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Exam Rept 50-245/86-21OL on 861215-19.Exam Results:All Candidates Passed Exams.Accuracy of Simulator Cause & Malfunction Book & Number of Errors Found in Training Matl Raised Concern
	ML20205F519
Person / Time
Site:	Millstone
Issue date:	03/12/1987
From:	Collins S, Howe A, Keller R, Lange D; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
Shared Package
ML20205F487	List: IR 05000245/1986021; ML20205F481;
References
	50-245-86-21OL, NUDOCS 8703310186
	Download: ML20205F519 (158)
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U. S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO. 86-21 (0L)

FACILITY DOCKET N FACILITY LICENSE N DPR-21 LICENSEE: Northeast Nuclear Energy Company P. O. Box 270 Hartford, Connecticut 06141-0270 ,

FACILITY: Millstone, Unit 1

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EXAMINATION DATES: December 15-19, 1986

l-CHIEF EXAMINER: A d Allen G. Howe, ReactoY Engineer Examiner Date 3 -7-ff7 REVIEWED BY: M[ -

David J. Lange,VLead M

R 9teactor Engineer Date 8- 9 - f ~/

Examiner REVIEWED BY: k Robdrt'M. Keller, Chief, Projects Date'~

3//0/e 7 I Section 1C APPROVED BY: b

' Samuel J. Collins, Deputy Director, ORP Date t 11 06 1

SUMMARY: Nine Operator and two Senior Operator license examinations were administered at Millstone Unit One the week of December 15, 1986. All candi-dates passed the examinations. A concern, expressed in the exit meeting, was the accuracy of the simulator cause and malfunction book. Another concern is the number of errors found in the training material and highlighted in the facility comments on the written examination These items are detailed in this repor PDR ADOCK 05000245 V PDR

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TYPE OF EXAM: Replacement EXAM RESULTS:

l R0 l SR0 l l Pass / Fail l Pass / Fail l

. I I I I I I I l Written Exam l 9/0 1 2/0 l l l l l

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10ral Exam l 9/0 l 2/0 l

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l l I I l Simulator Examl 9/0 l 2/0 l l l l 1

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1 I I I 10verall l 9/0 l 2/0 l l- 1 I I

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! CHIEF EXAMINER AT SITE: Allen Howe i OTHER EXAMINERS: D. Lange

L. Kolonauski

B. Hajek (U.S. NRC Consultant)

f 4 Summary of generic strengths or deficiencies noted on oral exams:

Strengths: Plant Familiarity, j Radiation Protection Practice '

Weaknesses: Knowledge of location and use of P&ID's and CWD's.

Summary of generic strengths or deficiencies noted from grading of written

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exams:

Reactor Operator Weaknesses: - Knowledge of detector overlap requirements during reactor startup.

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Strengths: - Plant design Senior Reactor Operator Weaknesses: -

Causes for " Gas Turbine Not Ready for Auto Start" alarm; Refuel interlocks-Use and interpretation of Technical Specifications j Strengths: -

Knowledge of reactivity effects LPCI, RPS, and recirculation systems i

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Normal operating procedures E0P entry criteria and Emergency plan Summary of strengths or deficiencies noted during simulator examinations:

a.) During the simulator scenerios, the examiners evaluated the candi-dates' ability to satisfactorily implement the Emergency Operating

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Procedures (EOPs). During emergency evolutions they were familiar with their individual and team responsibilities, they were able to ,

execute the E0Ps with the minimum shift staff identified in the facility Technical Specifications, the candidates did not physically interfere with each other nor did they duplicate efforts (unless required), and they were able to transition from one E0P to another and to enter the exit as required while assuring all necessary pre-cautions and steps were complete b.) During the administration of the simulator examinations several problems of concern were encountered and are explained below: The simulator cause and malfunction book provides base line, and sometimes erroneous, responses to the malfunctions. A highly

, detailed response is not required for simulator scenario pre-i paration but one case occurred when a malfunction caused a recirc pump to trip yet this effect was not mentioned in the cause and malfunction book. This limited detail contributed to lengthy scenario set-up time since the pre-administration review by facility representatives revealed numerious instances where

the pre planned event would not produce the desired effect and a rework was required. Improvements to this book would greatly

increase the quality of the simulator examinations.

t During the examinations a computer board failed producing a delay of more than two hours and ultimately the abandonment of simulator examinations for the last group of candidate This malfunction was repaired after all examinations were complet ,

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4 A fault in the power supply to the panel containing the electrical distribution system caused unusual indications and caused the examiners to abort one scenario. This problem is intermittent and should be corrected prior to the next examinatio . Summary of concerns on facility Lesson Plan material:

The facility comments on the written examinations reflect numerous examples where the facility provided training material is in disagreement witn other information, such as, operating procedures or Tech Specs or where the ma-terial is in error (as stated in comment). In other cases, the material is incomplete or the material initially supplied was inadequate and sup-plemental information was needed to validate the commen This is an area of concern because the large number of errors observed, relative to the small sample size used for testing objectives, indicates that the quality of the material, as a whole, is in doubt. The facility committed to amending some of the errors found during this examination and it is suggested that a review be performed to locate and correct any other errors which may exist, prior to the next NRC examinatio Details of the facility comments are attached to this report.

1 Personnel Present at Exit Interview:

NRC Personnel Allen Howe, Chief Examiner Dave Lange, Lead BWR Examiner Facility Personnel Harry Haynes, Manager, Operator Training Gregory Giles, Assistant Supervisor, Operator Training Michael Jensen, Assistant Supervisor, Operator Training Raymond Lueneneburg, Supervisor, Operator Training Ray Palmieri, Operations Supervisor Summary of NRC Comments made at exit interview:

The Chief Examiner discussed the generic weaknesses and strengths found during the oral examinations and the problems encountered with the simulator and simulator cause and malfunction documen The examination results were projected to be complete in four to six week _._,

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~5 The Chief Examiner requested input on how well the Knowledge and Abilities developed from NUREG 1123 and used for the SRO examination matched the facility learning objective The next examination, scheduled in May, was cancelled and the September examination was confirmed.

Attachments:

1. Written Examination and Answer Key (RO)

2. Written Examination and Answer Key (SRO)

3 .- Facility Comments and NRC Resolutions on Written Examinations made after Exam Review

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U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: MILLSTONE 1 REACTOR TYPE: BWR-GE3 DATE ADMINISTERED: 86/12/15 EXAMINER: B. K. HAJEK CANDIDATE: [NM 37[ INSTRUCTIONS TO CANDIDATE:

Use separate paper for the answers. Write answers on one side onl Staple question sheet on top of the answer sheet Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at 10ast 80%. Examination papers will be picked up six (6) hours after the examination start % OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 25.00 25.00 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 25.00 25.00 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 25.00 25.00 INSTRUMENTS AND CONTROLS 25.00 25.00 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 100.00 Totals Final Grade All work done on this examination is my ow I have neither given nor received ai Candidate's Signature

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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS ring the administration of this examination the following rules apply: Cheating on the examination means an automatic denial of your application and could result in more severe penalties. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating. Use black ink or dark pencil oniv to facilitate legible reproductions. Print your name in the blank provided on the cover sheet of the examination. Fill in the date on the cover sheet of the examination (if necessary). Use only the paper provided for answers. Print your name in the upper right-hand corner of the first page of each section of the answer sheet. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write oniv en one side of the paper, and write "Last Page" on the last answer shee . Number each answer as to category and number, for example, 1.4, .

20. Skip at least three lines between each answer.

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

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

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

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

15. Partial credit may be give Thereforo. ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLAN '

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

17. you must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been complete . _ _ _ _ _ ____-____ - _ _ __

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9. When you ccmplete your examination, you shall: Assemble your examination as follows:

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

' THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.01 (2.00)

Prerequisites in Procedure OP 335 require that LPCI not be operated for testing unless the keeptill system has been operating properl What adverse system transient is prevented by the LPCI keepfill system, and what might result if LPCI is operated with the keepfill system INOP7 (1.0) If the LPCI keeptill regulators had been isolated for maintenance, and you had returned them to service, what two observations could you make or parameters could you check to assure the system is properly fille (1.0)

QUESTION 1.02 (2.50)

When in cold shutdown with both recirculation system pumps shutdown, it is important to maintain flow through the core, Explain why it is necessary to maintain a flow path for natural circulation, and how this flow path will be assure (1.5) What two parameters can you observe to assure that natural circulation is occurring? (1.0)

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* THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.03 (2.00)

For each of the following, state whether the MAPLHGR /bE/

will increase, decrease, or remain the sam Near the beginning of life, cracks develop in the fuel, and the fuel comes into contact with the claddin (0.5) As fuel exposure increases, fission gases form and increase the gas pressure on the clad wal (0.5) As fuel exposure increases, the fission gases mix with the helium fill gas and cause a change in the heat transfer coefficient of the gases in the fuel pi (0.5) As the fuel ages, deposits form on the exterior surface of the fuel pin (0.5)

QUESTION 1.04 (2.00)

Concerning operations with the Rod Worth Minimizer System, p Y Early in the startup, the RWM maintains at least rd" ko one inserted rod between each two withdrawn rod '

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Explain how this rod pattern limits control rod 'J \,te#

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reactivity wort (1.0) 'l The RWM is not required above 20 percent powe Explain how operating at higher power levels has an effect on rod worth such that the RWM is not neede (1.0)

(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****) PRINCIPLES OF NUCLEAR POWER PLANT OPERATIO PAGE 4

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, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

QUESTION 1.05 (2.00)

State whether each of the following statements concerning suberitical multiplication is TRUE or FALS As Keff approaches unity, a larger change in neutron level occurs for a given change in Kef (0.5) If source neutrons are present in a reactor that is just critical (Keff = 1), count rate will increase at an exponential rat (0.5) When the count rate has doubled after a single notch withdrawal of a control rod, the reactor has l moved half way to critica (0.5) As Keff approaches unity, it takes longer for neutron level to reach equilibrium for a given change in Kef (0.5)

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QUESTION 1.06 (3.00)

Indicate whether the following will INCREASE or DECREASE reactivity during operation, and briefly EXPLAIN wh Moderator temperature increases during a reactor startu (0.75) Fuel temperature increase (0.75) A feedwater heater is los (0.75) Reactor primary system pressure suddenly decreases.(0.75)

QUESTION 1.07 (3.00)

Procedure OP 201, Approach to criticality, cautions that

"Certain xenon conditions and rod withdrawal sequences can result in extremely high rod notch worth as

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experienced here . . . Explain the specific operating conditions under which this will occur, where in the core these high notch worths may occur, and why the high notch worths will occur in these core location (3.0)

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1, PRINCIFLES OF NUCLEAR POWER PLANT OPERATION, PAGE 5 o aTHERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

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QUESTION 1.08 (3.00)

Pump shutoff head and pump runout must be considered whenever centrifugal pumps are operated. For each of the following, define the CAPITALIZED TERM, and Explain what adverse effects will happen if a pump is operated at its SHUTOFF HEAD for an extended period of time, how these effects will develop, and what methods are used to avoid those effect (1.5) Explain what conditions will cause a PUMP RUNOUT to occur, and how this adversely affects the pum (1.5)

QUESTION 1.09 (2.00) Define SHUTDOWN MARGI (0.5) Tell how the shutdown margin will change as a function of time after a reactor scram from a three month full power operation as a result of xenon poison effect (1.5)

QUESTION 1.10 (1.00)

A BWR ohould be operated in the Nucleate Boiling region of the Pool Boiling Curv Briefly describe what 10 meant by nucleate boiling.(0.75) Describe why the point of departure from nucleate boiling (DNB or OTB) should be avoide (0.25)

QUESTION 1.11 (1.50)

Peactivity additions to the core during reactor operation should avoid prompt critical, At what reactivity, ANY TIME during core life, would prompt critical be achieved? (0.5) Which time in core life (BOL, MOL, EOL) requiron the least amount of positive reactivity addition to achieve prompt criticality, and WHY? (1.0)

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1. _ERINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 6 o eTHERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUE3 TION 1.12 (1.00)

State whether each of the following statements concerning control rods is TRUE or FALS Withdrawal of a shallow control rod will affect the power in the fuel bundles around the control rod, but will have little effect on fuel bundles further awa (0.5) Withdrawal of a shallow control rod one or two notches can cause an overall bundle power decrease even though the bundle power increases at the bottom of the cor (0,5)

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(***** END OF CATEGORY 01 *****) P,LANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 7 QUESTION 2.01 (2.00)

Temperature control in the RBCCW System is performed by throttling the heat exchanger bypass valve RC-11 Why should the Service Waterl RBCCW he_at_ exchanger

~ outlet valves 4nauamba. thti.1.d7 (1.0)

@ t% .tNy. Ia. -(t k,e elm 7 Why must the Control Room be notified whenever RBCCW flow changes are made to the Radwaste "B" Concentrator Vapor Condenser? (1.0)

QUESTION 2.02 (3.00)

The differential pressure across a condensate domineralizer, and the minimum and maximum flow rates through the condensate dominera112ers must be monitored, especially during large reactor power changes, What are tho adverse consequences of an excessive dP or flow rate through a condensate demineralizer?(0.75) , What are the adverse consequences of an insufficient flow rate through a condensate demineralizer? (0.75) l What is the dP limit for the condensate demineralizers, and what actions are taken to assure operation within these limits during reactor power changos? (1.5)

QUESTION 2.03 (2.00)

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I Primary Containment System penetrations isolate on ,

i various signals and by valve groups dependent on the type of leak detected.

l What signals, including setpoints, will initiate a l

Group Two Isolation? (1.0) What type of leak (from where and to where) are the

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Group Two Isolation signals indicative of? (1.0)

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QUESTION 2.04 (2.00)

If low switchyard voltage occurs, excessively low voltage may occur in the Vital AC Systems, What minimum switchyard voltage must be maintained to preclude low voltage problems in the 120 V Vital AC Systems? (0.5) . How would you cause an increase in the switchyard voltage if it is determined that the voltage has fallen too low? (0.5) How would you determine (1) if switch,ard voltage i has fallen too low, and (2) if the RPS MG set has

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exceeded its low voltage limit? (1.0) M ~ ab d w M '.o % oI m "

QUESTION 2.05 (3.00)

i Four conditions are required for Automatic Pressure Reduction (APR) initiation.

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'st 4 d l What are these four conditions? ' g{ dJ (1.5)

i If these four conditions are pres'ent, and the t

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Drywell Pneumatic Air Supply to the APR valves has

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been lost due to a line brea , will the valves open automatically? Why or w y not? (0.75) If the Drywell Pneumatic Air Supply to the APR valves has been lost due to a line break, will

operation of the Remote Manual Switches in the Control Room have any effect on opening the valves?

Why or why not? (0.75) ,

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QUESTION 2.06 (2.50)

Three sources of makeup are available to the shell side l

of the Isolation condensers.

l Which source is preferred for initial filling? (0.5)

! EXPLAIN one reason why each of the alternate

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sources is not desireable as the preferred ,

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source for initial fili water.

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22__RLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 9 l

, QUESTION 2.07 (3.00)

Each Standby Gas Treatment train has several components to assure efficient removal of radioactive effluent What three components assure efficient operation of the two high efficiency filters? (1.5) - When and why might the SBGT filter train need to be

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cooled? (1.5)

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QUESTION 2.08 (2.00)

Concerning the Feedwater Coolant Injection (FWCI)

System, What is the purpose of the emergency condensate i l transfer pump, and when will it start? (1.0) Why is FWCI required when reactor temperature is above 330 degrees F7 (1.0)

i QUESTION 2.09 (1.00)

. The HP feedwater heaters are often removed from service near the end of core life. Why is it necessary to lower i reactor power before removing the HP heaters from j service? (1.0)

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QUESTION 2.10 (3.00) !

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With regard to the Low Pressure Coolant Injection (LPCI)

system, i What signale cause LPCI to initiate? (1.0)

i What interlocks must be satisfied to divert injection from the reactor to Containment Spray with a LPCI initiation signal present? (2.0) !

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. _ _ _ . . PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 10 QUESTION 2.11 (1.50)

The Instrument Air compressors have a four position control switch with the following switch positions:

STOP, NEUTRAL, STANDBY, and START, with a spring return to NEUTRA @sSTART With the control switch in NEUTRAhp what will '

control the running, loading, and unloading of the compressor? (0.75) Under what control switch and/or operating conditions will the Instrument Air compressor restart after a loss of normal power occurs and power is restored to the bus? (0.75)

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(***** END OF CATEGORY 02 *****) INSTRUMENTS AND CONTROLS PAGE 11 QUESTION 3.01 (2.00)

With the reactor operating and the turbine on the grid, a low condenser vacuum condition occurs. Describe how each of the following will respond to the low condenser vacuum trip (Vacuum Trip # 1): (1) Turbine Stop Valves, (2) Turbino Control Valves, (3) Turbine Bypass Valves, and (4) the Reactor Protection Syste (2.0)

QUESTION 3.02 (3.00)

During a reactor startup, you are advised to observe all nuclear instrumentation and to note proper instrument overlap. For the following two cases, state whether proper overlap would or would not be observed, and explain why or why not, During the early stagon of a startup, the SRMo have been periodically withdrawn on two occasions to maintain their readings betwoon 100 and 100,000 cps per OP 201, Approach to criticality. As the SRM readingo approach 20,000 cpo again, the IRMo are beginning to como on rang (1.5) During another reactor startup, the IRMo are reading about 20 on Range 9, and the APRMs are indicating about savon percen (1.5)

QUESTION 3.03 (2.00)

The plant 10 operating at full power, and the foodwater flow signal to the recirculation system falla to zero, How will the opoed of DOTil recire pumpo be affectert M4Y? (1.0) Ilow will the spood of BOTil recirc pumpo be af fected if the opued controller output on A MG sot also fails to zero? Wily? (1.0)

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QUESTION 3.04 (1.50)

The ATWS System is available to automatically shutdown the reactor 3n the event of an RPS failur Under what conditions will the ATHS System initiate? Include all setpoints knd time delay (1.0) What is the purpose of the 30 second time delay that maintains the ATHS scram valves open? (0.5)

QUESTION 3.05 (2.00)

Concerning the Reactor Manuni Control System, What two conditions could result in a " ROD DRIFT" alarm? (1.0) You are using the Notch override Switch to insert a control rod and the ROD DRIFT alarm does not come i Is this normal? Why or why not? (0.5) When using the Notch Override Switch to insert a control rod, when might you get a ROD DRIFT alarm? (0.5)

QUESTION 3.06 (3.00)

How will each of the following systemn or components be affected on a loss of the Vital AC Motor Generator and an automatic bus transfer to the emergency bus (IV-1),

and what operator action will be required? Reactor Pressure Controlle (1.0) Feedwater Regulating Valve (1.0) Recirculation Ma Set (1.0)

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QUESTION 3.07 (2.00)

CRD cooling water flow is normally 75 GPM. What is cooling water flow just after a scram? Explain wh (2.0)

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______.-W I,NSTRUMENTS AND CONTROLS PAGE 13 QUESTION 3.08 (3.00)

The Turbine control System contains a pressure control unit that consists of two independent pressure regulators and a bypass valve opening jack, With the reactor operating at full power, if an EPR LOSS OF POWER alarm occurs, explain what will happen to (1) the EPR and MPR servo position indications, (2) the red controller indication lights for the pressure regulators, (3) reactor pressure, and (4) turbine control valve positio (2.5)

!# # M ,T.nGEl'%h a % ek 3i+dtu Gha % % Q Toredjust-reactor pressure-es i... Gil. c --i =%

will you need to raise or lower the MPR setpoint?

Will this action raise or lower the servo position indication? (0.5)

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i QUESTION 3.09 (2.00)

Reactor Building air is monitored for high radiation levels to protect personnel. What four automatic actions will occur if a high radiation level is detected in the reactor building exhaust duct? (2.0)

QUESTION 3.10 (3.00)

The reactor is operating at 40 percent power with the Feedwater Control System in single element control, and Level Channel "A" selected for input. The reference leg isolation valve to the Channel "A" Narrow Range GEMAC develops a packing leak. For each of the following items, briefly explain the effects, indications, or actions which are caused by this failure if no operator action is taken and if the leak is not isolated, ACTUAL RPV water leve (0.5)

"" INDICATED RPV water level). (0.5) Feedwater Level Control Syste (0.8) Reactor Protection Syste (0.8) ECCS statu (0.5) PCI statu (0.8)

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INSTRUMENTS AND CONTROLS PAGE 14 QUESTION 3.11 (1.50)

MSIV closure while the reactor is operating may cause a reactor scram, If four MSIVs have closed, will a reactor scram occur as a function of valve closureA,(not pressure effects)? Why or why not? Ages (1.0) Under what conditions is the MSIV closure scram bypassed? (0,5)

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i f PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 15 o ' RADIOLOGICAL CONTROL QUESTION 4.01 (2.00)

State the reasons for the following two Immediate Actiong steps from ONP 502, Emergency Plant Shutdow Be sure also to include the reasons for the conditional parts of the action step Trip the turbine when below 50 MWe, (1.0) Transfer the Reactor Mode Switch to SHUTDOWN when it is determined the pressure regulating system is operating properl (1.0)

QUESTION 4.02 (1.00)

According to OP 301, Nuclear Steam Supply System (RR System), state the two reasons for the precaution,

" Recirculation System cross-tie valves will not be open unless both recirc pumps are secured." (1.0)

QUESTION 4.03 (3.00)

The reactor is operating at rated conditions when you receive the REACTOR BUILDING COOLING WATER PUMPS DISCHARGE PRESSURE LOW ard the REACTOR BUILDING COOLING WATER SURGE TANK LEVEL LOW .'larms, What is the likely erase of this set of alarms? (0.5) What are the immediate operator actions per the Off-Normal Procedure? (2.0) As a subsequent action, the procedure directs you to secure the nitrogen compressors if they had not trippe Why? (0.5)

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l PROCEDURES - NORMAL, ABNORMAL. EMERGENCY AND PAGE 16 i

. RADIOLOGICAL CONTROL

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o QUESTION 4.04 (1.50)

According to SHP 4902,. External Radiation Exposure Control and Dosimetry Issue, ,

1 What is the quarterly exposure limit for an operator with an up-to-date NRC Form 4 on file? (0,5) What is required to exceed the normal quarterly limit, and what is the next maximum exposure control beyond the normal linit? (1.0)

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QUESTION 4.05 (3.00)

While taking hourly instrument readings, you note that condenser vacuum is trending down (the condenser is losing vacuum), According to ONP 507, Loss of Vacuum, which of the following possible symptoms should you check to either confirm that vacuum is decreasing, or to identify the cause of the vacuum loss? (2.0) Off gas isolation valve closure Generator output decreasing Reactor pressure decreasing Circ water system high discharge pump trip Recombiner isolated alarm Low RBCCW discharge pressure SJAE inlet steam pressure decreasing High dP across intake screens Steam seal regulator malfunction 1 Reactor water level decreasing Explain why reducing reactor power can slow the decrease in the rate of loss of condenser vacuu (1.0)

-

(***** CATE00Hy 04 CONTINUED ON NEXT PAGE *****) ;

I

_ _ _ _ _ - _ _ _ _ _ ___ __ . _ . _ _ . _ _ _ _ _ _ _ _ - _ _ PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 17

' RADIOLOGICAL CONTROL QUESTION 4.06 (3.00)

General Caution # 2 in EOP 569, E0P Administration Procedure, states " Monitor RPV water level and pressure and primary containment temperatures and pressure from multiple indications." General Caution # 3 states "If a safety function initiates automatically, assume a true initiating event has occurred unless otherwise confirmed by at least two independent indications." In order to satisfy the requirements of General Caution # 2, explain with what frequency you must perform the monitorin (1.0) For these two cautions, explain what the difference is between " multiple" and " independent" indication (2.0)

QUESTION 4.07 (2.00)

According to the Precautions of OP 201, Approach to criticality (which are also applicable to OP 202 and to OP 203), At what power level must the Rod Worth Minimizer be operable? (0.5) How many Source Range channels are required to be operable and on scale? (0.5) What is the minimum permissible positive period? (0.5) When must the drywell be purged? (0.5)

QUESTION 4.08 (3.00)

According to ONP 5158, High Conductivity After Condensate Demineralizers, when should the Immediate Actions be taken as conductivity increases, and what are those Immediate Actions? (3.0)

(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****) PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 18

+ * RADIOLOGICAL CONTROL QUESTION 4.09 (2.00)

You have been designated as the " Operator in Attendance" for repair work on the Standby Gas Treatment Syste According to ACP-QA-2.06A, Station Tagging, What are your duties and responsibilities while on the job? (0.5) What actions must be taken if you leave the job sit Include time considerations in your answer. (1.5)

QUESTION 4.10 (2.50)

For each of the following conditions, indicate whether or not Emergency Operating Procedure entry is require If entry is required, state which procedure (s) to ente If entry is not required, state "Ncne." Consider each sub part as a separate item. Assume no additional conditions are present for each individual ite RPV level is 10 inche (0.25) Reactor power is 12 percent, Startup mod (0.25) Reactor power is 93 percent one minute after a load rejec (0.25) Power operations, Group I isolation occur (0.25) Suppression Pool level is 13.3 fee (0.25) Drywell pressure is 2.5 psi (0.25) Suppression Pool level is 13.7 fee (0.25) Suppression Pool temperature is 95 degrees (0.25) Reactor shutdown, RPV pressure is 1090 psi (0.25) Drywell temperature is 160 degrees (0.25)

QUESTION 4.11 (2.00)

In using OP 303, Reactor Cleanup System, you are cautioned twice to maintain pressure upstream of the drain flow regulator greater than 5 psi (1) What automatic action does maintaining system pressure prevent, (2) why does system pressure have to be monitored so carefully, and (3) what is the purpose of this system feature? (2.0)

(***** END OF CATEGORY 04 *****)

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

le PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 19

' * T'H"RMODTNAMICS , EEAT TRANSFER AND FLUID FLOW ANSWERS -- MILLSTONE 1 -86/12/15-B. K. HAJEK ANSWER 1.01 (2.00)- Water or fluid hammer [0.5] could cause damage to the pipes and other components in the system [0.5]. System pressure [must be above 35 psig (If it is at 45 psig, the primary PCV has failed and the backup is operating.)] [0.5]

Water leaving the high point vents. [0.5]

REFERENCE GE Heat Transfer & Fluid Flow, 2/85, pg. 6-53 Objectives 8.1, 1 Sys Text 1335, Section 6. Objectives 36, 3 K&As 293006-K1.05 (3.2), K1.12 (2.9).

ANSWER 1.02 (2.50)

' A natural circulation flowpath must be maintained

/ to prevent thermal stratification of the reactor coolant (0.5], which could lead to vessel repressurization (0.5]. A flowpath will be maintained if the water level is maintained at +50 inches (or above the first and second stages of the steam separators, or above the lowest moisture separator turnaround port). [0.5]

Core plate dP and jet pump flow) 24A},fempf y pua4W4.j2/A N

~ ,

F ^

REFERENCE M *1 O* 1, OE Heat Transfer & Fluid Flow, 2/85, pg. 8-55 - 8-5 Objectives 10.1, 10.2, 1 D) gg gjaggy,gf3 4/

SDC Text, Section 6.1. K&As 293008-K1.35 (3.1), K1.36 (3.1).

g gg );g

,

1,1)wedC.I(C, OP 2.06 , q 'l 5 '/.11, Flo-31 0.4b ,

b L u .n n o->. PRINCIPLES OF NUCLEAR POWER PLANT OPERATIO PAGE 20

' THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- MILLSTONE 1 -86/12/15-B. K. HAJEK ANSWER 1.03 (2.00) Increases (because the heat transfer coefficients at the points of contact are greater . . .] (0.5) Decreases [because the stress on the cladding is greater.) (0.5) Decreases [because the heat transfer coefficients of the fission gases are lower than that of helium.] (0.5) Decreases [because crud lowers the heat transfer rate from the surface.] (0.5)

REFERENCE GE Heat Transfer & Fluid Flow, 2/85, pg. 9-2 Objectives 4.2, 4.3, 4.4, and K&As 293009-K1.12 (2.9), K1.13 (3.1), K1.16 (2.4),

ANSWER 1.04 (2.00) The presence of the inserted rods depresses the flux around the withdrawn rod, and thus decreases the reactivity effect of the rod being withdraw [0.5) The in-between rods also decouple the withdrawn rods from each other, further reducing their potential reactivity wort [0.5) As power increases and voids form at the top of the core, [0.3) local flux peaking is limited because because the neutrons travel further in the voided area [0.3) and the effects of the dropped rod are spread over a larger area of the core. [0.4)

REFERENCE GE BWR Academic Series - Reactor Theory, ppg. 5-13, 16 -

1 Objectives 2.4, RWM Text, pg. K&As 292005-K1.09 (2.5), K1.10 (2.8).

. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 21

* THERMODYNAMICS , HEAT TRANSFER AND FLUID FLOW ANSWERS -- MILLSTONE 1 -86/12/15-B. K. HAJEK ANSWER 1.05 (2.00).

' True False True True REFERENCE GE BWR Academic Series - Reactor Theory, ppg. 3-5 - 1 Objectives 1.1, 1.2, 1.3, K&As 292003-K1.01 (2.9), K1.02 (2.1).

ANSWER 1.06 (3.00) Decreases (0.25] due to [the increase in neutron leakage from] the modcrator temperature coefficien [0.5] Decreases [0.25] due to the Doppler coefficient [and increased neutron resonance absorption]. [0.5] Increases [0.25] due to [ increased subcooling (or colder water going into the core) and] the moderator

temperature coefficient. [0.5] Decreases (0.25] due to [ swelling of the voids and]

the negative void coefficient. [0.5]

REFERENCE GE BWR Academic Series - Reactor Theory, ppg. 4 -8, 9, 16, 24, 34, 35, 3 Objectives 1, 2, 3, 4, 5, K&As 292004-K1.02 (2.5), K1.10 (3.2), K1.05 (2.9).

1- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 22

- * THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- MILLSTONE 1 -86/12/15-B. K. HAJEK ANSWER 1.07 (3.00).

l This will occur for a startup immediately after a shutdown or reactor scram under conditions of peak xenon (0.5]. I The high notch worths will occur in the periphery of the core (0.5] where the flux was previously the lowest

[0.5].

This is because xenon production is a function of the reactor neutron flux [0.5]. Therefore, the xenon concentration will be highest where the neutron flux was highest - that is, near the center of the core [0.5].

This will depress the flux in the center, forcing the high flux region to the exterior of the core [0.5].

REFERENCE GE BWR Academic Series - Reactor Theory, ppg. 6-1 Objectives 2.5.2, 2. K&As 292006-K1.07 (3.2),K1.08 (2.8), K1.10 (2.9), K1.14 (3.1)

ANSWER 1.08 (3.00) Shutoff head is the head developed when a pump is 3 h ^^

operated with its discharge valve closed. [0.5]

The pump will eventually add sufficient heat to the 1 ,Eto bMNg fluid from friction to cause cavitation leading to internal damage. [0.5] This is precluded by providing min flow valves or a trip if the discharge valve is not opened within a specified

, time period. [0.5)

Thwd h Q.r.0 &s*4lm 5 a. 4o n law dwshu~{6;scW) pursW4. Rynout;may' ,c9used-downstream o~f the ,pum;by,pympfng-to_a_

p'. [0.&T' This causes low pressure the pump flow to increase above design flow which causes the impeller to speed up (runout) [0.5] and the motor amps to increase, thus causing damage to the windings. [0.5]

REFERENCE GE Heat Transfer & Fluid Flow, 2/85, pg. 6-108 - 10 Objective 1 K&As 293006-Kl.17 (2.6), K1.19 (2.7).

l l

l

!

l

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1 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 23

* THERMODYNAMICS , HEAT TRANSFER AND FLUID FLOW ANSWERS -- MILLSTONE 1 -86/12/15-B. K. HAJEK ANSWER 1.09 (2.00). Shutdown margin is a measure of how suberitical a reactor is in terms of Kef Alternately SDM = 1 - Kef The shutdown margin will increase initially (0.5]

untill about 7 - 11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months after the scram (0.5].

It will then decrease until the xenon has essentially decayed about 70 hours8.101852e-4 days 0.0194 hours 1.157407e-4 weeks 2.6635e-5 months after the scra (0.5).

Alternately a curve may be drawn with the times indicate REFERENCE GE BWR Academic Series - Reactor Theory, pg. 1-36 and 6-11 Objectives 1:4.1 and 6: K&As 292002-K1.10 (3.2), 292006-K1.12 (2.8)

ANSWER 1.10 (1.00) Nucleate boiling is the process by which bubbles form at the surface of the fuel rod, and carry heat away from the rod as they grow and leave the surfac [0.75) At DNB, the heat transfer rate decreases. [0.25]

REFERENCE 1" GE Heat Transfer & Fluid Flow, 2/85, pg. 8-8 - 8-1 Objectives 2.2, 2.4, K&As 293008-K1.07 (2.8), K1.09 (3.0).

ANSWER 1.11 (1.50) When added reactivity exceeds beta effectiv EOL [0.5) because that is when beta effective is the smallest. [0.5)

l l

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a- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 24

. * THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- MILLSTONE 1 -86/12/15-B. K. HAJEK REFERENCE GE BWR Academic Series - Reactor Theory, ppg. 3-30 - 3 Objectives 3.5, K&As 292003-K1.04 (2.5), K1.07 (3.3).

ANSWER 1.12 (1.00) True True REFERENCE GE BWR Academic Series - Reactor Theory, pg. 5-2 Objectives 3.3, K&As 292005-K1.11 (2.4), K1.12 (2.6).

!

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.. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 25

. .

ANSWERS -- MILLSTONE 1 -86/12/15-B. K. HAJEK

.

ANSWER 2.01 (2.00) The piping makes a vertical drop at the exit of the heat exchanger, and any change in throttle valve position could result in draining the heat exchanger and actually reducing the coolin Temperature control is performed manually, and failure to adjust the heat exchanger valving could result in increased drywell temperatures and a possible reactor scra REFERENCE RBCCW Text, ppg. 13 - 1 Objectives 14, 1 K&As 223001-K2.10 (2.7) (Drywell Chillers), K3.02 (3.3),

K6.01 (3.6). (No K&As for RBCCW)

ANSWER 2.02 (3.00) Resin beads will fracture, escape through the effluent strainer, and enter the reacto Channeling may occur causing uneven distribution of condensate through the resin bed and therefore insufficient demineralizatio dp(max) = 40 psid [0.5]

Demins must be removed from service or placed into service as flow requirements change to maintain operation within the limits. [1.0)

REFERENCE CN Text, ppg. 12 - 1 Objectives 9, 10, 15, 16, 1 K&As 256000-K4.04 (2.7), A1.01 (2.9), A1.07 (3.1), A2.05 (2.9), A2.16 (2.8).

Note: This question is also related to FWCI operation since a minimum of three demins must be on line for FWC PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 26

-

ANSWERS -- MILLSTONE 1 ~ -86/12/15-B. K. HAJEK

_

ANSWER 2.03 (2.00) Low water level at +8 inches Hi Drywell pressure at 2 psig A reactor coolant leak to the primary containment (Drywell).

REFERENCE PC Text, pg. 5 Objectives 19, 2 K&As 223001-K4.03 (3.7), A2.01 (4.3).

ANSWER 2.04 (2.00) Above 3 5 KV gg g n 409 By contacting CONVEX and requesting a voltage } t ^^ *

increas ) S m+ tk * 907 F

b#k % '3'd bs.n h Check the electrical assembly relay drops [0.5] and note a loss of the associated RPS bus. [0.5] d** 'E

<a chio w ..-. M.

'

REFERENCE go gj wJ V&I AC Text, ppg. 9, 1 gg Objectives 17, 20, 2 K&As 262002-K6.01 (2.7), A1.02 (2.5), A2.01 (2.6). Il s T

.

t i

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2 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 27

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ANSWERS -- MILLSTONE 1 -86/12/15-B. K. HAJEK ANSWER 2.05 (3.00) . Drywell high pressure sealed in .

Rea,ctor lo-lo water level [-48"]

120fimedelayexpired p g gg yc.)

, , , At least one Core Spray or LCI pump running (loopsg A - f 58; 0.375 PER ITEM Yes, the valves will still open [0.25). The 3thS I'" b accumulators will provide nitrogen to open the g bt.q j kid. *4 valves (0.5). .

. Ye [0.25] Manual actuation 2,,,0 7 b + C- *

wisho-. 04 D,waar-the APR signal requested open gp A)kh h detowhm , [0.5), ,gp g , , y,g ;yqg,,

,

REFERENCE MN APR Text, ppg. 5, 10, Figure .jedeb5 4.-

Objective No objective for Part Q ph M h*% #

Part a. K&As 218000-K1.01 (4.0), K1. g.f g gy gu 644 ),

K4.03 (3.8), A3.01 (4.2), A4.02 (4.2,.

Part b K&As - K4.04 (3.5), K6.04 (3.6), A2.03 (3.4).

K&As 295019-AK2.18 (3.5)

ANSWER 2.06 (2.50) Demin water through the Condensate Transfer Syste The Condensate Storage Tank [0.25) contains radioactive contaminants [which would concentrate in the IC shell.due to water evaporation,) and would cause problems in distinguishing between background radiation levels and a U-tube leak. [0.75) 53*8 - % W j ,*

. ,x%g h t.aw.*j The Fire Water System [0.25) contains chlorides that will cause [ stress corrosion cracking)

corrosion of the stainless-steel U-tubes. [0.75)

REFERENCE IC Text, ppg. 2- Objectives 3, 4, K&As 207000-K4.03 (3.3), K4.05 (4.0), A2.01 (4.2).

. .-- - _ _ _ PLANT DESIGN INCLUDING SAFETY AND EMERG2NCY SYSTEMS PAGE 28

.

ANSWERS -- MILLSTONE 1 -86/12/15-B. K. HAJEK ANSWER 2.07 (3.00) . Demister kw heater (or htr before the first filter) kw heater (or htr before the second heater) The train may require cooling AFTER use (0.5] to remove the decay heat generated by fission products collected in the filter during operation. [1.0)

REFERENCE SBGT Text, ppg. 3, 6 - Objectives 5, lla, lib, lid, 1 K&As 261000-K4.02 (2.6), K4.03 (2.5), K5.01 (2.3*),

K6.01 (2.9), A1.07 (2.8), A2.03 (2.9), A2.04 (2.5),

A3.04 (3.0), A4.08 (2.6).

ANSWER 2.08 (2.00) Provides a means of rapidly providing makeup to the condenser from the CST. [0.5) It starts when FWCI initiates. [0.5] This is the temperature that corresponds to the saturation pressure above which LPCI and Core Spray will no longer deliver rated flo REFERENCE FWCI Text-1334, pp , 1 Objectives 2, 4, .

ANSWER 2.09 (1.00)

The operation requires closing of manual valves in the N5 58' "M b im pa d4 4 N heater bay [0.5], and this requires reactor power to be lowered to reduce radiation exposures to personnel. [0.5] c4N A kwM,f REFERENCE k *

ES Text 1346, ppg. 14 - 1 *1f b ^U""h* '#*/

Objective 1 Amd4eavoid K&As 239001-K1.16 (3.2), A1.10 (3.8), A3.04 (2.7), p4 cA4 *fJ%y 6 '/>

A4.06 (3.6). v fnqht No

_- -

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. . PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 29

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ANSWERS -- MILLSTONE 1 -86/12/15-B. K. HAJEK ANSWER 2.10 (3.00) Hi Drywell pressure (2.0 psig) [0.34)

OR LoLo Rx water level [-48"] [0.33] AND < 350 psig Rx pressure. (0.33)

[ Injection won't occur until Rx pressure falls to LPCI pump shutoff head pressure of 268 psig.) With an auto initiation signal present, Containment Spray first key in override. [0.5) Vessel water level is greater than 2/3 core height (0.5)

OR Containment Spray second key is in manual override. (0,5) Drywell pressure is greater than 5 psig. [0.5)

REFERENCE LPCI Text ppg. 10, 19, 2 Objectives 20, 29, 30, 31, 3 K&As 203000-K4.01 (4.2), 226001-K4.03 (3.1)

ANSWER 2.11 (1.50) The compressor will run continuously because the switch is in neutral (after start) [0.25). It will be loaded and unloaded by the unloading valve (which cycles with receiver pressure). [0.5)

% dor l&

' It will restart if the control switch is in/STANDBYrsC position (0.375), or if the compressor was running prior to the power loss. (0.375)

y p e 4e 14F & D/F *

QS Myb" P ext 1333, ppg. 13 - 1 + h EN ew M - ""

Objectives 8, 'U ' *

gj S K&As 264000-K3.03 (4.1).

K&As 295019-AK3.02 (3.5).

.. . INSTRUMENTS AND CONTROLS PAGE 30

. .

ANSWERS -- MILLSTONE 1 -86/12/15-B. K. HAJEK ANSWER 3.01 (2.00) Stop Valves shut (0.5) p*

fgb po Control valves shut (0.5) *

i i Bypass valves open (0.5) VU fI RPS initiates a stop valve closure scram if first ,'uf ghh, stage pressure is greater than 45 percent (0.5)

REFERENCE i TG Text-1314, pg. 27, 199 Objective 3 K&As 245000-K3.07 (3.6), K4.05 (2.9), A1.06 (3.3), A1.07 (2.8), A2.03 (3.5).

ANSWER 3.02 (3.00) No. [0.5] Overlap should b2 noted such that the Prou d * y IRMs are coming on scale with the SRMs indicating 10,000 to 100,000 cps while they are full in. (1.0] ,

y -

.g No . [0.5] 20 on Range 9 is equivalent to about 2 }

percent power since 100 percent on Range 10 is yf,m /oYo

'

equivalent to 10 percent power. (1.0) j

'

REFERENCE _____

OP 201, ppg. 5-6, Sections 5.6 - SRM Text-1401A, pg. 33. Objective 1 J RM Text-1402A, ppg. 27 - 28. Objectives 14, 1 K&As 215004 (SRMs)-K4.04 (2.8), K5.03 (2.8),

A1.02 (3.6), A2.05 (3.3), A4.07 (3.4).

t K&As 215003 (IRMs)-K1.06 (3.9), A1.02 (3.7),

A2.05 (3.3), A4.07 (3.6).

K&As 215005 (APRMs)-A1.01 (4.0), A2.08 (3.2),

A4.06 (3.6).

'

L--TecA %c5 pg 3/4 1-4 ; ddi I

s, m .to.O w m n o u m m"% &^

TM

% .A+Lk

<

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8. , INSTRUMENTS AND CONTROLS PAGE 31

,

ANSWERS -- MILLSTONE 1 -86/12/15-B. K. HAJEK ANSWER 3.03 (2.00) Both pumps will run back to 28 percent (0.33]

because the FW flow signal to the recirc system is less than 20 percent, (or the. flow limiter is not bypassed at less than 20 percent.] (0.67] The B pump will run back as in Part a. (0.25]

The A pump speed will remain the same (0.25]

because of a scoop tube lockup caused by the loss of speed signal. (0.5]

REFERENCE RRSC Text-1301B, ppg. 9, 10, 14, 1 Objectives 3, 7, 8c, 1 K&As 202002-K6.04 (3.5), K4.02 (3.0).

ANSWER 3.04 (1.50) RPV pressure above 1150 psig RPV level < -48" for > 9 sec ( n fala) To assure that the scram air header has completely depressurize REFERENCE ATHS Text-1409A, ppg.4 - Objectives 1, KAs 201001-K2.05 (4.5).

.

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ANSWERS -- MILLSTONE 1 -86/12/15-B. K. HAJEK ANSWER 3.05 (2.00) . When a nonselected rod leaves an even notch positio . When a selected rod passes an odd notch position when the automatic sequence timer is not in a cycl Yes. (0.2] Notch override bypasses the alarm. [0.3) If the selected rod passes an odd reed switch after the NOS is release REFERENCE RMCS Text-1302A, pg. 1 Objectives 2b, K&As 201002-K4.01 (2.7), K4.03 (3.6), K4.06 (3.5),

A2.02 (3.2), A3.03 (3.2), A4.02 (3.5).

ANSWER 3.06 (3.00) EPR will be lost (0.5], and pressure control will Ta UE*O i5 f **b' '

need to be switched to the MPR. [0.5] Opsb me W4 O The FRVs will lock up (0.5], and need to be rese (0.5] The scoop tube will lock up [0.5] and need to be reset. (0.5]

REFERENCE ,#'

V&I AC Text-1343,.pg. 1 I LP h Objective 2 (

K&As 262002-K3.02 (2.9), K3.03 ( 3 . O T,- K313.j( 2 . 7 ) ,

-

K4.01 (3.1), A3.01 (2.8).

JA

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_ __ INSTRUMENTS AND CONTROLS PAGE 33

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ANSWERS -- MILLSTONE 1 -86/12/15-B. K. HAJEK (b% *b flem kg A FCY *IM'**")5 a~,.. *

ANSWER 3.07 (2.00)

a.y_,w.. - .%

M* I4 'u m M Cooling water flow goes to azero (0.5]g because the -

g-charging water header goes to reactor pressure, and flow .

to the header increases to the max allowed by the & r4 J2c v h restricting orifice (180 GPM). (0.75] Since flow is 'v. M M sensed upstream of the charging water branch, the flow control valve closes to maintain system flow. (0.75]

REFERENCE

'

CRD Text-1302, ppg. 46 - 47, Figure Objective 3 K&As 201001-K4.12 (2.9), A1.01 (2.9).

ANSWER 3.08 (3.00) . EPR servo position goes to zero (0.5] N MPR servo position stays the same ( 0. 5 ] co bow.u4 dh I fit / l-' / EPR light goes out and MPR light comes on (0.5] Reactor pressure increases (by an amount equivalent to a 10 percent difference in servo CAF positions, about 100 psig (per LP)] (0.5] Control valves close down (0.5] Lower the setpoint to reduce reactor pressur This will raise the servo position indicatio REFERENCE TG Text-1314, ppg. 88, 89, 121, 166 - 168.

'

Objectives 29al, 73, 173, 175, 176, 177, 17 K&As 241000-K1.02 (3.9), K1.08 (3.6), K3.02 (4.2),

K3.08 (3.7), K4.01 (3.8), K6.01 (2.8), A1.01 (3.9),

A1.08 (3.3), A2.01 (3.5), A2.16 (3.4), A4.02 (4.1).

.

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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ - _ _ _ - _ _ _ _ _ _ INSTRUMENTS AND CONTROLS PAGE 34

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ANSWERS -- MILLSTONE 1 -86/12/15-B. K. HAJEK ANSWER 3.09 (2.00)

Reactor Building ventilation isolates (0,5)

Steam tunnel ventilation isolates (0.5)

SBGT initiates (0.5)

REFUEL FLOOR / REACTOR BLDG VENT HIGH RADIATION alarm (on CRP 903)Cf ll psyt/Ls] (0.5)

REFERENCE PRM Text-1406A, ppg. 25 - 2 Objectives 12, 13 K&As 288000 (Ventilation)-K1.02 (3.4), K1.03 (3.7),

K1.05 (3.3), K4.01 (3.7), K4.02 (3.7), K5.01 (3.1),

A2.04 (3.7).

K&As 272000 (Rad Monitoring)-K1.06 (3.2), K2.05 (2.6),

K4.02 (3.7),

ANSWER 3.10 (3.00) Actual water level will decrease, Indicated water level increas The FWLCS will call for less water (because of the increasing water level indication). A reactor scram will occur as water level drops to

+8 inche As water level drops to -48 inches, ECCS initiations will occur. ( FWCI, IC, and if other initiating conditions come in, LPCI, Core Spray, APR). l[ Q .nsd M J

- As water level drops to +8 inches Groups II and III isolations will occur. Group I will occur at -48 inche REFERENCE RVI Text-1300B, pg.18. FW Text-1316, ppg. 33 - 3 RPS Text-1408A, pg. 19. Objective 1 FWCI Text-1334, pg. 7. Objective PC Text-1311A, ppg. 49 - 5 K&As 216000-K1.12 (3.7). 259002-K1.01 (3.9), K1.03 (3.9), K5.08 (3.8). 259001-K1.08 (3.7), A2.07 (3.8).

223002-A2.05 (3.6).

,

,-, . , - , . - - . . , - , , , _ , , - - __ _ , . - - - _ . . , _ _ . . , , _ . . , _ . . - INSTRUMENTS AND CONTROLS PAGE 35

. .

' ANSWERS -- MILLSTONE 1 -86/12/15-B. K. HAJEK ANSWER 3.11 (1.50) The scram is a function of how many steam lines 1(**#d***"

have been closed rather than the number of valves Q _o),

that have gone closed. [0.5) Therefore, if the closed valves are in at least three steam lines, a J(6*N r**tf j

full scram will occur. (0.5) ffng- 44, If reactor pressure is less than 600 psig and the mode switch is not in RU REFERENCE RPS Text-1408, ppg. 22 - 2 Objectives 17d, 18e, 21, 2 K&As 212000-K1.14 (3.6), K4.02 (3.5), K4.12 (3.9),

A2.11 (4.0), A2.16 (4.0).

~

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

. '4 . PROCEDURES - NORMAL. ABNORMAL, EMERGENCY AND PAGE 36

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RADIOLOGICAL CONTROL ANSWERS -- MILLSTONE 1 -86/12/15-B. K. HAJEK

!

ANSWER 4.01 (2.00) This is to assure that the turbine does not become a system load [0.5), and is delayed to assure a premature tripadoes not occur.,[0.5)

cu m ~ p w.a.+ & This step must be accomplished before the MSIVs are closed by the 825 psig setpoint, possibly _.. -ht; f loMd jd#g* '

compounding existing problemsf. [075 F 1rdwever, if

, pressure control is malfunctioning, allowing the isolation will prevent a premature blowdown and cooldown. [0.5)

REFERENCE ONP Text-1500A, pg. 1 Objective 1 ONP 502 K&As 295006-AK2.01 (4.3), 2.07 (4.0), G7 (3.8).

.

ANSWER 4.02 (1.00) To comply with Tech Specs - _ tTo auur LPtI loop db lohc To prevent pump damag A

!

* J9 *L* 3 /3yten 4 E *6#, I#}^'

! REFERENCE 2M k 56 '*d ' RR Text, pg. 5 $cjo 3/'l b ' i Objective 30.

' OP 301, pg. KEAs 202001-K1.13 (3.1), GS (3.4), G10 (3.5).

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h i \ ! l

, , - _ . _ . , - . , . . . , - _ . _ . . _ . _ _ - . . . _ _ . . . __,--_,. . . , _ _ , . _ . _ _ _ , - _ _ _ _ _ _ _ . - - _ . _ - _ _ ,  , , _ _ _ _ _ _ _ . --_ _ . , . - - , ' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 37
*
* RADIOLOGICAL CONTROL ANSWERS -- MILLSTONE 1  -86/12/15-B. K. HAJEK ANSWER 4.03 (3.00). A rupture in RBCC . Attempt to isolate any system ruptur . Run recires to minimum spee . Scram the reacto . Trip turbine when below 50 MW . Shut MSIV . Initiate Isolation Condense . Start Standby Ga . Vent drywell and torus [through the 2 inch vents 1-AC-9 and -12.]
 (0.25 per numbered item) To prevent overheating and damag REFERENCE ONP Text-1500A, ppg. 173 - 17 Objectives 3, 10, (Etrapolated - 23)

ONP-524C K&As 295018-AK1.01 (3.5), AK2.01 (3.3), AK3.03 (3.1), G5 (3.5), G10 (3.4).

ANSWER 4.04 (1.50) mrem An " Increased Radiation Exposure Authorization" must be submitted and approved by the HP Supervisor or his designee. [0.5) The next limit is 2000 mrem. [0.5) - REFERENCE SHP 4902, pg. K&As 294001- K1.03 (3.3). PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND

PAGE 38

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RADIOLOGICAL CONTROL ANSWERS -- MILLSTONE 1 -86/12/15-B. K. HAJEK ANSWER 4.05 (3.00).

, . Yes 0.2 pts for each correct answe . Yes If candidate only lists the Yes No answers, No will be assumed for Yes the other . Yes No . Yes ;)y1 q f 7>gJ g J > Yes Yes 1 No Reducing power assists in maintaining vacuum by reducing the input of the non-condensible gasses (0.5] and lowering the condenser heat load. [0,5) REFERENCE ONP Text-1500A, ppg. 74 - 7 Objectives 58, 59, 60, 62, 64, 6 NP 507 K&As 295002-AA2.02 (3.2), GS (3.2).

ANSWER 4.06 (3.00) The term is understood to connote a frequency of observation sufficient to maintain an overall awareness of plant status. It does not imply uninterrupted activity, or require the dedicated attention of one individual, but is a general responsibility to be executed concurrently with other operational duties.

, Multiple means a number greater than one sufficient to assure the validity of the information. It may be defined as simply more than on Independent implies a more fundamental separateness than multiple in signal detection, processing, and display, minimizing the possibility of common mode failure REFERENCE EOP Text-1500B, ppg.5 - 1 Objectives 4, Sa, S . PROCEDURES - NORMAL. ABNORMAL, EMERGENCY AND PAGE 39

. o RADIOLOGICAL CONTROL AKSWERS -- MILLSTONE 1  -86/12/15-B. K. HAJEK K&As 295024-38, System Generic K/As 7 (3.6), 9 (3.7),

12 (3.9).

ANSWER 4.07 (2.00) Below 20 percent power (unless 12 or more rods have been withdrawn and a verifier is present.] Tw se Whenever primary containment integrity is required. - ri 70/1 M) REFERENCE L u l ka ophd a o W whn d o mscAICcel u ,d ,,Jn,* dut*F m x k,qQ L J ' 2.414 spa OP 201, pg. K&As 294001-K1.15 (3.4), 20 2. , Y T 7 1 d 11 6* 4 Rml

  % o f Lo C Ym 20s-t go b * * 311 T' 5. 3,4f-8 , 9 3.-), A,6,
     +a 6/

b-MW ANSWER 4.08 (3.00) When the conductivity of the feedwater system reaches 0.5 micromhos/cm (0.5], Scram the reactor, Close 1-CN-67 and 1-CN-69 (Hotwell Reject to CST) Trip all operating feed pumps Close 1-FW-4A, B, C (Feedwater blocking valves) Close all MSIVs and initiate Iso Condenser 0.5 for each numbered item. For items 2 and 4, valve numbers not required if description give REFERENCE ONP Text-1500A, pg. 126.

- Objective 100, 10 K&As 256000-A2.15 (2.8), A4.09 (2.9), G1 (3.5), G8 (3.4), G14 (3.6). (No AOP K&As)

    - - -  - -      _ _ - - PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND         PAGE 40
* *

R)DIOLOGICAL CONTROL ANSWERS -- MILLSTONE 1 -86/12/15-B. K. HAJEK ANSWER 4.09 (2.00). Responsible to position valves and/or breakers as listed on the Tag Log Sheet. [0.5) If leave for less than one hour (0.25), the job I5 '7 lW h shall be stopped [0.25), and the Work Order shalli be in the custody of the operator. (0.25] ) 49h g If leave for more than one hour (0.25), the valves and breakers shall be tagged as listed on the Tag M #.7 Log Sheet (0.25), and the Operator in Attendance is N P""' responsible to assure they remain us listed until % k.* permission to cancel is granted. (0.25) yy REFERENCE ACP-QA-2.06A, p . K&As 294001-Kl.02 (3.9).

ANSWER 4.10 (2.50) 1 None None b pL MdM"# E d W ED/k to W 9 j , 571, 572 o . (. / b , 571, 572 'e E69 M c.s upyn o.nte vu , 571, 572, 580 e .r/ if a ~ None cn ' e , 571, 572 a: f. * d =b6 REFERENCE EOP Text-1500B.

'

Objective E0Ps 570, 571, 572, 58 K&As 295024-38, System Generic K/As 11 (4.3).

, y -- - , , - . -- - ,-----~,_-.e-- ,_,,-,,r,_- - - , - - - - - - - , _ - - . - - - - . - , - . - - - . . ~ - . , - - .

           - , , , - ,-- ,,--, ,,. -

4 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 41

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RADIOLOGICAL CONTROL ANSWERS -- MILLSTONE 1 -86/12/15-B. K. HAJEK ANSWER 4.11 (2.00) The reject flow control valve (drain flow regulator) y, d d will automatically go shut (0.5] and not be annunciated (0.5] * D ,y.MY% g The automatic close feature limits drain flow to prevent 4,jr) ed* j+ draining system high points, which are above normal - reactor water level. (1.0) ,,__ REFERENCE OP 303, ppg. 11 - 1 RWCU Text, pg. 5 Objectives 43, 4 K&As 204000-K4.04 (3.5), A1.04 (2.8), A1.05 (2.6), A2.03 (2.9), A2.10 (2.7), A2.12 (2.7), A3.01 (3.3), G5 (2.9), GIO (3.2).

i

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 .

e - TEST CROSS REFERENCE PAGE 1 ESTION VALUE REFERENCE QUESTION VALUE REFERENCE _______ ______ __________ ________ ______ __________ 01.01 2.00 BRH0000230 03.01 2.00 BRH0000253 01.02 2.50 BRHOOOO231 03.02 3.00 BRH0000254 01.03 2.00 LRH0000232 03.03 2.00 BRH0000255 01.04 2.00 BRH0000233 03.04 1.50 BRH0000256 01.05 2.00 BRH0000234 03.05 2.00 BRH0000257 01.06 3.00 BRH0000235 03.06 3.00 BRH0000258 01.07 3.00 BRH0000236 03.07 2.00 BRH0000259 01.08 3.00 BRHOOOO237 03.08 3.00 BRH0000260 01.09 2.00 BRH0000238 03.09 2.00 BRH0000261 01.10 1.00 BRH0000239 03.10 3.00 BRH0000262 01.11 1.50 BRH0000240 03.11 1.50 BRH000026o 01.12 1.00 BRH0000241 ------

 ------

25.00 25.00 04.01 2.00 BRH0000264 02.01 2.00 BRH0000242 04.02 1.00 BRH0000265 02.02 3.00 BRH0000243 04.03 3.00 BRH0000266 02.03 2.00 BRH0000244 04.04 1.50 BRH0000267 02.04 2.00 BRH0000245 04.05 3.00 BRH0000268 02.05 3.00 BRH0000246 04.06 3.00 BRH0000269 02.06 2.50 BRH0000247 04.07 2.00 BRH0000270 02.07 3.00 BRH0000248 04.08 3.00 BRH0000271 02.08 2.00 BRH0000249 04.09 2.00 BRH0000272 02.09 1.00 BRH0000250 04.10 2.50 BRH0000273 02.10 3.00 BRH0000251 04.11 2.00 BRH0000274 02.11 1.50 BRH0000252 ------

 ------

25.00 25.00 ------ ______ 100.00 i.

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

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

hr*7*RCl1n1Ct7f $ NULLEAR REGULATORY COMMISSION

SENIOH REAClUR OPERAlOR L1 CENSE EXAMINATIUN FACILI1Y: _ M I LLSIO!)E _1_ _ _ _ _ _ _ __ _ _ _ _ REACIOR TYFE: _DWR-GE_3__________________ DATE ADMINISTERED: _@6[12/15_____________,__ __ EXAMINER: !j g EJE3_ A . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ CANDIDATE: __ . _ _k[[_k__,,,______

.I NSI6UCIJ ONS_IO_.CONDID61El Use separate paper for the answer Write answers on one sidt onl Staple question sheet   on top of the answer  sheet Points for each questinn are indatated in parenthenes aiter the questio The passing orade requiren at least 7 6). In each category and a final      orade of at 1eant OO El ami n a t 1on papert, wiii be picked up nl:    (6)  h ot ir s; at t er the ex smi nat i oi. st ar t , (J C Ahii l Di il F ' ".
   '

C AT Et'il lR Y O! Lo 1 Eht ih Y _ YblVE.. _1010L _ _ _bf Obl .._ _ WJI c.M _ _ _ _. _ _ _ _ _ . l t2 [E O Ob I ._ _ _ _ _._ - . _ _ _ i f' . 99 . .J' U . ' ' O . . _ lHEUhy (!F NUI,l.E sk FUWER FLAN 1 UFEh AT it p F Lu f DL. Af40 l HLhMt)D Y fMNI C b 2 % I" . _ lb . !'1' _ _ . _ . _ _ _ . ._._.._.._ FLANT SYSTEMu DEhlbN. f O'lT N UL., 4ND IN51 RUME N T AT I UN D. 99._ _ jb. !J9 . _ . _ . _ . . . . . _ _ _ __ _ _ . . . _ PH DC E.DUkf. 5 - NURMAL. AbNf H <hst. , EMEkbENCY AND RADIULOf>ICAt CONTROL .25299_ _2bz99 _ _ _ _ _ _ . _ _ _ __ __ ADMINISTRATIVE PROCEDURE CONDITIONS. AND LIMITATIONS 100.00 1otals Final brade alI wark done on th1c enamination in my ow I have nelther givFM nor received ai ________ _ ___ . ____ ____ Candidate's Signature

 -.  . . - _ - -_ . _ - _
     . _ - - - . _ _ . .
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply: Cheating on the examination means an automatic denial of your appl i cati on and could result in more severe penaltie . Restroom trips are to be limited and only one candidate at a time may leav You must avoid all contacts with anyone outside the examination - room to avoid even the appearance or possibility of cheatin . Use black ink or dark pencil gnly to facilitate legible reproduction ' 4 Print your name in the blank provided on the cover wheet of the examinatio Fill in the date on the cover sheet of the examination (if necessary). Use only the paper provided for answer . Print your name in the oppor right-hand corner of the first page of each section at the annwer shee Concecut i vel y number each answer sheet. Write "End of Category _" as appropriato, start each category an a npy page. write orRy on one side at the pepor, and write "Last Page" on the last answer shee Number each answer as to categor y and number, for example, 1.4, . Skip at l ea s t. tbreo lines between each answe . Separ ate answnr nheets from pad and place finished answer sheets fate down on voor d rm i or tabl . Use abbr eviat i ons only it they are commonly used in fatality }1t er3htr . The point valuo 'or each question is indicated in parentheses after the question and can be used as a guide for the depth of answer require . Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or no . Partial credit may be given. Therefore. ANSWER ALL PARTS OF THE DUESTION AND DO NOT LEAVE ANY ANSWER FLAN . If parts of the examination are not clear as to intent, ask questions of the egamingt onl . You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examinat io Thi s must be done after the examination has been complete '

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

(1) Exam questions on to (2) Exam aids - figures, tables, et (3) Answer pages including figures which are part of the answe Turn in your copy of the examination and all pages used to answer the examination question Turn in all scrap paper and the balance of the paper that you did not use for answering the question Leave the examination area, as defined by the examine If after leaving, you are found in t5is area while the examination is still in progress, your license may be denied or revoke .

_

51 __IHggBy_gE_Nyg6E88_EgWEB_PL@NI_QEgBBIlgN 2 _ELylpS 3 _8NQ PAGE 2 IbEBU99YN@dig QUESTION 5.01 (3.00) Following an automatic initiation of LPCI at a reactor pressure of 350 psig, reactor pressure decreases to 100 psig. For each of the parameters listed below, determine any change (i . e. 2ncrease, decrease, or rema2n the same). BRIEFLY EXPLAIN why the parameter changes or remains the sam , LPCI injection flow E1.03

t ! LPCI pump discharge head ( assume constart NPSH ) [1.O] e, i, LPCI pump power requirements .

01.03 't j. /, f

..

o QUESTION 5.02 (2.25) ,.

    -
       !

The react or 19 operat i ng at 7b*/. power when tnq EPR system ,' " power 1s Iost. How would the f011owing param.7ters

1NITIALLY change and WHv'? . ~: <

    .. .
          /q A. Reactor prenture    " , ,[.

_

         (O.75)

f B. Core flow ,- (* (0.75) ' C. Reactnr power ,- ,J Yu.75)

      /
      ~,m   . , . .
        .
          - x:;

QUESTION 5.03 (1.005 T

       ,
          ,,

The reactor scrams afte'r cperation at high power for a long tim Station management has determ.nnd that the plant sha!.T be cooled down as qui ckly as possible. What are three factors that " will contribute to a change in,'De SDM during and after c ool rjom? "f 41.0)

  ,
 .

k

d

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 (*4$$4 CA1FGORY 05 CONTINUE'.) DN I C T PAGF A****>      -
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L,___INEgBy_gE_ NUCLE 68_EgWEB_E66NI_QEE86Il982_ELUIDS _@ND 1 PAGE 3 IHESdggyN801CS DUESTION 5.04 .(2. 25) Consider two control rods. Both rods are at notch position 1 Rod A is located near the center of the core and rod B is located at the core edge. The reactor scrams after operating at high power for a long time. A hot startup was performed and power reached 10% about ten hours after the scram. To add the most reactivity ( at this time with a one notch withdrawl, WHICH rod would you choose and WHY? (1.25) Would a f ully inserted control rod have greater differential worth if it was next to a fully withdrawn control rod or next to a fully inserted control rod.? Explain your answe NOTE: Assume average cor er flux is constan (1.00)

 { ,  t OudSTfDN 5.05  (3.00)
' ,I.Yd2cate whether the fol1ow)ng wi11 INCREASE or DECREAEE reactivity
 ')Whing operation AND briefly EXPLAIN wh Moder ator ten.perature i nc r eases while below saturation temperatur (.75)
,
, F,uM - t emper atur e increase (.75) Lord of a f eedwatter heate (.75) A sudden reducf6nn in reactor primary system pressur (.75)
 , s
    '

..v . P QUESTION, 5.06 (1.50)

< '

What are ths three most limiting transients, at Millstone Unit 1, for MCPR consideraticn (1.5)

    !
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h

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,
   ,,

S< '

l

k , ljj

-,
,

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

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5:__IHEgBy_gE_NUg6E88_EgWEB_EL@NI_gEEBSIlgNz_ELUlpS 3_8NQ PAGE 4 l IMEBdQQyN8 DIGS j

QUESTION 5.07 (:2. 00 ) During your shift at full power , Power is reduced using control rods. If the Recirt. pumps remain at constant ) speed willscore flow change ? If so in what direction, l and why 7 If not why not ? (1.00) At low power operation , with the Recirc. pumps at minimum speed, power is increased with control rod withdrawal. Will core flow increase, decrease, or remain constant ? Explain your answe (1.00) l l l

QUESTION 5.08 (3.00)

You are operating under accident conditions per the EOP' l Onta of your main objectivos is to assure " adequate core colling". Etate the three (3) available methods, in their order at preference, to assure core cooling and briefly explain whv this order is preferre I

      ! How can each method of adequate core cooling be verifled?
      (3.0)

QUESTION s 5.09 (2.00) The reactor is operating at 100% power and flow. Condenser l Circulation water temperature now increases 10 deg F over a l

very short period of tim J l a) How would this affect condenser vaccum? Explai (1.0) b) How would this affect reactor power? Explai (1.0) .,

N 1 l , l

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s l l l l

5 __IHEggy_gE_NUCLEGB_EgWEB_EL8NI_QEEBBIl9dx_E69195 z_8ND PAGE 5 IHEBdgDyd8dlCS .OUESTION 5.10 (3.00) For each of the following events, or changes in the plant status, state wether the change will bring the recirc pumps CLOSER TO, FARTHER FROM, OR HAVE NO EFFECT on the point where the recirc pumps will cavitate. EXPLAIN EAC Vessel water level increas (1.0) Loss of a feedwater heate (1.0) Increase in retirc pump spee (1.0) QUESTION 5.11 (2.0 Cot ti d the samn react i vi t y change result in different periods in the same core? BRIEFLY EXPLAIN, (2.0)

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

6 __EL8NI_SYSIEUS_ DESIGNz _CgNIBg(1_8ND_INSIBUdENISIlgN PAGE 6 OUESTION 6.01 (3.00) The reactor is operating at 407. power with the Feedwater Control System in single element control, and level channel "A" selected for inpu The reference leg i sol ati on valve to the channel "A" NR GEMAC develops a packing lea BRIEFLY EXPLAIN the effects on the following which are caused by the above failure. Assume no operator action, and that the flow through the excess flow check valve does not shut the valve. Include applicable setpoint . ACTUAL RPV water level INDICATED RPV water level on channel "A" FWLC RPS ECES PCIS (6 @ 0.5 ea.)

QUESTION 6.02 (3.00) With regard to Low Pressure Coolant Injection (LPCI) syste What signalt+ cause the i ni ti at i on of LPCI? (1.0) DESCRIBE the interlocks which must be satisfied in order to divert injection from the reactor to containment apray with a LPCI initiation signal presen (2.0) OUESTION 6.03 (1.00) Why is it necessary to maintain the ESW system pressure above LPCI pressure? (1.00)

 (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
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bz__E68NI_SySIEDS_ DESIGN1 _CONIBQL2_8ND_INSIBydENI8IlgN PAGE 7 DUESTION 6.04 (2.00) Concerning a scram on low a i,r pressure in the scram pilot valve air heade What is the purpose of this scram? (1.00) When is this scram bypassed? (1.00)

QUESTION 6.05 (2.00) The plant is at full power and the feedwater flow signal to the recirc system fails to zer How will the speed of BOTH recirc pumps be affected? WHY? (1.0) How will the speeo ni BOTH recirc pumps be a+fected if the speed control 1 er output on A MB set had also failed to zero WHY?

      ~
      (1.0)

OUESTION 6.06 (3.00) With the plant operating at 100 7. (st eady state) power, the isolation condenner automatically initiates from a spurious signa ASSUME normal conditions up to this poin What three (3) isolation condenser system valves changed position and to what position (3.e. open. closed. throttle)? (U.75) The " ISOLATION CONDENSER VENT HIGH RADIATION" alarm has now alarme What event was the probable cause of this alarm and where in the control room can you read the vent line radiation levels? (2 panels) (0.75) What are the sources of makeup to the isolation condenser shell, what is the order of preference and why is this order preferred? (1.5) i (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)

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i s c __E(@NI_EygIEdS_QgglGN z _QON16g(z_@NQ_1N@lBydENl@IlON PAGE 8 OUESTION 6.07 (1.00) MCC-E5 has been deenergized due to an electrical faul This has a direct and an i n d'i r ec t effect on the station batteries. What are these effects? (1.0)

QUESTION 6.08 (2.50) Why are the diesel generator loads designed to energize sequentially? (1.0) You have just recieved the " GAS TURBlNE NOT READY FOR AUTO STAR ~I" alarm. What are five (S) possible causes? (1.5) OUESTION 6.09 (2.00) CRD cooling water flow is normally 75 gpm. What i s cooling water flow just after a SCRAM? Explain WH (2.0) OUESTION 6.10 (2.SU) While moving the refuel platform in t he rever se direction the platform stops. Assuming there are no power losses, list all conditions that could cause the refuel platform to stop? (2.5)

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6:__E66N1_@XSIEMS_QESIGNz_CQNIggLt_6UQ_INSIEUMENIBIlgN PAGE 4 OUESTION 6.11 (2.00) Consider all the following i.nf ormati on:

- the reactor is at 100% power
- APRM CHANNEL 1 is reading 102%
- FLOW SIGNAL CONVERTER UNIT 1 output is 90%
- FLOW SIGNAL CONVERTER UNIT 2 output is 102%
- 3 LPRM signals to APRM CHANNEL 1 are bypassed CHOOSE which of the following statements IS/ARE correct:
       ( RPS CHANNEL A tripped (1/2 scram)     - RPS CHANNEL B tripped (1/2 scram) Control Rod withdrawl block APRM HI-HI FLUX /INOP alarm APRM HI FLUX alar m CHANNEL A APRM TRIP SETDOWN alarm     ,

j FLOW BIAS OFF NORMAL. alarm (2.0; l l OUESTION 6.12 (1.00) The ATWS system innerts neoative reactivi ty by perf or ming two specific actions. What are these actions and what specific components operate to accomplish these actions? (1.0)

 (***** END OF CATEGORY 06 *****)
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2z-_BBQGEQQBES_ _NQBd@Lt_@@NQBd@Lx_EMEBGENQy_8ND PAC,E 10 BGD1960SICGL_CQNIBQL QUESTION 7.01 (3.00) OP-206, Plant Cooldown to Co'Id Shutdown, cautions that when both reactor recirculation pumps are off, reactor vessel level must be maintained above + 50 inche a. What is the reason for this precaution? (1.00) b. What are the additional actions, requirements, or conditions the operator must consider if level cannot be mai ntained above

+50 inches? (Five in total)    (2.00)

QUESTION 7.02 (2.00) During startup, the minimum admini strative positive period is (1) _____.__-.._ Reactor period may be calculated by multiplying by 10. S t he tine an (2) ________ required to increase reactor power by (3) _____ (iil1 in the blanks on answer sheet) (1.O) During plant heatup, WHY are you are cautioned against -operation of the mechanical vaccum pump when the reac t.or is above 5% thermal power ? (1.0) QUESTION 7.03 (2.00) ONP-506 " TOTAL LOSS OF STATION 125 VDC FOWER" has specific instructions to shut down the reactor or to stabalize conditions if a scram were to occur. Why do the actsons for this emergency plant shutdown differ from the actions of ONP-502 " EMERGENCY PLANT SHUTDOWN" ? ( (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)

Zz__P8ggg"UBES_ _NgSd6Lg_9BNgBd6Lx_EDEBGENgy_6Ng PAGE 11 689196991GBL_G.nNIBg6 OUESTION 7.04 (3.00) The plant is operating at 100 % power when you experience a loss of 345 kv transmission capability. The automatic actions that will occur are: 1. Select rod insert (0.75) 2. APRM high flux setdown (0.75) 3. Bypass valves will open (0.75) 4. The turbine control and intercept valves will throttle closed (0.75) Briefly explain the reason for each of the above actions.

QUESTION 7.05 (2.00) While removing a fuel bundle f r om the core, the grapple fails l releasing the fuel bundle. Per UNP-519 " DROPPED FUEL BUNDLE": What immediate actions shot t i d be taken? WHYi (1.5) If refuel radiation levels reach 100 mr/hr. what automatic actiont, will occurt' (0,5) DUESTION 7.06 (2.50) The reactor is operating at r ated conditions when you rec 2 eve the " REACTOR BUILDINb COOLING WATER PUMPS DISCHARGE PRES 5URE LOW" and the " REACTOR DUILDIND LOOLINb WA1ER SURGE TANK LEVEL LOW" alarm The off-normal procedure has three (3) immediate actions and five (5) immediate actions which are performed if conditions do NOT improve. What are these eight (8) immediate actions 9 (2.0) As a subsequent action the procedure directs you to isolate the (reactor water) clean-up system if not isolated. WHY? (0.5)

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

Z -_EBQGEQQBES_ _UQEU6Lt_@@UQBU@Lt_EUEEGEUGY_6ND PAGE 12 BOD 10LO51GOL_G991606 OUESTION 7.07 (2.00) Concerning the EMERGENCY OPERATING PROCEDURES (EOP's), define the following: a. stable conditions ( 0. 5 ) independant indications (0.5) c. concurrently (0.5) d. lineup for injection (0.5) OUESTION '7.00 (2.50) For each of the f ol l owi ng condi t i ons, determine wether or not emergency procedure entry tu required. If entry is required state which procedure (r> to enter. ]f no entry is required state "none". (2.5)

******* CONSIDER EACH ITEM SEPARATELY *********
******* ASSUME NO ADDITIONAL CONDITIONS ********* RFV level is to inches Reactor power is 12% GTARTUP MUDE Reactor power is 93'/. one minute after load reject Power operat2ans, GROUP 1 isolation occurs Supprension pool level is 13.5 feet )r yuel l pressure is 2.D pnig Suppression pont level is 13.7 feet Suppression pool temperature is 95 F i. Reactor shutdown, reactor pressure 1090 peig j. Drywell temperature 2s 160 F QUESTION 7.09 (2.00)

OP-329 "STANDT4Y GAS TREATMENT" has a caution which states

"Do not attempt to cool down both standby gas treatment trains at the same time". Why is the operator cautioned about this? (2.0)

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

l i L

_ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ , l l 22. _ _EBOGE D USE S _;,_U08d863, _,@E NOEU GL 3..,EME_BGEN G Y _A N D PAGE 13 BOD 10 LOO 1G66_CONIS06 OUESTION 7.10 (2.00) Regarding precautuons in the system operating procedures, Answer the f ol lowing TRUE or FALS NOTE: ALL OF THE STATEMENT MUST BE TRUE TO BE CONSIDERED TRU To protect a secured and isolated recirculation pump, ,g the pump seal supply must be secure (O at If a recirculation pump is inadvertantly tripped during power operations, two consecutive restart trys may be p attempte ( O'.,/> CRD pump opera in shall be maintained within t h e. following limits: maximum at temperature 150 manimum Drive Water temperature 150 g,g,% f /g ( U .)d Af ter a Core Sor m system keep f211 line has teen i sol at ed , its coro spray headsar must be vented to prevent damage to ,dt piping an hanger s Whr n the c or t' spray pump 25 starte ( O.pd DUESTION 7.11 (?.00) You discover a +1re wh11e on a p1 ant tour, per ONP-505 FIRE, what intormation (4 items) do you report to the control room? (1.0) What 2n t he mi ni miin numtwr of onni te per sonnel r e clu i r ed for the site fare brigadoP Inctude in your answer the restrictions associated with the stii f t operations cre (1.0)

        .
  (***** END OF CATEGORY 07 **$$$)

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. e m __eQMlN1GIB911ME_EBgGEDUBES z _GQNQlligN@t_6NQ_LldlI6IlgN@ PAGE 14 QUESTION B.01 (1.50) Concerning EPIP 4112 INCIDENT COMMUNICATIONS: Noti f i cati on of- the state is required within (1) _________ (a time) of the initiating event and within (2) _____________

  (a time) of event (3) ______________ (1.0) Notification of the NRC will be made within __________ (a time). ( 0. 5 )

OUESTION 8.02 (3.00) Concern 1 rig ACP-DA-9. 02 ( Station Surveillance Prooram ); What is the specified allowable time interval for the following test frequency . Dail y Surveill ance Tents (0,5) Monthly Surveillance Tests (0.5) Semiannual Surveillance Tests (v.5) What is the maxisrum combined interval time for three (3) consecutive surveil 1ance tecats 7 (O. 5) During your shitt an LCO occurs that requires the Core Spray system to be operable. Brief l y explai n What met hods are available to verify the system operabl e ? (1.0)

 (***** CATEGORY 08 CONTINUED ON NEXT PAGE 44***>

- _ _ _ . _ . L._ _0?b] NISIB811E E60G gpyBES2_GQUQ1IlgNL , 8N D_L IMII8IlgNE PAGE IS OUESTION 8.03 (3.00) Concerning procedur e ACP. -QA 2.06D Stati on Bypass / Jumper Control: List two methods to independantly verify a jumper is properly installed or remove (0. 5 ) An exception to performing an independant verification may be authorized under certain condition . WHO can grant this exception? (0.5) 2. For WHAT conditions may it be granted? (0.5) 3. Where shall this exception be documented? (0.5) Which of the f ollowing would be controlled by thi s procedure: (1.0) 1. gagging recirc MG lube oil relief valves 2. install ation of temporary shielding on RWCU pipino 3. plastic houe from i nter tumen t rac k drain to floor drain removal of LPRM t i rcui t. card for diagnostic test in shop i n stal 1 a t i on of ncaft01 ding ior upcomming repair job QUESTION B.04 (2.00) During a survot11ance cal 1bration test of the suppression chamber to reactor b u i l d i n t,i Vacuum Breakern you are informed that the "DRYWELL VACCUM RELIEF 14 UPEN' alarm will not clear, and the red and groen valve posit 2on indication light.s are li Using the attached copy of Tech Specs, answer the followin Ref erence any nect lons uue NOTE: PLANT in operating at 037. powo Are Pri mary Contai nment integrity requirements met? Justify your answer in terms of the TECH SPEC definitton (1.0) Can the plant continue to operate. If yes, under what conditions; if no, why not 7 (1.0)

 (***t* CATEGORY OH CONTINUED ON NEXT PAGE *****)
.. .

i . Oz__0DblN13188IIVg_BBgCgDUBE$2_gGUD111gNSg_@ND_LIUJIGI1995 PAGE 16 I l OUESTION 8.05 (1.50) Concerning FEDERAL radiation exposure control limits: An individual has a current NRC form 4 on file, he is 45 years 01d, his 1ifetime whole body exposure is 131 REM, and it is Jan. . What is his al1owable whole body exposure for the first quarter? (0,5) 2. What is his al1owable whole body exposure for the year? (0.5) What is the al1owable whole body exposure per quarter for an individual (45 years old) who does not have a current NRC form 4 nn file? (0,5) DUES' LION U.06 (l.50) What er e three (3) atems that you bhould chetF in a%ure that a procredure is a valid procedur6e before using tt to oporate plant equipmentP' (1.5) OUESTION U.07 (2.SO) Uutng the attached copy of Unit l's FJmer g enc y Act1on Leveln. r1 anni 4y the following events. Constdor each event independentl Main Steam Iine A has ruptured in the reactor building. MSIV's in C main stuam Inne are stuck ope (0,5) Severe winter ice storm on site, one source of offsite power los (0.5) Malfunction in radwaste facility resulting in release rate which corresponds to 2 REM /Hr. whole body at offuite location of maximum dos (0.5) Fuel bundl e droppted, refuel floor ARM reading 100 mr/h (0.5) Injured man on refuel floo (O.S)

 (***** CATEGORY 00 CONTINUED ON NEXT PAGE *****)

- .-.

,, .

Ot__60dlulGIEGIIME EBOGEQUBES t _GQUQlIlgUSt_6NQ_61dlIGIlONS PAGE 17 OUESTION U.OR (2.00) List the shift manning requirements (per tech specs) for the following situation , Shutdown Condition with coro alterationt in progres (1.O) Shutdown with avorago coolant temperature greater than 212 (1.0) DUESTION O.09 (2.50) The r eactor i t, at 207. power and containment anarting is in progress. The reactor oporator requests that PS-1621 A through D be manon)1y i sol at ett in ordFr to prevent a high drywell p r ene.i t r e ret r a m . Ut4 i ng the a t t ac hte rl 'l oc hn i c a l Specifications dotermino if these r ornponont u may b+ t ual at ed. Just i f y your answer end reierence tha noc ti onn unod t.s devel nn voor annwo **** Nu1E r tlNf:T IUNS ff F b- 1621 h - D Md. ATTACHID AS APikNDI) A 4444 (2.b) UUE*; TION 8.10 (2.50)

't he reettor in at ', u '4 power dur) ro; the f1rst stertup after a refueling outago. Tiu? hoactor Engineer informa you that t he reanc tor curront1y has ten (10) tror e iul1y withdrawn contrel rodn then his calculationu show aro re> quired for this pcwor lovel. 15 any action r n qui r eti? If so what, if not, why not ? NulEt Tech Speth anti Control Rod Wor th cur vian are attacherj. Refefenco any sections used to develop your annwer and SHOW 4LL- WOM . .    (2.b)
  ($$44* CAiliGURY 00 CONT 1NUED ON NEXi PAGE ***4$)

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e ,._,.epululs. lEGI1E _EBOG!iU Uh th _GO U Q 11 !U Uh _ .eU Q _L ldllOllO U S PAGE 18 OUESTION O.11 (3.00) During a PEFUELING OUTAGE fuel movement is in progress and it is determined that the emergency diesel generator in inoperabl Using the attached Tech Specs, answer the following and reference any sections used to develop your answe What actions, if any, apply to the low pressure ECCS systems? (1.0) Can refueling continue? WHY7 (1.CO What actions, if any, opply to the low pressure ECCS systems if the Emer. D/G becamo inop during startup at 3'/. power 7 (1.0)

      [
 ($4444 LND Uf- LHiltMIRY OH 44444)
 (4444$44444444 l N D Ol' EXAMINAllON 4*4644444444444)

-

..

., Dz__IHEQQy_QE_MUGLE@B_EQWEB_EL@UI_QEgh6IlQUt_E(QlD@t_6NQ PAGE 19 IHEBtjQQYU@t!1GS ANSWERS -- MILLSTONE 1 -86/12/15-HOWE, ANSWER 5.01 (3.00) increase [0.253 as the pressure of the system decreases the flow increases due to the centrifugal pump head / flow characteristic. [0.753 decrease [O.253, as the pressure of the system decreases the operating point on the pump characteristic curve is shifte.1 to a lower pump discharge preusure. [0.753 increano [0.251. from the pump charactersstic curve at the flow (capata t y) inc r eases t he power requirements al no increaura. [0.751 REFEF CNCE G.E. Heat Transfer and Fluid Flow. Ch. 6 pg. 6-95 6 6-9 SLil 6-1 VA 291004 K1.OS-2.9 ANSWER 5.02 (2.25) A. IncrensentO.26)due ta the control valves going shut i n remponse to the NFR boicoming tha controlling signal. (U.5) H. I n c r ea e rs ( 0. 25 ) ciur-. to ihr rednrtion in t he vold cont ent of the two phaso mixture in the core. (0.5) C. Inc r uabou (0. 25) due to the collapse of voids from the higher presnure which adds pouttive reacttvit (O.S) REFERENCE PNPS LP-MHC, Mechanical Hydraulic Control System, pg. MHC-10-1 G.E. Reactor Theory, ch. 4, pg. 4-24 SLO-4. G.E. Heat Transfer and Fluid flow, ch. 8, pg. 0-41. SLO-0. MILLSTONE LP 1314 P.121. 167 SLO-29al,177, 178, 17 KA 241000 N1.01-3.9,1.02 4.1,3.01-4.1, 3.02- . _ _ . . _ _ _ _

, .

Dz__1BE98Y_0E_UUGLE6B_E0 WEB _ELGUI_9EEBBIl002_ELulD@z_6ND PAGE 20 IUEBdQpyNSdJGS ANSWERS -- MILLSTONE 1 -86/12/15-HOWE, A.

ANSWER 5.03 (1.00) 1. fission product poisons ( may break down to Sm & Xo) LO.331 2. moderator temperature coefficient LO.33] 3. fuel temperature coefficient CO.33]

(Other factors will be considered if justified)
. REFERENCE G.E. Reactor Theory, ch.7. pg. 7-6. SLO-6.2.5.4, 4.2.3, 4. KA 292006 K1.12-2.0. KA 292004 K1.01-3.2, K1.05-2.9 ANSWER 5.04 (2.25) Hod 14 (0.2S) Upon t,crati. recovery. fistion produrt p oi son e, c a u r,"

a severe iluu dopresnton in what was the highest p o wer- producing regiori of the c or e. 1his results in a hight'r relative 41ux in regions of low poison concentration. Thoso shifts in the flux distribution increase the worth of peripherial rods and decreaso the worth of thoost in the centor of the core.(1.oO] The withdrawn rodIO.251 Flu:t is higher in this ares, thuc rod wnrth in greater.[0.75) REFERENCE G.E. Reactor l heor y, ch.5 p H,Y,10, ch. 6 pg. 12. SLO-5. KA 292005 F1.09- N1.12-2.9 ANSWER 5.05 (3.00) Adds nngative reactivity LO.2b] due to the increase in neutron 1cakage - Moderator temperature confficinnt. [0.503 Addu negative reactivity [0.253 dun to the increase in neutron capture in the funi - Doppler coefficient. [0.503 Addu positive reactivity 00.253 due to the decrease in noutron luakage - Moderator temperature confficient. [0.50] Adds negative reactivity LO.253 due to the increase in neutron leakago - Void coefficient. CO.50 REFERENCE G.E. Reactor Theory ch. 4 pg. U, 9, 16, 24. 34 35, 37. SLU-4.1 10 KA 292000 K1.11-3.G, K1.22-3.6, KA 292004 1.01-3.2,1.05-2.9 1.10- __ - _ - ___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ - . __ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ . - _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - - _ _ _ _ .

 .  ,

b.__10E98Y_0E_NUQLEQB_EQWE6_E(@NI_DEE8611QN t _E(ylDEg_6ND PAGE 21

   .IUEBU991N@d[QQ

. ANSWERS -- MILLSTONE-1 -86/12/15-HOWE, , ANSWER 5.06 (1.50) , Tr E . . 1 : ; ;- - generator 1 cad rejection without bypass. ( O.50 ) 2, TJrbis t }r,'o w .he.sk bf)**En

3/. Loss of FW heating. I ~ . " '

!  d,/. Feedwater controller failure ( maximum demand ) C        .S i   5, nesta fle.a co se a.< M y war ~ fle.a to 4 !** s .             )

REFERENCE ( f ,,4/, d 0,is de'/ a g Se.p 5 M 0.>. e.

1 GE Heat Transfer and Fluid Flow pg. 9-35 to 9-37. SLO KA 293009 K1.19-3.6. KA 295014 AK 1.05- ,

ANSWER D.07 (2.00) , i T

                  ' Core fIow wi)1 i ncresse rhte to ) ess t wo phat-e f 1ow resi nt erv e. ().001

! Core flow will increase due to greater natural circulation. (1.00)

REFERENCE r UE Heat Transfer and Fluid Flow pg. B-40 to 8-49. GLO-9.1, KA 293008 K1.28-2.S. K1.29-3.0.

4 ANSWER 5.OB (3.00) n IN ORDER CF PREFERENCE Core submergenceEO.333 verify by 2 means of level indication that level is above TAF E0.5] Spray coolingtO.333 verify by one core spray system operating at or above design conditions E0.53 , Steam cooling CO.33] verify by use of steam flow over fuel j and out of vessel via SRV'u as per EOP (cannot be verified

by direct indication) 00.53 , i t This ordnr preferred because this is the order from highest to lowest heat transfer coefficient CO.53 (NOTE: Reasons for preferenen given in Appendix B of the EOP'S

will be const dered f or full credit) ! a I ? l '

a

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

-- -

. ;

Uz__IHE9By_gE_UUCLE68_EgWEB_EL991_gEEB811ggi_ELg198 t_6ND PAGE 22 IHEBdgDYN9 digs ANSWERS -- MILLSTONE 1 -86/12/15-HOWE, . REFERENCE JAFNPP LP MIT 301.4 pg 15-18. SLO 1.03, 1.0 MILLSTONE LP 1500 D p. 26; APPENDIX B EOP's p.3-1 to 3- KA 295031 EK 1.01-4.7, 3.02-4.7, 3.03-4.4, 3.04-4.3, EA 2.04- ANSWER 5.09 (2.00) a) Decrease - the condenser works as a saturated system thus if cooling media temp increases, Tsat increases and Psat increasen thus vaccum decreases (1.0) b) Decrease - temperature of condensate /feedwater would increase and inaert negative reactivity due to moderator temp. coeffic1en (1.O) REFERENCE J4FNPP NET 237.4 pn.13 H228 G p H226.0 fig. 3-4, MET 222.9 po.22 GL O 237. 4. 3. 2. 229.O.1.19, 222.9.1 : Li . E . HEAT TH4NSFEh AND FLUID FLOW to 7-45' SLO- 7.o.4. ' 7. * 7.4.2, 7. KA 293007 K 1.06-2.8, 1.00-5.1; KH 29100t2 K 1.08-3.O, 1.10-2.8: KA 293008 K 1.09-2.7, 1.30- ANSWER 5.10 (3.00) Farther from cavitation (0.5). As the water l evel i ner eases, the static head of water component in NPSH determination is also increasing adding NPSH (0.S). Farther f rom cavitati on (0,5). When feedwater heating is lost inlet subcooling increases (inlet temp. . decreasing) which brings the water farther from naturation.(0.5) Closer t o cavitation (0,5). As pump speed increases the pressure in the eye of the impeller decreases, which will cause cavitation earlier for the same NPSH. (required NPSH increases) (0.5)

( Al ternate answer : Increased flow increases power and feedwater flow thus subcooling increases and this provides more NPSH)

REFERENCE JAFNPP LP MET-214.9 pg. 14-16. SLO 9.11, 9.1 G.E. HEAT TRANSFER AND FLUID FLOW P. 6-76 to 6-81, SLO-6.10.5. 6.1 KA 293006 K1.10-2.8, VA 291004 K1.06-3.3. KA 202001 K1.03- . . .

- _ _ _ ._ _____ _____ _ _ _ _ _ _ _ . _____-_ _ _ _  _ _ - _ _ _ _ _ _ _ _ _ - _ _

. . Dz__ISE98Y_DE_UUCLE88_EQWEB_EL891_9EEB811gN3_ELUIDS1_8Up PAGE 23 IbEBOODYUBUIGS ANSWERS -- MILLSTONE 1 -86/12/15-HOWE, A.

ANSWER 5.11 (2.00) YES.CO.53 The period depends upon Beto and reactivity.[0.5] Beta changen over cycle life, so the samo re. activity change produces one p=riod at BOC and a shorter period at EOC.E1.03 e REFERENCE G.E. REACTOR THEORY P. 3-29, 3-30, 3-34, SLO-3.3.5, 3. KA 292003 K 1.06- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _.

. 6. __eLeNI_@fSIE,t1S _DE_glGh_ CON 1R063,_6UtLINSIRUr1EU16110N

.        PAGE 24 ANSWERS -- MILLSTONE 1  -86/12/15-HOWE, A.

ANSWER 6.01 (3.00) Actual vessel level is decreasin (0.'u Lovnl channel "A" wt11 indicato increasing water leve (O.S) FWLCS will close the FRVs ta try to maintain leve (0.b) Reactor will scram G) +8" due to low reactor water leve (0.5) IC and FWCI initiation .D -48". (0.5) PCIS Group 2 and 3 isolatiens +8" also PCIS Group 1 isolation, "" _d r ' "' : ;-t

    '
    . 9 '"".   (0.$)

REFERENCE DC Reactor Level and Pressur e. LIC-0263, pp. 12 Millstone LP 1316 pg. 33-36: 1400 p.). 19.5LO-15, 1334 pg.7, SLO-6: 1311 A pg . 49- b U pg. l f KA 216000 L1.12 3.7: LA 259002 6 1.01-3. 9 K1.03 ". 9. 65. 00-3.1: KA259001 L1.OH-3.7.(d.07-3.0; ll(4 c23062 A2.US-3.6.

ANSWER 6.02 (3.00) H1 drywell prr"mure .O puig flh (.5 L.o Lo let water levol -4b" anti l o w- then t,Su oni g im pr essuro ( . 5) (1.0) With an auto inst 1ation sional pr omen t : 1. Cont t,pr av ist key 2o overr2cie (0.5) Voc c,o l water level tu creator than 2/3 corc= hot oht (0.5) UR Loni snr av 2nd Lev is in mar omrrido (0.*,) Drywell pressure munt be greater than 5 psig (0.5) REFERENCE LPCI LP 1335 pg.19, 29, SLO-20, 29, 30, 31, 32. KA 203000 F4.01- FA 226001 K4.93-3.1 ANSWER 6.03 (1.00) To prevent an unmonitored radioactive rel eaue from the LPCI HX (1.00) REFERENCE Malistone pg.10 SLO-13 KA226001 4.11- r , i 61__ELAUI_SySIEtlS_DED19Nz _990I6963_8Up_1NSIbgt%N161JON PAGE 25 ANLWERS -- MILLSTONE 1 -86/12/15-HOWE, A.

ANSWER 6.04 (2.00) To eliminate the conr.2equence ci e slow scram ( random rod insertion ) upon loss of air header proscure to the scram valves . (1.00) Mode switth in shutdown / refuel and the 5.D.V. bypans switch in bypas (1.00) REFERENCE Millstone L.P. 1400 pg. 35-3 SLO- 16, 29 30 KA 212000 K1.15-3.9 A?.10- ANSWER 6.0% (..Uus both p i e rc.p 9 will r unbact t V i r,. t i i . '. '. ) bre at me, t hea fW vinw eu gnal to the rectrc ovu u mi in less than '.? O (u. 6 /) (tho flow 12 mater 19 not b ypa s swd when thab signal as 1 s u ts t han 20'/. ) Ei pump runo bat i a s. tn p / > r i- a ii'.2b). A pump spend roowtnw tho l n nr#,e ( o. M. ) brc auw of a st o.)p tube inci<up cau wd by a l o n 's of speed signa (U.b) RF FE F, F N:F Mi l l nt onr. '.0111 p I bl U- 3 7 Lic , 1 K A20.2002 Lh. U4 -3. S . ) 4. n?--3. 0 6.7- ANSWER 6.06 (3.00) IC-3, open, IC-6 and 10-7. nhu (0.25 each correct combination) Indication- panol% 910 .and 902 (0.125 each) probable event- a U-tube lent (0.5)

'  *-  . L'"'  mt'-F*~- FIRE water- no cont ami nation, pa la*M*/ JWMr CONDENGATE transfer- han contamination    ( each)

! REFERENCE Mallatone L.P. # 1306 pg.2, 0, 14. GLO-3. 4, 5, O, 9c, 14e KA207000 V4.02-4.2. A2.02-4.7. A1.07-3.7, Kl.04-3.0. E1.0",- Lt.06- _ _ _ __ ___ _ _____ o 6:.__ELGUI 0XGIEUG_DE21h_EQUIEQLa._8UD_.10SIEudEUI61100 FAGE 2o ANSWERU -- MILISTONE 1 -86/12/1trHOWE. A.

ANSWER 6.07 (1.00) Deent 91; co the battery charger (101D) (0.5) and the battery room e:t h e tis t fa (0. 5 ) REFERENCE MILLSTONE LP 1341 p.17, SLO-17: LP 1344 . GLO-9 UA 262001 K3.03-3.2. K6.01-3.4 ANSWER 6. Ofi (2.S0) The D/G is e small power nitppl y and nequential loatli ng pr nt t.c t r t h r* genstrator f rom bet rio overlaatled (from u t a r t. i n g currentn) and trippin ( 1. D ) t l or: Lc ou t r e l .a v e not r esot

* loon tit 14. t on t r o n i o i he, tilb c t oit r ( > l rL * muslo ownteh to (H f 4 l t )'it; (#f r erli t l <t t t '(i f d. { #r@1 4 ovorupped of GIG or power turbino 4 ftre
* OTh blr. n t .1 i si e,p or nt irig r esni t s e in 6 i n% of DC c on t.r ol p ne' ;, t o ip.t.. bl (fr y test it orle n wt1.t h 1 ti oft normal (b r eq cl. e o.3 noth)

P f + f.h E NCli M I L L'i IL 6NI. l.P 13 M4 p. 4 bl ()- M i s i l' 1 T.W p . Str . bl.O -40 En 264000 6:S. 06-? . b 4 -;'. 6 ANSWER 6. IY? (2.OU) , m.o) Onolino water ()ow would be rprO YO ). b) boraune al1 t he Of<D r.yut em flow in darnctud to rochargo thee .a c c umul a t or u . This flow pausen throuoh thp f Iow plement cont r elIino t he i1ow contrel valve and menwed fIow Iw gruator than the mot flow t hou tho flow control valvn will clown and stop cooling water flow. (1.5) URADING NOtt'S Muut idont i f y act ual fIow, undormtend thn condttion of t ho accumul at or u. and be awer u of the phyuntal plp1nq arrangement cnd oporatton of thp fIow contral valv RCFCHENCY MillHTONE lF' 130',' p.46 , Ul. l b '.1. 6n ','01001 n 1. t i l - J . */ . i 4 .1,'- P . Y

_

6.i.__EL69I_SXDIEUS DES 1092 9991But-2 6NQ,10316UDEUlQ1190 PAGE 2/ ANSWERG -- MILLSIONE 1 -86/12/15-HOWE, ANSWEH 6.10 (2.50) 1. main grapple not full tip / > load (500H) on grapple load cell (0.33) OR load (400H) on trolley hotut loact cell (0.33) OR load (400H) on frame hosst innd coil (0.33) AND bridon over vonsul and ono rod withdrawn (0.5) brid 0o over venswl And enode swi tc h in startup/hnt standby (1.0)

l i REFERENCE MI LI.tiTONE pr oc . OP-320A p. */ , 10. LA 234000 Lt.04-3.6. L 4. 01 -4.1, f >7- 3. '/ . t AlEWEk 6.11 (W.v0s f [ . pr in)1 e r enn t $. ii. sv inr each etrrertiv or ro inc t tui chroc e, e t r e t r 1

l l bEF EhliNCI' MIL LiiiONI; i t' 14044 p.it 11. Ul f1- . S. 6 , U n 2 1 5 0 u 5 l a . oe, #,./.. hl.tH-4.I, let- '. . n i f

l r ANGWIH 6.Id (l. 6+>i i Altperudo tnd t oi r r t i nt ( O . . *2 7 tw opontho AlWh t,rr em vol vet. (i it d l kur.1re ptimp trip u t . Z Ti by tu mt i ng t h a' 11 0 tiold bro.4or ( O. No l l I4:FERENt.11 ! MILtETONE LP 1465/4 p . ti, t,LU- 3 th 202001 L1.27-4.3. 14.01- L3.v6- [

l l i l l r i I i

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

o 22._ Ebgg!?DDRES_ _ NOM 1/% _ ADNgPt;1(4 g, EtjLf3(j[NgY_ AND PAbE 20 64D1QL9!}ICAL,GONTROL ANGWERS -- MILLSTONC 1 -06/12/15-HOWE. A.

ANSWCH 7.01 (3.00) Maintain level / * Sh inchor, i n enhance nat ural circulation and that provunt thermal u t r a t 1 f I c: a t 1 on and huatup /protssur1:at1 o ( l . i sv) (lhe optarat or muut assume thAt t her mal strattiitation. heatup, and pronnurtzation may acror).

1. I'rininr y and we nottor v c on t ni nn,t nt mu%t be malntnined l en. h c.pocs for ahnvo .212 f: operotton munt bo aburar vai . Peruonnel working on primar y components muut be advised of ponestblo preusur17atio . Mon 1 t or ve+nel peramotert. 4or pr nmpt- 1 ntil c at 1 o n cH hreatu Hun shutdown cool 1su) 1n m.et, fIow. 2 pump %. (O.4 each ior 2.00) h!4 6 F+ NI:E UP -:'o6 hev. 9 po, I ei 9"ih 1 . *. e ". . ".

     . l . . ' t v- l. . '. . 1 . .' / " . < a ANSWtH 7.02 C . Go s o.<>i f.o i or 4 >ntl< (ii.:,). 6Ji . : r non > tii..!i!  . (..i 14 0. ( 0. ;'h )

h. ? > t o n t f i r. a n t. levoit ne hv<u uilon .o ut n'yonn wniil ti he pronent o t he c onderis er , t . '. i ono coul d r et ul t in o detonablo nit a turn of h ydr'n,p*o and ony pen (at atmonphorir pre.n*,nre) O '. , . I( t e .mh e v 4 ein mrs. to nr. cur. the eve nt. w oitl d poun a porunnruil h a .* a r < l to t u e tho p u nip and watrar noparat or nre nrit theni ont ci tn hi- n ott.o.4ti oti t ou t ut ant . Ui. 34) blil FRF f El

Mi l l t.t t,tio t it s Jh l . he v . !! .. pg. / t ' . . (If ~/ QW. ht . . L A ;W4001 i ' l . !!,-3. H . ANGWEh 7.v3 ( ,' ' . 00 ) On a 1 ot1 % of l i , VI)C . remot u e ont r01 + une t i onn ior 41/ O VAC and 4Ho VAC broatstru otr o los O> . 5 ) Aluo the statum inghts for pumpn cnd breakprn ere l on t . (0. !D* T hun. otittt i t onel pornotinel are requi red

.a t local statio#Us to perform .actinnu that would normally bo performed in the contrn! room. (1.0)

k A.elkok ens.aw ts teoMI reo ness ene.lskn are les RETERENCE Milletono OtD ~ teis. p . 1 '. . IA vviitus4 Al.;'.U.'.~3.. #ini.O.- An ! . n 3 .' . s 24 _BBQGEpuBES_ ,,NOBdOL,_OPN96MSL t EDEBgEUG,Y_ONp PAGE 29 BOD.I.gl O{ij g8k ggN1BO ANSWERS -- MILLSTONE 1 -86/12/15-HOWE, ANSWER 7.04 (3.00) 1. FW heattnq is lout and thin rod innertion decreasen reactor power to componnattr for the reactivity adde */ The notdown in to preservo MCPH margins d the select rod inser t M concurront with the lower FW inlot and the higher local power level . The bypass valves open as required to control reactor pressur . Tho control and inturcopt valvon will throttle to prevent a turbine overnpeed from the stored energy in the moisture soparatar (0.75 each) R EF t: f <LNC L till Lh l OtF OlJF -GU?.. IP j f ,0 0.'i . .6

    '
     .'i, blO 21 F A 29bt h fi Ai . 1.01- L . O - 3. ANfiWE h 7. t 85 (2.00) . evaruate thra rotuoi f l oot
. ervatuste t he tir ywt il
,

3. ovacuate the r ear t or tiin 1 <t no 3i a h a oli r adi at i nn al var m vaisto on iho refuol t l oor (U.33 each) l n (n s n i m t .? e r u li ni trin e a o sier e (ti.Di HDGl utertn onij rnactor hutIdtog vtsotilation i sol a t tna (U..9 Onrh) hrFFht NLl1 Ma i 1 %t ono ONt'-519 lP 1 1'. ? 9 p.2-3 tiLO-2, PA 29"aO23 AK 3 .181 - 4 . 3 HA1.0/- (11 0 - 3. 9 .

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Z, _EB0GEutJBEB_:_UGEt!OL&_0000BtiL_EMEO(jFNCY ANp P:,y 'a o _ 8001.0LO0lGel=_GQN1806 ANSWERS -- MILLSTONE 1 ~86/12/ l b -HUWE. . ft.

ANSWER 7.06 (2.50) , . determine the cause of the losa of HDCCW / attempt to tuolate tho l oal- attempt t o 5st ar t / restart any RbCC.J pump s If unable to restore flow 4 run recirc flow t o a minimum and scram 4 TRIP the main turbinn when < 50 MWo 4 shut tho M51V's -

* start the IC
* 5.t .ir t Sb6T nnd vont enntainment (U at 0.2S oach) To ennure i<WCi l r er.i n su ont over hent ec ( 0. b )
       '
-Ot- ~ro AtbucC RBu sa S/nt'" MM L CA k' .   -

fief ERENCE MILLBlONt UNT' -524( . . '. ' . In '.'@ n1H Al1.01-3.6. b'03.S. Ulc . .. ANSWER 7.0/ ( .2. 00 ) Par amur er t. ennut i t ut i tio i fif- ent t y 4, r e - wtthln nr cept ob! s' vnItu", and mem. uron talon to misure i fiat 'tho. 5't t I romein wnt.hin .rcoptablo valuec. (v.b) . I Intlient 1 c nn whi ch i;r e 4 r < un 6unn,woiat!y rin f f erent mour te+ . p r oc o n s.1 t u) anti d t c.p l . v . Mitii at ? 1 tin tho nowaibt11t. at r uinmon flint l 0 lal}ttrt"4 ( I 1. U ) ,

      , At i he teame timo (0.5) Ent abli th n suitabl6 t,ut t i nn put h anti d i . .: b .4 r o t- pot h t n tt.o. <<I Dous not mt.e a n to thtvet. e v. B)

F<EFERENCE MILLSTONE LP 150014 p. 9 10, 22, 76. Gl'. 0 - b . 12, 30. lii ?'/5031 012" . L . ,

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Za_.EBOEEQUBES_ _UOBd661_8BNOBDBL _EMEBGENCY_BNQg PAGE 31 BBD.IQLOGICBL_CONIBOL (\N6WERS -- MILLSTONE 1 -86/12/15-HOWE, , ;

~ 9 ee,rME;4 - '7.00 (2.50)

c. none r, o r.ie i.571, 572 ,, 571. 572 . 570s S71, 572, 580

g. rone

h. 530 t. S70 571, 572 3. 500 (0.25 each)

- REFEFENC M t t.Lii'l ONE FDP- D7 0. 571. ST2. 5HO, LP-1SOOB SLO-2. KA 294001 m 1.16- ANSWER 7 . O'/ (2.09)

Thir would opten ni can, tunnel d,mporr, (1,0) all owing a pat h for port >nt h il unt ant rol l t':1 radioactivo release. (1.0) liFVLHEud ' MILLUT(JNF UP-32*>, LF 1329. GLO-19. KA 261000 G10- ANSWEH 7.10 ( 2. 6KO T

- F  ,

w% T (o b (0,[each) liEFERENCE MILLSTONE OP-301, p. 5, 6; OP-302, p. 4; OP-336, p. VA 201001 010-3.3, LA 202001 G10-3.7. KA 209001 G10- i

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ZL__EBQGE.DUBEE_ _dQBU6bs_6hdQBd6Lt_EME8@ENQY_@NQ PAGE 32 B6Q10L90lG6L_GQNI@gL - ANSWERS -- MILLSTONE 1r -86/12/15-HOWE, ,

Q' _

 ( s
. ANSWER g7.11  (2.00)
'c ocati orv, type of *.ii re, sire, affetted. components    (0.25 each)
' members-(0.25). Shall not include 2 members of'the minimum shift crew required for the safe shutdown of the unit or for other essenttal functions ducirig the fire emergency. (0.75)

REFERENCE MILLSTONE ONP 505 p. 1; T/S p. 6-1. KA 294001 K1.16- * 1 -

t

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t

i V k i * ,, j 6 .

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

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

 .c 8 1 __89MINISTB8IlyE_BBgCEggBES z _CQNplIlgNg2_8Np_LIMIIBIlgNg      PAGE .33 fANSWERS~-- MILLSTONE'1'    --86/12/15-HOWE, s ANSWE .01 (1.50) .2 one hour.- 2. 15 minute . classification   (0.33 each)
 ' one hour;(0.5)

REFERENCE Millstone EPIP.4112 p. 2

ANSWER 8.02 (3.00) . Any hour of_the. day plus or minus six hours, not less than once/ da _2. The scheduled date plus or minus one week, not.less thanL12/ yea . The scheduled-date plus or'minus one month, not less than 2/ yea " l ( O.5 for each )- t f l b..A total maximum combined interval ~ time:for any three ( ;. i l consecutive tests not to exceed 3.25 ti mes the test. I nter val . (1.0) L c. (l') A-. system or component can be assumed operable if the associated surveillance; requirements have been satisfactorily performed within the-specified time interval, the system i s not having maintenance h perf or med and is in an operable statu ( O.75 )

 (2) Perform surveillance test to prove operability (0.25)

REFERENCE- ~ Millstone ACP.-DAa9.02-Station Survie11ance Program . ANSWER 8.03 (3.00)

 'a.~ visual ~(0.25) or functional (0.25)
 . .1. The Shift Supervisor (0.5)

2.-Where the verification would result in significant radiation exposure (0.5) 3.. ATTACHMENT 8.A or JUMPER-LIFTED-LEAD-BYPASS CONTROL SHEET (0.5)

 '
 ' c . ' 1',' 2, 4 (req'd for full credit 1.0)
~ REFERENCE Admin. Procedure ACP.-QA 2.06B Station Bypass / Jumper Control. P.1, 2,     1 'KA 294001 1.01-3.7, 1.02- , - , . . . . . . _ . _ _ _ . . _ _ . .. .. . _ . ..
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Oc__8DdlN1@IB6IlyE_88QGEQQ8E@t_GQNDillQNQt_6NQ_61dlI@IlONS PAGE 34 ANSWERS -- MILLSTONE 1 -86/12/15-HOWE, ANSWER RB.04 (2.00) Containment integrity requirements are not met.(0.5) By T/S definition the containment is not intact and an automatic containment isolation valve is inoperable and not closed. (0.5) T/S 3.7.A.3 applies for containment integrity requirement LCO to initiate an orderly shutdown and have the reactor in a cold shutdown condition in 24 hrs, per 3.7.A.7.(1.0) REFERENCE MILLSTONE T/S 3.7.A.3, 3.7. ANSWER 8.05 (1.50) . 3000 mrnm (0.5) mrem (0.5) mrem (0. 5 ) REFERENCE MILLSTONE Procedure SHP-4902 P.6, KA 294001 1.01-3.7. 1.02- ANSWER S.06 (1.5n) 1. controlled copy 2. approved 3. current revision 4. date of-use is after effective date (3 req' d 9 0.5 each) CAF-REFERENCE Admin. Procedure ACP.-QA 3.02 Station Procedures and Form P.2 KA 294001 1.01-3.7, 1.02-4.5 l

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i __GDMINISIB811ME_BBQCEQQBES x _CQNQlllQN@t_6bp 11MlIGIlgN@ PAGE 35 ANSWERS -- MILLSTON /12/15-HOWE, ANSWER- 8.07 (2.50) General Emergency (0.5)

' None required.however candidate may elect to classify as an unusual even (0.5)- General Emergency'     (0.5) Alert      (0,5)
- None required. Candisfals mer clauily as e s. Unescal sved ll k< (0.5)

, desem> +MF % n,as saas in10 red av M id- sJor$clo The naaseelby e4 +kf> floor is tensider e,/ e,,,4,,,;,y Q ,% % gyak,,,;4 a Marsa , , REFERENCE Millstone'EAL's Form 4701-1 Rev.3. KA294001 A1.16-4.7, KA295017 AK2.06- ANSWER 8.08 (2,00) sot's one dedi cat ed to supervision of core alterations (0.5) 1 O L- (0.25) 1 Non-1itensed operator (0.25)

' b. 2 SOL's (0.25) 2.0L (0.25) 2 NLO (0.25) 1'STA (0.25) REFERENCE Millstone table'6.2.1.

< ANSWER 8.09 (2.50) No (0.5) T/S 3.1 table 3.1.1 (RPS) and 3.2 table 3.2.1 (PCIS Inst.)

allow this action (1.0)-however, it is NOT authortred by table 3.2.2 (ECCS_ Inst.)- (thus i sol ati on would inop the ECCS start signal and the plant would be in a LCO requiring an orderly shutdown to cold condi t i ons. ) (1. 0) REFERENCE MILLSTONE T/S 3.1. table 3.1.1, T/S 3.2 tables 3.2.1 and 3. .-. -,- . . , . . .-.- ..-. . ..

. . 92__GDUINigIS@IlyE_PBgGEDyBES2 _CONQJIlgNg3_@ND_LJdlI@llgNS PAGE 36 ANSWERS -- MILLSTONE 1 -86/12/15-HOWE, A.

ANSWER 8.10 (2.50) ' WORTH = 13 X 10E-4 delta K (from curve) (0.5) Total reactivity = 10 X O.0013 delta K = 0.013 delta K (0.5) , From T/S 3.3.E, the max delta K difference allowed is 1% or 0.01 (0.5) ' Since the total reactivity dif f erence between the actual and calculated rod configuration is over 1% the reactor must be shutdown and the NRC must be notified within 24 hours. (1.0) REFERENCE MILLSTONE T/S 3.3.E ANSWER 8.11 ( 3. 00) a. No acti on required (0.25) 3.5.F.1 is not satisfied but F.2.3.7. and 8 are sati sf i ed. (0. 5) F.O references 3.7.b.4 which is satisfied (O.25) b. Yes.(0.5) The loss of the D/G did not place the plant in an LCO requiring refueling operations to be stopped. (0.5) c. 3.5.F.2 applies to.5) 7 davr, to operate in this condition provided GT and other EN::S and cont. cooling systems operable. (0.5) REFERENCE MILLSTONE T/S 3.5. ' I

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TABLE OF CONTENTS

Surveillance Page N .0 DEFINIT 10NS................................................................. 1-1

SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS

;   2. FUEL CL AD0l NG I NT EGRITY. . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.1. 2. . . . . . . . . . . . . . . . . . . 2-1
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2. REACTOR COOLANT SYSTEM............................. 2.2.2................... 2-6 l LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT

!
' REACTOR PROTECTION SYSTEM /4 1-1 PROTECTIVE INSTRUMENTATION Primary Conta inment I sola tion Funct ions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-1 Emergency Core Cooling Subsystems Actuation............................ 3/4 2-1

, Control Rod Block Actuation............................................ 3/4 2-1 . A i r Ej ec to r O f f- Ga s Sys t em. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-8 j Reactor Building Ventilation isolation and Standby Gas Treatment....... 3/4 2-8

i System Initiation I REACTIVITY CONTROL 4.3 ' Reactivity Limitations...........................A..................... 3/4 3-1 Con t rol Rod s W i thd rawa l . . . . . . . . . . . . . . . . . . . . . . . . . . B . . . . . . . . . . . . . . . . . . . . . 3/4 3-2

Sc ram i n s er t i on T i mes . . . . . . . . . . . . . . . . . . . . . . . . . . . . C . . . . . . . . . . . . . . . . . . . . . 3/4 3-4

! Control Rod Accumulators.........................D..................... 3/4 3-5 . Rea c t i v i ty An oma 1 1 e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E . . . . . . . . . . . . . . . . . . . . . 3/4 3-6 l Shutdown Requirement................................................... 3/4 3-6

Thermal Power - Core F10w.............................................. 3/4 3-6 i

l a )

i Amendment No. 43 December 16,1977

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Jrn. .y 1, 1986 Surveillance Page N Secondary Containment .............. ...................... C ..................... 3/4 7-13 P r i ma ry Con t a i nme n t I s o l a t i on Va lve s . . . . . . . . . . . . . . . . . . . . . . D . . . . . . . . . . . . . . . . . . . . . 3/4 7-14 RADIOACTIVE NATERIALS Radioactive Liquid Effluent Instrumentation ............... A ..................... 3/4 8-1 Radioactive Gaseous Ef fluent Instrum entation . . . . . . . . . . . . . . B . . . . . . . . . . . . . . . . . . . . . 3/4 8-6 Radioactive Liquid Effluents .............................. C ..................... 3/4 8-12 8-14 Radioactive Gaseous Effluents .......... .................. D ..................... 3/4 AUXILIARY ELECTRICAL SYSTEMS /4 9-1 3.10 REFUELING 4.10 Refueling Interlocks ...................................... A ..................... 3/4 10-1 Core Monitoring ........................................... B ..................... 3/4 10-2 Fuel Storage Pool Water Level ............................. C ..................... 3/4 10-2 Crane Operability ......................................... D ..................... 3/4 10-2 Crane Travel - Interlocks and Switches .................... E ..................... 3/4 10-2 3.11 REACTOR FUEL ASSEMBLY 4.11 Average Planar Linear Heat Generation Rate ................ A ..................... 3/4 11-1 Linear Heat Generation Rate ............................... B ..................... 3/4 11-8 Minimum Critical Power Ratio .............................. C ..................... 3/4 11-9 3.12 FIRE PROTECTION SYSTEMS 3/4 12-1 3.13 INSERVICE INSPECTION 4.13 3/4 13-1 3.14 PLANT SYSTEMS 4.14 3/4 14-1 DESIGN FEATURES I Millstone - Unit 1 iii Amendment N .;p(,pd,106

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May 23, 1984 SURVEILLANCE REQUIREMENT LIMITING CONDITION FOR OPERATION 4.0 GENERAL 3.0 GENERAL When a system, subsystem, train, component or Not applicabl device is determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided: (1) its corresponding i normal or emergency power source is OPERABLE; and '

(2) all of its redundant system (s), subsystem (s),

train (s), component (s) and device (s) are OPERABLE, l or likewise satisfy the requirements of this specification. If both conditions (1) and (2) are not satisfied, then the applicable action state-ments of the individual specifications must be taken. This specification is not applicable in Cold Shutdown or Refueling operational condition /4 0-1 1 Amendment No. 97

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Deccaber .985

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LIMITING CONDITION FOR OPERATION SURVEILLANCE REOUIREMENT q i . REACTOR, PROTECTION SYSTEM REACTOR PROTECTION SYSTEM < Applicability: Applicablity: Applies to the instrumentation and associated devices Applies to the surveillance of the instrumentation and

which initiate a reactor scram and provide automatic associated devices which initiate reactor scram and provide l ' isolation of the Reactor Protection System huses from automatic isolation of reactor protection system buses from their power supplies, their power supplies.

! ! Objective: Objective: ' To assure the operability of the Reactor P,rotection Syste To specify the type of frequency of surveillance to be , applied to the reactor protection instrumentatio Specification: Specification: The setpoints, minimum number of trip Instrumentation system shall be functionally tested and systems, and minimum number of instrument channels that must be operable for each position of the reactor mode calibrated as indicated in Tables 4.1.1 and 4.1.2, switch shall be as given in Table 3. respectivel Response Time Daily during reactor power operation, the maximum The time from initiation of any channel trip to the fraction of limiting power density shall be checked and de-energization of the scram solenoid relay shall not the APRM scram and rod block settings given by the exceed 50 millisecond equations in Specifications 2.1.2A and 2.1.2B shall be I determined to be vali r Reactor Protection System Power Monitoring Two RPS electric power monitoring channels for each The RPS electrical protection assemblies shall be inservice RPS MG set or alternate power supply shall be determined operable as follows: operable at all times except as follows: At least once per 6 months by performance of a CHANNEL FUNCTIONAL TEST, and With one RPS electric power monitoring channel for an inservice RPS MG set or alternate power supply inoperable, restore the inoperable channel to OPERABLE At least once per 18 months by demonstrating the OPERABILITY of over-voltage, under-voltage and status within 72 hours or remove the associated RPS NG under-frequency protective instrumentation by set or alternate power supply from service, performance of a CHANNEL CALIBRATION including With both RPS electric power monitoring channels for an simulated automatic actuation of the protective inservice HPS MG set or alternate power supply relays, tripping logic and output ctreuit breakers inoperable, restore at least one to OPERABLE status and verifying the following setpoints: within 30 minutes or remove the associated RPS MG set or alternate power supply from servic Over-voltage 5 (132)VAC, Under-voltage > (108) VAC, Under-frequency > (57)liz, and Amend No. ,?(,JWI, 107 3/4 1-1 Time-delay 5 (4.0) second _ _ _ - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ - _ -

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LIMITING CDOITICN FOR CPERATKN SURVEIIDNE REQUIMMNT When the reactor mode switch is in REFUEL or SHl1IDom and fuel is in the reactor vessel, no trip functions are required to be gerable provided that all control rods are fully inserted, and either electrically or hydraulimlly disarmed. 'Ihereafter, daily willance shall be perfomed to verify that all control rtxk rensin valved out or electrimlly disarmed.

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' 3/4 1-la Millstone - thit 1 Amen &ent No.X,110

JLme 14, 1984 ,

IOCIOR PRDIRT. IO4 SYSIDI (SGAM) DEIMMNrKrICN RBQUIRDENIS Mininun nacer cf Operable Modes in which Ebncticn Irst. Dennels Trip Ebncticm Trip Ievel Setting mst Be Operable Action * per Trip (1) RERH/ SINGUPAUT Systen SUITON4 (8,11) SDNBY RN 1 mde fheitch in SUIIDN X X X A

1 Manual Scran X X X A IIN: 3 High Flux 1120/125 cf full scale X X (5) A 3 Incperative A. HI W1tage < 80 volt DC X X X (10) A B. IIM ledule Urplugged C. Selectcr Ehtitch not in Operate Positicn APIN: 2 Flcw Blased High Flux See Sectico 2.1.2A X X X A cr B 2 Rechoed High Flux See Section 2.1.2A X X X A or B 2 In@erable A. >50% IPIN Irputs** X X X A or B B. Circuit Board kraved C. Selector fheitch not in Operate Rsition 2 High Reactor Pressure 11085psig X X X A 2 High Orywell Prmmre 12psig X (9) X (7) X (7) A ' 2 Reactor Irw mter tevel >1.0 inch *** X X X A , 2 Scran Discharge W inches above the center- X (2) X X A ' High level line of the Icuer erd cap to SDIV pipe weld 3/4 1-2 !

Millstone - tlnir 1 - hntrent tb. 98

- _ _ _ _ _ _ _ __

._ ____-___ -______- _ _ . _ _ _ _ _ _ _ .  . _- - - - _ _ _ _ _ _ .
   *
   ,~
            ,
  '
   .

Jme 14,1984 TABIE 3.1.1 (Ccntinued) REEIOR EPDIECTIm SYSID1 (SQW4) DEIRMNTATIm REUJIRENNIS ! Minima Mmber of Operable Modes in which Ebnction Inst. Ounnels Trip N ction Trip Imvel Settirg M st Be Operable Action * per Trip (1) RERE1/ SINtIUPAUT Systm SU1IDW (8,11) SDNBY RN , 2 'Ibrbine Cumh &-r Ior >23 in. hg. Vacun X (3) X (3) X -A cr C Vacuts i 2 Main Steamline Radiation _<7 x Nonnal Pbil Ibser X X X A or C in ayouand ! 4 (6) Main Steenline Isolation <10% Valve Closure X (3) X (3) X A or C Valve einsnee 2 'Ibrbine Omtrol Valve See Section 2.1.2 F X (4) X (4) X (4) A or C ' Fast Closure 2 'Ibrbine Stcp Wlve <10% valve closure X (4) X (4) X (4) A or C tetes: 1. 'Ihere shall be two cperable or tripped trip systems for each functio . Permissible to bypass, with control red block, for reactor guiic1.icn systen reet in REFUEL and SETIIDN positions of the reactor node switch.

i i 3. Bypm===i den reactor ptweene is <600 psig.

, 4. Bypmewr1 den first stage turbine prwa=we is less thsq that which uu+ds to 45% rated steen flow (generator ' output w ' ^_ely 307 Mee).

I , 3/4 1-3

' Millstone - thit 1 Anerrinent tb. 98

_ _ - - - _ _ _ - _ _ ___ _ _ _ _ _ _ _ _ - . . _ _ . . - _ - - .. - . , , April 40 1986 i

IHM's are typassed when xde switch is placed in run. The detector for each @erable Im channel shall he_ fully ;
inserted until the associated APIN channel is cperable ard indicatirg at least 3/125 full scale.**** '

i ! The desicp permits closure of any one valve without a scran beirg initiated.

i } May be bypameri when nemuunry by cimirg the marmal instrunent isolation valve for scran of PS-1621 A throucA D

daring purgirg for contairment inerting or deinertir . Wwn the reactor is suberitical and the reactor water taperature is less than 212%, only the follorirg trip l fmetions need to be cperable: ,

t i ! Pbde Switch in SUID0hN l Marmal Scran I Hi@ Flux 11M Scran Discharge Vohme High level

, APIM Ibrimrt Hi@ Flux .

) ftt required to be gerable den primary ocntairment integrity is not recpired.

? j 1 With the made switch in RN position an irrperative trip fmetion also requires an associated ARM "chenscale j alann." ' I I 1 Trip functions are not required to be cperable if all mntrol roch are fully inserted, ard either electrically or

{ hydraulically disanned in accordance with !%mcification 4. l'

 * Action: If the first colum cannot be met for me of the trip system, that trip systen shall be tripped. If the first colum  ;

camot be met for both trip systes, the gpropriate actions listed belor shall be taken: i

j Initiate insertion of operable rods and ocuplete insertion of all cperable rocks within four hour ; j Ihrta power level to IIM range and place mode switch in the SDRIUPAUT SDNM positim within ei@t hours.

4 Ihrt= turbine Iced and close noin steen line isolation valves within ei@t hour c

**

l An APIM will be considered incperable if there are less than two IRM inputs per level or there are less than 50% of the

nonmal ocupliment of IRM's to an API ***

!   One inch on the water level instrumentation is 127 above the tm <.f the active fue :
'
 ****

Ihr errata sheet dated 10-7-70 l i i

!       3/4 1-4 I  '
 *

Millstcne - thit 1 heenttnert It). g 110 r

!            !
           -
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l _ _ __ _ - _ - . _ _ - . _

    .
       .-
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.

TABLE 4. SCRAM INSTRUMENTATION FUNCTIONAL TESTS MINIMUM FUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTRUMENT AND CONTROL CIRCUITS IIstrument Channel Group (3) Functional Test Minimum Frequency (4)

High Reactor Pressure A Trip Channel and Alam (1) High Drywell Pressure A Trip Channel and Alam (1) Low Reactor Water Level A Trip Channel and Alam (2) (1) High Water Level in Scram Discharge A Trip Channel and Alam (1) C : denser Low Vacuum A Trip Channel and Alam (8) (1) Main Steam Line Isolation Valve Closure A Trip Channel and' Alam (1) Turbine Stop Valves Closure A Trip Alam (1) Manual Scram A Trip Channel and Alam (1) Tcrbine Control Valve Fast Closure A Trip Channel and Alam (6) (8) (1) Flow Blased High Flux APRM 8 Trip Output Relays (6) (7) (8) (1) Reduced High Flux APRM 8' Trip Output Relays (8) Before each startup (5)

IRM C Trip Channel and Alam (5) (8) Before each startup (5) High Steam Line Radiation 8 Trip Channel and Alam (2) (8) (1) Mode Switch in Shutdown A Place Mode Switch in Shutdown Each refueling outag Amendment 34 November 19, 1976 3/4 1-5

__ _ ___ ___ -___ -

f .' TABLE 4. stes: Initially once per month until exposure hours (M as defined in Figure 4.1.1) is 2.0 x 105 , thereafter according to figure 4. with an interval not less than one month nor more than three months. Millstone will use data compiled by Consnonwealth Edison on the Dresden 2 linit in addition to Millstone Unit I dat . An instrument check shall be performed on low reactor water level once per day and on high steamline radiation once per shif . A description of the three groups is included in the bases of this Specificatio . Functional tests are not required when the systems are not required to be operable or are tripped. If tests are missed, they shall be performed prior to returning the systems to an operable statu . Maximum test frequency required is once per wee . This test includes verification of time delay relay performanc . This test includes verification of 90% setdown in 1 30 second . This instrumentation is excepted from the functional test definition. The functional test will consist of injecting a simulated electrical signal into the measurement channel, i l l l l l i l l l

3/4 1-6 l

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      . . _ _

_ _ _._ _ ___ _ _ _ _ __ . . _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ . _____ _ _ .. .

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 .
          ;

TABLE 4. . SCRAM INSTRUMENTATION CALIBRATION MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS j Calibration Minimum Calibration , Group (1) Method Frequency (2) ! nstrument Channel B Heat Balance Once every 7 days \PRM Output Signal (5) B Standard Current Source Once every 3 months \PRM Flow Bias Trip B Standard Current Source Once every 3 months \PRM Reduced High Flux Trip 8 Standard Current Source Refueling IRM A Pressure Standard Every 3 months . ilgh Reactor Pressure , A Pressure Standard Every 3 months High Drywell Pressure Delta Pressure Standard Every 3 months Low R; actor Water A . A Vacuum Pump Every refueling ; Condenser Low Va ' Pressure Standard Every refueling ** Generator Load Rejection A Standard Current Source (4) Every 3 months High Steam Line Radiation C Water Level Every 3 months High Water Level in Scram Discharge A

Notes: A description of the three groups is included in the bases of this specificatio , Calibration tests are not required when the systems are not required to be operable or are trippe If tests are missed, they shall be performed prior to returning the systems to an operable statu . Maximum calibration frequency required is once per wee The current source provides an instrument channel alignment. Calibration using a radiation source

          . shall be made during each refueling outag . The heat balance method serves the calibration of the normal APRM high flux trip and the reduced APRM high flux tri ** Per erre  heet dated 10-7-70 m... ~.w, < in s q1 t,  .3/ar f-7    .m

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SilRVEILLANCE REI)UIREMENT LIMITING CONDITION FOR OPERATION .

4.2 PROTECTIVEINSTRUMENTATLON 3.2 PROTECTIVE INSTRUMENTATION ~ Appilcablit ty: Appitcability: Applies to the surveillance requirements of the Applies to the plant instrumentation which performs Instrumentation that performs a protective functio a protective functio *

, l Objective: Objective: To specify the type and frequency of surveillance To assure the operability of protective to be appIled to protective instrimentatio ; instrissentatio l Specification: Specification: The lastrumentation to be functionally tested and Primary Containment Isolation Functions calibrated as indicated in Table 4. l#ien primary containment integrity is required, the limiting conditions of operation for the l

instrumentation that initiates primary contain-ment isolation are given in Table 3. Emereency Core Cooline Subsystems Actuation

The limiting conditions for operation for the instrumentation which initiates the energency core cooling subsystems are given in Table 3. except as noted in Specification 3.5. Control Rod Block Actuation The limiting conditions of operation for the

 . Instrumentation that initiates control ~ rod block are given in Table 3. .
     ,ma er e

      . _  _ _ .

l . . I IAHlf 1. er o 1982 , I IN51RUMLNI Atl0N illAT INiil AILS PRIMANY CONTAthHtHI ISULATION . 1 ' l Minimum Number of ) Operable lustrument Channels Per Trip - f settips Tr_lp,tevel i Action DJ

         -
'    .'

System (1) . Instruments A

    ,,127 inches above top of active fuel  A 2 Reactor Low W ter i

79 (+4-0) Inches above top of active fuel A

Reactor low tow h ter High Drywe)1 Pressure

    ; 2 psig    5 -

f 2 (4) 120% of rated steam flow High Flow Main Steamline 2 (2) (5) 8 High Temperature Main

2 of 4 in each of Steenline Tunnel : 200*F 2 subchannels B l

High Radiation Main ; 7 times nomal rated power background Steaaline Tunnel 8 l

Low Pressere Main !

Steamlines t 825 psig C l 164 laches 3 trip setting (water differential 2$ High flow Isolation on steam linie) > 150 tache Condenser Line 44 inches > trip setting (water differential on water sTee) 35 fache .

.       d trip systems for each itle (1) thenever primary containment fategrf ty is required,
  .
   ~
       -

there i pe If shal Per each steamlin Action: the first column cannot be met for both trip systest, the ap Initiate an orderly shutdeun and have reector"In cold shutdown condiffen in 24 _ Initiate an orderly load reductien and have reactor in Mot Standby within 8 hour Close isolation valves in isolation condenser syste Atin g0,.

i

'
(4) May be bypassed when necessary by clostag the manuel    instrument isolatio have to be met for a steamline during purging for containment leerting or deinertin (5)

Minimum number of operable instrument channels per trip systen requirement does not if both contalvsment isolation valves in the line are close . 3/4 2-2

. kndmentNo./.!I;67   ,,
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      ,

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DBoEmber 6,1985 TARIE 3. corrected 2/12/86 l IPEIBLMNIATKN 'DAT INITIAIYS 100 BtfXX Miniimsn Ntcher of Operable Instnment Oiannels pe Trip Systen{II Instnsnent Trip Invel Settiro i 1( } AHN 14mcale (Flow Biased) See Specification 2.1.2B 1( APIN Downscalo 23/125 Fb11 Scale 1(6) Ibd Block Ptnitor tbscale (Flow Biased) I.65W+42_(2) 1(6) Ibd Block Penitor Ibwnscale 23/125 lb11 Scale 3 IIM thmscale(3) 23/125 Fb11 Scale 3 11M Urweale Il08/125 It11 Smle I4I

2 SIN Detector not in Startto Ibsition 2(5) m e I510comts/se Scran Discharge Witme - mter tevel High I14 inches above lower cap to SDIV pipe weld

! j 1 Scran Discharge Witme - Scran Trip Bypassed N/A (1) For the Startup/ Hot Starrby and Run positicrs of the Reactor Mode Selector Switch, there shall be two merable or trimed trip systems for each ftmetim except the SIN rod blocks; I1M downscale are not operable in the RN position and APfM downscale need not be (per&le in the Starttp4bt Starrty ntde. If the first coltsm cannot be met for one of the two trip systens, this conditicn may exist for to to seven days prwided that during that time the merable system is ftmetionally tested innediately and daily thereafter; if this condition lasts lanner than seven days, the systen shall be trimed. If the first coltan carmot be met for both trip system, the systems shall be trippe (2) W is the reciruilation flow required to achieve rated core flow ewess.si in percen (3) I1N dcunscale may be bypassed when it is on its lowest ranc (4) This function may be bypassed when the comt rate is >100 cps or when all ITN rance switches are above Ibsition ~

(5) One of these trips may be bypassed. The SIN ftmetion may be bypassed in the higher ITN rarpes when the IIN toscale rod block is rperabl Amerrinent No.)lqQrf,107   3/4 2-5

Table 3.2.3 Continued April 18,1983 Instnmentaticn 1 hat Initiates Rod Block (6) '!he trip may be bypassed den the reactor power is <30% of rated. An MM channel will be considerni irroerable if there are les than half the total ruter of normal irputs frun any IPIM leve (7) 1here nust be a total of at least four (4) cperable or geratirg APFN channels.

. Juerihent Nt ,87 3/4 2-Sa (Correction) _

- _ - - _ _ _ _ _ - _ _ _ _ _ _ - - - . ___ .
  ~    e-    .
-
.
         /
          *

r TABLE 4.2.1 l April 18. 1983 l MINIMUM TEST AND CALIBRATION FREQUENCY FOR CORE COOLING INSTRUMENTATION R00 BLOCKS AND ISOLATIONS -

..

Instrument Channel Instrument Functional Test (2) Ca11bration(2) Instrument Check (2) l ECCS Instrumentation l Reactor Low-Low Water Level (1 ) Once/3 Month's -- Drywell High Pressure (I) Once/3 Months -- ' Reactor Low Pressure (Pump Start) (1.) Once/3 Months -- Reactor Low Pressure (1) Once/3 Months --

  (Valve Permissive) APR LP Core Cooling Pump Interlock (1)  Once/3 Month ". Containment Spray Interlock  (1)  Once/3 Months -- toss of Normal Power Relays  Refueling Outage None  --
         '
; Power Available Relays  (1) (5)  None  --

Reactor High Pressure -

Once/3 Months -- Rod Blocks . . APRM Downscale (1) (3) Once/3 Months (1) APRM Flow Variable (1) (3) Dece/3 Months (1) IRM Upscale (6) (6) (6) IRM Downscale (6) (6) (6) RBM Upscale (1) (3) Once/3 Months (1 ) R8M Downscale (1) (3) Once/3 Months (1) i SRM Upscale (6) (6) SRM Detector not in Startup Position (6) (6)?-

       (6)  (6)

l' 9. Scrai. Discharge Volume - Water Level High Quarterly Refueling Outage -- 10. Scram Discharge Volume - Scram Trip Bypassed Quarterly None -- l Main Steam Line Isolation

! Steam Tunnel High Taperature  Refueling Outage Refueling Outage --

1 Steam Line High Flow (1) Once/3 Months Once/ Day Steam Line Low Pressure (1) Once/3 Months None i Steam Line High Radiation (1)(3) Once/3 Months (4) Once/ Day

l Amendment No. H . $1,IT. 86 3/4 2-6 (Correction)

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I I TABLE 4.2.1 (Continued) St.3tember 14, 1984 ! MININUM TEST AND CALIRRATION FREQUENCY FOR CORE COOLING INSTRUMENTATION ROD BLOCKS AND ISOLATIONS ! Instrument Channel Instrument Functional Test (2) Calibration (2) Instrument Check (2) i Isolation Condenser Isolation Steam Line High Flow (1) Once/3 Months (1) ' Condensate Line High Flow (1) Once/3 Months (1) ! Reactor Building Ventilation Steam Tunnel Ventilation and Standby Cas Treatment System Initiation - Ventilation Exhaust Duct, Steam Tunnel (1) (3) Once/3 Months Once/ Day j Ventilation and Refueling Floor Radiation i i Monitors ! l Air Ejector Off-Cas Isolation !

Radiation Monitors (1) (3) Once/3 Months (4) Once/ Day

Notes: 1) Initially once per month until exposure hours (M as defined on Figure 4.1.1) is 2.0 x 10 , thereafter according to Figure 4.1.1 with an interval not less than one month nor more than three months.

{
'

2) Functional test calibrations and instrument checks are not required when these instruments are not required to be operable or are tripped.

, 3) D is instrumentation is excepted from the functional test definition. W e functional test will consist of injecting j a simulated electrical signal into the measurement channel. See Note 4.

I 4) These instrument channels will he calibrated using simulated electrical signals once every three months. In addition, l calibration including the sensors will he performed during each refueling outag l 5) The individual power available on emergency bus relays will he functionally tested at the frequency specified'by (1)

,   above. A full functional test including the actuation of the permissives will be performed every refueling outag !

6) His instrumentation is excepted from the functional test definition. The functional test will consist of injecting a

j simulated electrical signal into the measurement channel. Functional tests shall be performed before each startup with a required frequency not to exceed once per wee Calibrations including the sensors will be performed during each refueling outage. Instrument checks shall be performed at least once per day during those periods when the I instruments are required to be operabl Amenelm.* n t Ne' l's, 77, Wn 1/ '. . 7

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

Srptember 14, 1984 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMElli The minimum number of operable instrument channels specified in Table 3.2.3 for the i Rod 8, lock Monitor may be reduced by one for maintenance and/or testing for periods not < l In excess of 24 hours in any 30-day perio k Air Ejector Off-Gas System 2- Except as specified in 3.2.0.2 below, both - ' air ejector off-gas system radiation monitors ' shall be operable during reactor power oper:- tion. The trip settings frer '.he monitors . I 55411 he set at a nine not ~to exceed the ' , a !

equivalcat of the instantaneous stack release ' limit specified in Specification 3.8. The tira Wley setting for closure of the steam

       ;'
!

Jet-air'ejectoreff-gas isolation ytive shall - i ! not exceed 15 minute ,. i l

          ,  +

2 From and after the date.that one of the tuo '

       '    .
        ; -

air ejector oft-tas system radiation _- I

     '
;
'   monitors is asade or found to be inonenbl reactor power operation is persistthic enly     s during the se:ceeding 24 hours, prpy!ded the  i

_ innocrable monitor is tripped, unless xc ' system I: sooner % operabl Reactor BuildigvcatT!atir,n Isolation, Steam Tu-ael_

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Fe~n~tT17tisi is~ c ladoiiR Standhy Gas Treatment

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g Sist T 8%!{l.ji.,-1 r , ' l Except as spectffed in 3.7 E.2 beloA six , -

radiation nr, nit: es shall be operable at all '

time , , I

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A ndment No. M 100 3/4 2-6- '

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! > I LIMITING CONDITION FOR OPERATION SURVEILLANCE REQU!RFMEP!fS - _ h 3 x -

         - One of the two radiation monitors in the reactor
           I

! building ventilation duct, one of the two . radiation monitors on the refueling floor and .

           -
           '

i one of the two Radiation monitors in the steam v ?

tunnel ventilation may be inoperable for 24 hr _ . " , '. s j If it is not restored to service in this tim j the twactor building ventilation system and steam 1 tunnel ventilation system shall be isolated and the standby gas treatment operated until repairs are complet . The radiation monitors shall be set to trip as follows: Ventilation duct - 11 mr/hr.

j Refueling floor - 100 mr/hr.

i I

l Steam tunnel ventilation - 12 mr/hr.

! I I i !

'

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. . I i 3/4 2-9

Anendment No. 100

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Fabruary 26, 1976 -

SilRVElitANCC R lillfMI NT LIMITIIIG CONDITION FOR OPERATION 4.3 RfACilVITY CONTR0t 3 RfACTIVITY CONTROL Agil icabiI iiy.: Appt icabillty: Applies to the surveillance requirements of the Applies to the operational status of the control rod control rod syste syste Ohicctive: Ob.jective: To verify the ability of the control roof system To assure the ability of the control rod system to to control reactivit control reactivity, Specificalion: I Specification: Reactivity Limitations _ Ileactivity Limitations _ Reactivity N rgin - Core I.oading_ _ Reactivity Margin - Core Leading sufficient control ends shall be with-The core loading shall be limited to that drawn following a refueling outage when core alterations were performed to which can be made subcritical in the most demonstrate with a margin of 0.3M AK reactive condition during the operating cycle with the strongest operable control that the core can be made subtritical rod in its full-out position and all other at any time in the subsequent fuel operable rods fully inserted, cycle with the strongest operable control rod fully withdrawn and all Reactivity Margin - Stuck Control Rods other operable rods fully laserte . Reactivity Wrgin - Stuck Control Rods _ Control rod drives which cannot be moved with control rod drive pressure shall be considered inoperable. The control rod Cach partially or fully wittwirawn directional control valves for inoperable operable control rod shall be esercisedThis control rods shall be disarmed electrically one notch at least once each wee and the rods shall be in such positions test stiall he leerformed at least once per l that Specification 3.3.A.1 is met. In no 74 hours in the event smwer operation is case shall the number of non-fully inserted continuing with three or paire Inoperable rods disarmed be greater than eight during 3/4 3-1 AmendmentNo.[,22 --

_ _ _ _ _ - . . __ _ _ _ .__ _ ~ : a .' December 6, 1985 - SURVEILLANCE REQUIREMENT i LIMITING CONDITION FOR OPERATION

control rods or in the event power opera-l power operatio If a partially or fully tion is continuing with one fully or withdrawn control rod drive cannot be partially withdrawn rod which cannot be i moved with drive or scram pressure the reactor shall be brought to a shutdown moved and for which control rod drive condition within 48 hours unless investiga- mechanism damage has not been ruled out.

i The surveillance need not be completed A tion demonstrates that the cause of the within 24 hours if the number of inoperable i failure is not due to a failed control rod rods has been reduced to less than three drive mechanism collet housing.

j and if it has been demonstrated that l control rod drive mechan. ism collet housing Control Rod Withdrawal i failure is not the cause of an immovable ! Each control rod shall be coupled to its control ro drive or completely inserted and the control rod directional control valves Control Rod Withdrawal disarmed electrically. However, for purposes of removal of a control rod drive, The coupling integrity shall be verified as many as one drive in each quadrant may for each withdrawn control rod as follows: i

be uncoupled from its control rod so long when the rod is fully withdrawn the as the reactor is in the shutdown or a.

refuel condition and Specification 3.3. first time subsequent to each l refueling outage or after maintenance, l is me observe that the drive does not go The control rod drive housing support to the overtravel position; and

system shall be in place during power when the rod is withdrawn the first operation and when the reactor coolant time subsequent to each refueling system is pressurized above atmospheric outage or after maintenance, observe pressure with fuel in the reactor vessel, discernible, response of the nuclear unless all control rods are fully inserted Instrumentation: However, for initial and Specification 3.3.A.1 is se rods when response is not discernible, subsequent exercising of these rods after the reactor is critical shall be performed to verify instrumenta-tion respons . The control rod drive housing support system shall be inspected after reassembly and the results of the inspection shall be recorde AmendmentNo.[, , , 107 3/4 3-2

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SURVEILLANCE REQUIRDIGIT-LIMITINE CONDITitul FOR OptRATICIL I I 4.3.8 Centrol Red Withdrawal 3.3.8 Centrol Red Utthdraus1_ 3. Whenever the reacter is in the startup 3. (a) To consider the red worth statutrer ? operable, the following steps west or run made below 205 rated thermal be performed: power. ne centrol rods shall be moved uniess the red werth alateIrer is (1) The control red vttheravel , operable er a second independent seguence for the red worth

operator er engineer verifles that slaintrer computer shall be the operator at the reacter censole verifled as correct.

4 is felleutag the centrol red pre-gram. The second operater may be (11) The rod work minimirer 'l used as a substitute for an temper- diagnostic test shall able red worth staletter during be sucessfully complete a startup only if the red worth alalmirer fatis after withdrawal , !

      (III) Proper annunctatten of the

, of at least twelve centrol rod select error of at least ena i out-of-segnance centrol red in 4. Centrol rods shall not be withdrawn each fully inserted group shall l for startup er refueling unless at least two source range channels

be vertfle have an ehserved count rate egual to er prester than three counts * (lv) The red black fonctlen of tl.a raJ worth statatzer shall he verlfted

per secon by attemptlag to withdrew an l out-of-sequence centeel red 64-yond the black pota (b) Ifuhtlethe red l Inc..rsble the worth reactereintelser is in the .s startup or run sede below 105 rated therwal su . and a second independent operator or

engineer is being used, he sh. ell verify that all red positions are correct prios to commenctag withdrawal of each i *

grou Amendaent No.11. AS. D . 76 ' April 16.1981, 3/4 3-3

..

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l *

N:vimber 12, 1982

]  LIMITING CONDITION FOR OPERATION   SURVEILLANCE REQUIREMENT ,

i During operation with limiting control ,. 4 Prior to control rod withdrawal for ! rod patterns, as determined by the startup or during refueling, verify

    ' {.

j reactor engineer, either: that at least two source range channels have an observed count Both RBM channels shall be operable; rate of at least three counts per

or , secon '

! ] Control rod withdrawal shall be When a limiting control rod pattern 1 blocked; or exists, an instrument functional test i of the ROM shall be perfomed prior The operating power level shall be to withdrawal of the designated

)   limited so that the MCPR will   rod (s) and daily thereafter.

i ' remain above 1.06 ' assuming a single error that results in complete Scram Insertion Times

,   withdrawal of any single operable      "

control ro . During each operating cycle, each

operable control rod shall be i

. Scram Insertion Times subjected to scram time tests from l the fully withdrawn positio If

: The average scram insertion time, based   testing is not accomplished during i   on the deenergization of the scram pliot   reactor power operation, the measured
,

valve solenoids as time zero, of all scram insertion times shall be l operable control rods in the reactor extrapolated to the reactor power j ' power operation condition shall be no operation condition utilizing greater than: previously determined correlation % Inserted From Average Scram The scram discharge volume drain and ,

Fully Withdrawn Insertion Times (Sec.) vent valves shall be verified open at least once per month.

' 5 0.375 20 0.900 The following conditions of opera-50 2.000 bility of the scram discharge volume 90 3.500 drain and vent valves shall be verified at least once per operating cycle in accordance with Section 3.13, l . Inservice Inspection: l

.-

Ame- int no. 15. U,) ( 6 6 3 3/4 3-4

   --
         .

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.      November 12, 1982 lMITING CONDITION FOR OPERATION  SURVEILLANCE REQUIRDiENT i
   , Closing time after sign ,
'    1  for control rods to scram and
    ,

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   - Verification of opening 4      when scram signal is reset l     . and when the scram discharge

, I

volume trip is bypasse * l

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    '   November fl982
-
.

LIMITING CON 0! TION FOR OPERATION SURVEILLANCE REQUIREMENT The average of the scram insertion time l D. Control Rod Accumulators for the three fastest control rods of a 1 I - groups of four control rods in a two by Once a shift. check the status in the two array shall be no greater than: control room of the pressure and level alarms for each accumulato '

 % Inserted from Average Scram fully Withdrawn Insertion Times (sec.)

. 5 0.398 20 0.954 50 2.120

3.800 1 The maximum scram insertion time for 90% l , insertion of any operable control rod

i shall not exceed 7.00 seconds.

l l The scram discharge volume drain and vent valves will close in less than 30 seconds after receipt of a signal , for control rods to scra Control Rod Accumulators At all reactor op' erating pressures, a rod accum slator may be inoperable provided that no other control rod in the nine-rod square array around this rod has a: Inoperable accumulato . Directional control valve electrically disarmed while in a non-fully inserted positio . Scram insertion greater than maximum , permission insertion tim .: l

j Amendment No. A, gg , 3/4 3-5

_

- -- -- . .- - . -- .__ - .. . . . . . - - .

November 12, 1982

) LIMITING CON 0lTION FOR OPERATION SURVE'IL' LANCE REQUIREMENT I < ' If a control rod with an inoperable accumulatr@ . is inserted " full-in" and its directional cont'rol valves are electrically disarmed, it shall not . . . be considered to have an inoperable accumulato l t

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November } 82 l1 l LIM NG CONDITION FOR OPERATION SIIRVEILLANCE REQUIREMENT

-
      -.

l .

Reactivity Anomalies Reactivity Anomalies

) The reactivity equivalent of the difference * During the startup test program and startups j following refueling outages, the critical

between the actual critical rod configuration k I and the expected configuration during power rod configurations will be compared to the

operation shall not exceed 11 AK. If this expected configurations at selected operating limit is exceeded, the reactor will be shutdown

conditions. These comparisons will be used , as base data for reactivity monitoring during

l until the cause has been determined and corrective actions have been taken if such subsequent power operation throughout the actions are appropriate. In accordance with fuel cycle. At specific power operating i

: Specification 6.6. the NRC shall be notified  conditions, the critical rod configuration i of this abnormal occurrence within 24 hour will be compared to the configuration expected l

based upon appropriately corrected past dat If Specification 3.3 A through D above are not This comparison will be made at least every l equivalent full power month.

- met, a normal orderly shutdown shall be Initiated and the reactor shall be in the cold l

shutdown condition within 24 hour . I Allowable combinations of thermal power and ! total core flow shall be restricted to Curve 1 j shown in Figure 3.3.1.

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! ! * j s . s l Amendment No. Jg. 2p. 25, 34 - '

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ALLOWABLE COMBINATIONS OF TOTAL CORE FLOW AND POWER LEVEL

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Amendment No. 52,8 i 'l - maf. wroo 24366 sept. 1981

  . CORE nm (3)    ,
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May 23, 1984 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.5 CORE AND CONTAINMENT COOLING SYSTEMS 4.5 CORE AND CONTAINMENT COOLING SYSTEMS Applicability: Applicability: Applies to the operational status of the emergency Applies to periodic testing of the emergency cooling subsystem cooling subsystem Objective: Objective: To assure adequate cooling capability for heat To verify the operability of the core and contain-removal in the event of a loss of coolant accident ment cooling subsystem or isolation from the normal reactor heat sin Specification: Specification: Surveillance of the Core Spray and LPCI Sub- Core Spray and LPCI Subsystems systems shall be performed as follows: l Except as specified in 3.5. A.2, Core Spray Subsystem Testing: 3.5.F.6, 3.5.F.7 and 3.5.F.8, both core spray subsystems shall be operable Item Frequency whenever irradiated fuel is in the reactor vessel, Simulated Automatic Each Refueling Actuation Test Outage Pump and Valve Per Surveillance Operability Requirement 4.13 Amendment No. 97 3/4 5-1

l May 23. 1984 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT From and after the date that one of the core spray subsystems is made or found to be inoperable for any reason, reactor operation is permissible only during the Core Spray header succeeding fif teen days unless such sub- Ap instrumentation system is sooner made operable, provided check Once/ day that during such fifteen days all active calibrate Once/3 months components of the other core spray sub- test Once/3 months system and the LPCI subsystem and both emergency power sources required for operation of such components if no external source of power were available shall be operabl . Except as specified in 3.5.A.4, 3.5.B.3,4,5, 3.5.F.6, 3.5.F.7 and 3.5.F.8 the LPCI 2. LPCI Subsystem Testing shall be as specified subsystem shall be operable whenever in 4.5. A.I.a. b and c except that irradiated fuel is in the reactor vesse three LPCI pumps shall deliver at least 15,000 gpa against a system head correspond-ing to a reactor vessel pressure of

    >14.7 psi Amendment No. 97   3/4 5-2
    .

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

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 .

May 23, 1984 LIMITING COWITION FOR OPERATION SURVEILLANCE REQUIREENT l From and after the date that one of the LPCI pumps is made or found to be inoper-able for any reason, reactor operation is periaissible only during the succeeding 30 days unless such pump is sooner made ! operable, provided that during such thirty days the remaining active components of the LPCI and containment cooling subsystem and all active components of both core spray subsystems and both emergency power sources required for operation of such components if no external source of power During each five year period, an air were available shall be operabl test shall be performed on the drywell ! spray headers and nozzle l I A maximum of one drywell spray loop may be inoperable for 30 days when reactor Surveillance of the Containment Cooling Sub-water temperature is greater than 212* systems shall be performed as follows: l If the requirements of 3.5.A cannot be met, Emergency Service Water Subsystem Testing: l j a. orderly shutdown of the reactor shall Frequency be iaitiated and the reactor shall be in Item ; the cold shutdown condition.within 24 Per Survel11ance hour Pump & Valve Operability Requirement 4.13

           ! Containment Coolina subsystems Except as specified in 3.5.B.2, 3.5.8.3,       .
           !

3.5.F.6, 3.5.F.7 and 3.5.F.8, both containment cooling subsystems shall be .; operable whenever irradiated fuel is in j the reactor vessel.

! . t l ! ! ! 1 ' ! Amendment No. 97 3/4 5-3 ,

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

i I

l LIMITIhr. ConcITION FOR OPERATION SMRVtiLLANCE REQtilREMENT April 16 1981

       , Iron and af ter the date that one of the *

! emergency service water (EW) subsystem pugs is made er found to be inoperable for any reason. reacter operation is ! pennissible only during the succeeding - thirty days unless pump is scener made eperable, provided that de-ing such thirty J j days all other active compecents of the i contalnuent cooling system are operable.

! From and after the date that one active i component in each containment cooling subsystes er a LKI and EW in one j containment caeling subsystem is made er

found to be Inoperable for any reaso l reacter operatten is pennissible only

during the succeeding 7 days provided the renslalog active tempements in each contalement caelleg subsystem. both core !

spray subsystems and both emergency power +

sources for operetten of such campensets If me enternal searce of power were available shall be operable.

! !

! Free and after the date that ese LKI i and one E5tl peep to each costalmeet l cooling subsystes is made er found to be ' lamperehte for any ressen, reacter opera-tien is permissible only daring the succeedlag fear days provided the renale- i

lag active campements of the contalement i coollag subsystems, both core sprey sub-j systems and both emergency power sources * ! for operstlen of such campenents if no enternal source of power were available, shall be operabl .

- Amenihnent flo. 76 April 16,1981  3/4 5-4
       :

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Jun , 1984

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LIMITING CONDITION FOR OPERATION

  . . _ _ _ _ . . . _ . . . SURVEILLANCE REQUIREMENT From and after the date that one contain-ment cooling subsystem is made or found to be inoperable for any reason, reasfor operation is permissible only during t he stecceeding tour days provided that all active components of the other contain-ment cooling subsystem, both core spray subsystems and both emergency power sources for operation of such components if no external source of power were i

available, shall be operahi . C. Surveillance of FWCI Subsystems shall be If the requirements of 3.5.B cannot be performed as tollows: met, an orderly shutdown shall be initiated and the reactor shall be in a Item Frequency cold shutdown condition within 24 hour Pump and valve Per Surveillance FWCl Subsystem aperability Requirement 4.13 Except as specif'-d in 3.5.C.3 below, the FWCI sybsystem shall be operable when-ever the reactor coolant temperature is greater than 330*F and irradiated fuel j ' is in the reactor vesse Simulated Automatic Every refueling Actuation Test outage , There shall be a minimum of 225,000 gallons I of water in the condensate storage tank Once a week the quantity of water in the for operation of the FWC condensate storage tank shall be logged.

l l i l l Amendment No. 98 3/4 5-5

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June 14,'1984 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT From and after the date that the FWCl subsystem is made or found to be inoperable for any reason, reactor operation is permissible only during the succeeding seven days unless such subsystem is sooner made operable, piovided that during such seven days all active com-ponents of the Automatic Pressure Relief Subsystem, the core spray subsystems, LPCI subsystem, and isolation condenser system are operabl . If the requirements of 3.5.C cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the cold shutdown condition within 24 hour D. Surveillance of the Automatic Pressure Relief Subsystem shall be performed as follows: Automatic Pressure Relief (APR) Subsystems During each operating cycle, the follow- Except as specified in 3.5.D.2 below, ing shall be performed: l the APR subsystem shall be operable when-ever the reactor coolant temperature is A simulated automatic initiation of greater than 330*F and irradiated fuel is the system throughout its operating in the reactor vesse sequence but excludes actual valve opening, and With the reactor at low pressure, each relief valve shall be manually opened until valve operability has been veri-fied by torus water level instrumenta-tion, or by an audible discharge detected by an individual located outside the torus in the vicinity of each relief line.

. Amendment No. 98 3/4 5-6

    - - _ _ _ _ _ _ _ _ _ _ - _
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Jurt 4 1984

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

LIMITING CONDITION FOR OPERATION

~
 - - - - -  -
      --

SURVEILIANCE REQUIREMENT

      ,  . .. . . _ _ _, From and af ter the date that one of the

__ When it is determined that one safety / four relief / safety valves of the auto- relief valve.of the automatic pressure matic pressure relief subsystem is maale or relief subsystem is inoperable the found to be inoperable when the reactor actuation logic of the remaining APR l coolant temperature is above 330*F with valves and FWCI subsystem shall be irradiated fuel in the reactor vessel, demonstrated to be operable immediately reactor operation is permissible only and daily thereafter, during the succeeding seven days unless l repairs are completed and the subsystem E. Surveillance of the Isolation Condenser made fully operabic and provided that during System shall he performed as follows: such time the remaining automatic pressure relief valves, FWCI subsystem, and gas Isolation Condensor System Testing: turbine generator are operable, The shell side water level and If the requirements of 3.5.D cannot be temperature shall be checked met, an orderly reactor shutdown shall daily, he initiated and the reactor shall bc l in a cold shutdown condition within 24 hour Isolation Condenser System Whenever the reactor coolant temperature is greater than 330*F and irradiated fuel is in the reactor vessel, the isolation con-denser shall be operable except as specified in 3.5.E.2 and the shell side water level l shall be greater than 66 inche Amendment No. 9s 3/4 5-7

-- ~ Jane 14,1984

--

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

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

_ _ _ . . . . _ __ ._ _ From and af ter the time that the Isola- Simulated automatic actuation and tion Condenser is made or found to he functional system testing shall be inoperable, for any reason, power performed during each refueling operation shall be restricted to a outage or whenever major repairs maximum of 40% of full power, i.e., are completed on the syste (804 MW ) within 24 hours until such time the Isolation Condenser The system heat removal capability is returned to service provided that shall be determined once every all active components of the core five year spray subsystems and LPCI subsystems are operabl Calibrate vent line radiation monitors quarterl . If the requirements of 3.5.E cannot be met, an orderly shutdown shall be initiated Notor operated valves shall be and the reactor shall be in a cold tested per surveillance requirement shutdown condition within 24 hour .1 Minimum Core and Containment Cooling System F. Surveillance of Core and Containment Coolina Availability System Except as specified in 3.5.F.2, 3.5.F.3, The surveillance requirements for normal 3.5.F.7 and 3.5.F.8 below, both emer- operation are in Section gency power sources shall be operable whenever irradiated fuel is in the reacto . Amendment No. 98 3/4 5-8

_ . - . _ _ _ _ . . - - - - - - _ - -.- - -. . . .-. . - - . - -- - . - _ I I --.

 :
 -
   . -
      *  5WRytlLLantt stettletMENT j  tlNITInE CONDITI M FW SetRAft s f      .

! From and after the date that the diesel generator is made er found to be inoper-able for any reason, continued reactor ; operation is permissible only during the : succeeding seven days provided that the get turbine generater. FUCI. Astematic - : Pressere nellef Subsysten, all components of the low pressere core cooling and the l contalement cooling subsystems shall be - eperabl . From-and after the date that the gas turbine generator is mode or found to be Inoperable for any ressen. continued reacter operetten is peculssible only - during the succeeding four days provided , that the diesel generator all components { ef the APR subsystem all components of ' the low pressere core cooling and contala- -

ment cooling subsystees shall be operable.

i If the regelresents of 3.5.F.1 connet j be est, an orderly sketdeun shall be j lattiated and the reactor shall be in

the cold shutdeun conditlen within 24 hours.

l Any com6tnetten of lamperable cessements la the core and contalement coollag l l systemis shall not defeat the capablitty ef the remalalag operable components to fulfill the core and containment cooling

  -
  -

functlen .

Amendment No. 79 76 April 16,1981

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         - ~ ~ ~ ' ~ ~ ~ ~ ~
     --
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Sultv[]tLAfgCL litigtllAtstglT t IMIIIssG LinultlGM ltNt Of titAllost

- *

_ ..__.___ lacept as specirled in 3.5.F.7 edica trradsated fuel is la tha vessel and the reacter is la the cold SInstdeum condittee, all law pressere core and castalement caelIog subsystems muy be laeperable - provided that me omrt is heleg done uhtch has the potential for draining the reacter vesse . Imea irradiated fuel is la the reacter vessel and the reacter is la the cold shutdemos canditlen er refueling candl- - ties, a slagte centrol red may he withdramm and the drive mechaelse removed if the fellemlag canditless are satisfie (a) feel reenval and replacement will met be dame without a full comple-meet of centrol red (b) Its mark will be perfensed la the reacter vessel other than fuel slpplag edille a centrol red drive hemslag is ape ,

      .
        .
   (c) either () hath care spray system II) hath les pressere coelaat injectlen systems, er lit) ene care spray system med one law pressere coolant injectlen system supplied by ladependent electrical poser shall be operabl ,
        *
   (d) With the torus dralmed. 1) the eperable les pressure coelaat injectten systems / core sprey systems will be aligned math the
          .
 -
.
.

3/4 5-10

%  .: yJ, yp, JJ. 46  Kirch 10.1978    .

. s' P

d' ._ - E

= A March 10, i
~   -     -

SURVEILLANCE REQUIREMENT LIMITING CONDITION FOR OPERATION condensate storage tank and the condensate storage tank section valve V7-58 locked upon, it) the condensate storage tank shall con-tain at least 414.000 gallons of - usable water and the refueling .

cavity shall contain at least 383.000 gallons of wate (e) The minimum electrical power source requirements shall be the same as specified in para-graph 3.7. . Except as specified in 3.5.F.7 when 46 trradiated feel is in the reactor ves-sel and the reactor is in the refuel condition, feel removal and replace- -

ment may be done provided the following conditions are satisfied:

 (a) Either1)bothcorespraysystem ) both low pressure coolant injection systems, or 111) one core spray and one low pressure coolant injection system supplied by Ll e i ;t electrical power shall be operabl (b) With the torus drained 1) the operabid low pressure coolant injection systems / core spray systems will be aligned with the condensate storage tank and the condensate storage tank section valve V7-58 locked open, 11) The condensate storage tank shall contain at least 414.000 gallons of usable unter and the refueling cavity shall contain at least 383.000 gallons of wate /4 5-11

_ _ _ _ _ - _ _ _ _ . . _ _ _ . _ _ _ ._. _ Narch 10. 19/8 LIMITING CONDITION FOR OPERATION SURVEILLA80CE RE()UIREMENT

         .
         .
  (c) The minimum electrical power source requirement shall be the same as

specified in paragraph 3.7. (d) No work is being done which has a potential for draining the vessel.

.

t 3/4 5-12

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e _ WY

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LIIllTING CONDITICII FM SptilATION SURVEILLANCE IltqulREM(NT t

7 Contalmeent Systees_ 4.7 ,Contalesment Systems Aspitcabt18ty Appi f cabliIty I

'Appites to the operating states of the primary and  Applies to the primary and secondary contate-secondary contatsument systen ment lategrit '

06.lective . Objective '

Ta assere the Integrity of the primary and secondary Te verify the lategrity of the primary and eentasmasat syste . secondary containmen ' Spectftcatten Spectftcatten

  ' Primary Contalsment
  * Primary Centalement 1. Suppressten Chenker lister Level and Temperature The suppressten chamber meter level and
 -

mulk temperature shall be checked once l , The volume and temperature of the meter in per shift. The interter pointed sur- ! the suppressten chanker shall he antatained faces above the water itne of the within the fellestag Ilmits iAenever pressure suppressten chamber shall primary containment is regstred: he Inspected at each refueling

outag t j a. Mentoun unter volesse 100.400 ft - l (correspondtag to a doencener Whenever there is Indicatten .l ,j ! submergency of 3.33 f t. at 1.0 psid) of reitef valve operetten uhtch adds heat to the suppression P881 , l b. Mlatunne unter velene 98.000 ft3 the bulk pool temperature shall l (corresponding to a doesnconer gehner* be centsneally menttered and ;

gence of 3.0 ft. at 1.0 psid) aise observed and legged every 5 mimetes matti the heat additten Maatsum unter temperature: Is teentested. ~

Amendment No. 13. 57. 73 March 11,1981

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

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

__

1 ListITlas CSIGITIOR FOR SptaATIOR btIRVEILLAhCE KgulatutCT -

;
     -

_ _

  (1) Durlag semel power operation - Ilhenever there is ladicattaa of 90* \     relief valve operetten with the local   .
   \ -    temperature of the suppressies (2) During testing which adds beat    peel reachlag 200*F er more
. to the suppresslee peel, the    en enternal visual emaninetten esoter temperatore shall not    of the suppressiem chanter shall    i
:   enceed le*F aheve the samal    be conducted before resseleg    '

I power operetten limit specified power operatie i le (1) aheve. In w tlee l . with such testleg. the pool , '

temperatore must he reduced to i

!   below the mamal power operetten         i l   Semit specified la (1) aheve
:-  wittle. 78 heers.

,.

l } L3) The reacter shall be sc:.mmed l l free any operetlag canditsee if ' " - l * the peel tee.serature react.es -

           ~
             ,

' Ile*F. Pemer operetten shall , !

 -
 '
  .

not he resumed eatsi the pool temperature is reduced les the '

ammal peuer operettee t !

.
 -

specified le 11) ahev ~ -

          ,
           !

i

.-

i

j (4) ^9u-segreestkiseletteecandi- ,

. .

Lloos. the rensar pressere ' i

~        -
*
 .

vessel-shali he depresserires;as t '

In; tem 200 psIg at mentd 1 caeldewa rete! !f 3. essi - i temperatore reaches 12h* 'd _

_

         ,    [
        ~

.

i d

        -
    -      .
            -
            *
  ,
        -
          ,    ,
, h at Ste. !). 57. 73 March II 381 3/4 ' '
          ' '
           ' '
    -

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          -
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  ,  .      ,   % ._
   ~
        ~'

_

     -
      -
        .-  ' July 6, 1978
;     _
    -
       :

l

       -
        .- _  , -

_ LIMITING COISITION ITR OPERATION SURVEILTANCE REQUIRDEkf i

! At least c=e of the two existing marrow Torus water level f astrumentatioe range torus water level monitoring     shall be calibrated once per 6 j  systems shall be operable whenever     months if both systees are primary containment is required, except     operabl l          *

as specified in 3.7.A. . J f If the torus water level monitoring system is disabled Torus water level instrumentation o and cannot be restored la six (6) hours, an orderly shall be calibrated once per month shutdown shall be initiated and the reactor shall be if only one system is operable.

j in cold shutdown within 24 hours unless the level ' monitoring system is made operable.

I

; Drywell to Suppression Chamber Differential Pressure Drywell to Suppression Chamber Differential Pressure

1

; Differential pressure between the drywell and The differential pressure between the drywell and i
$g  suppression chamber shall be esistained equal    suppression chamber shall be recorded once per ,

to or greater thaa 1.0 paid, except as specified shif below.

'l ! (1) The differential pressure shall be established withia 24 hours of entering the RIRI mode, and - ] ! may be reduced to less than 1.0 paid 24 hours l prior to a scheduled shutdow ! l (2) The differential pressure may be reduced to

less thaa 1.0 paid for a assimum of four (4)

hours during required operability testing of l the terus/ reactor building and drywell/torua

{ vacuum breakers, and during venting and i purging of the containment.

l j 3/4 7-3 i

! i

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

_

      ;  .
     +
    ~

Jus: 5, 1985 - '

._

LIMITillG CONDITION FOR OPERATION SURVEILIANCE REQUIRDENT ,

(3) Differential pressure may be less than 1.0 paid for a period not to exceed 48 hours for purposes of conducting a drywell entr _

4 . b. Drywell to torus differential pressere l instrumentation shall be calibrated once per month if only one system is operabl (4) If the provisions of (1) and (2) above cannot be met, the differential pressure shall _, ' be restored within the subsequent six (6)- c. The drywell to torna differr.atial pressure hour period or the provisions of 3.7. instrumentation shall be calibrated once per shall apply, six months if both systema are operabl , At least one (1) drywell to torus differdatial pressure monitoring system shall be operable whenever primary containment is required, except as specified in 3.7.A. If the drywell to torus differential pressure monitoring system is disabled and cannot be restored in sin (6) hours, an orderly shutdown shall be initiated and the reactor shall be in cold shutdown within 24 hours unless the differential pressure monitoring system is made operabl .

Amendment No. 102 3/4 7-4

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        -

, l

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    - -

w; , L

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       .

December 19, 1983 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT Primary containment integrity as defined 3. The primary containment integrity shall in Section 1.0 shall be maintained at all be demonstrated as follows: times when the reactor is critical or Integrated Primary containment Leak when the reactor water temperature is Test (IPCLT) i above 212*F and fuel is in the reactor vessel except while performing low power The containment leakage rates shall physics test at atmosphere pressure be demonstrated at the following test (during or after refueling')at power levels schedule and shall be determined in not to exceed 5 Mw(t). conformance with the criteria speci-fied in Appendix J of 10CFR50 using j the methods and provisioma of l ANSI N45.4-1972 and BN-TOF-1 1 Three Type A Overall Integrated Contaissent Leakage Rate tests shall be conducted at 40 i 10 <

month intervals during shutdown at P* (43 peig) during each ten-year service period. The third test of each set shall be conducted during the shutdown for the ten-year plant inservice inspectio . If any periodic Type A test fails to meet 0.75 L , the test schedule for subsequent * Type A tests shall be reviewed and approved by the Commission. If two consecutive Type A tests fail to meet 0.75 L,, a Type A test shall be performed at least every 18 months until two consecutive Type A tests meet 0.75 L , at which time the , above sched$le may be resume a Amendment No. M , k, 94 3/4 7-5

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

l t

December 19, 1983 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT The accuracy of each Type A test shall be verified by a supple-mental test which: Confirms the accuracy of the test by verifying that the difference between the supplemental data and the Type A test data is withis 0.25 L,. Mas duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplenestal tes Requires the quantity of gas injected into containment or ' bled from containment during the supplemental test to be . equivalent to at least 25 l percent of the total measured l leakage at P,. All test leakage rates shall be calculated using observed data converted to absolute value Error analyses shall be performed to select a balanced integrated leakage measurements syste . Amendment No. M 94 3/4 7-6 i l { l

      .- . . . _ _ _

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ______ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ . _ , . ____

  *
  .
;           December 19, 1983 l LIMITING CONDITION FOR OPERATION   SURVEILLANCE REQUIREMENT
!         b. Acceptance Criteria for IPCLT The maximum allowable leak rate at P shall not exceed L ( ^

weig$t percent of the co$tained air per 24 hours). The allowable operational leak ' rate, L , which shall be met l prior to increasing reactor

coolant system temperature above 212*F following a test (either as measured or following repairs and retest), shall not exceed I 0.75 L,.

c. Corrective Action for IPCLT ! j If leak repairs are necessary to meet ! the allowable operational leak rate.

l the integrated leak rate test need ! not be repeated provided local leak measurements are conducted and the leak rate differences prior to and j after repairs, when corrected to P and deducted from the integrated I$ak l ! I rate measurements, yield a leakage i ! rate value not in excess of the allowable operational leak rate L ,. ! d. (Intentionally Left Blank) g Amendment No. //*,94 3/4 7-7 g i

  * Correction issued July 1, 198 _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _  _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ __
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_. --- SURVEILLAleCE REQtilRtittill LIMillNG CONutil0N IUR C4'LRAll0N

_ _ _

         ._
    . Local Leak Rate Tests (LLAI)
            (1) Primary containment, testable pentratisses and isolatten valves shall be tested at a pressure of 43 psig except the main steam line isolatter valves shall ha tested at a pressere of 25 psig each operating cycle. selted double-gasketed

! seals shall be tested whenever the seal is closed after helag opened and at least once a during each operating cycle.

l 1 (2) personnel air lock door seals shall be tested at a pressure of 43 psig at least once every 6 eenths. If the airlock is openad when primary containment lategrity is resselred during the laterval heteneen the aheve tests, the air lack deer seals shall be tested at to psig within 72 hears of the first of a series of openin Acceptance criteria and corrective action for l LLRT: * .. If the total leakage rates listed below are 1 exceeded, repairs and retests shall be parfarmed

to correct the conditle (1) (a) A cashined leakage rate of gs 0.60 ty for all penetrattens l

and valves. encept for main steam

isolatten valves, subject to Type 5 and C tests ishen pressurtzed to $ .

             (b) Any one penetration er isolation l            .

valve encept main steam tseletten valves 51 tge (43). ,

No. p. )J. N . 77 August 8. 1981 3/4 7-8

Ainenement h \ p

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o* ,

_ January 12, 1979 ,

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 '

LilflTilNi CONDITIO'4 FOR Op[RATIU;! SURVElllANCE RE')UIREMENT ,

         . *ressure suppression chamber - reactor  -
       (c) Any one main steam building vacuum breakers:     line isolat.fon valve 11.5 scf/hr at 25 l Except as specified in 3.7. A. psi '

helow, two pressure suppression -

'
,

chamber-reactor building vacuum Continuous leak rate monitor: ! breakers shall be operabic at all l times when primary containment (1) When the primary containment

 . integrity is required. The set-    is inerted, the containment l   point of the dif ferential pressure    shall be continuously j   instrumentation which actuates the    monitored for gross  ,l

, pressure suppression chamber- leakage by review of the reactor building vacuum breakers ' inerting system makeup ! 58g shall be from 0.4 to 0.5 psi requirement ,

l From and after the date that one of (2) This monitoring system

the pressure suppression chamber- may be taken out of
reactor building vacuum breakers is service for the purpose made or found to be inoperabic for of maintenance or testing

,   any reason, reactor operation is    but shall be returned to
'

permissible only during the suc- service as soon as j ceeding seven days unless such practical.

vacuum breaker is sooner made i operable, provided that the repair The interior surfaces of the

procedure does not violate primary drywell shall be visually

;   containment integrit Inspected each operating cycle
   .

for evidence of deterioration.

! 4 Pressure suppression chamler - j reactor.hullding vacuum breakers: ,

, The pressure suppression chamber-reactor building
)

1 , vacuum breakers and associated Instrumentation including l

    ,
      '

setpoint shall be chect:cd for i proper operation every three month ! 3/47-9

%.....f. ann t IIF . 5R    -

_ _ _ - - _ _ _ _ . .__ . - _ _ _ l . _ _ _ . _ _ _ _____ July 6. 1978

I t lHillNG CONDI TION FOR Ol'titAlltWe SilRVIILLANC[ R[QtilRLMtNT = j 5, Prenure suppression chamber - drywell Pressu7esuppressionchamber-drywell j vacuum breakers! vacuum breakers:

' When primary a.ontairvnent is required, Periodic operability tests:
 ,  all suppreninn chaniher-drywell vacuum  .

j breakers shall be operable carept Once each month and following any j during testing and as stated in $ peri- release of energy to the suppres-l f icit s oei 3.7.A.S.h and c helow. Sup- sion chamber, each suppression pren ton chamber-drywell vat utan chamber-drywell vacuum breaker breakers shall be operable if: shall be. exercised through one ' open and close cycle and visually (1) The valve is demonstrated to npen ' inspected. Operability of valves, . fully with the applied force at' position switches and lasition

  ,  all valve positions not exceeding   indicators and alarms thall be that equivalent to 0.5 psi as ting  verifie on the face cf the valve dis Refueling outage tests:
   (2) The valve can be closed by gravi-

' ty, when released after being (1) All suppression chamber-manually open, to within not drywell vacuum breaters shall

greater than 0.075 inch or less - be tested to determine the - as. measured at the bottom of the . force required to open eac valve dis valve from fully closed to fully ope (3) Tlie position alahn system will ,

   , annunciate in the control room if   (2) All suppression chamber-l    any valve opening exceeds the   drywell vacuum breaker post-l    equivalent of 0.075 inch as   tion indication and alarm l    measured at the bottom of the   systems shall be calibrated, dis .

l ' . l

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Amendment No. f. 51 3/4 7-10 . ,

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LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT - September 21, 1978

p'to two (2) of the tes.(10) suppression (3) At least two (2) of the i chsaber-drywell vacuum breakers may be suppression chamber-drywell j determined to be inoperable provided that vacuum breakers shall be j they are secured, or known to be in the inspected including internal

! closed positio components. If seating or

friction deficiencies are * l If Specification 3.7.A.5.a or b cannot be noted such that Specification j met, the situation shall be corrected 3.7.A.S could not be me%, all

within 24 hours, or the reactor shall be vacuum breakers shall be j placed in a cold shutdown condition inspected including internal 1 within the subsequent 24 hour components and deficiencies correcte .

i

I i

l

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I i

Amendment No. 51 3/4 7-11

e

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,     - .

June 5 1985 LIMITING CONDITION FOR OPERATION

;      SURVEILLANCE REQUIREfENT

Orygen concentration: 6.

I Oxygen concentration: { After completion of the startup test Whenever inerting is required, the

' program and demonstration of plant primary containment oxygen concentra-electrical output, the primary contain- tion shall be measured and recorded ment atmosphere shall be reduced to on a weekly basi less than 4% oxygen with nitrogen gas whenever the reactor coolant pressure is greater than 90 psig and during

reactor power operation except as -

'l specified in 3.7. A.6.b or 3.7. A.6 Within the 24-hour period subsequent j  to placing the reactor in the Run mode following a shutdown, the containment atmosphere oxygen concentration shall be reduced to less than 4% by volume

and maintained la this conditio Deinerting may commence 24 hours ) prior to a shutdow Orygen concentration may be greater than j 41 by volume for a period not to exceed 48 bours for purposes of conducting i drywell entries relating to testing, surveillances, or maintenance on equipment

.-

important to safety.

! ' If the specifications of 3.7.A cannot be , met, initiate an orderly shutdown and have . i the reactor la a cold shutdown condition within 24 hour Amendment No. 102 3/4 7-12

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*

s-~ , f_ Janusry . 1986

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SURVEILLANCE REQIIIRENENT I

 {  LlHITING CONDITION FOR OPERATION
 "o      B. Standby Gas Treatment System
 , Stjndby Gas Treatment System c At least once per operating cycle, the

. o, Except as specified in Specifications i a 3.7.B.3 and 3.7.B.4 below, both circuits following conditions shall be demon-

 - '  of the standby gas treatment system and  strated:

j the emergency power sources required for operation of such circuits shall be Pressure drop across the combined HEPA j

operable at all times when secondary filters and charcoal adsorber banks is

containment integrity is require less than 7 inches of water at the

system design flow rate (1100 SCFM).

l The results of the in-place cold DOP and halogenated hydrocarbon tests, Inlet heater output is at least SkW.

i at minimum flow rate of 500 SCFM, on HEPA, filters and charcoal adsorber Air distribution is uniform with 120% of the averaged flow per unit across < banks shall show 1 99% DOP removal and 2 99% halogenated hydrocarbon the HEPA filters and charcoal

 ,,
 ~~  remova adsorber n The results of laboratory carbon The tests and sample analysis of i  [  sample analysis shall show 1 90%   Specification 3.7.B.2 shall be g>

radioactive methyl iodide removal at performed initially and at least once a velocity within 20% of actuag per year for standby service or af ter - , system design, 0.5 to 1.5 mg/m 720 hours of system operation and inlet methyl iodide concentration, following painting, fire, or chemical release in any ventila tion zone commu-1 95% R.H. and 2 190*F.

' nicating with the system that could Fans shall be shown to operate contaminate the HEPA filters or within 110% design flo charcoal adsorber , , Cold DOP testing shall be performed after each complete or partial g replacement of the HEPA filter bank or a a after any structural maintenance on i Z the system housin \ e

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J nuary I, 1986

1

o LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT E

, From and after the date that one circuit Halogenated hydrocarbon testing shall-i e of the standby gas treatment system is be performed af ter each complete or l

f

" made or found to be inoperable for any partial replacement of the charcoal reason, reactor operation and fuel
~

handling is permissible only during the adsorber bank or after any structural succeeding seven days unless such maintenance on the system housin '

circuit is sooner made operable,

provided that during such seven days all

active components of the other standby ( gas treatment. circuit shall be operabl . During fuel handling both circuits of the staridby gas treatment system shall be operable, except as stated in para-graph 3.7. In addition, there shall be operable either (a) two sources of t', offsite power (two 345kV or one 27.6kV

and one 345kV) and one emergency power y

source, or (b) one source of offsite

power (345kV or 27.6kV) and two emer-o' gency power sources to operate compo-nents required in paragraph 3.7.B.3.

, If the above cannot be met, procedures i shall be initiated immediately to I ! establish the conditions listed in 3.7.C.la through d and compliance shall

> be completed within 24 hours thereafte a a- 6.

! Primary containment shall be purged !

!" through the standby gas treatment system at all times when primary containment y integrity is required.

i . ' .. W ! O O'

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LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

- Secondary Containment Each circuit shall be operated at 1 east 15 minutes per month and

. l l Secondary containment integrity as defined operated as required to maintain the j ! in Section I shall be maintained during relative humidity in the charcoal i all modes of plant operation except when bed at or below 70 l all of the following conditions are me l At least once per operating cycle, I The reactor is in the cold shutdown automatic initiation of each branch condition and Specification 3. of the standby gas treatment system l 1s me shall be demonstrated, The reactor water temperature is At least once per operating cycle below 212*F and the reactor coolant manual operability of the bypass system is vente valve for filter cooling shall be i demonstrate i No activity is being perormed which I can reduce the shutdown margin below When one circuit of the standby gas that specified in Sepcification 3. treatment system becomes inoperable the other circuit shall be demonstrated

to be operable immediately and daily thereafte Secondary Containment Secondary containment surveillance shall be performed as indicated below: A secondary contalmnent capab111ty test shall be conducted after isolat-ing the reactor building and placing either standby gas treatment system filter train in operation. Such tests shall demonstrate the capability of the secondary containment to maintain a 1/4 inch of water vacumn with a filter train flow rate of not more than 1100 scfm. Secondary containment

 . . .. .e  ...i.,, ..,,  ,. , i,

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__

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SilHVilLLANCf RtQUIR(MtCT tlMlllNG CONDITION FOR OPtRAll0N - - _ . . lhe fuel cask or arradiated fuel is capability shall be desmonstrated at not lieing moved within the reactor three or more points within the buildin . contaisunent prior to fuel movement and may be demonstrated up to 10 days Primary Containment isolat ioni valves prior to fuel inoveinent. Secondary

containment capability need not be . During reactor power operating conditions, demonstrated more than once per i 1.

j all isolation valves listed in Table 3. operating cycle unless damage or and all instrument Iine flow check valves modifications to the secondary shall be operable except as specified in contaisument have violated the integrity-3.7. of the pressure retaining boundary of that structur ' Primary Containment isolation Valves

The primary containment isolation valves I

survelliance shall be performed as follows: 1 At least once per operating cycle l the operable isolation valves that are power operated and automatically initiated shall be tested for , simulated automatic initiation and closure time ~ At least once per operating cycle the instrument line flou check valves

'
.

shall be tested for proper operatio i At least once per quarter: 1) All normally open power-operated isolation valves (except for the main steam-line powet-operated isolation valves) shall be fully closed and reopened.

!

' Amendmentho./I,76 April 16,1981 3/4 1 14

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noe r s. r ., ruimnv runiAim.txt.ign A.r.iHn ,

          ,
        .
    '

nuadier of Power Isulatlen 7alve (Valve Operated Valves

Group I:lentification_ Number) Operating initiatin1

inboard Outboard Time (sec) Position p_I_

. ! 4 3<T<5 0 GC

neln Steau Line Isolatfon (MS .lA, 2A,18, 20, IC, 2C, 4 I - -

10,20) 1 35 C SC 1 MainSteamLineDrain(MS-5) 1 35 C SC l 1 Mein Steam Line Drain (MS-6) 1 5 C 5C j 1 Recirculation Loop Sample Line ($N-1, 2) 1 5 0 GC l

1 Isolation Condenser Vent to Main Steam Line(IC-6, 7) 20 0 GC ! 2 l 2 Drp ::11 Floor Drain (55-3, 4) 2 20 0 GC 2 Dryuell Egippent train (55-13,14) 1 10 C 50

, 2 DrytT11 Vent (AC-7)     1 15 C 50 l 2 Dryrell Vent pelief (AC-9)     1 10 C SC i 2 Dryn11 and Suppression Chamber Vent from Reactor    '

Dellding(AC-8) 1 10 C SC 2 Dryrell vent to standby Gas Troatment System (AC-10)

1 10 C SC l 2 Suppression Chamber Vent (AC-11) ' 1 15 C SC ,

2 SwPPetssfoi Chamber Vent Relief (AC;12) 1 10 C SC I 2 Swppression Chamber Supply (AC-6) / - 1 10 C SC j 2 Dryeell Supply (AC-5) -

1 10 C SC ' 2 Drywell and Suppression Chamber Supply (AC-4) 18 0 GC

3 Cleare ,Deminerailrer System (CU-2) 2 18 0 GC l i l 3 Cleanup Deminere11rer System (CU-3, ,28) 48 C SC '

] 3 50-1) 48 C SC

Shutdow ShutAcwn Cooling Cooling System System (iSD-2A,28,4A,43) 4 48 C SC

3 1 3 45 C SC

!

'3

Shutdown Cooling System [50-5) ) (IC 1) Reactor Head Cooling Line (HS-4 Isolation Condenser Steam Supply - 1

24

0 GC GC , i

i 4 isolation Condenser Steam Supply (IC-2) 19 C SC

1 4 Isolation Condenser Condensate Return (IC-3) 1 19 0 04 j 4 tsolation Condenser Condensate Return (IC-4) M 0 Process Fecs%ter Check Valves (FW-9A 104, 98,108) 2 2

M 0 l'rocru Centrol Rod Hydraulic Return Check Valves (301-g5, 98) 1 M C Proccu Reactor Head Cooling Check Valves (HS-5) 1 M C Process

 . Stam"ey 1.iwid Control Check Valves (5L-7, 8)  1 l
,

3 Clear-sp Donineralizer System (CU-5) I la C SC 3/4 7-15 l .-- ...- -. . # 24.76 anr,i m ing

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

i i TABLE 3.7.1 ! ! Key: 0 = Open C = Closed ! SC = Stays Closed ! GC = Goes Closed NA = Not Appifcable e i note: Isolation groupings are as follows: ! GROUP 1: The valves in Group 1 are closed upon any one of the following conditions: I Reactor low-low water level (This signal also trips the reactor recirculation pumps.)

! Main steam line high radiation.

l 3 Main steam line high flow.

l ' Main steam line tunnel high temperatur . Main steam line low pressure.

1 . I GROUP 2: The actions in Group 2 are initiated by any one of the following condittoas: Reactor Iow water'1eve . High drywell pressur i G400P 3: Reactor low water level alone initiates the following: Cleanup domineralizer system isolatio ' Shutdown cooling system isolatio . Reactor head cooling isolatio I GROUP 4: Isolation valves associated with the isolation condenser are closed upon indication of either high isolation t condenser steam or condensate flo l

l !

i

,
   -    3/4 7-16
          .
-

_ _ _

.- - -. . -  .. . . -     ,

4 ,'~~ , % }

.
.
    '    )
:
    -- -    .

i LIMITING CON 9fil0N FOR OPERATION , SURVilltANCE flEQUIRtt4NT . . In the event any isolation valve specifitd ?) Wis h the reactor power le u lis.in

in Table 3.7.1 becomes inoperable. scartor 75% of rated, trip main sle am
i'.silatinn valves (one aL .i time)

power operation may continue prowided at j least one valve in each line having an and verify closure tim i inoperable valve is in the mode correspond-ing to the isolated conditio At least onc e per month, the spain l

steam-line power-operated isolation l

If Specification 3.7.0 cannot be met, valves shall be exercised by partial j closure and subsequent reopening.

initiate an orderly shutdown and have reactor in the cold shutdown condition i

within 24 hour . Whenever an isolation valve listed in

! Table 3.7.1 is inoperable, the position of at least one other vdive in each line i ! having an inoperable valve shall be recorded daily.

I

l

! ! .

, l j .

, ! I '

4 ! I , Amendment No. 76 April 16,19R1 3g 7,;7 _

l l e

:>=

I

Z

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Z w @

 ~    -

u n Z N

 <    M J

CI3

>-

k baJ J I l

1 t l [ I

I l .

i !

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

_____ . _ _ _ _ _ . _ _ . _ _ . . _ ___ . .

f

*
 *

April .., 1981 , .. LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT-

..

3.9 AUXILIARY ELECTRICAL SYSTEM 4.9 AUXILIARY ELECTRICAL SYSTEM , Applicability: Applicability: Applies to the auxiliary electrical power syste Applies to the periodic testing requirements of l the auxiliary electrical system.

i Objective: Objective: I I To assure an adequate supply of electrical power during plant operatio Verify the operability of the auxiliary electrical * j syste Specification: , Specification: 1 The reactor shall not be made critical unless all Emeraency Power Sources ) of the following conditions are satisfied: ( Diesel Generator i One 345 kv line, associated switchgear, and i auxiliary startup transformer capable of The diesel generator shall be started j automatically supplying auxiliary powe and loaded once a month a demonstrate ! Both emergency power sources are operabl operational readiness. The test shall j

continue until the diesel engine and An additional source of power consisting of the generator are at equilibrium one of the following: temperature at full load output. Dur-ing this test, the diesel starting The 27.6 kv line, associated switch- air compressor will be checked for gear, shutdown transformer to supply operation and its ability to recharge j air receivers.

i power to the emergency 4160 volt buses.

1 One 345 kv line fully operational and During each refueling outage, the capable of carrying auxiliary power conditions under which the diesel i to the emergency buse generator is required will be simulated and test conducted to demonstrate that it will start and be ready to accept 4 volt buses five and six are energized and the associated 480 load with in 13 seconds'. volt buses are energized.

'3/4 9-1

.

*-

December 6, 1985 * LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT All station and switchycard 24 and 125 volt During the monthly generator test, batteries and associated battery chargers the diesel fuel oil transfer pumps are operabl shall be operate B. When the mode switch is in Run, the availability Cas Turbine Generator of power shall be as specified in 3.9.A, except as specified below: The gas turbine generator shall be fast started and the output breakers From and after the date that incoming power closed within 48 seconds once a month is available from only one 345 kv line, to demonstrate operational readiness, reactor operation is permissible only The test shall continue until the during the succeeding seven days unless an gas turbine and generator are at additional 345 kv line is sooner placed in equilibrium temperature at full load servic output. Use of this unit to supply power to the system electrical net- From and after the date that incoming power work shall constitute an acceptable is not available from any 345 kv line, demonstration of operabilit reactor operation shall be permitted pro-vided both emergency power sources are During each refueling outage, the operating and the isolation condenser system conditions under which the gas turbine is operable. The NRC shall be notified, generator is required will be simulated within 24 hours of the precautions to be and a test conducted to verify that taken during this situation and the plans it will start and be able to accept for restoration of incoming power. The emergency loads within 48 second minimum fuel supply for the gas turbine during this situation shall be maintained Batteries above 20,000 gallon . Station Batteries From and after the date that either emer-gency power source or its associated bus is Every week the specific gravity and made or found to be inoperable for any voltage of the pilot cell and tem-reason, reactor operation is permissible perature of adjacent cells and overall according to Specification 3.5.F/4.5F unless battery voltage shall be measure such emergency power source and its bus are sooner made operable, provided that Every three months the measurements during such time two offsite lines (345 shall be made of voltage of each or 27.6 kv) are operabl cell to nearest 0.01 volt, specific gravity of each cell and temperature

Amendment No. if , [, 107 3/4 9-2

-_________-__ _ _______-___ ___ _ _  _ _ _ _ _ _ _ _ _ - - _ _ _ _ _
  .

-

..

1 , LINITING CONDITION FOR OPERATION June 21, ,d4 ,

SURVEILIANCE REQUIREPENTS

! i

From and after the date that one of the c. The following tests will be performed in accord-i two 125 volt or 24 volt battery systems is ance with IEEE Standard 450-1975 "IEEE Recommend-

[ made or found to be inoperable for any ed Practice for Maintenance. Testing, and Replace I

j reason reactor operation is permissible i only during the succeeding seven days ment of Large Lead Storage Batteries for Generatingi Stations and Substations."

] unless such battery system is sooner made

operabl . ! At least once every refuel outage, a battery service test will be performed Diesel and Gas Turbine Fuel in accordance with section 5.6 of IEEE Standard 450-1975 to verify that the i ' There shall be a minimum of 20,000 gallons of battery capacity is adequate to supply diesel fuel supply on site for the diesel and and maintain in operable status all of a minimum of 35,000 gallone onsite for the gas the actual emergency loads for 2 bour ;

turbine, except as permitted in Specification 3.9. . At least once every 60 months, during shutdown, a performance discharge test will be performed in accordance with f Section 5.4 of IEEE Standard 450-1975 to verify that the battery capacity is j at least 80 percent of the manufacturer's ,

rating. Once per 60-month interval, this ! performance discharge test may be performed I in lieu of the battery serv $ce test.

l i

1 l

1 l , Amendment No. 99

)       3/4 9-3

  ,_
         .. ____

LIMITING CONDITION FOR OPERATION June 21, 1984 SURVEILLANCE REQUIREMENTS Switchyard Batteries Every week the specific gravity and voltage of the pilot cell and temper-ature of adjacent cells and overall battery voltage shall be measure Every three months the measurements shall be made of voltage of each cell to nearest 0.01 volt, specific gravity of each cell, and temperature of every fifth cel C. The quantity of gas turbine generator and diesel generator fuel shall be logged weekly and after each operation of the uni Once a month a sample of the diesel and gas turbine fuel shall be taken from the under-ground storage tanks and checked for quality.

i l

i

3/4 9-4 , Amendment " 99 ,

}

 - _ _ _ _ _ _ - - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _
,
'
' r #ffff5x rfve'd by Stpon Superintendent
           /2- Yl/

Effective Date DOCUMENT SUMWARY SHEET Revision No G Document Title thrat .I benco Nf>m leve/4 , Document N PIP Srm 4 DI-$? Originator LO. Buc l Summary: v Minor Hou'sekeeping - title changes, typographical errors, reference changes, et Change resulting from NRC regulations, bulletins, etc. (indicated in body of procedure per PAP 1.02). See action / description belo Change resulting from a Station CR (audit finding, program improvement, program expansion,etc.). See action / description belo Other See action /cescription belo AcMon: Personnel indicated on SF 330 read procedure as revise Personnel indicated on SF 330 read description below at a minimu (procedure may be reviewed as time permits).

riotion: _ CA -# u 2- *

         ,u=--

F (- UfA w=h Y

-.

Date

            #' ' / ~ ^ ^

reved: <'y ./ [K/

;            Effectise Station Su # ntencent SF-329 Rev. 1 Date: 12/4/E1
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varve l 8.4 PRIMARY CONTAINMENT HIGH PRESSURE Panel 905, Section A-2, Window 3-3 8. PS-1621A, B, C, . Initiating Device: I-Setpoint: psi . . Automatic Action: 8.4.4.1 Reactor Scram Gas Turbine Generator starts 8.4.4.2' 8.4.4.3 Diesel Generator start Core Spray System initiates 8.4. .4.4.5 LPCI system initiates 8.4.4.6 FWCI System initiates.

l 8.4.4.7 Group II Isolation.

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L APPENblX / QUESWDN 8.09

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;-c.; . 2:t :1 . TEST CROSS REFERENC '
-

PAGE 1-QUESTION VALUE . REFERENCE: ________ .______ __________

" LO5,01; -3.00 AXAOOOO220

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-

2.25 AXAOOOO224

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. a TEST CROSS REFERENCE PAGE 2
. QUESTION VALUE REFERENCE

________ ______ __________

~O8.08 2.00 AXAOOOO344 08.09 2.50 AXAOOOO338-08.10 2.50 AXAOOOO339 08.11- 3.00 AXAOOOO340

______ 25.00 _ _ _ _ . _ _ ______ 100.00

I ! ' . l

ATTACHMENT 3 Facility Comments and NRC Resolutions of Comments on Written Examinations The following represents the facility comments and the NRC resolution to those comments made as a result of the current exam review polic Only those comments resulting in significant changes to the master answer key, or were "not accepted" by the NRC, are listed and explained below. Comments made that were insignificant in nature and resolved to the satisfaction of both the examiner and the licensee during the post exam review are not liste i.e.: typo errors, relative acceptable terms, minor set point change SR0 EXAMINATION 5.06 FACILITY COMMENT: Candidates are not required by training objectives to know order of severity of MCPR related plant transients. Candidate is required to know the worst single transient, which is generator load rejection without bypass for this cycle. Also, the heat transfer text is a generic document and not necessarily plant specific on the order of transient severit Suggested Resolution: Accept for full credit generator load rejection without bypass and any two others listed in the H.T. Text on pages 9-36 through 9-39, with the excep-tion of inadvertent HPCI pump start. This includes recirculation flow controller failure since this transient is the most severe when initiated from low flo Resolved During Revie NRC Resolution: Comment accepted. Answer key will reflect suggested resolutio .08 FACILITY COMMENT: No comment on order of preference on methods to assure adequate core cooling. Concerning the reason for the order, however, there are several points which might be brought out to substantiate the answer.

' Specifically, candidates were trained that heat transfer coefficient alone was not solely the reason for the order of preference; and, in fact steam may actually have a higher heat transfer coefficient than spray cooling, but removes less heat because of the lower temperature difference between steam and the fue Suggested Resolution: In addition to reasons given in Appendix B to E0P's consider above fact in gradin . - .- __. - - - _ _ , _ -_ _. .-- . _ _ - . - . . - - . . . - . . . . --

     , ..  - .  .   . .. -

2 -

2, Attachment,3,.SR0 Examination-Reference:

   - Millstone I Requal Material, Mitigating Core Damage, Part I, Section F- (This material was used as reference documentation for E0P Training).

_ NRC Resolution:- ^ Suggested Resolution Accepted. This additional information was not sub-

mitted to the NRC as part of the training material for exam preparation.

t- ! -6.06.B.. FACILITY COMMENT: . - Depending on radioactivity of the primary coolant the isolation condenser . ' vent rad monitor may increase enough to cause the alarm due to the hig background. The vent alarm comes in at a low setpoint (10 mr).

Suggested Resolution: , Accept isolation condenser vent rad monitor as an answer.

! - NRC RESOLUTION: f Comment not accepte Isolation condenser text Section 1307 specifically-states the cause as a tube leak. This item is also specific training objective 9e.--Also, OP 307 " Isolation Condenser System" has notes referenc-ing the cause as a tube leak. No mention is made of high radioactivity levels in the primary coolant in any of .the reference material submitted , by the-facility.

. 6.06C. FACILITY COMMENT: Order of preference for make-up to Iso. Condenser is fire water first then condensate transfer. Demin water is preferred over the other ' i -two only initial fil l Suggested Resolution: 1 i

Since conditions stated in the question have the isolation condenser initiate at power the question is answered for the preferred makeup under ! these conditions. With fire water available, this source would be lined up and preferred.

4 Reference: 1 Text 1307 Page 2 & 3 , Procedure 307 Ops Form 307-1 (Valve line up) l NRC RESOLUTION: i l- Comment accepted.

! ! ! ! $ !

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

3 Attachment 3, SR0 Examination 6.07 FACILITY COMMENT: Battery Room exhaust fan is powered from MCC-F-5, not MCC-E-5. Text Erro Suggested Resolution: Omit Battery Room exhaust fan from answer ke Reference: 25202-3005 25202-3003 NRC Resolution: Comment not accepted. This question was developed from "125 VDC Elec-trical" Text B44A and specific training objective 9. The above references were not included in the material sent for exam preparation nor in the information supplied with the comments. Therefore, the " text error" the facility has cited cannot be verified.

6.09 and 3.07 FACILITY COMMENT: Comment 1 - CRD Cooling Water flow is normally around 55 gpm. System flow is around 75 gpm. Examinee may have been confused by the wording of the question and answered based on what happens to system flow following a scram. Consideration should be given to this possible interpretation in gradin Comment 2 - Concerning the response of Cooling water flow on a scram, cooling water flow actually goes to a minimum value instead of zer This is because the FCV does not physically close completely when a full close signal is applied to the controller. The valve passes 3-9 gpm through the disc in order to keep the lines ful (See the process flow diagram attached.) This information was not included in updated training materials even though it has been taught previously. Indicated cooling water flow would be practically zero, howeve Suggested Resolution: Consider this fact in grading as appropriat Reference: Attached FCD 29122 sh. 66

4 Attachment 3, SR0 Examination NRC RESOLUTION: Comment #1 not accepted. Although given flowrate is high, the intent of the question is clear. Candidates did not ask for clarification during the exa Comment #2 accepte Flow rates of 0-10 gpm will get full credit. This information not initially provided for exam preparation. Acceptance of comment is based on additional informatio .11 FACILITY COMMENT: The exam answer key was found in error, due to a calculation error in determining the APRM hi flux setpoin Suggested Resolution: Remove selection "e" as a correct answe Resolved During Revie NRC RESOLUTION: Comment accepte .03 FACILITY COMMENT: , On a loss of 125 VDC, some 480 VAC breaker control power is lost, not all. Also, the control room annunciators are lost.

"

Suggested Resolution: Mention of the control room annunciators should bo acceptable for credit in place of either the loss of indicating lights or the loss of some breaker control powe Reference: PNP-506 Page 3 NRC RESOLUTION: Comment partly accepte Loss of control room annunciators may be accepted as an alternate to loss of breaker indicating lights only. Also the answer key does not imply that all breaker control power is lost. It is simply a , direct quote from ONP-50 ,. _, _ _ _ . . _

m ! c l l I 5 Attachment 3, SR0 Examination 7.04-2. FACILITY COMMENT: The answer states that the APRM setdown following a load rejection is to preserve MCPR margins after the select rod insertion. Actually, the APRM setdown is done to preserve MCPR margins in the event of a failure of the select rod inser Suggested Resolution: Accept for credit that the APRM setdown is a backup to the select rod insert on a load reject.

> Reference: ONP text 1500A Page 27 Tech Spec Page B2-7 NRC RESOLUTION: Comment accepted. This comment highlights an inconsistency between the referenced lessen plan and the referenced Technical Specification paragrap .06 B. FACILITY COMMENT: The answer key states that the RWCU System is isolated as a subsequent action on a loss of RBCCW to protect the RWCU resin from overheatin Although this may be a concern, the system is designed to automatically isolate at 140 *F outlet temperature of the non-regenerative heat exchanger. The RWCU System is isolated under these conditions to reduce heat loads on the RBCCW Syste Suggested Resolution: Accept the answer on the answer key or that the RWCU System is isolated to reduce RBCCW heat load Reference: ONP Text 1500A Page 175 NRC RESOLUTION: Comment accepte _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _

. 6 Attachment 3, SR0 Examination 7.10 C. FACILITY COMMENT: This question requires the candidate to respond true or false to tempera-ture limits in the CRD System. These temperature limits are contained in the CRD System operating procedure, however they are not monitorable either in the control room or locally. These points are monitored via contact thermometers by the mainterance department after work has been done to the CRD pumps to ensure the work was done properly. However, there are no permanently installed temperature detectors at either point referenced in the questio Based upon the above facts and the true/ false format of the question, a candidate would not normally be exposed to this information and not have the opportunity to explain these points on the written examinatio Suggested Resolution: Remove part "C" from the question and redistribute the 2.0 points among the remaining three part NRC RESOLUTION: Comment accepted. Answer key will reflect above suggested resolutio It is suggested that due to the difficulty in monitoring these parameters a caution statement requiring these limits be followed should also indicate how they are monitore .11 FACILITY COMMENT: The question asks for four items of information a person should report to the control room if a fire is discovered. The second immediate action of this procedure is to " ensure there are no injured personnel". If there are injuries, this information should also be immediately transmitted to the control room such that medical assistance can be calle Suggested Resolution: Accept " notify the control room of any injured personnel" as one of the four correct immediate action Reference: ONP-505 Page NRC RESOLUTION: Comment not accepted. The question states what four (4) items are reported to the control room when a fire is discovere The answer key requires only what is explicitly stated in ONP 505 " FIRE" as an immediate action. ONP-505 does not explicitly require injured personnel data to be reported to the control roo , e

7 Attachment 3, SR0 Examination 8.04 FACILITY COMMENT: Valve numbers indicated by annunciator for Suppression Chamber to Reactor Building Vacuum Breakers are inconsistent with P&ID and electrical drawings. Numbers for these valves are AC-3A, AC-3 Candidate may have read question and discussed action based on a Suppression Chamber to Drywell Vacuum Breaker proble . Question did not specify that the vacuum breaker was not fully closed, only that the limit switches that give indication and alarm functions were not aligned to the " closed" position. Student may have assumed the valve closed and only a limit switch problem. (i.e., a detector alignment problem) Suggested Resolution: Credit should be given if student identified the vacuum breaker problem to be a Suppression Chamber to Drywell Vacuum Breaker pro-vided the student followed the correct action per Tech Spec Section 3.7. . Credit should be given to student who stated his assumption that valve was indeed closed and only a limit switch problem existed, provided his actions were in accordance with Tech Specs Section 3.7. Reference: P&ID 187477; CWD Sheet 813 NRC RESOLUTION: Comment #1 not accepted. Millstone Training text LP 1311 A " containment" P. 47 clearly states the alarm given in the question is for the reactor building to suppression chamber vacuum breakers. The question also states that the reactor building to suppression chamber vacuum breakers were being tested. Further, the candidates did not ask for clarification of what valve was affected during exam administration which indicates that the problem was understood. Therefore, answers relating to Tech Spec 3.7.A.5 are incorrec Comment #2 not accepte The question provided two clear indications that the valve was open or at best the actual position unknown. Without some other means of verification of actual valve position an assumption of an alignment problem with limit switches is non-conservative. Therefore, resolution of the problem per Tech Spec 3.7.A.4. is incorrec .07 Verbal comment that the injured man may be assumed contaminated since most of the refuel floor is contaminated. Thus an unusual event could be declared. Answer key modified to reflect this possibility.

C _ ___ _ _ _ _ _ _ _

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8 Attachment 3, SR0 Examination l 8.09 FACILITY COMMENT: Appendix A, handout given to students during exam, was found to be incorrec Pressure switch PS 1621A-D do not control auto start of Core Spray, LPCI, Gas Turbine or Diesel Generator. PS 1501-90A-D is utilized to perform this function.

Suggested Resolution: l Candidate answering the question with the response that PS 1621A-D may be ! isolated should be correct since Tech Specs does allow for isolation of this pressure switch during inerting.

I References: CWO Sheets 760, 785, 740, 751 l P&ID Sheets 25202-26009 Sheet 1 NRC RESOLUTION:

Comment not accepted. Appendix A given out during the examination was a l copy of the bottom portion of P.21 of OP-311 Revision 19, " Containment System." This was the only place this information was provided by the facility and was given to the candidates for reference in evaluating the problem. The candidates did not ask any questions or make any statements about the functions of these pressure switches during the exam. There-fore, allowing isolation of the pressure switches based on the information given in the problem is incorrect.

l l l l l I l !_____-______-_________-______-_______________________________________

y j- .

!

R0 EXAMINATION-1.01 b. FACILITY COMMENT: ! ' One additional observation that assures that the system is properly filled is the LPCI Heat Exchanger dp indication on panel 903. This normally indicates the difference between ESW pressure (which is service water pressure of approximately 50 psig when the ESW pumps i are not running) and the LPCI System pressure. If this indicates approximately 15 psid, then LPCI pressure is'approximately 35 psig, or is l properly filled.

L l Suggested Resolution:

l' , Accept LPCI/ESW Heat Exchanger dp as indication that system properly l filled.

.-

l Reference: LPCI P&ID 26008, J-18; SW Text 1321, Page 12, Section 6.1.3.1; LPCI Text 1335, Page 18, Section I NRC RESOLUTION: Since the dP is a comparison of the conditions in the two systems, the candidate must give an explanation of the varying conditions in the two systems for credit to be given for this answer. Pressure in LPCI could be high, or pressure in ESW could be low (as was noted during simulator operation). It also should be noted that the system could indicate 35 psig and still have an air bubble in it if it had not been vente The system text should be updated with the supplied informatio .02 b. FACILITY COMMENT: l Operator is directed by operating procedures referenced below to monitor j " vessel level, temperature, and pressure" as an indication of proper ! natural circulation without recirc pump operation (Lack of thermal i stratification, pressurization, and heatup indicate that proper natural circulation is present. Temperatures monitored may include coolant l -temperature or vessel skin temperatur Suggested Resolution: ! Accept vessel level, coolant temperature, skin temperature, and vessel , pressure as answer R_esolved e During Revie ' i , ! .

2 Attachment 3, R0 Examination

Reference: OP206, Step 4.11; OP301, Step ~ NRC RESOLUTION: The answer key was changed to include the parameters listed in the procedur .03 FACILITY COMMENT: Comment 1 - Question asks how "MAPLHGR" is affected by various situations, ; but intends to ask how "MAPLHGR" Limit" is affected. Comment Noted During Revie Comment 2 - Concerning answs s for parts b, c, d. The heat transfer text discusses the reasons for the shape of the MAPLHGR Limit Curve consistent with the answer key. However, the text points out that these phenomena . cause the limit to first " increase at a decreasing rate, then decrease."

l There is no differentiation as to when the decrease starts due to a particular effect. An examinee may have properly understood the impact of a particular effect on the MAPLHGR limit, but indicated that the limit is still increasing at a decreasing rate.

l Suggested Resolution, Comment 2: l Examinee should be given credit for " increasing" as an answer to Parts b, c, or d, if proper explanation of above thought process is given.

! NRC RESOLUTION: The suggested resolution would be acceptable if proper assumptions and support were given.

l 1.08 a. FACILITY COMMENT: ! l Pump shutoff head may be developed anytime a pump is operated against

a pressure in excess of its capability to provide flow, not just if the discharge valve is close Suggested Resolution

Accept a definition of shutoff head which describes pump being operated with no flow due to resistance in excess of its capability even if it does not specifically say " discharge valve closed". Resolved During Review.

l l

p-

-
 .

3 Attachment 3, R0 Examination , s NRC RESOLUTION: Alternate wording was noted to be acceptabl .04 FACILITY COMMENT: Question may lead examinee to explain how he can raise switchyard voltage.

Suggested Resolution

l \ Accept the following: ' increasing generator voltage / excitatio j J There are a variety of voltage indications available to the operato Suggested Resolution: Accept the following reading line voltages on GETAC (CRP909), Bus voltages on 908 and use 480V bus voltage to determine if RPS MG set has exceeded its low voltage limi NRC RESOLUTION: I For Part b, the resolution was accepted for partial credit. CONVEX must be notified before this action can be taken, and this was required. For

Part c, the facility's resolution was acceptable, plus several additional items given by the candidates were researched during the grading and accepted for credit. At least two items were required for full credi The system text should be updated with the supplied informatio .05 FACILITY COMMENT: b. & The part of this question worded "the drywell pneumatic air supply to the APR valves has been lost" can be interpreted two way Either there is no air supply to the valves or there is no air supply to the valves and accumulators Suggested Resolution: Accept as full credit the knowledge that air pressure is required to operate the valves electrically (ApR or switch) and if a student assumed the leak caused a loss of all air pressure for valve operation he would not consider the accumulators in his answe Resolved During Revie _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

.-

  ,
 .

4 Attachment 3 , RO Examination Reference:

i

APR Text 1337 Figure 5 NRC Resolution: This was acceptable if the candidate included a discussion of where the

. break could have occurred. It was required that the candidate include knowledge of the accumulator .09 FACILITY COMMENT:

The answer key lists only one reason for lowering power prior to removing HP heaters from service near end of core life. The procedure and the lesson plan list two reasons: To prevent exceeding fuel preconditioning limits and to minimize radiation exposur Suggested Resolution: l Accept'for full credit student knowledge of the power increase due to inlet subcooling increase, and radiation exposure, with half credit for eac Reference: . Procedure 346 Page 5 Lesson 1346 Page 14 Objective 15 NRC RESOLUTION: The answer key was changed and both answers were required for full credi .11 b. FACILITY COMMENT: The compressor control switch is always spring return to normal (neutral) and therefore switch position has no bearing on whether the compressor will start after an LNP or not. The operating conditions part of the question may be interpreted as plant conditions, i.e., diesel on 14F and 25 second time dela Suggested Resolution: Do not consider control switch position in the answer and consider plant conditions in lieu of previous compressor running condition Reference: Control wire diagram 25202-31001 Sheet 393 (attached)

.. ~

5 Attachment 3, R0 Examination NRC RESOLUTION: The accepted answer was diesel on 14F (0.375 pts) with a 25 see time delay (0.375).

3.01 FACILITY COMMENT: RPS will initiate a separate scram on low condenser vacuum at 23" Hg prior to the turbine trip /stop valve closure scram at 22.5" Hg. Examinee consi-dering integrated plant response to a loss of vacuum may discuss this scram function instead of or in addition to the stop valve scram since it occurs ( prior to and independent of the 22.5" Hg tri Suggested Resolution: For RPS response, accept scram initiated at 23" Hg vacuu Resolved During Revie Reference: RPS Text 1408 Page 30-31 Section 5. NRC RESOLUTION: Either answer was accepted for Item 4 of the answer key.

3.02 FACILITY COMMENT: Although one text reference, indicates that the IRM's should come on scale between 10 and 10 cps with SRM's full in (see Ref 1 below), other references (2) do not stipulate a lower limit of 10 cp Suggested Resolution: Do not require 10 lower limit on overlap condition, only upper limit of 10 * Resolved During Revie Reference: IRM Text 1402A, Page 27, Section 6.1. . SRM Text 1401A, Page 2, Section According to text references and Technical Specifications, proper IRM/APRM overlap requires only that the associated APRM channel is operable and greater than its 3/125 downscale trip prior to with-drawing the IRM's in order to ensure proper neutron flux indication and instrument overlap. IRM text also gives an example of proper overlap. IRM text also gives an example of proper overlap with approximate values for instrument readings. We feel that the example

_ _ _ ___ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ - _ - l l 6 Attachment 3, R0 Examination i given in this question (2% power IRM and 7% power APRM) meets the overlap criteria and that the values do not differ significantly j enough to warrant suspension of instrument malfunction by the opera-tor. This is especially true since the APRM average values are expected to be inaccurate at low power levels before reaching the target rod patter Suggested Resolution: Answer key should be changed to "Yes" for same reason as given. Note that during exam review, we stated that a one decade overlap is accepted practice. While this is true for SRM/IRM overlap (Reference 2 above), a reference for this requirement for IRM/APRM overlap could not be found. However, we feel that change in the key is justified based on the above explanatio Reference: IRM Text, Page 23, Section 6.1.2 and page 28 Obj 16 Tech Spec Table 3.1.1, Note 5 NRC RESOLUTION: This was accepted. The key point looked for in the grading was that the IRMs must be on scale prior to withdrawal of the SRM The system text should be updated with the supplied informatio Review of this comment and all the referenced material resulted in the accepted answer being taken from the Tech Spec statement that the APRMs must be above the downscale alarms prior to with-drawing the IRMs. Thus the correct answer is "yes" with this explanation. The system text should be updated with he supplied information.

3.06 a. FACILITY COMMENT: If the EPR is lost, the MPR will automatically assume control. No opera-tor action will be require Further, if the ABT occurs properly, the EPR will not be lost completely but will resume control after the transfer to the emergency source. In fact, if the transfer is fast enough, there may be no noticeable effect on the pressure control syste Suggested Resolution: Accept " automatic transfer to MPR" if power lost completely. Accept "EPR resumes or maintains control" if ABT assumed to be rapid. A change will be made to the Vital / Inst. AC Text to clarify these point Reference: Main Turbine Text 1314 Page 121, Sect 5.16.3 Page 167, Section 6.1.1 Attachment 3, R0 Examination NRC RESOLUTION: Comment accepted. Plant committed to change system tex .08 FACILITY COMMENT: Concerning the response of the MPR Servo to an EPR loss of power, the MPR stroke will increase slightly as reactor pressure increases during the changeover from EPR to MPR control. The MPR Servo position is determined by the difference between the MPR setpoint and reactor pressure. A change in either parameter will cause the Servo position to change, so that the MPR is calling for a different control valve position (whether or not it is in control). However, this change will be slight for the normal MPR pressure setpoint 10% higher than EPR and is not a practical consideration. The Turbine Text is vague on this fine point, but an examinee who has studied the system in more detail due to job background or a special interest in the turbine control system may have considered

it significant in his answe Suggested Resolution: Accept either "no change" or " slight increase" for MPR Servo response

during this event, f Reference

Simulator response to this event Turbine control manual GEK-17955 (available on request) ' NRC RESOLUTION: Comment accepted. Answer key updated. The system text should be updated if this information is considered significant as stated in the above commen .09 FACILITY COMMENT: Examinees may not consider annunciator alarms as " automatic actions". As a result of the actuation of Standby Gas Treatment System, the reactor building supply and transfer fans trip and their associated dampers clos The same is true of the Steam Tunnel Ventilation Supply and Exhaust Fan Suggested Resolution: In addition to answer key, accept the following as automatic actions: tripping of reactor building supply and transfer fans; closing of reactor building isolation dampers; tripping of steam tunnel ventilation supply and exhaust fans; an.1 closing of steam tunnel isolation damper _ _ _ _ _ _ _ _ _ _ _

r 8 Attachment 3, R0 Examination Reference: HVAC Tex 1327 Pages 38-40; 42-47-NRC RESOLUTION: In lieu of the candidate stating that the Reactor Building or Steam Tunnel Ventilation Systems " isolate", what happens within these. systems was accepted. This is just alternate wording. Also, since alarms weren't specifically requested, credit was given only for the first three items in the key at 0.67 points eac .11 FACILITY COMMENT: Wording of the question may have elicited different response from examinee than desired. The question may be interpreted as asking why a scram occurs on valve closure as opposed to reactor pressure if four MSIV's are close Tech Spec bases discuss the fact that it is necessary to scram the reactor on valve position in anticipation of the pressure and flux transient that will occur as a result of the closur Another possible scenario considered by the examinee could be that even if the valve closure results in isolation of only two steam lines, a subse-quent Hi steam line flow isolation of the other two lines will result, causing a scram on valve closure in that case. In discussing this question with the examinees, it appears that some individuals who asked a specific question on this point were told to consider only valve logic. However, a general announcement to the class was not mad Suggested Resolution: Consider the above points in the gradin Reference: Tech Specs. Bases, Section 3.1, Page 3/3 1-4 NRC RESOLUTION: This comment was not accepted. The question clearly states not to consider pressure effect .01 a. FACILITY COMMENT: The first part of this answer mentions that the reason the operator trips the turbine is to assure that the turbine does not become a system loa Another reason for tripping the turbine is to prevent it from tripping automatically on reverse power, that is, to take operator action rather than relying on an automatic interlock.

< __ _ ___ --___ _ _____ . ,. 9 Attachment 3, R0 Examination The reason for waiting until turbine load is less than 50 MWe is to mini-mize the resultant pressure transient when the turbine is tripped. How- l ever, since Millstone Unit _1 is designed to remain operational even after a full power load rejection, the system could easily withstand tripping the turbine at higher load Suggested Resolution: The answer key for the second part of this answer should be reworded to state: "the requirement to trip below 50 MWe prevents a premature trip by the operator which could result in unnecessary pressure transients".

Reference: ONP Text, Pages 17 and 18 NRC RESOLUTION: Comment accepte This increases the rigor of the answe .02 FACILITY COMMENT: There are some additional reasons for not having the recirculation loop cross-tie valves ope Suggested Resolution: If the valves were open and did not close on an LPCI signal, a major part of the LPCI flow during an accident could go out the break. This scenario would defeat the purpose of the LPCI loop selection featur . Having these valves open during operation could erroneously effect the LPCI loop selection instrumentation such that the wrong loop could be selected under accident condition Reference: Recirculation System Text, Page 47 LPCI Text, Page 24 NRC RESOLUTION: Item 1 in plant's resolution was added to the key. Item 2 replaced Item in the key. This was discussed during the review, and the plant stated that the candidate should know the specific aspect of the Tech Specs that should be complied with, even though the discussion is only found in the Bases. Two of the three items in the revised key were required for full credi F 10 Attachment 3, R0 Examination 4.04 a. FACILITY COMMENT: The question does not specifically ask for the station administrative quarterly exposure limit as is referenced in the answer key. SHP 4902 lists both the federal and administrative limits in this section of the procedur Suggested Resolution: The Federal Limits of 3000 mr/ quarter not to exceed 5)N-18) lifetime should also be acceptabl Reference: SHP 4902 Section 8.1. NRC RESOLUTION:

This was not accepted. The question clearly states according to proce-dure. The procedure is more restrictive than the CFR- . 4.07 FACILITY COMMENT: The answer key states that the RWM must be operable below 20% power. This is true. However, a further amplification is found in Tech Specs that says that no control rod movement will be made unless the RWM is operabl Suggested Resolution: An answer that mentions the RWM must be operable prior to any control rod movement and up to 20% power should be considered correc Reference: Tech Specs Section 3.3. The answer key states that the primary containment must be purged when primary containment integrity is require Suggested Resolution: The primary containment is required to be purged 24 hours after going into the RUN mode. It is not necessary to have the containment inerted whenever Primary Containment integrity is require Y 11 Attachment 3, R0 Examination Reference: OP202 Section 3.7.A. NRC RESOLUTION: The comment for Part a is wrong according to the copy of the Tech Specs supplied for exam prep. The 12 rod rule wasn't required for the answer in the original key, so tae comment is immateria The comment for Part d was accepte .11 FACILITY COMMENT: Parts (2) and (3) of this question seem to be asking for the same response. The reason to be cautious while draining from the RWCU System and the purpose of the valve isolation on the drain line is to prevent draining high points of the system. The fact that the valve closure is unannunciated is informational in nature only. Our objective for this question does not mention the fact that the isolation is unannounce Suggested Resolution: The answer for both parts (2) and (3) should be "to prevent draining the system high points". The fact that this is unannunciated should not be required for credi Reference: RWCU Text Page 59 OP303 NRC RESOLUTION: The question directed the candidates to the procedure. The lack of an annunciation is part of cautionary notes, and is therefore required. }}

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