ML20041F368
| ML20041F368 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 03/01/1982 |
| From: | CONNECTICUT YANKEE ATOMIC POWER CO. |
| To: | |
| Shared Package | |
| ML20041F363 | List: |
| References | |
| NUDOCS 8203160434 | |
| Download: ML20041F368 (69) | |
Text
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CONNECTICUT YANKEE 10CFR50.59 REPORT - 1981 8203160434 820301 PDR ADOCK 05000213 PDR R
I Summary As required by Code of Federal Regulations, Title 10, Part 50.59' entitled ~
Changes Tests and Experiments, the following report'for 1981 is' submitted.
Included in the report are the following:
1.
1981 Plant Design Changes 2.
1980 Plant Design Changes
- 3.
One Special Test conducted in 1981 4.
Annual Report of Challenges to Relief Valves
- The previously issued 1980 Plant Design Change Report omitted some design changes that were completed late in 1980. Those changes are included as an attachment to the 1981 Plant Design Change Report.
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i COMPLETED PLANT DESIGN CHANGES FOR 1981
O INDEX PLANT DESIGN CHANGES, 1981 PDCR NO.
PDC TITLE 424 Removal of Safety Injection Interlock on CH-FCV-110 and 110A 423 Auto Closure of S.G. Feedwater Isolation Valves on High Containment Pressure 421 Wide Range Gas Monitor 420 Cycle 11 Refueling 419 Installation of Flow Meters in Control Air System 418 PORV and Block Valve Logic Modification (2/3) 417 Degraded Voltage Protection Modifications 415 Steam Generator Blowdown Piping Modification 414 Condenser Retubing "B" Water box (1981) 411 Relocation of Category 1 Fuel Oil Valves and Rerouting of 1 " Fuel Oil Piping 410 TMI - Auxiliary Feed Pump Suction Bypass Valve a
409 TMI - Auxiliary Feed Pump Header Isolation Valve 407 Motor Operated Valves 567 and 569 - Control Switch Logic Modification 404 Removal of Service Water Lines,.1 " WS-21B-130 and 1 " WS-21B-262 402 Reroute RHR Control Air Line and Service Water Line to Sample Chiller 401 Safety Grade Auto-initiation of Auxilary Feedwater
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400 Station Battery Blown Fuse Indication 1
399 Connecticut Yankee Personnel Alert System 398 Power and Telecommunications Duct Line from Emergency Operations Center (EOC) to Connecticut Yankee Facility 396 Demineralized Water Storage Tank (DWST) Level Instru-mentation
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PDCR NO.
PDC TITLE 395 Resupport of Category 1 Electrical Equipment -
Part II 394 Resupport of Category 1 Electrical Equipment 1-Part'I 393 Modifications of Category 1 Building Structures 392 Redundant Fire Water Feed to Turbine Building i
387 Spent Resin Storage Facility 377 Containment High Range Radiation Monitors 372 CONVEX AGC/SCADA Remote Terminal Installation 370 TMI 2.1.3b Detection of Inadequate Core Cooling (subcooled margin monitor) 363 PORV Limit Switch Replacement 347 Reactor Coolant System Venting System 1
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PLANT DESIGN CHANGE NUMBER 424 Plant Design Change Number 424, entitled Removal of Safety Injection Interlock On CH-FCV-110 and 110A, is complete.
Description of Change Removes the Safety Injection Interlock on the charging system flow control valves (CH-FCV-110 and 110A). This will be accomplished by a
disconnecting leads for CH-FCV-110 (S-1) and 220A (S-1) at the SI WLs and MCB field terminal blocks.
i Reason for Change i
Upon receipt of a safety injection sig al, CH-FCV-110 and 110A will auto-matically open. These valves are maintained in their full open position until safety injection is reset. Thus, during safety injection, the operator is prevented from controlling charging flow by means of these valves. As a result, the operator must control charging flow by cycling the charging pumps on or off as required. This is an undesirable method for controlling flow.
This change enables the operator to control charging flow during safety injection by means of CH-FCV-110 and 110A vice cyclic operation of the charging pumps. Removal of the SI interlock is considered permissable since acceptable analytical results have been shown for LOCA and MSLB accidents without taking credit for charging pump operation.
Safety Evaluation CY is currently licensed on a Large Break LOCA ECCS evaluation docketed in 1971, and subsequently reevaluated and docketed in 1977. Neither analysis took credit for charging system flow. Even though the second charging pump starts, no water is assumed to be injected into the reactor coolant system because the hot leg injection lines are blocked, and the associated cold leg injection line is assumed to break, spilling the charging flow to containment. Thus, Large Break LOCA ECCS analyses will not be adversely affected by any changes made to the charging system.
The most recent Small Break LOCA analyses were produced as part of a post-THI cvaluation. Westinghouse issued a generic topical report (WCAP-9600, " Report on Small Break Accidents for Westinghouse NSSS System") which was reviewed by NUSCO and determined applicable to CY (NUSCO 126, " Evaluation of Westinghouse Topical Report WCAP 9600 to Determine Applicability to the Connecticut Yankee Reactor").
i In the NUSCO 126 evaluation, only section 4.2.3 is potentially impacted by PDCR - B231. Since the PDCR will now allow the operator to linearly vary charging flow, the range of break sizes addressed by NUSCO 126 Section 4.2.3 would be affected.
However, the conclusion of this section and the entire report would be unchanged.
The Steam Line Break analyses take no credit for charging system flow (any charging system flow would assist in mitigating the event).
It was thus concluded that the removal of the SI interlock on CH-FCV-110 and 110A would have no adverse affect on CY safety analyses, and was determined not to be an unreviewed safety question pursuant to 10CFR50.59.
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I PLANT DESIGN CHANGE NUMBER 423 Plant Design Change Number 423, entitled Auto Closure of SG Feedwater Isolation Valves on High Containment Pressure, is complete.
Description of Change l
This project provides for the interconnection of existing plant equipment at Connecticut Yankee to ef fect the automatic closure of the feedwater isolation valves on High Containment Pressure.
Reason for Change The impetus for the subject modification comes from a postulated main steam line break event inside containment with no loss of AC power. The single active failure assumed is a feedwater control valve failing wide open feeding the affected steam generator.
Calculated containment temperatures resulting from this scenario were undesirably high.
Safety Evaluation This design change, PDCR 423, provides for automatic closure of the main feedwater isolation valves on High Containment Pressure. This change is beneficial as it provides for safety grade isolation of the feedwater lines in the event of a steam line break inside containment, when single failure is postulated to occur in a feedwater control valve (valve failed wide open). Calculated containment pressure and temperature resulting from this scenario are undesirably high if the isolation scheme is not implemented.
The change was engineered in such a way that spurious actuation of this system would only result in the closure of 2 of the 4 MOVs. This is equivalent to a 50 percent loss of main feedwater. The consequences of this event are bound by the total loss of main feedwater event, which is part of the design basis of CY.
Based on these considerations, it was concluded that the design change did not involve an unreviewed safety question pursuant to 10CFR50.59.
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PLANT DESIGN CHANGE NUMBER 421 Plant Design Change Number 421, entitled CY - Wide-Range Gas Monitor, is complete.
Description of Change Install a General Atomics Wide Range Gas Monitor to sample the effluent through the stack during normal operatica and during accident conditions.
Two isokinetic nozzles will be installed at the stack. The sampling and computing skids will be installed in the new sampling room at the PAB. A read-out and control subpanel will be installed at the PAM Cabinet "FF" located in the control room.
Reason for Change To satisfy NUREG 0737 "TMI Action Plan Requirements", Item II.F.1, Attachment I, "Notie Gas Effluent Monitors".
Safety Evaluation This change was reviewed per the requirements of 10CFR 50.59.
The 3-inch core penetration through the floor at elevation 35'6", 3'6" west, and 12" north of column H 10k was made without reducing the capacity of the structure to meet its original design requirements provided:
1.
The penetration was drilled so that it did not cut any reinforcing.
Existing reinforcing is at 12" on center both ways top and bottom.
2.
The hole was located from below to insure the concrete spandel beam was not touched.
The change:
1.
Did not increase the probability of occurrence or the consequences of an accident previously evaluated in the safety analysis report.
2.
Did not introduce the possibility for an accident or malfunction of a dif ferent type than any evaluated previously in the safety analysis report.
3.
Did not reduce the margin of safety as defined in the basis for any technical specifications.
This change was not an unreviewed safety question.
PLANT DESIGN CHANGE NUMBER 420 Plant Design Change Number 420 entitled Cycle 11 Refueling is complete.
Description of Change The cycle 11 fuel loading was characterized as follows:
235 1.
A batch 13 feed of 52 assemblies at 4.0 wt %
U loaded on the core periphery.
2.
A mixture of 52 batch 12 once-burned assemblies at 4.00 wt %,
52 batch 11 twice-burned assemblies at 4.00 wt % and one batch 9 twice-burned assembly at 4.00 wt % as the center assembly of the core.
3.
All assemblies are stainless steel clad.
The cycle 11 fuel loading consisted of fuel from batches 9, 11, 12, and 13.
The 53 fuel assemblies discharged at EOL cycle 10 were from batches 7 and 10.
Reason for Change End of cycle 10.
Safety Evaluation PDCR 420 provided for replacement of Batch-10 fuel assemblies with fresh, Batch-13 fuel, and continued use of 52 Batch 12, 52 Batch 11, and one Batch 9 assembly.
The Cycle XI fuel nuclear and mechanical design was evaluated in BAW-1678 (June, 1981) and in REB file C2-517-278-RE.
Based on these evaluations, the Cycle XI fuel loading was determined not to be an unreviewed safety question pursuant to 10CFR50.59.
PLANT DESIGN CHANGE NUMBER 419 Plant Design Change Number 419 entitled Installation of Flow Meters in Control Air System is complete.
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Description of Change i
This change was the installation of six flow meters in the control air system.
Reason for Change In order to better define air usage in the control air system, a two part study is being conducted. The first part predicted the air usage of i
various portions of the system. based on draw!ngs and other sources of information. The second part is to measure actual air usages by portions-of the system. The installation of these flow meters permits actual flow r.easurements to be made. The flow meters are used as temporary test i
equipment. The electronics associated with them will only be hooked up a
when measurements are being taken.
l Safety Evaluation I
The installation of the six flow meters occurred in Noncategory I piping runs. Failure of these piping runs as a result of the installation of the flow meters would not affect safety equipment or cause radioactive l
gas releases..The change was not an unreviewed safety question per 10 CFR 50.59 because:
1.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FDSA was not increased, 2.
the possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report was not created, I
3.
the margin of safety as defined in the basis for any technical specification was not reduced.
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O PLANT DESIGN CHANGE NUMBER 418 Plant Design Change Number 418, entitled PORV and Block Valve Logic Modification 2/3, is complete.
Description of Change The high pressure interlocks for automatically opening the pressurizer PORVs 568 and 570, and motor operated block valves 567 and 569 were changed from a single high pressure signal to a coincident two out of three high pressure signal. This was done by replacing the Sigma meter relays formerly used for PC 401-1, PC 401-3, LIC 401-1 and LIC 401-3 with Foxboro model 63R alarm units and adding another alarm unit for PC 401-2.
The existing auxiliary relays (y type AF) were rewired and three new auxiliary relays (y type AR) were added to form the necessary two out of three logic.
The logic circuits are powered from the vital buses and designed such that loss of a vital bus will not initiate the automatic opening of the pressurizer PORVs or block valves.
Reason for Change i
Changing the pressurizer PORV and block valve opening interlock from a single signal to a coincident two out of three signal reduces the probability of a. spurious signal opening the valves.
The Foxboro model 63R alarm units were used to replace the Sigma meter relays because of the alarm units past experience of reliability and the fact that there was not any need for a meter indication for these functions.
Safety Evaluation This PDCR was reviewed with respect to the criteria con!ained in 10CFR50.59.
The change was determined to not involve an unreviewed safety question.
This evaluation was based on the following:
1.
the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report was not increased, i
2.
the possibility of an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report was not created, and 3.
the margin of safety as-defined in the basis for any technical specification was not reduced.
PLANT DESIGN CHANGE NUMBER 417 Plant Design Change Number 417, entitled Degraded Voltage Protection Modifications, is complete.
Description of Change The degraded voltage logic was modified to incorporate the new level two and level three logic. Level three logic provides an alarm calling for the operator to take steps to improve station voltage.
If level three has operated prior to a SI signal, of fsite power will be. tripped and station will shut down on the diesels. Level two also provides an alarm calling for the operator to improve the voltage but in significantly less time than level 3.
If an SI occurs accompanied at any time-by a level 2 operation, shut down will occur on the diesels.
Reason for Change The former degraded voltage scheme did not provide degraded voltage protection once emergency loads had sequenced onto the offsite supply.
Safety Evaluation This change was reviewed with respect to 10CFR50.59 and found:
1.
not to increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the SAR, 2.
not to create an accident or malfunction of a type note previously analyzed in the SAR, and 3.
not to reduce the margin of safety as defined in the basis for any technical specification.
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Pl. ANT DESIGN CHANGE NUMBER 415 Plant Design Change Number 415, entitled Steam Generator Blowdown Piping Modification, is complete.
Description of Change The scope of this project was to replace existing steam generator blowdown lines from the steam generator isolation valves to the blowdown flash tank with new piping, valves, supports, nozzles, and a flash tank wear-plate.
Reason for Change Erosion of the former steam generator blowdown lines, due to poor original design, caused the plant to experience numerous leaks in the two-phase flow portion of the piping. The blowdown piping modification improves the reliability of the system by reducing unscheduled maintenance on the blowdown lines and also minimizes any uncontrolled radioactive releases.
Safety Evaluation The blowdown system modifications; i.e., material changes and relocation of blowdown control valves, greatly increased the system integrity and significantly minimized the potential for uncontrolled releases of radioactivity. The change was not an unreviewed safety question per 10'CFR 50.59.a(2) because:
1.
The probability of occurrence-or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FDSA was not increased.
2.
The possibility for an accident or malfunction of a different type than any evaluated previously in the FDSA was not created.
i 3.
The margin of safety as defined in the basis for any technical specification was not reduced.
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PI. ANT DESIGN CHANGE NUMBER 414 Plant Design Change Number 414, entitled Condenser Retubing - B Waterbox (1981), is complete.
Description of Change Retubed "B" Waterbox in "A" Condenser. Replaced existing 18 guage Admiralty Brass tubes (6352 total) and 22 guage Stainless Steel 304 tubes (363 total) with 22 guage Trent " Sea Cure" tubes (6715 total). This was a QA Category II job.
Reason for Change Previously performed eddy current tests performed on the Admiralty Brass tubes formerly in service revealed extensive mechanical damage to the OD of the tubes. This damage resulted from an ammonia stress corrosion cracking phenomena. Approximately 8.8 percent of the total number of tubes had been plugged. Replacement of the affected tubes was necessary to maintain waterbox integrity and to preclude tube failures from occurring during normal plant operation.
Safety Evaluation This modification did not require a change to the Safety Technical Specifications and as such does not require NRC approval prior to imple-mentation.
The PDCR did not represent an unreviewed safety question in that the
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probability of a reviewed malfunction or the possibility of an unreviewed j
malfunction was not raised in implementing the change.
In addition, the margin of safety as defined in the basis for_ any Safety Technical Specifi-cation was not reduced in implementing the changes. The installation of these tubes in "B" condenser had no effect on plant safety.
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PLANT DESIGN CHANGE NUMBER 411 Plant Design Change Number 411, entitled Relocation of Cat 1 Fuel Oil.
Valves and Herouting of l\\" Fuel Oil Piping, is complete.
i Description of Change 4
I Relocation of Fuel Oil Valves FO-V-133A FO-V-133B FO-V-135 i
FO-V-136 Valves were relocated on east exterior wall of the new diesel generator building.
Fuel Oil piping will be rerouted under the floor slab from column line 12 of service building to the new valve station.
Reason for Change Increased fire protection. A source of combustible material was removed from an area adjacent to the cable spreading area.
j Increased plant safety. The possibility of a break in the fuel oil line J
was reduced by removing the valves and piping from a nonseismic masonry j
wall.
1 Safety Evaluation This modification was reviewed with respect to the requirements delineated i
by 10 CFR 50.59.
This change did not constitute an increase in the probability of a previously evaluated accident or malfunction.
In addition, the margin of safety as defined in the bases for any Safety Technical Specification was not reduced by implementing this change.
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PLANT DESIGN CHANGE NUMBER 410 Plant Design Change Number 410, entitled TMI - Aux. Feed Pump Suction Bypass Valve, is complete.
Description of Change Added manually operated valve FWV-210 and piping in parallel with valve FWV-150A on the DWST suction line..This work also included replacing a 6". tee and elbow with a 6" cross and tee, respectively. Work also included removal and reinstallation of pipe insulation (by a contractor)
. and removal / reinstallation of pipe heat tracing by CY maintenance personnel.
Mechanical and structural work was NUSCO QA Cat. I.
Also included was permanent mounting of the 6" pipe elbow spool' on the suction side of the motor driven auxiliary feed pump near the DWST enclosure wall.
Reason for Change i
A failure in valve FWV-150A would isolate the DWST from the suction side l
of the auxiliary feed pumps. Flow to the suction side of the auxiliary
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feed pumps with a failure in FWV-150A is not affected with these modifica-i tions. These modifications address concerns-described in CY letter Docket #50-213, Counsil to Eisenhunt dated December 4, 1979.
I A seismic analysis of suction piping from the DWST to the motor. driven auxiliary feed pump was performed. Design of the seismic pipe support (installed in this project) took into account loads generated both by piping from the DWST to the flanged wall penetration and piping downstream of this penetration to the ' suction side of the motor driven auxiliary feed pump, including the removable spool elbow. As a result of this analysis, the removable suction spool piece can remain connected during plant operation without compromising the system's safety.
System Evaluation This modification was reviewed with respect to the requirements delineated by 10 CFR 50.59.
This change did not constitute an increase in the probability of a previously evaluated accident and did not introduce the possibility of an unreviewed accident or malfunction.
In addition, the margin of safety, as defined in the basis for any Safety Technical 4 '
Specification, was not reduced in implementing these changes. The installation of this valve increased system reliability and lessened the possibility of a malfunction.
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PLANT DESIGN CHANGE NUMBER 409 Plant Design Change Number 409, entitled TMI Auxiliary Feed Pump Header Isolation Valve,.is complete.
Description of Change Installed a motor operated gate valve (FW-MOV-160) in the steam driven auxiliary feed pump discharge header piping along with its associated electrical power and controls required to remotely operate this valve from the control room. Modified all piping supports as required to seismically support all piping in this area to accommodatesvalve instal-lation and allow the removable piping spool piece required for motor driven auxiliary feed pump to remain installed during operation. All work and equipment associated with this modification was QA Category I.
Reason for Change 4
Installation of this isolation in the common discharge header for steam s
driven auxiliary feedwater pumps precludes a single failure from eliminating
- the ability to supply auxiliary feedwater to the steam generators.
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addition, resupporting of the piping in this area allows the removable
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piping spool piece required for the motor driven auxiliary feed pump to 4
remain installed during plant operation.
Safety Evaluation These modifications were re iewed with respect to the requirements delineated by 10 CFR 50.59.
This change did not constitute an increase in the probability of a previously evaluated accident and did not introduce the possibility of an unreviewed accident or malfunction.
In addition, the margin of safety as defined -in the basis for any Safety Technical -
Specification was not reduced in implementing these changes.
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The installation of this modification increased the reliability of the ts j
auxiliary feedwater-system. This change provides for two independent
-J flow paths to the steam generators.
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PI. ANT DESIGN CHANGE NUMBER 407 Plant Design Change Number 407, entitled MOV 567 & 569 - Control Switch and Logic, is complete.
Description of Change Replaces existing 2 position (close/open) control switch with 3 position i<
(close/ auto /open) control switch for pressurizer relief MOVs.
n c'h Modified wiring logic to:
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1.
Provide ability for manual override and closure of MOVs.
2.
Maintain or reclose HOVs when pressurizer pressure is normal an,d control switch is in "AUT0" position.
3.
Provide positive " light" indication that MOV has power and is properly lined up (including control switch in "AUT0").
Reason.for Change In the event of a spurious signal to operate,the Pressurizer Relief Valves a.id subsequent sticking of the PORVs, the former system would not j
permit convenient operator action (at the control board) for blocking the release path.
In addition, this system resets the MOVs in the event of an actual high pressure relief and a subsequent PORV sticking cuce the p
/f pressure returned to normal. This prevents an unnecessary depressurization lf/
and possible safety injection initiation.
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jafety Evaluation f
,i This PDCR was reviewed with respect to the criteria contained in 10CFR50.59.
The change was determined to not involve an unreviewed safety questian.
This evaluation was based on the following:
I' i 1.
the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in tbe safety analysis report was not increased,
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the possibility of an accident or malfunction of a different type than any evaluated previously in the safety analysis l
report was not created, 3.
the margin of safety as defined < in the basis for any technical specification was not reduced.
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c PLANT DESIGN CHANGE NUMBER 404 Plant Design Change Number 404, entitled Removal of Service Water Lines 1)" WS-21B-130 and th" WS-21B-262, is complete.
Description of Change 1
Removed service water lines li" WS-218-130, 1\\" WS-21B-262 and 3/4" WS-218-131 l
and associated valves SW-V-312, SW-V-240 and SW-V-241.
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l Lines 6" WS-151-250 and 6" WS-151-126 were capped at the tee off the i
6" service water line with 6" of Cat I carbon steel piping with a certi-fled cap.
l Reason for Change Increased the factor of safety for protection against a service water j
piping break.
Allowed erection of reinforcing steel to support masonry wall along i
column line 5' in new and spent fuel building.
Safety Evaluation This PDCR was reviewed with respect to the criteria contained in 10CFR50.59.
The change was determined to not involve an unreviewed safety question.
This evaluation was based on the following:
1.
the probability of occurrence or the consequences of an accident or malfunction of equipment.important to safety previously j
evaluated in the Safety Analysis Report was not increased, 2
the possibility of an accident or malfunction of a different I
type than any evaluated previously in the Safety Analysis
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Report was not created, 3.
the margin of safety as defined in the basis for any technical specification was not reduced.
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PLANT DESIGN CHANGE NUMBER 402 Plant Design Change Number 402, entitled Heroute of RHR Control Air Line and Service Water Line to Sample Chiller, la complete.
Description of Change Herouted I" control air line.
Line was remounted on the W12x50 steel columns erected on wall PAB 106. This line supplies the RHR system with control air.
Rerouted the service water line which supplies the sample chiller. This l\\" line was remounted on the W12x50 steel columns crected on wall PAB 106.
Retesting of both lines was required.
Reason for Change 1" control air line ran east to west along PAB 106. This line was mounted on the wall and had to be moved to erect the W12x50 columns which were required for seismic reinforcement.
I " service water line to the sample chiller ran east to west along PAB 106. This line was mounted on wall PAB 106 and had to be moved to erect the W12x50 columns which were required for seismic reinforcement.
Safety Evaluation These modifications were reviewed with respect to the requirements delineated by 10CFR50.59. These modifications were determined not to constitute an increase in the probability of a previously evaluated accident or malfunction and did not introduce the possibility of an unreviewed accident or malfunction.
In addition, the margin of safety as defined in the basis for the Technical Specifications was not reduced by implementing these changes.
O PLANT DESIGN CilANGE NUMBER 401 Plant Design Change Number 401, entitled Safety Grade Auto Initiation of Auxiliary Feedwater, is complete.
Description of Change The change results in the automatic initiation of auxiliary feedwater (AFW) whenever either of the two mode control switches is in the automatic position and either both main feedwater pump circuit breakers are open or (inclusive disjunction) there is a coincidence of any two out of four steam generator level signals in either of two redundant trains being below their setpoints. The initiating signals (auto mode) partially open tt.e two steam admission valves (PICV 1206A and B) for the redundant turbine driven AFW pumps (P32-1A and IB), and also fully open the four hypass valves (l! ICV 1301-1 thru 4) around the main feedwater regulating valves. These actions then initiate AFW flow to the steam generators.
Opening the valves is accomplished by de-energizing solenoids so as to exhaust compressed air from above the diaphragms of the air operated valves, thereby allowing the valve springs to open the valves and initiate the AFW systems.
Two new instrument transmitters were added to each of the plant's four steam generators. The two transmitters on each steam generator are powered from separate and redundant electrical power sources. The level transmitters are designated LT-1302-1A thru 4A and LT-1302-1B thru 4B.
The suffix "A" transmitters are in division or train one while the suffix "B" transmitters are in the second train. Except for the letter suffix, this nomenclature is identical to the existing wide range steam generator level instrumentation nomenclature; both sets of instrumentation are operating over the same range of steam generator levels.
The four level transmitters, in each train, are connected to a logic matrix which produces an output signal whenever any two of the four input signals are below their setpoints. The setpoints for the eight new instrument transmitters are 45 percent on the wide range scale, which corresponds to the normal lowlow level alarm point, and is 180 inches above the steam generator bottom support flange.
Output contacts from the redundant two out of three steam generator level matrices and the redundant main feed pump circuit breaker output contacts (two divisions, each with a "b" contact from each main feed pump breaker in series with a similar contact from the other main feed pump breaker) actuate two DC operated lockout relays to initiate AFW. One lockout relay is supplied from each of the redundant station batteries.
Two control switches have been located on Section G of the main control board near the existing controls for the AFW systems. These switches have two maintained positions: auto and nonauco. The two switches are redundant to each other, so that when either switch is in the auto mode, any of the previously described level or breaker position inputs will automatically initiate AFW including both pumps and all valves. Conversely, whenever both switches are in the nonauto position, AFW will not be automatically initiated regardless of SG levels or main feed pump breaker L
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positions. Therefore, when either of these auto /nonauto switches is in the nonauto position, an annunciator alarm will be present to provide a 1
status indication.
In addition, there is an annunciator alarm whenever either of the redundant AFW initiating lockout relays has tripped to initiate AFW. And, there is also an annunciator alarm whenever any one or more of the SG 1evel channels has tripped.
Auto initiation of AFW is accomplished by having each (either) of the two lockout relays deenergize six 3-way solenoid valves. Two of these cause steam to be admitted to the redundant AFW pumps and four of these valves open the bypass paths around the main feedwater regulating valves.
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Normally, when AFW isn't required, all six valves are held closed by maintaining air supply header pressure upon the valve diaphragms via 1
ports 1 and 2 of the six energized 3-way solenoid valves. Upon a demand for auto AFW, the six solenoid valves are deenergized, and in the case of the bypass valves they are allowed to go full open when port 2 of the valve is exhausted to atmosphere via port 3.
2 In the case of the steam admission valves, when the solenoids are energized (AFW initiation relays reset), the valves are held closed by PIC-1206A and B, as they were before this change, thru ports 1 (the valves) and 2 (the PIC controllers) of their respective solenoids. Upon deenergizing the solenoids, the valves are connected to port 3 of their respective solenoids. These, in turn, are connected to two new hand indicating controllers, IIICs 1206A and B, which are preset to position the steam admission valves for 450 psig of steam pressure at the terry turbine inlets given normal post trip (900 psig) steam generator conditions and an auto initiation signal. As the solenoids are deenergized, the 15-20 psig of air pressure above the diaphragms is bled off thru the HIC-1206A and B
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controllers. Thus the valves open to a satisfactory position to achieve AFW flow without popping the safety valves at the turbine inlets or 4
causing the turbines to go into an overspeed lockout condition.
2 All new circuitry was fused with 5 amp or 10 amp fuses in addition to the existing circuit breaker protection available at the distribution panelboards where these sources were derived. The 125 VDC sources were fused in both positive and negative legs and the 120 VAC vital (inverter) sources were fused in both legs also. The power supplies are as follows:
Relays or Devices Function Normal Power Supply 4AFW/A AFW Initiation Relay A DC pnl. A ekt. 7 i
4AFW/B AFW Initiation Relay B DC pnl. B ckt. 20 SOV 1301-1 Solenoid Valve Vital AC pnl. A ckt. 3 SOV 1206A Solenoid Valve Vital AC pnl. A ckt. 3 SOV 1301-2 Solenoid Valve Vital AC pnl. B ekt. 3 SOV 1301-3 Solenoid Valve Vital AC pnl. C ekt. 3 SOV 1206B Solenoid alve Vital AC pnl. C ekt. 3 SOV 1301-4 Solenoid Valve Vital AC pnl. D ckt. 3 SPEC 200 NB Electronic Process Inst.
Vital AC pnl. B ekt. 5 SPEC 200 NS Electronic Process Inst.
Vital AC pnl. C ckt. 5 i
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Reason for Change NRC requirements.
Safety Evaluation This change was reviewed with respect to requirements delineated in 10CFR50.59. The changes resulted in no increase in the probability of a previously evaluated accident, nor did they create the possibility for an unreviewed accident or malfunction. Additionally, the margin of safety as defined in the bases for the Technical Specifications was not reduced by the implementation of these changes.
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PLANT DESIGN CHANGE NUMBER 400 Plant Design Change Number 400, entitled Station Battery Blown Fuse Indication, is complete.
Description of Change Added an alarm (annunciation) actuating device to the station battery main in-line fuses.
(Both A and B banks.)
R_eason for Change e
The change provides immediate indication of a blown fuse, therefore alerting the operator who can immediately have the fuse replaced and/or cause investigated. Otherwise the operator would only be aware of the problem by abnormal voltage and current at the battery charger or when the battery charger loses its supply power.
Safety Evaluation The subject modification was reviewed against the requirements of 10CFR50.59 such that:
there has not been an increase in probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis, there was no increase in the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis; there was no decrease in the margin of safety as defined in the basis for any technical specification. Therefore, this system did not involve an unreviewed safety question.
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PLANT DESIGN CHANGE NUMBER 399 Plant Design Change Number 399, entitled CY:
Personnel Alert System, is complete.
Description of Change Installed a radio pager console in the Connecticut Yankee control room and call back recorders in the Connecticut Yankee control room viewing gallery. The work involved a seismically qualified conduit run from the control room to the PBX room; pulling a 100 pair telephone cable; penetra-tions in the control room and viewing room to connect telephone and AC power.
Reason for Change NUREG 0654 Appendix 3, required that all nuclear power plants must have a means of alerting public and technical personnel of plant accident conditions within 15 minutes after an accident. The personnel alert system meets this requirement by controlling 4 radio towers, spread throughout the system, which transmit a message to all remote pagers' units, selected groups, or individual units directly from the control room. Alerted personnel call back to receive a prerecorded message verifying alert status, thin leave a message identifying themselves and their action response to the situation.
Safety Evaluation This Plant Design Change Request was reviewed with respect to the criteria contained in 10CFR50.59. The change was determined not to involve an unreviewed safety question. This evaluation was based on the following:
1.
the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report was not increased, 2.
the possibility of an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report was been created, 3.
the margin of safety as defined in the basis for any technical specification was not reduced.
i PLANT DESIGN CHANGE NUMBER 398 l
Plant Design Change Number 398, entitled Power and Telecommunications From Emergency Operation Center (EOC) to Connecticut Yankee Facility, is complete.
Description of Change Three 4" steel conduits were adequately secured to the outer skin of the turbine building ending outside the switchgear room. Upon penetrating the outer skin, seismic supports were incorporated to secure a single 4" rigid steel conduit to cubicle #14 in the 4160V switchgear room.
Cubicle #14 furnishes power to the Emergency Operations Facility (EOF) building on the Connecticut Yankee access road.
The duct bank from the EOF to the plant consists of 12 PVC conduits with rigid steel at all areas covered with bituminous paving. Three of the conduits are utilized for telephone purposes.
Reason for Change In the initial stages of development for the EOF, an economic evaluation was conducted to determine if the building should be serviced from the local distribution or from the plant. It was determined to be more advantageous to draw power from the plant.
Load capabilities were evaluated and found to be acceptable.
Safety Evaluation 1.
The probability of an occurrence or malfunction of equipment important to safety was not increased as a result of adding a new feeder position to existing 4160V bus 1-2.
This modification involved a noncategory IE load being added to a noncategory lE bus, thus not impacting the existing plant category 1E systems, structures and equipments. As a precautionary measure, the raceway system associated with this inodification was seismically supported in those plant areas containing category IE systems.
2.
The modification did not create a new situation with regard to events previously analyzed in the FSDA, and therefore, the possibility for an accident or malfunction of a different type other than those previously evaluated was not created.
3.
The margin of safety as defined in the basis for the technical specifications was not reduced as a result of the contemplated modification which only involved noncategory lE systems.
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i PLANT DESIGN CHANGE NUMBER 396 Plant Design Change Number 396, entitled DWST Level Instrumentation, is complete.
Description of Change Installed a second level instrumentation channel for the demineralized water storage tank.
Reason for Change To provide redundant instrumentation for DWST Level as committed to in CYAPCO response to NRC Bulletins and Orders Task Force Review of the CY Auxiliary Feedwater System.
Safety Evaluation This change did not increase the probability or consequences of an j
accident or malfunction of equipment necessary for safety as previously evaluated in the FDSA.
It did not create a possibility for an accident or malfunction of a different type-than any previously evaluated in the FDSA. Neither did it reduce any safety margin as defined in the basis for any technical specifications.
Therefore, with respect to the criteria contained in 10CFR50.59, the change was not an unreviewed safety item.
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i PLANT DESIGN CHANGE NUMBER 395 Plant Design Change Number 395, entitled Resupport of Cat I Electrical Equipment - Part II, is complete.
Description of Change Modified or supplemented existing anchorage of Seismic Category I electrical equipment.
Reason for Change In a letter dated January 1, 1980, D. G. Eisenhut to W. G. Counsil, the NRC required an assessment of the seismic adequacy of the anchorage and support of safety related electrical equipment. The NRC letter also required implementation of all required modifications by September 1, 1980.
Calculations have been performed (NUSCO calculation number 79-113-166GM) to support the changes.
Safety Evaluation The modifications to the supports of the safety related electrical equipment identified in Table 1 have been evaluated and do not comprise an unreviewed safety question with respect to criteria set forth in 10CFR50.59, or require a change in the Connecticut Yankee Technical Specifications. The modificatiens are supported by Calculation GME 113-166CM and will increase the margin of safety of the equipment against effects of seismic induced loadings. As such, these changes did not increase the probability of occurrence of severity of any previously analyzed accident, did not create the potential for accidents not previously analyzed, or decrease the margin of safety of the affected systems.
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I PLANT DESIGN CHANGE NUMBER 394 Plant Design Change Number 394, entitled Resupport of Cat I Electrical Equipment Part I, is complete.
Description of Change Modified or supplemented existing anchorage of various Seismic Category I electrical equipment.
Reason for Change i
In a letter dated January 1, 1980, D. G. Eisenhut to W. G. Counsil, the i
NRC required an assessment of the seismic adequacy of the anchorage and i
support of safety related electrical equipment. The NRC letter also j~
required imp?nmentation of all required modifications by September 1, 1980. Calculations have been performed--NUSCO calculation number 79-113-t 166GM--to support the changes.
Safety Evaluation l
1 The modifications to the supports of the safety related electrical equipment identified in Table 1 have been evaluated and do not comprise an unreviewed safety question with respect to criteria set forth in 10CFR50.59, or require a change in the Connecticut Yankee Technical Specifications. The modifications are supported by calculation GME-79-113-166GM and will increase the margin of safety of the equipment against effects of seismic induced loadings. As such, these changes did not increase the probability of occurrence or severity of any previously analyzed accident, did not create the potential for accidents not previously analyzed, or decrease the margin of safety of the affected systems.
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PLANT uESIGN CHANGE NUMBER 393 Plant Design Change Number 393, entitled Modification of Category I Building Structures, is complete.
Description of Change Under the requirements of the Systematic Evaluation Program (SEP) and I&E 1
Bulletin 80-11 (masonry wall design), the building structures which are required to be seismically designed were reviewed for compliance with guidelines established for th'.s effort. Modification of existing structures, as required to meet the guidelines, was accomplished in accordance with 4
drawings reviewed and approved by NUSCO Engineering and CY Engineering.
Reason for Change NRC requirement.
Safety Evaluation The modifications to the screenwell, primary auxiliary building, turbine, and service buildings were evaluated and did not comprise an unreviewed safety question with cespect to criteria set forth in 10CFR50.59, or require a change in the Connecticut Yankee Technical Specifications. The modifications were implemented to increase the margin of safety of the structures against effects of seismic induced loadings. The changes result in structural designs consistent with acceptance criteria developed specifically for the SEP related seismic evaluation program. As such, these changes did not increase the probability of occurrence or severity of any preytcusly analyzed accident, did not create the potential for accidents not previously analyzed, or decrease the margin of safety of the affected structures.
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PLANT DESIGN CHANGE NUMBER 392 Plant Design Change Number 392, entitled Redundant Fire Water Feed to Turbine Building, is complete.
Description of Change Installation of a redundant fire water feed to the turbine building fire water manifold f rom the main underground yard loop.
Reason for Change To satisfy NRC's requirement as documented in letter of February 22, 1980, (D. L. Ziemann to W. G. Counsil).
Safety Evaluation The modification, which provides for a redundant 12" fire water supply to the Connecticut Yankee turbine building, will ensure the availability of fire water should a fire occur. The implementation of this additional fire water header increases the plant's operating margin of safety by providing system redundancy.
The system design changes have been reviewed with respect to 10CFR50.59 and determined not to constitute an unreviewed safety question. The probability of an occurrence or the consequences of an accident was not increased by these changes. The possibility of an accident not considered was not created and the margin of safety as defined in the basis of the technical specifications was not reduced.
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PLANT DESIGN CifANGE NUMBER 387 Plant Design Change Number 387, entitled Spent Resin Storage Facility, is complete.
Description of Change Constructed a reinforced concrete structure with dimensions of 22' x 29'x 10' high containing 11 cylindrical cells 5'10" in diameter, adjacent to the ion exchanger and the resin storage pit.
Each cell has a drain, a removable concrete cover (2' thick), and the sidewalls lined with stainless steel.
The structure was founded on rock or fill concrete approximately 3' below grade. Also, shield walls were provided for an additional 12' above the structure on the north and west sides. The drains were routed into the sump in the existing spent resin pit.
Drilled a 3" i diameter hole in the north wall of the existing spent resin pit to allow installation of drain line from storage cells.
Reason for Change To provide enough shielded storage for spent resin produced in one year.
Safety Evaluation The resin containers of approximately 126 cubic feet volume that could be stored in the facility are expected to contain less than one gallon of water and therefore can be considered solidified waste.
The site grade is elevation 21.5 feet, as documented in the FDSA, and provides sufficient flood protection for even the worst expected flood.
The 1936 flood of record was elevation 19.5 and even if a flood of this magnitude were to recur, the water surface elevation at the site would be only 15.1 feet, the reduction due to flood control projects that have been constructed upstream. The top of the structure is approximately elevation 29 and the bottom of each storage cell is approximately elevation 19.5 feet.
The design of the facility is essentially a reinforced mass concrete block 22' x 29' in plan with a separate cell of approximately 5 feet in diameter for each container. Each cell was be covered with a 2 foot thick reinforced concrete cap seated on a compressible sealing ring. The mass concrete and cap, in addition to providing shielding, protects against eny hypothetical missiles. The sealed cap also provides weather protection.
The mass concrete structure founded on rock is not susceptible to earth-quake damage.
Its design has been checked to verify that even the shield walls which extend to approximate elevation 41 can withstand the design basis earthquake of 0.17 g, zero period acceleration, using the spectra shape of Regulatory Guide 1.60.
There is little, if any, chance of fire since the container for solidified waste is of coated steel and sealed.
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Should any water be present in the cells, whether from the container or zhe outside, a drain from each cell to the radwaste treatment system is provided.
The storage facility was located adjacent to the existing resin storage pit and both the caps and containers are handled by the existing yard craneno change in container handling from the present procedures.
The fire protection line was modified to ensure the structure could not damage the line.
1 As summarized above, the design of the facility (1) ensured no increase in risk related to the handling and storage of the waste, (2) no change in the possibility of an accident, and (3) no change in Technical Spec-ification margin of safety.
In accordance with the requirements of 10CFR50.59, the proposed change was found not to be an unreviewed safety question.
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PLANT DESIGN CHANGE NUMBER 377 Plant Design Change Number 377, entitled Containment High Range Radiation Monitors, is complete.
Description of Change Installed two high range area radiation detectors on two steel I-beams supporting the control rod drive shield support structure in the con-tainment. The electronics / indicators were located in the post accident monitoring panel in the control room. Each detector has two coaxial cables connecting it to its associated electronics / indicator.
Reason for Change Item 2.1.8(b), of NUREG 0578 required the installation of two redundant safety grade high range area radiation monitors inside of containment.
7 The monitors must detect photon radiation up to 10 R/hr.
Safety Evaluation The subject modification has been reviewed against the requirements of 10CFR50.59 such that:
there has not been an increase in probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis; there was no increase in the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis; there has been no decrease in the margin of safety as defined in the basis for any technical specification. Therefore, this system does not involve an unreviewed safety question.
The above evaluation was based on the fact that the subject modification is independent of existing equipment such that a high range radiation monitor equipment failure will only effect the equipment in which the failure occurred. All high range radiation monitor equipment was seismically supported.
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l PLANT DESIGN CHANGE NUMiiER 372 Plant Design Change Number 372, entitled CONVEX AGC/SCADA Remote Terminal Installation, is complete.
i Description of Change i
The project entailed the addition of a data acquisition and supervisory control remote terminal unit (RTU) at Connecticut Yankee. This RTU j
automatically telemeters generator, station service, and 115-kV line NW j
and MVAR information to CONVEX. Generator and station service NWHR
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information is also telemetered to CONVEX. CONVEX also has control of the ll5-kV tie breaker (389T399) and the motor operated disconnect i
switches on lines 1206 and 1772. CONVEX control can be defeated, when j
necessary, by switches located on the substation control panel in the i
main control room.
Reason for Change The system project (PN //8013) CONVEX AGC/SCADA System has already been approved.
Initial scope calls for the installation of 60 remote terminal units in all the major NUSCO generating plants and substations.
Safety Evaluation.
This PDCR has been reviewed with respect to the criteria contained in 10CFR50.59. The change has been determined to not involve an unreviewed safety question. This evaluation was based on the following:
1.
the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report was not increased, l
2.
the possibility of an accident or malfunction of a different type than any evaluated previously in the safety analysis report was not created, l
3.
the margin of safety as defined in the basis for any technical specification was not reduced.
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PLANT DESIGN CHANGE NUMBER 370 Plant Design Change Number 370, entitled TMI 2.1.3b Det. of Inadequate Core Cooling (Subcooled Margin Monitor), is complete.
Description of Change Replaced PT403 and PT404 with post-LOCA qualified transmitters and replaced corresponding cables inside the containment.
Reason for Change To follow through with long term commitment to the NRC to upgrade the inputs to the subcooled margin monitor.
Safety Evaluation The change did not constitute an unreviewed safety question as outlined in 10CFR50.59.
1.
It did not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report.
2.
It did not create the possibility of a new type of accident not previously analyzed in the safety analysis report.
3.
It did not reduce the margin of safety that is defined in any technical specification.
The present and future proposed pressure transmitters perform no safety function and only serve to provide an input to the SSM and control board pressure indicator. The new transmitters have the capability to remain functional during and af ter a postulated LOCA and main steam line break.
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PLANT DESIGN Cl!ANGE NUtlBER 363 Plant Design Change Number 363, entitled pORV Limit Switch Replacement, is complete.
Description of Change Provided a Category IE environmentally qualified method of providing positive valve position indication for A0V 568 and A0V 570, (PORVs).
This was accomplished by replacement of the existing unqualified limit switches with environmentally qualified components, and the installation of new cable and conduit to each valve in accordance with seismic and separation criteria.
Reason for Change NRC requirement.
Safety Evaluation The modification to replace the limit switches for relief valves 568 and 570 provided a qualified positive position indication. This system has been reviewed against the requirementa of 10CFR50.59. The components used and method of installation was consistent with requirements for Category 1E equipment. There was not an increase in the probability of occurrence or the consequences of an accident or the malfunction of equipment important to safety as depicted by the Safety Analysis Report.
The possibility of an accident or malfunction of a different type than as defined in the basis for CY's Technical Specifications was not reduced or compromised. Therefore, the modification was not considered an unreviewed safety question.
PLANT DESIGN CHANGE NUMBER'347 Plant Design Change Number 347, entitled Reactor Coolant System Venting System, is complete. This system is not operational pending development of procedures and NRC approval.
Description of Change Installed the necessary valves and piping to permit remote manual venting capability of the reactor vessel and pressurizer from the control room for post LOCA noncondensible gas removal.
Reason for Change Provided the capability for removing noncondensible gases collected in the RCS in order to enhance and assure.long-term core cooling in accordance with the NRC requirements of NUREG-0578, TMI-2 Lessons Learned Task Force Status Report and Short Term Recommendations.
Safety Evaluation This Plant Design Change Request was reviewed with respect to the criteria contained in 10CFR50.59. The physical change was determined not to involve an unreviewed safety question; however, release of the system for operation is contingent upon final NRC evaluation and development of operating procedures. This evaluation was based on the following:
1.
the probability of occurrence or the consequences of an accident or malfunction of equipment important to. safety previously evaluated in the Safety Analysis Report was not increased, 2.
the possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report was not created, 3.
the margin of safety as defined in the basis for any technical specification was not reduced.
This evaluation was based on a review of the reactor coolant system design-criteria, design change calculations, and " Discussion of TMI Lesson criteria, design change calculations, and " Discussion of TMI Lessons Learned Short Term Requirements", which were used to develop the design change. Criteria for the design change has been reviewed from a mechanical engineering point of view and was judged to be appropriate.
This design change substantially increases the plant's ability to deal with large quantities of noncondensible gas without the loss of core cooling or containment integrity.
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i COMPLETED PLANT DESIGN CilANGES FOR 1980
INDEX PLANT DESIGN CHANGES, 1980 PDCR NO.
PDC TITLE 391 Refueling Water Storage Tank (RWST) Temporary Vent 388 Primary Vent Stack Spray Ring 385 Westinghouse Type W-2 Switch Neutral Position Indication Modification 384 Automatic Initiation of Auxilary Feedwater 383 Safety Injection Logic Modification 382 MW, Inputs to the Data Logger 380 RCP Component Cooling Water and Seal Water Return Isolation Modification 374 Safety System Lock Out Indication 367 Generator H2 Condition Remote Sample Modification 366 Guardhouse Air Conditioning System 359
. Containment Breathing Air System 357 Containment Electrical Penetration Replacement Phase C 356 Undervoltage Protection and Load Shedding Modification 355 New 319 Generator Step-up Transformer -
351 High Pressure Feedwater Heater Replacement 348 Containment Electrical Penetration Replacement Phase A & B Only 345 Level Alarm on Main Stack l
325 Instrument Air Supply and Low Pressure Alarm l
322 Master Cycler - Loss of Control Power Annunciator 312 Potential Transformer Installation i
309 Replacement of Loop Drain Valves i
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PLANT DESIGN CHANGE NUMBER 391 Plant Design Change Number 39;, entitled RWST Temporary Vent, is complete.
Description of Change Installed temporary vent on the RWST manway to prevent drawing a vacuum on the tank when it is pumped down at a high flow rate.
Permanent installation will consist of a hard piped overflow to the diked area and an enlarged tank top goose neck vent.
Reason for Change During a test a vacuum was drawn on the tank. This modification prevents this.
Safety Evaluation This plant design change request was reviewed with respect to the criteria contained.in 10 CFR 50.59.
The change was determined to not involve an unreviewed safety question. This evaluation was based on the following.
1.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated.in the Safety Analysis Report was not increased.
2.
The possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report was not created.
3.
The margin of safety as defined in the basis for any technical specification was not reduced.
This evaluation was based on a review of design calculation 78-79-177GM used to size the RWST temporary vent. Although the calculation specifies a minimum equivalent area of 12-inch Schedule 40 pipe, 14 inch piping and fittings were used for additional conservation.
PLANT DESIGN CilANGE NUMBER 388 Plant Design Change Number 388, entitled Primary Ventilation Stack Spray Ring, is complete.
Description of Change Installed a high pressure spray ring on top of the containment primary ventilation stack.
Reason for Change In the event of a release of radioactive material and the stack becomes contaminated, a high pressure spray ring would allow for washing down and decontaminating the interior of the stack without having to go through the expense of installing scaling ladders, a sky climber and cleaning it manually.
Safety Evaluation This PDCR has been reviewed with respect to the criteria contained in 10CFR50.59. The change has been determined to not involve an unreviewed safety question. This evaluation was based on the following:
1.
the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report was not increased, 2.
the possibility of an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report was not created, 3.
the margin of safety as defined in the basis for any technical specification was not reduced.
PLANT DESIGN CifANGE NUMBER 385 Plant Design Change Number 385, entitled Westinghouse Type W-2 Switch Neutral Position Indication Modification for Safety Related Systems, is complete.
Description of Chan g For the following list of equipment: modified the control switch (type W-2) and indicating light wiring such that the green light positively indicates only when both the control switch and its contacts are in the 4
Neutral or Automatic start position.
Containment Air Recirculation (CAR) Fans:
F-17-1, 2, 3, 4 Service Water pumps; P-37-1, 2, 3, 4 liigh Pressure Gafety Injection Pumps:
P-15-A & B 4
Low Pressure Safety Injection Pumps:
P-92-A & B 1
Charging Pumps:
P-18-A & B Emergency Diesel Generators:
E G 2A & 2B Boric Acid Pumps:
P-9-1A & IB Primary Water Transfer Pumps:
P-29-1A & IB Station Service Transformers:
485 and 496 Reason for Change i
As indicated in Westinghouse Technical Bulletin NSC-TB-80-9, investi-gations have revealed that the type W-2 switch can experience intermittent operation in the neutral (auto) position. This could prevent automatic starting of the equipment if the switch contacts are series connected to the actuation relay contacts. The change confirms the automatic contacts being in the proper position.
Safety Evaluation This PDCR has been reviewed with respect to the criteria contained in j
10CFR50.59. The change has been determined to not include an unreviewed safety question. This evaluation was based on the following:
1.
the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report was not increased, 2.
the possibility of an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report was not created, 3.
the margin of safety as defined in the basis for any technical specification was not reduced.
This PDCR complies with the recommended actions in Westinghouse Technical Bulletin NSC-TB-80-9.
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PLANT DESIGN CHANGE NUMBER 384 Plant Design Change Number 384 entitled Automatic Initiation Aux. Feedwater is complete.
Description of Change The change results in the automatic initiation of auxiliary feedwater whenever the control switches are in the automatic position and, either both main feedwater pump circuit breakers are open or (inclusive disjunction) there is a coincidence of any two of the four steam generator wide range signals being below its setpoint. The initiating signals (auto mode) partially open the two steam admission valves (PICV 1206A and B) for the redundant turbine driven auxiliary feedwater pumps (P32-1A and IB), and also fully open the four bypass valves (HICV 1301-1 through 4) around the main feedwater regulating valves. These actions then initiate AFW flow to the steam generators. Opening the valves is accomplished by deenergizing solenoids so as to exhaust compressed air from above the diaphragms of the air operated valves, thereby allowing the valve springs to open the valves and initiate the AFW systems.
The four level signals, in a logical two out of four matrix, are derived from the existing wide range steam generator level instrumentation (LT 1302-1 through 4) via bistables (LA 1302-lL through 4L) in the main control board. The wide range devices were set at 45 percent on the wide range scale and this corresponds to the normal low-low level alarm point or 180 inches distance above the steam generator bottom support flange.
It is noted that the wide range instrumentation is calibrated to be accurate for cold steam generator conditions; the error thus produced in initiating AFW with the wide range devices under hot SG conditions is thus on the conservative side, i.e.,
the hot water level which initiates AFW will, in fact, be above the 45 percent level discussed above.
Output contacts from the redundant two out of four steam generator level matrices and the redundant main feed pump circuit breaker output contacts (two divisions, each with a "B" contact from each main feed pump breaker in series with a similar contact from the other main feed pump breaker) actuate two DC operated lockout relays to initiate AFW. One lockout relay was supplied from each of the redundant station batteries.
Two control switches have been located on Section G of the main control i
board near the existing controls for the AFW systems. These switches have two maintained positions: auto and manual. The two switches are
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redundant to each other, so that when either switch is in the auto mode,
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any of the previously described level or breaker positica inputs will automatically initiate AFW including both pumps and all valves. Conversely, whenever both switches are in the manual position, AFW will not be automatically initiated regardless of SG levels or main feed pump breaker positions. Therefore, when either of these auto-manual switches is in the manual position, an annunciator alarm is present to provide a status indication.
In addition, there is an annunciator alarm whenever either of the redundant AFW initiating lockout relays has tripped to initiate AFW.
And, there is also an annunciator alarm whenever any one (or more) of the SG level channels has tripped.
Auto initiation of AFW is accomplished by having each (either) of the two lockout relays deenergize six 3-way solenoid valves..Two of these valves will admit steam to the redundant AFW pumps and four of these valves will open the bypass paths around the main feedwater regulating valves.
Normally, when AFW isn't required, all six valves are held closed by maintaining air supply header pressure upon the valve diaphrages via ports 1 and 2 of the six energized 3-way solenoid valves. Upon a demand for auto AVW the six solenoid valves are deenergized, and in the case of the bypass valves they are allowed to go full open when port 2-(HIC, air supply) is blocked off by the deenergized solenoid and port 1 (the valve) is exhausted to atmosphere via port 3.
In the case of the steam admission valves, when the solenoids are energized (AFW initiation relays reset), the valves are held closed by PIC-1206A and "B", as they were before this change, through ports 1 (the valves) and 2 (the PIC controllers) of their respective solenoids. Upon deenergizing the solenoids, the valves are connected to port 3 of their respective solenoids. These, in turn, are connected to two new hand indicating controllers, HICs 1206A and "B", which are preset to position the steam admission valves for 600 psig of steam pressure at the terry turbine inlets given normal steam generator conditions and an auto initiation signal. As the solenoids are deenergized, the 15-20 psig of air pressure above the diaphragms is bled off through the HIC-1206A and "B" controllers.
Thus the valves open to a satisfactory position to achieve AFW flow without popping the safety valves at the turbine inlets or causing the turbines to go into an overspeed lockout condition.
All new circuitry was fused with 3-amp fuses in addition to the existing circuit breaker protection available at the distribution panelboards where these sources were derived. The 125-VDC sources were fused in both positive and negative legs and the 120-VAC vital (inverter) sources were fused in both legs also. The power supplies are as follows:
Relays or Devices Function Normal Power Supply 4AFW/A AFW Initiation Relay A-DC pnl. A ckt. 7 4AFW/B AFW Initiation Relay B DC pal. B ekt. 20 SG1W Logic Relay Vital AC pnl. A ekt. 3 SOV 1301-1 Solenoid Valve Vital AC pnl. A ckt. 3 SOV 1206A Solenoid Valve Vital AC pal. A ekt. 3 SG2W Logic Relay Vital AC pnl. B ekt. 3 SOV 1301-2 Solenoid Valve Vital AC pnl. C ekt. 3 SOV 1301-3 Solenoid Valve Vital AC pal. C ekt. 3 SOV 1206B Solenoid Valve Vital AC pnl. C ekt. 3 SG3W Logic Relay Vital AC pnl. C ekt.~3 SG4W Logic Relay Vital AC pul. D ekt. 3 SOV 1301-4 Solenoid Valve Vital AC pnl. D ekt. 3 Lamp Test Circuit DC pnl. A ckt. 19 Reason for Change NRC requirement.
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Safety Evaluation The modifications, which provide automatic initiation of the auxiliary feedwater system, ensure the availability of the auxiliary feedwater i
during a loss of feedwater incident. These modifications provide for automatically starting both turbine driven AFW pumps and open all four steam generator feedwater bypass lines. The implementation of these j
initiation signals and circuits increased the plant's operating margin of safety by providing an automatically initiated system in lieu of relying solely on manual ir.itiation by an operator.
ThesesystemdIsignchangeshavebeenreviewedwithrespectto10CFR50.59 i
and det' ermined not to constitute an unreviewed safety question. The probability of an occurrence or the consequences of an accident not considered was'not created and the margin of safety as defined in the basis.af the technical specifications was not reduced.
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PLANT DESIGN CHANGE NUMBER 383 Plant Design Change Number 383, entitled CY - Safety Injection Logic Modification, is complete.
Description 'of Change
' Installed permanent modification to the safety injection logic (removal of pressurizer level coinciderice) and included means of manual Llocking and automatic / manual unblocking of safety injection.
Reason for Change To make the temporary modification permanent and provide means of blocking and automatic unblocking of safety injection.
Safety Evaluation The modification eliminates initiation of SIS on coincidence of pressurizer low level and pressurizer low pressure, and results in SIS initiation of pressurizer pressure only (1700 psig). Therefore, the probability of SIS initiation during a LOCA has been increased. The resulting mode of operation is such that:
a.
the probability of occurrence of any accident was not increased, b.
the possibility for an accident of a different nature than those previously analyzed was not created, c.
the margin to safety as defined in the basis for any Technical Specification was not reduced.
Based on the above evaluation,_the proposed design change did not involve an unreviewed safety question pursuant to 10CFR50.59.
1 PLANT DESIGN CHANGE NUMBER 382 Plant Design Change Number 382, entitled MWe Inputs to the Data Logger, is complete.
Description of Change Installed 20-ohm resistors in the current loop of the four SCADA Transducers.
Wired the voltage output into the data logger points already designated for the following inputs: Generator NWe, Station Service Transformers 309,
' 389 and 399 MWe.
Connections to the plant data logger are through States Terminal Blocks for isolation. The inputs permit the computation of gross MWe and net MWe output of CY.
i Reason for Change An accurate signal of plant MWe output was needed to provide performance data for Reactor Engineering.
i Safety Evaluation This PDCR has been reviewed with respect to the criteria contained in l
10CFR50.59. The change has been determined to not involve an unreviewed safety question. This evaluation was based on the following:
1.
the ' probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report was not increased, 2.
the possibility of an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report was not created, and 3.
the margin of safety as defined in the basis for any technical
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PLANT DESIGN CHANGE NUMitER 380 Plant Design Change Number 380, entitled RCP Component Cooling Water and Seal Water Return Isolation Modification, is complete.
Description of Change Removed the following seven valves from HCP and SIS actuation and placed on manual actuation only:
1.
Seal water return TV-334 2.
RCP thermal barrier component cooling return CC-FCV-608 3.
RCP oil cooler component cooling return CC-TV-1411 4.
Seal water return CH-MOV-311 5.
Seal water return CH-MOV-312 6.
Seal water return CH-MOV-313 7.
Seal water return CH-MOV-314.
Reason for Change Removal of the RCP auxiliaries from HCP and SIS circuits facilitates continued cperation of the RCPs as follows:
a 1.
Use during an accident l
2.
Post accident use 3
Protection of the RCPs on spurious HCP or SIS actuation 4.
Manual isolation of these systems to allow the operator sufficient time to evaluate the need for running or isolating the pumps.
Safety Evaluation This PDCR has been reviewed with respect to criteria contained in 10CFR50.59.
The proposed change was not an unreviewed safety question. That was because the continued operation of the reactor coolant pumps during an accident increases the margin of safe operation of the plant. The removal of the reactor coolant pump auxiliaries from the high containment pressure (HCP) and safety injection system (SIS) circuits to facilitate continued operation of the reactor coolant pumps during an accident, decreased the probability of an accident important to safety previously evaluated in the FDSA. This change also lowered the possibility for an accident of a different type than evaluated previously in the FDSA.
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PLANT DESIGN CHANGE NUMBER 33 Plant Design Change Number 374, entitled Safety System Lock Out Indication, is complete.
Description of Change At the equipment level, there were 18 piecee of safety-related equipment whose control scheme included a pull-to-lock feature on their respective main control board mounted hard switches. These control switches, when placed in the " lock out" position, rendered the respective piece of equipment unavailable for automatic operation. The purpose of this modification was to provide immediate and continuous indication of the lock out position of the hand switches for these equipments. This was accomplished by installation of a multiple window annunciator mounted in the control room. This indication identifies the safety system (s) impacted by the equipment which has been locked out.
Reason for Change This modification was being implemented in order to fulfill the require-ments of the commitment made to the NRC as described by W. G. Counsil's letter to D. L. Zieman dated January 3, 1979.
Safety Evaluation The modification to provide safety system equipment lock-out indication provides immediate and continuous indication of equipment locked out and the safety system impacted. This system has been reviewed in accordance with the requirements of 10CFR50.59. The components used and method of installation were consistent with the criteria set forth by Regulatory Geide 1.47 " Bypass and Inoperable Status Indication for Nuclear Plant Safety Systems" and Section 4.13 of IEEE Standard 279-1971 " Criteria for Protection Systems for Nuclear Power Generating Stations".
Hence, there was not an increase in the probability of occurrence or the consequences of an accident or malfunction of a different type than any previously evaluated, nor was the margin of safety defined in the basis for Connecticut Yankee's technical specifications been reduced or compromised. Therefore, this modification was not considered an unreviewed safety question.
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PLANT DESIGN CilANGE NUMBER 367 Plant Design Change Number 367, entitled Generator H2 Condition Monitor Remote Samplin_g Modification, is complete.
Description of Change Added wiring from remote panel in control room to generator condition monitor to allow remote sampling capability.
Reason for Change Allows the operators to start a sample immediately or automatically af ter an alarm occurs. The alarm can be verified (as per PDCR 324) and if an alarm is valid, the sample can be removed and shipped to Westinghoure for analysis.
If an invalid alarm occurs, no harm will be done to the sampling cartridge and it may be used again and again until a valid alarm is received.
Safety Evaluation This PDCR has been reviewed with respect to the criteria contained in 10CFR50.59. The change has been determined to not involve an unreviewed safety question. This evaluation was based on the following:
1.
the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report was not increased, 2.
the possibility of an accident or malfunction of a different type than any evaluated previously in the safety analysis report was not created, 3.
the margin of safety as defined in the basis for technical specification was not reduced.
PLANT DESIGN CHANGE NUMBER 366 Plant Design Change Number 366, entitled Guardhouse Air Conditioning System, is complete.
Description of Change Installed two-ton split air conditioning system to cool the console room and the guard station.
Reason for Change To provide separate and additional cooling to these areas.
Safety Evaluation This PDCR has been reviewed with respect to the criteria contained in i
10CFR50.59. The change has been determined to not involve an unreviewed safety question. This evaluation was based on the following:
1.
the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report was not increased, 2.
the possibility of an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report was not created, 3.
the margin of safety as defined in the basis for any technical specification was not reduceo.
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PLANT DESIGN CHANGE NUMBER 359 plant Design Change Number 359, entitled Containment Breathing Air System, is complete.
Description of Change This project involved the addition of provisions for supplying respirable air to workers in areas with a contaminated atmosphere, primarily in the containment. Modifications to the service air system included the following:
a.
replacement of the erirting service air compressor with two nonlubricated service air compressors, b.
addition of filtration and extension of the service air piping for expanded usage inside containment, and c.
addition of a set of check valves with a purge in between to prevent cross-contamination of the service air lines; e.g.,
primary side demineralizers for "fluf fing" of resin beds.
The air filtration units were connected to various points in the station air lines, at or near the work area. On the discharge of the filters are unique couplings for connection of service flexible breathing air hoses.
These hoses will be run to the " bubble hoods" worn by personnel.
Reason for Change Respiratory protection is necessary for personnel working in areas which have airborne contamination. The " bubble hoods" which are used with this system, have several advantages over the alternative full face mask respirators:
a.
the clear hood does not impair the wearer's field of vision, doesn't fog, and allows the use of normal corrective eyeglasses, b.
the clear hood is attached to a continuous supply of fresh respirable air eliminating the extra effort required by a conventional face mask respirator, and c.
mobility is increased because of the less restricted field of vision and the additional manifold stations.
Safety Evaluation These modifications have been reviewed with respect to the requirements delineated by 10CFR50.59. This change did not constitute an increase in the probability of a previously evaluated accident; did not introduce the possibility of a new type of accident or reduce any Technical Specification margin of safety. The mechanical aspects of this design change were not an unreviewed safety question.
PLANT DESIGN CHANGE NUMBER 357 Plant Design Change Number 357, entitled Containment Electrical Penetration Replacement Phase C, is complete.
Description of Change This PDCR was applicable to Phase C only, as described below.
Phase C consisted of the following:
Installation of new containment electrical penetrations manufactured a.
by Conax Corporation.
b.
Blanking of unused penetration nozzles.
c.
Completion of the new seismic raceway system inside containment.
d.
Installation and termination of new qualified safety related cable.
c.
Retermination of the existing cable feeding nonsafety related loads.
Reason for Change Qualified the new installation of penetrations, raceway, and associated cables to present day standards:
a.
Provides IEEE-317-1976 penetrations b.
Provides Reg. Guide 1.75 separation c.
Provides IEEE-383-1974 qualified cables.
Safety Evaluation This PDCR has been reviewed with respect to criteria contained in 10CFR50.59.
1.
The probability of an occurrence or malfunction of equipment important to safety was not significantly increased as a result of replacing
. existing containment electrical penetrations and class IE cables inside containment.
4 2.
The modification did not create a new situation with regard to events previously analyzed in the FDSA, and, therefore, the possibility for an accident or malfunction of a different type other than those previously evaluated was not created.
3.
The margin of safety as defined in the basis for the technical specifications was not reduced as a result of replacement of containment electrical penetrations and class IE cabling. Safety was improved as a result of this modification which in general meets all present day codes, regulations and standards.
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PLANT DESIGN CHANGE NUMBER 356 Plant Design Change Number 356, entitled Undervoltage Protection and Load Sheddir.g Modifications, is complete.
Description of Change Added two sets of " level two" undervoltage detection relays to each safety bus.
One set to identify a degraded voltage condition with two station service transformers in service and another set to identify a degraded voltage condition with only one station service transformer in service. Added overvoltage alarm relays to the 4160 and 480-volt buses.
Hoved the existing undervoltage tripping function from the nonsafeguards buses 1-2 and 1-3 to the safeguards buses 8 and 9.
Added a load shedding feature to strip each diesel generator of all loads (except one service water pump) if a loss-of-coolant accident occurs while the diesel generator is supplying loads subsequent to a loss-of-normal power. Many of the relays added to accomplish these changes were mounted on two new cabinets located next to the existing diesel panels 8 dbl and 9 dbl.
Reason for Change Hecent commitments to the NRC called for the installation of the undervoltage and overvoltage relays to prevent damage to any safeguards loads which j
may be operating or called upon to operate while the auxiliary bus voltages are too high or too low.
The background for these changes is provided by a letter from D. C. Switzer to A. Schwencer, dated July 21, 1977, and another letter from D. C. Switzer to D. L. Ziemann, dated November 15, 1979. The load shedding feature was added so that if a LOCA occurs while a diesel generator is carrying some auxiliary loads following a loss-of-normal power, the auxiliary loads will be shed so that required safety loads may be sequenced onto the diesel.
Safety Evaluation This PDCR has been reviewed with respect to criteria contained in 10CFR50.59.
1.
The probability of an occurrence or malfunction of equipment important to safety was not significantly increased as a result of the installation of undervoltage and load shedding circuitry. Safety was enhanced since this modification protects class IE systems and equipments from the effects of degraded voltage.
2.
The modification did not create a new situation with regard to events previously analyzed in the FDSA, and, therefore the possibility for an accident or malfunction of a different type other than those previously evaluated was not created.
3.
The margin of safety as defined in the basis for the technical specifications was not reduced as a result of the contemplated modification.
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PLANT DESIGN CHANGE NUMBER 355 Plant Design Change Number 355, entitled New 319 Generator Step-up Trans fo rme r, is complete.
Description of Change Replaced the existing (Westinghouse) generator step-up transformer (GSU) with a new (General Electric) GSU. This included removal of the existing sound wall, modifications to the deluge system, modifications of the existing foundation, a new oil sump / storage tank system, etc.
Reason for Change The Westinghouse GSU had experienced high temperatures and high gas-in-oil levels, indicating a relatively high potential for failure. This brought about the decision to purchase a spare. Since the new GSU has a lower noise level and lower electrical losses, it was decided to install it based on the fact that the installation cost will be recovered by the decrease in losses.
Safety Evaluation This PDCR has been reviewed in respect to the criteria contained in 10CFR50.59.
1.
The probability of an occurrence or malfunction of equipment important to safety was not significantly increased as a result of replacing the existing Westinghouse generator step-up transformer, a nonsafety grade equipment.
2.
The modification did not create a new situation with regard to events previously analyzed in the FDSA, and, therefore, the possibility for an accident or malfunction of a different type other than those previously evaluated was not created.
3.
The margin of safety as defined in the basis for che technical specifications was not reduced as a result of replacement of the generator step-up transformer.
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PLANT DESIGN CHANGE NUMBER 351 Plant Design Change Number 351, entitled Connecticut Yankee High Pressure Feedwater Heater Replacement, is complete.
Description of Change The change replaced Connecticut Yankee's lA and IB high pressure feedwater heaters with heaters of a new design which will achieve better system performance.
The new heaters have 439. stainless steel tubing material as a replacement of the original copper-nickel tube material in the original first point heaters.
Reason for Change NUSCO performed a detailed evaluation of Connecticut Yankee's high pressure feedwater heaters which indicated the following:
1.
heater 1A had 11 percent of its tubes plugged and heater IB had nine percent plugged, 2.
heat transfer reduction of 9.61 percent had resulted due to 11 percent of tubes plugged, 3.
projected tube failure would reach 15 percent through the middle of next core cycle (early 1981) which would result 'in a total capacity loss of 908 kW from both heaters.
Safety Evaluation This system modification replaced the first point feedwater heaters at Connecticut Yankee with new high pressure feedwater heaters of similar design. The pressure boundary was upgraded to present codes and standards which are equal to or exceed those used for the original fabrication and installation.
Based on the above, this system was not deemed to involve an unreviewed safety question as defined in 10CFR50.59.
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PLANT DESIGN CHANGE NUMBER 348 4
i This PDCR was applicable to Phase A and B, as described below, of this project only. A separate PDCR was filed for Phase C (work during the outage).
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Phase A - Consisted of the work in the cable vault which could be com-
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pleted before the refueling outage. This includes the installation of five large terminal boxes, conduit, conduit supports, new cable, cable termination in the boxes, and existing cable identification.
Phase B - Consisted of the work in the outer annulus of the containment which'could be completed before the refueling outage. This included the
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installation of four cable pull boxes at the catwalk, modification to the catwalk, installation of tray under the catwalk, conduit around the outer annulus, raceway supports and cable pulling.
Reason for Change 1
Qualified the new installation of penetrations and associar.ed cable to present' day standards:
Provides Reg. Guide 1.75 separation.
i-j Provides IEEE-383 qualified cables, i
This preoutage work minimized the impact of the electrical penetration l
replacement on the required outage down time and interference with adjacent work during the outage.
Safety Evaluation The probability of an occurrence of equipment important to safety was not significantly increased.as a result of the installation of portions of the raceway system associated with the project for replacing containment 4
electrical penetrations and Class-1E cabling inside containment.
In this respect it should be noted that the phase of the project covered by this PDCR does not include operative equipment.
r The margin of safety, as defined in the basis for the technical specifica-tions, was not reduced as a result of the modification. Ultimately, 4
j safety will be enhanced when the project is completed providing an f
installation which, in general, meets all present day codes, regulations and standards.
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PLANT DESIGN CHANGE NUMBER 345 Plant Design Change Number 345, entitled Level Alarm on, Main Stack, is complete.
Description of Change Added a sensor (s) which will alarm whenever a liquid level is reached in j
the main stack to alert plant personnel that a potential problem exists.
Heason for Change This change was being made as an aid to the operators, in order to allow early detection of liquids collecting in the base of the main stack.
Safety Evaluation This PDCR has been reviewed with respect to the criteria contained in 3
10CFR50.59. The change has been determined to not involve an unreviewed safety question. This evaluation was based on the following:
1.
the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report was not increased, 2.
the possibility of an accident or malfunction of a different type than any evaluated previously in the safety analysis report was not created,
-3.
the margin of safety as defined in the basis for any technical specification was not reduced.
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PLANT DESIGN CHANGE NUMBER 325 Plant Design Change Number 325, entitled Instrume:nt Air Supply and Lo Pressure Alarm, is complete.
Description of Change Changed the present instrument air supply to the terry turbine and MSIVs from 3/4" copper to 1" stainless steel. This pipe was pitched back to the turbine building to provide a natural drain path.
In addition, added an accumulator pressure switch which alarms on the main control board at or about 85 psig.
Reason for Change Previously there wasn't an alarm in the control room to alert the operators of a problem with low air pressure to the MSIVs.
Also, the addition of 1" stainless steel pipe added the necessary strength to minimize loss of air to the MSIVs.
Safety Evaluation The addition of stainless steel piping and a low pressure main control board alarm enhanced the quality and reliability of the control air system. This change was not an unreviewed safety question as contained in 10CFR50.59, as it did not:
a.
increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously analyzed in the Safety Analysis Report, b.
create the possibility for an accident or malfunction of a different type than previously evaluated in the Safety Analysis
- Report, c.
reduce the margin of safety as defined in the basis for any technical specification.
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e PLANT DESIGN CHANGE NUMBER 322 Plant Design Change Number 322, entitled Master Cycler - Loss of Control Power Annunciator, is complete.
Description of Change Added loss of control power auxiliary relays to the Master Cycler and pulser DC power supplies. Contacts of these relays operate a time delay alarm relay. The alarm relay inputs control room annunciator B1 point 2-3.
Additional contacts of these relays illuminate individual indicating lights on a local panel mounted in the master cycler cabinet.
Reason for Change A situation had been identified (loss of control power) where control power of control rod banks "A" and "B" could be lost without operator knowledge. This modification, although it will not prevent loss of control power, does provide for control room awareness of the situation.
Safety Evaluation The system was installed in a manner so that the electrical separation of existing redundant systems was not jeopardized. Further, it was installed in accordance with Regulatory Guide 1.75, " Physical Independence of Electric Systems". This change was not an unreviewed safety question since it did not:
a.
increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously analyzed in the Safety Analysis Report, b.
create the possibility for an accident or malfunction of a different type than previously avaluated in the Safety Analysis
- Report, c.
reduce the margin of safety as defined in the basis for any technical specification.
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o PLANT DESIGN CHANGE NUMBER 312 Plant Design Change Number 312, entitled Potential Transformer Installation, is complete.
Description of Change Installed a drawout type fused potential transformer complete with cubicle and accessories to compartment five of 4160-volt emergency diesel buses 8 and 9.
The primary high voltage windings of the potential transformers were connected through two fuses to phases one and three of their respective switchgear. The secondaries of these pts were connected to the loss of voltage and undervoltage trip initiation circuitry (1980 refueling outage). These transformers are supplied by the same manu-facturer as the existing switchgear and compatible with the existing switchgear. The new pts and their cubicles meet the same seismic specifica-tions as the existing switchgear. The PT High side fuses were not installed and thus do not expose the diesel buses to the pts until the job is completed next refueling.
Reason for Change A commitment was made to the NRC in D. C. Switzer's July 21, 1977 letter to A. Schwencer to change the loss of voltage trip initiation logic to two out of three logic, and to add undervoltage (low voltage) protection with the same logic.
In order to avoid tripping the plant due to a blown PT fuse, it is necessary that three independent pts be used to input the two out of three trip logic referenced ibove. This PDCR added the third independent PT to the two existing pts on each bus so as to meet the above requirement.
Installing pts without high side fuses permits con-siderable prework to be performed (under a future PDCR) on the total undervoltage protection modification. The system was installed during the 1980 and 1981 refueling outages.
Safety Evaluation This change has been reviewed with respect to 10CFR50.59 and found not to be an unreviewed safety question as it did not:
a.
increase the probability of occurrence or consequences of an accident, event or malfunction of equipment important to safety
- analysis, b.
create an accident or malfunction of a different type than previously evaluated in the safety analysis, c.
reduce the margin of safety as defined in the basis of any Technical Specifications.
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PLANT DESIGN CHANGE NUMBER 309 Plant Design Change Number 309, entitled Connecticut Yankee - Replacement of Drain Loop Valves, is complete.
Description of Change Replaced and relocate loop drain valves.
Reason for Change These valves are located in an area where it is difficult to reach.
By relocating these valves, it enabled Maintenance and Operations to perform their duties. The old valves had a history of leaking. The new valves reduce leakage.
Safety Evaluation The replacement of the RCS hot leg manual loop drain valves has been reviewed and did not comprise an unreviewed safety question as delineated in 10CFR50.59, or require a change in the technical specifications. No changes have been made which will effect the operation of the RUS loop drain piping as defined in the FDSA. A revised stress analysis of the modified portions of the loop drain piping has been performed to ensure that the changes do not degrade the capability of the piping to withstand all design basis loadings. These changes did not increase the probability of occurrence or the possibility of a new accident and did not decrease the margin of safety of the overall RCS loop drain system.
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RELIEF VALVE CHALLENGES Pursuant to the commitment made under Item II.K.3.3 of NUREG 0737 in the W. G. Counsil letter to D. G. Eisenhut, dated June 10, 1980, the following is the report of challenges to relief valves.
CYAPCO experienced a single occurrence of a challenge to the pressurizer power operated relief valves (PORVs) during 1981. On December 22, 1981, the reactor plant tripped on reactor coolant system (RCS) high pressure (2300 psig). This event was initiated by the failure of the control air line fitting to the i
High Level Dump Valve (HLDV) on the Heater Drains Tank (HDT). This in turn caused the HLDV on the HDT to fail open resulting in a low level in the HDT. Subsequently, Main Feed Pump "A" tripped due to low suction 1
pressure. An attempt was made to reduce main generator electrical load to compensate for the reduced feed flow. However, the mismatch between the primary and secondary systems caused a high RCS pressure reactor trip. The maximum RCS prest,ure during this event was 2305 psig.
Immediately prior to the reactor trip, the PORVs were observed, by valve position indication, to be partially open.
Because of short duration of the pressure spike, the PORVs never reached the-fully open position. The extremely short duration of time that the PORVs and blocking valves were partially open was confirmed by the absence of signals from the acoustic monitors and the lack of any noticeable temperature increase as measured by the pressurizer relief line downstream RTDs.
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SPECIAL TEST Description of Test A test was successfully conducted on October 29, 1981 to demonstrate the ability to recover from a station blockout by switching to an offsite power source as described in EOP 3.1-47, Section 6.0.
This test was conducted in accordance with Special Operating Procedure SPL 10.1-14, as reviewed and approved by the Plant Operations Review Committee and Station Superintendent on October 21, 1981.
Safety Evaluation The Station Blackout Recovery Procedure did not require a change in the plant Technical Specification because under Section 3.3, Subsection G.2, there is allowance made for operating with both residual heat removal loops operable but with the pumps deenergized for up to one hour provided:
1.
No operations were permitted that would cause dilution of the reactor coolant system boron concentration, and 2.
Core outlet temperature is maintained at least 10 F below saturation tempe ra ture.
Since in the blackout condition primary makeup is'not possible, and since the reactor coolant system temperature was less than 120'F at the beginning of the test, both of these parameters were easily met.
The test was deemed to not involve an unreviewed safety question because:
1.
it did not increase the probability of occurrence or the consequence
- of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report, 2.
it did not create a possibility for an accident or malfunction of a different type than previously evaluated in the safety analysis report, and 3.
it did not reduce the margin to safety as defined in the Technical Specification bases.
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