IR 05000354/1986034
| ML20214Q096 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 09/08/1986 |
| From: | Keller R, Kolonauski L, Lange D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20214Q077 | List: |
| References | |
| 50-354-86-34(OL, NUDOCS 8609240173 | |
| Download: ML20214Q096 (43) | |
Text
U.S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT N (OL)
FACILITY DOCKET N FACILITY LICENSE N NPF-50 LICENSEE: Public Service Electric and Gas Company Post Office Box 236 Hancocks Bridge, New Jersey 08038 FACILITY: Hope Creek Generating Station EXAMINATION DATES: July 7-11, 1986 CHIEF EXAMINER: bl@waAAMo 9[4 [ S(3
_Ly Kol'onauski, Reactor Engineer Examiner date REVIEWED BY: M '
d 9 '/ 96 David J. Ladfe, Lead R Examiner '
dafte REVIEWED BY: ff )
Robert M. Keller, Chief, Projects Section 1C
~
N'/[Y6 date APPROVED BY: -
HdrryB.K$er,Chlef,ProjectsBranch1 /d#te SUMMARY: Operator licensing examinations were administered to thirteen (13)
Senior Reactor Operator candidates during the week of July 7,1986. One candidate failed the written examination. All candidates passed their operating examinations.
A partial Licensed Operator Training inspection was conducted in accordance with Inspection Procedure 41701 of the Inspection and Enforcement Manual. No violations or deficiences were identifie PDR 860910 ADOCK 05000354 PDR
REPORT DETAILS TYPE OF EXAM: Initial EXAM RESULTS:
l SRO l l Pass / Fail l 1 l l 1 l l Written Exam l 9/1 l 1 l l I I l l Oral Exam l 12/0 l 1 I l l l l l Simulator Exam l 12/0 l 1 l l l l l l Overall l 12/1 l l l l Chief Examiner at Site: Lynn Kolonauski, NRC Other Examiners: David Lange, NRC Brian Hajek, NRC Consultant William Cliff, PNL Summary of generic strengths or deficiencies noted on oral exams:
Strengths were noted in the following areas:
The candidates were proficient in their use of piping and instrumen-tation diagram * All candidates displayed a responsible attitude toward the SRO positio *
The candidates were well-trained in the use of Emergency Operating Procedure Weaknesses were noted in the following areas:
Some candidates were unfamiliar with site personnel response during an emergency including activation of the TS *
A few candidates had difficulty in identifying the various fire extinguishers located throughout the plant (i.e., dry chemical, CO2, etc.) as well as what types of fires each was effective in suppressing. Personnel Present at Exit Interview:
NRC Personnel Lynn Kolonauski, Operator Licensing Examiner David Lange, Region I Lead BWR Examiner Richard W. Borchardt, Senior Resident Inspector David Allsopp, Resident Inspector Facility Personnel Stan LaBruna, Assistant General Manager, Hope Creek Operations H. Denis Hanson, Manager, Nuclear Training, PSE&G Calvin Vondra, Operating Engineer A. E. Giardino, Manager, Station QA Ralph Donges, Licensing Engineer Ron Schaffer, Assistant Manager, Operations Training Dennis Kabachinski, Training Supervisor, Hope Creek Simulator 3. Summary of NRC Comments Made at Exit Interview:
- The generic strengths and weaknesses noted in Paragraph I were discusse *
The personnel of the training and Operations departments were coopera-tive throughout the examination perio * The material provided by the training department for preparation of the examinations was much improved over that provided for previous examination * Although the plant access process for the examiners was unacceptably lengthy on the first day of plant walk-through examinations, the examiners recognized the prompt action taken by the Security, Health Physics, and Operations departments which prevented further access delay *
No violations or deficiencies were identified as a result of the Licensed Operator Training inspection.
4. Summary of Facility Comments and Commitments Made at Exit Interview:
The facility acknowledged the NRC comments noted in Paragraph * The facility personnel thanked the examiners for their cooperation throughout the week and apologized for the access delay . _
- The Training Department felt that the written examination was fair, challenging, and provided evidence of the continuing improvement in NRC licensing exams.
5. Changes Made to Written Exam During Examination Review:
All comments to the written examination were resolved either during the exam review or during the remainder of the week. The examiners returned to the regional office with no unresolved comments. The following list represents the significant changes made to the examinatio Question N Change Reason 6.05 a "No rod selected" keylock GEK reference provide in 0FF is an acceptable alternate answe .06 b "No short term effects" is Full flow would be an acceptable alternate reached in 25 second answe .07 a Graded according to the Question does not candidate's assumption give the status o or interpretation provided the incoming feeder by examiners during exa voltag .10 No rod block occurs; RWM Original answer '
moves to Group incorrec .10 a The original answer key Interpretation provided took the conservative by Hope Creek plant position that the startup managemen procedure steps should not be performed out of se-quenc However, the procedure allows this as long as the intent of the procedure is not violate .10 b It is permissible to Interpretation provided continue with the Startup by Hope Creek plant and increase reactor managemen pressure in order to perform the HPCI ST. If HPCI proves operable, tne RCIC action statement is changed from 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to 14 day Question N Change Reason 8.04 Candidates may consider The "C" ECCS jockey RHR "A" and "C" in answe pump also serves RHR
"A" & "C".
8.08 a RCIC, not HPCI, is The incorrect 250 V inoperabl battery was given on the tes .08 b 3.0.3 is NA because HPCI The incorrect 250 V is operabl TS 3.5.1.d.2 battery was given on applie the test.
Attachment: Written Examination and Answer Key (SRO)
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Attachm:nt 1 U. S. NUCLEAR REGULATORY COMMISSION
]f
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SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _HgPg_gBEEE___ _____,
REACTOR TYPE: _EMB-GE4_ ___ __ _______
DATE ADMINISTERED: _@6/9ZLg7 _ _____
EXAMINER: _EQLQNAUSEIi_6 ._ _ _
APPLICANT: _ _ _ _ _ _ _ _ _ _
_ _ _ _
INSIBUGIIONS_ID_ GEE 6100 nil Use separate paper for the answer Write answers on one side onl Staple question sheet on top of the answer sheet Points for each question are indicated in parentheses after the question. The passing trade requires at least 70% in each category and a final grade of at Icast 80%. Examination papers will be picked up six (6) hours after the examination start % OF CATEGORY % OF APPLICANT'S CATEGORY
__YCLUE_ _19106 ___SGQEE___ _Ye6UE__ ______________g81EGQBy_____________
24.00
,3Ef7U"__ _25tQg ___________ ________ THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS
_2Dt99__ _2Ergg ___________ ________ PLANT SYSTEMS OESIGN, CONTROL, AND INSTRUMENTATION
_3Dt99__ _25t99 ___________ ________ PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL
_25:99__ _25199 ___________ ________ ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 45 9__ 199199 ___________ ________ TOTALS FINAL GRADE _________________%
All work done on this examination is my own. I have neither givcn nor received ai EPPLiCEUYIE~5555diURE~~~~~~~~~~~~~~
_1bE061_0E_NQGLE@B_EQWE6=P(@N1_QBgBOllgNs_ELUlQSg_@NQ PAGE 2
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.IBEBdQDYNAdlGS QUESTION 5.01 (2.50) Define net positive suction head (NPSH). (0.5)
b. Briefly describe the major condition or effect that provides adequate NPSH for the recirculation pumps:
1. during low power saturated condition (0.75)
2. during full power operatio (0.75)
c. For part b, which (b.1 or b.2) provides the greatest value for (0.5)
Recirc pump NPSH?
QUESTION 5.02 (2.50)
c. Concerning control rod worth during a reactor startup with 100% (1.5)
peak Xenon versus a startup with Xenon free conditions, which statement below is correct? JUSTIFY YOUR CHOIC . Peripheral control rod worth will be lower during the 100%
peak Xenon startup than during the Xenon free startu . Central control rod worth will be higher during the 100% peak Xenon startup than during the Xenon free startu . Peripheral control rod worth will be higher during the 100%
peak Xenon startup than during the Xenon free startu . Both central and peripheral control rod worth will be the same regardless of core Xenon concentratio Answer the following questions as TRUE or FALSE, given that the (1.0)
unit is at rated conditions and a Reactor Scram occur . If the reactor in started up at the time of peak Xenon conditions, then the neutron thermal flux level will be located HIGHER in the core than if Xenon free conditions existe . At the time of peak Xenon conditions, the core is Iodine fre (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
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QUESTION 5.03 (3.00)
The reactor is operating at rated conditions when one (1) SRV inadvertently opens and stays open. State HOW the following parameters would change and WHY. (Assume NO scram occurs.)
c. Total INDICATED steam flow (1.0)
b. Turbine steam flow (1.0)
c. Reactor pressure (1.0)
QUESTION 5.04 (2.00)
c. Calculate the reactor cooldown rate for reactor pressure (1.5)
decreasing from 885 psig to 485 psig in one half hour. Show all wor b. Have you exceeded any reactor cooldown limits? Explai (0.5)
QUESTION 5.05 (2.00)
Using Figure 1 (the Power / Flow map) as a reference, answer the followings c. While at 100% power and 100% core flow, insertion of control rods (1.0)
with constant recirc pump speed results in an INCREASE in core flow. Briefly explain wh b. What damage could be expected if operation is allowed below the (0.5)
minimum power line?
c. What design feature prevents and precludes operation below the (0.5)
minimum power line?
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QUESTIDN 5.06 (3.00)
Your Tech Specs and procedures list strict limits for primary water purit c. Give two reasons for maintaining high water purity in the reactor (1.0)
coolan Why is conductivity the parameter which must be monitored contin- (1.0)
uously and upon which limits are placed? Why are chloride and conductivity limits allowed to change as a (1.0)
function of operational mode?
QUESTION 5.07 (2.00)
A periodic core performance edit (P-1) has just been completed by the *
Process Computer. After reviewing the output, you notice a MAPRAT value cqual to 1.00 c. Give the definition of MAPRA (1.0) Have any thermal limits been exceeded? Explain your answer full (1.0)
.
QUESTION 5.08 (2.00)
c. Briefly define subcritical multiplicatio (0.5)
b. A reactor has an initial Keff of 0.9. The count rate is increased (1.5)
by a factor of 20. What is the new Keff?
OUESTION 5.09 (2.00)
c. Explain the difference between the terms " excess reactivity" (1.0)
and " shutdown margin".
b. Explain the difference between the terms " void fraction" and (1.0)
" quality".
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QUESTION 5.10 (2.00)
FCllowing a single control rod withdrawal, the void fraction in the core increases by 1.5*/. and the fuel temperature increases by 30 deg F. Assuming no changes in the moderator temperature, what was the reactivity worth of this control rod? Show all work and ctate any assumptions that you mak DUESTION 5.11 (2.00)
Concerning centrifugal pumps:
c. What is the relationship between the input power to the pump (0.5)
and the mechanical efficiency?
b. What happens to the portion of the input pump power that is NOT (0.5)
converted to pump head (useful power)?
Compare the total amount of input power required for the following configurations to that required for a single pump.
{yypq a. Two identical centrifugal pumps in paralle (0.5)
b. Two identical centrifugal pumps in serie (0,5)
UW k 7/r7IK (***** END OF CATEGORY 05 *****)
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QUESTION 6.01 (2.00)
c. What is the purpose of installing an excess flow check valve in (0.67)
an RPV instrument line? How would indicated RPV level change if a leak developed in the (0.67)
reference leg? Why?
c. Caution 6 of EO 102, " Containment Control" states: (0.67)
"If drywell temperature exceeds 135 deg F use only channels A and B of the wide, narrow and upset RPV water level instruments."
What is the reason behind this caution?
QUESTION 6.02 (1.50)
The plant is operating at 1007. powe APRM channels A and C have failed hig Instrument technicians are investigating while you research Technical Specification A plant auxiliary operator wants to shift RPS B power supply to its alternate power supply for trainin Should you let him? Explain why or why no Direct your answer toward system response instead of administrative requirement QUESTION 6.03 (3.00) Which of the following provides the signal for a Turbine (0.5)
Control Valve (TCV) Fast Closure scram? TCV position limit switches " Rate of TCV position change Power to the TCV fast acting solenoids ETS oil pressure at the TCV List two turbine trips (NOT the scram) initiated from the EHC (2.0)
Hydraulic 011 System. Include parameters sensed. setpoints, and trip logic, if applicabl c. Assuming operating conditions, how will the TCVs fail (OPE CLOSED, or AS IS) for Loss of electrical signal to the servo valve coils (0.25) Loss of hydraulic fluid (FJS) to the control jet pipe (0.25)
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QUESTION 6.04 (2.50)
For EACH of the following conditions, state whether a scram, half-scram, rod bl ock, or no action is generated. For conditions that produce more than one action, state the more severe action (i . e. half-scram to more severe than a rod block). Loss of one RPS MG set Turbine trip at 20% power Main steam lines B and D isolate, Mode switch in RUN APRM B downscale, Mode switch in RUN Scram discharge volume level is at 40 gallons, Mode switch in STARTUP QUESTION 6.05 (2.50) List three (3) ways that the Rod Block Monitor (RBM) may be (1.5)
bypasse How does the RBM utilize the input from a LPRM detector that (1.0)
is failed HIGH or failed LOW 7 DISCUSS BOTH cases, but limit your answer to how the LPRM input is considered in the AVERAGING /
COUNTING circuit Assume the LPRM recently failed and has NOT been bypassed with its individual bypass switch.
QUESTION 6.06 (3.00)
For each of the HPCI (High Pressure Coolant Injection) System component fcilures listed below. STATE WHETHER OR NOT HPCI WILL AUTO INJECT into the reactor vessel, IF IT WILL NOT INJECT WHY, AND IF IT WILL INJEC provide ONE POTENTIAL ADVERSE EFFECT OR CONSEQUENCE of system operation with the failed componen Assume the plant is at full powe A:sume NO operator action and that the component is in the failed condition when HPCI receives the auto start signa c. The HPCI Aux Lube Oil pump fails to operat (1.0) The Minimum Flow Valve fails to auto open (stays nhut) when (1.0)
system conditons require it to be ope The HPCI discharge flow element (NOO7) output signal to the (1.0)
HPCI flow controller is failed at its maximum outpu (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
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QUESTION 6.07 (3.00)
c. The voltage of a 4.16 kV Class 1E bus dropped to 65% of norma (0.5)
What specific action woul d occur immediately?
b. When the Emergency Load Sequencer is activated upon an automatic (0.5)
Emergency Diesel Generator start signal, the PSIS is also activated. What is PSIS, and what is its purpose?
c. Assume a LOCA occurs with a simultaneous Loss of Offsite Power (1.0)
(LDP). After the vital 4160 volt bus power is restored, explain how load sequencing occurs in terms of the LOCA and LOP sequencers, d. If a loss of pressure in the main air system to the Emergency (1.0)
Diesel Generator occurs, would the E D/G still shutdown in response to an automatic shutdown signal? Explain.
QUESTION 6.08 (1.50) .
o. If the IRMs are indicating 20 on Range 4 and the an operator (0.75)
down ranged to Range 3, what trips if any, will occur? Why?
b. With the modo switch in startup, and IRM "C" reading 11 on (0.75)
Range 7, what trip (s) if any, would occur if IRM "C" was down ranged to Range 6? Why?
a (sts** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
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DUESTIDN 6.09 (3.00)
Civen the following EHC system conditions:
Reactor Power =100%
Reactor Pressure =1020 psig EHC Pressure Setpoint =920 psig Load Selector =100%
Load Limit =100%
Recirc Flow Control in Master Manual Cynchronous Speed Selected For the two situations described below Discuss the responses of both the EHC system and the plant (ie. power level, or pressure changes). The "EHC Logic Diagram" and the
" Pressure-Steam Flow Relationship" curve are attached for your us (Figures 2-3) Take your discussion to a point where the plant is in o stable condition. Assume NO operator actio The "4" pressure regulator fails upscale and the max combined (1.5)
flow limiter is set to 105%. The "A" pressure regulator fails upscale and the max combined (1.5)
flow limit is set to 130%.
QUESTION 6.10 (3.00)
c. Using the attached Figure 4 of the RSCS display unit, explain (1.0)
the function that each of the following Rod Pattern Controller <
Command pushbuttons perfor . ALL RODS 3. DYPASS FREE RODS 4. ABOVE LPSP Using the attached pull sheet (Figure 5). state whether or not (1.5)
a rod block is present f or the following situations. If a rod block is present, explain why and give the system that is applying the rod bloc . All rods in RSCS group 3 are at notch 08. Control rod 30-51 (step 110) is then pulled to notch 1 . All rods in RWM group 5 have been pulled to notch 12. The operator then pulls rod 30-27 (step 115) to notch 1 Describe the actions necessary to bypass control rod 14-27 (0.5)
at the Rod Bypass Switch Card (Figure 7, attached). Figure 6 contains the binary control rod coordinate (***** END OF CATEGORY 06 *****)
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QUESTIDN 7.01 (2.00)
DP-IO.ZZ-OO9, " Refueling", contains the following step
"Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to the start of CORE ALTERATIONS requiring control rod withdrawal, verify that the RPS shorting links are removed."
c. Note 5.1.9 states that: " Removal of the RPS shorting links is (1.0)
NOT required if adequate shutdown margin has been demonstrated IAW TS 3.1.1." Explain the reason behind this not b. Which of the following are core alterations as defined by TE7 (1.0)
Answer YES or N . Removal of startup sources 2. LPRM string removal 3. Relocation of fuel to the spent fuel pool QUESTION 7.02 (3.00)
For the following plant parameters, which Emergency Operating Procedures would be utilized? State the entry conditions met for each EO Assume all parameters exist simultaneousl RPV water levels -90 inches Drywell pressure +15 psig Raactor powers 25%
Drywell temperatures 245 deg F Suppression Pool temperature: 145 deg F ,
Raactor pressure: 1050 psig (88888 CATEGORY 07 CONTINUED ON NEXT PAGE *****)
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QUESTION 7.03 (3.00)
Concerning OP-EO.ZZ-102, " Containment Control - Drywell Pressure Control" procedures c. Why are Suppression Chamber Sprays NOT inititated if Suppression (1.0)
Pool pressure is greater than 14.8 psig?
b. Step DW/P-7 of EO 102 refers to the "Drywell Spray Initiation (1.0)
Pressure Limit" curve. Explain why you are allowed to spray the drywell only if you are below this curv c. Caution 23 of step DW/P-8 states: (1.0)
"DO NOT initiate Drywell Sprays if Suppression Pool Level is greater than 142.5 inches."
What is the reason for this caution statement?
NOTE: A copy of the applicable section of EOP 102 is attached as Figure 8 for your raferenc QUESTION 7.04 (3.00)
Concerning the recirculation and blowdown modes of the Reactor Water Cleanup (RWCU) Systems The operator is cautioned to place the RWCU system into blowdown (0.5)
mode prior to starting the CRD pumps. What is the reason for this caution? When operating in the blowdown mode why shouldn't you divert all (1.0)
the RWCU flow to Liquid Rad Waste or the Main Condenser? Under what operational conditions would the recirculation mode (1.5)
of the RWCU system most likely be used and why?
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QUESTIDN 7.05 (2.00)
The plant is operating at 100% power when the following alarms come in:
MN STM LINE RADIATION HI RADIATION MONITORING ALARM /TRBL The Chemistry department also notifies you about increasing reactor -
coolant conductivit c. As the SRO, what are your immediate actions? (Two required.) (1.0)
b. In the Hope Creek Tech Specs for Reactor Coolant Activity (1.0)
(3/4.4.5, attached), the LCD states that:
" The specific activity of the primary coolant shall be limited to less than or equal to 0.2 microcuries per gram dose equivalent I-131..."
Why does action statement 3.4.5.a.1 allow: " greater than microcuries... but less than 4.0 microcuries. . . for more than
. 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />..."7 OUESTION 7.06 (2.50)
A startup is in progress at Hope Creek. Reactor pressure is at 500 psig when BOTH CRD Hydraulic pumps are lost. Maintenance expects the repairs to take at least four hour a. What two problems will result within the CRD sytem if drive (1.0)
water is not restored?
b. What are your immediate actions in the situation described above (0.5)
according to OP-AD.ZZ-105, " Loss of CRD Regulating Function"7 (One required.)
c. List two possible causes of high CRD hydraulic unit temperature (1.0)
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QUESTION 7.07 (2.00)
c. Explain the difference between an RWP and an ERW (1.0)
b. The following statements pertain to Radiation Work Permit (1.0)
For each, state whether the sentence is TRUE or FALS If the sentence is false, be sure to explain wh . The SNSS may substitute continuous coverage by qualified Radiation Protection personnel in lieu of an RW . ERWPs may be issued for special maintenance work in airborne radioactivity area QUESTION 7.08 (2.00) Caution 5.3.9 of OP-SO.AC-OO1, " Main Turbine Operatica", (1.0)
discusses turbine critical speeds. Why is turbine operation at or close to a critical speed undesirable 7 b. When changing main turbine load, the time limits as given in (1.0)
the " Main Turbine Load Change Curve" (attached, Figure 9) must be observed. If load on the turbine changed from 100% to 22%
of rated within 10 minutes, would the time limit be violated?
Explain your answe QUESTION 7.09 (2.50)
Use the attached copy of OP-EO.ZZ-202 . " Emergency Depressurization",
(Figure 10) to answer the following question c. Briefly explain the reason for Caution 13. (attached to steps (1.25)
ED-06 and ED-11) Give the two bases behind step ED-4 of this procedur (1.25)
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QUESTION 7.10 (3.00)
A Hope Creek startup is in progress. OP-IO.ZZ-OO3, "Startup from Cold Shutdown to Rated Power" is in us Reactor pressure is 160 psi As the SNSS, you are informed that RCIC failed to produce >600 gpm in the test flow path during its operability surveillance tes The NSS has declared RCIC inoperable and has generated the proper paperwork for I&C and Maintenance to begin investigation and repai The NSS is also requesting your permission to continue with the ctartup in order to achieve 200 psig of reactor pressure so that the HPCI operability test may be perfomed in accordance with 10-003 ctep 5.3.14. His reasoning is that he has 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to fix RCIC and perform its operability test, OR once he has HPCI proven operable, he has 14 days (if needed) to repair the RCIC system. He also claims that the startup procedure does not require performance of procedure steps in order, as long as they are performed in a timely manne c. According to procedure OP-IO.ZZ-OO3, can the NSS bypass step (1.0)
5.3.12 (perform RCIC ST) and continue with the startup by going directly to step 5.3.14-(perform HPCI ST)?
b. Using this procedure along with the attached Tech Specs, decide (2.0)
whether or not you agree witn the recommendation of the NS )
Justify your answer, referencing all TS that you us (***** END OF CATEGORY 07 *****)
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QUESTIDN 8.01 (2.00)
List the three conditions which must be met in order to make on-the-spot changes to written procedures according to Tech Spec QUESTION 8.02 (2.00)
( 00 -o t00 )
It is discovered today, July 7 day shift (08-1600), phat a monthly curveillance item due on Monday, June 30 mid shift (kd-OBOO) was not performed. This item has been performed on time for the past six months. Has the specified time interval for this surveillance item been violated? (Yes or No) EXPLAIN your answe QUESTION 8.03 (3.00)
a. What is the purpose for establishing Safety Limits as found in (1.0)
the Hope Creek Tech Specs?
b. List the Hope Creek Safety limits as given in Section 2 of the (2.0)
Hope Creek Tech Spec "
kK M mboAd*- h'C}ockt p gryg DUESTION 8.04 (3.00) Cp M yfg g (a a. While in RUN, ECCS jockey pump "C" is declared inoperabl (1.5)
Repairs are expected to take approximately eight hours. Using the attached Tech Specs, evaluate the status of the system (s)
served by this jockey pump. Reference any Tech Specs you use to develop your answe b. How would your answer for "a." change if, in. addition to the (1.5)
jockey pump "C" failure, Core Spray valve HV-F015B (Loop B test return valve) f ails in the OPEN position?
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QUESTIDN 8.05 (2.50)
You are in the control room as the SNSS on the backshiftp power to at 100%. A small leak occurs in the common header steam piping ct the point it connects to the "D" sain steam lin The reactor is scrammed and all MSIV control switches are placed in CLOSE. At this point you determine that the "A" inboard and outboard MSIV's are STUCK ope a. Refer to the Emergency Classification Guide (ECG) chart attached (1.5)
(Figure 11) and determine the EAL, if any. Justify your choice, Answer TRUE or FALSE: You can upgrade the event as warranted (0.5)
but NRC concurrence is required to downgrade the even c. Answer TRUE or FALSE: .If an ALERT has been declared, the (0.5)
Emergency Coordinator may terminate the event only if none of the action levels defined in the ECG are applicable AND the plant is in a recovery mod QUESTION 8.06 (2.50)
The reactor was manually scrammed due to Suppression Chamber Water temperature >110 deg F in accordance with Tech Spec The reactor is now in hot shutdown, suppression pool cooling is on, and suppression chamber wager temperature is 97 deg Would it be a violation of Tech Specs to startup the reactor?
(i . e. enter Op Con 2)
OUESTION B.07 (2.00)
According to 10 CFR 50, each of the following events have a ONE HOUR rcporting requirement. (TRUE or FALSE)
c. The plant is in a condition NOT covered by operating and (0.5)
emergency procedure The loss of the offsite notification syste (0.5) A valid automatic initiation of the Reactor Protection Syste (0.5) A shutdown was commenced because the plant was in violation (0.5)
of the Technical Specification (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
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.
DUESTION 8.08 (2.50) /- Gn 0Ard 100A1\
While at power, 125 VDC charger 1AD413 and 250 V battery 6E946E become inoperabl c. Using the attached Tech Specs, discuss ALL applicable action (1.0)
statement b. Discuss all applicable action statements, if, in addition to (1.5)
the above malfunctions, the ADS Channel / was inop due to a fault on DC bus 1BD41 E
[AAArts3 ndy & AD fdW M ce'Nj t.gutyW+vd (03k TYDYn buJ Mk . )
DUESTION 8.09 (3.00)
1D10f CON L ...
Station service water (SSW) pump "A" has been out of service (1.0)
a.d f or two days. Maintenance continues, and the pump is expected to be returned to service in four day A sizeable pipe break occurs in the "B" and "D" SSW pump roo removing both pumps from service. Repairs on these pumps are expected to take a full wee Using the attached Tech Specs, determine what action (s) you will take and reference all TS that you use to develop your answe What would your actions be for the above scenario if instead (2.0)
of Op Con 1, the plant was in a Refuel? Assume heat losses are sufficient to maintain operational condition Reference all Tech Specs you use to determine your answe QUESTION 8.10 (2.50)
As the Emergency Coordinator, you have just declared a General Emergency at Hope Creek due to loss of the fuel clad boundary along with significant fission product releases to the Primary Containment. There is potential for a loss of Primary Containment as wel Using the Recommended Protective Actions Worksheets attached (Figures 12 and 13), designate the areas that require shelter or evacuation BY MARKING THE ATTACHED WORKSHEET WITH YOUR ANSWE (***** END OF CATEGORY OB *****)
(************* END OF EXAMINATION ***************)
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..
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ANSWERS -- HOPE CREEK -86/07/07-KOLONAUSKI, .. ,
--
t _
_
ANSWER 5.01 (2.50)
'
c. NPSH is the difference between the suction pressure and the (0.5)
saturation pressure of the fluid being pumpe . The recirc pumps are located 57 f eet below the normal water (0.75)
level; this difference in elevation provides adequate NPSH during low power saturated conditions. Expt _t- also - outyyW.fS4A Mtnte yq rpttd (0.75)
2. At full power, feedwater flow provides subcooling to the recirc pump suctio qy go . ogypg% NPSH is greater at full powe (b.2) (0.5)
REFERENCE a. HC Fluids LP FFOS, Instructional Objective 2 b,c. HC Reactor Recirc LP 19. pages 74-75 ANSWER 5.02 (2.50) "3" is the correct answer (0.5). The highest Xenon concentration will be in the center of the core (0.25), which is the high flux region from the-previous operating period (0.25). This will increase the flux in the area of the peripheral control rods (0.25) to increase their worth (O.25). . TRUE - - or FALSE , if et iop paktd docialRwx thM(24 (
'. FALSE g gg4 { .
REFERENCE Xenon LP, RXPH 33, pages 8-10
5t__1SE98Y_9E_NQC(g88,EQWE8_EL6N1_QEg8811gNt_E(ylDh3_6ND PAGE 19
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- ANSWERS -- HOPE CREEK -86/07/07-KOLONAUSKI, .
Olib dAA4. M (Ocatim of- df cL ANSWER 5.03 (3.00) daho.( ggy, c. As reactor pressure decreases from the SRV opening (0.5), total indicated steam flow will decrease (0.5). As pressure decreases (0.5), the EHC will shut down the TCVs and turbine steam flow will decrease (0.5). Reactor pressure will initially decrease (0.5); it will then stabilize at approximately the same value when the EHC shuts down the TCVs (0.5).
REFERENCE HC LP 106 Transient Analysis, pages 12-13 Lcarning Objective ANSWER 5.04 (2.00) . Convert pressure to psia: (0.25)
885 + 14.7 = 900 psia and 485 + 14.7 = 500 psia Obtain corresponding temperatures from steam tables: (1.00)
900 psia -> 532 F and 500 psia -> 467 F Determine temperature change: (0.25)
532-467 = 65 F in one half hour NO, the TS cooldown limit of 100 F/hr has not been exceeded, (0.50)
but it would be if this rate continued for 1 hou REFERENCE
)
argt exylQW Saturated Steam Tables, HC TS 3.4.6. d coot cLo w ( g ;9,
_ _ _ _ _
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.
- ANSWERS -- HOPE CREEK -86/07/07-KOLONAUSKI, ANSWER 5.05 (2.00)
c. The reduction in reactor power will reduce the void content in (1.0)
the core. Core flow increases because of this reduction in two phase flow resistanc Recirculation and Jet pump cavitatio (0.5) % Feedwater Flow interlock enforces the minimum power lin (0.5)
REFERENCE HC LP Reactor Recirc Control, LP 20-02, pages 51, 52 Lsarning Objective 14 ANSWER 5.06 (3.00) . Minimize deposition on f uel -> less heat transfer Minimize corrosion Minimize " crud" (activated corrosion products) buildup ->
reduce radiation hazard . Minimize radiation levels from carryover (2 at 0.5 each) Changes in conductivity indicate abnornmal conditions. If conduc- (1.0)
is within limits, then pH, chlorides, and other impurities must also be within limit o.33 Stress corrosion cracking requ[res high temperatures and high (1.0)
oxygen. Chloride and conductivity limits are less restrictiv at power because of the reduced oxygen concentration. Oxygen 0 33 (being a gas) is continuously leaving the RPV with the steam Cand is removed from the condenser by the SJAEs].
REFERENCE 0.33 TS Bases 3/4.4.4 Chemistry TUEQBY_QE_NgG6g@b_EQWgb_E6@d1_Qthb611gus_tLQlygg_ddQ Fabt mi
.
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- ANSWERS -- HOPE CREEK -86/07/07-KOLONAUSKI, ANSWER 5.07 (2.00)
c. MAPRAT = MAPLHGR actual /MAPLHGR-LCO (1.0)
MAPRAT is the Maximum Average Planar Ratio. It is the comparison between the actual APLHGR to the MAPLHGR limit as programmed into the Process Compute . 5 Yes 49,964 , we have exceeded the Tech Spec limit for APLHGR (0.5).
The value of MAPRAT should never be greater than 1.0 H5MP9).
REFERENCE TS 3/4.2.1, Average Planar Linear Heat Generation Rate HC Heat Transfer LP 16, pages 16-3 to 16-5 Process Computer LP 109, Figure 11, page 21, LO 4 ANSWER 5.08 (2.00) Subtritical multiplication is the process by which the source (0.5)
neutron level is increased as a result of the presence of fuel in a reactor with a Keff less than CR1 = CR2 if Keff < 1.0, CR = So/1-Keff (1.5)
Substitute:
20 (So/1-0.9) = So/1-Keff2
Keff2 = 0.995 REFERENCE Hope Creek Reactor Theory Section 21, Subcritical Multiplication Section 22, Count Rate Comparison
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.
ANSWERS -- HOPE CREEK -86/07/07-KOLONAUSKI, ANSWER 5.09 (2.00)
!
o. " Excess reactivity" is the reactivity associated with the excess fuel l that is added to the core beyond the minimum amount necessary for !
'
criticality. (This is done because Keff decreases over core life.) (0.5)
" Shutdown margin" is the amount of reactivity that the reactor is or could be made subcritical from its present positio (O.5)
VoluH.,
b. " Void fraction" is the ratio of the name of steam in a mixture to (0.5)
the tctal h 5 cf the mixtur " Duality" is the ratio of the veEUNd of steam in a mixture to the (0.5)
total velurr of the mixtur twMs REFERENCE Hope Creek Reactor Theory LP 19, page 4 H3at Transfer LP, page 10-4 ANSWER 5.10 (2.00)
Reactivity due to void change = (-1 x 10E-3 4k/k / % voids) (1.5%) (0.75)
= -1.5 x 10E-3 A k/k Reactivity due to fuel temp = (-1 x 10E-5 ak/k / deg F) (30 deg F) (0.75)
= -0.3 x 10E-3 ok/k Total worth = -1.8 x 10E '2 Ak/k (0.5)
REFERENCE Hope Creek Reactor Theory LP pages 26-4, 26-6
_ _ . _ _ _
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- ANSWERS -- HOPE CREEK -86/07/07-KOLONAUSKI, ANSWER 5.11 (2.00)
c. The mechanical efficiency of the pump is the ratio of the useful (0.5)
pump' power to the input pump powe b. Due to friction losses in the pump, not all input power is con- (0.5)
verted to useful power. The difference is converted into heat which is carried away with the fluid passing through the pum . 1. Total power required is less than twice the power required for (0.5)
a single pump.
MtNk 2. pump.Total power required is twice the power required for a single (O.5)
REFERENCE Hope Creek Fluid Flow LP, pages 6-7, 6-8 f
t
%
. . . _
_ _ _ _ _ _
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- ANSWERS -- HOPE CREEK -86/07/07-KOLONAUSKI, O ANSWER 6.01 (2.00)
c. The excess flow check valve is installed in an RPV instrument (0.67)
line to isolate the line if a break occurs downstream of the valv b. The leak would cause a decrease in reference leg density (and (O.67)
level if large enough). Indicated level would d;;.;;;;.
thCttas c. An increase in drywell temperature will heat up the water in the (0.67)
sensing lines, possibly causing erroneously high readings. These effects are minimized because the variable and reference legs have equivalent elevation drop REFERENCE HC Systems LP 2-01, pages 13,49,50 Instructional Objective 3a (LP 2-01)
Instructional Objective 1.5 (EOP 102 LP)
ANSWER 6.02 (1.50)
No (0.5). When transferring RPS power supplies, the RPS is momemtarily deenergized.This could result in a scram due to the half scram already present (1.0).
REFERENCE HC RPS LP O22-02, Insrtuctional Objective 9, page 54 ANSWER 6.03 (3.00) . or ETS oil pressure at the TCV (0.5) . FAS low oil pressure ' OI EltC Y M skIIChd] F"'I (0.5)
1100 psig OS M )
' --._ 2.__, ~'
- zt r' T ';,__ T ; -- t-ip (0.25) ETS low oil pressure- (0.5)
h .fqo,2;3
< 800 psig kglgg() r - - - - - . . - - -. .u u-- ,-- s a .- zu- 1 -- s ns,,a,,n ,g,3 ; . shut (0.25) shut (0.25)
62__Eb6bl_ ele 1EDE_9hM19Bi_YublbWLx_dbk_16MlbYDEbl61196 FH6L b ANSWERS -- HOPE CREEK -86/07/07-KDLONAUSKI, REFERENCE HC EHC LP 50-02, Instructional Objectives 13, 15, pages 17, 30, 31 ANSWER 6.04 (2.50) half-scram (0.5 each) no action half scram M A M JCrWYw mg occuh.di).Lh high rod block V hip yntiv)Ls. - ok (f- exylA( scram REFERENCE RPS LP O22-02, Instructional Objective 4, pages 10, 42, 44, 46 RBM LP 017-01, Instructional Objective 2 ANSWER 6.05 (2.50)
e. 1. Manual operation of the RBM BYPASS switch (0.5)
2. Reference APRM <30% (0.5)
Alto 4. "No ved ideded" kurtock J in off .
3. Edge rod selected (0.5) Failed Low: Removes the LPRM input from the xbb averaging circuit (0.25)
a n d e r= h 2 ; g,. J m counting circui ts (0. 25) .
Failed High: The higher input is averaged with the other inputs and processed as if it were a valid signal (0.5)
REFERENCE HC LP 017-01, RBM Instructional Objective 4, pages 9-14 ANSWER 6.06 (3.00) Will not inject (0.25). Turbine stop and control valves will not open (0.75). eHectr beccuvx ful.L flow ( Atso - No ther4 iemtWN bL HAckcL in 25 pc6 ) Will inject (0.25). Pump overheating and seal damage may result g during 1ow or no fIow. conditions (O.75). c,nect Atyd r . Will not inject (0.25). Maximum signal from the flow element will cause the turbine speed to remain at minimum (0.75).
l REFERENCE F 5C" 5Yd HC HPCI LP 26-02. Instructional Objective 3, pages 22-34
_ .__-. - _
0 __E66dl_h'SIEDE_DEEl@Nz_CONIB062_6dD_lN@lbyDEN1811gN I PH6E ao ANSWERS -- HOPE CREEK -86/07/07-KDLONAUSKI, L. (F inceig fuhr voHay < n? , L0f Jip - Loah rkd
- 2. IF wo M incami$ volby - farf- XFrt We land rM ANSWE 6.07 (3.00)
. All under voltage relays would trip. The motor circuit breakers (0.5)
on the associated bus would trip. (loads shed) PSIS refers to Process Start Inhibit Signals (0.25). This prevents automatic start of selected loads until the LDP or LOCA sequencer timer cycles (0. 25 ) ,,
c..When bus power is restored, the LOCA signal takes precedence (1.0)
and sequencing commences as in a LOCA only. ( Lo f A q v vv c v (d ll b 0N d % * * *O *" 'UkI'P"Afh*"'"AI' b "p*/
~
The E D/G can still be shutdown. A check valve retains a (1.0)
supply of air from the main system in a small tank located on the skid. When the shutdown solenoid valve is energized, air from the tank is directed to the servo fuel rack shutdown cylinder.
0 *qg ,, This overpowers the governor spring link and drives the injection
' pumps racks to the zero fuel positio REFERENCE 1E AC Distribution LP 066-01, Instructional Objectives 1.5, 1,7 Pages 40-43 E D/G LP 068-01, Instructional Objectives 3e, 9 Pages 96, 101, 102 ANSWER 6.08 (1.50) None (0.25). The IRMs would indicate 20 on Range 3; O-40 scale (0.5). An IRM upscale trip (rod block) would occur (0.25), because the IRM would read 110 on the 0-125 scale (0.5).
REFERENCE HC.IRM LP 014-01, Instructional Objectives 4a, 4b, 5 Pages 9-13
6 t__PLON1_@y@lEDS_QESIGNz_CQNlbgLs_QNQ_lN@lhyMEN1811QN FHbt ./
ANSWERS -- HOPE CREEK -86/07/07-KOLONAUSKI, .
s ANSWER 6.09 (3.00)
c'. The CVs will remain open at 100%; the BPV will receive a 5% (0.5)
signal. This will cause reactor pressure (and therefore power)
to decreas With the Max Combined flow limiter set down, a slower rate of RPV (1.0)
depressurization will occur. In this case, the main steam line pressure will drop and result in an NSSSS low pressure isolation and subsequent reactor scra The CVs will remain open at 100%; the BPVs will receive a (0.5)
30% signal. With all bypass valves full open, reactor pressure (and power) will decrease at a greater rate than in "a" abov oi F W 8
) jip LWith the Max Combined flow limit set to 130%, a higher rate of (1.0)
bL '
RPV depressurization will result. The RPV water level will swell I
g C to the Main Turbine and Rx Feed Pump'setpoint. A scram will result
,due to the turbine trip at this power leve REFERENCE gh HC EHC Control Logic LP 051-01, Instructional Objective 10
- p -
ANSWER 6.10 (3.00) . The amber lights display all of the rods which belong to the (1.0)
group determined by the selected ro . The amber lights display all of the rods which are permitted to move in the direction selected by the DIRECTION switc . BYPASS identifies all of the rods which have been bypasse . When illuminated, it indicates that the Law Power Setpoint (of 20%) has been exceede . Yes (0.25), RSCS enforces a rod block at notch 12 (0.5).
Yet (0.25), RWM will - '---- ' -ad hin"' ' ~~' " '4 (0.5). '
M W Vto 'tl of ) W 6 . Move toggle switches XO, X2, Y3 to ON, all other switches are (0.5)
at OFF. Move the toggle switch at the top of the card to O \ s REFERENCE 0.25 a.25 HC RSCS LP, Instructional Objectives 7, 9, 14, pages 10-14,21 HC Pull Sheet LP 008-01 HC RWM LP
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- ANSWERS -- HOPE CREEK -86/07/07-KOLONAUSKI, ANSWER 7.01 (2.00)
a. Removal of the RPS shorting links will causethe[SRMHi H( (1.0)
scram to be active. If adequate SDM is demonstrated, then there is no need for thislNMS % cram and the shorting links need not be remove b. Yes (0.33 each) Yes Yes REFERENCE HC SRM LP 013-01, page 28 OP-IO.ZZ-OO9, Refueling, page 6 ANSWER 7.02 (3.00)
bMIM 100 pr6 "
RPV water level < -38" YH LL d1 Lite 3 DW pressure > 1.68 psig (0.3)
Reactor power > 4% (0.3)
Reactor pressure > 1037 psig (0.3)
Containment Centrol, EO 102: (0.45)
DW pressure > 1.68 psig (0.3)
DW temperature > 135 deg F (0.3)
Suppression Pool temperature > deg F (0.3)
REFERENCE EO 101 LP, Instructional Objective 1.2, pages 5,6 EO 102 LP, Instructional Objective 1.3, page 5
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.
- ANSWERS -- HOPE CREEK -86/07/07-KOLONAUSKI, "chMT mq net k MM wiu yM cHg if d ka.it $7, 4 ANSWER 7.O3 (3.00) % dws. t N ejp, ,, g,, .tnwuy& -k
% rit o. Above 14.8 psig,'" chugging" (oscillating steam condensation) may (1.0)
-occur at the downcomer vent outlets. This cyclic stessing of the Cl6 vents could lead to f ailure and loss of the pressure suppression capabilit l 05 DW spray initiation above the limit may result in a containment (1.0)
depressurization rate which exceeds the capacity of the Suppression Chamber to Drywell and Reactor Building to Suppression Chamber vacuum breakers. This may lead to exceeding the design negative pressure of the containmen The Suppression Chamber to Drywell Vacuum Breakers cannot be (1.0)
relied on to relieve drywell negative pressure if they are submerge REFERENCE DW/P LP, Instructional Objectives 1.4 and Pages 8-12 ANSWER 7.04 (3.00) 1R.{W kW1CMl The CRD pump will increase water level (0.25) and there is no (0.5)
outlet flow path established. (0.25) Because cooling flow is lost to the regenerative heat exchanger (1.0)
(0.25) increasing the outlet temperature to the NRHX (0.25),
possibly causing an isolation of the system (0.5) Hot shutdown (0.5) with no recirculation pumps operating (0.5) (1.5)
to minimize thermal stratification of vessel water (0.5).
REFERENCE Prep for Plant Startup (ZZ-OO2) page 5 RWCU LP (21-01) pages 56,57
l
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- ANSWERS -- HOPE CREEK- -86/07/07-KOLONAUSKI, ANSWER 7.05 (2.00)
c. Drder power reduction as necessary to maintain activity levels (0.5)
on the main steam line radiation monitors less than the alarm setpoin . If a scram condition is reached, ensure the scram and implement (0.5)
OP-EO.ZZ-10 b. This action statement accommodates possible Iodine spiking which (1.0)
may occur following changes in reactor powe ( or yrCrf40tt)
REFERENCE OP-AB.ZZ-100 High Reactor Coolant Conductivity, Sections 3.0 and HC Module AB-100, Primary Objective 3 HC TS 3/4.4.5, Bases 3/4. ANSWER 7.06 (2.50) . Loss of cooling water to the CRD HCUs (0.5) The HCU Accumulators will depressuriz ( ie. l4ff of ICWW" (0.5)
cLyAbild) )
b. Manually scram the reacto (0.5)
c. Low cooling water flow (0.5 each, 2 required)
Defective thermocouple circuit Plugged CRD cooling water orafices Leaking Scram Discharge Volume REFERENCE OP-AB.ZZ-105 ANSWER 7.07 (2.00)
a. ERWPs are issued for routine, low exposure tasks of a repetitive (1.0)
nature. Radiological conditions in these work areas are not likely to change abruptly. The ERWP is reviewed periodically and reissued every six month RWPs are issued for specific tasks and time period . TRUE (0.5) FALSE. ERWPs are issued for routine activities only. In (0.5)
areas with high contamination, high radiation, or airborne radioactivity, an RWP would be issue ._____-
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- ANSWERS -- HOPE CREEK -86/07/07-KOLONAUSKI. REFERENCE SA-AP.ZZ-046, Rad. Access Control Program ANSWER 7.08 (2.00)
c. Because the turbine rotor is slightly unbalanced, as it passes (1.0)
through its critical speeds, large vibration amplitudes resul Yes (0.25). According to the curve, a change of this magnitude should take approximately 20 minutes (0.75).
w
'
REFERENCE HC LP O48-02, Instructional Objective 1.4, pages 13-14 OP-SO.AC-OO1 ANSWER 7.09 (2.50) M d.1 a.rtWL'ncA1.r oi Cori doe Q, The 100 deg F/hr limit is a TS LCO, but EOP actions take (1.25)
precedence over TS LCO . Injection sources are terminated in order to preclude power (0.625)
excursions and/or core damag . CRD and baron injection are not terminated because of their (0.625)
use to shutdown the reacto REFERENCE y05f th 3- bM M borrn M HC EOP 202, LP Instructional Objective 1.4, page 8 bE' b*"' '. E4k Oh (6 & K.FV ,
_
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-
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- ANSWERS -- HOPE CREEK -86/07/07-KOLONAUSKI, L.
ANSWER 7.10 (3.00) dk o. Admin step 3.1.1 does allow the NSS to perform steps out'of (1 sequence as long as they are perf ormed in a timely manner. -Gad ihM .0)
~
'Since RCIC failed its ST, however, continuing with the startup 00 P'8
. would not be conservative and should not be allowe * According to TS 3/4.7.4 RCIC must be operable at >150 psi (2.0)
With RCIC inop, HPCI must be proven operable or be <150 psig in 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> HPCI cannot be proven operable until >200 psig according to
'
TS 4.5.1. You should NOT allow the NSS to continue with the startup.],, q @
Maintain reactor pressure <150 psig until RCIC has been .
'
returned to servic Note- The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> specification applied to performing the ST does not apply (TS 4.0.4). This applies to changing Op Con while relying on performance of an ST. The operational condition of the plant is not changin as itgitd (n 7.10 (nkrpvtM yvnrid.ed REFERENCE bM OP-IO.ZZ-OO3, Admin Precautions and Limitations 3.1.1, Steps 5.3.12, 5.3.14 TS 3/4.7.4, 3.4.5.1 RCIC and HPCI cper. re TS 4.7.4.c.2- RCIC ST re TS 4.5.1.c.2- HPCI ST re Aikvwakt psuit r u stoit.d a S La Uvuma te.Wr- in D. Lg 7Ittli6 -
. wi4h WCl mt p hon:Md -tv k oyed - rc.ctc ac% sh d l'1 hau.n wowtd. oyp() . ,- Nova chamy of OPCm
.
lt (i not a. violaiton of tiG Tr b incytox Kx yrttsuN. to i b d W tWct rr cuwbe perford(.u)(%(n A il brr),
W Wc t prtw to kat operauts. (w hi.s 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> pen'ed , & tecic kchan th%wt is vtlo xed b i1 howra in 14 cL9,
_
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' ANSWERS -- HOPE CREEK -86/07/07-KOLONAUSKI,'L.
o ANSWER 8.01 (2.00)
c. The intent of the original procedure is not altere (0.67)
b. The change is approved by two members of the unit management (0.67)
staff, one of which holds an SRO license on the uni c. The change is documented and receives the same level of review (0.67)
and approval as the original procedure within 14 days of implemen-tatio PEFERENCE Hope Creek TS 6.8.3 ANSWER 8.02 (2.00)
-
No, (0.5) A maximum allowable extension is not to exceed 25% of the ,
eurveillance interval or 7.75 days (0.75) AND a total maximum combined' lT interval for any three consecutive tests not to exceed 3.25 times the j 4 , ql gpecified surveillance interva (0.75).
REFERENCE HC Tech Spec 4.0.2 Surveillance Requirements pg 3/4 0-2 ANSWER 8.03 (3.00) Safety limits are established to protect the integrity of the (1.0)
barriers (f uel cladding, RPV, primary system piping) to the release of radioactive materials to the environs during normal plant operations and anticipated transient . Thermal power shall not exceed 25% of rated thermal power (0.5)
with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flo . The MCPR shall not be less than 1.06 with the reactor vessel (0.5)
steam dome pressure greater than 785 psig and core flow greater than 10% of rate . The reactor vessel coolant system pressure as measured in (0.5)
the reactor vessel steam dome, shall not exceed 1325 psi . The reactor vessel water level shall be above the top of the (0.5)
active irradiated fue Gr__6EdiblElb611ME_Eb9EEM9bEEz_E9bMll19bEz_dbk_610llel190e FHdt 4
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ANSWERS -- HOPE CREEK -86/07/07-KOLONAUSKI, L.
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[)9 REFERENCE Hope Creek Tech Specs, Section 2 0 ICC Y M &lf0 MD A ( C M in atte y t 3. C. 2. 3. (fy Fool C)
3. C . 2. 2.. a. (Ay Ysel %)
ANSWER 8.04 (3.00)
G c. The "C" .)ockey pump serves Core Spray Headers A and C. The "A" (1.5)
Core Spray loop is still operable, as long as it is proven that the discharge piping is filled with water in accordance with TS 4.5.1. Because the test return valve is NC. the CS loop "B" is now (1.5)
inoperable because the valve is not in its correct positio Again, if the discharge piping of loop "A" remains full, it is still operable. TS 3.5.1.a.1 applies. (
0 it in 'a" c s CmuM6 Oy = NM 3 7 I 6 REFERENCE ayyg(p Hope Creek Tech Specs b y jn g o gy cmf(dtytc( ,
3. S. I. a . L o.yyh,o b M RHrt ccmn d1rtd y.o . 3 a.y y((v ANSWER 8.05 (2.50) Site Area Emergency (1.0). Steam line break outside the drywell with continuing leakage (0.5).
b FALSE (O 5) (
c FALSE (0.5) W~
fr( E b t' b r p g-fibald1 CIA N Ce REFERENCE HC Emergency Classification Guide ANSWER 8.06 (2.50)
Yes (0.5), suppression pool water. temperature must be less than 95 deg F in order to enter Op Con 2 by TS 3.6.2.1.a.2 (1.0). TS 3.0.4 does not allow entrance into an Op Con while relying on an action statement (1.0).
REFERENCE Hope Creek Tech Specs
dz__6DdlN1 SIB 611ME_EbugEQUEhes_ggBQlllydes_60Q_61011ellybg esoE ;s
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ANSWERS -- HOPE CREEK -86/07/07-KOLONAUSKI, O ANSWER B.07 (2.00)
c. True (0.5 each)
b. True c . Fa l se --- d. True
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REFERENCE 10 CFR 50.72 ANSWER 8.08 (2.50)
ECl( 3 .7. i lT c. 3.8.2.1.b -> HPG4 inoperable. Take 3. i- !" d2y (3.8.2.1.adoesnotapply.)
b. 5.5.2. nFCI OS ( f pm 3.8.3. c two hours to restore or SD in 12. 0 ' c; Because SPCT and ADS inop, 3. FCL C i REFERENCE 3 6 .(. d 2. g Hope Creek Tech Specs -
ANSWER 8.09 (3.00)
e.(None of the action statements listed in TS 3.7.1.2.a apply) (G.3)
TS 3.0.3 must therefore be applied. The unit must be placed 40r59 1. 0 in 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 />, in 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 in at least COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, According to TS 3.7.1.2.g, the associated SACS system must be (1.0)
declared inoperable and the action required by TS 3.7.1.1 must be take The associated safety related equipment must be declared (1.0)
inoperable as required by 3.7.1.1./.
REFERENCE 3 , 7. l . l . d - M Hope Creek Tech Specs -o2I3W 1 ANSWER O.10 (2.50)
See attached RPA workshee .- . . . -- . ..- _=_,. .. . ..... . _- - _-.. . . -. . - . . - .-. . . ~ . . .. . - . _. . _ _ .
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ANSWERS -- HOPE CREEK -86/07/07-KOLONAUSKI, '
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i REFERENCE
, HC ECG, Attachment 4 i
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RFIDMfNDED PPOTFLTIVE ACTIOJS TrDRKSilEET
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l10 Hill IU7 C litlE NH T
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~6' ttE REODRDER WIND FROM COWASS DIRECTION ETOM T
. gj HilH 348 3/4' to 11 1/4* N 11 1/4* to 33 3/4* NNE
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Et4E 33 3/4' to 561/4'
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56 1/4' to 78 3/4' ENE ~
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78 3/4' to 101 1/4' E H /
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101 1/4* to 123 3/4'
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d E 123 3/4' to 1461/4* SE 146 1/4* to 168 3/4'
168 3/4* to 191 1/4'
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191 1/4* to 213 3/4* SSW ,
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213 3/4' to 2361/4* SW HOW 236 1/4* to 258 3/4' Wsw M tf ESE 258 3/4' to 281 1/4* W 281 1/4* to 303 3/4' WJW 303 3/4* to 3261/4* NW - -- --
326 1/4* to 348 3/4* NNW GW - -
GE RFIDMM'NDED PR7IECTIVE ACTIOJS M)RKSillTP Designate areas and/or sectors with recomended protective actions GON USE E symbols as follcws:
S - Take shelter E - Evacuate mpm O - Other (specify)
v . Time Wind Direction lAlhN Wind Speed 30 vvvph
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< Ccmpleted by:
. Reviewed by: C SFPT/RAC/RSM Emer. Coord./SNSS/EIX)/EDM
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A, atoi enuad nLeEntWLL rset A QU STION VALUE REFERENCE
__E_____ ______ __________
05.01 2.50 BANOOOO279 05.02 2.50 BANOOOO281 05.03 3.00 BANOOOO283 05.04 2.00 BANOOOO284 05.05 2.00 BANOOOO286 05.06 3.00 BANOOOO289 05.07 2.00 BANOOOO292 05.08 2.00 BANOOOO293 05.09 2.00 BANOOOO294 05.10 2.00 BANOOOO295 05.11 2.00 BANOOOO296
______
25.00 06.01 2.00 BANOOOO297 06.02 1.50 BANOOOO298 06.03 3.00 BANOOOO301 06.04 2.50 BANOOOO303 06.05 2.50 BANOOOO312 06.06 3.00 BANOOOO313 06.07 3.00 BANOOOO316 06.08 1.50 BANOOOO317 06.09 3.00 BANOOOO331 06.10 3.00 BANOOOO342
______
25.00 07.01 2.00 BANOOOO304 07.02 3.00 BANOOOO310 07.03 3.00 BANOOOO311 07.04 3.00 BANOOOO320 07.05 2.00 BANOOOO325 07.06 2.50 BANOOOO327 07.07 2.00 BANOOOO328 07.08 2.00 BANOOOO334 07.09 2.50 BANOOOO335 07.10 3.00 BANOOOO344
______
25.00 08.01 2.00 BANOOOOO17 08.02 2.00 BANOOOO321 08.03 3.00 BANOOOO324 08.04 3.00 BANOOOO330 08.05 2.50 BANOOOO336 08.06 2.50 BANOOOO337 08.07 2.00 BANOOOO338 08.08 2.50 BANOOOO339 08.09 3.00 BANOOOO340 08.10 2.50 BANOOOO341
______
25.00
______
______
100.00
..