ML18078A237
| ML18078A237 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 09/26/1978 |
| From: | Librizzi F Public Service Enterprise Group |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML18078A239 | List: |
| References | |
| NUDOCS 7810110091 | |
| Download: ML18078A237 (11) | |
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0 PSllJ Public Service Electric and Gas Company 80 Park Place Newark, N.J. 07101 Phone 201/430-7000.
September 26, 1978 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.c.
20555 Attention:
Mr. A. ~chwencer, Chief Operating Reactors Branch l Division of Operating Reactors Gentlemen:
ECCS PERFORMANCE ANALYSIS NO. 1 UNIT SALEM NUCLEAR GENERATING STATION DOCKET NO. 50-272 In response to your letter of Septembe~ 13, 1978, we wish to inform you of our schedule for submittal of the ECCS performance analysis required by 10CFR50.46 and Appendix K to correct the errors identified as related to Zirc -
water reaction heat generation.
This analysis is being performed by Westinghouse Electric Corporation for both the No. 1 and No. 2 Units and is expected to be available for your review in November, 1978.
Should you have any further questions in this regard, please con-tact Mr. E. A. Liden, Project Licensing Manager, on 201/430-8048.
EAL/jr 6Pl 39 The Energy People Very truly yours, t;~~~ff General Manager -
Electric Production 95-0942
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, 0. C. 20555
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Docket No. 50-272 Public Service Electric and Gas Company September 13 firl,~}~ Manager Licensi~~ ~n~d ~~;!~1onment
- ttouo * -~A SEP 2 ~.fo, ATTN:
Mr. F. P. Librizzi General Manager - Electric Production Rtff.R iO __
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JCR IWD 80 Park Place, Room 72£1 Newark, New Jersey 07101 CCWIES --- FILE -----
Gentlemen:
Several months ago, you were issued an Order for Modification of License to the license for the Salem Generating Station, Unit No. 1, related to the requirements of 10 CFR 50.46(a)(l) that require ECCS perfonnance to be calculated in accordance with an acce;:itab le cal cul ational model which fu-lly conforms to the provisions of Appendix K, 10 CFR 50.
The previously approved model for your facility was found to contain errors and thus does not fully conform to the provisions of Appendix K.
As a coridition of that Order, as soon as possible, you are required to submit a revised analysis of ECCS cooling performance calculated in accordance with a revised Westinghouse evaluation model, corrected for the identified errors and approved by the NRC.
By letter to Westinghouse dated August 29, 1978, the NRG has approved a revised Westinghouse ECCS evaluation model which corrects the identified errors (associated with the zirconium-water.reaction heat generation).
Accordingly, you should proceed to submit the revised analysis for your facility as soon as possible. Within 20 days fran the date of this letter provide us with your schedule for submittal of this revised analysis.
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See next page
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Public Service Electric & Gas Company cc:
Mark J. Wetterhahn, Esquire Conner, Moore & Corber Suite 1050 1747 Pennsylvania Avenue, NW Washington, D.C.
20006 R_ichard Fryling, Jr. Esquire Assistant General Solicitor Public *service Electric & Gas Company*
80 Park Place Newark, New Jersey 07101 Gene Fisher Bureau Chief Bureau of Radiation Protection 380 Scotch Road Trenton, New Jersey 08628 Public Service Electric & Gas Company ATTN:
Herbert J. Heller Manager, Salem Nuclear Generating Station
~ancocks Bridge, New Jersey 08038 Public Service Electric & Gas Company ATTN:
Mr. R. L. Mittl General Manager - Licensing and Environment 80 Park Place Newark, New Jersey 07101 Salem Free Library 112 West Broadway Salem, New Jersey 08079
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Frederick W. Schneider Vice President Public Service Electric and Gas Company 80 Park Place Newark, N.J. 07101 201/622-7000 Production Mr. James P. O'Reilly Director of USNRC February 24, 1977
-latory Docket File 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;;t{Qni t No. 1 Technical Specification, Section 6.9.2, we are submitting ECCS Actuation Repdrt No. ECCS/77-01.
This report is required within 90 days of the occurrence.
Sincerely yours,
REPORT NUMBER:
REPORT DATE:
OCCURRENCE DATES:
FACILITY:
EVENT.
ECCS/77-01 February 15, 1977 See Attachment 1 Salem Generating Station Public Service Electric & Gas Company Hancock's Bridge, New Jersey 08038 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 six (6) such actuations at the Salem Unit No. 1 This report justifies the acceptability of the number of Safety Injection (S.I.) events to date relative to the stress imposed upon the 1-1/2 inch S.I. injection nozzles. of this report describes the circumstances surrounding each of the six (6)
ECCS actuations.
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 Accuation Report No. ECCS/77-01 Attachment 1 b)
Westinghouse Burl 3461 Letter, dated 12/13/76, Attachment 2 Prepared by T. L. Spencer
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SORC Meeting No.
17-77
ATTACHMENT NO. 1 TO ECCS ACTUATION REPORT NO. ECCS/77-01 SAFETY INJECTION NO. 1 At 2239 hours0.0259 days <br />0.622 hours <br />0.0037 weeks <br />8.519395e-4 months <br /> on 11-30-76, a Reactor Trip/Safety Injection signal was initiated due to high steam line flow in coincidence with low Tave.
The unit was in Mode 3 with 540°F Tave and 2235 psig RCS pressure.
Instrumentation personnel were working on No. 12 S/G Steam Flow Channel 1 which was in the test position.
With RCS temperature being less than 543°F, all four channels of low Tave were tripped.
A premature lifting of a No. 14 S/G safety valve resulted in a No. 14 S/G high steam flow signal.
With No. 12 S/G steam flow in test and a high steam flow signal on No. 14 S/G, the required two of four signals were available.
This in coincidence with an actual low Tave caused the Reactor Trip/
Safety Injection.
The duration of the actual water injection was less than two minutes.
SAFETY INJECTION NO. 2 At 2301 hours0.0266 days <br />0.639 hours <br />0.0038 weeks <br />8.755305e-4 months <br /> on 11-30-76, while recovering from the first inad-vertent safety injection actuation, a second SI actuation occurred.
The conditions prior to, cause, and duration of the second ECCS actuation were identical to the first event.
SAFETY INJECTION NO. 3 At 0238 on 12-1-76, the third Reactor Trip/Safety Injection was initiated due to high steam line differential pressure P4.
Prior to the event, the unit was in Mode 3 with Tave at 547°F and 2235 psig in the RCS.
Based on a review of recorder traces, a No. 14 S/G safety valve again lifted prematurely causing the steam line differential pressure P4 safety injection.
The duration of this safety injection was less than two minutes.
SAFETY INJECTION NO. 4 At 0315 on 12-3-76, the fourth inadvertent safety injection occurred.
Prior to the event, the unit was in Mode 3 with 537°F Tave and 2235 psig RCS pressure.
The duration of the actual injection was approx-imately two minutes.
No indication of the cause of this event appeared on the status panel.
The computer did not print out a sequence of events but did print out "12 Steam Generator ~P Pressure Reactor Trip SI".
Prior to the SI actuation, instrumen-tation personnel were working on No. 12 S/G Steam Flow Channel II (r.T-523).
They had hooked up a brush recorder to the Channel II transmitter output, channel output, and steam pressure channel (PT-525).
When hooking up to No. 12 S/G Steam Pressure Channel III (PT-526), the SI took place.
After the SI, it was discovered that the technician mistakenly used grounded test leads when hooking up the brush recorder to the various steam flow and steam pressure channels.
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ATTACHMENT NO. 1 TO ECCS ACTUATION REPORT NO. ECCS/77-01 SAFETY INJECTION NO. 4 (Continued)
On 12-22-76 while in Mode 5, a special test was conducted in order to demonstrate and verify that the inadvertent Safety Injection that occurred was in fact caused by a technician mistakenly using grounded test leads when hooking up a brush recorder.
Inputs to the affected channels were simulated in order to re-produce plant conditions that existed at the time of the inadvertent Safety Injection.
The Solid State Protection System output relays were placed in test, such that no actual actuation of the Safeguards Emergency System would take place.
A series of five separate tests were performed using various combinations of grounded and ungrounded leads.
Test results and circuit analysis show that a partial trip signal will be generated when a grounded lead is plugged into the PT-525 channel and either of the test points in the FT-523 channel.
These channels have a common tie point in the density compensation circuit already and introducing a ground into each channel will produce a parallel circuit which reduces the effective input to the Steam Line Differential Pressure bistable.
Test results also indicate that a momentary transient (10-15 msec duration) of approximately 1.5 VDC (equivalent to 450 psi) resulted each time a grounded lead was plugged into the PT-526 channel.
This transient was of sufficient magnitude and duration to trip another Steam Line Differential Pressure bistable.
This produced a second partial trip signal for the duration of the transient which resulted in a Reactor Trip/Safety Injection signal.
These test results coincide with the events that took place during the Safety Injection that occurred on 12-3-76.
Based on these results, the inadvertent Safety Injection on 12-3-76 was caused by a technician mistakenly using grounded leads.
S,A.FETY "INJECTION NO. 5 At 2113 on 12-25-76, a reactor trip occurred due to No. 12 S/G low level and low flow.
At 2114, a Safety Injection occurred due to high steam flow coincident with low Tave.
Prior to the event, the unit was at 17.5% reactor power with the turbine on line producing 70 MWe.
Tave was 552°F with 2235 psig pressure in the RCS.
The duration of injection flow was approximately three minute::;.
The steam dump controller was in the steam pressure mode and the setpoint was set for 980 psig instead of the required 1005 psig due to operator error.
Steam dump opened on the reactor trip which gave a high steam flow signal.
The high steam flow signal coincident with an actual low Tave caused the SI.
r ATTACHMENT NO. 1 TO ECCS ACTUATION REPORT NO. ECCS/77-01 SAFETY INJECTION NO. 6 At 0634 on 12-30-76, a Reactor Trip/Safety Injection signal was initiated due to high steam line differential pressure P2.
Prior to this event, the unit was in Mode 3 with Tave at 415°F and RCS pressure at 1500 psig.
Maintenance personnel were working on No. 12 MSIV (12MS167) which was inoperable due to a hydraulic line failure.
The foreman in charge of the maintenance personnel failed to place blocking tags on the hydraulic motor.
When maintenance filled 12MS167 with hydraulic oil, the valve cycled open resulting in a high steam line differential pressure P2 reactor trip/safety injection.
The duration of the actual injection was approximately three minutes.
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Electric Engineering Department Public Service Electric & Gas Company 60 Park Place Newark, New Jersey 07101
Dear Mr. Rippe:
SALEM NUCLEAR GENERATING STATION UNITS NUMBER 1 AND 2 Safety Injection (SI} Transient Design Basis ni:c i 6 1976 Sponsor _______ l Roulo,a Coples UlJ5... - -----
Due*________ ----- --
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During the recent pre-critical testing phas~, 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 tr~nsient 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 including p1p1ng 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 \\~ere 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.
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Attachment No. 2 R. Thus in spite of the original Salem design basis using 31.1 p1p1ng codes which did not specifically require transient design calculations for the subject transient, we believe that our more recent analysis provides a sound basis for acceptability of the Salem piping.
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R. D. Rippe, 3L D. J. Jagt, lL C
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- Bar c 1 ay, 1 L J. J. Do 1 an, 1 L Very truly yours, WESTINGHOUSE ELECTRIC CORPORATION
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J. P. Sluss, Manager Salem Project
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USNRC-RE:G. 1-1 MAR 77 12J 2*~
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