ML19319B829
| ML19319B829 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 02/24/1978 |
| From: | Grant T TOLEDO EDISON CO. |
| To: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| 1-14, NUDOCS 8001280660 | |
| Download: ML19319B829 (75) | |
Text
/
- ],R
'y
- .A REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
' DISTRIBUTION FOR INCOMING MATERIAL 50-346 REC: KEPPLER J G ORG: GRANT J S DOCDATE: 02/24/78 NRC (OLEDO EDISON DATE RCVD: 03/06/78 DOCTYPE: LETTER NOTARIZED: NO COPIES RECEIVED
SUBJECT:
LTR 1 ENCL 1 REPORT OF FACILITY CHANGE TESTS. AND EXPERIMENTS CONDUCTED WITHOUT PRIOR NNRC APPROVAL FOR 1977 EVENTS.
PLANT NAME: DAVIS BESSE - UNIT 1 REVIEWER INITI AL:
XEF DISTRIBUTOR INITIAL:
ococo************ DISTRIBUTION OF THIS MATERIAL IS AS FOLLOWS ******************
NOTES:
1.
SEND ALL AMENDMENTS TO J.
ROE ANNUAL & SEMI-ANNUAL OPERATING REPORTS (OL STAGE).
(DISTRIBUTION CODE AOOS)
FOR ACTION:
PRO.f "'" EZ-'"**W/ ENCL LIC ASST HYLTON**W/5 ENCL INTERNAL:
REG FILE **W/EN L NRC PDR**W/ ENCL I i - C r - U ' 2 C M C L- -
MIPC**W/2 ENCL HANAUER**W/ ENCL STELLO**W/ ENCL EISENHUT**W/ ENCL SHAO**W/ ENCL BAER**W/ ENCL BUTLER **W/ ENCL GRIMES **W/ ENCL P.
CHECK **W/ ENCL J.
COLLINS **W/ ENCL EXTERNAL:
LPDR'S PT. CLINTON, OH**W/ ENCL NAT LAB **W/ ENCL TIC **W/ ENCL NSIC**W/ ENCL ACRS CAT A**W/15 ENCL DISTRIBUTION:
LTR 40 ENCL 40 CONTROL NBR:
780650149 SIZE: 1P+31P M
T E EN.
8 0 01280 ko
?
.P u:
a r,
a.
w-&m eg.,. -
xA' L-g i N/,,
'--2 n
c. L) Toteca e EDISON February 24, 1978 JraEs S. GnaNT
- u. o
..a.
Docket Io. 50-346
[y'[*L License :To. IIPF-3 Serial !!o. 1-14
/
l Mr. James G. Keppler Regional Director, Region III y,
Office of Inspection and Enforcenent U. S. Ifuelear Regulatory Co= mission
'Q.
799 Koosevelt Road Glen Ellyn, Illinois 60137
Dear lir. Keppler:
Enclosed find two (2) copies of P-3-77-1, Report of Facility Change Tests, and Experinents Conducted Without Prior !!RC Approval, for the required 1977 events.
Very truly yours, Ca..:.a JSG/JRL/1jk Enclosure Dr. Ernst Vol;:enau, DiEector cc:
Office of Inspection and Enforcement Enc 1:
40 copies P-3-77-1 4
THE TO'IDO EUR.ON C~ APANY EOGON ?LAZA 300 MAOfSON AVENUE TOLEDO. CHIO 43652 5.):: T.:. a ~
\\
~
s
+-..
a 6
O SUMHARY OF 1977 COMPLETED FACILITY CHANGE REQUESTS FCR NO: 77-001 SYSTEM: Reactor Vessel COMPONENT: Reactor Vessel Head "0"-Ring Clips CHANGE, TEST, OR EXPERIMENT:
New design "0"-Ring backing plates and "0"-Ring retainer clips were installed on the reactor vessel head.
This work was com-pleted on May 7, 1977.
REASON FOR THE FCR: The above change was made with the recommendation of and with the assistance of the reactor vendor, Babcock and Wilcox,in order.to incor-porate the improved design clip.
SAFETY EVALUATION: The new design clip recommended by Babcock and Wilcox per-forms the same function and fit as the old wire clip. The form is slightly changed in that the wire clip was replaced by a stamped metal clip. The stamped clip will perform its required function as well or better than the wire clip. The material specified, AISI TP-304, presents no problem. in this application. Therefore, there are no unreviewed safety questions and no change was required to the Technical Specifications, i
=
i e
~
.~
&o
SUMMARY
OF 1977 COMPLETED FACILITY CHANGE REQUESTS FCR NO: 77-002 SYSTEM: Safety Features Actuation System (SFAS)
COMPONENT: SFAS Channel 1 and 4 K03 Relays CHANCE, TEST, OR EXPERIMENT: On May 11, 1977, work was completed which wired out unused contacts on the K03 relays in SFAS Channels 1 and 4.
Work was completed per a Field Change Proceduru (KCS136), supplied by and under the supervision of the vendor, Consolidated Controls Corporation.
REASON FOR THE FCR: These initially spare contacts will be used as part of the pressuriser heater control interlock, a licenring committment to the NRC.
SAFETY EVALL'ATION: This modification to the SFAS does not affect the safety of this system for the following reasons:
1.
Work was done under the Field Change Fackage KGS-136 which was prepared by Censolidated Controls Corporation under their QA pro-gram.
2.
Isolation between the 12 sections of the circuits and non-lE sec-tions of the circuit was performed in cabinets external to the SFAS cabinet.
3.
No other safety functions of the SFAS will be affected by this modi-fication.
h 4
f m
6
SUMMARY
OF 1977 COMPLETED F CILITY CHANCE REQUESTS _
FCR NO: 77-006 3YSTEM: Miscellaneous COMPCNENT: Control Room Vertical Panels CHANGE, TEST, OR EXPERIMENT: On July 19, 1977, covers were fabricated and installed on the east and south ends of the vertical panels in the Control Room.
REASON FOR THE FCR: These panels were installed to close the east and south ends of these panels.
SAFETY EVALUATION:
The added vertical panel covers will in no way affect the safety function of the Control Room panels, nor will the seismic qualification be affected.
^
I SINMARY OF 1977 COMPLETED
- FACILITY CHANGE REQUESTS FCR NO: 77-034 SYSTEM: Reactor Coolant System (RCS)
COMPONENT: Decay Heat Normal Suction Valve Controls CHANGd, TEST, OR EXPERIMENT: On October 10, 1977, the control scheme of valve DH-12, Decay Heat (DH) Pump Suction Isolation Valve was modified in conformance with drawings E-52B, Sheet 24C, Revision P3 and E-523, Sheet 24D, Revision 3.
Bechtel Wiring Diagrams E-200B; E-221, Sheet 20; E-286; E-528; E-529; E-557A, Sheet 40A and 40B; and E-6343, Sheet 5 were revised to reflect the change.
The valve was stroked from all control points after th-e modifications were completed to ensure operability.
REASON FOR THE FCR: This wiring change was completed to prevent inadvertent closure of DH-12 when power is removed and then restored to the control circuit.
These closures were being caused by a contact racing condition when power was applied, a design deficiency within the control circuit. Closure of DH-12 isolates the RCS from the pressure relieving capacity of the normal cooldown relief valve in the DHR System. This could result in exceeding the pressure temperature limits on the primary system.
SAFETY EVALUATION: This change to the control scheme for valve DR-12 does not change the safety function of the interlock on the valve.
The valve cannot be opened when the RCO pressure is above the setpoint on PSH-RC2B4. Also, if the valve is open and 480 V AC power-is available to the valve operator, the valve will automatically close if the RCS pressure raises above the setpoint on PSH-RC2B4. This change only prevents the valve from inadvertently clcring due to an auxiliary relay contact race when 480 V AC power is restored to the valve operator.
e.
...g
SUMMARY
OF 1977 COMPLETED FECILITY CHANGE REQUESTS FCR NO: 77-041 SYSTEM: Safety Features Actuation System (SFAS)
CCMPONENT: " SAM" Logic for AF 599 CHANGE, TEST, OR EXPERIMENT: This Facility Change Request was written to docu-ment the disconnecting of the wire used for the logic of a valve previously actuated by the SFAS. The Auxiliary Feedwater Discharge Valve to Steam Genera-tor 2 (AF 599) logic was disconnected from the SFAS by lif ting Wire W of Tecninal Board TB 60-8 in Cabinet C 5756C.
REASON FOR THE FCR: The Auxiliary Feedwater Discharge Valve to Steam Cenerator 2 was removed from SFAS control. This Facility Change Request documented the spare cable and verified all drawing changes required were completed.
SAFETY EVALUATION: Valve AF-599 does not require a SAM light, as it is not closed with a safety actuation (SA) signal from SFAS.
Redundant position indications have been provided (computer and indicating lights) using redundant position switches on this valve.
I
m I
SIMMARY OF 1977 COMPLETED FACILITY CHANGE REQUESTS O
FCR NO: 77-079 SYSTEM: Radiation Monitoring System COMPONENT: Radiation Detectors RE 5327 and RE 5328 CHANCE, TEST, OR EXPERIMENT: On November 25, 1977, noise suppression mufflers were installed on Control Room Ventilation Radiation Detectors RE-5326 and RE-5328. These muf flers were supplied by Victoreen, the radiation detector vendor and installed with their supervison.
REASON FOR T74 FCR: This FCR was initiated to reduce the noise level in the Control Room caused by these radiation detectors. These detectors are solid piped to the ductwork which is conducive to sound transmittal to the ductwork.
SAFETY EVALUATION: The noise suppression mufflers perform no safety function nor do they affect the safety function of the associated radiation monitors.
The investigation by Victoreen indicated that the seismic qualification for these radiation monitors'was not invalidated by the installation of the mufflers.
Therefore, t e FCR does not involve an unreviewed safety question.
t
(
SUMMARY
OF 1977 COMPLETED
- FACILITY CHANGE REQUESTS FCR NO:
77-108 SYSTEM: Electrical Penetrations COMPONENT: Sealant Specifications CHANGE, TEST, OR EXPERIMENT:
The requirement for shielding type sealant material installed to a tuli two feet six inch depth in electrical penetrations as shown on Drawing E-378B, Sheet I and Bechtel Specification 7749-M-255 was changed to a depth of two feet, O inches of shielding sealant. The new specification also calls for sealing the insides of the conduits to a two feet depth.
REASON FOR THE FCR: The original specification and drawing were intended to be used to finish sealing of the penetrations and covering the exposed ends of the conduits not filled with sealant.
However, the shielding sealant has been applied to a depth of two feet, O inches inside all conduits, and the sealant applied in this manner accomplishes the dual purpose of adequate personnel shielding and the establishment of a pressure boundary, making additional seal-ing unnecessary.
SAFETY EVALUATION: An analysis of the electrical penetration shlesding property by the architect-engineer, Bechtel Power Corporation, indicated that a 2'-0" depth of SF-150 silicone foam will reduce gamma radiation by a factor of 1.85 x 10 and will reduce the neutron dose rate to less than 0.03 mr/hr.
Therefore, the sealant applied in the manner described above provides adeuqate personnel radiation protection.
1
.x
SUMMARY
OF 1977 COMPLETED FACILITY CHANGE REQUESTS FCR NO: 77-119 SYSTEM: Diesel Generator Fuel Oil Transfer System COMPONENT:
Local Control Switches for Fuel Oil Pumps CHANGE, TEST, OR EXPERIMENT: On August 16, 1977, work was completed which relocated the local control switches (NP 1951, 1952) for the Diesel Generator Fuel Oil Pumps.
REASON FOR THE FCR: The initial installation location of these control switches was inaccessible.
These switches are the only way to manually start or stop these pumps since there is no remote control station.
SAFETY EVALUATION: Relocation of the local control stations will not alter their function, which is to meet Technical Specification 4.8.1.1.2.a.3 (verify-ing that the fuel transfer pumps can be started and transfer fuel from the secrage system to the day tank). The safety function of the fuel oil system is not altered by this facility change.
1
SUMMARY
OF 1977 COMPLETED FACILITY CHANGE REQUESTS FCR NO:
77-120 SYSTEM: Nuclear Instrumentation COMPONENT: Source Range Neutron Detectors NI-l and NI-2 CHANCE, TEST, OR EXPERIMENT: On July 27, 1977, new cable was pulled in spare conduits 39035B and 39010D for Source Range Neutron Detectors NI-l and NI-2.
The necessary terminations were completed.and the initial cables were spared.
REASON FOR THE FCR: This work was done to attempt to reduce the noise level by separation or the source range cable.
SAFETY EVALUATION: There will be no adverse impact on safety'since the new cable installation will only change the cable routing.
t
4
SUMMARY
OF 1977 COMPLETED F5CILITY CHANCE REQUESTS j
FCR NO: 77-134 Steam and Feedwater Rupture Control System (SFRCS)
SYSTEM:
COMPONENT: SFECS Steam Generator Level Bistable Cabinets CHANCE, TEST, OR EXPERIMENT:
On July 9, 1977, work was completed which added This capacitors to the four SFRCS Steam Generator Level Bistable Cabinets.
work was performed per a vendor supplied Field Change Package (Consolidated Controls Corporation Field Change Package KHD 136).
REASON FOR THE FCR: This change was necessary to filter the " noise" from the analog input to these cabinets.
The noise was coming from the field wir-ing and the level transmitters.
SAFETY EVALUATION: This change will not affect the safety or seismic qualifi-cation of these steam generator leval. bistable cabinets. Also, the response time of steam generator level inscrumentation will remain within the total response time required by the SFRCS. Therefore, the safety'of the-SFRCS will not be affected.
9
4
SUMMARY
OF 1977 COMPLETED FACILITY CHANGE REQUESTS FCR NO:
77-139 SYSTEM: Steam and Feedwater Rupture Control System (SFRCS)
COMPONENT: Terminals on the SFRCS Cabinets C5762A and C5792 CHANCE. TEST, OR EXPERIMENT: On September 9, 1977, studs were installed on the field wiring side oi Terminal Block TB 11, Terminals 9,,10, 12 and 13 in SFRCS Cabinets C5762A and C5792. These studs were installed with a locknut to facilitate the placement of jumpers on these terminals.
REASON FOR THE FCR:
During the performance of Surveillance Test ST 5031.14, the "SFRCS Monthly Test", startup feedwater valves SP7A and SP7B control signal must be disabled by jumpering these contacts.
The initial installa-tion consisted of screws at these terminals which woald not allow jumpering without disrupting the circuit continuity, tripping the startup feed water valves.
SAFETY EVALUATION:
The addition of the studs to ther terminals is only for the purpose of conducting the SFRCS Monthly Test.
This change will have no effect on the functioning of the system, and hence no impact on facility safety.
W s
i
m s
SUMMARY
OF 1977 COMPLETED. FACILITY CHANGE REQUEST FCR No: 77-147 SYSTEM: Control Rod Drive Control (CRDC) System COMPONENT: Asymmetric Runback CHANCE, TEST, OR EXPERIMENT: On September 12, 1977, work was completed on the Control Rod Drive Control' System to modify the asymmetric runback function of the CRDC. The asymmetric runback to 60% of full power was revised so that both an asymmetric rod condition and an in-limit occur in order to initiate the automatic runback. The work was completed and tested per Babcock and Wilcox Field Change Package #69.
REASON FOR THE FCR: This was to prevent unnecessary runbacks to 60% of full power.
i SAFETY EVALUATION: Only the trip portion of the CRDCS is Q-listed.
Since this change does not affect the trip portion of the CRDCS, an unreviewed safety question is not involved.
~ E " jA.',
I I
l 1
r1 i
N
'N
SUMMARY
OF 1977 COMPLETED FACILITY CHANGE REQUESTS FCR NO: 77-150 SYSTEM: Core Flood System COMPONENT: Core Flood Injection Line Check Valves, Bechtel No. B3-3 and B3-4 CHANCE, TEST, OR EXPERIMENT: On August 15, 1977, Bechtel Specification 7749-M-212, and Velan valve Drawing D00-166-S were revised to reflect "as built" condi-tions. The nipples and caps welded to Core Flood Valves Injection Line Check Valves CF31 (Bechtel No. B3-3) and CF30 (Bechtel No. 33-4) were SA-479-316 rather than SA-182-F304 as required by tha specifications.
REASON FOR THE FCR: To reflect "as built" conditions.
SAFETY EVALUATION:
The change of the nipple and cap material from SA-182-F304 (as required by Specification M-212) to SA-479-316 will have no effect on the functioning of the Core Flood Valves and hcnce no effect on the safety of the system.
t
O
.s
)
SUMMARY
OF 1977 COMPLETED
- FACILITY CHANGE REQUESTS FCR NO: 77-151 SYSTEM: Component Cooling Water System COMPONENT: Component Cooling Water Heat Exchangers CHANGE, TEST OR EXPERIMENT: The requirement for a stress report on the Com-ponent Cooling Water Heat Exchangers was deleted from Bechtel Specification 7749-M-23, Form ED-6058.
REASON FOR THE FCR: ASME Code Section VIII does not require such a report and the report was no. available.
SAFETY EVALUATION: This Facility Change Request changes Specification 7749-M-23, Form ED-6058 in particular, to reflect the actual documentation requirements of Section VIII of the ASME Code. An unreviewed safety question is not involved, f
k s-i l'
I
+=
..oe
O m,
e
SUMMARY
OF 1977 COMPLETED PACILI.TY CHANGE REQUESTS FCR NO:
77-152 SYSTEM: Emergency Ventilation System COMPONENT:
Energy Pipe Break Blowout Panels Pressure Setpoint CHANCE, TEST, OR EXPERIMENT: On July 16, 1977, work was completed which increased the blowout setpoint of the negative pressure boundary high energy break blowout panels from.5 psid to 1.0 psid. This change was implemented on all thirteen panels.
High energy line breaks were re-analyzed in the areas of the Auxiliary Build-ing affected by the change in the relief points.
It was determined that peak pressures resulting from any of these breaks were consistently below the design limits in the affected areas.
REASON FOR THE FCR: A Field Quality Control Engineer employed by Bechtel (the architect / engineer for Davis-Besse Unit 1) asked for comparison of the set pressure of the blowout panels to the pressure calculated to occur on the panels under a Loss o'f Coolant Accident (LOCA) condition.
The subsequent Bechtel inves-tigation showed the blowout panels were designed to blowout at 0.5 psid which is less than the 0.8 psid that was calculated to occur after a LOCA.
These blow-out panels are part of the negative pressure boundary wall and were installed to protect the penetration rooms from overpressurization if a steamline or feed-water break would occur. These panels should have been designed not to blowout under LOCA pressure conditions.
SAFETY EVALUATION: The increase in the relief point of the blowout panels to 1.0 psid will not change the analyses associated with a LOCA.
The peak pres-sures in the areas of the Auxiliary Building affected by this change in the relief point are still below the design limits (Table 3-6 sa of the FSAR, Compartment Pressurization).
t
0'
,/
SUMMARY
OF 1977 COMPLETED. FACILITY CHANGE REQUESTS 4
FCR NO: 77-158 and 77-159 SYSTEM: Miscellaneous Piping COMPONENT: Miscellaneous Piping Specifications M-190 and M-601 CHANGE, TEST. OR EXPERIMENT:
Specification M-190 and M-601 were changed to permit the use of.previously installed unions, screwed fittings, screwed pipe, and stainless steel pipe and swagelock fittings-in class GCB, RBD, HSC piping.
REASON FOR THE FCR: Piping class sheets in Specifications M-190 and M-601 did not allow the use of some installed piping specified in Bechtel Non-Conformance Reports 1222, 1223, 1224, 1225, 1226, 1228, 1229 and.1230.
SAFETY EVALUATION: This change permits the use of unions, screwed joints, swage-lock fittings and stainless steel material on certain HBD lines (150 lb, carbon steel, ANSI B31.1.0) as snown on Notes 10 and 11 of M-601, Sheet 27; unions on GCB-8 lines (300 lb, stainless steel, ASME Section III, Class 2) as shown on Note 4 of M-601, Sheet 17; and unions on HSC-77 lines (150 lb, stainless steel radwaste, ASME Section III, Class 3), that is shown on Note 3 of M-601, Sheet 38.
These screwed connections and materials are permitted by the applicable codes'and standards and will not adversely affect the safety function of the systems to which these lines are connected.
4 i
'I
-s
)
s
SUMMARY
OF 1977 COMPLETED FACILITY CHANGE REQUESTS FCR NO: 77-160 SYSTEM: Instrument AC Power COMPONENT: Fuse Ratings for the Containment Hydrogen Analyzers CHANCE, TEST, OR EXPERIMENT: On December 29, 1977, Bechtel Electrical Drawing E-641A, Sheets 'In and 2A were revised to show the fuse ra.'inas for the Contain-ment Hydrog'en Analyzers (circuit Y119 and Y218) as 10, amps and the load rating as 5 amps.
REASON FOR THE FCR:
This Facility Change Request was issued to correct a drawing error. 'dith one hydrogen dilution pump and analyzer in operation, each channel exceeded the capacity of the previously installed 4 amp fuse.
I SAFETY EVALUATION: The changing of the fuse size in these cabinets will in no way adversely affect the safety function or seismic qualification of these panels or the containment hydrogen analyzer.
0 4
~I
n p
9
SUMMARY
OF 1977 COMILETED FACILITY CHANGE REQUESTS FCR NO:
77-166 SYSTEM: Emergency Diesel Generators COMPONENT: Fuel Oil Supply Lines CHANGE, TEST, OR EXPERIMENT: A 1/2 inch thick steel load distribution plate was added over the top of the two 1 "-H3 C-86 supply lines from the Emergency Diesel Generator Fuel Oil Storage Tanks (T153-1, and T153-2).
REASON FOR THE FCR: The existing structural backfill did not meet the missile protection requirements shown on Drawing 7749-C-37, Revision 9 (Bechtel Non-Conformance Report 1232) which required 3 feet of backfill.
SAFETY E'!ALUATION: The original structural backfill design required 3'-6" minimum cover for missile protection of the 1 "-HBC-86 (supply line). The revised design provides equal or better missile protection in that it employs both a 1/2" steel plate and 21" of backfill over the supply lines.
Therefore, the Facility Change-Request involves no adverse effects on the functioning of the emergency diesel generator system.
i 4
't
~
4,
SUMMARY
OF 1977 COMPLETED FACILITY CHANGE REQUESTS FCR NO: 77-167 SYSTEM: Steam Generator COMPONENT: Steam Generator Level Instrumentation' Cabinet 9N41 CHANCE, TEST. OR EXPERIMENT: On July 25, 1977, work was completed by Instru-ment and Control Personnel which modified 24 volt and 15 volt power supplies (PL9N41-18 and 19) for A.C.-input operation. Work was completed and tested per Field Change Procedure KHN136 supplied by the vendor, Consolidated Controls Corporation, o
REASON FOR THE FCR: This FCR was performed to correct a design error. The initial 15 and 24 vole power supplies did act operate properly on A.C. voltage supply.
SAFETY EVALUATION: This change to modify the power supplies for AC input operation does not adversely affect the operatica of the Steam Generator level instrumentation. The change has been designed by the original equip-ment supplier and approved by Bechtel as adequate for the intended service.
i I
i i
l
'j
SUMMARY
OF 1977 COMPLETED' FACILITY CHANGE REQUESTS FCR NO:
77-181 SYSTEM:
Sampling System COMPONENT: Nuclear Area Sampling CHANGE, TEST. OR EXPERIMENT: Piping and Instrument Diagrams M-031, M-037A, M-039, and M-040A were revised to show the secondary root valves thht exist on the Nuclear Area Samples.
No physical work was required. Only the drawings were revised to 4
document the double valves as they exist.
REASON FOR THE FCR: The reason for the drawing change was to clarify the piping and instrument diagrams to prevent misinterpretation of the valve lineup required.
SAFETY EVALUATION: These piping and instrument diagram changes will not affect the function of the system and do not involve an unreviewed safety question.
The normally shown valve positions reflect operating procedures which consider high radiation in the area of some of the sampling system valves.
,s
SUMMARY
OF 1977 COMPLETEE' FACILITY CHANGE REQUESTS FCR NO: 77-188 SYSTEM: Spent Fuel Pool Cooling COMPONENT: Refueling Canal CHANGE, TEST, OR EXPERIMENT: Bechtel Piping and Instrument Diagram 7749-M-035 was revised to indicate the actual capacity of the refueling canal as 388,000 gallons. Prior to this revision, the capacity was incorrectly given as 355,000 gallons.
REASON FOR THE FCR: To correct an earlier capacity calculation.
SAFETY-EVALUATION: This FCR changes Piping and Instrument Diagram M-035 to reflect the actual capacity of the refueling canal.
Decay heat from stored fuel is removed by two pumps and two heat exchangers.
Speat fuel pool water tempera-ture is automatically controlled by temperature cone. rollers, which are not affected by the total amount of water in the Spent Fuel Pool. Boron concentra-tion in the Spent Fuel Pdol is maintained to equal to or greater than the boron concentration in the refueling canal (1800 ppm) during refueling operations.
Therefore, the safety function of the system is not altered by this change.
4 n
SUMMARY
OF 1977 COMPLETED FACILITY CHANGE REQLTSTS FCR NO: 77-162 SYSTEM: High Pressure Injection (HPI) System COMPONENT: HP1 Lube Oil Pumps Wiring Diagram ES2B CHANGE, TEST. OR EXPERIMENT: The HPI Lube Oil Pumps, Bechtel Wiring Diagram E523, sheets 63 and 64, were revised to correct an error on the drawings.
The local switches were changed to show the normally open contacts and Note 3 was changed to "PDS contact opens on flow frem the A.C. Pump and closes to start the D.C. Pump when flow is lost from the A.C. Pump".
REASON FOR THE FCR: The drawings were revised to correct a drawing error.
"as-built" SAFETY EVALUAfl0N: The subject drawing changes are needed to reflect conditions and do not affect the ability of the equipment to perform its safety function.
i l
J i
f t
4 7
~
~
SUMMARY
OF 1977 COMPLETED FACILITY CHANGE REQUESTS l
FCR No:
77-196 SYSTEM:
A.C. Instrument Power COMPONENT:
Inverter YV1 CHANCE, TEST, OR EXPERIMENT: On September 7, 1977, modifications vera made to Essential Instrument A.C. Power Supply Inverter YV-1 to perform a test on inverter operation during short circuits at various phase angles.
This was done to verify the modification was in conformance wi'h Bechtel Specification 7749-E-20 prior to permanent implementation. Testi-eas completed and inverter YV-1 restored to pre-test status.
REASON FOR THE FCR:
In May, 1977, difficulties with current limiting in Inver-ters YV-1 and YV-2 led to investigation by the vendor, Cyberex.
This test was done as part of the investigation of these difficulties.
SAFETY EVALUATION: The procedure which specifies the Inverter YV-1 modification and testing stipulates that Channels 2, 3 and 4 of essential 120 volt AC instru-mentation panels are energized from their normal source.
Channel 1 panel dur-ing the modification and most of Inverter YV-1 testing is energized from its alternate source. These conditions are allowed by Technical Specification.
Therefore, for the period of time covered by the above conditions, there is no nuclear safety problem.
O
SUMMARY
OF 1977 COMPLETED. FACILITY CHMICE REQUESTS FCR NO:
77-199 SYSTEM: Auxiliary Feedwater COMPONENT: Auxiliary Feed Pumps Speed Control Circuits CHANCE TEST, OR EXPERIMENT: On August 17, 1977, two additional relays were wired in on each Auxiliary Feed Pump Control Circuit. The affected drawings, 7749-E-285, 7749-E-289 Sheet 1, 7749-E-045B Sheet 11A will be revised to reflect the above change.
REASON FOR THE FCR:
It appeared that the original relays did not have a high enough current capacity to operat'e ths' speed changer motor.
In an attempt to remedy this problem, the above change was made.
In retrospect, this change did not solve the speed control circuit difficulty. Facility Change Request 77-221 located and resolved the circuit deficiency.
SAFETY EVALUATION: These two new contacts split the current carried by the original lone relay. This change does not change the speed control function of this circuit.
Therefore, it will not degrade the safety of the unit.
.e*
A*
L.
s
SUMMARY
OF 1977 COMPLETED FACILITY CHANGE REQUESTS FCR NO:
77-200 SYSTEM:. Auxiliary Feedwater COMPONENT: Auxiliary Feedwater Pumps 1 and 2 CHANGE, TEST, OR EXPERIMENT: On December 14, 1977, work was completed which removed the dynamic braking resistors from the speed control circuits of both Auxiliary Feedwater Pump Turbines.
REASON FOR CHANGE: The originally installed resistors had a power rating which was too low.
This caused the resistors to overheat and fail.
SAFETY EVALUATION: The dynamic braking resistor was installed to reduce the RPM / pulse change.
This is pump RPM change (either increase or decrease) per speed change demand pulse from Steam Generator (SG) level error signal.
Elmination of the dynamic, braking resistor from the control circuit will not impair the safety system design function of maintaining Steam Generator level with the Auxiliary Feed Pump.
Test results during Hot Functional Testing (HFT) in December 1976/ January 1977 and during mini HFT of August 1977 reveal that without the resistor and with a shorted (zero ohm', resistor, the SG level was maintained using the Auxiliary Feed Pump.
1
l
SUMMARY
OF 1977 COMPLETED FA'CILITY CHANGE REQUESTS FCR NO:
77-217 SYSTEM: Varicas Power Supplies COMPONENT: Various CHANGE, TEST, OR EXPERIMENT: On October 4,1977, work was completed as follows:
1.
The time delay of the undervoltage relays set at 90% of 4.16 kV safety related bus voltage would be changed from 10 seconds to 9 seconds and the time delay of the diesel generator breaker was changed from I second to.5 second.
2.
An additional overvoltage alarm setpoint on the 4.16kV safety related bus was added.
3.
Overcurrent relay setpoints for safety related 4kV motors was revised.
4.
The 1 a=p fuses in the secondary of.the control power transformers located in all safety related and non-safety related motor control centers was changed to 2.5 amp slow blow fuses.
5.
The undervoltage alarm setpoints of the plant were reset to further enhance the coordination between the alarm aystem and the protective system.
The af fected drawings are identified by the FCR and will be revised.
REASON FOR THE FCR: The reason for the above changes was to meet committments to the NRC contained in Toledo Edison Letter No. 293 (dated July 18, 1977).
These changes remove the station from the voltage restriction condition 2.c.(3)(q) of the operating license.
SAFETY EVALUATION: These modifcations have previously been evaluated by the NRC by the submittal of Toledo Edison Letter No. 293.
9
^
SUMMARY
OF 1977 COMPLETED
- FACILITY CHANCE REQUESTS FCR NO:
77-231 SYSTEM: Main Steam COMPONENT: Main Steam Isolation Valve (MSIV) MS-100 and MS-101 CHANGE, TEST, OR EXPERIMENT: The reason for this Facility Change Request was to approve the bypassing of the instrument air line filter with copper tubing.
This work was performed by Maintenance personnel with verbal approval of Power Engineering on August 15, 1977. Also, Drawing 7749-M-304-6 Revision 11 was issued to reflect this condition.
REASON FOR THE FCR: The plastic cover of the instrument air line filter failed at the threads. This component failure caused a loss of instrument air to the valve which closes (fails safe) without' air pressure acting on the operator.
See Licensee Event Report NP-33-77-55 for further details.
SAFETY EVALUATION:
The filters were not necessary since the air supply is instrument air which is dry and clean.
By replacing the filter with tubing, the possibility of the MSIV inadvertently closing due to failure of the filter is eliminated.
O
j i
i
SUMMARY
OF 1977 COMPLETED FACILITY CHANGE REQUESTS FCR NO: 77-251 SYSTEM: Emergency Diesel Generators COMPONENT: Emergency Diesel Generator Fuel Oil Transfer Pumps CRANGE, TEST, OR EXPERIMENT: This FCR was issued to revise Bechtel Drawing ESK-M-888 to show the recirculating siphon. break piping on both Emergency
~
Diesel Generator Fuel Oil Transfe'r Pumps.-
'$# 7. s.,.
REASON FOR THE FCR: The drawi'ng was changed to reflect the "as built".condi-tion.
No physical change was completed.
SAFETY EVALUATION: The drawing change per this FCR reflects piping added of equal class for this siphon break. The design requirements of the system are not altered.
I f
i l
i i
r
n o
SUMMARY
OF 1977 COMPLETED'F CILITY CHANGE REQUESTS FCR NO: 77-307 SYSTEM: Main Steam COMPONENT : Main Steam Safety Valves
. - 3 '... ;,
CHANGE, TEST, OR EXPERIMENT: On October 26, 1977, work was completed which modified the main steam safety valve drain lines and associated supports.
Additional supports were installed to prevent damaging piping reactions Jering relief valve actuations. All affected drawings were revised to reflect the modifications.
REASON FOR THE FCR: The dhanges to the supports and piping was made to prevent the drain lines from being damaged when the safety valves relieve as had pre-viously occurred.
SAFETY EVALUATION: Revising the main steam safety valve drain line supports will not adversely affect the functioning of this system.
I s
e s
W S
SUMMARY
OF 1977 COMPLETED FACILITY CHANGE REQUESTS FCR NO: 77-327 SYSTEM: 480 Volt Motor Control Centers (CC) EllC and Fila CHANCE, TEST, OR EXPERIMENT: On October 6, 1977, molded case breaker settings were verified to be set correctly for breakers BE 1168 and BE 1150 on MCC EllC and BF 1114 on MCC Fila. The breakers BF 1114 and BE 1168 supply containment lighting. Breaker BE 1150 is the feed to MCC EllE.
REASON FOR THE FCR: This was a new installation which required verification of the breaker setting.
SAFETY EVALUATION: Only one change applies to "Q" listed circuitry, that is I
on Breaker BE 1150 on MCC EllC This change is required to permit proper coordination. This change in no manner adversely affects safety aspects of Davis-Besse Unit 1.
1 l
1
,S,UMMARY OF 1977 COMPLETED FACILITY CHANGE REQUESTS FCR NO:
77-406 SYSTEM: Decay Heat / Low Pressure Injection COMPONENT: Decay Heat Isolation Valve Pit Watertight Cover CHANCE, TEST, OR EXPERIMENT: On November 8, 1977, Revision 9 of Bechtel Drawing C-635 was completed which reflected the "as-built" condition of the Decay Heat Valve Pit Watertight Cover on Valves DH 11 and DH 12.
REASON FOR THE FCR: The reason for this change was to make the drawing coincide with the "as-built" condition.
SAFETY EVALUATION: The installed fasteners, 1/2" carbon steel bolts partially welded around the underside of the flange, provide equal or greater holddown capacity compared to originally designed 3/8" carbon steel studs.
The function of the valve pit cover is to prevent water leakage into the valve pit rendering inoperable the decay heat valves DH 11 and DH 12.
Because the watertight seal is provided by a silicone
- rubber gasket compressed around this removable panel, the safety functioning of this equipment is not adversely affected by this facility change.
e
SUMMARY
OF 1977 COMPLETED EACILITY CHANGE REQUESTS FCR NO: 77-417 SYSTEM: Control Rod Drive Control System (CRDCS)
COMPONENT: Undervoltage Trip Units CHANGE, TEST, OR EXPERIMENT: The originally supplied undervoltage trip units on all four CRDCS trip breakers were replaced with trip units of identical identifi-cation and function. After the replacement units were installed, testing of l
the breakers demonstrated the proper operation of the undervoltage trip devices.
l REASON FOR THE FCR: As a result of General Electric (the device manufacturer),
cad Diamond Power (the CRDCS vendor), investigation into delayed opening prob-lems at other utilities, it was recommended that the original undervoltage trip units be replaced.
SAFETY EVALUATION: The originally supplied undervoltage trip units on all four control rod drive control system trip breakers were replaced with trip units of identical identification and function. The original General Electric supplied undervoltage trip units may have been manufactured incorrectly and may have not performed their intended function. The batch of trip units that were incorrectly manufactured could not be positively identified by the marking on the trip unit.
However, a physical inspection would immediately identify the unit as being defective. Because a physical inspection of the trip unit would require that it be removed frem the circuit breaker, it was decided that the unit be totally replaced with a known correctly manufactured unit. After the replacecent units are installed, testing of the undervoltage trip circuits of the circuit breakers further demonstrated that the new trip units are operating properly.
Because the original suspect undervoltage trip units are to be replaced with identical properly manufactured trip units, no adverse affect on safety exists.
l i
l I
l
SUMMARY
OF 1977 COMPLETED FACILITY CHANGE REQUESTS FCR NO: 77-450 SYSTEM: Auxiliary Feedwater i
COMPONENT: Auxiliary Feed CHANGE, TEST, OR EXPERIMENT: On November 22, 1977, work was completed which installed a spring to the upper valve lever and to the lower support bracket on both Auxiliary Feed Pamps Turbine Covernors. The spring which was provided with a tension adjustment should balance the governor intake. Terry Turb' i
the turbine vendor supplied the parts and supervised the installation.
i REASON FOR THE FCR: Vibrational closure of Auxiliary Feed Pump 1-2 Governor Valve had occurred rendering the Auxiliary Feed Pump inoperable. Since there was no force except friction to hold the valve open in the original design, the gravitational forces combined with the vibrational forces caused the valve to close. This spring should prevent vibrational closure.
SAFETY EVALUATICN: The Auxiliary Feed Pump Turbine Governor Valve controls the speed of the turbine. This change, which adds a sp. Tg between the upper valve lever and the lower support bracket of the governor, will alter the linkage such that vibrat.on should not close the AFPT governor valve.
This is a design improvement, because the governor valve will remain open when it should be open, even in the presence of vibration. Therefore, the safety function of the Auxiliary Feedwater System will be enhanced.
i a
m~
. ~
w ywy3 w-i-
---w
,----w-y---
-e-r-
-a y
i t
,O A
7
SUMMARY
OF 1977 COMPLETED FACILITY CHANGE REQUESTS FCR NO:
77-451
.i SYSTEM: Auxiliary Feedwater COMPONENT: Auxiliary Feed Pumps Turbines Governors CHAN0E, TEST, OR EXPERIMENT: On ctober 6, 1977, a modification was completed on both Auxiliary Feed Pump Turbine Governors by the vendor, Woodward Goveraor.
The work was done at the Woodward facilities with manufacturing, assembling, and testing done by Woodward. Terry Turbine, the turbine vendor, witnessed the test and supervised the necessary drawing and parts identification changes.
REASON FOR THE FCR: The governors were returned to Woodward to eliminate the binding of the governor that had been experienced in previous testing.
I SAFETY EVALUATION: The modifications were necessary to remove the adverse effect of the governor binding under certain conditions. The modification allows the pilot valve to overtravel allowing the pivot bearing to always remain in contact with the floating lever. This in turn removes the possibility that the turbine will be prevented by this binding from reaching design speed. This design im-provement permits the Auxiliary Feed Pump to perform its intended safety function 4
and removes a potential defect which could create a significant safety hazard.
i
l
~
D
SUMMARY
OF 1977 COMPLETED FACILITY CHANGE REQUESTS i
FCR NO: 77-501 SYSTEM: Auxiliary Feedwater COMPONENT: Auxiliary Feed Pumps Turbines (AFPT) Governors l
I CHANCE, TEST, OR EXPERIMENT: On December 14, 1977, the installation of an oil I
lite bushing on the governors for both Auxiliary Feed Pump Turbines was com-pleted. Woodward, the governor vendor supplied the material and supervised the installation. The governors were tested by running ten cycles on each AFPT.
g i
REASON FOR THE FCR: The oil lite bushing was decided upon when binding where the speed setting shaf t enters the receiver assembly was discovered.
I 4
l
.l SAFETY EVALUATION: The Auxiliary Feed Pump Turbine governor controls the speed 1
of the turbi.ne.
This change, which adds a self-oiling bushing on the speed t
setting shaft, is to further decrease the probability that the AFPT governor j
will bind. This is a design i=provement and hence the safety function of the 1
Auxiliary Feedwater System will be enhanced.
1 1
t 4
1 4
m
~
- ~. ;,i:.
t i
i l
4 1
I ii
A Q&al.)
~
c.
.m C.
- w TOLEDO a EDISON February 24, 1978 JAMES S. GRANT
%se Pees. seat Docket No. 50-346
,',7',['3, License No. NPF-3 Serial No. 1-14 Mr. fames G. Keppler Regional Director, Region III Office of Inspection and Enforcement U. S. Nuclear Regulatory Cor: mission 799 Roosevelt Road Glen Ellyn, Illinois 60137
Dear Mr. Keppler:
Enclosed find two (2) copies of P-3-77-1, Report of Facility Change Tests, and Experiments Conducted Without Prior URC Approval, for the required 1977 events.
Very truly yours.
h Cha JSG/JRL/1jk Enclosure cc:
Dr. Ernst Volgenau, Director Office of Inspection and Enforcement Enc 1: 40 copies P-3-77-1 N
j ee
.g t i. !Q
=-J
.a i
THE TOLLDO EDSON CCAPANY EDtSON PLAZA 300 MADISDN AVENUE TOLEDO. OHIO 43652
,.~
o 1
d
SUMMARY
OF 1977 COMPLETED FACILITY CHANGE REQUESTS FCR NO: 77-001 SYSTEM: Reactor Vessel COMPONENT: Reactor Vessel Head "0"-Ring Clips CHANCE, TEST, OR EXPERIMENT:
New design "0"-Ring backing plates and "0"-Ring retainer clips were installed on the reactor vessel head.
This work was com-pleted on May 7, 1977.
REASON FOR THE FCR: The above change was made with the recommendation of and with the assistance of the reactor vendor, Babcock and Wilcox, in order to incor-porate the improved design clip.
SAFETY EVALUATION: The new design clip recommended by Babcock and Wilcox per-forms the same function and fit as the old wire clip. The form is slightly changed in that the wire clip was replaced by a etamped metal clip. The stamped clip will perform its required function as well or better than the wire clip. The material specified, AISI TP-304, presents no problem in this application. Therefore, there are no unreviewed safety questions and no change was required to the Technical Specifications.
e
'E gg 3
(..
SUMMARY
OF 1977 COMPLETED FACILITY CHANCE REQUESTS FCR NO:
77-002 SYSTEM: Safety Features Actuation System (SFAS)
COMPONENT: SFAS Channel 1 and 4 K03 Relays CHANCE, TEST, OR EXPERIMENT: On May 11, 1977, work was completed which wired out unused contacts on the K03 relays in SFAS Channels 1 and 4 Work was completed per a Field Change Procedure (KCS136), supplied by and under the supervision of thc vendor, Consolidated Controls Corporation.
REASON FOR THE FCR: These initially spare contacts will be used as part of the pressurizer heater control interlock, a licensing committment to the NRC.
SAFETY EVALUATION: This modification to the SFAS does not affect the safety of this system for the following reasons:
1.
Work'was done under the Field Change Package KGS-136 which was prepared by Consolidated Controls Corporation under their QA pro-gram.
2.
Isolation between the IE sections of the circuits and non-lE sec-tions of the circuit was perfor=ed in cabinets external to the SFAS cabinet.
3.
No other safety functions of the SFAS will be affected by this modi-fication.
.. l.,
3
,<~.
SUMHARY OF 1977 COMPLETED FACILITY CHANGE REQUESTS FCR No:
77-006 SYSTEM: Mi'Jeellaneous COMPONENT: Control Room Vertical Panels CHANGE, TEST, OR EXPERIMENT: On July 19, 1977, covers were fabricated and installed on the east and south ends of the vertical panels in the Control Room.
REASON FOR THE FCR:
.hese panels were installed to close the east and south ends of these p nels.
SAFETY EVALUATION: The added vertical panel covers will in no way affect the saf ety function of the Control Room panels, nor will the seismic qualification be affected.
^ f,. f *l...
. y -t,. ;'.
9
,s e
c
SUMMARY
OF 1977 COMPLETED FACILITY CHANGE REQUESTS FCR NO:
77-034 SYSTEM: Reactor Coolant System (RCS)
COMPONENT:
Decay Heat Normal Suction Valve Controls CHANGE. TEST..OR EXPERIMENT: On October 10, 1977, the control scheme of valve DH-12, Decay Heat (DH) Pump Suction Isolation Valve was modified in conformance with drawings E-52B, Sheet 24C, Revision P3 and E-523, Sheet 24D, Revision 3.
Bechtel Wiring Diagrams E-200B; E-221, Sheet 20; E-286; E-528; E-529; E-557A, Sheet 40A and 40B; and E-634B, Sheet 5 were revised to raflect the change.
The valve was stroked from all control points after the modifications were completed to ensure operability.
REASON FOR THE FCR: This wiring change was completed to prevent inadvertent closure of DH-12 when power is removed and then restczed to the control circuit.
These closures were being caused by a contact racing condition when power was applied, a design deficiency within the control circuit. Closure of DH-12 isolates the RCS from the pressure relieving capacity of the normal cooldown relief valve in the DHR System. This could result in exceeding the pressure temperature limits on the primary system.
SAFETY EVALUATION: This change to the control scheme for valve DH-12 does not change the safety function of the interlock on the valve.
The valve cannot be opened when the RCS pressure is above the setpoint on PSH-RC234 Also, if the valve is open and 480 V AC power is available to the valve operator, the valve will automatically close if the RCS pressure raises above the setpoint on PSH-RC2B4 This change only prevents the valve from inadvertently closing due to an auxiliary relay contact race when 480 V AC power is restored to the valve operator.
i e
[
r'
SUMMARY
OF 1977 COMPLETED FACILITY CHANCE REQUESTS FCR NO: 77-041 SYSTEM: Safety Features Actuation System (SFAS)
COMPONENT: " SAM" Logic for AF 599 CHANCE, TEST, OR EXPERIMENT: This Facility Change Request was written to docu-ment the disconnecting of the wire used for the logic cf a valve previously actuated by the SFAS. The Auxiliary Feedwater Discharge Valve to Steam Genera-tor 2 (AF 599) logic was disconnected from the SFAS by lif ting Wire W of Terminal Board TB 60-8 in Cabinet C 5756C.
REASON FOR THE FCR:
The Auxiliary Feedwater Discharge Valve to Steam Generator 2 was removed from SFAS control. This Facility Change Request documented the spare cable and verified all drawing changes required were completed.
I SAFETY EVALUATION: Valve AF-599 does not require a SAM light, as it is not closed with a safety actuation (SA) signal from SFAS. Redundant position indications have been provided (computer and indicating lights) using redundant position switches on this valve.
9 G
G
r'
~
SUMMARY
OF 1977 COMPLETED FACILITY CHANGE REQUESTS FCR NO:
77-079 SYSTEM: Radiation Monitoring System COMPONENT: Radiation Detectors RE 5327 and RE 5328 CHANCE, TEST, OR EXPERIMENT: On November 25, 1977, noise suppression mufflers were installed on Control Room Ventilation Radiation Detectors RE-5326 and RE-5328. These mufflers were supplied by Victoreen, the radiation detector vendor and installed with their supervison.
REASON FOR THE FCR: This FCR was initiated to reduce the noise level in the Control Room caused by these radiation detectors. These detectors are solid piped to the ductwork which is conducive to sound transmittal to the ductwork.
SAFETY EVALUATION,: The noise suppression mufflers perform no safety function nor do they affect the safety function of the associated radiation monitors.
The investigation by Victoreen indicated that the seismic qualification for these radiation monitors was not invalidated by the installation of the mufflers.
Therefore, the FCR does not involve an unreviewed safety question.
l 4
7
?
SUMMARY
OF 1977 COMPLETED' FACILITY CHANGE REQUESTS FCR NO: 77-108 SYSTEM: Electrical Penetrations COMPONENT:
Sealant Specifications CHANCE. TEST, OR EXPERIMENT: The requirement for shielding type sealant material installed to a full two feet six inch depth in electrical penetrations as shown on Drawing E-378B, Sheet 1 and Bechtel Specification 7749-M-255 was changed to a depth of two feet, O inches of shielding sealant. The new specification also calls for sealing the insides of the conduits to a two feet depth.
REASON FOR THE FCR: The original specification and drawing were intended to be used to finish sealing of the penetrations and covering the exposed ends of the conduits not filled with sealant.
However, the shielding sealant has been applied to a depth of two feet, O inches inside all conduits, and the sealant applied in this manner accomplishes the dual purpose of adequate personnel shielding and the establishment of a pressure boundary, making additional seal-ing unnecessary.
SAFETY EVALUATION: An analysis of the electrical penetrction shielding property by the architect-engineer, Bechtel Power Co'-) 1 ration, indicated that a 2'-0" depth of SF-150 silicone foam will ree er ma radiation by a factor of 1.85 x 10 and will reduce the neutron dose to less than 0.03 mr/hr. Therefore, the sealant applied in the manner dest toed above pt3vides adeuqate personnel radiation protection.
k
f%,
e
SUMMARY
OF 1977 COMPLETED FACILITY CHANGE REQUESTS TCR NO: 77-119 SYSTEM: Diesel Generator Fuel Oil Transfer System COMPONENT:
Local Control Switches for Fuel Oil Pumps CHANCE, TEST, OR EXPERIMENT: On August 16. 1977, work was completed which relocated the local control switches (NP 1, 1952) for the Diesel Generator Fuel Oil Pumps.
REASON FOR THE FCR: The initial installation location of these control switches was inaccessible. These switches are the only way to manually start or stop these pumps since there is no remote control station.
SAFETY EVALUATION: Relocation of the local control stations will not alter their function, which is to meet Technical Specification 4.8.1.1.2.a.3 (verify-ing that the fuel transfer pumps can be started and transfer fuel from the storage system to the day tank). The safety function of the fuel oil oyetem is not altered by this facility change.
f 9
/%'
. -m SUNMARY OF 1977 COMPLETED ACILITY CHANGE REQUESTS FCR No: 77-120 SYSTEM: Nuclear Instrumentation COMPONENT: Source Range Neutron Detectors NI-1 and NI-2 CHANCE, TEST. OR EXPERIMENT: On July 27, 1977, new cable was pulled in spare conduits 39035B and 39010D for Source Fiange Neutron Detectors NI-1 and NI-2.
The necessary terminations were completed and the initial cables were spared.
REASON FOR THE FCR: This work was done eq fttempt to reduce the noise level by separation of the source range cable.
SA,FETY EVALUATION: There will be no adverse impact on safety 'since the new cable installation will only changc the cable routing.
l.
1 10
. ~.
iy P
,SUNMARY OF 1977 CCMPLETED FA'CILITY CHANGE REQUESTS d
FCR NO: 77-134 Steam and Feedwater Rupture, Control System (SFRCS)
SYSTEM:
STRCS Steam Genarator Level Bistable Cabinets COMPCNENT:
On July 9, 1977, work was completed which added CHANCE, TEST, OR EXPERIMENT:
This capacitors to the four SFRCS Steam Generator Level Bistable Cabinets.
work was performed per a vendor supplied Field Change Package (Consolidated Controls Corporation Field Change Package KHD 136).
This change was necessary to filter the " noise" from REASON FOR THE FCR:
the analog input to these cabinets.
The noise was coming from the field wir-ing and the level transmitters.
SAFETY EVALUATION:
This change will not affect the safety or seistaic qualifi-cation of these steam generator level bistable cabinets. Also, the response time of steam generator level instrumentation will remain within the total Therefore, the sar y of the SFRCS will response time required by the SFRCS.
not be affected.
0 1
II
p m
SUMMARY
OF 1977 COMPLETED FACILITY CHANCE REQUESTS FCR No: 77-139 SYSTEM: Steam and Feedwater Rupture Control System (SFRCS)
COMPOff!NT: Terminals on the SFRCS Cabinets C5762A and C5792 I
CH/ ACE, TEST, OR EXPERIMENT: On September 9, 1977, studs were installed on the field wiring cide of Terminal Block TB 11, Terminals 9,10,12 and 13 in SFRCS Cabinets C5762A and C5792. These studs were installed with a locknut to facilitate the placement of jumpers on these terminals.
REASON FOR THE FCR: During the performance of Surveillance Test ST 5031.14, the "SFRCS Monthly Test", startup feedwater valves SP7A and SP7B control signal must be disabled by jumpering these contacts. The initial installa-tion consisted of screws at these terminals which would not allow jumpering without disrupting the circuit continuity, tripping the startup feed water valves.
SAFETY EVALUATION: The addition of the studs to ther terminals is only for the purpose of conducting the SFRCS Monthly Test.
This change will have no efhet on the functioning of the system, and hence no impact on facility safety.
I1
<3
SUMMARY
OF 1977 COMPLETED
- FACILITY CHANGE REQUEST FCR NO: 77-147 SYSTEM: Control Rod Drive Control (CRDC) System COMPONENT: Asynsnetric Runback CHANCE, tist. OR EXPERIMENT: On September 12, 1977, work was completed on the Control Rod Drive Control System to modify the asymmetric runback function of the CRDC.
The asymmetric runback to 60% of full power was revised so that both an asymmetric rod condition and an in-limit occur in order to initiate the automatic runback. The work was completed and tested per Babcock and Wilcox Field Change Package #69.
s REASON FOR THE FCR: This was to prevent unnecessary runbacks to 60% of full power.
SAFETY EVALUATION: Only the trip portion of the CRDCS is Q-listed.
Since
. this change does not affect the trip portion of the CRDCS, an unreviewed safety question is not involved.
-I
im
~
SUMMARY
OF 1977 COMPLETED. FACILITY CHANCE REQUESTS FCR NO: 77-150 SYSTEM: Core Flood System COMPONENT: Core Flood Injection Line Check Valves, Bechtel No. B3-3 and B3-4 CHANCE, TEST, OR EXPERIMENT: On August 15, 1977, Bechtel Specification 7749-M-212, and Velan valve Drawing DC9-166-S were revised to reflect "as built" condi-tions. The nipples and caps welded to Core Flood Valves Injection Line Check Valses CF31 (Bechtel No. B3-3) ant' CF30 (Bechtel No. B3-4) were SA-479-316 rather than SA-182-F304 as required by the specifications.
REASON FOR THE FCR: To reflect "as built" conditions.
SAFETY EVALUATION:
The change of the nipple and cap material from SA-182-F304 (as required by Specification M-212) to SA-479-316 will have no effect on the functioning of the Core Flood Valves and hence no effect on the safety of the system.
e IY
ry
~
SUMMARY
07 1977 COMPLETED. FACILITY CHANGE REQUESTS s
FCR NO: 77-151 SYSTEM: Component Cooling Water System COMPONENT: Component Cooling Water Heat Exchangers CHANCE, TEST. OR EXPERIMENT: The requirement for a stress report on the Com-ponent Cooling Water Heat Exchangers was deleted from Bechtel Specification 7749-M-23, Form ED-6058.
REASON FOR THE FCR: ASME Code Section VIII does not require such a report and the report was not available.
SAFETY EVALUATTON: This Facility Change Request changes Specification 7749-M-23, Form ED-6058 in particular, to reflect the actual documentation requirements of Section VIII of the ASME Codes An unreviewed safety question is not involved.
a n ;$.
tS v
^
-.3 m,
SUMMARY
OF 1977 COMPLETED FACILITY CHANGE REQUESTS FCR NO: 77-152 SYSTEM: Emergency Ventilation System COMPONENT: Energy Pipe Break Blowout Panels Pressure Setpoint CHANCE, TEST, OR EXPERIMENT: On July 16, 1977, work was completed which increased the blowout setpoint of the negative pressure boundary high energy break blowout panels from.5 psil to 1.0 psid.
This change was implemented on all thirteen panels.
High energy line breaks were re-analyzed in the areas of the Auxiliary Build-ing affected by the change in the relief points.
It was determined that peak pressures resulting from any of these breaks were consistently below the design limits in the affected areas.
REASON FOR THE FCR: A Field Quality Control Engineer employed by Bechtel (the architect / engineer for Davis-Besse Unit 1) asked for comparison of the set pressure of the blowout panels to the pressure calculated to occur on the panels i
under a Loss of Coolant Accident (LOCA) condition. The subsequent Bechtel inves-tigation showed the blowout panels were. designed to blowout at 0.5 psid which is less than the 0.8 psid that was calculated to occur after a LOCA.
These blow-out panels are part of the negative pressure boundary wall and were installed to protect the penetration rooms from overpressurization if a steamline or feed-water break would occur. These panels should have been designed not to blowout under LOCA pressure conditions.
SAFETY EVALUATION: The increase in the relief point of the blowout panels to 1.0 psid will not change the analyses associated with a LOCA. The peak pres-sures in the areas of the' Auxiliary Building affected by this change in the relief point are still below the design limits (Table 3-6 aa of the FSAR, Compartment Pressurization).
i l
j
[b
SUMKARY OF 1977 COMPLETED FACILITY CHANGE REQUESTS FCR No: 77-158 and 77-159 SYSTEM: Miscellaneous Piping COMPONENT: Miscellaneous Piping Specifications M-190 and M-601 CHANGE, TEST, OR EXPERIMENT: Specification M-190 and M-601 were changed to permit the use of previously installed unions, screwed fittings, screwed pipe, and stainless steel pipe and swagelock fittings in class GCB, HBD, HSC piping.
REASON FOR THE FCR: Piping class sheets in Specifications M-190 and M-601 did not allow the use of some installed piping specified in Bechtel Non-Conformance Reports 1222, 1223, 1224, 1225, 1226, 1228, 1229 and 1230:
SAFETY EVALUATION: This change permits the use of unions, screwed joints, swage-lock fittings and stainless steel material on certain HBD lines (150 lb, carbon steel, ANSI B31.1.0) as shown on Notes 10 and 11 of M-601, Sheet 27; unions on CCB-8 lines (300 lb, stainless steel, ASME Section III, Class 2) as shewn on Note 4 of M-601, Sheet 17; and unions on HSC-77 lines (150 lb, stainless steel radwaste, ASME Section III, Class 3), that is shown on Note 3 of M-601, Sheet 38.
These screwed connections and materials are permitted by the applicable codes and standarda and will not adversely affect the safety function of the systems to which these lines are connected.
17
SUMMARY
OF 1977 COMPLETED NACILITY CHANCE REQUESTS FCR NO: 77-160 SYSTEM: Instrument AC Power COMPONENT: Fuse Ratings for the Containment Hydrogen Analyzers CHANCE. TEST. OR EXPERIMENT: On December 29, 1977, Bechtel Electrical Drawing E-641A, Sheets lA and 2A were revised to show the fuse ratings for the Contain-ment Hydrogen Analyzers (circuit Y119 and Y218) as 10. amps and the load rating as 5 amps.
REASON FOR THE FCR:
This Facility Change Request was issued to correct a drawing error. With one hydrogen dilution pump and analyzer in operation, each channel exceeded the capacity of the previously installed 4 amp fuse.
SAFETY EVALUATION: The changing of the fuse size in these cabinets will in no way adversely af fect the safety function or seismic qualification of these panels or the containment hydrogen analyzer.
l l
SUMMARY
OF 1977 COMPLETED EACILITY CHANCE REQUESTS FCR NO: 77-166 SYSTEM: Emergency Diesel Generators Cat?1NENT: Fuel Oil Supply Lines CHANCE, TEST, OR EXPERIMENT: A 1/2 inch thick steel load distribution place was added over the top of the two 1\\"-H3 C-86 supply lines from the Emergency Diesel Generator Fuel Oil Storage Tanks (T153-1, and T153-2).
REASON FOR THE FCR: The existing structural backfill did not meet the missile protection requirements shown on Drawing 7749-C-37, Revision 9 (Bechtel Non-Conformance Report 1232) which required 3 feet of backfill.
SAFETY EVALUATION: The original structural backfill design required 3'-6" minimum cover for missile protection of the 1 "-H3C-86 (supply line).
The revised design provides equal or better missile protection in that it employs both a 1/2" steel plate and 21" of backfill over the supply lines.
Therefore, the Facility Change involves no adverse ef fects on the functioning of the emergency diesel Request generator system.
SUMMARY
OF 1977 COMPLETED FACILITY CHANCE REQUESTS FCR N9:
77-167 SYSTEM: Steam Generator COMPONENT: Steam Generator Level Instrumentation Cabinet 9N41 CHANCE, TEST. OR EXPERIMENT: On July 25, 1977, work was completed by Instru-ment and Control Personnel which modified 24 volt and 15 voit power supplies (PL9N41-18 and 19) for A.C. input operation. Work was completed and tested per Field Change Procedure KHN136 supplied by the vendor, Consolidated Controls Corporation.
REASON FOR THE FCR: This FCR was performed to correct a design error. The initial 15 and 24 volt power supplies did not operate properly on A.C. voltage supply.
SAFETY EVALUATION: This change to modify the pcwer supplies for AC input operation does not adversely affect the operation of the Steam Generator level instru=entation. The change has been designed by the original equip-ment supplier and approved by Bechtel as adequate for the intended service.
ac
SUMMARY
OF 1977 COMPLETED FACILITY CHANGE REQUESTS FCR NO: 77-181 SYSTEM: Sampling System COMPONENT: Nuclear Area Sampling CHANCE. TEST, OR EXPERIMENT: Piping and Instrument Diagrams M-031, M-037A, M-039, and M-040A were revised to show the secondary root valves that exist on the l'uclear Area Samples. No physical work was required. Only the drawings were revised to document the double valves as they exist.
REASON FOR THE FCR: The reason for the drawing change was to clarify the piping and instrument diagrams to prevent misinterpretation of the valve lineup required.
SAFETY EVALUATION: These piping and instrument diagram changes will not affect the function of the system and do not involve an unreviewed safety question.
The normally shoen valve positions reflect operating procedures which consider high radiation in the area of some of the sampling system valves.
O
SUMMARY
OF 1977 COMPLETED [ACILITY CRANCE REQUESTS FCR NO: 77-182 SYSTEM: High Pressure Injection (HPI) System COMPONENT: HPI Lube Oil Pumps Wiring Diagram E52B CHANCE, TEST, OR EXPERIMENT: The HPI Lube Oil Purps, Bechtel Wiring Diagram
_E52B, Sheets 63 and 64, were revised to correct an error on the drawings.
The local switches were changed to show the normally open contacts and Note 3 was changed to "PDS contact opens on flow from the A.C. Pump and closes to start the D.C. Pump when flow is lost from the A.C. Pump".
REASON FOR THE FCR: The drawings were revised to correct a drawing error.
SAFETY EVALUATION: The subject drawing changes are needed to reflect "as-built" conditions and do not affect the ability of the equipment to perform its safety function.
4 I
SUMMARY
OF 1977 COMPLETED pACILITY CHANGE REQUESTS FCR NO:
77-188 SYSTEM:
Spent Fuel Pool Cooling COMPONENT: Refueling Canal CHANGE, TEST, OR EXPERIMENT:
Bechtel Piping and Instrument Diagram 7749-M-035 was revised to indicate the actual capacity of the refueling canal as 388,000 gallons. Prior to this revision, the capacity was incorrectly given as 355,000 gallons.
REASON FOR THE FCR: To correct an earlier capacity calculation.
SAFETY EVALUATION: This FCR changes Piping and Instrument Diag;am M-035 to reflect ene actual capacity of the refueling canal. Decay heat from stored fuel is removed by two pu=ps and two heat exchangers. Spent fuel pool water tempera-ture is automatically controlled by temperature controllers, which are not affected by the
-:stal amount of water in the Spent Fuel Pool.
Boron concentra-tion in the Spent Fuel Pool is maintained to equal to or greater than the boron concentration in the refueling canal (1800 ppm) during refueling operations.
Therefore, the safety function of the system is not altered by this change.
SUMMARY
OF 1977 COMPLETED FACILITY CHANGE REQUESTS FCR NO: 77-196 SYSTEM:
A.C. Instrument Power COMPONE!TT:
Inverter YV1 CHANCE, TEST, OR EXPERIMENT: On September 7, 1977, modifications were made to Essential Instrument A.C. Power Supply Inverter YV-1 to perform a test on inverter operation during short circuits at various phase angles.
This was done to verify the modification was in conformance with Bechtel Specification 7749-E-20 prior to permanent implementation. Testing was completed and inverter YV-1 restored to pre-test status.
REASON FOR THE FCR:
In May, 1977, difficulties with current limiting in Inver-ters YV-1 and YV-2 led to investigation by the vendor, Cyberex.
This test was done as part of the investigation of these difficulties.
S AFETY EV.41L'.". TION : The procedure which specifies the Inverter YV-1 modification and testing stipulates that Channels 2, 3 and 4 of essential 120 volt AC instru-mentation panels are energized from their normal source.
Channel 1 panel dur-ing the modificatiot.and most of Inverter YV-1 testing is energized from its alternate source. These conditions are allowed by Technical Specification.
Therefore, for the period of time covered by the above conditions, there is no nuclea, safety problem.
2Y
SUMMARY
OF 1977 COMPLETED FACILITY CHANGE REQUESTS FCR NO: 77-199 SYSTEM: Auxiliary Feedwater COMPONENT: Auxiliary Feed Pumps Speed Control Circuits CHANGE, TEST, OR EXPERIMENT: On August 17, 1977, two additional relays were wired in on each Auxiliary Feed Pump Control Circuit.
The affected drawings, 7749-E-285, 7749-E-289 Sheet 1, 7749-E-045B Sheet ilA will be revised to reflect the above change.
REASON FOR THE FCR:
It appeared that the original relays did not have a high enough current capacity to operate the speed changer motor.
In an attempt to remedy this problem, the above change was made.
In retrospect, this change did not solve the speed control circuit difficulty.
Facility Change Request 77-221 located and resolved the circuit deficiency.
SAFETY EVALUATION: These two new contacts split the current carried ' - the original lone relay. This change does not change the speed control f-necion of this circuit. Therefore, it will not degrade the safety of the unit.
SEKKARY OF 1977 COMPLETED FACILITY CHANGE REQUESTS FL*R NO:
77-200 SYSTEM: Auxiliary Feedwater COMPONENT: Auxiliary Feedwater Pumps 1 and 2 a
CHANCE, TEST, OR EXPERIMENT: On December 14, 1977, work was completed which removed the dynamic braking resistors from the speed control circuits of both Auxiliary Feedwater Pump Turbines.
REASON FOR CHANGE: The originally installed resistors had a power rating which was too low.
This caused the resistors to overheat and fail.
SAFETY EVALUATION: The dynamic braking resistor was installed to reduce the RPM / pulse change. This is pump RPM change (either increase or decrease) per speed change demand pulse from Steam Generator (SG) level error signal.
Elmination of the dynamic braking resistor from the control circuit will not imprir the safety system design function of maintaining Steam Generator level witu the Auxiliary Feed Pump.
Test results during Hr t Functional Testing (HFT) in December 1976/ January 1977 and during mini HFT of August 1977 reveal that without the resistor and with a shorted (zero ohm) resistor, the SG level was maintained using the Auxiliary Feed Pump.
SUMMARY
OF 1977 COMPLETED FA'CILITY CHANGE REQUESTS FCR NO: 77-217 SYSTEM: Various Power Supplies COMPONENT: Various CHANCE, TEST, OR EXPERIMENT: On October 4, 1977, work was completed as follows:
1.
The time delay of the undervoltage relays set at 90% of 4.16 LV safety related bus voltage would be changed from 10 seconds to 9 seconds and the time delay of the diesel generator breaker was changed from I second to.5 second.
2.
An additional overvoltage alarm setpoint on the 4.16kV safety related bus was added.
3.
Overcurrent relay setpoints for safety related 4kV motors was revised.
4.
The 1 amp fuses in the secondary of the control power transformers located in all safety related and non-safety related motor control centers was changed to 2.5 amp slow blow fuses.
5.
The undervoltage alarm setpoints of the plant were reset to further enhance the coordination between the alarm system and the protective system.
The affected drawings are identified by the FCR and will be revised.
REASON FOR THE FCR: The reason for the above changes was to meet committments to the NRC contained in Toledo Edison Letter No. 293 (dated July 18, 1977).
These changes remove the station from the voltage restriction condition 2.c.(3)(q) of the operating license.
SAFETY EVALUATION: These modifcations have previously been evaluated by the NRC by the submittal of Toledo Edison Letter No. 293.
s
SUMMARY
OF 1977 COMPLETED TACILITY CHANCE REQUESTS i
1 FCR NO: 77-231 SYSTEM: Main Steam COMPONENT: Main Steam Isolation Valve (MSIV) MS-100 and MS-101 CHANGE, TEST, OR EXPERIMENT: The reason for this Facility Change Request was to approve the bypassing of the instr' agent air line filter with copper tubing.
This work was performed by Maintenance personnel with verbal approval of Power Engineering on August 15, 1977. Also, Drawing 7749-M-304-6 Revision 11 was issued to reflect this condition.
REASON FOR THE FCR: The plastic cover of the instrument air line filter failed at the threads. This component failure caused a loss of instrument air to the valve which closes (fails safe) withaut air pressure acting on the operator.
See Licensee Event Report NP-33-77-55 for further details.
SAFETY EVALUATION: The filters were not necessary since the air supply is instrument air which is dry and clean. By replacing the filter with tubing, the possibility of the MSIV inadvertently closing due to failure of the filter is eliminated.
a
~
SUMMARY
OF 1977 COMPLETED FACILITY CRANGE REQUESTS FCR NO: 77-251 SYSTEM: Emergency Diesel Generators COMPONENT:
Emergency Diesel Generator Fuel Oil Transfer Pumps CHANCE, TEST, OR EXPERIMENT: This FCR was issued to revise Bechtel Drawing FSK-M-888 to show the recirculating siphon break piping on both Emergency Diesel Generator Fuel Oil Transfer Pumps.
REASON FOR THE FCR: The drawing was changed to reflect the "as built" condi-tion. No physical change was completed.
SAFETY EVALUATION:
The drawing change per this FCR reflects piping added of equal class for this siphon break. The design requirements of the system are not altered.
Ci
SUMMARY
OF 1977 COMPLETED FiCILITY CHANGE REQUESTS I
jFCRNO: 77-307 SYSTEM: Main Steam COMPONENT: Main Steam Safety Valves CHANCE, TEST, OR EXPERIMENT: On October 26, 1977, work was completed which modified the main steam safety valve drain lines and associated supports.
Additional supports were installed to prevent damaging piping reactions during relief valve actuations. All af fected drawings were revised to reflect the modifications.
REASON FOR THE FCR: The changes to the supports and piping was made to prevent the drain lines f rom being damaged when the safety valves relieve as had pre-viously occurred.
SAFETY EVALUATION: Revising the main steam safety valve drain line supports will not adversely af fect.the functioning of this system.
4 l
l l
l -
30 i
\\
SUMMARY
OF 1977 COMPLETED _ FACILITY CHANGE REQUESTS FCR NO: 77-327 SYSTEM: 480 Volt Motor Control Centers (CC) EllC and Fila CHANCE, TEST, OR EXPERIMENT: On October 6, 1977, molded case breaker settings were verified to be set correctly for breakers BE 1168 and BE 1150 on MCC EllC and BF 1114 on MCC Fila. The breakers BF 1114 and BE 1168 supply containment lighting. Breaker BE 1150 is the feed to MCC EllE.
REASON FOR THE FCR: This was a new installation which required verification of the breaker setting.
SAFETY EVALUATION: Only one change applies to "Q" listed circuitry, that is on Breaker BE 1150 on MCC EllC i
1 This change is required to pe.mit proper coordination. This change in no manner adversely affects safety aspects of Davis-Besse Unit 1.
l l
I i
3l
.s
SUMMARY
OF 1977 COMPLETED FACILITY CHANGE REQUESTS c.. -Q ;,.
.FCR NO: 77-406 4
SYSTEM: Decay Heat / Low Pressure Injection COMPONENT: Decay Heat Isolation Valve Pit Watertight Cover CHANCE, TEST, OR EXPERIMENT: On November 8,1977 Revision 9 of Bechtel Drawing C-635 was completed which reflected the "as-built" condition of the Decay Heat Valve Pit Watertight Cover on Valves DH 11 and DH 12.
REASON FOR THE FCR: The reason for this change was to make the drawing coincide with the "as-built" condition.
SAFETY EVALUATION: The installed fasteners, 1/2" carbon steel bolts partially welded around the underside of the flange, provide equal or greater holddown capacity compared to originally designed 3/S" carbon steel studs.
The function of the valve pit cover is to prevent water leakage into the valve pit rendering inoperable the decay heat valves DH 11 and DH 12.
Because the watertight seal is provided by a silicone rubber gasket compressed around this removable. panel, the safety functioning of this equipment is not adversely affected by this facility change.
D e
32
s
SUMMARY
OF 1977 COMPLETED FACILITY CHANGE REQUESTS FCR NO:
77-417 SYSTEM: Control Rod Drive Control System (CRDCS)
COMPONENT: Undervoltage Trip Units CHANCE, TEST. OR EXPERIMENT: The originally supplied undervoltage trip units on all four CRDCS trip breakers were replaced with trip units of identical identifi-cation and function. After the replacement units were installed, testing of the breakers demonstrated the proper operation of the undervoltage trip devices.
REASON FOR THE FCR: As a result of General Electric (the device manufacturer),
and Diamond Power (the CRDCS vendor),. investigation into delayed opening prob-lems at other utilities, it was recommended that the original undervoltage trip units be replaced.
SAFETY EVALUATION: The originally supplied undervoltage trip units on all four control rod drive control system trip breakers were replaced with trip units of identical identification and function. The original General Electric supplied undervoltage trip units may have been manufactured incorrectly and may have not performed their intended function. The batch of trip units that were incorrectly manufactured could r.ot be positively identified by the marking on the trip unit.
However, a physical inspection would immediately identify the unit as being defective. Because a physical inspection of the trip unit would require that it be removed from the circuit breaker, it was decided that the unit be totally replaced with a known correctly manufactured unit. After the replacement units are installed, testing of the undervoltage trip circuits of the circuit breakers further demonstrated that the new trip units are operating ycoperly.
Be cause the original raspect undervoltage trip units are to be repl sued with identical p sperly manuractured trip units, no adverse affect on saf aty exists.
.~
o l
?
SUMMARY
OF 1977 COMPLETED FACILITY CHANGE REQUESTS FCR NO: 77-450 SYSTEM: Auxiliary Feedwater COMPONENT: Auxiliary Feed CHANCE, TEST. OR EXPERIMENT: On November 22, 1977, work was completed which installed a spring to the upper valve lever and to the lower support bracket on both Auxiliary Feed Pumps Turbine Governors. The spring which was provided with a tension adjustment should balance the governor intake. Terry Turbine, the turbine vendor supplied the parts and supervised the installation.
REASON FOR THE FCR: Vibrational closure of Auxiliary Feed Pump 1-2 Governor Valve had occurred rendering the A -iliary, Feed Pu=p inoperable.
Since there was no force except friction to hcid the valve open in the original design, the gravitational forces combined with the vibrational forces caused the valve to close.
This spring should prevent vibrational closure.
SAFETY EVALUATION: The Auxiliary Feed Pump Turbi..e Governor Valve controls the speed of the turbine.
This change, which adds a spring between the upper valve lever and the lower support bracket of the governor, will alter the linkage such that vibration should not close the AFPT governor valve.
This is a design improvement, because the governor valve will remain open when it should be open, even in the presence of vibration. Therefore, the safety function of the Auxiliary Feedwater System will be enhanced.
i
+
~
L f
i
- s, 7
SUMMARY
OF 1977 COMPLETED FACILITY CHANGE REQUESTS FCR NO: 77-451 EYSTEM: Auxiliary Feedwater COMPONENT: Auxiliary Feed Pumps Turbines Governors CHANCE, TEST, OR EXPERIMENT: On October 6,1977, a modification was completed on both Auxiliary Feed Pump Turbine Governors by the vendor, Woodward Governor.
The work was done at the Woodward facilities with manufacturing, assembling, and testing done by Woodward. Tarry Turbine, the turbine vendor, witnessed the test and supervised the necessary drawing and parts identification changes.
REASON FOR THE FCR: The governors were returned to Woodward to eliminate the binding of the governor that had been experienced in previous testing.
SAFETY EVALUATION: The modifications were necessary to remove the adverse effect of the governor binding under certain conditions.
The modification allows the pilot valve to overtravel allowing the pivot bearing to always remain in contact with the floating lever. This in turn removes the possibility that the turbine will be prevented by this binding from reaching design speed.
This design im-provement permits the Auxiliary Feed Pump to perform its intended safety function and removes a potential defect which could create a significant safety hazard.
i l
1
~
3S
-,s
.s
,(
t
SUMMARY
OF 1977 COMPLETED FACILITY CHANGE REQUESTS 4
FCR NO: 77-501 SYSTEM: Auxiliary Feedwater COMPONENT: Auxiliary Feed Pumps Turbines (AFPT) Covernors CHANGE, TEST, OR EXPERIMENT: On December 14, 1977, the installation of an oil lite bushing on the governors for both Auxiliary Feed Pump Turbines was com-pleted. Woodward, the governor vend 8r supplied the material and supervised the installation. The governors were tested by running ten cycles on each AFPT.
REASON FOR THE FCR: The oil lite bushing was decided upon when binding where the speed setting shaft enters the receiver assembly was discovered.
SAFETY EVALUATION: The Auxiliary Feed Pump Turbine governor controls the speed of the turbine. This change, which adds a self-oiling bushing on the speed setting shaft, is to further decrease the probability that the AFPT governor will bind. This is a design i= prove =ent and hence the safety function of the Auxiliary, Feedwater System will be enhanced.
3 b