IR 05000272/1987026
| ML20236P516 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 11/09/1987 |
| From: | Keller R, David Silk NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20236P499 | List: |
| References | |
| 50-272-87-26OL, NUDOCS 8711180035 | |
| Download: ML20236P516 (64) | |
Text
{{#Wiki_filter:1 , c ' V. S. NUCLEAR REGULATOR COMMISSION REGION I l OPERATOR LICENSING EXAMINATION REPORT j EXAMINATION REPORT NO.
87-26 (OL) FACILITY DOCKET NO.
50-272/311 FACILITY LICENSE NO.
DPR-70/75 LICENSEE: Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, NJ 08038 FACILITY: Salem Units 1 and 2 EXAMINATION DATES: 9/15/87 - 9/17/87 boaA L //[l/7 7 CHIEF EXAMINER:
' D' te 0. Silk a Operations Engineer (Examiner), DRS APPROVED BY: ///7 /f7 R. M. Keller, Chief 'Date PWR Section, DRS SUMMARY: Written and operating examinations were administered to four instant Senior Reactor Operator candidates. All candidates passed "ooth portions of the examinations.
8711180035 871112 PDR ADOCK 0500
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REPORT DETAILS Replacement Exams Exam Results: SRO Pass / Fail l
Written Examination 4/0 Operating Examination 4/0 Overall 4/0
1.
Chief Examiner at Site: D. Silk 2.
Other Examiners: G. Weale (Sonalyst) 3.
Personnel Present at Exit Interview NRC Personnel D. Silk, Operations Engineer (Examiner) K. Gibson, Resident I-nspector Facility Personnel J. Zupko Jr., General Manager-Salem Operations J. Gueller, Operations Manager Salem D. Perkins, Station QA Manager G. Roggio, Station Licensing Engineer L. Catalfomo, Assistant Manager Operations Training J. Lloyd, Principal Training Supervisor-Salem Operations P. Lander, Principa; Training Supervisor-Salem Simulator 4.
Summary of NRC Comments made at exit interview: The NRC expressed appreciation to the training and operations departments for providing assistance in expediting the examination process. The NRC reviewed the number and type of examinations conducted during the week.
The NRC then presented the observations made during the week: 1) No generic weaknesses were noted during the operating examinations; 2) the candidates appeared to be knowledgeable about plant systems and procedures; 3) the candidates were fluent at performing board manipula-tions: 4) the roles of the candidates in the simulator were better defined than was observed during the June requalification examinations.
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JSummary. of' facility c'omments made at exit' interview: . The operations department expressed: appreciation for'the examiners -- [ . conducting the ' plant walk through p'ortion of the examination in such L a' manner!th'at:had minimal impact upon the operations crew.
Concern was expressed'regarding the length ofLthe simulator exam- - ination which was. conducted from 8:00 a.m. to 6:30 p.m.
The response to this is tha't the NRC was trying to accommodate the licensee's.
' assignment.of control room responsibility and th.e training depart-ment's request to exam the candidates within the team they'were . trained.
Thus, the. scenarios lengths and rotations were established accordingly.
6.
Examination Review: A review of the written' examination was conducted immediately following the examination'. Comments were resolved during the. examination review and Attachment 2 details the modifications to the answer key, 7.
Summary of-generic strengths or deficiencies noted from grading the written examinations- ! This information.is being.provided.to document areas of weakness which should aid the' licensee in upgrading replacement training' programs.
No reply is required.
Candidates did not know the effect the DROOP circuit would have on-l - ' diesel generator output voltage when reactive load increases during parallel operations.
' Most candidates thought that diesel generator load is adjusted by the - speed control when the bus is isolated.
- Candidates were unable to provide all system responses to a pressurizer level channel failing low.
Most candidates, with a list of reactor vessel components, were - unable to identify the correct sequence of re-assembly following refueling.
Candidates did not know when a soak time would be required per . - 2-EOP-FRTS-1, Response to Imminent Pressurized Thermal Shock ! Conditions.
Candidates were not fully aware of actions and reports required in - the event the plant exceeds a safety limit.
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NRC concerns 1regarding Emergency Operating Procedure usage.and format.
The'following are instance's of. candidates'not following procedures by taking premature actions and where procedural. steps were not correct under all plant conditions.
These examples'were-noted'from only four' simulator scenarios used to test the candidates.
These instances should be used as examples'when~ conducting future operator t'aining and the correctness of r the Emergency Operating Procedures to combat various casualties.
In one scenario, a steam generator tube rupture was in effect.. The - , shift supervisor correctly diagnosed the condition and ordered a reactor trip and safety injection when pressurizer pressure and leve11 were decreasing in'an uncontrolled manner. When the candidate came to step 36 of 2-EOP-TRIP-1, which asks "Is SG Blowdown Channel 2-R-19 A through D in. Warning or. Alarming?", he answered the question "Yes" when in fact no channels ~were.in Warning or Alarming.
By answering the, question "Yes".the. candidate was able to transition to 2-E0P-SGTR-1, Steam Generator Tube Rupture.
The licensee justified this' action by-stating that the operators have been trained to use discretion in following the procedures so that if the operators are certain of the casualty in affect,_they may maneuver through procedural steps in. order to transition to the desired procedure.
Even though the ~ candidate went to the correct procedure, he-violated a procedural step to do so.
In one scenario, a main steam line rupture occurred in containment which- - resulted in' cooling down the RCS to approximately 425 F.
When the shift supervisor came to step 24 of 2-EOP-TRIP-1 which asks "Is RCS Average Temp Stable at OR Trending to 547 F?" - he ordered the board operator'to " crack open the MS10s-(atmospheric relief valves)." This was a violation of. step 24 which directs-the operatorsEto "Stop Dumping Steam." The shift supervisor justified his action by stating that a small steam demand will lessen the swell effect of the reactor coolant when it heats up with safety injection adding inventory to the RCS.
Had the procedures been followed, the MS10s would have been opened several minutes later in step 12 of 2-EOP-LOSC-1 which states " Adjust intact'SG Atmospheric Relief Valves MS10s To Stabilize RCS Temperature." No adverse plant condition developod as a result of the candidate's order, however, he was in violation of procedure.
Step 14 of 2-EOP-TRIP-1 asks, "Are 2A Thru 2C Diesel Generators - Running?". If one diesel generator is out of service (as it was in two of the four scenarios), the above question would be answered "NO" since it asks if all diesel generators are running and thus time would be wasted as the operator performs steps to start a diesel generator that should have started.
The question appears to be asking if available or operable diesel generators are running.
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.5 i .. . Step 36 of 2-EOP-TRIP-l' asks, "Is SG Blowdown Channel 2-R-19 A- - Through D in Warning or Alarming?". If one channel were in Warning or Alarming the question would be answered "N0" since it asks if all channels are in Warning or Alarming and thus the transition to 2-EOP-SGTR-1 would be delayed.
The question appears to'be asking if any channels are in Warning or Alarming.
' Attachments: 1.
Written Examination and Answer Key (SRO) 2.
Resolution of facility comments from the Exam Review ! t
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NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION - F ACIL11 Y: _ SALEM _134_2_______________ l d REACTOR TYPE: _PWR-WEg4________________ DATE ADMINISTERED: _87/@9/15________________ . , EXAMINER: SILK _D.
CANDIDATE: _ _ {$_ __h_CO___________
i JNSIBUgIJONS_Ig_CBNDJDSIE1 l m Use separate paper for the answers.
Write answers on one side only.
' ' Staple question sheet on top of the answer sheets.
Points for each question are indicated in parentheses a+ter the question.
The passing grade requires at least 70% in each category and a final grade of at
least 80%. Examination papers will be picked up six (6) hours-after the examination starts.
. ' ' % OF ATEGORY % OF CANDIDATE'S CA1EGORY _, _MOL, lie _ _IDIGL ___.SCOBE__._ _YBL U E _ _ __ _ _ _ _ _ _._ _ _, _ _, C @IE G O R Y _ _ _ _ _ _ _ _ _, _ _ _, , h 00__ _2Dr@9 _ _ _ _.. _ _ _ _. _. _ . - _ _.. _ _ _ _ 5.
1HEORY OF NUCLEAR POWER FLANT
OPERATION, FLUIDS, AND ,y THERMUDiNAMICS 25a@9.__ 2h 99 ___ ________ .____.____6.
PLANT SYSTEMS DESIGN, CONTROLS AND INSTRUMENTATION . 25 9@___ __2M199 _ _ _ _ _ _ _ _ _ _.. __________7.
PROCEDURES - NORMAL, AbN~jRM AL, EMERGENCY AND RADIOLOGICAL COMTROL .2Dz99__ _29299 _ _ _ _ _ _ _ _. _ _ _ _ _ ___.._____ 8.
ADMINISTRATIVE FROCEDUHLS, CONDITIONS, AND L1MITAT]ONS 199 @9__ ___________ ________% Totals Final Grade All work done on this examination is my own.
I have neither given nor received aid.
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,e- . . NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS
.,, . Jug-i ng the administration of this examination the following rules apply: J.
Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
I ?. Restroom trips are to be. limited and only one r.andidate at a time may leave.
You must avoid al1 contacts with anyone outside the examination ! r'com to avoi d even the appearance or possibility of cheating.
' Use black ink or dark pencil only to facilitate legible reproductions.
,. 4.
Print your name in the blank provided on the cover sheet of the examination.
J.
Fill in the date on the cover sheet of the examination (if necessary).
Use only the paper provided f or answers.
. Pr i nt your name in the upper right-hand corner of the first page of tac i: . section of the answer sheet.
3.
Consecutively number each answer sheet, write "End of Category __' es j appropriate, start each category on a gym page, write only og opp sida of the paper, and write "Last Page" on t.n e last answer sheet.
+. . Number each answer as to category and number, for example, 1.4 6.'
10. Skip at least h +i agg lines between each answer.
' :f. Separate answer sheets from pad a n c' place finished answer sheets face down on your desk or table.
12. Use abbreviations only if they are commonly used in f aci l i t y 1 i,_t eratar p.
_ 13. The point value for each question is indicated in parentheses after the question and can be used an a guide for the depth of answer required.
.1 4. Show all calculations methods, or assumptions used to obtain an answer to mathemat3 cal prcilems whether ind1cated in the question or n o t., 15. Partial credit may be given.
Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
i6. If parts of the ex ami na t i on are not clear as to intent, ask questions cf the EEarntraer only.
17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in . compl eti ng the ex ami nat i on.
This must be done after the examination has been completed.
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,_ _ _ _ _ _ _ _ . _ _. _. _ _ _-_-__ - - _ -_ _ -1 ' A1 .. ,. d, -.4 ~ x8.DWhen' you complete your examination, you shall: ' Assembl e 'yourz examination as follows: a.
(1); Exam ouestions on top.
i I (2) Examfaids - figures, tables,.etc.
f 'I-(3) Answer'pages including. figures which are part of the answer.
j ' .1 b.
Turn in your' copy ofLthe. examination and all pages used to answer.
j the examination questions.
J c.
Turn'in all ' scrap paper 'and the balance of the paper that yeu-did-not use for answsring the questions.
i d.
Leave the examination area 3 as defined by the' examiner.
If after ] leaving,.you are found in this area while the examinationLis still
i n 'progr.ecs, your license'may be denied or revoked.
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- QUESTION 5.01 (1.50)
How will each of the following affect the results of a secondary calori-metric power calculation? Limit your answer to. CALCULATED LOWER THAN ' - ACTUAL, CALCULATED HIGHER THAN ACTUAL, or CALCULATED THE'SAME AS ACTUAL, Consider 1each case separately, a.
. Measured feed water temperature is 10 degrees lower than actual feed water temperature.
b.
_ Measured. steam generature pressure is 30 psig lower than actual-steam generator pressure.
c.
Measured > feed water flow is 1E5 lbm/hr hig"er than. actual feed water flow.
QUESTION. 5.02 (2.50) A centrifuge 1 charging pump is running with the discharge flow contral valve in mid position.
Indicate how each parameter will change (Increase,
- Decrease, or Remain the Same) if the discharge valve is fully opened, a.
Discharge flow b.
Pump discharge pr essur e upstream of the discharge valve c.
Motor amps d.
'Available NPSH to pump e.
Seal injection flow (Assume seal injection flow control valve does not change position) ' DUESTION 5.03 (2.50> How do each of the f oll owir a parameters change (Increase, Decrease, or .No change) if one main steam isolation valve closes with the plant at 25% load.
Assume all controls are in automatic and that no trip occur s.
a.
Affected loop steam generator level (initial change only) b.
Affected loop cold leg temperature c.
Unaffected loop steam generator level (i ni ti al change only) . f d.
Unaffected loop steam generator pressure e.
Unaffected loop cold leg temperature i
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N , - + n , - .o .r . (;GIUESTION:I[5[04) . (2.00) S '[a f (Compare' andlexplain.. the dif f erence in the reactivity worth' of a-rod thatJ i" gfik droppediwhileMat. power-to the reactivity worth-of.the same rod stuck , ", 04.; outTwhil e'all Lthe'.other. ' rods',are i nserted.. m; ~ t i 9 ..k'. ' (QUESTION '5105L L(2.10) y :9.; km;j ; Consider the following plant. conditions:
, N, ' iPower'= 100%: ' . h ., E ' iTave "=~571.5 F' W ~l T s t m e. = 513'.8 F: ' y , .h,rS 4, a CALCULATE 2 the - new st eam: prensure if 5%. of all.of the steam generator . ' itubesLare plugged and.the. plant is returned +to 100% power with Tave.o?.
, (571.5 F.- Stateiall assumptions, show all work and equations'used.
' , ;' < > \\ '. \\ RQUESTION ;5.06-3(2.40) < I ?! ., ::. the' fallowing: parameter changes affect DNBR 'How:does each'ofm < . g*n (Increase,.. Decrease?ar Remain the Same)? Briefly explain your answers ' c in. terms of J mar.gi nL to seturation.
+ + .a.
Pressurizer Temperature. Increases S degrees b.1 Grid,freqencyl decreases;to-58.8=H7.
, ac.--AFD' increases'to.+10% %,, ' Qi; 10UESTION 5.07- -(2.50) ?A' reactor at BOLJis' critical rt 10E(-9) amps.
Rods are withdrawn at 30 ,
- steps'per~ minute for 10 seconds in bank sequence.
+ a'. What is.tno :SUR 'immediately af ter the rod motion stops? Assume di+f- .- erentialtrod, worth 2s 20 pcm per step, effective delayed neutron
, f raction i s ' O.006'.and.-the ef f ective precursor decay const ant is ' 1.
- 0.1. set-1.
' State all' equations.used'and assumptions made.
(1.5) m 'bl ' . W ib.
1How end'why_would.the SUR change, if at all, if the initia] condit-i ions were at.EOL7 No calculations are necessary.
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OUESTION-5.08 (3.00) Listed below are conditions which will affect a f uel load 1/M pl ot. For each condition listed, INDICATE if predicted criticality will occur with FEWER or MORE assemblies than woul d actuall y be needed to go cri t i- ' cal and state a brief reason for each' answer.
a.
A fuel assembly is loaded near the detector.
l J b. The detector is too far f rom the newl y installed f uel. 'i c. The detector is too close to the source.
) d. The detector is too far from the source.
OtJESTION 5.09 (3.00) ! l Consider the following plant conditions: i l MODE 3, BOL Baron concentration is 900 ppm All shutdown banks withdrawn ! Actual reactivity present in the core is minus 4% delta-K/K Source range indication of 100 CP5 Ditf erenti al boron North is minus 10 pcm/ ppm f i A baron dilution to 750 ppm increases the source rance indication to J-132 CPS.
During the dilution, Xenon concentr at i on changes.
What was { Xenon's reacti vi ty contribution during this time? !
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y , .. . -QUESTION 5.10l (3.50). Assume Unit'1 has just tripped after operating'at.75% power for three days.
The control rods were at 154 steps on Bank.D, the boron concen-
tration~ remains constant during the event at 1200 ppm,. temperature is 'being controlled at 547 F,.and the core is near the beginning of cycle.
Using the graphs provided, answer the f ollowing questions.
State all
- assumptions and show all calculations.
a.
What is the actual margin (PCM) by which the reactor is shut down? (1.5) b.
What is the numerical difference between the margin by'which the reactor is shut'down and shutdown margin as defined in Technical" Specifications? (0,5) , o e f Q4Cp c.
ow niuch niakeup water must be added if the reactor is to be restarted 8 hours after the trip? Assume. control rods will be at at 154 st eps ~ on Bank D.
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_ 16? 1 PLANT'SY@TEd@_QE@l@N _QQNIRQL _8NQ_1N@lRUDENI611gN PAGE'
t t i,; i ! . OdESTION ' 6. 01 -- (2.40) j ' Answer theffollowing questions TRUE or FALSE regarding the diesel generators:
a.
In parallel operations, the DRODP circuit will act to maintain 1 generator output voltage constant when reactive load increases, b.
In parallel operations, if the voltage rheostat is raised to a higher value, the generator will pick up a larger share of the-reactive load.
'c.
If the diesel generator is carrying an isolated vital bus, the speed control is used to adjust bus load.
d.
With the synchroccope travelling slowly in the SLOW direction, the operator should correct the situation by adjusting the voltage-rheostat.
e.
The. Uni.t Trip Relay (DUTR) must be' Reset following a loss of offu to power to the-4160 V. vital busses.
f.
The f ollowing ~ trips remain operable after a diesel generator start from the SEC: Engine overspeed, Hi-lube oil temperature, Generator differential.
DUESTION 6.02 (2.50) Answer the f ollowing quest. ions TRUE or FALSE regarding the AFW synten a.
During normal operation, the AFW pump discharge headers should remain fl ooded to prevent runout, b.
The manual STOP and TRIP functions of the turbine driven AFW pump are disabl ed 2 4 a loss of 125 V DC occurs during pump operati on.
j l c.
The motor driven auxiliary feedwater pumps will not automatically J start in the LOCAL position.
d.
While in LOCAL control, all turbine and motor dri ven pun,p automat > c ] trips are still effective.
I l e.
Immediately after taking manual control of the steam generator inlet control valves (AF21s), with pump discharge pressure at 1250 psie, an AF21 will automatically close as a result of a downstream brent in the AFW line.
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.,,.. . QUESTION 16.03.
(2.00)
- During a Loss of Coolant Accident, all automati c saf ety-: i nj ecti on.
systems f unctioned properly, pressurizer level stabilized, and.RCS pressure stabilized at~1700 psig.
What.is the. approximate break flow rate?. Justify your answer and state your assumptions.
' - ' QUESTION 6.04 (2.40) . Answer the following questions regarding the radwaste systems: a.
Briefly describe how the VCT is used to degas the reactor coolant prior to refueling and what two gases are of most concern? (1.0/ b.
~If the radiation monitor R41A (f or the plant vent) signals en hig v l evel ' radi oac ti vi t y al arm, what eu,t omat i c actions, if any, occur?- f (0.4) c.
How and why will the RCDT pump discharge isolation valves, 2WL12,13, respond i' a SI signal end loss of offsite power si mul t aneousl y , occurs? ^ (G.4) '
'. DUESTION 6.05 (2.40) a.
How will containment f an cool er operation change, if at all, in a j response to a LOCA? ( 1.. ' ) j l b.
If automatic containment spray actuation is inoperable, what c ond i t i c., ' and coincidence requires menual containment spray actuation, and wh+t must be done to initiate containment spray? (0. 6 ) J
c.
What valve. realignments will occur when containment spray is actuatec? (0. 6 ) j i OUESTION 6.06 (2.40) e
- The plant is at approximately 10% power and all controls are in automat ic
except for rou control.
The controlling pressurizer leve] channel fails low.
Assuming no operator action, state how plant systems respond and specify-what reactor trip, if any, will occur.
State all assumptions.
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,. _ __ _ -- _ ,lbz_lEL6NI;5YSIEd5PEDIGN,_CONIggL,_eNp_IggIBUNgNIeIIgN- , k PAGE '8 .: , , ' '.+E , QUESTION 6.07-(2.40) L With the plantiat 75% power and all control systems in automatic, explain-the plant response to both VCT level control channels failing low.
Con- .tinue your' explanation until stable cor.ditions are reached or until: a reactor. trip occurs.
If the reactor will eventually' trip, state the - cause'of_the trip.
Assume no operator action is taken.
'OUESTION 6.08 (2.50) a.
State YES or NO as to whether the reactor will trip in response to a simultaneous failure low of both intermediate range channels during a reactor startup with power at 5%. Justify your answer.
fl.5) b.
Assume the reactor is at 100% power when one Intermediate Range channel fails high, followed immediately by a reactor trip (f rom other causes).
What additional action will have to be taken during emergency proced-ures to ensure proper operati on of the Nuclear Instrumentation Syster' (1.0) i OUESTION 6.09 (3.00) For each case beinw explein the'resulting method of reactor coolant test ' temperature control and indicate the approximate final RCS Tavg.
~ ' Assume all systems normal except as stated and no operator action.
Consider each. case separately.
Assume the pressure setpoints are for the Steam Dump system.
a.
The r ormal stean pressur e setpoint is reduced by 80 psi whilt in Hot Standby awaiting reactor startup.
b.
The Train A steam dump interlock pushbutton for 'OFF' is depreneu while at 5% reactor power awaiting turbine startup.
c.
Power is 100%, pressur e setpoint is raised to 1200 psig and Main steam pressure control is selected at the same time that a reactor trip occurs.
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6 ' QUESTION.~6.10 (3.00) . a.
The MSIV's'are closed on a main steam.line break to isolate the break and prevent blow down.of the other steam generators through the breet.
List two additional. reasons for closhng the MSIV's on a steam line . rupture.
(2.0) b.
The plant is at 100% power when a large main steam line break occurs downstream-of the MSIVs. 'What three safety injection signals.may.
actuate as a result of the steam li'ne break? (.l 01 l ! l (***** END OF CATEGORY 06
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+ 'BB91060ElcB6_GOUIE06 ) ! - ... QUc!STION 7.01 (1.00) 'Given the following' list', which sequence of events is the proper order . for reactor assembly-fallowing'a refueling? 1.. Reactor vessel headLreseated 2.
Control rod drive shz.fts connected to the spider assemblies 3.
Vessel cavity seal. removed 4.
Upper internals placed in the reactor 5.
New seal rings are installed on vessel' head a.
5,4,1,2,3 b.
2,4,5,1,3 c.
4,5,1,2,3 d.
4,2,5,1.3 e.
5,2,4,1,5 .DUESTION 7.02 (1.00)
What. is the reason for the. caution "Do not open 13 KV Disconnects 1T50 and 2T50" as stated in El I-4.9, " Blackout"? l DUESTION 7.03 (2.50) a.
An area'in.the auxiliary building has been r oped of f and postect with , a RADIATION AREA sign.
Five feet within the boundary is a valve ths/t ! produces a 2000 mrem /hr= field at 18 inches.
E:t p l ai n why or why net the area is properly posted.
( 1. U)
b.
A maintenance worker who is 24 years old and has a lifetime e>couure through'last quarter of 26 Rem on his NRC FORM 4.
He will be workinc
in an area that produces a E.6 Rem /hr field.
Assuming that the worke l has received no exposure thus far this quarter, what is the maximum time he may be permitted to wark is this area? Assume no emergt,ncy
condition exists.
(1.0) !
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7 .B8DI96901986_99dIB96 i-l; l
i ' QUESTION '7.04 (2.50) Answer. 'the' ' f ol l owi ng questions regarding OP II - 1.3.1, Reactor Coolant _ i LPump Operation.
i a.
With one or more of the RCS cold leg temperatures less than 312 F.
no RCP shall be started unless one of two conditions is met.
What i are these two conditions? (1.0' b.
Why must VCT pressure be maintained greater than 15 psig? (1.0) i c.
If it becomes necessary to open the No. 1 Seal Bypass Valve, why must all four Seal Return Valves (21-24 CV104) be opened first?(d.5) i 'OUESTION 7.05 (2,50)' < Answer the f ollowing questions regarding OF II - 6.3.2, Initiatino Resid- ! ual Heat' Removal.
I a.
'Why must the water level in the reactor vessel remain above the ret' er line of~the hot leg pipe while the RHR pumps are in operation? l (0.0) b.
Upon-starting an RHR pump while an RCP is running, why must the low pressure letdown control val ve (2CV18 ), be i mmedi atel y adj urt ed? (P. 79 c.
Why do the three CVCS Letdown Orifice Isolation Valves, (2CV3,4, ana 5) remain-open even though Ietdown Is from RHR? (O.7b)
i ,d.
W2th the RCS at 200 F, whv must the air suppl y be restored to the i RHR heat ex c h an g e> ' outlet valve (RH18), and the demand for RH1F j reduced to some minimum val ue? (0. 5 : !
j I
(3*444 CATEGORY 07 CONTINUED ON NEXT PAGE
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.1 \\ .. QUESTION.. 7.06 . (2.50) ) , ' Answer the'fc11owing' questions regarding EI-4.18 I, Complete Loss of- ~ Control' Air.
Assume the reactor ha-s already tripped.
a.
Why are the number.22 and 24 RCPs tripped? . ( 0. 5 ) k
i b.- With.the pressurizer level. at 92"/.,. expl ain why 'the running RCPs j
- SHOULD or SHOULD NOT be tripped if component cooling water i s. l ott.t c
'the thermal barriers.
(1.0) ! c.
.While maintaining the. steam generator levels,..why is the'AFW 23 I , l pump NOT used? ( 1. vu j l ! -QUESTION-7.07 . (2. 50) Answer,the following questions regarding I - 4.8,, Rod Control Svstem ) Malfunction.
'i a.
Following a decrease in turbine load or baron. concentration, what ) parameter changes and alarms would be lodicative of a failure of t h; rod control s y st ero in the Auto Mode' State four responses.
( 1. n, b.
Following a decrease in turbine load or baron concentrat1.on, c r cti . withdrawal stop accompanied by a turbine runback occurs as a cen-sequence of a failure of a control rod bank to insert.
What conm r-ion would cause the. rad withdrawal stop and the turbine runbecf , (0. 5: i c.
What are the l 'i rr i t s on Tave (high and low) whi ch requi re a reactc,r 'l trip? 13.e)
i
t
QUESTION 7.08 (2.50) I
1 ) Following a reactor trip, it is observed that two control rods have 4 all ed to insert into the care.
Explain how the operator would manipulate the I controls of the CVCR to perf or m an immediate boration for tr.i s si tuat y ca.
! i I I l (***** CATEGORY 07 CONTINUED ON NEXT PAGE 4****)
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58 pig 60GICAL_GONTROL
j . .. .OUESTION. 7.09 ' ( 3. 00)- During an emergency condition the STA reports the following: .' Core. Cooling - Orange Path
2.
Subtriticality - Yellow Path 3.. Integrity Orange Path '4. Heat. Sink - Red Path I a.- In what order should the above conditions be addressed? (1.0) .b.
Ten minutes later the STA reports that the above conditions still exist.except that Core Cooling.is on a Red Path.
What actions should j be taken and'why?' (1.0> j c.
With. conditions 1 through 4 present, what optimal or f uncti onal recovery. procedure should the operator fcllow if a loss of offsita power occurs and the diesel s i ai l to start? (b. E ' .d'. 'After a Peactor Trip, what two circumstances initiate monitoring the Status Treesr (O. M l OUESTION 7.10 (2.bO) Answer the following questions regarding the Emergency Operating Procedures.
a.
What is the criteria for tripping the RCPs? (0. S ) b.
How is.SI termination determined? (1.0) ) c.
How is natural circulation verified? (1.0) DUESTION 7.11 (2.50) Answer the following questions regarding 2-EOF-FRTS-1, Response to Imminent Pressurized Thermal Shock Conditions.
q a.
What are the r ed path conJitions for FRTS-19 - ( 1. 0 ) I b.
How is it determined if a soak tima it required? (0.t> l i c.
What restrictions, if any. are placed upon the performance of othe
procedures whil e prrf ormi ng a soak? (1.0i !
(44444 END OF CATEGORY O'7 41444)
_ _ _ _.. _ _ _ _ _ _ _ _ _ _ _ _ _ _
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i 92_iADMINISIBBIlYs_ES9GEDUBEgi_ggNp1IlgNg2_9Np_61dJISIlgNS L AGE '14 P , '
V
- s
[',
- DOESTION'- 8.01L (2.00)'
, L LIf.the specific ~ activity of'the.RCS is greater than 1.0 uCi/ gram dose ' ' equivalent I-131 for more than 48 hours during one continuous time inter-val, the plantLmust be placed'in at least hot standby with'RCS'Tave'less , .than'500 F.
What is the basis for reducing Tave to.less than 500 F7
- QUESTION 8.02'
( 2.' 00 ) State whether each of the following events requires a ONE HOUR notificat- ' ion.per 10CFR50.
Consider each separately.
' ' a.
.The plant is in'a. condition NOT covered by operating and emergency procedures, b.
The loss of the:offsite-notification system.
! I c.
A valid automatic initiation of the Reactor Protection System.
, d.
A shutdown was commenced-because the plant was in violatien of the Technical' Specifications.
. . . QUESTION 8.03 (2.50) A What actions and reports must be completed prior to allowing Unit 2 to j return to operations f ollowing a. lass of' f eedwater transient which reau2 ' 'l ed in'an indicated RCS pressure of 2775 psi g? I l l OUESTION 8.04 (2.50) a.
Who is in charge of issuing keys from the key cabinet? ( 0, ' )
b.
To what four duty stations are key rings issued? (1 0) I c.
Who should be noti fied in the event that a security key has been lost" ( 0. 93 ) d.
Is it permissible for an operator to lend the key ring to maintenance personnel in order to hasten the work being performed? ( O Sa
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) . i _____m______________-
c - [ - .,. ... . 'et_L89dlN1@lB611ME_BBQCEDUBESt_CQNDillQNS _@ND_61dlI@IlONS PAGE-15 u . m " % I QUESTION 8.05 (2.50) ' Answer the following questions'regarding Facility Staffing, i a.
.What restrictions are placed on the composition of the Fire Brigade? ( 1. 0.' b.
The plant is operating ' at 50% load when the on duty board operator (a reactor operator) becomes seriously ill and is sent to the hospital.
Explain what restrictions are in effect and what actions need to be taken, if any, in regard to shift staffing (1.5) -QUESTION 8.06 (2.50) a.
Is an on-the-spot change valid for both units? t Fi. 5 - b.
On-the-spot changen to procedures may be madre if what three prov2sinn are met? (1.0
! QUESTION 8.07 (2.50; a.
When is a Fire Protecti on impairment Per o.i t required to be issued and who is responsible f or notifying the Shift Supervisor regard 2ng i the items on the per mi t? (1.P) b., A maintenance supervisor, responsible for work in the Diesel Generator -{EK} area and the Nos. 21 and 22 Diesel Fuel Oil Storage Tanks and l Transfer Pumps {DC) area, established one individual to act as the ' fire watch f or continuous monitoring of both areas.
Explain why or why not this is adequate fire protetthon coverage.
(1.5) QUESTION 8.08 (2.50) The RCS is heating up at 50 F per hour with the RCS presently at 325 F.
Maintenance reports that Charging Pump 22 repairs will not be completed for one hour but that Charging Pump 21 1s operable.
Technical Specifirit-ions Action Statement allows 77 hours to repair an inoperable pump in Mode 3.
Assume Charging Pump 23 is inoperable.
What action, if any, should be taken and why? l ( l l l l (***44 CATEGORY 08 CONTINUED ON NEXT PAGE
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.4 i
L- . =.. ' ~ QUESTION 8.09 (2.50) The : plant' is operating at 75% power-and the latest leak rate data showc: i L , i' 13.2 GPM '- Total RCS. leakage rate-l 1.5 GPM' - Leakage. into the Pressuri zer Relief-Tank L 1.2 GPM- - Leakage into the Reactor Coolant Drain Tank ! 3.4 GPM - Leakage through 13SJ 156, loop 13, hot leg (Previous leakage rate was 1.6 GPM) 0.8 GPM - Total primary to secondary leakage , l 4'.2 GPM - Leakage past~RCP' seals ' l l What RCS leakage limits, if any, have been exceeded? Refer to attacheo-i Unit 1 Technical Specifications.
Show all work'and state all assumptions.
l QUESTION 8.10-(3.50) ! i a.
What is the f1INIMUM number of operable excore channels indicating l' Anial Flux Difference (AFD) outside the target band bef ore AFD i s ccv.- sidered outside its target band by Technical Specifications? b.
Assume the plant.is operating at full power and the AFD has been out- , side the target band for the last 5 minutes.
What are the TWO actions l; specified which you may choose between to meet the Technical Speca fi-cation regtd rements? I n r:1 ud e time 1 imitations.
( 3. lo > t l c.
Assume that it is 0310 on Ob/13/87 and the plant is presently .at~45% power.
Considering 'the AFD penalty history below, at what date and time may power be increased above 50%? Explain and show all work.
Assume no deviation outside the band after
0310 on 05/13/87.
TIME WENT GUT TIME BACK ' DATE OF BAND IN BAND POWER 05/12/87 0310 0318 85% 05/12/87 1557 1637 65% I 05/13/87 0148 0310 45% (?.E' I l'. .)
l e (***** END OF CATEBORY 08
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- - -_- _ . i
- g2;iTHEORY OF NUCLEAR POWER PLANT OPERATION _ELUlpS _AND PAGE 17'
R z
p IMEBdQDYUSDICS ! (
- ANSWERS -- ' SALEM 1&2-87/09/15-SILK, D.
.; i I ANSWER 5.01 (1.50)
a.
Calculated higher than actual CO.5] b.
Calculated higher than actual [0,5J c.
Calculated higher than actual CO.53 ! REFERENCE HTFF. Handout, Section III, Part B, Ch 2, pg 341 Heat transfer 191003 K 1.06 3.1/3.3 K 1.08 3.1/3.4 3.2 002 020 K 5.01 3. '2 / 3. 6 193007K106 002020K501 193007K108 ..(KA'S) I ANSWER 5.02 (2.50) i a.
Increase b.
Decrease c.
Increase d.
Decrease e.
" cr r [0,5 pts each] bM C REFERENCE HTFF Handout, Section III. Part B.
Ch 1, pgs 324-329 Components 191004 Pumps 1.05 2.3/2.4 1.06 3.2/3.3 ' 1.07 2.9/2.9 191001 Valves 1.03 2.7/2.9 191004K107 191004K106 191004K105 191001K103 ...(KA"S) { k s ANSWER 5.03 (2.50) ' a.
Decrease b.
Increase c.
Increase d.
Decrease e.
Decrease LO.5 pts each] ) REFERENCE ' ! HTFF Handout, Section II, Pcrt A.
Ch 3, pgs 99-119 > I 3.2 002 000 K 5.01 3.1/3.4 l l 5.09 5. 7/4.2 _ _ _ _ _ _ _. _
m ___ _
) r-5 } THEORY OF NUCLEAR POWER PLANT OPERATION _FLUIDSt_8ND, PAGE
u '1HE600Q1N8dlCS
. ANSWERS -- SALEM:12<2-87/09/15-SILK, D.
. 5.11 4.0/4.2 002000K511 002000K509 002000K501 ...(KA'S) i I
' ANSWER 5.04 (2.00) l The stuck rod would be worth more [0.53.
Reactivity worth is proportional to the relative flux squared [0.5]. For a dropped rod, the flux'is depressed adjacent to it [0.5J whereas if the same . rod was stuck out, while the others were inserted, it would be exposed to a much higher flux than the flux in the rest of the core [0.53.
-1 REFERENCE l Rx Theory Handout, pgs 202-204
'3. 1 000 003 EK 1.03 3.5/3.8 j 005 EK 1.05 J.3/4.1 i f LPD Session 35 # 3 000005K105 000003K103 ...(KA'S) l
1 ANSWER 5.05 (2.10) , 0 = U1 A1 (Tavg - Tst ml) U2 A2 (Tavg - 1 stm2) [0.5] =
U1 U2 ' = A2 =.95 A1 l (Tavg - Tetml) .95 (Tava - Tstm2) [0.53 i = (571.5 - 513.8) .95 (571.5 - Tstm2) = 571.5 - 57.7/.95 = Tstm = 510.7 F [0.6] I From steam tables Pstm 748.5 psia [0.53 =
i REFERENCE HTFF Handout, Section II, Port A, Ch 3, pg 118 l 3.5 041 020 K 5.02 2.5/2.8 ! Heat Exchangers 191006 K 1.03 2.2/2.3 j 191006K103 041020K502 ...(KA'S) !
i l i l a ? , I a
l I i
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) i C
- IBE8bgDYN6biQg-
ANSWERS -- SALEM 18<2-87/09/15-SILK, D.
- .
j . ANSWER 5.06- . 2.40) ( J l a.. Increases [0.33.
As PZR temperature rises, so does pressure, hence l the margin <to saturation' increases E0.53.
b.
Decreases E0.33.
The Delta T across the core will increase to pro-
duce the same power, so Th will increase and the margin to saturation decreases E0.53.
.i c.
Decreases E0.33.
Since more power is being produced in the top hal+
of the core the margin to saturation decreases CO.53 ', REFERENCE i I i HTFF Handout, Section II, Part B, Ch 4, pgs.225-231 ! 3.4 003 000 K 5.01 3.3/3.9 003000K501 ...(KA'5) I I . l ANSWER 5.07 (2.50) a.
Rho = (10 sec) (1 min /60 eec) (20 pcm/ step) (30 steps / min) ,
= 100 pcm LO.SJ j SUR = 26 (lambda (rho)) / beta ef f - rho) [P.53 L = 26 (1E-4 ) (5E-3) 0.52 dpm E0.53 ! = b.
At EOL, beta effective is less CO.53 which gives a higher SUR E0.03 i REFERENCE' Reactor Theory Handout, pas 271-285
3.1 00'; 000 K 5.02 2.9/3.4 .5.47 2.9/3.5 LPD Session 43 # 3,4 44 ft 3a 001000K502 001000K547 ...(KA'S) I t I i l l i i i I ' j - _ _ _ - _ _ _ _ -.. .O
.. . ._ - _ - _ _ _ - _ - _ _ _ _ _ j 'TQi. THEORY CN' NQGLE68_EQWEB_EL@NI_QEE86TlQNt_ELylQgu_6NQ PAGE-20- -1HEBt!QQ1N@tilGg + ANSWERS -- SALEM 1&2-87/09/15-SILK, D.
V
1 I l ANSWER 5.08 (3.00) l . a.
(Loading aLfuel assembly into the core close. to a neutron detector j increases-the fraction of neutrons in the core reaching the detector.)
The detector count rate' increases more than the core neutron popul'ation increases-EO.53. Fewer assemblies would be 'needed' E0.253.
b.
The detector will not see neutrons until'there are a great number.
- [0.5] More assemblies would be 'needed' E0.253.
i i c.-The initial. count rate is too high and the detector is insensitive to core. changes. 00.5] More assemblies would be 'needed' E0.253.
l d.
The detector will,not see neutrons until there are a greater number E O. 53.
More assemblies would be 'neeG d' E0.25]. REFERENCE Rx theory Handbook, pgs.251-257 3.1 001 010 K 5.16 2.'9/3.5-001010K516 ...(KA"S) ' i ANSWER 5.09 (3.00) 'Keft1 w 1/(1 --rhol) ' Keffi 1/((1-(-0.04)) = 0.9615 CO.53 = CR1(1 - Keffi) = CR2(1 - Keff2) 100(1 . 9615) 132(1 - Keff2) Keff2 = 0.9709 E0.53 = ,
rho 2 = (0.9709 -1)/0.9709-0.03 EO.53 r delta rho = rho 2 - rhol = -0.03 -(-0.04) = 0.01 1000 pcm EO.5] = Baron delta rho = -150 ppm n -10 pcm/ ppm = 1500 pcm E0.53 Xenon. delta rho = 1000pcm - 1500 pcm = ~500 pcm E0.5] REFERENCE Rx Theory Handout, pgs 119,120,128, 253-255 3.1 001 000 K 5.28 3.5/3.8 '001000K528 ...(KA"5) , l _ _ _ _. _ _ _ _ _. _ _ _ _ _ _ _ _ _ _
.. ._ L "%_2ISE_QELQE_NQQLE88_EQWEB_E(@NI_QEE88I1QNt_E6Q1Qjit_8NQ - PAGE
ISEBUQQYM@dlQS
- -
' ANSWERS -- SALEM 1&2-87/09/15-SILK, D.
l .
& . i ) ,
- ANSWER'
5.10 (3.50) . a.
Power Defect (Fig. 2) +1150 pcm EG.53 I IRW (Fig.
4, 15) -3460 pcm (-3740 + 280) E0.53 , SD rod worth (Fig. 16) -2650 pcm E0.53 ' _ _ ___ _ _. _ _ _ _ -5020 pcm 'b.
Differs by the worth of the most reactive rod being withdrawn.
From Fig. 17: 710 pcm.
E0.53
~ c.
Aenen wur di sF2y.
10 0 0 ,u_ a, 44b4] ue ppm cnange = xenon wortn + rower-v e i e r_ t < Lii. bur uii Wur Ui ( r~ 29 ici i)ylcfC 0dh - - - 12 3 n,,, 7. ~. p a,:.' p p - ! , - ' ~ '~ p ; : 4@.t] c,11 a r, s or ernin. c ris. ic14 7cre ga m. '-a ! ~ REFERENCE i l Rx ENG Man, Part 3, Sec 3.7'(b).
p 5,
1 3.1 001 000 K 5.08 3.9/4.4 5.09 3.5/.~3.o
5.17 4.2/4.? j 5.36 3.1/3.4 J 001000K508 001000K509 001000K517 001000K536 ...WA'b) U
, _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ -. - - _ - _ - _ _ _ - - _ - _ - - - - - -
=?6 _JEL8NI_SYSIEU@_psSigN3_ggNIBgst_8ND_JNSIBydgNIBIlgN
PAGE
- ANSWERS -- SALEM'l&2-87/09/15-SILK, D.
.. l ANSWER 6.01' (2.40) a.
False b.
True c.' False d.
False' e.
False j f.
False [0.4 pts each]
l REFERENCE SNV.VII, Diesel Generator, 'pgs 32,35,36,51,52,53 y OP IV 16. 3.1,. Emergency Power - Di esel Operation, pgs 3,4
3.7 064 000 K 4.02 3.9/4.2 K 4.06 2. 2 / 2. ~/ j A 2.02 2.//2.9 l A 2.03 3.1/3.1 I l LPO EDG # 1.3,1.7 J 064000A202 064000A207 064000K402 064000K406 (KA'S) ... i ANSWER 6.02 (2.50) a.
False b.
True c.
False l d.
True e.
False EO.5 pts each) i { REFERENCE SNV II, ch 11, pgs 23,24,29,30 3.5 061 000 K 4.04 3.1/3.4 K 6.01 2.5/2.8 { 3.7 063 000 K 2.01 2.9/3.1
PUMPS 191004 K 1.02 3.1/3.3 LPO AFW # 1,2.3,3/4.3,3/4.6 { 061000K404 061000K601 063000K201 191004K102 ...(KA"S) i l l l ! !
l.
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_ . _ _. -. _ ._____ -. ]6_286eNI_SySIgMS_pgSIGN_CgNIBg62_8Np_INSIByMENI9IION
2 PAGE
jANSWERS'-- SALEM 1&2-87/09/15-SILK, D.
l
, L l ! ) l i ' ANSWER 6.03 (2.00) Leakage would be the sum of the high head pumps discharge and positive I l displacement pump: [0.62 High head flow 2 X ~400 gpm = 800 gpm E0.63-Positive displacement pump 75 gpm C0.63 = _________ LEAK RATE 875 gpm (ce Yo0 (a J ftf v.n E0.2J =
- % t fu n e; t>3 )
ENOTE: acceptable value for high head flow is 300-500 gpm each pump] j REFERENCE SNV II, ch 10, pg 53, Table 2 (pg 86), Fig EC-3 3.2 006 000 K 5.06 3.b/3.9 1.03 4.3/4.6.
1.08 3.6/3.9 A 1.01 a.0/4.3 i LPO ECCS # 8.9.2 006000K103 006000K108 00600MK506 006050A101 (KA'S) .., l ! ANSWER 6.04 (2.40) a.
Open VCT vent to the vent header Raise VCT level to force gases out of the VCT Lower VCT 3 evel and fill gas space with. nitrogen Repeat until H-2 and Xe-135 concentrations are at desired level C[ Mh - Un$c3 > M' [0.4 pts each] b.
Isolates CNMT ventilation ""' t:
- P '
E0.4] .., c.
Closes because it is a CNMT isolation valve E0.43 ,. REFERENCE
- "#"*
YN 5 9 a I - " SNV V, ch 30, pgs 13,15,16
SNV lII, ch 17, pg 21 ! SNV V, ch 29, pg 14 i 4. 6 103 000 K 1.02 3.9/4.1 LPO LWDS # 5 K 4.06 3.1/3.7 RMS
- 1.2.5 3.11 029 000 K 1.01 3.4/3.7 CVCS # 18.2, 18.4 i
068 000 K 1.01 2.4/2.6 " 029000K101 068000K101 103000K102 103000K406 ...(ks"S) l
1 l ! I - _ _ _ _ _ i
-_ - - - -. _ - .- - -
. 1_jEL8NI_Sy@IEMS_DEg1GNi_CQUIBgk2_8Np_INSIBUMENIBIlgN PAGE '24 '6 . r j. ANSWERS -- SALEM 1&2-87/09/15-SILK, D.
I , i ' I e i I ' ANSWER '6.05-(2.40) ' a.
. Running CNMT fan cooling units trip All five cooling units restart (after a delay) in slow speed Service water flow increases (to 2500 gpm) to cooling. units CO.4 pts each3 b.
23.5 psig on 2/4 detectors Simultaneous operation of BOTH SPRAY ACT CONT ISOL PHASE B VENT ISOLATION key switches.
[0.3 pts eachl c.. Spray pump discharge i sol ati on valves (21CS2,22CS2) open-I Additive tank isolation valves (2CS16,17) open E0.3 pts each] j REFERENCE SNV II, ch 12, pgs 13,1fs 19,21
.3.6 022 000 K 1.01 S. 5/3.7-026 020 A 1.02 2.7/3.0 LPO CNMT Spray # 3.2,3,3 CNMT Vent
- 1.0,2.0 022000K101 026020A102
....(FA"5) ANSWER 6.06 (2.40) l PZR heaters trip off b ' val ve s cl oce b "2',, Letdown line inolation Letdown orifice isolation valves closel ' Charging flow increases D ~3 g,c VCT auto make up initiates at l J'" l evel I C l3 RCS inventory increase causing t he reactor to trip on high PZR level W if above P-7 (or P-10) CL L 1 LC.4 pia veto] l REFERENCE SNV VII, ch 45, pg 66 l 3.2 011 000 K 1.01 3.6/3.9 K 1.04 3.8/3.9 ' K 4.01 3.3/3.7 l K 6.04 3.1/3.1 , LPO PZR Level Control # 2.3,2.4,2.5 I I .011000K101 011000K104 011000K401 011000K604 ...'KA'S) i ! ! >
! > < l l.
. - _ - - - - _ -
- - _ _ 6; iPLANT SYSTEMS DESIGN _ggNIBgL _AND_ INSTRUMENTATION.
PAGE~ 25
3 a-. . , p yANSWERS--- SALEM 1&2-87/09/15-SILK, D.
F . l- .. I il
i ANSWER 6.07-(2.40) { < Charging pump suction switches to the RWST " Increased boron concentration in RCS causes rods to withdraw Tave starts to decrease after rods are f ull y wi thdrawn ] Stnnm hondnr nr ecen tro etnrte &m d ro r r n.m e m o, j Ge'ecr - l'~c cpr^ to try te -' nt r a l e ad -c l Reactor will trip f rom low pressurizer pressure as RCS inventory nhrinks ' L& 4 pts each] j O.b REFERENCE i SNV I,-ch 6, pgs 16,17, CV-6 SNV IV, ch 22, pgs 5.6, RS-11
3.1 001 000 K 4.03 3.5/3.8 LPO CVCS #12.1 ) K 1.06 3.6/3.7 l K 4.05 3.9/3.9 I 3.9 016 000 K 3.11 2.2/:1.2 I K 3.01 3.4/3.o K 3.'12 3.4/3.6 001000K405 001000K403 001000K106 016000K311 016000kJ1.
l ANSWER 6.08 (P.50) a.
YES LO.53. P-6 drops out (because JR 10 E-10 amps) LO.5] and SR NI reenergizes causing a l evel trip signal LO.5] l l b.
Manually reset SR t rip / block for both trains (to reactavate the SR high voltage and thus get Source Range ind) cation) L1.0]. REFERENCE SNV III, ch 20, pgs 9-21 3.9 015 000 K1.01 4.1/4.2 K3.01 3.9/4.3 i K4.01 3.1/3.3 K4.07 3.7/3.8 A2.02 3.1/3.5 LPO Excore N1 # 5,7,8 015000K401 015000K301 015000K104 0150004202 015000K407 - _ _ - _ - - - -
__- __ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ . _ _. _ _ _ _ _. _ _ _. _ _. + A _jEL8NI_SygIEUS_DESJQN _QgNIBg62_6ND_INSIBydENIGIJgN PAGE
2 . ANSWERS -- SALEM 1842-87/09/15-SILK,.D.
.. ) ANSWER.
6.09 (3.00) a.
The. normal steam pressure setpoint of.1005 psig maintains Tavg -~ at *547 F, a decrease in the setpoint to 920 psig would cause the dumps to open and' cool Tavg to '543 F where the P-12 interlock would close all steam dumps.
[1.0] b.
. Secondary pressure would rise to the setpoint of the secondary atmospheric relief valves which would maintain pressure at 1035 psig primary temperature 551 F.
01.03 x c.
.In pressure control mode, the output of the turbine trip and load rejection controllers is blocked.
Temperature would be controlled at 551 corresponding'to the relief valve setpoint.
li CD j ! REFERENCE Steam Tables SNV IV, ch 26, pgs 5, 15, SD-12 SNV V, ch 33, pg 13 3.5 041 020 K 3.02 3.8/3.9 K 4.09 3.0/7.3 l 039 000 K 6.01 2.1/2.4 i . Valves 191001 K 1.01' 3.3/3.4 LPO Steam Dump. System # 1.6,1.8,1.2 039000K601
) i l
h
l
i i ! l '1 L______.____.___
...- -. - --
i -f6t_2E66MI_SYSIEd@_QEgl@N _QQUIBQLu,_QNQ_lNEIBQUENIGIlgN PAGE-27 t l . ANSWERS -- SALEM 1&2-87/09/15-SILK, D.
, . , ANSWER 6.10 (3.00) a.
1.
Minimize positive reactivity ef f ects of RCS cooldown associ atert with the blowdown E1.0J 2.
Limit pressure rise within containment during a steam break in containment E1.0] b.. High steam. flow with low steam line pressure ' Low PZR pressure (SI) High steam line differential pressure E0.33 pts each] q e H,gi sh-Av - fh 4.w '7.q REFERENCE , SNV VII, ch 47, pgs 24-34 'TS_3/4.7.1.5, TS pg B 3/4 7-3 3.5 000 040 EK 1.05 4.1/4.4 LPD MSS
- 3.2 EK 2.01 2.6/2.S ECCS # 8/9.3 EK 3.01 4.2/4.5-EK 3.02 4.4/4.4 3.5 039 000 K 4.05 3.7/3.7 000040K105 000040K201 000040K301 000040K302 0390DOF406
...(KA'S) i ! - . ! l _ _ _ _ _ . _ _
Zi_iESQGEEV8EE_I_Ng8086t_6EUQBd86t_EMEggENGY_8NQ PAGE
88Q1QLQ@2 CAL CONTROL .,a ; ANSWERS)-. SALEM 1&2-87/09/15-SILK, D.
, ANSWER 7.01 (1.00) d'E1.0] ~ REFERENCE SNV VII, ch 43,.pg.9 3.11 034 000 K 1.01 2.5/3.2 K 6.01 2.1/3.0 LPO Rx Vessel & internals # 2.2,2.3 034000K101 034000K601 ...(KA's) ANSWER 7.02 (1.00) . Interlocks are such-that the Gas Turbine Freakers and 13KV breake rr-t"u s * be open to reclose IT50 and 2T50 E1.03 REFERENCE EI I-4.9, pg 4 / 3.7 000 056 EN c.02 4.4/4.7 i LPU 13 KV # 2.2.3 000056K302 ...(KA'S)
I l J ANSWER 7.03 (2.50) n i d (sq) l a.
I D(sq) i (2500 mrem /hr ) (2.25 f t sq)/(25 ft sq) = 225 mrem /hr [1.0] ! i = Area is NOT properly posted because dosage > 100 mrem /hr LO.5] b.
5(N - 18) = 5(24-18) = 30 Life time limit 30 - 28 = 2 Rem LO 5] = 2 Rem /0.6 rem /hr = 3. 33 hr s = 200 mi nutes E0.53 REFERENCE Radiation Protection Program Manual, pgs 6,18,29.30 10CFR20.201 10CFR20.203 194001 K 1.03 7.8/3.4 194001K103 ... (K W S ) ....._ _ _-__--_ - ___
'Zu_iES9REQQBEQ_;_NQBd8(t_6@NQBd8(t_EMEE@gNg1_GNQ PAGE
E60106951G86_kgNIBQL .- . " ANSWERS -- SALEM 1&2-87/09/15-SILK, D.
.- ANSWER 7.04 (2.50) 117.
a.
PZR vol ume l ess than "~. 2" E0.5] All SG temperatures < 50 F above any RCS cold leg temperature E0.51
b.
To ensure an effective back pressure on the RCP's No. & seal to maintain lubrication of thrt seal [1.0] N: L c.
To preclude possible cocking of the seals 00.5] REFERENCE OP II - 1.3.1, FCP operation, pgs 1,2 3.4 003 000 K 1.03 3.3/3.6 K 4.03 2.5/2.8 K 6.02 2.7/3.1 6 13 3.6/3.7 LPD RCP # 1.7,1.11,1.12 003000K403 003000K602 003000K103 003000G13 ... ( t1 A ' S i ANSWLP 7.07; (2.50) a.
Prevent RHR pump c ava t at 1 on E0.53 b.
To prevent RCS pressure from dropping below the minimum required for RCP operation [0.75] c.
To provide addi t i ona] relieving capacity in cace of an i nadve rt e nt RCE pressurizati on (0.753 d.
la prevent conldown 04 the FLE E0.5] REFERENCE OP II - 6.3.2 Initiating RHR pgs 2,4,5 3.4 005 000 K 1.04 2.9/3.1 K 1.09 3.6/3.9 G 13 3.3/3.4 LPD RHR tt 3.3,3.5,4/5.1 005000G13 005000K104 005000h.109 ... ( t.: A ' S )
- _ _ _ _ _ - _ _ - - - - - _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - '2A__EB99E996ES_ _U98U9k2_BENgBd863_EdgBggdCy_8Np PAGE
88DI969910BL_G991896 - ' ANSWERS -- SALEM 152-87/09/15-SILK, D.
. l l l l ANSWER 7.06 (2.50) a.
Reduces heat' input to the RCS [0.53 b.
The charging pump will be stopped at 90% PZR level, thus nt., seal in-jection flow E0.43.
With no CCW flow, the seal package could be dam-aged 4and RCS leakage could occur [0.43.
RCPs should be tripped LO.23, he,L~y necJ Isanny) c.
Using the 23 AFW pump niay reduce the pressure in the 21 and 23 EGs enough to cause a steam line differential pressure SI E1.03.
REFERENCE EI-4.18 I.
pgs 2,3 3.8 000 065 EK 3.03 3.7/3.9 078 000 K ?.0? 7.4/~.6 3.10 008 0:.80 h '. 01 3.4/3.5 . LPD Contral Air # 5.3 000065K308 000000K301 078000K302 ...(KA'D) ANSWER 7. 0 ^7 (2,50/ a.
Increasing Tave Tave-Tref Dev Al arm Hi-LD Tave Alarm 4. s, n. c Increasing PZR level E0,26 pts each3 A r >. 3.., f c. s,,,, b.
OT. del ta-T L LS. 51 Cf r/c k-T e-c.
581 F and 541 F LO.S pts each3 REFERENCE I - 4.8 Rod Control System Malfunction pgs 2,3,4 3.1 001 000 K 1.05 4.5/4.4 K 3.02 3.4/3.5 G 15 3.9/4.1 LPO Rod Contrel 4 1.4,1.6.4,17 001000G15 001000K105 001000K302 ...(UA'S)
_- . _ _ _ _ - _ _,
- ZL_LEBQGEQUBEQ_;_NQBd@(t_B@MQBd@Lt_EdESQENQY_ANQ PAGE
-- E82106001GGL_.G.QUIBQL ' .. & - ANSWERS -- SALEM 1&2-87 / 09 /15-S I L K, D.
.. I
ANSWER 7.08 (2.50) Start both. BAT pumps in Manual-fast (0. 0 7 CO 73 i Open rapid'baration stop valve (2CV175) u:. u ] C o ?) i Close. BAST'Recirc' valves (21 and 22 CV160) Fer67 (c. 7)
Maintain charging flow greater than 70 GPN [0.2J [vs] q Borate for 16 mi nutes en c, REFERENCE ) 2-EOP-TRIP-2 step 8 ) OP-II 3.3.8 Rapid Boration pg 2 ]
3.1 000 005 EK 3.01 4.0/4.3
EK 3.06 5.9/4.2 LPO CVCS #15.1 15.2 15.3
3
000005K301 ...(KA#5) ]
4 ANSWER 7.09 (U.00) { l a.
4,1,3,2 E1.03 l b.
The Core Cooling Red Esth thould be i mrned i at el y addreeted [ 0, 5: becaused it i s of higher priority than the Heat Sink Red P s.t h [0,S]. s c.
CC;. '.C [0.5) EC C-t c? '.-l d.
- i cct - ~ "r ~ ' ~ ' '- -_
- ' - r a,-, 9 aL m.- i u - E t-o '." -
m A~>altesio l i c tc+ [c i] g (n,W, v u,.,,) i t,, f., ch s. , REFERENCE (Usage Rules) ECA-0.0 ' 194001 Plant-Wide Generic A 1.02 4.1/3.9 LPO STA/SSE Emergency Responses N 1.2.1.3 .194001A102 ...(KA'E)
1 ' l-C__
_ _ _ _ _ _ _ _ _ _ _ _ - - -
- - - - - - - - - 'Zu_iC6QCEEUBEE_C_698U66t_6EUQBU66t_EdESGENCY_6NQ PAGE 32.
B601069EIG66_GQUIE96 -
- ANSWERS -- SALEM 1&2-87/09/15-SILK, D.
4 HO. r.h' c es..& L,-{c.rtn) M% S h.. h3 4. la ' T egl ANSWER 7.10 (2.50) I' 1 b e v O'
b M rh c I a.
PZR press < 1450 psig and At least 50 GPM BIT flow or 100 GPM SI flow CO.25 pts each] or Containment isolation ph'ase B [0.5] RCS subcooling(> 10 F) b.
RCS pressu(re stable or PZR level > 57.J increasing AFW flow (> 22E04 LB/HR)or OneSGNRlevel(>57) [0.25 pts each] c.
TCs stable or decreasing; WR hot leg temp stable or decreasing Intact SG pressures stable or decreasing Col d leg temp at saturation temp 4or SG pressure E0.25 pts each3 REFERENCE 2 EDP-TRIP 1 step 34, 40 '2 EOP-TRIP 2 step 24 3.4 000 017 G 11 3.5/3.6 3.3 000 009 EK 3.21 4.2/4.5 3.7 000 055 EK 3.02 4.3/4.6 LPD ECCS tt 10.1 RCP tt 3. 2 NC
- 1.3 000009K321 AN3WER 7.11 (2.50)
a.
RCS cooldown > 100 F in the last 60 minutes RCS press / temp to the left of limit A [0.5 pts each] b.
If RCS cooldown 5 100 F in any 60 minute period [0.53 c.
Nothing may be done to increase RCS pressure or decrease RCS temp 3r ature E1.0] REFERENCE 2-EOP-FRTS-1 step 23,32 Thermal Shock Status Tree 3.5 0C0 040 EU %. 04 4.5/ a Z__rE60tEQUEEQ_;_UQ8586t_8@NQBd86u_EMEBEENQX_GUQ PAGE
l ' *: 86019690lG86_GQNIBOL i
- '
.i ANSWERS.-- SALEM 1&2-87/09/15-SILK, D.
] +
l / LPO STA/SSE Emergency Response # 2.2 000040K304 ...(KA'S)
! i !
i i i l I I l ! ..--_--- -_- _.--__ _ -
, <-Lgz_rendidigIBellyg_EggCgnugget_C.guelIlgugm_eNQ_LIMlIGIlgNg PAGE
v, A fANSWERS -- SALEM 1842-87/09/15-SILK, D.
. ANSWER 8.01 (2.00) i < If a. tube rupture were to occur CO.5] a release will be prevented EO 53 since the SG atmospheric relief setpoint will be above the cc-responding . saturation. pressure for 500 F L1.0]. i REFERENCE TS 3/4.4.9 s 3.2 002 020 G 6 2.6/3.8 j LPO RCS # 12.1 .! 002020G6 ...(KA"S) l l
l l ANSWER' 8.02 (2.00) i i a.
Yes lo b.
Yes c.
No d.
Yes E0.5 pts each] REFERENCE 10CFR50.72 3.1 001 000 G 3 2.5/3.9 LPO Incident Reports # 3 001000G3 ...(KA'S) ANSWER 8.03 (2.50) Unit muwL ue p l ou_ c u tii iiu L m i cu i d u y vu u m i i
- i v uu
- 0.S vi i n W2Ui RC5 premmuu vu u u.. 1. nu L a ni tiu i
- .v u
LO.D3 e wi n: NRC is informed within one hour. LC.53 ECT 3 Safety Limit Vi ol ati on Report shall be prepared.
[0.
- U ] Report ~ is submitted to NRC and Upper Management (within 14 days) LC.L2- . REFERENCE Salem 2, TS 2.1.2 pg 2-1; T.S.
p 6-12a 3.3 010 000 G 3 2.5/3.6 G5 3.2/3.8 LPD Incident Report # 3.1 , ! l L- . - -. - - - - - - - -- - - - - - - - - --
_ _ _ _ - __ 'er _ rent!1NISIB611ME_EBQCEDQBEgt CQNDLIlQNSt_8UQ_ LIM 118IlgNQ' PAGE
.. .
- ANSWERS -- SALEM 162-87/09/15-SILK, D.
- -
RCS
- 11.2 010000G3 010000G5
...(KA'S) I . ANSWER 8.04 (2.50) b)oc4cce,d b'i"*y UNYI') E pece,y7 vi,+ I 4 L a.
Shift Superviser EO.53 b.
Primary unit Secondery unit e i 7 7 m,..., # gu A 6 )'wj e / servus v d u c p d v r,7 7 ;,3; mg, _,, Radwaste operator E0.25 pts each] l
, c.
Shift Supervisor E0.5] f s-9ec ~rdy j d.
NO CO.5] J l
REFERENCE i
AD-37 pgs 2,3
k 194001 K 1.05 3.1/3.4 ] l LPO Admin Requirements # S.2,5.3
194001K105 ...(KA'S) ) ANSWER 8.05 (2.50) i e i a.
The Fire Brigade shall not include iour members of the minimum sn i -f t crew necessary for the safe shutdown of the unit or any personnel ) required for other essential functions during a fire emergency.
f. 1. D l' ' I b.
You may operate for un to two hours with one less than the minimw c omp l i raen t (0.75] pr ovi ded that immediate action is taken to bring the comp 1iment up to minimun. L0.753.
REFERENCE TS pg 6-1 194001 K 1.16 3.5/4.2 A 1.03 2.S/3.4 LPO SMR 4t 1.1 194001A103 194001K116 ...(KA*S) l t _ _ _.
._ -- -- - - _ --_-_. - -- - - x,, ; .: . t-; g., - b A D M I N I S T R A T I V E P R O C E D U R E S i CONDITI.gNSi_AND LIMITATIONS PAGE-36 . ' .. ANSWERS ~-- SALEM 1&2-87/09/15-SILK,.D.
y l j' ( :- L ANSWER 8.06 (2.50)' a.. NO'E0.5] -b.1;The intent of the original procedure is not altered E 0. 6 3.
.2.The-change is-approved by two members of the plant management staff.. at least one of whom holds a SRO license CO.7]
' 3 T " _-b: ,. b,- g ' T: C. Y: C - ;
- ._
- md :;;.
.=d b; ' t ",- '. ' ~ ' - , - ~
- - --
-' n . g igg gi ..r e- .: . _q _1
- q7 ;q g
_; , , Tv ch y. : s "<{ev.,1.utei Od re a y y Th! wne Ie.d & rv Hew m la(Y w l n: 'il, [' ^- l C 7;)^ I g%cyluq [..nde, yet ;/,,,jg - (,y, y q c,% i9 s j f., ie.nf f fj,,, .,3
. REFERENCE
' .TS pg 6-13 AD-1 Section C pg 9 P1 ant-Wide'Generice ~ 194001 A 1.01 3.3/3.4 ' LPO Use and Control of Procedures # 3/4/5.1-194001A116 ... OtA'S) ANSWER 8.07 (2.50) a.. Wnenever a fire door, fire damper, fire barrier penetration or dire suppression or detection stem is made inoperable for any reason j [0.53.
The Fire Pratoct Supervi sor wi11 notify the SS E 0. S]. 'l b.
Thi s-i s not adequate firca protection coverage EO.5J because the arews t are on.different el evat i orm of the Aux building E0.63 thereby proh3h-iting continuous fire protection coverage as specified by TS (3. 7.1 W 3.a) [0.53 REFERENCE OD-50 pgs 2,3 LER 87-006 (Uni t 2) I 194001 K-1.16 3.5/4.2 LF'O Admin Requirements # 4.1,4.8 194001K116 ...(KA*S) l q . J ! I e
I ._ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _,
- _- __ "es__GDdlNISIB611VE_E8QQEDQBESt_QQNQlllQNQz_6NQ_Lidll@IlQNQ PAGE
.. ANSWERS -- SALEM 16.2-87/09/15-SILK, D.
. s ANSWER 8.08 (2.50) Since you are about to enter Mode 3 C0.53 the heatup must be discontinued [0.53 and Tave held at less than 350-F until Charging Pump $27 U2 2E is proven operable CO.53 because a made change cannot occur with reliance upon an action statement [1.03.
REFERENCE ! TS'3.5.2 pg 3/4 5-3 l TS pg 1-8 j TS 3.04 pg 3/4 0-1 j j f 3.4 003 000 G
3.4/3.8 194001 K 1.09 3.4/3.4 i ! LFO CVCS #21 003000G5 194001K109 ...(KA'S) ANSWER 8.09 (2.50) RCS Pressure Isolation Valve Limits exceeded.
[0.753 l (3.4-1.6)/(5.0-1.6) 1.8/3.4 : 50% [0.5] = UNIDENTIFIED Leakage l i m) + r exceeded.
00.753
13.2 - (1.5+1.2+3.4+ 0.8+4.2) 2.1 > 1.0 [0.5] = REFERENCE TS 3.4.6.2; TS ~.4.e'.3 3.2 002 020 6
3.6/4.1 LPO RCS # 4.1,4.2,4.3 002020G5 ...(K4'S; s ! i
, L-_-_____ - _ _ _ _.
_ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ - - _ _ _ _ _ _ _ - _ - _ _ _ - - - - - - _ - - - - _ _ _ - - - - - - - - - - - - - - - - - - - - - -- _ "et_:00dlNLEISGILME_EBQQEQQBEQt_GQUQlligNQu_8NQ_LidlI@IlgNQ PAGE
.. yANSWERS -- SALEM 1&2-87/09/15-SILK, D.
. i I ANSWER 8.10 '(3.50) a.
[0.53 b.
Within.15 minutes E0.23 1.
Restore the indicated AFD to within the target band [0.43, or 2.
Reduce the thermal power to <90% of rated thermal power. [0.4] c.
Accumulated penalty over the past 24 hours is 89 minutes C1.03.
The penalty will be reduced to 60 minutes at 1618 minutes on 05/13/87 and then power may be increased [1.0]. 85% 0318-0310 -
LO.25] 65% 1637-1557
[0.25] = 45% 0310-0148 82/2
[0.5] = = -. - 89 min, total penalty 05/13/87, from 1557; 81 min left -60 - 21 min -> 1618 05/13/B7 L1,C] REFERENCE TS 3.2.1; TS pg. B 3/4 2-1 2
3.9 015 000 G
3.3/1..E A 1.05 ' 7/3,9 .. LPD Excore N1 # 19,19/20.1,19/20.2, 19/20.4 015000A105 015000G5 ... ( K A ' E; )
.
- f'.'s ca v = s/t Cycle efficiency = (Newon.
, out)/(Energy in)
- <
. ,
.u = ag s = V,t + 1/2 at
E-= ac. . KE = 1/2 av a = (Vf - V )/t A = AN A=Ae**
g g PE = agn Vf = V, +-at w = a/t x = in2/tjjg=0.693/t1/2 ' 1/2'If * ES E1n)(t )3 t ~ h W = v AP [(t1/2) * (*b)3 4E = 931 an I'= I,e " ~ . . ' Q = aCpat d = UAat I=[0,"/WL
Pwr = W ah I=I 10-X y TVL = 1.3/u sur(t) p = p 10 HVL = -0.693/u
P'= P,e*/ ~ . SUR = 26.06/T.
SCR = S/(1 - K,ff) CR = S/(1 - K,gx) x > SUR = 26e/ t* + ( s - o )T CR (1 - K,ffj) = CR II ~ eff2) -
2
T = ( &*/s ) + [(s - o )/ o ] M = 1/(1 - X,ff) = G)/G o T = 1/(p - s) M * (1 - K,g,)/(I - K,g1) T = (s - o)/(la) SDM = (1 - K,g)/K,g 10-5 seconds a = (K,g-1)/K,g = 4K,ff/K,ff t* = T = 0.1 seconds-I , o = ((&*/(T K,g)] + [f,ff (1 + AT))
/ Idlj=Id P = (r4V)/(3 x 1010) I d) 2,2 2 gd j
2 I = oN R/hr a (0.5 CE)/d (,,g,73) R/hr = 6 CE/d2 (feet) . Water Parameters Miscellaneous Conversions j _ 1 gal. = 8.345 lbm.
1 curie = 3.7 x 1010dps 1 ga[. = 3.78 liters 1 kg = 2.21 lbm
3 'l ftt = 7.48 gal.
I hp = 2.54 x 10 Stu/hr Density = 62.4 lbm/ft3 1 rme = 3.41 x 106 8tu/hr Dansity = 1 gm/c:n3 lin = 2.54 cm Heat of vaporization = 970 Stu/lem 'F = 9/5'c + 32 Heat of fusion = 144 Stu/lbm 'C = 5/9 (*F-32) 1 Atm =-14.7 psi = 29.9 in. Hg.
1 BTU = 778 ft-lbf 1 ft. H 0.= 0.4335 lbf/in.2 p - _ --. - ._
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, . RpCTORCOOLANTSYSTEM OPERATIONAL LEAXAGE LIMITING CONDITION FOR.0PERATION l . 3.4.6.2 Reactor Coolant System leakage shall be limited to: a.
No PRESSURE BOUNDARY LEAKAGE, , b.
1 GPM UNIDENTIFIED LEAKAGE, c.
1 GPM total primary-to-secondary leakage through all steam generators and 500 gallons per day through any one staam generator, d.
10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and e.
40 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2230 + 20 psig.
~ APPLICABILITY: MODES 1, 2, 3'and 4 ACTION: a.
With any PRESSURE BOUNDARY LEAKAGE, be in at le'ast HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
b.
With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, reduce the leakage rate to within limits within 4 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS . 4.4.6.2 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by; a.
Monitoring the containment atmosphere particulate radioactivity monitor at least once per 12 hours, b.
Monitoring the containment sump inventory at least once per 12 hours.
SALEM - UNIT 1 3/4 4-15 .. _ _ _ _ _ _ _ _ -
l l' L a l . , ! REACTOR COOLANT SYSTEM . SURVEILLANCE REQUIREMENTS (Continued) ~ . c.
Measurement of the CONTROLLED LEAKAGE from the reactor coolant pump seals at least once per 31 days when the Reactor Coolant System pressure is 2230 + 20 psig and valve ICV 71 is fully { closed.
- d.
Performance of a Reactor Coolant System water inventory balance at least once per 72 hours. The water inventory balance shall be performed with the plant at steady state conditions. The provisions of specification'4.0.4 are not applicable for entry into Nde 4, and e.
Nnitoring the reactor head flange leakoff systen at least once per 24 hours.
f . . SALEM - UNIT 1 3/4 4-16 Amendment No. 58 . .. - - - - - _ - - _ - - - _ - - -. _.. _ - - -
_ . _ _ -. _ _ _ - __ _ _ _ - - ..- _ * OCO3G9 H ' 3.. . . , - j .
REACTOR-COOLANT SYSTEM a . PRIMARY C0OLANT SYSTEM PRESSURE ISOLATION VALVES LIMITING CONDITION FOR - , OPERATION , ' , 3.4.6.3 React'or Coolant,5ystem Pressure Isolation Yalves shall be j i operational.
i The integrity of all pressure isolation valves listed a.
in Table 4.4 4 shall have been demonstrated, except l ' as specified in' "b".
Valve leakage shall not exceed the amounts indicated.
- b.
In the event that the ' integrity of any pressure isolation
valve specified in Table 4.4 4 cannot be, demonstrated. - reactor operation may continue, provided that at
-
least two valves in each high; pressure line havino a non- - functional valve are in, an(aitemain in,the mode corresponding to the isolated condition.t -
AFPLICA!ILITY: MODIS 1, 2',' 3, and 4.
. ' If neither Condition "a" nor "b" can be met, an orderly ACTION: shutdown shall be initiated within one hour and the ) reactor shall be in at least HOT STANDBY within. 6 hours and in COLD SHUT 00WN within the following 30 hours.
I SU:VEILLANCE REQUIREMENTS Periodic leakage testing (b) on each valve 1,isted in Table 4.4.6.3 a.
4.4-4 shall be accomplished * 1.
Each time.the plant is placed in COLD SHUTDOWN condition for refueling.
2.
Each time the plant is placed in COLD SHUTDOWN condition for 72 hours if testing has not been accomplished in ' the preceding 12 months.
,
. . . . . (a) Motor operated valves shall be placed in the closed p~osition and power supplies deenergized.
(b{ To satisfy ALA*.A requirements, leakage may be measured indirectly (as l from the perfor,ance of pressure indicators) if accomplished in acccrdance with approved procedures and supported by computations showing that the l L cethod is capable of demonstrating valve compliance with the leakage criteria.
f . - , Order dated April 20, M 3 g 4 16a ' l , i - - .-_-_______________]
- - . m ue _ _ _ _ _, _ _ maintenance repair, or replacement work on the Valve.
,
- s.
4.- The provision of specification 4.0.4 is not applicable . for entry into Mode 3 or.4 ,- , b.
Whenever integrity of a pressbre isolation valve li.s'ted in Table 4.4-4 cannot be demonstrated, the integrity of the remaining valve.in each high pressure line having 'a leaking valve shall be determined and recorded daily.
In addition, .the position of one other valve located in the high pressure line shall be recorded daily.
, . I ' " . ,
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3/4 4-16b Order dated April 20, 1951 l . . _ - _ _ _ _ _ _ _. _
- - -~ , h * System . ' Valve No.
Allowable Leakace . .1 Pressure Safety Injection ' '45.0 GPM each Oal've .
- cop '11, cold leg 11SJ56
- I.50GPMeachvalve
. 11SJ43 55.0 GPM each valve - ' 5.0 GPM each valve + Loop 12, cold leg 12SJ56 12SJ43 15.0 GPM each valve , ' Loop 13, cold leg-h-j h ','$ V'){'
y e h ]Ch v Loop 13, hot leg 13SJ156 e c , 13RH27 - 15. 0 GPM each valve . Loop 14. cold le's D ' 15.0 GPM each valve Loop 14, hot leg 14SJ156 15.0 GPM each valve 15.0 GP/I each valve , 14RH27 , In$ermediate Pressure Safety Injection 15.0 GPM each valve Loop 11, cold leg IISJ144 25.0 GPM each valve Loep 11, het leg 11SJ156 15.0 GPM each valve 11SJ139 15.0 GPM each valve . Loep 12, cold leg 12SJ144 15.0 GPM each valve Loop 12, hot leg
<5.0 GPM each valve 55.0 GPM each valve
9 .oop' 13, col d l eg 135'J144 15.0 GPM each valve Loop 13. hot leg 13SJ156 15.0 GPM each valve - 135J139 16.0 GPM each valve 15. 0 GPM each valve Loop 14, cold leg 14SJ144 . <5.0 GPM each valve Loep 14, hot leg 14SJ156 E5.0GPMeachvalve 14SJ139 I"I. Leakage rates less than er ecual to 1.0 sps are considered acceptable.
I However, for initial tests, or tests following valve repair or replace-ment leakage ra,tes less than or equal to 5.0 gpm are considered acceptable.
2.
Leakage rates greater tha'n 1.0 g;e but less than or equal t.o 5.0 g;c are. considered acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum , permissible rate of 5.0 gpm by 50% or greater, . 3.
Leakage rates greater than 1.0 g;m but less than or equal to 5.0 gpm are considered unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the , margin between measured leakage rate and the maximum pemissible rate of 5,0 gpm by 50% or greater, ,
Leakage rates greater than 5.0 g;m are considered unacceptable.
, (b) Minimum differen'tial test pressure shall not be less than 150 psid.
. SALE'l-UNIT 1 3/4 4-16C Order dated April 20, 1981
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, ' i > Questio'n Number Modification 5.02c-Increase instead of' decrease due to-location of-charging' pump' discharge valve ~at Salem.
D > 5.10c-Question was misleading and therefore was deleted.
6.04a-Will accept Fission Gases or Xe-135 to' reflect . guidance given in system operating procedure.
6.04b-Remove: valve stated within the-parentheses from: key-. because it is incorrect.
6.06' -VCT auto' makeup occurs at.14% instead of 20% 6.06 & 6.07- .-Points in answer key were redistributed to be'tter " correlate to importance.
6.10b-Will accept' High' Steam Line Flow coincident with Low Steam Line Pressure or Low Tavg as two answers.
7.04a-Pre'ssurizer volume less than.92% instead of 93.2%. 7.04b-The answer should read: To ensure an effective back - Epressure on the RCPs' No. 1 Seals to maintain lubri-cation-of the No. 2-seals.
7.06b-Will accept radial. bearing'or seal package damage as part of the. answer to reflect a note in a system- ~ -operating procedure.
7.07a-Will accept increasing pressure as an alternate . answer.
7.07b~ -Will accept OP Delta-T as an alternate answer.
7.08-Points in answer key were redistributed to better correlate to importance.
7.09c-LOPA-1 instead'of ECA-0.0 to reflect specific facility terminology.
7.09d-There should only be one answer - as directed.by procedure in effect.
Candidates will not be penalized for providing a second' answer since the question asked for two.
7.10a-Agree to grade answer according to procedure referenced by the candidate.
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ATTACHMENT 2 7.10b-Setpoints are not required for full credit.
8.04b-Tagging and Surveillance Operator is now the Circulating Water / Service Water Operator - will accept either.
Will also accept distinction made between ' Unit 1 and Unit 2 Primary and Secondary Operators.
8.04c-Will accept either Shift Supervisor or Security.
8.06b.3-Answer should be "The change is documented and receives the same level of review and approval as the original procedure (under specification 6.5.3.2a) within 14 days of implementation" to reflect a change to Technical Specifications.
8.08-Both centrifugal charging pumps (21 and 22) are needed to satisfy the ECCS LCO.
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