25.00 3. INST JMENTS AND CONTROLS g g -- --_-__

T ' 25.00 4 PROCEDURES - NORMALr ABNORMAL.

EMERGENCY AND RADIOLOGICAL CONTRCL 9'7.oo-TOTA _5

.0' 100.00

'INAL GEADE __-___________-.I-A1; war- done on this eaan:.ction is u own. I M:ve ieither giver- r. :i - re: t : v r. o siu.

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lo- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 2

--- isisR657sisics- REKi isAssFEs As5 FEUi5 FE5E GUESTION 1.01 .( 2. 50 ) '

ao Tor an operator taking data for a 1/M plot, how will the Shut-down margin (SDM) affect the time elapsed before a stable count rate can'be obtained after withdrawing rods ? (0.75)

b. How will-the' initial count rate affect the count r ate at crit-icality ? (0.75)

c. If theJspeed of the control rods'were to somehow increase. What would be the effect be on' ,

1. Rod height at criticality ? (0.5)

2. Count rate at criticality ? (0 5)

GUESTION 1.02 (2.00)

a. Does Beta bar effective increase, decrease, or remain the same from BOL to EOL? Explain your answer. (1.5)

b. For two equivalent positive reactivity additions to a critical reactor, will the SUR be the same, larger, or smaller at EOL as compared to BOL? No explanation is necessary. (0.5)

GUESTION 1.03 (2.00)

The reactor is at 100% power at EOL. Rods are at 220 steps on Bank D.

Boron concentration is 300 ppm. Power must be-reduced to 75%. If rods

! will be inserted to 129 steps on Bank D, what will be the final boron con-centration at equilibrium conditions at 75% power 0 Use the Figures 50-6 50-7, 50-8, 50-10 attached. Show all work and state all sssumptions.

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

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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 3

--- isEER557sisics- sEAi iEAssFEs As5 FEUi5 FE5s QUESTION 1.04 (3.00)

Compare the Actual Critical Position (ACP) for a startup to be performed 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after a trip from 100% power. to the Estimated Critical Position (ECP) if the following events / conditions exist. Consider each separately and independently. Indicate whether the ACP is HIGHER than, LOWER than or the SAME as the estima+ed critical control rod position. Briefly explain-the reason for each of your answers.

a. The THIRD coolant pump is started two minutes prior to criticality,

b. The startup is delayed until 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after the trip.

c. The steam-dump Pressure setpoint is increased to a value just below the Steam Generator Atmospheric Dump (PORV) setpoint.

d. An entire string of.feedwater heaters is taken out of service just prior to criticality.

QUESTION 1.05 (2.50)

A clean core is started up and taken to 50 % powe'r, where it re-mains for 30 days'

a. Describe the reactivity changes the operator must compensate for due to fission product poisons. (1.5)

b. After 30 days power is increased to 100%. Explain any further reactivity changes required. tSpecific reactivity values are NOT reovired) (1.0)

NOTE! Indicate appronimate time and duration of reactivity changes.

GUESTION 1 06 (3.00)

How will the following paramater changes affect control rod worth? Explain.

A.Tave increases 5 degrees c (1.0)

B.Boren Concentration increases 50 PPM (1.0)

_, m .. . 7 _ _ _ _ . - ; #.

DELEED ..-

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

10 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 4

~~~~

Ts5Rs55YsAsics? SEAT T EAssrER~ 96'iLU 6~FEUs QUESTION 1.07 (2.00)

a. Why is the limit for the overtemperature Delta T trip based on not reaching saturation conditions in the hot less?

b. Refer to figure 2.1-1 attached.

Operation within the limits of the 2250 psia curve from "90%

power- ~622-F Tave to "120% power- 590-F Tave will prevent exceedir4s what specific minimum plant thermal criteria?

Include any applicable setpoints.

QUESTION 1.08 (2.50)

Assume one RCF trips at 30% power without a reactor protection system actuation or a change in turbine load. Briefly discuss HOW and WHY each of the following parameters will change.

a. Flow in the opereting Reactor Coolant Systemi loops. (0.5)

b. The ratio of core flow compared to the total loop flow. (0.5)

(Core flow / Total loop flow)

c. Reactor vessel delta-P. (0.5)

d. Actual Core delta-T. (0.5)

e. Steam temperature in a steam generator in an operating loop. (0.5)

GUESTION 1.09 (2.50)

Briefly explain how and why fuel centerline temperature is affect-ed by the fc11owing:

a. Fuel densification. (1.0)

b. Interstitial absorb 1on of fission gasses in the fuel pellets. (0.75)

c. Clad creep. (0.75)

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

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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 5

~~~~IEEE56DYU55 C5I~HEdY~iRdUEF5R~ U6~ FLU 56~FLBE QUESTION 1.10 (3.00)

a. A variable speed centrifugal pump is operating at 1/4 rated speed in a CLOSED system with the following parameters:

Power = 300 KW Pump delta P = 50 psid Flow = 880 spm What are the new values for these parameters when the pump speed is increased to full rated speed? (1.5)

o. Choose the answer that most correctly completes the sentence. (0.5)

"In a CLOSED system, two single stage centrifugal pumps operating in parallel will have--(choose-from-below)--. as compared to the same system with one single stage centrifugal pump operating wi,h one pump isolated.'

i. double the head and double the flow rate.

2. the same head and the same flow rate.

3. the same head and double the flow rate.

4 double the head and the same flow rate.

c. How does the available NPSH' change when system flowrate is increased ? (0.5)

d. Whv is cavitation undesirable? (0.5)

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

e- . . -

~2. ' PLANT DESIGN-INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 6 GUESTION 2.01- (2.00)

Match each RCS system penetration in Column A to its correct location in Column B.

COLUMN A COLUMN B

a. Excess' Letdown 1. Loop A Hot Les

2. Loop C Hot Les

b. Normal Letdown 3. Loop A Cold Leg.

4. Loop B Cold Les

c. RHR Cooldown Suction 5. Loop C Cold Les

6. Loop B Hot Les

d. P:t Surge Line

e. Normal Charging GUESTION 2.02 (2.50)

a. Explain why the flowrate through the RCP 41 seal is.not constant for all plant conditions. (1.0)

6. How is the ti seal return flowpath affected by a Containment (entmt) phase

A

isolation ' -(1.0)

c. What-determines the differential pressure across the RCP ti seal? (0.5)

QUESTION 2.03 (2.50)

3. How is the interlock associated with 4160 Bus 1A supply breakers ACB 41A&C bypassed so that a live bus transfer can be conducted? (0.5)

b. What plant condition will cause automatic closure of the Emergency Diesel Generator output breaker ? (0.5)

c. What are the thr ee 4160 volt loads (motors) ' capable of being supplied by either safeguards bus 1AE or 1DF? (1.5)

(****.* CATEGCRY 02 CONTINUED ON NEXT PAGE *****)

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2. PLANT DESIGN INCLUDING ~ SAFETY AND EMERGENCY SYSTEMS PAGE 7 QUESTION' 2.04 (2.50)

a. Describe the opening interlocks associated with the RHR inlet isolation valve-(MOV-1RH-701).Ine10de setpoints where applicable. (1.0)

b. State.the TWO conditions that must'be satisfied for Automatic Recirculation Sumo valv opening for old Les Recirculation.. (1.0)

ASWHE94L.E Sm MA SeeB How does RCS pressure erange if a RC is started when RHR is.

c.

controlling plant pressure and temperature under solid conditions ?

Assume a plant cooldown has just been completed using RHR from 250 F to 150 F. (0.5) i OUESTION- 2.05 (1.50)

Indicate whether the following statements concerning the Feedwater and Condensate system are TRUE or FALSE.

a. At 300 psis feed pump suction pressure (decreasing), a condensate

{)tWHd(b:::tr- pump-will_ auto start, if available.

~b. 70 % SG- level will trip a feedwater pump / turbine.

c. If.a hain Feedwater Pomp's recirculation valve doesn't open shutdown the pump.

GUESTION 2.06 (2 50)

a. State the. rated flow of each type of auxiliary feedwater pump below:

motor driven.

turbine driven. (1.0)

b. State FOUR conditions / signals that will automatically start-the MOTOR driven auxiliary feedwater pump. (Include coincidences) (1.0)

c. Why does the turbine driven auxiliarv feedwater pump's recirculation valve open at 255 gPm ' (0.5)

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

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2.- PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 8 QUESTION' 2.07 (3.00) ao -List 3 sources of hydrogen found in containment post-LOCA (1.5)-

bo .The containment air hydrogen concentration must be kept below percent by volume to prevent the possibility of an explosion in containment. (provide value) (0.5)

c. What two systems provide for hydrogen removal post-LOCA ? (1.0)

QUESTION ~2.08 (3.00)

For the following components, indicate whether they will receive an OPEN CLOSER or NO signal upon safety injection initiation due to low RCS pressure.

a. Control room supply and exhaust ducts

b. Main feed bypass valves

c. SI accumulator discharge isolation valves

d. Normal charging containment isolation valve

e. Main steam isolation valves

f. . RWST to charging Pump svetion valves 9 Baron recire to BIT isolation valves

- h '. VCT outlet isolation valves

.i. Component cooling isolation from letdown heat exchanger

j. Steam supply valves to turbine-driven feed pump

.GUESTION 2.09 (3.00)

a. Provide THREE Component Cooling (CCW) system alarms that could indicate a RCS to CCW leak.

b. Describer in detail, how the CCW system is protected against an overpressure condition if a RCS to CCW rupture occurred in the RCS Thermal Osrrier.

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

20 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 9 OUESTION 2.10 (2.50)

If a Recirculation Spray Heat E:: changer tube rupture were to occur dur:ng operation with a large break. LOCA in progress, what would be'

a. The indication to alert the operator of this occurence? (0.5)

b. The consequences of no operator action? (1.0)

c. Hcw would the operator recover from this event? (Give the basic steps that are necessary - valve and instrument numbers, etc. not required.) (1.0)

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

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

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= 3a INSTRUMENTS AND CONTROLS PAGE. 10

~ QUESTION 3.01 (3.00) Aswet. 2 i$ *

d " I The plant is operating at 80% power when agThat RTD fails high. Briefly EXPLAIN how this failure will affect the following. Consider-each item independently. Assume no operator action and all control systems are in automatic.

'a. Rod insertion limit setpoint (0.75)

b. Charging flow (initially) (0.75) l j c. Control rod bank position (0.75) i d. Steam dump control system (0.75)

GUESTION 3.02 (3.00)

Briefly describe any AUTOMATIC actions and slarms associated with the following Process Radiation Monitoring System channels.

a. Condenser Air Ejector. Vent Monitor (RM-1SV-100) (0.8)

T b. Gaseous Waste Particulate Monitor (RM-1GW-108A) (0.8)

c. Containment Purge Exhaust Monitor (RM-1VS-104A,B) (1.4) j during REFUELING operations.

i l . QUESTION 3.03 (3.50)

a. How might overcompensation of one intermediate range excore detector be recognized on the startup RATE meters after a reactor trip ? Consider response immediately after the trip and 5-10 minutes after the trip. (0.75) s

6. Intermediate range channel N35 is reading 10 -11 amps while channel N36 is reading 10 -10 amps. Using the Source Range, how would you verify which channel is reading correctly? -(0.75) 4 l c. 1. Explain the effect of an adjustment of the ' summing and

' level amp gain adjust pot' on the N50 current comparator .

and the power range (PR) level indication. (0.75) i

2. Why should caution 'oe used when adjusting this pot while et power? (0.75)

3. When is this pot adjusted while at power? (0,5)

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

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3. INSTRUMENTS AND-CONTROLS PAGE 11 j QUESTION 3.04 (2.00)

. Indicate whether the following situations will ARM ONLY, ARM AND ACTUATE or HAVE NO EFFECT on the steam dump system.

a) .50% power, 18% step load increase, Tavs is 6 F~less than Tref, steam dumps are in the Tavs mode of operation 1

1 b) 80% power, 7.5%/ min-ramp decrease in turbine load for 3 minutes, 1

Tavs is 7 greater than Tref, steam dumps are in the Tav3 mode-of operation j c) Hot Zero Power, Tavg=549 F, steam dumps are in the STM PRESS mode t with 985 psis set into the steam pressure controller d) Reactor trip, Tavg=549 degrees, steam dumps in Tavs mode 00ESTION 3.05 (2.50) ,

a. List two control rod interlocks that will prohibit rod withdraw-al in automatic only AND e:tplain the necessity for these inter-locks. (1.0)

b. List and explain the 5 functions of (or automatic actions resulting i from) an urgent failure in the Rod Control System.- (1.5) i j 00ESTION 3.06 (1.50)

TRUE or FALSE l s. The interceptor and govenor valves are the ONLY valves closed by the Overspeed Protection Controller.

l b. The throttle and reheat stop valves are th ONLY valves closed by the Overspeed Trip Mechanism.

t l .c . The reheat stop and intercepter valves are opened when the j turbine is latched.

I' OUESTION 3.07 (3.00)

List all the automatic signals tnat will cause a safety injection.

Include setpoints and coincidence / logic. Do not include manual SI.

(*r**m CATECORY 02 CONTINUED ON NEXT PAGE *****)

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30 INSTRUMENTS AND CONTROLS PAGE 12 QUESTION 3.08 (2.00)

Indicate whether the Over Temperature Delta Temperature (OT-DeltaT) trip setpoint will INCREASE, DECREASE, or REMAIN THE SAME for the following parameter changes. Consider each separately.

a. Increasing Tavg.

b. Tavs decreasing to less the than rated full power Tavg.

c. Delta I becoming more negative.

d. Pressort:er Pressure increasing.

GUESTION 3.09 (2.50)

a. Provide TWO additional (different/ separate) AUTOMATIC signals other than High-High S/G water level, which will generate a d feedwater isloation signal. (Setpoints are not required) (1.0)

b. List ALL the additional automatic actions associated with High-High SG water level, other than feedwater isolation valve closure. (MOV-FW-156 A,B,C) (1.5)

QUESTION 3.10 (2.00)

Plant load is 50% and the Chemical and Volume Control System (CVCS) in a normal lineup and controlling pressurizer level fails high. Assuming no operator action, state the sequence of events that would occur until stable plant conditions are reached or until the reactor trips. Include 3PP licable setpoints.

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

40 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 13

- -~~~~~~--'------------~~

~~~~RA5i6E55iCAt 55sTR5t QUESTION 4.01 (3.25)

a. .ist five conditions that require Emergency Boration of the i< C S . (2.25)

6. How is Emergency Boration initiated ? (1.0)

GUESTION 4.02 (3.00)

Answer the following concerning Adherence to Ope ating Procedures,

a. What two conditions do not require a procedure to be present (at the location) and open (readable)? (1.0)

b. There are two specific instancec that allow deviation from proce-deres, license conditions and/or Tech Specs. In one circumstance prior SRO approval is needed. In the other, no prior approval is required. What are these two circumstances? (2.0) 00ESTION 4.02 (3.00)

List the immediate action sub-steps from E-0, ' Reactor Trip or Safety Injection' that allow you to accomP lish the following immediate actions'

a. Check if SI is actuated (1.0)

6. Verify Generator Trip (1.0)

(1.0,*

c. Verify AFW pumps running

(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)

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40 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 14

'~~~~R 656L65iEAL E5 TR6L'-~~~~~~~~~~~~~~~~~~~~~

QUESTION 4.04 .(3.00)

Answer the following questions regarding E0P 35 FR-H.1, RESPONSE TO LOSS OF-SECONDARY HEAT SINK:-

a. List the two separate circumstances which warrant entry into this procedure. Include any applicable setpoints. (1.5)

6. What action is required if, during this procedure, the RWST level decreases to less than 20 feet ? (0.5)

c. What symptoms must-be present before a SG is considered

" (1.0) hot / dry .

QUESTION '4.05 (2.50)

During a serious emergency, operators may be called upon to assist in search and rescue or recovery operations in the plant.

a. In such cases, what dose could you receive

1) To bring an injured worker to safety? (0.5)

2) To eliminate the further escape of radioactive effluents ? (0.5)

b. What are the possible effects of receiving radiation exposures of the levels of 50 rem ? Include short-and long term effects. (1.0)

c. Who must authorize this voluntary radiation exposure up to the emergency limits ? (0.5)

(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)

40 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 15

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~~~~R d65UL55 feat C5sTRbl GUESTION 4.06 (3.00)

The following questions pertain to Procedure E-ir Loss of Reactor or Secondary Coolant.

a. If SI is terminated and RCS pressure decreases to less than 250 psis, then what action is required ? (0.5)

b. What are TWO circumstances per E-0 which require shutdown of the Reactor Coolant Pumps? One circumstance relates to the steam generators, the other relates to a RCP support system. (1.5)

c. The minimum pressuriner level required to terminate Safety Injection is not the same if containment conditions are adverse compared to normal. Which condition requires the higher level (adverse or normal) ? WHY ? (1.0)

GUESTION 4.07 (2.00)

a. Per AOP-43r High Reactor Coolant Activity, what three plant condit-lons may cause high RCS activity due to the release of irradiated corrosion products? (1.0)

b. In the event of high RC5 activity, what is the reason for securing the following containment sump pumps. primary drain pumps and their containment isolation valves, containment v a c uu ta pumpsr and contain-ment isolation valves for reactor plant sample systems? (0.5)

c. An increased concentration of what two gases sampled from the VCT gas sample space would be indicative of failed fuel? (0.5)

(***** CATECORY 04 CONTINUED ON NEXT PAGE *****)

Y -

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 16

~~~~R A6i6[66fEAL E6sTR6[~~~~~~~~~~~~~~~~~~~~~~~~

GUESTION 4.08 (3.00)

Answer -the followins questions concerning reactor startup.

a. What are the MINIMUM requirements for Source and Intermediate range operability (# required) prior to startup?

b. What is the MINIMUM temperature for criticality?

c. What is the MAXIMUM startup rate permitted under normal conditions?

d. What is the MINIMUM number of RCps required to be running prior to startup utilizing control rods?

e.: When is it permissible to block the Source Range (SR) high f lu:: trip?

GUESTION 4.09 (2.25)

a. State the Safety Injection termination criteria as presented in E-0. List setpoints, .if applicable. (1.25)

b. State the Safety Injection reinitiation criteria as presented in'E-0. List setpoints, if applicable. (1.0)

(~s * * *

END OF CATEGORY 04 *****)

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

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Figure 50-7 # -

-1800 -

Power Defect vs. Percent Power /

f J'

C:

Cycle 5 2

W - 300 ppet

' ,/ ,

T /

/ /

I /

/ /

-1600 -

/

i , /

f/

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/ J

/ /

/ / / - 900 ppm J / /

-1400 1 s'

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^ '

j/ / f j -1200 ppm i

V / / /

/ / / /

' ' 1500 ppm

/ / /

/

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-1200 1 /

/ /t

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i / / / /

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8, -1000 I / / / /

w / / / / /

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43 / / /

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4'

. / f Q / / / / /

" / / / f /

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i 0.

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/ / A / / .

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-600 ./ / i // n l n

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// /// \

e // /// ,

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i 0 20 40 60 80 100 ISSUE 2 Power Level (Percent of Full Power) REVISION 6

1500 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . .

'- _]

1400 \.' ; Figure 50-8 i N Integral Rod North vs. Steps Withdrawn i

~ .

'. Banks D and C Moving with 100 j 1300 -

Step overlap 3

- \ No Xenon, BOL, HZP E

~

Cycle 5

'.1 1200 _

..r .-

1100- ;

1000 i

1 900 8

800 '.

1 - .

.c 1 i

U 700 -'

& i.

'8 600 _ '-

m

~ i.

8 't E 500 _ 'n a

a

~

1.

400 x 1

300 ---- .x M'.

L.

200 ..

100 - _ __ __

-~Y N

O j j i j i j . i 6

l [

i j i l

i j 6 g i i g

0 20 40 60 80 100 .120 140 160 180 200 220 240 t

Control Bank D Position (Steps Withdrawn)

ISSUE 2 REVISION 6

0 0

8 1

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, L

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EQuA110N SHEET f = ma v = s/t Cycle efficiency o (Metwru cut)/(Energy in) 2 -

o = ag s = V ,t + 1/2 at 2

E = ac KE = 1/2 av a = (Vf - V,)/t A = AN A=Ae" g PE = agn Vf = V, + at * * */t A

an2/ti/2 = 0.693/t1/2 t

y ,y, 1/28##

E(*1/7)(th)3

((t1/2) * (*b)3 ~

aE = 931 m -

I'= I,e **

Q = nCpat d = UA&t I=1 0 Pwr = W7ah I = I,10-*/U L TVL = 1.3/u P = P 10sur(t)

HVL = -0.693/u p = p et /T .

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

CR, = S/(1 - Keffx)

SUR = 26o/ t* + (a - o)T CR j (1 - K,ff)) = CR 2 (1 - kW2) ' -

T = (t*/s) + ((s - o)/$ol M = 1/(1 - K,g) = CR j/CR, T = s/(o - s) M = (1 - K,ffa)/(1 - K,ffj)

T = (s - o)/(le) SDM = (1 - K,ff)/K,ff a = (K,ff-1)/K,ff = AK,ff/K eff t* = 10-5 seconds T = 0.1 seconds-I e = ((1*/(T K,ff)] + [i,ff (1

/ + 1T)]

Id P = (t+V)/(3 x 1010) Idli*Id2 jj

=2 Id 22 2

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

R/hr = 6 CE/d2 (feet)

Water Parameters Miscellaneous Conversions 1 gal. = 8.345 lbs.- I curie = 3.7 x 1010 dps 1 ga[. = 3.78 liters 1 kg = 2.21 lbm 1 ft* = 7.48 gal. 1 hp'= 2.54 x 10 3 Stu/hr Oensity = 62.4 lbm/ft3 1 av = 3.41 x 106 Stu/hr Density = 1 ge/cm3 ,

lin = 2.54 cm Heat of vaporization = 970 Btu /10m *F = 9/5'C + 32 Heat of fusion = 144 Stu/lbe *C = 5/9 (*F-32) 1 Atm = 14.7 psi = 29.9 in. Hg. 1 BTU = 778 ft-lbf 2

ES-201-2

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

1. Cheating on the examination means an automatic denial of your application

, and could result in more severe penalties.

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

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

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

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

6. Use only the paper provided for answers.

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

8. Consecutively number each answer sheet, write "End of Category " as

! appropriate, start each category on a new page, write _on1y one sTde of i the paper, and write "Last Page" on thTTast answer sheet.

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

10. Skip at least three lines between each answer.

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

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

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

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

15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND 00 NOT LEAVE ANY ANSWER' BLANK.

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

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

Examiner Standards 12 of 18

ES-201-2

18. When you complete your examination, you shall:

a. Assemble your examination as follows:

[ (1) Exam questions on top.

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

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

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

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

~

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

I i

1 1

I i

1 l

l

? -

l i

l l

Examiner Standards 13 of 18

.~. -

1. PRINCIPLES OF NUCLEAR-POWER PLANT OPERATION, PAGE 17

~~~~ - -

TsERs557sisics? SEAT TRAssrER As5 etUio rt5s ANSWERS -- BEAVER VALLEY 1&2- -86/07/22-BARBER, S.

ANSWER 1 01 (2.50)

a. The closer to criticality, (less SDM) the lanser time required to reach a stable count rate. (O'.75)

b. .A higher initial count rate will result in a higher count rate at criticality. (0.75)

c. 1. Critical rod height is not affected. (0.5)

2. Critical count rate will be lower. (0.5)

REFERENCE BVPS Reactor Theory Manual, Ch 5, ps 39-41 E0 11 001/010 K5.00 2.9/3.2 pg 3.1-7 ANSWER 1.02 (2.00)

a. Decreases (0.5) Pu 239 concentration increases and Pu 239 has a smaller beta. (1.0)

b. Larser SUR (0.5)

REFERENCE BVPS Reactor Theory hanual, Ch 5, pg 14-17 EO 3 001/000 K5.47 2.9/3.4 - ps 3.1-3

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, ~

PAGE 18

~~~~IUER566 Ud5 C5,"HEdT"TEdU5EER~ U6~iLUYD kl6U ANSWERS -- BEAVER VALLEY 1&2 -86/07/22-BARBER, S.

ANSWER 1.03 (2.00)

Power Defect: 1700 pcm - 1290 pcm =+ 410 pcm (0.5)

Xe: 2800 pcm - 2600 pcm = + 200 pcm (0.5)

Rods: 5 pcm -

405 pcm = - 400 pcm (0.5) 2.10

- Eib E --

Boron must supply ; ". pcm of hegative reactivity 210 pcm X (ppm /9.7 pcm) = 22 ppm (0.25) 300 ppm + 22 ppm = 322 ppm (0.25)

REFERENCE BVPS Reactor Theory Hanval, Ch 9, pg 3-8 EO 5 001/010 K5421 3.4 ANSWER 1.04 (3.00)

a. SAME (0.25) Steam dumps will compensate for any additional heat added by the fhd RCP. RCS temperature / reactivity unchanged. (0.75)

S

b. ACP HIGHER than ECP (0.25) Xenon will increase to near peak at 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after trip. Rods must be higher to compensate. (0.75)

c. ACP HIGHER than ECP (0.25) The corresponding temperature increase must e compensated by a hi aher critical rod (0 .

% W "pp,sition.TJe<O--HTM Nh.75)

- c r a c; - _ . c c, d ' crc; T;s

d. ACP to EP m ECP (0.25) ~ . ,_ - . _ _0:s

- ._.;;r t10 --;-

r

. . . . . .. u. _ _ ' ' . _ _ u. .,.. _.- .~

._ (0.75) 4 e tor Theor y rianual , Ch 9, pg 7-9 j M EO 4,6 004/000 A4.02 3.2/3.9 pg 3 1-18 ar<d & ss u reachA

1. PRINCIPLES OF NUCLEAR POWER PLANT OFERATION, PAGE 19

~~~~5U5R566 kd55C5~~55dT~5Rdh5F5E~dU6' FLUE 6~iLUU ANSWERS -- BEAVER VALLEY 1&2 -86/07/22-BARBER, S.

ANSWER 1.05 . (2.50)

a. The operator must 'i t M 2: c'- cr ':'.^ '? - -

compensate for the build up of Xenon to equilibrium in 40-50 hours (0.'?

and Samarium in 4 O V~ w. s0 (D.O O *i J3-sTdants eris' cars. 0 )

b. Again Xenon will increase to a new higher equilibrium value in 40-50 hours. (0.6) Samarium reactivity will not change. (0.4)

(Both may undergo a slight dip before increasing to or returning to equilibrium) ,.

REFERENCE BVPS Reactor Theory Manual, Ch 7, pg 13,22,23 EO 2 3,478,10 001/000 K5.33 3.2/3.5 pg 3.1-3 K5.35 2.1/2.5

'2. . CC ANSWER 1.06 : ,00'

a. Increases (0.25) moderator becomes less dense, neutrons can travel further higher probability of reaching a control rod, therefore rod worth increses. (0.75) ofgg g b.Ecc- ::r- (0.25) increased concentration ch;'t: 'l. ;fcctr;, ,cr ; t; ;hc

-..w,,,, -., 4 w ,, m ._., - . . _ . . , . . _ . . . . r-. _ , - _ - ___ u, _ _ , 1 - -_1 ____

NN

~' '

db N:S r hs c . (0.f5)

_!,;- :--- '^ .?"' ? > -- 1 :- r: : t i e r : , ;; ;; - t ~; c 'ONS QggQ REFERENCE GVPS Reactor Theory Manual, Ch 8, pg 14-16 EO 8 001/010 KS.04 2.2/2.8 pg 3.1-2 1

1 1

r I 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 20

--- isERR557sisics- sEEi isissFEE Es5 FEUi5 FE5s ANSWERS --- BE AVER V ALLEY 1&2 -86/07/22-BARBER, S.

].

4

' ANSWER 1.07 (2.00)

a. If core exit conditions reach saturation the enthalpy rise will no I

lonser be proportional to the delta T across the core. Therefore the OT Delta T trip no lonser provides adequate protection. (1.0)

b. Prevents e::ceeding the DNBR limit (0.75) of 13 (0.25).

1 REFERENCE BVPS Thermo. danual, Ch 7, ps 19-20 LP-SOS-1.1 EO 4

PWG 5 2.9/3.9 ps 2-1 ANSWER 1.08 (2.50)

a. Increase - due to reduction in back pressure from other loops.

b. Decrease .due the bacP. flow in the idle loop.

c. Decrease - due to les g g gq ee stance removal ac g kthe cor emperature.

d. Increase - less flow - ._--g gnisher

e. Decrease - increased delta - T~means lower Tc-and since S/ temp.

is always slightly < Tc, S/G temperature is less.

(0.2) direction (0.3) e:< p l a n a t i o n '

REFERENCE BVPS Thermo. Manual, Sect 4.8 Comp Pump-Centrifusal No. 7 2.3/2.3 ps A-10 l

4 4-

, , --,_.. ...-___-- - , - . - _ , , , , - , , . . , _ , _ _ , . . , _ . , _ _ _ _ . . _ _ _ . . . , , . . - - - - - - - - - - - - --=w -- - -~~r - = - - -

d

10 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 21

--- isEss567sAsiEs- REAi isAssFEE As5 FEUi5 FE6s ANSWERS -- BEAVER VALLEY 1&2 -86/07/22-BARBER, S.

! ANSWER 1.09 (2.50)

, a. Fuel centerline temperature (FCT) increases (0.5) due to densi-fication resulting in increased gap between fuel and clad. Therefore a larger delta T will be necessary to transfer the heat. (0,5)

b. FCT decreases (0.25) due to gradual swelling of fuel which reduces the gap between fuel and clad. So, a smaller delta T will transfer t. h e ,

heat. (0.5)

c. FCT will decrease (0.25) clad creep causes the gap to decrease.

So, a smaller delta T will transfer the heat. (0.5) 4

! REFERENCE i

BVPS Thermo. Manual, Ch 2, Sect 2.6 I

Comp HX and Cond. No. 9 2.4/2.5 pg A-17 f ANSWER 1.10 (3.00) 3 3 l a. Power (2) = Power (1) * (H2/N1) = 300 * (4) = 19.2 MW (0.5)

, 2 2 Delta P(2) = delta P(1) * (N2/N1) = 50 m (4) = 800 psid (0.5)

Flow (2) = Flow (1) * (N2/N1) = 880

4 = 3520 spm (0.5)

b. 3 (0.5)

c. DECREASES (0.5) j d. It causes pump damage (erosion, pitting and vibtration). (0.5)

REFERENCE BVPS Thermo. Manual, Ch 4. pg 31-35 Comp Pump-Centrifugal No.9 2.1/2.2 pg A-10 No.23 2.1/2.3 No.10 3.4/3.6

! No.29 3.0/3.1 l

4 i

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

~~. . - . . . -.

1 l 7. . PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 22 1 _______________________________________________________

ANSWERS -- BEAVER VALLEY 1&2 '- 86/07/22-BARBER, S.

1 d

ANSWER 2.01 (2.00) i

a. 5

b. 3

c. 1

d. 2 i e. 4 5 items (0.4) ea.

r REFERENCE BVPS Flow Diagrams, Fig. NS-3-1 l LP 2336 EO 3 002/000-K1.06 (3.7/4.0)

-K1.08 (4.'5/4.6)

-K1.09 (4.1/4.1) i ANSWER 2.02 (2.50) u 0.O) 4

a. As the plant pressure enanges so will the_celta-P across the a ti-seal thus changing the seal flowrate.,{.5: "

Ira i: hi;5 : ^_

'_;' p :: c '; r : :nd 1" :t lei r :: ; ::. . .52 (1.0)

b. Seal return entn.t isolation valves close. ( MOV-CH-378&381) (1.0)

c. RCS pressure compared to the backpressure created by the.VCT. (0.5)

! REFERENCE i BVPS OM 6, Sect 1.6.1, pg 24 OM 1, Sect 1.1.5, pg 9 OM 7, Sect 1.7.4, ps'62

LP 2336 EO 2 003/000 K6.04 2.8/3.1 pg 3.4-2

! A1.09 2.8/2.8 pg 3.4-3

! A2.01 3.5/3.9 3-L 4

I i

, , , , ,w.y._., _ , - . - - - - - . - - , . . , - , . , - . - . . - - - , . . , _ - - - . - ,, r-, -

4 t

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 23 ANSWERS ---BEAVER VALLEY 1&2 -86/07/22-BARBER, S.

ANSWER 2.03 (2.50)

a. The live bus transfer switch must be in the 'on' position. (0.5)

b. Emergency bus (IAE or 1DF) undervoltage signal. (0.5)

c. Reactor plant CCW pump (CC-P-1C) i Reactor plant river water pump (WR-P-1C) l Charging /HHSI pump (CH-P-1C) CO.5 ea.] (1.5)

REFERENCE BVPS OH 36, Sect 1.36.1, pg 7,8,26 Sect 1.36.4, pg 7 i

LP-SGS-36.1 EO 2r4,7 4

.062/000- K2.01 3.3/3.4 pg 3.7-1 4 K3.02 4.1/4.4 pg 3.7-2 K4.03 2.8/3.1 L

ANSWER 2.04 (2.50)

! a. 1. - RCS pressure < 430 psis. (auto close at 630 psis).

P:r. temp < 475 F (1.0)

b. 1. RWST Low-Low level (19'2.5' or 20' or 2/4 RWST low lvl alarm) l 2. SI signal present. (1.0)

c. RCS pressure (due to cold water from the RCP volute being ~

will increase rapidly heated in the steam generator CSGJ) (0.5) 1 REFERENCE 4

BVPS.0M 10 Sect 1.10.2, pg 3,7 ES-1.3, Attach. 2, pg i e

LP-SOS-10 1 EO 4,7 l

! 005/000 K4.07 3.2/3.5 pg 3.4-8 K5.05 2.7/3.1 4

i i

e

?

l

-~csu - . - - - , , , - - , - , - . -a ----..,.-,.gnn---. ,v, - ,--,em.g -,,,a-,- , , , - - ,--,_,-,--,,n,,n,,-r- . , - - ,----~--w.- ,n .e-

4

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 24 ANSWERS -- BEAVER VALLEY 1&2 -86/07/22-BARBER, S.

ANSWER 2.05 (1.50)

a. TRUE

b. FALSE

c. .TRUE REFERENCE BVPS OM 22, Sect 1.22.2, pg 5,6 LP-50S-22.1 EO 3 059/000-K4.03 (2.1/2.3) 4

-K4.16 (3.1/3.2) i -K4.14- (2.1/2.3) i ANSWER 2.06 (2.50)

a. Motor Driven- 350 GPM (at 2696 ft of head).

Turbine Driven- 700 GPM (at 2696 ft of head). (1.0) 3

b. low-low levels i n 2 /4- S / G ' s .

M ng h Any Main Feed pumps trip safety injection signal.

Start signal to FW-P-2 without it developing the required i discharge within a time period. 4 items (0.25) ea. (1.0)

c. To prevent pump overheating (0.5)

REFERENCE I BVPS OM 24, Sect 1.24.1, pg 6 Sect l '. 2 4 . 4 , pg 14,15 i LP-SOS-24.1 EO 3,7,9,10 061/000 K6.02 2.6/2.7 pg 3.5-42

K4.02 4.5/4.6 K4.08 2.7/2.9 9

I

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 25 ANSWERS -- BEAVER VALLEY 182 -86/07/22-BARBER, S.

ANSWER 2.07 (3 00)

a. Zirc-water reaction Radiolytic decomposition of ECCS fluids solutions used for entmt spray (1.5) grgsgogofmgt J Amy 3 items (0.57 ea'. 4

b. 4 (0.5) c .- Containment Atmosphere Purge Blower (1.0)

Hydrogen Recombiner REFERENCE BVPS OM 46, Sect 1.46.1, pg 2,8

'LP-SGS-46.1 EO 3,4,5

'028/000 K5.03 2.9/3.6 pg 3.6-23 K5.01 3.4/3.9 K1,01 2 5/7.5 ANSWER 2.08 (3.00)

a. NO l b. -4G- CL4SE

! c. OPEN

d. CLOSE

e. N0

f. OPEN 3 CLOSE

h. CLOSE

i. NO

j. NO 10 items (0.3) ea.

REFERENCE BVPS E-0, Attachment 2

, LP-SOS-11.1 EO 4,6 006/000 A3.03 4.1/4.1 pg 3.2-12 4

e n-- -- ,-n--,-e,. , . , - , . , , , - - - - . , , , - , . . , , - , , , a.,-,,, - - - , , - - - , - , , - - - - - . - - - - , - ~ - - a . - - > - <w- a

N 20 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 26 ANSWERS -- BEAVER VALLEY 1&2 -86/07/22-BARBER, S.

ANSWER 2.09 (3.00)

a. RCP Therm Bar Cool Wtr Disch Flow High RCP Therm Bar High Disch Temp High CCW Radiation level alarms Component Cooling Surse Tank Level alarms 3 items (0.5) ea.

bo High flow closes the respective trip valve (TV-1CC-107 A,B,C)

Check valve isolates on reverse flow Palief valve pr tests isolated piping AnV 3 items (0.5) ea.

Ys9 % \$ tb.

REFERENCE U 4ef~ % brt % dit

/

{

BVPS OM 6, Sect 1 6.1, pg 23 i Sect 1 6.4, ps 36,43,44 I 008/000 K3.01 3.4/3.5 ps 3.10-1 A3.01 3.2/3.0 pg 3.10-2

. ANGWER 2.10 (2.50)

a. Hi or Hi-Hi activity alarm from radiation monitor on discharge from the HX. (0.5)

b. Introduction of contaminated recirc spray water into river water return OR loss of recire water inventory from containment (either answer acceptable). (1.0)

c. (Verify alarm) i -Determine adequate recirc spray flow through non-faulted HX's

-Isolate faulted HX CO.5 each] (1.0) 1: REFERENCE GVPS OH 30, Sect 1.30.4, ps 61-72 LP-SGS-1.3.1 E0 2,5,6 026/000 K1.02 4.1/4.1 pg 3 6-13 I

l.

r . __

3. INSTRUMENTS AND CONTROLS PAGE 27

, ANSWERS -- BEAVER VALLEY 1&2 -86/07/22-BARBER, S.

ANSWER 3.01 (3.00)

a. Raises the limit, because high dT indicates a higher power. (0.75)

b. Increases to raise pressurizer level to 100% program, because of the higher Tave (0.75)

c. Rods move in, because of the Avet. Tave/ Tref mismatch (0.75)

d. No effect, the demand signal is present (Tave/ Tref) but there is no arming signal. (0.75)

REFERENCE BVPS nh 1, Sect 1.1.1, pg 12,20-Sect 1.1.5,-Fig 1-14 LP-SGS-1.3 E0 10 LP-SGS-1 4 E0 12 Pressurizer LP EO 3 016/000 K3.02 3.4/3.5 pg 3.9-11 K3.03 3.0/3.1 K3.01 3.4/3.6 ANSWER 3.02 (3.00)

a. High-High alarm (0.2) diverts air ejector discharsa to entmt (0.6)

b. High-High alarm (0.2) closes valves downstream of decay tanks (0.6)

c. High-high alarm (0.2) during refueling will automatically close the purge supply and e>:haus t damper s , ( 0. 4 ) activate the local fuel building and local entet evacuation alarms (0.4) and opens the main filter b a n k beek- d o m p e r s and closes the bypass dampers (0.4)

REFERENCE BVPS OM 43, Sect 1.43.1, pg 11,15,18 073/000 K4.01 4.0/4.3 pg 3.9-23

.=. _

3.. INSTRUMENTS AND CONTROLS PAGE 28 ANSWERS -- BEAVER VALLEY 1&2 -86/07/22-BARBER, S.

ANSWER 3.03 (3.50)

a. Its associated SUR indication would be more negative than the other channel when power is decreasins (0.5) and approximately 5-10 minutes after a trip it will indicate more negative than 1

-1/3 dpm (0.25). (0.75)

b. Compare SR level indication with IR level indication (IR 10 -10 amps = ~

10 4 CPS on the SR) (0.75)

c. 1. No effect on the current comparator, but it will cause the power ranse meter to move accordingly. (0.75)

2. Actual protective functions-(i.e. Ry . trip) may be initiated. (0.75)

3. After a calorimetric (if required). (0.5)

REFERENCE BVPS OM 2, Sect 1.2.1, ps 16,17 a Sect 1.2.4, ps 4 Sect 1.2.5, Fis 2-4 LP-SOS-2.1 EO 3,4,8 015/000 K1.01 4.1/4.2 ps 3.9-5 i

K3.01 3.9/4.3 K5.02 2.7/2.9 ANSWER -3.04 (2.00) a) No effect b' ^ ; cd ::t/:t: h(gM

- c) Arm and actuate d) Arm W cLM ( M (p, g Q REFERENCE BVPS OM 21, Sect 1.21.1, ps 13-17 System Flow Diagrams, Fig PGS-1-5 PGS-1-6 LP 2352 EO 3,7,S I

a I

l L.

30- . INSTRUMENTS AND CONTROLS PAGE 29 ANSWERS -- BEAVER VALLEY 1&2 ' -86/07/22-BARBER, S.

041/020 K4.11 2.8/3.1 K4.14 2.5/2.8 ANSWER 3.05 (2.50)

3. Law power (C-5) interlock (0.2) prevents auto rod motion when impulse power < 157. to preclude unstable operation. (0.3)

Hi3h bank 'D' rod stop (C-11) (0.2) prevents outward motion when the. bank is near the top to prevent system counter misalignment. (0.3)

-b. Deenergizes the lift coils (0.2)

Energizes the stationary and moving gripper coils (0.4)

Stops all automatic rod motion (0.5)

Energi=es the urgent failure alarm on the power cabinet (0.2)

Lights annuciator ' ROD CONTROL URGENT FAILURE '

(0.2)

REFERENCE

, BVPS OM 1, Sect 1.1.1, ps-18,50 Sect 1.1.5, pg 1,2 Sect 1.1.2, pg 3 LP-SGS-1.3 E0 12,14 001/010 K4.10 3.2/3.4 pg 3.1-7 001/050 A2.01 3.7/3.9 pg 3.1-12 ANSWER 3.06 (1.50)

a. TRUE

b. FALSE

c. TRUE (0.5 ea.)

l

! REFERENCE l BVPS 'M O 26, Sect 1.26.1- e3 14,26 Sect 1.26.4, ?g 5 l LP-SGS-26.6 EO 3 i 045/000 K4.13 2.6/2.8 pg 3.5-10 A4.02 2.7/2.6 pg 3.5-11 l

3o INSTRUMENTS AND CONTROLS PAGE 30 ANSWERS -- BEAVER VALLEY 1&2 -86/07/22-BARBER, S.

ANSWER 3.07 (3.00)

Low Pressuriner Pressure (0.5), <1845 psi 3 (0.3), 2/3 (0.2)

High Containment Pressure (0.5), 1.5 psis (0.3), 2/3 (0.2)

Low Steam Pressure (0.5), <510 psis (0.3), 2/3 detectors on 1/3 S/G (0.2)

REFERENCE BVPS OM-11, Sect 1.11.2, pg 5 LP-SOS-1.1 E09 013/000 A4.03 4.5/4.7 Ps 3.2-27 ANSWER 3.08 (2.00)

a. DECREASE

b. INCREASE

c. DECRE ASE ( In sho.lly (A)ill remciles hacnetd

d. INCREASE LU.d each)

REFERENCE BVPS OM 1, Sect 1 1.2, pg 9 LP-SGS-1 1 EO 8 012/000 K6.11 2.9/2.9 PG 3.9-2 A1.01 2.9/3.4 AMSWER 3.09 (2 50)

a. 1. SI

2. R:< trip coincident with Low Tavs (1.0)

6. 1. Both Feedwater pumps trip ( FF-P-1 A ,18 ) 613) -

2. All FW res valves shut (FCV-FW-478,488,498)(44) g*FW b ass valves shut ( FCV-FW-479,489,499(0.4)~ i t ; n _ ' 0 '- ' .:.

REFERENCE iP @%

BVPS OM 1, Sect 1.1.5, pg 23 LP-SGS-1.1 EO 5

i' 3. INSTRUMENTS AND CONTROLS- PAGE 31 ANSWERS -- BEAVER VALLEY 1&2 -86/07/22-BARBER, S.

059/000 K4.19 3.2/3.4 pg 3.5-38 A3.06 3.2/3.3

' ANSWER 3.10 (2.00)

- h-r. level will. decrease due to charging < letdown.

- 6 % P:r. level isolates letdown.

- P:r. level increases due to charging > letdown.

- At 92% reactor trip occurs. (No credit for trip on high press due to spray valves' ability to control 4 items 0.5 es. pressure increase)

REFERENCE BVPS OM 6, Sect 1.6.1, pg 58 BVPS.FSAR, Sect 7, pg-7 7-11 & Fig 7.7-5

? Pressuricer LP EO 3 i

011/000 K3.02 3.5/3.7 pg 3.2-21 K3.03 3.2/3.7 I ~ 01'6/000 K3.02- 3.4/3.3 pg 3.2-21

?

i 4

f I

i 3

4

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 32

~~~~R565bl6GEUdL C5sTR5L ANSWERS -- BEAVER VALLEY 1&2 -86/07/22-BARBER, S.

ANSWER 4.01 (3.25) loul-lon)

a. 1. Control rod height below the ginsertion limit.

2. Failure of any control rod to drop following a reactor trip.

3. Uncontrolled reactor cooldown following a reactor trip.

4. SDM less than requirements of Technical Specifications.

5. Une:<plained or uncontrolled reactivity increase.

6. Plant shutdown required from emergency shutdown panel Any 5 items (0.45) ea. 7 AN*06

b. 1. Open emergency boration valve (MOV-1CH-350)

2. Start a boric acid transfer pump in fast 3 Take manual control of the charging flow control valve and establish maximum flow.

3 items ' .5) ea.

REFERENCE BVPS OM 7, Sect 1.7.4, pg 47 LP 2337 EO 9,10 004/010 A2.07 3.8/3.9 pg 3.1-21 000/024 EA1.17 3.9/3.9 pg 3.1-45 SWG11 4.0/4.0 pg 3.1-47 (t,00)

ANSWER 4.02 . 0^'

a. Emergency procedure immediate act'on steps (0.5) .

Routine procedures that are frequently repeated (0.5) ci "c':1 i

" : reded t dc;;ct: " :- ?? cr 1, c c r. : : : c r. d i t ; ; r. :

6. c^

h_, : :ctic: ::r;;;tc^t :*' th: 72 pro..de c-, : - 1 c r. t 7c r t e c t i c --

-d _- i~-ediately crr- ert 't O' ": ;pp :c;; i; ;;cd:d to t;,.:

h:t-- :- ::t:ct i; crcssor to p-_;c^t p ;cr-nr1 : r. , c y .- c, i p .cr t

'c :3 ^'

Qgg REFERENCE BVPS Station Admin Proc, Ch. 4, pg 40, 41 PWG 23 2.8/3.5 pg 2-2 L

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 33

'~~~~~~~~~~~~~~~~~~~~~~~

~~~~R d6I6LUUI6 L"66 TRUL ANSWERS -- BEAVER VALLEY 1&2 -86/07/22-BARBER, S.

ANSWER 4.03 (3.00)

a. Any SI annunciator-LIT (0.5)

SI actuation status light -LIT (0.5)

b. Main generator curit:r breaker-OPEN (0.5)

Exciter circuit breaker-OPEN (0.5)

c. Motor driven. pumps-RUNNING (0.3)

Turbine driven pump-RUNNING-(if necessary) (0.3)

Verify AFW discharge valves-FULL OPEN (0.4)

REFERENCE BVPS E-0, pg 4-6 000/007 SWG-11 4.4/4.5 ANSWER 4.04 (3.00)

a. From E-0 REACTOR TRIP OR SI when minimum AFW flow is not verified. (0.5)

All S/Gs' NR level < 5% with total FW flow to S/G's < 350 spm (red path condition). (1.0) ,

b. CCS should be aligned for cold les recire. (0.5)

J

-e, r a 550 F $ d

, SG WR level < 10% (pJI)

REFERENCE BVPS FR-H.1, pg. 2,15 000/054 SWG7 3.6/3.7 pg 3.5-57 SWG10 4.1/4.2

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 34

~~~~ - ------------------------

RA6i6t6EIEst EEsTR6L ANSWERS -- BEAVER VALLEY 1&2 -86/07/22-BARBER, S.

ANSWER 4.05 (2.50)

a. 1) 75 REM

2) 25 REM (0.5) each

b. Increased liklihood of cancer, particularly leukemia. Short term somatic effects include-blood changes. (1.0)

c. Emergency Director (or authority as delesated by the ED) (0.5)

REFERENCE BVPS RCM, Chapter 1, p.?

PWG15 - 3.4/3.9 ps 2-2 PWG16 3.4/3.7 ANSWER 4.06 (3.00)

e. Manually rastart LHSI pumps (0.5)

b. High head SI pumps in operation (0.5) and RCS/ Highest SG DP

'~

less than-145 psid (0.5) or CCW to RCP-NO FLOW INDICATED (0.5)

c. The level must be higher (50% vs 5%) if adverse containment conditions e>:ist (0,5) due to potential reference les heating which causes indicated level to be higher than actual. (0.5) 4 REFERENCE

_BVPS E-1, ps 8,11 000/011 Gen K/A 7 3.7/4.2 pg 3.3-13 EK3.12 4.4/4.6 ps 3.3-12 1

i 4

1 s

1

c- _

4.- PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 35

- -~~~----~~--------------

--- EA5i5E55iBEL 55sTR5t ANSWERS -- BEAVER VALLEY 1&2 -86/07/22-BARBER, S.

ANSWER 4.07 (2.00)

a. Plant heatup, plant cooldown, abnormal pressure / temperature transients (0.33 each)

b. Preclude potential high airborne and increased radiation levels in the auxiliary bu' gns (0.5) c.NQMA gxenon and iodine 4 0.25 each)

REFERENCE BVPS AOP-43, ps 1, 2 000/076- EK 3.01 3.1 EK 3.05 36 Sys sen 5 3.6 LP-2336 EO 5,8 ANSWER 4.00 (3.00)

a. 2 SR 2 IR

.b. 541 F

c. ,br dpm

d. 2 RCPs operating 4 .

e. When power is above the P-t+ permissive (1 Em10 amps) 5 items (0.6) ea.

REFERENCE BVPS OM 50, Sect 1.50.4, PS 26-28 OM 1, Sect 1.1.2, ps 3 PWG 12c- 3.5/3.4 pg 2-1

[

'4 . PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 36

~~~~~~~~~~~~~~~~~~~~~~~~

~~~~RE656E655 EEL'66NTRUL ANSWERS -- BEAVER VALLEY 1&2 -86/07/22-BARBER, S.

ANSWER .4.09 (2.25)

.a. RCS subcoo'.ing adequate, greater thanSCM(nAttachment5)

Pressurizer level > 5%

Auxiliary. feed flow of at least 350 spm or Level in at least one S/G > 5%

RCS pressure stable or increasing 5 items 0.25 ea.

b. RCS.SCM lessthanthatrequired(pyAttachment6)

Pressurizer level cannot be maintained > 5% (1.0)

REFERENCE BVPS.0M 53, Sect 1.53.A.1, pg 14,15 013/000 A4.03 4.5/4.7 pg 3.2-27 i

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ML20210R276

Text

SUMMARY

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