ML20126L736
| ML20126L736 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 06/04/1981 |
| From: | Colbert W DETROIT EDISON CO. |
| To: | Kintner L Office of Nuclear Reactor Regulation |
| References | |
| RTR-REGGD-01.075, RTR-REGGD-1.075 EF2-53-455, NUDOCS 8106080346 | |
| Download: ML20126L736 (30) | |
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Deholt Edison f5E!!!ENa June 4, 1981 EF2 - 53,455 I
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Mr. L. L. Kintner Division of Project Management
,D Office of Nuclear Regulation 6-L,[]
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4 JUN O51981hL U. S. Nuclear Regulatory Ccamtission Washington, D. C.
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Dear Mr. Kintner:
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Reference:
Enrico Fermi Atomic Power Plant Unit 2 NRC Docket 50-341
Subject:
Responses to Questions frcm Gerry Mauch i
Please find enclosed responses to several questions asked by Gerry Mauch. This infouration will be incorporated as appro-priate into a forthecming amendment to the FSAR.
Subjects: e Regulatory Guide 1.75 e A'IWS "Monticello Fix" e Item II.D.3 - S/R Valve Ta11 pipe Thermocouples i
e ADS Option 2 e Seismic Recorder o Bypass Valve Surveillance e Level 8 Trip Surveillance e FSAR 7.3 Omissions Yours truly,
/
S William F. Colbert Technical Director Enrico Fermi Unit 2 WEC:RMB/cIm
- 1S" TfllS DOCUMENT CONTAINS POOR QUAllTY PAGES d'I 8106080}Uh 9 ',
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June 4, 1981 2-EF2 - 53,455 Mr. L. L. Kintner bec:
R. M. Berg F. E. Gregor J. W. Honkala E. Lusis L. E. Schuennan A. E. Wegele L. F. Wooden Document Control l
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SUBJECT:
Regulatory Guide 1.75 Compliance In reply to the.NRC Staff question on the applicability of Regulatory Guide 1.75 to the Fermi 2 design, Edison believes that adequate discussion of the existing design and its acceptability exists presently in the body of the FSAR.
The compliance statement can be found in Appendix A.
A summary of.the physical separation'can be found in Section 3.12.3.2.3.
A discussion of system independence is located in Section
. 8.3.1.4.
' Edison's reply to a formal staff question which required that the specific Fermi 2 design aspects which do not meet the regulatory position of Regulatory Guide 1.75, Revision I, be justified is found in Appendix E on page E.2.222-4c as the response to question 222.2 (3.12).
L. F. Wooden
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3 s r;.& increate in recirculation f * :w e.
of the nederator by increaning thete.?orarily reduces the void
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confen flow of coolant through the core.
the reactivity of the core,The additional neutron moderation increases causing the increase.
The reactor power level to etcam volume in the core with a consequentincreaacd steam genera the
- effect, When recirculation floand a new steady-stato power level is established. negative reac in the reverse manner.w is reduced, the power level is reduced Figure 7.7-13 illustrates how the RFCS operates with the turbine controls for automatic load following.
in conjunction Each recirculation pump motor has its own motor-generator set for a power supply.
between the motor and generator of the motor generator setA vari To change the speed of the reactor recirculation pung variabic speed converter varies the generator speed which the changes the frequency and magnitude of the voltage supplied the pump motor so.that the desired pump speed is attained.
to The RFCS uses a demand signal from either the operator or the main plant turbine-generator speed governing mechanism demand signal is supplied to the master con troller The from the master controller adjusts A signal epced controller for each motor generator setthe speed setting of the converter.
1 variable speed converter.The master controller signal adjusts each motor g The master controller signal is compared with the actual speed of the generator by the speed controller.
The speed controller signal causes adjustment the speed converter, of epeed until the feedback from the generator equalsresulting in a change controller signal.
the master j
The reactor power change resulting from the change in
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recirculation flow causes the initial pressure regulator to raposition the turbine control valves.
cignal was a turbine load / speed error signal,If the original demand responds to the change in reactor power level by adjusting the turbine turbine control valves until the load / speed error signal isthe reduced to zero, l(p5gg,)u 7.7.1.3.3.2
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Each of the two motor generator sets and its controls iside motor generator set.
Figure 7.7-12 shows the general
' crrangement and rating of the motor generator set.
generator set the motor-et any speed between approximately 19 percentcan continuously supply powe
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of the drive motor speed.
and 96 percent The motor generator set is capable
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A In order to mitigate the gf fects of an ATWS event, a provi-sion has been included in the Fermi 2 design to trip the re-circulation pump motor generator field breakers using a specific logic philosophy defined as'the "Monticello fix."
This logic 6 includes teactor pressure and level r initiating signals W
which wSEE interf ace with redundant breaker trip coils in each MG set.
Suitable redundant time delay relays will be Provided to maximize the single failure.W withstand capacility of the design.
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',g H.II.o.3 Valve Position Ind(2af.icn
([f H.II.D.3.1 Statement of Concern The operator needs a posi'tive ind ica tion of power-opera t ed relief-valve and safety-valve positions to provide additional a.c arance that the operator will correctly diagnose plant transients that potentially involve the opening of relief or safety valves.
8 II.D.3.2 NRC Position Reactor coolant system relief and safety valves shall be provided with a positive indication in the control room derived from a re-liable valve-position-detection device or a reliable indication of flow in the discharge pipe, which meet the following' requirements:
a.
The basic requirement is to provide the operator with unambiguous indication of valvo position (open nr
. closed) so that appropriate operator actions can be taken.
b.
The valve position should be indicated in the control room.
An alarm should be provided in conjunction vith this indication.
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c.
The valve-position indication may be safety grade.
.If the position indication is not safety grade, a t.
reliable single-channel direct ind ica t ion pcwe r ed from a vital instrument bus may be provided
.f tackup methods of determining valve position are ava t '.2cle and are discussed in the emergency procedures as an aid to operator diagnosis of an action.
d.
The valve position indication should be seismically qualified consistent with the component or system to which it is attached.
- e.
The position indication should be qualified for its appropriate environment (any transient or accident that would cause the relief or safety valve to lift) and in accordance with NRC Order, May 23, 1980 (CLI-20-81).
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It is important that the displays and controls added to the control room as a result of this requirement w
0 not increase the potential for operator er ror.
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human-f actors analysis should be per formed tak ing into consideration--
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The use of this information by an operator during' both normal and abnormal plant conditions j(%
2.
Integration into emergency procedures E.II.D.3-1 Amendment 33 - March 1931 I
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Other alarms during emergency and need for priori-tization of alarms H.II.D.3.3 octroit Edison Position The Fermi 2 plant has a valve-position-indication system that uses one pressure switch on each of the 15 safety and relief valves.
The :M.
tailpipe temperature-monitoring system provides a backup $i ]', monitors valve,1eakage as its normal function.
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H.II.D.3.1 Modifications H.II.D.3.4.1 Desien Basis An in-containment, tailpipe-mounted, pressure-switch system provides status information via control room indicating lights of the relief valves and safety valves.
This system provides the following information to the plant' operator during normal and abnormal operating conditions:
a.
Positive indication of valve position, including the e
stuck-open valve condition b.
Positive identification of the specific valve or valves that c e open c.
Annunciation of the activation of the automatic depressurization system (ADS) in the control room By being provided with the immediate indication and annunciation of the valve opening and the identification of the valve, the plant operator can initiate recommended actions to control or 1
rectify the situation.
The NRC han specified in NUREG-0 578 and NUREG-07 37 (References 1 and 2) that components of the safety / relief valve (S RV) monitor system must be quali f ied for the appropriate environment 3',
condi-i tions to be experienced under normal and abnormal conditions of Plant operation.
These environmental conditions include temper-atore, pressure, and humidity, and also the seismic acceleration of the component or system to which the components of the SRV
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' system are attached.
, The system instruments are quall'fied to IEEE 323-1974 and l-IEEE 344-1975.
.The power for this system comes from a reliable source that is not affected by the loss of offsite power.
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SUBJECT:
OPEN I&C ITD1 ON ADS IDGIC CHANGE At the-I&C exit interview, the staff reviewer requested that the specific design nodification to the ADS initiation logic be identified, the attachment indicates clearly the methcd being employed to mitigate this concern.
L..F. Wooden
/dk-Attachment 4-81
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EF-2-FSAR H.II.K.3.18 Modification of Automatic Dooressurization System Logic--Feasibility Study and Modification for Increased Diversity for Some Event Sequences H.II.K.3.18.1 Statement of Concern
.The automatic depressurization system (ADS) requires manual actu-ation in some' event sequences to ensure adequate core cooling.
H.II.K.3.18.2 NRC Position 33 The ADS actuation logic should be modified to eliminate the need for manual' actuation to ensure adequate core cooling.
A feasibil-ity and' risk assessment study-is required to determine the optimum approach.
One-possible scheme that should be considered is ADS actuation on low reactor-vessel-water level provided no high-pressure coolant injection (HPCI) or high-pressure coolant system
-(HPCS ) flow: exists and a low-pressure emergency core' cooling (ECC) system is running.
This logic would complement, not replace, the existing ADS actuation logic (Reference 1).
H.II.K.3.18.3 Detroit Edison Position (a) l35 The ADS logic modifications that will eliminate the need for man-ual actuation. to ensure adequate core cooling have been studied by General Electric as a generic item on behalf of the BWR Owners' 33 Group.
A feasibility and risk assessment study has been performed to determine the optimum approach.
The results of this study have been provided to the NRC.
Five options, including retaining the current design, were considered.
The results showed that the addition of a bypass of the high drywell pressure trip if the reactor water level remains below the low-pressure ECCS initiation 35 setpoint for a sustained period, or the elimination of 5.he high drywell pressure trip, are the preferred concepts.
Based on our i
evaluation of these recommended changes and associated risks, the optimum approach for' Fermi 2 is described below.
l33 Of the two' alternatives recommended by the BWR Owners ' Group study, the addition of a brjass to the high drywell pressure trip if the reactor water level remains below the low-pressure ECCS initiation setpoint for a sustained period, is judged to be the preferred solution.
This.will be-accomplished by installing a " bypass" timer activated on low water level (Level 1).
When this timer runs out, the high 35 i
'drywell pressure trip is bypassed and the ADS is initiated on l-I
,a.
The Detroit Edison position was revised in response to a ques-tion from the NRC staff that was transmitted informally to
-Edison in-a meeting held on April 22, 1981.
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H.II.K.3.18-1 Amendment 35 - May 1981 l
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EF-2-FSAR water level alone.
This additional logic does not affect the high drywell pressure - low water level initiation sequence for pipe breaks inside the drywell.
Figures H.II.K.3.18-1 and H.II.K.3.18-2 show the logic for this alternative, with the bypass timer started on Level 1.
A time delay of 8 minutes Mas been chosen for the Fermi 2 plant from the analyses presented in NEDO-24708 (Figure Group 3.5.2.1-33).
The results of these analyses demonstrate that adequate core cool-35 ing is ensured for isolation events, even with the ADS blowdown delayed 10 minutes after Level 1.
Figure Group 3-1 shows the i
same analysis assuming a stuck-open relief valve.
Starting the bypass timer at Level 1 allows the operator enough time to control the system manually and still ensure automatic depressurization in time to prevent excessive fuel heat-up, even under the worst case conditions described above.
H.II.K.3.18.4 Modifications The descriptions of proposed modifications will be provided no later than 4 months before the scheduled issuance of an operating license for Fermi 2.
H.II.K.3.18.5 Schedule The modifications will be completed by the end of the first 33 refueling, which is at least 6 months af ter the approval of the proposed changes by the NRC.
H.II.K.3.18.6 References 1.
U.S. Nuclear Regulatory Commission, NRC Action Plan Developed as a Result of the TMI-2 Accident, NUREG-0660, May 1980; Revi-sion 1, August 1580.
2.
U.S.
Nuclear Regulatory Commission, Clarification of TMI Action Plan Requirements, NUREG-0737, October 1980.
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i H.II.K.3.18-2 Amendment 35 - May 1981 1
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Low Water Level (Lavelil Seal m
In if High Orywell 0
Pressurs O W ***
Timer Seal in Y
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Low Water L.svel (Level 1) 1f Confirm Water Level is Below Scram Level i
lf Note:
The 120-second actuation timer 20,Seco
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,,,a,v i i,io Actuation ecovers cefore it time out. The Timer timer will rutart if ene low level signal occurs esain.
If Low-Pressure ECCS Pumos Running ENRICO FERMI ATOMIC POWER PLANT e
3r UNIT 2 FINAL SAFETY ANALYSIS REPORT AOS FIGURE H.ll.K.3.18-2 Actuation ADO DELAYED BYPASS OF DRYWELL PRESSURE TRIP WITH BYPASS TIMER STARTED AT LOW WATER LEVEL (LEVEL 1)
AMENDMENT 35 - MAY 1981
9 i
SUBJECT:
Fermi 2 Recorder Seismic Qualification An informal request that Edison resubmit copies of the recorder seismic qualification was made by the staff
- reviewer at the I&C exit interview.
The information requested is attached as agreed.
L/ F/ Wooden
/dk Attachment 6-4-81 i
Chernolt EC! son ____.
ENRICO FERMI UNIT 2 PROJECT Instrumentation & Controls FF2 - 53,110 Date:
June 4, 1981 To:
File - H11 F rc.n:
R. W. Bart' kW bCh MM.
Supervising Engineer
Subject:
Shake Test of Hil-P602 and Recorder T50-R800-B.-Trip Report This letter documents witness of a Seismic Shake Test by C. A. Schulte, August 10, 1977, at the General Electric facility at San Jose, Californi.i.
As part of the Seismic Shake Test for Control Room Panel till-P602, a type "W", L/N Speedomax, 12 point recorder, was energized and monitored during the shake testing.
This recorder, "Drywell and Torus Wall and Air Temperature Recorder, T50-R800-B", mounted on panel insert 602C520 was energized with a simulated signal to read approximately h scale.
Readings were compare.1 before and after the test.
It was found'that after the shake test, the recorder continued to record the proper value as called for by the simulated input.
No damage to the recorder was noted after the test was completed.
This letter replaces the original trip report, written in September 1977, which is now not available.
General Electric has documented the seismic test as follows:
1.
Seismic Test Report - Summary Benchboard Hil-P602 - TDCC 3478, 8/30/78.
2.
Complete Test Report - ATR - 510 - EAD05-01, 8-77, available at General Electric.
l Written by:
G. A. Schulte
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, m.- y p -. -....,.,,e.-.,.- ....4 ..+,,y [ ,..+. +. +.., -. +,9.-,.... .+...e.. 6w++. v4.. 04ps YZ SH Sgs &&+ ..., 7 .+-+~... W. .++49 .+ +. 6 +... +. +. ..,.,+ 10g5 YZ $ H2 140gs ..-4: Q ,W-L M..- -. ~... L. ....+-4. . ~... ~..3ogs YZ 5 H2 30 0 g s
- * - - + + * * ~ -
f a,.... 6 60 0495 YZ 24 H .10 0 g s t o ~ ,q +-- 1 C' g s YZ 24 H 27 g s
- ~ ~ * - ~ ~ -
..+: - + h. 30gs YZ 24 H: 50 g s a- - + + - + - - - - - * - w .+... -. -. ...~. O49s XZ 13 H2 9 9s .'-.~ ~.-.' I . M .L -+ n. y" .... l l .+..
- + *
+. .. +.... g t0gs xZ 13 Ht 18Qgs U +++-+M+-*- - + - + + ~ * - - - - + - - +++ - - -. 30gs XZ 13 Hg 60 0 9s e .-.5-+ + y. >.. ...+..... ,e. 40 ~~. 6..... +. + + + -. +.. . -. ~... ~ . y + -+ 7. -. .t e- .. +. - 7 +, :: + -- A... Yd AA .7 i '20 6 f.. 4. .m.
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- 3. Speedomax H Multipoint
[ Cat. No. 315101-000 0044 6 001095170 381-038 Serial No. C 69 M543 41 h" r. Range 010 mv i Accuracy : 15% of span a) X.Z Axis &i Resonance was ceserved at 13 Hz 5 Endurance Testing: I 13 Hz @ 4 g-No incident Accuracy : 15% Ly 13 Hz @ t 0 g-No incident Accuracy : 15% l 13 H2 @ 3 0 g s- 'cout evd snorted to case. Olomng fuse i, instanedelectricaitaceoncaseanc compieledrun Accuracy : 50 % b) Y Z Axis Resonance was coser.ec as 5 Hz & 24 M: ( Endurance Testing: 1 5 H & 24 H2 @ 4 g-NC ccident Accuracy 1.15% 5 H2 & 24 H2 @ t 0 ;)~No ac; cent Accuracy 115% 5 Hz @ 3 g s -ir out card fed out ireciaced card) Accuracy 5% 24 H2 @ 3 g s-incut card tell out (replaced card) Accuracy : 5% c) Accurecy after testing : 15% of span Leeds & Northrup 4 F e P ,m-... -,. - -
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,+; = t-++-* j..,..+ f+; 7. p ...-.-t. y g-g.. 4-+-+.% -y ...n. . -.-. -.... y y e i '*-~ 3 ! l i t i i 1 i i i-i i i! I I 80 -.-.t-*A-,+-- i i - SPEEDOMAX W STRIP CHART 5 4 (0 2 g input) .. 3. +- ..-+.! ..--..-..o..
- 7.., - +-f..
.- + .-.. +...., .... a4-L- "G* FORCE "G* FORCE r... A, .-.L H.:: lNPUT AXIS FREQ. OUTPUT .. + -. ,[- e~ ...n. 5 60 ..-_.L .~.....w.- .m+-.--,--, l 04g YZ t$ H2 299'S -...p.. -.. . + - ~ i-l. ,,d ~ ~ ^ * ' ' * * * + ~ 10 9 YZ 15 H2 7gS ' *T. I T.
- Y,-
i -5 3 0 g's YZ 15 Hz 13 g s
- + M+-
04g XZ 10 Hz l$gs !i![ ( '!r - i i 4 r N. t~ 10g XZ 10 Hz 4gs - i. g 30gs XZ 10 Hz 10 g4 i ' 04g XZ 20 Hz 1 $ g's ,e 10g XZ 20 Hg 33gs 7~~.~..-..__., 30gs XZ 20 H: 98gs 'i l -.- 4 4 -+ w +- 1 i f t u i .- j -.4. h .,Y.l AAlf. ...j g .........,..... - +...-~.i .. 5.... d A4..-.tt.-.Ot- * ?? ' ...-.). .../. s.. y.- *t-* ....a .,..\\.. y s+ +- ....+. ...s f.......,r. g......-... .. ~ ~. ..-.p. - a : ......-.},........-.~..+.-+-.+.6- -..+-.-+p----4 i 'O 20 2$ 30 35 FAEOUENCY(HZ)
- 4. Speedomax W Strip Chart Cat. No. W t t COC-0CJ4 6-06C{JO 033 60t Serial No. D 6a 64809 4.t Range' 10 ~y Accuracy I.15% of Scan a) X.Z Axis
% c m was oese'ved at 10 H: 5 20 Hz Endurance Testing: ! *.. ', > < / @ 4 g -No inc.deat Ac %'1c, : 15% i J HJ & i '; g -NO ric cMf ACCuF, *5% l 7: **/ 4.d t 0 a Pac + ' 'r,J 'ed o't i!:gnt&1 sc' gi Arc a ac,. 25% r '. ~; M 31 s - No nc de' i ract i1Sh .o "i G 3 : s-No ec ceN Arn'ac, t 5% b) Y Z Asis ' r " was (0ser ed at 15 Hz v Endurance Testing: + G.
- 1.No,c.cee-t A.
nc, r 25% - Hl ti ! '. g - % nc cent A, ac, ; 5% ea f.'a, O s. P a;,. * 'o a We ot* A.-, e y., ,5% c) Accuracy af ter testing
- 15 4
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- G*FOACE e
. i.. INPUT Axis F Ar.0. OUTPuf W_ ........ +. I ., 4 L 4..-. --...4-, . p 9.. 0 4 ~ 04gs - YZ 26 H2 4gs 5 60 l '~.*..*.*.*.-~**M.'.-.'..*..-.'. ..j4+... ..e-..,* togs YZ 26 Hz to g s 30gs YZ 26 Hz 30 g s -+++..?....f....... 4'j--, 04q$ YZ 8 Hg 2g5 w ay }.. _ togs YZ 8 He 4gs ~- - ~ + - * - - i 30gs Y2 8 Hz 30 g s o I. 04gs x2 9H 6gs I' 4
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- 5. Speedomax W Multipoint Cal No. 513-10t.000 0044 6 003 038 60t Serial No. C 69 69829 t l Range 0. 0 mv Accuracy : t5% of scan a) X.Z Axis Resonance was coserved at 9 HZ Endurance Testing:
Accuracy : !$4 9 H2 @ 4 g-No incident Accuracy 254 9 Hz @ t 0 g-No incident Accuracy Il ?- 9 Hz @ 3 0 g s-No incident b) Y.Z Aufs Resona::e was ceservec at 3 H2. 3 HZ & 26 HZ Endurance Testing: Accuracy : I54 3 HZ. 8 Hz & 26 H @ 4 g-No inc dent Accuracy : 25. 3 Hz. 8 Hz & 26 Hz @ l 0 g-No mc: dent Accuracy : 5% 3 Hz @ 3 g s-No incident Accuracy : 75% 8 Hz @ 3 g s-incut card fell out Accuracy : 5% 26 Hz @ 3 g s Aear cover teil off ' c) Accuracy af ter testing : 15% of span v FA & H Mit A Lic< 0 ro. N W A trD) d:C Z H 3 2 I T-M 3 - ?C:0 .o -,Leed.s & Northrup
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- 6. Speedomax W/ L Cat.No. *c0 30 A t 34131 1029 4359 6.r0000.C01772 038 712 722 Senal No, C *2 U+i49 Range'tC.
Accuracy : 1% :.' gan a) X Z Axis p..e f r. .as of..;; a ) Ht Endurance Testing: Accu'acy : 15% . Jd 0 J '; - W "". deaf Accu x< : 25% a
- r (dt01-Ne co.-t Accuracy - 50%
~.. 9 "; A 3 0 g s-Na eccert b) Y.Z Axis .. W a a ~ o /. n :CseNeJ Endurance Testing: Ac:vx v : 25"o l u.. gi 4 '.c mcfor AW AC /
- 57" 3a "i s t J g '. cc4e !
a j 33 M; (;f 3 y.;5. O p p."'S '" Off 3.5 Accu acv 75'o r l uo ; 'O : p..r. 'a.'Or i aa j C) Accuracy af ter testing
- 15% c,! mn l
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- 7. Speedomax G Strip Chart Cat. No. 60f 01 Serial No. 346 51412 1 1 I
Range 0.tC rv Accuracy :.15% of span i a) X.Z Axis i Aesonance was cosemed at 5 H2 I Endurance Testing: 5 Hz @ 4 g-No incident Accuracy : 15% 5 Hz @ 10 g-No incider't Accuracy 25% 5 Hz @ 3 0 g s-Connect on la amchfier encocer orone a 1 after 25 see Replaced enccoer and continvec teshng Accuracy : 755 b) Y.Z Axis Resonance was coserved at 5 Hz Endurance Testing: 5 Hz @ 4 g-No incident Accuracy : 15% Hz @ 10 g-No incident Accuracy : 25% 5 Hz @ 3 0 g s-No incident Accuracy 115% c) Accuracy af ter testing : 15% of scan I. I Leeds & Northrup '3
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- 8. Speedomax Mark lli M Cat. No. $ %~
Serial No. P- 'ct.ce Range Accuracy. 'Y: cf span a) X.Z Axis..is. r 7,ric at 6 Hi &,'O Hz Endurance Testing: Ac'./ac,
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- 9. Speedomax 660 cat.No.%6 33 30 30 30-ISO 180180 000 0210-0240 0240 6 055 Serial No. M t I t 1 1 1 Range 010 rv Accuracy : 15% of scan l
l a) X.Z Axis WI Den cecks trea cown l Resonance was coservec at t5 HZ Endurance Tesung-l 15 H'. @ 4 g-No.ncident Accuracy : 25% 15 Hz @ t ') g.-NO cccent. Accuracy : 75 % b) Y.Z Axis A ;,en cecks trec down 1 Resonacce was ocserved at 15 HZ & 18 HZ [ Endurance Testing: 15 HZ @ 4 g-No encic?nt Accuracy : 25% 15 H2 @ t 0 g-No ccicent Accuracy : 75% 18 HZ @ 4 g-No.ncicent Accuracy 25% D 18.Hz @ 10 g-No incicent Accuracy : 75% c) Accuracy after testing : 15% Leeds & Northrup 10
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e h 4 I t 4 .h-SUBJFCf:- TURBINE BYPASS VALVE SURVEHLANCE i I ' This subject is addressed in response to NBC' Staff C*Jeetion 212.64A. t V 1 r i C L. F. Wooden /dk 6-4-81 e I h i L L t 1 Y \\ e i, .= I 1 !( i f. 4_ ..,,....r..~ .,..,,..,,_,-.-,.~.....,_.~.,._..m_,_,........,_.-, .m -
~ W 4 g i. -F SUELTECT: FEEIMATER LEVEL 8 TRIP SURVEILLANCE This subject is addressed in response' to MC Staff Question 212.64A. i. n L. F. Wooden -/dk r 6-4-81 y I s W c l 6 h E k i .?e<,"
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SUBJECT:
Disposition of Certain Ommissions in i Section 7.3'of the Fermi 2 FSAR Edison considered the staff reviewer 3' requests that both the Containment Spray mode of.the RHR System and the Main l Steam Isclation Valve Leakage Control System be added to Sectioni.3 of the FSAR. The Containment Spray mode of the RHR System is discussed-in FSAR Sections 5.5.7.3.3 and 6.2.1.3.2. No credit has been taken for this system in any safety analysis, there-fore, a description of the system should not be included in'Section 7.3. F The MSIV LCS was..added subsequent to the initial authorship of the FSAR. A fairly complete discussion of the system and its safety design criteria can be found in Appendix 9A. Direct formal reference to Appendix 9A will be made in i Section 7.3 of the FSAR as shown on the attachment. L. F. Wooden l~ /dk' Attachment 6-4-81 \\~ i l _,.._._a. _ _._._. _ _ _. _. _. _. _ _.. _, _ _ _. _. _.. _,, _ _ _ _ -..,,., _. -, _ _ _ _.,,, _. -..,
EFo 2-FS AR 7.3 ENGINEERED SAFETY FEATURE SYSTEMS di Included in this section are descriptions and analyses of the instrumentation and controls for the following ESF systems: a. Emergency core cooling system b. Primary containment and reactor vessel isolation control system c. Emergency core cooling auxiliary system d. Emergency equipment cooling water system e. Main control room atmospheric control system f. Standby gas treatment system g. Standby power system - h; Post-LOCA combustible gas control system. - The format of this section departs from the Standard Format Guide in that the description and analysis are grouped together under each system heading rather than by descriptions and by analyse f 7.3.1 Emergency Core Cooling System 7.3.1.1 Design Basis Information l The design basis information for the ECCS, required by Section 3 of IEEE 279-1971, is provided in Subsection 7.1.2.1.3. 7.3.1.2
System Description
The ECCS includes the following subsystems: a. High pressure coolant injection system (HPCI) b. Automatic depressurization system (ADS) c. Core spray system d. Low pressure coolant injection (LPCI) mode of the RHR system. N ,The purpose of ECCS instrumentation' and control is to initiate appropriate responses from the ECCS to ensure that the fuel is adequately cooled in the event of a design basis -LOCA accident. The cooling provided by the system restricts the release of radioactive materials from the fuel by preventing or limiting the extent of fuel damage following situations in which reac, tor coolant is lost from the NSSS. 7.3-1 ~ _ _.. _ _.. _.
y mSERP 4 A discussion of the Main Steam Isolation Valve Leakage Control System is incorporated in Appendix 9A. L. F. R:cden /dk 6-4-81 ,-}}