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Exam Rept 50-245/87-23OL on 870921-25.Exam Results:Five Senior Reactor Operators Passed Exams.Two Senior Reactor Operators Failed.Related Info Encl
	ML20236W136
Person / Time
Site:	Millstone
Issue date:	11/19/1987
From:	Lange D, Lumb T; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
Shared Package
ML20236W127	List: IR 05000245/1987023; ML20236W125;
References
	50-245-87-23OL, NUDOCS 8712070297
	Download: ML20236W136 (135)
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{{#Wiki_filter:_ _ _ _ - _ _ - _ U.S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT N (OL) 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 EXAMINATION DATES: September 21 - 25, 1987

CHIEF EXAMINER: /z p/j2[yy T( Lu. ,O ratio s Engineer Date

APPROVED BY: . 2[[ f) Dasid J. Lange, Chief, BWR Section _]/l a [ Operations Branch, Division of Reactor Safety '

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SUMMARY: W.itten examinations and operating tests were administered to j seven (7) senior reactor operator (SRO) candidates. Five (5) - SR0s passed the examinations. All others failed the examination I

l l l l l 8712070297 871203 PDR ADOCK 05000245 V PDR I l _ _ _ _ _ _ _ _ _ l

_-_ DETAILS TYPE OF EXAMINATIONS: Replacement EXAMINATION RESULTS: l R0 l SR0 l l Pass / Fail l Pass / Fail l 1 I I I I I I l Written l N/A l 6 / 1 l 1 l l l 1 I I I I Operating l N/A l 6 / 1 l l l l l l 1 I i l Overall l N/A I 5 / 2 l

l- I i CHIEF EXAMINER AT SITE: T. Lumb, Operations Engineer OTHER EXAMINERS: A. Howe, Senior Operations Engineer B. Hajek, NRC Consulatant (Examiner) D. Florek, Senior Operations Engineer The following is a summary of generic strengths and deficiencies noted on the operating tests administered September 22 - 24, 198 This information is being provided to aid the licensee in upgrading license and requalification training programs. No licensee response is require STRENGTHS Use of the Emergency Operating Procedures, Off Normal Procedures and Alarm Response Procedures Interpretation and use of Technical Specifications Control Board Knowledge DEFICIENCIES Lack of familiarity with the new Process Computer Crew communications in the simulator portion of the examination (wher, an instructor was playing the role of an operator) _ _ _ _ _ _ _

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     -3- The following is a summary of generic strengths and deficiencies noted from the grading of the written examinations. This information is being provided to aid the licensee in upgrading license and requalification training programs. No licensee response is require STRENGTHS Understanding of plant parameter response on a loss of EPR -

Question 5.08 Knowledge of APR System logic - Question 6.07 Understanding of RWCU System valve control - Question 7.01 Knowledge of indications of a Stuck Open Relief Valve - Question 7.10 Ability to use and interpret Technical Specifications - Questions 8.03 & 8.07 Knowledge of RPV cooldown limits and their bases - Question 8.09 DEFICIENCIES Understanding of the effect of increasing core age on the void coefficient of reactivity - Question 5.03 Understanding of the effects of failures in the Recirc Flow Control System - Question 6.09 Understanding of the effects of Main Steam Line radiation mon 1 tor failures - Question 6.11 Understanding of the purposes of steps in the Emergency Operating Procedures - Questions 7.03 & 7.06 Knowledge of concerns when implementing a work order - Question 8.02 Personnel Present at Exit Interview, September 25, 1987: NRC Personnel T. Lumb, Operations Engineer W. Raymond, Senior Resident Inspector

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J QM-4-Facility Personnel R. Palmieri, Operations Supervisor B. Ruth, Manager, Operator Training R. Lueneberg, Supervisor, Operator Training M. Jensen, Assistant Supervisor, Optrator Training G. Giles, Assistant Supervisor, Operator-Training Summary of NRC Comments Made at Exit Interview: Security and facilities for the written examination were adequat The review of the written examination lasted three hours and was i very thorough. Facility personnel were reminded of the requirements for submitting written comment The training material submitted for exam preparation was well organized and easy to use. The NRC appreciated the material on the recent modifications. A few lesson plans were missing and no material on radwaste was included. The reactor theory material had limited learning objective The generic strengths and weaknesses noted on the operating tests (see section 3 of this report) were presented and the performance of the simulator (see Attachnent 4) was discusse During the simulator portion of the operating examinations, the simulator instructors had many questions concerning what the examiners were expecting from the candidates and comments on the prepared scenarios. The guidance for scenario preparation and operating examination administration is contained in NUREG 1021, Operator Licensing Examiner Standard The results of the examinations would not be discussed at the exit meeting but would be contained in the Examination Report. Every effort would be made to send the candidate's their results in approxi-mately 30 working day ' Attachments: Written Examination and Answer Key Facility Comments on Written Examination after Facility Review NRC Response to Facility Comments Simulation Facility Fidelity Report

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2 " NUCLEAR REGULA10RY COMMISSION i SENIOR REAC1OR OPERA 1OR L1 CENSE EXAMINAT1DN i ' FACIL11Y: _M _ .I L_ L _ T U N E_._ - _ _ __1 _ . _ _ . . _ _ _ . t ' REAC1OR 1YPE: ,,9 W R - G E 3 _ _ _ _ _ _ _ __ _ _._. _ _ _ _ _ DAIE ADMINISTERED: _B7/09/_21 _ _ _ , _ _ _ . _ _ _ _ _ _ EXAMINER: _LUMh_T.___________,_____ CANDIDAIE: ___ ._h_h_3 b ._ ________ INSlguCllONS,_lg__CONDlDgl@1 Use separate paper for the answer Write answers on one side onl Staple question sheet on top of the answer sheet Points for each question are indicated in parentheses af t.er the question, lhe passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be pickt?d up six (6) hours after the examination start % OF CATEGORY % OF CANDIDATE'S CAIEGORY

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zs. zs _Ist.99 ._ SNr.99 __. .______ _ . _ . _ . _ _ HEORY OF NUCLEAR POWER PLAN 1 OPERATION, FLUIDS, AND 1 THERMODYNAMICS , 2.4.co 24.25 PLANT SYSTEMS DESIGN, CONTROL, M ._ _$ 1. '.Y . _. .. . _ . ___ . AND INSTRUMENIAIIUN 25.7 7 _ . . _ _ . _ _ /. PROCEDURES - NORMAL, ABNCRMAL, _T5 99__ _Dr.SM _ . _ . . . _ . _ . . . . EMERGENCY AND RADIOLOGICAL CON 1RO . *LE

    - _ - _ . . _ - _ . ADMINISTRATIVE PROCEDURES,
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_ - . _ ._- -___ g , M. co Totalu ME_ _ . . . _ _ . _ _ _ _ . . . _ __ _._ _ % Final Grade All work done on this examination is my ow I have neither given nor received ai __ .___ __________ __ ___.______ ______ Candidat e's Signature d g % dels. hen ob jM ^* & _

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - _ ._. . _ _ _ _ _ _ - _ _ _ - 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 application and could result in more severe penalties. Restroom trips are to be limited and only one candidate at a time may Ieav You must avoid alI contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating. Une black ink or dark penci1 on!y to facilitate 1egible reproductions. Print your name in the blank provided on the cover sheet of the q l examinatio l Fill in the date on the cover sheet of the examination (if nec ese.ar y ) . l Use onty the paper provaded for answer : Print your name in the upper right-hand corner of the ftrst page of each settion of the answer shee i ' Consecutively number each answer sheet, write "End of Category __"*,a_d as e appr opri ate, start each category on a new page, write only on ty n e of the paper, and write "Last Page" on the last answer shee .4, 6.3. Number each answer as to category anti number, ior example, 10. Skip at 1 east t_hr_tg 1 i nes between each answe . Separate answer sheets from pad and place finished answer sheets face down on your desk or tabl . Use abbr evi at s ons onl y if they are commonly used in facility 1,tteratur . The point val uta for each question is indicated in parentheses after the question and can be useo as a guide f or the depth ci answer require . Show alI calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the ques t i on or no < 15. Partaal credit may be give Therefore, ANSWER ALL PARfS OF THE (2UEG1 ION AND DO NOI LEAVE ANY ANSWER BLAN . If parts of the examination are not clear as to intent, ask questions of the ex ami_ner; an] . 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 examinatio This must be done after the examination has l been complete . i _-- -

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10. When you complete your examination, you shalI: Assemble your examination as f ollows:

   (1) Exam questians on to (2) Exam aids - fxqures, tables, etc (3) Answer pages including figures which are part of the answer, Turn in your copy of the examination and all pages used to answer the examination questian lurn in all scrap paper and the balance of the paper that you did not use for answering the questions, Leave the examination area, as defined by the examine If after leaving, you are iound in this area while the examination is stil1 in progr ess, ynur license may be denied or revoked.

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! l } GUESTION 5.01 (2.SO) l DEFINE "the point of adding heat " (PAH), (0.b) o

 ' EXPLAIN WHY the PAH is expected to occur in the Intermediate Range of the Neutron Monitoring Syste (1.0)

!0 n EXPLAIN WHY the PAH does not occur at the same power level

0 on each startu ( 1. 0)
l UUESTIUN S.02 (3.00)

] STATE whether each of the fal1owinq changes, would INCREASE, DECREASE or HAVE NO EFFECI on the heat transfer rate in the RUCCW heat. exchangers? Assume all other parameters are held constant and )

l JUS 11FY your answer.

) state any additinnal assumption (!.0) p lube Fanlure L Lph ry, ) An i ncr ease in Service Water system flow (1.0) A decrease in Service Water temperature (1.0) 0 QUESIION S.03 (2.50)

Does the void coef4icient of reactivaty become MORE or LESS NEGAlIVE

with INCREASING core age? EXPLAI He sure to include each factor which affects the coefficient and identiiy the dominant IatLo (2.5) b o ,

QUESTION S . O tl (2.00)

e _ Concerning RAPID RPV DEPHESSURIZATION: WHAT happens (INCREASE, DECREASE or REMAINS THE SAME) to ACTUAL RPV water 1evel during a rapid RPV depressurization with no (1.0) i n.i ec t i on flow to the vessel? EXPLAIN WH WHY must. an adequate shutdown margin be established prior to rapi d RPV deprensuri z ati on*? (!.O)

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l QUESiluN U.05 (3.00) I i Use the attached BWR Power to Flow Map (F i gure 1) to armwer the J ialIownn l IDENTIFY Line A and EXPLAIN why core i1ow ini ti al 1 y increases then decreases as control rods e.re pulle ( 1. 0.)

( 1.. O ) IDENTIFY Line D and EXPLAIN why the line is slightly concav F and G and EXPLAIN what they signif (1.0) IDENTIFY Lines E, QUESIION 5.06 (2.00) . For each of the pairs of conditians 1isted below, st ate WHICH condi ti on would have the GRrATER differential rod worth and briefly, EXPLAIN WH Reactor moderator temperature of ISO ^F or 500^ (1.0) For a rod at posttion 10 or position 40 of a core operating (1.0) at 1007. powe i GUESTION S.07 (2.SO) l The Reattor Engineer iniorms you that a "coastdown" is planned to extend l i the plant's operattnq cycle due to system grid requirement WHAl is a "coastdown" and how does it al1ow plant operation ta (1.O) continue? I DESCRIBE two (2) other methods that can be used to lengthen the operating cycle and EXPLAIN why they are not the preferred ( 1. 5 ) method (or cyc1e extension?

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_ .. . _ l QUESTION 5.08 (3.00) l o r,t The reactor is operatsaq at 90% power when EPR system power 1 <3 HOW would the following parameters INITIALLY change and WHY? l (1.0) Reactor pressure

        { t. 0) Reactor power (1.0)

l Core flow ! l GUESTION 5.09 (3.00) . For each of the thermal limits listed below STALE:

 - the Ismiting parameter it is associated with
 - the cause of cladding ianlure as<;ociated with it
 - the Iimsting condition (1.0) CMFLPD (1.O) M"FHAI (1.0) MFLCPR l

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1_H_E_.RM_O_ D_.YN A._M_ _I CS QUES 110N b.10 (1.50) A reactor heat bal ance was per f ormed (by hand) during the midnight to 0000 shift due to the Process Computer being out of commission. The gain adjustment iactors were computed, t>u t the APRM gain adjustments have not been mad TRUE or FALSE? If the feedwater flow rate used in the heat balance calculation was LOWER than the actual feedwater flow rate, then '..he ac tual power is HIGHER than the currentiy (0.S) calculated powe b, TRUE or FALSE? 14 the reactor recirculation pump heat input used in the heat balance calculation was OMITI'ED, then the (0.5) actual power is HIGHER than the currently calculated powe IRUE or FALSE'? If the RWCU return temperature used in the heat balance calculation was HIGHER than the actual RWCU return temperature, then the actual power is LOWER than (0,5) the currently calculated power.

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l QUES 11DN 6.01 (1.SO) 1 For each of the fol1owing situatians, state whether the INDICAlED vessel 1 1evel wi11 INCREASE, DECREASE, or REMAIN THE SAME for the specified 1evel {

instrumen ACTUAL veur.el level REMAINS THE HAM { i How wil1 / The DrywelI temperature increases from 80^F to 230^ (O. S) the NARROW RANGE GEMAC level instrumentation respond? A reactor startup is in progres The head vent has been close Vessel temperature and pressure are increased from atmospheric and 190^F to 800 psig and Sla^ How will the ROSEMOUNT level (0.5) instrumentation respond? The reactor is shutdown and a cooldown is in progres Shutdown Cooling is initiated. How will the WIDE RANGE YARWAY level (O.S) instrumentation respond? QUESTION 6.02 (3.00) Nillstone Unit 1 is operating at 100% power &dien the vital AC bus tr ansf ers from its normal to alternate power sourc DESCRIDE how each of the a f f ect ed by the transfer, folinwing components and systems are (0.5) Reactor Pressure Controller (O.S) Reactor Huiidiog Ven t.1 1 at i on l

      (0.9) Feedwater System i
      (0.5) l Recirculation System (0.S) l Reactor Pr ot ecti on System (O. S) Rud il1ock Monitor
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OUESTION 6.03 (2.00) For each condition given bestow STA1E whether t. h e affected APRM channel channel will have an INOP trip? (YES or NO?) l APRM channel 5 has 4 LPRM toputs in level "A" level; 3 LPRM inputs and i LPRM input in in "U" level;-3 LPRM inputs in "C" (0.5)

 "D" level, channel not bypassed APRt1 channel 3 (with function switch in COUNT) meter    (0.5)-

indicates 50%, channel not bypassed APRM channel 4 APRM mode swatch in POWER, channel not (0.5) bypanned (0.5) APRM channel 1 APRM mode switch in ZERO, channel bypassed GUES110N 6.04 (2.00) Which of the f ol l owi ng is TRUE concerning the Standby Liquid (1.0) Control System:: (CHOOSE ONE) 1) If injected, the SLC system will provide at least a 2.6% ; delta k shutdown margin and 660 ppm baron concentration in the reactor vesse ) Ni t rogen--charged accumul ators assure adequate suction pressure f or the pump ) In the event a remote (outside cant.rol room) reactor shuttlawn is required, SLC injection can be ar_ tuated by the local pump START switc if necessary to - 4) The pumps may be operated simultaneously shutdown the reactor in an ATW Mi1istone Unit 1 Technical Specifications contatn requirements on neutron absorber volume and SLC pump capacity which result in a maximum injection time for the Standby Liquid Control System, (1.0) WHAl is the maximum injection time and WHAT is the bases ior it? l

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_, _ ___ _ PAGE O b _ b10d1_E151EdE.iEE31Ed1_E99109ht_QUp_lyglguggdlgllgN 1 l i - DUES 11ON 6.OS (2.00) WHA 1 AU10 mal lC ACllONS occur within the FWCI system as a result of :

a FWCl im tl ati on signal to ' limit inventory loss and ensure that (1.5) suff-icient ilow is provioed to the RFPs? IRUE or FALSE? The FRVs will not return to automatic level cont rol . unt il the initiation signal has cleared and flow control (0,5) is j rese s& UUESTION 6.06 (1.00)

Uy procedure, pressure across the MSIVs must be equalized prior to j opening when the vessel i s pr essur i z r WHAT are two ( '2 ) potential hazards ai opening the MSIVs wzth an (1.0) ex cessi ve differential pressure across them? DUESTION 6.07 (2.50) l For the f ollowing si t uations, state whether the Automatic Pressure Relief System (APR) relief valves will OPEN, ClJISE or REMAIN AS 1 Consider each set of condttions separatel l I

     . . . . reactor APR anittating signal sealed an, APR valves open    (O.S)

water level then rises to -30 inche f.PR APR i ni t i at i rsg signal sealed in, APR val ves open . . . .

       (0.5)

timer reset button is t hcfn depresse {

APR initiating signal sealed in, APR valves open . . . . then a - DC power iaiiure occurs that affects alI busses supplytog APR valve (O.5) )

a loss of the Drywell Atmo- i d. APR initiating parameters present, l spheric Compressor System has occurred, 120 second t i mer f.iming (0.5) aut . . . . then the 120 second timer times ou APR i ni ti ati ng parameter s present; all LPCI and Core Spray pumps are secured except for CS pump D whit.h i s running with a discharge then pressure of 195 psig, 120 second timer timing out . . . .

       (O. 5) l the 120 second timer times ou l
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QUESilON 6.00 (3.00) \ , ] \

9S~/. power when a Loss of Normal Power I 1he reacter i s oper ating at (LNP) signal is recesve ( l . O)' l WHAT COND11 IONS wi 11 generate a LNP signal? liOW does the 4160V Electrical System realign to supp.1y power to (2 O) essential Icads when a LNP occurs? ! UUESIION 6.09 (3.00) ihe reactor is operating at 100~/. power with recirculation flow control in master manual. Explain HOW and WtiY the recirculation pumps r espond ta the f011owing conditians. Where applicable, provide spect f ic value (1.O) Master Contral1er output iaiIs LOW Full open indication on recirculation pump A discharge valve ( 1. O ) is Iost at the va1v U , , iu ,- b ypuw.> voivua ;: 0;mn 4 (1.0) U Recirc t'.G Set tachometer output iails t ,zer QUESTION 6.10 (3.00) LPCI loop select. logic has been initiated and has determined that only one Reci rcul at i on pump is rutenin WHAT ACTIONS must occur or WilA T CONDITIONS must be met prior (1.0) to loop selection? WiiY must toese candillors be established? F Briefly DESCRIDE HOW the LPCI loop select logic sel ects the unbroken loop and lines up the RecirculationIeclude Systemsetpoints after at has determined how many pumps are runnin and/or time delays and all Recirculation system components (2.0) ; that are affecte i

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_ _ _ _ - - . - . PAGE to 6_ _.P_L_ A_N._ T_ S_Y_S..T_E_M_S_ _D_E_S__I.G_N_g_C_O_N_ T R._O_L_g_AN_ D __I N_GT.R_.U_M_E_N_T_A_T _I O _ _ _ . ! OUESTIUN 6.I1 (2.00)

I Main Steam Line 'A' radialion monitor iails downscal WHAT ( 1. O) AUTOMATIC ACTIONS and ALARMS, 1i any, wi11 occur? While attempting to repair the downstale instrument, the 1 &C technician inadvertent 1y pulIs the power plug ior NSL 'C'

r arti a t i on monito WHA 1 AUTOMAI1C ACTIDNS and ALARMS, i4 any, ,

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66 i 1 1 occury i i i

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'OUES11ON 7.O! (1.00)

l OP-303, REAClOR CLEANUP SYSTEM, states that the system is operated with the Pressure Control WHY Valve (PCV) in AUID and the Flew Cont.rcJI are they operated in this arrangement? Valve (FCV) in Manua ( 1. 0) UULS1!ON 7.02 <3.00) i Concerning ONP-512, ItAPID AND TOTAL LOSS OF INSTRUMENT AIH: l What is the back-up air supply to the FHVs and how long can (1.0) they be operated with this supply? Would all of the MDIVs be expected to close on a complete (0.75) loss of instrument air? WHY or WHY NOT? WHY is the operator instructed to start the Standby Gas Treatment System on a loss of instrument air? (0.75) Wily is the h'gh level alarm for the RBCCW surge t ,to k expected to occur on a loss of instrument air? ( 0. 5 ) UUESfIUN 7.03 (2.SO) If RPV water l ev.?! cannot he determined, EDP-S78, EMiiRUI- NCY RPV FLOODING, _ directs the operator-to commence and raise injection ilow until:

 - At l eas t 4 SRVs are open, and
 - RPV pressure is not lowering, and
 - HPV pressure is at least 50 psig above suppr essi on   -

chamber pressur WHY must each of these condi tions he sat i sf i ed to ver i f y HPV FLOODI NG (2.5) without level indication? l l l l l l l (***** CAlEGORY 07 CONTINUED ON NEX1 PAGE *****)

PAGE 12 Zi_ C099999999_~_U90UOh2_O9d99UOb1_ggggggggy_gND 00919h99190b_99NIggy i QUEST 1DN 7. 04 ( l .. SO ) f OP-324C, Augmented Off--Gas System, caut1ons that "during unit startup, off gas flow should be through the delay line until the recombiner is up to operating temperature".

WHA 1 are two (2) adverse consequences of placing the recombiner in service when at less than operating temperature? EXPLAIN WHY they (l.S) occu .

QUES 1IUN 7.OS (2.00) ONP-SO2, EMERGENCY PLnNT SHUTDUWhl, has the operator transfer t.he modo suitch to SHuiDOWN after determining that the pressure regulating system is operating properly, WHAT are three purposes of placing the mode switch to SflulDOWN? (1.5) WHY is it important that the pressure regulating system be ( 0. 5 ) operati ng properl y bef are pl acing the mode switch to SHUTDOWN': F QUESilON 7.06 (3.00)

EOP-574, RPV STCAM COOLING, has been en t.er ed and the isolation Condenser is unavailable, therefore, steam cooling must be performed by opening one safety / relief valv The procedure directs the operat or ta wai unti1 RPV water ievel reaches -220" to open the SR WHAT is the purpose of waiting untii this ievel ta commence ( l . 5) j

steam cooling with an SRV? If during steam cooling with one SRV, RPV pressure drops below 700 puto, the procedure directs the operator to emergency depressurize. WHAT is significant about 700 psig and WHV is (l.5) RPV depressuri zati on required below 700 psio?

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, QUEST 1UN 7.07 (l.SO) You are supervising fuel movements i n accordance wi th OP-320U, FUEL LOADING / UNLOADING / SHUFFLING., A bundle is be2ng moved from the core to the spent fuel pool and is currentl y locatco in the transfer chute between the reactor cavity and the spent fuel poo You are noti 4ied that the drywel1 seal is leaking and the refuel iloor must be evacuate WHAT ACTIONS, ii any, do you take prior to evacuation? EXPLAI ( 1. 5) GUESTION 7.00 (2.50) The plant is operating at 100 7. power when you experience a loss of 345 KV t ransmi *.,si on capabi 1 i ty (1 cad reject). The automatic actions that wiil occur are:

   - Select rod insert
   - APRM high flux setdown
   - Bypass valves wil1 open
   - The turbine control and intercept valves will throttle closed Uriefly EXPLAIN the reason fop each of the above action (2.5)

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QUEST 10N 7.09 (3.00) ! OP-206, PLANT COOLDOWN TO COL.) SHUTDOWN, cautions that reactor vessel , l level must be maintained above 450" to prevent thermal stratificatio WHAT is significant about a reactor vessel water level of 4bO" (0. 5 ) with respect to prevention of thermal stratificatio EXPLAIN HOW thermal stratification could occur in each of the

     +50".

ial1owinq situatians if vessel 1evel is below 1) Hoth Reactor Recirculation pumpt, are secure No Shutdown (0.5) Cooling fIow establishe 'H' Recirc pump running, *A* Recirc pump secure Shutdown 2)

      (O. 5)

Cooliog i1aw establishe WilAl are the adverse consequences of thermal stratification if it occurs when:

      ( 0. 5 )

1) Primary and Gecondary Containment are not established?

      (0. 5)

2) the reactor vessel is ont vented?

      (0.5)

3) the reactor vessel is open for maintenance? QUES 1 ION /.1O (2.00) lhe reactor is at 100% power and the following alarms are received:

 - SAFEIY AND BLOWDOWN VALVE LEAKAGE
 - SAFETY / RELIEF VALVE OPEN De spectii WHAI condition (s) have caused these alarms?   (1.0)

loclude setpoin WHAT are four (4) control room indications that the operator (1.0) can use to veri f y that a SRV tu stuck ope *****)

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

- _ _ - _ - - _ - _

_ _ _ _ PAGE 15

?.__C09EEE90EE._~ d90UOh2. OOdSOUOh1_ggggggggy_gND RADIOLO.G.ICAL
 ---- - - CONTROL
  ---------

DUESTION 7.11 (3.00) For each of the following situations SlAIE which, if any, Of f If -Normal none, Procedures and/or Emergency Procedure (s) you would ente state NON A procedure index is provided for your referenc (O.S) Scram Pilot Air Header Pressure - 50 psig NRHX High temperature Alarm, Cleanup Pump A f r i pped , High Area Radiation Alarm (RWCU) and High Area Temperature Alarm (O.S)

 (RWCU)
      (0.5) Heactor Pressure - 1O94 psig Path itecirculation Pumps tripped, power st.eady at 45% one (1) (0.5)

minute later 13.6 feet (0.9) Suppre% ion Pool Level - Condensate Pump Dascharge Conductivity High High Alarm ( 0. 5) f.

,

i

1

k i

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

_-_ PAGE 16 W _0901N19100II R PU9CEgggEQ _ggggillgNQ,,9Ng_Ligilg}lggg j j l i UUESI1UN U.O1 (2.SO) l l j What is the federal quarter 1y exposu.e 1amit 1or occupational radiat1on exposure ta the whole body Ior a 24 year ofd, male (1.05 < employee with an up-to-date NRC Form-4? l What would the same emptayee's administrative quarter 1y exposure if his current quarterly radiation limit he at Millstone Unit 1, (0.5) expouure was unknown? What is the initial exposure contral Iimit ior a calender quarter (0.S) for personnel at Millst.one Unit I with completed NRC-Form 4s?

      (0,5) What is required to extend the initial exposure control limit?

GUESTION H.02 f1.00) What are ttree major conter on (i dent i f i ed in ACP-RA-2.02C, WORK URDERS) of t b r' Sh11t Supervisor / Supervisory Contral Oper a t or when impiement.ing ().b) a war k order that wi1I place a Category I system out. of service? UUESTIUN U.03 (3.00) The instrument Matotainence Supervisor informs you that while checking the Loss of Nnrmal Power (LNP) relay Icgic, he found one of the under-voltage relays inoperative, the exact cause is unknown. The loss of this under val tage rel ay will disable the LNP relay logi The reactor is operating at 957. powe Can operation continue to accordance with the Technical Specifications? If sn, under what . conditions? If not, why not?

      (3.0) l l
    ******************************** j
********************************************ECHNICAL
* NOTE: USE THE ATTACHED SEC110f4S OF lilE T SPECIFICATIONS
      *
*  TO ANSWER THIS QUESTIO FULLY REFERENCE ALL APPLICABLE
      *
*  SEC1 IONS OF THE TECHNICAL SPECIFICATION r:s*************************************************************************

i I i (***** CAlEGORY 00 CONTINUED UN NEX1 PAGE *****) l _ _ - _ _ . _

- - _ _ _ _ - _ . PAGE 17 9N._._OEUlylglGGllW_PGQCEpyGEg1_C969]Ilgypt_9Bp_llD1I9IJgyp UUES110N U.04 (2.50) In accordance with ACP-6.01, Control Room Procedure: WHA 1 rule must the operators follow regarding belief of instrument readings? WHAl is the reason for this rule? (1.25) b, in WHAT two (2) situations can remote operations be performed without first requesting permission from the (O.5) Contral Room? HOW OFTEN does the Shift. Supervi sor/ Supervisory Cont rol Operator rewiew operator round sheets'? WHAT are two (2)

      (0.75)

purposes of this review? DUEU1 ION U.OS (2.SC) During what operating condi ti ons i s Primary Containment ( 1. O ) Intearity required in accordance wtth rechnical Spect( cat 2ons? WHAT five (b) condi tions must be met to establish Primary (1.3) Cont. Inment integrity? GUESTIDN G.06 (3.00) State whether each oi the loilowing changes would he a NON INTENT change or an INIENT chang . A change to the acceptance criteria of the Core Spray (O.5) System Operabi1ity Tes . A change to the Fuel Handitog Procedure which adds (0.5)

steps to clarify the data taking requirement . A change ta the Condensate Demineralized Hackwash and (0,5) Precoat Procedure to use a special precoat proces What i s required for FINAL approval and implementation of: A NON INTENT change? (0.75) An INTENT change? (O.75) (***** CATEGORY 00 CONIINUED ON NEX1 PAGE *****)

.
 -_ -  _

__ _ _ _ ___ _

i PAGE 10 L O . _ _ g e g i n i S i g g y l y E e R O C e o u R E S ,,_ C O N g i l l g N g g,, j N g J : l d l l O I I O N S ..

-DUEST1DN .D.07 (3.00)
         ;

l When you t ur'nover at the' start of the mis.inight to eight AM shiit the i plant i s a t' 100% power and all conditions are'nermal with the (allowing- j

 - except i ons:
 '
 -- APRM channel .1 is bypassed for maintenanc APRM channel 5 is failed high and bypasse Two hours into the shift APRM channel 2 f ails downscale, in accordance with Technical Specifications, WHAT ACTIONS, if any,    (3.0)

are required in this.. situation?

 **************************************4************************************
 *. NOTE: USE 1HE ATT ACHED SECTIONS (3F THE TECHNICAL SPECIFICAT IONS    *
 * TO ANSWER THIS QUESTION. FULLY REFERENCE ALL APPLICAULE    *
 * SECTIONS OF THE lECHNICAL SPECIFICATIONS.

!

 '**************************************11***********************************

l QUESTJON U.OH (2.00)

        ~
 - Concerning surveill ance test f requenci es and time intervals in accordance with ACP-DA-9.02, STAllON SURVEILLANCE PROGRAM: WHA 1 are t he requi r ements. f.or the time interval during wftirh a   (1.0)

scheduled monthly survei l l ance rnuht be performed? WHAT iG the maximum allowable extension that can be granted ior 10.5) perf ormance of a scheduled survei'llance? ~WHAT is the maximum comhined inte val iime al1 owed for tbree 40.S) consecutive periarmances of any survei 11 ance test'? l-DUEST1DN O.09 (2.00) WHAT is the Tech Spec thermal limit for RPV couldown rate? WHAT (l.0) i s the bases for this Iimit? WHAT is the MilIstone Unit i adn inistrati ve limit for RPV (2.O) cooldown rate? WHAT is the bases ior this 1imit? i

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

l' _ _ _ _ _ _ _ _ .-. 1

- _ - - _ _ _ - _      -  -  _

PAGE 19 9 __0901619100IlyE_PggCEgggEg1_ggNgiligNgt_gNp_LigilgligNg l l

1 0

UUEG11UN U.10 (3.00) Classi f y the f ol l owi ng evente. in accordance with EPIP 4701 (attached).

In your classifstation statement also include the Key Conditions and the Emer gency Action Level . Chemistry reports that two stack gas camples, drawn 1/2 hour (0.7S) apart, exceed the Tech Spec instantaneous limits by IS time A break has occurred in the Recirculation System Sample Line an the react or buiIdin Group 1 valves have failed to (O.75) asolat Ihe Reser ve Station Serva ce Transformer is tagged out Ior (0.7S) ' maintenance and 4KV bus 14D trip Galety Reisef Valve 11G-3B anadvertantly opens and cannot (O.7S) be recIose .

  (***** END OF Call ~ GORY 00 *****)
 (**wt********* END OF EXAt11NAT ION
    ***************)

_ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ - - _ _ _ _ _ -

.
 . . _ . _ _ _ _

7- . PAGE 20 9 ._1dE995_.9E_d99hE90 E9dE0_EbOU1_9EEE011991_Ebulppt,,,gND IUEBD99YN9919@

  - MILISTONE 1  -07/09/21-LUMB, ANSWERS i

l l ANSWER S.01 (2.50) PAH is the power level at which heat from fission is sufficient to cause observable reactor coolant temperature increase. (O.S) Or' had &bbb cn.>trc.omas & io%es At 1ower power s heat 1osses are 1arger than the heat input i r om i i st.i on (0,5) so no temperature change is observable even though the i a ssi on process is adding heat to the coolant. (0,5) or had .s p %% % e vceek es. 3- l an u + % s eq , The specifsc power level at which the PAH is expected to occur varies with initial coolant temperature and drywell temperature (0.S) since (O. S) both t temperatures determine the rate of energy loss to ambien REFERENCE MP1 Operator Training - Procedures, TX-1200 - General Operating Procedurer., pg 11, ubiectives IS & 16 ' as q - 74 0 K/A 292000 Kl.1S (3.7/3.7) 2Y2OOOK115 ...(KA"S) ANSWER S.02 (3.00) Decrease (O,25). Iligh prensure fIuid wi!I mix with Iow pressur e f1uid, lowering the temperature diIference between the cooling medium and the cooled medium (0.7S cr* red d n w 44,.3 c4 HP (L :.h Increase (0.2S). An increase in velocity will increase the mass f l ow rate of the cooling iluid, nncreasing the heat transfer rate (0.75). Increase (0.2S). A decrease in the temperature of t he cooling iluid causes g dou rr - in delta T which duct. ..- t.he heat transier - t h SC- merc4ses rate (0.75) f(CFERENCE 9- Heat Mi l l st.one Uni t 1 Operator Training - thermodynamics, Chapt er Transfer, pg 9-40, Ub.i ec t i ve 16 K/A 291006 K1.04 (2.0/2.0), K/A 291006 Kl.OG (2.9/3.0), K/A 273007 K1.06 (2.7/2.0) ...(KA*S) 291006K104 291006K100 293OO7K106 _ _ _ _ - - _ _ _ _ _ _ - - _ _ - _ _ _ _ - _ -

- _ _ _ _ _ _ _ _ _ _ PAGE 21 S. _ _IUEggy_gF _NgCLEOB_EgWEO_E(gN1_gfgG911gN,,_F(ylggg_gNg IUtiBM09YNOM1CS

   - MILLSTONE 1   -87/09/21-LUMU, ANSWERS ANSWER   5.03  (2.50) % ,c. g , g ,4
     . __ . - - -

MLCi c d.L o b clo m . u fu t.,,r. (_ C. S') Be tumt u u.-m o uy u u v w. wi

 - Control rod density decreases so there is less loss of neutr ons to control rods which compensates for the offect of vonds making the void coefficient Iess negative (0.5).

- As gadolinium burns out the average neutron energy decreases which increases the number of thermal neutrons making the void coefficient more negative (0.5).

- Pu240 huiIdup over core li4e reduces the resonance escape probabi1ity, increasing the availability of thermal neutrons and making the void c oe f -f i c i en t more negative (0.5).

- U235 hornup over core life increases the thermal flux which makes the void coef f i ci ont more negati ve (0.5).

+-m .m - r e 1, .r , , ,, a g - -g to itj , -=ch y , ,, t g ...; c _t3 r ,g _e g ,, _ , _;a

     : - .1 1 :c, .t y gg . , i o cy, n n 7a ; m ,, e n 29.) ,

REFERENCE Reactor lheory, Chapter 4 - Coef f icients of Reactivity, pgs 4-22, 4-23 &

- 4-24 K/A 292004 K1.13 (2.1/2.2), K/A 292005 K1.09 (2.5/2.6)

292OOSK109 292OO4K113 ...(KA'S) ANSWER 5.04 (2.00) DECREASES (0.5) reduces the vessel inventory due to steam - Iass (O.25) and normal shriok on cooldown (O 25) e,c im 4% % saca - om>e A AP hb o o4% %, wAA chab*h postLive reactivity inserted during cooldown (0.5) could result in {

criticality if inadequate SDM (0.5) l i j REFERENCE I MP1 Operator Training - Pr oc c.d ur es , TX-15008 - Emergency Operating Procedures, pg 35, Objectives 19 L 20 l l K/A 293003 K1.01 (2.3/2.4), K/A 292004 K1.01 (3.2/3.2) 293OO3K101 292OO4K101 ...(KA'S) l _ _ _ - _ _ - _ - _ _ _ __

_ _ . . ._____ _____ i PAGE -22 5:__1HE901_9E_@.lEbE90._E9dEE EhgNT__ OPERATION t lL_UIDSt_AND THERMODYNAMICS

-- -----------

t i ' ANSWERS - MILLSTONE 1 -87/09/21-LUMB, i i i I I ANSWER S.OS (3.00) Natural Circulation Line (0.5). The increase in core ilow at low power levels is due to heating of the moderator which promotes natural circulation (0.25). As power increases more voids are produced which increases two phase ilow resistance and thus decreases core fIow (O.25).

orAMA N 8takL*6 0.75-The % Flow Control or 100% Rod Pattern or 100% Load Line (Art 1 line is slightly wacave because core inlet subcocling decreases as ilow increases (pM . O. "

 -C .> U iu duL Pu..p Cm . LoL4un L-  ' " . 12 ) . L . U - % ' :- Li      'O.;%
  ' O . 2--) . -+_ i . L, Uiu ;lu- litcoluc' L:'

Rwap-CovaLalaun L 4 u_ 4 2 Operation above lines E & F is required to prevent cavitation of the jet pumps and the reci rc pumps -ec.pmt! 'l >/ (4hrff.# Cavitation will not occur at any core flow as long as feedwater flow is above 20%. 1his is the basis for the Flow Interlock Line to d .# ('the recirc pumps are interlocked so that they cannot be run above minimum flow when ieedwater i1aw i; below 20%.) RElERENCE Millstone Unit 1 Operator ' training - thermodynamics, Chapter 10 - HWR Thermal Hydraulics, pgs 10-62, 10-63, Objectives 12 & 21 MP1 Operator 1 raining - Procedures, TX-12OO - General Operating Procedures,

Ob.3cctive 37 K/A 293000 Kl.19 (2.6/2.0), K/A 293000 K1.37 (3.2/3.4), X/A 291004 K1.01 (3.2/3.2) ...(KA'S) 293OOSK137 293OOOK119 291004KlOl

_ _ . _ _ _ _ _ _ - _ _ - _ - ___ _ - . ___ . - - - _ _ _ - FlulDS d ND PAGE 23 E _ IDEOB LgF_NyC!. Egg _ POWER PLANT _gPERATION n IUfjRDODYNOUICQ ANSWERS -- MILLSTONE 1 -07/09/21-LUMB, ANSWER S.06 (2.00) At SOO^F (0.25) As moderator temperature increases, neutron leakage out of the f uel . bundl es is increased, thus the control rod is exposed to higher neutron flux and rod worth increases. (0.7S) L, 4 q p tg MSW ob ted '-y- At 10 (0.25) The void content of the upper portion af the core is high at operat.ing conditions, the offects of deep control rod withdrawal can be substantial radiall The negative power eflect of increased voids above the control rud is not seen because the channel length above the control rod i s. rel ati vel y short. (O./S) REFERENCE Reactor Theory, Chapter S - Control Rods, pqs 5-8, 5-9, 5-12, S-13, S-16, S-17 & S-lO,-Objectives 2.2 b K/A 292005 Kl.09 (2.5/2.6), K/A 292005 K1.10 (2.0/3.3), K/A 292005 Kl.12 (2.6/2.9) ...(KA'S) 292OOSK112 292OOSK110 292OOSK109 AMSWER S.07 (2.SO) A coastdown occurs at end of cycle when power is allowed to decrease naturally as U235 burns out. (0.5) The reactor continues to operate at a Keff = 1 because the negative power caeIficient contributes positive reactivity to compensate for the fuel 1 asses (primariiy decrease in voids and fuel temperature are the major contributors).

(0.b) g r. y ecg g ag,3 9 fr,p gg ogg,,3 ,,,,, y_ 4 g , L . Feedwater temperature reduction by isolation of FW heaters (O.S) - decreased plant efficiency (0.25) Excess core f low - operating above rated core flow (0.5) recirc system severe duty, vibration of pump internals, or high core differential pressure (0.25) Derat i ng - continued operat.i on at lower than rated power level (0.5) plant is operated at lower than rated power level (0.25) d*J'W hd -

 (any 2 of the above for full credit)  b b56 !<oe.\ h'ksf *s M k m u cu es Ww e> c a RLFERENCE     pt~-p ba4 A fv s om ssed. (e.S)

N < ac e Reactor Theory, Chapter 7 - Reactor Operational Physics, pg 7-20, Mel<. Objectave (0.LT) i

 ---        1

_ PAGE 24 91__ly[ggy_gg_Nyg(EAR PgWER_ PLANT _OPERATIgNt _ffLUIDS t_AND 1HERM0 DYNAMICS

--------------

ANSWERS -- MILLSTONE 1 -07/09/21-LUMB, K/A 292000 Kl.19 (3.1/3.2), K/A 292000 K1.20 (3.3/3.4) 292OOBK120 292OOOK119 ...(KA'S) At2SWER- S.00 (3.00) Increases (0.25) due to the control valves (nresponse to the MPR becomi ng thecontrol1angsignal)goingshut (O./h) Increases (0.25) due to the collapse of voids from the higher pressure which adds positive reactivity (0.75). Increases (0.2S) due to the reduction in the void content of the two phase mixture in the core (O.73).

REl:ERENCE Reactor lheory, Chapter 4 - Coefficients of React i vi t y, pg 4- 24 MP1 Operator Training - Systams Volume 3, TX-1314 - Turbine-Generator System, pgs 121 L 167, Objective ~~de P W~~

- cladding iailure due to lack of adequate cooling following a LOCA (0.33) - 22OO^F peak clad temperature (0.33) or Lo . MCPR (O.34) o - M k p>c u r n - cladding failure due to inadequate cooling caused by 1 ass of nucleate boiiing (O.33) - onset of transition boiling (0.33) oc REFERENCE Millstone Unit 1 Operator Training - Thermodynamics, Chapter 11 - Thermal Limits, pgs 11-3 thru 11-32, Objectives 3, 4, 7, 0, 10, 14 & 15 K/A 293009 K1.07-(2.8/3.6), K/A 293009 Kl.OB (3.0/3.4),
K/A 293009 K1.09 (3.1/3.7), K/A 293009 Kl.11 (2.8/3.6), K/A 293009 K1.12 (2.9/3.5), K/A 293009 K1.19 (2.8/3.6), K/A 293009 K1.20 (3.1/3.6), K/A 293009 K1.28 (3.0/3.5) 293OO9K128 293OO9K120 293OO9K119 293Oo9K112 293OO9K111 293OO9K109 293OO9K100 293OO9K107 ...(KA'S) _-
_ _ _ _ _ _ _ _ _ . _ . _ _ _ _ . _ _ _ _ _ _ . _ . _ _ _ _ _ _ 7. . PAGE 25-5_ _. _ __TH_E.O_R_Y .O_F _N_ U_C_L_E_A_R__P_
~~- _ _ _ _ OWE _R_P_L_AN.__T _O_P__E_R_ AT_I O_N_1._F_L_U__I DS_g_A_N_~~
__ _ 1UEggggyNQQlCS-07/09/21-LUMU, ANSWERS -- MILLSTONE 1 ANSWER S.10 (1.50) f^U "P S I' 0" fl (0. 5 ) fftfE-- PAu6 Lac.c.y\- TRu g i(- e p 6 , go,5) FALSE (0.5) FALSE REFERENCE Millstone Unit 1 Operator Training - T hermaclynami cs , Chapter 10 - HWR Thermal Hydraulics, pg 10-7, Objective 2 K/A 293007 K1.13 (2.3/2.9) 293OO7Kil3 ...(KA'S)
~~. - - - - - - - - _ _ _ _ _ _ _ _ _ _ _ _~~
PAGE 26 6 __Ph0NI_gySIEMg_pE@l@yt_CQNIBQLg_QNQ_lgglBUMENIgligg-87/09/21-LUMD, ANSWERS -- MILLSTONE 1 1.00 ANSWER 6.01 W Increase (0.S) Decrease (0.5) I .- m , uusu 'O.Si REFERENCE MP1 Operator 1 raining - Systems Vol ume 1, TX-13 OOH - Reactor Vessel Instrumentation System, pgu 7, 0, 14 - 17 K/A 291002 K1.OB (2.0/2.9), K/A 216000 KS.07 (3.6/3.8), K/A 216000 KS.10 (3.1/3.3), K/A 216000 KS.09 (2.9/2.9), ...(KA'S) 216000KSO7 216000KSO9 216000Kb10 291002K100 ANSWER 6.02 (3.00) EPR remains in control with only a minor perturbat i on is system pressure (0.S) solates (0.5) FRVs lockup (fail as is) (0.b) Scoop tubes lockup (0.5) HPS channel O trips (O.S) no affect (0.S) REFERENCE MP1 Operator 1 raining - Systems Volume 6, 1X-1343, Vital and I nst r umen t AC Systems, pgs 14, 15 & 16, Objective 22 K/A 262002 A3.01 (2.8/3 1), K/A 262002 A4.01 (2.8/J ), K/A 262002 (3.1/3.4)
~~...(Kn'U)~~
~~262002G001 262OO2A401 262OO2A301~~
,
ANSWER 6.03 (2.00) NO (0.S) YES (0.5) YES (0.5) NO (0.5) REFERENCE MP1 Operator l' raining - Systems Volume 6, 1X-1404A, Average Power Range Monitor System, pgs 9, 14 & 15, Objective 4a K/A 215005 K1.04 (3.6/3.6), K/A 215005 A2.03 (3.6/3.0) 21SOOSK104 21SOOSA203 ...(KA'G) _ _ _ _ _ _ _ _ _ _
! FEE 27 6_ _;_ _P_L_ A_N_ _T _S_Y_S_T_E_M_S _D_E_S__I G_N__,_ C O N T R_ O_ L_ E _ A_ N _D _ _I M S_ T R_ U_ M_ E_ _ .
~~- _ _ _~~
t-87/09/21-LUMB, ANSWERS -- MILLSTONE 1 ANSWER 6.04 (2.00) ) (1.0) or Q) Maximum time: 125 minutes (0.25) suf -f i c i ent to overcome the positive reactivity addition rate (0.2S) caused by the decay of peak xenon (0.25) and moderator cooldown (0.25)
~~' REFERENCE l MP1 Operator Training - Systems Volume 2, 1X-1304 - St andby Li qui d Control ,~~
6 + l+ pgs 4, 7, O& 19, Objectives 1 & 13 < 99 K/A 211000 K4.09 (2.5/2.5), K/A 211000 KS.03 (3.2/3.5), K/A 211000 A1.09 (4.0/4.1) ...(KA'S) 211000A109 211000K409 211000KSO3 1.50 ANSWEH 6.OS G' . 0 0 3 c. 3 Emergency Condensate Transfer Pump starts and disch valve opens ( e-ebt SJAE min fIow valve shuts (0..'bi . FRV and Blocker receive shut signal fC . ..; / o. 3 Main FRV Hlocker receives open signal ( h P4i) o. 3 Recombtner i sol a t i on valves shut 64 -26) O. J A4,, FMVs moi..to;. ': v ut iv o. iv.ZG)
~~(, t,o%cht. p H **A,tooo h y sbA-)~~
M 'iL U E -it7-179 HEFERENCE MP1 Operator 1 raining - Systems Volume 4, 1X-1334 - Feedwater Coolant Injection System, pqs 17 - 20, Objective 4 K/A 206000 A1.OO (4.1/4.0), K/A 206000 U.7 (4.1/4.2), K/A 259001 K1.00 (3.6/3.7) ...(KA'S) 2S9001K108 206000 GOO 7 206000A100 . ANSWER 6.06 (1.00)
~~- [f the DP is Iarge enough the valve may not actuate - May result in a large steam hammer causing hydraulic shock to cbwnstream components - Hay cause a pressure transient - (2 9 0.5 each)~~
~~REFERENCE OP-317, Main Steam System, pg 4 TX-1317 - Main Steam System, M/' t Operator Training - Systems Volume 3, pas O & 35, Objectives 13 L 33~~
~~.- -- _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _~~
- - _ -_ _- _ - - _ . . _ - _ .__ . _ _ . _ - - _ . - . _ - - - - _ - _ - - _ - _ _ _ _ _ - - _ _ _ _ PAGE 20 b __Eh0U1_E191EUS_95E1EUL_EEU199ht_QMg_lgglggggylOllg@ ANSWERS -- MILLSTONE 1 -07/09/21-LUMB, K/A 239001 G.10 (3.5/3.7), K/A 239001 KS.02 (2.9/3.1) 239001KSO2 239001G010 ...(KA*S)
{
ANSWER 6.07 (2.50) APR val ves remain as i s (0.5) APR val ves cl ose (O.S) APR val ves cl ose (CAF) (0.5) APR valves open (0.5) APR valves open (CAF) (O.5) REFERENCE MP1 Operator Training - Systems Volume 5, TX-1337 -U&Automatic Pressure Relief System, pgs S & 10 - 14, Objectives K/A 210000 KS.01 (3.3/3.0), K/A 218000 K4.04 (3.5/3.6), X/A 218000 A2.OS (3.4/3.6), K/A 210000 K6.02 (4.1/4.1), K/A 210000 A1.OS (4.1/4.1) 218000KSO1 21BOOOK602 21GOOOA105 218000A205 21BOOOK404
~~...(KA'S)~~
~~ANSWER 6.00 (3.00) D power is not available from both NSST and RSST (1.0) Joad shedding f rom 4160 VAC (and 480 VAC busses) (0.S)~~
~~- breakers trip to align 4160 VAC Distribution System to receive power from diesel and gas turbine (0. 5)~~
~~gas turbine auto starts and reenergizes 14A, 14C & 14Eg(0.S)~~
- dienel auto starts and reenergizes 14F (0,5) g REFERENCL MP1 Operator Training - Systems Volume S, 1X-1341 - 4160 VAC Distribution System, pgs 30 & 31 K/A 262001 K3.02 (3.8/4.2), K/A 262001 A3,03 (3.4/3.5),
~~K/A 262001 A3.04 (3.4/3.6), K/A 264000 K4.OO (3.8/3.7),~~
-
~~K/A 264000 K6.08 (3.6/3.7) 262OO1A304 262OO1A303 264000K600 264000K408 262OO1K302~~
~~...(KA*S) , (,~~
c________. . _ _ i
, - _ - - _ _ - _ _ _ _ _ __ PAGE 29 6. . ' PLANT SYSTEMS DESIGNL.CONTRO_L_LAN D _.I.N_ST_R_ U_M_ E_N_ T A_ T__I O_N _ _
~~-07/09/21-LUMB, ANSWERS .- - MILLSTONE l'~~
ENSWER 6.09 (3.00) Hotn recirculation pumps run back to 20% (0.5) as limited by the dual )imiter on the output of te master controller (O.S).
or tvwasr ' c r%ektr oJ pJ- M Mg t' % m^~. W A#.M6 de.a N be W ,.(cip b.- . Reci rcul ati on pump A ,-=c ' cc'; t o 2"fj.p( 0. 5 ) due'to thc disch2rge
..
v ra i u n n n +- t,,ii mm - 3 , g, o, c' t h ""rd 1 ! -i i .. d .c t (O.3).
Recirculation pump B speed will be unaffected (0.2).
%,Q sgu.A 5%f c.' Recirculation pump B will speed up until it reaches the-li.Et c' ihe ve nc ! mit:- 2-- t 'r cpccd cc::tra! ! r (0.5) because the speed' controller b im5 t c, etch ;c n;ccd vi Uu yunw h4w do i rnd
.
mpond c : gmri (0.3). - CAF M \* O b s f u d 64 d 1= " L S g ^< Recirculation pump A speed will be unaffected. (0.2) REFERENCE MP1 Operator Training - Systems Volume 1, 1 X- 130111, Reactor Recirculation Speed. Control System, pgc36, 7, 9, 10,$12 & F1gure 3, Otsjectives 9 & 6 K/A 202002 K3.OS (3.2/3.3), K/A 202000 Kl.OS (3.S/3.5) 202OO2KlOS 202OO2K3OS ...(KA*S) ANSWE .1O (3.00) a. (- both pumps are tripped 44r't))
~~(- RR cross-tie or cross-tie bypass val ves are cl osed -44rP4) - reactor. pressure must decrease below 900 psig (O.g)g This ensures that flow imbalances do not confuse the logic ( 0.[)~~
is a 2 second After ip determines how many pumps are running.there TD (0. 27 then jet pump riser pressures are compared 0.' manifold H is > 1 psid lower than manifald A ( loop A is selected, atherwise Icop B is selected ( O. 2 (0.2(, loop selection
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~~After a 1/2 secon TD to allow for relay action is sealed in ( . A close signal is sent to the selected loop discharge valves (4ve+~~
~~(0. 2 . %i u m a n"1 * i;c . m.c pump to trip 1 i it nuu i us u a . .g n=n then ir apu is il now aloning s v. a T.h e e t %-~~
A cl ose si gnal is sent to the RR cross-tie or cross-tie bypass ( O . [. REFERENCE MP1 Operator Training - Systems Volume 5, TX- 1335 -- Low Pressure Coolant injection System, pgs 22 - 26, Objectives 24 & 27 K/A 203000 K4.11 (4.O/4.0), K/A 20300 A3.07 (4.2/4.6) 203OOOK411 203OOOA307 ...(KA'S) l
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PAGE 30 bi__EhOdl_E151EUE_EEE1Edi_E9dlO9hE_Qgp_1 INSTRUMENTATION ANSWERS --; MILLSTONE 1 -87/09/21-LUMB, T.
ANSWER 6.11 (2.00) MAIN STEAM LINE RADI AT ION MONITOR DOWNSCALE al arm (1.0) hall scram channel A (0.34)
~~- hall Group I isolation (0.33) - half mechanical vacuum pumpisolation (0.33)~~
REFERENCE MP1 Operator Training - Systems Volume 7, TX-1406A - Process Radiation Monitoring Systems, pgs 19 & 20, Objectives 6, Bb &9 K/A 272OOO K1.OG (3.6/3.9), K/A 272000 K1.09 (3.6/3.8), K/A 272000 K4.03 (3.6/3.9) ...(KA'S) 272OOOK403 272OOOK109 272OOOK100
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~~Zs__P99EEU90E9_~_U90UOht OEU90UOb2_ggggggggy_QyD R A D- I O L O. G.I..C~~
~~- ---- A L C O. N T R O L - --- ----- ~B7/09/21-LUMD, ANSWERS -- MILLSTONE 1 i~~
l l ANSWliR 7.01 (1.00) This mode prevents the constant pressure and flow adjustments that would g i result if both controllers were in automat ic since the controllers oppose cach other. (1.0) REFERENCE OP-303, Reactor Cleanup System, pg 7 MP1 Operator 1 raining - Systems Volume 2, TX-1303 - Reactor Water Cleanup System, pqs 51 L 52, Objective 34 K/A 204000 K5.05 (2.6/2.6), K/A 204000 KS.06 (2.6/2.6) 204000K506 2040COK505 ...(KA'S) ANSWER 7.02 (3.00) backup air botties (0.5) can operate a variable 1ength of time depending on the 1eakage throngh the FRV and the number of valve cycles (0.5) No. (0.25) Only the outboard valves would be expected to close because the inbnard valves are supplied with nitrogen from the Drywell Instrument Air System. (0.5) SHGTS is started to maintain reactor building pressure (0.25) since the buiIding venti 1ation dampers faiI closed and ventilation ians will trip on a loss of instrument air. (0.5) The RBCCW surge tank fi1I valve f ai i s open on a 1oss of i t est rumen t ai (0.5) REFERENCE ONP-512, Rapid and lotal Loss of Instrument Air, pg 2
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MP1 Operator Training - Procedures, TX-1500A, Off Normal Procedures, pgs 107 - 109, Objectives 91, 92, 93 L 94 K/A 295019 AK3.01 (3.3/3.4), K/A 295019 AK2.02 (2.9/3.0), K/A 295019 AK2.05 (3.4/3.4), K/A 295019 AK2.08 (2.0/2.9) 295019K205 295019K2OO 295019K301 295019K202 ...(KA'S)
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~~. - MILLSTONE 1 -07 / 09 / 21 -L.UMB , ANSWEHS l )~~
~~ANSWEH /. 03 (2.50)~~
- 4 is the minimum number of SHVs that wi11 generate sui 1icient st earn flow . (0.S) to assure adequate core cooling (f ewer would require a higher RPV pressure). (O.S) - Decreasing RPV pressure while injecting w o u 3. d indicate a net inventory los (O.S)
~~tasa M SPAst%s' &~~
- SO psi pressure di f ierenti al cm:I i ^ :' " + " ~ ' " ' ' * 5"2 i d '" ' " ' ' ' ' ' " " " " .m.,mt r o, ur<vs open) (075) .ca 8 ' i r- i St t a remove the deca */ heat (, generated 10 minutes after a scr am f r om full power) (015)
HEFERENCE MP1 Operator T rai ni ng -- Procedur es, T X-- 1SOOB - Emergency Operating Procedur es , pgs 109 & 110, Objective 40 K/A 293031 EK2.01 (4.4/4.4), K/A 29b031 EK'2. 02 (3.8/3.9), K/A 2VSO31 Enl.13 (4.3/4.5), K/A 293031 EA2.04 (4.6/4.0) ...(KA'S!
~~'295031K201 295031A204 295031A113 295031K202 ANSWEH 7.04 (1.30) <O.S)~~
II temperature is too 1ow the oxygen and hydrogen wi11 not recombine un _l.caustng an increase in total f1ow (due ta increase in vol urre ) (O.5) 2 2, a potent s al l y explosive mixture in the aff-gau line (O. S ) G C,5 ca A 3.M5de..were eb e et.4.b Ch. L HEFERENCE OP-324C, Augmented Off-Gas System, pg b Augmeated O f f --Gas MPt Operator Training - Systems Volume 14, TX-1324C Sys t ern , pgs S7 L SU, Ubjective 31 K/A 271000 K4.04 (3.3/3.6), K/A 271000 KS.09 (2.6/2.0), - K/A 271000 A2.13 (2.4/2.0) ...(KO'S) 271000KSO9 271000K404 271000A213 l _ _ _ ______
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~~PAGE 33 2 __I.39EES90E$_'__d9000bL OEd93UOhL. EMERGENCY AND RA I OL.OG.I CAL __CO_N_T ROL_~~
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ANGWERS - MILLSTONE 1 -07/09/21-LUMB, ~ g\w Suca Et S O _ g% Ng + (USA O nu hn,x2 SW ANSWEft 7.05 (2.00) .ra ,%b(w s c.rw s d h f , 3 faking the mode switch to SHUTDOWN supplies additional scram signal (0.5), seals in a rod block ( 0. 5 ) and bypasses the o r ecc U25 psig MSIV cl osure signal to maintain steam flow path to condenser (O.5). If the EPR system is not functioning properly MSIV isolation at 025 psig will prevent an uncontrolled vessel blowdown and cooldow (O.5) REFERENCE MP1 Operator training - Procedures, TX-1500A - Off Normal Procedures, pq 18, Objective 11 K/A 295006 AK2.07 (4.0/4.1), K/A 295006 AA1.01 (4.2/4.2) 295006K207 295006Al01 ...(KA'S) ANSWEll /.06 (3.00) lii gher fuel temperatures are required to produce the thermal gradient oticessary to remove decay heat from the core with steam. (0.75) This water level corresponds to a fuel uncovery time that will allow the fuel to reach a temperature that will generate sufficient heat transfer yet masotain an adequate margin to thermal limits. (O.75) or t.ac.,L ero." M A3A ho -41 E.~ 6 e 4 [4WL ioh~ N L o r) /00 psig is the mintmum RPV pressure that will produce adequate steam ilow through one URV to limit peak clad temperature to 22OO^F (or to assure adequate core cooling). (0.75) The surge of steam flow resulting from rapid depressurization will lower iuel temperatures provi di ng addi ti onal time to establich a source of coolant snjection. (0.75) REFERENCE MP1 Operator training - Procedures, TX-15008 - Emergency Operat ing Procedures, pgs 49 & 50, Objective 24 K/A 295031 LKl.01 (4.6/4.7), K/A 295031 G.07 (3.7/4.0) 2950310007 295031K101 ...(KA"S) l
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~~PROCEDURES - NORMAL._ggNORMOLh_ EMERGENCY _AND PAGE 34 R.ADIOLOGIC .-- ----- CONTROL~~
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ANSWERS -- MILLSTONE 3 -87/09/21-LUMD, ANSWEf( 7.O/ (l.SO) The bundle shall be located in.the nearest available pool rack (or reactor care location) prior to evacuation (0.75) to prevent fuel uncovery (0.7S) REFERENCE step 6.10 OP-320C, Fuel Loading / Unloading /Shuf f li ng, change 1, pg 1, K/A 295023 AK3.01 (3.6/4.3), K/A 295023 G.10 (3.9/4.2) 295023K301 295023G010 ...(KA'S) ANSWER 7.00 (2.SO) Select rod insert - FW heating is 1 ant due ta 1 cad reductann atidi og poui t i ve reat t i vi t y. (O.25) Rod insertion decreases reactor power to compensate ior the reactivity additio (O.S) APRM high ilux setdown - to preserve MCPR margins (0.25) ii the select rod insert f ai l s concurrent with the lower FW inlet temperature and higher 1ocal power 1evels. (O.5) t o control reactor pressur (O. 5) 1)ypass valves open - Control and Jntercept valves throttle - tothe prevent a turbine overspeed from stored energy in the moistur e separators. (0.5) REFERENCE ONP-SO3A, Loss of 34SKV Transmission (Load Reject), pg 4 MP1 Operator Training - Procedures, TX-lbOOA - Off Normal Procedures, . pg 26, 27, Objective 21 K/A 2'75005 AKl.01 (4.0/4.1), K/A 295005 AKl.02 (3.2/3.6) 29500SK101 29SOOSK102 ...(KA'S)
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-- __ . __ __ 'P AGE ' 35 1 PROCEDURES - NORMAL g_ADNORMAL u EMERGENCY ANDl BOD 196991C06_CgNIBQL 1- -07/09/21-LUMB,. ANSWERS -- MILLSIONE ANSWER 7.09' (3.00). Natural circulation will not occur ii . water level i s below +50" ( 0. 5 ) ' -b.1)' Without adequate recirculation flow, er. Jwu tc T1a -i x U: n;; . ..o m ter ca" c- " m:. man , ivi --J tc.pcr2t r ci tc !:c eiq"i'ic M 1y I um 'A tc r m t P"" in *"" cccc c--1- (0.5)corehMru p bg M t* N <**
level 2) Operation -af only H recirc pump and Shutdown Cooling with N^d '" below +50" can-shart circuit the core. (0.5) be bej"'.J: ) could viol ate Tech Spec containment r requirements. (0.5)- Hcb 3) hazards to personnel in'the area above the 'aactor vessel'(0.S)- E I -2) could Iead t.o repressurization (O.5) c *d REFERENCE-OP-206, P1 ant Cool down to Cold Shutdown, pg MP1 Operator Training - Procedures, IX-12OO --General Operating Procedures, pqs 69, 70.& 71,-Objective 63 L'6'1
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~~K/A 293000 K1.33 (3.1/3.3), K/A 293000 K1.37 (3.2/3.4) ,~~
~~. 2V3OOOK135 293OOOK137 . ...(KA'G)~~
~~ANSWER 7.1O (2.00) !~~
~~(0,5)~~
~~or more GRV taiIpipes~~
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~~l, SAFETY . AND BLOWDOWN LEAKAGE - 260^F in ( 0. 5 ) SAFETY / RELIEF VALVE OPEN - 10 psig in 1 or more SRV tailpipes Verify that the relief is stuck upen by observing:~~
~~- yetlow 1ight on ARP section of CRP indicating steam -~~
~~flow in tro % ko el osc. . dc ks~s tai 1 pipe (0.25)~~
- temperature recorder i nd i ca t i on (0.25) - h ,, Nu 9tcJ flew Ara Mk - SRV switch position / position i n J i ca ti on (0.25) - Rg gw ch y , - Torus temperature (i ncrease) (0.25) _e g 6 Q.g ,J eccJ,e - Generator electrical output (decrease) (O.25) _, g ,,p,g g - RPV pressure and water level oscillations (0.25) . (any 4 L) 0.25 each)
REFERENCE MP1 Annunciator Alarm and Response Procedures, CCP 903 A-1 window 6-2 and CRP 903 A3-2 window 4-0 MP1 Operator Training - Systems Volume 5, TX-1337 - Auto Pressure Relief System, pgs 7 & O, Objectives 2a & 2h K/A 239002 A2.03 (4.1/4.2), K/A 239002 G.B (3.9/3.7), K/A 239002 G.15 (4.3/4.3) ...(KA'S) 239002G01S 239002GOOB 239002A203
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~~??__C09EE990EE_~_U90UOh1_OEU99UOha_gDggGENCY gNp _~~
B0919699190L_ggNIBgL-07/09/21-LUMB, f ANSWERS -- MILLGTONE 1 ANSWER 7.11 (3.00) 1 a.(ONP 512) ONP 502 (0.J/5 erretrt-b. ONP 516, ONP 509 (0.25 each) c. ONP-502, EDP 570, EOP 571, EUP 572 (0.125 each) g,g,y ( 9r+- each ) d. ONP 504, ONP 502,(EOP 570, EOP.571, EOP572) NONE (0.5) ib "' ONP SISA't om a i REFERENCE MP1 Operator Training - Proceduren, T X-1500A Of f -Nortnal Procedures, pas 6, 7, 62, 63, 103, 123, 136, 137 Objectives 6 & 10 TX-1SOOD - Emergency Operating Procedures, pg 4 Object.iven 2 h 47
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~~K/A 295006 G.11 (4. 3/ 4. 5) , K/A 295006 G.12 (3.8/4.4) 295006G012 295006GO11 ...(KA'S)~~
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~~PAGE 37 9. .._._ggg191gIggllyE_ggggEpygEgg_ggNg}IlgNg,__Ogg, LigilgJ1gNg~~
~~--87 / 09 / 21 -LUMU , ANSWERS - MILLSTONE 1 I~~
ANSWEli O.01 (2.50) mrem / calendar quarter (0.5) provided that his lifetime dose does not exceed S(N-10) Rem where N in the age in years. (O.D) mrem / quarter.(0,5) mr etn/ quar ter (O.S) foe m compl eti o;, oi an "Irtcreased Radiation Exposure Authorization"^(HP I'orm 4902-D) -4 0 :25+(and Health Physics Supervisor approval)(0./S) REFERENCE SHP 4'? 2, Exter nal Radivition Exposure Cont n1 and Dosimetry lusue, pgs 7, O L 10 K/A 294001 K1.03 (3.3/3.0) g
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294001K103 ...(KA'S) ANSWER H.02 (1.50) lagging requi r ements (0.5) Technical Specifications t i dersti f y any LCO requirements and per f or m ' any necescisry surveillance testiteg of redunaant safety systems) (0.S) Tie t est requirremen t s (approve or amend retest recommendations) (0.S) 4 I4WP ( O 5) 3 O Q T each REFERENCE ACP-DA-2.02C, Work Orders, Rev 13, sections 3.10 & 6.42, pas 11, 12 L 23 K/A 244001 K1.02 (3.9/4.5), K/A 294001 A1.03 (2.7/3.7) ! 294001K102 294001A103 ...(KA'S)
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~~1 J ION _S_,- __A_.N_D_ PAGE 38 l 8 ADMIN --.I ST-_ . -_I-V_ E PRO _CE.-DU_RE_S_p RAT - - . _ _ _ C_ON_.D_I _ __ M.-_I T A_T__I O__NS _ -~~
~~- v.---. i ANSWERS -- MILLSTONE 1 -87/09/21-LUMB, T.~~
L I l i
i J ANSWER U.03 (3.00) l No, operation can't continu ( 0. 5 ) The diesel generator and gas turbine must be declared inoperable because j this woul.d disable the auto start feature of the diesel generator and gas turbine. ( 0. 5 ) T/S 3.9.A requires both emergency power sources ne operao19 when the
~~. reactor is critical (0.5)~~
1/S 3.9.B.4 ril l ows reuctor operation as long as two offsite lines (345 or 2-' . 6 k v) are operable and T/S 3.5.F is met. (0.5) T/S 3.5.F.1 requires that both emer gencv power suorces be cperable whenever irr adi ated iuel is in thrs reactor (0.5).
(T/S 3. 5. F. 2 and 3.5. F.3 al so cannot. be met due t o the EDG and gas turbine inoperabiiity.)
Since the requirements of I /S 3. 5.F cannot be met, an or derl y shutdown muf5 t be initiated and.the reactor must be in the COLD SHUTDOWN or refuel condition withan 24 hours. (O.5) REFERENCE MP1, Technical Speciiicatians, pas 3/4 5-8,TX-1330 9-3 , 5 t d 12, TO 3. "L L , , g 2 ~k l MP1 Ope: ator Training - Systems Volume 5, - Diesel Generator, Objective 27 T X -- 133(7 - Gas !urbine Generator, MP1 Operator Training - Systems Volume 5, Ob.j ec t i ve fl9 K/A 264000 G4 11 (3.4/4.1) 264000G011 ...(KA'S1 DY TS Tc k 3.2 7 - w % ck m er .5%% L e n k LoS A *
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~~PAGE- 39 ADMINISTRATIVE PROCEDURFSg _ CONDITIONSg _AND_ LIMITATIONS-87/09/21-LUMB, ANSWERS -- MILLSTONE 1 ANSWER O.04 42.50)~~
~ Operat ors should believe instrumentation unt11 it is proven faulty by rilrect comparison with another instrument reading the same oriable, b'y iunctianal test or by calibration. (O.75) The purpose of this rule is to ensu:e that psychological rejection does not interfere with operating decision (0.5) to take emergency remedial action (0.23) - when periorming Shutdown from Outside the Control Room (0.2S) once per shift (0.25) -- t o ideatify t. rends or abnormal readings (0.2S) and to verify that data has been properl m specy recorded)(0.23) (c.2 7 2 @ C. 2 F t qA "
ucr pcads REFERENCE 6.1.7 & 6.1.0, ACP-6.01 Control Room Procedure, Rev 10, sections 6.1.4, ! pgs 6 & 7 MP1 Operator 1 raining - Systems Volume 7, 1X-1100 - Administrative Procedures, pgs 57 L SO, Objectives 59, 67 L 60 K/A 294001 A1.11 (3.3/4.3), K/A 294001 A1.12 (3.S/4.2;
~~'74001A112 294001A111 ...(KA'S)~~
~~l ANSWER O.OS (2.50) whenever the reactor is critical (O.S) J-OR-~~
~~- whenever the rx plant > 212^F (0.25) ,~~
~~with fuel in the vessel (0.2S) ( All manual containment isolation valves not r equi red it be opcn deari ng accirjent condi ti ons are cl used (O.3) -~~
- At least one door in the airlock is closed ano sealed (0.3) - All autom at i c con t ai nraen t i; sol a t i ott val ves are operabl e or at e deactivated in the isol at iore posi t i on (0.3) - All blind flanges and manways are closed (0.3) - Dr ywel l and suppression Chamber are intact (0.3)
REPERCNGE MP1 Tecnnical Specifications, pgs 1 -2, 1 - 3, 3/4 7-4 MP1 Operator Training - Systems Volume 2, TX-1311A - Primary Containment ) System, pgs 64 L 6$, Objectives 31A L 32 K/A 223001 G.11 (3.3/4.2) 2230010011 . .(KA'S) i i
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_ _ _ _ _ _ O*-__gDMlNiglRgllyi,,, PROCEDURES n 'ONDI_TIUNS i AMD_t_ IMITATIONS PAGE 40 ANSWERS - MILLSTONE 1 -87/09/22-LUMD, ANSWER U.06 (J.00) . INIENT (0.5) NON INTENI (O.S) INTEN1 (O. S) . approval of 2 11 ceased fsiROs of which 1 must be the on 37 g g shift SS for implementation (0.23) Of#te+PORC review if required (0.125) -AND_ O b E MW Station / Unit Superintendent approval (O.125) Of- A/W- h within 14 day *5 of implementation (0.25) ,gg (f. , d (.c d i S tto , MGHtWPORC revi ew :i required (0.25) - fin D-E'a':m / Unit Superintendent approval (O.2S) Ior implementation and final approval (0.2S) REFEHENCE ACP-DA-3.02, Station Procedures and Irorms, Rev 40, sections 4.7, 4. 8 < 6. 9.1. 3 & 6.9.2.4, pgs 4, '.5 , 30 & 33 MPI' Operator Training - Systems Volume 7, 1 X-1100 - Admi ni strat i ve Procedures, pg 44,, Objective 48 K/A 2Y4001 A1.01 (2.9/3.4) 294001A101 ...(KA'S) ANSWU4 0,07 (3.00) T/S 'lable 3.1.1 requires that there be a minimum of two (2) operable channels per t r- i p system for the APRM scram functions. (O.S) RPS ch,tnnel A does nct meet this requirement, therefore the system must be t ri pped (half scram inserted on RPS channel A). (O. S) T/S Table 3.2.3 requires that there be a r:inimum of one (1) operable channel pi r trip 1;ystem for the APRM rod block f unc t.i ons. (0.5) Both channels meet !.hi s r equi rement, but Note (7) requires at ! east four (4) operab1e APRM channels (O.3 +rrs-+xuuM44m--may ex i - t ! 0; .m. v , '/) de .,w --+ tn u; mr-U "2 md b ! n'- P J u nct.4m-<d-t+1t- up er a b i t-- .irystem---frh ai m u i bi 2 ; bcm t ~1 ; u.nu& ,t :1 -y 'nd daily t he r a f t. u . (0.L). !l - tl,m i: LN e,xi.s14 -aMui _.u v ui . (7) days"ttierr tne inuptrnble syntci. innannes FT) .. .m , i Lt-
~~' I'I 5U* UQ A,t b.a A ft' ue* gg d 4 im G (.a (D1 d61 y . AwQ s REFERENCE '~~
D6IO MP1 Tuchnical Specifications, pgs 3/4 1-3, 1-5, 2-7, 2-0 K/A 215005 K4.01 (3.7/3.7), K/A 215005 K4.02, (4.1/4.2), K/A 215005 G.11 (3.4/4.)) 21SOOSK402 21SOOSG011 21SOOSK401 ...(KA?S)
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___-_- ._ - t 0 _0901NiglggllNg_PPOCEQUREg,,_ggNgillgNgy,_@Ng_LidlI@llggy PAGE 41 O ANSWERS -~ t11LLSTONE 1 -07 / 09 / 21 --LUM B , ANSWER U ., 00 '(2.00)
~~; SchedulIed date 4/- I week (O.S)~~
~~not less than 12 times / year (0.5) Not i.o exceed 25% of the test interval ( 0. S) Not to exceed 3.25 times the test interval (0.5) ,a REFERENCE~~
~~ACP-DA-9.02, Station Survei11 ance Program, Re IS, section 4.5, pgs 4 &S K/A 294001 A1.03 (2.7/3.7)~~
294001A103 ...(KA'S) ANSWER U.09 (2.00) IOO^l'/ hour (0.S) To ensure stress i n t.ensi t y and fatigue limits for the reactor vessel are not exceeded and react or vessel int egri ty will be maintained for its design 1ifetime. ( 0. 5 ) l l ^F/ hour (O. S) To ensure that the T/S limi t is not violated (0.S) REFERENCE MP1 lechnical Specifications, pgs 3/4 6-1 & H 3/4 6-1
~~MP1 Operator Iraining - Procedures, TX-12OO - General Operating Procedures, l pgs 66 L 67, Objectives 20, S9 & 61 l~~
~~K/A 2V3010 Kl.04 (2.9/3.2) 293010K104 ...(KA'U)~~
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~~'S OSU.Id1E1I_ 11EE_. PROCEDURES n CONDIT1DNSg_AND_LI_MIT_ATIONS-87/09/21-LUMB, f ANSWERS'-- MILLSTONE 1-l l~~
~~ANSWER U.10 (3.00) Key Conditions Emergency Act. ion Level C1assi4ication __.___ .______ _..______ k _ _ . _ _ _ _ _ _ _ _ . _ _ _~~
~~. _...___________~~
Radioactive GE low range stack monitor j Alert (0.25) Release (0.25) reads > 10 times TS limits i for > 15 minutes (0.25) i Barrier Failure Rx Bldg area temp or rad General Emergency (0.2S) or Imminent alarms and CST level Barrier Failure decreasing and valve open (0.25) indication (0.25) l Loss of Power ( 0 . 2'5) 4KV bus trips and RSS1 linusual Event (0.2S) undervoltage alarms (O.25) ,
~~.i Relief Valve discharge j Unusual liven t (0.25) Barrier Failure '~~
[ or I mmi nen t high temp alarm or torus flarr i er Failure water high temp alarm or (0.25) safety / relief valve open i indication (0.25) i i l HEFERENCE l l EI'I P 4 701, Unit locident Assessment, C1 ossification and Repor t abi l i t.y, 1 F or m 4701-1, pgs 1, 2, 3& 4 K/A 294001 A 1. I t2 (2.9/4.7) 29',6001 A116 ., (KA'S)
~~. _- _. _ __ _ _ _ _ _ _ -~~
~~- - - - - _ - _ - _ - - - - - - -~~
~~TEST CROSS REFERENCE PAGE 1 UUESTION' VALUE REFERENCE~~
~~.------ - -------- --------~~
l 05.01 2.50 TELOOO1101 05.02 3.00 TELOOO1071 05.03 2.50 1ELOOO1133 05.04 2.00 lELOOO1102 03.05 3.00 TELOOO1100 l 05.06 2.00 TELOOO1072 I 05.07 2.S0 1ELOOO1179 05.08 3.00 TELOOO117S 05.09 3.00 TELOOO1170 05.10 1.50 IELOOO1139
~~------~~
25.00 06.01 1.50 IELOOO1172 06.02 3.00 TELOOO1198 06.03 2.00 TELOOO1157 06.04 2.00 IELOOO1210 06.0S 2.00 TELOOO1204 06.06 1.00 TELOOO1199 06.07 2.S0 TELOOO1166 06.00 3.00 IELOOO1202 06.09 3.00 TELOOO1170 06.10 ,5.00 TELOOO1201 06.11 2.00 TELOOO1211
~~------~~
2S.00 07.01 1.00 TELOOO1104 07.02 3.00 TELOOO1120 07.03 2. S0 IELOOO1100 07.04 1.S0 IELOOO11GS 07.0S 2.00 TELOOO1103 07.06 3.00 TELOOO1143 07.07 1.50 TELOOO1212 07.00 2.50 IELOOO1177 07.09 3.00 TELOOO1103 07.10 2.00 TELOOO112S < 07.11 3.00 TELOOO1107
~~------~~
25.00 00.01 2.50 TELOOO1206 08.02 1.50 TELOOO1191 00.03 3.00 TELOOO1082 00.04 2.50 TELOOO1143 08.0S 2.50 TELOOO1190 08.06 3.00 TELOOO1192 00.07 3.00 TELOOO1004 08.08 2.00 TELOOO1144 00.09 2.00 TELOOO1195 _ _ _ _ _ _
~~.. - _ - - _ _ _ - _ - - _ _ - _ _~~
~~PAGE 2 TEST CROSS REFERENCE QUESTION . VALUE REFERENCE~~
~~-- --- --.------- - - - - - - - - -~~
~~00.10 3.00 TELOOO1197~~
~~. - - - . - -~~
~~25.00~~
~~-_----~~
~~goe-m # dels - 100.00 i~~
.
~~. - . _ ----_- _---- _ --- _ ___ _~~
f j -- L 120 -
, 11 0
~~- ggg" l 90%~~
PUMP SPEED 80% 100 - 70% 90 - 607 50% 80 - 40% C E 70 - y 30%
g 60 - 20%[/ 5 / * 50 -
~~/[ !~~
~~40 -~~
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~~' + i I- * " '[ ' I' l I J 10 20 30 40 53 6C 70 83 90 100 i '; O O :~~
~~PE:: CENT CORE rLOW FlGURE 1 BWR POWER FLOW MAP~~
~~.. ,~~
l
~~-- . - - - - - - . - - - _ - - _ _ . - , _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ______~~
~~_ - _ - - _ _ _ - _ _ - - _ _ _ _ _ - - _ . - _ _ -~~
~~. . .x, * ,* , , .. . 500 UNIT 1 EMERGENCY PROCEDURES INDEX ... (Procedures require PORC review) ( ,~~
~~OFF WORMAL PROCEDURES . , I~~
^
NUMBER TITLE RE EFF. DATE ONP 502 Emergency Plant' Shutdown 2 08/13/86 ONP 503A Loss of 345 KV Transmission.(Load Reject) 2 05/04/87 l ONP 503B Loss of All Station A.C. Power (LNP) 1 05/24/85 ONP 503C Loss of Off-Site and On-Site /24/85 Power (Station Blackout) ONP 503D Response to a Request for Generation Reduction 0 04/25/86 1 I ONP 504 Recirculation System Failures 1 05/24/85 i
~~'ONP 505/ Fire 1 05/30/85 A0P 2559 ONP o06 Total Loss of Station 125 VDC Power 0 12/31/85 ;~~
JNP 507 Loss of Vacuum 2 04/15/87 i ONP 508 Fuel Cladding Failure 1 05/24/85 ONP 509 Excessive Radioactive Levels 1 05/24/85 ONP 511 Plant Shutdown from Outside the Control Room 1 05/24/85 ONP 512 Rapid and Total Loss of Instrument Air 2 02/11/87 ONP 514A Natural Occurrences 2 04/02/86 ONP 514B Freezing Temperatures 2 05/24/85 ONP 514C Earthquake 1 05/24/85 ONP 515A High Conductivity Prior to Condensate 0 05/24/85 Demineralized i l
~~, !~~
~~Rev. 111 04/28/87 Page 1 of 3 ; l l~~
'
~~m _ _____~~
~~, , , ., g; t.' . . , *- 500 UNIT 1 EMERGENCY PROCEDURES INDEX (Procedures require PORC review) (- -~~
~~EMERGENCY OPERATING PROCEDURES~~
'
NUMBER TITLE RE EFF. DATE DNP 515B High Conductivity After Condensate .0 05/24/85 Demineralizers ONP 515C High Conductivity Reactor Water 0 05/24/85 l ONP 516 High Energy Pipe Rupture 0 05/24/85 ONP 518 Inadvertent Criticality 1 05/24/85 ONP.519 Dropped Fuel Bundle 2 01/28/87 l ONP 521 Loss of Water Inventory in Reactor 3 11/21/86 Cavity or Fuel Pool ONP 522 Detonation in the Off Gas System 1 05/24/85 ONP 523 Loss of Feedwater Heating 1 05/24/85 ( SNP 524A Loss of Turbine Building Closed 2 12/10/86 Cooling Water System ONP 5248 Loss of Turbine Building Secondary 1 12/31/86 Closed Cooling Water ONP 524C Loss of Reactor Building Closed 2 12/31/86 Cooling Water ONP 524D Loss of Service Water 0 02/12/86
-
,
,
~~Rev; 110 4/7/8'7 , Page 2 of 3 cf' a___:______ _ _ _ _ _ .~~
~~, ,. _ _ - _ - -- - - . _ . .. ~ . , , . , .- ' 500 UNIT 1 EMERGENCY PROCEDURES INDEX (Procedures require PORC review) [7 >- ,~~
EMERGENCY OPERATING PROCEDURES . NUMBER TITLE RE EFF. DAT .E0P 569 E0P Administration Procedure 1 03/20/84 E0P 570 RPV Level Control Emergency Procedure 2 01/28/87 E0P 571 RPV Pressure Control Emergency Procedure 1 03/20/84 E0P 572 Reactor Power Control Emergency 2 01/28/87 Operating Procedure E0P 573 RPV Spray Cooling Emergency Procedure 1 03/20/84 E0P 574 RPV Steam Cooling Emergency Procedure 1 03/20/84 E0? 575 RPV Level /Rx Power Control Emergency 1 03/20/84 Procedure E0P 576 RPV Level Restoration Emergency Procedure 1 03/20/84 E0P 577 Emergency RPV Oepressurizatfon Emergency 1 03/20/84 ( Procedure
'
~~E0P 578 RPif Flooding Emergency Procedure 1 03/20/84 E0P 579 Alternate Shutdown Cooling Emergency 1 03/20/84 Procedure E0P 580 Containment Control Emergency Procedure 1 03/20/84~~
~~. . . .;~~
~~Rev. 110 4/7/87~~
~~- Page 3 of 3 L-_---_____-~~
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~~TECHNICAL SPECIFICATIONS~~
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~~MILLSTONE NUCLEAR POWER STATION ,~~
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UNIT NO. 1 , i i Docket No. 50-245 Appendix A to License No. DPR-21
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~~JAN 2 01987-~~
.c TABLE OF CONTENTS Page N .0 DEFINITIONS ...................................'............ 1-1 SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS 2.1.1 FUEL CLADDING INTEGRITY............. 2.1.2 ................ 2-1 2.2.1 REACTOR'C00LANT SYSTEM.............. 2.2.2 ................ 2-7 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT . GENERAL ..................................... 4,0 ......... 3/4 0-1 3.1- REACTOR PROTECTION SYSTEM ................... 4.1 ......... 3/4 1-1 : PROTECTIVE INSTRUMENTATION . . . . . . . . . . . . . . . . . . 4.2 . . . . . . . . . 3/4 2-1 A. Primary Containment Isolation Functions ................. 3/4 2-1 B. Emergency Core Cooling Subsys tems Actuation . . . . . . . . . . . . . 3/4 2-1 ( C. Control Rod Block Actuation ............................. 3/4 2-1 D. Air Ejector Off-Gas System............................... 3/4 2-12 E. Reactor Building Ventilation, Steam Tunnel Ventilation' Isolation, and Standby Gas Treatment .
System Initiation ...................................... 3/4 2-12 1 REACTIVITY CONTROL .......................... 4.3 ......... 3/4 3-1 z I A. Reactivity Limitations ..................... A ........- 3/4 3-1 ] B. Control Rod Wi thdrawal . . . . . . . . . . . . . . . . . . . . . B ........ 3/4 3-2 j C. Scram Insertion Times ...................... C ........ 3/4 3-5 j D. Control Rod Accumulators ................... 0 ........ 's/4 3-7 ) E. Reactivity Anomalies ....................... E ........ 3/4 3-8 F. Shutdown Requirements ................................... 3/4 3-8 G. Thermal Power - Core Flow ............................... 3/4 3-8 STANDBY LIQUID CONTROL SYSTEM ................ 4.4 ........ 3/4 4-1 A. Normal Operation ....................... A ............ 3/4 4-1 : B. Operation with Inoperable Components .................... 3/4 4-3 I C. Boron Requirements ..................... C ............ 3/4 4-4 D. Shutdown Requirement .................................... 3/4 4-4-( i Amendment No. 1 I _ _ _ _ _ _
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~~i - JAN 2 91987 l . Surveillance Pace N j { CORE AND CONTAINMENT COOLING SYSTEMS /4 5-1~~
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A. Core Spray and LPCI Subsystems . . . . . . . . . . . . . . A . . . . . . 3/4 5-1 B. Containment Cooling Subsystems............... B ...... 3/4 5-3 C. FWCI Subsystem .............................. C ...... 3/4 5-5 D. Automatic Pressure Relief Subsystems ........ D....... 3/4 5-6 E. Isolation Condenser System .................. E ...... 3/4 5-7 F. Minimum Core and Containment Cooling System Availability ....................... F ...... 3/4 5-8 )1 PRIMARY SYSTEM B0UNDARY /4 6-1 A. Thermal Limitations ......................... A ...... 3/4 6-1 B. Pressurization Temperature .................. B ...... 3/4 6-2 C. Coolant Chemistry ........................... C ...... 3/4 6-5 D. Coolant Leakage.............................. D ...... 3/4 6-11
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E. Safety and Relief Valves..................... E ...... 3/4 6-11 F. Structural Integrity ........................ F ...... 3/4 6-12 G. Jet Pumps ....................................G ...... 3/4 6-13 H. Reci rculation Pump Flow Mi smatch . . . . . . . . . . . . H . . . . . . , 3/4 6-14 1. Snubbers .................................... I ... .. 3/4 6-15 J. Condensate Demineralizers ................... J ...... 3/4 6-21 CONTAINMENT SYSTEMS /4 7-1 A . P rima ry C on t a i nme n t . . . . . . . . . . . . . . . . . . . . . . . . . A . . . . . . 3/4 7-1 B. Standby Gas Treatment System ................ B ...... 3/4 7-10 C. Secondary Containment ....................... C ...... 3/4 7-13 D. Primary Containment I sol ation Valves . . . . . . . . D . . . . . . 3/4 7-14 RADI0 ACTIVE MATERIALS /4 8-1 A. Radioactive Liquid Effluent Instrumentation . A ...... 3/4 8-1 B. Radioactive Gcseous Effluent Instrumentation. B ...... 3/4 8-6 C. Radioactive Liquid Effluents ................ C ...... 3/4 8-12 D. Radioactive Gasious Effluents................ D ...... 3/4 8-14 AUXILIARY ELECTRICAL SYSTEM /4 9-1 3.10 REFUELING 4.10 3/4 10-1 A. Refuel i ng Interl ocks . . . . . . . . . . . . . . . . . . . . . . . . A . . . . . . 3/4 10-1 B. Core Monitoring ............................. B ...... 3/4 10-2 C. Fuel Storage Pool Water Level . . . . . . . . . . . . . . . C . . . . . . 3/4 10-3 D. Crane Operability ........................... D ...... 3/4 10-3 E. Crane Interlocks and Switches................ E ...... 3/4 10-3 ( ii Amendment No. 1 ! _ - __ - __ - _
. UAN 2 91987 Surveillance Page N ' .11 REACTOR FUEL ASSEMBLY 4.11 3/4 11-1 A. Averaae Planar Linear Heat Generation Rate (APLHGR)................... A ...... 3/4 11-1 B.' Linear Heat Generation Rate (LHGR)........... B ...... 3/4 11-5 C. Minimum Critical Power Ratio (MCPR). . . . . . . . . . C . . . . . . 3/4 11-6 3.12 FIRE PROTECTION SYSTEMS 3/4 12-1 A. Fire Suppression Water System ................ A ...... 3/4 12-1 B. Spray and/or Sprinkler Systems . . . . . . .. . . . . . . B . . . . . . 3/4 12-5 C. Carbon Dioxide and Halon 1301 Systems ....... C_...... 3/4 12-7 D. Fire Hose Stations .......................... D ...... 3/4 12-9 E. Fire Protection Instrumentation ............. E ...... 3/4 12-13 F. Penetration Fire Barriers.................... F ...... 3/4-12-17
3.13 INSERVICE INSPECTION 4.13 3/4 13-1 3.14 PLANT SYSTEMS 4.14 3/4 14-1
; DESIGN FEATURES ; 5-1 ADMINISTRATIVE CONTROLS 6-1 Pespor.sibility .................................. 6-1 1 Organization .................................... 6-1 . Facility Staff Qualifications ......,............ 6-4 Training ........................................ 6-6 Review and Audit .........................a...... 6-6 , Reportabl e Event A cti on . . . . . . . . . . . . . . . . . . . . . . . . . . 6-15 afety Limit Violation .......................... 6-15 j Procedures ...................................... -6-16 !
~~6.9- Reporting Requirements .......................... 6-17 . 6.10 Record Retention .......................... ..... 6-19 -~~
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6.11 Radiation Protection Program .................... 6-20 6.12 High Radiation Area ............................. 6-21 6.13 Systems Integrity ............................... 6-21 6.14 todine Monitoring ........................ ...... 6-22 6.15 Radiological Effluent Monitoring and Offsite Dose Calculation Manual............ 6-22 6.16 Radioactive Waste Treatment ..................... 6-22 f iii Amendment No. 1 l I l
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~~OCT 311986 LIMITING CONDITION FOR OPERATION-l * 3.1 REACTOR PROTECTION SYSTEM Applicability'y:~~
~~.. Applies to the instrumentation and associated devices which initiate a reactor scram and provide automatic isolation of the Reactor Protection System ouses from their power supplie Objective:~~
To assure the operability of the Reactor Protection Syste Specification: The setpoints, minimum number of trip systems, and minimum number of instrument channels that must be operable for each position of the reactor mode switch shall be as given in Table 3. '
. Response Time-The time from initiation of any chanr.el trip to the de-energization of the scram solenoid relay shall not exceed 50 millisecond Reactor Protection System Power Monitoring 1 Two RPS electric. power monitoring channels for each inservice RPS MG set, or alternate power supply, shall be operable at all . times except as follows: With one RPS electric power monitoring channel for an inservice RPS MG set or alternate power supply inoperable, restore the inoperable channel to OPERABLE status within 72 hours or remove the associated ~
~~RP5 MG set or alternate power supply from servic I~~
/ With both RPS electric power monitoring channels for an inservice RPS MG set or alteraate power supply inoperable, restere at least one to OPERABLE Status within 30 minutes or remove the associated RPS MG set or alternate power supply from servic SURVEILLANCE REQUIREMENT 4.1 REACTOR PROTECTION SYSTEM Applicabilit Applies to the surveillance of the instrumentation and associated devices which initiate reactor scram and provide automatic isolation of reactor protection system buses from their power supplie Millstone Unit 1 3/4 1-1
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~~OCT 311986 SURVEILLANCE REQUIREMENT (Continued) l~~
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l 4.1 REACTOR PROTECTION SYSTEM ( l l l Objective: To specify the type and frequency of surveillance to be applied to the { reactor protection instrumentatio Specification: Instrumentation system shall be functionally tested and calibrated as indicated in Tables 4.1.1 and 4.1.2, respectivel Daily during reactor power operation, the maximum fraction of limiting ) power density shall be checked and the APRM scram and rod block settings 1 given by the equations in Specifications 2.1.2A and 2.1.28 shall be determined to be vali The RPS electrical protection assemblies shall be determined operable as follows: I At least once per 6 months by performance of a CHANNEL FUNCTIONAL j TEST, and J At least once per 18 months by demonstrating the OPERABILITY of over-voltage, under-voltage and under-frequency protective ! instrumentation by performance of a CHANNEL CALIBRATION including ( j simulated automatic actuation of the protective relays, tripping logic ar.d output circuit breakers, and verifying the following setpoints:
' Over voltage 5 (132)VAC, Under-voltage 2 (108)VAC, Under-frequency 2 (57)Hz, and d .' Time-delay 5 (4.0) second When the reactor mode switch is in REFUEL or SHUTOOWN and fuel is in the j reactor vessel, no trip functions are required to be operable provided i that all control rods are fully inserted, and either electrically or i hydraulically disarmed. Thereafter, daily surveillance shall be performed to verify that all control rods remain valved out or electrically disarme l l )
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~~t2 c ep h u me l t s _ eef ah o e n nn t pb u _ _ el l dt c h w uo H i a m _ _ hbb e t o fi fR rr _ _ t aah nyt d t oI t e e _ rrt el s s si p r ree vir i M is ee eo e _ _ _~~
:
s e o p pf eai F ooI sdf W R I h o TP Ot nh h n Ti h T _ t ) ) ) ) ) ) ) _ _ o 1 2 3 4 5 6 7 _ _ N ( ( ( ( ( ( ( _ _ _ _ -
_ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ - _ _ _ _
. *
l
~~' JAN 2 01987 '~~
I
LIMITING CONDITION FOR OPERATION
~~. ( .~~
~~l 3.2 PROTECTIVE INSTRUMEN.TATION~~
.+
C. The minimum number of operable instrument channels specified in Table 3.2.3 for the Rod Block Monitor may,,b l maintenance and/or testing for periods not,e in excess reducedofby 24one for in hours ' any 30-day perio ,
- ;*
~~3 /; Air Ejector Off-Gas System l'~~
- Except as specified in 3.2.D.2 below, both air ejector off gas system radiation monitors shall be operable during reacter power operation. The trip settings for the.conitors shall be set at a value not to exceed the equivalent of the instantaneous stack release limit specified in Specificatiut; The time delay setting for closure of the steam jet mir ejector off gas isolation valve shall not exceed 15 minute . From and af ter the date that one of the two air ejector off gas'
system radiation monitors is made or found to be inoperable, reactor power operation is permissible only during the succeeding'24 hours, provided the inoperable monitor is tripped, unless such system is sooner made operabl , Reactor Building Ventilation Isolation, ~ Ste.9m Tunnel Ventitatiorv Isolation and Standby Gas Treatment System Initiation Except as specified in 3.2.E.2 below,.six radiation monitors shall ; be operable at all time . . One of the two radiation monitors in the reactor building ventilation ouct, one of the two. radiation monitors on the refueling floor and one of the two radiation monitors in the steam tunnel l ventilation may be inoperable for 24 hr If it is not restored to service in this time, the reactor building ventilation system and 1 steam tunnel ventilation system shall be isolated and the standby l gas treatment operated until repairs are complet . The radiation monitors shall be set to trip as follows: Ventilation duct -.11 mr/h Refueling floor - 100 mr/h Steam tunnel ventilation - 12 mr/hr.
1 ,.; > w
* l ; /
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I
' A**"d*'"t "*
~~Millstone Unit 1 3/4 2-12 1~~
~~- . , _ . , , _ - ._ -_- _ . - _ ,~~
a J,
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~~i JAN 2 91997~~
~~, LIMITING CONDITION FOR OPERATION 'd 3.5 CORE AND CONTAINMENT COOLING SYSTEMS j L Aglicability: ~~~
~~Applies to the operational'scatus of the emergency cooling subsystem "~~
~~. Objective:~~
To assure adequate cooling ca;nbility for heat removal in the event of a loss of coolant-accident or isolation from the normal reactor heat sin S_oecification: .g Core Spray and LPCI Subsystems , Except as specified in 3.5.A.2, 3.5.F.6, 3.5.F.7 and 3.5.F.8,
'
both core sprr; subsystems shall be operable whenever irradiated l ' fuel is inNhefeactor vesse . From and ufter the date that one of the core spray subsystems is made or feand to be inoperaDit for any reason, reactor operation is; pers ysible only during the succeeding fifteen days unless such subsystem is sooner made operaust, provided that curing such fifteen dayr, all active components-:f the other core spray subsystem and the LPCI subsystem and both emergency power sources required for . operation of such componi.its,uif no external source of power were
~~' (~~
available,shallbeopt6;sbl . Except as specified in 3 5.A A; 3.5.B.3, 4, 5; 3.5.F.6; 3.5.F.7 and 3.5.F.8, the LPCI subsystem shall be operable whenever irradiated , ' y T fuel is in the reactor vesse . From and after the date that one of the LPCI pumps is made or found
~~[ to be inoperable, for any reason, reactor operation is permissible , {h '~~
~~only during the succeeding 30 cays unless such pump is sooner made operable, provided that during such thirty 0;ys the remaining active~~
~~, - ' components of the LPCI and containment cooling subt,ystem 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 were availatale. shall be operabl i A maximum of one drywell spray loop may be inoperable for 30 days~~
~~. ,~~
~~when reactor water temperature is greatr than 212* i~~
'
~~f' 6, If the requirements of 3.5. A cannot be met, an orderly shutdown of~~
*}' the reactor shall be initiated and the reactor shy [ be in the COLD SHUTDOWN or REFUEL condition within 24 hour Q ,
~~9 ,, 1 %~~
~~, ,~~
j
~~, l t'~~
~~L' ,~~
~~' Amendment No I '~~
~~fhN 'c d .~~
~~/14111 stone Unit 1 > 3/4 5-1 ; '4 )4 , -~~
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~~%,- ,;,;:.y~~
~~- ;3 - \ ' . ,~~
~~JAN 2 91937 '~~
~~. . M t!EILLANCE REQUIREME,NT (' ,~~
~~V '~~
~~' 4.3 CORE AND CONTAll' MENT.CO.OLING a, ,, SYSTEMS Applicability: q l Applies to periodic testing of the emergency cooling subsystems.~~
~~m Ob.iective: q To verify the operability of the core and ccatainment cooling subsystem Specification: , 7- A U Surveillance of the Core Spray and LPCI Subsystems shall be performed as~~
~
~~follows: ! Core Spray Subsystem Testing:~~
) i . i gg ' Frequency / ', i Simulated Automatic 3 Each Refueling Actuation Test Outage , Pump and Valve Per Surveillance Operability Requirement 4.13 Core Spray header a p instrumentation check , Once/ day calibrate i Once/3 months test Once/3 months ;r .
~~3,. LPCI Subsystem Testing shall be as specified in 4.5.A.1.a, b and c except that three LPCI pumps shall deliver at least 15,000 gpm~~
~~. against a system head corresponding to a reactor vessel pressure ,~~
~~of 214.7 psi . During each five year period, an air test shall be performed on the~~
-
~~dry ell spray headers aid nozzle ;~~
~~' )~~
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~~Amendment No. 1 e e Millstone Unit 1 ' 3/4 5-2 y 8 y~~
~~, +ll l'~~
~~ev 2 ,1~~
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JAN 2 9 ggg7 LIMITING CONDITION FOR OPERATION 3.5 CORE AND CONTAINMENT COOLING SYSTEMS Containment Cooling Subsystems Except as specified in 3.5.B.2, 3.5.B.3, 3.5.F.6, 3.5.F.7 and 3.5.F.3, both containment cooling subsystems,shall be operable whenever irradiated fuel is in the reactor vesse . From and after the date that one of the emergency service water (ESW) subsystem pumps is made or found to be inoperable for any reason, reactor operation is permissible only during the 7,ucceeding i thirty days unless pump is sooner made operable, provided that during such thirty days all other active components of the containment cooling system are operabl . From and after the date that one active component in each containment cooling subsystem or a LPCI and ESW in one containment cooling subsystem is made or found to be inoperable for any reason, reactor operation is permissible only during the succeeding 7 days provided the remaining active components in each containment cooling subsystem, both core spray subsystems and both emergency power sources for operation of such components, if no external source of power were available, shall be operabl . From and after the date that one LPCI and one ESW pump in each containment cooling subsystem is made or found to be inoperable for any reason, reactor operation is permissible only during the succeeding four days provided the remaining active components of the containment cooling subsystems, both core spray subsystems and both emergency power sources for operation of such components, if no external source of power were available, shall be operabl . From and after the date that one containment cooling subsystem is made or found to be inoperable for any reason, reactor operation is permissible only during the succeeding four days provided that all active components of the other containment cooling subsystem, both core spray subsystems and both emergency power sources for operation of such components, if no external source of power were available, shall be operabl . If the requirements of 3.5.B cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the COLD SHUTDOWN or REFUEL C0hDITION within 24 hour ,
)
~~l Amendment No. 1 Millstone Unit 1 3/4 5-3 l~~
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~~. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ -_ , . . .~~
~~dSN 2 91987~~
.
~~SURVEILLANCE REQUIREMENT 4.5 CORE AND' CONTAINMENT COOLING SYSTEMS Surveillance of the containment cooling subsystems sh'all be performed as follows:~~
~~' Emergency Service Water Subsystem Testing: ~ Item- Frequency Pump & Valve Per Surveillance Operability Requirement 4.13 . **" '~~
Millstone Unit 1 3/4 5-4 ! u_____________u.._.._ I
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.
~~JAN 2 91987~~
.
LIMITING CONDITION FOR OPERATION 3.5 CORE AND CONTAINMENT COOLING SYSTEMS FWCI Subsystem Except as specified in 3.5.C.3 below, the FWCI subsystem shall be operable whenever the reactor coolant temperature is greater than 330*F and irradiated fuel is in the reactor. vesse . There shall be a minimum of 225,000 gallons'of water in the condensate storage tank for operation of the FWC . From and after the date that the FWCI 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, provided that during such seven days all active components 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 shut'down shall be initiated and the reactor coolant temperature shall be less than 330 F within 24 hour i SURVEILLANCE REQUIREMENT 4.5 CORE r.ND CONTAINMENT COOLING SYSTEMS Surveillance of FWCI Subsystems shall be performed as follows: Frequency Item Pump and valve Per Surveillance operability Requirement 4.13 b, Simulated Automatic Every refueling Actuation Test outage Once a week the quantity of water in the condensate storage tank shall be logge I
1
Amendment No. 1 Millstone Unit 1 3/4 5-5 L >
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* . .
JAN 2 91987 LIMITING CONDITION FOR OPERATION ( 3.5 CORE AND CONTAINMENT COOLING SYSTEMS Automatic Pressure Relief (APR) Subsystems Except as specified in 3.5.D.2 below, the APR subsystem shall be operable whenever the reactor coolant temperature is greater than 330'F and irradiated fuel is in the reactor vesse . From and after the date that one of the four relief / safety valves of the automatic pressure relief subsystem is made or found to be inoperable when the reactor coolant temperature is above 330*F with irradiated fuel in the reactor vessel, reactor operation is permissible only during the succeeding seven days unless repairs are completed and the subsystem made fully operable and provided that during such time the remaining automatic pressure relief valves, FWCI subsystem, and gas turbine generator are operabl . If the requirements of 3.5.D cannot be met, an orderly reactor-shutdown shall be initiated and the coolant temperature shall be less than 330 F within 24 hour ;
.
~~SURVEILLANCE REQUIREMENT~~
.
4.5 CORE AND CONTAINMENT COOLING SYSTEMS Surveillance of the Automatic Pressure Relief Subsystem shall be performed as follows: During each operating cycle, the following shall be performed: A simulated automatic initiation of the system throughout its operating sequence, but excluding actual valve opening, and With the reactor at low pressure, each relief valve shall be manually opened until valve operability has been verified by torus water level instrumentation, or by an audiole discharge detected by an individual located outside the torus in the vicinity of each relief lin . When it is determined that one safety / relief valve of the automatic ; pressure relief subsystem is inoperable, the actuation logic of the i I remaining APR valves and FWCI subsystem shall be demonstrated to be operable immediately and daily thereafte Amendment No. 1 Millstone Unit 1 3/4 5-6 . .
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~~JAN 2 91987~~
~~. LIMITING CONDITION FOR OPERATION h~~
i 3.5 CORE AND CONTAINMENT COOLING SYSTEMS Isolation-Condenser System Whenever the reactor coolant temperature is greater than 330*F and irradiated fuel is in the reactor vessel, the isolation condenser shall be operable except as specified in 3.5.E.2, and the shell side water level shall be greater than 66 inche . From and after the time that the Isolation Condenser is made or found to be inoperable, for any reason, power operation shall be restricted to a maximum of 40% of full power, i.e. , (804 MWth) within 24 hours until such time the Isolation Condenser is returned to service provided that all active components of the core spray {
'
~~subsystems and LPCI subsystems are operabl . If the requirements of 3.5.E cannot be met, an orderly shutdown shall be initiated and the reactor coolant temperature shall be~~
~~+ less than 330*F within 24 hour , -~~
SURVEILLANCE REQUIREMENT 4.5 CORE AND CONTAINMENT COOLING SYSTEMS Surveillance of the Isolation Condenser System shall be performed as follows: Isolation Condensor System Testing: The shell side water level and temperature shall be checked dail , i Simulated automatic actuation and functional system testing shall be performed during each refueling outage or whenever major repairs are completed on the system, The system heat removal capability shall be determined once every five year l Calibrate vent line radiation monitors quarterly, Motor operated valves shall be tested per surveillance I requirement 4.1 l Amendment No. 1 Millstone Unit 1 3/4 5-7 I
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~~JAN 2 91987 LIMITING CONDITION FOR OPERATION h~~
(
3.5 CORE AND CONTAINMENT COOLING SYSTEMS Minimum Core and Containment Cooling System Availability Except as specified in 3.5.F.2, 3.5.F.3, 3.5.F.7 and 3.5.F.8 below, j both emergency power sources shall be operable whenever irradiated j
fuel is in the reacto . From and after the date that the diesel generator is made or foJnd to be inoperable, for any reason, continued reactor operation is permissible only during the succeeding seven days provided that the gas turbine generator, FWCI, Automatic Pressure Relief Subsystem, 3 i all components of the low pressure core cooling and the containment cooling subsystems shall be operabl . From and after the date that the gas turbine generator is made or found to be inoperable, for any reason, continued reactor operation is permissible only during the succeeding four days provided that the diesel generator, all components of the APR subsystem, all components of the low pressure core cooling and the containment cooling subsystems shall be operabl . If the rt;quirements of 3.5.F.1 cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the COLD SHUT 00WN or REFUEL CONDITION within 24 hour . Any combination of ir. operable components in the core and containment cooling systems shall not defeat the capability of the remaining operable components to fulfill the core aiid containment cooling function . Except as specified in 3.5.F.7, when irradiated fuel is in the vessel and the reactor is in the COLD SHUTOOWN CONDITION all low pressure l core and containment cooling subsystems may be inoperable provided that no work is being done which has the potential for draining the reactor vesse SURVEILLANCE REQUIREMENT 4.5 CORE AND CONTAINMENT COOLING SYSTEMS Surveillance of Core and Containment Cooling System The surveillance requirements for normal operation are in l Section 4.9.
1 k l Amendment No. 1 Millstone Unit 1 3/4 5-8 _ _ - _ _ - _ _ _ i
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JAN 2 91987 LIMITING CONDITION FOR OPERATION 3.5 CORE AND CONTAINMENT COOLING SYSTEMS 1 3.5. When irradiated fuel is in the reactor vessel and the reactor is in the REFUEL CONDITION a single control rod may be withdrawn and the drive mechanism replaced or fuel removal and replacement may be conducted provided that the following conditions are satisfie (a) The reactor vessel head is remove .
(b) The cavity is flooded.' (c) The spent fuel pool gates are remove l : (d) Water level is maintained within the limits of specification 3.1 l (e) Either (i) both core spray systems, (ii) both low pressure !
~~coolant injection systems, or (iii) one core spray system ,~~
'
and one low pressure coolant injection system, each combination being supplied by independent electrical. power, shall be operable or available for operation with the respective 4160 volt supply breaker (s) racked ou (f) With tne torus drained,(i) the ECCS configuration required in 3.5.F.7(e) shall be aligned with the condensate storage tank and the condensate storage tank suction valve V7-58 locked open, (ii) the condensate storage tank shall contain at least 414,000 gallons of usable wate shall contain at least 383,000 gallons of wate (g) The minimum electrical power source requirements shall be the same as specified in paragraph 3.7. (h) No work will be performed in the reactor vessel other than fuel sipping while a control rod drive housing is ope (i) Fuel removal and replacement will not be done without ' a full complement of control rod (j) During fuel movement no work being done which has the potential for draining the vesse , I k l l l k Amendment No. 1 Millstone Unit 1 3/4 5-9 _ _ _ . . ..
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~~<d OCT 311966~~
~~. LIMITING CONDITION FOR OPERATION .~~
3.9 AUXILIARY ELECTRICAL SYSTEM Applicability: Applies to the auxiliary electrical power syste Objective: To assure an adequate supply of electrical power during plent operatio Specification: The reactor shall not be made critical unless all of the following conditions are satisfied: One-345 kv line, associated switchgear, and auxiliary startup transformer capable of automatically supplying auxiliary powe . Both emergency power sources are operabl ! An additional source of power consisting of one of the following:
* The 27.6 kv line, associated switch gear, shutdown transformer to supply power to the emergency 4160 vo.it buse ( One 345 kv line fully operational and capable of carrying auxiliary power to the emergency buse . 4160 volt buses five and six are energized and the associated 480 volt buses are energize . All station and switchyard 24 and 125 volt batteries and associated battery chargers are operabl SURVEILLANCE REQUIREMENT 4.9 AUXILIARY ELECTRICAL SYSTEM Applicability:
~~Applies to the periodic testing requirements of the auxiliary electrical syste Objective: Verify the operability of the auxiliary electrical syste s Millstone Unit 1 3/4 9-1 _ _ - - - - - _~~
{
i -
JAN 2 91987 . SURVEILLANCE REQUIREMENT (Continued) i 4.9 AUXILIARY ELECTRICAL SYSTEM Specification: Emergency Power Sources , i Diesel Generator The diesel generator shall be started and loaded once a month i to demonstrate operational readiness. The test shall continue until the diesel engine and the generator are at equilibrium J temperature at full load outpu During this tes.t, tF.e diesel l starting air compressor will be checked for operation and its ability to recharge air receiver During each refueling outage, the conditions under which the diesel generator is required will be simulated and a tes conducted to demonstrate that it will start and be ready to I
'
accept load within 13 second During the monthly generator tert, the diesel fuel oil transfer pumps shall be operate I Gas Turbine Generator The gas turbine generator shall be fast started and the output breaker closed within 48 seconds once a month to demonstrate l operational readines The test shall continue until the gas turbine and generator are at equilibrium temperature at full { load output. Use of this unit to supply power to the systtm
]
electrical network shall constitute an acceptable demonstration of operabilit During each refueling outage, the conditions under which the gas turbine generator is required will be simulated and a test l conducted to verify that it will start and be able to accept emergency loads within 48 second I I Millstone Unit 1 3/4 9-2 _ _ _ _ _ - _ - -
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~~l'.~~
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~~JAN 2 91987~~
~~- LIMITING CONDI_ T ION _F03 0FERATION -~~
_ 3,9 AUXILIARY ELECTRICAL SYSTEM When the mode switch is in RUN, the availability of power shall be os ! spcified in 3.9. A, except as sptcf fied below: l From and after the date that power is available from only one 345 kv line, reactor operation is permissible only during the succeeding seven days unless :n additional 345 kv ',ine is sooner placed in servic . From and after the date thdt irsoming power is not available from any 345 kv line, reactor operation shall be Thepermitted provided hRC shall both be notified, ; emergency power sources are pge,rabl i within 24 hcurs, of the precautions to be taken during this The situation and the planc ice restoration of incoming powe minimum fuel supply fop the gas turbine during this situation shall be maintained above 20,000 gallon ;
~~' From and af ter the date that power cannot b6 made available frem the i RS3T, +,he p'lant shall be isolated from the grid within the next '~~
72 hours after which time reactor operation is permissible according to specification 3.9.8.2. During the 72 hour period, both emergency The minimum fuel supply for the power sources shall be operabl gas turbine during this situation shall be maintained above 20,000 gallons. If during the 72 hour period it is determined that the plant cannot be operated iso?ated from the grid, be in at least HOT STANDBY within the next six (6) hours and in COLD SHUTDOWN w the following thirty (30) hour . From and after the date that either emergency power source or its I l ascociated bus is made or found to be inoperable for any reason, F/ g' l reactor operation is pt:1tissible according to Specification .5.F unless such emergency power source and its bus are sooner mace operabh., provid2.d that during such time two offsite lines 045 l or 27.6 kv) are operabl l
{
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' Frem and after the date that one of the two 125 volt or 24 voit battery systems is made or found to be inoperable for any reason ieactor opeiation is permissible on?y during the succeeding seven days unlesc such battery system is sooner made operabl SURVEILLANCE REQUIREMENT 4.9 AUXILIARY ELECTRICAL SYSTEM B_atteries Station Batteriec Amendment No. 1 Millstone Unit 1 3/4 9-3
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~~I OCif 311986 i~~
~~. SURVEILLANCE REQUIREMENT (Continumd) -~~
4.9 AUXILIARY ELECTRICAL SYSTEM 4.9.B. Every week the specific gravity and voltage of the pilot cell and temperature of adjacent cells and oserall 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, end temperature of every fifth cel c The following tests will be performed in accordance with IEEE Standard 450-1975 "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Generating Stations and Substations."
(1) At least once every refuel outage, a battery service test l will be parformed in acccrdance with section 5.6 of IEEE Standard 450-1975 to verify that the rattery capacity is adeq'este to supply and maintain in operable status all of the actual emergency loads for 2 hour '
~~. (2) At least once every 60 months, during shutdown, a performance discharge test will be performed in accordance with Section 5.4 of IEEE Standard 450-1975 to verify that ! '~~
the battery capacity is 3t least 80 percent of the manufacturer's rating. Once per 60-month interval, this (' performance discharge test may be performed in lieu of the battery service tes . Switchysrd Catteries a, Every week the specific gravity and voltage of the pilot cell and temperature of adjacent cells and overall battery voltage shall be measured, 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 Millstone Unit 1 3/4 9-4 _ _ _ - _ _ _ _ _ - _ - _ _ - - . _ _ _ _ _ - -
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~~o K}T 3I !bbb~~
~~.,: LIMITIf;G' CONDITION FOR OPERATION .~~
3.9 AUXILIARY ELECTRICAL SYSTEM Diesel and Gas Turbine Fuel There shall be a minimum of 20,000 gallons of diesel fuel supply on site for the diesel and a minimum of 35,000 gallons on site for the gas turbine, except as permitted in Specification 3.9. SURVEILLANCE REQUIREMENT 4.9 AUXILIARY ELECTRICAL SYSTEM ~ _
.c . ' The quantity of gas turbine generator and diesel generator fuel shall be logged weekly'and af ter each operation of the uni Once a month a' sample _of the. diesel and gas turbine fuel shall be taken from the underground storage tanks and checked for qualit a
! l l l l i Millstone Unit 1 3/4 9-5 _ _ - _ _ _
- _ _ _ _ . - _ _ b k ' NORTHEAST UTILITIES .
~~,,,, u ,u m u, y,, .,.. a a we, cene,ai omce . seicen si,eet, se,i,n. connecticut 1 s ..s .m m a n=.a-a P o BOX 270 ..m..,.w.u,**"' ,,..,u..w * * '~~
HARTFOnD. CONNECTICUT 00141-0270 L J ,,,, _ a .. ". .". .- (203) 665-5000 September 24, 1987 MP-10C82 Re: NUREG-1021/ES-201/ para i U. S. Nuclear Reg.latory Commission Document Control Desk f Waohington, i i
;
Reference: Facility Operating License No. DPR-21 Docket No. 50-245 September 21, 1987 NRC License Examination Comments Gentlemen: { Attached is the compilation of comments on the written examinations administered to Millstonc Unit 1 Senior Lic6nse Candidates on September 21, 198 These comments were the result of a review of the examinations conducted by members of the Millst6ne Unit 1 training staff and Unit One Operationc personne Included are both the comments discussed during the exam review meeting of September 21, 1987, plus additional comments resulting from reviews conducted subsequent to this meetin Attendees at the September 21, 1987 meeting were: Reviewer Participants Lumb - NRC D. Reed - NUSCo G. Sturgeon - NUSCo D. Meekhoff - NUSCO R. Payton - NUSCO M. Jensen - NUSCo J. Geary - Instructor K. Walker - Instructor
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N. Berg - MP1 Assistant Operations Supervisor Observers R. Lueneburg - MP1 Supervisor Operator Training R. Palmieri - MP1 Operations Supervisor l _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - -
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~~Page 2 of 2 MP-10882 U. S. Nuclear Regulatory Commission~~
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The exam reviews were conducted considering the following: . Does the question elicit the correct response? L Is the Key answer Correct? L Is there potential for additional correct responses? Is the question appropriate? ! References are provided, where necessary, to substantiate the ; comment Please contact Ray Lueneburg, Supervisor, Operator Training, Millstone Unit 1 with any questions concerning our comment Yours Truly, NORTHEAST NUCLEAR ENERGY COMPANY
~~! j'Step hen th E. (n< Scace Station Superintendent Millstone Nuclear Power. Station (~~
SES/MCJ: sam Attachment: Senior Reactor Operator Exam Comments and applicable references e
_ _ _ _ _ _ . _ _ . _
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e t NRC EXAMINATION REVIEW COMMENTS - 9/22/87 5.01 FACILITY COMMENT: Reactor Theory Text defines PAH somewhat differently (heat additions overcome heat losses). Consider alternate wording i.e. a concise answer which expresses the intent of objective is 'PAH is expected to occur in IRM range because core heat from fission begins tc exceed ambient losses in this range." l 1 No Commen RECOMMENDATION: Accept other definitio Consider alternate wordin REFERENCES: Reactor Theory Text, page 7-10 (previously supplied). N/A
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5.02 FACILITY COMMENT: Decrease in heat transfer may also be explained using a reduced mess flow argumen High pressurc fluid will leak into low pressure resulting in reduced mass flow of high pressure fluid Also, both downstream, thus reduced hest transfe arguments could be combine None None RECOMMENDitTION : Consider above explanation as wel N/A N/A REFERENCES: HTFF Text pp 9-49 Eq. 9. 22 ( pr<n iously supplied ) .
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s . 5.03 FACILITY COMMENT: There 16 no objective for this information, nor is a candidate required to master this knowledge during training. This is because the void coefficient is operationally beyond the operatot's influence, and always remains dominantly negativ . Text does not stipulate how void coefficient varies at MP1 with core ag Fig. 4-10 is only an example of possible behavior. Text does not identify which factor always dominate l R8 COMMENDATION:
~~; Accept INCREASE or DECREASE with appropriate l !~~
justification of factor . Consider lack of training objectives and emphasis in grading, especially concerning discussion of factor REFERENCES: Reactor Theory Text page 4.22, 23, 24 (previously i supplied).
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5.04 FACILITY COMMENT: 1 Although the long term effects on both inventory and level are as described in the key, the short term (initial) behavior of level will be differen i Initially due to the rapid pressure decrease, voiding will result causing the annulus level to j increase due to increased dP inside the core ( Thus, actual level will increase ]
~~" swell" effect) '~~
for some time until inventory reduction is sufficient to cause decreas ., None 1 i RECOMMENDATION: I j Accept above description of level behavior if time i frame is specifie ) f REFERENCES: ~ The (attached) excerpt from NEDO 24862-B, Abnormal Event Analysis, General Electric Co., describes the swell effect during a stuck open relief valve even While this is not the idefitical situation in the question, the initial effect is the same, (For APR actuation, it would be even more pronounced.) This literature is used in STA training and is available as reference material for the operator (
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~~___ 5.05 FACILITY COMMENT: None~~
~~< There is no specific objective requiring this knowledge at MPl. The reason for the concavity of j '~~
line 'D' is not operationally significant, and is therefore not emphasized. The cavitation lines (E,F, and G) are not present on the allowed operating region map contained in Tech. Spec (attached). Operation below line 'G' is prevented by the low feedwater flow interlock, and therefore these lines are unnecessar The. text discussion is for background information onl i Same as abov \ RECOMMENDATION: Consider these facts in grading. Accept a reasonable explanation of the necessity to prevent , cavitation in those regions without specific names '- of each lin Same as b REFERENCES: T.S. Figure 3.3.1 (attached) i i
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5.06 FACILITY COMMENT: Text discussion of reasons for highet rodText worth at uses 500 F is slightly different than key' L, & L g to explain phenomenon, not " l e a k a g e . l None
RECOMMENDATION: 1 Consider alternate wording as in tex REFERENCES: Reactor Theory: page 5-13 (previously supplied).
, 5.07 FACILITY COMMENT: I l The key's answer was taken from an exam bank _ question ke The text answer is much briefe The exam bank answer was more detailed than required for examination purpose Another method used at MP1 to extend cycle length is to raise reactor level higher in the allowed operating band (i.e. 35"-36"). This increases core j flow due to increased pump head for the same spee This is not mentioned in the text, but is common { practic i I l l
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RECOMMENDATION: ! Consider text description of "coastdown". ! Accept " raising level" as method of extending cycle lengt REFERENCES: Reactor Theory Text, page 7-2 " This method is.used per management directive as allowed in the operating level band of 20"-40" per OP-20 I 5.08 FACILITY COMMENT: Answer Key includes MPR taking control as part of answer. The question does not ask for MPR response, only pressure respons The pressure ; behavior is the same initially regardless of whether the MPR is available or not (i.e. control valves will go closed). None None RECOMMENDATION: Accept answer without reference to MPR.
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REFERENCES: N/A i 5.09 FACILITY COMMENT: The form of the question could lead to some confusion by the candidate. The three terms given (MFLPD, MAPRAT, MFLCPR) are rot " thermal limits" as stated, but are operational ratios used to monitor the thermal limits.For This may affect the interpretation of the questio example, the " limiting condition for this ratio is in all cases as far as Tech. Specs. are concerned. The
~~" limiting parameter" sought in the key was really the thermal limit itsel A parameter might also be thought to be pin power (LHGR), node power (APLHGR), or bundle power (MCPR).~~
RECOMMENDATION: Consider candidates interpretation of question when comparing his answer to the key. We feel that a person knowledgeable in the thermal limits and their bases may not have responded as desired due to the form of the questio REFERENCES: Same as Ke ! I
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~~___ _-____ . . 5.10 FACILITY COMMENT: The answer in the Key for a). is incorrec ;~~
~~=Q +Q + Qc -Q CTPc ,3cm =~~
~~( h f w),- -- > b,e,c r e a s e,s,; - Qc ,1 ore there 6th "*** - Q, yis Q~~
~~Tbb HIGH, or actual power < calc. power None None I RECOMMENDATION: Change Answer Key to FALS , REFERENCES: Same as Answer Ke .01A NO COMMENTS~~
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6.01B NO COMMENTS f
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_- _ ___ --- . . 6.01C FACILITY COMMENT: Initiating shutdown cooling will have 2 effects, increasing cooldown rate, as well as increasing pressure 1
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in the jet pum These effects will have opposite impacts on the wide range Yarways. Additional cooldown will cause indicated W.R. Yarway level to decrease, but increasing pressure in the jet pump due to shutdown cooling flow will cause W.R. Yarway indicated level to increase, similar to recirc system flow. The question does not specify which effect to consider RECOMMENDATION: Accept either increase or decrease as a correct 4
,
response, since the question did not require an explanatio i l REFERENCES: MP1 Operator Training - Syrtems Volume 1, TX1300B, RVI System pages 7,'8, 15-16 i j RESOLUTION: 1 Unresolved .
1 I NO COMMENT j 6.02A-F
~~!1 6.03A-D NO COMMENT I~~
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~~, * s 6.04A FACILITY COMMENT: 1 Responses 1 and 4 are tru each SBLC pump is equipped with a local test pushbutton which, if depressed and ;~~
held, will run the associated SBLC pump regardless of SDLC initiating control switch position on CRP 90 Therefore, the pumps may be operated simultaneousl RECOMMENDATION: Accept either 1 or 4 as a correct response, since the question directed the student to " choose one" respons REFERENCES: MP1 Control Wiring Diagrams, Sh-665 and 666 MP1 Operator , Training Systems, volume 2, TX 1304, pages 15 & 1 l RESOLUTION: Unresolved
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6.043 NO COMMENTS l j d
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~~M f f. 6.05A FACILITY COMMENT: h / The Answer Key lists " Main FRV maintains flow at 105%" ., N~~
~~. . .~~
~~To limit inventory loss See and ensure attached sufficientMP1 excerpt'from flow j {~~
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~~is provided to the RFP' operator training volume 4, TX 1334. The FRV's shift to~~
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~~f. tow control to prevent RFP runout, notflow to limit to RP inventory; losses and provide sufficient~~
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Additionally, the question does not state whether If the the FWCI studentinitiation assumes is FWCI with orinvitational without a with LNP.LNP, he will also list condensate pump start,and condensate booster pump
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~~star < b RECOMMENDATION:~~
~~- Delete " Main FRV maintain flow at 105%" as a correct hI .,a '~~
~~response from.the Answer Key I ,A~~
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~~- Do not remove credit if the examinees lists cendc\ 3: 'l ntp t pump start and booster pump star ('~~
~~i:' ki\ " REFERENCES: ,~~
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~~MP1 Systems, Volume 4, TX 1334 pgs 17 & 18, a t ta chr6.II~~
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~~Y \ p[ Unresolved / u l ?~~
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~~< .6.0,58 FACILITY COMMENT: >+~~
~~See attached shee ' h~~
~~.,ts' .RECOf@SNDATION:~~
s
~~, Eliminate the questicn.~~
~~I r ,' r REFERENCES: ) ' MP1 Operator Training Systems, Feedwater System, pages 40-4d. Volume 3J,TX 1316 ,~~
~~/ " ' , \ , I ') e RESOLUT}Qh: ,~~
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i } ATTACMENT FOR 6.05B a s , The question is unclear in 2 areas: It does not specify which initiation signal has cleared , FWCI or Flow Contro sThe FWCI initiatin signal has no first order effect on the L 3
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FR If the flow control initiations signal (105% flow for RFP's pump) has not cleared, the FRV will shift to level control while the reset pushbuttons is depressed, but will shift back to flow control when the button is release , ,,
~~' If the flow' control initiation signal has cleared after s~~
transfer to flow control, the FRV will not shift to level control unless level >+37 or RFP discharge pressure <300' or l PS depressed - (flow control " seals in"). l l The question does not specify how flow control is reset: The FRV will reset from flow to level control regardless of feed flow if level is >+37" or RFP discharge pressure drops l ' i <300 ps FRV will reset from level to flow control manually whenever the pushbutton is depressed, but will shift back to flow control unless flow has dropped <105% for the running pump combinatio Clearing the signal will have no effect if the initiation signal of concern is FwCI, but will affect system operation if the initiation signal is flow control.
l , Reset will occur automatically regardless of both initiation signal Reset will occur in all cases manually, but will stay reset when the operator releases the pushbutton and if the flow control initiation signal is cleare i Because of the complexity of the system the general wording The l of the question will not ilicit the required respons j student is simply not provided sufficient informatio l
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i 6.08A FACILITY COMMENT: o A logic diagram which schematically represents the conditions which cause the LNP relays to actuate as described in MP1 Operator Systems, Volume 5, TX 1341, page 30, (attached) is attached, o Additionally, during the 1987 Outage, changes were implemented to 4160VAC bus 24F(14H) to accommodate
" Appendix R" requirement These changes necessitated changes to the LNP LOGIC, and 24F was renamed 14H. These changes have not been officially incorporated into the MP1 Operator Training Systems handout, but may show up as an examinees respons RECOMMENDATION:
Accept the Logic Diagram as an adequate substitute for the words written in the Answer Ke RCFERENCES: MP1 Operator Training, Systems Volume 5, TX 1341, Page 3 MP1 Operator Training, Appendix R Mods, LP 1332 page 20-25 (attached).
RESOLUTION: Unresolved
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_ _ _ _ _ _ _ _ _ _ _ . l l , 6.08B FACILITY COMMENT:
Question 6.08B requests 4160V Electrical System response ' to an LN The Answer Key in part states: " Load shedding from 4160VAC and 480VAC busses."
RECOMMENDATION: Delete the reference to 480 VAC busses since it was not l specified in the questio REFERENCES: l N/A RESOLUTION: Unresolved i 6.09A FACILITY COMMENT: l As worded, " Explain how and why the Recircs Respond . .
~~. . .", the question elicits a response from the examinee describing why the recircs have runback, not what stops them at 28% speed as indicated on the Answer Ke RECOMMENDATION:~~
~~Accept "because the master controller output has dropped to minimum" as a correct respons _ _ _ _ - _ _ _ _ _~~
~~, , , , , _ , - - . - - . - _ - - _ _ - _ - _ - _ - _ . o REFERENCES:~~
~~TX 1301B, pages MP1 Operator Training Systems Volume 1, 3& RESOLUTION:~~
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~~Unresolved !~~
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~~J 6.09B FACILITY COMMENT: Loss of full open indication will also result in a drive motor breaker tri Additionally, the statement,~~
" assume bypass valves are open" is confusing to MP1 examinees, since our bypass valves were removed and caps were welded in place years ag RECOMMENDATION: " RRMG In addition to the answer on the key, accept 'A' t '
~~Drive Motor Breaker will trip open" as a correct respons REFERENCES: l~~
~~- MPl. Control Wiring Diagrams, Sh 430 - MP1 Systems, Volume 1 TX 1301A, Reactor Recirculation System RESOLUTION: I Unresolved :~~
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n.: ;. ,cr~ a:: = = == = = ~ = - - --- 6.09C FACILITY COMMENT: o Error limiter does not function as described in the Answer Ke of error o Error limiter functions to limit size Pumps will signal thereby limiting acceleration. limiter until rate set by error speed up at a max it reaches the high speed sto o Pump accelerates to high speed stop because the speed controller has lost the speed feedback signal, resulting in maximum error signa RECOMMENDATION: Accept as a correct response: " Pump will speed up until the speed it reaches the high speed stop because signal."
controller has lost the speed feedback REFERENCES: TX 1301B, page 11 & 1 MP1 Operator Training, Volume 1, RESOLUTION: Unresolved
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6.10A FACILITY COMMENT: Question asks "What actions must occur or what conditions must be met prior to loop selection." Logic does not require both pumps to be off and the selected valves to be closed prior to proceeding, it merely sends a close signal to the selected valves and a trip signal to the running pum LOGIC never goes back to check that the valves have closed or the pumps have tripped prior to proceeding. The only MUST is reactor pressure must drop to 900 psi RECOMMENDATION: Modify the Answer Key to: Delete: RR cross tie or cross tie bypass valves are close Substitute: Selected RR cross tie valve and cross tie bypass valv REFERENCES: See attached excerpts from Systems Volume 1, TX 1335, page 25 & 2 RESOLUTION: Unresolved 1 I I l l
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, . . L-6.10B FACILITY COMMENT: Opening statement sets condition of 1 Recirc Pumprunning pump tripped Running, part 6.10A has the that no RR pump are running for 6.10 The Answer Key for 6.10B requires the following statements: "This causes RR pump to trip if it was running" and "The other pump then trips if it was running", but both pumps are already of RECOMMENDATION: Delete the above responses from the Answer Ke REFERENCES: N/A RESOLUTION: Unresolved 6.11 NO COMMENTS 6.11A&B NO COMMENTS i 7.01 NO COMMENTS
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7.02 NO COMMENTS ! l i 7.03 FACILITY COMMENT: ! The answer key should include as correct that the reactor is maintained 50 psig above the suppression Also, chamber pressure to ensure the SRV's remain ope the portion of the answer key "... ten minutes after a scram..." should not be required since the basis for , these conditions was not asked for in the questio l
RECOMMENDATION: Accept the above comments for credi REFERENCES: Tx 15008 - Emergency Operating Procedures - p l (previously supplied).
RESOLUTION: t Lead examiner agreed with both comments and requested that a comment sheet be made reflecting these item I l
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7.04 FACILITY COMMENT: The question asks for two (2) adverse consequences of operating the AOG system and with the Hydrogen Recombiner at an explanation of why each too low of a temperature occur The answer key requires three responses, one of which (increased system flow) would not be observable at Millstone 1 since we have to add air to the system normally to have sufficient system flo RECOMMENDATION: Accept any two consequences as correct: (2) (1) the insufficient potential recombination of hydrogen and oxygen, increased system flo l for an explosive mixture, and (3) i These consequences occur because the hydrogen and oxygen ' entering the Recombiner will not recombine at a low temperatur Note that neither the referenced text pages-nor procedure (OP 324C) make mention of the increase in system flow under the stated conditions of the questio ItEFERENCES: Text (Tx-1324C) Augmented Off Gas System pgs. 57 & 58 OP 324C - page 5 (previously supplied). 1 d i4 l l
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7.05 FACILITY COMMENT: ( There are additional purposes for placing the mode switch in shutdown following a scra RECOMMENDATION: Accept the following additional reasons for placing the mode switch in shutdown following a scram:
-1- Placing the mode switch to shutdown allows bypassing the scram discharge volume high level scram and scram air header low pressura scram such that the scram can be rese Placing the mode switch to shutdown removes the APRM flow biased scram and rod block setpoints and inserts lower scram (15%) and rod block (12%)
setpoint REFERENCES: RPS Text (Tx-1408) pgs. 33, 39, 57 ONP Text (Tx-lbOOA) pg. 11 (previously supplied) ; i
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~~i 7.06A FACILITY COMMENT: i The answer on the answer key is absolutely correct for , this question, however the operational reason for~~
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waitirig until -220" to blowdown should also be considere That is, waiting until water level reaches-220", rather than commencing the blowdown at a higher level buys time to re-establish some injection flow to the reactor and at the same time not allowing fuel clad temperature to reach 2200' .06B Please consider the following answer in addition to the answer key for this part of the questio What is significant about 700 psig? Above 700 psig one open SRV will provide sufficient steam flow to assure adequate core coolin Why is depressurization required below 700 psig? Below 700 psig the flow through one SRV is not sufficient to assure adequate core cooling. Therefore depressurization through the four APR valves is required to provide sufficient steam flow to assure adequate core coolin RECOMMENDATION: Consider. accepting the time aspect explained above as credit for this question.
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7.07 NO COMMENTS 7.08 NO COMMENTS 7.09 FACILITY COMMENT: 7.09 The question asks how thermal stratification can occur if level is below 50" and no forced cooling is available. The answer key refers to the fact that monitored temperatures can be lower than the core aren temperature This is true, but does not answer the questio Thermal stratification occurs under the above conditions because the core is still generating heat to the coolant which will result in thermal layers being formed due to the density difference between the layer RECOMMENDATION: Accept this description of thermal stratification for the answer key since the question actually asks how thermal stratification can occur under these condition REFERENCES: Millstone Text (Tx-1200 - General Operating Procedures) page 69, Objective 64 (previously supplied).
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~~e 7.10 FACILITY COMMENT: 7.10D Other answers should be considered correct for this questio RECOMMENDATION: Accept the following in addition to those listed in the answer key:~~
-
~~Torus level oscillations~~
~~- Reactor power change - Control valve position change - Change in steam flow indicators -~~
~~Steam flow / feed flow mismatc RESOLUTION: Agreed to by examine Submit comment sheet to this effec i i~~
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__. .- , I . . 7.11 FACILITY COMMENT: This condition would not necessarily warrant entry into ONP 512 because Suchitan could have event hasresulted recentlyfrom a local transien occurred at Millstone ONP's 504 and 502 should be entered but the EOP's would not be entered unless the candidate assumed This that the scram caused an entry conditio question does not imply that a failure to scram has occurred as mentioned in the post exam revie Also requires a reactor scram and entry into ONP 50 RECOMMENDATION: Only ONP 502 needed for full credi No reference need to be made to the EOP's unless an explanation follows as to why they were entere Accept ONP 502 and 515A for full credi GENERAL COMMENT: If an assumption was made that caused the candidate to enter the EOP's, he may mention EOP 569 as one of the procedures used since this is theoretically the controlling EO !
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I 8.03D FACILITY COMMENT: Answer key requires two answer for full credi . Completion of HP form 490 . H.P. Supervisors Approval H.P. Supervisors signature as on HP Form 490 , therefore the form is not complete without his signatur l RECOMMENDATION: Full credit should be given to candidate if they simply state " completion of HP Form 4902.5" or " Completion of HP Form 4902.5 and Supervisors Approval".
REFERENCES: HP Form 4902.5 (attached) RESOLUTION: NRC examiner will review commen i
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~~l 8.02 FACILITY COMMENT: l Question asks for "rajot" concerns of the SS/SCO when implementing a work order in accordance with ACP-QA-2.02C. ACP-QA-2.02C lists eightNo~~
~~" items" the reference is made SS/SCO must concern himself wit to major or minor concern RECOMMENDATION:~~
~~Answer key should be modified to accept any of the 8~~
~~" items of concern" as defined by ACP-QA-2.02C provides they do not duplicate one anothe REFERENCES:~~
l ACP-QA-2.02C pg, 10 & 11
8.03 FACILITY COMMENT: Question is worded in such a way that the candidate would not specify the Action statements that were given in the Answer Ke Candidates would go directly to Tech. Specs. Section 3.2 task 3. RECOMMENDATION: Since the candidate was told that the failure of the undervoltage relay results in a Total loss of LNP relay logic, the correct answer should be per Table 3.2.2 of Tech. Spec. section 3.2 " Initiate orderly shutdown to cold shutdown".
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, REFERENCES: Tech. Specs. Section 3.2 table 3. (previously supplied) RESOLUTION: Examiner will review reference and modify hnswer Key accordingl G.04A NO COMMENTS 8.040 FACILITY COMMENT: I Another viable answer would be during routine rounds performed by the Plant Equipment Operator (PEG), the PEO is directed to maintain various parameters in spec per the PEO rounds sheet. This would involve valve manipulations for routine operations such as filling surge tank RECOMMENDATIONS: Also accept " Maintaining varicus plant equipment parameters in spec; in accordance with Operations Department Instruction (0.09 (PEO Rounds).
REFERENCES: Operations Department Instruction 10.0 _ - _ - _ - _ _
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~~' RESOLUTION:~~
None 8.04C NO COMMENTS 8.05 NO COMMENTS 8.06 A FACILITY COMMENT: PhBT A&B: The question doesn't specify "per ACP 3.02 or per Tech. Specs. Section 6".
PART B: Since all the procedures listed in past are unit specific questions, SORC and Station Superintendent would not be required for final approval., RECOMMENDATIONS: Accept either correct answet from ACP-3.02 or Tech. Spec Section PART B: Do not take credit off for not putting SORC/ Station Superintendent for Final Approva I
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f: i , REFERENCES: ACP QA-3.02 SECTION 4.7, 4.8, 6.9.1.3, 6.9. Technical Specifications Section 6, page 6-16 (previously supplied).
8.07 FACILITY COMMENT: 1
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~~1) Question only asks.for actions to be taken based on information give The Answer Key requires the~~
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i candidate to write out what is contained in the applicable tables from Tech Spec i 2) For the Rod Block function there are no " Actions" to take because the Rod Block has already been inserted automatically due to the failure of APRM channel Therefore, the student shouldn't have to specify the action statement of table 3. since the action has been take RECOMMENDATION: Correct answer should be per Tech. Spec 3.1 TABLE 3.1.1for Note 1, Trip the associated trip system (1/2 scram) Red Block Function - No actions since Rod Block is already inserte J REFERENCES: Tech. Spec Sections 3.1, 3.2 Table 3.1.1, 3. (previously supplied).

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. . . ) RESOLUTION: l Will change Answer Key, not requiring candidate to l specify number of instruments require ! Rod Block Function - NO RESOLUTION 8,08A NO COMMENTS i 8 . 0 815 NO COMMENTS I 8.08C NO COMMEN .09A NO COMMENTS 8.09B NO COMMENTS 8.10A NO COMMENTS
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8.10B 'NO COMMENTS 1
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0.10C NO COMMENTS 8.10D NO COMMENTS f l
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5, i l ATTACHMENT 3 NRC RESPONSE TO FACILITY COMMENTS The following represents the NRC resolution to the facility comments (listed in Attachment 2) made as a result of the current examination review policy . Only those comments resulting in significant changes to the master answer key, or those that 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 examination review are not listed (i.e.:
typographical errors, relative acceptable terms, minor set point changes).
~ Question 5.01a: Alternate wording is acceptabl Question 5.016: Alternate wording is acceptabl Question 5.02a: Alternate wording is acceptabl Question 5.0 Comments accepte Question _5.04a: Comment accepte i Question 5.05b: Redistributed partial credit point value ' Question 5.05c: Comment accepte Question 5.06a: Alternate wording is acceptabl Question 5.07a: Alternate wording is acceptabl . Question 5.07b: Additional answer is acceptable with reasonable description of method and drawback Question 5.00: Comment accepte Questiqn 5.09: Comment accepte Question 5.10a: Comment accepted < The equation supplied by the facility is not the same as the referenced equatio Question 6.01c: Comment accepte Deleted part c of questio Question 6.04a: Comment accepte ;
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i I f-ATTACHMENT 3 -2-
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l i L Question 6.05a: Comment accepte { l Question 6.05b: Comment accepte Deleted part b of questio I Question 6.08a: 1.ogic diagram is an acceptable answe Question 6.08b: Comment accepte Question 6.09a: Alternate wording is acceptable, , l Question 6.09b: Comment partially accepted. Accepted "A" RRMG drive motor breaker trip as the only correct answer. The reference material supplied did not indicate that the bypass valves had been remove Question 6.09c: Alternate wording is acceptable. Reference material did not give a clear explanation of the error limite Question 6.10a: Comment partially accepte Deleted to trip pumps and close valves as part of required answer and redistributed point Question 6.10b: Comment accepted. Redistributed partial credi . Question 7.03: Comment accepte Question 7.04: Comment accepte Question 7.05: Additional answers are acceptabl Question 7.06a: Partial credit given for time aspec Question 7.06b: Comment not accepte Reference material supplied does not support alternate answe Question 7.09b: Alternate wording is acceptabl Question 7.10b: Additional answers are acceptabl Question 7.11: Comments accepted. A typographical error in the lesson plan for the Condensate System caused the change to part Question 8.01d: Comment accepte Question 8.02: Comment accepte Question 8.03: Alternate answer is acceptable.
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ATTACHMENT 3 -3-i i Question ~8.04b:. Additional answer is acceptable, i Question 8.06: . Comment accepte Question 8.07: Comme'nt partially accepted. All applicable Tech Spec sections must be referenced for a complete answe l
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_ . .. I 1.- F L e ATTACHMENT 4 SIMULATION FACILITY FIDELITY REPORT Facility Licensee: Northeast Nuclear Energy Company Facility Licensee Docket No.: 50-245 Facility Licensee No.: DPR-21 Operating Tests administered at: Millstone Unit 1 Operating Tests Given On: September 22 - 24, 1987
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~~During the conduct-of the simulator portion of the operating tests administered September 22 - 24, 1987, the following apparent performance and/or human factors discrepancies were observed:~~
- One control rod inadvertently scrammed when turnover for a scenario was in progress. -The problem was immediately corrected and only caused a short dela Standard forms and miscellaneous equipment such as stop watches were not available for the:candidatcs' use. This detracted from the plant fidelity .of the simulato Only one phone line to the instructor's booth can be used at a tim This also detracts from the candidates' concept of the simulator as the real plan .
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IR 05000245/1987023

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