ML20207T409

Forwards Comments on Written Exam Administered to 30 License Candidates on 861021.Exam Questions Clear,Concise & Appropriate

ML20207T409
Person / Time
Site:	Vogtle
Issue date:	10/24/1986
From:	Bockhold G GEORGIA POWER CO.
To:	Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
Shared Package
ML20207T381	List: ML20207T380 ML20207T403 ML20207T409
References
NOT-00923, NOT-923, NUDOCS 8703230542
Download: ML20207T409 (100)

Text

{{#Wiki_filter:- - - - _. _ - _ _ - _ -. Georgta Power Company. Nuclear Operations - ENCLOSURE 3 ~5. Post Office Box 1600

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y Waynesboro. Georgia 30830 Telephone 404 724-8114 404 554-9961 GeorgiaPower Vogtle Project the SOU'%m E.4KinC SV:i!Om October 24, 1986 Plant Vogtle - Units 1 & 2 NRC Written Examination File: X7BP03 Log: NOT-00923 Mr. John F. Munro, Acting Chief Regional Operator Licensing Section U. S. Nuclear Regulatory Commission 101 Marietta Street, Suite 2900 Atlanta, Georgia 30303

Dear Mr. Munro:

This letter provides our comments regarding the written examination which was administered to 30 license candidates on October 21, 1986 at Vogtle Electric Generating Plant. The enclosure contains specific comments and recommendations based on a technical review of the questions and answers. The examination questions were all clear, concise and appropriate. Comments are forwarded to provide the examiners with additional information that could effect answers to questions. Sincerely, te George Bockhold General Manager KR!/PDR/msh Enclosure cc: Mr. Anthony Vinola P. O. Box 1625 1520 Sawtelle Drive Idaho Falls, Idaho 83415 Mr. J. J. Badgett NO-NORMS 8703230542 870223 PDR ADOCK 0500 4 V J

.oe. ,-?. QUESTION: 1.02a and 5.13a ' COMMENTS: The E0P training text has additional reasons for maintaining pressurizer level. RECOMMENDATION: Accept as an additional answer: ~Provides sufficient inventory to maintain pressurizer pressure control.

REFERENCES:

E0P Training Text, Pg 4-17

e. Make continual efforts to do whatever is necessary to restart o a reactor coolant pump. The level of safety of the plant - increases greatly when forced circulation flow is established.- The RCP that provides the best pressurizer spray should be the first started if possible: RCP No. 4 then RCP-No. I as a second choice. 2 E o Maintain steam generator levels. Feedwater additions will have a momentary transient effect. However, natural circulation should re-establish itself when the feedwater addition stops. No degradation of heat transfer occurs so long as three steam generators maintain a wide range level greater than 40%. Massive additions of feedwater are unnecessary. Operate steam dumps or steam generator atmospheric relief o valves with slow, deliberate actions. In order to maintain natural circulation, steam must be dumped. Steam generator safety valves will adequately remove heat, but should only be used as a last resort. Maintain at least a 50*F subcooling margin if the Con.tro'l Rod o Drive Mechanism fans are in-service (100*F if not in-service) by raising pressurizer pressure or dumping more steam. [ o Maintain pressurizer level greater than the heater cut-off i setpoint (17%) at all times. This allows pressurizer pressure control, which is important to controlling natural circulation. The procedure allows a minimum of 4% pzr level-and 28'F subcooling margin before SI must be reinitiated but these should not be used as operational limits. j 1 l r-4-17 _,3 . ' % + -,1 A f_. < A* W

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g., 6i-6 -QUESTION: 1.10a and-5.03a COMMENTS: Accept alternate definitions for suberitical multiplication other than mathematical formula RECOMMENDATIONS: Accept the following as alternate definitions: Neutron level remains at a constant value, greater than the source strength, due to having source neutrons in the vicinity of fuel, even though Keff is less than unity. Neutron level increases due to adding positive reactivity and having source. neutron in the vicinity of fuel, even-though Keff is maintained at a value of less than unity.

REFERENCES:

Vogtle Text Chapter 21,- Page 40 F e.

i \\ REACTOR PHYSICS REVIEW The two major aspects of suberitical multiplication discussed thus far are used in the following definitions of suberitical multipheation: Suberitical Multiplication E that process by which: Neutron level remains at a constant value, greater than the source strength, due to having source neutrons in the vicinity of fuel, even though K,gg s less than unity. i Neutron level increases due to adding positive reactivity and having source neutrons in the vicinity of fuel, even though K,fr is maintained at a value of less than unity. l 1. Count Rate Versus Neutron Level The suberitical multiplication processes discussed thus far have dealt with the actual neutron level in the core. The area of interest, however, is the indicated neutron level or count rate. Neutron leakage is proportional to the neutron level in the core, and the number of neutrons detected by the ex-core nuclear detectors is proportional to neutron leakage: therefore, the indicated count rate is directly related to the in-core neutron level by the detectors' efficiency, # ndet. The equilibrium count rate is represented by: CR = Ng,x # ndet on 1 CR = S (1-K) ndet cps where ndet = n/sec 2. Count Rate Comparison The most common use of suberitical multiplication is the comparison of the count rates prior to and following changes in K,gg. This comparing is accomplished by using a count rate ratio: 21-40 Origirni 4 m % + wM me mM k a een-m'a L eMaM.s w w +awa-Am. w. uw - -Rn -~ -'"-mAra-a ^ a= 1 1 -as 1a

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2.01 COMMENTS: For each room noun name given in the answer there also exists an associated. room number to identify the same room (s). These room numbers.are shown on P&ID listed in the references. ' RECOMMENDATION: Accept the following answers as correct: 1. Low pressure' in letdown'line. l 2. High temperature in l'etdown HX room (R-A07) 3. High temperature in CVCS valve. gallery (R-A08) 4. High ' temperature in auxiliary building piping penetration room (R-A09)

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k 1 .e QUESTION:- 2.03 and 6.01 f COMMENTS: 'VEGP has a safety grade cold shutdown design. Going to the cold shutdown condition requires use of the reactor vessel head vent for two additional reasons. RECOMMENDATION: Accept as alternate answers: a. A safety grade letdown path or b. Letdown path to provide for head cooling

REFERENCES:

Westinghouse letter of July 22, 1986 LO-LP-09202, Page 13 2 1 i - i t t l r l I i i i e

LO-LP-09202-00 'lil.

• LESSON OUTLINE:

NOTES C. Safety Gride Letdown: Reactor,VesselHea{lVent

11. "' Purpose-Objective 8

fl\\1 (Providesasafetygradef}Alsoallowsletdown fne'forvent-a. ingl vessel head to PRT.- to exceds letdown path, ~ b.$hentnoncondensiblegases 2. Head Vent Letdown Flowpath a. Exits RCS through on inch vent line b. Moves through either of two sets of valves in parallel 1) 2 series valves per set 2) SOV's: HV-8095A and 8096A or HV-8095B and -8096B c. Pathe join and mqve through temperature and flow elements d. Path divides to either PRT or excess letdown e. PRT Path 1) HV-442A and B with Hand Controllers on QMCB (and PSD?) 2) SOV in parallel 3) Bypass line with sight glass 4) Enters PRT with line from pressurizer f. Encess Letdown Path 1) Through MOV-8098 2) Enters excess LTDN HX and follows excess Ictdown flow path 3. Components, Instrumentation and Control a. Safety Grade Vessel Head Vent Isolation' Valves l 1) HV-8095A and -8096A in parallel with HV-8095B and 80968 13 l L

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..n ElectricCorporation ' Divisions 8o: 355 T ~'.;y a y PittsburghPennsylvania15230 ,.i of $;., 2,#l, GP-11251 ^..: ..,\\', & ' .-i./, July 22, 1986 ng;, yg; ....s'" ' ';N Mr. F. B. Marsh S. O. GAE-280 id".ProjectEngineeringManager FSD/SS-GAE-4354 ' Bechtel Power Corporation Plant Vogtle ? ' P. O. Box 282 "#[Waynesboro, Georgia 30830 'H ' e " ' V0GTLE ELECTRIC GENERATING PLANT Hr. -rl UNITS 1 AND 2 SYSTEMS AND OPERATIONS FOR jf's, SAFETY GRADE COLD SHUTDOWN

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Dear Mr. Marsh:

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c. Attached for your information and use is a technical description for the y),.%. Safety Grade Cold Shutdown Design (SGCSD) at plant Vogtle.

s. Please note that l sg.(this description is not a detailed operating procedure; however it provides Nfinformation regarding the operational considerations that are necessary to .f.,g;;. achieve a safety grade cold shutdown. P- ~. 'lg@ThiscompletestheWestinghouseactiononOpenItemNumber2131N. If there are any questions, please call. 3

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+" GENERAL DESIGN BASIS IThe primary objective of the Safety Grade Cold Shutdown Design (SGCSD) is ,')toenableNSSStobetakentocoldshutdownconditions(I)within36 hours " following certain postulated events. These events encompass any Condition II, III or IV event including a loss of offsite power after a reactor trip. The SGCSD systems are designed to a single failure criteria, utilizing only safety grade systems. In the event of control room evacuation SGCSD can be achieved by using an emergency shutdown panel, located outside the control room. For design purposes, all SGCSD components should be designed under the emergency condition guidelines for transient stress analy, sis, by assuming a single occurrence over plant life. To achieve a safety grade cold shutdown, three critical functions must be performed. (1) residual and sensible heat removal (2) boration (3) depressurization These functions are performed during three basic phases of plant recovery. ,7 Phase 1 - Hot Standby ) g ' During Phase 1 the plant is placed (and maintained) in a hot standby (2) condition. For design purposes, the SGCSD enables the operator to maintain the plant in hot standby for about four hours. During this time at hot standby, residual heat is removed by utilization of the pc auxiliary feedwater system and the steam generators; natural circulation of the reactor coolant is relied upon to transfer heat from the core to the steam generators. For purposes of establishing the requirements for auxiliary feedwater capacity, it is assumed that the RCS is borated to cold shutdown boron concentration prior to cooling of the RCS. g, _... ( M e. g a # a. _-. c ~.... - w L="__,. %.,,.3A. g m, t ,c. , +.ww.. mew - ~.~ - -- -w A

O e Normally, the RCS is borated using the Chemical and Volume Control System (CVCS)toinjectconcentratedboricacidwhilelettingdownreactorcoolant f to the boron recycle system. Given the redundancy of the CVCS in terms of multiple pumps and charging and letdown paths it is likely that it would remain available, even following an initiating event and single failure. However, in the event that all control grade equipment is unavailable, a safety grade means of boration is required. The safety grade means of injecting boron into the RCS is provided by using the boric acid transfer pumps and the centrifugal charging pumps to inject 4 wt.% boric acid from the boric acid tank. Also, charging pump flow control. capability is provided along with high energy line break l protection for the boration system; reactor coolant pump seal injection is l also maintained. The safety grade means of letdown is provided by a ( l letdown line connection from the reactor vessel head vent line. This letdown line connection provides a path through which the RCS fluid can be passed to the pressurizer relief tank and provides benef_its in. vessel head cool g and reduction of thermal stresses during a cooldown without offsite power. (1) Cold Shutdown Conditions reactor subcritical RCS temperature 1 200*F RCS Pressure 1 ppendix G A Requirements s (2) Hot Standby Conditions reactor subcritical 3 RCS temperature 1 350*F RCS pressure 1 ppendix G A Requirements m = l) c, .l.} E r .:.? u x / u,.4%,%.... _.-,;.L.W.'_re M=QMNT a n

i o o . Phase 2 i e, x Phase 2 occurs following boration of the RCS to cold shutdown c'oncentrations; the RCS is cooled at an approximate rate of 35'F/hr by , increasing the steam dump from the steam generator power operated relief valves. In addition, the charging pumps are used to weliver refueling water to makeup for primary contraction due to cooling. Makeup is also provided for the RCS inventory discharged when the reactor vessel head letdown path is periodically opened to provided head cooling. Upon approaching the end of this phase of cooldown, the RCS is depressurized by venting the pressurizer through use of the safety-grade pressurizer power-operated relief valves. To ensure that the accumulators do not repressurize the RCS, the accumulator discharge valves are closed prior to the RCS pressure dropping below the accumulator discharge pressure. As an additional design feature, each of the accumulator nitrogen supply lines is equipped with a parallel arrangement of two Class IE solenoid operated valves. These valves provide a means through which the accumulators can be vented should the accumulator discharge valves fail to close. l Phase 3 'The third phase of SGCSD occurs when the RHRS cut-in conditions have been reached.(3). Residualheintfromthecoreisremovedbypassingthereactor DE coolant through the RHR heat exchanger and discharging the fluid back into the RCS. The RHRS suction isolation valves and their associated electric power supplies are designed to allow the initiation of one of the RHRS subsystems while retaining the capability of isolating the RHRS from the RCS even with the most limiting single failure. Separate Class 1E power 23 supplies are provided to each of the four RHRS suction isolation valves. Also, four protection grade RCS narrow range pressure transmitters are provided to maintain interlocking with the four independently powered RHR isolation valves. (3) Residual Heat Removal System Operating Conditions: (1) RCS pressure 400 psig (2) RCS temperature < 350*F p 7 ,E:- y, k: ... 2.; L l: M % p r w W q, ~ .n. ~ :, n.

. =. QUESTION: 2.17b and 6.17b 4 COMMENTS: This part of the question asked how the setpoint was determined. Only the wide range loop temperature (auctioneered 4 low) is used to establish the setpoint. Wide range pressure is compared to the setpoint to trip bistable. 1 RECOMMENDATION: Accept the following as the correct answer: Wide range loop temperature (auctioneered low). i.

REFERENCES:

LO-LP-16f01, Page 5 and 6 4 l i } 1 4 i I i. I, I i l I i } E ,o.,e -,..m y.-. ,7 ,y _m.,_- c_, _,,..

1 LO-LP-16501-00 Ill. LESSON OUTLINE: NOTES b. Generates RCS pressure setpoint at low temperatures. c. Notifies operator to arm circuit. d. Automatically open PORV if necessary to pre-vent exceeding pressure limits (if armed). 3. Input Parameters Objective 4 a. Wide Range RCS Pressure LO-TP-16501-00-002 1) PT-403, RCS loop 1 2) PT-405, RCS loop 4 3) Inputs to a) Pressure Indicators, QMCB-C Range 0-3000 psig (PI-403, - 405) b) Pressure Recorder, QMCB-C (PR-403) c) Pressure Indicators on Remote Shutdown Panels d) Computers e) COMS b. Wide Range RCS Temperature LO-TP-16501-00-003 1) Hot and cold legs for each loop 2) T - Protection set I (TR A) hot 3) T - Protection set II (TR B) cold 4) Auctioneered Low signal inputs to COMS 4. COMS Circuit a.- Auctioneered low temp input LO-TP-16501-00-004 1) TR A uses Thot 2) TR B used T Id 3) Used to develop pressure setpoint l 4) Used to actuate " arming required" Objective 6 LO TEMP alarm l 5 l ., -.-..-.... w...u a : - - .u a

v LO-LP-16501-00 Ill. LESSON OUTLINE: NOTES b. Wide Range Pressure Input 1) TR A uses loop 4 2) TR B Uses loop 1 3) Compared to press setpoint (developed from auct. temp). a) Trips bistable to send open signal Objective 5 to associated PORV. b) Cenerates HI PRESS DEV alarm if setpoint exceeded. 5. Arming Circuit a. Opening of PORV requires that operator LO-TP-16501-00-005 manually ARM circuit, b. ARM / BLOCK handswitches. Objective 6 1) QMCB-C 2) Locally at shutdown panel. 3) Control station selected by LOCAL / REMOTE transfer switch, Arming signal f,eeds into PORV and PORV Block c. Valve logic. 1) PORV LO-TP-16501-00-006 a) Allows Coms actuation signal to be West. Logic, sheet passed on to PORV. 18, 19 of 20 l b) Opens PORV if in AUTO and in REMOTE control. c. Review PORV logic for PORV not in t AUTO, REMOTE. 2) Block Valve a) Allow BLOCK valve to open. l l b) Must be in AUTO and REMOTE control. c) Review Block Valve logic for con-ditions where block valve is not in AUTO, remote. 6 p 4 e g m.m t. . - + - - 1 -asm. sm-'W+,a h - 4 b 's N- .*-.--Jhgm6f h I d4 %M-l- 4m E

QUESTION: 2.20 COMMENTS: There are two MSIVs-in series on each SG steam outlet line. If one valve fails to shut the other MSIV will still provide isolation of the affected SG. RECOMMENDATION: Accept the following as an alternate answer: The second MSIV in that steamline will shut and isolate the affected steam generator.

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'O' O o. QUESTION: 2.24 t. COMMENTS: The VECP technical specifications provide the latest information on coincidences and setpoints concerning the SG hi-hi level protective circuit, r RECOMMENDATION: Accept tihe setpoint and' coincidence for hi-hi SG 1evel as 78%, 2 out of 4. i

REFERENCES:

VEGP Technical Specifications. Tables 3.3-3 and 3.3-4 i I l i i l i l 4 1 .I ? j ? 4 j 4 1 .,..-.,_-.-,.~,m.,_. ,,_.___..__,.___-.,_,---m,,_,m-

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• ofS 2.

LI-CSS 3 LT-o6S 4 h ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRUNENTATION TRIP SETPOINTS { r-SENSOR TOTAL ERROR E FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE I -a H 5. Turbine Trip and Feedwater Isolation a. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A. .1 Actuation Relays ( 61 J J 79 79 7 b. Steam Generator Water E5:6-] 2.18J' 1.5T <Es2A 3x of <E84 41% of narrow .j . Level--High-High (P-14) narrow range range 1,nstrument

c. ' Safe /y.T"/ecNon See hocfias,a/ t/ nit.t above-

'"'LT****L

• F**-

) ,p span. po, a y Sg, //o,,Aley,,,;,,, y*, lues. 3 ) sety,ints aa,,L wg 6. Auxiliary Feedwater y u__..i. _v a._ a_ _.._. -.. .u...a.. .u.. a - a..-b, Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A. ) and Actuation Relays j 17.0 1.1. /f ) 17 /S.3 l b..c. Steam Generator Water E39re] [27.10] 1.5J 1 E32n.]E of 1 E3Gr4]E of narrow t ,J.evel--Low-Low narrow range range instrument 'il Star / Notor-instrument span. Dr/ver, Pkmps span. 4. 'f i r;.l'_:;: "C" ' D"E E" '[S"E E" ~ "^ "^ Ex: : 1t:::. Ex: ::!t:;.

2) Sfart G rbine -

1 *1. 0 /Z./ W l.$ 2 /7'lo o/narra.! .> /S 3 'la of Driver, Pu.w.ps ra.nge instnurent,w r,w ra ny e sydn. in sf rameos/ span. 7 ? i I I l V J

L1 - QUESTION:. 3.10 COMMENTS: The latest setpoints and coincidences for answer 2 and 6 are provided in LO-LP-27102-00 (Rod Control System) RECOMMENDATION: Accept the following as correct: 2. C-2: 1/4 PR's greater than or equal to 103%, both 6. C-11: Control bank D at 223 steps, auto

REFERENCES:

LO-LP-27102-00, Pages 10 and 11 F l l n

5 ,/ LO-LP-27102-00 7 111'. ' LESSON OUTLINE: NOTES c) The generator has a reverse current condition d) An output phase develops a ground e) The generator output breaker trips . on overcurrent Also alarms for RIL B. Other Functions of Rod Control System computer discussed later 1. Rod Stops LO-TP-27102-00-011 V.T Fig 7.2-1-1 a. Rod stops prevent rod motion before a core (SH. 9 Of 20) limit is exceeded or when rod motion is not Objective 4 possible or not desirable b. Automatic / Manual Rod Stops 1) C-1 a) 1/2 IR is greater than 20% current equivalent to reactor power ", If g,eA td NIS N M g' b) Manually blocked from main control ' g te,f u, g board at greater than 10% 2) C-2 a) 1/4 PR is greater than 103% b) May be bypassed at NIS racks 3) C-3 l l a) 2/4 OT Delta T - 3% below trip ~ setpoint f b) Also initiat'es a turbine runback 4) C-4 l a) 2/4 OP Delta T - 3% below trip setpoint b) Also initiates turbine runback l 4 I c. Automatic Only Rod Stops = 1 1) C-5 i a) Less than 15% turbine power Turbine Impulse l-press PT-505 [l t 10 a

LO-LP-27102-00 fl. LESSON OUTLINE: NOTES 2) C-11 Manual control will allow CBD to be [ fem M cJmverv'r withdrawn to 228 a) CBD at 223 steps steps 2. Data logging circuits LO-TP-27102-00-012 a. Bank data logging circuit Objective 9 Two separate circuits 1) Receives signal from bank selector, Part of Logic cabinet bank overlap unit, and slave cycler concerning beginning or ending of step 2) Information relayed to: a) Computer for rod deviation monitoring need s.p. b) P to A converters in DC hold supply cabinet for use in RIL. b. Group data logging circuit (MCB demand counters) 1) Receives signals from selector switch, bank overlap unit; slave cycler at end of each step 2) Provides demanded position derives step counter c. DC Hold Supply and P/A Converter Objective 8 1) DC Hold Supply provides +120 VDC Maintenance of cir-and +70 VDC hold power to power cuits cabinets (a) +120 VDC to close, +70 VDC to hold 2) P/A converter changes bank demand pulser to analog signals to be used in the process I & C system (a) RIL computer and alarm (b) Bank D withdrawal. limit and alarm (c) Multi-pen recorder. ~ d. Rod Insertion Limits (RIL) LO-TP-27102-00-013 1) RIL Computer Comparator Inputs (a) Rod control Data Logging via P-A Converter 11 =--p4e

• -.e

-g...cg m

i ._q j. i + QUESTION: 3.12 ~ COMMENTS: The latest setpoints are provide'd in tNA?VEGP technical specifications. RECOMMENDATION: Accept the following as correct: 3. 1960 psig. 6. 1850 psig 3

REFERENCES:

..VEGP Technical Specifications, Table 2. -1 and Table 3.3-4 2 \\ s s i .fs s 7 h;,_ i,. 3 s f { M-i r t k y e..- ,n rw- --,w,- c-, ..n, ~6,

.O D I TA8LE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS j s i G J N-oo */, N-oo

• z, N-oo +3, N-oo++ )

SENSOR TOTAL ERROR FUNCTIONAL UNIT ALLOWANCE (TA) Z_ (S) TRIP SETPOINT ALLOWABLE VALUE c z L. Q 1. Manual Reactor Trip M.A. N.A. N.A. N.A. N.A. L w 2. Power Range, Neutron Flux y/ / / / y' y' fif. f 'a. High Setpoint [7.5] {4.56] O 1{109]% of RTP** 1[111.23% of RTP** f4.56[0 15 of RTP** of RTP** b. Low Setpoint .3 S h of RTP** with 1E h of RTP** with 3. Power Range, Neutron Flux, M f .5 0 a tip constant a Aipe constant High Positive Rate j (N-oo # /, #-oo +2, N-oo +5, N-oo++ I 1.12fseconds >[2. seconds $b of RTP** 6 tith S h of RTP** with i 4. Power Range, Neutron Flux, N 5 0 High Negative Rate af.iy constant a,tl constant [ (N-oo ps, Al-co 4z, N'.co +3, A -co*4 ) >t2il seconds $ seconds l _h5hofRTP** of RTP** 5. Intermediate Range, 7.0 8.41 0 f Neutron Flux iN-co36 M-co36 ) ) y/ y J (l.nfe.r) (Lrfe.r-) e $E4G 3 cps $[1. ' :: 19' cps IL i 6. Source' Range, Neutron Flux (17.03 {10.013 0 Al-co s 2 ) 4S 34 I J l

7. (M-co3/Overtedperature AT E6-73 Ee-793 $0.87 See Note 1 See Note 2 3

f rf-co-// A + 8, ff-092 /A 48, TE-o+3/d

• 8, 7E-044r A rA )

l ) See Note 4 8. Overpower AT E*B3ro /+3 E+-33418) f0.2y See Note 3 i (rr -ort / sq +8, TE-o*2 / At8, YE-or31 A v6, TE o+4/y } 19so: 19 I 9. Pressurizer Pressure-Low! E6-83 5./ 10.71 1.5V ~>E19983 psig ~ psig N 7 iPf-o4ff ),8 +C, ff-o454 A,8,K, PT-o457),8,9c, kr-o45g A g v c) h385[psig i 39f psig 1 Pressurizerfressure-High $3.1F 0.71F'f.8)/ 1 I 10. +C. PI-orsy A,8 +C,18Pfl.5f,$[92]%)of instrument ${93.83% of instrument fr-ovst A sitc s .j ( (PT-o pSS J,8, vC, Level-HighP f-o +S 4 ), 8,t5.,0] u d2. Pressurizer Water 11. 'N " 'U** ( (L7-o f59, LT-0440, L 1-oV6/ /1.5[ _, 0 of loop of loop 12. Reactor Coolant Flow-Low 5 design flow

• design flow *
• Loopdesignilow=k95,700fgpa Lee o 1 Loor z Loon s Loos,4

[

• • RTP = RATED THERMAL POWER f7-oA//V fr-oV2V fr-0Y14 ffovyy FT-c y/f f r-cy2 f f 7-o V36 fr-cyyC fr-o Y/4 f7-c Y ? h ff-oYf4 f r-c ys,o S

[ t lT i TABLE 3.3-4 I 6 ij S ' ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUNENTATION TRIP SETPOINTS i1 m i l, e SENSOR 5 TOTAL ERROR i FUNCTIONAL UNIT ALLOWANCE (TA) Z s (S) TRIP SETPOINT ALLOWA8LE VALUE l 1. Safety injection (Reactor Trip, i Feedwater Iso 1ation, Control Room Emerjency Mode Actuffon, fornporreistfooA*r,y stafer-

I

?::: helet4on, Start Diesel

i Generators, Containment Cooling

\\

Fans,

.if;;gry Service,Materf, do,,pa;,,,,,,,,+fe,tara,, p;,a ), 'p,,,ja,,,,,,,,,,e g,,;;/,/1,,,Ise Au%, Aa akry ~ ll

Caohay,

/~ es.iafer-a. Manual Initiation N.A. N.A. N.A. N.A. N.A. N,4r_ pe,wn e b. Automatic Actuation fogic N.A. N.A. N.A. N.A. N.A. [0.71 [ h.5f $ h psig 5[3.b1psig c. Containment Pressure--Nigh 1 .0 1 -

i. PI-o 43 4, Pr-ogas, Pr-os3 t, PI-

'1 li! . d. ^ PressurizerPressure--Lowl 3.f M0.71J fl.5[ $f1850kys1g: $ M83 psig / b ( PI-0455,ial P :::, PI-o45'I, Pr-o45! PI-o 4 5t, 1 l, Different [3.D] [Gr87] [1.5/ c. ~ [97] psi - [100] psi l [. Betu::7. St:= LS:;--High 1r19 c.f. Steam Line Pressure--Low /20.0/ f10.71/p.5/ S k psig S k psig* ,q- /3. / SIS St 7 'p 2. Containment Spray a. Manual Inftiatton N.A. N.A. N.A. N.A. N.A. l l b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A. 'p and Actuation Relays } / / 2 2. 22 8 c. Containment Pressure--High-3 73.0 fd.717 .57 5 [12.05] psig 5 [12.31} psig 9 i PE-0934, PE-of3 G, PE-o936, PT-o937 Loop 1 Loop 2 Loop 3 Loop 4 F.Z~- o fi y PE-osz +e fr-oss e fr osyv d Pr-OS/6 PE OSid fr-of.9f PC cGyr PE-oss ia fr oczc fr-os3s f.r-o n s t 5

,7 QUESTION: 4.14 COMMENTS: Procedure 19211-1, Step 2, has two RNO actions.related to the turbine. Only one action is required if the turbine will not trip. RECOMMENDATION: Accept for full credit: (Manually) run back the turbine r

REFERENCES:

19211-1, Step 2' t 4 1 1 f 4 1 i b 2 t 4 i 4 4 f' i 1 l 1 .,,www-- .-,wr g , ram-n n,,, m w, ..g--en,n., w, w-,-, nw .an,-v,ne,g..m,,.e,.,. ,w,.g,r.,,,,.e-v,

PROCEDURE No. REVISloN PAGE No. 19211-1 0 2 of 6 f, s ACTION / EXPECTED RESPONSE , RESPONSE NOT OBTAINED IMMEDIATE OPERATOR ACTIONS 1. Verify Reactor Trip: 1. Manually trip reactor. Rod bottom lights - LIT. IF reactor will NOT trip, e Reactor trip and bypass THEN manually open supply e breakers - OPEN. Tiid breaker to INB08 and e Neutron flux - LOWERING. INB09: e INB08-01 e 1NB09-01 IF reactor NOT tripped, TREN manually insert control rods. 2. Verify Turbine Trip:

a. All turbine stop

a. Manually trip turbine, valves - SHUT.

IF:turbineLwill NOT. trip, THEN manually _run back ' turbine. IF turbine can NOT be run

Eack, THEN shut main steam line isolation valves and bypass valves.

3. Verify AFW Pumps Running:

a. MDAFW pumps - RUNNING.

a. Manually start MDAFW pumps.

b. TDAFW pump - RUNNING IF

b. Manually open steam NECESSARY.

supply valve HV-5106. 4. Check Reactor Power: ~ 4. Return to procedure and and step in effect. e GREATER THAN 5%. -OR-e e IR SUR - GREATER THAN 0 DPM. CONTINUED m..s \\ .k- -ndi-.

• ws k'.lrY'*.

a -.s's es.' ~ .. ur -.

-QUESTION: 4.20 ' COMMENTS:- -The E0P Training Text instructs operators to follow E0P actions even if Tech Spec limits are exceeded. RECOMMENDATION: _ Accept as an additional explanation: Technical Specification limits may be exceeded if directed by-E0P actions.

REFERENCES:

VEGP E0P Training Text 1-14, 15

I i Safety Function Restoration Concept working together to direct the operator's i action when the operation of the plant exceeds the bounds of the protective system limits. Figure 3. EMERGENCY OPERATIONS CONCEPTS s Normal Operation c Ne Yes Abnormal Operation U" O oiseasses n, l g . Emergency Operation h E, { ir No /SI Ne ...C.S..F., Optimal j Recovery yee 1 Concept CSF c Evene No Restoration l Concept l o v.. 7 y R.C.r? l 4 i7.. ~ [ App'1 cation of the" Plant Tech' Specs 3 ~.. ~,.. Oprrating the~ plant in'an'energency condition may cause the operator-to i t " violate'planttechl specs; The plant Technical Specifications'contain'the 11miting conditions."for. plant norma 1' operation ^in the applicable modes. By abidingbytheseconditions,theNiant'soperationwouldbeconductedinasafe manner and the design safety features would be ready to respond if a design basis accident were to occur. One could consider that the Technical Specifications play a preventative and preparative role in ensuring plant safety. If an accident does occur, the Emergency Response Procedures play a responsive role in dealing with the accidents. The Energency' Response i " Procedures provide the actions ;to be performed and para: eters to be monitored to, i maintain' plant ' safety and to achieve optimal recovery., ~ t l I-1-14 rev. I 4 ,,,.n ....., ~., . ~.. - -_ _. ~ _ ~, _ ~ _ ~ _ _ - -

.- ~. When the safety systems are actuated and performing their role during an ) accident, many of?the'prepara ,.m _tive Technical Specifications.will-bel i lated *due 7

m. - -

to the action of the safety systems (e.g., af ter the RWST is injected into the RCS, its level will be below its Technical Specification limit). The accident j itself could be a violation of the preventative Technical Specifications (e.g., a LOCA will exceed RCS leak limits). The Emergency Response Procedures were developed to respond to accident conditions and are supported by extensive analytical background, in most cases best estimate. Theagtions' delineated'in}theprocedures'are,thoseactions ,necessary to~ deal with:the' accident',in order to maintain or restore the plant in + ~ a safe condition. In general, the Technical Specification limitations are considered in developing the emergency response actions in the procedures. However, the procedures contain actions which wil1~~1ead'to Technical ~. c Specificationviolationyinordertomaintainplantsafety(e.g., opening pressurizer PORVs during a complete loss of secondary heat sink will violate RCS leak limitations, but is necessary to provide for core cooling and prevent more severe consequences). I r Althoughit1s[desirabletoremainwithinTechnicalSpecificationl limits?at all times, one must keep in mind that the overall' objective is to protect theT 1 health'and~safetyfof the public} This may require' procedurally. violating'a ~.. particular Technical Specification.in response to an accident. s. i r ] t s-I-1-15 rev. 1 ni ...- L A --~ -~.- -.--..c n.,

[

. QUESTION:

4.22 COMMENTS: Step 8 of Procedure 19013-1 requires realignment of. containment spray when RWST level less than 16%. The caution note on page 4 of Procedure 19013-1 requires realignment upon receiving the RWST EMPTY LEVEL alarm. Also Tab 4.9 of the Plant Technical Data Book shows RWST EMPTY LEVEL alarm corresponding to 9%_RWST level. ' RECOMMENDATION: Accept the following answer as correct: RWST EMPTY LEVEL alarm (9%) 'OR RWST level less than 16%

REFERENCES:

1. 19013-1,.Pages 4 and 9 2. Plant Technical Data Book, Tab 4.9

s PROCEDURE No. -- REVISION PAGE No. 19013-1 0 4 of 12 ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTION e Any; pumps.i.taking? suction'from"the RWST~should be r

stopped _.upon; receiving RWST_ EMPTY LEVELyalarm~.' 'Then complete the transfer to recirculation and restart the ECCS pumps in the following order:

a. RHR pumps.

b. SI pumps.

c. CCPs.

.d.yCS; pumps.] e The SI pumps should be stopped if RCS pressure is-greater than their shutoff lead pressure of 1625 psig since their miniflow is isolated during this procedure. 3. Align ECCS For Cold Leg Recirculation:

a. Verify RHR pumps -

RUNNING.

b. Verify RHR hot leg suction valve positions:

e HV-8701A - SHUT e HV-8701B - SHUT e HV-8702A - SHUT e HV-8702B'- SHUT l

c. Place the control switches for the l

following equipment l. to the position i indicated:

1) Lockout selector switches for SIP RWST suction j

isolation and SIP miniflow isolation: i l e HV-8806 - ON e HV-8813 - ON CONTINUED um

e ' PROCEDURE No. -- REVISloN PAGE NO. 19013-1 0 9 of 12 ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED 6. Verify other flow path 6. Recheck valve alignment and from CNMT sump to RCS pump status. cold legs: IF one complete injection

a. RCS cold leg inj ection path cannot be established from CCPs:

between the CNMT sump and the RCS cold legs, o RHR supplying CCP THEN go to ECA-1.1 LOSS OF suction header. M KGENCY COOLANT e One CCP injecting RECIRCULATION. through the BIT. -OR-

b. RCS cold leg injection i

from SIPS: e RHR supplying SIP suction header. e One SIP injecting into RCS cold legs. 7. Start ECCS Pumps As Necessary. 8. Check RWST Level - 16% 8. WHEN RWST: level lowers.to 4 l OR LESS. T5f7T M 'erfory Step 9.) P Return to Procedure and Step in effect until RWST level lowers to 16%. 1 i l-CONTINUED l w

4 )RoCEDURE No. REVISloN PAGE No. 19013-1 0 10 of 12 ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE Spray additive flowrate is used in this step to confirm-satisfactory CS pump operation after alignment for recirculation. 19 L "?fAlign icontainment: Spray, ' System Foritacirculation: -}

a. Open CS pump 1 suction valves from containment emergency sump:

e HV-9002A e HV-9003A

b. Shut CS pump 1 suction valve from RWST:

e HV-9017A

c. Verify continued spray

c. Verify spray additive additive flowrate of tank level and proper approximately 40 gpm valve alignment.

per operating CS pump if running. IF valve alignment verified correct, THEN establish communications with an operator at local CS pump suction and discharge pressure gauges to verify proper operation.

d. Open CS pump 2 suction valves from containment emergency sump:

e HV-9002B e HV-9003B CONTINUED 103445

a 4 es au 23 55 S-TG s REFUELING WATER STORAGE TANK ~f Es 'ns i e!! ) P m EQUIP. # 1-1204-T4-001 4 3 900 4v3 h j N A D E MO g=z< " E 94 700 PJ ^ L" T SIT E'?3S94:0 9*

j m E LCl ALARM

.4 65 l I)-- n 0 1 S UMIT 14 f Ey G / SD !2 / $y E !o _J 500 u g = E =5-400 2,6 / c) a S 300 - nin As au m nr I manniani) O m 200 h.,- ' --"** E ( "- Q^^') 100 / 0 0 2 = 0 10 30 50 70 90 100 E E "E INDICATED (%) Ll-990.991,992,993 i U" S EE ~ M n

QUESTION: 6.05 COMMENTS: The ESF sequencer lesson plan provides a more detailed discussion of sequencer operations. The only time a " load shed" occurs is on an under voltage condition. A load shed will remove all loads. If an SI occurs af ter under voltage sequencing is complete then SI loads will be sequenced on and an ACCW pump will be running. The ACCW pump was sequenced on with the U/V sequencer and is not a safety related load. If an SI occurs while the U/V sequencer is still sequencing loads (and before the ACCW pump starts) then SI loads will sequence on; all non safety related loads will remain de-energized. RECOMMENDATION: Accept the following as the correct answer - loss of power with an SI before the U/V load sequercing is completed.

REFERENCES:

LO-LP-01401-00, Pages 7-13

ilj. LESSON OUTLINE: N E D. Sequencers are located in control building - level B in ESF switchgear rooms E. Sequencers are powered from 1E power sources 120V AC distribution panels 1. LAY 2A - Train A 2. 1BY2B - Train B III. UNDERVOLTAGE CONDITION ON A SAFEGUARDS BUS A. Basic Sequencers function upon sensing undervoltage conditions 1. Shed all loads; load center transformers remain connected. 2. Start Diesel Generator 3. Trip 4160VAC preferred power supply breaker 4. Shut DG BRKR (when pehnissives satisfied) 5. Start timed incremental load sequence B. Undervoltage Sensing LO-TP-01401-00-002 1. Four (4) u/v relays on each bus monitoring for three decreased levels of bus voltage. a. located at the sequencer panel. b. feeds three sets of four bistables 2. Third level voltage a. A less than or equal to 88.5 percent for greater than equal to 10 seconds b. "4160V SWGR (Bus) TROUBLE" -- QEAB-ALB-36/37 3. Second level voltage a. less than or equal to 86.5 percen*. for greater than or equal to 20 seco h 7

LO-LP-01401-00 111. LESSON OUTLINE: NOTES b. "4160V SWGR (Bus) TROUBLE " QEAB-ALB-36/37 c. Start Sequencer Operation 4 First level voltage a. less than or equal to 70 percent for greater than or equal to.75 sec. b. "4160V SWGR (Bus) TPOUBLE" QEAB-ALB-36/37 c. Start Sequence Operation. C. Detailed Sequencer Operation 1. Simultaneously a. DG start signal LO-TP-01401-00-003 LO-TP-014.01-00-004 b. Trip preferred normal and alternate incoming breaker to 1E bus (2 see TDPU) c. Shed all bus loaas exc'ept load center trans formers (.2 see TDPU) 2. Shut DG breaker if following permissive satisfied a. DG ready to load MM greater than 440*and greater than 90 percent rated voltage (less than 9.5 sec) b. Preferred incoming breakers open (1.2 sec) c. No bus faults as indicated by no tus lockouts on preferred incoming' breakers d. DG close permissive from sequencer (indicates load shed completed) e.- No lockouts on Diesel engine or generator. 3. Start Sequencer loading a. .5 second delay - allows on line transformers to receive initial surge upon bus energisation. 8

T LO-LP-01401-00 ); 'L'ESSON OUTLINE: NOTES [ b. sequencer loads at 5 second intervals sequencing completed at 30 seconds after start c. 4. When last timer activates (30 sec) - electric input automatically resets u/v activation circuits 5. BLOCK AUT0/ MANUAL SIGNAL CLEARS at 36 seconds after sequencer start to allow operator induced load manipulation. D. Loads sequenced during first 30.1 seconds on uv signal (* required knowledge for students) - Refer to H0 page 3 E. Non-1E loads powered from INB01 or INB10 will be reenergized (if an SIS does not exist) when the bus feeder breaker is reshut by the sequencer. The busses may be reenergized with an SI signal present by pushing the "SI Override" pushbutton and then closing the breaker. F. Protection against excessive load cycling. 1. Logic provided to prevent more than two under-voltage conditions from being recognized within a two hour period and starting loads. limitation is provided to prevent a. automatically dxceeding manufacturer's re-commendations concerning motor start capability of two successive starts wittiin a two-hour period. LO-TP-01401-00-004 b. Logic Operation 1) The let and 2nd'UV signal vill initiate a load shed and resequence of the re-l quired loads. 2) The third UV signal will initiate a load shed only 2. Reestablishment of sequencing may be accomplished by t manually resetting the timer at the sequencer. It will automatically reset after two hours. G. Points to stress 1. Only the diesel and sequencer associated with the affected bus will operate. 9 ~ --~

LO-LP-01401-00 fil'. LESSON OUTLINE: NOTES 2. If the normal IE bus feeder breakers fail to open within 1.2 seconds after the UV condition is sensed a. " UNIT RES AUX XFMR FDR SUP BKR FAIL TO OPEN" 4 on QEAB-ALB-36 b. DG breaker will not shut and sequencer will not load the bus. 3. If any equipment fails to start after the last load is sequenced on: " SEQ ( ) SAFETY EQPT FAIL TO START" a. QEAB-ALB-36/37 b. Operators must search for the faulty equipment and attempt to start or operate as required c. Operator aids include 1) MLB on QMCB 1A1 and 1A2 2) E0P attachments 4. 36 seconds after t'he sequencer starts the auto / manual block signal is removed to allow component control switch manipulation a. This block exists to prevent on operator from stopping sequenced equipment during the petiod the sequencer is timing out. b. This prevents the operator from ad-vertantly providing a momentary stop or open signal to the selected load. Without this block circuit, if the op-erator realigned equipment during se-quencer timing, the equipment would restart or reshut once the switch was i realigned to the neutral position. However, the sequencer logic would sense the equipment failed to operate properly and would generate a " SEQ SAFETY EQUIP FAIL TO START" alarm needlessly. This could result in operator confusion over equipment operability. 1 i l 10

LO-LP-01401-00 ~ LESSON OUTLINE: NOTES gy. SAFETY INJECTION SIGNAL CONDITION A. The Safeguards Sequencer monitors for an SIS from either train A or train B Solid State Protection System. B. Basic sequencer functions upon sensing the SIS 1. Starts both Diesel Generators 2. Starts timed incremental load sequence on SI 3. Blocks U/V sequence C. Detailed sequencer operation. LO-IP-01401-00-006 LO-IP-01401-00-003

• 1.

Both DG receive start signal thru the sequencer.

• Stress this point.

a. opposite train SIS input is electrically isolated to maintain train separation for fault isolation. the DG breaker will not close if no U/V detected on respective safety bus. b. "DG READY FOR LOADING" indications after g F 9.5 seconds. c. Diesel Generator Breaker 1) Breaker will not close if no U/V is detected on each respect'ive safety. bus. 2) Breaker will open on SIS if the gen-erator is in parallel with the preferred or alternate normal power source. 2. Sequencing operation a.- Sequencing starts after the following delays 1) 90 ms delay - for step timing bus reset 2) 0.5 sec. delay - to start load sequencing (hold over from UV circuit - see III -C-3-A.) ) 11

i LC-LP-01401-00 ~ ll1. LESSON OUTLINE: NOTES b. U/V logic is blocked by SI to establish appropriate loads to be sequenced. Both sequencers actuate to sequence SI loads, c. however only the train that received the SIS signal will respond. i.e., Inadvertent " Train A" actuation 1) Train A and B DG's will start 2) Train A and B Sequencers will operate 3) Only " Train A" loads will actually sequence on. a) The sequencer contact is in series with an SIS contact in the start circuits of the various components, b) The SIS contact is train related. c) Therefore since both contacts must be shut to start the component only the components associated with the actuatsd SIS train will start even ~ though both sequencers are operating. D. Loads sequenced during first 30.5 seconds on SIS (* required knowledge for students) (Refer to handout Page 3) E. Points to Stress 1. Both diesels and both sequencers will respond to a SIS on either safety train. a. Diesels cannot be secured until the associated train SIS is reset. (ie. Train B diesel may be stopped when Train B SIS is reset Train A diesel cannot be stopped) f b. The DG breaker receives a trip signal directly from the SIS train - not l l via the sequencer. (see drawing (1X3D-BA-XO9A) 2. Diesel engines do not have to be running to l start sequencing if normal power is available I when an SI is received. 12 l

LO-LP-01401-00 lif. ' LESSON OUTLINE: NOTES V. SAFETY INJECTION WITH UNDERVOLTAGE A. Cases Considered 1. SIS and UV simultaneously 2. SIS following UV B. .ilS and UV simultaneously 1. Common signals (immediate) a. DG receive start signals b. Auto / Manual operation of ESF equipment blocked 2. U/V outputs a. Load shed IE & non IE loads (.2 secs) b. Trip RAT normal supply feeder breaker (.2 sec) c. DG auto closure permissive to DG breaker (.8 sec) 3. SIS outputs a. Block U/V signal to load sequencer. b. Activate step timing bus (.5 secs) after .1) DG started (9.5 see) 2) DG breaker shut 4. Auto / Manual operation of ESF equipment unblocked at 36 seconds after signal receipt. C. SIS following U/V 1. If sequencer is sequencing on UV loads when SI received; a. Sequencer resets to SI mode (90 ma delay) b. SI loads will sequence on as required 2. If U/V sequencer has timed out, receipt of SIS will be same as for SI without U/V condition. 13

QUESTION: 6.08 COMMENTS: A more complete list of temporary monitoring equipment is contained in the plant startup procedure. RECOMMENDATION: Any four of the following 9 components to be acceptable for full credit 1. Portable submersible neutron detectors 2. Scaler / timer 3. Ratemeter 4. HV power supply 5. (2 pen) strip chart recorder 6. Amplifier 7. Discriminator 8. Audible count rate control 9. Preamplifier

REFERENCES:

1. VG Startup Procedure, 1-500-01, Rev. 1 " Initial Fuel Load Test Sequence", Part 3.0, Test Equipment, Pages 6, 7 and 8 (attached)

(sacccou a n 1-500-01 1 6 ./ p ~ 2.8.2.13 93110-C, "New Fuel Assembly Handling Fixture Operating Instructions" 2.8.2.14 93210-C, "New Fuel Elevator Operating Instructions" 2.8.2.15 93220-C, " Fuel Handling Machine Operating Instructions 2.8.2.16 93240-C, Reactor Vessel Assembly / Disassembly 2.8.2.17 93260-C, " Fuel Transfer System Operating Instructions" 2.8.2.18 93270-C, " Sigma Refueling Machine Operating Instructions" 2.8.2.19 93300-C, " Conduct of Refueling Operations" 2.8.2.20 93320-C, " Preparation for Refueling" 2.8.2.21 93340-C, " Refueling" 2.8.2.22 93360-C, " Limitations and Precautions for Handling New and Partially Spent i i Fuel Assemblies" V 2.8.2.23 93250-C, Lower Internals Installation Instruction 2.8.2.24 93370-C, " Verification of Core Loading Pattern" 2.8.2.25 93420-C, " Primary Source Installation cuide 2.8.2.26 93500-C, Manual Operation of Fuel Handling Equipment. 3.0 TEST EQUIPMENT 3.1 PORTABLE SUBMERSIBLE NEUTRON DETECTORS Hake: WL6998A or equivalent No. Reqd: 3 (1 is a apare) 3.2 SCALER / TIMER g Hodel M/722 or M/778 or equivalent No. Reqd: 3 CONTINUED n..s e

' poccou..E n 1-500-01 1 7 (- RATEMETER s Model: M/441 or equivalent No. Reqd 3 3.4 HV POWER SUPPLY Model: M/456 or M/109C or equivalent No. Reqd 3 3.5 2 PEN STRIP CHART RECORDER Hodel: LATER No. Reqd 1 3.6 AMPLIFIER Model: M/490 or equivalent No Reqd: 3 n 3.7 DISCRIMINATOR t ) Make N/421 No Reqd 3 3.8 AUDIBLE COUNT RATE CONTROL No. Reqd: 1 3.9 PORTABLE NEUTRON SOURCE (1-5 CURIE) No. Reqd 1 3.10 . FUEL LOCATION STATUS BOARD No Reqd 1 3.11 BINOCULARS g No. Reqd 2 3.12 BATTERY POWERED llEADSETS No. Reqd 3 O v CONTINUED m...

^ T,accEDU..E No 1-500-61 l 8 LE M ON NE NO l c) 3.13 ' PREAMPLIFIER Model: M/142IH or equivalent No. Reqd 1 4.0 NOTES AND PRECAUTIONS 4.1 GENERAL 4.1.1 Positive identification must be maintained on all fuel assemblies and inserts throughout the fueling operation. 4.1.2 The RCS boron concentration should not be greater than 150 ppm more than the minimum required boron concentration to prevent unnecessary reductions in the response of the neutron detectors. 4.1.3 If the Inverse Countrate Ratio (ICRR) data for any nuclear monitoring channel reaches a value of 0.1 renormalize the ICRR data to 1.0. 4.1.4 After a temporary neutron detector has been positioned within the reactor vessel, the detector position should-(~J not be altered until required by the loading sequence. N, ( Signal cables and tie-down ropes should be fastened to a stationary object to prevent inadvertent movement of a detector. If a detector is moved for any reason (such as hitting a tie-down rope), reposition detector and a new baseline count rate shall be obtained per 1-500-02, 4.1.5 If a fuel assembly is to be temporarily stored in a safe core location, it must be placed adjacent to the core baffle wall and be separated from the nearest single fuel assembly by a minimum of one fuel assembly width and two fuel assembly widths from the nearest cluster of fuel assemblies. If these geometry conditions cannot be achieved the fuel assembly shall not be stored in the reactor vessel. 4.1.6 Do not remove underwater lights from the core while energized. They will shatter within seconds of not being submerged. 4.1.7 ICRR plots are not required for the first een (10) assemblies. Criticality is not possible while loading these first fuel bundles. lh* CONTINUED m.., i k

3. QUESTION: '6.09 COMMENTS: ' Tech Spec Table 3.3-9 provides the list of remote shutdown panel instrumentation. Eight indications are available on the shutdown panels. RECOMMENDATION: ' Accept any 6 of the following 8: .1. SR neutron flux 2. Extended range neutron flux 3. T wide range 4. T wide range h 5. Core exit thermocouples 6. RCS wide range pressure 7. Wide range SG 1evel 8. Pressurizer level

REFERENCES:

VECP Technieni Specifications, 3.3.3.5.a " Remote Shutdown System Instrumentation", Table 3.3-9, Page 3/4.3-58

T2e d b 4 ' E # f 3, 33. 3. 5, c. TABLE 3.3-9 REMOTE SHUT 00WN SYSTEM i [, I Minimum ,3 Readout Channels Channels 'estru ent Function Location Available Ooerable i 1. Source Range Neutron Flux A 1 NI 31E 1 2. Extended Range Neutron Flux B 1 NI 13135 C&D 1 3. RCS LP Cold Leg Temperature A, B 1/ Loop 1/ Loop Loop 1 TI 04130 (Panel A) Loop 2 TI 04230 (Panel B) Loop 3 TI 04330 (Parel B) Loop 4 TI 0443D (Panel A) 4 RCS Hot Leg Temperature A 2 2 Looo 1.TI 0413L Loop 4 TI 0443C 5. Core Exit Thermocouples B 2 2 Loop 2 Core Quadrant TI 10055 Loop 3 Core Quadrant TI 10056 6. RCS Wide Range Pressare A, B 2 2 Panel A PI 405A Panel B PI 403A I Steam Generator Level A, B 1/ Loop 1/ Loco Wide Range Loop L LI 501B IPanel A) Loop 2 LI502B(PanelB) Loop 3 LI 5036 (Penel B) Loop 4 LI 5048 (Psnel A) 8. Pressurizer Level A, B 2 2 Panel A LI 459C Panel B LI 460C 9. RWST Level L 1 LI 0990C 13

10. BAS.T Level L

1 PI-101152 13

11. CST Level L

2 23 Tank 1 l LI 5100 Tank k' l LI Sil! + 1. A. Remote Shutdown Panel PSDA B

• Re*ote Shutdown Panel PSDB L

Local Indication L 2. l Gra::h will be provided to detennine level from pressure reading. 3. Alternate local level indication may be established to fulfill the minimum channels OPERABLE. ,) l V0GTLE - UNIT 1 L/' 3 58 \\

(y < : 7 ;,. ,:y[ ?_g ~' e .y !p^ W r.. m ~ p,.. x QUESTION: ..r - 6.10 c ,sa ,a. ', t X y.- 'The/: Interlocks specified in the Vogtle Training Text were based i< , COMMENTS: unon!n generic SIGMA refueling machine before Vogtle's was ' ' 't.. M l purchased. Subsequently a machine has been purchased with the d.nterlocks listed in the attached lesson plan. a, h,f RE(OMMENDAT10h: Use the!corrher interlocks listed in lesson plan LO-LP-25101 to d

p forimilato the ; correct answer for the question given.

kI, ,a + E

REFERENCES:

LO-LP-25101, Pages 16 and 17 <7 T ? 4 r J l / ? m t .F r,. ,g. -.g .e '.g o .. e ; e E we .,e-- \\' i s i-M y 4 J 6 ?, s 4 Y ( g ( ,\\ ll + f k 1 ,.i i c ! r.. t ,9

• I Q: --

s, J'

i LO-LP-25101-01-C 111. LESSON OUTLINE: NOTES

11) Indicating Lamps a)

HEATER ON (side of console) b) INTERLOCK BYPASSED c) FRAME DOWN d) FRAME UP e) TRAVERSE CONTROL (Lit indicates Rx side switch is on) f) CAR AT PIT g) CAR AT REACTOR h) VALVE OPEN d. Emergency Pull Out Cable System 1) Provision for failure of transfer drive system during refueling 2) Allows transfer ca'rt to be physically e pulled back into SFP building by cable. E. Fuel Handling Safety Interlocks 1. The refueling machine can only place a fuel can be bypassed assembly in the core or FTS. 2. When the refueling machine gripper is supporting a fuel assembly, the machine can only traverse when the fuel assembly is fully withdrawn into the mast. 3. When the refueling machine gripper is not supporting a fuel assembly, the nachine can can.only traverse when the gripper is withdrawn to a safe height above the fuel assembly. 4. Simultane'ous traversing and hoisting operations are prevented. s 5. The refueling machine is restricted to raising a fuel assembly or core component to a height at which the water provides a safe radiation shield. s-16 3 ' mren M 6 wJ w D++A ort - N M M We - ee n - ' -e m ow4 Amap-m kap w...#g 44.h a.e Jo g hah [ W mh'M4m ' - ' ' ' ' ' - ~ -

LO-LP-25101-01-C Ill. LESSON OUTLINE: NOTES 6. When a fuel assembly is raised or lowered, ( interlocks ensure that the refueling machine can only apply loads which are within safe operating limits. Hoist load cell limits can be bypassed. 7. The fuel gripper is monitored'by limit switches to confirm operation to the fully engaged or fully disengaged position. An audible and visual alarm is actuated if both engage ar.d disengage switches are actuated at the same time or if neither is actuated. 8. When the refueling machine is supporting an RCCA, it can only traverse when the RCCA is fully withdrawn into the guide tube and the guide tube is withdrawn to a safe height above the fuel assembly. 9. Lowering of the guide tube is not permitted if slack cable exists in the hoist. 10. The guide tube is prevented from lowering completely out of the mast. e 11. Before the fuel gripper can release a fuel ( assembly, the fuel gripper must be in its down position in the core or in the FTS. 12. The weight of the fuel assembly must be off the gripper before the fuel gripper can release a fuel assembly. 13. The FTS container is prevented from moving unless the engaged gripper is in the full up position or the disengaged gripper is withdrawn into the mast or unless the refueling machine is out of the fuel transfer zone. An interlock is pro-l vided from the refueling machine to the FTS to accomplish this. l III. Summary l Review the objectives I A. I 17 ~- 4 1 -- = .. -. ~.-- -

o

o..

QUESTION: 6.12 COMMENTS: Additional signals will stop blowdown flow. RECOMMENDATION: Accept the following as additional answers: a. High temperature (TE-1150, closes FV-1150) b.- High pressure (at demin inlet, shuts FV-1150) -c. High flow'(FT-15212 A, B, C and D will close HV-15212A, B, C and D respectively)

REFERENCES:

Reference VEGP Training Text 13c-4, 5

h l (s1 s s/ STEAM GENERATOR BLOWDOWN SYSTEM on the condensate inlet lines. Indication, a variable control set point, and variable high and low alarm are provided on the PSBP panel. TI-1158 - TRIM HEAT EXCHANGER INLET TDIPERATURE This instrument indicates on the PSBP panel the temperature of the blowdown entering the heat exebanger. =TE-1150'-l BLOWDOWN TRIM HEAT EXCHANGER BLOWDOWN OUTLET TEMPERATURE ~ This instrument indicates.the, temperature of the blowdown flow and closes,FV-1150_on.a high signal to protect l downstream demineralizer resin. Indication and a high alarm areprovided on the PSBP panel.

• 0ne instrument per blowdown connection.

AUX BUILDING LINE BREAK MONITORS This system serves to isolate SGBD when a line break is detected in the aux bldg rooms. Valves 15212A,B C.D and 15216A,B,C,D will automatically close on high temperature from these area monitors. This temperature is displayed on the QPCP (TI-15212, 15216). Should this temp. rise high enough to cause closure of 15212 and/or 15216 a QMCB alarm will set (ALB63F01) to alert the operator of the safeguards actuation. HS-15212E g-~s & 15216E allow the operator to select the controlling temperature t I channel for each train. HV-15212 serves as the train A protection and \\' HV-15216 serves as the train B protection. Note if the safeguard closure actuation occurs all train associated valves will close: thereby isolating all blowdown. 7.2 PRESSURE PIC-1151 - MAIN PRESSURE LETDOWN This instrument maintains a constant backpressure on the high pressure section of the SGBDS via PV-1151 modulation. Pressure changes due to load fluctuations are compensated by individual blowdown flow control valves. Indication, a variable control set point, and a variable high alarm and low are found on the PSBP panel. PN11537-PRE-FILTERINLETPRESSURE ~ This instrument indicates the inlet pressure to the pre-filter and provides a high' pressure signal-to close FV-1150. Indication is local and'on the PSBP Panel. 'A high alarm is also located on the PSBP Panel. PI-1154 - PRE-FILTER OUTLET PRESSURE This instrument locally indicates the outlet pressure from the pre-filter and the inlet pressure to the mixed bed demineralizers. In conjunction with PT-1153, it may be used to determine the pressure drop across the pre-filter. Filter cartridges should be {v~'} changed when the Delta-P exceeds 20 psi. 13c-4 Revision 1

( \\ STEAM GENERATOR BLOWDOWN SYSTEM ~ PDS-1218 OUTLET. FILTER DIFFERENTIAL PRESSURE This detector measures the common Delta-P across the parallel outlet filters and alarms at the PSBP when the Delta-P exceeds 20 l psi._ PDI-1156 and PDI-1184 locally indicate the Delta-P across outlet filters 1 and 2, respectively. PDI-1168 SPENT RESIN SLUICE PU)(P DISCHARGE FILTER DIFFERENTIAL l' PRESSURE This instrument locally indicates the differential pressure across the spent resin sluice filter. This filter cartridge should be changed when the Delta-P exceeds 20 poi. 7.3 LEVEL LI-1165 - SPENT RESIN STORAGE TANK LEVEL This instrument indicates the total level in the spent resin storage tank, i.e., combined resin slurry and sluice water level. High and_ low level alarms are provided, as well as a low-level interlock which shuts off the spent resin sluice pump.. Level indication and alarms are located on the PSBP panel. Indication is also on the Solid Waste Disposal Panel. 7.4 FLOW FI-1152 - INLET FLOW l This instrument indicates the total blowdown flow rate through the system. Indication is provided on both the Main Control Board and the PSBP. A high alarm is also provided on the PSBP Panel. FI-1160 - DISCHARGE FLOW l This instrument indicates the blowdown flow rate to the W.W.R.B. Indication is on the PSBP Panel. 1 FI-1169 - SPENT RESIN SLUICE PUMP FLOW This instrument provides indication of sluice pump flow. A low flow interlock shuts off the spent resin sluice pump in case the suction line screen or pipe becomes resin-plugged. Indication and a low alarm are located on the PSBP Panel. PT-15212, 15216, AUX BLDG _LINE_ BREAK _ PROTECTION _,,,., Righlflow.in the SGBD linelwill isolate the blowdown'line.' FT-15212A,B C and D for loops 1 through 4 respectively, will auto-close HV-15212A,B,C and D, respectively. Similar isolation will occur with FT-15216A,B,C and D. HC-1171, 1172, 1173, 1174 - STEAM GENERATOR BLOWDOWN FLOW This instrument maintains a constant blowdown flow rate from each steam generator connection via HV-1171, 1172, 1173, 1174 modulation. A high alarm is provided on the PSBP Panel while indication is provided on the QPCP and the PSBP Panel. 13c-5 Revision 1 r

+ I - QUESTION: 6.16 ~ COMMENTS: Power range high negative rate and high positive rate trips are 2 separate trips. Each has its own basis and each is a 4 separate line item for required operability in Table 3.3-1 of the VEGP technical specifications. RECOMMENDATION: Accept the following as two separate non-blockable reactor trips if specified: 1. Power range neutron flux, high' negative rate (2/4). 2. Power range neutron flux, high positive rate (2/4).

REFERENCES:

VEGP Technical Specification 3.3.1. Table 3.3-1 l l-L i ( i l [

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERA 8LE with RESPONSE TIMES as shown in Table 3.3-2. APPLICABILITY: As shown in Table 3.3-1. ~ ACTION: As shown in Table 3.3-1. SURVEILLANCE REQUIREMENTS 4.3.1.1 Each Reactor Trip System instrumentation channel and interlock &nd the automatic trip logic shall be demonstrated OPERABLE by the performance of the Reactor Trip System Instrumentation Surveillance Requirements specified in Table 4.3-1. 4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be demonstrated to be within its limit at least once per 18 months. Each test shall include at least one train such that both trains are tested at least once per 36 months and one channel per function such that all channels i are tested at least once every N times 18 months where N is the total number of redundant channels in a specific Reactor trip function as shown in the " Total No. of Channels" column of Table 3.3-1. V0GTLE - UNIT 1 3/4 3-1

TA8LE 3.3-1 h REACTOR TRIP SYSTEM INSTRUMENTATION m MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICA8LE c$ FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERA 8LE MODES ACTION -e 1,f4k57 2 g. 1 H 1. Manual Reactor Trip 2 1 2 3f 11 9 2 1 2 Power Range, Neutron Flux (A/ood/, A/oo#2, 4 b 2. 2 3 1, 2A 2f'b a. Hfgh Setpoint g g, g,my b. Low Setpoint 4 2 3 IW, 2 2f-b 3. Power Range, Neutron Flux 4 2 3 1, 2 2f High Positive Rate Moo + 4) b (Noo+/ Noo+2 Hoof 3' Flux, Power Wange, kutron 4 2. 3 1, 2 2f 4. w1 High Negative Rate Woo 43 Woo 4+] (Noo+1, Moo 4 z, Range,, Neutron Flux w Intermediate 2 1 2 1""', 2 3 4 5. ~ (Woo 3S, Nc' ox)

6. ~ Source Range, Neutron Flux / Noose, Noo32) f e'

[3,4,5 4 a. Startup 2 1 4 5 b. Shutdown 2 1 2 7. Overtemperature AT 4 2-3 /, 2 t, 6 [TE-c 9//,g-04f/, 7(-o+3/, ff-cyg/) "fi: Tied 5:'r::5: 1 2 3 1, 2 se Thr : Le-- epe etten i 1** 3 1, 2 0 b. Thr : L::;.*1: t T:;r : L::; 0;:::tt:= 3 2 2 1, 2 IC T;: L::; 0;;r:t!:= 3 1** 2 1, 2 0 A.a. ' 3

g. /kse r Ope, a fio n z

1 2 1 t

b. Sfa y fup 2-

/ / 2 3

~ oo. -..e- -QUESTION: 6.17a and 2.17a COMMENTS:' At normal operating pressure the. cold overpressure mitigation. system (COMS) is blocked by switches on the QMCB, panel C. RECOMMENDATION: Accept the following as correct answer: COMS is blocked or,COMS has no effect.

REFERENCES:

FSAR Logic Diagram, Figure 7.2.1-1 sent as Chapter 25 to VEGP-Training Text i L l r

0. - -- t',c m ~.ns h- -mP ~~-~ (*

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== s ss.v s ows e er am swa ma o i.f.v :*x ". '.E l,".! 'e'*od"^o.f/;'s.6. 4 w an s e.~s. =a ussa s e, e-we .-.no.

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6. Tat mCs e ).*.a=C ac? -Cr. i.e LM.e.O

(- 4.T% V M wwCC #A 4 Es 6.atow s e es. to.s m.s.vrermet.tr a. ms m fg w et e { t na a =::e.ws. h .se/t b e. s 'i Amend. 11 11/84 f PRESSURIZER PRESSURE ~ voorts _ ttscTaicosesenavneoptaary RELIEF SYSTEM (TRAIN B) IMI3M unnt t ase unsT FIGURE 7.2.1-1 (SHEET 19 OF 20) m..

+ - A av, , ' QUESTION: 7.06. ~ COMMENTS: The foldout page,.which includes SI actuation criteria,_ is applicable throughout E-0. In addition step 4 RNO contains 4 items requiring manual actuation.of SI. -Two sets:of indications relating to SI are in 19000-1. ~ ~ RECOMMENDATION: Accept as the: correct-ansver: 1. Subcooling monitor less than 28'F. 2. Pzr level cannot be maintained greater than 4%. OR 1. Pzr pressure less than or equal to 1850 psig. 2. Steam line' pressure less.than or equal to_585-psig or steam line.depressurization rate greater than or equal to 110 psig/2 sec. -3. Ctat press greater than or equal to 4 psig. .4. Any ' automatic alignment of ECCS equipment to the injection -phase.

REFERENCES:

19000-1, Rev. 0' E0P Training Text, Vol. I,_ Chapter 2. Page 3 i i I' l l I I i

PROCEDURE No. REVISloN PAGE NO. ,19000-1 0 3 of 26 3 ( ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED 4. Check If SI Is Actuated: 4. Check if SI is required: IF one or more of the e Any SI annunciator - LIT. Tdllowing conditions is met: 'PRZR7pressuie[ lass [than'Mr1 e SI ACTUATED BPLB e window - LIT. 'e q u a l -;;t o 1 8 0 0 t p s i g ;- s Steam:linelpressure less, e than or equalito 585.psig', ~ e Containment; pre'ssure Jreater;than_orgequalito, SPsig,j e! ' Any: automatic ~alignmetit; ~ ~ ' fof ECCS;~ equipment;_to:* s : ~.j ection: phase.- J in ..a a-THEN SI is required. IF SI is required, THEN manually actuate. - ) IF SI is NOT required, b TREN go to 19001-1, ES-0.1 REA'CTOR TRIP RESPONSE. 5. Verify FW Isolation: 5. Manually shut valves as necessary. o MFIVs - SHUT. e BFIVs - SHUT. e MFRVs - SHUT. e BFRVs - SHUT. 6. Verify Proper ESFAS 6. Manually actuate SI. Actuation: IF proper ESFAS actuation

a. MLB indications can NOT be verified, correct for injection THEN continue with phase after load Step 7.

sequencing.

b. Go to Step 13.

CONTINUED wm

PAGE NO. REVISION 26 of 26 p PROCEDURE NO. 0 19000-1 1 Sheet 1 of 1 7 r \\). jotDOUTPAGE 1. RCP TRIP CRITERIA Trip all RCPs if BOTH conditions listed below occur: CCPs or SI pumps - AT LEAST ONE RUNNING. a. RCP Trip Parameter - RCS PRESSURE LESS THAN 1375 psig. b. 2. SI ACTUATION CF.ITERIA 19000-1, E-0 REACTOR TRIP OR SAFETY Actuate SI and go to INJECTION, Step 1, if EITHF4 condition listed below occurs: RCS subcooling monitor, indication - LESS1THAN 28'Fi e PRZR levell-CANNOT BE MAINTAINED GREATER TEAN 4% ~ e [34% FOR ADVERSE CNMT!. 3. RED PATH

SUMMARY

SUBCRITICALITY - Nuclear power greater than 5%. a. CORE COOLING - Core exit TCs greater than 1200*F. b. h -OR-Core exit TCs greater than 700*F AND RVLIS full range less than T9I with no RCPs running. HEAT SINK - Narrow range level in all SGs less than l 27% AND total available feedwater flow c. less chan 550 gpm. INTEGRITY - Cold leg temperature lowers more than 100*F d. in last 60 minutes AND RCS cold leg temperature less than 248'F. CONTAINMENT - Containment pressure greater than 52 e. Psig. f 4. AFW SUPPLY SWITCHOVER CRITERION Switch AFW suction to alternate CST when operating CST lowers to less than 15%. J L 1 i t 4 F0344% d

1. QUESTION: 7.08D' COMMENTS: Both 2.2.3 and 2.2.4 in 12000-1 discuss RHR operations in mode 5. - RECOMMENDATION:- Accept the following as the answer: "No time limit (applicable if all RCS loops are filled, at least one RHR train OPERABLE and in operation, and the secondary side water level in at least 2 SG's is greater than 17% wide range.") OR 2 hours

REFERENCES:

.12000-1

Approval Procedura No. Vogtl] Electric Generating Plant k 12000-1 NUCLEAR OPERATIONS g, g,,,,,,, g,. Date 1 Georgia Power g Unit ease no. REFUELING RECOVERY (MODE 6 TO MODE 5) l.0 PURPOSE This procedure provides instructions for taking the unit from a refueling condition (Mode 6) to cold shutdown (Mode 5). 2.0 PPICAUTIONS AND LIMITATIONS 2,1 PRfCAUTIONS 2.1.1 If the count rate on either source range channel increases unexpectedly by a factor of two or more during any operation, the operation shall be suspended ( immediately until a satisfactory evaluation of the situation has been performed. 2.2 LIMITATIONS 2.2.1 The RCS pressure and temperature should not exceed 450 psig and 350 degrees when open to the RHR System. 2.2.2 In Mode 5, shutdown margin shall be greater than or equal to 1.5% A k/k. (Technical Specification 3.1.1.2) -2.2.3 When in Mode 5, with loops filled, at least one residual heat removal train (RER) shall be operable and in operation, and either: a. One additional RHR train shall be operable, or I b. The secondary side water level of at least two steam generators shall be greater than 17% of wide range level. (Technical Specification 3.4.1.4.1) ' 2.2.4 While in Mode 5, one RHR train may be inoperable for up to 2 hours for surveillance testing provided the other RHR train is operable and in operation. (Technical Specification 3.4.1.4.1) WCI A MC4130 ~.. ~ n w sw.. w sw a....... s,. s a.. ~.

QUESTION: 7.19 COMMENTS:' Step 2'of 18038-1 has a RNO column with different actions than the action / expected response column based on the presence of a fire. - Thus the presence of a fire also dictates the number of-actions completed. RECOMMENDATION: Accept as an alternate answer: Control room fire

REFERENCES:

18038-1, Rev. 1 i l I l t { I l [ b t [ l' i

^ h_~ PROCEDURE NO. REVISION PAGE No. 18038-1 1 2 of 32 { 1 ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED IMMEDIATE OPERATOR ACTIONS '.1. Manually trip the reactor NOTE Step 2 should be done prior to Control Room evacuation;IFn. ~* personne1"safetyris,NOT hindered.1 am.~. uaa u,,;_. u...aa. 2 SUBSEQUENT OPERATOR ACTIONS

2. IF a Control Room fire, and 2.

IF NOT a Control Room fire at discretion of Shif t THEN: Supervisor THEN:

a. Shift CCP suction to

a. Verify-turbine trip.

RWST - OPEN: e LV-112E -OR-e LV-112D

b. Shut PRZR PORV block

b. Align VCT-for AUTO valves.

makeup: e HS-4001A in AUTO e HS-4001B in AUTO-AFTER-START e Pot set at 5.3 e LV-112A in AUTO e-PV-131 in AUTO

c. Stop RCPs 1 and 4.

c. Place PRZR pressure control in AUTO.

Tm 703445 l

QUESTION: 7.20 COMMENTS: Procedure 19200-1;has'been revised and changed setpoints; referenced in this question (19000-1 foldout page supports 19200-1). RECOMMENDATION: Accept the following as the correct answer: Narrow range level in all'SGs less than 27% and total available ~ l feedwater flow greater than 550 gpm.

REFERENCES:

1. E0P foldout page red path summary for heat sink. 2. 19200-1, Page 5, Heat Sink Status Tree 4 I-f l (.-

PAGE NO. REVISloN 26 of 26~ 4 PROCEDURE No. 0 i 19000-1 1 lj Sheet 1 of 1 FOLDOUT PAGE \\;. - e 1. RCP TRIP CRITERIA Trip all RCPs if BOTH conditions listed below occur: CCPs or SI pumps - AT LEAST ONE RUNNING. a. RCP Trip Parameter - RCS PRESSURE LESS THAN 1375 psig. b. 2. SI ACTUATION CRITERIA 19000-1, E-0 REACTOR TRIP OR SAFETY e Actuate SI and go to INJECTION, Step 1, if EITHF4 condition listed below occurs: RCS subcooling monitor indication - LESS THAN 28*F. e PRZR level - CANNOT BE MAINTAINED GREATER THAN 4% e [34% FOR ADVERSE CNMT1 3. RED PATH

SUMMARY

SUBCRITICALITY - Nuclear power greater than 5%. a. CORE COOLING - Core exit TCs greater than 1200*F. b. A -OR- ' Core exit TCs greater than 700*F AND RVLIS full range less than T91 with no RCPs running. ' HEAT. SINK - Narrow range level in all SGs less than 27% AND total available feedwater flow c. less than 550 gpm. INTEGRITY - Cold leg temperature lowers more than 100*F d. in last 60 minutes AND_ RCS cold leg temperature less than 248'F. CONTAINMENT - Containment pressure greater than 52 e. Psig. 4. AFW SUPPLY SWITCHOVER CRITERION Switch AFW suction to alternate CST when operating CST lowers to less than 15%. d 9 ,, up e g ~~ 1.- m%. g _ a.

PROCEDURENO. REVISION PAGE NO. 19200-1 I 'O 5 of 9 r f HEAT SUE - F-0. 3 1 23 -l' TOTAL AVAILABLE NO FEDWATER FLOWTO SGs GREATER I t THAN 550 YES GPM 99GSS99960e 'GO TO e 15i232-1 NARROW ~ NO N NO LEVELIN AT PRESSURE IN LEAST ONE SG 4tt3g,gg33 GREATER WM THAN 1235 YES 27% PSIG YES 999999998 GO TO g 19233-1 NARROW NO ~ CA RANGE LEVELIN ALL SGs LESS THAN 78". YES r t-GO TO O,OOOOOO 19234-1 NO PRESSURE IN ALL SGs LESS THANil85 PSIG YES eseece GO TO e 19235-1 l NARROW NO RANGE LEVELLN

i

_ _ y = = L u D%lt % SG8 g

y 27%

w SEr 10MLS e 'i, Y. '.*A-N Aw -- N'^~ ' _._.3.__. ,--m-- - - ~ "'- "-- ~ ~ " '"~

QUESTION: 7.21 COMMENTS: Cooldown rate is a technical specification LCO. The E0P training text instructs operators to follow the E0P actions even if tech spec LCO actions are entered. RECOMMENDATION: Accept as an additional explanation; "The RCS cooldown should be accomplished as quickly as possible, and should not be limited by technical specification limits."

REFERENCES:

VEGP E0P Text, Chapter 1, Page 14 and 15 9 t k

O. Safety Function Restoration Concept working together to direct the operator's i (j' action.when the operation of the plant exceeds the bounds of the protective system limits. Figure 3. EMERGENCY OPERATIONS CONCEPTS 4 Normal Operation A6.ren? Abnormal Operation Ru No g._tre7.s v Emergency Operation h f 1' ..$.df Set et Optimal Recovery y.. Concept CSF e No ROStorstlon D..,enes.c R.e., Concept g v.. u .R. Telp .. 58 I 4 7. Application of the Plant Tech Specs Operating the~ plant in an emergency condition may cause the operator to violate plant' tech specs.:- The plant Technical Specifications contain the limiting conditions for plant normal operation in the applicable modes. By abiding by these conditions, the plant's operation would be conducted in a safe manner and the design safety features would be ready to respond if a design basis accident were to occur. One could consider that the Technical Specifications play a preventative and preparative role in ensuring plant safety. If an accident does occur, the Emergency Response Procedures play a responsive role in dealing with the accidents. The Emergency Response ' Procedures' provide the actions to be performed and parameters to be monitored to maintain plant safety and to achieve optimal recovery. 1-1-14 rev. 1 i ~ __K_ $d .......-...i~. -21 . ;. ___4 _.

Mien the safety systems are actuated and performing their role during an r^N accident, many of the preparative Technical Specifications will be violated due ~ to the action of the safety systems (e.g., af ter the RWST is injected into the RCS, its level will be below its Technical SpecLfication limit). The accident itself could be a violation of the preventative Technical Specifications (e.g., a LOCA will exceed RCS leak limits). The Emergency Response Procedures were developed to respond to accident conditions and are supported by extensive analytical background, in most cases best estimate. The actions delineated in the procedures are those actions necessary to deal with the accident in order to maintain or restore the plant in a safe condition. In general, the Technical Specification limitations are considered in developing the emergency response actions in the procedures. Itawever, the procedures ~contain. actions wiiich will lead to Technicall

Specification' violat' ions"in ' orderf to meintain plant' ~saf ety (e.g., opening pressurizer PORVs during a complete loss of secondary heat sink will violate RCS leak limitations, but is necessary to provide for core cooling and prevent more severe consequences).

bG Although it is desirabin to remain within Technical Specification limits at all times, one must keep in mind that the overall objective is to protect the health and safety of the public. This may require procedurally violating a particular Technical Specification in response to an accident. (V I-1-15 rev. 1 a.... i.).:a dkss:a

s.. i QUESTION: 8.03B I: COMMENTS: The answer mentions a five (5) hour cooldown to RHR. The bases of the Vogtle Tech Specs has no time associated with cooldown. ' RECOMMENDATION: Delete the five (5) hour number in the answer.

REFERENCES:

Tech Spec bases for condensate storage tank 3/4.7.1.3, Page B 3/4 7-2 l ) s' J i

N PLANT SYSTEMS BASES SAFETY VALVES (Continued) 09 Power Range Neutron Flux-High Trip Setpoint for loop operation, L S.; p;rn at ;f "^TED E "*L 80"" "" " " W " T [%] = . 1 O t;: ht f:r [S 1] ? 0; 0?: Pet 10, Total relieving capacity of all safety valves per steam X = line in 1bs/ hour, and Maximum relieving capacity of any one safety valve in Y = 1bs/ hour AUXILIARYFdEDWATERSYSTEM 3/4.7.1.2 The OPERABILITY of the Auxiliary Feedwateri ystpa ensures that the Reactor S Coolant System can be coo' led down to less thanf350fF from normal operating conditions in the event of a total loss-of-offsite power.

d h

' d:1'x r-

- ('- Exh :? ntrS =tr-dr';= =r!: y "xt :t'r ;7 ':c' [250] g :t ; pr; n r; ef (1100] p;;; t. ..m tat?' 'rd ter '!r: 1m t : r c' t'e ete r ;r :te c. The ete: -d in : ~* y 'e + t Vit - 9't c' fe!' n *r; tat!! fr $> t 'l r e' [??O] g it : prn::r: :' [11 03 p;is te th; entrance ef the et;;; gene.aters. Thie cepecity ia e.ff! th;t :d:;=t; 'nictr 'ha n ;;;ihth t; re;;;; d;;;j h;;t 0'=t t: =:r: M ret r the ".=:tr C ri r t Sy:t= t =;; ;tr; : k n th;a [000]*r hca \\th; "n'd=1 Int Sn;;l Cy:t= ny t; pknd St; :;r;t'=. 3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITV of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT hours with steam discharge to the atmosphere STAND 8Y conditions for 4 concurrent with total loss-of-offsite powe% The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics. % lo// owed by a. eco/J.own +o AHA I"i Yi" 4 /8" c* M iYie"5 3/4. 7.1. 4 SPECIFIC ACTIVITY The limitations on Secondary Coolant System specific activity ensure that l - the resultant offsite radiation dose will be Ifatted to a small fraction of 10 CFR Part 100 dose guideline values in the event of a steam line rupture. This dose also includes the effects of a coincident 1 gpm reactor-to-secondary tube leak in the steam generator of the affected steam line. These values are I i consistent with the assumptions used in the safety analyses. .m 03/47-2y/*"N0R INFORMATION ON = ( MLE - UNIT 1 1 = h

QUESTION: 8.08(a) COMMENTS: The reactor trip uccurred because of I&C shorting RTB U.V. coil. A reactor trip signal is then produced due to a turbine trip and a negative rate due to rods dropping into the core which should have tripped RTA. If RTA remained closed than this becomes a condition II event. In addition the failure of this intercept valve is a tech spec item which provides protection to safety related components. RECOMMENDATION: Answer should be Condition II

REFERENCES:

Operation Admin Procedure 10006 Tech Specs 3/4 3.4 turbine overspeed protection and bases

Vogtis Electric Gener: ting Pfark -g g, 10006-C T ld.Y.IGdAsu NUCLEAR OPERATIONS h 1 9/stes una Georgia Power m.* connon 1 of 19 i y REACTOR TRIP REVIEW d U 1.O PURPOSE This procedure provides instructions to collect and information concerning Reactor document pertinentto determine their cause, and to--ascercain the

Trips, status and proper functioning of safety-related '

i' equipment necessary to make the determination that the affected Unit can be restarted safely. / 2.0 DEFINITIONS 2.1 Condition I The cause of the trip is known and has been corrected: all safety-related equipment functioned prope.rly during ( I the trip. j 2.2 Condition II The cause of the trip is not known and/o'r some safety-related ec.uipment functianed in an abnormal or 9 degraded manner during the trip. 3.0 INSTRUCTIONS that may be used as a Figure 1 provides a flow chart general guide for completing a " Reactor Trip Report", the 1, and making the determination that Data Sheet affected unit can be restarted safely. a n ? 0 ja a a l 'i j i FOR INFORMAfl0N>0NLY s 1 F03101A Wim t~ ~ - ' ' ~ " ' ' ' - - -. - _., _, _ _

gi y

3CEDURE No.

RE!V SicN j PAGE No. i 16006-C 1 3 of 19 i r i 3 b. Is it is adeeptable to restart the reactor? { i

They shallLlook beyond.the obvious indications to it n

/ ~ O diagnose the cause of the event and evaluate the s plant. response. They shall review available O inforsstion throughly, looking for: (1) abnormal !9 indications o:e degraded trends in equipment J norail or i.nticipated sequence, g out of the ]h parformance, (2) events occurrin (3) failed or u degraded response of equipment to control signals, l, "(4) unusual chemistry results or radiation l readings, and (5) unanticipated alarms. i 1 1 3.3.3 Reactor Trip Personnel Statementa, Data Sheet 2 should D be completed br the, persons involved with the trip and O incluaed with# he Reactor Trip Report, Data Sheet 1. t Reactor Trip Personnel _ Statements may be completed by c0 individuals or groups of individuals as long as it promotes an accurate event raconstruction, p 3.3.4 (The Shift Technical Advisor and the On-Shift Operations t l Supervisor are reaponsible for event classification (the Operatibus S'.arerintendent will resolve 4 t Ldisagreements). Fer Condition I trips the On-Shift l N Operations Supervisor has authority to restart the O reactor. For Condition II trips only the General l O Manager, or his designee, gan authorize reactor Q rastart. 3.3.5' If the cause of the trip or significant aspects of the trip recoveryfare not understood, a more thorough investigation must be made>until the cause of the trip -is known~or has been investigated to the fullest reasonable extent before permission to startup shall be authorized. 'if the cause is not known, only the N General Manager, or his designee, can authorize unic O .startup. The On-Shift Operations Supervisor has the D authority to contact other departments to aid in the D { investigation. N NOTE L O The STA and On-Shif t! Operations l Supervisor may change their initial classific'ation of~a i reactor-trip if mere JLnformation becomes'available. O Oo to 1 0 1 p FOR INFORMAT10NDONLY w _ ~ _

I' 4oCEDURE NO. " E" " 5 of 19 10')06-c 1 l Oc urs 1 V i 1 i y [ STA i I 9F \\ x le e STA and 0503 3 i Event Condition I a w Condition !! Classified By Q Event STA and OSOS CV'"8 Q i 1 P t 19 0 General nager ~ PRB Review ' O C5 1 r 7 Plan Cond ion )e 3 9F Ene'ra! Manger REACTOR TRIP PIPORT FLOW CHART Q Figure 1 FOR INFORMATION ONL) s

40CEDLtRE No., Revision 10006-C 1 "* " " ~ 7 d M i Sheet 2 of 11 DATA SHEET 1 ) s b. Off Normal Status Of Plant Systems Of Safety Systems: l i l l l l '?l l! L c. Tests and Surveillances in Progress: }; d. Operations in progress at time of Reactor Trip: I I 3. POST TRIP CONDITIONS } a. RPS Operation First out annunciator List of RPS channels actuated l [ t C FOR INFORMATION ONLY a l L

UGcEDURE No. REWSloN NE No. ,, 10006-C 1 8 of 19

• 0 Sheet 3 of 11 DATA SHEET 1

-l Did the Reactor Trip result from an automatic or O manual trip? (Circle one): Auto Manual i f Comment: i l t Did all Reactor Trip Breakers open, and all rod banks ~. fully insert? (Circle one) o Yes No If no explain: f l l Did RPS channels actuate conservatively with respect to th'eir intended set points? (Circle one) l s Yes - No If no, explain and describe instruments: fd O f Based on available information and above evaluation did the RPS function correctly? (Circle one) QD Yes No O N I If no, describe corrective action required and I reference any supporting documentation: p 1 o 0 o ? O i i FOR INFORMATION ONU ~

---

w_x

. _zw._ d s

INSTRUMENTATION 3/4.3.4 TURBINE OVERSPEED PROTECTION LIMITING CONDITION FOR OPERATION 3.3.4 At least one Turbine Overspeed Protection System shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTION With one stop valve or one governor valve per high pressure turbine a. steam line inoperable and/or with one reheat stop valve or one reheat intercept valve per low pressure turbine steam line inoper-able, restore the inoperable valve (s) to OPERABLE status within 72 hours, or close at least one valve in the affected steam lines, or isolate the turbine from the steam supply within the next 6 hours. b. With the above required Turbine Overspeed Protection System other-wise inoperable, within 6 hours isolate the turbine from the steam j supply. \\ l l SURVEILLANCE REQUIREMENTS 4.3.4.1 The provisions of Specification 4.0.4 are not applicable. 4.3.4.2 The above required Turbine Overspeed Protection System shall be main-tained, calibrated, tested, and inspected in accordance with the VEGP Turbine Overspeed Protection Reliability Program. Adherence to this program shall demonstrate OPERABILITY of this system. The program and any revisions should be reviewed and approved in accordance with Specification 6.5.1.6.

• Revisions shall be made in accordance with the provisions of 10 CFR 50.59.

l

• Not applicable in MODE 2 or 3 with all main steam line isolation valves and associated bypass valves in the closed position and all other steam flow paths to the turbine isolated.

67 V0GTLE - UNIT 1 3/4 3-84 \\

i f INSTRUMENTATION j BASES 3/4.3.4 TURBINE OVERSPEED FROTECTION This specification is provided to ensure that the turbine overspeed protcetion instrumentation and the turbine speed control valves ar' OPERABLE e and will protect the turbine from excessive overspeed. Protection from turbine excessive overspeed is required since excessive overspeed of the turbine could gen 3 rate potentially damaging missiles which could impact and damage safety-related components, equipment or structures. i I' ~ ~ YOGTLE - UNIT 1 B 3/4 3-7 /FOR INFORMATION ONLY

e .e

s.

.e. . QUESTION: 8.08B COMMENTS: The -response to the question is a double jeopardy question.. Depending on how the examinee answered part "a" determined his response to part "b. RECOMMENDATION: If Condition I was selected then allow.OSOS for part "b". If Condition II was selected then allow Vogtle General Manager..

REFERENCES:

OPS Admin Procedure 10006 t 1 I t l l l I l i I I l

m-s.

• .s.;

yo QUESTION: 8.14 COMMENTS: There is only one condition that allows the OSOS oj; the department supervisor to-release a subclearance when the subclearance holder is not on site. That is when the subclearance holder is contacted by phone and gives his permission. RECOMMENDATION: Accept " contacted by phone" as.a complete answer _and disregard other student answer (s).

REFERENCES:

Plant Admin Procedure 00304-C' 4 i e d i a I i 6 i l L l l

{ Qy > ?, Q.[" *j 7' f f 4 $3 j. .g, .J;6;

j l}ENO,f4joff35.!h 7

PAG REVislON. m, l 00304 kql(Qg l7

f,(

i } h d. N u 'l 1 i. p ~p TAGGING: DESK;UN17 > { .o o t

t.

'I Alocationhiniornear,theControlRoom,underthe3 -~ ~ direction of.the Shift Supervisor, where active CLEARANCE books, UNIT CLEARANCE LOGS, and.other related ( -forms are'kept.:.1 i TAGGINGDEbK0PERATOR, i A' qu'a'lihieb dke$ationsiDepar'ement employhwho j J coordinatesiactivities: as described in this procedure. ~ UNITCY.hRINCh$0G k:- ,.,.i ,n :300. An index of CLEARANCES (Figure 4) which contains the following informations. CLEARANCE NUMBERo Equipment b3-- Description, Breaker Identification Number, Type of l Lc-Protection, Afforded'by the Clearance, Installed and Released Dates, and any QUARTERLY CHECKS. f RESPdNSIBILITIEb 5 RESPONSIBLE, POSITIONS FOR RELEASE OF CLEARANCES l a m... u > t m.a General Manager, Manager Unit Operations and che L ManagarjTest and Outage u r- ~ TheiGeneral, Manager, the Manager, Unit Operations, or t the Manager, Test and Outage is responsible for 4 i releasing (,SUBCLEARANCES if the SUBC1EARANCE HOLDERScannot. be con J individual when he/she returns to work. q e n..o n, 4 a s ;w i.. ith 6 : Unitt I or:tUnit;II Field Construction Managers / Project o $ !.j - k Construction Manager 7, f@'DurihghhE!InitialTestProgramtheUnitIorUnitII Field Construction Managers or the Project Construction a l 1 e 14 Manager l1s; responsible for releasing subclearances of ,;l;, construction personnel,.if the SUBCLEARANCE HOLDER can .y not be gontacted and,providing notification to that a [., individuala hen he/she returns to work. w n ( TSiNdb$tf0haki$nh'SUjefyhs%r3050S1Srth'e( h ~ d 3 , 'iDepartment,Supervisorjmay release a.SUBCLEIRANCE if the ( , SUBCLEARANCE HOLDER is off-site, and gives permission d by. phone.g ThSOSOSM$kbepartmentSupervisormayreleasea 4 SUBCLEARANCE if the SUBCLEARANCE HOLDER is off-site, and cannot be. contacted. They are also resaonsible for providing notification to that individual w hen he/she l returns to work. i $0RINNdRMATION 0Ri i rA

T^.T. 4 c. .,s 's ., e v ' QUESTION: '8.16 COMMENTS: Procedure 00052-C list six (6) items that constitute a procedure intent change. RECOMMENDATION: Accept any 5 of the following: 1. Change in purpose 2. Change in scope 3. . Reduction in controls established in plant admin proc. 4. Reduction in level of safety as addressed in Tech Specs and FSAR 5.- Change in acceptance criteria that is less conservative 6. Change to plant administrative procedures

REFERENCES:

Procedure 00052-C, Rev. 2 l L

,, v f.,* bpprov._I Proc:' dura No. i i Vogtle Electric Generating Plant b1 ' Mh, NUCLEAR OPERATIONS Rev si n No L oate 2 (> 5/26/g6 Georgia Power twt c-n ease No TEMPORARY CHANGES TO PROCEDURES 1;i n M i l 9 ! OCT . :lt W1 UI % -n\\ 1.0 PURPOSE 1.1 This procedure describes the method to control, request, review, and approve temporary changes to Vogtle Electric Generating Plant (VEGP) procedures (TCP). 2.0 DEFINITIONS 2.1 TEMPORARY CHANGE A procedural modification which does not change the original intent of an existing procedure and which is intended for temporary use. () 2.2 PROCEDURE INTENT CHANGE A proposed change that involves any of the following: l 2.2.1 A change in the purpose of the affected procedure. l { 2.2.2 A change in the scope of the affected procedure. l l 2.2.3 A reduction in the administrative control established in the Plant Administrative Procedures, i 2.2.4 A reduction in the established level of safety of the plant as addressed in the Technical Specifications and the FSAR. 2.2.5 A change in acceptance criteria that is less conservative than previously established. 2.2.5 'A change to Plant Administrative Proc d e ures. i l !b ,,,,,4 ew $ % = a e6 Jew as.a=*- 4}}