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Exam Repts 50-272/86-09 & 50-311/86-09 on 860408-10.Exam Results:One Senior Reactor Operator & Three Reactor Operators Received Licenses
	ML20198S460
Person / Time
Site:	Salem
Issue date:	05/21/1986
From:	Keller R, Kister H, Norris B; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
Shared Package
ML20198S436	List: IR 05000272/1986009; ML20198S432;
References
	50-272-86-09, 50-272-86-9, 50-311-86-09, 50-311-86-9, 860512, NUDOCS 8606100438
	Download: ML20198S460 (50)
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s O U.S. NUCLEAR REGULATORY COMMISSION

REGION I

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EXAMINATION REPORT NO /86-09(OL)

50-311/86-09(0L)

i FACILITY DOCKET NO FACILITY LICENSE NO DPR-70 DPR-75 LICENSEE: Public Service Electric and Gas Company P. O. Box 236 Hancock's Bridge, New Jersey 08038 FACILITY: Salem 1 and 2

EXAMINATION DATES: April 8-10, 1986 p CHIEF EXAMINER: I- Nor71s

[Barry/ ,

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Y-React 6r Engineer ( aminer)

REVIEWED BY:

Robert M. Keller, Chief

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[ fate Projects Section 1C APPROVED: ( 2/ fC f, Harry B. Kistes? Chief Date Projects Branch No. 1 SUMMARY: One Senior Reactor Operator (SRO) and three Reactor Operator (RO)

candidates were examined during this period; all received their licenses.

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8606100438 860527 i PDR ADOCK 05000272 V PDR

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O REPORT DETAILS

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TYPE OF EXAMS: Replacement EXAM RESULTS:

l RO l SR0 l l l Pass / Fail l Pass / Fail l l- l l _I l Written Exam l 3/0 l 0/0 l l l l l _l

l Oral Exam l 2/0 l 1/0 l

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l Simulator Exami 2/0 l 1/0 l l l l _l l0verall l 3/0 l 1/0 l l l l l l

l CHIEF EXAMINER AT SITE: B. S. Norris (NRC)

OTHER EXAMINERS: N. F. Dudley (NRC)

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D. M. Silk (NRC)

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] Summary of generic deficiencies noted from grading of RO written exam:

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! Question N Comment

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i 1.0 Candidates had difficulty in predicting how parameters j would trend on a loss of natural circulation flo !

2.0 Candidates did not know the design requirements for minimum flow for the Auxiliary Feed Water Syste .08 a Candidates did not know the requirements for termination

of the injection phase for the Residual Heat Removal Syste .0 Candidates did not know the meaning of the phrase " lock-up ,

of a power cabinet" with respect to the Rod Control syste .0 Candidates could not explain the significance of both

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Intermediate Range detectors being overcompensate l i

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3.1 Candidates were not familiar with how the Letdown line radiation monitor could differentiate between increased activity due to core age and fuel failur .11 Candidates were unfamiliar with the electrical interlocks associated with various RHR valve .05 Candidates were not able to state the reasons for controlling feed flow when all steam generators are

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depressurize .0 Candidates did not know whose approval was required to increase an individual's exposure limi .1 Candidates were not familiar with the reason for securing Number 21 and 22 AFW pumps on a loss of Control Ai . Personnel present at Entrance Meeting:

NRC Personnel B. S. Norris - Chief Examiner Facility Personnel

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J. K. Lloyd - Principal Training Supervisor, Salem Operations A. Orticelle - Operations Engineer, Salem R. Best - Simulator Training Supervisor K. Moore - Operations Training Supervisor Personnel present at Exit Meeting:

NRC Personnel B. S. Norris - Chief Examiner D. M. Silk - Examiner Facility Personnel J. K. Lloyd - Principal Training Supervisor, Salem Operations R. Best - Simulator Training Supervisor K. Moore - Operations Training Supervisor R. Schaeffer - Assistant Manager, Operations Training H. D. Hanson - Manager, Nuclear Training J. Gueller - Operations Manager, Salem

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O Summary of Comments Made at Exit Meeting: Concern was expressed by the examiners for the condition when a casualty situation is addressed by both an Emergency Operating Proce-dure (EOP) and an Emergency Instruction (EI); in that, the lower tier

, procedure did not reference the higher tier E0Ps. Example: Loss of all Feedwater (0 pen item 86-09-01)

The facility recognized the problem and stated that all of the EI's were undergoing review to eliminate redundant procedures. The Operations Manager committed that this effort would be accomplished within the next yea The facility expressed a concern over the complexity of the simulator scenarios; in that, the candidates were placed in a situation beyond the " single-fault" FSAR design accident '

The examiners stated that the symptom based E0Ps were written to prioritize unanalyzed combinations of casualties, and the scenarios were written to examine the candidate's ability to follow procedures and use their judgment when conditions did not exactly match the procedure The facility expressed appreciation for the time spent by the

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examiners for review / resolution of comments on the RO written examinatio The examiners noted the assistance of the simulator instructor . Facility comments on'the R0 written examination are listed in Attach-ment 2, with the NRC resolution detailed in Attachment 3.

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Attachments: Written Examination with Answer Key (RO) Facility comments on the written examination NRC resolution of facility comments l

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U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION

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~3 FACILITY: SALEM 1&2

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j REACTOR TYPE: PWR-WEC4 DATE ADMINISTERED: 86/04/07 EXAMINER: SILK, APPLICANT: dE f#~ f E

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INSTRUCTIONS TO APPLICANT:

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

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25.00 25.00 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 25.00 25.00 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 25.00 25 3 3. INSTRUMENTS AND CONTROLS 2 S.5 *

25.00 46-09 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 98.50 J00.00 MO-00 TOTALS FINAL GRADE % ..

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All work done on this examination is my own. I have neither . #h given nor received ai APPLICANT'S SIGNATURE '"# #

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. PRINCIPLES OF NUCLEAR POWER PLANT OPERATIO PAGE 2

THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW t

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QUESTION 1.01 (1.50)

Describe the conditions under which cavitation occur (1.0) What are two noticeable effects of cavitation? (0.5)

QUESTION 1.02 (1.50)

Describe how a water hammer physically occurs in terms of fluid dynamic (1.5)

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I QUESTION 1.03 (1.25)

The reactor is at 10% power. What is the initial effect of an increase in steam demand on steam generator level (increase, decrease, no effect)? Ex-plain your answer assuming feedwater flow remains constan %

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, QUESTION 1.04 (1.50)

Explain why the Axial Flux Difference would become more negative or less negative for the following conditions. Consider each separatel OTDelta-T runback from 100% power with rods in automatic (0.5)

i Feed flow increases to the steam generators with rods in manual (0.5) Xenon is building in to the bottom of the core (0.5)

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

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PRINCIPLES OF NUCLEAR POWER PLANT OPERATIO PAGE 3

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i THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

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QUESTION 1.05 (2.25) After operating at 100% power for three months, power is suddenly lost to all of the reactor coolant pumps. - Below are three things that can be done to enhance natural circulation. Why is each done? (1.5) Pressurizer level should be maintained at 50% or greater aintain at least 15 F subcooling in RCS Haintain heat sink Briefly explain how the following parameters will be trending if natur-al circulation is LOST: RCS differential temperature (.25) Steam generator steam pressure (.25) Steam generator level (.25)

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QUESTION 1.06 (2.25)

a.._ State how the following parameter changes affect the critical heat

, flux (increase, decrease, no effect): ' Reactor coolant flow-decreases . . ....= (.25) Reactor coolant temperature increases .- -- S'* (.25) ~

i Reactor coolant system pressure increas'es - (.25)

! Sketch the approximate coolant temperature profile across the two flow ,

i channels shown in Figure 1. Which channel is operating closer to De-

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parture from Nucleate Boiling (DNB)? Justify your answe #

QUESTION 1.07 (2.00)

The reactor is at 30% of full power. Briefly explain how and why each of .. <.4

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the following parameters will change if one reactor coolant pump (RCP) is N # # # #

shut off with rods in manual and the turbine ~is in P-impulse mod ~

my Reactor coolant flow in the unaffected loops (0.5) x Indicated reactor coolant flow in affected loop , ( 0.5) g ,%g Steam' flow in the unaffected loops (0:5) T-ave (0.5) 940r:2

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-.----._ = ~ ~ ~ . PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 4

THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

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k QUESTION 1.08 (2.25) Define Shutdown Margin (SDM). (1.25)

l Give three reasons for Rod Insertion Limit (1.0)

QUESTION 1.09 (2.25)

The reactor is at 100% power at EO Rods are at 220 steps on Bank Boron concentration is 300 ppm. Power must be reduced to 30%. If rods will be inserted to 100 steps on Bank D, what will be the final boron con-centration at equilibrium conditions at 30% power? Use the attached fig-ures. Show all work and state all assumption QUESTION 1.10 (2.00)

O If rods are in manual and a steam generator safety valve opens, explain ( how and why -reactor power responds. Assume reactor is at 75% power. (2.0)

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QUESTION 1.11 (3.25) riefly explain the production and removal of Xenon (Xe) in the core?

(1.25) After threa months of steady state 100% power operations, the reactor trips. A two month outage follows. After the trip:

1. When will Xe reach its peak concentration in the core and what will be its maximum reactivity worth? (1.0)

2. When will Samarium (Sm) reach its peak concentration *in the core and what will be its maximum reactivity worth? (1.0)

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QUESTION 1.12 (3.00) .

Give two reasons why the Doppler Coefficient is of importance to

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reactor safety. (1.5)

O Explain how the Moderator Temperature Coefficient could become ( positive during high boron concentrations? (1.5)

l (***** END OF CATEGORY 01 *****) ,

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,. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 5 p

G QUESTION 2.01 (1.00)

The plant is at 75% power and all systems are in automatic. What systems will help the plant stabilize itself after a main feedwater pump tripped because of low feed pump turbine bearing oil pressure? Assume no reactor tri (1.0)

QUESTION 2.02 (1.70) The plant is at 100% power. The high flow alarm is actuated by high flow in the discharge header of the Component Cooling Water System (CCWS). The surge tank level is rising and radiation monitors on the CCWS are alarming. What should the operator suspect as the cause of the event? (0.5) What are the safety-related loads serviced by the CCWS7 (1.20)

QUESTION 2.03 (2.00) ,

State six of the places where the Chemical and Volume Control' System

(CVCS) interfaces with the Reactor Coolant System (RCS). (2.0)

QUESTION 2.04 (2.40) -

Describe the flow path used during Rapid Boration. Begin at the source of the boron, and mention major components along the flow path and end when the boron reaches the RCS. (2.4)

l QUESTION 2.05 (2.00)

l What effect, if any, does a Phase-A Containment Isolation Signal have on seal leakage flow? (1.0)

I What damage, if any, would result to the RCP's if they were left pan ,m *

ning following a Phase-B Containment Isolation Signal? Explain. (1.0)

.

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

-

,

.re:L%e .

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

.I

PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 6

.

.g QUESTION 2.06 (3.00) Give two purposes of the containment spray syste (1.0) How is hydrogen concentration controlled in containment and what are three potential sources of hydrogen following a LOCA? (2.0)

QUESTION 2.07 (3.60) From which busses do the 11 and 12 Motor Driven Auxiliary Feedwater Pumps draw their power? (0.6) Which steam generator (s) supplies steam to the 13 Turbine Drive Auxil-iary Feedwater Pump? (0.5) List the four sources of water that can be supplied to the Auxiliary Feedwater System (AFWS) in the preferred order of us (1.0) For a transient or accident condition, the flow provided by the AFWS must meet what requirements in addition to the minimum flow rate?(1.5)

'

.

,

QUESTION 2.08 (3.10) What busses feed Unit l's Residual Heat Removal (RHR) pumps? (0.6) What functions do the RHR system provide as part of the Emergency Core Cooling System (ECCS)? (1.5) ,

l

' When and why is the Injection Phase for the RHR system terminated?

(1.0) ..y

.

QUESTION 2.09 (3.20)

e What'are four conditions that will generate a Safety Injection Signal?.+ w % *

Include setpoints and coincidence (1.2) _

' ' During a continued RCS depressurization caused by a LOCA, indicate the .,

order in which the ECCS subsystems will inject into the RCS and .the 6me66Mw

.

pressure at which each will injec (2.0) ...

1 ',918d@'iEdv y ;.,

,

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9 AMY[k,p (***** CATEGORY 02 CONTINUED ON NEXT PAGE *****) . .;m W .- . _ _______ _ _ - .__ _ . . _ _ ~ .. _ . . . _ _ _

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

.

'I

/ PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 7

.

i ,

l QUESTION 2.10 (3.00)

<

Starting from a diesel generator, explain how power would be supplied to a Vital Instrument Bu Include major components and voltage (3.0)

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.

INSTRUMENTS AND CONTROLS PAGE 8

)

!

QUESTION 3.01 (1.50)

The plant is at 100% power. Explain the response of the Steam Dump System (SDS) if PT-506 fails low?

QUESTION 3.02 (1.70) Will the plant depressurize from a PORV opening if a high failure of the master control pressure channel occurs ? Explain (1.0). With the control bezel selected to automatic, when will the air-oper-ated relief valves open and what is the reset value? (0.7)

QUESTION 3.03 (1.20) Why are the wide-range RTD's not used for protection related functions?

O

\ / Why are the narrow-range RTD's not used during natural circulation?

V .

.

QUESTION 3.04 (2.50) Explain the purpose of the variable gain unit in the pcwer mismatch circuit? (1.0) , What are the six rod control interlocks and in what modes (automatic and/or manual) do they function? (1.5) '

QUESTION 3.05 (2.50)

l (' During a logic cabinet urgent failure, hc,v is it possible to regain .

.

,"

control of the rods?

~

~'(0.5)~ ~ What is meant by a " lock-up of the power cabinet"? (2.0)

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,

(***** CATEGORY 03 CONTINUED ON NEKT PAGE *****) J$$$I

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

, ,

.. - - ...

.

l

. INSTRUMENTS AND CONTROLS PAGE 9 (

QUESTION 3.06 (2.25)

The selected channel to the pressurizer level control system fails lo What component responses will be initiated by the channel failure?

(1.75) If no operator action is taken, what wili eventually trip the plant?

.(0.5)

t QUESTION 3.07 (2.05)

' If the level in the reference leg of the level transmitters on a steam generator decreases, will the control system cause actual steam gener-

=ator level to increase, decrease, or remain the same? Explain. (1.0)

b. .What three conditions cause automatic closure of the feedwater regulating valves? (1.05)

.

-

QUESTION 3.08 (2.30) What is the purpose of the Overtemperature Delta-T and the Overpower

'

Delta-T trips? (1.3)

I Explain how the Overtemperature Delta-T Turbine Runback work (1.5)

QUESTION 3.09 (2.50) If both Intermediate Range Detectors are overcompensated, what could j happen during a reactor shutdown regarding the source lange detectors

and the Reactor Protection System? Explain. (1.0)

'

-

l

~ Howisnuclearpowocindicationadjusted'tomatchbecondarynower?(0.5) What does the detector current comparator do and when will it give an

, . alarm? (1.0) . ,

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

'

- _ _ _ - -

i

.

l INSTRUMENTS AND CONTROLS PAGE 10 I'mT

\v)

QUESTION 3.10 (3.00) What three process radiation monitors can be indicative of primary to secondary leakage? (1.8) As the reactor ages, RCS activity increases. How does the process radition monitoring system distinguish between RCS activity from reactor age and fuel failure? (1.2)

QUESTION 3.11 (3.00) What electrical interlock pre <ents the operator from opening the con-tainment sump to RHR pump isolation valves (21SJ44 and 22SJ44)7 (0.5) What electrical interlocks prevent the operator from opening the RHR to Safety injection pump and charging pump suction isolation valves (21SJ45 and 22SJ45)? (1.5) What electrical interlocks prevent the operator from opening the RER to containment spray isolation valves (21CS36 and 22CS36)? (1.0)

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END OF CATECORY 03

.~ . . - . . - _

,

'

. . PO'cFTUEFS - NORMA *. \BNCHMAL. FMFRCrNr.' AND PAGE

.

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.'

.

QUESTION 4.01 (1.00) A shift crew may be one less than the minimum requirement for a period of time of two hours. If your replacement is late, can you leave when your shift is over? (0.5) What is the maximum amount of time an operator can be on shift during a 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period? (0.5)

!

i QUESTION 4.02 (1.00)

What is the limit for a cooldown rate of the RCS and what is its basis?

(1.0)

,

QUESTION 4.03 (1.50)

i

'

a. In accordance with E0P-LOCA-2, Post LOCA Cooldown and Depressurization, why might pressurizer level rapidly increase? (0.5).

b. In accordance with E0P-LOCA-3, Transfer to Cold Lag Recirculation, Charging and Safety Injection Pump flow are monitored when closing RWST suction valves. What actions should be taken if a large flow decrease occurs? (1.0)

"

!

QUESTION 4.04 (2.00)

The plant is at 100% power and both Main Feed Pumps trip. 21 and 22 auxil-iary feedwater pumps have started. No flow is indicated and pump discharge

_

pressure is greater than 1350 psig. In accordance with E0P-TRIP-1, Reactor l Trip or Safety Injection, and EI-I 4.12, Loss of Feedwater:

l a. Why depress the PRESS OVERRIDE DEFEAT push button? (1.0)

'

l b. What protective feature is removed by the PRESS OVERRIDE DEFEAT 7 (0.5)

I

! c. When auxiliary feedwater flow is initiated to the steam generators, what condition takes precedence over water hammer considerations?(0.5)

.

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

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

'

.

,4 . FROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND FAGE

RADIOLOGICAL CCNTROL ) ',,

i

~

\

QUESTION 4.05 (2.50)

For a multiple steam Fenerator depressurization, what are three reasons for carefully controlling feed flow to the steam generators?

,

QUESTION 4.06 (2.50)

A 24 year old, with a life time exposure through the last quarter of 23 ram, will be working in a 200 mrea/hr radiation field during a planned re-fueling outage. In addition to his life time exposure, he has received 2100 mrem in the present quarte What provisions must be met to allow an individual, in non-emergency situations, to exceed the quarterly regulatory limit? (1.0) Who's approval is needed, in this case, to increase this individual's exposure limit?

(1.0) How long may this individual work before he reaches the maximum quarterly limit allowed at Salem under thhse conditions? (0.5)

' QUESTION 4.07 (2.50)

) -

The plant is being maintained in hot standb ,

, How often must the shutdown margin be checked? What is its limit?

(1.0) - If the shutdown margin is required to be increased, how is this done and what action should follow to ensure safe plant conditions? (1.5) ,

,

W; QUESTION 4.08 (2.00) ' Once in the E0P's, what are three conditions that result in trans- '

itioning to EOP-SGTR-1, Steam Generator Tube Rupture? (1.0) - aM- - -~ rs a %.g' s+they'

What is the principle behind isolating the ruptured steam generator? '

'

(1.0) ' ' "

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

.

s PROCEDURES - NORMAL. ABNORMAL, EMERGENCY AND PAGE 13

RADIOLOGICAL CONTROL o

QUESTION 4.09 (3.50) List the bases for the following the below statement out of IOP-3,

" Hot Standby to Minimum Load"? (2.5)

"Within fifteen minutes prior to criticality, verify RCS Tavg to be greater than or equal to 541 F" If, during power operation, Tavg drops below 541 F, what actions must be taken as per the Technical Specifications? (1.0)

QUESTION 4.10 (3.00)

A Control Air Header Pressure Low alarm comes in. Attempts to increase header pressure failed and the reactor was manually tripped when pressure in the instrument air header fell to 65 psi It has been decided to go to cold shutdow Why does EI-I-4.18, Loss of Control Air, stop the No. 22 and 24 reactor coolant pumps? (1.5)

N Why does EI-I-4.18 use only No. 21 and 22 AFW pumps to.' maintain steam

,

generator level at approximately 33%? (1.5)

QUESTION 4.11 ( 3. 50)[2,0) Besides the overhead annunicators alarming, what other indications

? will the control room operator have that 2C 115V vital instrument bus l

has been lost? (1.0)

t affect will the loss of 2C 115V vitsa instrument bus have on 9[4 N ssurizer pressure? Explain. (0.5) What manual actions will be taken to regain pressurizer pressure control? 44,&Fi[3t A (f.o)

.

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

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. 9

. <

U PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 14

~~~~isERs55isEsiCs, sEET-TRAssFER"dU5~FLUi5 FL5s

____________________________________________

ANSWERS -- SALEM 182 -86/04/07-SILK, ArJSWER 1.01 (1.50) Cavitation occurs when the pressure of the fluid is reduced to a press-ure below that of its saturation pressure for a given temperature which will allow boiling to occu (1.0) Noise, encess vibrations, and reduced pump capacity. (Any two .25 each)

REFERENCE Heat Transfer, Thermodynamics, and Fluid Flow (HTTFF) pgs. 319, 320

__________________________________________________________________________

Appendix Components: Pump Centrifugal ANSWER 1.02 (1.50)

Water (0,5).

hammers occur when slugs of, liquid are separated by vapor pocke'ts When flow is suddenly i rte r e s s e d , the slugs of liquid are accel'

[ erated thr.ough the system (0.5). When the liquid is suddenly forced to (' }% j change direction, it imparts a great deal of its momerftum to the con-straining elements (0.5).

mas -

REFERENCE ~

HTTFF pg. 346

___________________________________________________________________________

3.5 059 000 E 5.04 <

ANSWER 1 03 (1.25)

,

'

Increase (.25). As steam flow increases the number and size of steam bubbles in the steam generator increases (0.5). This reduces the density, thus increasing the specific volume of the mass in the steam generator, which causes a swell in level (0.5).

REFERENCE

- HTTFF pgs. 46-50 --* 9* ' +C

,

IOP-3 pg. 14

__________________________________ ________________________________________m

.

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x PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 15

~~~~ihikR55795 hie 5~~9Ehi~ikih5Fik~595~FCGiB~iC50

____________________________________________

ANSWERS -- SALEM 1&2 -86/04/07-SILK, ANSWER 1.04 (1.50) More negative (.25) because rods are inserted and push the flux to the bottom of the core (.25) More negative (.25) because more moderation will occur in the bottom of the core due to T-cold decreasing (.25) Less negative (.25) because Xe inserts negative reactivity in the bottom of the core and thus flux moves to the top of the core (.25)

REFERENCE Rx Th pgs. 196-210 232-233 T5 pgs. 1-1 and B 3/4 2-1

___________________________________________________________________________

3.1 001 000 K 5.06 K 5.38 ..

NSWER 1.05 (2.25)

J l To'e'nsure that no vapor pockets form in the loops (0.5)

To prevent steam pocket formation (0.5) To help thermal driving head . (0.5) . Will exceed 100% full power value (.25)

i Pressure will decrease (.25) Level will increase (.25) y l REFERENCE l HTTFF pgs. 356, 357 l ___________________________________________________________________________

!

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i a PRINCIPLES OF NUCLEAR POWER PLANT OPERATION,

- ~ ~

PAGE- 16 j

~~~~iU5Rb66Y d55C5I~55dT TR5U5F5R dU5~ FLUID FL5s

_________________..__________________________

ANSWERS -- SALEM 1&2 -86/04/07-SILK, ANSWER 1.06 (2.25) . Decrease (.25) Decrease t.25) Increase (.25) The control rod channel will operate closer to DNB (.25) because the

cross-sectional flow area is less while the fuel rods are producing almost the same amount of heat as fuel rods in the unit channel (.75).

0.5 for sketc REFERENCE HTTFF pgs. 226, 231-234

___________________________________________________________________________

!

3.4 003 000 K 5.01 ANSWER 1.07 (2.00) Flow will increase (.25) due to head,4 css caused by lost RCP (.25) ^ * ' Indicated flow will initially decrease to zero as the~~~ RCP coasts down -

.

(.25) but will increase due to badk' flow (.25) Steam flow will increase (.25) to account for affected steam generator (.25) Tave will increase initiallv then decrease (.25) because load remains constant (.25>

.

REFERENCE HTTFF pgs. 322-329, 264-266

__________________________________________________________________________

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

.

v PRINCIPLES OF NUCLEAR POWER PLANT OPERATION,

~

PAGE 17

~~~~ TUER566YUd5EC57~55dT ~TEEU5EER dU6~ELU 6 FLUU

____________________________________________

ANSWERS -- SALEM 1&2 -86/04/07-SILK, ANSWER 1.08 (2.25) SDH shall be the instantaneous amount of reactivity by which the reactor is soberitical or would be suberitical from its present cond-

,

' ition assuming all full length rod cluster assemblies are fully in-serted except for the single rod cluster assembly of the highest re-i activity worth which is assumed to be fully withdraw (1 25) . To ensure that acceptable power distributions are maintained (.34) To ensure that the minimum SDN is maintained (.33) To limit the potential effects of a rod ejection accident (.33)

REFERENCE '

Rx Th pg 323 Technical Specifications (TS) Unit II pgs 1-6 and B 3/4 1-4

________-_-_______-______-____-_--_-_________--_--________--__________,____

3.1 001 000 K 5.08 s ,,NSWER 1.09 (2.25)

Power Defect: 2425 pcm - 800 pcm = +1425 pcm (0.5) '*

Xe: 3200 pcm - 2000 pcm = +1120 pcm (0.5)

Rods: ; O pcm - 650 pem = - 65Q,pcm (0.5)

_

m. . ,r.ny

- is95 e---

E:oron must supplv 1900 pcm of negative reactivity 1900 pcm X (ppm /10 pcm) = 190 ppm (0.5)

G66 ppm + 190 ppm = &+9 ppm (.25)

90 0 wPr REFERENCE 490 Reactor Engineering Manual (REN) Part 1 ECP

_____________________________________________________-_____________________

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, PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, i

_ ______________________________________________ PAGE 18 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

____________________________________________

ANSWERS -- SALEM 1&2 -86/04/07-SILK, D.

) ANSWER 1.10 (2,00)

j When steam decrease (0.5). demand increases Teold will decrease which will cause T-ave to When T-ave decreases Moderator Temperature Coefficient cdds positive reactivity which causes reactor power to increase (0,5).

When reactor power increases fuel temperature increases which adds neSative reactivity from the doppler coefficient (0.5). Reactor power will increase until the reactivity changes from MTC and doppler are equal (0.5).

REFERENCE Rx Th pg. 308

_________________________________. ________________________________________

3.1 001 000 K 5.29 ANSWER 1 11 (3.25) Xe is produced directly as a fission product (.25) and also indirectly

['*g from the decay of Te-135 to I-135 uhich decays to Xe-135 (0.5). Xe is b / removed by decay *and by burnout (0.5). ' . 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months 5000 pcm

days 1050 pcm REFERENCE Reactor Theory (Rx Th) TP 37.2 to 3 ______________________________________________________________________

3.1 001 000 K 5.13 *

5.38 3.5

1 ANSWER 1.12 (3.00)

i . It is always negative and thus provides a negative reactivity insertion when fuel temperature rises (.75) It acts immediately to inhibit a power increase (.75) At high RC5 temperatures (0.5) the moderator expands and displaces

! boron form the core which is a positive reactivity insertion (1 0) .

i REFERENCE i Rx Th pg. 170, 152 -

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-

PAGE. 19

--- isEss557sisiCs? SEAT iRissFEE As5 FL5i5 FL5s

____________________________________________

ANSWERS -- SALEM 1&2 -86/04/07-SILK, s K 5.20 r

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'f PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE- 20

_______________________________________________________

ANSWERS -- SALEH 1&2 -86/04/07-SILK, ANSWER 2.01 (1.00)

Steam dump system (.33)

Rod control system (.33)

Auxiliary feedwater system (.34)

REFERENCE LP SDS pg 8 Emergeny Instruction EI I-4 12 pgs 2, 5

___________________________________________________________________________

3.5 039 000 system generic 4 ANSWER 2.02 (1.70) Thermal barrier cooling coil rupture (0.5) RHR Hx RHRP seal Hu SIP seal Hx g

-

Charging SIP seal Hx (0.3 each)

,

REFERENCE .

SN Vol 2 CCNS pgs 13, 8, Fig CC-1

__ ________________________________________________________________________

3.10 000 000 K L.02 K 1.04 ANSWER 2.03 (2.00) .

Letdown from cold les 4 4

<

Excess letdown from cold les 4 3 l Charging to cold les 4 3

Alternate charstng to cold les 4 4

, Auxiliary spray l RCP seal injection ,

( . 55 w.wiii i REFERENCE .I he le*p .c .

I SN Vol le CVCS Fig CV-4 13 6e aft ?

___________________________________________________________'________u__"_____,,

j 3.1 004 000 K 4.05 '

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s V PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 21

_______________________________________________________

ANSWERS -- SALEn la2 -86/04/07-SILK, ANSWER 2.04 (2.40)

.e4 Boric acid is pumped frgm the boric acid tanks (0.8) through the boric acid transfer pumps ( 0 .* a)

the suction of the chargingthrough pumps CV1,Jp).

( bypassing the blender going;to

":;- ;-;t; t '. ; pr; : t h: r- ;: : :

+"- ~;;" 'let cer.tr:1 c:2v: C"55 '0.2}- seal pressure ( 0 $h) , and charging line isolation valves CV68,69 ( 0 .* ntrol .

valve CV71 Flow then goes through the regenatgve heat exchangers (O'Y) and Soes into the RCS on the 43 cold leg ( 0 .'#) .

REFERENCE St Vol 1, CVCS pg 38, Fig CV-4, CV-8

___________________________________________________________________________

3.1 004 010 K 6.09 ANSWER 2.05 (2.00) Seal leakof f ca,nnot ei<it containment (0.5). It exits through a relief valve upstream of the isolation valves and 90es to the PRT (0.5).

[

g (J S)

Damage could result to the upper bearing (W) , lower bearing 1.55) (*) ,

N

.

leur- radi:1 b r:r i rq ing(.F0

'

.25} c r. d ;ccis ,.25; due to lack of proper cool-REFERENCE SN Vol 1, RCP pas 23, 15

___________________________________________________________________________

3.10 008 000 L 3.01 .4 003 000 K 4.11 30 ANSWER 2.06 (3.00) .

Spray cool water into the containment atmosphere in the event of a LOCA to ensure that containment pressure does not exceed its de-sign pressur (0.5) Remove elemental iodine from containment atmospher (0 5) Hydrogen recombiners (0.5) reduce h'ydrogen that comes from zircopium water reaction (0 5), radiolytic decomposition of emergency'ecre' cool ' ,~

,

ing solution (0.5) and corrosion of construction materials (0.'5)'.

REFERENCE '

SN Vol 2 Containment and Containment Spray, - 22e 23 . 2- E

______________________________________________pgs 20,___________________________~+ __

3.6 026 000 K 4.04 ' % '^P

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_______________________________________________________

j ANSWERS -- SALEN 1&2 -86/04/07-SILK, .6 028 000 System generie 4 K 5.03 29 ANSWER 2.07 (3.60) A Vital 4KV Bus (0.3)

12 - 1B Vital 4KV Bus (0.3) and 13 (0.5) . AFW storage tank Fresh water and Fire Protection Tanks

- DM water storage tanks Service water (.25 each) The minimum flow (must be delivered within one minute)to at least two effective steam generator (.75) and capable (of maintaining the required

]

.

flow) to attain (stable zero-load) hot standby for at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (.75).

, REFERENCE

Lesson Plan (LP) AFWS pgs 8, 21, 18, 19, 15 Student Notebook (SM)' Aux Feed System pg 14

[ 3 5 061 000 ( 2 02 37 ( K 1.03 I.' 4.01 .9

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Sys Gen 5 3.3 i

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Nj PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE' 23

ANSWERS -- SALEh 1&2 -86/04/07-SILK, ANSWER 2.08 (3.10) A 4 KV Vital Bus (0.3)

12 - 1B 4 KV Vital Bus (0.3) Low head injection during injection phase (.3)

Recirculate water from the containment sump back to the RCS during recirculation phase (.3)

Provide suction to the: High head SI pumps (.3)

Intermediate head SI pumps (.3)

Provide flow to the Containment spray headers (.3) RWST (0 5).

low level alarm indicates the switch to the recirculation mode At this time sufficient water level should exist in the con-tainment sump to provide the required net positive suction head for the RHR pumps (0.5).

REFERENCE SN Vol 2, RHRS, pgs 19, 20 LP RHR5 pg 13 (\ -------------------------------------------------------- ,-..--------------

3.4 005 000 K 2.01 *

System generic "O K 4.02 ANSWER 2.09 (3.20) Low pressuri:er pressure 1765 psig 2/3 High contaiment pressure 4.0 psig 2/3

,

'

High steam 12ne differential pressure 100 psid 2/3 High steamline flow, 2/4, with low low Tave, 543 F, or low steamline pressure, 500 psis (0.3 each) .s, High head injection 1765 psig Intermediate head injection 1520 psis Accumulators 650 psig Low head injection 170 p. sis ~. -

,

- (.25 for each respons'

REFERENCE * * *

"" ~*' *

SN Vol 2 ECCS, pgs 8, 9, 27

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

-

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K 6.02 .

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. PLANT DESIGN INCLUDING SAFETY AND ENERGENCY SYSTEMS PAGE 24

-_---_-________________________________________________

ANSWERS -- SALEM 182 -86/04/07-SILK, D.

'

ANSWER 2.10 (3.0 )

,

The diesel nerators te into the 4160V vital busses (.33). The voltage is stepped do n by a 160/240 transformer (.33) to the 230V vital busses (.33). The 23 V A is transformed and rectified to approximately 125V DC

,

'

in the power su y cabinets (1.0) where the rectified voltage is compared with an infeed m a 125V DC bus (.33). The higher voltage is sent to an inverter (.33) whi feeds a vital instrument bus (.33).

REFERENCE SN Vol 7 Electrical Distribution, pg 35, Fig ED-6, 8

_-___________________________________________________________________

3.7 064 000 K 1.01 4.1

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_-------------- PAGE- 25 ANSWERS -- SALEM 1&2 -86/04/07-SILK, ANSWER 3.01 (1.50)

There util be no response by the SDS (0.5). PT506 feeds the arming signal (0.5) but there is no demand signal because PT505 provides Tref input for the demand signal (0.5).

REFERENCE LPe SDS pg 15; SN Vol 4 Fig SD-4, 10

3.9 016 000 K 3.03 ANSWER 3.02 (1.70) A pressure channel, set at 2335 psis, provides an interlock to prevent plant depressurination if the master control pressure channel failed j high (1.0). Open at 2335 psis (.35) and close at 2315 psig (.35).

REFERENCE --

}SNVol4PZRPressandLevelControlpgs 12 - 13 ,

_,n--------------------------------------------------------------------------

3.3 010 000 K 4 03 . -

ANSWER 3.03 (1.20) The wide-range RTO's have a relativelv slow response tim At low flow rates the narrow-range RTD's are inaccurat <

REFERENCE SN Vol 4 RCSTIS pgs 5 - 7

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

3.9 012 000 h o.06 '

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____________________________

ANSWERS -- SALEM 1&2 -86/04/07-SILK, ANSWER 3.04 (2.50) The variable gain at high power prevents the power mismatch circuit from demanding rapid rod withdrawal which could result in an excess power overshoot (1.0).

b. Auto and Manual IR Nuclear Overpower PR (High Ran3e) Nuclear Overpower OTdT OPdT Auto C WJ Turbine first stage pressure less than 15%

Control Bank D Withdrawal Limit (.25'each)

REFERENCE SN Vol 4 Rod Control System pgs 22 - 24, Table 1 '

LP Rod Control System pg 16

___________________________________________________________________________

3.1 001 000 K 4.08 *

K 4.07 .

ANSWER 3 05 (2 50) Use individual bank selection on the Bank Selector switc (0.5) An ' Inhibit' signal is generated whidh prevents a power cabinet from responding to any more current orders from its slave' cycler (1.0). *

All groups on that cabinet receive orders for reduced current to both stationary and. movable gripper coils, as well as zero current to;all ,

lift coils (1.0).

_

'

REFERENCE SN Vol 4 Rod Control System pgs 48 - 50

- -- - --- --

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____________________________ PAGE 27 ANSWERS -- SALEM la2 -86/04/07-SILK, ANSWER 3.06 (2.25) CVCS letdown isolation valve closes (.93)

All CVCS letdown orifice isolation valves close(43J All pressuricer heater groups are turned off(43>

"2r 3;:: i: "uil : : : -i &lm ;%)

C'J C C ' I s a c a r,t : s i s;1c; Iserrastel c he ge*J. 25 ::(:: '

F c .; 'ully ' Reactor will trip on high pressurizer level at 92%. (0.5)

REFFRENCE SN Vol 4 PZR Press and Level Control pgs 26, 27, 29

___________________________________________________________________________

3.2 011 000 K 3.01 K 4.05 ANSWER 3.07 (2.05) The indicated level will be higher th n actual, causing the feed water control signal to decrease feed flod k-til indierted ::: 27; 1 t:

t 2 ' . +Ana br e, aclua l le vef w[, jf ofec rea.ge f.5)

7: 257--

. Steam generator Hi Hi level of 67%

'SI signal gg ye g Reactor trip coincident with low Tavs 646 F , (.35 each)

REFERENCE SN Vol 4 SGWLC ogs 19, 20, 21, 34

___________________________________________________________________________ '

3.9 015 000 K 3.12 .5 059 000 k 4.19 3.2

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____________________________

ANSWERS -- SALEM 1&2 -86/04/07-SILK, ANSWER 3.08 (2 80) OTdT - To ensure operation within the DNB criteria (.65)

OpdT - To protect the reactor core from overpower conditions (.65) The signal is initiated at 3% below the reactor trip set point (0.5).

We t= hwe-een4+ol -systen-decreases turbine-4eed -;t ; .- ; ; ;;; ;7 := __.._i; :^.5? by running the turbine back at 200% per minute for 1.5 seconds A with a 28.5 second interval between load reduct' ions (0.5).

g REFERENCE SN Vol 4 RCSTIL pgs 9, 12

___________________________________________________________________________

3.9 012 000 K 4.02 K 1.03 K 1.06 ANSWER 3.09 (2.50)

N Overcompensation results in a lower indicated flux which could cause s_ the source r . . ige detectors to be reinstated too early (0.5) thus caus-

,, ing a reactor trip from source range high flux trip (0.5). Adjust the sairs of the summing and level amplifier (0.5). Signal outputs from the detectors are compared with the average of the correspond 2ng signal f r o ni the appropriate detector sections (0.5). An alarm actuates from a 2% ceviation from average when power is greater than 50% (0.5). ,

REFERENCE SN Vol 3 Excore Nuclear Instrumentation pgs 16 - 19

___________________________________________________________________________

3.9 015 000 K 3.01 System generic h 4 36

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____________________________ PAGE 29 ANSWERS -- SALEM 1&2 -86/04/07-SILK, .

ANSWER 3.10 (3.00) .- . , Condenser Air Removal Gas Monitors (0.6)

Steam Generator Blowdown Liquid Monitors (0 6)

Condensate Polishing Filters Monitors (0.6) The letdown lane monitors look at iodine and gross gamma activity (0.6). Fuel failures cause the ratio of iodine to gamma activity to increase (0.6).

REFERENCE _

SN Vol 3 RMS pgs 22 - 25 EI I-4.7 pg i gFI2-4.t6, psi

___________________________________________________________________________

3.9 073 000 K 1.01 K 4.02 ANSWER 3.11 (3.00) BOTH of the RWST to RHR pump isolation valves (21RH4 and 22RH4) are CLOSED (0.5) *

.

. . Either 2RH1 OR 2RH2 i s CLOSED (RCS 421 hot les suction isolation

.

.

valves) (0.5)

2. BOTH 21SJ44 AND 22SJ44 are OPEN (Containment sump isolation valves)

(0 5I 3. Either 2SJ67 OR 2SJ68 are CLOSED (Safety Injection miniflow i solat-ton valves) (0.5)

,

c. Either 2RH1 OR 2RH2 is CLOSED AND 21SJ44 AND 22SJ44 are OPEN (1.0)

REFERENCE SN Vol 2 ECCS pg. 43

________________-__________________________________________________________-.c 3.2 006 000 h 4.08 K 4.06 .6 026 020 K 4.03 .

Vf r

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r :;..p<y?'rrW .,-

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

~~~~

RA6 ULUU55 L'C5sTR5t

____________________

ANSWERS -- SALEM 1&2 -86/04/07-SILK, ANSWER 4.01 (1.00) No (0.5) hours (0.5)

REFERENCE TS P3s 6-1, 6-5

__________________________________________________..________________________

Generic Knowledges 1 ANSWER 4.02 (1.00)

The RCS cooldown rate shall not exceed 100 F/hr (0.5). This limit is in-posed to ,17; cit _. ------.-- m v v ;- . ... t._; c m;J ^

...v., ... ACS (0 5).

rninieras e * rwe/ 3 -fersit*1, REFERENCE TS pg 3/4 4-27 TS pg B 3/4 4-8 ( ___________________________________________________________________________

3.2 002 020 Sys Gen 5 ANSWER 4.03 (1.50) Voiding in RCS (0.5) Stop affected pumps and verify proper valve alignment for rectreulat- ,

ion (1.0).

REFERENCE EOP-LOCA-2 pg. 6 E0P-LOCA-3 pg. 10

___________________________________________________________________________

3.3 000 009 EK 3.10 .3 000 011 EH 3.12 l o- ,

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t PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 31

~~~~ -

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RA65UL65iCAt C NTRUL

____________________

ANSWERS -- SALEN 1&2 -86/04/07-SILK, ANSWER 4 04 (2.00) Overrides a failure in the pump run out protection circuitry that pre-venting the dischar3e valves from opening (1.0) Pump run out protection (0.5) Conditions requiring maximum flow for decay heat removal (0.5)

REFERENCE EI-I 4.12 pg , 5

__________________________________________________________________________

3.5 000 054 EK 3.04 ANSWER 4.05 (2.50) Minimize any additional cooldown of the RCS (.83) Keep steam generator tubes ??*@U:d (.83)

') 3 . Minimi=e source of steam flow to containment (.83) .

V REFERENCE +

__ ___ _ ._ _ __ _ b_'_ "_ _ _T_"_T _ _#___________________________ EK 3.04 ANSWER 4.06 (2.50) Dose to whole body should not exceed 3 rem per quarter (.33)

'

The 5(N - 18) limit must not be exceeded (.33)

The individual's exposure history is documented on NRC Form 4 (.34)

j Senior supervisor-RP (0.5) and the individual's senior supervisor.(0.5) mrem /0 = 2100 mrem + 200 arem/hr X T ; T = 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (0.5)

REFERENCE

_10 CFR 20.101 b AP-24 P3 13

___________________________________________________________________________

System wide and plant wide generic knowledge 15 35

.

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- PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 32 RADIOLOGICAL CONTROL r

,' ANSWERS -- SALEM 1&2 -86/04/07-SILK, ANSWER 4.07 (2.50) Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (0.5) and it is limited to greater than 1.6%

delta k/k (0.5).

Rept+ boraten - --- - ' '-

-'",.-"",-

,

( 2 0 _ _ _ _..l ; _ ; ;'.. t _ _ " ".

s .

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. . . - - .

r r-i --- ---- --- - - ,- - - -- w "9t" Th r* * * *P Ensure pressurizer and RCS boron concentration are within 50 ppm (0.5)

. , _ . . _ . . , . . . . . . . . . . . . y...... .. ....s... .ua .y.. in .o w sv.ar REFERENCE IOP 8 pg 5 ue - ^.^.; ys 2 o y _r 7 - 3. 3 6 > (xj b

-

________. __________________._____. ______________.... _____________.______

3.1 004 020 Sys Gen 5 K 6.01 ANSWER 4.08 (2.00)

3 Uncontolled increase in steam generator level (.34) Secondary radiation alarms (.33)

Boron and activity present in steam generator (.33) ',"- ir.: :::: th; r.._:_-._ i .. th; ._r ...d CC (,30) ;- ..i.---- , , .. , ..s . ,

.......

,

.... .. m . .. y . .- y .y.- s.ss, ........,.........s.ssi.-

"Ts m m,se ,.gg<, (o.oj REFERENCE E0P-LOSC-1 pg. 6 E0P-FRHS-3 pg. 3 E0P SGTR-1 pgs. 3, 13

..__... _________ ...________ .____ .___..... _____...___.. ________.......

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, PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 33

-

~~~~R d55 L55iEst E5NTs5t

____________________

ANSWERS -- SALEM 142 -86/04/07-SILK, ANSWER 4.09 (3.50) . MTC is within the analyzed temperature range 2. Protective instrumentation is within normal operating range P-12 (Tavg > 543 F) above its setpoint Pressurizer is operable Reactor vessel is above minimum temperature (RT-NDT) (0.5 each) . Restore Tavs > 541 F within 15 minutes, or (0.5) Be in hot standby within next 15 minutes (0.5)

REFERENCE IOP-3, Hot Standby to Minimum Load, pg. 6 TS, pgs. 3/4 1-6 6 D3/4 1-2

___________________________________________________________________________

3.1 001 000 K 5.15 K 5.16 SWER 4.10 (3.00) '

Reduces heat input to RCS (0.5). Leaving 21 and 23 RCP's in service gives best RCS pressure control (1 0). Using No. 23 AFW pJmp may reduce the pressure in 21 and 23 steam gen-erators enough to cause a Steam Line Differeritial Pressure Safety Injeetion (1.5).

REFERENCE EI-I-4 10 Loss of Control Air pg 2

___________________________________________________________________________

3.0 000 065 CH 3.08 .

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ANSWERS -- SALEM 1&2 -86/04/07-SILK, ANSWER 4.11 ;0.50p.o) There will be a midscale indication on all instruments with indicators powered from the bus (0 5), but a zero indication on all instruments with only transmitters powered from the bus (0.5).

s. A-b ' ::: - - ill d : = = = Y ~ E != = = = : ! 17 : :',, =t:.; 'i== :: 1 ^:: : ::: p:: 0,5)

a Control charging to establish pressurizer level at 20% (-er51 Establish excess letdown and balance chargin3 and letdown &&T97 Energize the pressurizer den (y sprey enters Jdsia5 heaters from the emergency Power supply ;^.-)

REFERENCE

%e erecl, o.3 c*c h)

AOP-ELEC-VIB-C pgs le 2

3.7 000 057 EK 3.01 , , , ,

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O Attachment 2 Facility Comments on the R0 Written Examination Question N Comment 1.0 Possibly also fluctuating ampsj flow oscillation .02 Key should also reference pressure decrease .03 What would happen if examinee correctly explains the wrong answer?

1.05. Key - incorrect and not applicable. E0P no longer calls for 50%

PL. The reference is not the technical document utilized for procedure developmen .0 The question asks for a trend - the key gives a value for Delta Also, pre-EOP referenc SG level is not an absolute guarantee of natural circulatio .0 Drawing is hard to decipher; especially on a timed examination. The actual flow channels are shown as the same size. Question "a" O discusses CHF and "b" switches to DNB - this adds a minor degree of confusion and is not used at tne NTC as an acceptable way to develop exam In addition, the term "almost" is used in the ke "Almost" is subjectiv We do not feel the "b" is a suitable question for R0's and that the drawing adds additional confusion.

,

1.0 Eliminate short term Tavg perturbation. Which Tavg?

1.08 b These are T.S. Basis - beyond R0 level, i

1.09 If curve is used boron wort is not exactly 10. There are other questions which have aumerical values but no latitude for answers on the ke .10 The final (.5) answer may not be stated in those word .1 Is Te required?

1.1 R0's are not required to recall curves from renory. They are

'

available in the Control Rocm. There s..:i.: :e an acceptable range of answer as the Rx Eng. curves snouic ce the referenc .01 Ouestion does no: state whether or n:t ners is operator action.

l- p How many answers are required - that should be stated up fron I

\

AFW would not be utilized if the reacter did not trip. Other

acceptable answers - EHC, SGWLC, SGFF;, PP, FL.

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2.0 The only Hi flow alarm is RCP Thermal Barrier. This is the key

,

answer but the stated alarm (high flow) does not exis :

l 2.03 Does loop have to be identified?

i

' '

2.04 CV-55 is not in,PDP flow path. Should CV-71 be required. It would be easier to answer, grade, etc., if a drawing and hi-liter were provide ,

,

2.0 Should read RCP seal leakoff flow.

2.0 Some of key assumes loss of normal seal suppl ' '

2.0 Sources: Ha inventory in RC ,

'

u l 2.0 High point value and questionable R0 knowledge level.

I 2.0 Could add automatic somewhere in question.

.

-

2.0 Band of accepUble answers, HHSI-1765 assumes an auto S .10 The examinee may assume that the DG is the source and might not mention battery or auctioneerin .0 The examinee's answer may be longer and associated with

'

Delta K/K.

3.0 You do not regain conrol of the " failed" rods, i

! s 3.0 Is key answer appropriate for RO? We do not use the term

,

" lock-up of the power cabinet." ,

3.0 Key calls for PDP and CV-55 chang PDP and CCP are not run at same tim .0 . .The key doesn't really answer the questio .0 Low Trip - 554*F not 543 F.

] -

3.0 Do noi expect 10%/ minut .

'

3.1 Some more acceptable answers: R-46, Steam line, all. vent channel s 3.1 Unit 2 has semi-automatic switchove .0 This is an Administrative / Supervisory functio Limit thermal stresses - fracture is not'

'

4.02 Bases for R ( guarantee ;

,

.

-- .,-m.-. w- , ---4.-, - , . - - , r .-mmg -.r- - - - .,g--. ----w--,

--

--,..-y , -

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

, ,

, ?

!

.e

-

4

.

l 4.0 There is no ref. to water hammer in E0P-Trip .

R0 scope? Bases behind steps. Answer 2 - tubes don't really

'

4.05 need to be covered.

i The stated reference does not contain these answers.

! 4.06 This is really a supervisory and HP function. An NCO is no

,

more responsible for this than any other radworker at the

'

plant.

'

4.0 There are no stated condition Key answer is too specific for general questio <

, 4.0 The bottom line is to minimize the release.

. 4.0 Bases and very high point value for R0' .10 Again, very high point value, t

4.1 Only the first half is aske Looks for recall from memory i of subsequent actions, comments, etc. Subsequent actions are performed with procedure in han ) 4.1 Same as above (a).

4.11 There are 0.5 missing. A0P's are not memorize They are ABNORMAL OPERATING PROCEDURE.

I These are performed " procedure in hand." There is one for each bus, each is different, et It is not fair game frorn memory.

i

O . .- _ -

- _ .

. .

.;

,y .

J.*

!

"

Attachment 3 i

NRC Resolution of Facility Comments l QuestionN Resolution

,

1.0 Subjective commen j

! 1.02 Subjective comment.

,

! 1.03 Subjective comment.

!

1.05. Answer key is based on some level in the pressurizer, not specifically 50% . ,

1.0 With respect to Delta T, consideration will be given during i grading; the question asked how parameters would trend, it did ,

i not imply that the parameters were an indication of natural j circulatio [ Note: with respect to the Facility's comment

about the references, they need to ensure that generic theory texts do not contain information that conflicts with their-

'

facility.]

t 1.0 Subjective commen s.

l 1.0 Subjective comment, i

l 1.0 Not accepted, K&A Catalog gives an Importance Factor of 3.9.

l l 1.09 Subjective commen ;

1.10 Subjective commen .1 No.

l

'

. 1.1 Subjective commen .01 Subjective commen '

I l 2.0 The question did not quote an alarm window but posed a problem I with a high flow alarm as one of the indications.

l

, 2.03 Ye I l

2.04 Answer key modifie .0 Will be considered for future examination v I

.

--ac-1e i-et -p.,y,.*-------% c- s.-w-e y -- - =e . . e-- ----,=.,---,,,w,m m%m-,,u, -r.----,-e---v w.---, ., +

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

__ _ _- _ . _ . _ _ _

<

..

.

'%

.o

2 -

U 2.0 Answer key corrected.

.

1 2.0 Considered in gradin .0 Subjective comment.

2.0 Will be considered for future examination l 2.0 Subjective commen i l 2.10 Points redistributed on answer ke I

3.0 Subjective commen .0 Subjective comment.

I 3.0 Subjective comment. The terminology is directly from the

.

reference materia .0 Answer key modified.

+

3.0 Answer key correcte .0 Answer key correcte .0 Answer key modifie .1 Will be considered during grading, i

3.1 Will be considered during grading.

!

i 4.0 Not accepted, NCO's need to know how long they can safely be on shif ; e

'

4.02 Subjective comment. Answer key modified, j

4.0 Water hammer is referenced in EI I-4.1 .05 Answer key modified; additional reference is the ERG Background

'

Documents. K&A catalog gives rating of .06 Not accepted. K&A catalog gives this concept a rating of 3.5.

j 4.0 Initial conditions were given as Hot Standby; however, answer key modified to be more genera .0 Answer key correcte .0 Subjective comment.

!O

1

_ _ _ _ _ _ _

. _ . . . _ . _ _ _ . . _ _ _ _ . _ _ _ . _ ._ . _ . . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . . _ __. . . _ __ ____ __ .

'

.

,

, .

.

l 3 !

l 4.10 Subjective comment. The question did not ask for recall of the

, steps; instead it stated what the procedural steps were and asked for the reason for these step ;

4.11 Part b deleted and points redistribute i

' t

i

)

!

@

.

d

l l l l

l l

l

,

I

!

'

.

,_

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

_-

_

NRC:1 Form 6a R v. Oct 80 1p Transaction Type

' New Item N#'S O' 5' Y " S' # '

Originator Reviewin'g' Supervisor Name Modify O Deiete Docket i Docket i Docket i Docket i L51ol-bl?laj IIIIIII IIIIIII I I I I l_L1 isto l-13lf lel l I l l l l--I IIIIIII IIIIIII IIIIIil lIIIIII IIIllO I I I I l__lj i i IiIIi lIIIIII I I I I I I~l lIIIIl_l )

IIIIIII IIIIIII IIIIIII IIIIIl1 -

l lllllI IIIllll llIIIII 'lIIIIII i IIIIIll IIIIIll llll1II IIIIIII i llllll IT lill lilllli 111llll l

l l

Item Number Type Module # Area Resp Action Due Date Updt/Cisout Rpt # Date 0/M/ Closed l

$6 -l0l7 l-l ol f l llll, P 3lcl 2V oli - of _

s l7 ll- l l-l l l llllllq ll l MM DD YY MM DD YY Ori ginator - . Modiffer/C' lose- .

!

laloisIn , silIII I III I II Descriptive Title e a is Edf + v .s i i u o l i - > r a e r a d de e s s e d i u & - FN e n E c r r u e m e v l 6 - e i n i- ,

's, e P r o o e d , -

e a n d L ,, a n G e r - < < v / n r i - - c 4 i n . +1's #

f m . - _ , u - , , . - , - . t ,<,,-y,..,,,w

M-7 i

_ _ _ _ _ _ _ _ _ _ _

l

v t e Inspection Report - Salem - 1986009
80 \| 81 \| 82 \| 83 \| 84 \| 85 \| 86 \| 87 \| 88 \| 89 \| 90 \| 91 \| 92 \| 93 \| 94 \| 95 \| 96 \| 97 \| 98 \| 99 00 \| 01 \| 02 \| 03 \| 04 \| 05 \| 06 \| 07 \| 08 \| 09 \| 10 \| 11 \| 12 \| 13 \| 14 \| 15 \| 16 \| 17 \| 18 \| 19 \| 20 \| 21 \| 22 \| 23 \| 24 \|
1986 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 7(0) 9(0) 22(0) 23(0) 25(0) 26(0) 27(0) 28(0) 32(0) 33(0) 99(0) 26(0) 35(0) 99(0) Security Operator Exams Emergency Planning Other

IR 05000272/1986009

Text

REGION I

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