ML20133K283

Exam Rept 50-354/OL-85-29 on 850708-17.Exam results:8 Reactor operators,11 Senior Reactor Operators & 1 Instructor Certification Candidate Passed Oral & Written Exams.Three Candidates Failed Simulator Exam

ML20133K283
Person / Time
Site:	Hope Creek
Issue date:	10/12/1985
From:	Joshua Berry, Keller R, Kister H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
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ML20133K269	List: ML20133K266 ML20133K283
References
50-354-OL-85-29, NUDOCS 8510210154
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Text

i U. S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPCRT NO. 85-29 (0L)

FACILITY DOCKET NO. 50-354 FACILITY LICENSE NO.

LICENSEE: Public Service Electric and Gas Company 80 Park Plaza - 17C Neward, New Jersey 07101 FACILITY: Hope Creek Generating Station EXAMINATION DATES: July 8, 1985, to July 17, 1985 CHIEF EXAMINER: / Am u I f.or'Jingineer J PBerry,' Lead Re 4,m br.t-r Examihler

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REVIEWED BY:  ! /Bkk5 Robert M. KelTeY, Chief, Projects Section IC Date APPROVED BY [O f/ [I H(rry B. IRster, Chief, Projects Branch No. 1 Dat'e SUKvARY: Operator and Senior Operator Initial Cold License Examinations were conducted at the Hope Creek Generating Station from July 8 to July 17, 1985. .

Eight'(8) Reactor Operators, eleven (11) Senior Reactor Operators, and one (1) l Instructor Certification candidates were examined. All candidates passed the written and oral examinations. Three (3) Senior Reactor Operator candidates failed the Simulator examination. The candidates failing the Simulator examin-ation showed deficiencies in communication, plant direction and control, and the use of Emergency Operating Procedures.

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

1 R0 l SRO. l- Inst. Cert l l Pass / Fail i Pass / Fail i Pass / Fail I l I l l i I I I I IWritten Exam I 8/0 l 11/0 1 1/0 l l l l l 1 1 I I I l 10ral Exam -l 8/0 1 11/0 1 1/0 1 1 1 I I I

) i I I I I i ISimulator Examl 8/0 1 8/3 1 1/0 l l l l l l 4

l l 1 l l 10verall I 8/0 l 8/3 1 -1/0 I i

! I I- 1 I I l' I I i

1. CHIEF EXAMINER AT SITE: D. Lange,1st week, J. Berry, 2nd week i -

2. OTHER EXAMINERS: 'F. Crescenzo, USNRC A. Howe, USNRC (Trainee)

L. Banavitch, USNRC (Trainee)-

4 G. Robinson, USNRC (Consultant)

G. Sly, USNRC (Contractor, P.N.L.)

W. Cliff, USNRC (Contractor, P.N.L.)

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1. Summary of generic strengths or deficiencies noted:

Generic strengths were recognized by all examiners in the candidates' familizaration of control panels and inplant components.

This overall strength was also noted during the grading of the R0/SRO written examination.

2. All candidates were generally familiar with plant operating procedures and operating surveillance test requirements.

3. Personnel Present at Exit Interview:

NRC Personnel J. Berry, Chief Examiner D. Lange, Region I BWR Examiner L. Banavitch, Region I BWR Examiner (In Training)

A. Howe, Region I BWR Examiner (In Training)

F. Crescenzo, Region I BWR Examiner J. Lyash, Resident Inspector (Hope Creek)

Facility Personnel G. C. Connor, Operations Manager, PSE&G G. Meecht, Principal Nuclear Training Supervisor (Simulator)

S. LaBruna, Assistant General Manager, Hope Creek R. Salvesn, General Manager, Hope Creek R. Schaffer, Assistant Manager, Operations Training W. Gott, Principal Training Supervisor, Hope Creek

4. At the exit meeting, the Chief Examiner noted deficiencies in the lack of Surveillance tests that are approved for training on the simulator and the number of unsatisfactory simulator malfunctions.

Prior to the next set of cold license exams, to be conducted in October, these deficient areas were identified to be corrected.

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5. CHANGES MADE TO WRITTEN EXAM DURING EXAMINATION REVIEW:

See Attachment Attachments:

1. Written Examination and Answer Key (RO)

2. Written Examination and Answer Key (SRO)

3. Changes made to Written Examination during Examination Review i

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U.S. NUCLEAR REGULATORY CONI!55!0N REACTOR OPERATOR LICENSE EXAMINATION facility: umm. ce,ak Reactor Type: nun Date Administered: 7/ ins Examiner: r.nrane r- nnhin=nn Candidate: ______.

INSTRUCTIONS TO CANDIDATE:

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

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

% of Category  % of Candidate's Category Value Total Score Value Category oc n 25_a_ 1. Principles of Nuclear Power Plant Operation Thermo-dynamics, Heat Transfer and fluid Flow os n es n 2. Plant Design Including Safety and Emergency Systems 25 0 25.0 3. Instruments and Controls 25.0 _25.0 4. Procedures - Normal, Abnormal, Emergency, and -

Radiglogical Control 100.0 100.0 TOTALS Final Grade  %

All work done on this examination is my own. I have neither given nor received aid. ,

l Candidate's 5ignature

1.' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW (25)

'1.01 Using the attached " ROD WORTH COLD TO HOT FULL POWER" Figure:

a) Give two reasons why the rod worth increases from cold to point "A". (1.0) b) Give the major reason why rod worth increases from point "A" to point "B" (0.5) c) Oive the three parameters af fecting the curve from point "B" to point "C". (1.0) d) Indicate which parameter given in part c causes ,

the rod worth to decrease. (0.5) l 1.02 Indicate whether the following statements concerning Xenon behavior are TRUE or FALSE for your reactor.

(If any part of the statement is not true, mark the W

statement false.)  ;

a) Equilibriu= Xenon is reached faster at higher powers than at lower powers. (0.5) b) Equilibrium Xenon levels are directly proportional to power level. (0.5) >

c) The time to reach peak Xenon af ter shutdown var $es from about 6 to 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, taking longer to reach  !

peak Xenon for scrams from lower previous power levels. (0.5)-

d) An approximate value for peak Xenon after shutdown -

from 1000 power is 2.7% ak/k. (0,5) 1.03 Energy from fissions occurring in the fuel is transferred  ;

to saturated liquid water while the pressure is held constant. Indicate whether the following increase, decrease, or remain the same. Briefly explain your answer a) enthalpy of the coolant . (1.0) b) temperature of the coolant (1.0) ,

c) steac quality of the coolant (1.0)

CATEGORY CONTINUED ON NEXT PAGE t

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. 1 1.04 Your reactor has just reached criticality af ter a refueling outage. Because of a rod drive problem, one rod is then fully inserted and disarmed. Although the reactor is suberitical, the count rate does not fall to the original count rate prior to startup (i.e., when all rods were inserted). Briefly explain why. (2.0) 1.05 a) Briefly indicate the major mechanises which provide adequate Net Positive Suction Head (NPSH) for the Recirculation Pumps

1) during low power saturated operation (0.75)

11) during 100% of full power operation (0.75) b) For part b, which operating condition will provide the greatest NPSH? (0.5) 1 1.06 The Doppler coefficient of reactivity is important in the limiting of transients during reactor operation. For each of the conditions given below, indicate whether Doppler coef ficient becomes more negative, less negative, or remains the same. Briefly explain each answer.

a) Fuel Temperature increases (1.0) b) Core Void Fraction decreases (1. 0) 1.07 Assure the reactor is operating at 100% power and a trip of the recirculation pumps occurs. Indicate for each parameter listed below the initial change (increase, decrease, none) and briefly explain why the change occurs.

a) Reactor Power (one reason) (1.0) b) Reactor Water Level (two reasons) (1.0) c) Reactor Steam Dome Pressure (one reason) (1.0)

CATEGORY CONTINUED ON NEXT PAGE e

L 1.08 Your reactor has just reached criticality af ter a refueling outage and is placed on a hundred (100) second period.

Thereaf ter no rod movement or recirculation flow changes occur. As indicated below, several power increases with corresponding times were recorded. Indicate which (if any) were taken prior to reaching the heating range and which (if any) were taken if ter the heating range. Show all work.

a) reactor power increased by a factor of 2 in 80.5 seconds (1.0) b) reactor power increased by a factor of 5 in 160.9 seconds (1.0) 1.09 If you double the speed of a centrifugal pump, will the following increase by a factor of two, four, six or eight?

a) Discharge head (0.5) b) Power required by pump actor (0.5) c) Flow rate (0.5) 1.10 a) Why is the Linear Heat Generation Rate (LHGR) an important parameter during operation at high power levels? (1.0) ,

L b) Why have limits been prescribed for the Average Planer LHGR? (1.0) l

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t 1.11 a) As the production of Pu-239 increases in the core, ,

does the delayed neutron fraction increase or decrease? .

Briefly explain your answer. (1.5) i i

b) Briefly explain the major effect the change in the '

delayed neutron fraction has on reactor operation. (1.0)

I END OF CATEGORY i

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3 2 .' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS (25) 2.01 Consider the Reactor Water Cleanup (RWCU) System a) Indicate the three Reactor Coolant System locations that supply reactor water to this system (1.0) b) List four conditions that automatically close the Containment Isolation Valves (F001 and F004) (2.0) 2.02 Indicate whether the following statements concerning the High Pressure Coolant Injection (HPCI) System are TRUE or FALSE. (If any part of the statement is not true, mark the statement false),

a) Upon receipt of a HPCI initiation signal, the auxiliary oil pump starts supplying oil pressure to open the Turbine Control Valves once the Steam Supply Valve to Turbine (HV F001) is open. (0.5) b) The HPCI Booster pump provides NPSH for the HPCI main pump and provides cooling water for the Main Puep Seals and the Lube Oil Cooler. (0.5) c) When the HFCI Turbine Trip Push button is depressed, the RPCI turbine is tripped. When the push button is released and an initiation signal is present, the HPCI system will restart and align t' tnject into the core. (0.5) d) The HPCI suction will be automatically aligned to the suppression pool if there exists a low level in the Condensate Storage Tank or a high level in the Suppression Pool. (0.5) 2.03 Consider the Safety and Turbine Auxiliary Cooling System (STACs) a) List three different RHR System components cooled by the Safety Auxiliary Cooling Loops (SACS) (1.0) b) Indicate the normal makeup water supply source and the emergency makeup water supply" source (1.0) c) Besides low loop expansion tank levels and pump trips, list three conditions that cause the Turbine Auxiliary Cooling loop isolation valves to automatically close (1.0)

CATEGORY CONTINUED ON NEXT PAGE ,

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2.04 Consider the Emergency Diesel Generation a) Briefly explain the difference between the " NORMAL" mode of operation and the " EMERGENCY TAKE0VER" mode of operation. (1.0) b) State the four conditions and setpoints that will automatically start the Emergency Diesels provided the start permissives exist. (2.0) 2.05 Consider the Control Rod Drive Mechanisms a) Briefly describe how the velocity limiter accomplishes its purpose. (1,0) b) Is the accumulator needed to scram the drives when the reactor is at rated pressure? Briefly explain. (1.0) c) List two internal control rod drive problems that can occur from scramming a' control rod too f ast. (1.0) 2.06 There are four Reactor Feedwater Turbine runbacks.

For each runback a) Indicate the initiating event or events. (2.0) b) For each event, indicate the resulting feedwater flow status after the system has stabilized. (1.0) 2.07 Briefly describe how non-condensibles are removed from j the reactor head area a) during operation (0.5) b) during startup (0.5) 2.08 Consider the Core Spray System a) What function does the ECCS Jockey Pumps provide for the Core Spray System? (1.0) b) Assume that the core spray system is in Standby mode when the proper auto initiation signals are received. List the sequence of autoratic actions which then occur. Include any time delays or relevant pressures (assume off-site power is available) (1.5)

CATEGORY CONTINUED ON NEXT PAGE

2.09 Consider the Standby Liquid Control (SLC) System.

a) Where is the sodium pentaborate solution injected into the vessel? (0.5)

, b) Briefly describe what effect loss of instrument air would have on the SLC System. (1.0) c) Once the SLC System is initiated, injection of the entire contents of the SLC storage tank must be allowed to continue to completion (TRUE or FALSE 7) (0.5) 2.10 Consider the Containment Inerting and Purge System l a) What condition must be satisfied for your containment to be considered inerted?

1 (0.5) b) Af ter a LOCA there are two major processes that generate hydrogen. Briefly explain each process (Include necessary chemical equations and initiating events such as j temperature, etc.)

a (2.0)

I END OF CATEGORY i

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Y 3: INSTRUMENTS AND CONTROLS (25) 3.01 Consider the Wide Range Reactor Water Level Indicators a) For what reactor pressure vessel and drywell conditions are these indicators calibrated? (0.5) b) Would the water level reading increase, decrease, or remain the same, if a leak occurred in the Transmitter Equalizing Valve. Briefly explain (1.25) c) Briefly explain how and why a decrease in incoming feedwater temperature would affect level indications (1.25) 3.02 Match each of the items listed in Colurm I with the appropriate setpoint from Column II (2.0)

Column I Column II 5 : .t.

a. LPCI Injection Permissive 1. 82

b. NS 4-MSIV Low Pressure Closure 2. 380

c. SR/V set point 3. 460 (Hi ghes t Pressure)

d. RPS High Pressure Scram 4. WEE ~iC0

5. 1037

6. 1071

7. 1120

6. 1130 3.03 C* nsider the Source Range Monitors (SRMs) t) Briefly describe how a SRM channel discriminates against gamma's (1.5) b) When should the SRMs be fully withdrawn?

Briefly explain why this is necessary (1.5)

CATEGORY CONTINUED ON NEXT PAGE

3.04 Consider the Average Power Range Monitors (APRMs) a) Power is supplied to the APRM's by what system and at what voltage? (0.5) b) List the three rod blocks (include set points) associated with the Flow Unit. (1.5) c) Briefly explain why recirculation loop flow rather than core flow is used in the Flow Units. (1.0) 3.05 Consider the Reactor Building Ventilation Exhaust Radiation Monitor a) What type of detector is it and what type of radiation does it detect; (0.5) b) Indicate the three control functions it automatically performs upon reaching a high-high condition. (1.5) 3.06 For each of the lettered conditions given below, indicate which of the following occur: (If core than one action occurs, state the more severe action, i.e., half-scram is more severe than a rod block).

1) scram

11) half scram iii) rod block iv) no action a) one MSIV closes while reactor is at 50% power (0.5) b) During initial fuel loading, one SRM reads 8 x 105 eps (0.5) c) one Main Steam Line High Radiation Monitor inoperative (0.5) d) Recirculation Flow at 50% and reactor power at 80% (0.5) e) Loss of one RPS MG Set (0.5) 3.07 a) A manual keylock switch is provided in order to bypass the " Scram Discharge Volune High Level Trip." In what operational modes is this switch functional? (0.5) b) If a CRD accumulator trouble alarm occurs, what two conditions could this signify? (1.0) c) How can you distinguish which condition in part "b" is causing the alarm? (0.5)

CATEGORY CONTINUED ON NEXT PACE

/6 3.08 Consider the Nuclear Steam Supply Shutoff System (NSSSS) a) State the two conditions and their set points that autocatically cause isolation of the Reactor Water Sample Valves. Indicate whether one or both conditions are necessary to cause isolation. (1.5) b) A high drywell pressure will cause the NSS?S to initiate closure of valves in what two systems? (1.0) 3.09 As indicated in the procedure entitled Radiological Protection Program, the portable radiation instruments listed in Column I are available for use. Match the instrument in Column I with its proper use in Column II. (2.0)

Column I Col umn II a) Portable Ion Chamber 1. Beta-Gamma High Dose

'(RO-2A) Rate Surveys b) Portable Ion Chamber 2. Beta-Ganza Dose Rate (PIC-6A) Surveys (0-50 R/HR) c) Portable G-M Detector 3. Personnel Monitoring (Teletector) (Contamination) d) Portable G-M Detector 4 Bets-Garsa Low Dose Rate (RM-14) Surveys (0-50 Mr/Hr)

5. Gama-Dose Rate Surveys (0-1000 RHr) 3.10 Consider the Reactor Recirculation Flow Control System a) Briefly explain why the scoop tube must be properly positioned (between 18 to 281 travel) for the start sequence. (1.0) b) Before attempting to start the recirculation pump several electrical interlocks must be satisfied.

In addition to the scoop tube e, name two other interlocks that must be satisfied. (1.0) c) The Error Limiting Network prevents rapid changes in recirculation pump speed. Briefly explain how this is accomplished. (1.0)

END OF CATEGORY

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l 4 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY, AND RADIOLOGICAL CONTROL (25) 4.01 You are operating at full power and receive the following alarms:

CAS RADW CHAR 'IltTMT PNL 00367 and RADIATION HONITORING ALARM /TRBL In accordance with OP-AB.ZZ-127Q, OFF GAS SYSTEM - HIGH RADIATION a) State 3 probable symptoms (other than the alarms given above) (1.5) b) What immediate operator action must be performed?

(Assume a scram condition is not reached) (1.0) 4.02 For the case of a jet pump failure, list three symptocs that would be apparent on console indicators. (1.5) 4.03 During a norcal startup from cold shutdown to rated power, the following cautions are given. Briefly explain the reason for each crution.

a) During low flow conditions, feedwater flow to the reactor should be maintained relatively constant (0.5) b) When placing a RTP in service, a Turbine Bypass Valve should be open (0.5) c) As the FVH #6 outlet valves are opened, allow time between each valve opening (0.5) 4.0' Consider a Loss of Off Site Power a) What three systems initially are used for pressure and level control? (1.5) b) What is the source of power for the instrument bus? (0.5) c) What is the minimum length of time the instrument bus will be powered by the source of part (b)? (0.5) 4.05 Consider a Reactor Scram (OP-EO.ZZ-100) a) List four different means of verifying that all rods are inserted (2.0) b) What two functions are accomplished by switching the mode switch to SHUTDOWN? (1.0)

CATEGORY CONTINUED ON NEXT PAGE

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i 4.06 Several of the Emergency Operator Procedures define when

" Adequate Core Cooling" is assured and give three viable methods to accomplish this. List these three methods giving a brief explanation of each method. (3.0) 4.07 In accordance with SA-AP.ZZ-024(Q) RADIOLOGICAL PROTECTION PROGRAM a) What is the initial quarterly whole body dose administrative limit for a person over 21 years old with a complete NRC-4 form? (0.5) b) What is the recommended upper dose lindt for emergency exposure?

i) to save a life (0.5)

II) to save station equipment (0.5) c) In accordance with 10 CFR 20, what two conditions must be met to extend quarterly whole body dose limits to 3 REM? (1.0) 4.08 During refueling the following alarms occur:

FUEL POOL LEVEL HI/LO FUEL POOL COOLING SYS LEAKAGE HI FUEL POOL COOLING SYS TROUBLE a) List the four automatic actions which would occur (assume pool level continues to decrease) (2.0, b) In addition to ensuring that all appropriate autocatic actions are completed, what immediate operator actions are necessary? (1.0) 4.09 The reactor is at 10Z power during a reactor startup. The operator suspects that the rod he is presently withdrawing is uncoupled. What three possible indications does the operator have available to indicate an uncoupled control rod? (1.5) 4.10 Consider a condition of REACTOR HIGH WATER LEVEL (OP AB.ZZ-117Q) Assuming the reactor does not scram a) Will reactor power increase or decrease? Briefly explain.

(1.5) b) What immediate operator actions are required? (1.5) l l

CATEGORY CONTINUED ON NEXT PAGE

i3 4.11 Give two separate examples which define misoperation of an ECCS system and would justify placing an ECCS in manual mode.

(1.0)

END OF EXAM i

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ROD WORTH COLD TO HOT FULL POWER LM aC30 -

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a Ib CGlin T l oti GlICC T g 's ma v . 5/t Cycle ef ficiency * (;!ctwork out)/(Energy in) 2

, mg s = Y,t + 1/2 at 2 A = All A = A,e E = mc KE = 1/2 my a = (Vg - V,)/t PE = m9h i= en2/t1/2 = 0.693/t 1/2 v7 = Y, + a t " " '/t t

1/2erf = ((tjp)(t3 ))

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aE = 931 om I = 1,e-"*

P = P,10 g gy

. . I=I o 10-x/TVL Q = mCpat P = Po et /T g ge TYL = 1.3/u o Q = UAe T SUR = 26.06/T HVL = -0.693/u

?wr = M gan ,

SCR = 5/(1 - K,77)

SUR = 28p/t= + (s - p)T CR, = S/(1 - K,gg,)

T = (t*/o) + [(a - p)/fp) CR)(1 - K,773) = CR 2 (1 - k ff2)

T = t/(o - 8) '

M = 1/(1 - X,77) = CR;/CR, T = (s - o)/(to)

M = ( 1 - K,7 7,) /( 1 - K,77 3 )

• IKeff-II/Keff ' #'eff/K eff SOM = (1 - K,77)/K,77 t- = 10- seconds

, = ((t=/(T K,77)] + (i,f f/ (1 + T)) T = 0.1 seconds- ,

P = ( n y)/(3 x 1010) t = eN Idj=1d22 Id gj =1d22

. 2 R/hr = (0.5 CE)/d (meters)

Water Parameters Miscellaneous Conversions _

1 gal. 8.345 lbm.

1gaj.=3.78 liters 1 curie = 3.7 x 1010dps 1 ft = 7.48 gal. 1 kg = 2.21 lbm ,

Density = 62.4 lbm/ft3 I hp = 2.54 x 103 8tu/nr Density = 1 gm/c.d 1 nw = 3.41 x 106 8tu/hr Heat of vaportration = 970 8tu/10m lin = 2.54 cm Heat of fusion = 144 8tu/lbm *F = 9/5'C + 32 1 Atm = 14.7 psi = 29.9 in. Hg. *C = S/9 ( *F-32) i k

G. E. Robin:on />F C 7/85 ANSb'ER SHEET HOPE CREEK R.O. EXAM

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSTER AND FLUID FLO (25) 1.01

Reference:

Reactor Theory, Vol. 3, Control Rod Worth, LP RXPH31-01, pg. 14 (1.0) a) From cold to point "A" rod worth increases due to the neutron flux and thermal dif fusion length increasing.

(0.5) b) Tro: points "A" to "B" (the reactor heacup is in its later stages therefore the neutron flux is relatively constant) the thermal diffusion length is still increasing (therefore rod worth increases but at a decreasing rate).

i (1.0) c) From points "B" to "C" neutron flux, diffusion length and pitch affect curve

(0.5) d) pitch (the neutron flux and thermal diffusion length increase at a slower rate than the increase in pitch) 1.02

Reference:

Reactor Theory, Vol. 3, Xenon, LP RXPH33-01, pgs. 7 and E (0.5) a) True (0.5 b) False (0.5) c) False (0.5) d) h fegri (1 7 */ 18 h Il M4*"}

1.03 Referenec: Thermodynamics, Vol. 1, Phases of Matter, LF HT and T22-01, pgs.10.11, and 12 (1.0) a) er.thalpy increases since energy is being added te the coclant.

1 (1.0) b) temperature remains the same since pressure is held constant.

(1.0) c) steae quality increases since the added energy is transferming liquid to vapor.

1.04

Reference:

Reactor Theory, Vol. 3. Suberitical Multiplication, LP RXPH21 - 01, pg. 7 (1.0) E-effective is less than one, but net as lov as it was when all rods were inserted.

(1.0) Therefore, the suberitical multiplication factor (1/1-K-effective) is higher for the one rod inserted case, thus the count rate is l

higher.

k

17 1.05

Reference:

Licensed Operator's System, Vol. 1, Reactor Recirculation System, LP 019-01, pgs. 73 and 74 (0.75) al 1) Recirculation pumps are located (57 ft) below the normal reactor water level. This provides adequate NPSH during low power saturated operation.

(0.75) ii) Feedvater flow provides the subcooling to the recirculation pump suction when operating at higher power levels.

(0,5) b) At Full Power, NPSH is greater (437 f t.)

1.06

Reference:

Reactor Theory, Vol. 3, Doppler coefficient, LP RXPH 30-01, pgs. 10 and 11 (1.0) a) The Doppler coefficient becomes less negative as the fuel temperature increases because the change in atomic motion is less per degree change or the chanFe in broadeninE of the peaks is less for higher fuel te=perature changes (either one is acceptable).

(1.0) b) The Doppler coefficient is less negative at lower void fractions because of shorter slowing down lengths thus neutrons spend less time in the resonance absorption region.

1.07

Reference:

Mitigating Core Damage Lesson Plane, Transient and Accident Analyses, LP 106-00, pgs. 27 and Fig. 15.3-1 (Also Q and A, No. I for LP-106)

(0.5) a) Reactor power decreases (0.5) Due to increasing voids (0.34) b) Reacter water level initially increases (0.33) Due to Swell caused by voiding and (0.33) Due to recirculation pump no longer taking suction on annulus (0.5) c) Reactor steam dome pressure decreases (0.5) Due to action of EHC with decrease in power 1.06

Reference:

Reactor Theory, Vol. 3, Reactor Period, L .P. RXPH24-01, pg . 4 r

=

t/In j-o 80.5 (0.5) a) T =

2

= 116 see In (0,5) T greater than 100 second therefore in heating range l

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

/ f' 160.0 (0.5) 99.97 sees b) T =

In 5

=

(0.5) T equal to 100 seconds, therefore below heating range 1.09

Reference:

Fluid Flow, Pump Characteristics, Pump Head, Pump Laws LP-FF06-01, pgs.13 and 14 ,

i (0.5) a) four  :

(0.5) b) eight (0.5) c) two  ;

1.10

Reference:

Heat Transfer, Linear Heat Generation Rate and ,

Bases, LP HT 6 T15, pg. 5 and Average Linear Heat Generation Rate, HT and T16, pg. 3 (1.0) a) LHGR is the parameter used to monitor the internal cladding ,

stresses and thereby keep them at an acceptable level lt L

I (1.0) b) Limiting the Average Planer LHGR ensures that for a postulated LOCA a peak cladding tetperature of 2200 des F is not reached 1.11

Reference:

Reactor Theory, Vol. 3, Prompt and Delayed Neutrons, LP RXPH 23-01, pg. 9 and Reactor Period RXPH 24-01, pg. 9 I (0.5) 'a) decreases (1.0) The delayed neutron fraction for Pu-239 is less than the delayed neutron fraction for U-235. As Pu increases in the core, the percentage of fissions of Pu increases and the percentage of U-235 fissions decrease, the delayed neutron fraction ,

decreases ',

(1.0) b) The decrease in delayed neutron fraction results in a shorter reactor period for the same amount of reactivity inserted 1

- . - - , . - . ,,,..,-.,' ---;n . - , ------ --,-... . . - -

If

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS (25) 2.01

Reference:

Licensed Operator's Systems, Vol. 7. Reactor Water Cleanup, LP 21-01, pgs. 8, 9, and 66,dho pp Ji Ara 13 (0.33) a) Recirculation Loop A (0.33) Recirculation Loop B (0. 34 ) Bottom Head Drain Line (2.0) b) Any four (0.5 pts. each)

.-38 inches reactor water level , ,g gg /,,

Equipment Room High Ambient Temperature q y ey, Equipment Roon High Dif ferential Temperature ' pq re,e ed y desa RWCU High Differential Flow RRCS Initiation 2.02

Reference:

Licensed Operator's Systems, Vol. 3 High Pressure Coolant Injection System LP 26-01, pgs. 33, 39, 68 and 79 (0.5) a) False (Turbine Stop Valve not Steam Supply Valve)

(0.5) b) False (Barametric Condenser not main pu=p seals)

(0.5) c) True (0.5) d) True 2.03

Reference:

Licensed Operator's System, Vol. 6, Safety and Turbine Auxiliary Cooling System LP 80-01, pgs. 10, 12, and 14 ,ff EI (1.0) a) Any three of the items below RER Heat Exchangers (2)

RHR Seal Coolers (4)

RHR Motor Bearing Coolers (4)

RER Pump Room Coolers (4)

(0.5) b) Normal Makeup - Demineralized Water System (0 . .' ) emerEency makeup - Service Water System ' fdfE l Nf c fe5 8n C%r W ]

(0.34) c) JLOCA  ; e I, g e v (Solooes tac 5 Me (0.34) Auj j LOP L sm (0.33) from SACS

% vet ()Lov Operating La L. Ace v m Loop e tat evFlow Pres to

s. TACs l 2.04

Reference:

Licensed Operator's System, Vol. 6, l Emergency Diesel Generators, LP 68-01, pgs. 56, 71 (1.0) a) " NORMAL" mode allows DG control selections from the contrel race. "D'F.RGENCY TAKE0VER" locks out contrel from the control rooc and allows the remete generator control panel to take over control if necessary.

l l (0.5) t) Reactor Water Level; -129 inches (0.5) High Dryvell Pressure; +1.68 psig (0.5) Both normal and alternate feeder voltage 1 94% and bus voltage 1 70%

(0.5) Normal and alternate feeder i 92%

a

. y'0

}

2.05 Refarence: Lic2nstd Oparator 's Systems , Vol .1, Control Rod Drive Pkchanist, LP-05-01, pgs . 10i, 22, and 30 (1.0) a) The shape of the velocity lindter is such that the falling action of the blade creates a large pressure drop across the velocity limiter j (1.0) b) (0.25) NO (0.75) Internal ball / check valve directs reactor water to bottom of piston (1.0) c) '(0,5) Index Tube Damage ,, g,, ,p (0.5) (Bellville) washer and(graphitar) seal damageJ rtese tteec 2.06

Reference:

Licensed Operator's System, Vol. 4 '[

Reactor Feedwater System, LP-56-01, pgs. 49, 50, and 51 ,

(0.5) One RFFT, Trips (0.25) RTPT Runback to 85%

(0.5) One Secondary Condensate (0.25) RFPT Runback to 801 (0.5) One Pricary Condensate (0.25) RTPT Runback to 701 (0.5) ATWS Signal (0.25) RFPT Runback to 01 or F'""'"""

(High Pressure 1071 psig specs ,p 9,vena 29ee8#" l and APRM on scale after (LPz4 we py 25) ,

25 seconds) b 2.07

Reference:

Licensed Operator's Syste=, Vol. 1 Reactor Vessel and Internals, LP 01-00, pg. 38 (0.5) a) From Heat Vent te main steam line (0.5) b) Fron Head Vent to RAD Waste Syste= Cbelow ***f) 2.05

Reference:

Licensed Operator's Syste=s, Vol. 3 Core Spray Systen, LP 27-01, pgs.14 and 20 Eces (1.0) a) hMR Jockey Pu=ps ensure the CS piping remains full, thereby preventing water hammer upon initiation (0.5) b) Pumps start after a 10 second delay (0.5) Injection valves (F004 and F005) receive oper signals but will not open until 461 psig r-pr:g g (0,5) Kini'ait va' e (F031) closes after flow is abeve low flow se tia i ,1 O

DI 2.09

Reference:

Licensed Operator's System, Vol. 12 Standby Liquid Control, LP 23-01, pgs. 29, 37, and 39 (0.5) a) "A" Core Spray Header (1.0) b) Loss of SLC level indication in the control room and level annunciation via "SLC TANK TROUBLE" (Recraf cam e teact 'a#'**""

na dp ce tt smel * *r (0.5) (uk L P capedd by sef) b* bbl

• c) 3eme lii k ********,9 sq c orgo ye.a ctcc) Sep pc /t-s) 2.10

Reference:

Licensed Operator's Systems, Vol. 3, Containment inerting and Purge System, pgs. 8 and 9 (0.5) a) Concentration of oxygen must he less than or equal to 4% by volume (1.0) b) Zircaloy - Water Reaction Zr + 2H y 0 ZrO2 + 2ii2 + Heat greater than 1600'T +- seee

• f (1.0) Radiolytic Decomposition of Water radiation 2HO 2H2+O2 i

i l

29

3. INSTRUMENTS AND CONTROLS 3.01

Reference:

Licensed Operator's Systems, Vol.1 Reactor Vessel Instrumentation LP 02-01, pgs 24, 4 7 and 48 (0.5) a) Calibrated for 1000 psig RPV and 135*P in dryvell (0.25) b) Increase (1.0) An Equalizing Valve leak causes the reference and variable legs to equalize. The D-P cell senses zero D-P and yields

.a high level signal (0.25) c) Indication would read higher (1.0) Colder Water causes a density increase in the downcomer region (below the elevation of the feedwater spargers) 3.02

Reference:

Licensed Operator's System, Vol. 1 Reactor Vessel Instrumentation LP02-01, pg. 39

Reference:

Licensed Operator's System, Vol. 3 Residual' Heat Removal, LP 28-02, pg. 80

Reference:

Licensed Operator's Systen, Vol. 4 Main Stea: Systec, LP 46-01, Pig.1 (0.5) a) LPCI Injection Fercissive (3) 460 psig (0.5) b) NS4-MSIV Low Press. Closure (4) bbfpsig (0.5) c). SRV set peint (Highest Press.) (E) 1130 psig (0.5) d) RPS High Pressure Scra: (5) 1037 psig 3.03

Reference:

Licensed Operator 's Syste=, Vol . :

Source Range Monitering, LP 13-01, pgs. 10, 27, and 31 (0.75) a) Gac=a pulses are usually much smaller than those produced by neutrons (0.75) Therefore, a discriminator counts pulses of.a given magnitude

, (neutron produced pulses) and does not count lover voltage l pulses (gamma produced)

(0,5) b) When all IRMs are on Range 3 or greater l (1.0) To prevent damage to the fission chamber caused by heat

frc the pewer produced inside the detector er es kudk 'dekekv hfa t

y w c-~ - n -

-4 m w w n ,---

0 3.04

Reference:

Licensed Operator's System, Vol. 2 Average Power Range Manitor, LP 16-01, pgs.13,15 and 31 (0.5) a) RPS 120 Volt Distribution (0.5) b) Upscale - 108* of rated flow (0,5) Comparator OFF Normal - 10% difference between flow unit outputs (0.5) Inoperative - Unit mode switch out of operate or module unplugged (1.0) c) Recirculation loop flow much steadier so chance of spurious scrams reduced 3.05

Reference:

Mitigating Core Damage Lesson, Plans Process and Area Radiation Monitor Response LP 105-00, Table 1 and Table 2 (0.5) a) Beta - Scintillator (0.5) b) $solate nor=&1 ventilation of reactor building (0.5) se: t ' Isolate primarv containment purge exhausts (0.5) Initiates FRYS 9.s h s acs pasp+ A r 0 $P -E e f t fG T Y .

em am fcos 39 f* +, w ( L P s4 sTan kt. lo S 4 Y b1 Ie S* Y) 3.06

Reference:

Licensed Operator's Systen, Vol. 2 Reactor Protection Syster, LP 22-02, pgs. 9, 37, 45, and 49

Reference:

Licensed Operator's Systems, Vol. 2 Rod Elock Monitor Syste: LF 17-01, Table 1 (0.5) a) No Action (0.5) b) Full Scra:

(0.5) c) Half-Scrat (0.5) d) Red Eleck (0.5) e) Half Scra:

3.07

Reference:

Licensed Operator 's Sys tems , Vol. 2 Reactor Protection System LP 22-02, pg. 32

Reference:

Licensed Operator's Systec, Vol. 1 Reactor Manual Control System LP 07-01, pg.10 (0.5) a) SHUTD0'4 OR REFUEL (1.0) b) Low nitrege pressure or high water levc1 1r. the accu =ulator (0.5) c) of Gy carceu gee vn p a,,, f,c, , zg p x,, ma u p c. S S . u sag...e s . .., w 9 lea Lo g c. 25 /k cv9 fn44 o5 kss 1/ar &fu ', &

frest luo s Casa. o f skam -

i

M 3.08 RafGrcnc2: Licsniad Oparctor's Systca, Vol. 3  ;

Nuclear Steam Supply Shutoff System LP 45-01, pgs. 19, 36, and 37 (1.5) a) Either low reactor water level (0.4),

-38 inches (0.2) g (0.2) High main steam line radiation  !

(0.4) 3 x hTPE (0.3)

(0.5) b) Residual Heat Removal System (0.5) Traversing In-core Probe System 3.09

Reference:

Procedures, Vol. 1 SA-AP.ZZ-024(Q)

Radiological Protection Program, Fig. 2 (0.5) a) RO-2A (2) Beta Gamma (0-50 R/Hr)

(0.5) b) PIC-6A (5) Gamma (0-1000 R/Hr) ,

(0.5) c) Teletector (1) High Dose Rate (0.5) d) RM-14 (3) Personnel Monitoring 3.10

Reference:

Licensed Operator's System, Vol. 1 Recirculation Flow Control, LP-20-01, pgs.19 and 43 (1.0) a) This position results in the 60% unload speed condition to .

Eenerate " break away" starting torque An bue m Cl*W O N ' M*(

(0.5) b) M/G set drive motor breaker sn.e/,r>, de let, e y e m. ,

'* 2 (0.5) Generator field breaker d<sd.og e. ata clestt i est pne e,i hp s bnatr> h *!**4g c,.p /4) < 2 m W r

(1.0) c) The errer network compares the demand signal to the actual generator speed (Tachometer output)

If the error signal is large, it limits the magnitude

of the error signal to the speed controller (range of 6.25 to 8%)

l

<A c%

4 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY, AND RADIOLOGIVAL C0h7ROL (25) 4.01

Reference:

OP-AB.ZZ-127(Q)0FF-GAS-HIGH RADIATION, pg. 1 (0.5) a) Main steam line radiation monitor activity increasing (0.5) Off gas activity increasing (0.5) North plant vent stack radiation monitor activity increasing (1.0) b) Reduce reactor power as necessary to maintain the off gas activity less than the high high alarm setpoint.

4.02

Reference:

OP-AB.ZZ - lll(Q) JET PUMP FAILURE, pg.1 (0.5) The indicated recirculation loop flow dif f ers by more than 101 from the established pu=p speed-loop flow characteristics (0.5) The indicated total core flow differs by more than 10% from the established total core flow value derived froc recirculation loop flow measurements (0.5) The indicated diffuser-to-lower plenue differential pressure of any individual jet pu=p differs from the established pattern by more than 101 4.03

Reference:

OP-IO.ZZ - 003Q Startup froc Cold Shutdown to Rated Power, pgs. 10, 13, and 20 (0.5) a) To minimize the thermal transients on the Reactor Vessel (0.5) b) To minimize the drop in Reactor Pressure when rolling the RFP Turbine (0.5) c) To let reactor vessel level return to normal 4.0;

Reference:

OP-AE,ZZ-135Q Loss o f Of f site Power , pgs . 2 and 3 (0.5) a) EPIC level and pressure (0.5) R/'lClevel C

(0.5) SRV's pressure (0.5) b) From 125 VDC via UPS (0.5) c) 4 hrs 1

. M 4.05

Reference:

Emergency Operating Procedures Lesson Plans OP-EO.ZZ-100 SCRAM, pgs. 5 and 6 (0,5) a) RSCS (0.5) Fuel Core Display (0.5) CRIDS (0.5) NSSS Computer edit (0.5) b) " backs up" scram in the RPS Logic (0.5) precludes MSIV isolation in C. ^^ ' ;' -

en 4=> pmm 4.06

Reference:

OP-EO.ZZ-102(Q) Containment Control, pg. 2 and 3 (1.0) Core Submergence - each fuel element is covered with water (1.0) Spray Cooling - each fuel element is sprayed with sufficient flow to remove all heat generated in the fuel bundle (1.0) Stea= Cooling - steam updraft up through the uncovered portion of each fuel bundle is sufficient to remove all heat generated in the fuel bundle 4.07

Reference:

Procedures, Vol. 1, SA-AP.ZZ-024(Q)

Radiological Protection Prograc, Fig. 4, pps. 39 and 46 (0.5) a) 1000 mre=/QTR (0.5) b) i 75 REM (0.5) 11 25 REM (0.5) c) The co=pany has on file a complete Forr NRC-4 for the individual containing all prior exposure history (0.5) The dose allowed by the extension when added to the individual's accumulated occupational dose does not exceed 5(N-18) REM where N is the individual's age in years.

4.08

Reference:

Procedures, Vol. 5, OP-Ah.ZZ-144 (Q)

Loss of Fuel Pool Inventory, pp. I 4sse (0.5) a) Skinner Surge Tank Makeup valve (HV-660e$ opens (0.5) Fuel Fool Pump (s) trip (0.5) Reactor Building Ventilation Suster isolates (0.5) FRVS starts i

(0.5) b) Terminate refueling operations (0.5) Evacuate the refuel floor l

I

4.09

Reference:

Procedures, Vol. 5, OP-AB.ZZ-103(Q)

Uncoupled Control Rod, pg. 1 (0.5) a) Rod overtravel alarm (0.5) CRD travels past position "48" when fully withdrawn (0.5) No response observed on the nuclear instrumentation when the CRD is being withdrawn 4.10

Reference:

Procedures, Vol. 5, OP-AB.ZZ-117(Q)

Reactor High Level, pg.1 (0.5) a) Increase in power (1.0) Increase in feedvater results in lower temperature water and i

thus more dense water entering core. This results in better moderation thus an increase in power (1.0) b) Transfer the level controller or RTP Turbine Controller to manual and restore vessel level to between Level 4 and Level 7 (0.5) Ensure all appropriate automatic actions are complete (Zer NPc & s s de M k thpo l'o re L Kio n, w d* 4s r n 11Le s p k )

4.11

Reference:

Emergency Operating Procedures Lesson Plans OF-EO.ZZ-100 SRAM, pg. 7 (0.5) inappropriate initiation av mser e 'ad'* '" "d ** 0' M (0.5) continued operation beyond the trip setpoints l

l l

27rac h thent .s Z

U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION Facility: Hope Creek Reactor Type: BWR Date Administered: 7/ /85 Examiner: C. E. Robinson Candidate:

INSTRUCTIONS TO CANDIDATE:

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

% of Catego ry  % of Candidate's Catego ry Value Total Score Value Cateoory ps.o 7c,n 5. Theory of Nuclear Power Plant Operation, Fluids, and Thermo-dynamics 25.0 25.0 6. Plant Systems Design,

. Control, and Instrumentation 25.0 25.0 7. Procedures - Norr.al ,

Abnorcal, Emergency, and Radiological Control 25.0 25.0 8. Administrative Pro-cedures, Conditions, and Limitations 100 Totals Final Grade All work done on this exa-ination is my own, I have neither given nor received aid.

Candidate's Signature i

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l

v

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS (25) 5.01 Using the enclosed Power / Flow (operating) Map, answer the following:

a) Insertion of control rods with constant pumps speed results in an increase in core flow. Briefly explain why. (1.0) b) What damage could be expected if operation is allowed below the Minimum Power Line? (1.0) c) What design feature precludes and prevents operation below the Minimum Power Line? (0.5) 5.02 Briefly explain how and why an increase in indicated water level would affect core flow when the core is being cooled by natural circulation. (1.5) 5.03 Your plant has been operating at 95% power for one month when the reactor scrams. You begin startup within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Briefly explain how and why control rod worths have changed when compared with a hot startup 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the scram. (2.0) 5.04 The reactor is being started up for the first time and is on a 100 second period. Moderator Temperature is at 170'F.

With no operator action, what will be the moderator temperature when the reactor period becores infinite. State any assumptions you make and SHOW ALL WORK. (2.5) 5.05 During a cooldown of the reactor vessel during a small break LOCA, the reactor pressure decreased from 885 psig to 535 psig in one half hour. Has your reactor cooldown lindt been exceeded? SHOW ALL WORK (2.0) 5.06 Consider the Minimum Critical Power Rate (MCPR) i

a) What is the purpose of the flow biasing factor, KF '

' (include the most limiting transient) (1.0) b) A limiting condition for operation is that the MCPR aust be greater than 1.20 x Kf when in operational condition 1 and thermal power is greater than 25%.

State the two transients that make this LCO necessary (1.5)

CATEGORY CONTINUED ON NEXT PAGE i

L.

bl 5.07 Consider an actual plant system with two pumps in parallel, one of which is running. The second pump is started. What will happen to the flow rate? Briefly explain your answer (1.5) 5.08 Consider the case where a control rod is dropped at 1%

power from full in to full out a) Which reactivity coef ficient will first cause the rste of power increase to slow down. Briefly explain < (1.0) b) Would the peak power be greater at beginning of core life or end of core life? Briefly explain (1.5) 5.09 a) Briefly explain how an increase in the temperature of the Circulating Water System affects condenser vacuum (1.0) b) According to Thermodynamic Theory, would the above increase in temperature af fect the plant thermodynamic efficiency? BRIEFLY EXPLAIN (1.25) c) List three other factors that a' Iso affect condenser vacuum (0.75) 5.10 Consider the indicated reactor water level a) Is the indicated level higher, lower or the saee as the actual water level directly above the core (inside the skirt)? BRIEFLY EXPLAIN (1.0) b) Briefly explain how and why indicated water level varies with a rapid increase in steam flow (2.0) 5.11 During a routine startup, a control rod is partially withdra n adding a given amount of reactivity changing K-eff from 0.90 to 0.93. Another rod is partially with-drawn and adds the same amount of reactivity again a) is the second increase in neutron population more, less or the same as the first increase in neutron population? Breifly explain (1.0) b) Is the time to reach equilibrium in the second K-eff increase greater, less or the same when compared to the time to reach equilibrium in the first increase?

Briefly explain (1.0)

END OF CATEGORY 5

d6 i

6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION (25) 6.01 Consider the Emergency Diesel Generator System a) In addition to the Emergency Stop Push Button, list four conditions that will initiate an emergency diesel generator trip under any conditions. (2.0) b) Besides the Demineralized Water Makeup and Transfer System, what system is required for support of the Emergency Diesel Generator System. Include what functions this system provides. (e dr. Tier 'e*'*, (1.0) 6.02 Consider the Redundant Reactivity Control System (RRCS) a) Both logics (of either A or B channel) must trip to initiate the RRCS protective actions (TRUE or FALSE) (0,5) b) The Alternate Rod Insertion parameters are sealed in for ten minutes af ter initiation (TRUE or FALSE) (0.5) c) Which of the four integrated logics are activated when the operator manually initiates the RRCS (1.0)

AFL d) What is the significance of an illuminated AAR VALVE OPEN status light? (1.0) 6.03 Consider the Main Steam Line Radiation Monitoring System a) In addition to the channel mode switch not in operate, name two conditions which would cause a channel to be inoperative (1.0) b) List four automatic actions which occur upon receipt of a high radiation trip signal (2.0) 6.04 Consider the Automatic Depressurization System a) During a small break loss of coolant accident 'all signals are present and valid for auto depressurization acuation. Water level then starts to increase above the low-low set point. What affect does this have on auto depressurization initiation if:

1) The timer has not timed out. BRIEFLY EXPLAIN (1.0)

11) The timer has timed out. BRIEFLY EXPLAIN (1.0) b) The vacuum breakers prevent condensation of steam in the SRV tail pipes from drawing torus water up into tail pipe. Give two reasons why drawing torus water up into the . tail pipe must be prevented (1.0)

CATEGORY CONTINUED ON NEXT PACE L

t' 6.05 Consider the Average Power Range Monitors a) What are the two requirements regarding LPRM inputs to an APRM for the APRM to be considered operable? (1.0) b) Briefly describe why a LPRM's sensitivity decreases with use and describe the design feature which reduces the sensitivity loss rate (1.5) c) If one flow unit in one trip system is bypassed, a rod block will occur (TRUE or FALSE) (0.5) 6.06 Consider the Residual Heat Removal System (RHR) a) What action must be taken to enable operation of "B" RHR pump from the Remote Control Panel? (0.5) b) List three different components that are cooled by water from the Safety Auxiliary Cooling Water System (1.0) c) When an LPCI initiation signal is received, the

- . RHR Head Exchanger Valves automatically open F,

(TRUE or FALSE) (0.5) d) What m=> valves are automatically closed by the NSSSS upon receiving an 82 psig (increasing) reactor pressure signal? t t c e n) (1.0) 6.07 Consider the Reactor Feedwater Pump Turbine (RFPT) a) Normal steam supply is from the moisture separator e xhaus t . Excluding the Auxiliary Boiler steam, what other source of steam is available to supply the RFPT's? (0.5) b) List two different operating conditions for which the steam supply referred to in the answer to part "A" would be used (1.0) c) During all of the four Reactor Feedwater Pump runbacks, if the R7PT reset pushbutton is pushed, the ranback will not stop (TRUE or FALSE) (0,5) 6.08 While operating at 90% power with the Feedwater System in three element control, one SRV fa'ils open. If no operator action is taken, will reactor vessel level be greater, less, or the same af ter the initial transient effects have settled when compared to the water level prior to the SRV failing to open? BRIEFLY EXPLAIN (3.0)

CATEGORY CONTI11UED ON NEXT PACE l

S 6.09 Air pressure is normally supplied to the Scram Inlet and Exhaust Valves to hold them closed to prevent a reactor scram from occurring. Scram pilot valves and Backup Scram Pilot Valves are provided to cause air to be removed from the Scram Inlet and Exhaust Valves upon receipt of a trip signals from the RPS System a) Hust the Backup Scram Pilot Valves be energized or deenergized to cause a scram? (0.5) b) From what source is the Backup Scram Pilot Valves powered? (0.5) c) If one of the Backup Scram Valves should fail to operate on receipt of the appropriate signal, what assures that the other valve can perform the necessary function? (1.0)

END OF CATEGORY 6 6

7 ., PROCEDURES - NORMAL, ABNORMAL, DiERCENCY, AND RADIOLOGICAL CONTROL (25) 7.01 Consider OP-ED.ZZ-102, Containment Control Procedure a) List the four entry conditions for this pro edure (?.0) b) What adverse effect would you expect if you operate outside the limits of the " Heat Capacity Temperature Limit Curve?" (See SC/L-1 curve at back of exam) ,

(1.0) 7.02 You are in the process of executing an emergency procedure  !

using the Procedure Flow Chart a) Another entry condition for that procedure occurs.

Briefly describe what action you must take and give two reasons why this action should be taken (2.0) b) An action step cannot be performed. What must you do? (1.0) 7.03 Briefly explain under what conditions an Emergency Core e Cooling System can be shutdown or placed in manual (1.5) 7.04 While operating at full power, a fire in the Control Room necessitates a Control Room Evacuation i a) List the three immediate operator actions required by the Control Room Evacuation Procedure (1.0) b) List the four items that should be checked before evacuating the control room if time permits (2.0) 7.05 According to Loss of Instrument Air (OP-AB7.Z-131(Q) a) Identify three autoratic actions that should have occurred as instrument air header pressure decreases to below 70 psig (1.5) b) Does a loss of instrument air require a manual scram? (0.5) 4 i CATEGORY CONTINUED ON NEXT PAGE r

1

7.06 In accordance with the Hope Creek Generating Station Radiological Protection Program (SA-AP.ZZ-024(Q) a) What is the recommended upper limit (WHOLE BODY) dose for the following situations:

i) Emergency dose to save a life (0.5)

11) Emergency dose to save station equipment (0.5) b) What is the maximum quarterly dose a person is allowed to receive according to 10 CFR Part 20? (Include all relevant requirements associated with receiving this quarterly dose) (1.5) c) Whose apptoval (two pecple) must be obtained prior to reaching the quarterly dose indicated in part "b"? (0.5) 7.07 During refueling operations at least two source range monitor (SRM) channels must be operable a) Where cust the required SRM detectors be located? (1.0) b) For initial loading and startup the count rate may be less than 3 CPS if what two conditions are cet? (1.0) 7.08 Consider OP-IO.ZZ-007(Q), Operations f rom Hot Standby (MSIV's closed) a) Briefly explain why during low flow conditions the reactor makeup should be maintained relatively constant (0.5) b) Briefly explain why there should be no vacuut on the Main Condenser p rior to removing the Main Turbine Steae Seals (0.5) c) What two reactor conditions define Cold Shutdown?

(Operational Condition 4) (1.0) 7.09 Consider the case of a Loss of an RPS Channel (OP-AB.ZZ-110(Q) a) What four operational parameters *must immediately be monitored for any change by the operator? (2.0) b) When positioning the "RPS HC SET TRANSTER SWITCH," what caution must be observed and why? (1.0)

CATEGORY CONTINUED ON NEXT PAGE

___.-___a

7.10 In accordance with OP-AP.ZZ-002(Q), Conduct of Operators, the shift complement can be reduced by one less than the minimum a) except for what position? (0.5) b) This does not permit shif t crew positions to be unmanned at shif t change (TRUE or FALSE) (0,5) 7.11 While operating at 70% of full power, the pressure regulator f ails upscale. Assume a scram condition is not reached and all appropriate automatic actions are complete. According to OP-AB.ZZ-120(Z),

Reactor Pressure Control System Halfunction, what immediate operator action must be performed and why is this action necessary? (1.5)

END OF CATEGORY 7 i

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8 ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS (25) 8.01 In accordance with the Hope Creek Cenerating Station Technical Specifications, define the following terms:

a) Shutdown Margin (1.5) b) Fraction of Limiting Power Density (FLPD) (1.0) 8.02 In accordance with OP-AP.ZZ-992(Q), Conduct of Operation:

a) Under what conditions may an unlicensed person manipulate the controls that directly affect reactivity? (0.5) b) Under what conditions may a licensed operator take reasonable action that departs from a license condition or technit c) specification? (0,5) c) Under what conditions is a second licensed senior reactor operator not required to be in the control room? (0.5) 8.03 According to the Hope Creek Generating Station Technical Specifications, what two restrictions apply to extending the time interval allowed between required surveillance tests? (2.0) 8.04 Section 3.4 of Technical Specifications provides guidance for Limiting Conditions for Operation concerning inoperable control rods. One specific action statement says "With more than eight (8) control rods inoperable, be in at Icast HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />." Assuming an adequate shutdown margin can still be maintained, what is the bases for this action statement? (2.0) 8.05 Technical Specifications required the Rod Worth Minimizer (RWM) to be operable below 20% thermal power a) What is the basis for not requiring the RWM above 20%

power? (2.0) b) Besides a manual scram, under what conditions may rods be moved in operational conditions 1 and 2 when thermal power is less than 20% and the RWM is inoperable? (1.0) 6.06 Briefly explain the dif ference between Safety Limits and Limiting Safety System Settinga (2.0)

CATEGORY CONTINUED ON NEXT PAGE l L

[

Ib 8.07 In accordance with OP-AP.ZZ-002(Q), Conduct of Operations, list two criteria that determine when procedures that control operations shall be followed step by step with the procedure present (1.0) 8.08 NOTE: USE THE ATTACHED SECTIONS OF THE TECHNICAL SPECIFICATIONS TO ANSWER THIS QUESTION. FULLY REFERENCE ALL SECTIONS YOU USE During the Midnight to Eight shift, with the reactor operating at 95% power, you notice that the "B" Recirculation Pump speed is approximately 8% faster than the "A" pump. In attempting to match the two speeds, you receive a scoop tube lock-out on the "B" Hotor Generator set. An attempt to reset the lock-out fails and the pump speeds are now 12% apa t. What must be done in accordance with the technical specification if this problem continues? (3.0) 8.09 NOTE: USE THE ATTACHED SECTIONS OF THE TECHNICAL SPECIFICATIONS TO ANSWER THIS QUESTION. FULLY REFERENCE ALL SECTIONS YOU USE During your shift, with the reactor operating at 95% power (steady ctate), Control rod 26-23 which is fully withdrawn cannot be moved with normal CRD pressure. Indicate what actions are required initially and if the probice continues (3.0) 8.10 NOTE: USE THE ATTACHED SECTIONS OF THE TECHNICAL SPECIFICATIONS TO ANSWER THIS QUESTION. FULLY REFERENCE ALL SECTIONS YOU USE.

With the reactor operating at 90% power, one of the acoustic monitors on an ADS safety relief valve fails.

Indicate what actions are required (2.0) 8.11 NOTE: USE THE ATTACHED SECTIONS OF THE TECHNICAL SPECIFICATIONS TO ANSWER THIS QUESTION. FULLY REFERENCE ALL SECTIONS YOU USE The Division 1 diesel is operational and is 30 minutes into a surveillance test when the air starting receiver fails. The maintenance repair team estimates a two day repair time a) Is the Diesel Generator inoperable according to your Technical Specifiestions? Briefly explain (1.0) b) Are all Division 1 ECCS Systems inoperable because of the Diesel Generator problem? Briefly explain (1.0) c) If at the same tire a Division 2 core Spray pump is out of service, what added implication does this have on your Technical Specification position? (1.0)

END OF EIAMINATION l

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I

p/s TABLE II-3-1 PROPERTIES DE SAWRATED STEM AND SAWRATED WATER (TE.9ERAWRE)

{'

voheme. ft've E ntnator. 8tw/tc (nt ocy. 8:w/te a r Waff e (vap 5ttom Watt' (vep $ttam Wa t t' (v4D Sitam g

er 'o r he h

'a is h, sq s,, s, 32 0 06S59 001632 3305 3305 -0 02 1075 5 1075.5 0 0000 2.1873 2 1873 32 35 0.09991 0 01602 2946 2948 3 00 1073 8 1076 8 0 0061 2.1706 40 0.12163 0 01602 2446 Pub 2.1767 35 8 C3 1071.0 1079 0 0 0162 2 1432 45 0.147u 001602 2037 7 2037.8 2.1594 40 13 04 10681 1061.2 0 0262 2.1164 53 0.17795 C.016:2 1704 5 1704 8 2.1426 45 18 C5 1055 3 1063 4 0 0311 2 0931 60 0.2541 0 016*,3 1207.6 1207.6 2.1252 50 2S 06 1059 7 ICS 7.7 0 0155 2.0391 2 0946 60 73 0.3629 C01601 815 3 85E 4 E; 38 C5 1054 0 1092 1 0 0745 1.9900 C$*{t C 016*.7 633 3 633 3 2 C6641 70 S*

48 C4 104! 4  !?f 5 4 0 0932 19436 2 C31F Otill 0 7.(13 4t31 451 1 60 10 55 C2 104* 7 1133 6 01115 1.89?3 09412 C0;E13 3b3 4 35; 4 2 C*.E 6 ' 93 l110 62 03 IC371 11051 0 1295 1.8533 1.9121 100 1.2753 C01617 265 4 2t! 4 77.95 1C314 1101 3 0.1472 1.6105 1.9577 110 123 1.E927 C01(23 203 25 233 25 87 97 1025 6 1113 5 0 1645

' 133 2.2233 C 01625 157 32 1.7653 1.9339 120 I

140 157.33 97 H 10198 1117.6 0.1817 1.7295 1.9112 133 2.5672 C.01629 122 95 122 00 107.95 150 1014.0 1122 0 01955 1.6910 1.8 D 5 3.718 0 01634 9705 97.07 140 160 117.95 100S 2 1126 1 0.2150 1.6536 4 741 0 01643 77.27 77 29 1.0665 150 127.96 10C2 2 11332 0 2313 1.6174 17; 1.64!7 160 5 913 C01645 62 C4 180 7.! ! !

62 06 13'97 996 2 1134 2 0 2473 1.5822 18295 001E51 50 21 50 22 170 ISO 146 CC 99*,2 1135 2 0 2631 1.5453 9.340 0 01657 40 94 40 95 1.8111 160 200 155 04 964 1 1142 1 0 2757 1 5145 11.525 C 01654 33 62 33 64 I.7934 190 210 16EOs 977 9 1lat 0 0 2540 14tia 14 123 C01671 27 $; 27 E2 1.7764 20c 17E 15 971 5 I145 7 02091 14539 1.76:3 210

. 212 14 696 0 01672 26 78 PE 53 '.83 17 9703  !!$3 $

220 17.165 0 0167E 23 13 0.3121 1(U7 1.7568 212 23 23 15 '.68 23 9652 lit) 4 0.3241 20.779 0.01625 19 364 1.4231 1.7442 220 243 19 381 198 33 958 7 1157.1 0.3385 24 968 001693 IE 304 1.39 2 1.7293 230 250 16 321 20645 952I 1163 6 0.3533 1.3609 29.825 0 01701 13 832 13 819 1.7142 240 218 59 945 4 1164 0 0 3677 1.3323 2i; 1.7000 250

!! 427 } * * !"Oi 11 74! I1 712 i 221 7{

270

• 41.656 931 f 11(7 4 C 3!if 13:43 1$!f? 260 CC1716 10C42 10 06; 235 95 9317 280 49 200 C 31725 6 627 8 644

!! 7C i 8 0 3560 1.27ti 1 6729 270 a 293 249 17 924 6 1173 8 04035 '

57.550 0 01735 7u3 7463 1.2531 1.6595 280 300 259 4 917.4 11768 04236 E7 005 CC1741 6 4A5 12238 1.6473 290 6 4CE 26i 7 91C 0 1175.7 04372 1.1979 310 1.6351 303 77.67 0 01755 5 609 5 626 320 283 0 902 5 1882 5 04506 89 64 0 01766 4 896 1.1726 1.6232 310 340 4.914 2904 894 8 1185.2 0.4640 1.1477 117.99 0 01767 3.770 3185 1.6116 320 360 311 3 878 8 1190 1 0 4932 153 01 0 01811 2 939 1.0990 1.5892 340 2 957 3323 8621 1194 4 380 195 73 0 01636 2 317 0.5161 1.0517 1.5678 360 2.335 353 6 844.5 1198 0 0 5416 1.0057 1.5473 3s0 400 247.26 0 01864 iSul 420 18633 3751 825 9 12010 0 5647 306 78 0 01894 I a.60$ 1.4997 09637 1.5274 400 440 396 9 806 2 1203 1 0 5315 381.54 0 01926 11976 1.2169 0.9165 1.53S0 420 460 419 0 785 4, 1204 4 0.6161 466.9 0.0196 0 9746 0 9942 08729 I4890 440 440 441.5 763 2 1204 8 06405 566.2 0 0203 *- 0.7972 464,5 0 8299 84704 460 0 8172 739 6 1204.1 0 6648 0.7871 1.4518 440 500 680 9 0.C204 06545 0.6749 48 7.9 714.3 1202 2 520 812 5 0 6E93 0fu3 14333 540 0 023; 0535( 05595 512 C 657 0 1199 0 500 962 8 0 0215 0 u37 0 7133 0 7313 14146 520 560 ' 1I33 4 04651 5306 657.5 1194 3 07378 0 0221 0.3651 06577 1.3954 540 540 0.3871 562 4 625 3 1187 7 0.7625 13262 0 0228 0 2994 0 6132 1.3757 540 03222 5893 589 9 1179 0 03676 05673 1.3553 580 600 1543 2 0 0236 02438 02675 6171 550 6 1167 7 620 1766 9 0.0247 01962 0 8134 05106 1.3330 600 0 220E 646 9 506 3 640 , 2059 9 0 0260 0.1543 1153 2 08403 0 46E9 1.3092 620 0.1802 679.1 e54 6 1133 7 640 2365 7 0 0277 01166 08666 0 4134 12821 640 640 0.1443 714.9 3921 1107 0 0 8995 2708.6 0.0304 0.0808 0 3502 1.2498 460 0.1112 758 5 310.I 1068 5 0 9365 02720 1.2086 500 i 700 ,

3034.3 0.0366 0 0386 0 0752 822 4 172.7 9952 0.9901 0I490 L 705.5 < 32D8.2 0.0508 0 0.0508 906 0 0 9060 10612 0, 1.I390 700 1.0612 705.5 28

[

! pl EQUAT10t4 SHEET C = me v= s/t cycle efficiency = (,e:-cra out)/(Energy in) w = eg s=vt=

• 1/2 at2 E = mc2 KE = 1/2 mv a=(vf - V,)/t A = 1N -a t A=Ae PE = mgh V =V + at w e/t f o t = en2/t = 0.692/t 1/2 1/2 NPSM = Pin - I sat I I/2 eff = [(t 1/2)(tb)]

m. .AV

[(t gfy) + (t b)$

AE = 931 am I = ;o e-Ea Q = k:at Q = UAah I = [ e'"*

Pwr = W ah f

! = !o 10'*/

• T'.'; = 1. 3 /w .

F

• FoICsur(t) p = p e /It HVL = -C.693/w o

SUR = 26.06/T SCR = 5/(1 - Kgf)

CR, = 5/(1 - Kgf,)

SUR = 2 6e / t = + (s - c)7 cg (1 . ggf]) = c;p() . kg, )

g

~

T = ( t */c ) + ((e - e } /2 e ) p=

jf(1 , geff, = gg,j;p c i

I = 1/(s -8) s M = (1 - Agf }/(1 - K gg,

~= (f e)/Dc) SDv =

' '. - A . , , ,* t e ,. ,.

. c.

c *

(K ,,l)/Kgf = at,,,/t gf ta = 10'0 seccads a = 0.1 secones" c =

[(t=/(7K df !

• E8df/II*AII)

P = (:ev)/(2 x 1010) i gc; 2* 1 2C' t = oN gi e), , ; '2 2

R/hr = (0.5 CE)/c (meters)

NPSH = Statit heae - h1-F ga. R/hr = 6 CE/c2 ('eet';

Wa:er Parameters Miscellareous Conversions 1 gal. =

Igaj.=8.345lbe.

2.78 liters I curie = 3.7 x icICc s 1 ft- = 7.4E get. .

1 9 = g,;) l ar ,

Density = 52.4 Ic. ./f t- i np == 2.i

• IC6- E t.ti n e 1 ma Density = 1 ge/cm3 2.41 : 10 Btu /nr Heat of vaporization = 970 Btv/;be Itn = 2.5c cm

• eat of fusice = la: Bit /lbm =F = s/i=: . 32 l A t e. = 14.7 psi = H.i in. Hg. = C = 5 /9 ' = F . 3 2

y C. E. Robinson f

7/ /85 ANSWER SHEET SRO EXAM HOPE CREEK PLANT

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS AND THERMODYNAMICS (25) 5.01

Reference:

Licensed operator Systens, Vol.1, Reactor Recirculation Flow Control LP-20-01, pas. 51, 53 (1.0) a) Reduced power lowers the resistance to flow by lowering the core and separator pressure drops, therefore core flow increases (1.0) b) Recirculation Pump and Jet Pump Suction Cavitation (0.5) c) 20% feedwater flow interlock enforces the minimum power line 5.02 R(ference: Mitigating Core Damage, Recognizing Inadequate Core Coolling LP-100-00, pg. 8 (0.5) Increase in water level will increase core flow (1.0) As annulus level is raised the natural circulation flow driving head increases and thus core flow increases 5.03

Reference:

Reactor Theory, Vol. 3, control Rod Worth LP RXPH 31-01, pg. 14 and Xenon LP RXPH-33-01, pg. 5

( 0. 5) The rods near the center of the core will be reduced in worth (0.5) The edge rods will be increased in worth (0.5) After sustained 95% power operation, Xenon builds up in areas where flux was highest thus essentially poisoning the center portion of the core. Therefore during startup, the flux is highest at edges of core (0.5) Rod worth is proportional to (the squar'e of the) flux.

Therefore rod worth is highest for the edge rods and conversely lower for the center rods

5.04

Reference:

Reactor Theory, Vol. 3 Prompt and Delayed Neutrons LP RXPH 23-01, pg. 8 i and Transparency E Reactor Period LP RXPH 24-01 Transparency 6 Reactivity Coefficients and Defects RXPH 26-01, pg. 5 (0.5) At BOL Beff = 0.0068 A = 0.08 sec ~1 a(mod) = -1 x 10-4 ok/k/degF Seff-p 6*ff

, p =

Ap 1+AT l

~

(1.0) p =

0.0068/(1+0.08x100) = 7.6x10 ' ok/k Note: Second part will be graded independent of the first part Change in mod. temp. = 7.6 x 10-4 x (+1x10-4) = 7.6 degF l

(1.0) Moderator Temp. = 170 + 7.6 = 177.6 degF 5.05

Reference:

Thermodynamics, Vol. 1, Steam Tables HT&T25-01

Reference:

Tech. Specs., pg. 3/4 4-19 Referring to the Steam Tables (1.5) 885 psig = 900 psia = 532 degF Sat. Temp

$35 f.fo 417 56m3 psig = les psia = 4Wt degF su St //c (532 - 420) = R degT/ half hour or &M degF/hr (0.5) Yes, the cooldown limit of 100 degT/hr has been exceeded

! (Sk , t%. os s e ,,s of e arko=ed tse e tec "r drop hos ste! /t/ he*~ t* " *0*0 5.06

Reference:

Heat Transfer LSSS, LCO, ar.d Bases Associated with Critical Power, LPHT&T 14-00, pgs. 4 and 5

Reference:

Tech. Specs., pg. B 3/4 2-4 (1.0) a) The purpose of the Ky factor is to define operating limits at other than rated core flow conditions and to assure that the Safety limit MCPR will not be violated during a flow increase transient resulting from a motor-generator speed control failure (0.75) b) 100*F loss of feedwater heating at high never i

(0.75) rod withdrawal error at power l -

[

b

- 5.07 Reference Fluid Flow, Beonoulli's Equation, Fluid Friction, Head Loss LP FF03-01, pg. 15 i

(0.5) the flow rate is less than twice the original flow (1.0) The frictional head loss increases (as the velocity squared).

Therefore the resistance to flow increases and reduces the flow rate to below twice the initial flow rate i

5.08

Reference:

Reactor Theory, Vol. 3, Doppler Coefficient LPRXPH 30-01, pgs. 8 and 12 (0.25) a) Doppler coefficient or fuel temperature coefficient i

(0.75) An increase in fuel temperature causes an increase in i the probability of (U-238) absorption, thus reducing the probability of a neutron causing fission. NOTE: Any reasonable answer will be accepted (0.5) b) Beginning of Core Life (1.0) The doppler coefficient becoces more negative froe BOL to EOL due to buildup of Pu-240 (and fission products with pronounced resonances) 5.09

Reference:

Th(rnodynamics Student Handout, Section 8, Application of General Energy Equation and Heat Exchange Operation, Section 14, Plant Efficiency (1.0) a) The condenser acts as a saturation system, therefore the higher the temperature, the higher the pressure, the poorer the vacuu=

4 (0.25) b) The plant thermodynamic ef ficiency goes down (1.0) Because of the increased condenser temperature, the amount of work obtained from the steam as it passes through the turbine is reduced. Therefore, ef ficiency is reduced (0.75) c) Any three of the below (other appropriate answers will be accepted)

Non-condensable gases Circulating Water flow rate .

Back pressure in the condenser exhaust system In leakage I

1 l

l f

i

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

+

5.10

Reference:

Licensed Operator's System, Vol. 1 Reactor Vessel Instrumentation LP 02-01, pgs. 45 and 50 (0.25) a) indicated level (7") higher than inside the skirt (0.75) At 100ll steam flow there is a (7" H O) 2 pressure drop across the steam dryers (0.5) b) As steam flow increases, pressure decreases (0.5) The saturated water in vessel is suddenly superheated which results in an increase in voids (0.5) The mass of water above the variable leg tap increases (total mass of water remains constant) causing an increase in variable leg head (0.5) Indicated level increases (SWELL) 5.11

Reference:

Reactor Theory, Vol. 3, Suberitical Hultiplication, RXPH 21-01, pgs. 6 and 7 (0.25) a) more (0.75 core neutrons are continuously available to amplify the source (1/1 K-ef f) since' K-ef f is larger in the second case (0.25) b) greater (0.75) a longer time is required to reach equilibrium because more previous , generations of neutrons are significant thus requiring a longer wait time (NOTE: any reasonable answer will be accepted) i I

l I

T l

l

[

6.. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION (25) 6.01

Reference:

Licensed Operator's System, Vol. 6, Emergency Diesel Generators, LP 68-01, pgs. 72, 89 and 90 (0.5) a) fcenerator differential overcurrent (0.5) g, j Engine overspeed (0.5) E**" Low lube oil pressure i

(0.5) Bus differential overcurrent

% e e n e va f,v o mcuvume ( Goct op wieg) q 73 L P U ~

(0.25) b) Safety Auxiliaries Cooling System (0.75) Cooling medium to the D.G. intercooler water heat exchanger, jacket water heat exchanger and lube oil heat exchanger.

Also fill water to the fuel oil transfer system fuel oil day tank loop seal 6.02

Reference:

Licensed Operator's Systems, Vol. 2, Redundant Reactivity Control, LP 24-01, pgs. 9, 13, 46, 48 (0.5) a) TRUE *

(0,5) b) FALSE (0.5) c) Alternate Rod Insertion (0.5) Standby Liquid Control (1.0) d) At least one of the 4 ARI valves assigned to a RRCS channel solenoids is energized 6.03

Reference:

Licensed Operator's System, Vol. 3, Nuclear Stea= Supply Shutoff System, LP 45-01, pgs. 19 and 20  !

4 (0.5) a) low detector voltage

! (0.5) module unpluSged (0.5) b) Scram Reactor j (0.5) MSIV closure l

(0.5) I*1 Main Steam Line Drain Valve cl*5'rt (0.5) Gun Mechanical vacuum pump start interlock [$perationaltrip ga .,, a s n c we, n sor) n e< < < e. esa c ta se e, u o

\

~

6.04

Reference:

Licensed Operator's Sfstem, Vol. 3,

. Automatic Depressurization System, LP-29-02, pgs.16,17 and 21 (0,5) a) i) initiation will not occur (0,5) Because timer will reset and not restart until level drops below the set point (0.5) ii) Autodepressurization will continue (0.5) This condition will continue until completion or reset by operator (0.5) b) To prevent water hammer of SRV exhaust pipe and T-Quencher (0.5) To prevent pressure oscillations on exhaust side of SRV which could result in improper valve operation 6.05

Reference:

Licensed Operator's System, Vol. 2, Average Power Range Monitor, LP 16-01, pgs. 11, 14 and Table 1

Reference:

Tech. Specs. 3/4 3-5

Reference:

Licensed Operator's Systems, Vol. 2, Local Power Range Monitor System, LP 15-01, pg. 7 (0.5) a) 14 or more LPRM in an APRM channel '

(0.5) At least 2 LPRM inputs per level (0.5) b) Depletion of U-235 (1.0) U-234 absorbs neutron and produces U-235 to reduce the sensitivity loss rate i (0.5) c) FALSE f

6.06 keference: Licensed Operator's System, Vol. 3 Residual Heat Removal LP 28-02, pgs. 18, 80, 83 (0,5) a) The control Selector Switch must'be placed from NORM to EMERG (O.34) b) RHR Heat Exchanger A /g , g g, pu,,,p h (;,,,/4w (-F Fo -el  ;

(0.33) RER Pump Seal Coolers pek ,q ka, cC % % D' (0.33) RRR Motor Bearing Oil Coolers (0,5) c) TRUE (0.5) d) Shutdown Cooling Inboard suction valves (HV-F009) f # v -/ar#

(0.5) Head Spray inboard isolation valves (HV-F022)e4 #v r*

• 3 1/o % / ny b% F etS A T &

{%C ( T.S. Takte 37E~4 pp 3/4 3-C1]

i

6.07

Reference:

Licensed Operator's System, Vol. 4 Reactor Feedwater System, LP-58-01, pgs.13, 27, 28, 29 and 56 (0.5) a) Main Steam P

(0.5) b) During Startup (0.5) With two or more RFPT's operating near rated conditions, LP steam supplemented by HP steam (0.5) c) FALSE 6.08

Reference:

Licensed Operator's Manual, Vol. 5 j Reactor Water Level Control, LP $9-01, pgs. 10 and 11 (1.0) Decrease e, /c g,3 (1.0) The relief valves are upstream of the Main Steam Line Flow Restrictors so the eteam flow signal is reduced. This causes a permanent mismatch between steam flow and feedwater flow (1.0) This mismatch creates a (voltage) signal that is balanced by another si Enal produced by the desired water level minus ,j actual water level I

6.09

Reference:

Licensed Operator's Systems, Vol. 2, Reactor Protection System, LP 22-02, pgs. 12 and 13

(0.5) a) must be energized to cause a scram (Nek yms4 A m Ceu (4 k  !

en fta pukt se tl F * *

• Is a.  ?

(0.5) b) 125 VDC Station Battery g , ,. e 4 m m ) l (1.0) c) check valve installed around one valve to prevent s , a valve failure inhibiting the desired scram action l ev ,ne & s. o.c 6. Muy K5 *J le hk elec hecal l pews c t v. e n r c.r-e n )  ;

k i

I s * '

t r

r f

?

r I

l i

i i

I  ?

L

7 .* PROCEDURES - NORMAL, ABNORMAL, EMERGENCY, AND RADIOLOGICAL CONTROL (25) 7.01

Reference:

Emergency Operating Procedures Lesson Plans OP- EO . ZZ-102 Suppresion Chamber Level and Temperature Control Sections, LP 125-00, pgs. 5 and 10 (0.5) a) Suppression Chamber Water Temperature above 95'F (0.5) Suppression Chamber Water level outside 168.5 in-172.5 in (0.5) Drywell temperature above 135'F (0.5) Drywell pressure above 1.68 psig (1.0) b) Actuation of the ADS valves will result in probably unstable steam condensation at or near the SRV discharge 7.02

Reference:

Emergency Operating Procedures Lesson Plans Introduction to Emergency Operating Procedures LP 121-00, pg. 12 (1.0) a) Return to the beginning of the procedure and start through the flowchart again (0.5) The reason for this is plant conditions may have changed sufficiently to warrant a change in the response to a decision step or (0.5) the ability to continue performing a step previously requiring some action (1.0) b) Continue in the procedure 7.03 Re f e rence : Flow Chart OP-EQ. ZZ-100 Scram (0.5) By two independent indications (0.5) 1) misoperation in automatic mode is confirmed or (0.5) 11) adequate core cooling is assured 7.04

Reference:

Procedures, Vol. 16 OP-AB.ZZ-130(Q)

Control Room Evacuation, pgs. I and 2 (0.34) a) Scram the reactor (0.33) Trip the turbine (0.33) Ensure all appropriate automatic actions are complete (0.5) b) Check for neutron flux (decreasing,)

(0.5) Vessellevel(ncreasing)

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

4 1

I (0,5) Vessel pressure decreasing, (0.5) Main Generator tijtt breaker / open I 7.05

Reference:

Procedures, Vol . 5 , OP-AB.ZZ-13(Q)

Loss of Instrument Air, pg. 1 i (0.5) a) Standby Service Air Compressor starts ( 92 psig )

(0.5) Emergency Instrument Air Compressor starts (70 psig)

(0.5) service Air Supply Header Isolation Valve

! (HV-7595) closes (70 psig Instrument Air Pressure) -

1 (0.5) b) e ye s (Ja bejut.J dc /rk) [W no -E t '3' SN Y#

7.06

Reference:

Procedures, Vol. 1, SA-AP.ZZ-024(Q) i Radiological Protection Program, pgs. 44 (Fig. 4) and 46 (0.5) a) i) 75 REM

\

(0.5) ii) 25 REM ,

(0.5) b) 3 REM / quarter (0.5) does not exceed 5(N-18) where N is age of person i (0.5) and has a valid NRC-4 form i

i (0.5) Station Genersi Manager and VP-Nuclear-av

• . * " sad Rad,Refeder W P j 7.07

Reference:

Tech. Specs., pgs. 3/4 9-3 and 3/4 9-4 (0.5) a) One of the required SRM detectors located in the quadrant where core a?.terations are being performed (0.5) and the other located in an adjacent quadrant (0.5) b) The signal is greater than 0.5 CPS (0.5) The signal to noise ratio is gre'ater than 2.0 b

l g yen 1 - - - r g e - yy-----yey-p-t- -+%&_. -T*m--6 e--rD -' * -++"T e g ew-- ie+- e- g- ,- oc

jbb

' 7.08

Reference:

Procedures , Vol . 5, OP-IO.ZZ-007(Q)

Operation from Hot Standby (MSIV's Closed), pgs. 6,15, and 17 (0.5) a) To minimize the thermal transients on the Reactor Vessel (0.5) b) To prevent pulling in cold air along the Main Turbine shaft (0.5) c) Reactor Coolant Temperature less_than 200 *F (0.5) RPS Mode Switch in shutdown 1

7.09

Reference:

' Procedures , Vol . 5, OP-AB.ZZ-110(Q)

Loss of an RPS Channel, pgs. I and 2 (0.5) a) reactor power (0.5) reacter pressure (0.5) water level (0.5) Main Generator output (0.5) b) Do not switch to wrong alternate feed (energized RPS bus)

(0.5) A full scram will occur 7.10

Reference:

Procedures, Vol.1, OP-AP.ZZ-002(Q)

Conduct of Operations, pg.15 (0.5) a) SNSS .

(0.5) b) TRUE 7.11

Reference:

Procedures, Vol. 5, OP-AB.ZZ-120(Z)

Reactor Pressure Control System Malfunction, pg. 2 (0.5) Reduce the setpoint of the Maximum Combined Flow Limit Potentiometer, I as necessary (0.5) to close the By Pass Valves (0.5) in order to control steam flow and re, actor pressure i

4

, , , -- =~

. /

8.

• ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS (25) 8.01

Reference:

Tech. Specs., pgs. 1-3 and 1-6 (0.5) a) Shutdown margin shall be the amount of reactivity by which the reactor is suberitical or would be suberitical (0.5) Assuming all control rods are fully inserted except for the single rod of highest reactivity worth which is assumed to be fully withdrawn (0.5) and the reactor is in the shutdown condition; cold (i.e.,

68'F) and Xenon free (1.0) b) The FLPD shall be the LHGR existing at a given location divided by the specified LHGR limit for the bundle type 8.02

Reference:

Procedures, Vol. 1, OP-AP.ZZ-002(Q)

Conduct of Operations, pg. 12 (0.5) a) for training purposes under the direction of a licensed operator (0.5) b) in an emergency when this action is immediately needed to protect the public health and safety (0.5) c) cold Shutdown or Refueling Condit!o::

n/se f uts l<e, car k pagle,p n,4/ ge-fief "geen* u u w a s s

I

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8.03

Reference:

Tech. Specs, pg. 3/4 0-2 (1.0) A maximum allowable extension not to exceed 25% of the surveillance interval but (1.0) The combined time interval for any three consecutive surveillance intervals shall not exceed 3.25 times the specified surveillance interval 8.04

Reference:

Tech. Specs, pg. B 3/4 1-2 (2.0) The occurrance of eight inoperable control rods could be indicative of a generic problem and the reactor must be shutdown for investigation and resolution of the problem 8.05

Reference:

Tech. Specs, pgs B 3/4 1-3 and 3/4 1-16 (2.0) a) When thermal power is greater than 20% of rated thermal power, there is no possible rod worth which if dropped at the design rate of the velocity limiter could result in a peak enthalpy of 280 cal /gm (1.0) b) By verifying control rod movement and compliance with the prescribed control rod pattern by a second licensed operator or other technically qualified member of the unit technical staff who is present at the reactor controls 8.06

Reference:

Heat Transfer, Critical Power Safety Limits and Bases LP HT and T 13-00, pg. 4 (1.0) Safety Limits are established to protect the integrity of the fuel cladding, reactor pressure vessel and primary system piping (1.0) Limited Safety System Settings are devised to ensure that a Safety Limit is never reached 8.07

Reference:

Procedures, Vol. 1, OP-AP.ZZ-002(Q),

Conduct of Operations, pg. 19 (0.5) Where reliance on memory cannot be trusted and (0.5) Where operations must be perforved in a specified sequence 8.08 Reference Tech. Specs.

(1.5) Tech. Spec. 3.4.13a (pg. 3/4 4-5)

If the speed of the recirculation pumps cannot be restored within two hours to the specified limits (5%) then declare the recirculation loop of the pump with the slower speed not in operation and take the ACTION required by Spec. 3.4.1.1 1

v .

@l 3 RO dec let s ** B " hep /**f C* 'l b'f ** ' ' O* /' 5G "'N ~" 'S chsked (% 'inoy is see h .re y 1 % W **

• 5d M r ~ 0'O }

SR c now as eev co wd woh Tf. 3 9. t. I.nc k J h w .'e Au- SC g 44 power /Stev ny nort usJs ef F 1 SA.I.I-l 641 12 AM b* y 6Y (1.5) Since loop "A" is the slower of the two loops, it is not operable, while loop "B" is not operable because of the lock-out.

Therefore, section "b" of Spec. 3.4.1.1 is more limiting, thus the unit shall be at least in STARTUP within six hours and in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 8.09 Reference Tech. Spec.

Tech. Spec. 3.1.3.1 (pg. 3/4 1-3)

(0.5) Within one hour

1) verify that the rod is separated from all other inoperable control rods by at least two control cells in all directions (0.5) 11) Disarm the associated direction control valves hydraulically by closing the drive water and exhaust water isolation valves (1.0 ) Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> the SHUTDOWN MARGIN shall be verified acceptable with an increase allowance for the withdrawn worth of the unmovable rod. Surveillance 4.1.1.C '

e+Me se % ,,, e 4 /ro s J k( SA s4dcus wu&n "'d '2 ",

(1.0) Restore the unoperable control rod to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT SHUTDO*a'N within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 8.10 Tech. Spec. 3.4.2.1d (2.0) With one or more safety / relief valve acoustic monitors inoperable, restore the inoperable monitor (s) to operable status within 7 days or be in at least HOT SHUTDO*a'N within the ravt 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hou rs G'OTE : the fact that the valve is ADS is not the

. limiting cciteria) l l

f i

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- 8.11

Reference:

Technical Specifications (0.5) a) YES

./

(0.5) air pressure not greaterthan380psig(Tech. Spec.4.8.(.2.7)

(0.5) b) NO (0.5) When a component is determined to be inoperable solely because its emergency power is inoperable, it may be considered to be operable for purposes of satisfying its LCO if its corresponding normal power source is operable and all its redundant systems, etc. are operable (Tech. Spec. 3.0.5)

(1.0) c) In violation of Tech. Specs 3.0.5 Must perform the action statement of

1. At least STARTUP within the next six hours and

2. At least SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and ,

3. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> i

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10 JAN 583 i

15.E LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION f,,,,((

A 3/4 0-1 3/4.0 APPLICA8I(ITY............................................. ,

3/4.1 REACTIVITY CONTROL SYSTEMS .

3/4 1-1 3/4.1.1 SHUTDOWN MARGIN........................................

3/4.1.2 REAC TIVITY AN0MALIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . .'. . . . . . 3/4 1-2 3/4.1.3 CONTROL 2005 Control Rod Operability................................ 3/4 1-3 Control Rod Maximum Scras Insertion Times.............. 3/4 1-6 Control Rod Average Scram Insertion Times..... . ... .. ... 3/4 1-7 -

Four Control Rod Group Scram Insertion Times........... 3/4 1-8 Control Rod Scras Accumulators......................... 3/4 1-9 J Control Rod Drive coupling............................. 3/4 1-11 Control Rod Positi on Indi cation. . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-13 Control Rod Drive Housing Support...................... 3/4 1-15 l

3/4.1.4 CONTROL 200 PROGRAM CONTROLS Red wo rth Mi ni mi :e r. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-16 Rod Sequence Control Systas............................ 3/4 1-17 Mad Block Monitar...................................... 3/4 1-18 3/4.1.5 3TANCRY LIQUID CONTRCL SY57EM. . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-19 3/4.2 PmrYR DISTRIBUTION LIMITS 6 AVE 1tAGE PLAMAR LINEAR HEAT GENERATION RATE.............

3/4.2.1 3/4 2-1 3/4 2.2 APRM SETP01NTS......................................... 3/4 2-f.3 3/4.2.3 MINIMUM CRITICAL POWER RATI0...........................

3/4 2-# Y ~

3/4.2.4 LINEAR HEAT GENERATION RATE. . . . . . . . . . . . . . . . . . . . . . . . . l. . . 3/4 2 "f- N 7'* ")

iv I

%ePT. cAhw

10 JAN $83 E i! .

. LINTTING CONDITIONS FOR OPERATION AND SURVEILLANC{ RTQUIREMENTS SECTION PAGE a

3/4.3 INSTRUMENTATION .

3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION............ 3/4 3-1 3/4J.2 ISCLATION ACTUATION INSTRUMENTATION.................. 3/4 3-9 3/4.3.3 EMERGENCY CORE C00 LING SYSTEM ACTUATION INSTRUMENTATION...................................... 3/4 3-f7 3r l{  ;

3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION ,

ou ATW5 Recirculation Pump Trip Systas Instrumentation.. 3/4 3-46 11 Ene-of-Cycle Recirculation Pump Trip system ,

Instroentation...................................... 3/4 3 p ll 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION sv

{N5TRUMENTATION...................................... 3/4 3 ll cv 1/4.3.6 CONTROL R00 BLDCK INSTRUMENTATION.................... 3/4 3-H- ll v

l 3/4.3.7 MONITORING INSTRUMENTATION i Radiation Monitoring Instrumentation................. 3/4 3-57 Gr ll 72 Seismic Monitoring Instr oentation................... 3/4 3 ll w

. Meteorol ogi cal Monitaring Instrumentati on. . . . . . . . . . . . 3/4 3 ll

- Remote Shutdown Monitoring Instroentation. . . . . . . . . . . 3/4 3-70w ll r Accident Monitaring Instruentation. . . . . . . . . . . . . . . . . . 3/4 3d ll

~

' Sevrce Range Moni tars. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

3/4 34 ll Traversing In-Care Probe 5ystaa............... ...... 3/4 34 ll

• M r ' = ' r d ?c- ~ i d h t; = i-- Sy;t r ..............

O  ;-70 [{

' CL . : . L. ^.. .. ' . . " c. R ..................,........ /* " !!

. n ,e Cetecti on instroentat4 cn. . : . . . . . . . . . . . . . . . . . . . m 2 4 r ii so *~

3/4 3 ll .

i 1

• Laose-Pa

.Kasos.c.. Detection Wt %='4 Systes........ bhae N M h...in........ e;tew== A. n/4 3-17 '

E 3/4.3.8 TURRINE OVERSPEED PROTECTION SY57EM................... 3/4 3-g ll FTurwAM/MAant med W W a t pa,.%,mw%%- u w 3.n l 3/4. 3. s - ACTuAn0N INSTRuMENTAn0N. . . . . . . . . . . . . . . 3/4 3-er l V

a.. v-...s ,, ,

w cseew t

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

3

./ .

i

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i 1 E. E. I i

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS i t

SECTION fgg[

i 3/4.4 REACTOR CDOLANT SYSTEM +

i 3/4.4.1 RECIRCULATION SYSTEM  !

Recirculation Laeps.................................. 3/4 4-1 [

Jet Pumps.......................................*..... 3/4 4 d }

Recirculation Puers.................................. 3/4 4-JS i Idl e Reci rculation Loop Startup. . . . . . . . . . . . . . . . . . . . . . 3/4 4- M [

3/4.4.2 SAFETY / RELIEF VALVE 5 i

safety / Relief Va1ves................................. 3/4 4 47 -

sr,r4 ..,_...,.,._,,_, .,

- - ,_-..- ._ ,_. . ...__......... ... y, g ,

i 3/4 4.3 REACTDR CDCLANT SYSTEM LIAKAGE  ;

Leakage Detection Systans............................ 3/44Y N l Ope rati ona l La a kage. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4 i rs J 3/4.4.4 CNEMIST1Y............................................ 3/4 4-ht-3/4.4.5 SPECIFIC ACTIVITY....................................

3/44-h 3/4.4.5 PRCESURE/ TEMPERATURE LIMITS

['

M Reactor Csolant 5ystas............................... 3/4 4-38 l I

teactar Steam Dome................................... 3/44-h  ;

l 29 -

. 3/4.4.7 MAIM STEAM LINE I5CLATICM VALVE 5. . . . . . . . . . . . . . . . . . . . . 3/4 4  !

)

3/4.4.8 STRUCTURAL INTEGRITY. . . . .'. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/44M j

~

3/4.4.9 RESIDUAL EAT REMOVAL - ,

21 -

l mbt 5hutdown......................................... 3/4 4 n [

l i

c.id Shot.o ........................................ 3/44-8  :

l t 3/4.5 EMERGEMCY CORE CDCLING SYSTEMS j 3/4.5.1 ECCS - 0PERATING..................................... 3/45-1 .  !

t 3/4.5.2 ECCS - SMU1Tl0WN...................................... 3/4 5-4 t 3/4.5.3 SUPPRESSION CHAMBER.................................. 3/4 5-4 I, s.

?

. . . . . . . , t

. . . . . . - , - - V,

. Y 10 JAN1903 j

~ '

INDEX i

LIMITING CONDITIONS FOR OPERATION"AND SURVEILLANCE REQUIREMENTS ,

PAGE SECTION 3/4.8 ELECTRICAL P0WER SYSTEMS

> \

3/4.8.1 A.C. SOURCES  ;

A. C. Sources-0perating. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 8-1  !

A.C. Sources-Shutdown................................ 3/4 8 1 3/4.8.2 D.C. 50URCES ,

D.C. Sources-0perating............................... 3/4 8 h l $f 11 Jr #-

l D. C. Source s-$hutdown. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-id -1.l

' l 3/4.8.3 ONSITE POWER DISTRIBUTION SYSTEMS -

I 3/4 8. W

,O Distribution - Operating............................. l i

.ms/Y Di stributi on - Shutdown. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 8 .48/

\

d 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES

~

A.C. Cir;.it; S: M: H= j Cet:!=rt. . . N4 H S Primary Containment Penetration Conductor Overcurrent AS l Protective Devices................................. 3/4 8-JP- l Motor Operated Valve Thermal Overlopd Protection.-Areaste3/4 37 i nehr oper ted wl4 Tw l oveel..e Pr.tec+:., -mr evraun 3/9 t-sm : n j Reactor Protection Systen Electric Power Monitoring.. 3/4 8-etsm23 i

3/4.9 REFUELING OPERATIONS .

1 3/4.9.1 REACTOR EDE' SWITCH. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-1 l 3/4.9.2 INSTRhENTATI0N......................................3/4t-3 l

3/4.9.3 CC[TRC L ROD F0SITION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-5

(

! 3/4.9.4 DECAY TIME........................................... 3/4 9-6  ;

l l 3/4.9.5 C0 MINI CATI ONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-7 {

3/4.9.6 REFUELING PLATF0RM................................... 3/4 9-8 l i

b i

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ves ems ~!

L l . _ . . . _ - . . _ __.-_ -_ ._ _ ._. ._ ________.I

/

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• 3/4.0 APPLICA8ILITY LIMITING CON 0! TION FOR OPERATION 3.0.1 Compliance with the Limiting Conditions for Operation' contained in the succeeding Specifications is requi md during the OPERATIONAL CONDITIONS or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.

3.0.2 Moncompliance with a Specification shall exist when the requirements of the Limiting Candition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Candition for Operation is restored prior to expiration of the specified time intervals, completion of the Action requirements is not required. .

3.0.3 When a Limiting Condition for Operation is not met, except as provided '

in the associated ACTION requirements, within one hour action shall be initi- .'

ated to place the unit in an OPERATIONAL CDNOITION in which the Specification .-

does not apply by placing it, as applicable, in:

~

, ' 1. At least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, -

l

' . 2. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and  :

3. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. ,

- Where corrective seasures are concleted that permit operation under the ACTION I

i requirements, the ACTION may be taken in SCcordance with the specified time

~ limits as seasured from the time of failun to meet the Limiting Condition for coeration. Exceptions to these requirements are statec in the indivicual

- Specifications.

This Specification is not applicable in CPERATIONAL CONDITIONS 4 or 5. l l l

3.0.4 Entry into an CPERATIONAL CONDITION or other specified condition shall not ce sace unless the concitions for the Limiting Concition for Coeration are  :

set without reliance on provisions contained in the ACTION requirements. This '

provision shall not prevent passage througn or to CPERATIONAL CONDITIONS as l l required to comply with ACTION requirements. Exceptions to these requirements t are stated in the individual Specifications. .

l L

%) /#U(47 4 f  ;

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3/40-1 I

- ~~

l.

/AJSGAT A 70 P &. 3l'l0-I

l 3.0.5 be inoperaale solely because its esorgency power , or source is OPERABLE for the purpose of satisfying the requireme limiting condition for Operation provided:

(1) its corresponding normal or emergency power source is CPEAA8LE; and (2) all of its redundant systas(s),

sesystes(s), train (s), component (s) .nd devica(s) are OPERA 8LE, or likowise satisfy the requirements of this specification.

Unless both conditions (1) unit in an OPERATIONAL CONDITION in which the app for Operation does not apply by placing it, as applicanie, in:

1.

2. At least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,

3. At least HOT SWTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and -

At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. <

This specification is not apolicable in OPERATIONAL CONDITION 4 or 5 .'.

7.;

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B 3/a.1 REACTIVITY C xT20L SYSTEMS 3/a.1.1 SHUTDC'N wARGIN t.IMITING CONDITION FOR OPERATICM SURV!ILLANCE RECUIRO80C5 4.1.1 The SWTDCW MJtGIN shall be determined ta be equal to or gnatar than specified at any time during the fuel cycle:

a. By seasurement, prior ta or during tae first start:.:o after eaca nfueling.

b. By sessurweent, witnin 500 M/T orier to the can avenge exposure at mica the predic.ad SWTDOW MRGIN, including uncartainties and calculation 31ases, is equal to the specified limit.

cr.

c. Witnin pp nearf after detection of a vitadtrwn control nd tnat is immevaale, as a nsult of escassive friction er sechanical inter-farenca, or is untriocaale, exceot that the move neutred $wlTOCW MRGIM sna11 he verified accrotaole with an incnased allowance for the withdrawn wrta of the f amovaale or untrippanle control rod. .

"tzcept. novement of IRMs, SRMs or special movanle detectan.

Seef cactw

'  :! .O ;;'_:."', 3/4 1-1 i

L i

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. I C J4N 19c3 l REACTIVITY CONTRot. SYSTEMS i 3/4.1.2 REACTIVITY ANOMALIES LIMITING CONDITION FOR OPERATION 3.1.2 The Mactivity equelence of the difference between the actual R00

DENSITY and the predicted 200 DENSITY shall not exceed 1% delta Uk.

APPLICA8ILITY: OPERATIONAL CONDITIONS 1 and 2. I ACTION:

With the nactivity equivalence difference exceeding 1% delta k/k: l [

4. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> perform an analysis to determine and explain the cause  !

of the reactivity difference; operation any continue if .the difference [

is explained and corrected. l

b. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

J i

SURVEILLANCE REDUIREw!NTS

}

)

i 4.1.2 The reactivity,.ecuivalence of the difference between the actual RCD l

! DENSITY and the. predicted RCD DENSITY shall be verified.to be less than or L equal to 1% delta Uk

a. During the first startup following CORE ALTERATIONS, and j

b. At least once per 31 effective full power days during POWER OPERATION. . [

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(nT./t; 3/4 1-2  !

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. REACTIVITY CONTROL SYSTDts i 3/4.1.3 CDNTROL 4005 CONTRCL 800 OPfRASILITY LIMITING CDef7 ION FOR OPERATION ,

3.1.3.1 All control rods shall be OPERA 8LE.

APoLICAs!LITY: OPERATIONAL Com ITIONS 1 and 2.

ACTION:

a. With one control red inoperable due to being immovable, as a result of ascessive friction er mechanical interference, or known to be untrippaale:

1. Within one hour:

a) Verify that the inoperable control rod. if withdrawn, is separated free all other inoperaale control rods by at least two control calls in all directions.

b)- Disarm the associated directional control valves" e64mem l' M Ma51IcalN h closing the drive water and exhaust -

water isolation valves.

c) Camply with Surveillance Requirement 4.1.Lc.

crise,' be in at least NOT wmow within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. l m _

n s ..a .- ..,

) 2. Restore the,inoperaale contrei roep ortxAsLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />

/ l or be in at least MOT $HUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. With one or more control rods trippable but inoperable for causes other l than aderessed in ACTI'* a, above:

1. If the inoperable control rod (s) is witherawn, within one hour:

a) Verify that the inoperable withdrawn control rod (s) is separated from all ather inoperaale withdrawn control rods by at least two I control calls in all directions, and b) Demonstrata the it:sertion capability of the inoperable withdrawn l l control rod (s) by inserting the control rod (s) at least one l not.ch by drive water pressure within the neraal operating range".

l Otherwise, insert the inoperable withdrawn control red (s) and disarm the associated directional control valves" either:

a) Electrically, er l b) Mydraulically by closing the drive water ano exhaust water jt l

1 solation valves.

l l

l "The inoperacle control rod any then be withdrawn to a position ne further l withdrawn than its position when found to be inoperable. '

"May be rearmed intermittently, under aesinistrative control, to perwit testing associated with restaring the control rod to OPERA 8LE status.

herit catu.

0:' ;*: ' n ) 3/4 1-3 -

3 0 MAR 1983 e . .

• l0 i;

REACTIVTTY CONTROL SYSTDt5 LIMITING CONDITTON FOR OPERATTON (Continued) -

ACTTON (Continued)

2. If the inoperable control rod (s) is inserted, within one hour disarm I the associated directional control valves ** either:

a) Electrically, or b) Nydraulically by closing the drive wetar and exhaust water l 1selation valves.

Otherwise, be in at least MUT SWTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. l

3. The provisions of Specification 3.0.4 are net applicable. ll

c. With more than 8 control rods inoperable, be in at least HOT SHUTDOWN

,vithin 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. -

i 4 , /hUW SURVEILLANCE REQUIREwPITS

• 4.L 3. L 1 The serem discharge volm drain and wat valves shall be l demonstrated CPEAA8LE by:

1 24 ks .

~

a. At least once per % verifying each valw to be open," and l

b. At least once per h days cycling each valve through at least one l completa cycle of full travel.

l 4.1.3.1.2 when aben the (;mt ;r- in?) llow power setpoint) of the RW l 1 and RSCS, all withdrawn control rods not required ta have their directional

control valves disarmed electrically or hydraulically shall be demonstrated l OPERA 8LE by moving each control red at least one notch

a. At least once per 7 days, and .

l

b. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when any control rod is immovante as a '

result of escassive friction er mechanical interference.

4.5.3.L3 All control rods sha11 he despnstrated CPERABLE by performance of l Surveillanca Requirements 4.L3.2. ,4.L3.4, 4.L 3.5, 4.L3.5 and 4.L3.7. l

=These valves may be ciesed intermittently for tasting under administrative controls.

""May be rearmed intermittaertly, under administratin control, ta perwit tasting associated with restoring the control rod to OPERABLE status. -

E.et cattw

';;';A'O 3/4 1-4 30WE t

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_ REACTIVITY CO WROL SYSTEMS SURVETLLANCE REQUfttwENTS (Continued 1  !

i 4.1.3.1.4 demonstrating:The seras disenerge volume shall be detamined OPERA 8LE by a.

The scram discharge vel mo drain and went valves OPERA 8LE, when N tien of less than or equal tacontrol rods are scram testes free a norma 2f'aonths, by verifying that the drain and vent valves:50 E ROC DEMS -

D 1. Close within 30 '

rods to scram, andseconds after receipt of s signal for control 2.

Open when the : eras sfgnal is reset. .

sI

b. . Proper floa** ht

• i.w o og feef" p,'fl .n..%),  :

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.Ja.'h .anv /  :: ram di:chaeae ve/urne tonr* vaive (:) and/ or J

e y :! ~"'Tm dif:he 2:e - s:/u #c de:In .e \"e U.: : . 'leew;:e i .i

,: n peesbi} , re:l:re :5 !e : ' en 1 e ~ ? va!.'e =# d :n

• drain gt.~.'e  ?: CSMAsTLE :.L:!us ,.) < 5 ; n i h :u r: or ce in af.**1:5 Her :Hur:owN wiih;e ae nex+ 1: n s es .

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l REACTTVITY CDNTROL $YSTEMS CONTROL R00 MAXIMJN $ CRAM IN$ERTION TTMES LIMTTING CONOTTION FOR OPERATION i t 3.1.3.2 The anximum scram insertion time of each control rod from the fully J witherewn position to notch position $51. based on de-energizatten of the ('

serem pilot valve solenetes as ties zero, shall not exceed (7.01 seconds.

APPLICA8ILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTTON:

a. With the maximus scram insertion time of one or more control rods 0 a eeding (7) seconds:

,, j

1. Declare the control rod (s) with the slow insertion time inoperable. O and f

., ' 2. Perfore the Surveillance Requirements of Specification 4.1.3.2.c at U least once per 60 days when operation is continued with three or *

. more control rods with anximum scras insertion times in excess of l 47.01 seconds.  ;

i Othenrise, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.  !

-- l

] b. The provisions of Specification 3.0.4 are not applicable. l

SURVEILLANCE REQUIREWENTS l

' i I

4.1. 3. 2 The maximum scram insertion time of the control rods shall be demon-  !

strated through asasurement with reactor coolant pressurs greater than or  !

equal to 950 psig and, during single control rod scras time tasts, the control  !

rod drive p eps isolated from the accumulators: I

a. For all control rods prior to THERMAL POWER exceeding 40% of RATED THERMAL POWER following CDRE ALTERATION 5* or after a reactor . ,

shutdown that is greater than 120 days.

b. For specifically affected individual control rods fellowing maintenance en er modification ta the control rod or control red drive systan which could affect the scras insertion time of those specific control rods, and .

c. For at least 155 of the centrol rods, en a rotating basis, at least once per 120 days of POWER OPERATICM. l l

"Except movement of SRM, IRM, or special movable detectars or normal control rod movement.

NOPE CREtt

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=: ; .";4 3/4 1-4

O

. /.3 REACTTVTTY CONT 110L SYSTtMS CONT 110L WO Avt1tAGE SC1 TAM INSE1rTION TTMf5 LIMITING CONOTTION FOR OPERATTON 3.L3.3 The everage scres insertion time of all CPDABLE control roes from the fully witherman position, bases en ee-energization of the scras pilot velve solencias as time zero, shall not onceed any of the following:

Position Inserted From Average Scram Inser-Fully witherswa tien Time (Secones)

~

j4 ' . 431

0.46

!L 93

3.49 AP't.!CA8!LI"Y

OPEJtATICMAL CONDITICn$ 1 and 2. .

AL"ICM:

With the awe ige scram insertion time exceeding any of the aeove limita, De in at least NCT SWT:CWN witnin 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

\

%I SURVEILLANCY RECU!REwtW*$

A. L 3. 3 All centrol rods shall be eenenstrated OPERA 8LE ty seras time i testing free the fully witherewn positten as requirse Dy Surveillance Aequiresent 4.L 3.2.

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REACTIVITY CONTROL SYSTEM 5 .

POUR CONT 1t0L 200 GROUP SCRAM IN$tRTTON TTMES LIMITING CopeITION FOR OptRATION

• 3.1.3.4 The average scran insertion time, free the fully withdrown position, for the three fastest control rods in each group of four centrol rods arranged in a two-hy-two array, based on doenergization of the seres pilot valve o solencias as time zero, shall not anceed any of the fo11 swing:

Position Inserted From Averece Scram Inser- >

Fully witherawn tion Time (Secones)

'0.01oAS

'". "'; o .%

. '1.  :: 2. .o s (2. ""; 3. 'lo APPLICA8ILITY: OPERATICNAL CONDITIONS 1 and 2.

ACTION: ,

a. With the average scram insertion times of control rods exceeding the Il above limits

'S 1. Declare the control rods with the slower than average seras ll j insertion times inoperaale until an analysis is performed to determine that required scras reactivity remair>s for the slow four control rod group, and

2. PerformthisurveillanceRequirementsofSpecification4.1.3.2. cat ll 1 east once per 60 days when operation is continued with an average scras insertion time (s) in escass of the average scram insertion time limit. .

Otherwise, be in at least NOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />,

~

b. The provisions of Specification 3.0.4 are not applicable. ll SURytILLANCE REQUTREMENTS 4.1. 3. 4 All :entrol mds shall te demonstrated OPERA 8LE try r,cres time tasting l from the fully withdrawn position as required Dy Surveillance Requirement 4.1. 3. 2. >

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REACTIVITY CONT 1t0L SYSTDt$

CONTROL E30 SCRAM ACCtMILATORS LIMITING eammITION POR OPERATION 3.1.3.5 All control rod scram accumulators shall be OPERA 8LE. l ApptIP.ASILITY: OPERATIONAL CON 0!TIONS 1, 2 and 5".

ACTION:

a. In opERATICNAL CONDITIONS 1 er 2:

1. With one control rod scras accumulatar inoperable, within 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />s:

a) Restore the inoperable accumulater to OPERA 8LE status, or

~

b) Declare the control rod associated with the inoperable l ,

accoulator inoperable.

Otherwise, be in at least NOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. l

2. With more than one control red scres accumulatar inoperable, ,

declare the associated control rods inoperable and: i a) f the control red associated with any inoperable ser accumulater is withdrawn, immediately verify that at least  !

one control red drive pump is operating by inserting at  :

least one withdrawn control red at least (ne notch or place j App pesaT)'(the reactae = *= u ' '- W wewn sosition. J  ;

c g) Insert the inoperable control rods and disare the associated i d control valves either:

1) Electrically, or l

2) Hydeculically by closing the drive water and exhaust  !

wetar isolation valves.  !

Otherwise, he in at least MOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. l

b. In OPERATIONAL COM3ITION 5": l With one withdrawn control rod with its associated scram ,I

1.  :

accumulator inoperable, insert the affected control rod and -

disam the associated directional control valves within one [

hour, either. .

I a) Electrically, or l i

b) Hydraulically by closing the drive water and exhaust watar isolation valves..

f With more than one withdrawn control red with the associated l 2.

scras accumulator inoperable er no control M drive pump oper-  !

ating, immediately plass'the reactor made switch .in the Shutdown  ;

positica. ,

t

c. The provisions of Specification 3.0.4 are not applicable. -

lf i j

=At leest tne accumulator associated with each withdrawn control rod. Not applicable to control rods removed per Specification 3.9.10.1 er 3.9.10.2.  ;

-- - = m m 3/4 1-9 <r- -

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S pesect To e+. 3/4 l- 1 :

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4) L=ediately verify that at 1'

• pu=; is opert:in b i -

'"' "I' IP 8 drive C:n:PCI red one not.h. And nd 4:hcrawinF 4 wi:hdra i

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l REACTIVITY CONTROL SYSTEMS l l

i M ILLANCE REQUIREMENTS l l

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4.1.3.5 Each control rod scras accumulator shall be detamined OPERA 8LE:

a. At least'once per 7 days by verifying that the indicated pressure h

2. 8He h IOIO) ' CO) . -(0) psig unless the control rod is inserted e and disarmed er scrammed.  !

a. At least once per 18 months by:  !

l

1. Perfomance of a:  !

a) CHANNEL FUNCTIONAL TEST of the lest detectors, and  ;

b) CHANNEL CALIBRATION of the pressure detectors, and I verifying an alare setpeint of (940)

• 4101 -jol psig on l decreasing prossure. j I

2. '": N} (Measuring and recording the time for up to I 10 minutes) that (each individual 1 accumulater check valve  !

maintains (the associated { accumulator pressure aeove the alars set point,(':  ;-;:t-- "2: :r ::::' 2 ,10 c'  ::)  !

- with no control rod erive pump operating. I j M a.k SCM"  % M

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yyy CONTROL SYSTEM '

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Fg R00 ORIVE COUPLI.NG

• I g!LyffMG CONDITION FOR OPERATION 3,1,3. 6 All control rocs shall be coupled to their crive mechanisms.

gicASILITY: OPERATIONAL. CONDITIONS 1, 2 p c 5".

N a.

In OPERATIONAL COM0! TION 1 and 2 with one control red not coucled to i

its associatec crive mechanism:

1. Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, either:

a) If pemitted by the RWM and RSCS, insert the control rod crive sechanise to accomplish recoupling and verify recoupling by withcrawing the control rod, anc:

1). Observing any indicated response of the nuclear instrumentation, and ter 1

2) Demonstrating that the control red will not go to the overtravel position. '

b) If recoupling is not accomolishec on the first attemot or, if not ermittec cy the RhN or RSCS then until permit *.ac by the RWM anc RSCS, declare the control roc inoperacle and insert the control roc anc disars the associated

, cirectional control valves"" either:

1) Electrically, or

2) Hycraulically by closing the crive water anc exhaust water isolation valves.

2.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, t.

In CPERATIONAL CONDIT!CN 5" with a withdrawn control roc not couplec  !

to its associatec crive sechanism, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, either:

1.

Insert the control rod to acccmolish receucling and verify receucling by witacrawing the centrol roc anc comonstrating that the control rod will not go to the overtravel position, or 1

! 2. If recoupling is not ac:omplished, insert tne control rod and cisam the associatec cirectional control valves ** either:

a) Electrically, or "

b) Mycraulically by closing the drive water and exhaust water isolation valves. i c . 711 pnJrms e 4; Sp-s he ad n~ 3. C. 9 a rs net

• sy lis.44 "At least eacn withdrawn control rod. Not applJcable to control rocs removed per Specification 3.9.10.1 or 3.9.10.2. '~

• "May be rearmed intermittently, under administrative control, to permit tasting associated with restoring the control red to OPERABLE status. l'

/f8M CAEtt- 3/4 1-11 8

1 ,

lL k

AN 4 c

% 4 REACTIVITY CONTROL SYSTEMS l'l l

l SUtvf1LLANCE REQUIREMENTS 4.1.3.6 fach affected control rod shall be demonstrated to be coupled to its drive mechanism by observing any indicated response of the nuclear instrumen- l tation while withdrawing the control rod to the fully withdrawn position and then verifying that the control rod drive does not go to the overtravel l position: .

a. Prior to reactar criticality after completing CORE ALTERATION $ that could have affected the control rod drive coupling integrity,

b. Anytime the control rod *is withdrawn to the " Full out" position in subsequent opertrion, and

c. Following maintenance on or modification to the control red or control rod drive system which could have affected the control rod drive coupling integrity.

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3/4 1-12 l t

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REACTIVITY CONTROL SYSTEM CONTROL ROC POSITION INDICATION I

. LIMITING CONDITION FOR OPERATION 3.1.3.7 The control roc position indication system shall be OPERA 8LE. .

APeLICA8ILITY: OPERATIONAL CONDITIONS 1, 2 and 5".

ACTION:

a. In OPERATIONAL CONDITION 1 or 2:

1. With one or more control roc position indicators incoeratie, '

except for the "Ful.1-in" or " Full-out" indicators: ,

. i a) Within one hour:  ;

1) Determine the position of the control roc ey:

(a) Moving the control roc, by single noten movement, to a position with an CPERAELE position indicator, (b) Returning tne control roc, by single noten govement, i to its original position, anc j (c) Verifying no control roc crift alarn at least once i I

per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, or

2) Move the control red to a position with an  ;

OPERABLE position indicator, or j r

3) when THERMAL PCWER is witnin the low power setpoint of the RSCS, ceclare the control roc inoperacle, or

4) When THERMAL POWER is greater inan :ne low power i setnoint of the RSM, ceclare the control roc  !

inoceracle, insert the control rod anggisarm the j associated cirectional control valversacher: j t

(a) Electrically, or  !

(b) Mycraulically by closing the crive water  :

and exhaust water isolation valves.  ;. l I

b) Otherwise, ne in at least HOT SHUTOOWN within the next l 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.  ;

"At least eacn witacrawn control rod. Not applicable to control rods removed  !

per Specification  !

~~

4 3r Pfay k reav me)3 Ver- 9.10.1 or 3.9.10.2.

e t&ntyN un der isfr animo h, w ca f*$ fo MEGAk1ciuNesl,+b)W nszas sissec o~f!

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i REACTIVITY CONTROL SYSTEM ,

LIMITING CDN0! TION FOR OPERATION (Continued)

ACTION: (Continued) )

2. With one or acre control red " Full-in" and " Full-out" position indicators inoperacle:

a) Either:

l') When THERMAL POWER is within the low power setpoint '

of the RSCS: '

(a) Within one hour:

(1) Determine the position of the control rod (s) by:

(a). Moving the control roc, by single noten movement, to a positior wita an OPERA 8LE position indicator, (b) . Returning the control rod, by single noten movement, to its original position, anc (c) Verifying no control rod crift alam at least per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, or (2) Move the control roc to a position with an CPERA8LE position incicator, or (3) Declare the control rod inoceratie.

(b) Verify the position and bypassing of control rocs with }

inoperaale " Full-in" and/or " Full-out" position incica- /

tors by a second licensec coerator or other tecnnically

qualified memoer of the unit tocanical staff.

2) When THERMAL POWER is greater taan the low power setpoint of the RSCS, detemine the position of the control roc (s) per ACTION a.2.a) 1)(a)(1), 40cve, b) Otherwise, be in at least HOT SHUTOCWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />,

b. In CPERATIONAL COMOITICH 5" with a withdrawn control od position 1

indicator inoperacle, move the control rod to a position with an CPERA8LE position indicator or insert the control rec.

c. The provisions of Scecification 3.0.4 are not a:clicanle. .

l SURVEILLANCE REQUIREMENTS 4.1.3.7 The control roc position indication system snall be cetermined OPERA 8LE by verifying:

a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> that the position of esca control rol is i-indicated, -

h. That the indicated control rod position changes during the movement of the control rod drive when perfoming Surveillance Recuirement i 4.1. 3.1.2, and

c. That.tne control red position indicator corresponds to the control rod position indicated by the " Full out" position incicator wnen -

performing Surveillance Recuirement 4.1.3.6b.

"At least esca withdrawn control rod not amplicaele to control rods removed I

per Specifications 3.9.10.1 or 3.9.10.2.

l(OM CAE(2 3/4 1-14 ,

l

Mb REACTIVITY CONTRob SYSTEMS CONTROL 200 0 RIVE HOUSING SUPPORT I

' i LIMITING CONDITION FOR OPERATION 3.1.3.8 The control rod drive housing support shall be in place. l AP9(!CA811.ITY: OPERATIONAL CON 0!TIONS 1, 2 and 3.

ACTTON: l With the control rod drive housing support not in place, be in at least NOT  ;

$NUTDCWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTCCWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. l 1

~ .

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i 5

i 5

$URVEILLANCE REQUIREMENTS  !

r 4.1.3.8 The control rod drive housing support shall be verified to be in place l by a visual inspection prior to startup any time it has been disasseecled or when maintenance has been performed in the control rod drive housing support ,

area.

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. . . . . 3/a.4 REACTOR COOLANT Sv! TEM

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3/4.4.1 RfCIRCULATION SYSTEM

...._ RfCIRCULATION LOOPS _

LIMIT 1NC CONDITION FOR OPERATION

]

3.4.1.1 with: Two reactor coolant systes recirculation loops shall M in operation -

. . a. . f

+

Total core flow greater than or equal to 45% of rated core flow, or*i b.

THERMAL POWER less than or equal to the Ifnit specified in Figure:

3.4.1.1-1.

APDLICABILITY:  !

OPERATIONAL CONDITIONS 1* and 2*.

_ ACTION:

4.

With one reactor coolant system recirculation loop not in operation

+ imeediately initiate action to reeuce THERMAL POWER to less than! ,

egual to the limit specified in Figure 3.4.1.1-1 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and  !

initiate 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. measures to ' place the unit in at least HOT SMUTDCWN witi i

b. i With no reactor coolant system recirculation loops in operation ,

+ equal to the limit specified in Figure 3.4.1.1-1 i initiate measures to place the unit in at least STARTUP witnin urs 6 n( '

anc in NOT SMUTDCWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

c. +

i With two reactor coolant systes recirculation loops in operation an total core greater flow than the less limit thanin45%

specified Figure of3.4.1.1-1:

rated core flow and TH 1.

Determine the APRM and LPRM** noise levels a)

1. At 1 east once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and I B)  ;

Within 30 minutes after the completion of a THERMAL POWER

• increase of at least 5% of RATED THERMAL POWER. '

2. -

With the APRM or LPRM"" neutron flux noise levels greater than i three times taefe established baseline noise levels, immediately(

l

" initiate the corrective required limits within action 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> tobyrestore increasingthecore noise flow levels to to {

greater than 45% of rated core flow or by reducing THERMAL POWER l to less than or equal to the limit specified in 'igure 3.4.1 1-1 l,

1 i

"See Special Test Eiception 3.10.4. . . . ..

Y ** Detector levels A and C of one LPRM stri {

! and C of one LPRM string in the center of the core should be monitores,ng pe!

nets cus%

3/4 4-1 i

}

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l

,- . s REACTOR COOLANT SYSTEM i

(

SURVEILLANCE REQUIREMENTS  !

i l

4.4.1.1.1 Each pump discharge valve shall. be demonstrated OPERA 8LE by cycling each valve through at least one complete cycle of full travel during each >

STARTUP' prior to THERMAL POWEA exceeding 25% of AATED THERMAL POWEA. .

4.4.1.1.2 Each puso MG set scoop tiume mechanical and electrical stop shall be demonstrated OPERABLE with overspeed setpoints less than or equal to 105% and 102.5%, respectively, of rated core flow, at least once per is months.

4.4.1.1.3 Establish a baseline APRM and LPRM** neutron flux noise value within the regions for wnich monitoring is required (Specification 3.4.1.1, ACTION c) p within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of entering the region for which monitoring is required unless l baselining has previously been performed in the region since the last REFUELING t l CUTAGE.  ;

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• If not performed within the previous 31 days.  !

~

l ** Detector levels A and C of one LPRM string per core octant plus detectors A anc C of one LPRM string in the center of the core shoulc te acnitored.  !

H0PE CREEK, 3/4 ** W '

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REACTOR COOLANT $YSTEM ,

JET PUMP 5 .

LIMITING COMOITTON FOR OPERATION 3.4.1.2 All jet pumps shall be OPERABLE. -

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With one or more jet pumps inoperacle, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. ..

SURVE*LLAN:[ REOUIREWEN'$ /

4. 4.1. 2 Each of the above required jet pumos shall be demonst. rated CPERA8LE prior to THERMAL POWER ascoecing 25% of RATED THERMAL PowtR and at least once -

per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by cetermining recirculation loop flow, total core flow and aiffuser-to-lower plenus cifferential pressure for each jet pump and verifying that no two of the following conditions occur when the recirculation pumos are -

operating .1 L .. . - . . . . . . In accordance we% SaeetA wrt. 3. V.1. 3

a. The indicated recirculation loco flow differs by sors than 1C1 from the estaaltsnec pump speec-loop flow characteristics.

b. The indicated total core flow differs by more can ICE from the estaclisnei: total core flow value derives from recirculation loop flow sensurements.

c. The incicated diffuse-to-lower plenue difverential pressure of any individual jet pumo atffers from the es'.aalisnec patterns ey more l than ICE. .

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10 JAN 513 M REACTOR COOLANT SYSTEM RECIRCULATION PUMPS

. - ,.s...-

~ s . .. .. .

LIMITING CONDITION FOR OPERATION 3.4.1.3 Rectrevlation pump speed shall be maintained within: l 55 of each other with core flow greater t$an or equal to 7CE of l a.

rated core flow. **

l

b. 10E of each other with core flow less than 705 of rated core flow.

l APPLICASILITY: OPERATIONAL CONDlTIDMS 1* and 28 ACTION:

With the recirculation pump speeds different by more' than the specified limits ,. either:

e

a. Restore the recirculation pump speeds to within the specified limit [

wittiin 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or

, b. Declare the recirculation loco of the pump with the slower speed not l s

in operation and take the ACTION required by Specification 3.4.1.1.

N w .

SURVEILLANCE REOUIREMENTS Q

4.4.1.3 Recirculation pump speed shall be verified to be within the limits at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. ' s I

a:me special Test Lucepti'on 3.10.4.

(En rt arse l.h* f =f , Hit npiec~~tmy k ssspes,Ja/

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i RfACTOR CDOLANT $YSTEM IDLE RfCIRCU1.ATION LOOP STaRTUp LIMITING CONCITION FOR OPERATION .

3.4.1.4 An idle recirculation leen s'ha11 not be started unless the temperature sifferential between the reactor pressure vessel staas space eselant and the w. l bottaa head erain line coolant is less than er oogt-"to "_^^!", r:. MT

• F,4 T d . _ When _ _ _' c . _ _ _ . . . '.'.i f-r1'^^,- .

- r,f-

a. both loops have been idle, unless the tamperature sifferential between the reactor coolant witnin the idle lose to be started us and the coolant in the reactar pressure vessel is less than or equal i to}50$'F,or ,

b. When only one looo has been idle, unless the temperatun differential between the reactor coolant within the idle and operating mcircula-tion loops is less than or equal ta *F and the operating looo i flow rata is less than or equal to 01K of rated loop flow. ,

i AP8LICA8IL P : CPERATIONAL CONDITICMS 1, 2. 3 and 4. (

AC"0N:

With tescersture differences and/or flow rates exceeding the above limits, i.

sussene startue of any idle recirculation loop. [

I t

?

i l

$URVU LLANCE REOUIRESENTS

. t 4.4.1.4 The tamperature diffematials and flow rata shall be dotamined ta be 1 within the limits within 15 minutas prior to starsue of an idle recirculation l l

loop.

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- I REACTOR CDOLANT SYSTEM ,

3/4.4.2 SAFFTY/ RELIEF VALVES .

! SAFETY / RELIEF VALVES

• LIMITING CONDITION FOR 0PERATION jc

. '3.4.2.1 v.. . _ _ _ . , _ _ _ ,

._ _ ___IS;t :;:- x : - ':t; ' :n r:} ghe safety valve function of at least 441$E(of the following1' reactor coolant ' '

i systas safety / relief valves shall be OPERA 8LE with the specified code safety valve function lift settings:"

$kk b.3.'._ b..I55.

.. .$N ,5e b . . _,_

4 TW safety-reliefvalves9fHef. psia alt ~m .

.5 TN safety-relief valves 9 fMSH"'lisig fl1 ..

5IS4 safety-relief valves 9 f4006-) psig 115 - - -

Lt w APPOCABILITY

OPERATIONAL CONDITIDMS 1, 2 and 3. - ' -

mSC s" . I A ACTION: - t

_i_a.1 l

_ _ _ _ _ _, l'r_ __,__m .___... . 3... ......_ __m_ -_

f. ym .m_ _

?.x : c2} the sar.9"" . ; ' - - funct W ' r . M rs of the above 13 recutree#safetv/=14 ' ci.. inoperan e, .. ". :* i==<t MOT SM TTDOWN

1. _'-..o u nours and in COLD $HUTDOWN within the next 2a houn.---

$ With one or more- :: : n ':t; r f:n :-) safety / relief valves stuck open,/# f provised that suppression pool average water tamperature is 1est than 444?F,  ;

close the stuck open (nr n':t; : E r :-1':-) safety relief ialve(s); i

if unaale to close the stuct open valve (s) within 2 minutes or if sun-l pression pool average water temperature is 49 H'F or greater, p' ace the reactor nose switch in the Shutdown position. Inc

h. Wy one or more safety / relief valve (t: ;' ; nn ; ;2 2.;)

• i

! (1gous 1c ttorsfinoperable, restore the inoperable (; 't:(n); . l 1 . tsionitay.h OPERA 8LE status within 7 days or be in at least NOT i

. SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> anc in COLD SHUTDOWN within the '

1 i fellowing 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />.

]

latti $~feie}7 /telief fw. shoe. e f tgree. er seere, a f y 4 , fe + , .

b-Above Med ombs3 o'ngerQ' be in sf* / rent h@T* M1 l ; % Iz heases a n d in c ela s w powo) ~ttnin t>, e. nas +' 19 kests .

pson? -

i g .

"The lift setting pressure shall correspond to ambient conditions of the valves [

at nominal operating tamperatures and pressures. . j W

• 1  !

Eces ca.e w

^'"T",(7g,'!)

3/4 4-X I

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M L~

M INSERT A TO PG. 3/4 4-7:

a. With the safety / relief function of two of the 14 above listed valves inoperable, restore at least one inoperable valve to an OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

W T .

INSERT S TO PG. 3/4 4-7: 5 ,

• • SRVs which perform an ADS function must also satisfy the

- OPEP.A3:~ !""? requ: rements O f Technical Specificat:,.cn 3/4.5.1, ECOS - OPERATING.

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NEACTOR COOLANT SYSTEM l i i SURVEILLANCE REQU12J'Df75 I

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4.4.2.1.2 The '*_ e';: ; x:_n ; ^; M facoustic monitar{ for each safety /

relief valve shall be demonstrated OPERABLE with the setpoint verified te be l

".'""" : O! ;;'O by performance of a: l ,,,

of t-te A,il egen noise leve l pog) CHANNEL a. jFUNCTIDMAL TEST 1 '~'**' at least once per 31 days, and a L. CHAMNEL CALIBRATION at least once per la months ("f.

l

! i one provisions of Specification 4.0.4 are not applicable provided the y Surveillanca is perfornec within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> af*.ar reactor staan pressure is

! adequate to perfors the tast.{  !

< .d "n +. l setPS *4=Il ke la au. l*> ' * ' tk % maa 4<.ba's remend +. n. A JJ.stment h th vs/n -All oy<~ neh <. le ve l ,

s L.11 hc. aca..y ks4 e d J ,,3 fte .stao t.y % t y ,3 n .

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, 10 JAN 383 l REACTOR COOLANT SYSTEM wgr.vfd.o,h, .

. I SAFETY / RELIEF VALVES LDW-LOW SET FUNCTION , , {

i i

LIMITING CONDITTON FOR OPERATION 3.4.2.2 The n1ief valve function and the low-low set function of the following l reactor coolant system safety / relief valves shall be OPERABLE with the following i

settings:  ;

l ""

Low-Low Set Functioa Relief Function l Setooint" (osio)

• 3 Setooint" (psic)

• 3  !

valve No. ,

ggn, close gg,n Close f I (1033) ($26) - ['

l (1073) (536)

(1113) (946)

(1113) (946)

(1113) (946) [

ApoLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.

. s

• ACTION: ,

a.' With the relief valve function anri/or the low-low set function of one of I the above recuired reactor coolant system safety / relief valves inoperable,

- P -

restore the inoperable relief valve function and low-low set function l  ;

to OPERABLE status within 14 days or be in at least HDT SHUTDOWN within j the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SWTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.  !

9

. b. With the relief valve function and/or the low-low set funct'on of more l than one of the above required reactor coolant system safety / relief valves .

inocerable, be in at least MOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24' hours.

SURVEILLANCE REQUIREMENTS i .

l 4.4.2.2.1 The relief valve function and the low-low set function pressure l l actuation instrumentation shall be demonstrated OPERA 8LE by performance of a- . ,

i a. CHANNEL FUNCTIONAL TEST, including calibration of the trip unit, at least j ence per 31 days. . i

b. CHANNEL CALIBRATION, LDGIC SYSTIM FUNCTIONAL TEST and simulated autamatic l

operation of the entire system at least once per la months.

1 .

"The lift setting pressure shall correspond ta ambient conditions of the w valves at nominal operating temperatures and pressures. ,

we sanex 9 S !*! (*" ") 3/4 4 9

-__9, r y_-- - - - , , - - , , . .__--,-,-.--.----,,r. . , w, ,-,- .,,.c.

- - . _ , , , - . , - . . , _ . - - , - _ . ,__.__.m .____._.__mm_

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS - OPERATING LIMITING CONDITION FOR OPERATION 3.5.1 The emergency con cooling systems shall be OPERA 8LE with:

a. The core spray systas1456-) consisting of two subsystans with each suesystas comprised of:

enee s&9

1. $Twel CPERABLE h p M sJ, and l

' ~

2. An CPERABLE flo'w path capable of taking suction from the suppression chamber and transferring tne water through the spray sparger to the reactor vessel.

b. The low pressun coolant injection (LPCI) systes'of the residual heat removal systes consisting of e subsystans with each sucsystes comprised of: w .

Ces

1. MCPERABLE LPCI pump (s), and l

2. An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water to the reactor l vessel.

c. The high pressun cooling infection (HPCI) systan consisting of:

1. One OPERABLE HPCI p op, and

2. An CPERABLE flow path capable of taking suction from the suppnssion chamber and transferring the water to the react:r vessel.

d. The aut:matic depressurization system (ADS) with :t '$= 0 (;i:} 4 4 IE OPERA 8LE ADS valves.

Apot.ICABILITY: OPERATIONAL CONDITION 1, 2* , ** f , and 3* , **,8# l "The MPCI systes is not requind 'a be OPERA 8LE when nactor staan done pressure is less than or equal to - en- g. l

==

The ADS is not neutred to be OPERA 8LE en nactor steam done onssun is ,

' less than or equal to {1001 psig.  :

I

1. $ee Special Test Exception 3.10.6.

em e3heRBRS1sie w be. inovemble iw W.h

• *ow,.

• t i isL.Pt1 a.\M,4s,Asg%4 in shdtsoA coo b en ve.ath e vess d

?*=swcl h less A %e AWR. cJ.g made-k TeeWss.*wesM. % .

C7 (L?./d; 3/4 5-1 we atek 30 MAP $8?

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

7 h'

EMERGENCY CDRE CDCLING SYSTtus l

LIMITING CONOITTON FOR OPERATION (Continued) ,

ACTION:

a. For the core spray systas:

C.see SM

1. With one 486 suasystem inooernale, proviese that (at least ene M

' T" :-_-- ' ' :-- LPCI suasystas OPEAASLE, restare the

!' C8%" inoperaalea486 suasystes to OPEAASLI status within 7 says er be in at least MOT SHUTDOWN wiuin the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in CDLD j $NUTDOWN within the following 24 houn.

' cessspee3 t

2. With both G&& suasystass inocereale, be in at least MCT SHUTDOWN witnin 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 houn. [

b. For the LPCI systas:

With we==-e x: '": ; r ' :: h r - be

[,,,,,

, n r LPCI suasystans inoperaale, 1.

i previsee snat at least one 484 sussystem is OPEAASL1, restare .

' ' to OPERA 8LE status within 7 days or

& & taa inoperaale LPCI.5 i

1 to in at least HOT SHUT 00WN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD l

SWJTCCWN witnin the following 24 houn.

. -'r (;' ( ;' 2C: ;, nr ; x--t'; ::' :: c' r M (:- ' t' b- :-

--t -_:r'rt: c're : :x-t': ::'~: x : :t: ', i: '

it

' a* -

r:

.'._ __"_.u!"J'.". . __*' : ' r ' '_ r - : c ' - N 'l != '**' r' t' ' -

?  % +e R2

} 2.k With

"" ene LPCI suasystasseehee *ee inoperaale, '

n s';--- 'n: ' "Ils r % system i .,n, nz: -; 0"!"'2' ', restare r:

j to CPEAA8LE status within heeye or be in at least NOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> andjin COLD SMJTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.*

g%

4-3 %. With teen LPCI suasystaes :rr ':: inoperaale, be in at least

• l HOT SHUTDOWN witnin 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in CDLD SHUTDOWN within the .

next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />."

Core S .

c. For the MPCI system, provided the 499, the LPCI M systas, the A05 and tne RCIC system are OPEAASLE:

1. With the MPCI system inopertale, restart the HPCI system to OPEAASLE status witnin 14 days er be in at least NOT SHUTDCWN within the next 12 houn ane reduce reactar staan eene pressure -

to < ft069 psig within the fellowing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. l l I .

f I Nnenever two or een RMR suasystees are inopersale, if unaale to attain COLD SHUTDOWN as reeutred by this ACTICM, maintain reactor coolant temperature ,,

as low as practical ny use of alternate heat removal setness.

Eeet etttu.

I p / ' 3/4 5-2

o h'

EMERGENCY CDRE Co0 LING SYSTEMS LIMTING CON 0! TION FOR OPERATION (Continuedi ACTION: (Continued)

d. For the ADS: ,, gh

1. With one of the above utred ADS valves inoperable, provided the.HPCI systas, t W and the LPCI system are OPERA 8LE .

restore the inoperaale A05 valve to OPERA 8LE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />  ;

and reduce reactor steam done pressure to i {1001 psig within l the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. ,

2. Wi,th two or more of the above required ADS valves inoperable, i be in at least NOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce reactor l staan does pressure to 1 }1001psig within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. [

l

e. In the event an ECCS systas is actuated and injects water into the l Reactor Coolant Systas, a special Report shall be prepared and sub- l

," mitted to the Commission pursuant to Specification 6.9.2 within i 90 days describing the circumstances of the actuation and the total I accumulated actuation cycles to data. The current value of the .  !

useage factor for each affected safety injection nozzle shall be l provided in this-Special Report whenever its value exceeds 0.70.

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EMERGINCY CORE C30 LING SYSTEUS

! SUtVIILLANCE Rfou!REMENTS i I

l 4.5.1 Theemergencycorecoodli systaas shall anhbe demonstrated OPERA 4LE by: -

l a. At least once per[31 day %s: * '

1. For the tet,d the LPCI system, and the MPCI systas:

a) Verifying by venting at the high point vents that the [

system piping free the pump discharge valve to the systas i isolatipa valve is ' filled with water.

b) Verifyfng that each valve, manual, power operated or j automatic, in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct" position.

2, 'r te L e rj:* . :r'^j' I tut (te) ( t '- : z:) Lae:

,;^.-

C:~i ;"- z;:- : C C f: (:;:-j ' 9 ::I ri 2 ;r -:-

-_- : :d ' z tt n't :;:--t--).

l 2.'1, For the MPCI system, verifying that the HPCI pure,t flow controller i is in the correct position.

4.-= mmem Verifying that,(when tasted pursuant to Specification 4.0.5:

D.

1. The two 46643umos in each subsystem together develop a flow of at least 163501. gym against a test line pressure of greater than , !

or equal to (t$ psig, r - x::.dfr; M : rtr nz:1 pr::=r:

f ? , ;;i;. I

2. L %(""LPCI

pumps in each subsystaa 2;A. developsa flow at leas gpa against a test line pnssun of l le>ooo x- m;; 2; e

:-- r ==? e prf rrj r-ef x_1 M t psig, Ci'f rxW'. ;r== . f ' =(20) ;:'t. l sw

3. The HPCI puso develops a flow of at least 4GGG.gpa against a l I i test line pressure of >*SeGG.) psig when staas is being supplied i

3/

to the tureine at 000", -l'O , - 20* psig.** l 4

c. At least onceqpe w18 w months:

tooep ao,-oo . .

! 1. For the 4&fr. the LPCI systas, and the HPCI systas, performing a i systes functional test whien includes simulated automatic actuation of the systas throughout its emergency operating -

sequence and verifying that each automatic valve in the flow path actuates to its correct position. Actual injection of -

- coolant into the reactor vessel any be excluded from this test.

l "tacast snas an automatic valve capable of automatic return to its ECC3 l position when an ECCS signal is present any be in po/sition for anotner mode l ef operation.  !

• • The provisions of Specification 4.0.4 are not applicable provided.the '

surveillance is performed vithin 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor staas pressure is adequate to perform the test.

1. Wke b b4. deier,Wnel Mag Pee-c? k'ECN- ,

1 bra emasw.

O! 070 (!Z /') 3/4 $-4 3 0 MAR 583 41 .

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!"f1GENCY CORE C00t.!*iG SYSTEMS ,

S'.'*v!!'. LANCE REEUIR!wfMTS (Continued)

2. For tne HPCI systes, verifying that:

a) The systes develops a f of at least '*"") gpa against a test line pressu n of psig, :;- ;:;:-r' - 1: : n::::-

gl , , _

.;;;;: -- ::_ : :" ' l"

-^

t ' . when staam ' b.. .e.. .'p__s.e ' ; W h a15 gig.

b) The suction is automatically transferred from sne condensate storage tank,to the suppression channer on a condensate storage tant water level - low signal anc on a suppression ,

enamner,- water level high signal.

d. For the ACS: .

1. At least once per 31 days, performing a CHANNEL FUNCTIONAL TEST l

o f g3, .. .. a .. . 9_ m . -

- n:: m as system lo p essure

' alars systes. p m .-** = . 4 la nd= = .e s

2. ' At lea'st one's per 18 aanths:

a).. Per.f.gcLing a system functional test which includes l l

simula(ed. automatic actuation of the system throughout its emergency operating sequence, but excluding actual valve ~

i actuation.

i l

b) Manually opening each ADS valve when the reactor steam done pressure is greater than or equal to 100 psig(**) and l observing that either:

1) -The control valve or bypass valve position responds l accordingly, or

2) There is a corresponding change in the asasured steam l 1

f1ow. Q w.:. J l,sM d M **-- "-"

c) Performing a CHANNEL CALIBRATION of theA_---

n

-' gas systen loJfessun alars system and verifying an alarm setpoint of {tf') * -js.} psig on secreasing pressure.

I .

. ""Tae provisions of Specification 4.0.4 are not applicanle provided the -

surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactar steam pressure is l

adequate to perfors the test.

N v.t.e 6. W / dei ' m,d J ,'$ . 7 % , u .,3 ,

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

...,_,y_, , _ _ . _.__...._-,_-,,,._.m.. . . - . . . _ - - _ - - - , _ .

E. met 3ENCY CORE COOLING SYSTEWS

. . 3/4.5.2 ECCS - SMUT 00W '

I LIMITING CONDITION FOR OPERATION As a msnimarry 98x emergersy cevs. caeling M vp4 ems 3.5.2 ?; y two of the followingshall be OPERA 8LE: l e.J l

a.Twe/ ore spray systen 4466) subsystems,with e subsystem comprised of:

l cm .

1. (?t ':::t :::) ITwo] OPERA 8LEMM pump si, anc l

2. An CPERA8LE flow path capable of taking suction from at least one of the following water sources and transferring the water through the spray sparger to the reactor vessel: l a) From the suppression chamber, or b) When the suppression chamber water level is less than the limit or is drained, from the concensate storage tank containing at least t 3 available gallons of water, ,l r- equivalenttoaleveljof12$1.

115.000 enek .

b. A Low pressure coolant injection (LPCI) system subsystemsj with e- - ,l subsystem comprised of:  :,

]

1. At ':::: OneOPERABLELPCIpump,and -

2. An QPERABLE flow path capable of taking suction from the suppression chancer and transferring the water to the reactor s' ll vessel.

~

i ApoLICABILIW: OPERATIONAL CONOITION 4 and 5". l ACTION:

4. Withoneoftheaboverequiredsubsystem(s) inoperable,restoreat least two subsystem (si to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or suscend j l , all operations with a potential for draining the reactor vessel. ,

i l b. W'ith both of the above required subsystems inoperable, suspend CORE l

AL48ATIONS and all operations with a potential for draining the reactor vessel. Restore at least one subsystem to OPERABLE status

}

withid 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or establish 0':COCi ;;TGiai INTEGRITY within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.  %'M*' ** Mis'"' ~ ~=")

"Tne ECC5 is not required to be OPERABLE provided that the reactor vessel head  ;

is removed, the cavity is flooded, the spent fuel pool gates are removed, and water level is maintained within the limits of Specification 3.9.8 and 3.9.9.

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D(RGENCY CORE COOLING SYSTEMS

$URVE!LLANCE RECUIREMENTS 4.5.2.1 At least the above required ECOS shall be esmonstrated CPERA8LE per Surveillance Requirement 4.5.1.

f 4.5.2.2 The core spray systas shalf be determinel0PEAASLE at least ones per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying the condensata storage tant required volume when the J condensata storage tant is required to be OPEAASLE per Specification 3.5.2.a.2.D). l l l

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~~.i! '17/2} 3/45-7 3 0 t'AP 1983 1

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. 10 JAN 883 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES l A.C. SOURCES - OPERATING LIMITING CONDITION FOR OPERATION I

As a minimum, the following A.C. electrical power sources shall be j 3.8.1.1 j OPERA 8LE:  ;

I

a. Two physically independent circuits between the offsite transmission networt and the onsite class 1E distribution system, and l

b. -/[o separate and independent diosal generators, each with; f

.a y .l

1. A separate fuelltankt containing a minimum . j of y gallons of fuel, g g g,7jg , ,f j

_;M N

. 2. A separate fuel storage systeng ^_ ' 'n;p ::...- of , -  ;

gallons of fuel, and

3. A separate fuel transfer pump er-e.b E c g e. b b .

Appt.fCAntti;Y: OPERATIONAL CONDITIONS 1, 2, and 3.

a

\_, ACTION: ._

g*

a. With either one offsite circuit or one diesel generator of the above  ;

cuired A.C. electr' cal power sources inoperable, demonstrate th O P ., ILITY of tne r:maining A.C. scurces by perforcing Su'ved 'h n v f Requir nts 4.8.1.1.1..a and 4.8.1.1.2.a.4, f or one diesel narat:1 ', !L hereafter; at a time, itnin one hour and at least once per 8 hourabove required 0 f restore at le

• two offsite circuits and 4ep"- of t  !

t diesel generator o OPERABLE status w' thin 72 rs or be in at least HOT SHUTDOWN w in the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> in COLD 5HUTDOWN wi' thin the following 24 urs. ,

t

b. With one offsite circuit and o sal generator of the above required A.C. electrical power sources .p le, demonstrate the OPERABILITY of theirosaining A.C. Sou by per ing Surveillance Requirements e and at least once per j

4. 8.1.1.1. a and 4. 8.1.1 .a.4 within one t 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter; store at least one of above required inoperable A.C. es to 0PERA8LE status with 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in OWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and COLD SHUTDOWN at least HOT within t andallowj Restore at least two o ite ,

circu tows pg 24 hours.bf the above. required diesel generators. (

in OP LE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from time of initial loss or t least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTD t

' within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. '

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INSERT A TO PG. 3/4 8-1:

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i For two diesel generators, a fuel oil transfer pump may be '

inoperable for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> without declaring the asso-  !

ciated diesel inoperable provided the companion fuel oil  ;

storage tank and transfer pump are OPERABLE. This pro- '

vision does not apply to fuel oil transfer pumps which have been tagged out for tank sediment settling following filling.

t Following fuel oil storage tank filling, the tank and I associated fuel oil transfer pump shall be tagged out for up to 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> to allow for sediment settling. However, the associated diesel shall not be declared inoperable provided the companion fuel oil transfer pump and storage

• tank are OPERABLE.

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

L v a i lAISCAT A To M- 3 Y f ~ l .'

A cTie AJ!

of the above reeuired desel ear +*

• .Witn '

ent.-

inspenaht. demonstrata me OPOASILITY of the ec=~a. - ~'

seurtes 99 pe,rfereing Surveillance Requimment -

A.C.

4.8.1.1.1.3 wipin 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 8 hears 111anca Resul thereafter; and 1esel generstar to OP DASLI status witnin t 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; restsor to in at least NOT amntow witnin the nort it. ~

, l 3 9L) says-hours and in COLD $NUTD308 witnin the following 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />. ,

- l i

h k h[tn two of tne above required diesel generators ineoenble,

...% A.C. seus ey desenstrate the OPGA41LM of. Oe. ft .

per*sming Surveillans Recutrwent 4.8.1.1.1.a 4restare witnin at least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> one and at least once see 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> enereafter:

of tne inocersele diesel genersta-s to 07 Bast.L status 1 win -l TL*ours or te in at least & sum witnin sne next it neurs and in COL 3 $HUTOCWM witnin ue following to Meurs.  !

• ~

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SewaA l 4 A u _1. a . <  !

uutw 24 W l i

i Wita tRree diesel generstars of the aseve reeuired'A.C. electrical i peer _sourgas tamperusle, demonstrate tas OPOASILITY of tan remaining f L C. seurges by Wrfefeine Surveillants teouirements 4.8.1.1.1.a. and witAin 1 Roerf l

4. 8.LL- 2a. 4. =

restsee at least one of tas inse-l ernste eiesel geneesters ts OPDASLI status within 2 tours er to in  !

st 1sent, iST SetTD0hel wittie tan meet 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and is CDL3 Se#Whm l witate the fellowing 24 heers. t l

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A.ui$fr an offsite circuit of the above recuired A.C. electrical ~~ '- - -

power sourtes incoertale, comonstrate the OPGA8!LITY of tne reesining A.C. sourcer by performing Surveillance --

Recuirement 4.8.1.1.1.a witnin 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once par 8

• hours tnereafter; and Surveillance Reevirement 4.8.1.1.2.a.4 -

witnin 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />; restore at least tese offsite circuits 1 to OPGA8LE status witnin 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or te in . --

i at least HOT Swoow,4 within tne next f?. nours and in COLD SWTDOWN witnin tne following 3,4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. . . . . . .

e.,Wita two of the acave reevired offsite A.C. ci rcui ts , _ _ _

incoennie, demonstrate the CPERASILITY of Bt diesel ,

generators b, perforwing Surveillance Recuirement . , , _

~ 4.8.1.1.2.a. A witnin 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> unless tne diesel gsnerators are alevacy coerating; restore at least one of -

tae incoer:51e offsita sources to CPDARLE status wicin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least NOT Samoav witnin the next r?. hours Q,r=le b O*. S C.,cw 4 A. ( % setow... ;4

• hours .

J

._ .._ _ _ "sm With one offsite cirtvit and one diesel generator o* the above rTouired A.C. electrical power sour:ss incoerable, demonstrata ce OP RA81LITY of tne maaining A.C. Sourte by .

perderming Surveillance Recuirm 4.8.1.1.1.a witnin 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> I.

ane at least once eer 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafterjano Surveillance

'I. - ~ ".

Recuirerent 4.8.1.1.2.4.4 witnin 8 nours; restore at least one . '

1 of the inoce-tele sourtes to CPGABLE status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> q

or be in at least NOT o N witnin tne next a.nours and in

• COL 3 SW":CwM witnin the following ;4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

a .

M SevtJ!ce Y**.,

ye Wita one offsita circuit and the dissal generstars of the aseve eveuf fic A.C. electrical power sources ineserar te, semenstrats the OPDA8!LITY 4 N 'd- of the remaining A.C. saurtas by performing $seve111ance Asquirements witain 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least ense poF 4.4.1.1.la.

eN.1 1.6rij a % u af tar-Armstare at least one of the aseve roeutree g inesereste A.C. seurtes. ta OPtAASLA status vitain 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> se tea in ,

}

at.least ICT SWTOObst vitain the nest.12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLO 941712s>

within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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10 JAN 883 ELEC*RICAL POWER SYSTEMS SutvEILLANCE REQUIREMENTS i

I 4.8.1.1.1 Each of the above required independent circuits between the offsite transmission network and the onsite Class 1E distribution system shall be:

a. Determined OPERA 8LE at least-once per 7 days by verifying correct breaker alignments and indicated power availability, and

b. Demonstrated OPERA 8LE at least once per 18 months during shutdown by transferring, manually and automatically, unit power supply from the normal circuit to the alternate circuit.

l 4.8.1.1.2 Each of the abov's required diesel generators shall be demonstratec l CPERA8LE:

a. In accordance with the frequency specified in Table 4.8.1.1.2-1 on a -

STAGGERED TEST SASIS by: a4

1. Verifying the fuel level in the =; rd r;' rru-M fue1Atankt. l

.it

2. Verifying the fuel level in the fuelAsterage tanks. ,

3. Verifying the fuel transfer pump starts and transfers fuel fres the storage system to the ::3  : x;in x ;nt:0 fue1Vtanks. I m .a ,

4. Verifying the diesel sta from ameient condition and accelerates to at least rpm in less than or equal to g . ,

j, to t-rer seconcsl The generator voltage and frequency sh 1 be i

CT N Ma#-

secones 4.41601 : 142Cf volts anc J60) 211.2{ H: within "

Camel. after the start signal. The diesel generator sna11 he started

~

for th'is test by using one of the following signals:

I a) Manual.

b) Simulated loss of offsite power by itself.

c) Simulated loss of offsite power in conjunction with an ESF actuation test signal. .

d) An ESF actuation test signal by itself.

• 5.

44sc Verifying the diesel generator is/synenronized, loaded to g# r greater than or ecual to ( r n' _;: W :;; kw '- 1:n 2:n r 5+.* --- , ; ;21 = ( M :r rc:, and operates with this load for at least 60 minutes.

E. Verifying the diesel generator is aligned to provide staneby power to the associated emergency busses.

T7. Verifying the pressure in all diesel generator air start ,

receivers to be greater than or equal to-(46& psig.{

leggar T F. *

• g mp aso .

i c.p. At least once per 31 days and after each operation of the diesel where the period of operation was greater than or equal to I hour by checkTng i for and removing accumulated water from the day 2 x;i n x .nt:0 l

. fuelytanks.

le. : + en e s.

$ % 4,o.nl vt.cf s#r== as ik*at . 4,6.s s Lil ,.k 1. n . .s

+.

ufor-e/

. r,.Le,. n . ~ . .+g , . e,# ,.,.,.ar,.r.5 a/,- n , .

e~ ~rr. 4 1 +,.w. g. -h- a e ~.w+-A.

21'*".% ';;;;;r & ,f 3 3 :wl,'5.';:n v *- -

-- - - ~ u w a,w,,,7,4,*2s

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. v! 'a INSERT TO PAGE 3/4 8 - 2.

8. Verifying the lube oil pressure, temperature and differential pressure across the lube oil filters to be within manufac-turer's specifications.

b. At least once per 31 days by visually examining a sample of lube oil from the diesel engine to verify absence of water and by verifying a minimum of fee +y. 55-gallon drums of lube oil are stored onsite. t t

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ELECTRICAL POWER SYSTEMS  ;

SURVEILLANCE REQUIREMENTS (Continued) ,

d.*g. 5t 1elit'orIce"pe'r~( d D ' ;?$d f d*:r '- M h : :1 ': :: .M;tr '- -

^22: '2 : M"- Of it ' i3) 1921 days by removing accumulated water from the fuelvstorage tankisl.

i M*9 6 .d g At least once per 92 days and from new fuel oil p 'rior to addition to

.f* the storage tanks by obtaining a sample in accordance with ASTM-D270-1975 and by verifying that the sample . wets the following 1 minimum requirements and is tested within the specified time limits:

, 1. As soon as sample is taken or from new fuel prior to addition to the storage tank, as applicable, verify in accordance with the tests spectfied in ASTM-0975-77 that the sample has:

a) A water and sediment content of less than or equal to ,

0.05 volume percent. -

b) A kinematic vis&cosity 9 40*C of greater than or equal to 1.9 centistakes, but less than or equal to 4.1 centistokes.or .

t+.

c) A specific gravity as specified by tYhs@tt mIdiI5actur'eI-um.a wuq5'"e too'r

'0/P'e :f ; ;:t:r the :r - r! u ht ?:n tMr :- - r1 e :r as AP1 gravity 9 60*F of

"~ greater than or equai to es cagrees but less than s or soual to 4 :L degrees.

2. dithin one week after obtaining the sample, verify an impurity level of less than 2 se of insolubles per 100 ml. when tested in accordance with ASTM-D2274-70.

1

3. Within two weeks after obtaining the sample. verify that the '

other properties specified in Table 1 of ASTM-0975-77 and Regulatory Guide 1.137, Position 2.a. are met when tasted ws47" in accordance with ASTM-0975-77. . ,

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i 4,s. At*1 east once pier 18 months, during shutdown, by: ,

1. Subjecting the diesel to an inspectien in accordance with prbcedures prepared in conjunction with its manufacturer's recommendations for this class of standby service.

2. Verifying the diesel generator capability to reject a load of n-

' greater

..n:1 ;;;r than;"Orors1*) egual .-d to (i n ,.;t

,..:^^r "--; h; - ^^-- h r:r;I 'j (th.rd; -

..:tt"9.-

f:r % W r:r;; ny h:d) h" fr c h ni ;;..;r;^.n (10) while maintaining ,h (voltage at (41601

• 14201 volts and frequency at (60) : fi4. )3Hzl (engine speed $ 75% of the ' difference betwen nominal speed and the overspeed triqsetpoint or 15% above nominal, whichever is less]

at nem6 d5 h He9s csrtw 3 .

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L-INSERT TO PAGE 3/4 8-3

e. At least once per 31 days by performing a functional test on the emergency load sequencer to verify operability.

g. At least once per two months by verifying the buried fuel oil transfer piping's cathodic protection system is operable and at least once per year by subjecting the cathodic pro-taction system to a performance test.

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! O JAN 1983 d I

i ELECTRICAL power SYSTEMS J

l

- SURVEILLANCE REQUIREMENTS (Continu~ed) l

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3. Verifying the diesel generator capability to reject a lead of 443o ' :-t' cr =t' ;) kW without tripping. The* generator voltage shall not exceed ?") volts during and following the load  ;

rejection. 4 80 l

4. Simulating a loss of offsite power by itself, and:

less

" 3 "S) Verifyingldoenergization of the emerg'ency busses and load shedding from the emergency busses. Jg. % ht sg b) Verifying the diesel generator starts on the auto-start [

signal, energizes the emergency busses with permanently '

connected loads within S89 seconds.Aenergizes the auto-connectec (shutdown) loads through the load sequencer O l and operates for greater than or equal to 5 minutes while its generator is loaded with the stutdown loadse Afs,7

. energization, the steady stata voltage and frequency of i the emergency busses shall be saintained at {41607 2 (

34202 volts and (60J z J1.2) Hz during this test.

5. Verifying that on an ECC5 actuation test signal, without loss of offsita power, the diesel generator starts on the auto-start l' signal and operates on standby for greater than or ecual to 5 minutes. The generator voltage and frequenc ll be (41601, 2 (4201 volts and1601:T1.21.Hz within4&49'y seconds shaafter the

. . ,. auto-start signal; the steady stata generator voltage and fre- g

. . - ouency shall be saintained within these limits during this tast. t

' 6. e

• on a simulated loss of the diesel en ' * . . . t offsite power no the 1 ros the emergency i busses and that diosal generator is ,

. _ ...___ __ - r , with design requirements. t 6,Z Simulating a loss of offsite power in conjunction with a'n ECCS '1 actuation test signal, and: i u s p...e is deluLf . l a) Verifyinghee,,ergizationoftheemergencybussesandIcad.

n l  :

shedding from the emergency busses. .w M dad P,

. g .

Verifying the diesel generatoe starts on the auto-start j b) signal, energizes the emergenr.y busses with pemanently .

connected loads within SDFseconds energizes the l auto-connected shutdown loads through the load sequencer and operates for greater than or equal to 5 minutes whileitsgeneratorisloadedwiththeemergencyleadI.g

- After energization, the steady state voltage and frequency  !

of the emergency busses shall be maintained at (4160L2 j

$4201voltsand$601231.27Mzduringthistest. .;

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-7,/. Verifying 'that all automatic diesel generator trips, except l j

1 engine overspeed,and, generator differential current 3vare t automatically bypassed upon loss of voltage on the emergency )

bus concurrent with an ECCS actuation signal. - I

&.4 s dea u n ts s r- en me.wwe.h ,

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k<.er:sfie

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10 JAN $83 ELECTRICAL POWER SYSTEMS

• i i

SURVEILLANCE REQUIREMENTS (Continu~ed

. J, f. Verifying the diesel generator operates for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. (

During the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of this test, the diesel generator 46U shall be loaded to greater than or equal to ( -r. a ;;ing)/ kW and during the remaining 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> uof tgis test, the diesel r:t'-- los. The generator shall be loaded to (;;ntinu::: generator voltage and {601211.21 Hz within ftW seconds after the start signal; I the steady stata generator voltage and frequency shall be main- I tained within these limits auring this test. Within 5 minutes after completing this 24-hour test, perform Surveillance g l

, Requirement 4.8.1.1.2.g.4.b)."

l

1. K. Verifying that the auto-connected loads to each diesel generator l do not exceed the 2000-hour rating of ( ) kW. i 4737 l

fg. )s. Verifying the diesel generator's capability to:

a) Synchronize with the offsite power source while the -

generator is loaded with its emergency leads upon a simulated restoration of offsite power, l U) l b) Transfer its loads to the offsite power source, and ,

" c) 8e restored to its standby status,and

\

d) Diessi s & *e samtf bivaker is &w L')

//. W. Verifying that with the diesel generator operating in a test '

mode and connected to its bus, a simulated ECC5 actuation signal overrides the test mode by (1) returning the diesel generator l to stancty operation, and (2) autos.atically energizes the emergency loads with offsite power. _

• • t with all diesel generator [ air '

elvers "I

{13.Ve . .

e ecu psig and the pressurizee to . ,

compressors secured * . ter starts at least 5 times * = ient conditions and ac:eie ..;: " 49964 51'i- 0 Min less than or ecual to '")* seconds.Y e.L st

/2.Jr. -Verifying that the fue1Atransfer pump transfers fuellfrom each l fue17 storage tank to the day :: : ;'n; r-tM tanks of eacn L l diesel via the installed cross connection lines.

M..g. Verifying that the automatic load sequence timer is OPERABLE l l

- with the interval between each load block within 210ll of its l design interval. , ,

"If Surveillance Requirement 4.8.1.1.2. 4.b)is not satisi.Gw-ily Instead, completed ~

the l

it is not necessary to repeat the preceding 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> tast. I diesel generator may be operated at (.; tict:::+tM :t' ;) kw for one hour o until operating temperature has. stabilized.

~~% INSST 0~ 07 (b/5) 3/48-[

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SUPERSCRIPT NOTES FOR PAGE 3/4 8 - f*

(1) Test is per IEEE 308-1974, Section 6.4, to demonstrate that the Class II loads can operate on the preferred power supply. , ,

(2) I"Tr 308-1974, Section 6.4, requires demonstration tha+ the standby power supply is independent of the preferred power supply. The surveillance requirement for verifying that the diesel generator circuit breaker is open after the diesel generator had been synchronized. with the of fsite power source, transferred its leads to that source and re-stored to its standby status provides this demonstration.

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l P ECTRICAL POWER SYSTEMS f

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l RVEILLANCE REQUIREMENTS (Continued)  !

N'M. Verifying that the following diesel generator, lockout features,w h w l

4.ged, prevent diesel generator starting: =b tr - ; '-= l s re a - . .... -_. a s [

p N ) ; ' ' ' ' _ 'l' "' n . '". Y, ~ " ' - ' '  ;

-, n....., .,

t I

/#. At least once per 10 years or after any modifications which could {l affect diesel generator interdependence by starting both diesel generators simultaneously, during shutdown, and verifying that both I diesel generators accelerate to at least (9eet.rpa 3**

in less taan I or equal to-96-y* seconds.

At least once per 10 years by: l l jj.

l

1. Draining each fuel oil storage tank, removing the accumulated

'sedimentandcleaningthetankusinga(sodiumhypochlorite) ,

L i

solution, and l i

2. Performing a pressure ' test of those portions of the diesel fuel

• oil system designed to Section III, subsection N0 of the ASME i I l' Coce in accorcance with ASME Code Section 11 Article IWD-5000.

l 4.8.1.1.3 Reeerts - All diesel generator failures, valid or non-valic, shall be esported to tne Constission pursuant t.o Specification 6.9.1. Reports of l diesel generator failures shall include the information recesumended in j Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977. i If the nuncer of failures in the last 100 valid tests, on a per nuclear unit I basis, is greater than or equal to 7, the report shall be supplemented to

/

incluce the additional information reconsended in Regulatory Position C.3.b of .

l [

J Regulatory Guice 1.108, Revision 1, August 1977. i ch, hp eoecs,eeA , geweed ev- dMece d.u.A , a.sA \ow 6%,,

en premy e, ( repke \ecked re.\a,3 ,(,0 84Q. .

Q3a.c.k.q gewerdov- 4'3=cew M .emA ge%,d.',me

- (h69 be.keJ ec.bj ,M e6a ). l e)Ae.we.eche- gmyl ud \eeked ye.\.9-.Trrecp\w, f f s ,b u. t.c.w . 4 - l WL ? u A b t , ewer y.e.1 (. bee 6 .c l

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b TA8LE 4.8.1.1.2-1 OIESEL GENERATOR TEST SCHEDULE

, Number of Failures in Last 100 Valic Tests" Test Frecuency

<1 At least once per 31 days ,

2 At least once per 14 days 3 At least once per 7 cays

>4 At least once per 3 days ,.

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"Crl erla for oe:er:1ning nuncer of failures anc numoer of valic tests snail me in accorcance with Regulatory Position C.2.e of Regulatory Guice 1.1C8, Revision 1, August 1977, wnere tne last 100 tests are determinec on a per nuclear unit easis. For the pura.,oses of this test senecule, only valic tests concucted after tne OL issuance cate snall ne incluced in :ne concutation of the "last 100 valic tests." Entry into this test senecule shall be sace at the 31 cay tast frequency. .

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l 10 JAN 1983 .

ELECTRICAL POWER SYSTEMS A.C. SOURCES - $HUTDOWN ,

LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be OPERA 8LE: .

a. One circuit between the offsite transmission network and the onsite Class 1E distribution system, and

%e er e.h

b. -Gae diesel generatorsArith:

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1. A t yser..  :-d ts.  :- ' : xxt:d fuelltanks containing a sinimus of ,

f956-) gallons of fuel. ,.

A80 c. air e r., .f + tsa 3-lu'*'. "w" ^ "*t.1 n L s at

2. A fuel storage system ---"' "- - -' '

4p,tu ,=4=4 of ' fuel, A

3. A3 transf er. pumpA (,e sad st.c. ha E .

AP8LICA!ILITY: OPERATIONAL CONDITIONS 4, 5 and *.

ACT!fN:

, si'5'

a. With less Ahan the above required A.C. electrical power sources  !

n.4f OPERABLE / suspend CORE ALTERATIONS, handling of irradiated fuel in g,g;,D@the reactor vessel and crane operations over the spent fuel storagej::

pool wnen fuel assemblies are stored therein. In addition, when in.

OPERATIONAL CONDITION 5 with the water level less. than ( :' 'n; n'-2" above the reactor pressure vessel flange, immediately initiate cor:rective action to restore the required power sources to OPERABLE status as soon as, practical. ,

b. The provisions s

of Specification 3.0.3 are not applicable.

SURVEILLANCE REOUIREMENTS i

4.8.1.2 At least the above required A.C. electrical power sources shall be

- demonstrated OPERA 8tE per Surveillance Requires,ints 4.8.1.1.lg 4.8.,1.1.2x

.c.: '.".1.1.2, except for the requirement of 4.8.1.1.2.a.5. awa b @ IM f=j.

l Nhen nancling irraciated fuel in the s  :::rdrj ::-t n~

.b %. M.. -IM

. , ens 3/4 3.g* ll

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s INSERT A TO PG. 3/4 8- g :

Following fuel oil storage tank filling, the tank and associated fuel oil transfer pump shall be tagged out for up to 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> to allow for sediment settling. However, the associated diesel shall not be declared inoperable provided the companion fuel oil transfer pump and storage tank are OPERABLE.

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Af7NJbnrrirt j NRC RESOLUTION TO HOPE CREEK WRITTEN COMMENTS ON R0 AND SRO WRITTEN EXAMS (Letter to John Berry dated July 12, 1985)

Hope Creek Comment on Question 2.016 QUESTION: State (the four) or: State four (4) signals that will isolate in RWCU F001 and F004.

CONCERN:

There are actually five signals that will provide this response. Any four of the five should be credited for full response. They are:

1. RRCS initiation

2. Hi RWCU different flow 60 gpm w/45 see T.D.

3. -38" RPV level.

4. Hi area temperature

5. Hi area differential temperature RECOMMENDA'..'I ON : Accept any four of the above listed five for full credit.

REFERENCE:

LPs 21, Pg 66, Rev 1 NRC Resolution: The answer key indicated that four out of the above listed 5 was all that was required.

Hope Creek Comment on Question 2.09b QUESTION:

What effect would a loss of instrument air have on the Standby Liquid Control System?

ANSWER: Loss of level indicator CONCERN:

The equipment installed are differential pressure cells and is supported by a recent documentation change. The lesson plan does mention a bubble-type level indicator device.

RECOMMENDATION: Consider an answer of no effect if response states differential pressure cells as correct.

REFERENCE:

LP# 23, Pg 11, Rev 1 NRC Resolution: Hope Creek Recor:mendation accepted L

Hope Creek Coment on Question 2.09c QUESTION: To the effect: When can SBLC injection be terminated?

CONCERN:

Answer key states as per LP, the SBLC injection is to go to completion. Subsequent to systems '

training, Emergency Operating Procedure training stated SBLC injection is to be terminated when all rods are verified to be at least position 02 or beyond (00).

RECOMMENDATION:

Both termination and completion should be acceptable 2

' answers with supporting qualifications stated by the examinee.

REFERENCE:

EOP:

OP-EO.ZZ-101(Q) Rx Pressure Vessel Control, Step RC/Q-2 i

EOP LP: LP# 302HC-000.00-124-00 RPV Control, Pg 8, Step 6 i

NRC Resolution: This question was as follows:

"Once the SLC System is initiated, injection of the entire contents of the SLC storage tank must be allowed to continue to completion (TRUE or FALSE) ."

The answer was changed to " FALSE" since E0P's contain the more up to date information.

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Hope Creek Comment on Question 3.05b QUESTION: To the effect: A Reactor Building Hi Hi rad isolation signal is received. What three (3) things isolate?

CONCERN: There are many more than three (3) actions that occur due to RxBldg Hi Hi.

s

1. RxBldg Supply & Exh. System isolates

2. FRVS starts

3. Control Room Emerg. Filt. System starts s

4.

SACS A&B Pumps start (if not already running)

5. Service Water Pumps A&B Start.

6. Isolation Dampers Close HV-9370A&B, 9414A/B

7. PCIS Influenced Systems A.

B.

PCIG (Prim. Containment Inst. Gas)

CACS (Containment Atmos. Control System) l C. MSIV Sealing System l'

D. D/W Floor & Eq. Drains E. Torus Water Cleanup F. Containment H2 Recombiner System (CHRS)

G. Emg. Inst. Air

REFERENCES:

A.

Abnormal Procedure OP-AB.ZZ-126(Q)

Concerns 1,2,3 on Pg 1, 2.0 Auto Actions LP# 80, Pg 18 B. LP* 42 (RBVS), Pg 24/25, Item 3.b.1-5 (SACS)

C. LP# 44 (PCIS), Table #2,3,4 & Fig # 2,3,4 I

D.

SOP'st OP.-SO.EA-00l(Q) Service Water, Pg 3 E. P&ID# M-ll-1 Sht 1, @Ccor. E-4, SACS RECOMMENDATION: Expand your answer key to accept this wider rance of accepted responses.

NRC Resolution: Hope Creek Recommendation Accepted.

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Hope Creek Comment on Question 4.10 QUESTION:

Regarding AB-OP/ZZ-ll7, Reactor High Water Level Procedure stated that high water level occurred.

ANSWER:

Key was looking for a response of reactor power increasing.

CONCERN:

Information was given to the effect that the reactor did not scram. If one were to assume this means that the reactor was greater than 30% power as determined by first stage pressure and reactor level of 54" was not reached, then power would increase. On the other hand, if one were to assume that the reactor was greater than 30% power as determined by first stage pressure and reactor level of 54" was exceeded and the reactor did not scram but the turbines tripped, power would increase on the ATWS.

One could also assume that reactor power was less than 30% as determined by first stage pressure and reactor level exceeds 54". Then turbines trip, power would decrease due to reduction / stoppage of flow through FW heaters.

This is a fairly open-ended question.

RECOMMENDATION:

This question could have gone various pathways to an end point. Required numerous assumptions.

Consider since only alternate answers one symptom with 3ustifications was stated. The high water level could occur for numerous reasons which include reactor recirc pump trip, SRV opens, by pass valves opens in which case power would have decreased. Consider alternate answers based upon assumptions.

NRC Resolution: Hope Creek Recon:mendation Accepted; however, wasn't necessary since candidate's answers were, in general, in line with answer key.

Hope Creek Comment on Question 4.11

QUESTION

To the effect: When are you allowed to override an ECCS initiation signal and secure the ECCS systems?

CONCERN

Answer key' states

1. Inappropriate initiation

2. Operation beyond trip pts.

f RECOMMENDATION: Also allow "misoperation of auto mode"

REFERENCE:

EOP-ZZ-101. Step RC/L-3, Operator Caution #10 NRC Resolutic,n: Hope Creek Recommendation Accepted.

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L Hope Creek Comment on Question 5.05 i

QUESTION: .

To the effect: The reactor was depressurized from 8350 to 5000 in 30 minutes. This resulted  ;

in a 62*F temperature drop in 30 minutes. i Was or is the cool down rate exceeded? ,

CONCERN:

Not sufficient information provided for either a yes or no answer.  !

Operators can with qualification justify both yes/no responses. l T.S. 3.4.6.1.a requires a H/U-C/D rate of <100*F  !

• in one hour's time. Therefore, operator i must consult vessel temperature / pressure  ;

history for previous 30 minutes and control i vessel temp / press cool down rate for subsequent 30 minutes.

r RECOMMENDATION:

Both yes and no responses are (should) be accepted.  ;

Point value should support the amount of justification desired.

t

REFERENCE:

Hope Creek T.S. 3.4.6.1.a Pg 3/4.4.-19 T.S. 4.4.6.1.2 Pg 3/4.4.-20  ;

f NRC Resolution: Hope Creek Recorr.endation Accepted.

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4 k Hope Creek Consnent on Question 6.03b I QUESTION: To the effect: What conditions will make the  !

main steamline radiation monitor inop and what four (4)

I-auto station responses will occur? -

CONCERN:

i If Group I is not subdivided, there are four '

responses; responses occur if Group I is subdivided then five  ;

1. MSIV's close - Group I

2. MSL Drains Close - Group I '

3. Rx Scram t 4.

S. HOGGER - Mech. Vac. Pump Trip 4310, 4311 - Sample valve on recirc close ,

i RECOMMENDATION:

Accept any four of the five listed.

P

REFERENCE:

LP# 45, NS4, Pg 20

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NRC Resolution: Hope Creek Recommendation Accepted. 5 i

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Hope Creek Comment on Question 7.05b

. i QUESTION:

To the effect: What are the auto station responses t to a loss of instrument air? Is a manual scram required?

CONCERN: A manual scram as per a loss of instrument air abnormal is a subsequent action and as per OP-AP-ZZ-102, Step 5.1.1.3 (attached);

subsequent actions are not required to be held to memory.

RECOMMENDATION: On loss of air, the AB requires a manual scram only if "muliple rod drifts" begin to occur.  ;

REFERENCE:

OP-AB-ZZ-131 "

Loss of Inst. Air", Step 4.8 OP-AP-ZZ-102 Use of procedure, Step 5.1.1.3 l (attached) y i

NRC Resolution: Answer key changed to indicate manual scram necessary l (if multiple rod drifts occur) l 4

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Hope Creek Comment on Question 7.06b and c QUESTION: Three part question on maximum dose limits.

CONCERN: The answer to part "c" was dependent on the answer to part "b" of the question. This could be a double jeopardy question.

RECOMMENDATION: Consider deleting part "c" of the question.

REFRENCE

OP-AP ZZ-024, Rev J. Pg 41 i

NRC Resolution: Hope Creek Recommendation not accepted. The question was divided into parts b and c for clarity. Point count of part c was only 0.5.

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Hope Creek Conenent on Question 8.02c-QUESTION:

In accordance with OP-AP.ZZ-992 (assume OP-AP.ZZ-002),

when is a second SRO not required in the control room? i ANSWER:

Key stated operational conditions 4&5, Section 5.6.4.2 CONCERN:

This section of AP-AP.ZZ-002 is out of context  !

and must be considered with Section 5.6.4.1.

If "second" was understood to mean two SRO's j

in the control room, "never" could be a correct ^

answer. However, if taken in context then

] " operational conditions 4&5" would be the only answer.

RECOMMENDATION: Consideration should be given to both answers  ;

of "never" and " operational conditions 4&S". l

REFERENCE:

OP-AP.ZZ-002, Rev 7, Section 5.6.4.1 and 5.6.4.2  !

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l NRC Resolution: Hope Creek Reco:mendation accepted.

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Hone Creek Comment on Question 8.07 QUESTION:

To the effect: According to OP-AP.ZZ-992, when must the procedure be present and followed step-by-step.

ANSWER:

Key referenced OP-AP.ZZ-002, Section 5.15.1 as the correct response.

CONCERN:

There is no OP-AP ZZ-992 and a candidate could assume this procedure to be OP-AP.ZZ-102. Then, steps 5.1.8.2.a for Integrated Operating Procedures and 5.1.ll.2.a for Surveillance Test Procedures would be the correct answer. If OP-AP.ZZ-002 was assumed, the key would be correct.

RECOMMENDATION: Accept both answers, sections 5.1.8.2.a and 5.1.ll.2.a for OP-AP.ZZ-102 and 5.15.1 for OP-AP.ZZ-002.

REFERENCE:

OP-AP.ZZ-002, Rev g OP-AP.ZZ-102, Rev $

NRC Resolution: Hope Creek Recommendation not accepted. Although number of procedure was incorrect, the correct name of the procedure was given. Therefore confusion with OP-AP.ZZ-102 is not likely.

Mope Creek Cement on Question 8.08 QUESTION:

To the effect: Plant is at 95% with Rx Recirc pump "B" speed 8% greater than "A".

An attempt is made to match pump speeds by reducing (assumed on my part) the speed of B. The "B* pump scoop tube subsequently locks up and upon "B" tube reset the "B" pump now is 12% speed scoop greater than "A". What is your actions?

CONCERN:

The initial concern is the speed mismatch of 8% and T.S. 3.4.1.3 allows < 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to correct, declare "A" (slower speed pump) inop. But due to corrective actions taken, "B" in fact becomes inop due to the scoop tube lockup and as per T.S. 3.4.1.1.a, $2 hours are allowed to comply with Fig 3.4.1.1-1 (45% core flow /PWR curve) and/or correct.

In the two-hour grace period, the speed mismatch must be corrected and the scoop tube lockup must be corrected. Now depending on SRO interruption, one of two answers should be allowed with supporting justification.

1. SRO declares "B" inop and trips T.S 3.4.1.3.a action is satisfied ("A" pump is no longer the slower speed pump - "B" is)

SRO now is concerned with T.S. 3.4.1.1.a and has 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to satisfy power / flow requirements of Fig 3.4.1.1-1 and $12 hours for controlled S/D.

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2. SRO declares both recirc loops inop, "A" due to slow speed loop as per T.S. 3.4.1.3.a and "B" inop due to loss of control capability (definition of operability' and therefore is responsibile to comply with T.S. 3.4.1.1, action b.

Bottom line - either T.S. 3.4.1.1, action "A" or "B" can be justified and should be accepted.

REFERENCE:

T.S. 3.4.1.1 & 3.4.1.3 NRC Resolution: Hope Creek Recomendation accepted as long as candidate (if he chose answer 1 above) indicated he declared "B" inop and' tripped pump.