B16240, Annual Repts for 10CFR50.59 PORV & SV Challenges Primary Coolant Specific Activity Analysis Occupational Radiation Exposure for 1996
| ML20135E311 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 12/31/1996 |
| From: | Laplatney J CONNECTICUT YANKEE ATOMIC POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| B16240, NUDOCS 9703060425 | |
| Download: ML20135E311 (157) | |
Text
CONNECTICUT YANKEE ATOMIC POWER COMPANY HADDAM NECK PLANT 362 INJUN HOLLOW ROAD e EAST HAMPTON, CT 06424-3099 February 28,1997 j i
Docket No. 50-213 l
B16240 i Re: 10CFR60.59 U.S Nuclear Regulatory Commission Attention: Document Control Desk !
Washington, DC 20555 Haddam Neck Plant Annual Reoorts Pursuant to the provisions of 10CFR50.59 and Sections 6.9.1.4 and 6.9.1.5 of the i Haddam Neck Plant Technical Specifications, this report is submitted covering ;
operation at the Haddam Neck Plant for the period January 31,1996, to December 31, 1996.
If you have any questions, please contact Mr. G. P. van Noordennen at (860) 267-3938.
l Very truly yours, CONNECTICUT YANKEE A FOMIC POWER COMPANY l
.l4k gg) J. J. LaPlatnyh Unit Director
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Enclosure /
cc: H. J. Miller, Region l Administrator i M. B. Fairtile, NRC Project Manager, Haddam Neck Plant k W. J. Raymond, Senior Resident inspector, Haddam Neck Plant REIRS Project Manager, NRC Office Of Nuclear Regulatory Research l
9703060825 961231 PDR ADOCK 05000213 R. . .. . . . PDR
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- Docket No. 50-213 i
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1 HADDAM NECK PLANT l
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Annual Reports for 10CFR50.59 i
PORV and SV Challenges Primary Coolant Specific Activity Analysis
! Occupational Radiation Exposure 3
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~; January 1,1996 through December 31,1996
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HADDAM NECK PLANT CONTENTS j TITLE SECTION J
. Plant Design Change Records l l
Procedure Changes II l
Jumper-Lifted Lead and Jumper Bypass Changes ill Tests IV l
l Experiments V :
! FSAR Changes VI Technical Requirements Manual Changes Vil Technical Specification Bases Changes Vill j General IX Primary Coolant Specific Activity Analysis X l
- Challenges to Relief Valves XI l Occupational Radiation Exposure Xil j i
HADDAM NECK PLANT SECTION I Plant Design Change Records (PDCRs)
(Page 1 of 1) i PDCR Number lille j 1007- Upgrade of Fire Water Supply Pressure Maintenance System l
1146 R1 Addition of High Containment Pressure Trip to SW-TV-2365A
& B and the Addition of Steam Generator Blowdown i Radiation Monitor Trip to the Steam Generator Blowdown Trip Valves !
1411 R1 Containment Service Air-Breathing Air Station Removal 1479 Removal of Service Water Elbow Tap Flow Indicators for !
RHR Heat Exchangers 1510 MOV Torque Switch Bypass /MOV Torque Switch Removal 1550 R1 Radiation Monitoring System Upgrade 1553 Service Water MIC Chemical injection 1556 Permanent Installation of Diesel Generator EG-7 1575 Fuel Handling System Modification / Upgrade 1586 Upgrade the Fuel Storage Building Crane CR5-1A from 5 Tons to 6 Tons 1587 Control Room Modifications 1592 Spent Fuel Pool Rerack Project
1 PDCR 1007
- 1. PDCR Number: 1Q01
Title:
Uograde of Fire Water Supply Pressure Maintenance Svstem
- 2. Description of Change: !
This change replaced the hydropneumatic tank / pressure maintenance pump with a vertical submersible pressure maintenance pump. The change also modified the pressure control band from a range of 105psig - 125psig to 100psig - 125 psig. This change is complete.
- 3. Reason for the Change:
The existing system had been unreliable and this change improves reliability.
- 4. Safety Evaluation:
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- a. This change was safe for the following reasons:
This change has no impact on existing design basis accidents or i malfunctions of equipment and poses no threat of creating new accidents I or new malfunctions of existing safety equipment.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
This change modified the pressure maintenance system. It did not effect the ability of the Fire Protection Water System to supply water during a fire, nor did it affect any equipment or components important to safety.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT T(PE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
The change did not affect any systems or components other than the pressure maintenance system. It did not create any new malfunctions or accident possibilities.
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PDCR 1007- !
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY s TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis !
for this statement is:
l The change was limited to the pressure maintenance system. It did not i directly or indirectly affect any margins of safety. ,
- c. This cliange did require a change to the Technical Specifications. l t
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PDCR 1146 R1
- 1. PDCR Number: 1146 R1 '
Title:
Addition of High Containment Pressure Trio to SW-TV-2365A & B and the Addition of Steam Generator Blow-down Radiation Monitor Trio to the Steam Generator Blow-down Trio Valves.
- 2. Description of Change:
l This change identifies in applicable locations, that automatic trip of the I
Blowdown Trip valves occurs when a high radiation condition is experienced by both radiation channels, R-16A and R-188, at the same time. l
- 3. Reason for the Change:
This change was implemented to isolate Blowdown upon a high radiation alarm I in accordance with Technical Specifications and to isolate non-essential service water loads in the event of high containment pressure or loss of the 4160 V busses. This will prevent cavitation of the single running service water pump and ensure adequate service water cooling to the Containment Air Coolers and the Emergency Diesel Generators.
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- 4. Safety Evaluation:
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- a. This change was safe for the following reasons: )l This revision correctly documents that an automatic trip of Blowdown Trip valves occurs when a high radiation signal is detected. The installed change did not impact the performance of the modified systems and did not impact the consequences of the accidents,
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF l EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
The failure modes associated with the change can not be an initiating j event for the plant design basis accidents and as such does not increase 1 the probability of occurrence or the consequences of any previously evaluated accidents or malfunctions of equipment important to safety. l l
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A i DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE !
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PDCR 1146 R1 SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is: !
The new failure modes were not common mode failures and the failure modes result in the desired safety actions. Thus the plant response was not impacted to where it can be considered a new accident.
l This design change resulted in SW-TV-2365A & B and BD-TV-1312-1,2,3,4 failing to the safe position (closed) of the original design. Thus the probability of failure of the proposed modifications is no greater than the probability of failure of the original design.
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis +
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, The changes made did not impact the performance of the modified systems and did not impact the consequences of the accidents. Thus, the changes made do not impact the margin of safety as defined in the Technical Specifications.
- c. This change did not require a change to the Technical Specifications.
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PDCR 1411
- 1. PDCR Number: 1411
Title:
Containment Service Air Breathing Air Station Removal
- 2. Description of Change:
The Containment Service Air Breathing System originally supplied air to twelve stationary breathing air units located inside the Reactor Containment building for supplying breathing air to personnel performing maintenance activities during plant shutdowns. The modification disconnected and removed the breathing air stations. The modification permits the connection of portable breathing apparatus to the Containment Service Air Breathing System.
- 3. Reason for the Change:
l The use of portable breathing air units permits preventive maintenance of this apparatus outside the containment building, thereby reducing man-rem exposure. The modification is considered an enhancement to the system.
- 4. Safety Evaluation: l l
- a. This change was safe for the following reasons:
The Containment Service Air Breathing System is not a safety-related !
system and does not perform any safety-related or shutdown function. ;
The modification did not alter the function of the system. The modification !
did not have a safety impact on the plant or increase the risk to the public health or safety.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
This modification did not impact the safety of the Containment Service Air Breathing System or other interfacing systems. The Containment Service Air Breathing Syshm is not an accident initiator nor does it perform an accident mitigatic runction. Its failure causes no increase in the risk to the public health or safety and was determined not to be a unreviewed safety question.
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l PDCR 1411 ,
l THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A i DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for i this statement is:
The Containment Service Air Breathing System and breathing air piping remained as designed with the stationary units being removed and i connections added for portable units. It does not interact with safety-related systems during it use. In addition the breathing air piping remains isolated and deprest,urized during plant operation and is not relied upon ,
for safe shutdown or emergency functions.
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY l TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
The design function of the Containment Servics !st t"4 thing System is ;
not affected by this modification and has no irnp.4. ,n the system's margin of safety. The modification has not altered the function of the system. Maintenance personnel are still supplied with breathing air during i in-containment activities. j
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- c. This change did not require a change to the Technical Specifications.
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PDCR 1479
- 1. PDCR Number: 14ZB
Title:
Removal of Service Water Elbow Tao Flow Indicators for the RHR Heat Exchangers
- 2. Description of Change:
The Service Water System was modified by removing elbow tap flow indicators F1-1401A and B. The modification also removed all associated tubing, tubing supports and instrument valves.
- 3. Reason for the Change:
The removed flow indicators we,e considered obsolete and inaccurate and not relied upon. More accurate readings are obtained from flow element FE-250.
- 4. Safety Evaluation:
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- a. This change was safe for the following reasons: l I
The removed flow irdicators (FI-1401A & B) provided flow readings l locally. Local flow indimtion is now provided with the existing flow l element (FE-250) at the discharge header for the combined SWS flow !
from the RHR heat exchangers. Sufficient isolation valves exist such that flow through either or both RHR heat exchangers can be measured.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because: 1 THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
The SWS pipe affected by this modification is not postulated to be an initiator of any previously evaluated accident. The replaced SWS lines were evaluated to be a like-for-like replacement and therefore meet the original design requirements of the system. The portion of the SWS piping replaced with a like-for-like replacement did not introduce any new equipment malfunctions. The local flow indicators that were removed did not affect the Main Control Room panels or displays and were not associated with the Emergency Operating Procedures. This modification did not degrade the performance of any safety system, since flow measurement capabilities remain, and it did not prevent any actions assumed in the accident analysis. Normal operation of the SWS is Page 1 of 2
PDCR 1479 unaffected by this modification and the RHR heat exchangers will continue tc perform the intended design function of heat removal.
THE POSSIB!LITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE 4
SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is.
1 The elimination of the two elbow tap flow indicators, and associated l j valves, tubing and supports, removed potential failure modes or !
mechanisms. Since the modification used like-for-like piping and l
equipment for replacement and eliminated components, it did not l introduce any new failure modes or mechanisms. '
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY
! TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis i
1 for this statement is:
The SWS design function is to supply water flow to the RHR heat i
exchangers. The modification did not alter the system's ability to meet its design function, flow rate parameters, or the ability to measure this parameter.
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- c. This change did not require a change to the Technical Specifications.
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PDCR 1510
- 1. PDCR Number: 151Q
Title:
MOV Toraue Switch Byoass/MOV Toraue Switch Removal
- 2. Description of Change:
This change altered the configuration of the valve operator control logic to eliminate the use of a torque switch to backup the limit switches for all open and closed strokes on valves SW-MOV-3 and 4 and SW-MOV-837A and B.
- 3. Reason for the Change:
The change eliminated the possibility of the torque switch prematurely tripping and disabling the motor during periods of high torque requirements.
- 4. Safety Evaluation:
- a. This change was safe for the following reasons:
The change does not cause an increase in risk to the public; adversely affect the probability or consequences of previously evaluated accidents or malfunctions; create a new accident or malfunctions; and does not decrease the existing margin of safety, l
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION ;
because:
THERE IS NO INCREASE IN THE PF.0BABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN !
THE SAFETY ANALYSIS REPORT. The basis for this statement is:
The accidents applicable to the modification are the LOCA and a HELB inside containment. The function of the subject valves is to assist in the mitigation of these accidents and a failure of the valves, in itself, will not initiate either of the accidents. The modification increases the probability that the valves will function as required following the applicable accidents.
The malfunctions were evaluated and found acceptable far this I modification: leak tightness of the valves; binding, blockage, or other !
overload during rotation of the disc; inadequate torque capability of the l actuator to a fully open or close the MOV against flow; failure of the limit i switches or an improperly set limit switch which could result in an J overtorque of the actuator and/or 90 gearbox at stall; and evaluation of )
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PDCR 1510 1
I the valve weak links. The evaluation of the valve weak links indicates that the pressure boundary will not be violated in the event of a stall.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE
- SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
No new equipment was added by this modification and no existing equipment was altered in any way that has the potential of creating a 5
different type of accident or malfunction.
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY
- TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is
The removal of the torque switch provides greater assurance of valve ;
operation under accident conditions. It has been demonstrated that there l j are no credible failure modes associated with the modification that would
, affect the performance of any safety system during accident conditions. I
- c. This change did not require a change to the Technical Specifications. !
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PDCR 1550 R1
- 1. PDCR Number: 1550 R1
Title:
Radiation Monitoring Svstem Uograde
- 2. Description of Change:
This change replaced existing Radiation Monitoring System (RMS) monitors R-11 (Containment Particulate), R-14A (Main Stack), R-15 (Air Ejector), R-16A/B (Steam Generator), R-18 (Service Water Discharge), and R-22 (Waste Test l Tank Discharge) with open frame skid-mounted sample systems. Each new !
channel consists of a new detector and associated input / output piping, valves, '
pumps and local UDR controls as required. All mounting of new equipment was evaluated for seismic 2 over 1 concerns. All new equipment is non-QA.
- 3. Reason for the Change:
The equipment replacement was made to implement recommended system improvements and to correct identified operability problems.
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- 4. Safety Evaluation:
- a. This change was safe for the following reasons: '
The change does not cause an increase in risk to the public; adversely affect the probability or consequences of previously evaluated accidents or malfunctions; create a new accident or malfunctions; and does not decrease the existing margin of safety.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
1 The only accidents applicable to this modification are the Steam l Generator Tube Rupture and Radioactive Waste System Failure :
accidents. No part of this modification changed, degraded, or prevented !
any actions described or assumed in the UFSAR for the subject '
accidents. The RMS is not used to mitigate the consequences of any accident, nor is the RMS used as a fission product barrier.
The entire RMS is non-safety related and cannot initiate the failure of a safety function or safety-related component. All of the new RMS Page 1 of 2
PDCR 1550 R1 equipment has been analyzed for seismic 2 over 1 concerns and have been appropriately mounted.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
This modification did not alter the function or operation of the RMS equipment. Since operation of the equipment is not changed, no new accidents or malfunctions were created by these modifications. The modifications consisted of replacing existing equipment with equipment of equal or greater caliber.
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
This modification did not alter the methodology used to calculate and adjust alarm / trip setpoints for radioactive liquid or gaseous effluent monitoring instrumentation. The new equipment meets all applicable l regulatory requirements and assures that limits of 10CFR20 are not exceeded. As a result, the existing margin of safety is maintained,
- c. This change did not require a change to the Technical Specifications.
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PDCR 1553
- 1. PDCR Number: 1553
Title:
Service Water MIC Chemical iniection
- 2. Description of Change:
This change installed a new chemical injection system for the addition of a bio-penetrant (BULAB 8007) and bio-disperant (BULAB 7005) into the service water system.
- 3. Reason for the Change:
The change was made to reduce and inhibit tuberculation and to ensure that biological matter will be suspended in the flow stream.
- 4. Safety Evaluation:
- a. This change was safe for the following reasons:
The new chemical injection system has no impact on existing accidents or malfunctions of equipment and does not pose a threat of creating new accidents or new malfunctions of existing safety equipment.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
The new chemical injection system is connected to the service water 1 system. There is not a connection between the new chemical injection I system and the Reactor Coolant System or safety-related equipment. l Any failure or malfunction of the new chemical injection system can not initiate an accident nor can it result in a malfunction of equipment important to safety previously evaluated in the safety analysis report.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
Installation of the new chemical injection system and injection of the chemicals only impacts the service water system. Any failure or Page 1 of 2
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PDCR 1553 malfunction of the new chemical injection system can not create the possibility of a new accident or malfunction of equipment important to safety.
THE MARGIN OF SAFETY AS DEFINED IN THE BASIG FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
The installation of the chemical injection system and addition of the BULAB chemicals does not affect the flowrates supplied to safety related equipment cooled by the service water system.
- c. This change did not require a change to the Technical Specifications.
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PDCR 1555
- 1. PDCR Number: .1555
Title:
Permanent Installation of Diesel Generator EG-7
- 2. Description of Change:
This change involved permanently installing the non-QA Diesel Generator, ,
EG-7, and tying its output to the Class 1E 4911 transformer and Bus 9 through i a normally open Class 1E isolation device. The diesel generator was permanently installed in a location that affords protection from the effects of a tornado and design features were added to the diesel's enclosure to provide l further hardening.
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- 3. Reason for the Change:
The diesel generator is an additional source of on-site power to be used to support shutdown risk initiatives in the event of a tornado. The change reduces the reliance on equipment that depends on service water cooling.
Regulatory Guides recommend that, in order to assure a non-QA piece of equipment cannot impact the operability of a QA device, a Qualified Isolation Device must be installed between Bus 11 and EG-7. Therefore, it is through the installation of the isolation device, that EG-7 was permanently installed.
- 4. Safety Evaluation:
- a. This change was safe for the following reasons:
The change does not cause an increase in risk to the public; adversely affect the probability or consequences of previously evaluated accidents i or malfunctions; create a new accident or malfunctions; and does not !
decrease the existing margin of safety.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
The Qualified Isolation Device will normally be open and will be in the open position for plant responses to any design basis accident. Since there are no changes to the power source, the malfunctions evaluated for the isolation device are the same as those that exist for the Bus 11 feeder circuit.
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PDCR 1555 1 THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
The isolation device is seismically qualified which ensures that the device will not change state during a seismic event, thereby not introducing new failure modes to Bus 11. In addition the isolation device will normally be ;
open, precluding the balance of the generator's distribution circuit l affecting Bus 11 operation. l THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY !
TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis ,
for this statement is: l 1
The isolation device and the associated raceway are seismically qualified I and the switch will normally be in the open position.
- c. This change did not require a change to the Technical Specifications.
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PDCR 1575
- 1. PDCR Number: 15Z5.
Title:
Fuel Handling System Modification / Upgrade
- 2. Description of Change:
This change upgraded the fuel transfer system. The upgrade included providing a replacement winch with one with a clutch mechanism that prevents overtensioning of the cable; replacing the existing floor mounted sheave with a sheave that has a load cell; mounting a position encoder on each winch; replacing the existing motor with continuous duty single speed motor with a variable speed control drive; and installing two dead end cable connections.
- 3. Reason for the Change:
l This change was made to improve the safety, reliability and operation of the system.
- 4. Safety Evaluation:
- a. This change was safe for the following reasons:
The change does not cause an increase in risk to the public; adversely affect the probability or consequences of previously evaluated accidents or malfunctions; create a new accident or malfunctions; and does not decrease the existing margin of safety.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION 07 EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED THE SAFETY ANALYSIS REPORT. The basis for this statement is:
The modification improved the reliability, safety and operation of the fuel handling system. The system interfaces remain unchanged. The fuel handling system is not required to mitigate the consequences of any accident and does not interface with any system in a manner that could adversely affect the consequences or probability of any previously evaluated accident or malfunction. The probability of occurrence and consequences of the applicable accident, i.e., fuel handling, is unaffected by this plant modification.
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i PDCR 1575 THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN -THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
The change installed new equipment in the same location as the existing equipment and did not create any additional interfaces. All failure modes that could cause an accident - or malfunction have been identified,-
evaluated and found to not create the possibility of different type of accident or malfunction than has been previously evaluated.
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
The modified system performs all the same functions as the existing system through the use of more reliable and enhanced safety features. All postulated failures are bounded by the existing fuel handling accident.
- c. This change did not require a change to the Technical Specifications.
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PDCR-1586
- 1. PDCR Number: 153fi
Title:
Upgrading of Fuel Storage Building Crane CRS-1 A from 5 Tons to 6 IQDa
- 2. Description of Change:
The capacity of the spent fuel building crane CRS-1A was increased from 5 Tons to 6 Tons. The seismic and structural adequacy of the fuel building and support structures were evaluated for the revised loading conditions.
- 3. Reason for the Change:
This change was made to support the lifting and installation of the new high density spent fuel storage racks.
- 4. Safety Evaluation:
- a. This change was safe for the following reasons:
The adequacy of the spent fuel building, the bridge girders, runway beams, and their support bracket structures was evaluated by the crane vendor, who certified that the crane can safely carry a load of 6 tons.
Additional structural analyses were performed which showed that the change was safe. These are identified in the safety evaluation.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
The change was determined to be within the structural capability of the installed system and therefore would not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
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PDCR-1586 1 i
The change reviewed the structural adequacy of an installed system and did not make physical changes to the system and therefore, does not increase the possibility of an accident or malfunction of a different type l than any evaluated previously in the safety analysis report. l 3
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TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis i 4 for this statement is:.
While the increase in usable capacity of the crane from 5 tons to 6 tons is a change from the original design, the impact has been reviewed and meets the acceptance criteria stipulated by the applicable industry codes.
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- c. This change did not require a change to the Technical Specifications. ;
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PDCR 1587
- 1. PDCR Number: 153Z
Title:
Control Room Modifications 1
- 2. Description of Change:
r This change included remodeling of office space, carpeting and installation of a new control room lighting system. New operator's consoles, a new shift manager console and a new drawing table were added. The annunciator silence switch was relocated to the Control Operator's desks.
- 3. Reason for the Change:
The modifications provide a general refurbishment of the control room, provide ,
better viewing of the control panels, increased console and office space, '
reduced noise levels and an improved lighting system to enhance operator's ability to biologically adjust to rotational shiftwork. '
- 4. Safety Evaluation:
- a. This change was safe for the following reasons:
A human factors and ergonomics review has determined that this i modification will not have a detrimental effect on the operator's ability to :
perform their job duties. The lighting system improvements have been l shown to enhance operator awareness and response time, especially during the third shift. l I
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION l because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
This modification will not impact the probability of occurrence of previously analyzed events in a negative manner. In fact, this modification could improve the ability of the operators to recognize and respond to the symptoms of an accident.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
Page 1 of 2 l
h PDCR 1587 l
This modification includes remodeling of office space, carpeting and
- installation of a new control room lighting system. The new lighting l system and the new operator consoles will be mounted to address seismic II/I concerns. The modification does not create the possibility for a malfunction or accident of a different type than previously evaluated.
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
l This modification includes remodeling of office space, carpeting and installation of anew control room lighting system. This modification does not impact the margin of safety.
- c. This change did not require a change to the Technical Specifications.
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PDCR 1592
- 1. PDCR Number: 1592
Title:
Spent Fuel Pool Rerack Project
- 2. Description of Change:
This evaluation addresses the CY Spent Fuel Pool Cooling System and its capability to accommodate the increased storage capacity of the spent fuel pool.
- 3. Reason for the Change:
In order to maintain full core discharge capability through the CY license validity date of 2007, it is necessary to partially rerack the spent fuel pool using maximum density storage modules.
- 4. Safety Evaluation:
- a. This change was safe for the following reasons:
The SFP cooling system has not been modified by this change. The SFP cooling system has been re-evaluated for postulated malfunction and accidet scenarios. The results of these evaluations demonstrate that the proposed change is safe.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
The Spent Fuel Pool Cooling System has not been modified by this l change. The structural aspects of the Fuel Storage Building, the !
structural / material compatibility of the racks, and the ability to store fuel to ensure Kerr remains s 0.95 during all aspects of the rerack project. The results of these evaluations conclude that no Unrevawed Safety Question exists.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE :
SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for I this statement is:
Page 1 of 2
1 PDCR ;502 i
The equipment in the SFP has not been modified in any way ay this change, and as such, does not create the possibility for an accident or malfunction of a different type than previously analyzed.
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis !
for this statement is: * '
The design basis fuel handling accident has been evaluated and was determined to be unaffected by the change. Although additional fuel will be stored in the SFP, the initiation of a Fuel Handling Accident is unrelated to this proposed change. Therefore, this change does not decrease the margin of safety in the Technical Specifications. ;
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- c. The CY Rerack Project did require a change to the Technical Specifications. CYAPCO License Amendment 188 was implemented to l facilitate completion of PDCR 1592.
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HADDAM NECK PLANT SECTION ll Procedure Chanaes (Page 1 of 2) i Procedure Number Iltle ANN 4.5-39 R6 Maintaining Pressure in the Pressurizer Relief Tank AOP 3.2-12 Pressurizer Safety Spoolpiece with Valve AOP 3.2-50 Plant Operations Outside the Control Room
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CMP 8.2-94.3 Procedure on Clearing #3 CAR Fan Cooling Coil Service f
- Water Instrumentation Sensing Line Blockage
. 16"-AC-152N-20 EOP E-0 Modifications to Emergency Operating Procedure E0 to Trip the Main Feedwater Pumps for Failure of a Feedline Isolation
) Valve EOP 3.1-10 Modifications to EOP 3.1-10, Partial Loss of AC EOP ES-1.3 Modifications to EOP ES-1.3 to Stop One HPSI Pump if Two 1
Are Running NOP 2.3-4 Pressurizer Safety Spoolpiece with Valve NOP 2.4-1 Pressurizer Safety Spoolpiece with Valve NOP 2.4-5A R0 Establishing Alternate Isolated Loop Overpressure Protection 4
NOP 2.5-1 R14 Maintaining Pressure in the Pressurizer Relief Tank i NOP 2.6-1 A R4 Mode 5 or Mode 6 RCP Seal Water Supply !
l NOP 2.6-7 R8 Maintaining Pressurizer in the Pressurizer Relief Tank i
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NOP 2.6-12 Draining the RCS in Mode 5 and 6 l NOP 2.7-3 R25 Purification System Resin Replacement )
NOP 2.9.1 R32 Placing the Residual Heat Removal System in Service NOP 2.24-3 Filtered SWS and Adams Filter Operation
HADDAM NECK PLANT SECTION ll Procedure Changes (Page 2 of 2)
Procedure Number Ill!g PMP 9.1-8 Low Pressure Steam Dump Test PMP 9.1-54 Testing of 480 Volt Breakers SNM 1.4-5 In-Plant Transfer of Fuel (Excluding Refueling)
TPC 96-309 Allows Bypassing of Up Travel Limit Switch TPC 96-311 Addresses Acceptability of Moving Fuel Assemblies Under Seven Feet of Water TPC 96-320 Revised the Fuel Lifting Load Limit SPL 10.3-32 Moving Spent Fuel Assembly R45 in Spent Fuel Pool SUR 5.1-0 R30 Maintaining Pressure in the Pressurizer Relief Tank SUR 5.5-70 Crosby Main Steam Safety Valve Surveillance Testing SUR 5.7-66 R9 Local Leak Rate and Pressure Isolation of Safety Injection Recirculation P-24 Isolation Valves VP-737 On-Site Handling and Installation / Removal Procedure VP-798 Evaluation to Begin Core Offload With Leakage
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i ANN 4.5-39 R6 l
! 1. Procedure Number: ANN 4.5-39 R6 1
Title:
Maintainirgfressure in the Pressurizer Relief Tank i l
- 2. Description of Change:
l The UFSAR states incorrectly that the Pressurizer Relief Tank (PRT) is l
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maintained at a pressure of 3 psig using the nitrogen regulator. The pressure i
maintained in the PRT is typically 5_ to 13 psig, sometimes as high as 15 psig.
The procedure provides guidance on maintaining the PRT pressure. 1 1
- 3. Reason for the Change: I
)
The procedure was modified to reflect maintaining 5 to 13 psig in the PRT and i to address the differences between operating as described in this procedure and as described in the UFSAR. l
- 4. Safety Evaluation .
- a. Th;s change was safe for the following reasons: ,
l Operating the PRT in the described manner is clearly within the design !
parameters of the tank and cannot affect the operat;on of other l equipment, specifically the pressurizer safeties and the power operated !
relief valves. l
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION ;
because: '
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE i OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN l THE SAFETY ANALYSIS REPORT. The basis for this statement is: !
Initial tank pressure cannot have any effect on the initiation of any accident. Maintaining the tank pressure as high as 15 psig will allow the post transient pressure to remain well below the tank's design pressure.
Since the tank has no safety function, this pressure can have no effect on ,
the probability of occurrence or the consequences of a malfunction of equipment important to safety.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A I
DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
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ANN 4.5-39 R6 :
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- Operating the PRT in the described manner is clearly within the design i
parameters of the tank and cannot affect the operation of any other equipment, specifically the pressurizer safeties and pressure operated relief valves.
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY i TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis !
for this statement is:
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Operation of the PRT is not addressed in the Technical Specification or in their bases.
i c. This change did not require a change to the Technical Specifications.
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l AOP 3.2-12
- 1. Procedure Number: AOP 3.2-12
Title:
Pressurizer Safety Soooloiece with Valve
- 2. Description of Change:
This evaluation addresses the use of a substitute valve for one of the i pressurizer code safety valves under certain conditions.
I' 3. Reason for the Change:
- This valve will allow RCS pressurization during a loss of decay heat removal to permit natural circulation cooling using steam generators.
1-l 4. Safety Evaluation:
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- a. This change was safe for the following reasons
- This valve has been evaluated with respect to its ability to provide the i required RCS vent path.
l lt has also been evaluated with respect to its ability to allow RCS pressurization during a loss of RHR.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN-THE SAFETY ANALYSIS REPORT. The basis for this statement is:
The maximum pressure the valve can see during its operation is 425 psig and it has been tested to 638 psig shell and 468 psig seat. Therefore, !
there is high confidence in the integrity of the valve and it does not !
increase the probability of occurrence or the consequences of a previously evaluated accident or malfunction. !
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A !
DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is: 1 This valve is a simple manual valve. It has no other failure mechanism I than has been previously evaluated.
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AOP 3.2-12 THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
The conditions specified for use of this valve ensures that the ability to prevent overpressurization is intact. The valve provides less flow resistance than one train of LTOP and therefore does not reduce the margin of safet'/.
- c. This change did not require a change to the Technical Specifications.
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l AOP 3.2-50 !
- 1. Procedure Number: AOP 3.2-50
Title:
Plant Ooerations Outside the Control Room
- 2. Description of Change:
Add a step to the procedure to Isolate the fire protection header to containment spray cross connect valve, RH-MOV-31, during an Appendix R shutdown under AOP 3.2-50.
- 3. Reason for the Change:
Prevent inadvertent initiation of containment spray via the fire protection headers due to spurious opening of RH-MOV-31 during an Appendix R shutdown.
- 4. Safety Evaluation:
- a. This change was safe for the following reasons:
This change prevents the inadvertent initiation of the Containment Spray via the fire protection system and adds to plant safety.
The change has no impact on existing design basis accidents.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because: )
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE !
OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is. I The change isolates RH-MOV-31 by shutting FP-V-119 and FP-V-136 l during an Appendix R shutdown. No other piping systems or branches i are affected. The initial conditions for an Appendix R shutdown are ;
bounded by existing accident analyses.
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THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
Page 1 of 2
AOP 3.2-50 The change does not affect any systems or components other than RH-
- MOV-31. It does not create the possibility any new accident or
, malfunction.
- I THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY ,
TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis i for this statement is:
- 1 Postulated accidents under which this valve would be used are bounded 1 by existing accident analyses and therefore the margin of safety is not ;
} reduced. '
i c. This change did not require a change to the Technical Specifications.
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CMP 8.2-94.3 ;
l 1. Procedure Number: CMP 8.2-94.3 i
Title:
Procedure on Clearing #3 CAR Fan Cooling Coil Service Water ,
Instrumentation Sensing Line Blockage l 5
- 2. Description of Change: ;
This change is to allow cleaning the #3 Containment Air Recirculation (CAR) f Fan Cooling Coil Service Water differential pressure instrument and tubing. ;
- . The activity involves a' temporary breach of containment integrity in that a small
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portion of the service water system inside containment will be open. !
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- 3. Reason for the Change: i l
- Over time the instrument lines build up silt and require cleaning.
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- 4. Safety Evaluation. 'i
- a. This change is safe for the following reasons:
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l The activity is conducted within the bounds of the action statement, j
Technical Specification 3.6.1.1.
A dedicated operator will be stationed at the Service Water outboard containment isolation valve in communication with the control room, should it become necessary to isolate the penetration.
A procedure prerequisite was provided specifying that no safety related equipment be taken out of service or high risk surveillance be performed during the evolution without a probabilistic evaluation.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is: j l
The system is not connected to the reactor coolant system and the change maintains positive controls over the activity within the bounds of Technical Specification action statement 3.6.1.1.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE Page 1 of 2
CMP 8.2-94.3 SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
There are no changes to equipment or to the containment boundary.
Also, integrity will be reestablished within the action statement time.
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
The activity was conducted within the bounds of the Technical Specifications and compensatory actions were in place to re-establish containment integrity in the event of an accident.
- c. This change did not require a change to the Technical Specifications.
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i CMP 8.2-94.3
- 1. Procedure Number: CMP 8.2-94.3
{
Title:
Contingency Plan Action
- 2. Description of Change:
This safety evaluation addresses the acceptability, from a containment isolation,10CFR50 Appendix A and J standpoint, of reconfiguring the Closed System Containment Penetration #53 boundary in the event that the original configuration boundary cannot be restored within one hour.
- 3. Reason for the Change:
This evaluation is part of a contingency plan to define a reconfigured containment boundary that re-establishes containment integrity of the Service Water Closed System piping within the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Technical Specification action statement.
- 4. Safety Evaluation:
- a. This change was safe for the following reasons:
This contingency plan was reviewed against the requirements of 10CFR50 Appendix A and J and found to be acceptable.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
Containment integrity for the instrument line is re-established within one hour consistent with the Technical Specifications.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVICUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
This contingency plan provides for the re-establishment of the containment integrity for an instrument line and does not create the possibility of an accident or malfunction different than any evaluated previously.
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CMP 8.2-94.3 THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY i l
TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
Containment integrity will be maintained using leak tight, passive j components and the change complies with the intent of 10CFR50 !
Appendix A and J design and test requirements.
- c. This change did not require a change to the Technical Specifications.
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i j CMP 8.5-41
- 1. Procedure Number: CMP 8.5-41 k
Title:
Freeze Sealing of Pining Systems / Freeze Seal of 16"-AC-152N-20
- 2. Description of Change:
This evaluation addressed the effect of a postulated failure of Component i
Cooling Water pipe 16"-AC-152N-20 during defueled liquid nitrogen freeze sealing of the pipe for removal, testing and replacement of CC-RV-763A and j CC-RV-763B, RHR heat exchanger outlet relief valves.
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- 3. Reason for the Change:
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- The freeze sealing of the pipe is necessary to remove, test and replace the i
affected valves. Both a' technical evaluation and a safety evaluation were performed to address the affects ..? the freeze seal on the pipe and the affect of ;
i a pipe failure.
- 4. Safety Evaluation:
I a. This change was safe for the following reasons:
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- The change has no impact on existing design basis accidents nor j malfunctions of equipment and does not create new accidents or new 1
malfunctions of exist ing safety equipment.
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- b. This change does not constitute an UNREVIEWED SAFETY QUESTION ,
! because: !
! l l THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE !
, OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF j EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN j~ THE SAFETY ANALYSIS REPORT. The basis for this statement is- )
! i The postulated pipe failure considered in the proposed change could
! adversely impact the CCR and RHR systems, however, neither of these
! systems will be required to be operable when the freeze seal is applied.
l The plant will be in a defueled condition. Therefore, there is no increase i in the probability of occurrence or in the consequences of any previously evaluated accident or malfunction.
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- THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A t DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE i SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for j i
this statement is:
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CMP 8.5-41 ,
The postulated _ pipe failure would only affect the RHR and CCR systems, neither of which will be in operation or required to be operable when the freeze seal is applied. Therefore, the possibility of a new accident or malfunction will not be created.
i-THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BELN REDUCED. The basis ,
for this statement is:
}i neither of which will be in operation or required to be operable when the freeze seal is applied. Any margins of safety that are affected by this change applies to equipment not required to be operable during the 7
maintenance of the freeze seal. ,
, c. This change did not require a change to the Technical Specifications.
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EO P E-0
- 1. Procedure Number: EOP E-0
Title:
Modifications to Emergency Ooerating Procedure E-0 to Trio the Main Feedwater Pumos for Failure of a Feedline isolation Valy_e
- 2. Description of Change:
The change modified Emergency Operating Procedure, E-0, Reactor Trip or ,
Safety injection. The change provided instructions to the operator to trip the l feedwater pumps and close their discharge MOVs if after a safety injection I signal a feedline isolation valve failed to close and the associated steam i generator pressure was less than 800 psig. l
- 3. Reason for the Change:
The design basis steam line break analyses credit the closure of the feedwater regulating valve if the feedline isolation valve fails to close. Engineering '
analysis concluded that the feedwater regulating valve may not close against a high differential pressure. In the event that a feedline isolation valve failed to close, instructions need to be provided to trip the pumps and close their MOVs.
- 4. Safety Evaluation:
- a. This change was safe for the following reasons:
The operator actions required were proceduralized and the dedicated operator had no concurrent duties. The tripping of the main feedwater pumps and closing of their discharge MOVs for a steam line break with failure of the associated feedline isolation valve ensures that the design basis assumptions remain valid.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
This change modified the procedure that is used after safety injection actuation and therefore could not cause an accident The tripping of the pumps and closing of their discharge MOVs provides assurance that feedwater flow is terminated. Therefore the probability of occurrence or the consequences of a previously evaluated accident will not occur.
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EOP E-0 THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for i this statement is:
The revised procedural actions will not create the possibility for a different accident or malfunction.
i THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY l TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
The change used a dedicated operator to assure that the design basis i
analyses are valid. The containment temperature and pressure are not impacted by the change. Therefore, the existing margin of safety has.
been maintained.
- . c. This change did not require a change to the Technical Specifications.
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EOP 3.1-10 {
L 1. Procedure Number: EOP 3.1-10 i !
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Title:
Modifications to EOP 3.1-10. Partial Loss of AC
- 2. - Description of Change:
- The change revised the Partial Loss of AC procedure. The changes improve the guidance provided to the operators for restoring power to spent fuel pool cooling.
- 3. - Reason for the Change:
l For a full core offload, the minimum time to spent fuel pool boiling, following j loss of cooling, is seven hours. The modification provides guidance for a more 3
timely; use of B electrical train to power the B spent fuel pool cooling pump.
This restoration has been validated and can be performed within seven hours.
1 l 4. Safety Evaluation:
- a. This change was safe for the following reasons:
i i The procedure provides appropriate guidance for a partial loss of AC.
i The procedure changes provide greater assurance that spent fuel pool cooling will be restored within seven hours. The changes also' provide
- greater assurance that a diesel generator will not be damaged during j attempts to start it from the control room.
h b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
! THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE i OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF t
EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN
, THE SAFETY ANALYSIS REPORT. The basis for this statement is:
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The procedure is used for a partial loss of AC. The cross-tying of electrical trains will only be performed in Modes 5 and below. this means that it can not impact accidents in Modes 1 through 4. The changes do i not affect fuel handling equipment nor equipment that can cause boron dilution. The changes provide additional guidance to assure either
- successful diesel generator operation or timely diesel generator shutdown. The changes use equipment within their design capabilities.
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l THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A i DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE
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EOP 3.1-10 SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
The procedure uses equipment with their capabilities. No hardware modifications were made. i THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
The EOP changes enhance the guidance provided to the operators for restoring spent fuel pool cooling for a partial loss of AC. Based on the validation, the restoration of spent fuel pool boiling. This means that the margin of safety is not adversely impacted.
- c. This change did not require a change to the Technical Specifications.
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Title:
Modifications to EOP ES-1.3 to Stoo One HPSI Pumo if Two Are Running
- 2. Description of Change:
Emergency Operating Procedure ES-1.3, Transfer to Sump Recirculation, was modified to stop one HPSI pump if two are running at the start of the switchover to sump recirculation.
- 3. Reason for the Change:
- The operators must perform the switchover to sump recirculation prior to the RWST inventory dropping below the level that ensures successful operation of the HPSI pump (s). The change was made to reduce the drain down rate during the switchover to sump recirculation. This provides additional time to the operators to complete the necessary actions.
- 4. Safety Evaluation:
- a. This change was safe for the following reasons:
The change provides additional time for the operator to perform the required actions during transfer to sump recirculation while maintaining adequate ECCS flow for LOCA mitigation.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
The procedure is used after a LOCA occurs. It can not cause a LOCA.
The procedure change provided additional assurance that the transfer to i sump recirculation can be performed prior to the RWST level dropping to the point where damage to the HPSI pump (s) may occur. The change does not reduce injection flow below that required for LOCA mitigation.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A {
DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
Page 1 of 2
l There was no change to any plant hardware. The procedure change improved the ability of the operator to mitigate all design basis LOCAs while meeting the applicable acceptance criteria. The change does not ,
alter the mitigation strategy. j l
THE MARGIN OF SAFEW AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
The revised procedure provides assurance that there is adequate core cooling flow for LOCA mitigation while increasing the time that the operators have in ES-1.3 to establish sump recirculation prior to damaging i the HPSI pump (s). j
- c. This change did not require a change to the Technical Specifications.
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i NOP 2.3-4 l
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- 1. Procedure Number: NOP 2.3-4 l
Title:
Pressurizer Safety Soooloiece with Valve
- 2. Description of Change: .
This evaluation addresses the use of a substitute valve for one of the !
pressurizer code safety valves under certain conditions. !
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- 3. Reason for the Change: :
This valve will allow RCS pressurization during a loss of decay heat removal to f permit natural circulation cooling using steam generators.
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- 4. Safety Evaluation: ;
- a. This change was safe for the following reasons: '
This valve has been evaluated with respect to its ability to provide the required RCS vent path. '
i lt has also been evaluated with respect to its ability to allow RCS pressurization during a loss of RHR.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
The maximum pressure the valve can see during its operation is 425 psig and it has been tested to 638 psig shell and 468 psig seat. Therefore, there is high confidence in the integrity of the valve and it does not increase the probability of occurrence or the consequences of a previously evaluated accident or malfunction.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
This valve is a simple manual valve. It has no other failure mechanism than has been previously evaluated.
Page 1 of 2
. - . . . . - . - - - - ~ - - - . _. . . - . - . _ ~ _ .
1 NOP 2.3-4 l l
THE MARGIN OF SAFETY AS DEFINED 'lN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is: I J
The conditions specified for use.of this valve ensures that the ability to prevent overpressurization is intact. The valve provides less flow resistance than one train of LTOP and therefore does not reduce the margin of safety.
- c. This change did not require a change to the Technical Specifications.
l 1
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l Page 2 of 2
I NOP 2.4-1
\
- 1. Procedure Number: NOP 2.4-1 I j.
Title:
Efessurizer Safety Snooloiece with Valve
- 2. Description of Change
- This evaluation addresses the use of a substitute valve for one of the
- pressurizer code safety valves under certain conditions.
l 3. Reason for the Change:
b \
i This valve will allow RCS pressurization during a loss of decay heat removal to !
! permit natural circulation cooling using steam generators.
- 4. Safety Evaluation.
l
- a. This change was safe for the following reasons- I This valve has been evaluated with respect to its ability to provide the required RCS vent path. l it has also been evaluated with respect to its ability to allow RCS pressurization during a loss of RHR.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFEW PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
The maximum pressure the valve can see during its operation is 425 psig and it has been tested to 638 psig shell and 468 psig seat. Therefore, l there is high confidence in the integrity of the valve and it does not '
increase the probability of occurrence or the consequences of a ;
previously evaluated accident or malfunction.
J THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFEW ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for i this statement is:
This valve is a simple manual valve. It has no other failure mechamsm than has been previously evaluated.
Page 1 of 2 l
4 I
NOP 2.4-1 ;
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY ,
TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
, The conditions specified for use of this valve ensures that the ability to ;
prevent overpressurization is intact. The valve provides less flow
' resistance than one train of LTOP and therefore does not reduce the margin of safety. j
- c. This change did not require a change to the Technical Specifications.
]
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1 i 2
4 Page 2 of 2
NOP 2.4-5A R0 i
l 1. Procedure Number: NOP 2.4-5A RO i l
Title:
Estahlishing Alternate Isolated Looo Overoressure Protection l 2. Description of Change:
3 j
This change adds a new procedure.
i !
. 3. Reason for the Change: !
!- i It provides a means of overpressure protection for loops which are completely ,
i isolated from the vessel by aligning them to the drain header and relying on the I
- drain header relief valve. I l l l 4. Safety Evaluation:
l 1
! a. This change was safe for the following reasons: i i
\
i This procedure has been evaluated to determine if the alternate means of I loop overpressure protection is adequate to protect the equipment in the j isolated loop. '
No new accidents can be created by this procedure, as it affects only isolated equipment. Similarly, no new malfunctions can be created.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
)
This procedure shifts reliance for loop overpressure protection from the l overpressure protection check valves to the drain header relief valves. l Failure of a check valve to open and failure of a relief valve to open are both considered very low probability failures. In this mode, and under the guidance of thw procedure, potential sources of over pressurization are !
minimal. Sources of high pressure water are isolated. The RCP motor i breaker is racked out, and there is.no significant heat source available. I The only way to increase loop pressure would be by ambient temperature increases or valve leakage. Either of these is a very slow process, such that the relief valve has adequate capacity (the check valves are 1.5" and )
, the relief valve is 3"). Therefore, there is no increase in the probability of occurrence of a malfunction of equipment important to safe , . i Page 1 of 2
NOP 2.4-5A R0 THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
Failure of the relief valve to open when required could lead to over pressurization of the isolated loop, which is no different than failure of the check valve to open. Failure of the relief valve to reclose would lead to draining of the isolated loop which has no safety significance.
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNIC #,L SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
_ The Technical Specifications limit pressure of the primary side of the i
steam generators to 500 psig when less than 70F. Since the drain header relief is set to 150 psig, this change cannot affect the margin of safety as defined in the basis of any Technical Specification.
1
- c. This change did not require a change to the Technical Specifications.
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i Page 2 of 2
i j- NOP 2.5-1 R14
, 1. Procedure Number- NOP 2.5-1 R14 i
Title:
Maintaining Pressure in the Pressurizer Relief Tank
, 2. Description of Change:
i The UFSAR states incorrectly that the Pressurizer Relief Tank (PRT) is I
, maintained at a pressure of 3 psig using the nitrogen regulator. The pressure !
maintained in the PRT is typically 5 to 13 psig, sometimes as high as 15 psig.
l The procedure provides guidance on maintaining the PRT pressure.
l 3. Reason for the Change:
l The procedure was modified to reflect maintaining 5 to 13 psig in the PRT and
- to address the differences between operating as described in this procedure and as described in the UFSAR.
I
- 4. Safety Evaluation: i 4
- a. This change was safe for the following reasons:
- Operating the PRT in the described manner is clearly within the design parameters of the tank and cannot affect the operation of other I
equipment, specifically the pressurizer safeties and the power operated relief valves.
j b. This change does not constitute an UNREVIEWED SAFETY QUESTION
- because
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
Initial tank pressure cannot have any effect on the initiation of any accident. Maintaining the tank pressure as high as 15 psig will allow the post transient pressure to remain well below the tank's design pressure.
Since the tank has no safety function, this pressure can have no effect on the probability of occurrence or the consequences of a malfunction of equipment important to safety.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
Page 1 of 2
s NOP 2.5-1 R14 Operating the PRT in the described manner is claarly within the design parameters of the tank and cannot affect the operation of any other equipment, specifically the pressurizer safeties and pressure operated j relief valves.
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICrilON HAS NOT BEEN REDUCED. The basis for this statement is:
$ Operation of the PRT is not addressed in the Technical Specification or in i their bases.
- c. This change did not require a change to the Technical Specifications.
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Page 2 of 2
i NOP 2.6-1A R4 l l
- 1. Procedure Number: NOP 2.6-1 A R4 +
Title:
Mode 5 or Mode 6 RCP Seal Water Sunoly
- 2. Description of Change:
{
This procedure provides three methods for supplying seal water to the RCP l
during shutdown Modes 5 and 6 when RCS pressure is insufficient to provide ;
fluid return to the VCT. To minimize the production of radioactive waste water j it is desirable to use a recirculation water source (i.e. the RCS) rather than an ;
external water source.
i Method A aligns a flowpath from the discharge of the RHR pumps (in addition 1 to the normal return flowpath to the Loop 2 cold leg) to the suction of the '
charging pumps via RH-MOV-33A or 33B. l Method B aligns a slipstream flowpath from the purification system (with RHR i or cavity purification in service) to the VCT Method C aligns a slipstream flowpath from the purification system (with RHR ;
or cavity purification in service) to the RWST. l t
These above three methods for supplying RCP seal water are not specifically ,
discussed in the CY FSAR.
- 3. Reason for the Change:
The procedure was updated as part of the normal station procedure revicion process. ,
- 4. Safety Evaluation: !
i
- a. This change was safe for the following reasons: -
l This procedure does not subject any of the involved systems to conditions for which they were not designed nor does it place any systems in a ,
configuration not originally intencuJ. The two postulated failures (a failure -
of an RHR pump or pressure boundary leakage from any of the involved t systems) are anticipated transients and procedures already exist for t coping with such transients. AOP 3.2-12, " Loss of RHR" and AOP 3.2-1 31A, " Reactor Coolant System / Refueling Cavity Leak (Modes 5 and 6)
! contain guidance for mitigating these transients. :
i b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
Page 1 of 2
NOP 2.6-1 A R4 THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
The design basis boron dilution accident assumes a dilution flow rate of 180 gpm from the primary water pumps. The results of this analysis show that the operators have at least 15 minutes in Modes 1 through 5 and 30 minutes in Mode 6 to terminate the inadvertent dilution before a loss of shutdown margin occurs. Any portion of the piping system which is not already in equilibrium with RCS boron concentration which could be flushed to the RCS during this procedure is small compared to the 2700 gallons (180 gpm x 15 min) of primary water (0 ppm boron) assumed to be input to the RCS in the boron dilution analysis.
This procedure has no affect directly or indirectly on fuel handling equipment or the fuel handling process.
The equipment needed to detect and mitigate a boron dihition accident is the Nuclear Instrument source range monitors (used to cause a boron dilution alarm). This procedure has no effect directly or indirectly on these instruments.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
This procedure does not subject any of the involved systems to conditions for which they were not designed nor does it place any systems in a configuration not originally intended. Appropriate limits and precautions have been specified in the procedure to ensure all equipment is operated within it's design limits.
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
The procedure contains the necessary precautions and guidance to preclude an adverse impact on any of the physical protective boundaries:
- c. This change did not require a change to the Technical Specifications.
1 Page 2 of 2
NOP 2.6-7 R8 i
- 1. Procedure Number: NOP 2.8-7 R8
Title:
Maintainino Pressure in the Pressurizer Relief Tank j
- 2. Description of Change:
The UFSAR states incorrectly that the Pressurizer Relief Tank (PRT) is maintained at a pressure of 3 psig using the nitrogen regulator. The pressure maintained in the PRT is typically 5 to 13 psig, sometimes as high as 15 psig.
The procedure provides guidance on maintaining the PRT pressure.
- 3. Reason for the Change: i The procedure was modified to reflect maintaining 5 to 13 psig in the PRT and to address the differences between operating as described in this procedure and as described in the UFSAR.
- 4. Safety Evaluation:
- a. This change was safe for the following reasons:
Operating the PRT in the described manner is clearly within the design parameters of the tank and cannot affect the operation of other equipment, specifically the pressurizer safeties and the power operated i relief valves.
i
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
l THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE l OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF
, EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is: i initial tank pressure cannot have any effect on the initiation of any accident. Maintaining the tank pressure as high as 15 psig will allow the post transient pressure to remain well below the tank's design pressure.
- Since the tank has no safety function, this pressure can have no effect on the probability of occurrence or the consequences of a malfunction of equipment important to safety.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
Page 1 of 2
NOP 2.6-7 R8 4
t Operating the PRT in the described manner is clearly within the design ;
parameters of the tank and cannot affect the operation of any other i
, equipment,'specifically the pressurizer safeties and pressure operated l relief valves. '
s THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY I TECHNICAL SPECIFICATION HAS NO7 BEEN REDUCED. The basis i for this statement is: ;
Operation of the PRT is not addressed in the Technical Specification or in their bases. (
- c. This change did not require a change to the Technical Specifications.
l l
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l Page 2 ui 2
NOP 2.612 :
l i
- 1. Procedure Number: NOP 2.6-12 '
Title:
Drainina the RCS in Mode 5 and 6
~
l l l 2. Description of Change:
This change is a new procedure used to reduce inventory in the reactor coolant ;
system. It provides several alternate means of draining the RCS while in Mode !
l 5 or 6. Those methods are: draining through the drain header to the PDT, l diverting RHR purification to the RWST or the ADT, and diverting RHR flow to the RWST.
- 3. Reason for the Change:
This procedure formalizes processes which have been in use which were not previc :', covered by a procedure.
- 4. Safety Evaluation:
l
- a. This change was safe for the following reasons: l
! Since the RCS will be cooled down and depressurized prior to performing this evolution, p: essure and temperature will be low enough for any of the involved systems. For any of the flowpaths, indication and interpretation of reactor vessel and pressurizer level is critical. The procedure has been l determined to provide adequate reference information and redundant l inventory checks to ensure that level changes are done in a controlled ,
fashion with reasonable assurance that the desired level can be achieved.
l Multiple backup isolation is available to prevent overdraining in the event the normalisolation valve fails.
, b. This change does not constitute an UNREVIEWED SAFETY QUESTION I
because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
The boron dilution and fuel handling accidents (the only accidents
, evaluated in Mode 5 and 6) were reviewed and found to be unaffected by this change. This is a process used when the unit is in Mode 5 or 6 only.
It affects only the draining of the reactor coolant system. It therefore has no effect on any accident analyses.
Page 1 of 2
NOP 2.6-12 ;
Malfunction of any of the involved components could lead to overdraining of the reactor coolant system. This can lead to loss of residual heat removal. The methods involved can drain the RCS to the bottom of the nozzles, which is still well above the fuel. Since the process is better controlled, and indication has been improved, and adequate information is now available to verify inventory changes, and multiple backup isolation is available, the consequences of any malfunction are reduced.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
The only event which can possibly occur as a direct result of using this procedure is overdraining the RCS. This can occur only by equipment malfunction or by operation inattention. The result of this event is similar to a loss of coolant accident, but is much less severe due to the low i temperature, the lack of pressure, the lower decay heat rate, and the
- ability to isolate the draining flowpaths by multiple methods. Although the injection capability may be reduced in this mode, it is clearly adequate to restore the unit to stable conditions in conjunction with flowpath isolation.
. Therefore, it is concluded that this is not a new accident as it is clearly
- bounded by the existing LOCA analysis.
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
Neither the Technical Specifications nor the UFSAR address the draining l process. The integrity of any fission product barrier is not degraded by this process, as described above. Therefore, this change cannot affect the margin of safety as defined in the basis of any Technical Specification.
^
- c. This change did not require a change to the Technical Specifications.
i i
Page 2 of 2
NOP 2.7-3 R25 1;
i
- 1. Procedure Number: NOP 2.7-3 R25 '
Title:
Purification System Resin Reolacement l
-2. Description of Change: !
The spent resin tank level detector is inoperable requiring a temporary !
procedure change to sluice any resin. The Waste Liquid Polishing Demineralizer (1-9) whose total capacity is 40 cubic feet is scheduled for resin !
i replacement to accommodate processing of the Aerated Drain Holdup Tank l contents (approximately 60% full). The NOP has been modified to decant the l
SR tank prior to commencing any sluicing operation.
jl
- 3. Reason for the Change:
I This SE evaluates the safety impact of transferring spent resin to the spent resin tank without operable level indication and high level alarm on the tank. l i
This temporary procedure change assumes no operable level detection while ;
performing the sluicing evolution. it clearly defines the steps necessary to ;
sluice the Waste Liquid Polishing Demineralizer, decanting after every flush, to j prevent overfilling the SR tank. In addition, it mandates continuous monitoring of the vent line and STOP all operations at the first sign of an overfill.
- 4. Safety Evaluation:
- a. This change was safe for the following reasons:
The change does not cause an increase in risk to the public; adversely affect the probability or consequences of previously evaluated accidents or malfunctions; create a new accident or malfunctions; and does not decrease the existing margin of safety.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE i OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF l EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN I THE SAFETY ANALYSIS REPORT. The basis for this statement is l
This temporary procedure change increases the probability of overflowing !
the SR tank to the IX pit sump due to inoperable level indication. Several l measures to modify the system operation have been included in this TPC l to minimize potential overflow. They include: decanting the SR tank after :
Page 1 of 2
I i- I j NOP 2.7-3 R25 each flush to increase available SR tank volume, continuous monitoring of j the vent line and halting sluicing operations at the first sign of overflow.
i
! All other probabilities of occurrence are unaffected by this change since all other aspects of the Spent Resin System operation will remain
[ unchanged.
The consequences of a SR tank rupture have already been evaluated and are far more severe than an SR tank overflow. Thus, this temporary procedure change has no impact on the consequences of this previously l i
evaluated malfunction of equipment important to safety. l i This TPC has no impact on the consequences of the other previously ;
evaluated malfunctions of equipment important to safety sincc he Spent l Resin System operation is essentially unchanged. i i
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A I DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE l SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
This temporary procedure change will have no impact on any design basis accident nor does it create the possibility of a new unanalyzed accident. l This temporary procedure change will not cause a malfunction of a i different type than previously evaluated because other than inoperable ;
level indication which has been compensated for via this temporary i procedure change, the Spent Resin System operation is essentially i unchanged.
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
This temporary procedure change has no impact on the margin of safety inherent in the physical protective boundaries (spent resin building).
- c. This change did not require a change to the Technical Specifications.
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4 Page 2 of 2
NOP 2.9-1 R32 l
l 1. Procedure Number: NOP 2.9-1 R32
Title:
Placing the Residual Heat Removal System in Service -
l
- 2. Description of Change:
- This change revised the procedure as follows
i
, Equipment specification limitatiens have been proceduralized in regards:to
- maximum allowable pump and I; eat exchanger flow rates. The change to the l procedure involves the use of new vent valves installed on the RHR pump L casing and seal water cooler. Verifications have been added to ensure proper i pump and motor oil levels prior to starting the RHR pumps. Guidance has i been given on starting a second RHR pump, including use of a pump specific l valve alignment check list. Guidance has been given to increase the allowable j difference between RHR process and RCS bulk temperatures from 20 to 30 j deg. F (prior to initiating shutdown cooling).
f
- 3. Reason for the Change:
1 I
This change was implemented to provide additional guidance to the operator regarding RHR pump operation. Vent valves were added to the system and subsequently to the procedure to permit more thorough venting of the pump and seal water cooler prior to startup.
- 4. Safety Evaluation:
- a. This change was safe for the following reasons:
The changes provide additional guidance to the operator intended to enhance the reliability of the RHR system and ensure _ that system components are operated within their design limits. The operation of the RHR system remains unchanged. As such, there are no changes to any of the key parameters associated with the accident analyses. These procedure changes will not affect any of the parameters associated with ensuring the integrity of the physical protective boundaries to the release of radioactivity.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN 1 THE SAFETY ANALYSIS REPORT. The basis for this statement is:
Page 1 of 2
l NOP 2.9-1 R32 The changes provide additional guidance to the operator intended to enhance the reliability of the RHR system and ensure that system components are operated within their design limits. The operation of the RHR system remains unchanged. As such, there are no changes to any of the key parameters associated with the accident analyses. These procedure changes will not affect any of the parameters associated with ensuring the inteody of the physical protective boundaries to the release ,
of radioactivity. !
1 THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statemer.t is:
The changes provide additional guidance to the operator intended to enhance the reliability of the RHR system and ensure that system components are operated within their design limits. The operation of the RHR system remains unchanged. As such, there are no changes to any of the key parameters associated with the accident analyses. These procedure changes will not affect any of the parcmeters associated with ensuring the integrity of the physical protective boundaries to the release of radioactivity.
-THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
The changes provide additional guidance to the operator intended to l enhance the reliability of the RHR system and ensure that system '
components are operated within their design limits. The operation of the RHR system remains unchanged. As such, there are no changes to any of the key parameters associated with the accident analyses. These procedure changes will not affect any of the parameters associated with ensuring the integrity of the physical protective boundaries to the release of radioactivity.
- c. This change did not require a change to the Technical Specifications.
Page 2 of 2
.- . - . . . . - ~ - . - - - . . - - - - - - . - - - - - - - . - . . - - - . . -
1 NOP 2.24-3
- 1. Procedure Number: NOP 2.24-3
Title:
Filtered SWS and Adams Filter Ooeration
- 2. Description of Change:
!~
- The change to procedure NOP 2.24-3 will allow parallel operation _ of the
. primary Adams filters (FL-53-1A/B) during periods of flooding and high debris i suspension in the Cur.necticut River. It will also allow restricting Service Water i
flow to the CAR fan coolers to less than 400 gpm for short periods (up to a l
li week) by throttling the filter outlet valves SW-V-838A & B.
- 3. Reason for the Change:
l The change was to allow parallel operation of the primary Adams filters during
- - periods of flooding and high debris suspension in the Connecticut River.
- 4. Safety Evaluation: j j a. This change was safe for the following reasons:
1
- The hydraulic performance of the SWS will not be affected by parallel flow
[ through the filters since the filters are bypassed during a LOCA and the i' bypass lines are sized to pass full required flowrate.
l
[ b. This change does not constitute an UNREVIEWED SAFETY QUESTION l because:
i THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE j OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF
! EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN
{ THE SAFETY ANALYSIS REPORT. The basis for this statement is:
l The probability of occurrence of previously evaluated accidents are unaffected by the proposed change which allow parallel operation and i throttling of the Adams Filters during river flooding conditions. This change can not cause an LNP or LOCA and can not adversely effect the 1 system hydraulic model assumptions and requirements since no credited automatic or manual valves are changing as a result of this procedure i change.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE i I
SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
Page 1 of 2 ,
NOP 2.24-3 The possibility of an accident of a different type than previously evaluated is not created by the proposed procedure change based on Section 2.1.3 .
and the fact that Service Water System failures can not initiate a DBA.
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
The hydraulic performance of the SWS will not be affected by parallel flow through the filters since the filters are bypassed during a LOCA and the
- bypass lines are sized to pass full required flowrate.
- c. This change did not require a change to the Technical Specifications.
l 4
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Page 2 of 2
l PMP 9.1-8 :
i
- 1. Procedure Number: PMP 9.1-8 ,
i i
Title:
Low Pressure Steam Dumo Test I
- 2. Description of Change: !
The proposed procedure change allows operations to close the motor-operated isolation valve (SD-MOV-108) after testing of the associated steam dump valve ;
(SD-V-104) due to steam leakage by the valve. By closing the motor-operated l
, valve, seat leakage past the associated air-operated valve will be minimized >
- and plant efficiency will be increased.
- 3. Reason for the Change:
- The procedure revision will allow MOV (SD-MOV-108) to be closed after testing 3 due to steam leakage past SD-V-104 to increase plant efficiency. )
i l
- 4. Safety Evaluation- '
- 1 i a. This change was safe for the following reasons:
l j The change does not cause an increase in risk to the public; adversely l
affect the probability or consequences of previously evaluated accidents !
or malfunctions; create a new accident or malfunctions; and doei not decrease the existing margin of safety. i
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION l
,i because: 1 1
l i THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE ,
! OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF I
) EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN i
- THE SAFETY ANALYSIS REPORT. The basis for this statement is: )
l l The low pressure steam dump system is designed to limit peak overspeed
- following full load rejection. There are no licensing basis accidents l l associated with this evolution. Therefore, the change will not have any l l effect on the probability of occurrence or consequences of any previously i
! evaluated accidents or malfunctions, j i i i THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A (
i DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE i j SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for l 1 this statement is I
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PMP 9.1-8 The proposed change only isolates one low pressure steam dump valve, thus remaining within the design of the system. Therefore, the change will not create the possibility of a new accident or malfunction. ;
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY -
TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis j for this statement is:
The proposed LPSD valve line-up will not impact the margin of safety as defined in the Technical Specifications.
- c. This change did not require a change to the Technical Specifications.
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PMP 9.1-54
- 1. Procedure Number: PMP 91-34 !
Title:
Testing of 480 Volt Breakers i
- 2. Description of Change:
l Preventive maintenance will be performed on #3 and #4 CAR fan's from its !
emergency power supply (Bus 11) breakus. The work will be performed on l one fan supply breaker at a time, j
- 3. Reason for the Change:
I The change is being made to allow preventive maintenance to be performed on
- 3 and #4 CAR fan's.
- 4. Safety Evaluation:
I
- a. This change was safe for the following reasons:
The change does not cause an increase in risk to the public; adversely affect the probability or consequences of previously evaluated accidents or malfunctions; create a new accident or malfunctions; and does not decrease the existing margin of safety.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
l The CAR fans mitigate the consequences of an accident and cannot cause an accident. Removing / installing the CAR fan emergency power supply normal breaker or spare breaker and testing of that fan, using the I emergency power supply (Bus 11) can not cause a steam line break, LOCA, or RCCA ejection. In addition, removing and installing open breakers will not induce an electrical transient or partial loss of power.
Therefore, there is no increase in the probability of occurrence or consequence to previously evaluated accidents or malfunctions.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A ,
DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
Page 1 of 2
PMP 9.1-54 Removing / installing the CAR fan emergency power supply normal breaker or spare breaker and testing that fan using the emergency power supply will not change the way the breakers fail. A different kind of accident or malfunction can not be caused by having one fan out of service.
Therefore, there is no impact on the possibility of an accident or malfunction of a different type than previously evaluated.
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
During the removal / installation of the CAR fans emergency power supply normal breaker or spare breaker and testing of that fan using the emergency power supply, three CAR fans will remain operable and, therefore, the design basis accident containment temperature and pressure requirements will be met. In accordance with the plant Technical Specifications, three fans will remain operable and there will be no impact on the margin of safety.
- c. This change did not require a change to the Technical Specifications.
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I SNM 1.4-5 l
- 1. Procedure Number: SNM 1.4-5 !
Title:
In Plant Transfer of Fuel Excluding Refueling) f i i j 2. Description of Change: l
} TPC 96-309 to SNM 1.4-5 al!ows bypassing of the up travel limit switch on the !
{ spent fuel crane, substituting procedural and additional operational controls l j and reliance on a backup switch to prevent excessive up travel, to relocate a :
j fuel assembly which was resting on the top of a fuel rack. t 1 3. Reason for the Change:
i
- The UFSAR states that fuel is moved under eight feet of water, however actual i
- i. plant configuration made it necessary to reduce the pool level to prevent l
l submergence of a valve operator. This change allows bypassing of the up ;
- travel limit switch on the spent fuel crane, substituting procedural and l
! additional operational controls and reliance on a backup switch to prevent l l excessive up travel, to relocate a fuel assembly which was resting on the top of l 3
a fuel rack. {
1
- 4. Safety Evaluation: !
- a. This change was safe for the following reasons: )
i This change does not cause the consequences of any accident to be l; greater than previously evaluated.
The fuel assemblies will remain cooled with the same volume of water even though the fuel is moved under seven feet. The shielding provided }
by the seven feet is more than adequate. i
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION I because: i l
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE ;
OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF l EQUlPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
There are no failure mechanisms of any equipment which can be affected by this change. Therefore, there is no increase in the probability of occurrence of the fuel handling accident. The. fuel handling accident, as evaluated, assumes all the fue' pins in the dropped assembly are ruptured. This change cannot make that condition worse.
Page 1 of 2 J
SNM 1.4-5 i
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A i DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE i SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for i this statement is:
- This change cannot create the possibility of a new accident. The worst
, case accident has already been evaluated.
l THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY i TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis j for this statement is:
There is no change to the margin of safety as defined in the basis for any
- technical specification. This change does not cause the consequences of j any accident to be greater.
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- c. This change did not require a change to the Technical Specifications. '
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i SNM 1.4-5
- 1. Procedure Number: SNM 1.4-5 i
Title:
In Plant Transfer of Fuel (Excludina Refuelina) i
- 2. Description of Change:
) TPC 96-311 to SNM 1.4-5 addresses the acceptability of moving fuel i assemblies under seven feet of water, i
l 3. Reason for the Change:
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j The UFSAR states that fuel is moved under eight feet of water, however actual plant configuration made it necessary to reduce the pool level to prevent 1
- - submergance of a valve operator. This change evaluates the acceptability of I i moving fuel assemblies under seven feet.
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j 4. Safety Evaluation: l l 1 l a. This change was safe for the following reasons: l This change does not cause the consequences of any accident to be ,
greater than previously evaluated. )
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i I The fuel assemblies will remain cooled with the same volume of water j even though the fuel is moved under seven feet. The shielding provided l i
by the seven feet is more than adequate. Extensive surveys have verified
- the acceptability of the dose rates.
I
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION l
[
because- l
} THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF
! EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN l j THE SAFETY ANALYSIS REPORT. The basis for this statement is:
- (
There are no failure mechanisms of any equipment which can be affected j by this change. Therefore, there is no increase in the probability of j occurrence of the fuel handling accident. The fuel handling accident, as evaluated, assumes all the fuel pins in the dropped assembly are ruptured. This change cannot make that condition worse.
L THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A l DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE l 1
Page 1 of 2 l
SNM 1.4-5 SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
This change cannot create the possibility of a new accident. The worst case accident has already been evaluated. ,
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is: ,
There is no change to the_ margin of safety as defined in the basis for any technical specification. This change does not cause the consequences of ._
any accident to be greater.
- c. This change dd not require a change to the Technical Specifications. ;
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SNM 1.4-5
- 1. Procedure Number: SNM 1.4-5
Title:
In Plant Transfer of Fuel (Excluding RefirSog) i
- 2. Description of Change:
l 4
TPC 96-320 to SNM 1.4-5 revised the fuel lifting load limit specified in the procedure from 1800 pounds to 2000 pounds for fuel assemblies A52 and E31.
- 3. Reason for the Change:
f i
Fuel assemblies A52 and E31 were exhibiting high drag within the Spent Fuel l Storage Racks during movement required to perform a partial re-rack of the .
Haddam Neck Plant Spent Fuel Storage Pool.
j
- 4. Safety Evaluation: !
- a. This change was safe for the following reasons:
The proposed operation of the fuel handling equipment does not j invalidate any of the assumptions or conclusions of the fuel handling i accident and is bounding for this change. The proposed operation of the !
fuel handling equipment at a load limit of 2000 pounds did not change the {
consequences of the Fuel Handling Accident. :
i
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
i THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE ;
OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF i EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN !
THE SAFETY ANALYSIS REPORT. The basis for this statement is- l There was no increase in the probability of occurrence of a fuel handling accident by increasing the lifting force limit to 2000 pounds. The structural integrity of the Spent Fuel Handling Tool and the fuel assemblies would !
have been maintained at a load of at least 2500 pounds. There is no ;
increase in the consequences of the fuel handling accident, as evaluated, [
since it assumes that all the fuel pins in the dropped assembly are ;
ruptured. This change cannot make that condition worse.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A -
DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE l SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for !
this statement is:
Page 1 of 2 l I
SNM 1.4-5 Failure of the fuel assembly or the fuel handling tool could not cause an accident of a different type than previously evaluated. Although a postulated failure of a storage rack module is a malfunction of a different type, it was shown that this rack module was not important to safety. The worst case accident has already been evaluated.
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis ;
for this statement is: ,
There is no change to the margin of safety as defined in the basis for any i technical specification. i TS 3.9.11 This change does not change the water depth available.
TS 3.9.12 The radiological release postulated has remained unchanged.
TS 3.9.13 The boron concentration was administratively maintained at a much greater value than was required.
- c. This change did not require a change to the Technical Specifications.
Page 2 of 2
SPL 10.3-32 l
- 1. Procedure Number: SPL 10.3-32 l
l
Title:
Moving Soent Fuel Assembly R45 in Soent Fuel Pool
- 2. Description of Change:
This procedure was written to provide instructions for the operation of moving spent fuel assembly R45 from one storage rack location to another. This fuel handling operation was different than described in the UFSAR in that slings were used to move fuel assembly R45 instead of normal fuel handling equipment.
- 3. Reason for the Change:
This procedure was prepared in order to move assembly R45, using two slings.
- 4. Safety Evaluation:
- a. This change was safe for the following reasons: ;
While this method of moving fuel was different than that described in the l UFSAR, it was safe since two slings were used, each of which had been load tested to accommodate the entire load. The specific fuel assembly being moved had only one cycle of exposure and has decayed for ten years. The existing fuel handling accident clearly bounds this activity.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because: l THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE l OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
l The special procedure addressed the use of load tested slings versus the normal spent fuel handling tool and since the load lift capability of the two slings exceeded that of the fuel handling tool, there is not increase in the probability of occurrence or the consequences of an accident or malfunction of equipment previously analyzed.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
Page 1 of 2
SPL 10.3-32 '
This procedure provides a safe load path and protection across the ,
exposed pool floor liner and does not create the possibility of a new type '
of accident.
l THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis ,
for this statement is: .
l The minimum water level required by the Technical Specifications is maintained. The movement of assembly R45 does not affect the pool l water level and therefore the margin of safety is maintained. .;
- c. This change did not require a change to the Technical Specifications.
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SUR 5.1-0 R30
- 1. Procedure Number: SUR 5.1-0 R30 5
Title:
Maintaining Pressure in the Pressurizer Relief Tank
- 2. Description of Change
The UFSAR states incorrectly that the Pressurizer Relief Tank (PRT) is maintained at a pressure of 3 psig using the nitrogen regulator. The pressure maintained in the PRT is typically 5 to 13 psig, sometimes as high as 15 psig.
The procedure provides guidance on maintaining the PRT pressure.
- 3. Reason for the Change:
The procedure was modified to reflect maintaining 5 to 13 psig in the PRT and .
to address the differences between operating as described in this procedure and as described in the UFSAR.
- 4. Safety Evaluation: -
- a. This change was safe for the following reasons: -
Operating the PRT in the described manner is clearly within the design parameters of the tank and cannot affect the operation of other
- equipment, specifically the pressurizer safeties and the power operated relief valves.
- b. This change does not constitute an UNREVIEWED SAFET/ QUESTION because:
a THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT iMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is: .
2 Initial tank pressure cannot have any effect on the initiation of any accident. Maintaining the tank pressure as high as 15 psig will allow the post transient pressure to remain well below the tank's design pressure.
, Since the tank has no safety function, this pressure can have no effect on the probability of occurrence or the consequences of a malfur:ction of i
equipment important to safety.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
Page 1 of 2
SUR 5.1-0 R30 l
Operating the PRT in the described manner is clearly within the design !
parameters of the tank and cannot affect the operation of any other ,
equipment, specifically the pressurizer safeties and pressure operated relief valves. l 1
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
Operation of the PRT is not addressed in the Technical Specification or in !
their bases. '
- c. This change did not require a change to the Technical Specifications.
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SUR 5.5-70
- 1. Procedure Number: SUR 5.5-70 {
Title:
Crosbv Main Steam Safety Valve Surveillance Testing i l
- 2. Description of Change:
The change allows for Mode 1 testing of the Crosby Main Steam Safety Valves (MSSVs) at reduced power levels for verification of valve setpoints. This surveillance only affects the self-actuation function of the MSSVs at Haddam Neck Plant (HNP). This change differs from past testing which verified valve setpoints in Mode 3.
- 3. Reason for the Change:
This change will allow testing of the HNP MSSVs at power to reduce the time spent in Mode 3 during scheduled plant refuelings.
- 4. Safety Evaluation: ;
- a. This change was safe for the following reasons:
Testing of a main steam valve at no time renders the valve inoperable. l The ability to provide over pressure protection is not jeopardized.
i The procedure provides instruction on crucial installation steps to j minimize the chance of hanging up a safety valve.
Testing is bounded by previous analyses.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
Testing of the main safety valves at no time renders a valve inoperable.
The ability to provide overpressure protection is not jeopardized. Testing of these valves is bounded by the existing loss of load analysis.
Therefore, the testing of these valves does not increase the probability of occurrence or the consequence of an accident or malfunction previously evaluated.
Page 1 of 2
__ _._ . ._. _ __ _ . _ _ . _ . _ __ ._.. _ ~. _.- _ _. _ - _ . . _ _ .. . . _ -
I SUR 5.5-70 -
I THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The bash for
- this statement is:
1 1
1 A stuck open safety valve and a safety valve which fails to open on l 2
demand have been analyzed. These events are bounded by the existing i accident analysis. No accidents or malfunctions of a different type than '
, previously evaluated are introduced.
l THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
The repair will not impact the margin of safety as defined in the basis for any Technical Specifications (TS), since there are no TSs associated with the feedwater regulator valves at the HNP.
- c. This change did not require a change to the Technical Specifications, i
Page 2 of 2
SUR 5.7-66 R9
- 1. Procedure Number: SUR 5.7-66 R9 l
Title:
Local Leak Rate and Pressure isolation' of Safety Iniection l Recirculation P-24 Isolation Valves r i
- 2. Description _of Change:
i I i l This procedure change addresses the filling of the safety injection measuring tank, TK-100-1 A, and the HPSI recirculation lines with primary water for testing j of safety injection valves SI-V-863A, B, C, and D. !
i
- 3. Reason for the Change: )
The change was to permit the testing of safety injection valves SI-V-863A, R, ;
C, and D. :
- 4. Safety Evaluation:
- a. This change was safe for the following reasons:
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l The change does not cause an increase in risk to the public; adversely j affect the probability or consequences of previously evaluated accidents l or malfunctions; create a new accident or malfunctions; and does not i decrease the existing margin of safety. i l !
- b. This change ooes not constitute an UNREVIEWED SAFET( QUESTION ;
i because-i i
THERE IS NO INCREASE IN THE PROBAB'LITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF i EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN I THE SAFETY ANALYSIS REPORT. The basis for this statement is: l The design basis accident has been evaluated and determined to be unaffected by the performance of the surveillance procedure. The boron dilution accident analysis demonstrates that the operators have adequate time to respond to an inadvertent dilution. Therefore, the change does not increase the probability of occurrence or the consequences of an accident or malfunction previously evaluated.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A l DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for l this statement is:
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Page 1 of 2 !
- SUR 5.7-66 R9 4
All failure modes that could cause an accident have been identified and
- evaluated. The change to this surveillance procedure does not create the !
- potential for a new unanalyzed accident or malfunction. I THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY l TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is-t 1
- The postulated malfunction of any one of the safety injection valves is bounded by the existing analysis. Therefore, the change does not reduce !
the margin of safety as defined in the basis for any Technical !
Specification.
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- c. This change did not require a change to the Technical Specifications. !
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__ _ _ - .,.m - _ _ - _ ._-. _ _ -._ _ _ _. - _ _ - . . _ _ _ . . . . . . -
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- VP-737 i
- 1. Procedure Number: VP-737 ;
Title:
On-Site Handling and inctallation/ Removal Procedure. Haddam Neck I Rerack Project. Holtec International Procedure No. HPP-50124-10 1
- 2. Description of Change- ,
l This evaluation addresses the procedure to be performed in support of the j removal and installation activities associated with the Haddam Neck Rerack ,
Project. The items to be removed by this procedure are: 30 existing rack I modules,13 seismic restraints, the cask pit frame and the fuel barrier frame. !
This procedure will also install 13 new high density rack modules. I 1
- 3. Reason for the Change:
{
This evaluation was performed in support of the removal and installation !
activities associated with the Haddam Neck Rerack Project. )
l
- 4. Safety Evaluation: I l
- a. This change was safe for the following reasons: )
The change does not cause an increase in risk to the putJic; adversely affect the probability or consequences of previously evaluated accidents i or malfunctions; create a new accident or malfunctions; and does not decrease the existing margin of safety.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
The design basis accident has been evaluated and determined to be unaffected by this procedure. Although a new rack removal and installation sequence has been developed, it is still bounded by the existing analysis. A fuel handling accident is unrelated to this proposed i procedure. Therefore, this change does not increase the probability of occurrence or the consequences of an accident or malfunction previously evaluated.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVICUSLY IN THE 1
Page 1 of 2 j l
VP-737 i l
SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
All failure modes that could cause an accident have been identified and evaluated. The installation and removal of the racks from the spent fuel !
pool does not create the possibility of a fuel handling accident. Therefore, <
the change does not create the potential for a new unanalyzed accident or malfunction.
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
Accidents concerning a rack drop, a dropped too!, a seismic event during the rerack project and load paths have been evaluated and are bounded by previous analyses. Therefore, this change does not involve a :
significant reduction to the spent fuel pool or spent fuel building's margin ;
of safety. t
- c. This change did not require a change to the Technical Specifications.
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VP-798 1
- 1. Procedure Number: VP-798
Title:
Evaluation to Begin Core Offload with Existing Cavity Seal Leakage
- 2. Description of Change:
This evaluation addresses the acceptability of changing the allowable cavity seal leakrate from 16 ml/ min. to 2000 ml/ min.
- 3. Reason for the Change:
This evaluation was performed to permit core offload with cavity seal leakage greater than the previously accepted value.
- 4. Safety Evaluation:
- a. This change was safe for the following reasons:
The change does not cause an increase in risk to the public; adversely affect the probability or conseq0ences of previously evaluated accidents or malfunctions; create a new accident or malfunctions; and does not decrease the existing margin of safety.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO . CREASE IN THE PROBABILITY OF OCCURRENCE l OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is: 1 The boron dilution and fuel handling accidents were reviewed and found to be unaffected by this change, since a fuel assembly cannot be dropped on the affected section of seal. There is no increase in the probability of a cavity seal failure. Therefore, the change does not increase the probability of occurrence or the consequences of an accident or l malfunction of equipment previously evaluated.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A l DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE l SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
Cavity seal failure has been previously evaluated. No new accidents or
! malfunctions are caused by this change.
Page 1 of 2
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VP-798 l
I THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY j TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis i for this statement is:
The Technical Specifications do not address cavity seal failure. Since there is no increase in the probability of a cavity seal failure, there is no !
reduction in the margin of safety. I
- c. This change did not require a change to the Technical Specifications.
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! HADDAM NECK PLANT SECTION lli Jumoer-Lifted Lead and Byoass (J-LL-B) Changes (Page 1 of 1)
J-LL-B Number lille 96-03 Temporary Air Ejector Vent Path 96-15 CH-RV-332 Leak Sea!
96-31 Spent Fuel Building Crane CR5-1-1 A Hoist (See SNM 1.4-5 TPC-309 in Section ll) 96-55 RVLIS B Train 40 Conductor Cable
J-LL-B 96-03 j
- 1. J-LL-B Number: 96-03 i
Title:
Temocrarv Air Elector Vent Path
- 2. Description of Change:
This J-LL-B has been removed. This J-LL-B allowed the installation of a hose
! and portable exhaust fan during a period when the Primary Auxiliary Building purge fans were shutdown and the Spent Fuel building exhaust fan was shutdown or not available to cross-connect to the process plenum and l
provided the exhaust path for the condenser air ejectors. The air ejector gases were discharged to atmosphere.
- 3. Reason for the Change:
The J-LL-B was to provide a means to throttle air ejector flow to enable welding on air ejector piping to implement a design change.
- 4. Safety Evaluation:
- a. This J-LL-B was safe for the following reasons:
This J-LL-B was safe because it did not violate the intent or practices of R the Radioactive Effluent Monitoring Manual /Offsite Dose Calculation Manual (REMM/ODCM). This J-LL-B did not violate any Technical Requirements Manual items since all activity was quantified. This J-LL-B did not constitute any added discharge to the environment, and so did not increase the dose consequences to the public. Installation of this J-LL-B l was safe.
- b. This J-LL-B does not constitute an UNREVIEWED SAFETY QUESTION i because:
i THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
Installation and removal of a temporary air ejector vent path did not l increase the probability of occurrence of a Steam Generator Tube Rupture. This J-LL-B was installed and removed on the air ejector system which is not safety related equipment. This J-LL-B did not violate the intent or practices of REMM/ODCM or the Technical Requirements Manual and did not increase the dose consequences to the public. The air ejector monitor was in service monitoring the release. The net effect of Page 1 of 2
J-LL-B 96-03 this J-LL-B was the loss of dilution from gas emissions through the stack.
i Since the amount of iodine released in the condenser is small, the loss of stack dilution has an insignificant impact on offsite dosage during the design basis tube rupture when compared to the discharge from the code safety valves and Auxiliary Feedwater steam exhaust.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE l SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for ,
this statement is: l 1
Installation and removal of a temporary air ejector vent path did not create ;
an accident of a different type previously evaluated. This J-LL-B did not l increase the possibility of a malfunction of a different type than previously l evaluated. This J-LL-B was installed and removed on non-safety related l equipment. It did not affect any safety-related equipment.
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY {
I TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis 1 for this statement is:
N j installation and removal of a temporary air ejector vent path did not !
decrease the margin of safety. This J-LL-B did not violate the intent or practices of the REMM/ODCM or the Technical Requirements Manual and did not increase the dose consequences to the public. The net effect of
- this J-LL-B was the loss of dilution from gas emissions through the stack.
Since the amount of iodine released in the condenser is small, the loss of stack dilution has an insignificant impact on offsite dosage during the design basis tube rupture
- c. This change did not require a change to the Technical Specifications.
Page 2 of 2
J-LL-B 96-15
- 1. J-LL-B Number: 96-15
Title:
CH-RV-332 Inlection Leak Seal
- 2. Description of Change:
- This change consisted of making an injection leak seal on the body to bonnet joint of Reactor Coolant Pump (RCP) seal water return relief valve CH-RV-332.
The J-LL-B will be removed during RFO-19.
i
- 3. Reason for the Change: !
l A leak seal was necessary as a temporary repair of a body to bonnet joint leak on CH-RV-332. '
- 4. Safety Evaluation:
- a. This change was safe for the following reasons:
Compliance with the limits provided in Haddam Neck Plant procedure CMP 8.5-149, " Leak Sealing" and the associated technical evaluation for performance of the leak sealing activity effectively prevented the gross failure of the valve pressure boundary. '
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
Compliance with the limits provided by CMP 8.5-149 and the associated technical evaluation effectively prevented the possibility of a gross failure of the valve pressure boundary (body to bonnet joint). Thus, the work did not increase the probability of any design basis accidents. Even if gross failure of the valve pressure boundary had occurred, it would not result in a loss of coolant from the Reactor Coolant System (RCS) (i.e., will not cause a LOCA), since the amount of seal water returned from the Reactor Coolant Pumps (RCPs), approximately 8 gallons per minute (gpm) represents that amount of seal supply water which is not injected into the RCS.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A
{ DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE Page 1 of 2 1
l J-LL-B 96-15 i
SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for '
this statement is:
i While a pressure boundary failure of this valve could have posed a transient to the chemical and volume control system, it would not represent a new or different type of accident.
l l
A postulated pressure boundary failure of this valve is isolable on the i upstream side (l. e., from the RCPs). Leakage from the Volume Control l Tank (VCT) side is not isolable. The VCT makeup system may or may not be able to keep up with the postulated leak depending on its magnitude.
In the worst case, VCT level would continue to drop until either operator action is taken or automatic charging pump suction switchover to the Refueling Water Storage Tank (RWST) _ occurs (at 10% VCT level).
Following switchover of charging pump suction to the RWST, inputs to the i VCT would need to be stopped or redirected to stop the leak (i.e. letdown, !
makeup, and hydrogen) A postulated release of letdown and VCT )
radioactive gas contents into the PAB during this scenario would not result in radiological consequences worse than those predicted in the analysis of a letdown line break outside of containment that was performed as part of the Systematic Evaluation Program (SEP Topic XV-16) and is documented in a letter from D. Crutchfield to W. Counsil, dated July 10, 1981.
i THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
A pressure boundary failure of this valve will not affect any of the parameters associated with ensuring the htegrity of the physical protective boundaries to the ielease of rau;oactivity, namely the fuel cladding, the RCS pressure boundary, or the containment boundary. !
Therefore, there is no reduction in the margin of safety by performing the leak seal. '
- c. This change did not require a change to the Technical Specifications.
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Page 2 of 2 m - _ . _ - - . _ - . . - . _ - _ , _. :
J-LL-B 96-55
- 1. J-LL-B Number: 96-55
Title:
RVLIS B Train 40 Conductor Cable
- 2. Description of Change:
This evaluation addresses the use of a 40 conductor cable, manufactured by ABB, to temporarily connect the 'B' Heated Junction Thermocouple (HJTC) probe to the Refuel disconnect Panel (RDP) in order to restore the 'B' train of RVLIS.
- 3. Reason for the Change:
This change is being made to provide enhance ability to determine Reactor Vessellevel while in Mode 5.
- 4. Safety Evaluation:
- a. -This change was safe for the following reasons:
The change does not cause an increase in risk to the public; adversely affect the probability or consequences of previously evaluated accidents or malfunctions; create a new accident or malfunctions; and does not decrease the existing margin of safety.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
This change does not affect any of the design basis accidents and therefore does not increase the probability of occurrence or the consequences of a previously evaluated accident or malfunction.-
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
The failure of this change would result in the lost capability to read Reactor Vessel level. This is not an indication normally provided during Page 1 of 2
J-LL-B 96-55 Mode 5 and is not required by Technical Specifications. Therefore, this change cannot cause a new accident.
3
^
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
This cnange does not alter the function of the RVLIS system nor is the
. RVLIS system required by Technical Specifications in Mode 5. The t
indication provided will enhance the operator's ability to determine Reactor Vessel level while in Mode 5. The change will have no impact on
, the margin of safety.
- c. This change did not require a change to the Technical Specifications.
i J
r Page 2 of 2 I
4 HADDAM NECK PLANT SECTION IV Tests (Page 1 of 1)
The following tests were performed under the provisions of Title 10, Code of Federal l Regulations, Section 50.59 during 1996.
Procedure Number litig ST 11.7-135 Special Setpoint Testing for Main Steam Safety Relief
! talves, MS-SV-14, 24, 34, and 44
. ST 11.7-167 Load Test Spent Fuel Building Crane (CRS-1 A)
ST 11.7-168 SWS Acute Treatment for MIC, Functional Test of PDCR 1553 l ST 11.7-172 DWST Emergency Fill from PWST and RPWST Test ST 11.7-193 Low Pressure Safety injection System Operability Test ST 11.7-194 South Service Water Header Simulated LNP Test l
, ST 11.7-196 Reactor Head Vent System Functional Test ST 11.7-200 Underwater Reactor Cavity Hatch Seal Troubleshooting k
ST 11.7-135
- 1. Test Number: ST 11.7-135
Title:
Soecial Setooint Testina for Main Steam Safety Relief Valves.
MS-SV-14. 24. 34. and 44
- 2. Description of Change:
This change allows the removal of a blowdown , atrictors following blowdown l
of an isolated pilot valve. This will ensure that when the inactive pilot valve is !
made active, opening of the blowdown valve will not lead to lifting of the main '
valve. ,
l
- 3. Reason for the Change: ;
I With the blowdown restrictors installed, setpoint verification could not be properly performed with an improved testing apparatus. By removing the restrictor from an isolated pilot valve, setpoint verification will be facilitated, and pilot valve life improved. ;
- 4. Safety Evaluation:
- a. This change was safe for the following reasons:
Inadvertent actuation of the main valve is precluded by the complete blowdown of the pilot valve with the restrictor still installed. After the pilot l valve is positively checked as isolated, the restrictor is removed to ;
facilitate testing. Before being repressurized, the flow restrictor is placed back on the pilot valve.
4
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because: l THERE IS NO INCFsEASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
The accident evaluated under the proposed change is the spurious opening of a main steam safety valve. Inadvertent actuation of the main valve is precluded by the complete blowdown of the pilot valve with the restrictor still installed. The malfunction of equipment important to safety is failure of the safety valve to open on demand. At no time during the testing of a particular pilot valve is the ability to provide overpressure 1 protection jeopardized.
Page 1 of 2
_ . -. . .- - - . - . . - ~- - - . . .- . . . . - --
t ST 11.7-135 THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A
{ DIFFERENT TYPE THAN ' ANY EVALUATED PREVIOUSLY IN THE I SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for !
j this statement is: i
! This special test will not create any accident or malfunction different than any evaluated previously. Spurious opening of a main steam safety valve is bounded by the excess load analysis contained in the UFSAR.
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY
- TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
- i. The loss of load analysis shows that the safety valve with the highest set l- pressure (the main steam safety valves being tested) would not be challenged during this event unless some other failure were to occur.
', Should another safety valve failure occur, the ability of the safety valves i being tested to actuate would not be lose. Therefore the margin of safety
,_ is not reduced during this test.
- c. This change did not require a change to the Technical Specifications.
l 1
l 3
Page 2 of 2
- ST 11.7-167 a j
- 1. Test Number: ST 11.7-167 l
i
Title:
ST 11.7-167 for Load Test Soent Fuel Buildina Crane (CR5-1 A) l 2. Description of Change:
i This change tests crane CR5-1 A which will be loaded with a 7.5 ton test weight
- in the fuel Storage BuilAg (FSB) buggy bag. This ST will load test the crane to 125% (7.5 tons) as is required for the upgrade of cranes by ASME B30.2
" Overhead and Gantry Cranes".
l 3. Reason for the Change:
- To test crane CR5-1A after implementation of PDCR 1586 which upgraded the crane loading from 5 tons to 6 tons.
I 4. Safety Evaluation:
i a. This change was safe for the following reasons:
l During the load test of crane CR5-1 A, there will be no crane movement or load over the spent fuel pool.
T l b. This change does not constitute an UNREVIEWED SAFETY QUESTION 1 because:
l THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE i OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
The ST will be performed to the south of the spent fuel pool (SFP). At no time will the crane or load be over the pool. At_all times there is at least one floor of separation between the test weight and the spent fuel pool cooling equipment located on elevation 21'-6". The maximum spacing of the building steel in the north / south direction if 5'-7" for floor elevations ,
47'-0" and 35'-0". During the walkdown of the load path it was determined i qualitatively that a drop of the test weight would not have a sufficient i impact force to adversely affect the spent fuel pool cooling equipment l located on elevation 21'-6". l THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A l DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE l SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
Page 1 of 2
l ST 11.7-167 l The spent fuel pump suction lines which penetrate the south end of the SFP at elevation 41'-0" are exposed at the New Fuel Vault. These two I lines extend from the wall approximately 2 feet. During the walkdown of the load test path it was determined the test weights would be approximately 5 feet to the south cf these lines. If there was a malfunction of the crane and these lines were impacted, the resultant loss I of spent fuel pool level would be bounded by the conclusions of the l l
UFSAR in chapters 9.1.3 and 15.5.2 l
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
The crane and structure have a minimum margin of safety of at least 3 times the intended load of 6 tons. The test weight to be used for the load test has been certified by the Vendor for the 7.5 ton load. The postulated malfunction of the SFP cooling system is within the bounded analysis I contained in the present CY UFSAR.
- c. This change did not require a change to the Technical Specifications.
Page 2 of 2
i ST 11.7-168 i' \
- 1. Test Number. ST 11.7-168 i
Title:
SWS Acute Treatment for MIC. Functional Test of PDCR 15fia
- 2. Description of Change: '
The proposed test procedure (ST 11.7-168) will inject Bulab 7005 and 8007 ;
into the Service Water System as designed and evaluated by PDCR 1553 for i an acute continuous treatment period of 45 days. The use of the Bulab chemicals, which are designed to combat Microbiologically influenced '
Corrosion (MIC), has been evaluated for its safety impact by Safety Evaluation [
CY-SE-1553k "SW MIC Chemical Injection".
- 3. Reason for the Change:
To perform a functional test of PDCR 1553.
- 4. Safety Evaluation:
- a. This change was safe for the following reasons: ;
ST 11.7-168 is not a USQ since it puts safety-related valves and l components in their fail-safe position or condition for responding to a DBA. The EDG room and lube oil temperatures will not be adversely ,
impacted. The fouling levels in the EDG heat exchangers will bo monitored and will eliminate the potential for any common mode or single failure of SW-FCV-129 & 130 to open. In addition, the use of the chemicals in this Special Test will not only reduce the potential for MIC, ,
but will also reduce siltation deposition in the Service Water System. !
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
The probability of occurrence of previously evaluated accidents are unaffected by the performance of ST 11.7-168, which changes the position of valves and standby equipment in the SWS to allow chemical treatment while maintaining the system hydraulic model assumptions.
Performance of the ST can not cause an LNP or LOCA.
Page 1 of 2
)
i ST 11.7-168 l
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A
- DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is
The possibility of an accident of a different type than previously evaluated ;
is not created by performance of ST 11.7-168 since the test only effects '
Service Water cooling and the loss of SW is not an accident initiating event. Loss of the SWS has also been evaluated and an AOP created to cope with this unlikely event.
, The possibility of a malfunction of a different type than previously I evaluated is not created by performance of ST 11.7-168 since the test l simply aligns flow to standby components, which if fouled can be cleaned ;
online. /Any failure of the non-QA chemical treatment injection l piping / tubing will not result in a malfunction of the SWS.
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY ;
TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is- l The margin of safety is not affected by estformance ST 11.7-168 based upon the fact that any failure of the ran-QA chemical treatment injection piping / tubing will not result in a malfunction of the SWS.
- c. This change did not require a change to the Technical Specifications. i i
t i
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! Page 2 of 2
l ST11.7-172 l 1. Test Number: ST 11.7-172 ,
i
Title:
DWST Emergency Fill from PWST and RPWST Test
- 2. Description of Change-4
(
The Primary Water Storage Tank (PWST) and Recycled Primary Water 4 Storage Tank (RPWST) system will be operated as stated in existing plant 3 l
procedures and line-ups as specified in existing Emergency Operating i
- Procedures to determine the transfer capability of the transfer pumps.
4 !
! 3. Reason for the Change:
i l The purpose of this test is to verify the capability of transferring water (at 200 ,
gpm) from the PWST and RPWST to the DWST as stated in various Haddam Neck Plant design documents.
- 4. Safety Evaluation: ;
- a. This change was safe for the following reasons:
The PWST and RPWST systems will be operated in accordance with r existing normal and emergency operating procedures. The Demineralized i Water Storage Tank (DWST) is the safety-related Auxiliary Feedwater i (AFW) suction source. All tanks will be maintained above Technical Specification and low-level alarm setpoints. These systems discharge into the top of the DWST such that a break or malfunction in the transferring system will not result in loss of DWST inventory.
- b. This change dws not constitute an UNREVIEWED SAFETY QUESTION became:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
The accidents evaluated under the proposed change are loss of feedwater or loss of normal power. Transferring water from the PWST and RPWST will not affect or cause these accidents nor cause the malfunction of equipment important to safety.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE Page 1 of 2
_. . _ - . . .- _. -__ = . - _ - .-
i
. l ST11.7-172 1 SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
Transferring water into the DWST will not create the possibility of an I accident or malfunction different than previously evaluated. l THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEH REDUCED. The basis for this statement is:
The Technical Specifica#' e quire a minimum level in the DWST and the PWST. The water se lowered below these lower limits or the lo level alarm level. .. ' s#WST level will be maintained above this limit.
- c. This change did not require a change to the Technical Specifications.
l l
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i Page 2 of 2
ST 11.7-193
- 1. Test Number: ST 11.7-193
Title:
Low Pressure Safety Iniection System Ooerability Test
- 2. Description of Change: l l
ST 11.7-193 Revision 0 is an integrated flow test that aligns the RWST to the LPSI pumps and injects into the reactor vessel through the core deluge valve train. The test will be accomplished in Mode 5 with the reactor vessel head bolted in place and the reactor coolant temperature maintained below 140 F at all times. Since the RHR system shares common piping with the LPSI piping, the RHR system must be isolated during the test. The associated Technical Specification 3.4.1.4.2 allows the RHR pumps to be de-energized for up to 1 ;
hour provided no operations are permitted which would dilute the boron concentration, and the core outlet temperature is maintained at least 10 F below the saturation temperature.
l
- 3. Reason for the Change:
To perform a low pressure safety injection system operability test.
- 4. Safety Evaluation:
- a. This change was safe for the following reasons:
i The test procedure contains the necessary guidance to operate the plant in a safe manner. Plant systems and components will be operated within design limitations. Additionally, the test limitations on subcooling, heat-up rate, cool down rate, and pressurizer levels are conservative and can accommodate the anticipated heat up rate with the RHR system secured.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPOP.T. The basis for this statement is:
Due to procedure requirements, the test procedure does not increase the probability of occurrence of a boron dilution accident. The procedure requires that: no evolution involving the addition of positive reactivity is in progress, the RWST temperature is greater than 70 F, and the RWST boron concentration is greater than or equal to the reactor coolant system
/;
Page 1 of 2
. - - . - - . . - . - . - . - _ . - - . ~ . _ . . . - . - ..
J ST 11.7-193 boron concentration. Therefore, there is no increase in the probability of occurrence of a previously evaluated accident.
The test will not affect the performance or availability of safety systems j that mitigate the consequences of accidents. The fission product
- boundaries will not be degraded. Hence, there is no increase in the radiological consequences of a postulated boron dilution accident.
4 The test will have no adverse impact on the physical protective i
boundaries including fuel, cladding, RCS pressure boundary, and j containment. Additionally, plant systems will not be operated outside technical specifications or design basis allowances. Therefore, the consequences of previously evaluated malfunctions are unchanged.
r THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE l SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for
- this statement is:
} The performance of the test procedure does not create the possibility of j an accident of a different type than previously evaluated.
- THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY l TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is
l The performance of the test procedure will have no adeerse impact on the I physical protective boundaries nor will any plant system be operated j outside of Technical Specification limits or design basis.
i
- c. This change did not require a change to the Technical Specifications.
l I
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Page 2 of 2 l
ST 11.7-194 i 1. Test Number: ST 11.7-194
Title:
South Service Water Header Simulated LNP Test
- 2. Description of Change:
} This change performs a simulated Loss-of-Normal Power (LNP) test which will j allow data acquisition on CAR fan pressure decay and Service Water pump coastdown after tripping the South SW header pumps (P-37-1C & D).
i 3. Reason for the Change:
i This test will allow analysis of actual conditions in the CAR fan service water system after an LNP. Data will also be acquired on pump coastdown.
- 4. Safety Evaluation:
- a. This change was safe for the following reasons:
f The valve alignments will not cause any credited automatic safety
- features in the SWS to be defeated. No component malfunctions can 1 occur. No Technical Specification action statement will be entered. No i credible accidents in this Mode are being affected.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION 1
because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE
- OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF l EQUIPMENT IMPORTANT TO SAFETY PREVIOllSLY EVALUATED IN l l THE SAFETY ANALYSIS REPORT. The basis for iis statement is:
\
This test is being performed in Mode 5. The only credible accident in this mode, boron dilution, is unaffected by this test. No equipment is being !
placed in a position which could cause a malfunction.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
This test does not create a new accident since the loss of the SWS can not initiate a boron dilution event. The potential malfunctions associated with this test are being prevented by procedural precautions and steps.
Page 1 of 2
ST 11.7-194 THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
The margin of s9fety is not affected by this test since the probability of occurrence or the consequences of previously evaluated accidents are unaffected. This test does not create the possibility of a new accident or malfunction in equipment important to safety.
- c. This change did not require a change to the Technical Specifications.
1 I
Page 2 of 2
_ _ _ _ _ _ _ _ _ _ . - _ _ _ _ _ _ _ ~ _ _ . _ _ . . _ _ . _ _ _ . .. _.
l ST 11.7-196
- 1. Test Number: ST 11.7-196
Title:
Reactor Head Vent System Functional Test
- 2. Description of Change:
This special test involves connecting a hydro pump to the downstream side of the reactor head vent valves RC-V-500 and 597 with the plant in Mode 5 and attempting to pump a limited quantity of water into the reactor vessel head to determine if the valves are opening properly.
- 3. Reason for the Change: I i
This test is to verify the proper operation of reactor head vent valves RC-V-500 1 and RC-V-597. l
- 4. Safety Evaluation. ;
l
- a. This change was safe for the following reasons.
This test is judged to be safe and does not pose an Unreviewed Safety ,
Question based on the following discussion. ST 11.7-196 calls for the l hydro pump to be used for the test to be equipped with a relief valve set at 200 psig. Even if the procedure instruction to stop the hydro pump if a rapid pressure increase is observed is not carried out, the relief valve will limit the potential pressurization to a fraction of the reactor head vent are designed for RCS design pressure and temperature (2485 psig and 650 F). The vessel itself is prevented from over-pressurization by the LTOP System whose capacity far exceeds that of the hydro purnp.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
The design basis boron dilution accident assumes a dilution flow rate of 180 gpm from the primary water pumps. The results of this analysis show that the operators have at least 15 minutes in Modes 1 through 6 to terminate the inadvertent dilution before a loss of shutdown margin occurs, in this test, a hydro pump is used which has a very small flow capacity and a limited water supply volume (5 gallons) limiting the Page 1 of 2
ST 11.7-196 i
maximum total water injection to 5 gallons at a time. The total volume of unborated water which can be added to the RCS during the test is limited i by the procedure to 20 gallons.
l l
The consequences of the maximum dilution resulting from this test is well !
within the bounds of the design basis boron dilution accident. I THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for ,
- this statement is
- !
All failure modes analyzed are either bounded by existing accident analyses, or do not result in equipment malfunctions.
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
i The amount of unborated water added as a result of this test is small in 4
comparison to the design basis boron dilution accident for which tiiere has been shown an acceptable margin of safety. The dilution resulting from the addition of 20 gallons of unbcrated water to the RCS is insignificant
- and thus there is no impact on the margin of safety caused by this test.
l c. This change did not require a change to the Technical Specifications. )
l i
l Page 2 of 2
! I ST 11.7-200 l
! 1. Test Number: ST 11.7-200 I
1
Title:
Underwater Reactor Cavity Hatch Seal Troubleshooting
- 2. Description of Change:
I ,
! The reactor vessel seal ring has four hinged hatchway-type openings that are l
- closed during refueling and sealed with a double gasket arrangement. The !
! hatchways permit water in the upper reactor cavity to drain onto the top of the l supplementary neutron streaming shield. Presently, there is over 200 ml/ min of reactor coolant collecting out of the 1" drip pan drain lines. The suspected j
! cause is believed to be one or both of the North side cavity seal hatches. In !
I order to verify this, the test fittings located between the grooved seal rings will !
I be utilized in a similar manner to the Westinghouse seal test procedure. The
- special test will pressurize this connection with service air at between 45 and l 50 psig. If the subject seal is faulted, the air pressure (being greater than the ~
j 8 psig of cavity water level static pressure) will effectively stop water from i collecting out of the drain line. This test will be done on the seals, one after i another, in order to assess if a seal is leaking. A check of the bolt torque will !
j be done on any hatches found to be leaking. I i
l 3. Reason for the Change:
To troubleshoot the underwater reactor cavity hatch seal. ;
. l j 4. Safety Evaluation:
I
- a. This change was safe for the following reasons:
This test is judged to be safe and does not pose an Unreviewed Safety question. The test will not affect the consequences of a boron dilution accident. Additionally, fuel handling will not occur until the leakage source j is identified and repaired orjustified as being acceptable.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is: !
l The design basis boron dilution accident assumes a dilution flow rate of l 180 gpm from the primary water pumps. The results of this analysis show Page 1 of 2
- , . - - -- --- - . ~. - . . - . . .. .. - . . -
i
- ST 11.7-200
- 5 that the operators have at least 30 minutes in Modes 6 to terminate the i
- inadvertent dilution before a loss of shutdown margin occurs. In this test, .[
j service air will be used to pressurize the seals. Therefore, the l
- introduction of air, possibly into and out of the cavity, does not affect the t
. probability of a boron dilution accident occurring. Addition of water to
- support diving operations is insignificant. Since fuel handling will not
- occur during this test, the fuel handling accident is unaffected. There is therefore no effect on any Chapter 15 analysis. '
i The special test will pressurize the seals to 50 psig. The seals are 4 normally tested on a refueling outage basis during a pressure decay test
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at 45 to 50 psig. The seals are designed to handle this pressurization and can withstand full service air pressure without bolt failure. Therefore, the j probability of hatch seal failure is not increased.
i The test does not introduce any new consequences to a previously evaluated malfunction. Even if a gasket pair were to fail, leakage would be minor due to the extremely small available flow area formed by overlapping metal to metal contact between the hatchway and the seal ring. The hydrostatic pressure caused by the water in the refueling cavity l
would also tend to seat the hatchway against the seal body.
1 i THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A l DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE -
SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for
)
4 this statement is:
The test will not compromise the hatch seals. Therefore, any significant
- increase in leakage resulting from the testing process is considered unlikely. Failure is therefore highly improbable. Therefore, the test does not introduce a new possible malfunction of the seal.
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statemer.t is:
The test does not have any affect on any Chapter 15 accident analysis.
Additionally, the test does not introduce a new malfunction or accident of a different type than previously evaluated. Therefore, the margin of safety is unaffected by the test.
- c. This change did not require a change to the Technical Specifications.
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HADDAM NECK PLANT i SECTION V Exoeriments '
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a There were no experiments performed under the provisions of Title 10, Code of Federal Regulations, Section 50.59 during 1996. :
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HADDAM NECK PLANT SECTION VI Final Safety Analysis Reoort Changes (Page 1 of 1)
- FSARCR Number lille 96-CY-06 Auxiliary Feedwater System Editorial Changes and ,
Clarifications 96-CY-07 Main Steam Line Break Containment Temperature Increase 96-CY-11 Containment isolation Valves identified on the Main Steam l System Outside Containment 96-CY-17 Large Break LOCA Containment Reanalysis 96-CY-18 UFSAR Change Section 13.1 Organizational Structure 96-CY-28 Fuel Handling System Description l
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FSARCR 96-CY-06
- 1. FSARCR Number: 96-CY-06
Title:
Auxiliarv Feedwater System Editorial Changes and Clarifications ,
- 2. Description of Change:
This change revises the Haddam Neck Plant UFSAR for the Auxiliary Feedwater System to incorporate editorial changes into the UFSAR and to clarify inconsistencies in the UFSAR.
- 3. Reason for the Change:
This change was necessary to incorporate editorial changes and clarify inconsistencies in the UFSAR.
i i 4. Safety Evaluation:
- a. This change was safe for the following reasons:
The change was not a hardware change and resulted in the UFSAR being updated to be consistent with the installed system.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF i EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
There is no change to any hardware or operating procedures and therefore does not increase the probability of occurrence or the consequences of any previously evaluated accident.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE )
SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
There is no change to any hardware or operating procedures and therefore does not create the possibility of a new accident. ,
i THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis l for this statement is:
Page 1 of 2
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- FSARCR 96-CY-06 l
1 There is no change to any hardware or operating procedures and }
} therefore does not reduce the margin of safety.
- c. This change did not require a change to the Technical Specifications.
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FSARCR 96-CY-07
- 1. FSARCR Number: ' 96-CY-07
Title:
Main Steam Line Break Containment Temnerature Increnne
- 2. Description of Change:
As part of a containment analysis methodology upgrade program, a change has been made in the amount of metal structures assumed in the containment.
The surface area has been reduced by a factor of approximately eight. This assumption has a significant impact on the main steam line break containment i response. While the containment peak pressure remains below the design pressure, the peak containment temperature has increased from 280*F to ;
366'F. This increase in peak containment temperature has been evaluated for j impact.
- 3. Reason for the Change:
The revised containment analysis changed the equipment qualification temperature profile and needed to be evaluated.
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- 4. Safety Evaluation: ;
- a. This change was safe for the following reasons:
Since containment integrity is still assured and all of the equipment required to mitigate a steam line break inside containment remains operable, the increase in peak containment temperature is safe.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
Equipment qualification was evaluated using a conservative containment temperature profile. The peak containment temperature is assumed to be 3 367 F to bound the accident analysis results. All of the required electric equipment has been shown to be operable with minor exceptions. The cables of concern have been replaced with qualified cables. Since the high temperature peak is of a short duration, the containment structures will remain below the design temperature of 260*F. Therefore there is no Page 1 of 2 l l
FSARCR 96-CY-07 i increase in the probability of occurrence or the consequences of an ,
accident or malfunction previously analyzed. '
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
All of the equipment required to mitigate a steam line break inside
- containment has been shown to be operable.
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY
' TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis l for this statement is:
The containment design requirements for pressure and temperature are still met even with the higher peak containment temperature following a steam line break inside containment. In addition, all of the equipment j required to mitigate the steam line break will continue to be operable.
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- c. This change did not require a change to the Technical Specifications.
i Page 2 of 2
FSARCR 96-CY-11
- 1. FSARCR Number: 96-CY-11
Title:
Containment isolation Valves Identified on the Main Steam System Outside Containment
- 2. Description of Change:
Containment isolation Valve FSAR Table 7.3-1 was revised to include Main Steam and Service Water isolation valves that had been previously omitted by mistake and to correct miscellaneous typographical errors.
- 3. Reason for the Change:
FSAR Table 7.3-1 was revised to more closely meet the intent of 10CFR50 Append:x A GDC 56 & 57.
- 4. Safety Evaluation:
- a. This change was safe for the following reasons:
There was no physical changes to the plant design. The table was updated to reflect actual plant conditions.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
The change did not modify any plant equipment or procedure and therefore did not increase the probability of occurrence or the consequences of an accident previously evaluated.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
This change did not result in any modification to equipment or changes in containment boundary.
Page 1 of 2
FSARCR 96-CY-11 THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis -
for this statement is:
The change did not alter any existing plant design or equipment. The Table documented the existing containment isolation boundary design.
- c. This change did not require a change to the,_ Technical Specifications. l 1
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I FSARCR 96-CY-17 )
- 1. FSARCR Number: 96-CY-17
Title:
Large Break LOCA Containment Reanalvsis I l
- 2. Description of Change: I 4
- The containment analysis methodology upgrade program was performed to l l address a) change in the amount of the metal structure assumed in the l l containment b) justify the reduction of the air side acceptance criteria for the j CAR fans and c) the documentation of the maximum sump temperature used for
[ the RHR heat exchanger performance evaluation. It was determined that while l l the containment peak pressure and temperature (liner temperature) remain l l below the design values of 40 psig and 261*F, the peak containment sump
- temperature has increased from 232*F to 251.2*F. The peak containment
! pressure was calculated to remain at 54.5 psia. The containment building is i designed for a maximum operating pressure of 54.7 psia. Therefore, the peak j pressure remains below the design value with no adverse impact. The peak
- atmosphere temperature was also increased from 271*F to 271.4 F. The peak i atmosphere temperature is bounded by higher allowable temperature associated i with the Main Steam Line Break Analysis. Therefore, only the increase in peak l containment sump temperature has been evaluated for impact.
i 3. Reason for the Change:
- To address component performance at the higher sump temperature.
i
) 4. Safety Evaluation:
I
- a. This change was safe for the following reasons:
Containment integrity is still assured even with the higher containment sump temperature. This is because all of the equipment required to mitigate the consequences of the LOCA inside the containment continue to be operable and the containment peak pressure and temperature (liner temperature) remain below the design value of 40 psig and 261*F. Thus, there is no impact on offsite doses associated with LOCA inside the containment.
- b. This change does not constitute an UNREVIEWED SAFETY GUdSTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF Page 1 of 2
j FSARCR 96-CY-17 l EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
i The containment design requirements for pressure and temperature are still met following the reanalysis of the large break LOCA inside the containment. The new . analysis results essentially confirmed the i
. conservatism of the previous results for peak pressure and temperature.
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The sump temperature increase has been evaluated and found to be
- acceptable. In addition, all of the equipment required to mitigate the LOCA j inside the containment will continue to be operable.
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' THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A i
DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE
! SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for i j this statement is:
i
! Since the change involves an accident analysis assumption only and does
{ not involve any hardware changes, the change cannot create the possibility l
- of an accident of different type.
l THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY 4 l TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis l for this statement is:
. l i The containment design requirements for pressure and temperature are still ,
j met following the reanalysis of the large break LOCA inside the i
- containment. The new analysis results are confirmed to the previous ;
t results for peak pressure and temperature and verified the conservatism of .l that analysis. The sump temperature increase has been evaluated and
. found to be acceptable. In addition, all of the -equipment required to I
mitigate the LOCA inside the containment will continue to be operable.
Thus, there is no reduction in the margin of safety as defined on the basis of anyTechnicalSpecification.
- c. This change did not require a change to the Technical Specifications.
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J FSARCR 96-CY-18
- 1. FSARCR Number: 96-CY-18
Title:
UFSAR Change Section 13.1 Organizational Structure
- 2. Description of Change:
- This summary addresses organizational changes described in Section 13.1 of ,
the Updated Final Safety Analysis Report. The changes were made to reflect I 9
Northeast Utilities organizational changes which have been implemented and I were submitted to the NRC in Revision 19 of the NU Quality Assurance Program.
- 3. Reason for the Change:
The organizational changes were made to integrate the senior management of Northeast Utill ties to provide a more effective organizational and support structure for each of the three (3) Northeast Utilities sites. Additional resources
, were made available to the Haddam Neck Plant through the other Northeast Utilities sites of Seabrook Station and Millstone Point. A further purpose for the change was to provide increased depth of experience -to each site. :
!_ Organizations were integrated around function, rather than being site specific. {
) 4. Safety Evaluation:
1 a. This change was safe for the following reasons:
This change was reviewed for the impact of integrating the overall l corporate organization. The availability of a greater resource and l
- experience base made the change beneficial from a resource utilization perspective and allowed the sharing of experience among the various plants. The organizational change did not make any faciiity change or create the potential for an unreviewed safety question.
f b. This change does not constitute an UNREVIEWED SAFETY OUESTION j because:
i THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF
! EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
The organizational changes contained in UFSAR Section 13.1 do not result in any physical change to the Haddam Neck Plant or change any operating procedure and thus does not impact previously evaluated accidents. The changes have been implemented based on organizational Page 1 of 2
FSARCR 96-CY-18 effectiveness studies and represent an enhancement to the overall organization.
THE POSSIBILITY -FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
The organizational changes did not make changes to the physical plant or to operating procedures and thus do not create the potential for a new unanalyzed accident.
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
The organizational changes made did not make changes to the physical plant or to operating procedures and thus do not have an impact on the margin of safety.
- c. This change did not require a change to the Technical Specifications.
Page 2 of 2
FSARCR 96-CY-28
- 1. FSARCR Number: 96-CY-28
Title:
Fuel Handling System Descriotion
- 2. Description of Change:
This safety evaluation addresses the operation of the Fuel Handling System differently than is described in the FSAR. The FSAR states that fuel is handled under 8 feet of water and the current spent fuel handling tool requires an additional sling to meet the required 8 feet. This evaluation addresses the use of a sling between the hook and the bail of the crane.
- 3. Reason for the Change:
This evaluation addresses the use of a sling between the hook and the bail of the fuel handling crane to maintain the minimum 8 feet of water above the fuel during transit.
- 4. Safety Evaluation:
- a. This change was safe for the following reasons:
The sling to be used for the operation is rated at 4000 pounds, more than twice the weight of the fuel assembly. The handling of the fuel does not introduce a new accident.
The use of the sling has been evaluated against existing accidents and there is no impact.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
The use of a sling does not affect the fuel handling accident previously evaluated.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for j this statement is: 1 Page 1 of 2 l l
FSARCR 96-CY-28 All of the design basis accidents were reviewed for impact by this change.
Component fa!!ure due to either the sling or the spent fuel handling tool will not create an accident which has not been previously evaiuated.
)
l THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY !
TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
The existing accident analysis bounds the use of a sling and the margin of safety has not been reduced.
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- c. This change did not require a change to the Technical Specifications.
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HADDAM NECK PLANT SECTION Vll Technical Reauirements Manual Changes (Page 1 of 1)
TRMCR Numb 2I Iltle C-95-1 Section ll-2 Appendix R Shutdown Related Components (Tables 1 & 3 modifications)
C-95-6 Section ll-2 Appendix R Shutdown Related Components I
(Table 2 modifications) l C-95-8 Modification of the Operability Notes and Compensatory Measures for WR-4, NIS Wide Range Monitor i
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TRMCR C-95-1
- 1. TRMCR Number: C-95-1
Title:
Section 11-2 Aooendix R Shutdown Related Comoonents (Tables 1 & 3 Modfications)
- 2. Description of Change:
This change modifies Section ll-2 Table 1 and Table 3. In Table 1 change the compensatory requirement for verifying the operability and positioning of backup lanterns for the permanent Appendix R emergency lighting units listed in Table 2 from " daily" to " monthly". In Table 3 " Manual Action Equipment Inventory for Appendix R Shutdown", add the following to Control Room '
inventory: "three operable portable lanterns (with shoulder strap).
- 3. Reason for the Change:
Performing a monthly verification of lanterns is a standard test for other units in the plant and is considered adequate.
- 4. Safety Evaluation:
- a. This change was safe for the following reasons: ;
The addition of three portable lanterns with shoulder straps enhances the permanently installed Appendix R lighting system. I
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- b. This change does not constitute an UNREVIEWED SAFETY QUESTION i because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF 1 EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN l THE SAFETY ANALYSIS REPORT. The basis for this statement is:
I The addition of three portable lanterns with shoulder straps enhances the i permanently installed Appendix R lighting system and does not increase I the prolability of occurrence or the consequences of an accident or 1 malfunction of equipment important to safety.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT T(PE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
Page 1 of 2
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TRMCR C-95-1 i
l possibility for an accident or malfunction of a different type than any '
evaluated previously.
l THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
l-I The addition of three portable lanterns with shoulder straps enhances the permanently installed Appendix R lighting system and does not reduce the margin of safety as defined in the Basis for any Technical SpecificaDon.
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- c. This change did not require a change to the Technical Specifications.
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l TRMCR C-95-6 j
- 1. TRMCR Number: C-95-6 1
Title:
Section 11-2 Accendix R Shutdown Related Comoonents (Table 2 Modifications)
- 2. Description of Change:
This change modifies Section Il-2 Table 2 by adding two new additional Appendix R emergency lighting units at " Appendix R - ELU Location". <
- 3. Reason for the Change:
The lighting units were added based on the relocation of feedwater valves FW- ,
V-137-1, 2, 3, and 4 under PDCR 1449 " Modernize Feedwater Controls".
These valves must be accessible during certain postulated Appendix R fire :
scenarios. The lights provide adequate illumination for the operator to perform the task, given a loss of offsite power.
- 4. Safety Evaluation:
- a. This change was safe for the following reasons:
The change updates a table to add two lighting units.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN .
THE SAFETY ANALYSIS REPORT. The basis for this statement is.
)
The updating of Table 2 does not increase the probability of occurrence or 1 the consequences of an accident or malfunction of equipment important to safety.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
The updating of Table 2 does not increase the possibility for an accident or malfunction of a different type than any evaluated previously.
Page 1 of 2
TRMCR C-95-6 THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
The updating of Table 2 does not reduce the margin of safety as defined in the Basis for any Technical Specification.
- c. This change did not require a change to the Technical Specifications.
1 Page 2 of 2
TRMCR C-95-8
- 1. TRMCR Number: C-95-8
Title:
Modification of the Operability Notes and Compensatorv Measures for WR-4 _NIS Wide Range Monitor
- 2. Description of Change:
This change modifies the operability notes and compensatory measures required for WR-4, NIS Wide Range Monitor (source range portion).
- 3. Reason for the Change:
This change was initiated to provide clarification to the operability notes and compensatory measures as defined in Table 1 of Appendix R Shutdown Related Components Instrumentation on page 11.2-13 of the TRM. This change provides clarification of the required actions to be performed when WR-4R is inoperable.
- 4. Safety Evaluation:
- a. This change was safe for the following reasons:
This change does not modify, add, or alter any plant equipment. Proper understanding and evaluation of fires in applicable area have resulted in the simplification of the TRM compensatory measures.
The change does not cause an increase in risk to the public; adversely affect the probability or consequences of previously evaluated accidents or malfunctions; create a new accident or malfunctions; and does not decrease the existing margin of safety.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION j because: !
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is: I This change does not modify, add, or alter any plant equipment.
Therefore, the change does not increase the probability of occurrence or the consequences of an accident or malfunction previously evaluated.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE l Page 1 of 2
TRMCR C-95-8 SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for i this statement is:
The clarifications and compensatory measures do not increase the possibility for an accident or malfunction of a different type than any !
evaluated previously.
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY l TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis )
for this statement is: i
)
The clarifications and compensatory measures do not reduce the margin of safety as defined in the Basis for any Technical Specification.
- c. This change did not require a change to the Technical Specifications.
l Page 2 of 2
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HADDAM NECK PLANT !
i SECTION Vill !
l Technical Soecification Bases Changes t
(Page 1 of 1) ,
PTSCR Number litle C-17-96 Bases Change 3/4.9.15 Spent Fuel Pool Cooling Operation l l
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PTSCR C-17-96 i
- l. 1. PTSCR Number: C-17-96
Title:
3/4.9.15 Scent Fuel Pool Cooling. Refueling Operations Bases-i- Revision
! 2. Description of Change:
! An evaluation of the Spent Fuel Pool Cooling System (SFPCS) was performed i
for CYAPCO License Amendment 188 and determined the cooling system has
- sufficient capacity to maintain bulk pool temperature at, or below 150'F for any i postulated discharge scenario, including a failure of the most efficient spent i fuel pool cooling pump. As a result of this analysis, a fuel handling delay time after shutdown must be imposed before moving fuel assemblies into the spent l fuel pool. -The bases was modified to clarify section 3/4.9.15 regarding single
[ failure considerations.
- 3. Reason for the Change:
This evaluation was performed to support the CY Spent Fuel Pool Rerack Project. A Technical Specification bases clarification was needed concerning
. single failure considerations.
- 4. Safety Evaluation:
- a. This change was safe for the following reasons:
The change does not cause an increase in risk to the public; adversely affect the probability or consequences of previously evaluated accidents or malfunctions; create a new accident or malfunctions; and does not decrease the existing margin of safety.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFEW PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
A TS bases change clarifying single failure considerations for the SFPCS and the related in-core hold time versus pool temperature does not increase the probability of occurrence or the consequence of an accident or malfunction of equipment previously evaluated.
Page 1 of 2 l
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PTSCR C-17-96 THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A
' DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
i
-The change does not introduce the possibility of a new accident or malfunction not previously evaluated. No physical modification to the l facility is made by this change.
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis
] for this statement is:
The spent fuel pool cooling system analysis has demonstrated its ability to accommodate the increased heat load due to the larger storage capacity 1 resulting from the rerack project. Therefore, this change does not involve a reduction to the spent fuel pool's margin of safety.
! c. This change did not require a change to the Technical Specifications.
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Page 2 of 2
4 HADDAM NECK PLANT
- SECTION IX General (Page 1 of 1) _
- GEN Number Title s
- 02 Simultaneous Use of Fuel Building and Yard Cranes
) 03 NUC-EN 104H Testing Motor-Operated Valves using Teledyne QUICKLOOK System j 04 On-Site Storage Container (OSSC) 05 Modification of Spent Fuel Pool Cooling Pump Motors 3
- Thermal Overload Relay Heaters and Circuit Breakers i
, 06 Modification to the Blowdown Flow Calculation on the Plant I
j Process Computer (PPC)
I 07 ABBATTRY-1409-EY R0 Battery Loading I
GEN-01
- 1. General Number: Q1 '
Title:
Continuous Ooeration of the AFW System DC Hydraulic Punig '
- 2. Description of Change:
This change will provide for continuously running of the DC powered AFW hydraulic pumps to supply the motive force for starting the Terry turbine driven AFW pumps, P-32-1 A, B.
- 3. Reason for the Change: !
The purpose for the change is to remove the necessity of considering the starting current of the DC hydraulic pumps in the first minute of any loss of power condition or a station blackout.
- 4. Safety Evaluation: 4 I
- a. This change was safe for the following reasons:
Operation of the DC hydraulic pumps continuously is within the design )
basis and has no adverse affect on safety. The change does not involve a physical modification to the AFW system or the plan as represented in the design basis documents.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because: l THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
i The proposed change in the operating philosophy of running the DC hydraulic pump is within its design basis and has no affect on the possibility or the consequences of previously evaluated accidents or equipment malfunctions.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
There are no new failure modes introduced by this change in operating philosophy.
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GEN-01 ,
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY :
TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is: ;
I The proposed change in operating philosophy is within the capabilities of l r the equipment as documented in the design basis information in PDCR !
1127. The proposed change has no impact on the margin of safety i
because it does not reduce the ability of the AFW system to mitigate an accident.
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! c. This change did not require a change to the Technical Specifications.
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- GEN-02
- 1. General Number: 02
Title:
Simultaneous Use of Fuel Building and Yard Cranes
- 2. Description of Change:
The purpose of this change to demonstrate that the structural integrity of the 1
cranes and support structure will be maintained for an increased load rating of the (10) tons for the Yard Crane, six (6) tons for the North Crane, and three (3) tons each for the South Crane and for the Utility Crane.
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- 3. Reason for the Change:
, This evaluation was prepared to support the crane upgrade needed to lift the
, heaviest high density rack during the proposed spent fuel pool rack project.
i 4. Safety Evaluation:
- a. This change was safe for the following reasons:
The results of the Yard Crane structural analysis confirmed that the :
stresses in the structural components will remain within the required code limits, for the maximum lifting loads considered without any crane
, movement restrictions.
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- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF l EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
l The four cranes affected by the changes proposed under this safety evaluation perform no safety-related function, and failure of these j structures have no impact on previously evaluated accidents or i
malfunctions of equipment important to safety.
- THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is
- '
The changes proposed do not impact movement of heavy loads over existing fuel. The affected cranes do not perform any safety-related
! Page 1 of 2
GEN-02 I
functions and the new load arrangements meet the design structural !
requirements. !
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY -
4 TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis ;
for this statement is: ;
9 The change does not adversely impact the margin of safety as defined in the basis of any technical specification, since the new loading arrangement has been verified to comply with design basis structural requirements. l j
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- c. This change did not require a change to the Technical Specifications.
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l GEN-03 j
- 1. General Number: 03
Title:
NUC-EN 104H RO. Testing Motor-Ocerated Valves using Teledyne QUICKLOOK Svstem i
- 2. Description of Change:
The use of the Teledyne QUICKLOOK software and engineering procedure <
NUC-EN 104H provides another method to acquire and record test data during l diagnostic testing of Motor-Operated Valves (MOVs) and other plant I components. The software provides the ability to view the output of the sensors versus time.
- 3. Reason for the Change:
The software and engineering procedure are being used to improve the capability of the testing program by providing additional test data. :
- 4. Safety Evaluation:
- a. This change was safe for the following reasons: !
4 The QUICKLOOK System components are calibrated and are maintained calibrated in accordance with PORC approved procedures. This system does not cause the MOV or test components to operate differently than described in plant design basis documentation since the system output is not directly connected to plant circuitry. ;
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- b. This change does not constitute an UNREVIEWED SAFETY QUESTION l because:
- THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
I The use of this software does not increase the probability of an accident previously evaluated because the system and procedure do not challenge
- any primary system boundaries, nor does it cause the component under i test to operate in any manner other than as designed, unless the component is placed out of service and operated by PORC approved procedures. The use of this software will not cause any operating system to operate other than according to its design basis.
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GEN-03 THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
The use of this software will not create an accident different than those previously analyzed since the function of the valves, pumps and circuits in the plant are not affected by the use of the diagnostic system.
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
The use of the software and its associated procedure will not reduce the margin of safety as defined in the Technical Specifications since its use does not alter the operation of any system or component different from its design basis.
- c. This change did not require a change to the Technical Specifications.
Page 2 of 2
- _ _ _ . _ .- _ _ . . . _ _ __ . _ _ _ _ _ _ - _ _ _ _ . _ _ . _._. . . m.
GEN-04 i
- 1. General Number: M
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Title:
On-Site Storage Container (OSSC) i
- 2. Description of Change:
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l j This safety evaluation considers the safety impact of storing dry radioactive
{ waste in an On-Site Storage Container, referred to as Culvert 12. This i ;
container is used to temporarily to store spent filter cartridges prior to shipment >
! off-site for disposal.
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- 3. Reason for the Change:
- l
! As a result of a recent NRC audit at Seabrook Nuclear Station, it was i l requested that the practice of storing dry active waste (DAW) in OSSC at CY I i be reviewed in accordance with 10CFR50.59. I I i
- 4. Safety Evaluation: I i {
j a. This change was safe for the following reasons- '
y !
The On-Site Storage Container is a self-contained storage unit with no equipment important to safety to ir.itiate an accident sequence or {
malfunction. The potential release of material stored in the OSSC is j bounded by existing design basis analyses. l
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION i because:
]
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
The On-Site Storage Container is a self-contained storage unit with no equipment important to safety to initiate an accident sequence or malfunction. The potential release of material stored in the OSSC is bounded by existing design basis analyses.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for I this statement is:
Page 1 of 2 j
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GEN-04 .
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" I There are no additional accidents or malfunctions initiated by storage of
- spent filter cartridges in the OSSC.
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- l. THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is: l
. i Since there are no credible accidents that can release radioactivity from the'OSSC, the margin of safety is unaffected. Radiological releases from i - a gaseous radwaste failure accident remains the bounding accident.
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- c. This change did not require a change to the Technical Specifications.
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GEN-05 l l
- 1. General Number: Ei j
Title:
Modifications of Soent Fuel Pool Coolina Pumo Motors Thermal Overload Relav Heaters and Circuit Breakers
- 2. Description of Change:
This evaluation considers the replacement of type 84FH Thermal Overload ;
Relay Heaters with type 82FH heaters, replacement of one 100 ampere type !
HFB breaker with a type HFD breaker, and replacement of a 150 ampere type KB breaker with one 100 ampere HFD.
- 3. Reason for the Change:
This change is being made to resolve concerns over breaker obsolescence and functional degradation and to increase equipment protection and i i
availability without adverse affects on operability margin. !
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- 4. Safety Evaluation:
- a. This change was safe for the following reasons:
J Failure of the equipment affected by this change will not initiate or contribute to the Loss of Offsite Power or Fuel Handling Accidents. The new protective devices adequately protect the pump motor while favoring pump availability.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT.OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is: l i
Failure of the equipment affected by this change will not initiate or contribute to the Loss of Offsite Power or Fuel Handling Accidents. This change enhances equipment circuit protection of the spent fuel pool cooling pump motors. ;
1 THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A l DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE i SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is: j i
Page 1 of 2
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GEN-05 !
I There are no new system level failure modes associated with this change. I This change involves the replacement of protective devices which sense an abnormal motor / feeder condition.
4 THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis i for this statement is:
The margin of safety is not reduced, since the safety limits, as described in the Technical Specifications, are unchanged,
- c. This change did not require a change to the Technical Specifications.
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GEN-06
- 1. General Number: .011
Title:
Modification to the Blowdown Flow Calculation on th? Plant Process Comouter (PPC)
- 2. Description of Change:
A modification to the blowdown flow calculated by the PPC and used in the Calorimetric program has been evaluated for use. Based on actual testing, the measured blowdown flow has higher than previously assumed for a given valve position.
- 3. Reason for the Change:
Measured blowdown flow was greater than assumed for a given valve position.
Higher blowdown flow corresponds to lower actual plant power. Correcting the blowdown flow correlation will allow plant operation at rated thermal power.
- 4. Safety Evaluation:
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- a. This change was safe for the following reasons: ,
The modification to the correlation between blowdown valve position and flow rate does not constitute a change to the previously validated calculational methodology.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION because:
THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQU!PMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
The change has no impact on the design or operation of systems important to safety, nor does it increase the challenges to or degrade performance of safety systems. The change will not alter any assumptions made in the safety analysis.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
Page 1 of 2
. .. . . - . . - - . _ _ . = - - - - - _ . . . . . .- . .-
GEN-06 No new failure modes are introduced by this change, since there is no physical change or procedure change. The methodology used in the determination of core power has not changed.
THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis
- for this statement is
j Changing the blowdown flow control valve position to flow rate correlation does not allow reactor power to exceed its acceptsace limit since the new correlation more accurately reflects actual flow conditions, therefore, the margin of safety is not reduced.
- c. This change did not require a change to the Technical Specifications.
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GEN-07
- 1. General Number: QZ
Title:
ABBATTRY-1409-EY R0 Batterv Loading
- 2. Description of Change:
Calculation No. ABBATTRY-1409-EY, R0 supersedes the station battery "A" and "B" portion of Calculation No. PA-LOE-1171-GE, Rev. 2 and supersedes PA-LOE-1172, Rev. 2 entirely. These calculations address the design basis loading for the CY batteries for LNP/LOCA and SBO scenarios.
- 3. Reason for the Change:
The revised calculation was performed to address the design basis loading for the CY batteries for LNP/LOCA and SBO scenarios.
- 4. Safety Evaluation:
- a. This change was safe for the following reasons:
The change in Connecticut Yankee's station battery loading for the LNP/LOCA and SBO loading profiles was acceptable because the batteries are capable of performing their safety function for any of the !
increased loading intervals as determined by the calculation. 1 Additionally, the 2-hour LNP/LOCA loading profile was bounded by a previously performed surveillance test.
- b. This change does not constitute an UNREVIEWED SAFETY QUESTION ;
because: 1 THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE l OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF ;
EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:
There was no increase in the probability of occurrence of previously evaluated accidents c. malfunction of equipment during the implementation or as a result of this change since this safety evaluation i addressed calculational changes only.
THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:
Page 1 of 2
GEN-07 4
This safety evaluation addressed calculational changes only. The changes in the calculated battery loading was within the capability of the battery to perform its safety function and was bounded by previously performed surveillance testing. This change did not create an accident or malfunction of equipment of a different type than previously evaluated.
" THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:
This change did not negatively impact the basis of the Technical Specifications. The margin of safety was not reduced, since the safety limits and the parameters of the protective boundaries, as described in the j Technical Specifications are unchanged.
4
- c. This change did not require a change to the Technical Specifications.
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i HADDAM NECK PLANT '
SECTION X 4
, Primary Coolant Soecific Activity Analvsis
. There was no primary coolant iodine spiking in 1996 that exceeded the one microcurie l per gram limit set in the Technical Specifications.
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4 HADDAM NECK PLANT l
SECTION XI Challenges to Relief Valves 4
In accordance with the commitment made under item II.K.3.3 of NUREG 0737 (TMI Action Plan) in the W. G. Counsil letter to D. G. Eisenhut, dated June 10,1980, the following is a report of Challenges to Relief / Safety Valves during 1996.
There were no challenges made to the Primary Relief / Safety Valves in 1996.
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HADDAM NECK PLANT SECTION Xil OCCUPATIONAL RADIATION EXPOSURE Regulatory Guide 1.16 Report - 1996 '
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. - - ~ _ _
.- - - .._ - -. - _- - ~.~ - - - . _ . . - . -_ .. - -
i !
PEwPDD20 l Conn. Yankee Atomic Power Co. 26.P b.1997 sweenam j ABG. QUIDS 1.16 REPORT Page Reporting Year s 1996 1 Unit i 1
. 8 Of Personnel a 100 mRes. ....... Total Man.mpen......-
Work & Job runction <
...............................................................................y Station Utilit Contract Station Utility Contract
)
In. Service Inspection ;
- Engineering 1 1 21 3 131 12773 Health Physics 3 0 6 1 0 145
- Maintenance 4 1 75 Operations 0 297 39157 i 1 0 0 6 0 0 3 Supervisory 0 l
1 i
.............................................................................,.............................. 4 Reactor Operations and Surve'111ance 3 0 22 784 Engineering 12 7 41 1644 403 627 '
Health Physics 19 1 34 7653 164 13079 Maintenance 51 110 l 4 7688 108 3328 Operations 36 1 0 12744 172 0 Supervisory 1 3
[
3 139 185 Refueling 85 .... l
, Engineering 4 1 7 79 34 993 Health Physics 5 0 1 88 Maintenance 0 119 33 1 25 3198 1 9551 Operations 8 1 I 0 188 4 0 Supervisory 0 1 1 0 i
0 i
...................................................................................................................... 9 Routine Maintenance t
t Engineering 10 5 19 I
669 346 1443 +
1 Health Physics 15 1 25 145 320 1159 Maintenance 50 95 6863 1 147 1802 j Operations 10 0 0 368
- 0 0 Supervisory 1 3 3 7 217 114 d' ..........................................................................................................................
special Maintenance
{
l Engineering 7 1 18 628
}
I 43 4251 Health Physics 18 0 17 3349 0 1155 Maintenance 31 2 71
{
1195 368 3325
, Operations 5 0 0 206 0 0 j Supervisory 0 3 2 0 378 65 Waste Processing p
t I
i Engineering 0 0 1 0
" 0 0 !
Health Physics 14 1 27 3893 1 4778 ,
i Maintenance 4 0 3 442 0 34 !
Operations 0 0 0
- 0 0 0 !
Supervisory 0 1 0 0 1 0
.......................................................................................................................... {
Job Totals Engineering i 34 15 107 3023 957 20087 Health Physics j 74 3 116 15129 485 20435 Maintenance {
173 9 379 19386 921 63197 1 Operations 60 3
2 0 13512 176 0
, Supervisory 2 12 12 146 1005 e e e e e e el
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e...... ,
- Unit 1 Total 341 41 614 51196 3544 104776 !
........................................eeeeeeeeeeeeeeeeeeeeeeeeeee....eeeeeeeeeeeeeeeeeeeeeeeeeeeeeee............ ........ j
)
Notes Report contains unofficial dose only Dose of persons with c100 mrem 18 NOT INCLUDED 1
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