IR 05000333/1986009
| ML20215N691 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 10/29/1986 |
| From: | Howe A, Keller R, Kister H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20215N681 | List: |
| References | |
| 50-333-86-09OL, 50-333-86-9OL, NUDOCS 8611070167 | |
| Download: ML20215N691 (153) | |
Text
{{#Wiki_filter:e U. S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO. 86-09 (OL) FACILITY DOCKET NO. 50-333' FACILITY LICENSE NO. DPR-59 LICENSEE: Power Authority of the State of New York P.O. Box 41 Lycoming, New York 13093 FACILITY: FitzPatrick Nuclear Power Plant EXAMINATION DATES: July 28-31, 1986 CHIEF EXAMINER: d'. /8-d 7-86 A. Howe, Reactor En' i'neer (Examiner) Date g REVIEWED BY: J /U# [/Y8 /8'2 7' 86 R. Keller, Chief, Proj4 cts Section 1C Date APPROVED BY: /0 L7 M Har'ryl.' KfQer, Chief, Dat6 [ Projects Branch No. 1 . SUMMARY: Operator License examinations were conducted at the James A.
FitzPatrick Nuclear Power Plant during the week of July 28, 1986.
Examinations were administered to three reactor operator candidates, two senior reactor operator candidates and one instructor certification candidate. All candidates passed the examinations with the exception of one reactor operator candidate who failed the oral examiration and one reactor operator candidate who failed the written and oral examinations.
8611070167 861029 PDR ADOCK 05000333 V PDR , _, _. _ - -. _ _ _ - _ _ - _ _ - . - . -., _ - _ ____
REPORT DETAILS TYPE OF EXAMS: Replacement EXAM RESULTS: l R0 l SR0 l Inst. Cert I-l Pass / Fail l Pass / Fail l Pass / Fail l l l l l l l
I I l Written Exam l 2/1
2/0 l 1/0 l l l l l l
I I I i 10ral Exam l 1/2 l 2/0 l 1/0 l l l __.
l_ l i I I I I i l0verall l 1/2 l 2/0
1/0 l
I I I I CHIEF EXAMINER AT SITE: A. Howe OTHER EXAMINERS: D. Lange F. Crescenzo 1.
Summary of generic strengths or deficiencies noted from grading of written exams: Strengths SR0 - Nuclear power plant theory; RPS indications; Technical Specification requirements on surveillance intervals and single loop operation; Emergency Plan.
R0 - Nuclear power plant theory; ADS, Fire Protection and feedwater control systems; abnormal procedures for reactor scram and loss of 10500 bus; entry conditions to Emergency procedures.
Weaknesses SR0 - Abnormal procedures for loss of UPS and for feedwater flow con-troller failure; Technical Specifications on ECCS requirements during maintenance, and Shift Supervisors log entries.
RO - SBGT and ESW systems, refuel system interlocks; APRM and RWM; abnormal procedures for loss of instrument air and for loss of UP, . ! .
__ ,, 2.
~ Personnel Present at Exit Inte' view: r NRC Personnel A. Howe, Chief Examiner D. Lange, Lead BWR Examiner A. Luptak, Senior Resident Inspector Facility Personnel R. Converse, Resident Manager R. Locy, Assistant Operations Superintendent D. Simpson, Training Superintendent F. Catella, Training Coordinator G. Fronk, Nuclear Training Specialist 3.
Summary of NRC Comments made at exit interview: The Chief Examiner extended his appreciation of the cooperation from the operations staff and the training staff.
The facility was informed that the examination results will be forwarded.
as soon as reasonable and with a goal of results within thirty (30) days.
Generic strengths and weaknesses of the oral examinations were discussed with emphasis that the SRO candidates performed generally better than the RO candidates. Also, a weakness was noted with regard to knowing the controls available in each safe shutdown panel.
i a The exam review went well and only a few unresolved comments are to be forwarded to Region I.
During the exam review, some lesson plan changes i l and procedure changes were agreed upon by the facility staff.
! It was noted that several sections of the facility provided training material were vague and of poor quality. At.the NRC's request, better , l training material was provided.
Improved screening by the facility of future examination material was recommended.
- The facility was informed of some inconsistencies in NRC grading of l I facility provided requalification examinations which will require a more I detailed review. Notification was given that a training program inspec-tion will be conducted to follow up this effort.
I ' i
Attachments:
j 1.
Written Examination and Answer Key (RO) i 2.
Written Examination and Answer Key (SRO) ) ' 3.
Facility Comments on Written Examinations and NRC Resolution ' of Comments
. _ _. _ _.., _ _.. _.. _. _ _... - -, _ _ _ _. _, - - _ _ _.. _ _ _ -,... _ _. _... _. - - _ _ - . -... - - - . - - ,.
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NUCLEAR REGULATORY COMMISSION , t REACTOR OPERATOR LICENSE EXAMINATION , F.ACILITY: _FITZPATRICK_____________ REACTOR TYPE: _ B_ W R _- G_ E 4 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ __ DATE ADMINISTERED _@6f9Z42@__________===____ _ EXAMINER: _LANGE _Qz_______________ t APPLICANT: ________ _________ IN@l@UCllgNE_Ig_@PPLIC@ nil Uso separate paper for the answers.
Write answers on one side only.
Stcple question sheet on top of the answer sheets.
Points for each qu stion are indicated in parentheses after the question. The passing grcde requires at least 70% in each category and a final grade of at lecst 80%. Examination papers will be picked up six (6) hours after . tho examination starts.
% OF CATEGORY % OF APPLICANT'S CATEGORY __Y8LUE_ _lQI@L ___@CQBE___ _y@LUE__ ______________C@IE@Q@Y_____________ _2E 99__ _2E 99 1.
PRINCIPLES OF NUCLEAR POWER ___________ ________ PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW _2E 99__ _Z9399 ___________ ________ 2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS _2E 99__ _2@z99 ________ 3.
INSTRUMENTS AND CONTROLS ___________ _32:99__ _32:99 .________ 4.
PROCEDURES - NORMAL, ABNORMAL, ___________ EMERGENCY AND RADIOLOGICAL CONTROL i ! i 199 99__ 199 99 ________ TOTALS ___________ I FINAL GRADE ___,_____________*/. All work done on this examination is my own. I have neither givcn nor received aid.
APPL 5Cd5T 5~555NkiUR5~~~~~~~~~~~~~~ I l
l i - .-. - _ -., _. - . - _ - _. - - -. -, _ _ _ _ - -. -. - - - - - -. - _ -.. -
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PRINCIPLE @_QE_ NUCLEAR _PQWER_ PLANT _QPERATigN PAGE
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s QUESTION 1.01 (2.50) The concept of Subtritical Multiplication is used to describe the behavior of the reactor during refueling operations or startup.
.c.
In a suberitical reactor, if the source level doubles, what will happen to the neutron level? (0.50) b.
What three variables affect the suberitical neutron level? (0.75) c.
In a subtritical reactor, if a reactivity of 0.003 dk/k is added to the reactor, will it take longer to reach equilibrium if the initial K eff is O.92 or if K eff is O.992 7 Explain the reason for your answer.
(1.25) QUESTION 1.02 (3.00) Ansume-the reactor is operating'at 100% power and one recirculation pump trips. Indicate how each listed indicated parameter would first change (Increase or Decrease) and briefly explain why the change
REGION OF o , REQUIRED "->-
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SOOlUM PENTA 80R ATE SOLUTION ,, VOLUME-CONCENTR ATION R ECulREMENT S
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JA?lTP , - i > , 88. 5 SURVEIT.IJLNCE REOUIREMENTS l , 3.5 f.IMITING CONDITIONS FOR OPERATIOff 4.5 mRE AND MMAIWW COOI,M 3.5 Cone A;m CONTAITNE?IT COOLIfG SYSTalS
SYSTI~MS ,
. Applicability: Applicability: , , Applies to periodic testing of the Deer-Applies to the operational status of the gency Core Cooling Syst4ms, the , . Emergericy' Cor'e Cooling Systems,. the suppression pool cooling and contaisseent suppression pool cooling, and - spray mode of the Residual.lleat ' Removal ' l containment spray saodes of the Residual (RHR) System.
, lleist Removal (RllR) System.
I
I Obiective: , ' g.jecegye of the Core operability $teh To verify the i To assure operability of the Core and and Contairunent Cooling Sy under all conditions for which operability is , containment Cooling Systems under all < conditions for which this cooling essential.
< capability is an essential response to.
' ' , ' -
plant abnormalities., -l
l t soicification: . Specifications . ' l ' . '
A'. Core' Spray System and Low Pressure - j Coolant Iniection (LPCI) Mode of the A.
Core Spray System and Low Pressure Coolant Iniection (LPCI) Mode of the-i s RIIR System .
- j (RllR) System I
1.
Doth Core Spray Systems shall 1.
Surveillance of the' Core Spray bit operable whenevaar irradiated system shall be performed as . l fuel is in the reactor vessel follows: j
and prior to reactor startup from a cold condition, except I.t.em,. Frequency as specified below: a. Simulated Each operating Automatic cycle i I Actuation ' . - - , Test ' ] i ' -
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i ' - . @2 - - - - - - - - - - - - - - - ---
- - - -- - ,f-., , -. .; ,.., , . , - -
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s i ! ' t I l 3.5 (cont'd) . JAFNFP . 4.5 (cont'd)
. .
J b.
Flow Rate Once/3 Months j Test - core . I
spray pumps I - shall deliver , .at least i , l 4,625 gym l .against a,sys-l ! l tem head cor-l i , ) responding to . ' - , a total pump ' l , . . developed. head
of A113 peig . ', t ' . . - 1 ' ,nce month O .o.
Fump, Opera- , bility
, I ,
! d.
Motor Oper-Oges/ponth ? sted Valve ' ' i t , e.
Core, spray g..' c. , I Needer oy JI Tl h ;.T } U " ' - l f - ' . - Instrumenta- - g - . . tion-6- 'Once/ day , Check , 'I once/3 monthe , , . onoe ! cal *1brate t /3 months Test , '- , e e Once)each j . ~ ' . ) - f.
Logie system l Functional ,opecatling ' ' . ' Test cyclLe
i ,
- g.
Testable Tosked for i i Check Valves operab!!1ty . any' time thy l lr'eactor is in j the cold con ' dition esseed-
' , I Ing^48 leours, ' if o,perability . test s have not - ' * ~ been performed ,
i . 'de ing the pro- , Amondsont 40 -, ao ng 31 days.
113 - - - - -- ,, - _ _
.. _.. _ _, _.. _ . _ ._.
.. _ .?-..g ' g'%' , . - 4.5 (cont'd) ! , & , 3.5 (cont'd) . ' J ATHPP-b.
Flow Rate Once/3 Months I Test - Core - spray pumps f ,shall deliver , .at least I i l 4,625 gym a g a l,n s t a,sys-l tem head cor-responding to ,i - s ' a total pump' ' , . developed. head e of A 113 psig . i .. ', " t ' . ~ .c.
Pump,0pera-Once month , bility , , - k .d.
Motor Oper-Ogce/ month ' ated Valve e.
Core,5 pray g., lj ..TU* ,, , f l , llender ay . . Instrumenta- .- . - tion-Once/; day , i Check , 'Ionce/3 menthe - Cal'1brate onoe /3 montha Test , ' I e' ' on*ce. t/en9h ' . , . f.
Logio system ' . Functional ,opeesting Test cyc.Le
. , ' Testable Teshed for 'g.
- operability . check valves any time the fe'esotor is in the cold con ' dition exceed 4 ' ing 48 Igours, + if operability test) have not ' ' , been performed , ' A sien,f eia e
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j 'h j' , , D%.., ' ' ' ' 3. 's (Cont'd) 4.5 (Cont'd) '. '.. ', d?4 ) , , ..A... i< ,s , , .i.
,! isdeterminedthattheLPCImode'fq ! t.. From the time that the I.pCI mode is made or h.
k' hen it l found to be' inoperable for any reason, inoperable, both Core Spray Systems.'and the - ~~ . ! continued reactor operation is permissible containment spray subsystem shall be demonstrated to l l during the succeeding 7 days unless the I.PCI he operable immediately and daily thereafter.
' , ' c, , i mode is made operable earlier provided that ~
I during these 7 days all active components of [- , ' hoth Core Spray Systems and the containment l spray subsystem (including two RilR pumps) shall p ! , '
' he operable.
,]_ [" , ', , ! "'. .,, ,, - Q W., ' , , s .- ., . c.
k' hen,the reactor water temperature is greater c.
The power source disconnect and chain' lock're motor 1/.
than 21L*F, the motor operator for the RilR operated RHR cross-tie valve, and lock on sanually r ' cross-tie valve (HOV20) shall be maintained operated gate valve shall be inspected once each . i disconnected from its electric power source, operating cycle to verify that both valves are ' l It shall be maintained chain-locked in the closed and locked.
'I ' closed position. The manually operated gate l , . ] valve (10-RHR-09) in the cross-tie line, in
! ',l series with t'he motor operated valve, shall be . maintained locked in the closed position.
- ' I
, . l 4.a.
The reactor shall not be started up with the , l RilR System supplying cooling to the fuel pool.
.
.
i.II i ,. The RI'R System shall not supply cooling to t'he }., j, ".,?. i b.
i , , i , ] spent fuel pool when the reactor coolant temperature is above 212*F.
. .
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- f
,y. g,k '.. . k< . g ,1 . -p;., - - , .+1., , . [' [sfae nd ent No.
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JAFNPP f.
It i ..i.. ' ' t te e ., , 1.5 (Cunt'd) 4.5 (cont'd) ' %. All recirculation pump discharge valves and 5.
All recirculation pump discharge and bypass valves hypass valves shall 1+ crerable prior to shall he tested for operability any time the reactor reactor startup (or closed if permitted in in the cold condition exceeding 48 hours, if
l elsewhere in these spec if icat ions). operability tests have not,hcen performed during the l preceding 31 days.
l
- '
' e . , i ) fi. If the requirements of 1.5 A cannot be met, the reactor, shall he placcel in the cold condition ., ! within 24 hrs., - , , ! , , .. . .. l ' 15. CONTAINMENT COOI.ING SullSYSTl H HODE (OF 11. CONTAINMENT COOLING SUBSYSTEM MODE (OF l Tile RIIR SYSTEM) Tile RilR SYSTEM) l l.
l 1.
Both subsystems of the containment cooling 1.
Subsystems of the containment cooling mode are i mode, each including two RIIR, one ESW pump and ' tested in conjunction with the test per(ormed on the I two RHRSW pumps shall he operable whenever I.PCISystemandgivenin4.5.A.I.a.b{e,andd.. - ,
there is irradiated fuel in the reactor vessel, Residual heat removal service water pumps, edch loop l prior to'startup from a cold condition, and consisting of two pu'aps operating in parallel, will , j reactor coola'nt temperature > 212*F except as he included in testing, supplying 8,000 ppsk The ! specified below: Emergency Service Water System, each loop of which ! consists of a single operating emergency nervice '. waterpump,of3,700spswillbetestedinaccqtdance with Section 4.llD.
' i g puring each five-year period, an air test shall be - performed on the containment spray he&ders and nozzles.
! , . l 2.
Continued reactor operation is permissible for 2.
When it is determined that one RNR pump an for one I 30 days with one spray loop inoperable and with RHRSW pump of the components required in 3.5.3.1 ) reactor water temperature freater than 212*F.
above are inoperable, the remaining redundant' active,
! components of the containment cooling'uode , j subsystems shall be demonstrated to be operable , ! i ' immediately and daily thereafter.
, , i . Amen.Iment.No. X 95 - IlSa .- ' ' > . ,.,
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. . - .- .. - ,.w m-ww p-e.
~ 4*. w - .. g.
V ,I o k,) ' v ~ JAFNPP ' . I
- Y 1. 's (Cont'd)
4.5 (Cont'd) I _ ' t.
<l..uld one PHR pump and/or one RIIRStl pump of l.
lihen one containment cooling subsystem loopibecomes, i , tobeoperableimmediatelyanddailythdr%strated t he component s requi red in 't.5.R. I above be inoperable, the operable, loop shall be de , ' ' ' ' eafker.
m.t.le or found inoperable, continued reactor ' operation is permissible nuly during the
succeeding 30 days provide.1 that during siich 30 days all remaining active components of the - ' ccutainment cooling mode are operable.
I a.
St.ould one of the containment cooling '
, , , ' subsystems become inoperable, continued reactor I operation is permissible for a period not to l exceed 7 days, unless such subsystem is sooner
made operable provtoed that during such 7 days ' ., all active,copponents of the other containment - - cooling subsystem are operable.
, , ,. . .. 5.
If the requirements of 3.5.R cannot be met, the reactor shall be placed in a cold condition ' within 24 hr.
' - . ' I.
- 6.
Low power physics testing and reactor operator I ' ' training shall be permitted with reactor
.. l l coolant temperature < 212*F with an inoperable . , component (s) as specified in 3.5.8 above.
' - , ' s ,
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Amendment No.)(95 I 116 j e i I ' .. ' , , ,
__ . _. _ ._ ._ _ _ _ -. - - - , _ 't l - W. %w h ' i s- ' I
'. i JfkFNPP ., 3.5 (cont'd) 4.5 (cont'd) I , ! I . DELETED . - , , C.
IIIGil PRESSURE COOLANT INJECTION C.
IIIGil PRESSURE COOLANT INJECTION (IIPCI S YSTf'll) (llPCI SYSTEM) I i Surveillance of HPCI System shall.be . performed as follows provided reactor .- steam supply is available.
If steam is* 4-not available at the time the surveillance..., ' i , 79.
tes't is scheduled to be perf6rmed,'thetest shall be p .
continuous operation from~theiti'me steam: becomes available.
,lpn.g.
_ ,. ' '
,- 4 1.
The HPCI System shall be operable 1.
IIPCI System testing shall be.,. xr.as specified y: L - ' .
whenever the reactor pressure is in 4.5.A.1.a, b, c, d, f, and g except.s... greater than 150 psig and irradiated that the llPCI pump shall deliver at least.
4,250 gpm.against,a system dea ding to a reactor vessel press $ correspon fuel is in the reactor vessel and ure of 1120 .!:n.
- prior to reactor 6t.artup from a
"'" cold condition, except as specified psig to 150 psig.
below: .' i 'i ' ,. , sc l - ' , - . j . ,
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l, Amendment No. pf 95 ,, . . ,
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v , - , , ' , l, ' 3.5(cont'd) JAFNPP 4.5(cont'd),
. I - . . , \\.
I - a.
From and after the date a.
When it is determined that ' ,. _.... that the HPCI System is the HPCI subsys. tem is in-_ . _.,
made or found to be operable the RCICi the ' ' . l inoperable for any reason, LPCI subsystem, both core ) continued reactor spray subsystems, arid the l operation is permissible ADS subsystem iactuation
only during the succeeding logic shall be demori-
- ,. i. '
7 days unless such system strated to-be operable ' ' ' ;; ' ' is sooner made operable, immediately.
The RCIC W-
provided that during such system and ADS subsystem [ ' be f
days all active logic shall- - components of the Auto-demonstrated to be matic Depressurization operable daily thereafter.
i , ' l System, the Core Spray ,, l Syste'm, LPCI System, and f.
Reactor Core Isolation ! Cooling System are j operable.
, b.
If the requirements of 3.5.C cannot be met, the j,j ,. reactor shall be placed in
s9-' - the cold condition and~ vi.
' pressure less than ' i#.!! MY 'g.i'.. #.- ' ' 'd . e.. 2- . - ~, - ' ' ' ) 150 psig within 24 hrs.
y :: 5, % -:.
l i 2.
Low power ~ physics testing and , 'j.~.f.> ~ ,' ~ - N ! ' reactor operator training shall
- fMN
'?- ' ' ~ i' ' il t- -*y,T j be ' permitted with reactor , . ' coolant temperature 5212*F with . -!
I an inoperable component (s) as .. ,a
specified in 3.5.C.2 above.
~ L ' - . ' . ! {! . j i .
Amentiment Nn.
118 , ' - i .
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3.5 (cont'd) JAFilPP 4.5 (cont'd) I
, , ! , - D.
Automatic Depressurization System D. Automatic Depressurization System l (ADS) (ADS) - 1.
The ADS shall be operable whenever 1.
Surveillance of the Automatic j the ' reactor pressure is greater Depressurization System shall + than 100 psig, and irradiated fuel be performed during each ' perating o is in the reactor vessel and prior cycle as follows: ' to reactor startup from a cold - ' ' ' condition, except as specified a.
A simulated, automatic initiation
, , belows which opens all pilot valves.
.,, .l r a.
From and af ter the date b.
Manually open each relief / safety ' that one of the seven valve while bypassing steam to , relief / safety valves of the condenser and observe a 2 1036 I . the ADS is made or found closure of the turbine bypassn..;.. g f.
l to be inoperable for any valves, to verify that the.
- M ai
reason while it is required, relief / safety valve h4s ned.
',# '
continued reactor operation is permissible only during c.
A simulated automatic initiation , , the succeeding 30 days unless which is inhibited by the override i repairs are made and provided switches.
r . , , ' that during such time the IIPCI .I.
System is operable.
- .
,. g ' . ! [ ' j'y,c.
'b.
Froin the time that more than , , p ,
one of the seven relief / safety
I ;. . , valves of the ADS are made or found to be inoperable for any reason, continued reactor opera ' tion is permissible during the ,i succeeding 24 hrs. unless repairs-are made and provided, that l Js . e f I . .l .. . . f Amendment :.c..,,4 84 119 , I l . --. _ l ,
.,_.a.
._m- -- -- - m - , , i ~ ' '.
- 7 T'
' ' . ' l t ., . I ' 3.5 (cont'd) JAFNPP 4 5 (cont'd) . I! . . during auch time, the IIPCI ' System is operabic.
. . i t . 2.
If the requirements of 2.
A logie system functional test.
s l 3.5.D.1 cannot be met, the ..When it"is deterinined th'af.
' re.ictor shall be placed in the' a.
' , cold condition and pressure one valve of the AUS is -
loos than 100 psig, within 24 inoperable, the .. DS. - hr.
nuhsystem actuatirin 'l IIc , for the operable.. ADS -
valves and the itPCI *sub-system shall 4 i be demonstrated to J.
be operable inanediately. and . at 1gast weekly.
y thereafter.. - - l ' b.
When it is detersmined that . . than , i ne n - relief / safety valve *,of t'hq, . '
snore ' ads,is ' inoperable, kthet ' '
- t.
, . . IIPCI System ashall be - deinonatrated to lie , , ' operable insnediately.
. , . . , . . , 3.
14w power physica tenting,and ? .- . reactor operat.or training chall I - - I '[.:- I' he perr.iltted, wit.h inoperable w - . co.ng ont nts as specified in 'l ' , ' . 3.5.1.a and 3.5.1.h nhove, ' provided that reactor coolant ?- - - , ,temperaturo is s212*r and the i reactor vessel is vented or rehet'or vossol head is removed.
, 1.. ... i . , , - . L6 .. . . 'Amendmenr. No. 33 120 l,...,.h- . , '
. , , ._s.
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-.. _ - - _ _ - - 3' ' $*,.q.%') ~ - - -...: --_ - ~:3 M.v. 'Ti'f ' ! . .
- v ,. ' ' ' s. .- r.,. .; ) Q , I .
! , JAFNPP 4.5 (Cont'd) !
3.5 (Cont,'d) Reactor Core Isolation Coollas ,'
, E.
i i E.
Reactor core Isolation Cooling _ l (RCIC) System _ ,. , i ) (RCIC) :s y s t em I ! 1.
RCIC S'ystem testingshallbe .
! 1.
The RCIC System shall be oper* performed as follows provided l able whenever there is'Irradi-a reactor steam supply is i sted fuel in the reactor vessel ' available..If stemis is not .f and tiie reactor pressure is available at the time the I greater than 150 psig and prior .survelliance test li's scheduled reactor startup from a . to be performed, the. test shallt l to a ! cold condition, except from the
be 'perforsted within ten days, s of continuous operation fgem _, ' ' f time that the RCIC System is r.a d e or found to be inoperable the time steam becomes available
continued , if . ' r
i for.any reason, .. . reactor popeer operation is Frequency Item por$ilssible during the suo- ,l
a. Simulated Once[ opera' tin'y - l ceeding 7 days unless the I system is made operable ear 11er Automatic-cycle.
I provided that during these 7 Actuation [ days the llPCI System is operable.
b.
Pump Oper-One.e/ month 2.
If the requirements of 3.5.B ability ' cas$ot be met, the reactor 'l
i shall be placed in tlie cold c. Motor Oper-Once/ month ' condition and pressure less ated Valve than 150 psig within 24 hours, Operability ,,,g, , ,, ,,,
g [ 3.
f.o w power physics testing and d. Flow Rate Once/3 months' h ' reactor ' operator training shall I be permit. Led with inoperable e. Testable Tested for , specified in ! components as Check Valves operability ! 3.5.E.2 above, provided that 'any time'the ' reactor-coolant, temperature is reactor is in " l 4 2120r.
.the cold cond! . , l tien esceading l 40' hours,'If
operah!!!ty ' tests have not been performed
' i ] during the pre cedleiy 31 days . . , , , , , , 'Amendeont 40 1'21' . '
- . ..... -,_ __ . ' h v i. :i, . ... k l ' ' ,. . ' JAFHPP 4.5 (cont'd) 3.5 (cont'd) The RCIC pump shall. deliver ' least 400 gym for a syste's at head corresponding to a. reactor pressure of 1,120 peig,to . 150 psig.
i '. ( . 2.
When it is determined that the ~ , RCIC system Ks inoperable st_a . time when it is required.to'be . operable, the IIPCI, System shall
, , he demonstrated to be' operable . d ' Immediately and daily thereatter.
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Minimm me m ncy Core and dinfrm:m nr.cracncy Core and contain-Containment Cooling System k' l __ - ' i.
. > ment coolino System Availabilitig Availability
1.
Any combination of inoperable Not Applicable.
components in the Core and Con- _ ' taimnent Cooling Systema shall . . not defeat the capability of the rem'aining operable - ' components to f ulfill the core i ,and containment cooling , . . ' < functions . . - . - . 2.
When the irradiated fuel is in , ' the' reactor vessel and the reactor is. _. in the cold ., ' condition, all LPCI, core , , ,, . .. ' spray, and containment c6oling - subsystemn may he inoperable - provided no work is beiny done i which hha the potential f9r I . draining the reactor vessel.
- G.
aintenance of Filled Discharge Pine G.
Maintenance of Filled Discharac! Pipe ' .
- l
. ' Whenever core spray subsystems, LPCI The following surveillance 1 - . requirements shall be adhered to,linI ' subsystems, IIPCI, or P.CIC are
. ' required to be operable, the . order to assure that the discharge: - discharge piping from the pump piping of the core spray subsystem,- .- discharge of these systems to the LPCI subsyntcm, IIPCI, and RC1C are-
s last block valve shall be filled.
filled: mi - a.
From and after the time 1.
Every month pr, lor to.'the, - , that the pump discharga testing of the LPCI,subspatem . piping of the llPCI, RCIC, and core spray subsystem, the ,l u- .- LPCI, or Core, Spray discharge piping of these,' Systems cannot be systems shall be vented from ,' ' . maintained in a filled the high point, and water flow' i ' observed.
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.. . . . . . . ' Amendment No. 4 122 l, .. . ..
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JaffPP , ' ! . >\\ ! 4,5 (cont'd) .) . , 3.5 (coat'd) - , . . - - i +, . i , , condition, that pump sinalt be considered inoper-2.
Followlag any period'where (Ae LPCI setsys -
,] able fo:. purposes satisfyles specifications tems or core spray subsystems have act been required to be-operable, the ' discharge '3'.5.4, 3.5.C. and'3.5.E.
i piplag of the inoperable system, shall be i , M.
Averste Planar Linear Nest Generation Rate vented. from the high polat prior to the , return of the system to service..'
- I
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J I (APLHCR) ' f,W. m ' j The APLHCR for each type of fuel as a function of 3.
Whenever the NPCI, ECIC 'er~ Core Syray Systes
- averags planar exposure shall not exceed the is 11aed ap to take section from the edaden- ' . limiting value shown in Figures 3.5-9 through sate storage tank, the discharge piping of j 3.5-11 for two loop operation.
Foe single loop the NPCI, RCIC, and Core Spray shall be vented from the high potat of the system, , j operation these values are reduced by multiplying and water flow observed os a mostkly basis.
by 0.84 (see specification 3.5.K.
Reference 1).
' If anytime during reactor power operation greater The level switches located ca the Core Spray , than 25% of rated power it is determined that the 4.
limiting value for APLHCR is being exceeded, and RNE system discharge piplag high points action shall then be initiated within 15 minutes which monitor these 11aes to lasure they are to restore operation to within the prescribed full shall be functionally tobt each month;' 't., s., W e.g $Z Q}% Rate
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7-limits.
If the APLHCR is not returned to withis Meat G meestica 9- ! the prescribed limits within two (2) hours, as R.
Averste Flamana Linear
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The reactor power shall TheAPLNGRforeachtypeoffuel'!.as/al,fusottomief'[[/. , I be reduced to less than 25% or rated power within j average planar esposure shall.be deterstaodidaily% ' - } .the next,four hours, or untti the APLHGR 'is returned to within the prescribed limits.
escing reactor operation at h25%.rety ; thermalN l p- - . , owe- . t i ' ! - . lt - ! i ,I ' .
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I j.,. . ..> , / .e.... Amendment No, p4I, p4, J, pf, 98,
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Qucar lleat-Gerv rat lon Itala (IllGn) . ...w ~
- tles linear leat generat.lon rato (IllGit) of any I. T.incar lleat Generation flate (likIR) ' -
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' , ,.;.gl ]. . mi liiany (nut masodily at any axial location ' ., : ;..., 11e IllGR shall be checked daily darinri
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reactor operation at ),25% rated thermal W ' ,, , power.
.
It anyt ine: storlevj scactor gnser oteration greater ' than 2S1 of rotal xwr it lu detenninal that Lie E , , limit ing value for IllGIt ju bcIng exceotlal, action i ' .19:11 then lo initiatol within 15 minul,:a lo re - I utore <p: rat.lon to ulthin the prescrlini limitu.
If the IIIGit la int returnal to ulthin (1i0 pre-t.crilot limitu within,tm (2) loirn, an onterly l' - -..... , reactor trust rolactlon t hall lx comencal inne-j,Y D ' ellatuly. *1150 reac' tor to.Jer slusil le rohunt to
. g '. '. leau thin ~2St et ratal loasr within tim su:xt four l !I c,- ,i , Iniru, or until the tilGlt la returtal in within '
' ' s.).c/ - , tlu prescrilol limitu.
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Thermal Hydraulle Stability J.
Thermal Nrdeaulle Stab 111tv , , 1.
Whenever the reactor is la the startup or 1.
Establish base 11ae APEN and LPEN neutros flex noise values within 2 hours of enterlag the [. run modes, two Reactor Coolant Systes ,, l recirculation loops shall be la operation, region for which monitorlag is required a - unless base 11 stas has been performed elece s . with: the last refueling outage. DeteM or levels A - i - ! and C of one LPEN string per core octant plus a. Total core flow greater than or l detectors A and C of one LPEN string la the , equal to 45 percent of rated, or i, - conter of the core should be moattored.
- i ' b. Thermal power less than or a,3ual to the limit specified in ,l , Figure 3.5-1 (Line A), i i , ! except as specified in Specifications I , - l 3.5.J.2 and 3.5.J.3.
I 2.
With two Reactor Coolant System recircula-e
, tion loops in operation and total core flow ) e less than 45 percent of rated, and thermal , power greater, than the limit specified in ,,' - . Q '
- Figure' 3.5-1 (Line A); or with one Reactor l.
f Coolant System loop, operating and thermal power greater than the limit speelfied la.
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Figure 3.5-1 (Line A): t.
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,s. Determine the APRM and LPRM noise levels: , ,i .,, i ' ,
1. Within 2 hours af ter reaching steady- , i_."
state within the regions of Figure - 3.5-1 where monitoring is required, l , ' and at least once per
hours . thereafter; and . g . ' - -
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discharge valve of the low-speed pump the f, l f aster pump is less than 50 percent of it's . . . ' l rated speed.'
With no Reactor Coolant System Recirculation i I 6, . loop in service, the reactor shall be placed ,'
,i In Hot Shutdown within 12 he.us.
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l. ::;V ydf 2. Within 2 hours after completing ' p.f.i_.:, *llU[ ' $h fd" # ) - ' .
' an increase in thermal power of 5 - @.+, k< W-j . ., percent or more of rated thermal ' 7" ? ??
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I.~^d;2x$ '.l .' ' - .ra? -.6 b. If the APRM and LPRN neutron flux noise - ! levels are greater than 5 percent and.
t ' 'd"df[ . . 4ddF greater than three times their . - '"'" - ' 8 '.*". established baseline noise levels, . J initiate corrective action within
' ! minutes to restore the noise levels to within the required limits within
y, hours, by increasing core flow and/or reducing thermal power.
, 3.
If during single-loop operation, core thermal power is greater than the limit
" defined by line A of Figure 3.5-1, and core ' ' flow is less than 39 percent, inuned t ately initiate corrective action to restore core thermal power and/or core flow to within the l ' l _ limits, spec'ified in Figure 3.5-1, by increasing core flow and/or initiating an I ' i orderly reduction of core thermal power by 'l ' inserting control rods.
' '
,i . ' ! 4.
The requirements applicable to single-loop , operation in Specifications 1.1.A.
2.1.A.
l 3.1.A.
3.1.B. 3.2.C and 3.5.H shall be.(n effect within 8 hours following the removal j , of one recirculation loop from service, or j ' * ' ' ' ' the reactor shall be placed in the hot l shutdown condition.
i l' , ' ' . . i , ' Amendment No. 98, , ' 124b
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Specifications 3.5.J.1, 3.5.J.2 and 3.5.J.3 ,
Stability Stability Monitoring Mmitoring (APRM and LPRM) Required Stability Monitoring (AP M and For Single Loop ( ' ' ' During Two-Loop Operation Required d li , ', ' (APM and LPM) Required um) Operation , 60 - ' gW Line A - b Single and ' I ho-Icop Operation b 50 - y Single-Loop Operation u Prohibited ' $ i >
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,' , 70 CORE FLOW (PERCENT RATED) AmendmentNo.)(,g,g,p 134 , '
- - - - - ,
ATTACHMENT 3 . FITZPATRICK R0 EXAMINATION Facility Comments and NRC Resolution to Reactor Operator Examination The following represent those facility comments that were not resolved during the post examination review or resulted in a significant change to the answer key.
Comments that were minor in nature, i.e., typos, set pt. changes, are reflected on the master answer key.
Question 1.07 Facility Comment: Typically, a restart will cause an increase in the removal rate of Xenon from the core.
It is difficult to determine at what power level the newly intro-duced burnout rate will overcome the production of Xenon from Iodine left after the scram without some rather complex mathematical manipulations.
It was felt, by some, that this question was misleading and/or impossible to answer unless given further information.
NRC Resolution: Comment not accepted.
If further information was needed or a more direct interpretation of the question was desired, it should have been expressed during the examination. The examiner received no questions on this during the examination. The four choices.given clearly reflect wide variances in the effect of Xenon behavior following a restart of the reactor after a scram from full power.
It was not the intent of the examiner to have mathematical justification expressed as part of the answer key.
Question 3.01 Facility Comment: ! Question wording does not seem to elicit the response listed in the key. Also, problems with key answers as follows: a.
Second half concerning 44% runback should be omitted as this would not correspond to normal response, d.
There should be no requirement for memorizing where approximately 30' instruments are powered from, or not required powered from, as in this , case.
- . k i
Fitzpatrick R0 Examination 2- . NRC Resolution: (a) Comment accepted. Candidates'will be. graded to the recirculation pump runback of 26% with < 20% F.W. flow.
(b)(c)(d) Candidates will be graded on overall system or component response.
Answer key will not change.
Question 3.10 Facility Comment: Since answering all three parts of this question depends upon one knowledge item, we feel that it is a triple-jeopardy situation and at best is too-heavily weighted.
NRC Resolution: Comment will be considered during grading.
If candidate selects an incorrect rod group for the initial condition, the remaining parts of the answer required will be based on this assumption. A clear understanding of what should be displayed in the RWM window, based on candidates assumption, is requi red.- Question 4.09 Facility Comment: This question is inappropriate and should be removed from the examination.
The candidate should know that, under certain conditions of level and temperature level, indications may not be accurate and that there is guidance given in the E0Ps. They should not be held responsible for memorizing tables of numbers that they would normally be able to look at in the daily operation of the plant.
Part d answer should be OK E0P caution #6 list if level <70" and Drywell temperature >169 F then indication is unreliable.
NRC Resolution: The JAF training material clearly emphasizes that one of the student learning objectives.is to be able to recognize certain adverse conditions when level indication may not be reliable. The candidates will not be graded on memort-zation of numbers required for level indication reliability. The required answers for A-D will be to state if the level instrument is or is not reliable only, given the conditions of level and temperature indication.
The training department should clarify the student learning objective, SLO, F-EOP-1.01.01, to more accurately reflect what is required.
Part d answer is changed to OK.
- -. . -- -. - - ..
Fitzpatrick SR0 Examination Facility Comments and NRC Resolution to Senior Reactor Operator Examination The following represent those facility comments that were not resolved during the past exam review or resulted in a significant change to the answer key comments that were minor in nature,.i.e. typo's, set pt. changes'are reflected on the master answer key.
Question 5.03-a Facility Comment: Could include "Not all rods inserted" as a reactivity NRC Resolution: Accepted "Not all rods in" added and facility agreed to modify lesson plan from which this answer was taken.
Question 5.07-a - Facility Comment: .This question may not elicit the response listed ~1n the answer key.
The candidate may interpret the question as to the thermal response of the reactor and not the reactivity effects in response to a transient.
NRC Resolution: Accepted. Alternate answers will be evaluated during grading.
. Question 5.10-c Facility Comment: At power as pump speed increases, reactor power increases, causing feedwater flow to increase. As feedwater. flow increases available NPSH will also increase due to increased sub cooling in the downcomer region. This will override the lower pump suction pressure at the eye.
NRC Resolution: Accepted. Alternate answer reflects a "real plant" approach and demonstrates an understanding of the effect of subcooling on NPSH available.
Question 6.03 - See R0 comment 3.10 ._ _ _ _ .
Fitzpatrick SR0 Examination
. Question 6.06-b Facility Comment: .The answer to this question is not easily attained. With some research into this matter, it now appears that one emergency diesel generator supplying the emergency bus alone may in fact be overic ided.
One emergency generator will be fully loaded with just the ECCS pumps running at rated conditions. The deciding factor, if the generator is overloaded or not, will be the remaining loads which could be running on the emergency Motor Control Centers. So'at this time we've concluded that operator action would probably be required to limit the load on the operating emergency generator.
This action would include close monitoring of generator load and selective tripping of unnecessary loads to prevent generator loss.
This questior, has highlighted a training deficiency, so candidate response will probably not reflect this information.
NRC Resolution: . Accepted. This situation is possible yet not adequately covered in the training' material. Answers will be considered during grading.
Question 6.07 Facility Comment: Candidates. required to know low pressure pump running at approx. 100 psi don't necessarily distinguish between pumps.
-NRC Resolution: ~ Not accepted. JAFNP Enabling Objectives 025-1.21 in the ADS Lesson Plan requires candidates to know ADS initiating signals including setpoints.
Question 7.02-a Facility Comment: No immediate operator actions --- Candidate not required to memorize procedures.
NRC Resolution: Accepted. Although not labeled as immediate, immediate operator action is necessary to control vessel level when this transient occurs. Answer key modified to reflect that level control is required by manual feed pump control.
. _ _ _ __
~ Fitzpatrick SRO Examination 3~ Question 7.05 Facility Comment: Could reference operator aid on 09-04 panel instead of E0P's caution #16.
NRC Resolution: - Accepted. This fact is not indicated in training material.
Question 7.06 - See R0 comment 4.09.
. Question 7.10 Facility Comment: Scram solenoid indicating lights can be checked on 09-05 panel NRC Resolution: Accepted.
Panel 9-5 added, although not mentioned in F-0P-18 as an acceptable method.
Question 8.06 Facility Comment: Student may assume RAP is in accordance with technical specification and answer in accordance with technical specification.
NRC Resolution: Accepted.
Facility will revise RAP to agree with Technical Specifications.
Question 8.08 Facility Comment: This question calls for the candidate to memorize Tech Spec which is not required. Also wording (interpret) could lead the candidate to statements 1 and 2 under that policy statement.
NRC Resolution: Accepted. Wording of question could lead candidate to other areas of policy statement. Reference to these areas is acceptable. }}