ML18078A958
| ML18078A958 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 03/06/1979 |
| From: | Schneider F Public Service Enterprise Group |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7903120205 | |
| Download: ML18078A958 (5) | |
Text
1-Frederick W.. Schneider Vice President P.ublic Service.Electric.and Gas Company 80 Par.k Place Newark, N.J. 07101.201 /430-7373 Production March 6, 1979
- u. S. Nuclear Regulatory Commission Director, Division of Operating Reactors Office of Nuclear Reactor Regulation Washington, D. c.
20555 NRG IE BULLETIN NO. 79-01 NO. 1 AND 2 UNITS SALEM GENERATING STATION In response to your letter of February 8, 1979, transmitting NRG IE Bulletin No. 79-01 which was received on February 16, 1979, we have investigated this matter with respect to appli-cability to Salem No. 1 and 2 Units.
The following are the results of OU! investigation:
The subject bulletin upgrades the status of NRG IE Circular 78-08, "Environmental Qualification of Safety Related Electrical Equipment at Nuclear Power Plants."
We have already addressed some of the items mentioned as a result of IE Bulletins 77-05,77-05A, 77-06, 78-02, and 78-04.
During. the conduct of NRG Inspection No. 50-272/78-23, we indicated to Mr. L.J. Norrholm, Resident Reactor Inspector, that the review of environmental qualification of safety related equipment was in progress as part of the Unit 2 licensing effort and our evaluation would be presented in response to FSAR questions by the NRG staff.
Data on the Salem environmental qualification program was pre-sented in response to FSAR questions Q5.62, Q5.63, Q7.30, Q7.33, Q7.35 and Q7.41.
The information and additional supplemental data packages are under NRG staff review.
Although part of the Unit 2 license review, the information is also applicable for Unit 1.
As a result of this review, a number of components will be replaced due to a redefining of the environmental parameters for the containment because of a main steam line break and NRG positions on post accident instrumentation.
The replaced instru-ments were originally supplied with qualification data for less severe environmental conditions.
The following information, in addition to Table 1, describes the changes, follow-up action and justification for continued operation on Unit 1.
7903120205"
Director 3/6/79 Div. of Operating Reactors TRANSMITTERS Pressurizer Level, Steam Generator Level, Reactor Coolant System Pressure, Steam Flow and Pressurizer Pressure have been dete~mined to be required by the NRC staff to be qualified to higher steam line environmental parameters.
The present instruments installed on Unit 1 are qualified for LOCA environments (peak temperature of 271°F).
These instruments will be replaced with components having a higher qualifi~ation temperature.
SOLENOIDS Solenoids for in-containment isolation valves were found not to have qualification data for LOCA/MSLB environments.
These solenoids will be replaced with ASCO NP Series solenoid valves which have acceptable qualification data for the con-tainment accident environment.
LIMIT SWITCHES We have previously responded to this item in our response to NRC IE Bulletin 78-04 which indicated our use of NAMCO D2400X Limit Switches for in-containment isolation valve seal-in circuits and our intent to either replace the limit switches or modify the circuits prior to the end of the first refueling outage for Unit 1.
The replacement NAMCO limit switches are qualified for LOCA/MSLB environments.
Our response to FSAR question Q.7.35 elaborates on the design changes.
INSTRUMENT PANELS The containment safety related panels were tested to verify their adequacy for LOCA/MSLB environment.
The panels were tested with a small louvered pressure relief section in the door.
This modification assures that there is no structural damage to panels under LOCA/MSLB conditions.
The above items are the known changes due to environmental quali-fication review.
These changes will be made during the Unit 1 first refueling crutage.
All other safety related components are provided with adequate documentation which is presently under review by the NRC staff.
The staff review should be applicable
Director 3/6/79 Div. of Operating Reactors for both units since the design is the same.
Qualification documentation is available for audit at the PSE&G offices.
A detailed list of all documentation for environmental qualifi-cation as required by item 3 of the Bulletin will be prepared when the NRC staff has agreed on our data for Unit 2.
The response to FSAR question Q7.30 provides this information.
Attachment CC Mr. Boyce H. Grier NRC, King of Prussia, Pa.
Mr. L. J. Norrholm NRC, Hancocks Bridge, N.J.
Very truly yours,
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SALEM NUCLEAR GENERATING STATION NRC IE BULLETIN 79-01 RESPONSE TABLE 1 FUNCTION/OPERATION Pressurizer Level LOCA(ST)
MSLB(LT)
Steam Generator Level-MSLB(LT)
Reactor Coolant System Wide Range Pressure-MSLB(LT)
Steam Flow-MSLB(ST)
Pressurizer Preeeure-LOCA(ST)
Isolation Valves-LOCA(ST)
Solenoids PRESENT COMPONENT Barton 386/351 Barton 386 F&P 50EP1041 F&P 10B2496 F&P 50EP1041 ASCO Solenoids LB83146 LBX8316 HT834475 HT834477 8344B31 JX8342A3 REPLACEMENT COMPONENT Barton 764 Barton 764 Barton 763 Rosemount 1153A Rosemount 1153A ASCO Solenoids NP Serles 206-380-6F 206-381-3F 206-381-6F NP831654E NP8316A74E NP8344A74E NP8344A76E NP8344A78E NEW QUALIFICATION DOCUMENTATION W letter report NS-TM-1950 (proprietary)
Rosemount Report No. 3788 ASCO Report AQ521678/TR JUSTIFICATION FOR CONTINUED OPERATION The present transmitters have qualifica-tion documentation for containment LOCA conditions of 271°F.
The instrument panel tests/analyses performed by PSE&G show that transmitters enclosed in the panel experience a surface temperature of around 220°F-250°F when subjected to a MSLB simu-lated environment (peak temperature of 350°F).
The transmitters in question are enclosed in instrument panels which will maintain their temperatures below their qualification tem-peratures.
The transmitters are being re-*
placed with components having sequential test data to higher environmental conditions.
The present solenoids do not have adequate
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qualification documentation.
The new ASCO solenoid valves have undergone sequential ~!
testing of radiation, environment and seism11lf Ii applicable to the Salem plant.
These sole-1 noids and associated isolation valves are located inside the containment.
The outside containment isolation valves will operate thereby isolating the containment.
W has performed an analysis on solenoids indicating that their failure mechanism will cause the solenoids to fail such that the isolation valves will close.
The isolation valves re-quire control air to remain open.
Control air ls isolated to the containment following a LOCA which will cause the isolation valves to fail close.
Administrative procedures will require that the operator assure that the outside containment valves are closed and control air is isolated to the containment following a LOCA.
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SALEM NUCLEAR GENERATING STATION NRC IE BULLETIN 79-01 RESPONSE TABLE 1 (Cont'd)
PRESENT FUNCTION/OPERATION COMPONENT REPLACEMENT COMPONENT NEW QUALIFICATION DOCUMENTATION JUSTIFICATION FOR CONTINUED OPERATION
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Limit Swltches-LOCA(LT)
NAMCO D2400X on Isolation Valves Instrument Panels-LOCA(LT)/
PSE&G Design MSLB(LT)
LOCA Condltlo~s 271°F 43*.2 PSIG MSLB Conditions 350°F 42.8 PSIG (LT)
Long Term (ST)
Short Term NAMCO EA-180 PSE&G Design with small louvered door opening (approx.
80 sq. in *. or less)
NAMCO Report.
Rev. 1, 9/5/78 Wyle Labs Report (later)
Preliminary data in response to Ques 7.41 of FSAR This matter was addressed in response to IE Btilletin 78-04.
The present switches could fall after the protection system ls reset in such a fashion that the isolation valves will open.
Administrative procedures will require that the isolation valves both inside and outside are closed (by depressing the closed pushbuttons) and that the inside iso-lation valve control circuits are de-energ~zed thereby preventing valve operation prior to reset of the protection system.
The louvered section for panel door was to assure that no structural damage would occur during simulated accident conditions.
Panels without louvered sections could suffer some structural damage but there ls no evidence to suggest that equipment damage would occur */IA It is postulated that sufficient inleakage~
air exists in the panels presently to avoid structural damage.
The changes described above will be completed during the first refueling outage for Unit 1.
The refueling outage ls scheduled for March 31, 1979.
Salem Unit 1 may continue to operate based on the short time frame involved prior to refueling and the additional comments provided above which indicate that safety functions will be performed.
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