ML19225A221
| ML19225A221 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 07/12/1979 |
| From: | Gahm J PUBLIC SERVICE CO. OF COLORADO |
| To: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| Shared Package | |
| ML19225A215 | List: |
| References | |
| NUDOCS 7907180575 | |
| Download: ML19225A221 (6) | |
Text
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LICENSEE EVENT REi: ORT CONTACT. SLCCK l l l l l (ELEASE P AINT ALL "EQUIRED 1.*JFC AM ATICN 1 S LCENSEE NAME LCENSE Nr, vee A tcEssi Ev!NT iv es !vo!
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25 2S 20 31 02 CATECCav Tvp7 si,[.,AN OccxET NuvCEA E' VENT OATE REPCAT CATE l0,[1] CON"7 1 7 8 57 l l '" l l Ll l 0l 5l 01 -I Ol 2l 617I O 15 IO I6 l 7 l 7 l l0 l7 l1 l2 l 7 19 58 59 50 51 68 59 74 75 aC EVENT CESCP:PT:CN h l Durine olant 7 89 startuo at 15% cover, an incorrect feedwater flow trio existed in 1 of 2 l CL3 l PPS locics due to 3 .al fun c tion . This sinele failure would have crevented all four 7 89 SC l ara,- v,re A u- o "71"es fro- onenine autcraticallv. The locic odule was recaired.
7 89 l BC (C3] l tested, and reins talled. Failure nodes and ef fects analyzed. Relay interlocks revised l 7 89
@Ie} [ to eliminate undesireable sing' e f ailure inhibit problem. 0 RO 77-17A l 7 89 sysTtu cAusE
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{dI5) lsidered to be of a random nature. The single. failure inhibit was due to a desien 7 89 l SO d loroblem.
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REPORT DATE: Julv 12, 19 79 REPORTABLE OCCURRENCE 77-17A Page 1 of 5 0CCURRENCE DATE: May 6, 1977 FORT ST. VRAIN NUCLEAR GE'IERATING STATION PL3LIC SERVICE COMPANY OF COLORADO P. O. BOX 361 PLATTEVILLE, COLORADO 80651 REPORT NO. 50-267/ 77-17(14) A Final IDENTIFICATION OF OCCURRENCE:
On May 6, 1977, an incorrect feedwater flow trip existed in one of two Plant Protective System logics due to a malfunction. This single failure would have prevented all four steam / water du=p valves from opening if a condition existed which called for them to open. This situation has been identified as a reportable occurrence per Fort St. Vrain Technical Specification AC 7.5.2 (a) 5.
EVENT DESCRIPTION:
During plant startup with the reactor power at 15% on May 6,1977, it was determined that stea=/ water du=p valve HV-2218 could not be opened. HV-2218 was being exercised as part of a checkout after safety related maintenance had been performed on the valve. Investigation found a Plant Protective Sys-tem feedwater flow low relay (CR-174-13) had incorrectly tripped and inhibited HV-2218 from opening.
The low feedwater relay was tripped because of a failure in the CT-2B2 logic module. It was determined that this malfunction would have inhibited all four steam / water durp valves (HV-2215, HV-2216, HV-2217, and HV-2218) from opening automatically. However, the steam / water du=p valves could have been opened manually by the operator from the control room fellowing a loop shut-down, if required.
The faulty CT-232 logic module was repaired and the plant startup continued.
CAUSE DESCRIPTION:
The circuitry of HV-2213 (see Figure 1), which was typical of each of the steam / water du=p valves, contained among other interlocks, two contacts in series from the feedwater flow low relays (CR-174-1A and CR-174-1B) . Feed-water flow is monitored by the Plant Protective System on each secondary f coolant loop and when flow is below 20% on either loop, CR-174-1A is tripped (energized) by the A logic and CR-174-1B is tripped (energized) by the B logic thus preventing the steam water du=p valves from opening.
346 111 -
REPORTABLE OCCURFINCE 77-17A Page 2 of 5 CAUS E LESCR FTION:
Failure of three integrated circuit logic chips in logic module CT-2B2 caused CR-174-13 to energize opening the series contact on all du=p valves, even though the feedwater flow was greater than 20% on both loops. See Figure 1 for HV-2218 (typical) .
As a result , attention was drawn to the details of the interlocks in the s team / water du=p circuitry. The logic chip failures caused XCR-93174B output relay CR-174-1B to be energized. This relay interlock was for the purpose of inhibiting du= ping of either steam generator loop if feedwater flow was I
less thaa 20% in either loop and to facilitate safe shutdown cooling by main-l taining forced circulation cooling. The circuit is described in Final Safety 4 Analysis Report Section 7.1.2.5. Normally closed contacts from the output I relays of CR-174-1A and CR-174-13 were in the series circuit of each of the four steam water du=p valves. The interlocks were redundant in that a single interlock failure would not allow du=p of a steam generator loop if feedwater j flow was less than 20%. The problem was that a single failare (accidental l energi:ation of XCR-93174A or XCR-931743 control relay) could inhibit both the A logic and B logic deep capability of a loop thereby adversely affecting the steam water dump capability if requircd when both loops were operating.
l Interruption of continuity to the A logic dump capability (and the redundant i B logic dump capability) are monitored by continuous current monitors and ac-l tuate annunciators.
APFARENT CAUSE OF OCCURRENCE:
The cause of the logic module malfunction was a logic chip failure considered to be of a random nature.
The single f ailure which could have prevented all four steam / water dump valves f rom opening if a condition existed w tich required them to open was due to a design problem.
CORRICTIVE ACTICN:
The following actions have been taken:
- 1. The CR-2B2 logic module was repaired by replacing three logic chips, tes ted, and reins talled.
I
- 2. Failure modes and effect analysis was performed. For the s team / water dump circuit as it existed, the analysis did not identify any other unacceptabla system consequences than were known to exist. These were (1) the potential for a concurr-nt du=p of Loops 1 and 2 with the per=issible number of moisture tunitors out of service and tripped accompanied by an instrument bus f ailure, and (2) dump inhibit of Loops 1 and 2 due to single integrated circuit chip f ailures. For the proposed modified design, the analysis indicated both the unaccept-able system consequences stated above were eliminated and no others were introduced.
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REPORTABLE OCCURRENCE 77-17A Page 3 of 5 QORRECTIVE ACTION (continued):
- 2. (continued)
Based on the results of these analyses, it was concluded that the design modifications to the steam water dump system satisfy all Plant Protection System requirements and are consistent with the original desiga. basis of the plant.
- 3. With Nuclear Regulatory Co==1ssion concttrrence, the relay interlocks have been revised to eliminate an undesirable single f ailure inhibit p roblem.
4 The interlock circuits have been functionally tested successfully.
8 I
The interi= measure of recording the dump logic trouble alarm condi-tions on a once per shift basis when feedwater flow is greater than 20% is no longer required.
j No f urther corrective action is anticipated or required.
FAILURE DATA /SIMILAR REPORTED OCCURRENCES:
Reportable Occurrence Report No. 50-267/76-01 reported a condition affecting all steam water du p valves, but due to a different cau:.e.
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REPORTABLE OCCURRENCE 77-17A Page 5 of 5 Prepared by:///& ' _
1
( e f. . IC Mi(hael TecrtnicalJ.Serviceu Ferrib/) Engineer
/ /
Reviewed by: (1, If d [,
J/ b' . Gdhm Te inical Services Supervisor Reviewed by: A Frank M. Mathie Operations Manager Approved by: / 777, ode Don ^iarembourg [
Manager, Nuclear Production 346 115 .