ML18151A111

From kanterella
Jump to navigation Jump to search
LER 87-011-01:on 870516,low Flow RCS (EIIS-AB) Flow Reduced to 47%,following Reactor Trip.Caused by Failure of RCS a Hot Leg Loop Stop Valve.Samples of Loop Stop Valve Stems Will Undergo Ultrasonic testing.W/871012 Ltr
ML18151A111
Person / Time
Site: Surry Dominion icon.png
Issue date: 10/12/1987
From: Benson D
VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
87-013A, 87-13A, LER-87-011, LER-87-11, NUDOCS 8710200010
Download: ML18151A111 (4)
v • d • e


Text

-i UPDATED. REPORT - PREVIOUS ~ T DATED 6/12/87 NRC Fo,'m 3M 19,831 e POW 28-06-01

  • U.S. NUCLEAR REGULATORY COMMISSION APPROVED 0MB NO. 3150,0lOC EXPIRES: 8/31/88 LICENSEE EVENT REPORT (LERI FACILITY NAME 11 l IOOCKET NUMBER 121 I PAGE 131 SURRY POWER STATION, UNIT 1 TITLE 141 o 1s101010121810 1 loF O l3 REACTOR TRIP ON LOW RCS FLOW DUE TO FAILURE OF LOOP STOP VALVE EVENT DATE 151 LEA NUMBER 161 REPORT DATE l7r OTHER FACILITIES INVOLVED 181 MONTH DAY YEAR YEAR k< SEQUENTIAL NUMBER REVISION NUMBER MONTH DAY YEAR FACILITY NAMES DOCKET NUMBERISI 01s1010,0 1 I I ols 116 8 7 817 - 0 jl 11 - oI1 110 112 8 I1 o,s1010101 I I THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIRE~ENTS OF 10 CFR §: /Chock on* or mo,. of rh* follow;ng/ 111 I

_ OPERATING MODE 1111 I 11010 -- 20.402lbl Z0.40!ilcl 

                                                                                                                                            --X 50.7311Jl2Jliy)
                                                                                                                                                                                  ---     73.711b)

POWER. 20.411111111 llll 50.381cll1l 50.73111121M 73.711cl LEVEL 1101 20.4111(111111111 50.311cll21 50.731oll2l1Yiil 

                                                                                                                                                                                   -    . OTHER /Spec;fy ;n Absrracr b1tow and in Taxr. NRC Form 118Jtliili----                                                          -

20.411111111 lliiil !50.73ioll2llil 60.73loll2llviiil1Al. 366AI 20.411111111 lllvl ---- 50.731111:Zlliil 

                                                                                                                                            ,__ 50.731111211YiiillBI 20.4111111111M                                  so.131,H211;m                                               50.73111l2llxl LICENSEE_ CONTACT FOR THIS LER (121 NAME                                                                                                                                                                                TELEPHONE NUMBER AREA CODE D. L. Benson, Station Manager 81   o, 4     3 15    I  71-13111814 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (131 MANUFAC-       REPORTABLE                                                                                                MANUFAC-        REPORTABLE                   :::::::::::::::::::::       -:-:-

CAUSE SYSTEM COMPONENT TURER TO II/PROS CAUSE SYSTEM COMPONENT TURER TO II/PROS FLU /:{:\)) <? y i:::::::: *.- :*, -:-:--:-:- .....,. *:*,:-:- 1 :*::::::::

i
  • /i:1**::*:*r::::::: 1!::**,,,,-,.,.

X AIB II S IV I Ai3 I 911 1:-::,:,:_::: :::::;-;:: :-:-:-:-:-:- -:-:--:-: I I I I I I I >)'? l\]j )){) :-:-:-:-:-:-.-:-:-: f]  :-_:-: 

                                                                                                                                                                                                                \i,::,>t/):

I I I I I I I  :-;:::;::: ::::,:;.::;:::::: I I I I I I I  !> .:-*- ... SUPPLEMENTAL REPORT EXPECTED 1141 MONTH DAY YEAR EXPECTED n YES (!I yn, compl*r, EXPE~TED SUBMISSION DATE/ ABSTRACT (Limit to '400 11>>c11, i.,.. *pproxim*rtly fifr,,n 1ing1,-,,,.ce ryp,wrir~n li~eiJ (111 

                                                                                               ~                   NO SUBMISSIOI\/

DATE 1151 I I I On May 16, 1987 at 0824 hours. with Unit 1 at 100% power. a low flow reactor trip occurred when II A" loop reactor coolant system (RCS) (EIIS-AB) flow decreased to 47%. Following the reactor trip, the source range channels (EI_IS-DET) did riot automatically reinstate. All other protection and* control systems functioned properly. Operators followed appropriate 

                    .plant procedures and stabilized the plant following the reactor trip.

This event occurred when the II A" hot leg loop stop valve (EIIS-ISV) stem failed, permitting the disc to drop, partially blocking loop flow. A metallurgical analysis was performed which determined that the stem fail-ure was attributed to-stress corrosion cracking. The stress corrosion cracking was due to a combination of excessive backs eating force and a 

                    *process of thermal aging. At the next outages of sufficient duration, samples from the Unit 1. and Unit 2 stems will underg'? metallurgical analy-sis, and the Unit 2 stems will undergo ultrason_ic testing as was previous-ly performed on Unit 1. The failure of source range channels to reinstate was due to the under compensation of the intermediate range channel NI-36.

The source range channels were manually reinstated, and technicians read-justed the intermediate range compensating voltage, 8710200010 871012 ,;;a PDR ADOCK 05000280 j,f 1/, s PDR NAC Form 386 

!9 eJ,

t- ) 

  ~'IC Form ~A                                                                                                    U.S. NUCLEAR REQULATOIIY COMMIIIION 19-831 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION                                      APPROVEO 0MB NO 3150-010*

EXPIRES. 8/31/BB FACILITY NAME 111 DOCKET NUMBER 121 LER NUMBER 1111 'AQE 131 SECUE'NTIAL . "EVISION 

                                                                                                         ~UMl!IE,rll         NUM9ER SURRY POWER STATION, UNIT 1 o Is I o I o I o 12 I s Io sI7 -    o I 1I 1               0 11      0 12 OF      Oj 3 TEXT / I f " ' ° " ' ~ ; , ~   u* - - NRC Fann .111SA'1/ 117l

_ 1.0 Description of the Event On May 16, 1987 at 0824 hours, *with Unit 1 at 100% power, a low flow reactor coolant system (EIIS-AB) (RCS) flow decreased to 47%. Follow-ing the reactor trip, the source range channels (EIIS-DET) did not automatically reinstate. All other protection and control systems functioned properly. Operators followe*d* appropriate plant procedures and stabilized the plant follow-ing the reactor trip. 2.0 Safety Consequences and Implications The low flow reactor trip automatically trips the reactor to maintain sufficient margin above a DNBR of*l.3 with a.loss of RCS flow. The com-plete loss of flow in one loop from a reactor power of 100% (2441 MWt) with three loops operating is an analyzed event. During the event, "A" loop flow was maintained at approximately 47% and total core flow remained at 84%.

  • A confirmatory analysis performed by Nuclear Engineering con-cluded that DNBR was maintained above the accident analysis value of 1.3.

In addition, all other safety related systems remained operable during the event, and plant parameters remained well wit,hin the bounds of the acci- 

                             .. dent analysis. Therefore, this event did not. constitute an unreviewed safety question, and the health and safety of the public were not affect-
                               *ed.

3.0 Cause The cause of this event was a failure of the 11 A" RCS hot leg loop stop valve (MOV-1590) (EIIS-ISV). The valve stem failed, permitting the disc to drqp, partially blocking the loop flow. The valve stem failure was attributed to stress corrosion cracking. The stem material was analyzed by a contractor laboratory and determined that it was of a metallurgical composi_tion which was conducive to the process of thermal aging when exposed to elevated temperatures (above 550 degrees Fahrenheit). The material was 17-4 PH stainless steel 

                              *heat treated to H-1100. The r.esulting embrittlement due to thermal aging, coupled with high stress due to excessive backseating force, re-duced the resistance to stress corrosion cracking.

The failure of the source range channels to reinstate was due to the under.compensation of the intermediate range channel NI-36.

UPDATED REPORT - PREVIOUS REPORT DATED 6/12/87 POW-28-06-01 

 "!RC For11*,'d611A (9-831 LICENSEE EVENT R.EPORT (LER) TEXT CONTINUATION U.S. NUCL\:AR REGULATORY COMMl~ION APPROVED 0MB NO. 3150-0104 EXPIRES: *e/31/88 FACILITY NAME (1 I                                                    DOCKET NUMBER (21                   LER NUMBER (61                    PAGE (31 YEAR SURRY POWER STATION, UNIT 1 o Is I o I o I o I 2 I8 I o 8 I 7 -     oI 1 I1 -         ol 1 o I 3    OF    o I3 TEXT /If mar.   -ca la rw,uiml, u* IHlditioMI NRC Form 31515A 'al (171 4.0        Immediate Corrective* Actions The Operators performed all appropriate emergency procedures and function restoration procedures to ensure the plant was returned.to a stable con-dition. This included manually reinstating the source range channels.

Also, the STA performed the critical safety function status tree review to ensure specific plant parameters were noted and that those parameters remained within_ safe bounds. 5.0 *. Additional Corrective Actions The unit was placed in the cold shutdown condition, and the stem of the hot leg loop valve was replace.d. The stems of the other f-ive Unit 1 loop stop valves were ultrasonically tested and found to be satisfactory. Technicians readjusted the intermediate range compensating voltage, 6.0 Actions Taken to Prevent Recurrence At the next outages of sufficient duration; samples of the loop stop

  • valve stems from Unit 1 and Unit 2 will undergo metallurgical analysis, Additionally, the Unit 2 stems wil_l undergo ultrasonic testing as was previously perforµied on Unit 1. * - * -

During* the Unit 1 start-up following the hot leg stem replacement, the loop stop valves were not placed on their backseats, Due to leakage through the packing, the valves were later backseated to a maximum de-flection of.1/16 inch according to the manufacturer's recommendations. 7 *0 .Similar Even ts A similar failure occurred on the Unit 1 "B" loop Hot.Leg Isolation Valve on December 1, 1973. , 8.0 Manufacturer/Model Number Anchor Darling/Drawing Nos. 95-11778 and 95-11779. 

 *,rR~ f.'OAM 3&6A

. -~ 

                            -           USNRC-DS l'l81 OCT Iq A lfr. 0 J   VIRGINIA ELECTRIC AND POWER COMPANY Surry Power Station P. 0. Box 316 Surry, Virginia 23883 October 12,. 1987 U.S. Nuclear Regulatory Commission                     Serial No.:          87-013A Document Control Desk                                 Docket No.:           50-280 016 Phillips Building                                  Licensee No.:        DPR-32 Washington, D.C. 20555 Gentlemen:

Pursuant to Surry Power Station Technical Specifications, Virginia Electric and Power Company hereby submits the following updated Licensee Event Report for Surry Unit 1. 

                                          . REPORT NUMBER 87-011-01 This report has been reviewed by the Station Nuclear Safety and Operating Committee and will be reviewed by Safety Evaluation and Control.

Very truly yours, 

    . .li -    f{~

hDa:.;;z. Benson 

    . Station Manager Enclosure cc:  Dr. J. Nelson Grace Regional Adainistrator Suite 2900 101 Marietta Street,. NW Atlanta, Georgia 30323}}
Retrieved from "https://kanterella.com/w/index.php?title=ML18151A111&oldid=2504678"