ML20028E250
| ML20028E250 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 01/14/1983 |
| From: | Lentine F COMMONWEALTH EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| 5806N, NUDOCS 8301210136 | |
| Download: ML20028E250 (12) | |
Text
Commonwrith Edison C
One First National Plaza, Chicago, Illinois O
Address Reply to: Post Office Box 767 Chicago. Illinois 60690 January 14, 1983 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S.
Nuclear Regulatory Commission Washington, DC 20555
Subject:
Zion Station Units 1 and 2 Environmental Qualification of Electrical Equipment NRC Docket Nos. 50-295 and 50-304 Reference (a):
December 14, 1982, letter from S. A. Varga to L.
O. DelGeorge.
Dear Mr. Denton:
Reference (a) transmitted the NRC's Safety Evaluation Report (SER) for the Environmental Qualification of Safety-Related Electrical Equipment for Zion Station.
This is to the provide information that reaffirms our justifiestion fcr continued operation regarding the items id ent i fi ed in the SER.
The Attachment to this letter provides information regarding Fisher & Porter transmitters and Marathon terminal blocks.
We have identifled no items in categories 1B, 2A, and 2B for which justification for continued operation was not previously submitted.
Resolution of aging deficiencies will be addressed in our 90-day submittal.
To the best of my knowledge and belief, the statements contained in this letter and the Attachment are true and correct.
In some respects, these statements are not based on my personal knowledge but upon information furnished by contractors and other Commonwealth Edison personnel.
Such information has been reviewed in accordance with Company practice and I believe it to be reliable.
One (1) signed original and forty (40) copies of this transmittal are provided for your use.
Please address questions regarding this matter to this office.
Very truly yours, M.
F.
G. Lentine Nuclear Licensing Administrator im Attachment
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5806N 8301210136 830114 DR ADOCK 050002 fC(
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ATTACHMENT COMMONWEALTH EDISON COMPANY ZION UNITS 1 ar.d 2 JUSTIFICATION FOR CONTINUED OPERATION FOR ITEMS IDENTIFIED IN FCR's TEP 5806N
SARGENT & LUNDY ENClNEERS CHICAGO s
In the recently issued Franklin Institkite Technical Evaluation Report, the main item of concern was the short-term operability of the certain safety-related transmitters.
The following are FRC TER item numbers and equipment:
1.
FRC Item 36 LT-517 LT-527 LT-537 LT-547 LT-518 LT-528 LT-538 LT-548 LT-519 LT-529 LT-539 LT-549 These are the narrow range steam generator level transmit-ters which are located in containment.
2.
FRC Item 38 PT-455 PT-456 PT-457 PT-458 These are the pressuri2er pressure transmitters located in containment.
3.
FRC Item 39 FT-512 FT-532 FT-513 FT-533 FT-522 FT-542 FT-523 FT-543 These are the steam 1ine flow transmitters located in con-tainment.
4.
FRC Item 40 PT-514 PT-524 PT-534 PT-544 PT-515 PT-525 PT-535 PT-545 PT-516 PT-526 PT-536 PT-546 These are the steam line pressure transmitters located in the lower safety valve room.
The licensee has prepared an accident analysis on the following three accidents:
Loss of Coolant Accident (LOCA), Mainsteam Line Dreak (MSLB) and Fe Iwater Line Break (FLB).
The licensee has reviewed the line breaks analyzed and deternined which trans-mitter will cause the required safeguards actuation (safety injection and/or reactor trip).
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CARGENT & LUNDY ENGINEERS CHICA2O b,
A.
LOCA Section 14.3.1.1 of the Zion FSAR details the LOCA break cases.
It states that, should a break occur, depressur-ization of the Reactor Coolant System causes fluid to flow to it from the pressurizer, resulting in a press'ure decrease in the pressurizer.
Reactor trip and safety injection occur when the pressurizer low pressure set points are reached (1825 and 1815 psig respectively).
Safety injection actuation / reactor trip are also provided by a high containment pressure signal.
Table 14.3.4-3 of the Zion FSAR presents a LOCA event chronology for all cases analyzed.
For dhe event chron-ology, the safety injection signal is assumed to be gen-erated by pressurizer pressure.
Also, a 25 second delay in starting of the ECCS System is assumed in the analysis.
The design maximum delay time between SI signal and the time that the Safety Injection System is ready to deliver water is 22 seconds.
Note 4 of the same table identifies the delay time in receiving the safety injection signal f rom the containrent pressure as a maximum of 2.0 seconds for a three foot 2 equivalent break.
Using the 22 second maximum response time with the two second delay yields 24 seconds, which is still less than the 25 seconds as-sumed in the analysis for ECCS starting.
Thus, the two acceptable means of safety injection actuation are:
1.
Pressurizer Pressure PT-455, 456, 457 and 458 located inside containment.
These transmitters would provide the safety injection signal on low pressurizer pressure.
However, the harsh environment may render these transmitters in-operable.
2.
Containment Pressure PT-CS19, 20, 21 and 22 located outside containment in Zone A7 (a nonharsh environment throughout any acci-dent).
These transmitters would cause a safety in-jection on high containment pressure.
Figure 14.3.4-2 presents the results of the pressure transients for the four break sizes analyzed.
The containment pressure set point of 4.5 psig is reached in less than two seconds for all cases.
SARGENT QLUNDY ENGINEERS CHIC A 2O A.
LOCA (cont' d)
Although the pressurizer pressure transmitters are located in containment and are subjected to a harsh environment fol-lowing a LOCA, the secondary means of providing the safety injection and reactor trip (high containment pressure) is located in a nonharsh environment of the plant and would be available.
The responsc time delay between the pressur-izer pressure and containment high pressure signals is less than two seconds and results in ECCS flow being delivered within the time assumed in the analysis.
8 Justification for Continued Operation:
)
The analysis presented indicates that there will always be at least one means available for detecting a LOCA and in-itiating the safety injection.
This means is provided by the containment pressure transnitters located in a non-harsh environment of the plant.
Qualification of this equipment is by experience.
B, MSLB The Zion FSAR, Section 14.2, details tha action and safo-guards protection following a MSL.B bcth in and cutside containment.
A s afety injection signal is received from any of the following:
1.
Two cut of three low pressurizer pressure transmitters.
PT-455, 56, 57 located in Zone C1.
2.
Two out of three differential pressure signals between a steam line and remaining steam lines PT-514, 24, 34, 44, PT-515, 25, 35, 45, PT-516, 26, 36, 46 located in Zone T3.
High steam flow in two out of four main steam lines 3.
in coincidence with either low low reactor coolant system average temperature or low main steam line pressure.
FT-512, 22, 32, 42 Zone C1 FT-513, 23, 33, 43 Zone Cl PT-516, 26, 36, 46 Zone T3 TE-411A&B, 421A&B, Zone Cl TE-431A&B, 441A&B Zone Cl 4.
Two out of four high containment pressure transmitters PT-CS19, 20, 21, 22 located in Zone A7..
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SARGENT & LUNDY E N GIN E E R S CHICAGO 4
B.
MSLB (cont'd)
Steam Line Break In Containment:
The worst case evaluated is the complete severance of,a pipe inside containment at the outlet of a steam generator at no-load conditions and all reactor coolant pumps running..
For this case of a steam line break in containment, Figure 14.2.5-9 of the Zion FSAR shows the containment pressure.
as a function of time following the accident.
From the curve it is evident that the containment pressure exceeds the safety injection set point of 4.5 psig in less than one second.
Thus, the indication which produces the primary safety injection signal is containment pressure.
As dis-cussed previously in the LOCA analysis, the containment pressure transmitters are located outside containment in a nonharsh zone.
One of the available backup actuations, which also provides the necessary safety injection, is the differential pres-sure signal between one steam line and the remaining steam.
lines.
These transmitters are also located outside con-tainment and for a MSLB in containment, this plant zone is nonharsh.
Justification for Continued Operation:
For a MSLB in containment, the primary means of initiating a safety injection is the containment pressure transmitters located in a nonharsh environment of the plant.
These trans-mitters will be unaf fected by the accident and will provide their necessary function.
Steam Line Break Outside Containment:
The Zion FSAR analysis of a MSLB outside containment was conducted on a transient arising as a result of a break equivalent to a steam flow in two out of four main steam lines in coicidence with low low Reactor Coolant System average temperature.
These instruments are located in containment and are unaffected by the break.
Thus, the primary means of detecting a line break outside contain-ment as analyzed in the FSAR, will be unaffected by the accident and will provide its necessary function.
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SARGENT & LUNDY EN GIN E E RO CH3CAGO B.
MSLB (cont ' d)
Justification for Continued Operation:
For the MSLB outside containment, the primary means of-in-itiating the safety injection signal, as analyzed in the Zion FSAR are transmitters located in a nonharsh zone of f
the plant following the accident.
These transmitters being unaffected by the accident will be capable of providing their necessary safety function.
C.
FLB A loss of the normal feedwater results in a reduction in capability of the secondary system to remove the heat gen-erated in the reactor core.
Of primary importance follow-Ing this accident is a reactor trip which, if not initiated, could result in primary plant damage from increasing temp-erature and pressure.
Since there is no inventory loss from the reactor coolant system, or pouitive reactivity inserted, safety injection actuation is not required to mitigate this accident.
However, a safety injection would provide the necessary reactor trip.
The following provides the necessary protection against a loss of normal feedwater:
1.
Reactor trip on any of the following:
a.
Two out of four high pressurizer pressure PT-455, 56, 57, 58 located in Zone C1.
b.
Two out of four over temperature d T signals TE-411A&B, 421A&B, 431A&B, 441A&B located in Zone C1.
c.
Two out of three low low steam generator water level transmitters in any steam generator LT-517A, 27C, 37D, 47B LT-518A, 28C, 38D, 48B LT-519A, 29C, 39D, 49B located in Zone Cl d.
S' team flow /feedwater flow mismatch with coincident low steam generator level.
Feedwater flow transmitters are in the steam tunnel, steam flow transmitters and steam gen-erator level transmitters are located in con-tainment. J
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CARGENT & LUNDY EN GIN E E RS CHICA2O 4
C.
FLB (cont'd)
OR 2.
Safety injection signal from any of the following:
a.
Two out of three differential pressure signals between a steam line and remaining steam lines PT-514 PT-524 PT-534 PT-544 PT-515 PT-525 PT-535 PT-545 PT-516 PT-526 PT-536 PT-546 located in Zone T3 b.
Two out of four high containment pressure trans-mitters PT-CS19, 20, 21, 22 located in Zone A7.
In Containment:
Should a FLB occur in containment, the primary means of providing a reactor trip as identified in the FSAR is the low low steam generator level *.ransmitters.
This occurs at 27 acconds as determined by the analysis.
These trans-mitters are located in containment and would thus be sub-jected to the harsh environment.
However, the FSAR analysis did not take into account the potential for providing the necessary safeguards actuation cauced by high containment pressure transmitters, which wculd provide safety injection as well as the required teactor trip.
The licensee has made an assessment of the Zion contain-ment's short-term pressure response due to a break of the main feedwater line.
The results indicate that the time required for the containment pressure to reach the 4.5 psig set point is conservatively estimated to be less thah 20 seconds.
This response in containment pressure indicates that the required reactor trip would first be initiated by the high containment pressure transmitters which are located outside containment in a nonharsh zone of the plant.
Therefore, reactor trip is provided by transmitters which will be unaffected by the accident and will be capable of providing the necessary function.
Out of Containment:
In the event of a loss of feedwater accident out of contain-ment, the primary means of providing the reactor trip and subsequent auxiliary feedwater flow initiation is available and will be in a nonharsh environment (low low steam gen-erator level transmitters). --
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CARGENT O LUNDY ENGlNEERS CHICAGO C.
FLB (cont'd)
Justification for Continued Operation:
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The licensee has concluded that for each of the FLB l'oca-tions, transmitters are available in nonharsh environments of the plant to provide the necessary safeguards actuation.
There fore, the licensee concludes that justification for continued operation has been provided.
FRC Item 37 LT-501 LT-502 LT-503 LT-504 These are the wide range steam generator level transnitters
.which are being replaced with qualified transmitters.
These are located in containment.
FRC Item 42 FT-FWO3 This is the auxiliary feedwater flow transmitter located in the Auxiliary Building.
This transmitter is being replaced with a qualified transmitter.
Justification for Continued Operation:
These transmitters are required to operate following a High Energy Line Break (HELB) for maintenance of auxiliary feedwater flow to restore and maintain normal steam generator level.
For a HELB inside containment, the feedwater flow transmitters will be in a nonharsh environment and can be used as a guide for maintaining long-term steam generator level.
For a HELB outside containment, both the narrow range and wide range steam generator level transmitters will be available and will be in a nonharsh environment of the plant.
Also, if a HELB occurs in the Auxiliary Building, only FWO3 will be in a harsh environment.
The three other auxiliary feedwater transmitters remain in nonharsh areas of the plant.
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CARGENT & LUNDY ENGINEERO CHICAGD s
Terminal Blocks Items 55, 66, 68, 69, 70, 71, 72, 86 FRC Identified Concern of Licensee's Response:
Qualification Not Established.
Wyle Test Report-45611-1 documents testing performed on Marathon fixed Barrier terminal blocks, Series 6000 and 1600, both horizontally and vertically mounted in enclosures.
The 6000 series blocks were furnished from the Zion Station stock, the enclosures were standard terminal boxes, manufactured per CECO S td. EM47150, thus, duplicating those in use at the Zion Station.
Both insulated and uninsulated barrel ring-tongue, terminal lugs were installed on the terminals with Okonite cable forming two separate series circuits on alternate ter-minals utilizing all terminals on each block.
Curcuit resistance (continuity) was tested throughout the test program.
The total circuit resistance through a single terminal block was well below the 10 OHM criteria.
Insulation resistance between circuits ar.d between each circuit and ground was tested at 500 ydc for one minute wid1 a lx106 OHM minimum requirement.
The terminal block assemblies were irradiated for 448 hours0.00519 days <br />0.124 hours <br />7.407407e-4 weeks <br />1.70464e-4 months <br /> at a dose rate of 4.6x105 rads / hour for a total integrated dose of 2.06x108 rads.
The reported minimum dose is 2.0x108 rads.
1 The terminal block assemblies were aged for 20 and 40 years.
One terminal block assembly consisting of sheet steel enclo-sure and four terminal blocks, wt i aged for 466 hours0.00539 days <br />0.129 hours <br />7.705026e-4 weeks <br />1.77313e-4 months <br /> at 120*C for an equivalent life of 20 years.
The other identi-cal assembly was aged for 932 hours0.0108 days <br />0.259 hours <br />0.00154 weeks <br />3.54626e-4 months <br /> at 120 C for an equivalent life of 40 years.
The aging times and temperatures are based on a normal operating ambient of 122*F (50 *C) which exceeds the worst case normal ambient at the Zion Station.
Post radiation and thermal aging functional tests equalled or bettered the results of the baseline functional tests in reagrd to insulation resistance with the continuity test howing a slight increase in circuit resistance after thermal aging but still a factor of _000 below the allowable 10 OHM's.
l The seismic qualification tests consisted of biaxial resonance search and random multifrequency tests in each of two test orientations; vertical and horizontal.
The seismic tests demonstrated that the terr.inal block assemblies possessed sufficient integrity to withstand, without compromise of structures or electrical functions, the simulated seisnic environment.
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4 CARGENT & LUNDY ENGINEERS CHICASO 4
For the LOCA test the terminal block assemblies _were verti-cally mounted within the test chamber simulating field condi-tions.
A voltage potential of 175 Vac and a current of 15 amperes was applied to each terminal block assembly to monitor current leakage between the circuits and ground.
Arrheni.us
-methodology was used to compress the one year accident period to 34 hours3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br />.
From initial conditions c:f 135*F and atmospheric pressure the chamber temperature and pressure was increased to 270*F and 30 psig in 20 seconds which was maintained for approximately four minutes.
The chamber was allowed to cool to 140*F and a second transient was begun; raising the temperature and pressure to 380*F at 50 psig maintaining this for approx-imately 20 seconds.
The temperature was dropped to 345*F and maintained for the remainder of the test; the pressure was decreased to 25 psig 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> into the test and at 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> dropped to 22 psig for the remainder of the test.
During the first six hours of the accident exposure, a chemical spray of 0.04 g/ min /ft2 and a ph of 8.5 to 10 was maintained, for the next three hours the spray rate was increased to 0.5 g/nin/f t2, Approximately one hour and 50 minutes into the LOCA test a 6000 series terminal block (#2) arrangeri horizontally in essembly
- 1, which was preased for 20 years, shorted co grcund.
This was attributed to a random failure of the wiring rather than the termi.nal block itself.
All other terminal blocks saccess-fully met the requirements.
Post LOCA functional te s e.L cho we d.
a slight increase in resistance during the continuity test, although still only 0.1% of the maximum allowed..
The incula-tion resistance tests showed a slight decrease in resistance, but were above the minimum requirements.
Worthy of note is the fact that the terminal block that apparently shorted t.o ground during the LOCA test passed the post LOCA functional tests.
Based upon the results of this test and operating experience, we conclude that the terminal block assemblies in use at the Zion Station are fully qualified for their 40 year design life plus accident conditions.
The licensee maintained a copy of the proprietary test report in their files should the staff wish to review the particulars of the test.
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The NRC has asked that the licensee resolve the aging deficiencies identified in the Franklin Technical Evaluation Report (TER).
The licensee has been compiling aging information on all equip-ment identified in the TER.
This work has been completed; however, due to the magnitude of the aging concerns identified, the licensee felt that this issue was better responsed to on a case-by-case basis rather than in general terms.
Because of this, the licensee will provide a resolution to all aging concerns-when it resubmits the Zion Environmental Qualification to the NRC.
This is scheduled to be provided within 90 days of issuance of 10CFR50.49, " Environmental Qualification of Safety-Related Electrical Equipment for Nuclear Power Plants."
Finally, the licensee has reviewed the items in NRC Categories 1B,.2A and 2B and concluded that no additional items (other than those previously discussed), exist for which justification for continued operation was not previously submitted.
Although
'the terminal blocks did not contain justification for continued
, operation in the licensee's 90 day submittal, the qualification which the licensee has now provided shows full qualification of this item.
Justification for continued operation for the short-term safety-related transmitters was originally provided for in the licensee's 90 day submittal; however, the licensee has pro-vided an expanded discussion on their operation to. ensure that justification for continued operation can be ensured.
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