ML19031B175

05/10/1977 Letter ECCS Actuation Report No. 77-26/990

ML19031B175
Person / Time
Site:	Salem
Issue date:	05/10/1977
From:	Schneider F Public Service Electric & Gas Co
To:	O'Reilly J NRC/IE, NRC Region 1
References
LER 1977-026-01
Download: ML19031B175 (6)

Text

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Frederick W. Schneider Public Service Electric and Gas Company 80 Park Place Newark, N.J. 07101 201/622-7000 Vice President Production May 10, 1977

-R~Jat~

Mr. James P. O'Reilly Director of USNRC Off ice of Inspections and Enforcements Region 1 631 Park Avenue King of Prussia, Pennsylvania 19406

Dear Mr. O'Reilly:

LICENSE NO. DPR-70 DOCKET NO. 50-272 Pursuant to the requirements of Salem Generating Station Unit No. 1 Techn~cal Specifications, Section 6.9.2, we are submitting ECCS Actuation Report No. 77-26/990. This report is required within ninety (90) days of the occurrence.

Sincerely yours, 7713701 :::.~.

_J

Re'port Number: 77-26/990 Report Date: 4/27/77 Occurrence Dates: See Attachment 1 Facility: Salem Generating Station Public Service Electric & Gas Company Hancocks Bridge, New Jersey 08038 Event Appendix A Technical Specifications, Section 6.9.2 requires the reporting of Emergency Core Cooling System (ECCS) Actuations within 90 days of their occurrence. To date, we have experienced eight (8) such actuations .. The purpose of this report is to describe the circumstances surrounding Safety Injection Nos. 7 and 8. Details of Safety Injections Nos. 1 thru 6 are contained in ECCS Actuation Report No. ECCS/77~01, previously submitted.

Discussion/Conclusion The referenced.Westinghouse letter documents the acceptability of

. fifty (50 ). safety injection transients at a RWST temperature of 40 °F.

As the lowest RWST temperature in any of the subject transients was 61.5°F, none of .the subject transients approaches the severity of the design basis transients and, as such, are acceptable.

References a) ECCS Actuation Report No. 77-26/990, Attachment 1 b) Westinghouse Burl 3461 letter, dated 12/13/76, Attachment 2.

Prepared by T. L. Spencer

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Manager -

/

SORC Meeting No. 46-77

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ATTACHMENT NO. 1 TO ECCS ACTUATION REPORT NO. 77-26/990 SAFETY INJECTION NO. 7 At 1308 hours0.0151 days <br />0.363 hours <br />0.00216 weeks <br />4.97694e-4 months <br /> on 3/30/77, a Safety Injection/Reactor and Turbine Trip signal was initiated due to Steam Differential Pressure Low Pl. Prior to this event, the unit was in Mode 1 at 75% reactor power, 880 MWe generator load *. Operations Department surveillance procedures SP(O) 4.3.2.1.l(E) for Technical Specification surveillance 4.3.2.1.l was in progress. Step 5.2.34A states "Open and then close the following valves to clear the status panel indicating lights: 13MS7, 14MS7, 13MS18~ 14MS18". The bezel from which these valves are operated also contains the open/close control pushbuttons for 13MS167 and 14MS167 (13 and 14 Main Steam Isolation Valves). Attempting to perform step 5.2.34A, an. operator mistakenly.pushed the close pushbuttons for 13 and 14MS167. The closing of the 13 and 14 MSIV's caused the Safety Injection. The cause of this event was operator error.

SAFETY INJECTION NO. 8 At 1001 hours0.0116 days <br />0.278 hours <br />0.00166 weeks <br />3.808805e-4 months <br /> on 4/12/77, a loss of the lG 4kV Group Bus caused the loss of No. 14 RCP. Plant conditions were stabilized at 25% reactor power, Tave at 566°F, Rod Control System in manual, three (3) loop operation.

(ref. LER 77~25/0lT). At 1032 hours0.0119 days <br />0.287 hours <br />0.00171 weeks <br />3.92676e-4 months <br />, the lG Group Bus had been re-energized.and an attempt was made to restart No. 14 RCP. The resultant S/G water level transient caused a No. 14 S/G Hi-Hi level Turbine/Reactor Trip. Immediately following the trip, a Safety Injection occurred.

With Tave at 566°F, a turbine trip caused the Steam Dump - Turbine Trip Controller to.initiate a Hi Tave steam dump sequence. The steam dump sequence resulted in six (6) steam dump valves*opening fully and three (3) of the remaining six (6) to modulate to a 1/2 open position. The resultant. steam flow through the steam dump valves combined with a 4-5% flow. spike was sufficient to cause Nos. 11, 12 and 13 S/G Hi Steam Flow alarms to actuate. When steam dump was initiated, Tave decreased to less than 543°F causing a low Tave signal. The cause of the Safety Injection was Hi Steam.Flow.coincident with low Tave.

Attachment No. 2 R. D. Rippe Thus in spite of the original Salem design basis using 31.1 piping codes which did not specifically require transient design calculations for the subject transient, we believe that our more recent analysis provides ~

sound basis for acceptability of the Salem piping. ~

Very truly yours, WESTINGHOUSE ELECTRIC CORPORATION 9~

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• P* S1us s , *Man ager .

Salem Project U /hs

--- cc: R. D. Rippe, 3L D. J. Jagt, ll C. f. Barclay, ll J . J . Do 1 an , 1L

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CHIEF MECHANICAL ENGINEER ENGINEERING DEPT,

• Mr. R. D. Rippe Chief Mechanical Engineer --~ I

Noted------------------------

Electric Engineering Department Public Service Electric &Gas Company n~c 1 6 _1976 60 Park Place Newark, New Jersey 07101 Sponsor______ Rout~ Coples Lff..1:s..- ----

Due.__ _ _ _ ----- - - * -

Dear Mr. Rippe:

File _______ _

---~--t SALEM NUCLEAR GENERATING STATION UNITS NUMBER 1 AND 2 Safety Injection (SI} Transient Design Basis During the recent pre-critical testing phase, the plant was subjected to three (3) inadvertent Safety Injection (SI) initiation events, which we understand resulted in some water being injected into the Reactor Coolant Loop. We also understand that the NRC has verbally asked for the design transient basis for Salem for this type of event.

While we have not specifically analyzed Salem for this type of transient, we are confident that our ongoing plant analysis associated with ASME Section III more than demonstrates that the recent three (3) SI's will have no detrimental effect on Salem. Our conclusion is based on the fol-lowing rationale.

We have analyzed sufficient Section III piping systems includin~ piping similar to yours with the 1-1/2 11 SI nozzles to show that fifty (50.) such SI events can be accomodated without exceeding the appropriate stress.

limits at the SI nozzle.

• These analysis were based on the nozzles being subjected to a 40°F water transient which is probably far worse than the actual transient.seen at Salem.

The results of these analysis are in the process of being formalized for submittals to the NRC for Section III plants.

RECE!Vf.O 00CUM£i{T PROCESSING /Jin[

1Y/f MAY 16 AM fl 52