ML18227D918
| ML18227D918 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 11/04/1976 |
| From: | Florida Power & Light Co |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| Download: ML18227D918 (69) | |
Text
TURKEY POINT PLANT UNITS 3 AND 0 DOCKET NUMBERS 50-250 AND 50-251 CHANGES, TESTS, AND EXPERIMENTS MADE WITHOUT PRIOR COMMISSION APPROVAL FOR PERIOD JANUARY 1, 1976, THROUGH JUNE 30, 1976 IN COMPLIANCE WITH TITLE 10, SECTION 50.59 (s)
CODE OF FEDERAL REGULATIONS
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ROU 0 This report is submitted in accordance with 10 CFR 50.59 (b), which requires that reports of:
i) changes in the facility as described in the FSAR, ii) changes in the procedures as described in the FSAR, and iii) tests and experiments not described in the FSAR which are conducted without prior commission approval be reported to the Commission at least annually. This report is intended to meet this require-ment for the period January 1, 1976, through June 30, 1976. Changes, tests, and experiments prior to January 1, 1976, were reported in the appropriate Semi-annual Operating Report in accordance with the Technical Specifications in effect at the time. Future reports submitted pursuant to 10 CFR 50.59 (b) are intended to be submitted annually and will cover the preceding July 1 to June 30 period.
This report is divided into three "sections: the first, Plant Change/Modifi-cations, covering changes in the facility as described in the FSAR; the second, Procedure Changes, covering changes in the procedures as described in the FSAR; and the third, Tests and Experiments, covering tests and exper-Ments not described in the FSAR.
B 'F CONTEl'TS EKE IllTRODUCTION e s ~ s t s s e e e e e t e s e e e e s s t s 1 TABLE OF CONTENTS '...,...,,...,,,...,, 2 PLANT CHANGE/NOD I F I CAT IONS, PROCEDURE CHANGES . . . 24 TESTS AND EXPERINENTS . . . , . . . ~ ~ ~
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<6 0 0 S During the reporting period, the below listed Plant Changes and Modifi-cations (PC/Ms) were completed. They are split into the following two groups:
I. PC/Ms which were partially completed prior to this reporting period and which were previously reported in Semiannual Oper-ating Reports.
II. PC/Ms not previously reported.
The first group is listed below with the current status. The second group is briefly described with a summary of the safety evaluation.
GROUP I PC/Ms PREVIOUSLY REPORTED
- 1) 74-75 RCC Change Fixture Limit Switch Modification (Completed on Unit 4).
- 2) 74-90 Containment Pressure Switch with Smaller Range (Completed on Unit 4).
- 3) 75-15 Turbine Cross-over and Under Piping Modification (Completed on Unit 3).
- 4) 75-50 Reactor Coolant Pump Vibration System Installation (Com-pleted on Unit 4).
- 5) 75-55 New Fuel Elevator - Add Limit Switch (Completed on Unit 4).
- 6) 75-63 Fuel Transfer Cart Add Redundant Limit Switch (Completed on Unit 4).
- 7) 75-68 Reactor Coolant Pump Modification of No. 3 Seal Ring Clamp (Completed on Unit 4).
- 8) 75-70 Spent Fuel Pit Bridge Crane Hoist (Completed on Unit 4).
- 9) 75-77 Turbine Low Load Limit Switch Installation (Completed on Unit 4).
I GROUP II PC/Ms NOT PREVIOUSLY REPORTED.
- 1) 74-47 Install Check Valves Emergency Diesel Air Start System.
- 2) 75-8 Startup Transformer Lockout Relay Circuit Modification.
- 3) 75-11 Boron Injection Tank Redundant Recirculation System.
- 4) 75-37 Charging Pump Suction Drain
- 5) 75-51 Modification of 240 KV Switchgear Breaker Motor Operator.
- 6) 75-61 Replace Lube Oil and Hydrogen Coolers'ontrol Valves.
- 7) 75-64 Process Radiation Monitoring System, R-20, Letdown Monitor-Recorder Meter Agreement.
- 8) 75-65 Process Radiation Monitoring System, Installation of Digital Ratemeter.
- 9) 75-67 Crosby Relief Valve Modification.
- 10) 75-69 Improve Ground Bus.
- 11) 75-78 RHR Valves, MOV-750 & 751, Interlock Limitation.
- 12) 75-79 Containment Sump-Add Redundant Level Transmitter.
- 13) 76-2 Control Rod Drive Mechanisms - Install Connectors.
- 14) 76-4 Integrated Leak Rate Test Relative Humidity Monitoring System.
- 15) 76-10 Modification of Steam Generator Blowdown Piping.
- 16) 76-13 Retube Condenser Water Boxes with Cu-Ni and Titanium Tubes.
- 17) 76-15 Remove Wall between Nuclear Plant Supervisor's Office and Emergency Exit Hallway.
- 18) 76-16 Replace Auxiliary Transformer's Radiator.
- 19) 76-20 Charging Pump Material Changes.
t Plant Change/Modification 74-47 Units 3 and 4 "Install Check Valves Emergency Diesel Air Start System'he check valves in the air start system were changed to lift type check their seats valves. The swing check valves which were installed hammered against due to pulsating pressure created by the air compressor. The lift check valves have an air dashpot above the piston to cushion the disc.
..This change does not pose any unreviewed safety questions because:
(1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report (FSAR) has not been increased.
FSAR Section 8.2.3 states that one of the two Emergency Diesel Generators must be operable in order to reduce the probability of occurrence and consequences of a maximum hypothetical accident (MHA).
Changing the check valves will not result in an interconnection between the two diesel generators and thus will not increase the probability of a simulataneous failure.
Thus, the probability of occurrence or the consequences of an accident or previously evaluated in the t
malfunction of equipment important to safety FSAR has not been increased.
(2) The possibility for an accident or malfunction'f a different type than any evaluated previously in the FSAR has not been created.
The FSAR has already considered the failure of an emergency diesel generator to operate properly. A failure of the new style check valve would have no more serious consequences than a failure of the swing check type.
In fact, the failure of the lift checkfortype valve is much less likely, because it was designed specifically pulsating air service. In addition, the lift check valve is designed for pressure up to 1000 psi compared to only 400 psi for the present swing check.
(3) The margin of safety as defined in the basis for any technical specification has not been decreased.
Section B 3.7 of the Technical Specifications states that each diesel unit has the capability to start and run the required engineered safeguards for an MHA in one unit and to safely shutdown the second unit..
Changing the check valves will in no way effect the above stated basis for the Technical Specifications.
- 2. Plant Change Modification 75-51 Units 3 and 4 "Modify MG-9 Motor Operators in the Plant Switchyard" All the Turkey Point Plant Switchyard McGraw-Edison 240 KV disconnect switch motor operator engaging devices were replaced with I.T.E. devices.
Also, the tops of the motor operator cabinets were replaced.
The pops were replaced due to excessive rusting and water leakage. The I.T.E. disengaging devices can be more easily operated, locked open or locked closed, and visually checked for position.
This change does not pose any unreviewed safety questions because:
(1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has'ot been increased.
This change is essentially a modification of the FPL transmission system and does not constitute a change to any nuclear safety-related system or component.
(2) The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.
(3) The margin of safety as defined in the basis for any technical speci-fication has not been decreased.
I Units 3and 4 "Modification of the Startup Transformer Lockout Relay Circuit" d
One isolation diode was added in the Startup Transformer Lockout relay circuit. The purpose of the diode is to prevent "sneak" circuit signals from actuating the S. U. Transformer control circuit backup relay. Actuation of the backup relay can result in the tripping of several breaker backup protections.
This "sneak" circuit signal would be present only under special testing condi-tions when the lockout relay lead is lifted.
This change does not pose any unreviewed safety questions because:
(1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.
The ~orst related malfunction that could conceivably occur would be loss of external electrical load, which has been analyzed in the FSAR. This change decreases the probability of an inadvertant loss of electrical load.
(2) The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.
(3) The margin of safety as defined in the basis for any technical specifi-cation has not been decreased.
0 4. Plant Change/Modffftation 75-11 Unit 4 "Boron Injection Tank Redundant Recirculation System" A redundant piping system has been installed to ensure the ability to recirculate the Unit 4 Boron Injection Tank.
,This change does not pose any unreviewed safety questions because:
(l) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.
The loss of the recirculation line during accident conditions would not affect the accident consequences as evaluated in the FSAR. The modification added redundancy in that it duplicated the function of the original recirculation line.
(2) The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.
The function and design are identical to the original recirculation line.
(3) The margin of safety as defined in the basis for any technical specification has not been decreased.
The margin of safety is increased by having a second recirculation path available.
Unit 4 "Charging Pump Suction Piping Drains" In order to provide a means for draining the charging pumps for maintenance, a drain was installed on the suction piping.
This change does not pose any unreviewed safety questions because:
(1) The probability of occurrence or the consequences of an accident or malfunction of equipment important'to safety previously evaluated in the Final Safety Analysis Report has not been increased.
The worst accident that would be created would be inadvertent release of the contents of the Volume Control Tank into the Charging Pump Room.
Liquid spillage in the Auxiliary Building was evaluated in the FSAR and the consequences and probability of an inadvertent release of the con-tents of the Volume Control Tank would be less than the analyzed condition.
(2) The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.
(3) The margin of safety as defined in the basis for any technical specifi-cation has not been decreased.
- 6. Plant Change/Modification 75-61 Units 3 and 4 "Replace Lube Oil and Hydrogen Coolers Temperature Control Valves" The originally installed control valves were oversized, and proper control in automatic was not possible. Without automatic control, excessive turbine and generator vibration could occur.
The replacements for CV-2200 and CV-2203 are Fisher "V" Vee-ball valves.
The new valves are sized for design flow under actual, measured conditions.
The new Vee-ball valves have a much greater turn down ratio enabling auto-matic control at considerably less flow than butterfly valves of similiar size.
This change does not pose any unreviewed safety questions. This system is for the lubricating oil system cooler for the turbine and the hydrogen cooler for the generator, neither of which can affect the safety of the nuclear reactor.
- 7. Plant Change/Modification 75-64 Units 3 and 4 "Process Radiation Monitoring System R-20 Meter-Recorder Agreement" With the original circuit the R-20 (Letdown Line Monitor) chart recorder printed out a value which was 25 percent higher than its ratemeter indication.
To correct this a 400 ohm resistor was installed in the drawer output to the chart recorder. This corrected the discrepancy between the two readings.
The accuracy of the channel was not changed by this modification.
This change does not pose any unreviewed safety questions because:
(1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.
This monitor has no automatic functions associated with it. Its prime purpose is to alert the operator to a fuel cladding leak.
(2) The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.
(3) The margin of safety as defined in the basis for any technical specifi-cation has not been decreased.
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- 8. P1ant 'Chango/Modification 75-65 Units 3 and 4 "Process Radiation Monitoring System Installation of Digital Ratemeters" This change provided for replacing the analog ratemeters with digital readouts. It facilitated adjustment of the alarma maintenance setpoint and allowed department operators to adjust the setpoint whereas before representative was required to adjust the setpoint.
The digital ratemeters are adaptable to all Process Radiation Monitoring System channels except R-20, Reactor Coolant System letdown flow monitor.
The ratemeters were installed in channels R-14, stack effluent, and R-18, liquid release.
This change does not pose any unreviewed safety questions because:
(1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.
The change does not affect the operation or functions of the Process Radiation Monitoring System instruments.
(2) The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.
(3) The margin of safety as defined in the basis for any technical speci-fication has not been decreased.
- 9. Plant Change/Modification 75-67 Units 3 and 4 "Substitute Crosby Relief Valve No. H-51341-7 for No. H-51341-2" The subject Crosby type relief valve is used in several plant systems.
The manufacturer made some minor modifications in the design of his replacement valves in order to simplify manufacture and assembly of the valves. The changes were of a minor nature and did not affect the dimensions or functions of the valve. The change was necessary because when ordering replacements only the latest model is available.
This change does not pose any unreviewed safety questions because:
(1) The probability of occurrence or the consequences of an accident or
. malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.
These valves are manufacturer's replacement valves for the same service.
As such, they are functionally and dimensionally equivalent.
(2) The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.
(3) The margin of safety as defined in the basis for any technical speci-fication has not been decreased.
0 10. Plant Change/Modification 75-69 Units 3 and 4 "Additional Ground Bus" Additional grounding conductors were added between the plant grid and the switchyard, and also between all line coupling capacitors. The purpose was to minimize the ohmic resistance to ground.
This addition does not pose any unreviewed safety questions because:
(1) (a) With respect to the probability of occurrence of an accident previously evaluated in the FSAR: The addition of more ground bus can only result in improving the presently installed system.
(b) With respect to the consequences of an accident previously evaluated in the FSAR: The consequences remain unchanged, since the original ground bus will remain.
(c) With respect to the probability and consequences of malfunctions of equipment important to safety previously evaluated in the FSAR: Safety evaluation of original equipment is unaffected by these changes.
(2) (a) With respect to the possibility of an accident of a different type than any analyzed in the FSAR: The installation will improve the present ground bus system and therefore will not introduce the possibility of a different accident.
(b) With respect to the possibility of a malfunction of a different type than any analyzed in the FSAR: The original installation will not be disturbed and the addition will only improve it.
(3) With respect to the margin of safety as defined in the basis for any technical specification:
(a) With the addition of another ground bus, increased integrity is achieved.
(b) With the addition of another ground bus, the consequences will be much less severe.
(c) The probability of a malfunction of the ground bus is de-creased because there will be three tie busses instead of only one.
(d) The consequences of a malfunction will not be worse because fundamentally the system has not been changed.
Units 3 and 4 "MOV-*-750 and MOV-*-751 Closure" After weeks of operation at Reactor Coolant System (RCS) operating pressure, approximately 2000 psig, the narrow range pressure transmitters, P403 and P405 (0-1000 psi), remained at saturated output when the RCS was depressur-ized during a normal cooldown. These pressure transmitters function 'to prevent opening the Residual Heat Removal valves MOV-*-750 and MOV-*-751 until RCS pressure has decreased below 465 psig. Since the pressure trans-mitters'utput was at saturation the valves could only be opened by bypassing the pressure interlock.
To correct the above situation the range of P403 and P405 was changed to 0-3000 psig. To provide narrow range indication P402 was rescaled to 0-1000 psig.
This change does not pose any unreviewed safety questions because:
(1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.
Accidents such as in FSAR Section 14.2.2, Accidental Release - Recycle or waste liquid, will be made less likely. By making the MOV-*-750 and MOV-*-751 interlock more reliable, there is less chance of over-pressurizing the Residual Heat Removal Heat Exchanger, which could result in an accidental release of radioactive liquid.
(2) The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.
e All equipment to be used in the PC/M was previously in service. Therefore, all common failure modes of the transmitter remained the same. However, this PC/M will lessen the probability of malfunctions where the transmitter "hangs up" at 1000 psig. The transmitter is now linear to 3000 psig, well above normal operating pressure.
(3) The margin of safety as defined in the basis for any technical specifi-cation has not been decreased.
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- 12. Plant Change/Modification 75-79 Units 3 and 4 "Containment Sump Level Add Redundant Level Transmitter" The sump level transmitter located in the reactor pit trench is inaccessible during operation and for most of the refueling period due to high radiation level. The transmitter occasionally fails and because it is one of the backup reactor coolant leakage detectors, a second transmitter which can easily be remotely connected was installed.
This change does not pose any unreviewed safety questions because:
(1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.
This change improves the reliability for backup Reactor Coolant System leakage detection and is functionally equivalent to the original equipment.
(2) The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.
The transmitter and circuitry are the same as the original equipment.
(3) The margin of safety as defined in the basis for any technical specifi-cation has not been decreased.
The change provides for redundancy of Reactor Coolant System leakage detection.
V 13 'lant Change/Modification 76-2 Unit 4 "Control Rod" Drive Mechanism-Head Cables/Connectors" The original CRDM cables were deteriorating due to the high temperatures at the reactor head. The cables were rated 0 C, at 90oC and operating temperatures were in the, range of 90oC to 95 The replacement cables are rated at 150 C. The new cables were attached to the existing cables near the 58 foot level curb box with Pyle National connectors.
This change did not pose any unreviewed safety questions because; (1) The probability of occurrence or the consequence of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report'as not been increased.
.Accidents such as in FSAR Section 14.1.4, RCCA drop, will be made less likely by using CRDM head cables rated for 150 C instead of 90'C.
(2) The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.
Accidents such as a misalignment in reconnecting the cables after refueling are made less likely by positive post-refueling tests.
(3) The margin of safety as defined in the basis for any technical specification has not been decreased.
The new cables will not increase the required 1.8 seconds rod drop time in Section B3.2.
Unit 4 "Integrated Leak Rate Test (ILRT)
Relative Humidity Monitoring System" The originally installed relative humidity monitoring system had many operating difficulties during the first operational ILRT (on Unit 3 during 1975) . To preclude the same problems a new type system similiar to that which was satisfactorily used at the St. Lucie Plant was installed.
This change did not pose any unreviewed safety questions because:
(1) The probability of occurrence or the consequences, of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.
Neither the consequences nor the probability of the loss of coolant accident (FSAR Section 14.3.5) will be increased.
Accident analysis assumptions of a leak rate of 0.25% per day will be better assured by using this new relative humidity equipment.
(2) The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.
The addition of the system will reduce the possibility of accidentally moisture saturating the containment during the ILRT and damaging delicate electronic instrumentation and other equipment.
(3) The margin of safety as defined in the basis for any technical specification has not been decreased.
The addition of this new system will insure the bases Section B4.4, Containment Tests, can be met. Relative humidity is a
, variable used in the ILRT Computer Program. This new system will insure this variable is available and accurate. The old system used only six detectors; the new system consists of ten detectors.
- 15. Plant Change/Modification 76-10 Unit 4 "Steam Generator Blowdown Piping Modification" The piping between the steam generator blowdown motor operated valves and the blowdown flash tank experienced excessive erosion. To remedy this, this section of the blowdown system piping was changed to stain-less steel using schedule 80 pipe and 1500 psi rated globe valves.
This change does not pose any unreviewed safety questions because:
(1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.
The change improves the reliability of the blowdown piping, making failure of the piping less likely and hence reducing the probability of an accidental release of potentially radioactive steam generator blowdown fluid to atmosphere.
(2) The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.
(3) The margin of safety as defined in the basis for any technical specification has not been decreased.
~ 16. Plant Change/Modification 76-13 Unit 4 "Retubing Condenser Water Boxes" Due to excessive pitting of many of the condenser tubes, the 4B North and 4B South waterboxes have been retubed with titanium (22 B.W.G.) and 70/30 Cu Ni (18 B.W.G.) respectively. The original tube material was aluminum brass.
Additionally, the Unit,4 Condenser foundation support has been reinforced to take increased upward thrust loads caused by the lesser weight of the titanium and copper nickel tubing.
This change does not pose any unreviewed safety questions because the new condenser tube material should be less susceptible to corrosion and hence less likely to leak and allow contamination of the steam generators with high chlorides concentrations.
~ 17. Plant Change/ModdH.catdon 76-15 Units 3 and 4 "Removal of Control Room Wall Between Nuclear Plant Supervisor's Office and Emergency Exit Hallway" The wall between the Nuclear Plant Supervisor's office and the Emergency Exit hallway was removed in order to enlarge the Nuclear Plant In addition, the Emergency Exit doorway was walled up is Supervisors'ffice.
and now part of the office.
This change does not pose any unreviewed safety questions because:
(1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.
Although the wall is a part of a Class I seismic structure, removal of it had no affect on the analysis of any accidents already in the FSAR.
(2) The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.
Relocation of the wall was such that the wall could not fall on any nuclear safety related equipment and could not cause any malfunction not previously analyzed in the FSAR.
(3) The margin of safety as defined in the basis for any technical specification has not been decreased.
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- 8. Plant Change/Modification 76-16 Units 3 ant 4 "Replace Cooling Radiators on Auxiliary Transformers" To improve corrosion resistance and heat transfer characteristics a new fin radiator was installed as a replacement type radiator for the Auxiliary Transformer.
This change did not pose any unreviewed safety questions because:,
(1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.
The loss of off-site power accident is discussed in the FSAR. This change did not increase the severity or probability of occurrence of this accident. The change improved the cooling and corrosion resistance of the radiator.
(2) The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.
The probability of damage due to fire from oil spill and damage to other safety related equipment was not altered by the change. Adequate protection from a fire protection deluge system and equipment location has already been provided.
(3) The margin of safety as defined in the basis for any technical specifi-cation has not been decreased..
The margin of safety has been increased because of the more efficient cooling of the auxiliary transformer provided.
- 19. Plant Change/Modification 76-20 Units 3 and 4 "Charging Pump Material Changes" Several material changes, as delineated below, were made to the charging pumps. These changes were recommended by the pump manufacturer, Union Pump Company.
- 1. The gland material was changed from SB-144 to SB-150 to eliminate lead bearing material from this pump end part.
- 2. The plunger base material was changed from A-479 type 304 to type 316L. Union Carbide coats the .plungers with LC-4, a chrome oxide, to produce a packing/plunger combination with acceptable life. Type 316L has a higher bond strength than the 304 to LC-4 as originally furnished.
- 3. The original gland/box equipment was furnished with eight threads per inch between gland and stuffing box. Maintenance experience has shown that the new gland/box equipment with twelve threads per inch makes assembly easier after repacking the stuffing box.
This change does not pose any unreviewed safety questions because:
(1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.
The material change was an improvement and thereby reduces the possibility of failure of these pumps and of contamination of other components which contain lead sensitive alloys.
(2) The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.
Any failure with these new materials would be similiar to failures with the original materials.'hese new materials should extend the life of the pumps.
(3) The margin of safety as defined in the basis for any technical speci-fication has not been decreased.
The margin of safety will be increased because of lower wear rates, higher strength and more adherent coatings. The possibility of adding lead to the system has been eliminated.
0 (ii) PROCEDURE CHANGES The following procedures were changed, reviewed, approved, and re-issued during the reporting period. The procedure changes are as described below and only those procedure changes constituting changes in the facility as described in the final safety analysis report are reported.
Minor or routine procedure changes not affecting procedures as described in the safety analysis report are not reported.
- 1. ADMINISTRATIVE PROCEDURE 0103. 6, REPORTABLE OCCURRENCES, CHANGE DATED FEBRUARY 3, 1976, CHANGED THIS PROCEDURE AS FOLLOWS.
A. This procedure was extensively revised to comply with change 26 to the Technical Specifications (Amendment 13 to Facility Operating License DPR-31 and Amendment 14 to Facility Operating License DPR-41) which went into effect on January 1, 1976, and effected changes in the reporting requirements.
A copy of the safety evaluation for change 26 of the Technical Specifications accompanied change 26 and had received prior approval by the Nuclear Regulatory Commission.
- 2. ADMINISTRATIVE PROCEDURE 0103.7, REPORTS REQUIRED BY TECHNICAL SPECIFICATIONS AND 10 CFR, CHANGE DATED JUNE 9, 1976, CHANGED THIS PROCEDURE AS FOLLOWS.
A. Added to Step 8.2 delination of involvement that off-site individuals or groups may play in fulfillment of the reporting requirements.
B. This procedure was revised to comply with change 26 to the Technical Specifications which went into effect on January 1, 1976, and effected changes to the reporting requirements.
A copy of the, safety evaluation for change 26 to the Technical Specification accompanied change 26 and had received prior approval by the Nuclear Regulatory Commission.
- 3. ADMINISTRATIVE PROCEDURE 0109.1, PREPARATION, REVISION, AND APPROVAL OF PROCEDURES, CHANGE DATED APRIL 16, 1976, CHANGED THIS PROCEDURE AS FOLLOWS.
A. Added reference to ANSI N18.7-1972 in Section 3.3, and throughout the procedure. I
(ii) PROCEDURE CHANGES Page 2 B. Corrected reference to applicable sections of Technical Speci-fications to comply with change 26 to the Technical Specifications.
C. Added "Administrative Provisions" under section 3.4.4; Security Procedures.
D. Added section 3.4.5; Emergency Procedures.
E. Added section 4.0; Precautions.
F. Added step 5.7 to comply with change 26 to the Technical Specifications.
G. Added reference to the following section 6.0, References.
- 1. Regulatory Guide 1.33, Quality Control Program Requirements.
- 2. USAEC Guide to the Preparation of Emergency Plans for Production and Utilization Facilities - December 1970.
- 3. 10 CFR 73.
H. Extensively revised section 8.0, Instructions to comply with change 26 to the Technical Specifications and the added references listed "G" above.
A copy of the safety evaluation for change 26 to the Technical Specification accompanied change 26 and had received prior approval by the Nuclear Regulatory Commission.
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- 4. ADMINISTRATIVE PROCEDURE 0110.4, PLANT NUCLEAR SAFETY COMMITTEE GENERAL PROCEDURES WAS REVISED ON FEBRUARY 3, 1976, AS FOLLOWS:
A. This procedure was extensively revised to implement change 26 to the Technical Specifications.
A copy of the safety evaluation for change 26 to the Technical Speci-fication accompanied change 26 and had received prior approval by the Nuclear Regulatory Commission.
- 5. OFF-NORMAL OPERATING PROCEDURE 0208.4, ANNUNCIATOR LIST PANEL B-REACTOR, CHANGE DATED APRIL 16, 1976, CHANGED THIS PROCEDURE AS FOLLOWS:
A. Inserted new window number one in vertical column two to reflect incorporation of Plant Change/Modification (PC/M) 73-60, Permissive Status Lights Addition of Audible Alarms.
B. Revised operator action required for window number five, vertical column five.
(ii) PROCEDURE CHANGES Page 3 C. Added additional operator required action as follows for window two, vertical column eight which reads "(F) if control rod malfunction, refer to O. P. 1608.1".
D. Setpoint for window number two for vertical column eight was revised from 10% to 15% difference between upper and lower detectors on any power range channel.
A safety evaluation for PC/M 73-60 was performed prior to implement-ation and a description and summary of the safety evaluation was pre-viously reported in Semi-annual Operating Report 5 on February 20, 1975, and is currently on file in Document Control at Turk'ey Point.
- 6. OFF-NORMAL OPERATING PROCEDURE 0208.9, ARKNCIATOR LIST - PANEL G-MISCELLANEOUS, CHANGE DATED APRIL 18, 1976, CHANGED THIS PROCEDURE AS FOLLOWS:
A. Revised Unit 3 set points for windows one, two, and three of vertical column number three to comply with change 27 to the Technical Speci-fications. (Amendment 15 to Facility Operating License DPR-31 and Amendment 14 to Facility Operating License DPR-41).
B. Added windows four, five, and six to vertical column four to annun-ciate condensate pump low flow for condensate pumps 4A, 4B, and 4C respectively to reflect incorporation of Plant Change/Modification (PC/M) 74-79 (Addition of a third condensate pump and individual pump recirculation lines).
C. Added windows one thru six on vertical column five to reflect incor-poration of PC/M 74-4 (Installation of Digital Data Processing System) .
D. Added windows one thru five on vertical column seven to reflect incorporation of PC/M 74-4.
E. Added windows three, four and five to annunciate condensate pump low recirculation flow for pumps A, B, and C respective'ly as a result of incoporating PC/M 74-79.
The safety evaluations for PC/Ms 74-79 and 74-4 were performed prior to implementation. A description and summary of the safety evaluations was previously reported in Semi-annual Operating Reports 5 (2/20/75) and 6 (8/28/75) and is currently on file in Document Control at Turkey Point.
A copy of the safety evaluation for change 27 to the Technical Speci-fication accompanied change 27 and had received prior approval by the Nuclear Regulatory Commission.
, (AA) PROCEDGRE CHANGES Page 4
- 7. OPERATING PROCEDURE 1502. 1, STEAM GENERATOR PLACING IN SERVICE, CHANGE DATED APRIL 16, 1976, CHANGED THIS PROCEDURE AS FOLLOWS:
A. Revised precaution number two in section 4.1 to require two of three auxiliary feedwater'umps operable for one unit operation, and three of four operable for dual unit operation to comply with change 27 of the Technical Specifications.
B. In section 7.0, added a requirement to complete the valve line-up as is specified and attached to the procedure.
A copy of the safety evaluation for change 27 to the Technical Specification accompanied change 27 and had received prior approval by the Nuclear Regulatory Commission.
- 8. OFF-NORMAL OPERATING PROCEDURE 1708.1, PART LENGTH RCC MALFUNCTION, CHANGE DATED APRIL 16, 1976, CHANGED THIS PROCEDURE AS FOLLOWS:
I A. Added steps 4.1 and 4.2 to incorporate limiting conditions delineated in sections 3.2.1.e and 3.2.6.(d) to comply with change 29 to the Technical Specifications (Amendment 17 to Facility Operating License DPR-31 and Amendment 16 to Facility Operating License DPR-41).
B. Added section 5.0 References.
C. Deleted requirement to fill 5464 throughout this procedure.
out trouble and incident report form A copy of the safety evaluation for change 29 of the Technical Speci-fication accompanied change 29 and had received prior approval by the Nuclear Regulatory Commission.
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Special Test One Unit 3 "At Power Temperature Coefficient" On January 20, 1976, a measurement of the at-power temperature coefficient.
was conducted. Its objective was to me'asure the change in reactivity due to a change in reactor coolant temperature while at hot full power. This measurement was conducted without the use of a reactivity computer.
The detailed data taken during this measurement was lost in transit to Westinghouse, so the power history from the NCCO log was used to determine burnup, while only the 'initial and final temperatures were used to determine hT In order to perform this measurement, the volume control tank level was raised to the high end and throughout the measurement no dilution or boration to the reactor was made. Control rods were stationary at 220 steps during this period while turbine load was periodically adjusted -to maintain a constant power level. During the measurement T vgand AT were monitored and the RCS boron concentration was periodically analyzed.
Theoretically, if power, rods, and boron are held constant, then T vgmust decrease with burnup.
, The reactivity loss due to burnup over the duration of the measurement is obtained from the following equation:
aCb hP (AC ) (ABU (~BU) b Where: ~~ ~ Design inverse boron worth from WCAP 8630 d,Cb
<Cb ~ Slope of measured boron letdown curve
~BU ABU ~ Burnup during period of measurement as determined from power history.
The at power temperature coefficient is then ho ATavg The above temperature coefficient is a combination of the moderator and temperature fuel effect.
Data used in determination of the temperature coefficient is presented in Table l.
The measured boron letdown curve is shown in Figure 1.
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Anal sis of'ata
~m
~hC ABU 0 ':M) /.'fZU IP l0 39 ~cm ECb ',
ppm hBU ~ 23.75 EFPH = 30.97 !AND/ZZU 3.45'F avg'..
dp ~290pcm The at power temperature coefficient is then ho gem hT ~ 8.39 'F Subtracting out the doppler contribution of 1.4 pcm/'F, the measured isothermal moderator coefficient is 6.99 pcm/'F. The design value from Figure 5.4 of klCAP 8630 is 10.9 pcm/'F for the isothermal moderator at po:ver temperature coef ficient.
Table 1
. The at-power. temnerature .coefficient measurement, was commenced. at..1103 January.. 20.
and completed at 1115, January 21.
('F)
Initial 562.42 55 Final 558.97 RCS Boron Concentration T'me (Local Tine) ( n>
January 20, 1976 1115 731 1150 728 1245 731 1345 727 1445 731 1545 727 1645 723 1745 706 1845 706 1945 731 2045 733 2145 730 2245 733 2340 731 30-
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Time (Local Tine) ~(DPD) 0 January 21, 1976 0110 720 0200 726 0400 729 0600 724.
0900 730 1100 733
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I The at-power temperature coefficient measurement does not pose any unreviewed safety questions because:
- 1. With respect to an accident previously evaluated in the FSAR: The loss of full load accident with the turbine in manual is discussed in Chapter 14 of the FSAR and concludes that the integrity of the RCS and Steam System is maintained. Therefore, the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
- 2. With respect to the possibility of an accident of a 'different type than analyzed in the FSAR: This test involves placing the turbine and rod control system in manual. All plant parameters are main-
'tained within Technical Specification Limits. Therefore. no condi-tions are set up to create the possibility of an accident other than those discussed in Chapter 14 of the FSAR.
- 3. With respect to the margin of safety as defined in the basis for any Technical Specification: This test requires that all Technical Specification Limits be maintained. Thus, the margin of safety as discussed in Technical Specifications is not decreased.
- 2. Special Test Two Unit 4 "Stationary Gripper Fuses" The Stationary Gripper Fuses in the Control Rod Drive System have been responsible for several "Dropped Rods" on Units 3 and 4. Investigation revealed that the resistance of the Shawmut fuse type A25 x 10, used in the Stationary Gripper Circuit, 'increases with time.
The resistance increases from 0.008 ohms to approximately 1.0 ohms, due to the inability to fully solder the fuse element to the end caps. A fiberglass clad material, bonded to the fuse element, extends into the end cap slot and prevents a good fuse element to end cap bond.
Westinghouse supplied 96 Bussman TRONKAX-10 fuses as a potential sub-stitute. These were installed in all 45 Stationary Gripper Circuits on Unit 4.
Prior to installation the fuses were X-rayed to verify the end cap solder connections and the resistance of the fuses was recorded and compared with the resistance at 'the end of the test.
This special test, which is also designated Special Test IGC /310, does not involve an unreviewed safety question because:
- 1. a. The probability of occurrence of an accident previously evalu-ated in the FSAR is not made more likely by the test. Accidents such as in FSAR Section'4.14, RCCA Drop, are less likely when the Bussman fuses are x-rayed to verify proper soldering.
- b. The consequences of an accident previously evaluated in the FSAR are not made more serious by the test since the fuse value is the same as the original fuse.
- c. No equipment discussed in the accident analysis of Chapter 14 of the FSAR is made more likely to malfunction by the test because the fuse has a high interupting capacity that will limit short circuit current.
- d. The consequences of malfunction of equipment important to safety previously evaluated in the FSAR are not increased by the test for the same reason mentioned above.
- 2. a. The possibility of an accident of a different type than any analyzed in the FSAR is not created since the original fuses and fuses inserted for the test'are 10 Amp.
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- b. The possibility of malfunction of a different type than any analyzed in the FSAR is not created because this test only refers to the Stationary Gripper circuit.
3~ a. The margin of safety as defined in the basis for any tech-nical specification is not decreased.
The new fuses will not increase the required 1.8 seconds drop time discussed in basis B3.2.
The test fuses were removed from the control rod drive mechanisms after the start of the Unit 4 Cycle II III refueling outage and were returned to Westinghouse for purposes of obtaining X-rays of the fuses after use.
On June 3, 1976, Westinghouse transmitted to Florida Power and Light Company, positives of the X-rays of these fuses taken before and after use. The X-rays revealed no further deterioration after use.
Westinghouse further stated that they anticipate issuance of a technical bulletin within a few months that will identify new fuse specifications and a recommended model and supplier.
Thus, the results of this special test are inconclusive at this writing.
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- 3. Special Test Three Units 3 and 4 "Capacitor Addition to Rod Control System" This test consisted of the addition of a capacitor to SM-3-408B (R28-4C) (Auto rod speed control, 5960D19). The purpose of the capacitor was to stop SM3-408B from satuzating, and thereby causing an urgent failure alarm in the CRDM. SC-3-408B is operated with high gain (~8), and at random times, due to randomr"noise", poor amplifier construction/materials and poor design, SM-3-408B saturates.
Westinghouse attempted to solve this problem by filtering the final 4 to 20 milliamp output, but the first stage amplifier still saturated.
This test filtered the amplifier itself. The capacitor was added to the feedback circuit (FLAG + 9 ms).
This test, which is also designated special test I & C II, does not pose any unreviewed safety questions because:
(1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report is not increased.
Accidents involving the rods (FSAR Sections 14.1.1, 14.1.2, and 14.1.4) are made less probable. The addition of the capacitor makes less probable the failure of SM3-408B.
(2) The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report is not created.
- a. If the capa'citor opens, the result is as if it were not in the circuit. Conditions would revert as originally designed.
- b. If the capacitor develops a short circuit, the amplifier gain goes low (rod speed would remain at minimum 8 step/
minute). Only the response of correcting a Tavg vs. Tzef deviation would be impaired. The operator would note the slow response and put the rods in manual and manually restore Tavg to Tzef.
(3) The margin of safety as defined in the basis for any technical specification is not decreased.
The addition of this capacitor does not decrease bases B 3.2-2 specifying the maximum allowable rod drop time of 1.8 seconds.
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This test is currently still in effect and results are still being monitored and are being evaluated.
Thus, the results of this test are inconclusive at this writing.
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- 4. Special Test Four Unit 4 "Reactor Coolant Pump Motor Upper Oil Pot Tubing" The 3/8" diameter copper tubing supplying oil to the reactor coolant pump upper oil pot recently experienced a puncture on reactor coolant pump 4B as a result of crimping by the tube clamps.
Special test procedure ST-76-6 was issued to accomplish removal of the softer copper tubing and replace it with more durable AISI Type 304 stainless steel tubing backed with oil resistant neoprene wedges between the tubing and the clamps for reactor coolant pump 4B.
This test does not pose any unreviewed safety questions because:
(1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report is not increased.
(2) The possibility for an accident or malfunction of a diff-erent type than any evaluated previously in the Fianl Safety Analysis Report is not created.
(3) The margin of safety as defined in the basis for any tech-nical sp'ecification is not decreased.
The test as performed on Reactor Coolant Pump 4B is considered a success and Florida Power and Light Company plans to make the installation permanent on Reactor Coolant Pump 4B, and modify the remaining Unit 3 and 4 Reactor Coolant Pumps accordingly.
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