ML20077C996
| ML20077C996 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 07/21/1983 |
| From: | Schroeder C COMMONWEALTH EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| 6999N, NUDOCS 8307260314 | |
| Download: ML20077C996 (5) | |
Text
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Commonwealth Edison 3i.7 Address Reply to: Post O$ce Box 767 one First National Plaza Chicsgo, Ittinois
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Chicago. Illinois 60690 July 21, 1983 Mr.' Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 Subjects LaSalle County Station Units 1 and 2 Revision to FSAR Table 14.2-123, Relief Valves Startup Test NRC Docket Nos. 50-373 and 50-374-Reference (a):
License NPF-ll, LaSalle County Station Unit 1.
Dear Mr. Denton:
Reference (a) states, in part,:
"2.C.(28)
Initial Test Program (Section 14, SER)
The licensee shall conduct the post-fuel-loading initial test program (set forth in Section 14 of the licensee's Final Safety Analysis Report, as amended) without making any major modifications of this program unless inodifica-tions have been identified and have received prior NRC approval.
Major modifications are defined as:
(a):
Elimination of any test identified in Section 14 of the licensee's Final' Safety Analysis Report, as amended as being essential; (b):
Modification of test objectives,-methods or acceptance criteria for any test identified in Section 14 of the licensee's Final Safety Analysis Report,.as amended as being essential; (c):
Performance of any test at a power level different from that described in the program, and (d):
Failure to complete any tests included in the described program (planned or scheduled for power levels up to the authorized power level)."
The purpose of this letter is to confirm " prior NRC approval" to modify FSAR Table 14.2-123.
Enclosed please find a revised copy of FSAR Table 14.2-123 as discussed with, and agreed to by Messrs. Bournia and Long of your staff during a telecon with Messrs. Bishop and Renwick of LaSalle County Station on July 6, 1983.
As was discussed at that time, this revision deletes reference to measurement of relief valve capacity.
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H.-R.[Denton July 21, 1983 The measurement of relief valve capacity during startup' testing is no longer recommended by General Electric, who is now revising the LaSalle Startup Test Instruction to reflect this.. Experience at other recent:startups, and at LaSalle, has shown_that the measurement gives
- inconclusive-results~.
General Electric now suggests reliance on the ASME method for determining 1 relief valve. flow in accordance with " Pressure
. Relieving' Device Certification" as specified by the-National Board.of Boiler and Pressure Vessel Inspectors.
This documentation for.LaSalle is on. file with_ General Electric
' San Jose and certifies the relief capacity'to be consistent with FSAR Table 5.2-12.
The attached FSAR Table 14.2-123 revision-has received review and approval by On-Site:and Off-Site review.
This change and changes to any other affected sections of-the FSAR are expected to-be included in FSAR Amendment 64.
This change is provided-by letter at this time, per the requestof Messrsi Bournia and Long.
To the best of my knowledge and belief the statements contained herein1and in the attachment are true and correct.
In some respects
.these statements are not based on my personal knowledge but upon infor-
-mation furnished by other Commonwealth Edison and contractor employees.
Such information has.been reviewed in accordance with Company practice and-I believe it to be reliable.
Enclosed for your use are one (1) signed original and forty (40) copies of this letter and enclosure.
l If there-are any further questions in this matter, please
. contact this office.
Very truly yours, 4
1)11)V) i C. W. Schroeder
]
Nuclear Licensing Administrator 1m 4-cc:_ NRC Resident Inspector - LSCS 6999N i
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LSCS-FSAR AMENDMENT 57 JULY 1981 7
TABLE 14.2-123 RELIEF VALVES STARTUP TEST PURPOSE The purposes of this test are (1) to verify the proper operation of the primar'y system relief valves, (2)'to verify that the discharge piping is not blocked, (3) to verify proper seating of each relief valve following operation, (4) to obtain signature-information of relief valve operation for subsequent comparisons, (5' to scrify cach clief velve's cap; city, and_ f6t+to confirm j.;
gh proper overall functioning of the low-low set pressure relief logic.
DESCRIPTION The main steam relief valves will each be opened using the
" manual" control mode so that at any time only one valve is open.
During heatup at 250 psig, each valve will be opened and closed to demonstrate proper functioning.
Flow verifi-cation of each relief valve will be determined at rated pressure by observing bypass o.r control va.1ve motion and by observing a change-in-discharge' thermocouple readings.
Proper reseating j%g:
of each relief valve will be verified by observation of temper-atures in the relief valve discharge piping.
At selected test conditions each valve will be manually actuate.d and appro-priate system parameters recorded during the transient.
Data analysis includes a comparison of relief valve response times to design values for each relief valve.. To verify relief l
valve accumulator capacity, the MSRV with the highest nominal safety spring setting will be actuated while the MSRV accumu-lator air supply is isolated.
Thc capacity cf cach cli f cal;; will b; dct;;;incd at ;ct:d p c;cu;c by thc cacunt of
)h bypcc; c: cc..t;;l ecl/; cicourc : quir:d to maintain rccctor p cccure.
The response time on at least one of the valves will be recorded.
The backpressure of the discharge lines having the l
highest calculated dp and the lowest calculated dp will be measured.
During pressurization transients such as MSIV full l
closur.es and turbine trips / generator load rejection the opera-
' tion of the safety grade low-low pressure relief logic. system i
will be monitored.
A comparison between the reactor pressure behavior and SRV actuations will be made to confirm open/close j
setpoints and containment load mitigation'through the preven-tion of subsequent simultaneous SRV actuations.
l i' G 14.2-258
a LSCS-FSAR RMENDMENT 57 JULY 1981 TABLE 14.2-123 (Cont'd)
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ACCEPTANCE CRITERIA Level 1 There should be positive indication of steam discharge during the manual actuation of each valve.
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Thc num totcl of the pcrccntc;c correctcd flcu ratc muot-bc g catcr than 111.51 cf thc nucicar bcilcr warranted Otc =
)h ficw, ct 1031 cf th: :pring ctpcint prcosurc of 1155 poig.
The low-low set pressure relief logic shall fun'ction to preclude subsequent simultaneous SRV actuations following the initial SRV actuation due to the original pressurization transient.
Level 2 No observable leakage shall exist following reclosure.
The pressure regulator must satisfactorily control the reactor transient and close the control and/or bypass valves.by an gg crount equiv'. lent to the relicf calvc stccm fic,.
The transient IM recorded signatures for each valve must be analyzed for a.
M relativersystem response comparison.
The delay time (be't E F: " "'" "
trip and motion) shal1~be 1 0.1 seconds and the response time (main disc stroke time) shall be 1 0.15 seconds.
Mc individual relicf ;;17: may hav; a ficw rat
'cc :ceted to the actpcint prcacurc; that, considcrscg =c Curencat uncer-9, taintica, is ic:a than 001 cr grc;tcr than 122.33 cf its cxpccted ri m_
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.em um prc aurc cf 11'5 paig.
l Ne more than 251 of the inct:ll;d rclicf calvc may h:v: an-individual cc rccted fica ratc that is bctwcen 001 and 1909 g
y l
cf their cxpcctcd ficw ratcc.
l Thc total ficw : pacity of thc : fcty/rclicf-valvc used in-the 7.uten tic Ocprc cugination Cyatcm must bc equal to or I
grectcr than 1.8 x 105 lb/b r at 1125 poig when the calv:
jp having thc highcat mccoured capacity in 2:aumed to bc-cut cf acr :icci The selected MSRV with the highest nominal safety spring setting must indicate fully open when manually actuated with air supply to its accumulator isolated and vented.
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Discharge line back pressore shall be comparable with informa-
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tion presented on the Nuclear Boiler Process Flow Diagrams.
p1ea su re ) }
1 rot k eack rc /ie,l ya h e feet llaW a n el )Jhaa/
1:. 2 =s -
as hypa s, ya/yepo,;f,'l
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be l e s s tA o st /0 7; of yo/ve posi; tion he g the over-ase tef Il va /vt n sition re w m aez
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LSCS-PSAR AMENDMENT 57 JULY 1981 TABLE-14.2-123 (Cont'd) b When.'he. low-low pressure relief logic functions, the open/close t
actions of ~ the SRV's shall occur within -+ ~13 psi and + 20. psi
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aaf their design points, respectively.
INITIAL CONDITIONS s
1.
All construction tests and preoperational tests are com-pleted and approved.-
2.
Instrumentation has been installed and calibrated.
3.
All systems are operational to'the extent required to
. conduct the test.
i s
O 14.2-259a
,