ML22238A085
| ML22238A085 | |
| Person / Time | |
|---|---|
| Issue date: | 08/24/2022 |
| From: | NRC/OCIO |
| To: | |
| Shared Package | |
| ML22238A081 | List: |
| References | |
| FOIA, NRC-2022-000208 | |
| Download: ML22238A085 (22) | |
Text
Li en see:
Dat e of Inspection:
U. S. ATOMIC ENERGY COMMISS ION REGION II DIV JS ION OF COMPLIANCE Report of Inspection CO Report No. 50-231/70-4 General Electric Company and SAEA License No. DR-15 Category B October 6-9, 1970 Date of Previous Inspection:
August 10- 12, 1970 _
Reviewed By:
Proprietary Information:
None SOOPE
//-I 7-70 Date A. routine, announced inspection was made at the fast spectrum, sodium-
On October 6-7, 1970, V. D. Thomas, Compliance
_ Headquarters, reviewed the licensee' s corrective act ions following a
- ttpartial" scram which occurred on September 12, 1970.
The nfeede~*
report covering Thomas' portion of the inspection is attached as Exhibit B.
SUMM~RY Safety Items - None
!oncompliance Items - Automatic flow control valves were installed in the discharge lines fro~ the redundant argon vent vacuum pumps.
These pumps a::*e used to refill the auxiliary primary system if there is a 1 ine break
No, 50-231/70-l; CO Rpt,
- n the main primary coolant loop.
Contrary to 1.0 CfR 50.59, the liceasee
~id *not have a written safety evaluation to determine that this change did not involve an unreviewed safety question.
A Form AEC-592 will be issued for this item of noncompliance.. (See Section M.)
Unusual Occurrences - The "part ial11 scram reported in an Inquiry l-~emorandum dated September 18, 1970, was reviewed during this inspection.
The licensee reported the malfunction to DRL in a letter dated September 21, 1970, as required by Section 3,C.(1) of the operating license.
(Sec Section F,3,)
Overinflation of an access suit caused a mechanic to loose his bal.ance and fall across the reactor vessel head during a refueling cell entry.
Use of the man access su:i.t.s has been suspended by the facility manager as recommended by the Site Safety Committee until corrective action.has been completed to prevent similar occurrences.
(See Section U.)
Status of Previously Reported Problems - The licensee's response to the Form AEC-592, issued for two items of noncompliance reported in CO Report No. 50-231/70-~, is considered unsatisfactory by Region II.
Region 11's comments on the response were forwarded to Compliance Headquarters on October 6, 1970.
The noncompliance items,,,ere concerned with hiri r:6 a maintenance supervisor and instrument engineer who did not meet the experience requirements for these positions as specified in the Technical Specifications.
Modifications have been completed to prevent failure of the containment ventilation isolation valves to open on demand.l/
Valve operation was transferred from the instrument air system to the instrument nitrogen systan to prevent freezing of moisture in outside lines during cold weather.
(See Section K.2,)
Fabrication of the new argon purification system has been completed and has been shipped from California.
The new bubbler system will be installed during the next maintenance outage of the facility.
Other Significant Items -
- 1. Wide-Range Monitor (WRM) - Heat Balance Correlation Data obtained during S Mw static tests revealed that agreement between WRM's and reactor heat balances was dependent upon core average temperature.
At any given thennal power level, WRM indications are
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proportional to core average temperature. Diagnostic tests recommended,~
by the G-E. engineering group in Sunnyvale are being performed to evaluate the problem.
(See Section F.l.)
l/Report ed in Inquiry Memorandum dated January 9, 1970, and in CO Report No. 50-231/70-1.
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llllii1 CO r.pt, No. 50-221/70-.'.. 2, cover Gas Monitor Sensitivity Data was reviewed which demonstrates that the cover gas gamma monitor will detect a fission gas release equal to 1% of that contained in one fuel rod.
Paragraph 3. 3.L of the Technical Specifications requires such a demonstration prior to operating above 10 Mw.
(See Se*~t ion E. 3.)
3, Fuel Surveillance Program Examination of. guinea *pig rods has been completed at 5, 10, l.'.>, and 17.5 Mw during the power ascension program.
No evidence of fuel cladding failure has been observed during any part of the fuel surveil-lance program.
Paragraph 3.10.C of the Technical Specifications states that power shall not be increased above 15 or 17. 5 Mw until guinea pig rod examinations are satisfactory.
(See Section G.)
4, Core Physics Requirements Shutdown margin and core excess reactivity were reviewed and found to be as required by applicable sections of the Technical Specifications.
(See Sec t io n F
- 2 * )
- 5. Primary Coolant Samples Primary system sodium samples have been taken as required by paragraph 4.4.Q of the Technical Specifications.
(See Section E.2.)
- 6.
Nitrogen Cooling System Drver Condensate Sampling of the nitrogen cooling system dryer condensate was in accord-ance with the req~irements of paragraph 4.5.K of the Technical Specifi-cations.
(See Sect ion K. 1.)
- 7. Partial Pluggage of Cover Gas Supoly Line Partial pluggage in the reactor cover gas ~rgon supply line resulted in the inability to maintain the desired cover gas pressure.
Normal opera-tion was restored after clearing the pluggage with a "rotary snake. 11 (See Section E.1.)
Followup Items - See ~xhibit A for the current status of outstanding items.
Management Interview - The following items were discussed with Arterburn at the conclusion of the inspection:
- 1. Partial Scram (See Section F.3.)
The inspector stated that the time response data indicates that the cause of the partial scram had been identified and that the system as presently
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50-: 31;10-4 co f.p *
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adjusted should correct che problem.
Since routine surveillance tests failed to detect this Pfoblem, consideration should be given to revising the tests to verify that a scram cannot be reset either manually or by a change in the adjustment of the auxiliary lockout contacts. Arterburn
&tated that the method and frequency of performing such tests would be considered based on the recomrnendat ions of the G-E engineering group in Sunnyvale.
- 2. Argon Vent Vacuum Pumo Discharge Valves (See Section MJ.
Arterburn was informed that a written safety evaluation was not available for the installation of automatic flow control valves in the discharge lines from the redundant argon vent vacuum pumps.
A written-saf*ety evaluation is required by 10 CFR 50.59 to demonstrate that the modifica-tion did not involve an unreviewed safety quest:ion.
Arterburn stated that members of the Site Safety Committee had reviewed the changes individually in the course of approving the modification request form.
The inspector informed Arterburn that Region II would issue a Form AEC-592 for failure to maintain a written safety evaluation of this modification.
- 3. Discrepancies Between WRM' s and Heat Balances (See Section F. l}
Arterburn stated that diagnostic tests would be conducted based on recommendations from the G-E engineering group in Sunnyvale.
The hi&h flux level scram point will be lowered to 95% to prevent exceeding the Technical Specification limit at low core average temperatures.
Tests will be conducted at 5 Mw to determine the relationship between WRM indications and reflector position, sodium flow rate, and core inlet temperature.
Core average temperature will be varied from 400° to 760°F.
Arterburn stated that the problem will be reported to DRL in accordance with Section 3. C(2) of the operating license.
- 4.
Power Ascension Program Arterburn stated that all Group II power ascension tests have been com-pleted through the 17. 5 t1w level. Several of the Group III tests have been completed at 15 Mw.
Only a few t ests remain to be completed at 19 and 20 Mw before starting the balanced oscillator tests. Electronic equipment for these tests is currently being checked out for proper operation.
Out-of-cell and in-cell testing of the Fast Reactivity_Excursio" Device (FRED) is complete and the test data sent to Sunnyvale for analysis.
There was good comparison between actual and predicted results according to Arterburn.
CO Rpt. No. 50-231/70- :1 The reflector configur..::tiori at 17.26 l1w (September 5, 1970) and the associated worths were as follows:
Reflector No.
3 8
Position 61%
70%
Worth 96c llOc 206c Total Measured Worth 132c 133c 265c The difference between the total available excess and the required reactivity at 17.26 Mw is 59c.
The additional reactivity required to reach 20 Mw is (20 - 17.26) Mw x 6.5c/Mw = 18c.
This results in 41c exces s reactivity at 20 Mw.
The excess reactivity can be expected to decrease at a rate of le per month due to decay of Pu241 and 0.75c per full power day.
Technical Specifications 4.2.A states that Technical Specifications 3.3.A and 3.3.B shall be demonstrated at least once every four months after achieving 10 Mw.
Becker stated that this requirement is fulfilled each time the reactor is taken critical. The reactor operators are given the estimated critical reflector position prior to each startup and are instructed to notify the shift physicist if the actual critical position differs by more than 6 cm from the predicted value.
With a reflector in its position of maximum worth, 6 cm is equal to lOc and, therefore, would detect a change in core or reflector reactivity status greater than llOc.
Technical Specification 3.3.C states, "The reactor power coefficient of reactivity at constant inlet temperature and constant coolant flow rate shall be negative."
Values obtained from the power ascension data indicate that the power coefficient is equal to -6.5c/Mw.
Technical Specification 3.3.F states, "The reactor shall have a phase margin of at least 30 degrees at the point where the Nyquist plot crosses the unit circ le." The evaluation of the Nyquist plots using on-line Fourier data for flow of 30, 50 and 100% indicate phase margins ranging from 74-77° at the Nyquist unit circle intercept.
Technical Specifications 3.!0.A r.nd 3.10.B are satisfied by the preceding results.
- 3. Partial Scram Arterburn notified Region 11 that the reflector system had failed to function a~ designed on September 12, 1970, following a short duration scram signal.l/ The licensee's evauation and corrective actions YReported in Inquiry-Memoranch.::*!l c~ted Sept ember 18, 1970.
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p llpt, No, 50-231/70-4 following the occurrence were reviewed by V. D. Thomas, Complicnce Headquarters.
Thomas' review is contained in Exhibit B.
Swartz stated that he would write a letter to Arterburn recommending that routine surveillance tests be revised to preclude this type of failure in the future.
This item was added to the Outstanding Items List.
G.
Core and Internals Fuel Surveillance Program Results of the fuel surveillance program were discussed with Johnson.
Technical Specification 3.3.K states that fuel rods having defects defined below shall not be reinserted in the core:
- 1.
Cladding rupture, perforation, or other observable defects which may cast reasonable doubt on the integrity of the rods.
- 2. Local swelling of the cladding in excess of 10 mils or bowing of the rod sufficient to prevent reinsertion into the core.
- 3. An increase of more than one-half inch in the column height of either fuel segment.
Johnson stated that Item l above was confirmed by visual inspection.
Item 2 was checked by a diameter spiral trace and diameter trace at fixed azimuth in the fuel inspection station with a measuring sensitivity of 0.0002 inch.
Bowing and overall rod l ength are measured with indicators having a sensitivity of 0.004 inch.
Item 3 was checked by both trans-mission and emission gamma scans having a sensitivity of 1/32 inch.
Both scans produced traces of good quality and the emission scan showed sharp peaks at the pellet interfaces over almost the full length of the fuel column.
The results of all guinea pig rod inspections conducted to date are listed below.
Changes in pellet column length and cladding diameter are referenced to preirradiation data.
Rod Al6 Observable Defects Change in Pellet Column Length Cladding Diametral Change Bowing S Mw None
<1/16"
£0.00111 0.036" 15 Mw None
<l /1611
<O.OOltt 0.036" 17.S Mw None 1/16"
<:'O. 001 tt 0.036"
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Licensee:
Docket No. 50-231 Dates of Inspection:
October 6-7, 1970 Inspected By:
Vincent D. Thomas, Instrumentation Engineer Reviewed* By:
H. R. Denton, Chief, Technical Support Branch Proprietary Information:
None Int rod uct ion
/1/~*l?c:
Date
//;/;~:
1Date Ao announced inspection was performed at the SEFOR Nuclear Plant of the G*E Company and Southwest Atomic Energy Associates located in Cove Creek Township, Arkansas. This report is limited to the events surrounding the incomplete scram that occurred on September 12, 1970.
- The inspection was made to review the procedures, test data, and corrective action resulting from the subsequent investigation performed by G-E following the occurrence.
Summary
- 1. Following the incomplete scram event on September 12 and during the investigation that followed, the actuator tabs on all 18 DC contactors were adjusted s o that the auxiliary contacts would operate with or before the main contacts. Paul Swartz, G-E, stated that he will recommend that the site personne l establish a testing program so the timing sequence of contact operation is always known.
He aiso stated that the scram reset circuit will also be included as part of the te s t
- program,
- 2. Examination of t es t data revealed that nine of 12 safety systems and all six contai~ment building isolation DC contactors were not operating properly. Main contacts of the DC contactors wer e operating before the auxiliary contacts operated, Proper operation requires the opposite sequence of cont3ct action.
- 3. No at~empt was.. ~de to s imulate the conditions that caused the incident prior to adjus tment of the auxiliary contact operation. A short duration pulse test was completed efter the fix.
Exhibit B Page 1 of 9 I
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4* Failure in the scran re set circuit, which could have caused the occurrence, was not cons idered during the invc 1,tigation. Examination of the electrical circuits and test data revealed that of the four DC contactors directly involved in the incident, more than one and poss\\~!y four could have failed to operate properly at the time of the ~ident,
- 5.
Inspection of the main secondary flow controller set point station on September 14 revealed the adjustable slide wire-driven by the operator's thumbwhcel to be erratic on the low end and had a short in the area of 3000-4000 gp~ *.
6, Mr. Swartz stated that EVESR nuclear plant in Vallecitos, California is the only other nuclear facility that utilized this type contactor in the safety system.
He also stated that the EVESR facility is no longer operating.
7, G-E performed all testing requirements for the safety system, secondary flow monitoring system, and the reactor core outlet temperature moni~oring system set forth in the technical specifications.
DETAILS I.
PERSONS CONTACTED The foilowing G-E and Compliance personnel accompanied and/or were contacted during the inspection:
General Electric Company - SEFOR site Jesse Auterburn - Site Manager Mel Mathis - Instrumentation Engineer Dan Gellerman - Senior Instrumentation Technician General Electric Company - Sunnyvale Mel Muir - Manager, Control and Electrical System Unit Paul Swartz - Electrical Engineer Division of Comoliance Charles Upright - Reactor Inspector, CO:II Vincent Thomas - Instrumentation Engineer, CO:HQ II, RESULTS OF INSPECTION A,
Description of Events Surrounding Incomplete Scram On September 12, 1970, while attempting to increase the main secondary sodium flow, a short duration low flow trip signal caused the Exhibit B Pnge 2 of 9 1
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~itiation of a reactor scram action; however, en incomi>lcte scram occurred.
he events surrounding the "incomplete scram" *,1erc ns fol10~*1s:
- 1. "D" scram bus was in a trip condition due to a secured freon unit which caused a "low freon pressure" trip condition to
- exist,
- 2. The reactor was operating at 5 NW with an established flow rate in the main secondary sodium system of 1400 gpm.
- 3. The reactor operator was in the process of attempting to adjust to the secondary flow controller sc~ point station to increase the system flow rate to 4500 gpm.
- 4. An annunciator alarm - "low Na flow main secondary loop" -
was received and all reflector upper limit lights were extinguished, carriage separation had occurred on reflectors No. 3 and No. 8, and all position indicators were some 5 CM below full-up position. The power level on the wide range monitor (WRM) also indicated a decrease from 5 MW to approxi-mately 3 MW because of the reflector movement and bus "D" scram bus indicator was the only light lit. All scram solenoid position indicator s were extinghushed indicating no scram solenoids were deenergized.
- 5. The operator immediately scrammed the reactor.
B. Background The SEFOR safety system is a mult iple bus (A through F) type made up of r edundant electric magnetic relays (DC contact or). Fast acting primary r elays (four pole mercury-wetted contact type) are operated directly from trouble function trips, such as flow, pressure, and flux.
Contacts of these primary r elays are wired in series or series-paralle l arrangements, depending upon the trip logic, to make up the safety chain that operates the DC contactors. Contacts of the DC contactors arc wired in redundant one of three (A through C contactors) logic in series with redundant two of three (D through F) loeic matrices. The two logic matrices in turn operate two solenoid valves that are located in the hydraulic con-trol system of the reflector rod drives.
C.
Licensee's Conclusions as of September 14 1 1970 Review of the report dated September 14 to the site safety committee covering the events and subsequent investigation r evealed the following possible problem areas were considered and investigated:
- 1. Intermittent loss of 125VDC power to the hydraulic scram solenoids - Ruled out because of indication on event recorder pen No. 24 low flow ma.in secondary and the associated low flow alarm. Contacts 12 and three of K3 in any one relay chassis
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Exhibit B Pase 3 of 9
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(scram drawer s D, E, or F) had to close to de f l ect the pen on the event recorder.
- 2. Intermittent loss of 26.5VDC power to the safe ty control center - DC cohtactors - Ruled out prin~rily because or the indication on the event recor~er,
Loss of 26,5 volt power to the DC contactor coil circuitry would hot have produced a pen deflection on the event recorder and loss of 26.5 volt power to the scram drawers A through F would have deflected all of the pens associated with the affected chass is plus all associated annunciators would have been in the alarm condition.
In either case, DC contactors AlKl/A2Kl through C1Kl/C2Kl and E1Kl/E2Kl through F1Kl/F2Kl should have rer..a.ined deenergized (locked out).
DC contactors DlKl and D2Kl were not considered since this drawer (D)was in a scram (deenergized) condition.
- 3. Temporary loss of 125VDC to any one of the scram drawers A through C or E through F was also considered as a possible cause ; however, it was ruled out due to the deflection of pen No. 24 on the event recorder, The 125VDC power is not in the primary relay coil circuit it is in the relay contact circuit so that the pen deflection must have been an inter-ruption of the fast acting primary relay located in the E and/or F scram drawers.
This still does not explain the lack of a scram contactor's lockout feature. Fa ilcre of l ockout (auxiliary) contacts on Al/A2Kl through Fl/F2Kl to break was considered and it was determined that it could cause the affected contactor to reenergize if a short duration trip occurs.
- 4. The site personnel used two Simpson Voltohmists (VOH's) to determine if the hydraulic scram solenoid logic circuitry could be interrupted (main contacts of the DC contactor) without the lockout (auxiliary contacts from some DC contactors) ever opening.
One VOM was placed across one set of main contacts of the scram contact and the other was placed across the auxiliary contacts. The relay actuation was manually moved against its mechanical stop (fully energized) and slowly rcieas ed until one or both VOM's indicated open circuit. Movement was then continued until the second contact (if only one opened)VOH indicated open circuit. The test was repeated with the s econd set of main contacts of the scram contactor.
The results of the 125VDC.scram relay contact operation are listed below:
Exhibi t B Page 4 of 9 I,
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/,lKl A2Kl BlKl B2Kl ClKl C2Kl DlKl ElKl E2Kl FlKl F2Kl D2Kl
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Hain X
X X
X X
X X
X X
Auxiliary X
X X
Note:
X indicates which opened first It was felt from the information taken above that since one trip was already in the system (D bus) and the event recorder indicated a short duration trip from pen No. 24 that busses E and/or F received the trip causing interruption* of 125VDCt the main contacts of the scram relay open deenergizing the scram solenoids before the auxiliary contacts locked out the DC contactor coil.
The site safety committee then reviewed the findings above and agreed that the 12 contactors should be adjusted so that the auxiliary contacts would release before its main contacts.
This adjustment was *made by the site personnel on September 13.
No attempt was made to duplicate
- the initial conditions,
The adjustment consisted of repositioning the actuator tab so that the desired contact sequence would occur.
-Tests on the reactor control system (Q0-5) and ~onthly t e ~t s of the reactor safety system (M0-9) we re run. The licensee test procedures were comple ted in order to satis fy the r equire-ments set forth in the technical specifications and to de termine if the systems functioned properly following the adjustment.
On September 14 Dan Gellerman, Instrumentation Technician, discovered the cause of the short low ~low trip was due to the erratic behavior on the low end of the set point adjust station of the main secondary flow controller. It was found that a short in the ar~a of 3000-4000 gpm had exis t ed.- The nlldc wire wan rcpnirc<l and placed bnck in ncrvice ofter functional testo (W0-5) were performed on the set point stnt ion.
Exhibit B Page 5 of 9
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The report states that ::he <!Vents and investigation were discussed with the site personnel (Mathis) and Jim Haar of Sunnyvale.
D. Licensee's Cone lus ions as of September 21. 1970 Review of another report to the Site Safety Committee dated September 21,relating to the investigation of the incomplete scram event revealed the following:
- l. A series of tests were run to obtain drop-out t imcs of the scrams solenoid, the DC contactor and finally, the combination of both as a typical scram unit.
- 2. In the first series of tests 125VDC was applied to a spare scram solenoid via a spare primary relay. The relay was pulsed by a signal generator to determine the solenoid drop-out and pick-up times. A small voltage (1.5VDC) sign~l indi-cating contacts of the relay opening and closing and also the position of the scram solenoid were recorded on a multi ch~nnel high speed oscillograph. The results of approximately 50 recorded pulses show that the scram solenoid r equires a loss of power for at least 12-15 milliseconds before it will drop out and 8-20 milliseconds to pull back in.
- 3. The second series of tests were run to obtain information of the time relationship be tween the operation of the DC con-tactor (whose auxiliary contacts would open after the main contacts) and the solenoid scram valve. Using the same spa re components and t est setup used in the solenoid scram valve drop-out a nd pick-up test, the results revealed the following:
(a) Applying a series of varied pulse widths the DC contactor coi l circuit could be made to hold in while o~in contacts would open._ Pulses of 65-70 milliseconds in duration to the coi l circuit, both main contacts opened between 10-60 milliseconds. The DC contactor on several occasions dropped out.
(b)
Applying a series (11 recorded) of the 65-70 millisecond pulse widths show that the scram solenoid opened for 20-60 Qilliseconds ten times. One pulse width of 80 milliseconds deenergizcd the scram contactor and scram solenoid.
(c)
A 70-75 millinccond pu~,w would opr.n the " 11p,1rc tr:;t" scram relay 100'1. of the time while n 60-65 millisecond pulse width, will open the scram solenoid a.bout 40"1. of the time.
No attempt to further destort the actU.'.ltion tab on the DC contactor under test was made.
Exhibit B Page 6 of 9 I
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The report stat es that the events and investigation were discussed with the site personnel (Mathis ) and Jim Haar of Sunnyvale.
D. Licensee's Conclusio~s as of September 21, 1970 Review of another report to the Site Safety Committee dated September 21,relating to the investigation of the incomplete scram event revealed the following:
- 1. A series of tes ts were run to obtain drop-out times of the scrams solenoid, the DC contactor and finally, the combination of both as a typical scram unit.
2.
In the first series of tests 12SVDC was applied to a spare scram solenoid via a spare primary r e lay.
The relay was pulsed by a signal generator to det ermine the solenoid drop-out and pick-up times.
A small voltage (l.SVDC) s ign~l ind i-cating contacts of the relay opening and closing and also the positi on of the scram solenoid were recorded on a multi channe l high speed oscillogra ph. The r esults of approximately 50 recorded pulses show tha t the scram so lenoid requires a loss of power for at l eas t 12-15 milliseconds before it will drop out and 8-20 milliseconds to pull back in.
- 3. The second series of tests were run to obtain information of the time relationship between the operation of t he DC con-tactor (whose auxiliary contacts would open after the main contacts) and the solenoid scram va lve. Using t he same spar e components and test setup used in the sol enoid scram valve drop-out and pick-up test, the results revealed the following :
(a)
Applying a series of varied pulse widths the DC cont~ctor coil circuit could be made to hold in while main contacts would open._ Pulses of 65-70 milliseconds in duracion t o the coil circuit, both main contacts opened between 10-60 milliseconds. The DC contactor on several occas ions dropped out *
(b) Applying a series (11 recorded) of the 65-70 millisecond pulse widths show that the scr am solenoid opened for 20-60 cilliseconds t en times. One pulse width of 80 milliseconds deenergized the scram contactor and scram solenoid.
(c)
A 70-75 milli~c-cond pul_uc woul d opr:n the 11:;p,,rc t c-:;t" scrnr.1 rclo.y 100'7. of the time while a 60-65 mil 1 isccond pulse width, wil 1 open the scram sol enoid ilbout 40'7. o(
the time. No at t empt to further destort the actuation tab on the DC contactor under test was made.
Exhibit B A second inspection of the DC contnctors (12) Al/A2Kl through Fl/F2Kl along with the DC cont nctors (six) used in the containment isol~tion sys t em was made on September 20. Using the sam::: inspection techniques which employed the two VOM ' s previous ly discussed, it was found that all six containment i solation DC cont~ctors contact timing relationship was faulty, The site per ~onne l then proceeded to adjust the DC contactors i.n the same manne r as previously described until the auxiliary contacts r e -
leased before the main contacts, E. Licensee's Safety System Response Tes t September 24, 1970.
After the above adjustments and series of t ests were completed, the site personne l then performed a test to determine the safety systems '
(A through F) responses to simulated short durntion trips.
A set of contacts from the spa re primary relay was placed be twP.en each relay scram drawer and a selected trip unit. The primary r e lay was then pulsed with a series of varied pulse widths until one or both scram busses (solenoid dcene rgized) opened. Recordings were made of each short duration pulse showing its effect on the safety system. The test was repeated for all 12 DC contactors that make up the six safety s ystems of A through F.
The pulse widths in milliseconds at which point the DC contactor deenergized are tabulated as follows :
- AlKl-70 BlKl-45 ClKl-67 DlKl-75 ElKl-65 FlKl-55 A2Kl-65 B2Kl-55 C2Kl-45 D2Kl-75 E2Kl-60 F2Kl.:6o In a series of 72 r ecorded 45-75 milliseconds short duration trip signals sent to the six safety sys t em s cram drawers, it was found that i n all cases where the pulse was of suffic i ent duration to break the voltage to the scram solenoids, that scram went to completion, that i s, the auxiliary contacts on the DC contactor opened and the coil circuit deenergizcd.
Upon completion of the safety syst em response, short duration trip tests and two license te st procedures were performed as required by the technical specifications. They were as fo l lows:
W-0 Test of the Main Secondary F low Monitor W-0 Test of the Reactor Core Outle t Temperature Monitor F. Compliance Examination of Licensee Investigation G-E had stated in an earl ier telecon between the s ite personnel, CO:II and CO:HQ that one relay had failed, Initial r eview of the electrical Exhibit B I l I',
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circuits r evealed the poss ibility Lhat two. and possibly four, relays had failed. After d i scuss ing this matter with the s ite per sonnel, thr.y concurred with our observation. During the visit to the oite and after reviewing the data take n relating to the timing rclationshlp be tween the auxil i.ary and ma in contacts, the r.e lay which initiated the scram :i long with its scram drawer was tentative ly ide ntified as FlKl.
Since one scram drawer ("D" bus ) w.:is in a scram condition, any trip signal coming from E and/or F drawer s would ~ave caused a scr am to occur. Since each drawer has two DC contactors connected in para l lc 1, both will respond to the same trip signal; therefore, two and possib ly four had to malfunction.
G-E stated that the failure of the contacts within the same relay to operate in the proper sequence as being t he cause of the event. Review of the settings taken before the initia l co~
tact adjustments were made revealed that of the three " good" contactors,
scram drawers E and F had one each. This means that even if a short dura-tion trip occurred then the scram would have gone to completion since the auxiliary contacts would have released on the "good" scram relay before the main contacts. G-E s tated that the contactors do not have the same release and pick-up characteristics, even though they are identical.
However, further examination of their tes t data showed that E2Kl had a shorter tim~ response characteris tic than its redundant counterpart ElKl.
G-E then s tated that the response time data taken was an ave rage and not absolute accurate measurement.
This inspector then observed t hat "F" drawer had one "good" and one faulty r e lay where the faulty r elay had a shorter release time characteristic than its counterpart F2Kl.
This indicates that FlKl was the relay that "failed."
G-E did not disagree with this observation.
During the meeting held in Auterburn's office this inspector had asked if the scram reset circuit had been considered suspect at any time during the investigation and G-E stated that it had not.
G-E agreed tha t a failure in the reset circuit would produce the same res ults tha t occurred.
Additional discussion on this matter r evealed that the last time the r ese t circuit has been tested was during the initial pla nt preops (two years ago).
- . Paul Swartz then stated that a letter would be sent to the site r ecommending testing of t h is circuit at a more frequent rate. Swartz also s tated that.,,
in ad-~ition to this recomme ndation, periodic testing of the timing rela-tionship between all DC contactors used in both safety and containment isolation sys tems will be t ested. Auterburn st ated that he would wait until a 11 t he in format ion r elating to t he r ecommended t ests was reviewed before he would make a statement of agreement with these recommendations.
The review and examination of t he DC contactor t hat i s being uti-lized a s p:,rt of the safety and safegu:ird system s crnm and actu.1tinr. cir cuit~
t:,.v,.nl,*<I to tld.11 i111:pr,ctnr th:1L th,* lll<'thrnl nf: :itl_l1111t 111 ~~ th,, :imd I i11 ry c-011-t11c_l11 Jor II dlfl,-r,*111. Ll111l11~ 1*1*l11Ll1111,,lilp w:111 t:1!1* 0111-'.I* 11i:111111tl li,*11oll11i~ ol tnl>s.
Al 110 the u:i,~ of the auxiliary contact:i a s 11 H<\\111-111 control LOL i.Ls own coil circuit when not mounted on the same actuator l>oard as the main Exhibit B Page 8 of 9 i: I
. I I
I I l
' t
I contacts is of questionable practice.
This was discu~sed,,ith G-E personnel without their Cvlt'ments.
The s ite peL*sonnel follm,ed all testing r equirem':!nts set forth in the technical specifications for the safety system (M0-9) secondary flow monitoring system (W0-15). Test No. M0 S.:ifety and Safeguards S)6 terns was run after the initial DC contactor actuator ctrm tab adjust-ment on September 13, after the containment isolation building DC contactor tab adjustQent performed on September 20, and finally a ft er the s hort duration pulse tes ts. Test W0 Reactor Cor e Outlet Temperatur~
Monitors was r un following the short duration pulse test.
The flow set point adjustrrent signal ori&inating from the main secondary flow controller is common to the three low flow safety trips units. Since this i s a deviation from the separation (channel indepe ndence) requirements set forth in IBEE-279, the matter was brought to the at tention of DRS, DRS stated that the item had been r eviewed during the design r eview stages and was accepted.
No further examination of this item was made by the writer.
Mr. Swartz stated that EVESR nuclear plant is the only other nuclear facility that utili~ed this type contactor in the safety system He also stated that EVESR facility is no longer in operation.
Exhibit B Page 9 of 9
U. S. ATOMIC ENERGY COMMISSION REGION II DIVISION OF COMPLIANCE Report of Inspection CO Report No. 50-231/71-2 jcensee:
General Electric Company and Southwest
)ates of Inspection:
Jates of Previous Inspection:
Atomic Energy Associates (SEFOR)
License No. DR-15 Category B March 17-19, 1971 January 6-8, 1971 Inspected By:
C.
~ -Z.0-7/
Reviewed By*
C. M. Upri (In Charge o(?_Y. ~
Inspector (Operations)
R. )l:*
C bitt, Re tor Inspector (Operations)
~
- ~~~
W. C. Seidle, Proprietary Information:
Reactor Inspector None SCOPE A routine, announced inspection was conducted at the fast spectrum, sodium cooled SEFOR reactor located approximately 20 miles southwest of Fayetteville, Arkansas.
SW-iHARY Safety Items - None
!_oncompliance Items -
A Form AEC-592 was issued for the following items of noncompliance:
Date
t No. 50-231/ 71-2
,p * ~eactivity Control and Core Physics 1,
Failure of Core Outlet Temperature Scram.!/
Arterburn notified Region II by telephone on February 25, 1971, that the outlet temperature scram had failed during a routine surveillance test conducted on February 21, 1971.
A scram signal was not produced when a test switch was actuated to simulate high core outlet temperature.
Three temperature inputs are connected to 1/3 scram l ogic and the other two trips functioned properly.
G-E reported the failure to DRL in a letter dated March 2, 1971.
The failure was discussed with Mathis and found to be correct as reported.
The failure was caused by relay contacts which failed to open and deenergize 125 VDC relays that initiate scram action.
The contacts were believed to be damaged by a current surge back through scram chassis C when 125 VDC power to the scram relays tripped off during a checkout of system modifications.
(See Section F.4.)
Scram busses A and B were deenergized during the checkout and experienced no contact damage.
All other relay contacts in scram chassis C were checked for proper operation.
The chassis was removed and inspected for damage to components or wiring.
The faulty relay was replaced and satisfactorily tested.
To prevent a recurrence of this failure, Zener diodes were installed in parallel with the coils of all 18 Kl relays used for reactor scram and containment isolation.
The Kl relays are shown on Drawing No. 197R234 in FDSAR Supplement 2.
Proper system opera-tion was checked by LTP M-0-9, "Reactor Safety System Test,"
after completing the modifications.
The failure and subsequent modifications were reviewed and approved by the Site Safety ColIUllittee as required by the Technical Specifications.
- 2.
WRM - Heat Balance Correlation~/
Administrative control of high flux trip points was reviewed as specified in Engelken's memorandum to Skovholt dated November 27, Yrnquiry Memorandum dated March 2, 1971.
Yeo Report Nos. 50-231/70-4 and 71-1, Section F, and letter from G-E to Region II dated November 3, 1970.
I 11
ljlllsee:
J,.S of Inspection :
U.S. ATOMIC ENERGY COMHISSION REGION II DIVISION OF COMPLIANCE Report of Inspection CO Report No. 50-231/71-1 General Electric Company and Southwest Atomic Energy Associates (SEFOR)
License No. DR-15 Category B of Pr evious Inspection:
January 6-8, 1971 December 16, 1970 ected By:
Inspector (Operations) irkparick,Reactor Inspector (Operations) ewed By:
ffietary Information:
Specialist Reactor Inspector None SCOPE z-6 -7/
Date
~
~
Date
~fo/'II Date utine, announced inspection was conducted at the fast spectrum, um cooled SEFOR reactor located ap9roximately 20 miles southeast 1ayet teville, Ar kansas.
SUl1HARY
~ty I terns - None I
I ii 11 11
- 1
'I
~t. No. 50-231/71-1
~eactivity Control and Core Physics
- 1.
Wide-Range Monitor - Heat Balance Correlation.!./
Arterburn informed the inspector during the previous site inspec-tion that discrepancies between WRM indications and reactor heat balances would be reported to DRL within 30 days in accordance with paragraph 3. C. (2) of the license.
A report was submitted on November 3, 1970, but was sent to the Region II Compliance office instead of DRL as originally stated.
A copy of the report was forwarded to Compliance Headquarters on November 1q, 1970.
Administrative controls regarding adjustment of the high flux trip settings were reviewed as specified by Engelken's letter to Skovholt dated November 27, 1970.
WRM trip points were at 95%
and have been left set down since the discrepancies were first observed according to Arterburn and Becker.
Tr ip points will not be reset to 105% until reactor is increased above 17.5 Mw (87.5%) at which time reactor core average temperature would be maintained at 760°F.
Only one series of tests remain to be completed with core temperature below 760°F and the method of adjusting WRM trip points will be specified by the test pro-cedure.
Arterburn and Becker stated that the decision had not been made on whether to recalibrate to the lower temperature or to set the trip points down based on existing data.
Becker stated that the No. l WRM has been generally indicating higher than No. 2 or No. 3 for a given power level.
The difference appears to be drift in the No. l system but bench tests have not revealed any component changes that could cause drift.
There is no reason to believe that the detector gamma compensation is changing.
Becker stated that a program is being formulated which will determine if there is drift in the detector or in the amplifier.
Shift personnel have been instruct ed to notify Becker anytime a WRM is not within 2% of the reactor heat balance.
WRM operation and high flux trip setting will be reviewed during the next inspection.
- 2.
Partial S~ram'l-_/
Arterburn stated in the Management Interview following the pre-vious site inspection that consideration would be given to CO Report No. 50-231/70-4.
CO Report No. 50-231/70-4, Section F.3.
I.
1:
C p.pt. No. 50-231/71-1 revising existing surveillance tests to assure proper adjus t-ment of the auxiliary lockout contacts. Discussion of proposed tests with Mathis indicates that satisfactory testing will be performed.
An annual system test will require a complete inspec-tion of the scram relay (K-1) cabinets as well as insertion of a short duration scram signal while measuring the relative opening time of the main and auxiliary contacts.
This type of test was conducted to assure proper system operation following the partial scram discussed by V. C. Thomas of Compliance Headquarters in Appendix B of CO Report No. 50-231/70-4.
Mathis stated that the test would be used following modification of the scram relays to provide the 400 msec delay required by operation of the Fast Reactivity Excursion Device (FRED).
The procedures to accomplish the testing are in preparation and will be reviewed during the next inspection.
Auxiliary Systems
- 1. Failure of Main Secondary Expansion Tank Level Trip Arterburn notified the inspector by telephone on January 4, 1971, that one of the three low level trips on the main secondary expansion tank had failed to trip during monthly surveillance testing on January 3, 1971.
The other two channels functioned properly and provided a scram signal from the two-out-of-three logic.
The occurrence was reviewed during this inspection.
During the completion of monthly surveillance test M-0-3,
lowering the expansion tank level failed to cause a low level trip input from safety chassis F.
Safety chassises D and E tripped and provided a scram signal through the two-out-of-three logic circuit shown in FDSAR Supplement No. 2, diagram No. 197R234 of the reactor safety system.
When the instrument mechanic moved the trip unit chassis associated with L.S.
244-6, one of the relays apparently dropped out giving a trip input from safety chassis F.
Either one of two relays in the trip unit failing to drop out or having contacts that failed to open could have caused the malfunction.
One of the relays uses mercury wetted contacts and also provides alarm functions as well as the low level trip signal.
This relay is controlled by the other relay in the unit which operates on a 50 Mv signal from the level probe and is considered to be more susceptable to such a failure.
According to Hixson and Mathis, this is the first failure involving either type of relay.
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