CY-98-010, Annual Rept for 10CFR50.59,Jan-Dec,1997

Annual Rept for 10CFR50.59,Jan-Dec,1997

ML20203E713
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	12/31/1997
From:	Mellor R CONNECTICUT YANKEE ATOMIC POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
CY-98-010, CY-98-10, NUDOCS 9802270114
Download: ML20203E713 (233)

Text

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l ONNECTICUT YANKEE ATOMIC POWER COMPANY HADDAM NECK PLANT 302 INJUN Hollow ROAD e EAST HAMPToN. CT '10424-30D3 February 23,1998 Docket No. 50 213 CY 98 010 Re: 10CFR50.59 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Haddam Neck Plant AnnuaLRepod Pursuant to the provisions of the 10CFR50.59(b)(2) and 10CFR50.71(e)(2)(ii), this report contains a brief description of any changes, tests, and experiments, including a surnmary of the safety evaluation of each. This report covers operation at the Haddam Neck Plant for the period of January 1,1997 to December 31,1997.

's if you have any questions, please contact Mr. G. P. van Noordertnen at (860) 267 3938.

Very truly yours, CONNECTIC T YANKEE ATOMIC POWER COMPANY

<Q% . k. $

.M J R. A. Mellor \

Vice President - Operations and Decommissioning Enclosure cc: H. J. Miller, Region i Administrator T.L. Frederichs. NRC Project Manager, Haddam Neck Plant /

W. J. Raymond, Senior Resident inspector, Haddam Neck Plant /

9802270114 971231 PDR ADOCK 05000213 I[' /

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Docket No. 50 213 HADDAM NECK PLANT Annual Report l

for 10CFR50.59 r

January 1,1997 through December 31,1997 l

HADDAM NECK PIRI CONTENTS TITLE SECTION Design Change Records (DCRs) l Procedure Changes ll j Jumper Lifted Lead and-Jumper Bypass Changes lll Tests IV Experiments V Final Safety Analysis Report Changes VI Technical Requirements Manual Changes Vil Technical Specification Bases Changes Vill General IX

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HADDAM NECK PLANT SECTION I Dealgn Change Records (DCRs)

(Page 1 of 6)

Safetv Evaluation QCR Number Number lilla DCN DCY 00 045 97 SY EV 97-0008 System Category Determination in a Decommissioned Plant - Turbine Lube Oil System l DCN DCY 00 046 97 SY EV 97 0010 System Category Determination in a Decommissioned Plant - Well Water And Water Treatment System DCN DCY 00-047 97 SY-EV 97-0030 System Category Determination in a Decommissioned Plant - Boron Recovery System DCN DCY-00-048 97 SY EV-97 0029 System Category Determination in a Decommissioned Plant - Leak Monitoring

& Miscellaneous Systems DCN DCY-00 049 97 SY EV 97 0034 System Category Determination in a Decommissioned Plant - Reactor Coolant System DCN DCY-00-050-97 SY EV 97 0033 System Cate9ory Determination in a Decommissioned Plant - Component Cooling System DCN DCY 00 05197 SY EV 97-0027 System Category Determination in a Decommissioned Plant - Primary Sample System DCN DCY-00-052 97 SY EV 97-0032 System Category Determination in a Revs. O and 1 Decommissioned Plant - Safety injection System DCN DCY-01052 97 SY-EV 97 0032 System Categosy Determination in a Rev.1 Decommissioned Plant - Safety injection System

l HADDAM NECK PLANT SECTION I Qasign Change Records (DCRs)

(Page 2 of 6)

Safetv Evaluation DCR Number Number Iltle DCN DCY 00 053 97 SY EV-97 0060 System Category Determination in a Decommissioned Plant - Main Steam System DCN DCY-00-054 97 SY-EV 97 0048 System Category Determination in a Revs. O and i Decommissioned Plant Feedwater &

Condensate System DCN DCY-01-054 97 SY-EV-97-0048 System Category Determination in a Rev.1 Decommissioned Plant - Feedwater &

Condensate System I

DCN DCY-00 055 97 SY EV-97-0050 System Category Deteimination in a Decommissioned Plant Service Water System DCN DCY-00 056 97 SY EV-97-0017 System Category Determination in a Decommissioned Plant - Feedwater &

Condensate Sample System DCN DCY-00 057 97 SY EV-97 0016 System Category Determination in a Decommissioned Plant Circulating Water and Vacuum Priming System DCN DCY-00-058 97 SY EV-97-0037 System Category Determination in a Decommissioned Plant - Chemical &

Volume Control System DCN DCY-00-059 97 SY EV-97-0026 System Category Determination in a Decommissioned Plant - Fuel Oil to Diesel Generator & Auxiliary Steam Boller System DCN DCY-00-060 97 SY-EV-97 0025- System Category Determination in a Decommissioned Plant - Diesel

. Generator Compressed Air System

HADDAM NECK PLANT SECTION I Design Change Records (DCRs)

(Page 3 of 6)

Safety Evaluation DCR Number Number Ilite DCN DCY-00-061-97 SY-EV 97 0055 System Category Determination in a Decommissioned Plant - Primary Ventilation System DCN DCY-00-062 97 SY EV 97-0018 System Category Determination in a Decommissioned Plant - Closed Cooling Water System DCN DCY-00-063-97 SY EV 97-0036 System Category Determination in a Decommissioned Plant - Residual Heat Removal System DCN DCY-00-064 97 SY EV 97 0059 System Category Determination in a Decommissioned Plant - Liquid Waste System DCN DCY-00 065-97 SY-EV-97-0071 System Category Determination in a Decommissioned Plant - Gaseous Waste System DCN DCY-00-066 97 SY-EV 97-0043 System Category Determination in a Decommissioned Plant - Primary Water System DCN DCY-00-067-97 SY-EV 97-0022 System Category Determination in a Decommissioned Plant - Turbine Building Waste Water Treatment System DCN DCY-00-068 97 SY-EV 97-0046 System Category Determination in a Decommissioned Plant - Spent Fuel Pit Cooling System DCN DCY-00-069 97 SY-EV 97-0041 System Category Determination in a Decommissioned Plant - Service Air System

i HADDAM NECK PLANT j SECTION I j Design Change Records (DCRs) i j (Page 4 of 6) l Safety Evaluation

[ DCR Number Number Iltle DCN DCY 00 070 97 SY EV 97 0052 System Category Determination in a

} Decommlasioned Plant - Control Air -

System I

DCN DCY-00-07197 SY EV 97-0040 Gystem Category Determination in a

Decommissioned Plant Reactor Containment Control Air System

! DCN DCY-00 072 97 SY EV 97 0042 System Category Determination in a ,

Dewmmissioned Plant - Nitropen System i DCN DCY-00-073 97 SY-EV 97-Ofl12 System Category Determination in a Decommissioned Plant - Hydrogen Gas i System DCN DCY-00-074 97 SY EV-97-0014 System Category Determination in a

Decommissioned Plant - Miscellaneous Bottled Gas System DCN DCY-00-075-97 SY-EV-97 0013 System Category Determination in a

Decommissioned Plant - Hydrogen Dryer System 1

! DCN DCY-00-076 97 SY-EV 97-0045 System Category Determination in a Decommissioned Plant - Fire Protection System DCN DCY-00 077-97 SY-EV 97-0015 System Category Dettermination in a Decommissioned Plant - Post Accident Sampling System

DCN DCY-00-078-97 SY-EV 97 0062 System Category Determination in a

! Decommissioned Plant- Heating Steam

{ & Condensate System n

3

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1 inDDAM NECK PLANT

$ SECTION I Design Change Records (DCRs) 1 (Page 5 of 6)

Safety Evaluation l

Q.CR Number Number Iltle

' l I

DCN DCY-00-079 97 SY-EV 97-0064 System Category Determination in a Decommissioned Plant - Switchgear Building HVAC System DCN DCY 00-080 97 SY EV 97-0049 System Category Determination in a Decommissioned Plant - Floor, Roof and Equipment Drain System i

DCN DCY 00 08197 SY EV-97 0063 System Category Determination in a Decommissioned Plant - Septic System DCN DCY 00-082 97 SY EV 97 0024 System Category Determination in a Decommissioned Plant - Domestic Water System

DCN DCY-00-083 97 SY EV 97-0010 System Category Determination in a Decommissioned Plant - Hydrazine Feed System PDCR 1435 SE 96-009 Replacement of Battery Charger BC 1-1 A; Revised Safety Evaluation PDCR 1459 CY-SE 97-004 Remove SW V-849 and SW-V-850 PDCR 1553 CY-SE-1553 Service Water MIC Chemical Injection POCR 1587 SE 1587a & b Control Room Modifications PDCR 1598 SY-EV-97-0005 Main Feedwater Pump Motor Replacement - Early Release B DCR CY-97001 SY EV-97-0002 Removal of Main Feedwater Isolation Valves, Revision A DCR CY-97002 SV EV 07-0003 Check Valve Addition to Service Water

_ Supply to Spent Fuel Heat Exchangers l

HADDAM NECK PLANT SECTION I Design Change Records (DCRs)

(Page 6 of 6)

Safety Evaluation DCR Number Number lille DCR CY 97003 CY-SE.97 008 Deletion of Reactor Related Accidents from the Licensing and Design Basis of CY DCR CY 97004 SY EV 97-0009 Domestic Water pH Treatment System DCR CY 97005 SY-EV 97-0031 Replacement of the Above Ground Diesel Rev.O Storage Tank DCR CY 97006 SY EV 97-0038 Installation of Test Holes M SFB, PAB &

NSGB Ventilation Ductwoik DCR CY 97007 SY-EV 97 0044 Revision to CYAPCO HNP UFSAR Section 15,5 DCR CY 97-008 SY EV 97 0054 Auxiliary Particulate and lodine Sample System DCR CY 97013 SY-EV-97-076 Sodium Hypochlorite Tank Replacement DCR CY 97013 SY-EV 97-076 Sodium Hypochlorite Tank Replacement Rev.1 DCR CY 97992 SV-EV 97-0003 Check Valve Addition to Service Water Supply to Spent Fuel Heat Exchangers

N OCW DCY 00 o45 97 sY- wN ooos

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Attachment 1 I"S" I 'I I NADDAM NECK PLANT Q Annual Report Summary of Changes Mark the Appropriate choice:

Design Change Y Setpoint change f Test Tech Requirements Manual Change Experiment i Procedure Change _

Jumper Bypass

1. Change Number: Revision Nwaber:

Title:

System catecery Determinatien in a Decem4mmiened Plant -

Turbine tube Oil System ,

2. Description of Change:

This Safety Evaluation addresses the system category determination  ;

for the Turbin) Lube 011 System, as shown on P&ID No. 16103-26001, Sheets 1 to 5. The majority of this system is being " Abandoned *,

however, portions of the system will remain *Available".

3. Reason for the Change:

The Turbine Lube oil System provides lubricating oil for the main turbine. Since the main turbine is no longer being used, this system is being " Abandoned". The only portions of the system that are not being "Aban nned" are components within the turbine oil tank room, which are currently being used for storage and transport of waste oil, and adjacent outside oil truck conaectionc. These components must remain *Available" until all was.te oil disposal _needs are complete.

4. Safety Evaluation

a. This change was safe for the following reasons:

Per this 10CFR50.59 Safety Evaluation Review, the change does not ,

affect the probability or consequences of previously evaluated accidents or malfunctions of equipment important to safety.

There is no reduction in the margin of plant safety.

b. This change does not constitute AN UNREVIEWED SAFETY QUESTION because '

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I DCN pc6 00- 045 -97 s 3y.w .97- oooa ik9 s 2 4 0

[- THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CCNSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT l TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is: ,

CY is in a defueled condition and has notified the NRC of its  !

decision to permanently cease power operation. Fuel has been permanently removed from the reactor vessel. As such, much of the equipment which previously had a safety related and/or operational function during normal plant operation, may no longer be important to safety or required to be operable. This includes the Turbine Lube Oil System. Previously during normal plant operation, low turbine oil pressure resulted in a turbine autostop signal (AST) which also resulted in a reactor trip. With the plant in a defueled condition, this function is no longer required and a malfunction of the AST is of no consequence. In addition, the Turbine Lube Oil System does not directly communicate in any way with any system containing radioactivity, the 8 pent Fuel Pool or the Spent Fuel Handling System. Therefore, the probability of occurrence or consequences of a fuel handling accident or a radwaste system failure, which are the only accidents applicable to the defueled condition, is unaffected.

THE POSSIBILITY FOR AN ACCIDF.NT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT HEEN CREATED. The basis for this statement is:

The operation and required function of the *Available" Turbine Lube Oil components is unchanged from previous operation and function. Therefcre, there is no possibility of a different type of malfunction. The " Abandoned" components no longer have a required function in the plant and, therefore, cannot malfunction.

Also, there is no possibility of an accident of a different type created because the Turbine Lube Oil System does not directly.

communicate in any way with any system containing radioactivity, the Spent Fuel Pool or the Spent Fuel Handling System. The abandoned portion of the system includes the turbine lube oil coolers which were cooled by Service Water (SW), with fire protection (FP) water as emergency backup, during normal plant operation. SW presently provides cooling to the Spent Fuel Pool (SFP). However, a postulated failure of thase coolers er the interfacing SW lines, which are isolated free the SW headers and fire main.,could not affect the SW line, which prerently provides flow to the SF heat exchanger, or the FP system.

THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The tasis for this statement ist

% Q(.Y . 00 045

• 97

.< -kV = 97* 0008 l  %>o "Available" and ' Abandoned" syntems are those which no longer have j/n j Tech Spec requirements associated with them. The Turbine Lube oil

(.) system, never had any direct Tech Spec requirements associated with it for any mode of plant operation. Indirectly, low turbine oil pressure would result in a turbine trip signal which would also trip the reactor. Tech Spec 3/4.3.1 includes reference to a turbine trip, but is applicable to Mode 1, i.e., it has no applicability for a defueled reactor. Therefore, there is no impact on the margin of safety as defined in the Bases of any Tech Spec.

c. This change did not require a change to the Technical Specificatic.ns .

In addition, a 10CFR50.82 Decommissioning Review was also performed that concluded that this change does not constitute an unreviewed decommissioning question for the follovin, reasons:

Placing the Turbine Lube Oil System in the "Available" and

" Abandoned" category does not foreclose (preclude) release of the site for possible unrestricted use. There is no potential failure mode which could create a major spill or rupture, causing a significant radioactive contamination of the soil or buildings which could not easily be decontaminated. It also does not permanently install or retain major structures or equipment on site, or render it difficult to remove systems, structures or components from the site.

Placing the Turbine Lube oil System in the "Available" and

' Abandoned" category does not result in a significant environmental L impact not previously reviewed. It does not result in the discharge of radion,:tivity or chemicalo, either by a controlled release, or due to a malfunction or failure of equipment. It also does not affect the site environment (e.g., terrain, noise, visual appearance, transmission lines) in any way.

Placing the Turbine Lube oil System in the *Available" and

' Abandoned" category does not remove reasonable _ assurance that adequate funds will be available for decommissioning. The cost of procedure revisions, UFSAR revisions, etc., taken in total, are not likely to jeopardize decommissioning funds.

There is no applicability to the PSDAR of placing the Turbine Lube oil System in the 'Available" and ' Abandoned" category since that document has not yet been developed for CY. Rather, the system categorization process will be used as input to the PSDAR development.

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t;cw pcY. 00 - 046- 97 sY- EV oot9 ,

Attachment 1 "0"'*

HADDAM NECK PLANT Annual Report Summary of Changes Mark the Appropriate Choice Dr. sign Change Y Serpoint Change Test Tech Requirements Manual Change Experiment ProceLore change Jumper Bypass

1. . Change Number: Revision Number:

Title:

System Category Determination in a Decomminaioned Plant -

Well Water and Water Treatment synlam i

2. Description of Change:

This Safety Evaluation addresses the system category determination i for the Well Water and Water Treatment System, as shown on P&ID No.

16103-26003, Sheets 1 and 2. The majority of this system will remain

• Available" except for

• Abandoned" items which include: 1) the *C"

- and *D" wells and pumps; 2) water supply lines to the hydrazine & ETA

./ feed-tanks, to the caustic system, to _ the service water system, and to the circulating water pumps emergency seal water supplys and 3) an abandoned water supply line from the condensate pumps.

3. Reason for the Change:

The Well Water and Water Treatment System will remain 'Available" to supply water for both plant and personnel activities. The wells and water treatment truck supply purified water to various tanks. This water is *Available v for spent fuel pool makeup, boiler makeup, and other' decommissioning activities such as reactor coolant system (RCS) decontamination. This system is also *Available" for alternative cooling of the closed cooling water heat exchangers. Furthermore, this system supplies all domestic water needs such as showers, toilets, and sinks. The *C" and *D" wells and pumps are being

• Abandoned" because they have experienced either operational or water impurity problems in the past and are not needed. The *A" and *B" wells, which will remain 'Available", provide sufficient water capacity for plant needs including peak manpower levels. Other portions of the system are being

• Abandoned" because its use is no longer required.

4. SafetyEva}uations

a. This change was safe for the following reasons:

e kcN pcY 046 - 9 7 SY- EV 0019

, Pkge 2 el 3 5 Per this 10CFR50.59 Safety evaluation Review, ,

hange does not affect the probability er consequences of prevfously evaluated accidents or malfunctions of equipment important to safety.

There is no reduction in the margin of plant safety.

b. This change does not constitute AN UNRTVIEWED SAFETY QUESTION because THERE IS NO INOREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MAtFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is CY is in a defueled condition and has notified tha NRC of its decision to permanently cease power operation. Fuel has been permanently removed from the reactor vessel. As such, much cf the equipment which previously had a safety related and/or operational function during normal plant. operation, may no longer be important to safety or required to be operable. This includes the Well Water and Water Treatment System. The operation and required function of the system components which will remain 'Available" is unchanged from previous operation and function. Therefore, the probability of occurrence of a previously evaluated malfunction is also unchanged. The

• Abandoned" components no longer have a g required function in the plant and, therefore, cannot malfunction.

In addition, the Well Water and Water Treatment System does not directly communicate in any way with any system containing radioactivity, the Spent Fuel 'ool or the Spent Fuel Handling System. Therefore, the probability of occurrence or consequences of a fuel handling accident or a radwaste system failure, which are the only accidents applicable to the defueled condition, is unaffected.

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 ist The operation and required function of the Well Water and Water Treatment System components which will remain "Available" is unchanged from previous operation and function. Therefore, there is no possibility of a different type of malfunction. The

" Abandoned" components no longer have a required function in the plant and,'therefore, cannot malfunction. Also, there is no possibil'ity of an accident of a different type created because the Well Water and Water Treatment System does not directly communicate in'any way with any system containing radioactivity, the Spent Fuel Pool or the Spent Fuel Handling System. The

'g " Abandoned" Water Treatment cross-tie to the service water (SW) system will be isolated. This portion of the SW system, which supplies cooling water to the main exciter coolers,.is also being

DcN DcY 046 - 97 sy - EV oot9 Ne3 JS

" Abandoned" and is not required for spent fuel pool (SFP) heat exchanger cooling. Therefore, there is no postulated failure of a this Water Treatment to SW interface that could adversely affect

. 7 the SW flow to the SFP heat exchanger.

b]

THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION RAS NOT BEEN REDUCED. The basis for this statement is:

L -

"Available" and " Abandoned" systems are those which no longer have Technical Specification requirements associated with them. The Well Water and Water Treatment System never had any Technical Specification requirements associated with it for any mode of 5 plant operation. Therefore, there is no impact on the margin of safety as defined in the Bases of any Technical Specification,

c. This change did not require a change to the Technical Specifications.

In addition, a 10CFR50.82 Decommissioning Review was also performed h that concluded that this change does not constitute an unreviewed decommissioning question for the following reasons:

Placing the Well Water and Water Treatment System in the "Available" and " Abandoned" category dos.' not foreclose (preclude) release of the p site for possible unrestrie rem use. There is no potential failure mode which could create a majo; spill or rupture, causing a significant radioactive contamination of the soil or buildings which could not ca:!Iy be decontaminated. It also does not permanently install or retain major structure s or equipment on site, or render it difficult to reme-'u sya ems, structures or components from the site.

_- Placing the Well Wat and Water Treatment System in the "Available" and " Abandoned" category does not result in a significant environmental impact not previously reviewed. It does not result in the discharge of radioactivity or chemicals, either by a controlled release, or uue to a malfunction or failure of equipment. It also does not affect the site environment (e.g., terrain, noise, visual appearance, transmiss on lines) in any way.

Placing the Well Water and Water Treatment System in the Available" and " Abandoned" category does not remove reasonable assurance that adequate funds will be available for decommissioning. The cost of procedure revisions, UFSAR revisions, etc., taken in total, are not likely to jeopardize decommissioning funds.

There is no applicability to the PSDAR of placing the Well Water and Water Treatment System in the "Available" and " Abandoned" category

'g since that document has not yet been developed for CY. Rather, the system categorization process will be used as input to the PSDAR

,_ development.

_ _ . . . . 2

, DcN pcY- 00 047- 97 ACP 1.2 2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Repor.

Page 1 ofI Safety Evaluation Nutnber: SY EV.97 0030 Revision: O Document Number: ENG 17156 At mehment IM for Baron Recoverv tvrtem Revision: N/A Document

Title:

Svitem c ,ecorv Determination in a Decommiednned Plant Boron Reenverv System Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Brief Description of Change and Safety Evaluation Sommary:

This Safety Evaluation addresses the system category determination for the Boron Recovery System, as shown on P&lD No.

16103 26004, Sheets I through 4. Ponions of this system will either remain "Available" or be placed in " Lay up" to suppen plant decommissioning activities and reactor coolant system (RCS) decontamination. Portions of this system are being

" Abandoned" because they are no longer needed.

The operation and required function of the "Available" components is unchanged from previous operation and function.

Herefore, the probability of occurrence or consequences of a previously evaluated malfunction is unchanged and there is no possibility for a malfunction of a different type. ne " Lay-up" components will not be functioning while in that state and, therefore, a malfunction is not possible. The "Acandoned" components no longer have a required function in the plant and cannot t.'W,metion. The Primary Sample System does not directly communicate with the Spent Fuel Pool or the Spent Fuel

. H.cim W em. The "Available" or " Lay up" portions of the system which are connected to a system containing C radcemir re unchanged. The " Abandoned" ponions of the system will be isolated from any system containing rademniw Therefore, the probability of occurrence of a fuel handling accident or a radwaste system failure, which are the only accioents applicable to the defueled state,is unaffected and there is no possibility for an accident of a different type.

This change does not foreclose (preclude) release of the site for possible unrestricted use. It does not result in a significant environmental impact not previously reviewed. It does not remove reasonable assurance that adequate funds will be available for decommissioning.

This change is safe and does not involve an unreviewed safety question. The change does not affect the probaNiity or consequences of previously evaluated accidents or malfunctions of equipment important to safety. There is in, r. .ction in the margin of plant safety, in addition. this change does not involve an unresolved decommissioning question.

2. Reason for the Change:

The "Available" ponions of the Boron Recovery System will be used for processing boric acid mix tank inventory and inventory from the RCS and the cavity, in addition, interfaces with bordering systems (primary water, venti!ation and liquid waste) will remain "Available" for consistency with the categories of the interfacing systems. The only portion of the Borc,n Recovery System which is being placed in " Lay-up" is an interface with the gaseous waste system, which is in " Lay up" to suppon the RCS decontamination. The boron recovery evaporators, coolers, rebollers and associated pumps and interconnecting piping are no longer needed and are being " Abandoned". In the defueled condition, the boron recovery process is no longer needed and is not performed. Boron has not been recovered for a number of years and has been discharged according to permitting laws. Ponions of the Boron Recovery System that are being "A&doned" have not been used since 1991. The boron waste storage tank heaters were previously " Abandoned"in place.

Preparer hA sa ara (J / AA A. m Date OS - / 5 - 9 '7 26000. DOC

- . .. -. - - - .- - - . - - . - - - - - - ~

~

pcN ocy- oo- o48 -97 ACP 1.2 2.42 Rev.1 MAJOR O ror. 3. io cra so.5, <2) a.go,<

Page1of1 Safety Evaluation Number: SY.EV.97 0029 Revision: 0 Document Number: ENG 1.7156 Attachment 112 for I enk Monitorine and Miscellaneous Svitems Revision: N/A Document

Title:

. System Catenorv Determination in a Decommits}; ,ed Plant . I enk Monitorine & Miscellaneous Systemt i

d k rovide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Brief Description of Change and Safety Evaluation Summary:

nis Safety Evaluation addresses the system category determination for the Leak Monitoring and Miscellaneous Systems, as shown on P&lD No. 16103 26005. Portions of this system will remain "Available" to support plant decor.imissioning

activities and other portions of this system are being " Abandoned" because they are no Ic.yer needed.

The operation and required function of the "Available" components is unchanged from previous operation and function.

Therefore, the probability of occurrence or consequences of a previously evaluated malfunction is unchanged and there is no possibility for a malfunction of a different type. The " Abandoned" components no longer have a required function in the plant and cannot malfunction. The Leak Monitoring and Miscellaneous Systems does not directly communicate with the Spent Fuel Pool or the Spent Fuel Handling System. The "Avallable" portions of the system which are connected to a system containing radioactivity are unchsnged. The " Abandoned" portions of the system will not directly communicate with p any system containing radioactiviry. Therefore, the probability of occurrence of a fuel handling accident or a radwaste

, v system failure, which are the only accidents applicable to the defueled state,is unaffected and there is no possibility for an

accident of a different type.

This change does not foreciose (preclude) release of the site for possible unrestricted use, it does not result in a significant environmental impact not previously reviewed, it does not remove reasonable assurance that adequate funds will be available for decommissioning.

This change is safe and does not involve an unreviewed safety question. The change does not affect the probability or conseque:nces of previously evaluated accidents or malfunctions of equipment important to safety. There is no reduction in

, the margin of plant safety, in addition, this change does not involve an unresolved decommissioning question.

2. Reason for the Change:

The reactor cavity seal tell-tale drain system is being maintained "Available" to allow prompt detection of reactor cavity seal leakage. This system will remain "Available" until the reactor cavity is drained down. The reactor containment sumps and in. core instrumentation sumps collect miscellaneous leakage, drainage or spills in containment. These sumps and associated pumps and interconnecting piping will remain "Available" to provide a means for collecting this leakage and transferring it to the PAB. The neutron water shield tank system will remain "Available" to provide shielding during the decommissioning effort. The P.72 reactor containment pressure sensing lines will remain "Available" for containment pressure monitoring in the event the equipment hatch is reinstalled. The Leak Monitoring System monitors containment pressure which, in tum, actuates a reactor trip. Since reator trip actuation is no longer required, most of the Leak M(.nitoring System, with the exception of the P 72 reactor containment pressure sensing lines is being " Abandoned". The Containment Valve Stem Leakoff System was previously " Abandoned" in place.

Preparer D.A .s A n rA AJ / h & 'm Date #S=G~97 260oS3. DOC

ACP 1.2 2.42 Rev.1 MAJOR p Form 3 - 10 CFR 50.59 (b)(2) Report b Page1of1 Safety Evaluation Number: SY EV.97 0034 Revision: 0 Document Number: ENG 17.1M Attachment 12 2 for Reactor Coolant System Revision: WA Document

Title:

Svstem Catecorv Determintion in a Decommittiened Plant . Reactor Coolant System Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Brief Description of Change and Safety Evaluation Summary:

This Safety Evaluation addresses the system category determination for the Reactor Coolant System (RCS), as shown on P&!D No.16103 26007, Sheets I through 3. De majority of this system will be placed in " Lay.u,," to support plant decommissioning activities and RCS decontamination. A small portion of the system will remain "Available". Portions of this system are being " Abandoned" because they are no longer needed.

De operation and required function of the "Available" components is unchanged from previous operation and function.

Therefore, the probability of occurrence or consequences of a previously evaluated malfunction is unnanged and there is no possibility for a malfunction of a different type, ne "Available" portions of the RCS, which intrface with the sampling, gaseous waste, and nitrogen supply systems, do not change the operation of those interfacing systems in any way. De

" Abandoned" portions of the RCS will either be isolated from the rest of the system, which contains radioactivity, or will be j " Abandoned" in place to maintain RCS pressure boundary, he majority of the RCS will be maintained in wet " Lay.up" and will not be functic g. This will not change the processing of any system leakage and w::1 have no affect on the radioactive l

g)

) (~ waste system. Therefore, the probability of occurrence of a fuel handling accident or a radwaste system failure, which are l the only accidents applicable to the defueled state, is unaffected and there is no possibility for an accident of a different type, l

i This change does not foreclose (preclude) release of the site for possible unrestricted use, it does not result in a significant environmental impact not previously reviewed. It does not remove reasonable assurance that adequate funds will be available for decommissioning.

His change is safe and does not involve an unreviewed safety question. De change does not affect the probability or consequences of previously evaluated accidents or malfunctions of equipment important to safety. There is no reduction in the margin of p . safety. In addition, this change does not involve an unresolved decommissioning question.

2. Reason for the Change:

The RCS previously circulated pressurized water through the reactor vessel and the four loops, being heated as it pures through the core. It next flowed to the steam generators, where the heat was transferred to the secondary (steam) syswn , and then back to the pumps to repeat the cycle. With the plant in its current state, the function of the RCS is no longer required.

He majority of the RCS is being maintained in " Lay.up" to permit use during the decentamination process. The only portions of the RCS that will remain "Available" are interfaces with the sampling, gaseous waste, and nitrogen supply systems. He hot leg sample line will remain "Available" to support water chemistry monitoring of the RCS while in " Lay-up". The nitrogen system will remain "Available" for adding nitrogen Whe RCS as needed. The interface with the gaseous waste system will remain "Available" for venting. The " Abandoned" portions of the RCS include selected instrumentation, the head and pressurizer vent SOVs, and the air supply to the PORVs. Rese items will not be used during the RCS decontamination process.

[ ] Preparer

'D . A . s A A F A o / hfh Date c6-o9-97 260073. DOC l

I l

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_____J

DCN pcY CEO - 97

[ . ACP 1.2 2A2 Rey,1 MAJOR Form 3 10 CFR 50.59 (b)(2) Report r~.

Page1of1 Safety Evaluation Number: EY.EV.97.0011 hlevision: 0 DoeumentNumber: ENG 1.7.1 M A* ak= a* 19 9 for ra-aaa a* caalia. E*- Revision: N/A DoctEDent

Title:

Eve' = c

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Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Brief Description of Change and Safety Evaluation Summary:

his Safety Evaluation addresses the system category determination for the Component Cooling System, as shown on P&ID No.16103 2600s, Sheets 3 through 7. Portions of this system will either remain "Available" or be placed in " Lay.up" to support plant decommissioning activities and reactor coolant system (RCS) decontamination. Portions of this system are being " Abandoned" because they are no longer needed.

De operation and required function of the "Ava!!able" componenu is unchanged from previous operation and function.

Derefore, the probability of occurrence or conuquences of a previously evaluated malfunction is unchanged and there is no possibility for a malfunction of a different type. De " Lay up" components will not be functioning while in that state and, therefore, a malfunction is not possible, ne " Abandoned" components no longer have a required function in the plant and cannot malfunction. The Component Cooling System does not directly communicate with the Spent Fuel Pool or the Spent Fuel Handling System. De "Available" or" Lay up" portions of the system which are connected to a system containing i radioactivity are unchanged and are already addressed in the radweste system failure event. De " Abandoned" portions of the system will be isolated from any system containing radioactivity. Therefore, the probability of occurrence of a fuel handling accident or a radwaste system failure, which are the only accidents applicable to the defueled sute, is unaffected and there is no possibility for an accident of a different type, his change does not foreclose (preclude) release of the site for possible unrestricted use, it does not result in a significant environmental impact not previously reviewed. It does not remove reasonable assurance that adequate funds will be available for decommissioning.

His change is safe and does not involve an unreviewed safety question. De change does not affect the probability or consequences of previously evaluated accidents or malfunctions of equipment important to safety. Dere is no reduction in the margin of plant safety, in addition, this change does not involve r.n unresolved decommissioning question.

2. Reason for the Change:

The Component Cooling System is a closed loop cooling system that transfers heat from reactor auxiliaries to the service water system during plant operation and during normal cooldown/ shutdown modes. Most of the system will be placed in

" Lay up" to support RCS decontamination, ne only portions that are "Available" are the component cooling tank relief valve discharge path to the reactor containment sump and interfaces with the primary water and primary ventilation systems.

' The " Abandoned" portions of the system include cooling of the boric acid evaporator components, waste evaporator components, containment penetrations, neutron shield tank cooler, and the component cooling surge tank overflow line. This portion of the system is being " Abandoned" since the associated components will either no longer be used or no longer require cooling, ne balance of the system is being maintained in " Lay up" to facilitate operation of other" Lay up" systems that will be used during the future RCS decontamination.

Preparer h A 5 A rU"M / -'e"-4-, Date o -to - 9 7 260083. DOC

r DCN pCf 091- 97

/ ACP 1.2 2.42 Rev. I MAJOR

(

"] Form 3 - 10 CFR 50.59 (b)(2) Report Page1of1 Safety Evaluation Number: SY-EV 07.0027 Revision: 0 Document Number: ENG t 71% Atinehment 112 for Primary Samole System Revision: N/A Document

Title:

Svttem Catecorv Determination in a Decommittiened Plant Primprv Samole Svstem Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Brief Description of Change and Safety Evaluation Summary:

Dis Safety Evaluation addresses the system category determination for the Primary Sample System, as show 16103 26009, Sheets I and 2. Portions of this system will either remain "Available" or be placed in " Lay plant decommissioning activities and reactor coolant system (RCS) decontamination. Portions of this sys

" Abandoned" because they are no longer needed.

The operation and rmuired function of the "Available" components is unchanged from previous operation and fun Derefore, the probability of occurrence or consequences of a previously evaluated malfunction is unchanged an possibility for a malfunction of a different type. The " Lay up" components will not be functioning while in that state therefore, a tr'alfunction is not possible, The " Abandoned" components no longer have a required function in cannot malfunction. The Primary Sample System does not directly communicate with the Spent Fuel Pool or the

/_) Handling System. De "Available" or" Lay up" portions of the system which are connected to a system conta

" radioactivity are unchanged. The " Abandoned" portions of the system will be isolated trom any system con radioactivity. Therefore, the probability of occurrence of a fuel handling accident or a radwaste system failure, whi only accidents applicable to the defueled state, is unaffecte3 and there is no possibility for an accident of a This change does not foreclose (preclude) release of the site for possible unrestricted use it does not result in environmental impact not previously reviewed. It does not remove reasonable assurance that adequate funds will be available for decommissioning.

His change is safe and does not involve an unreviewed safety question. The change does not affect the proba consequences of previously evaluated accidents or malfunctions of equipment important to safety. There is no reduction in the margin of plant safety. In addition, this change Joes not involve an unresolved decommissioning question.

2. Reason for the Change:

The Primary Sample System provides for laboratory analysis to evaluate the chemical and radiological makeup of plant liquids and gases, he system is designed to be operated manually, on an intermittent basis, under conditions from full power operations to cold shutdown. The "Available" portions of the system include the hot leg sample line and the neutron water shield tank sample line. De hot leg sample line will remain "Available" to support water chemistry monitoring RCS while in " Lay up" The neutron water shield tank sample line is "Available" to support water chemistry samp neutron water shield tank, which was classified as "Available" to provide additional shielding during the initial phases of decommissioning. The " Abandoned" portions of the system include the steam generator chemistry monitoring pa chemical pump and tanicthat were previously abandoned in place. Steam generator water chemistry monitoring is required. Therefore, the steam generator chemistry monitoring panels are no longer needed. The balance of the sys being maintained in " Lay up" to permit future sampling during the RCS decontamination process.

Preparer bA < ARE A LI l M: n Date 0 9 -I2 ~97 260091 DOC

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pChi DCY cS2- 97 ACP 1.2 2A2 Rev.1 MAJOR

{ Form 3 - 10 CFR 50.59 (b)(2) Report Page l of 1 Safety Evaluation Number: S%EV 97-00M Revision: 0 Document Number: ENG 1.7-1M Attachment IM for Saferv iniection Svetem . Revision: N/A Document

Title:

Svstem r='enorv Determination in a Decommittiened Plant - Safety Iniection System Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Brief Description of Change and Safety Evaluation Summary; nis Safety Evaluation addresses the system category determination for the Safety injection System, as shown on P&lD No.

16103 26010, Sheets I and 2. Ponions of this system will remain " Operable","Available" or be placed in " Lay up" to support plant decommissioning activities and reactor coolant system (RCS) decontamination. Ponions of this system are being " Abandoned" because they are no longer needed.

The operation and required function of the " Operable" and "Available" components is unchanged from previous operation and function. Therefore, the probability of occurrence or consequences of a previously evaluated malfunction is unchanged and there is no possibility for a malfunction of a different type, ne only " Operable" ponion of the Safety injection System is the RWST which is an attemate source of makeup water to the spent fuel pool. The functions of the "Available" ponions of the system do not change. The " Abandoned" ponions of the system will be isolated from any system containing

. radioactivity. The portion of the system being maintained in " Lay up" will not be used until the RCS decontamination Q, process. Therefore, the probability of occurrence of a fuel handling accident or a radwaste system failure, which are the only accidents applicable to the defueled state,is unaffected and there is no possibility for an accident of a different type.

This change does not foreclose (preclude) release of the site for possible unrestricted use. it does not result in a significant

environmental impact not previously reviewed. It does not remove reasonable assurance that adequate funds will be available for decommissioning.

This change is safe and does not involve an unreviewed safety question. The change does not affect the probability or consequences of previously evaluated accidents or malfunctions of equipment imponant to safety. There is no reduction in the margin of plant safety, in addition, this change does not involve an unresolved decommissioning question.

I

2. Reason for the Change:

The Safety injection System, which is part of the ECCS previously delivered borated water to the reactor vessel for cooling .

the core following a LOCA. The only peition of the Safety injection System being maintained " Operable"is the RWST which is an alternate source of makeup water to the spent fuel pool. The NRC's SER for approval of License Amendment 188 stated that makeup water to the spent fuel pool can be provided from the RWST. The "Available" ponion of the system includes the portions of the system that will be used to provide spent fuel pool and reactor cavity purification functions and to provide avalable flowpaths for reactor cavity draindown activities. The " Lay-up" portion of the system includes the portion of the system that will be used to facilitate future RCS decontamination and water processing. This portion of the system includes the LPSI pumps and safety injection pump seal leakoff pump, and interconnecting piping; piping to and from the HPS! pumps; and interconnecting piping with CVCS and RCS Core Deluge. The " Abandoned" portions of the system include the HPSI pumps, the safety injection measuring tank, and interconnecting piping.

Preparer B . A . S A A P A AJ / M"-MM Date e 7- U-9 7 260101 DOC i

4 W W(- 01-052 - 97 ALP 1.2-2A2 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page l of t Safety Evaluation Number: SY E V 97-00 M Revision: 1 Document Number: ENG 1.71R Attachment 1M for Safety Inlectinn Sv= tam Revision: N/A Document

Title:

. Svetem r'atenorv Detenninstinn in a Deenmmlatinned Plant - Saferv Iniectinn Svitem .

Provide a brief description of the change and a summary of ths Safety Evaluation in the format below.

1. Brief Description of Change and Safety Evaluation Summary:

This Safety Evaluation addresses the system category determination for the Safety Idection System, as shown on P&ID No.

16103 26010, Sheets 1 and 2. Portions of this system will either remain "Available" or be placed in " Lay-up" to suppon l

plant decommissioning activities and reactor coolant system (RCS) decontamination. Pordons of this system are being

" Abandoned" because they are no longer needed.

The operation and required function of the "Available" components is unchanged from previous operation and function. The RWST will be drained and repaired prior to being used in support of cavity draindown and other decommissionia activities.

Therefore, the probability of occurrence or consequences of a previously evaluated malfunction is unchanged and tnere is no possibility for a malfunction of a different type. The " Lay-up" components will not be functioning while in that : tate ano, therefore, a malfunction is not possible. The " Abandoned" components no longer have a required function in the plant and cannot malfunction. The Safety injection System does not directly communicate with the Spent Fuel Pool or the Spent Fuel

. Handling System. The "Available" or " Lay up" ponions of the system which are connected to a system containing radioactivity are unchanged. The " Abandoned" portions of the system will be isolated from any system containing radioactivity. Therefore, the probability of occurrence of a fuel handling accident or a radwaste system failure, which are the only accidents applicable to the defueled state, is unaffected and there is no possibility for an accident of a different type.

This changc does not foreclose (p eclude) release of the site for possible unrestricted use. It does not result in a significant environmentalimpact not previously reviewed. It does not remove reasonable assurance that adequate funds will be available for decommissioning. This change is safe and does not involve an unreviewed safety question. The change does not affect the probability or consequences of previously evaluated accidents or malfunctions of equipment important to safety. There is no reduction in the margin of plant safety. In addition, this change does not involve an unresolved decommissioning question.

2. Reason for the Change:

The Safe.ty Irdection System, which is part of the ECCS, previously delivered borated water to the reactor vessel for cooling the core following a LOCA. The RWST, previously maintained as " Operable" as an attemate source of makeup water to the Spent Fuel Pool (SFP), has been recategorized as "Available". Per SY-EV 97-0103, the DWST is now the source of makeup water to the SFP for evaporative losses when PWST or RPWST are unavailable. Procedure AOP 3.2 59 has been revised to delete the RWST as a source of makeup water and to add the DWST in its place. The RWST will be used in support of cavity drain down and other decommissioning activities. The other "Available" portion of the system will be used to provide spent fuel pool and reactor cavity purification functions and to provide available flowpaths for reactor cavity draindown activities. The " Lay-up" portion of the system will be used to facili' ate future RCS decontamination and water processing. This portion of the system includes the LPSI pumps and safety injection pump sealleakoff pump and piping; piping to and from the HPSI pumps; and interconnecting piping with CVCS and RCS Core Deluge. The " Abandoned" portions of the system include the HPSI pumps, the safety injection measuring tank, and interconnecting piping.

Preparer . t 4 s 4 a r4 e / > m '_---_ - Date d 2. - M - 9 7 260103-t. DOC

__ _ _ _ _ __.. _ _ _ 7 _ _ _ .

DC N DC.Y- 00 093-97

. ACP 1.2 2.42 Rev.1 MAJOR I'

Form 3'- 10 CFR 50.59 (b)(2) Report Page1of1 l Safety Evaluation Number: EY EV 07 00'o Revision: 0

! Document Number: ENG 1%114 A*=^=' tS 9 far u.6 *^ te Revision: N/A Document Thie: Ev==='a c=*a-e-v Th'-laadaa in a TS= =!==!=C:-.: . u.6 e -- . s.n.

i Provide a brief i W of the change and a summary of the Safety Evaluation in the format below.

1. Brief Description of Change and Safety Evaluation Summary:

Dis Safety Evaluation addresses the system category determination for the Main Steam System, as shown on P&ID No.

16103 26012, Sheets 1 through 11. Portions of this system will either remain "Available" or be placed in " Lay up" to i support plant decommissioning activities and reactor coolant system (RCS) decontamination. Most of this system is be

" Abandoned" because it is no longer needed.

De operation and required function of the "Available" components is unchanged from previous operation and function, ne

• Lay up" components will not be functioning while in that state and, therefore, a malfunction is not possible, he

" Abandoned" components no longer have a required function in the plant and cannot malfunction. Therefore, the pr

of occurrence or consequences of a previously evaluated malfunction is unchanged and there is no possibility for a malfunction of a different type, ne Main Steam System does not directly communicate in any way with any system containing radioactivity, the Spent Fuel Pool or the Spent Fuel Handling System. Derefore, the probability of occurrence of a fbel handling accident or a radwaste system failure, which are the only accidents applicable to the defbeled state,is unaffected and there is no possibility for an accident of a different type, his change does not foreclose (preclude) release of the site for possible unrestricted use, it does not result in a significant environmental impact not previously reviewed, it does not remove reasonable assurance that adequate funds will be j

available for decommissioning.

This change is safe and does not involve an unreviewed safety question. De change does not affect the probability or consequences of previously evaluated accidents or malfunctions of equipment important to safety. Dere is no reduction in the margin of plant safety. In addition, this change does not involve an unresolved decommissioning question.

2. Reason for the Change:

2 1

-- Most of the system is being " Abandoned". De system is no longer needed since there will be no further power generation.

Mg)cr components being "Abendoned" include: the main steam piping downstream of the non ruturn valves, except for i annospheric steam dump and atmospheric vent lines; the main steam drum (36" SHP-601 7); the turbine main stop trip b valves; the high pressure turbine; the steam crossover piping; the moisture separator reheaters; the low pressure turbines; the high pressure steam dump spargers; the low pressure steam dump manifold; the steam generator blowdown cooler; the Tetry l

turbines; the steam air e,)ection subsystem; and the turbine shaft gland sealing subsystem. De *Available" portions of the i

system include boundaries with the following "Available" systems control air, beating steam & condensate, liquid waste, primary ventilation, feedwater, circulating water, and service water. The " Lay-up" portion includes the main steam lines hom the steam generators to the non-return valves, the atmospheric vent lines, the atmospheric steam dump, the steam 4

' generator sample coolers and part of the steam generator blowdown system. The secondary side of the steam generators will be kept filled during the decontamination process.

1 i

preparer b. A . sAMAx / ' ^ ^ W -- Date es-ke-*?

• 1 260123. DOC 1

i

E

, pcN pcy oM - 97 ACP 1.2 2.42 Rev.1 MAJOR

.C Form 3 - 10 CFR 50.59 (b)(2) Report Page1of1 Safety Evaluation Number: sv.Fv.97.oo4a Revision: o Document Number: ENO 1.7 1 M Ave = Ament 12 2 for Feedwater & Cand.name. Sven*= Revision: N/A Document

Title:

Svitem ra'annrv Determinatinn in a Decommia<innad Plant Feed-ater a cand nene. Svetem Provide a brief description of the change and a summary of the Safety Evaluation in the format below,

1. Brief Description of Change and Safety Evaluation Summary:

his Safety Evaluation addresses the system category determination for the Fudwater & Condensate System, as P&lD No.16103 26013, Sheeu 1 through 13. Portions of this system will either remain "Available" or be p up" to support plant decommissioning activities and reactor coolant system (RCS) decontamine:

being " Abandoned" because it is no longer needed.

' ion. Mos The operation and required function of the "Available" componenu is unchanged from previous operation and fu

- " Lay up" components will not be functioning while in that state and, thmfore, a malfunction is not possible, ne

" Abandoned" components no longer have a required function in the plant and cannot malfunction. Der of occurrence or consequences of a previously evaluated malfunction is unchanged and there is no possibility malfunction of a different type. De Feedwater & Condensate System does not directly communicate in an system containing radioactivity, the Spent Fuel Pool or the Spent Fuel Handling System. Derefore, the probability Q

L occurrence of a fuel handling accident or a radwaste system failure, which are the only accidents applicable state, is unaffected and there is no possibility for an accident of a different type.

Dis change does not foreclose (preclude) release of the site for possible unrestricted use, it does not result in a

- environmental impact not previously reviewed. It does not remove reasonable assurance that adequate funds will be available for decommissioning.

His change is safe and does not involve an unreviewed safety question. De change does not affect the probab consequences of previously evaluated accidents or malfunctions of equipment important to safety. Dere is no reduction in the margin of plant safety. In addition, this change does not involve an unresolved decommissioning question.

2. Reason forthe Change:

Most of the system is being " Abandoned". He system is no longer needed since there will be no further power gen Major components being "Abendoned" include: low pressure feedwater heaters, steam jet air ejecLrs, gland steam condenser, condensate pumps, gland seal water pumps, main feedwater pumps, feedwater heater drains tank, feedwater heater drain purnps, moisture separator rebesters, and reheater drains receiver tanks, he "Available" portions of include the feedwater heater relief valve vent stack and drain, DWST, CST, a nitrogen supply line to the CST, and boundaries with other"Available" systems, ne vent stack and drain provides a relief valve discharge path for the hot w storage tank, and a flowpath to the discharge tunnel for service water and well water exiting the closed cooling water heat exchangers. The DWST and CST will store purified water to be used throughout the decommissioning effort. De ni supply line to the CSTwill provide corrosion protection of supplied systems. He " Lay up" portion includes the electric auxiliary feedwater pump, the chemical addition tank, and the feedwater hnes from the electric auxiliary feedwater to and including the secondary side of the steam generators. De secondary side of the steam generators will be ke during the decontamination process. He electric auxiliary steam generator feedwater pump will provide steam gener GVand makeup; and the chemical addition tank will be used for water chemistry cont ol as necessary.

Preparer + h@ =

B. A TAM 4 AJ / Date od-to-97 260133 DOC

,t

/ PM DcY 654-97 ACP 12-2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page 1 of1 Safety Evaluation Number: sv Ev.97 004s Revision: 1 Document Number: ENG 1.7-15& Artschment 1M for Feedwater & Condennte Svetem Revision: WA l Document

Title:

. System mienorv Determ'mation in a Deenmmhtinned Plant Feedwater & Cntdemate Svstem Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Brief Description of Change and Safety Evaluation Summary:

This Safety Evaluation addresses the system category determination for the Feedwater A Condensate System, as shown on P&lD No.16103 26013, Sheets I through 13. Portions of this system will either remain " Operable","Available" or be l placed in " Lay up" to support plant decommissioning activities and reactor coolant system (RCS) decontamination. Most of this system is being " Abandoned" because it is no longer needed.

The demineralized water storage tank (DWST) and its level instrumentation is being maintained " Operable" to allow the tank accomplish iu function as an attemate source of makeup water to the spent fuel pool (SFP). His portion of the system will be maintained in full compliance with pre-established procedures, surveillances, preventative maintenance procedures and programs, as applicable. The operation and required function of the "Available" components is unchanged from previous operation and function. The " Lay up" components will not be functioning while in that state and, therefore, a malfunction is not possible. The " Abandoned" components no longer have a required function in the plant and cannot malfunction.

Therefore, the probability of occurrence or consequences of a previously evaluated malfunction is unchanged and there is no possibility for a malfunction of a different type. The Feedwater & Condensate System does not directly communicate in any way with any system containing radioactivity, the SFP of the Spent Fuel Handling System, oder than the DWST makeup to the SFP. Therefore, the probability of occurrence of a fuel handling accident or a radwute system failure, which are the only accidenu applicable to the defueled state, is unaffected and there is no possibility for an accident of a different type.

This change does not foreclose (preclude) release of the site for possible unrestricted use. It does not result in a significant environmental impact not previously reviewed. It does not remove reasonable assurance that adequate funds will be available for decommissioning. This change is safe and does not involve an unreviewed safety question. The change does not affect the probability or consequences of previously evaluated accidents or malfunctions of equipment important to -

safety. There is no reduction in the margin of plant safety. In addition, this change does not involve an unresolved decommissioning question.

2. Reason for the Change:

The majority of the Feedwater & Condensate System will be " Abandoned". The system is no longer needed since there will be no further power ge tration. He portions of the system which will remain "Available" include the feedwater heater relief valve vent stack and drain, condensate storage tank (CST), and boundaries with other systems. The CST will store purifie,d water to be used throughout the decommissioning effort. The nitrogen supply line to the CST will provide corrosion protection of supplied systems. He " Lay-up" portion includes the electric auxiliary feedwater pump, the chemical addition tank, and the feedwater lines from the electric auxiliary feedwater pump up to and including the seconday side of the steam generators. The electric auxiliary steam generator feedwater pump will provide steam generator fill and makeup; and the chemical addition tank will be used for water chemistry control as necessary. The DWST has been recategorized from "Available" to " Operable" to wpport it's use as an attemate source of makeup water to the SFP, per SY EV 97 0103.

Procedure AOP 3.2 59 has been revised to delete the RWST as a source of makeup water and to add the DWST in its p' ace.

Preparer - D. A . c4 m*4 AJ / / e ? Im Date / 2 9 ~7 26ol331. DOC 1

- _ m m

pc N pcy - Oo - 055 97 ACP 1.2 2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page1of1 Safety Evaluanon Number: EY EV 97 0050 Revision: 0 Document Number: ENG 17-156 A**lu===' 17 $ for Service Water Sv="

Revision: N/A Document

Title:

Eva'a N=arv Deter'ala='i= In a Ne- inianed Plant - Eervice Water Svm Provide a brief description of the change and a summary o'tv 2afety Evaluation in the format below,

l. Brief Description of Change and Safety Evaluation Summary:

his Safety Evaluation addresses the system category determination for the Service Water System, as shown on P&!D No.

16103 26014, Sheets I through 9. Portions of this system will remain " Operable", "Available" or be placed in " Lay up" to support plant decommissioning activities and reactor coolant system (RCS) decontamination. Portions of this system are being " Abandoned" because they are no longer needed, r i

The operation and required function of the " Operable" and "Available" components is unchanged from previous operation and function. Derefore, the probability of occurrence or consequences of a previously evaluated malfunction is unchanged i

l and there is no possibility for a malfunction of a different type. De " Operable" portion of the Service Water System will support maintenance of the spent fbel in a safe condition, ne functions of the "Available" portions of the system do not change. De " Abandoned" portions of the system will be isolated from any system containing radioactivity, ne portions of the system being maintained in " Lay up" will not be used until the RCH.acontamination process. Derefore, the probability cf occurrence of a fuel handling accident or a radweste system failure, which are the only accidents applicable to the defueled state, is unaffected and there is no possibility for an accident of a different type. This change does not foreclose (preclude) release of the site for pouible unrestricted use. It does not result in a significant environmental impact not i

previously reviewed. It does not remove reasonable assurance that adequate funds will be available for decommissioning.

This change is safe and does not involve an unreviewed safety question, ne change does not affect the probability or consequences of previously evaluated accidents or malfunctions of equipment important to safety. Dere is no reduction in

. the margin of plant safety, in addition, this change does not involve an unresolved decommissioning question.

2. Reason for the Change:

De Service Water System supplies cooling water to both the primary and secondary plants. The " Operable" portions of the system include: the four service water pumps; both service water supply headers; the main return header; the spent fuel pool cooling heat exchanger supply and return lines; the emergency diesel supply and return lines; a supply line to the fire pump diesel; the Adams Filters; supply and return lines for radiation monitors R 1g and R-19; and a supply line from fire

• protection which is " Operable". The portions of the system that will remain "Available" include the screenwash subsystem; the hypochlorite subsystem; the circulating water purip seal supply piping; the Bulab chemical injection subsystem; the closed cooling heat exchanger supply and return piping; the turbine building supply headers; the air conditioning units supply and return piping; the component cooling water heat exchanger supply and return piping; the boric acid mix tank vent eaadaaa=* supply and return piping; the primary drains tank vent candaawr supply and return piping; and the supply and return lines for CAR Fans 1 and 2. De portion of the synom that is in " Lay-up" includes the boundary with the component cooling system which has also been categorized as " Lay up". De portions of the system being " Abandoned" include: the circulating water hypochlorite dilution lines; a chemical addition line; selected instrumentation; a boundary with service air;

' wi supply and return piping for the turbine oil coolers, the hydrogen seal oil coolers, the generator hydrogen coolers, the -

,olated phase bus coolers, the main exciter coolers and associated Kinney filter, the boren recovery overhead condenser, the steam generator blowdown tank condensers, the chlorine analyzer, the waste wr,i~,ew overhead condenser, the steam generator sample chiller condensers, CAR Fans 3 and 4 cooling coils and motor coolers, and the main steam sample cooler.

Peuparer n & -CLw Date eA-2o97 260143. DOC -

gN pC,Y - 00 e o66 - 97 sy. ev pi7 Attachment 1 Pas s I af 3 RADDAM NECK PLANT s

Annual Report Summary of Changes Mark the Appropriate Choice:

Design Ch uge Y Setpoint Change Test Experiment Tech Requirements Manual Change Procedure Change Jumper Bypass

1. Change Number:

Revisien Number: - -

Titles svatem catecrorv Determinatien in a Do ct mi amiened plant -

Egedwater and cendenante Emmnle Evatem

2. Description of change:

This Safety Evaluation addresses the system category determination for the Feedwater and Condensate Sample System, aw shown on P&ID No.

I 16103-26016, Sheet 1.

The majority of this system is being

' Abandoned", however, portions of the system will remain "Available".

3. Reason for the Change:

The Feedwater and Condensate Sample System is used to sample water impurities and corrosion products within the feedwater and condensate system.

The feedwater and condensate system is being ' Abandoned",

consequently there is no further need for the Feedwater and condensate Sample System. . Therefore, the entire Feedwater and condensate Sample System is being ' Abandoned" except for the main condensers which will remain "Available" for atructural integrity of the circulating water system.

4. Safety Evaluation:

a. This change was safe for the following reasons:

Per this 10CFR50.59 Safsty Evaluation Review, the change does not affect the probability or consequences of previously evaluated accidents or malfunctions of equipment important to safety.

There is no reduction in the margin of plant safety.

b. This change does not constitute AN UNREVIEWED SAFETY QUESTION because o

DCW PCT- 00 -0% -97 SY - DV 0017 '

Phge 2 ef 7

, THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE

, , CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPOR TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:

CY is in a defueled condition and has notified the NRC of its decision to permanently cease power operation. Fuel has been permanently removed from the reactor vessel.. As such, much of the equipment which previously had a safety related and/or operational function during nomal plant operation, may no longer be important to safety or required to be operable. This includes the Feedwater

, and Condensate Sample System. The operation and required function of the system components which will remain 'Available" is unchanged from prsvious operation and function. Therefore, the probability of occurrence of a previously evaluated malfunction is also unchanged. The ' Abandoned" components no longer have a required function in the plant and, therefore, cannot malfunction.

In addition, the Feedwater and Condensate Sample System does not i

directly communicate in any way with any system containing radioactivity, the Spent Fuel Pool or the Spent Fuel Handling System.

Therefore, the probability of occurrence or consequences of a fuel handling accident or a radwaste system failure, which are the enly accidents applicable to the defueled condition, is unaffected.

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 operation and required function of the 'Available" Feedwater and Condensate Sample components is unchanged from previous operation and function.

different type of malfunction.

Therefore, there is no possibility of a The ' Abandoned" components no longer have a required function in the plant and, therefore, cannot malfuncticn. Also, there is no possibility of an accident of a different type created because the Feedwater and Condensate Samplo System does not directly comunicate in any way with any system containing radioactivity, the Spent Fuel Pool or the Spent Fuel Handling System.

THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement isi

'Available" and ' Abandoned" systems are those which no longer have Technical Specification requirements associated with them. The Feedwater and Condensate Sample System, dever had any Technical

' Specification requirements associated with it for any mode of plant operation. Therefore, there is no impact on the margin of safety at defined in the Bases of any Technical Specification.

.---- --- - - - - - - - - - - - - - - - - - ~ ~

4 pcW ocY- oo - 056 - 97 W- av 0087 Pa y 3d3

/ c. This change did not require a change to the Technical Specifications.

1 In additi0n, a 10CFR50.82 Decornissioning Review was also performed that concluded that this change does not constitute an unreviewed-decommissioning question for the following reasons:

Placing the Feedwater and Condensate Sample System in the 'Available" and ' Abandoned" category does not foreclose (preclude) release of the site for possible unrestricted use. There is no potential failure mode which could create a major spill or rupture, causing a significant radioactive contamination of the soil or buildings which could not easily be decontaminated.- It also does not permanently install or retain major structures or equipment on site, or render it difficult to remove systems, sentetures or components from the site.

Placing the Feedrater and Condensate Sample System in the *Available" and ' Abandoned" category does not result in a significant environmental impact not previously reviewed. It does not result in the discharge of radioactivity or chemicals, either by a controlled release, or due to a malfunction or failure of equipment. It also does not affect the site environment (e.g. , terrain, noise, visual appearance, transmission lines) in any way, f Placing the Feedwater and Condensate Sample System in the 'Available" and ' Abandoned" category does not remove reasonable assurance that adequate funds will be available for decomissioning. The cost of procedura revisions, UFSAR revisions, etc., taken in tocal, are not likely to jeopardize decomissioning funds.

There is no applicability to the PSDAR of placing the Feedwater and condensate Sample System in the 'Available" and ' Abandoned" catego.mf since that document has not yet been developed for CY. Rather, the system categorization process will be used as input to the PSDAR development.

O' b 4 4 g #

• 0c9 Dr.y 057 - 97 g . av - 97 . co4

/ Attachment 1 ge i *I 3

/ XADDAM NECK PLANT m Annual Report U Summary of Changes Mark the Ap n opriate Choice:

Design Change Y Setpoint Change Test Tech Requirements Manual Change Experiment Procedure Change Jumper Bypass

1. Change Number: Revision Number

Title:

Syntam cat accrv Det emination in a Decanuni smiened Plant -

Circulating Water and Vacuum Driming system

2. Description of Change:

This Safety Evaluation addresses the system category determination for the Circulating Water and Vacuum Priming, as shown on PAID No.

15103-26017. The majority of this system will remain "Available",

however, portions of the system are being " Abandoned".

3. Reason for the Change:

The majority of the Circulating Water System will remain 'Available" since it is still required for dilution flow to support releases controlled by the REMODCM. The system will be operated intermittently and most often with only two pumps operating at once.

The vacuum Priming System is required to support circulating water pump operation and likewise will be operated intermittently. The only portions of the Circulating Water and Vacuum Prining System that are being " Abandoned" are two lines to the air ejectors. These lines are being " Abandoned" since the air ejectors, which are part of the main steam system, are being

• Abandoned".

4. Safety Evaluations

a. This change was safe for the following reasons:

Per this 10CFR50.59 Safety Evaluation Review, the change does not affect the probability or consequences of previously evaluated accidents or malfunctions of equipment important to safety.

There is no reduction in the margin of plant safety.

l

b. This change does not constitute AN UNREVIEWED SAFETY QUESTION because ,

ocN pcY - oo - os7- 97 sY - EV - 9 7 - o o n G Mu3a 2 .I 3 THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE D CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement ist CY is in a defueled condition and has notified the NRC of its l decision to permanently cease power operation. Fuel has been permanently removed from the reactor vessel. As such, much of the

, equipment which previously had a safety related and/or operational function during normal plant operation, may no longer be important to safety or required to be operable. This includes the circulating Water and Vacuum Priming System. The operation of the system components which will remain 'Available" is unchanged from previous operation and function. Therefore, the probability of occurrence of a previously evaluated malfunction is also unchanged. The

• Abandoned" components no longer have a required function in the plant and, therefore, cannot malfunction. In addition, the Circulating Water and Vacuum Priming System does not directly communicate in any way with any system containing radioactivity, the Spent Fuel Pool or th9 # pent Fuel Handling System. Therefore, the probability of ocepr ance or consequences

of a fuel handling accident or a radwaste (fr.2cm f ailure, which l are the only accidents applicable to the defueled condition, is unaffected.

) 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 operation and required function of the Circulating Water and vacuum Priming System components which will remain 'Available* is unchanged from previous operation and function. Therefore, there is no possibility of a different type of malfunction. The

" Abandoned" components no longer have a required function in the plant and, therefore, cannot malfunction. Also, there is no pcssibility of an accident of a different type created because the Circulating Water and Vacuum Priming System does not directly communicate in any way with any system containing radioactivity, the Spent Fuel Pool or the Spent Fuel Handling System. The gland seal water supply to the circulating water pumps is from the service water (SW) system through 1-in lines. This portion of the SW system is 'Available" and is not required for spent fuel pool (SFP) heat exchanger cooling. In the current defueled state, the SW system flow requirements are so low that the component cooling heat exchangers are needed as phantom loads to allow the SW pumps to operate in the preferred range above 2000 gpm. Therefore, any postulated failure of these 1-in SW lines or the circulating water pumps could not adversely affect the required flow rates to the

(~)

V. SFP heat exchanger.

l

Dct4 9CY 057 - 97 Sy . EV - 9 7 - 0 014

,. Pago 3 d3

/ t THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL L SPECIFICATION MAS MOT BE2N REDUCED. The basis for this statement is:

'Available" and ' Abandoned" systems are those which no longer have Technical Specification requirements associated with them. The Circulating Water and vacuum Priming System, never had any Technical Specification requirements associated with it for any mode of plant operation. Therefore, there is no impact on the margin of safety as defined in the Bases of any Technical Specif, cation.

c. This change did not require a change to the Technical l

Specifications.

In addition, a 10CFR50.82 Decommissioning Review was also performed that concluded that this change does not constitute an unreviewed decommissioning question for the following reasons:

Placing the Circulating Water and Vacuum Priming System in the

• Available" and ' Abandoned" category does not foreclose (preclude) release of the site for possible unrestricted use. There is no potential failure mode which could create a major spill or rupture,

, causing a significant radioactive contamination of the soil or

, buildings which could not easily be decontaminated. It also does not permanently install or retain major structures or equipment on site, or render it difficult to remove systems, structures or components from the site.

Placing the Circulating Water and Vacuum Priming System in the "Available" and " Abandoned" category does not result in a significant environmental impact not previously reviewed. It does not result in the discharge of radioactivity or chemicals, either by a controlled release, or due to a malfunction or failure of equipment. It also does not affect the site environment (e.g., terrain, noise, visual appearance, transmission lines) in any way. I Placing the Circulating Water and Vacuum Priming System in the "Available" and ' Abandoned" category does not remove reasonable assurance that-adequate funds will be available for decommissioning.

The cost of procedure revisions, UFSAR revisions, etc., taken in total, are not likely to jeopardize decommissioning funds.

There is no applicability to the PSDAR of placing the Circulating Water and Vacuum Priming System in the 'Available" and ' Abandoned" category since that document has not yet been developed for CY.

Rather, the system categorization process will be used as input to the PSDAR development.

OcN pcY o58 - 97

, ACP 1.2 2.42 Rev.1 MAJOR f' Form 3 10 CFR 50.59 (b)(2) Report # * "'

Page 1 of1 '

SafetyEvaluation Number: SY EV 07-0017 Revision: 0 Document Number: ENG 171M A~h==a' 13 9 for ch mint a vn6m. r2. w1 sv=.. Revision: WA Document

Title:

Evem r~enev N-rmiamian in a P= mi==taa d Plant rh=i=1 & Vab= raaerni tvm Provide a brief description of the change and a summary cf the Safety Evaluation in the format below.

1. Brief Description of Change and Safety Evaluation Summary:

Disshown as Safety Evaluation on P&ID No, addresses the system category determination for the Chemical and Volume Control Syste 16103 26018 Sheets 1 through 6. Portions of this system will nmain " Operable","Available" or be placed in " lay up" to support plant decommissioning activities and reactor coolant system (RCS) decontamination. Por of this system are being " Abandoned" because they are no longer needed.

De operation and required function of the " Operable" and "Available" components is unchanged from previou and function. brefore, the probability of occurrence or consequences of a previously evaluated malfunction is un and there is no possibility for a malfunction of a different type. De " Operable" portion of CVCS will support mainten of the spent fuelin a safe condition, he functions of the "Available" portions of CVCS do not change, ne " Abandoned" portions of CVCS will be isolated from the rest of the system, which contains radioactivity. De remainder of the CVCS wil be maintained in " Lay up" and will not be used until the RCS deconomination process. Thenfore, the probability

+ occurrence of a fuel handling accident or a radwaste system failure, which are the only accidents applicable to the defueled state,is unaffected and there is no possibility for an accident of a different type.

This change does not foreclose (preclude) release of the site for possible unrestricted use. It does not result in a signi environmental impact not previously nyiewed. It does not remove reasonable assurance that adequate funds will be available for decommissioning.

His change i. safe and does not involve an unreviewed safety question. The change does not affect the probability consequences of previously evaluated accidents or malfunctions of equipment important to safety. There is no reduction in the margin of plant nfety, in addition, this change does not involve an unresolved decommissioning question.

2. Reason for the Change:

De CVCS maintains the required volume of borated water, maintains the correct boron concentration, and maintains the desired chemistry and purity in the RCS to accomodate all modes of operation. CVCS also provided a means for injecting borated water into the RCS following a LOCA. The only portion of the CVCS that will be " Operable" is the purification flow loop to the spent fud pit cooling system. The portions of the CVCS that will remain "Available" include a purifi sti flowpath to the RCS and cavity via the safety injection system. The mmaining portions of the CVCS that are "Available" are boundaries with adjacent "Available" systems including primary water, primary ventilation, gaseous waste, and liquid De " Abandoned" portion of the CVCS includes the compressed air bottle for LD TV 230 and the charging makeup flow transminers which are no longer needed. He balance of the CVCS, which includes the "A" and "B" charging pumps, metering pump, volume control tank, regenerative and non-regenerative heat exchangers, reactor coolant pump seal wate er beat exchang# , chemical addition tank, boric acid mixing subsystem, and associated piping, is being maintained in " L to support the future RCS decontamination.

V Pnparer -.

b.A 54me w / ^=k, Date os- 2 9-9 7 26oll3. DOC 4

Deu pcy. 60 o59 - 97 sy. cv 0026

'1 Attachment 1 Pa98 Id 4 BADDAM NECK PLANT

.-[ Annual Report Summary of Changes Mark the Appropriate Choice:

Pesign Change Y Setpoint Change Test Experiment Tech Requirements Manual Change Procedure Change Jumper Bypass

1. Change Number:

Revision Number:

Title glem Catecorv Detemination in a Decommismiened Plant -

Fuel Oi1 to Diesel Generater & Auv411aq' 9 team BM 1er Q'ntem

2. Description of Change:

This for Safety Evaluation addresses the system category detemination the Fuel Oil to Diesel Generator and Auxiliary Steam Boiler System, as shown on P&ID No. 16103-26020, Sheet 1 The fuel oil supply to the two emergency diesel generators (EG-2A and EG-2B) will remain " Operable" and the fuel oil supply to the air cooled diesel (EG */) and to the auxiliary steam boilers will be *Available".

3. Reason for the Change:

Technical Specifications 3.B.1.1 and 3.8.1.2 apply to the diese l generators for normal operation (Modes 1 through 4) and shutdown conditions (Modes 5 and 6), respectively. These Tech Specs did not specify the defueled condition and did not require the diesel generators to be " Operable" with the plant in its current state.

However, Tech Spec Clarification No. C-TSC-0BB has been issued to require performance of the surveillance requirements of Tech Spec 4.B.1.2 while the plant is in a defueled condition. This invokes the Mode 6 requirements for the current defueled condition, which are that one diesel generator be ' Operable". This, in turn, requires the fuel oil supply system to the diesel generators to be

• Operable".

Maintaining the fuel oil supply to the air cooled diesel generator

'Available" will not compromise use of this diesel generator in the future as a possible alternate source of power for the nuclear island. The fuel oil supply system to the auxiliary steam boilers will be used for building heating throughout the decommispioning effort. Without heating, necessary freeze protection for indoor components would not be present. Therefore, this system must remain "Available".

008 pcy 009 - 97 Sy 3.y - 9 7 - 0 0 Z (e

4. Safety Evaluation: Sp 2 .i 4

.)

g a. This change was safe for the following reasons:

Per this 10CFR50.59 Safety Evaluation Review, the change does not affect the probability or consequences of previoucly evaluated accidents or malfunctions of equipment important to safety.

There is no reduction in the margin of plant safety.

b. This change does not constitute AN UNREVIEWFD SAFFTY QUESTION because THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRl"NCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALLATED IN THE SAFETY ANALYSIS REPORT. The basis for this-statement is:

CY is in a defueled condition and has notified the NRC of its

' . decision to pemanently cease power operation. Fuel has been permanently removed from the reactor vessel. However, the fuel oil supply eye";em to the diesel generators is still required te be fully " Operable" in order to comply with plant Tech Specs and the TRM. The fuel oil supply system to the air cooled diesel generator and to the auxiliary steem boilers will be "Available".

The operation and required function of the " Operable" and A

s/

"Available" system components is unchanged from previous operation and function. Therefore, the probability of occurrence of a previously evaluated malfunction is also unchanged. In addition, the Fuel Oil to Diesel Generator and Auxiliary Steam Boiler System does not directly communicate in any way with any system containing radioactivity, the Spent Fuel Pool or the Spent Fuel Handling System. Therefore, the probability of occurrence or consequences of a fuel handling accident or a radwaste system failure, which are the only accidents applicable to the defueled condition, is unaffected.

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 operation and required function of the Fuel Oil to Diesel Generator and Auxiliary Steam Boiler System components which will be maintained " Operable" and 'Available" is unchanged from previous operation and function. Therefore, there is no possibility of a different type of malfunction. Also,,there is no possibility of an accident of a different type created because the system does not directly communicate in any way with any system

• containing radioactivity, the Spent Fuel Pool or the Spent Fuel Handling System.

s#

e _ _ _ _ _ _ _ _ _ .

b6N DCY 00 . oS9 - 97 W - cv . 97 - oozG Pb9 e >d4 y

1 THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement

/ khP

• Per ENG 1.7-156, an " Operable" system is one which has a safety related function in the decommissioned mode (i.e. , maintenance of the spent fuel in a safe condition) or is required to be fully operational in order to comply with plant Tech specs, TRM, or other regulatory requirements. The fuel oil supply system to the diesel generators has been evaluated and determined to be an

' Operable" system. The system continues to be maintained in full compliance with Tech Spec 4.8.1.2 per C-TSC-080 while the plant is in a defueled condition. 'Available" systems are those which no longer have Tech Spec requirements associated with them. The fuel oil supply system to the air cooled diesel generator and to the auxiliary steam boilers never had any Tech Spec requirements associated with it for any mode of plant operation. Therefore, there is no impact on the margin of safety as defined in the Bases of any Technical Specification,

c. This change did not require a change to the Technical Specifications.

In addition, a 10CFRSO.82 Decommissioning Review was also performed that concluded that this change does not constitute an unreviewed g decommissioning question for the following reasons:

Placing the Fuel Oil to Diesel Generator and Auxiliary Steam Boiler System in the " Operable" and "Available" categories does not foreclose (preclude) release of the site for possible unrestricted use. =There is no potential failure mode which could create a major spill or rupture, causing a significant radioactive contamination of the soil or buildings which could not easily be decontaminated. It also does not permanently install or retain major structures or equipment on site, or render it difficult to remove systems, structures or components from the site.

Placing the Fuel Oil to Diesel Generator and Auxiliary Steam Boiler .

System in the " Operable" and 'Available" categories does not result in a significant environmental impact not previously reviewed. It does not result in the discharge of radioactivity or chemicals, either by a controlled release, or due to a malfunction or failure of equipment. It also does not affect the site environment (e.g.,

terrain, noise, visual appearance, transmission lines) in any way.

Placing the Fuel Oil to Diesel Generator and Auxiliary St$ss Boiler System in the " Operable" and "Available" categories does not remove reasonable assurance that adequate funds will be available for decommissioning. The cost of procedure revisions, UFSAR revisions, O' etc., taken in total, are not likely to jeopardize decommissioning funds.

PCN pcY 059; M SY 97 - 002 feqe 4 cf 4 I.

There is no applicability to the PSDAR of placing the Puel Oil to O oi 2 o a rator ne Auxt21 ry st = not2 r Svat = in the Oger hie-and "Available" been categories developed for CY. category since that document has not yet Rather, the system categorization process will be used as input to the PSDAR development.

E.

pcN DcY 060 9'i SY 97 - 0026 Page i d3 Attachment 1 dADDAM NECK PLANT Annual Report Summary of Changes Mark the Appropriate Choice:

1 Design Change X Setpoint Change Test Tech Requirements Manual Change

Experiment Procedure Change . . _ .

Jumper Bypass

1. Change Number: Revision Number

Title:

System Categerv Determinatien in a Decemmismiened plant -

Diesel Generator temeremmed Air System

2. Description of Change:

This Safety Evaluation addresses the system category determination for the Diesel Generator Compressed Air System, as shown on P&ID No.

. 16103-26020, Sheet 2. The entire system will remain " Operable" to support diesel-generator operation.

3. Reason for the Change:

Technical Specifications 3.8.1.1 and 3.8.1.2 apply to the diesel generators for normal operation (Modes 1, 2, 3 and 4) and shutdown conditions (Modes 5 and 6), respectively. These Tech Specs did not specify the defueled condition and did not require the diesel generators to be *0perable" with the plant in its current state.

However, Tech Spec Clarification No. C-TSC-088 has been issued to

, require performance of the surveillance requirements of Tech Spec 4.8.1.2 while the plant is in a defueled condition. This invokes the Mode 6 requirements for the current defueled condition, which are that one diesel generator be " Operable". This, in turn, requires the Diesel Generator Compressed Air System to be " Operable".

4. Safety Evaluation:

a. This change was safe for the following reasons:

Per this 10CFR50.S9 Safety Evaluation Review, the change does not affect the probability or consequences of previously evaluated accidents or malfunctions of equipment important to safety.

There is no reduction in the margin of plant safety.

1

T

- OcW pcy-00 oco-97 sy av oozs Bage 2 of 3 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:

CY is in a defueled condition and has notified the NRC of its decision to permanently cease power operation. Fuel has been permanently removed from the reactor veusel. However, the Diesel Generator Compressed Air System is still required to be fully

" Operable" in order to comply with plant Tech Specs and 1 TRM.

The system continues to be maintained in full compliance 5ich pre-established procedures, Tech Specs, TRM, UFSAR, surveillances, preventative maintenance procedures and programs, as applicable.

In addition, the Diesel Generator Compressed Air System does not directly communicate in any way with any system containing radioactivity, the Spent Fuel Pool or the Spent Fuel Handling System. Therefore, the probability of occurrence or consequences of a fuel handling accident or a radwaste system failure, which are the only accidents applicable to the defueled condition, is unaffected.

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 operation and required function of the Diesel Generator Compressed Air System components which will be maintained

" Operable" is unchanged from previous operation and function.

Therefore, there is no possibility of a different type of malfunction. Also, there is no possibility of an accident of a different type created because the Diesel Generator Compressed Air System does not directly communicate in any way with any system containing radioactivity, the Spent Fuel Pool or the Spent Fuel Handling System.

THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION RAS NOT BEEN REDUCED. The basis for this statement is:

Per ENG 1.7-156, an " Operable" system is one which has a safety related function in the decommissioned mode (i.e., maintenance of the spent fuel in a safe condition) or is required to be fully operational in order to comply with plant Tech Specs, TRM, or other regulatory requirements. The Diesel Generator Compressed Air System has been evaluated and determined to be an " Operable" system. ;he system continues to be maintained in full compliance with Tech Spec 4.8.1.2 per C-TSC-088 while the plant is in a

T pos pcy- oo - 0(oo - 97 sY - e.V cc25 Tagd 1 of 3 defueled condition. Therefore, there is no impact on the margin of safety as defined in the Bases of any Technical Specification.

c. This change did not require a change to the Technical Specifications.

In addition, a 10CFR50.80 Decommissioning Review was also performed that concluded that this change does not constitute an unreviewed decommissioning question for the following reasons:

Plac'^g the Diesel Generator Compressed Air System in the " Operable" category does not foreclose (preclude) release of the nite fc. '

possible unrestricted use. There is no potential failure mode which could create a major spill or rupture, causing a significant radioactive contamination of the soil or buildings which could not easily be decontaminated. It also does not permanently _ install or retain major structures or equipment on site, or render it difficult to remove systems, structures or components from the site.

Placing the Diesel Generator Compressed Air System in the " Operable" does not result in a significant environmental impact not previously reviewed. It does not result in the discharge of radioactivity or i

chemicals, either by a controlled release, or due to a malfunction or failure of equipment. It also does not affect the site environment (e.g., terrain, noise, visual appearance, transmission lines) in any way.

Placing the Diesel Generator Compressed Air System in the " Operable" does not remove reasonable assurance that adequate funds will be available for decommissioning. The cost of procedure revisions, UFSAR revisions, etc., taken in total, are not likely to jeopardize decommissioning funds.

There is no applicability to the PSDAR of placing the- Diesel Generator Compressed Air System in the " Operable" since that document has nct yet been developed for CY. Rather, the system categorization process will be used as input to the PSDAR development.

pc14 Oc'( 061 - 97

/ ACP 1.2 2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report l.

Page l of t Safety Evaluation Number: sY-EV 97-0055 Revision: 0 Document Numb r:.E J 131M Attarhment 122 for Primnry Ventilation S" stem Revisiom WA Document

Title:

__. System Catego-v Determination in a Decommittioned Plant Primnev Ventilation System Provide a bricf description of the change and a summary of the Safety Evaluation in the format below.

1. Brief Description of Cb.vge and Safety Evaluation Summary:

This Safety Evaluation addresses the system category determination for the Primary Ventilation System, as shown on P&lD No.1610*, 26024, Sheets I through 5. Ponions of this system will remain " Operable","Available" or be pitced in " Lay up" to suppon plant decommissioning activities and reactor coolant system (RCS) d%ontamination. Portions of this system are being " Abandoned" because they are no longer needed.

The operation and required function of the " Operable" and "Available" cornponents is unchanged from previous operation and function. Therefore, the probability of occurrence or consequences of a previously evaluated malfunction is unchanged and there is no possibility for a tnalfunction of a difTerent type. The " Operable" ponion of the Primary Ventilation System will support maintenance of the spent fuel in a safe condition. The functions of the "Available" ponions of the system do not change. The " Abandoned" ponions of the system will be isolated from any system containing radioactivity. The ponion of q the system being maintained in " Lay up" will not be used until the RCS decontamination process. Therefore, the probability

~; of occurrence of a fuel handling accident or a radwaste system failure, which are the only accidents applicable to the j

defueled state, is unatTected and there is no possibility for an accident of a different type. This change does not foreclose (preclude) release of the site for possible unrestricted use. it does not result in a significant environmental impact not previously reviewed. It does not remove reasonable assurance that adequate funds will be available for deco nmissioning.

This change is safe and does not involve an unreviewed safety question. The change does not affect the probability or consequences of previously evaluated accidents or malfunctions of equipment imponant to safety. There is no reduction in the margin of plant safety, in addition, this change does not involve an unresolved decommissioning question.

2. Reason for the Change:

The " Operable" portions of the Primary Ventilation System include the spent fuel building exhaust system and primary ctack radiation monitor R 14A. The spent fuel building exhaust system is a safety related system which is required by Tech Spec 3.9.12 to be " Operable" during movement of fuel within the spent fuel pool or during crane operation with loads over the spent fuel pool. The primary stack radiation monitor R 14A is required by Tech Spec 3.3.3.8.1 to be " Operable" at all times.

Radiation monitor R 14B must remain " Operable"as a viable backup to R 14A as defined by C TSC-46. The PAB purge duct to the primary vent stack and associated boundaries with the waste gas system remains " Operable" to provide a path for monitoring discharges to the stack. The "Available" portions of the system include the PAB ventilation & purge system, the PAB supply ventilation system, the spent fuel buildine mpply and recirculation system, the waste disposal building ventilation system, and CAR Fans I and 2. The vem ion systems and CAR Fans are needed to provide a suitable and safe environment for personnel and equipment. The only ponion of the system that is in " Lay up" is primary vent header moisture drain tank TK 86-1 A and interconnecthg piping with the RCS. The " Abandoned" portions of the system include the CRDM cooling fans and ductwork, the iodine removal unit, the air tank receis er for containment pressurization (TK 81-l A), the containment air monitoring system i cluding udiation monitor R-l1/12, CAR Fans 3 and 4, reactor containment control air to CAR Fans 1 and 2, a boundary with the boron recovery system, the modified blind flanges, and the spent fuel building decon room exhaust fan which was previously " Abandoned"in place.

Preparer A 4. 54 bra A) / /-*c-8d m Date od-f 7-97 260243. DOC

peu ocY o62 - 97

, S y-lY 9740/7 fs u 9 7-6 0 Pkge i *f }

Attachment 1 EADDAM NECK PLANT Annual Report Summary of Changes Mark the Appropriata Choice:

Design Change _Y Setpoint Change Test Tech Requirements Manual Change Experiment Procedure Change Jumper Bypass

1. Change Number: Revision Number:

Title:

System catecorv Determinarien in a Decommi ssioned. Pl ant -

Closed cooline Water Rystem

2. Description of Change:

This Safety Evaluation addresses the system category determination for the Closed Cooling Water System, as shown on PLID No. 16103-26027, Sheets 1 and 2. Portions of this system will remain "Available" to support operation of other systems which are also "Available". Other portions of the system are being " Abandoned" since its use is no longer required.

3. Reason for the Change:

The Closed Cooling Water System provides cooling for the control air and service air compressors, the vacuum deaerator pumps, the :: n feed pump lube oil coolers, the feedwater heater drain pumps, the feedwater sample coolers and the condensate sample coolers. Portions of this system will remain "Available" to support operation of the control air, service air and water treatment systems which are also "Available". Other portions of the system are being " Abandoned" since its use is no longer required.

4. Safety Evaluation:

a. This change was safe for the following reasons:

Per this 10CFR50.59 Safety Evaluation Review, the change does not affect the probability or consequences of previously evaluated accidents or malfunctions of equipment important to safety.

There is no reduction in the margin of plant safety.

b. This change does not constitute AN UNREVIEWED SAFETY QUESTION because

pc4 Ocy . 00- 062 - 97 SY - EV 0018 Pbge I J1 THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTA!rt TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is CY is in a defueled condition and has notif.ud the NRC of its decision to permanently cease power operation. Fuel has been permanently removed from the reactor vessel. As such, much of the equipment which previously had a safety related and/or operational function during normal plant operation, may no longer be important to safety or required to be operable. This includes the closed Cooling 9ater System. The operation and required function of the system components which will remain 'Available* is unchanged from previous operation and function. Therefore, the probability of occurrence of a previously evaluated malfunction is also unchanged. The

• Abandoned" components no longer have a required tunction in the plant and, therefore, cannot malfunction. In addition, the closed Cooling Water System does not directly roenunicate in any way with any system containing radioactivity, eta Spent Fuel Pool or the Spent Fuel Handling System. Therefore, the probability of occurrence or consequences of a fuel handling accident or a radwaste system failun which are the only accidents applicable to the defueled condition, is unaffected.

THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE TRAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:

The operation and required function of the Closed Cooling Water System components which will remain *Available" is unchanged from crevious operation and function. Therefore, there is no possibility of a different type of malfunction. The ' Abandoned" components no longer have a required function in the plant and, therefore, cannot malfunction. Also, there is no possibility of an accident of a different type created becauss the closed Cooling l

Water System does not directly communicate in '.ny way with any system containing radioactivity, the Spent Fuel Pool or the Spent Fuel Handling System. Cooling water is supplied to the Closed cooling Water heat exchangers from the service water (SW) system through 2-in lines. This portion of the SW system is "Available" I and is not required for spent fuel pool (SFP) heat exchanger cooling.' In the current defueled state, the SW system flow requirements are so low that the component cooling heat exchangers are needed as' phantom loads to allow the SW pumps to operate in '

the preferred range above 2000 gptn. _ _

Therefore, any postulated failure of these 2 in SW lines or the closed Cooling Water heat exchangers could not adversely affect the required flow rates to the SFP heat exchanger.

~ . .

- . - --- . _ - - .. - - - . - . _ . - . . ~. - - -

p ocT , 00- OM - 97 g . 0V 97 = 0088 bee 5 AS l/'

1 TH1l MARGIN OF SAFETY AS DEFINED IN THE BASI!! FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:

I

• Available" and ' Abandoned

• systems are those which

• o longer have Technical Specification requirementa associated with ahem. The closed Cooling Water System, never had any Technical specification requirements associated with it for any mode of plant operation.

Therefore, there is no impact on the margin of safety as defined

in the Bases of any Technical Specification.

c. Thio changs did not require a change to the Technical Specifications.

In aduition, a 10CFRSO.82 Decommissioning Review was also performed that concluded that this change does not constitute an unreviewed decommissioning question for the following reasons:

Placing the closed Cooling Water System in the "Available" and

" Abandoned

• category does not foreclose (preclude) release of the

site for possible unrzstricted use. There is no potential failure mode which could create a major spill or rupture, causing a significant radioactive contamination of the soil or buildings which could not easily be decontaminated. It also does not permanently i

install or retain major structures or equipment on site, or render it j dif ficult to remove systems, structures or components from the site.

i, Placing the closed cooling Water system in the "Available" and

" Abandoned

• category does not result in a significant environmental impact not previously reviewed. It does not result in the discharge of radioactivity or chemicals, either by a controlled release, or due to a malfunction or failure of equipment. It also does not affect the site environment (e.g., terrain, noise, visual appearance, transmission lines) in any way .

Flacing the Closed Cooling Water System in the "Available" and

" Abandoned" category does not remove reasonatie assurance that

adequate funds will be available for decommissioning. The cost et

! procedure revisions, UFSAR revisions, etc., taken in total, are not

, likely to jeopardize decommissioning funds.

There is no applicability to the PSDAR of placing the Closed Cooling Water System in the "Available" and " Abandoned" category since that document has not yet been developed for CY. Rather, the system categorization process will be used as input to the PSDAR development.

1

. - - , . --- -- ,,.w-

-- - ru-'* ---

DcH DCT - 00 o 063 " 97 ACP 1.2 2.42 Rev.1 MAJOR Forta 3 10 CFR 50.59 (b)(2) Report 7'* '

Page1of1 Safety Evaluation Number:- EY.EV 97 00$6 Revision:- 0 Document Nusnber:ING 17.M6. Attm4 ment 1$ $ int Roaldont Heat Remnval Evetem Revision: N/A .

Document

Title:

System Pateenev Detenninntinn le a Deenmmittinned Plant . Retidon1 Heat Removat Kvrtem Provide a brief descripion of the change and a summary of the Safety Evaluation in the format below, i

1. Brief Lescription of Change and Safety Evaluation Summary:

His Safety Evaluation addresses the system category determination for the Residual Heat Removal (RHR) S shown on P&!D No.16103 26023.

Portions of thl: rystem will remain " Operable","Available" or be placed in " Lay.up" to suppen plant decommissioning activities and ructor coolant rystem (RCS) decontamination. Portions of this sys being " Abandoned" because they are no longer needed.

The operation and required function of the " Operable" and "Available" componema is unchanged from previ and func-ii.w. Therefore, the probability of occurrence or consequences of a previously evaluated malfuncti and there is no possibility for a malfunction of a different type. He "Opersble" portion of RHR continues to be maintain in order to comply with the plant TRM. He function of the "Available" portions of RHR do not change. The "Abnadone portions of RHR will be isolated from the rest of the system, which contains radioactivity. Tbc remainder of the RHR System will be maintained in " Lay up" and will not be used until the RCS decontamination process. Derefore, the probability of occurrence of a fuel handling accident or a radwaste system failure, which are the only accidents the defueled state, is unaffected and there is no possibility for an accident of a different type, nis change does not foreclose (preclude) release of the site for possible unrestricted use. it does not result in a si environmentalimpact not previously reviewed. It does not remove reasonable assurance that adequate funds will be available for decommissioning. -

His change is safe and does not involve an unreviewed safety question. De change does not affect the probabil consequences of previously evaluated accidents or malfunctions of equipment important to safety. Dere is no reduction in the margin of plant safety, in addition, this char;t does not involve an unresolved decommissioning question.

2. Reason for tbc Change:

He RHR System removed residual beat from the reactor core and reduced the temperatare of the RCS durirg plan cooldown, it also transferred water betwwn the Refueling Water Storage Tank and the refueling cavity during rtfuel operations. The system was also to be used following a LOCA to cool ned circulate spilled water from the teactor containment sump back to the RCS. With the plant in its current state, the function of the RHR is no longer required. He only portion of the RHR System that will be " Operable"is the interface with the fire protection system which is bein maintained " Operable"in order to comply with the plant TRM, Section 11.1. De portions of the RHR System that will remain "Available" include tbc reactor containment sump, the I g.in RHR pump suction line, the component cooling tank relicf valve discharge line, and an interface with the PAB ventilation system. He " Abandoned" portion of the RHR System includes the containment spray header, the charcoal filter spray lines, and piping to the cone deluge lines. De balance of the RHR System, wbleb includes the RHR pumps, beat exchangers and interconnecting piping, is bein maintained in " Lay up" to facilrtate the future RCS decontaminazion.

Preparer _ b. A . SA A FA Al[  ?"D^ -

Date OS

• I7

• 9 7 ma-

)

- ~ - ~ - -

h i

btN tW- 00 OM- 97 ACP 1.2 2.42 Rev.1 MAJOR f .

Form 3 10 CFR 50.59 (b)(2) Report Page1of1 Safety Evaluation Number: EY EV 97 0059 Revision: 0 Document Number ENO 17136. Attehmer t 1M for Lieuld Warte System Revision: N/A Document

Title:

Svrtem cetecorv Determinatinn in a Decommluioned Plant Linuld w aste svrtem Provide a brief description of the change and a sumicary of the Safety Evaluation in the fonnat below,

l. Brief Description of Change and Safety Evaluation Summary:

This Safety Evaluation addresses the system category determination for the Liquid Waste System, as shown on P&ID No.

16103 26030, Sheets 1 through 5. Portions of this system will remain " Operable","Available" or be placed in " Lay up" to support plant decommissioning activities and reactor coolant system (RCS) decontamination. Portions of this system are being "Abandoced" because they are no longer needed.

The operation and required function of the " Operable" and "Available" coroponents is unchanged from previous operation an't function. Therefore, the probability of occurrence or consequences of a previously evaluated malfunction is unchanged l and there is no possibility for a malfunction of a different type. De " Operable" portion of the Liquid Waste System will support Tech Spec requirements for effluent discharge monitoring. De functions of the "Available" portions of the system do not change. ne "Abanhoed" ponions of the system will be isolated from any system containias radioactive liquid. The portion of the system being maintained in " Lay up" will not be used until the RCS decontamination process. Therefore, the probability of occurrence of a fuel handling accident u a radwaste system failure, which are the only accidents applicable to the defueled state,is unaffected and there is no possibility for an accident of a different type. His change does not foreclose (preclude) release of the site for possible unrestricted use, it does not result in a significant environmental impact not previously reviewed it does not remove reuenable assurance that adequate funds will be available for decommissioning.

This change is safe and does not involve an unreviewed safety question. De change does not affect the probability or consequences of previously evaluated accidents or malfunctions of equipment important to safety. Dere is no reduction in the margin of plant safety. In addition, this change does not involve an unresolved decommissioning question.

2. Reason for the Change:

ne Liquid Waste System collects, holds, processes and disposes of all potentially radioactive atrated liquids generated by the plant, ne " Operable" portion of the system is the waste test tank effluent monitor (R 22), the waste liquid effluent discharge flow recorder / controller (FRC 1003), and associated instrutnentation loop to and including air operated valve WD-FRCW1003, nis valve fails closed on a loss of control air and on a loss of power. R 22 monitors all waste liquid effluents to the service water discharge beader. In the event that radiation levels are above the acceptence level, R 22 will automatically stop the waste effluent release by closing FRC%1003. Currently, all liquid waste is processed by the Chem Nuclear Mobile Demineralizer System. Portions of the Liquid Wute System that are needed and must remain "Available" include: the aerated drains tanks; the aerated drains boldup tartk; the wute test tanks; the spent resin tank; the waste evaporator distillate F.lter; the waste liquid polishing demineralizer; and associated pumps, equipment and interconnecting piping. Also, the PAB and ion exchanger pit sumps and pumps are needed to collect and transport miscellaneous leakage, drainage and spills. De portions of the Liquid Weste System being maintained in " Lay up" include the steam generstor blowdown ank cooler and interconnecting piping since their use may be required during the future RCS decontamination.

Since allliquid waste is processed by the Chem Nuclear Mobile Demineralizer System, the Liquid Wute System process components are no longer needed and are being " Abandoned". Rese components include the aerated drains demineralizer

~ g _ and filter, the steam generator blowdown condensers, the waste evaporator, and associated components and piping.

Preparer b. A.5AB EAM / h *:' m Date CB 19 = 9 7 26030LDOC

o Otp (W- 00 ' 06 " 97 ACP 1.2 2A2 Rev.1 MAJOR

(~ ~

Form 3 10 CFR 50.59 (b)(2) Report Page1of1 Safety Evaluation Number: 1%EW97 0011 0 Revision:

Docuraent Number: ENO t 7 t M A" *k=*a' 112 for A***^"' Waa'* Sva'*= Revision: N'A Document

Title:

Sva'*= t'a'aaarv tha**=la=*iaa in a D ra==l**laaad Plant na* aue Waa'* *; vata =

Provide a brief description of the change and a summary of the Safety Evalution in the fonnat below,

l. Brief Description of Change and Safety Evaluation Summary:

his Safety Evaluation addresses the system category ceterinination for the Gaseous Wate System, as shown on P&iD No.

16103 26031, Sheets I through 3. Ponions of this system will reinain " Operable","Available" or be placed in " Lay up" to support plant decommissioning activities and reactor coolant system (RCS) decontamination. Ponions of this system are being " Abandoned" because they are no longer needed. ,

• Ihe operation and required function of the " Operable" and "Available" components is unchanged from previous operation and function. Therefore, the probability of occurrence or consequences of a previously evaluated malfunction is unchanged i and there is no possibility for a malfunction of a different type. De " Operable" portion of the Gaseous Wasts System is the boundary with the primary vent stack header which is within the " Operable" pottion of the primary ventilation system. De functions of the "Available" portions of the system do not change ne " Abandoned" portions of the system will be isolated from any system containing radioactivity, ne portion of the system being maintained in " Lay up" will not be used until the RCS decontamination process. Derefore, the probability of occurrence of a fuel handling accident or a radwaste system failure, which are the only accidenu applicable to the defueled state, is unaffected and there is no possibility for an accident of a different type. This change does not foreclose (preclude) relene of the site for possible unrestricted use. it does not result in a significant environmental impact not previously reviewed it does not remove reasonable assurance that adequate funds will be available for decommissioning. This change is safe and does not involve an unreviewed safety question. The change does not affect the probability or consequences of previously evaluated accidents or math.nctions of equipment important to safety. Dere is no reduction in the margin of plant safety, in addition, this change does not involve an

- untnolved decommissioning question.

2. Reason for the Change:

ne Ganous Waste System is designed to collect, hold and dispose of all radioactive gueous wastes generated by the plant.

The " Operable" portion of the systers is the boundary with the primary vent stack header which is within the " Operable" portion of the primary ventilation system. ne "Available" pottions of the system include: the waste gas surge tank, the equipment and floor drains, the drain tanks and pumps, the primary drains tank and pumps, the primary drains tank vent condenut, and associated piping. ney also include interfaces with the primary water, liquid waste, chemical and volume conwot, primary sample, reactor coolant, boron recovery, prirnary ventilation, and nitrogen systems. De primary drains tank and associated pumps must remain "Available" for draining primary systems. De waste gas surge tank must remain "Available" to provide a vent path from the primary drain tank to the primary vent stack. De primary drains tank vent condenwr must remain "Available" to remove condensable vapor prior to venting to the stack, ne equipment and floor drains and associated components must remain "Available" to collect liquid waste. De portions of the systern being maintained in" Lay up" include: the waste gas compressors, the degasifier vent cooler, the waste gas decay tanks, the sample cylinder station, the deguifier prefilter, the degasifier vent condennt, the degasifier tank, the degasifier liquid transfer pumps, the degulfier effluent cooler, and the degasifier preheater. De " Abandoned" portions of the system include the l J valve stem leak off cooler and two previously " Abandoned" control air lines.

Preparer D.4.

• A atAAJ / 2=*O tn Date .. 07

• iE *II 260313.00C

-: 3- -

=- :- __

f f bN bef 066- 97 ACP 1.2 2.42 4

, Rev.1 MAJOR

() Form 3 10 CFR 50.59 (b)(2) Report Page 1 ofI Safety Evaluation Number: EM%07-00d1 Revision: 0 Document Number: Jg 1.7.t $6 Attachment 1M for Primary Water Svitem Revision: WA Document

Title:

Svatem Catennrv Determinatinn in a beenmmittinned Plant . Primary Water Evitem l

Provide a brief description of the change and a suma .-ev of the Safety Evaluation in the fortnat below.

1. Brief Description of Change and Safety Evalurw WWy:

l This Safety Evaluation addrenes the system category dmtmination for the Primary Water System, as shown on P&lD No.

16103 26046, Sheets I through 3. Ponions of this system will rrnain " Operable","Avallable" or be placed in " Lay up" tc suppon plant decommissioning activities and reactor coolant system (RCS) decontamination. Ponions of this system are being " Abandoned" because they are no longer needed.

Tht operation and required function of the " Operable" and "Available" components is unchanged from previous operation and function. Therefore, the probability of occurrence or consequences of a previously evaluated malfunction is unchanged and there is no ponibility for a malfunction of a different type. The " Operable" ponion of the Primary Water System will support maintenance of the spent fuel in a safe condition. The functions of the "Available" ponions of the system do not change. The " Abandoned" ponions of the system will be isolated from the rest of the system, which contains radioactivity, he portione cf the system being maintained in " Lay.up" will not be used until the RC3 decontaminatiori procen.

Therefore, the probability of occurrence of a fuel handling accident or a radwaste system failure, which are the oniv O accidents applicable to the defueled state, is unaffected and there is no poss This change does not foreclose (preclude) release of the site for ponible unrestricted use, it does not result in a significant j environmentalimpact not previously reviewed. It does not remove reasonable assurance that adequate funds will be .

, available for decomminioning.

This change is safe and does not involve an unreviewed safety question. The change does not affect the probability or consequences of previously evaluated accidents or malfunctions of equipment important to safety. There is no reduction in

, the margin of plant safety. In addition. this change does not involve an unresolved decommissioning question.

, 2. Reason for the Change:

he Primary Water System supplies primary grade water as needed to various plant systems for makeup, filling, venting, flushing, or special testing that requires high purity water. De majority of the system is "Available" except for one

" Operable" water supply, and several" Lay up" and " Abandoned" water supplies. Primary water can be used for spent fuel pool makeup. The only portion of this system which is " Operable" is the boundary with the spent fuel pit cooling system which is consistent with the corresponding spent fuel pit cooling system classification. The Primary Water System is the i

only source of clean, domineralized water on the hot side of the plant and will be used for flushing, rinsing and numerous other activities during decomminioning. For this reason, the majority of the system will remain "Available". De

" Abandoned" ponion of thesystem includes the water supplies to the steam generator sample coolers and the waste evaporator pumps. These components are being " Abandoned" and no longer rec. Je a water supply. The only portions of the system that are being placed in " Lay up" are boundaries with the RCS, whieti is also in " Lay up".

i Preparer D. A. sA nE AU / A =-Hv Date at-o1-97 C

l 26046tDoc

- - , - ,_,, ,- -- - - . , . - .-.r.,, ,- - y, , - , w, ,,.ev,,, - w-

OcW De.Y 067- 97 sy - tv oott P49e t ef 3

/ Attachment i HADDAM NECK PLANT Annual Report Summary of Changes Mark the Appropriate Choices Design change Y Setpoint Change Test Tech Requirements Manual Change Experiment Procedure Change Jumper Bypass

1. Change Number Revision Number

Title:

Eystem Categely Determination ina Decommi nalened Plant .

Turbine Building Waste Water Treatment system

2. Description of Change:

This Safety Evaluation addresses the system category determination for the Turbine Building Waste Water Treatment System, as shown on P&ID No. 16103-26047. The entire system will remain 'Available",

except for the turbine building sump overflow lines to the discharge canal which will be ' Abandoned".

O 3. Reason for the Change:

The function of the Turbine Building Waste Water Treatment System is to process and cleanup all water that enters the turbine building drains. This system must remain 'Available" to prevent the release of oil or other pollutants to the discharge canal. In addition, keeping the Turbine Building Waste Water Treatment System 'Available" will help satisfy applicable NPDES regulations which limit release of pollutants to the river.

4. Safety Evaluations

a. This change was safe for the following reasons:

Per this 10CFR50.59 Safety Evaluation Review, the change does not affect the probability or consequences of previously evaluated accidents or malfunctions of equipment important to saiety.

There is no reduction in the margin of plant safety.

b. This change does not constitute AN UNREVIEWED SAFETY QUESTION because

(~) -

s.- . ,. .--r.. , - - . , - - - . . - - - - - . . - - . ,------,,,-- ~ . . - ---.me u-- --%m-. - - - - - , . . - . , , - , - . - - - ..--,,,r- -

e

Ocw ocY 067 - 97 97- 002L ST-CV=d Pae 2 3 THERE IS NC INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE

! CONSEQUENCES OF AN ACCIDE!U OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFE?Y PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The

() basis for this statement ist CY is in a defueled condition and has notified the NRC of its decision to permanently cease power operation. Puel has been permanently removed from the reactor vessel. As such, much of the equipment which previously had a safety related and/or operational function during normal plant operation, may no longer be important to safety or required to be operable. This includes the Turbine Building Waste Water Treatment System. The operation and required function of the system components which will remain *Available" is unchanged from previous operation and function. Therefore, the probability of occurrence of a previously evaluated malfunction is also unchanged. Releases from the Waste Neutralization Tank and the Turbine Building Sump are controlled by the REMODCM. However, in the defueled state, there is no longer any possibility of a l

system in the Turbine Building becoming contaminated. The i " Abandoned" components no longer have a required function in the plant and, therefore, cannot malfunction. In addition, the Turbine Building Waste Water Treatment System does not directly communicate in any way with any system containing radioactivity, the Spent Fuel Pool or the Spent Fuel Handling System. Therefore, the probability of occurrence or consequences of a fuel handling accident or a radwaste system failure, which are the only

() accidents applicable to the defueled condition, is unaffected.

THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT RAS NOT BEEN CREATED. The basis for this statement is:

The operation and required function of the Turbine Building Waste Water Treatment System components which will remain *Available" is unchanged from previous operation and fu.iction. Therefore, there is no possibility of a different type of malfunction. The

" Abandoned" components no longer have a required function in the plant and, therefore, cannot malfunction. Also, there is no possibility of an accident of a different type created because the Turbine Building Waste Wacer Treatment System does not directly communicate in any way with any system containing radioactivity, the Spent Fuel Fool or the Spent Fuel Handling System.

THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION RAS NOT BEEN REDUCED. The basis for this statement is: .

• Available".and ' Abandoned" systems are those which no longer have Technical Specification requirements associated with them. The

() Turbine Building Waste Water Treatment System never had any

! DOM pcf -- 17 00 -- 067".97 s y . Ev 00U Pa ): 3 *t 3

, Technical Specification requirements associated with it for any mode of plant operation. Therefore, there is no impact on the 0_ rei# r r tv e ri= a '# tb er av z camic Specification.

c. This change did not require a change to the Technical Specifications.

In addition, a 10CTRSO.82 Dicommissioning Review was also performed that concluded that this change does not constitute an unreviewed decommissioning question for the folloving reasons:

Placing the Turbine Building Waste Water Treatment System in the "Available" and ' Abandoned" category does not foreclose (preclude) release of the site for possible unrestricted use. There is no potential failure mode which could create a major spill or rupture, causing a significant radioactive contamination of the soil or buildings which could not easily be decontaminated. It also does not permanently install or retain major structures or equipment on site, or render it difficult to remove systems, structures or components from the site.

Placing the Turbine Building Waste Water Treatment System in the "Available" and " Abandoned" category does not result in a significant environmental impact not previously reviewed. It does not result in the discharge of radioactivity or chemicals, either by a controlled release, or due to a malfunction or failure of equipment. It also does not affect the site environment (e .g. , terrain, noire, visual l appearance, transmission lines) in any way. Keeping entire system

'Available" except for the turbine building sump overflow lines to the discharge canal, which will be " Abandoned", will help satisfy applicable NPDES regulations which limit release of pollutants to the river.

Placing the Turbine Building Waste Water Treatment System in the "Available" and ' Abandoned" category does not remove reasonable assurance that adequate funds will be available for decommissioning.

The cost of procedure revisions, UFSAR revisions, etc., taken in total, are not likely to jeopardize decommissioning funds.

There is no applicability to the PSDAR ci placing the Turbine Building Waste Water Treatment System in the "Available" and

' Abandoned" category since that d cument has not yet been developed for CY. Rather, ?.he system categorization process will be used as input to the PSDAR development.

O V

l

oc w pcy oo . o&B . 97 ACP 1.2 2.42 Rev.1 MAJOR g Form 3 10 CFR 50.59 (b)(2) Report Page1of1 Safety Evaluation Number: SY EV 97-0046 Revision: 0 Document Number: ENO 111%. Aftnehment IM for Snent Fuel Pit Cootine Svitem Revision: N/A Document

Title:

Svitem catenerv Determination in a becommittlened Plant Snent Fuel Pit Cooline Svitem Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Bricf Description of Change and Safety Evaluation Summary:

his Safety Evaluation addresses the system category detennination for the Spent Fuel Pit Cooling System, as shown on P&lD No.16103 26049. Portions of this system will remain " Operable" or "Available" to suppon plant decommissioning activities. One portion of this system is being " Abandoned" because it is no longer used.

The operation and required function of the " Operable" and "Available" components is unchanged from previous operation and function. Therefore, the probability of occurrence or consequences of a previously evaluated malfunction is unchanged and there is no possibility for a malfunction of a different type. Tech Spec 3/4.9.11, which stipulates minimum water levels f:t the Spent Fuel Pool, is applicable whenever fuel is in the pool. Also, the Mode 6 requirements of Tech Spec 3/4.9.15 have been invoked for the defueled condition at CY by C.TSC.093. nese include maximum pool temperature limits and operability requirements for system pumps and the plate heat exchanger. In order to meet these requirements, portions of the

, Spent Fuel Pit Cooling Sys:em must remain " Operable". The functions of the "Available" portions of the system do not g gehange. The only portion of the system being " Abandoned" was removed and" Abandoned" previously. Therefore, the probability of occurrence of a fuel handling accident or a radwaste system failure, which are the only accidents applicable to the defueled state, is unaffected and there is no possibility for an accident of a different type.

This change does not foreclose (preclude) release of the site for possible unrestricted use. it does not result in a significant environmental impact not previously reviewed. It does not remove reasonable assurance that adequate funds will be available for decommissioning.

This change is safe and does not, involve an unreviewed safety question. The change does not affect the probability or c:nsequences of previously evaluated accidents or malfunctions of equipment important to safety. There is no reduction in the margin of plant safety, in addition, this change does not involve an unresolved decommissioning question.

2. Reason for the Change:

De Spent Fuel Pit Cooling System removes residual heat from the spent fuel stored in the pool. The entire system will remain either " Operable" or "Available" except for one section of piping which was previously removed and " Abandoned.

The " Operable" portions of the system include le pool, cooling loops, filtration loops, and the fuel transfer canal sluice gate to meet the requirements of Tech Specs 3/4.9.11 and 3/4.9.15. The "Available" portions of the system include the spent fuel building floor drains, sump, sump pumps, interconnecting piping to the aerated drains tank, ano the service air line to the fuel transfer canal. The only portion of the Spent Fuel Pit Cooling System being" Abandoned"is the removable poollow suction line which was previously removed arid " Abandoned".

2-Preparer

n. A 34a rad /

• 4e - - Date o&-od-97

/c ,

DcH Dc6 00 0M - 97 ACP 1.2 2.42 Rev.1 MAJOR Forrn 3 10 CFR $0.59 (b)(2) Report Page l of t Safety Evaluation Number: s%Ev.97 004i Revision: 0 Document Number: ENO 17.146. Attachment IM for Service Air %vitem Revision: _ WA Document

Title:

Svitem Ccteen Determinatinn in a Decommittioned Wnt . t,ervlee Air Kvitem Provide a brief description of the change and a summary of the Safety Evaluation in the format below,

l. Brief Description of Change and Safety Evaluation Summary:

his Safety Evaluation addresses the system category determination for the Service Alt System, as shown on P&lD No.

1610b26051, Sheets 1 through 4. Ponlons of this system will remain "Available" to support plant decommission activities and other portions of this system ele being " Abandoned" because they are no longer needed.

De operation and required fnetion of the "Available" components is unchanged from previous operation and function.

Derefore, the probability of occurrence or consequences of a previously evaluated malfunction is unchanged and there possibility for a malfunction of a different type. The " Abandoned" components no longer have a required function in the plant and cannot malfunction. The Service Air System does not directly communicate in any way with any system l

containing radioactivity, the Spent Fuel Pool or the Spent Fuel Handling System. Therefore, the probability of occurrenc I

a fuel handhng accident or a radwaste system failure, which are the only accidents applicable to the defueled state,is unaffected and there is no possibility for an accident of a different r>pe.

His change does not foreclose (preclude) release of the site for possible unrestricted use. It does not result in a sig environmental impact not previously reviewed. It does not remove reasonable assurance that adequate funds will bc available (c,r decommissioning.

I This change is safe and does not involve an unreviewed safety question. The change does not affect the probability o consequences of previously evaluated accidents or malfunctions of equipment important to safety. There is no reduction in the margin of plant safety. In addition, this change does not involve an unresolved decommissioning question.

2. Reason for the Change:

The Service Air System supplies compressed air throughout the plant for breathing air, operation of pneumatic tools, refue!ing tools, and other miscellaneous compressed air loads. He majority of the Service Air System will remain "Available" to perform these functions. The " Abandoned" portions of the Service Air System include air supplies to the water treatment demirertlizers, the rod control cluster (RCC) change fixture gripper, and the service water filters; and the air bottle rack. The water treatment demineralizers and RCC change fixture gripper are no longer used; hence their air can be " Abandoned". De air supply to the service water filters (w hich are "Available")is no longer used and can be

" Abandoned", he use of the eight service air bottles is not credited in control room dose calculations for the defueled condition. Without use of service air bottles, the control room doses arc within the limits of GDC 19. Therefore, the air bottle rack can be " Abandoned".

?reparer

b. A . x A A CA AJ / & Mh* = _

Date _ 0 7-l 7- 9 7 U

26051). DOC A

Ocw DcT - 00 070- 97 o ACP 1.2 2.42

Rev,1 MAJOR Form 3 10 CFR 50,59 (b)(2) Report Page l of t Safety Evaluatice Number: *:Y.EV 07 oo!7 Revision: _ o Document Number: ENG 1.7.!!6 Aftehment 12 7 for Cnn'm1 Air System Revision: __ N/A _

Document

Title:

_ Svetem ra'etorv Determination in a DecommMnned Plant . Control Air Svrtem Provide a brief description of the chaege and a sum' gary of tbc Safety Evaluation in the format below.

1. Brief Description of Cbange and Safety Evaluation Summary:

Dis Safety Evaluation addrenes the system category determination for the Control Air System, as shown on P 16103 26052, Sheets I through 9. Portions of this syste.n will el:her remain "Avallable" or be place plant decommissioning activities and reactor coolant system (RCS) decontamination. Portions of th

" Abandoned" because they are no longer needed.

The operation and required function of the "Available" components is unchanged from previous operation and fun Herefore, the probability of occurrence or consequences of a previously evaluated malfunction is unchanged a possibility for a malfunction of a different type. He " Lay.up" componenu will not be functionin5 while in that sta therefore, a malfunction is not pouible. The " Abandoned" components no longer have a required function in cannot malfunction. The Control Air System does not directly communicate with the Spent Fuel Pool or the Spent

, Handling System. The portions of the system being maintained "Available" or placed in " Lay.up" which sup to a system containing radioactivity are unchanged by this categorization process. De " Abandoned" portions of the will be isolated from any system containing rsdioactivity. Derefore, the probability of occurreece of a fuel han accidert or a radwaste system failure, which are the only accidents applicable to the defueled state, is unaffected and th t.c possibility for an accident of a different type.

His change does not foreclose (preclude) release of the site for possible unrestricted use. It does not result in a s environmental impact not previously reviewed. It does not remove reascnable assurance that adequaie funds will be available for decomminioning.

This change is safe and does not involve an unreviewed safety ques: ion. The change does not affect the probabil consequences of previously evaluated accidents or malfanctions of equipment important to safety. There is no reduction in the margin of plant safety, in addition, this change does not involve an unresolved decommissioning question.

2. Reason for the Change:

A large portion of the Control Air System will remain "Available". %c portions of the Coraal Air System that will remain "Available" include the three compressors, two headers, and interconnecting lines to the various plant components b served. No portions of the system are categorized as " Operable", since there are no Technical Specifications or Technical Requirements applicable to this >ystem. During plant operation, loss of Control Air is a significant transient as it results in challenges to safety equipment. In the present defueled condition, Control Air is used only for house leads such as h and ventilation. Lou of Control Air in this condition presents no challenges to safety equipment. Portions of the Control Air System that are not eat rized as "Available" include portions serving components that are in " Lay up", or portions serving components that are be f' Abandoned". In these instances, the categorization of the Control Air portion matches the cctegory of the component being served.

i k Y '

Prepner D. A . .s A n CAs) / 2 - k-- Date e L ~ IL - 9 7 260121 DOC

1 l

I pcw pc.Y 071-97

, ACP 1.2 2.42 Rev.1 MAJOR l

( Form 3 - 10 CFR 50.59 (b)(2) Report Page1of1 Safety Evaluation Number: _Syfy.v.97 0040 Revision: 0 Document Number: ENG 1.71R Attmehment 1M for Reactor Containment Control Air Svttem Revision: N 'A Document

Title:

Svitem Cateeorv Dete mlnation in a Deenmmittlened Plant . Reactor Containment Control Air Svttem Provide a brief description of the change and a summary of the Safety Evaluation in the fonnat below,

l. Brief Description of Change and Safety Evaluation Summary:

His Safety Evaluation addresses the system category determinatier for the Reactor Containment Control Air System, as shown on P&lD No.16103 26054. Ponions of this system will be placed in " Lay.up" to support plant decommissioning activities and reactor coolant system (RCS) decontamination. Ponions of this system are being " Abandoned" because they are no longer needed he " Lay up" components will not be functioning while in that state and, therefort, a malfunctio.,12 not possible, ne

" Abandoned" components no longer have a required function in the plant and cannot malfunction, he Reactor Containtnent Control Air System does not directly communicate in any way with any system containing radioactivity, the Spent Fuel Pool or the Spent Fuel Handling System. Herefore, the probability of occunence of a fuel handling accident or a radwaste system failure, which are the only accidents applicable to the defueled state, is unaffected and there is no possibility for an accident of a different type.

His cht.nge does not foreclose (preclude) release of the site for possible unrestricted use it does not result in a significant environmental impact not previously reviewed. It does not remove reasonable assurance that adequate funds will be available for decommissioning.

e his change is safe and does not involve an unreviewed safety question. De change do;;s not affect the probability or consequences of previously evaluated accidents or malfunctions of equipment important to safety. Dere is no reduction in the margin of plant safety, in addition, this change does not involve an unresolved decommissioning question.

2. Reason for the Change:

The Reactor Containment Control Air System provides air for actuation of LD.TV.230 and the pressurizer spray valves.

Dese 5alves are within the RCS decontamiaation Dowpath and need control air for cycling. In order to suppon the decontamination effort, the majority of the Reactor Containment Control Air System will be placed in " Lay.up". De only ponions of the Reactor Containment Control Air System that are being " Abandoned" are the control air lines to the CAR fans and PORVs. De PORVs and CAR fans 3 and 4 are being " Abandoned", consequently control air to these components is no longer needed. Although CAR fans 1 and 2 will remain "Available", the dampers are spring loaded on lots of control air such that air Oow will pass through the CAR fan charcoal and HEPA filters. His is precisely the air Howpath needed during decommissioning; hence control air lines to CAR fans I and 2 are no longer needed, ne HEPA filters will be "Available" for non design basis paniculate Oltration during decomminioning. The charcoal filters, which were originally installed for removal ofiodine following a design basis accident, will also be useful for removing noxious vapors from the air during decommissioning.

Preparer b, A . S A AEA AJ / / d cM - Date M 70 *97 g .

26054 DOC

pc N DcY . 00 - 072 - 97 ACP 1.2 2 42 Rev. I MAJOR

(]

v Forrn 3 10 CFR 50.59 (b)(2) Report Page1of1 Safety Evaluation Number: SY EV 07-0042 0 Revision:

Document Number: rNG 1713& Attachment 123 for Nitrocen $vitem Revision: __ N/A Document

Title:

Svstem Catecorv Determination in a becommittiened Plant Nitrocen svitem Provide a brief description of the change and a summary of the Safety Evaluation in the fonnat below,

l. Brief Description of Change and Safety Evaluation Summary:

i nis Safety Evaluation addresses the system category determination for the Nitrogen System, as shown on P&lD No.16103-26055, Sheet 1 Ponions of this system will either remain "Available" or be placed in " Lay up" to suppon plant decommissioning activities and reactoc coolant system (RCS) decontamination. Ponions of this system are being

" Abandoned" because they are no longer needed.

The operation and required function of the "Available" components is unchanged from previous operation and function.

Therefore, the probability of occurrence or consequences of a previously evaluated malfunction is unchant,ed and there is no possibility for a maifunction of a different type. The " Lay up" components will not be functioning while in that state and, therefore, a malfunction is not possible. The " Abandoned" components no longer have a required function in the plant and cannot malfunction. The Nitrogen System does not directly communicate with the Spent Fuel Pool or the Spent Fuel Handling System. The "Available" or " Lay up" ponions of the system which are connected to a system containing

(]

v radioactivity are anchanged. The "Abr.ndoned" ponions c f the system will be isolated from any system containing radioactivity. Therefore, the probability of occurrence of a fuel handling accident c,r a radwaste system failure, which are the only accidents applicable to the defueled state, is unaffected and there is no possibility for an accident of a different type.

This change does not foreclose (preclude) releue of the site for possible unrestricted use.11 does not result in a significant.

1 environmental impact not previously reviewed. It does not remove reasonable assurance that adequate funds will be "

available for decotomissioning.

This change is ufe and does not involve an unreviewed safety question. The change does not affect the probability or consequences of previously evaluated accidents or malfunctions of equipment ;mponant to ..fety. There is no reduction in the margin of plant safety, in addition, this change does not involve an unresolved dec' mmissioning question.

2. Renon for the Change:

• The condensate storage tank is needed as a source of purified wattr throughout the decommissioning effon. Nitrogen is supplied to this tank to inhibit corrosion. The Nitrogen System must remain "Avallable" to supply the nitrogen and thereby inhibit tank corrosion and supplied systems corrosion. ne interface with the gaseous waste system will also remain "Available"in support of the gaseous waste system category, ne only portions of the Nitrogen System which are being placed in " Lay up" are the boundaries with the chemical volume and control system, w hich is in " Lay up" to support the RCS decontamination. The only portions of the Nitrogen System that are being " Abandoned" are a nitrogen supply to the s;.am generators and a boundary with the hydrogen system. The nitrogen supply to the steam generators is used for nitrogen blanket corrosion protection. This protection is no longer needed and the supply line can be " Abandoned". The hydrogen supply bounday with the nitrogen syster i is being " Abandoned" since the entire hydrogen system is being " Abandoned".

e kIef i bM :_ " M.1 ;1 Ob*D ste

• 2M5513 DOC l

NN

/ W.DCV 00- 07 3 fdlC 9743' Attachment 1 PMs t .t 5 HADDAM NECK PIANT Annual Report Summary of Changes Mark the Appropriate Choice:

Design Change Y Setpoint Change Test Tech Requirements Manual Change Experiment Procedure Changa Jumper Bypass

1. Change Number: Revision Number:

Title:

Evstem categerv Deteminatien in a Decomimmiened plant .

}{MdIDgen can system

2. Description of Change:

This Safety Evaluation addresses the system ategory determination for the Hydrogen Gas System, as shown on P&ID No. 16103-26055, Sheet

2. The entire system is being ' Abandoned".

3. Reason for the Change:

The Hydrogen Gas System supplies hydrogen to the main generator and volume control tank (VCT). Since the main generator is no longer being used, its hydrogen supply is being ' Abandoned". Although the VCT is in lay-up, its associated hydrogen gas supply is not presently needed, nor will it be needed during the reactor coolant system decontamination. Therefore, this supply is also being ' Abandoned".

4. Safety Evaluations

a. This change was safe for the following reasons:

Per this 10CFR50.59 Safety Evaluation Review, the change does not affect the probability or consequences of previously evaluated accidents or malfur.,tions of equipment important to sefety.

There is no reduction in the margin of plant safety,

b. This change does not c.onstitute AN UNREVIEWED SAFETY QUESTION because THEFE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPCRTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:

pcN Ocy- 00 073

• 97 SY-IW 97-0082 Page 243 CY is in a defueled condition and has notified the NRC of its decision to permanently cease power operation, puel has been permanently removed from the reactor vessel. As such, much of the equipment which previously hsd a safety related and/or operational l

function during normal plant operation, may no longer be important to safety or required to be operable. This includes the Hydrogen Gas System. Previously, during normal plant operation the Hydrogen Gas System supplied hydrogen to the main generator and volume control tank (VCT) . With the plant in a defueled condition, these functions are no longer required and a malfunction of the subject equipment is of no consequence. In addition, the Hydrogen Gas System will be isolated from any system containing radioactivity and does not directly communicate in any way with the Spent Fuel Pool or the Spent Fuel Handling System.

Therefore, the probability of occurrence or consequences of a fuel handling accident or a radwaste system failure, which are tha only accidents applicable to the defueled condition, is unaffected.

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 " Abandoned" Hydrogen Gas System components no longer have a required function in the plant and, therefore, cannot malfunction.

Also, there is no possibility of an accident of a different type created because the Hydrogen Gas System will be isolated from any system containing radioactivity and does not directly communicate in any way with the Spent Fue*. Pool or the Spent Fuel Handling System.

THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement ist

• Abandoned" systems are those which no longer have Technical Specification requirements associated with them. The Hydrocen Gas System never had any Technical Specification requirements associated with it for any mode of plant operation. Thecefore, there is no impact on the margin of safety as defined in the Bases of any Technical Specification.

c. This change did not require a chango to the Technical Specifications.

In addition, a 10CFR30.82 Decommissioning Review was also performed

.that concluded that this change does not constitute an unreviewed

-decommissioni.g question for the following reasons:

pcN pcy. oo - 075 - 97

',' Tf. EV

• 97- 0011 Poge 1 of 1 Placing the Hydrogen Ges system in the ' Abandoned" category does not foreclose (preclude) release of the site for possible unrestricted use. There is no potential failure mode which could create a major l

spill or rupture, causing a significant radioactive contamination of the soil or buildings which could not easily be decentsainated. It also does not permanently install or retain major strut.ures or equipment on site, or render it difficult to remove systems, structures or components from the site.

Placing the Hydrogen Gas system in the ' Abandoned" category does not result in a t,ignificant environmental impact not previously reviewed.

It does not result in the discharge of radioactivity or chemicals, either by a controlled release, or due to a malfunction or failure of equipment. It also does not affect the site environment (e.g.,

terrain, noise, visual appearance, transmission lines) in any way.

Placing the Hydrogen Gas system in the " Abandoned" category does not remove reasonable assurance that adequate funds will be available for decommissioning. The cost of procedure revisions, ITFSAR revisions, etc., taken in total, are not likely to jeopardize decommissioning funds.

Thert is no applicability to the PSDAR of placing the Hydrogen Gas system in the " Abandoned" category since that document has not yet been developed for CY. Rather, the system categorization process will be used as input to the PSDAR development.

.= _

_ = = = _ = _

e

~

k - h d ~/ -Od/ 4

/ e , i et 3 Attachment 1

• EADDAN NECK PLANT '

Annual Report Summary of Changes Mark the Appropriate Choices Design Change M Setpoint Change Test Tech Requirements Manual Change Experiment Procedure Change Jumper Bypass

1. Change Number Revision Numbers .

Title:

Eystem cataeory Determinatien in a Beenmmiamiened plant -

Mimcellaneous Bottled Can Evatam

2. Description of Change This Safety Evaluation addresses the system category determination for the Miscellaneous Bottled Gas System, as shown on P&ID No.16103-26055, Sheet 1. The majority of this system will remain 'Available",

however, portions of the system are being

• Abandoned".

3. Reason for the Change:

The Miscellaneous Bottled Gas System.provides a supply of gases (oxygen, helium, argon and acetylene) to the chemical laboratory; a supply of propane gas as an ignition source for the auxiliary boilers; and a nitrogen gas supply dedicated to the post-accider,t sampling system. The entire Miscellaneous Bottled Gas System will remain 'Available", except for the nitrogen gas supply. Since the post-accident sampling system is no longer needed and is being

• Abandoned", the dedicated nitrogen gas supply to this system is likewise being 'Atandoned".

4. Safety Evaluation:

a. This change was safe for the following reasons:

Per this 10CFR50.59 Safety Evaluation Review, the change does not affect the probability or consequences of previously evaluated accidents or malfunctions of equipment important to safety.

There is no reduction in the margin of plant safety,

b. This change does not constitute AN UNREVIEWED SAFETY QUESTION because

/ pcy pcv - oo - 07+ -

gy , GV . 9 7 - 00 t {7 Pote 7 *f 3 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:

CY is in a defueled condition and has notified the NRC of its decision to permanently cease power operation. Puel has been permanently removed from the reactor vessel. As such, much of the equipment which previously had a safety related and/or operational function durjng normal plant operation, may no longer be important to safety or required to be operable. This includes the Misec11aneous Bottled Gas System. The operation and required function of the system components which will remain 'Available" is unchanged from previous operation and function. Therefore, the probability of occurrence of a previously evaluated malfunction is also unchanged. The " Abandoned" components no longer have a required function in the plant and, therefore, cannot malfunction.. In addition, the Miscellaneous Bottled Gas System does not directly communicate in any way with any system containing radioactivity, the Spent Puel Pool or the Spent Fuel Handling System. Therefore, the probability of occurrence or consequences of a fuel handling accident or a radwaste system failure, whfeh are the only accidents applicable to the defueled condition, is unaffected.

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 operation and required function of the Miscellaneous Bottled Gas System components which will remain 'Available" is unchanged from previous operation and function. Therefore, there is no possibility of a different type of malfunction. The " Abandoned" components no longer have a required function in the plant and, therefore, cannot malfunction. Also, there is no possibility of an accident of a different type created because the Miscellaneous Dottled Gas System does not directly communicate in any way with any system containing radioactivity, the Spent Puol Pool or the Spent Fuel Handling System.

THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:

• Available" and ' Abandoned" systems are those which no longer have Technical Specification requirements associated with them. The Miscellaneous Bottled Gas System, never had any Technical Specification requirements associated with it for any mode of plant operation. Therefore, there is no impact on the margin of safety as defined in the Bases of any Technical Specification.

pcw Dcf 074 97 cf . fiV 004

y h se 3 d' 1

c. This change did not require a change to the Technical Specifications.

In addition, a 10CFR$0.82 Decommissioning Review was also performed that concluded that this change does not constitute an unreviewed decommissioning question for the following reasons:

Placing the Miscellaneous Bottled Gas System in the *Available" and

' Abandoned" category does not foreclose (preclude) release of the site for possible unrestricted use. There is no potential failure mode which could create a major spill or rupture, causing a significant radioactive contamination of the soil c. buildings which could not easily be decontaminate'd. It also does not permanently install or retain major structures or equipment on site, or render it difficult to remove systems, structures or components from the site.

? lacing the Miscellaneous Bottled Gas System in the *Available" and

' Abandoned" category does not result in a significant environmental impact not previously reviewed. It does not result in the discharge of radioactivity or chemicals, either by a controlled release, or due to a malfunction or failure of equipment. It also does not affect the site environment (e.g. , terrain, noise, visual appearance, transmission lines) in any way.

Placing the Miscellaneous Bottled Gas System in the 'Available" and

' Abandoned" category does not remove reasonable assurance that adequate funds will be available for decommissioning. The cost of procedure revisions, UFSAR revisions, etc., taken in total, are not likely to jeopardize decommissioning funds.

There is no applicability to the PSDAR of placing the Miscellaneous Bottled Gas System in the 'Available" and ' Abandoned" category since that document has not yet been developed for CY. Rather, the system categorization process will be used as input to the PSDAR development.

l

DC CY ~ 00 *

"O G

• 97 4

/ L. < R Attachment 1

• 1 MADDAM NECK PLANT .

Annual Report i

Summary of Changes Mark the Appropriate Choice Design Change Y Setpoint Change '

Test Tech Requirements Manual Change  !

Experiment Procedure Change Jumper Bypass

1. Change Number: Revision Number:

Title:

Eystem C*atagerv Determination in a Decommimmiened Plant -

Mydrogen D Wer system

2. Description of Change j This Safety Evaluation addresses the system category determination j for the Hydrogen Dryer System, as shown on P&ID No. 16103-26055, Sheet 4. The entire system is being ' Abandoned".

J

3. Reason for the Change:

The Hydrogen Dryer System circulates moisture-free hydrogen to the main generator. Since the main generator is no longer being used, this system is being ' Abandoned",

4. Safety Evaluations

a. This change was safe for the following reasons:

Per this 10CFR50.59 Safety Evaluation Review, the change does not affect the probability or consequences of previously evaluated accidents or malfunctions of equipment important to safety.

There is no reduction in the margin of plant safety.

b. This change does not constitute AN UNREVIEWED SAFETY QUESTION because

- THERE IS NO INCREASE IN THE PROBABIL'ATY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION 0:' EQUIPMENT IMPORtANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is CY is in a defueled condition and has notified the NRC of its decision to permanently cease power operation, ruel has been vfy--mg -es--e-n-. m- -c-ww y- m--=---r -m -ma- gaaa-m-r - - - - y---y--re,v_-v.-tr- r ---i- ---s-- -u----v- m 1

Dc4 bcY 075- 97 9(- CV 00 63 Pbse 2 ef 5 permanently removed from the reactor vessel. As such, much of the equipment which previously had a safety related and/or operational function during nornal plant operatien, may no longer be important to safety or required to be operable. This includes the Hydrogen Dryer System. Previously, during normal plant operation the Hydrogen Dryer System circulated moisture-free hydrogen to the main generator. With the plant in a defueled condition, this function is no longer required and a malfunction of the subject equipment is of no consequence. In addition, the Hydrog2n Dryer System does not directly communicate in any way with any system containing radioactivity, with the Spent Fuel Pool or the Spent Fuel Handling System. Therefore, the probability of occurrence or consequences of a fuel handling accident or a radwaste system failure, which are the only accidents applicable to the defueled condition, is unaffected.

THE POSSIBILITY FOR AN ACCIDENT OR MALFUNC'. ION OF A DIFFERENT TYPE THAN A!Fl EVALUATED PREVIOUSLY IN THE SAFETY MIALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:

The " Abandoned" Hydrogen Dryer System components no longer have a required function in the plant and, therefore, cannot malfunction.

Alco, there is no possibility of an accident of a different type created because the Hydrogen Dryer System does not directly communicate in any way with any system containing radioactivity, the Spent Fuel Pool or the Spent Fuel Handling System.

THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION RAS NOT BEEN REDUCED. The basis for this statement is:

' Abandoned" systems are those which no longer have Technical Specification requirements associated with them. The Hydrogen Dryer System never had any Technical Specification requirements associated with it for any mode of plant operation. Therefore, there is no impact on the margin of safety as defined in ths Bases of any Technical Specification.

c. This change did not require a change to the Technical Specifications.

In addition, a 10CFR50.82 Decommissioning Review was also performed that concluded thas this change does not constitute an unreviewed decommissioning question for the following reasons:

Placing the Hydrogen Dryer System in the " Abandoned" category does not foreclose (preclude) release of the site for possible unrestricted use. There is no potential failure mode which could create a major spill or rupture, causing a significknt radioactive contamination of the soil or buildings which could not easily be

l

/ pcN pcy 00- 070-97 Sy - Ov 000 l

Page 3 of 3 Y

decontaminated.- It also does not permanently install or retain major structures or equipment on site, or render it difficult to remove systems, structures or components from the site.

- Placing the Hydrogen Dryer System in the ' Abandoned" category does not result in a significant environmental impact not previously-reviewed. It does not result in the discharge of radioactivity or chemicals, either by a controlled release, or due to a malfunction or failure of equipment. It also does not-affect the site environment (e.g., terrain, noise, visual appearance,-transmission lines) in any way.  !

Placing the Hydrogen Dryer System in the ' Abandoned

• category does not remove reasonable assurance that adequate funds will be available for decommissioning. The cost of procedure revisions, UFSAR revisions, etc., taken in total, are not likely to jeopardize decommissioning funds.

There is no applicability to the PSDAR of placing the Hydrogen Dryer system in the ' Abandoned" category since that document has not yet been developed for CY. Rather, the system categorization process will be used as input to the PSDAR development.

m = = ==

Y M

~b F peu bc'( op 97 ACP 1.2 2.42

/ .

Rev.1 MAJOR Form 3 10 CFR 50.59 (b)(2) Report Page1of1 Safety Evaluation Number: SY.EV.97 0045 Revision: 0 Document Number: ENO 17136_ Attmehment 12 $ for Fire Prnteetinn Svetem Revisfor.: N/A Document

Title:

Svetem cateenrv Detervnination in a Deenmmittinne'* Nnt . Fire Pretectinn Svetem Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Brief Description of Change and Safety Evaluation Summary:

nis Safety Evaluation addresses the system category determination for the Fire Protection System, as shown on P&lD No.

16103 26056, Sheets 1 through 16. Portions of this system will remain " Operable" or "Available' to support plant decommissioning activities. A portion of this system was previously" Abandoned"in place.

De operation and required function of the " Operable" and "Available" components is unchanged from previous operation and function. Tht.cfore, the probability of occurrence or consequences of a previously evaluated malfunction is unchanged and there is no possibility for a malfunction of a different type. Maintaining portions of the Fire Protection System

" Operable" will satisfy all applicable requirements of Technical Requirements Manual (TRM) Se tions 11.1.B. C. F and G.

De functions of the "Available" portions of the system do not change, ne Fire Protection System does not directly 4

communicate in any way with any system containing radioactivity, the Spent Fuel Pool or the Spent Fuel Handling System.

Therefore, the probability of occurrence of a fuel handling accident or a radwaste system failure, which are the only accidents applicable to the defueled state,is unaffected and there is no possibility for an accident of a different type.

nis change does not foreclose (preclude) release of the site for possible unrestricted use, it does not result in a significant environmental impact rot previously reviewed, it does not remove reasonable assurance that adequate funds will be availabie for decommissioning.

nis change is safe and does not involve an unreviewed safety question, ne change does not affect the probability or consequences of previously evaluated accidents or malfunctions of equipment important to safety. Dere is no reduction in the margin of plant safety, in addition, this change does not involve an unresolved decommissioning question.

2. Reason for the Change:

The Fire Protection System supplies timely protection against fire in various plant areas. De entire system will remain either

" Operable" or "Available" except for one underground section of piping previously " Abandoned" in place. The " Operable' ponions include the electric fire pump; the diesel fire pump and fuel tank; strainers; supply headers and valves for the turbine

, building fire hoses, service building sprinklers, diesel generator deluge, turbine building sprinklers, auxiliary boiler room sprinklers, turbine oil tanks room sprinklers, cable spreading area sprinklers in the PAB, service building and turbine building; screenwell building sprinklers, maintenance shop sprinklers, and deluge spray headers associated with the turbine.

generator, turbine building hall crane bay deluge, and cable chase sprinklers. Also remaining " Operable" is a water supply to the service water system which is used for loss of service water flow scenarios. The "Available" portions are the fire protection pressure maintenance pump and supply headers and valves for the turbine building outside areas, I&C building, areas outside the maintenance shop, transformer deluge, switchgear building sprinkler systems and hose reels (except for cable chase sprinklers which are classified as " Operable"), and warehouse 2. It also includes fire protection for the engineering building, containment entry building HP count lab and maintenance module b

Preparer D. A. 5A B 6AM / h ! : ?w Date V e

• M

• 9 '1 26050 DOC

• ~

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IS M Id)

Attachment 1

• EADDAM NECK PLANT ,

Annual Report '

Summa n of Changes Mark the Appropriate Choice Design Change s setpoint Change Test Tech Requirements Manual Change Experiment Procedure Change Jumper Bypass

1. Change Number: Revision Number:

Title:

Svatem cateceW Determinatien in a Becommittiened Plant -

Peat-accidet.* Emmnline system

2. Description of Change: I This safety Evaluation addresses the system category determination for the Post-Accident Sampling System, as shown on P&ID No. 16103-26057 The majority of this system is being " Abandoned", however, small portions of the system which interface and comprise the boundaries with other systems will either remain 'Available" or be placed in

• Lay-up".

3. Reason for the Change:

The Post-Accident Sampling System is used for reactor coolant system (RCS) sampling and containment sampling following an accident. Since the reactor is defueled, and will remain so, an RCS accident is not possible and the system is no longer required. Therefore, the Post-Accident sampling System is no longer needed and is being

" Abandoned'.

4. Safety Evaluations

a. This change was safe for the following reasons:

Per this 20CFR50.59 Safety Evaluation Review, the change does not affect the probability or consequences of previously evaluated accidents or malfunctions of equipment important to safety.

There is no reduction in the margin of plant safety,

b. This change does not constitute AN UNREVIEWED SAFETY QUESTION because ,

.- DCN DC W - CV "T.

97~000040- 07T- 97

,- D98 2d3 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 ist CY is in a defueled condition and has notified the NRC of its decision to permanently cease power operation. Fuel-has been pomanently removed from the reactor vessel. As such, much of the equipment _which previously had a safety related and/or operational

-function during normal plant operation, may no longer be important i

to safety or required to be operable. This includes the Post-Accident Sampling System. The operation and required function of the system components which will remain 'Available" is unchanged ,

from previous operation and function. Therefore, the probability of occurrence of a previously evaluated malfunction is also- ,

unchanged. The coniponents in

• Lay-up* will not be functioning while in that state and, theirefore, a malfunction of the subject equipment is not possible. The

• Abandoned" components no longer have a requi-

• function in the plant and, therefore, cannot malfunction.. In addition, the Post Accident Sampling System does not directly communicate 6n any way with the Spent ruel Pool or the Spent Fuel Handling system. Once the system is placed in the

'Available*,

• Lay-up" and " Abandoned" categories, it will be isolated from any system containing radioactivity. Therefore, the probability of occurrence or consequences of a fuel handling accident or a radwaste system failure, which are the only accidents applicable to the defueled condition, is unaffacted.

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 operation and required function of the *Available" Post-Accident Sampling System components is unchanged from previous operation and function. Therefore, there is no possibility of a ditferent type of malfunction. The components in

• Lay-up" will not_be_ functioning while in that state and, therefore, a malfunction of the subject equipment is not possible. The

' Abandoned" components no longer have a required function in the plant and, therefore, cannot malfunction. Also, there is no possibility of an accident of a different type created because the Post-Accident Sampling System will be isolated from any system containing radioactivity and does not directly comunicate in any way with the Spent Fuel Pool or the Spent Fuel Handling System.

THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:

,e ' pc4 Oc.T on - 97 gy - CV 0010 Phge 3 of 7

/

// 'Available",

• Lay-up" and ' Abandoned" systems are those which no longer have Technical Specification requirements associated with l them. Previously administrative controls were placed on the Post-i Accident sampling System by Tech Spec 6.15 to reduce leakage from systems outside containment that could contain highly radioactive fluids during a serious transient or accident, and by Tech Spec 6.16 to maintain the capability to obtain and analyze RCS and containment atmosphere samples under accident conditions. The Post-Accident sampling System does not have any Technical Specification requirements associated with it for the defueled condition. Therefore, there is no impact on the margin of safety as defined in the Bases of any Technical Specification.

c.- This change did not require a change to the Technical Specifications, a In addition, a 10CFR50.82 Decommission 1pc Review was also performed that concluded that this change does not constitute an unreviewed decommissioning question for the following rearons:

l Placing the Post-Accident Sampling System in the 'Available",

• Lay-up" and ' Abandoned" categories does not foreclose (preclude) release of the site for possible unrestricted uss. There is no potential failure mode which could create a major spill or rupture, causing a significant radioactive contamination of the soil or buildings which could not easily be decontaminated. It also does not permanently install or retain major structures or equipment on site, or render it difficult to remove systems, structures or components from the site.

Placing the Post-Accident Sampling System in the 'Available",

• Lay-4 up" and ' Abandoned" categories does not result in a signsficant environmental impact not previously reviewed. It does not result in the discharge of radioactivity or chemicals, either by a controlled release, or due to a malfunction or failure of equipment. It also does not affect the site environment (e.g., terrain, noise, visual appearance, transmission lines) in any way .

Placing the Post-Accident Sampling System in the 'Available",

• Lay-up" and " Abandoned" categories does not remove reasonable assurance that adequate funds will be available for decommissioning. The cost of procedure revisions, UFSAR revisions, etc., taken in total, are not likely to jeopardize decommissioning funds.

There is no applicability to the PSDAR of placing the Post-Accident Sampling System in the *Available",

• Lay-up" and ' Abandoned" categories since that document has not yet been developed for CY.

Rather, the system categorization process will be used as input to the PSDAR development.

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pew ocy - oo- 07e-97 ACP 1.2 2.42 Rev,1 MAJOR Form 3 e 10 CFR $0.59 (b)(2) Report Page l of t safety Evaluation Number: E%EW97 0062 bvision: 0 Document Number _ENG 1.7156 n t*ehment e 12 2 for He**Ine 0**m & Paadaamate Eve *= Revision: N'A DocutneDt Title! .E.'d" F*'*WV Det*f"Ia*'Ian If1 A Ne^""I"*I^^*d Plaaf H**'ia" *"- A Pandaa**'* Svetem i

l Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

l. Brief Description of Change and Safety Evaluation Summary:

Ws Safety Evaluation addresses the system category determination for the Heating Steam & Condensate System, as shown on P&iD No.16103 26052, Sheets I through 8. Portions of this system will remain "Available" to support plant decommissioning activities and other portions of this system are being " Abandoned" beause they are no longw needed.

The operation and required function of the "Available" components is unchanged from previous operation and function.

Derefore, the probability of occurrence or consequences of a previously evaluated malfunction is unchanged and there is no possibility for a malfunction of a different type. The " Abandoned" components no longer ha"e a required function in the plant and cannot malfunction. The Heating Steam & Condensate System does not directly communicate in any way with any system containing radioactivity, the Spent Fuel Pool or the Spent Fuel Handling System. Thwefore, the probability of occurrence of a fuel handling accident or a radwaste system failure, which are the only accidents applicable to the defueled state, is unaffected and there is no possibility for an accident of a different type.

The change does not foreclose (preclude) release of the site for possible unrestricted use, it does not result in a significant l enviro.unental impact not previously reviewed, it does not remove reuonable assurance that adequate funds will be

available for decommissioning.

1 This change is safe and does not involve an unreviewed safety question, ne change does not affect the probability or consequences of previously evaluated accidents or malfunctions of eqWoment imponant to safety. Then is no reduction in the margin of plant safety. In addition, this change does not involve an unruolved decomminioning question.

2. Reason for the Change:

The Heating Steam & Condensate System employs an oil fired boiler to produce heating steam. This steam is used to supply various heating and hot water loads throughout the plant. Therefore, this system must remain "Available". The majority of the system will remain "Available", however, there are several portions of the system that are being " Abandoned". The

" Abandoned" portions of the Heating Stsam & Condensate System include steam supplies to the boron evaporators; the degasifier pre heater; the waste evaporator reboller; the caustic dilution water heater; and the containment humidifier; a boundary with the " Abandoned" feedwater & condensate system; and a 10" steam line that was previously " Abandoned"in place.

Preparw -- p ram h / 2- = ie r ~ Date - ed-22-1r

= =_

260$t) DOC

DcN pcf-00 079 97 ACP 1.2 2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page l of t Safety Evalaation Number: SY EV-97 00M htevision: 0 Document Number: ENG 1%156. Attachment IM for Switcheer Buildine MVAC System Revision: N'A Document

Title:

.Svetem ratecorv Determination in a Drenmmittiened Plant Switcheese Buildine MVAC System Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Brief Description of Change and Safety Evaluation Summary:

his Safery Evaluation addresses the system category determination for the Switchgear Building HVAC System, as shon on P&ID No. 16103 26063, Sheets I and 2. Portions of this system will remain " Operable" or "Available' to support plant decommissioning activities.

The o3..- tion and required function of the " Operable" and "Available" components is unchanged trom previous operation and function. Therefore, th: probability of occurrence or consequences of a previous'j evaluated malfunction is unchanged and there is no possibility for a malfunction of a different type. Maintaining ponions of the Switchgear Building HVAC Syrtem " Operable" will support operation of equipment needed to meet the requirements of 10CFR50A8 for fire protection and to suppon " Operable equipment being used to maintain spent fuel cooling. De functions of the "Available" ponions of the system do not change. The Switchgear Building HVAC System does not directly communicate in any way with any system containing radioactivity, the Spent Fuel Pool or the Spent Fuel Handling System. Therefore, the probability of occunence of a fuel handling accident or a radwaste system failure, which are the only accidents applicable to the defueled state, is unwTected and there is no possibility for an accident of a different type.

This change does not foreclose (preclude) release of the site for possible unrestricted use. It does not result in a significant environmental impact not previously revie n,1. It does not remove reasonable assurance that adequate funds will be available for decommissioning.

His change is safe and does not involve an unreviewed safety question. The change does not affect the probability or consequences of previously evaluated accidents or malfunctions of eqtipment important to safety. There is nueduction in the margin of plant safety, in addition, this chang;. does not involve an unresolved decommissioning question.

2. Reason for the Change:

De Switchgear Building HVAC System provides a suitable environment for personnel and equipment operation. The entire system will remain either " Operable" or "Available". The " Operable" penions of the system include the switchgear building 3rd floor HVAC subsystem with 'B'switchgear room supp!y fan and battery room heater and exhaust fan. The 'B' switchgear ventilation subsystem is required to be " Operable" to support operation of'B' switchgear room components. The

'D' service water pump is powered from Bus 11 which 1.; L,cated in ,he 'B'switchgear room. The 'B' battery supports EDG 2B start, breaker operation, and arxiliary components requi+ed for diesel operation. fire protection for the engineering building, containment entry building. HP ccent lab and maintenance module. The "Available" portions of the system include the switchgear building ist and 2nd floor HVAC subsystems; and the heating, cooling and filtration po'tions of the 3rd floor HVAC subsystem. This equipment must remain "Available" to support a suitable and safe environment for personnel and equipment.

Preparer _ b A . .sA B CA A) / & " le m Date . OT-e9-97 260631 DOC g .. . - -

DcN DcY-00-080 97 ACP 1.2 2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) .'leport Page 1 of1 Safety Evaluation Number: SY.FV.97 0049 Midon: 0 Document Number: ENG 1.71% Attmehment 12.2 for Floor. Roof & Fouirment Dnin System Revision: N/A Document

Title:

_ System catecory Determination in a Decomminioned Plant . Floor. Roof & Enninment Dnin Svsem u

Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

l. Brief Description of Change and Safety Evaluation Summary:

ThL Safety Evaluation addresses the system category determination for the Floor, Roof & Equipment Dra shown on P&lD No.

16103 26064, Sheets 1 through 7. His entire system will remain"Available' to support plant decommissioning activities.

The operation and required function of the "Available" components is unchanged from previous operation and fu Therefore, the probability of occurrence or consequences of a previously es aluated malfunction is unchange possibility for a malfunction of a different type. De Floor, Roof & Equipment Drain Syste'n does not directly c in any way with any system containing radioactivity, the Spent Fuel Pool or the Spent Fuel Handling System probability of occurrence of a fuel handling accident or a radwaste system failure, which are the only acciden the defueled state, is unaffected and there is no possibility for an acci 'ent of a different ype. -

This change does not foreclose (preclude) release of the site for possible unrestricted use. It does not resul environmental impact not previously reviewed. It does not remove reasonable assurance that adegrr e funds will be i available for decommissioning.

This change is safe and does not involve an unreviewed safety question. The change does not affect the pr consequences of previously evaluated accidents or malfunctions of equipment important to safety. There is no reduction in the margin of plant safety. In addition, this change does not involve an unresolved decommissioning question.

2. Reason for the Change:

The enti y Floor, Roof & Equipment Drain System will remain "Available" These drains will provide building dr throughu 6e decommissioning effort and will be one of the last systems in each building to be dismantled.

Preparer

b. A SA R EA AJ / An ?=h: a Date . o 6 - l2 - 1'1 d

260643 t0C s

- i mi

OcM DcT- 00-081- 97 ACP 1.2 2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Rerort Page1of1 ,

Safety Evaluation Number: SV-EV 97-0063 Revision: 0 Document Number: ENG 17156 A"=rkment 12 ? for ':aatie Svetem , 8tevision: N/A Document

Title:

Svetem (%'..rrv neterminatinn in a naramml=elaned Plant - Eentic Svetem Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Brief Desetiption of Change and Safety Evaluation Summary:

This Safety Evalus'!on addresses the system category determination for the Sepic System, as shown on P&ID No.16103 26065, Sheets 1 through 5. He majority of this system will remain "Available" m support plant decommissioning ectivties.

Portions of this system are being " Abandoned" because they are no longer needed.

He operation and required function of the "Available" components is unchanged from presious operation and function.

Derefore, the probability of occwrence or consequences of a previously evaluated malfunt. tion is unchanged and there is no 4 possibility for a analfunction of a different type. De " Abandoned" components no longer save a required function in the plant and cannot malfunction. De Septic System does not directly communicate in any way with any sptem containing radioactivity, the Spent Fuel Pool or the Spent Fuel Handling System. Therefore, the probability of occurrence of a fuel handling accident or a radwaste system failure, which are the only accidents applicable to the defueled state, is unaffected and there is no possibility for an accident of a different type.

His change does not foreclose (preclude) release of the site for possible unrestricted use. It does not result in a significant environmental impact not previously reviewed it does not remove reasonable assurance that adequate funds will be available for decommissioning.

His change is safe and does not involve an unreviewed safety question. De change does not affect the probability or consequences of previously evaluated accidents or malfunctions of equipment important to safety. Here is no reduction in the margin of plant safety, In addition, this change does not involve an unresolved decommissioning question.

2. Reason for the Change:

The normal plant Septk System must remain "Available" to provide sanitary disposal and processing for plant showers, tollets, and sinks. De only item being " Abandoned"is septic tank TK 89-1 A, which was previously " Abandoned"in place.

Preparer *b a KA B e A n / & + 8: Am Date _ 0 7- 0 9 - 9 ~7 26o653. DOC

don DcY 082 -97

/ 5Y-O-f7@99' Ibge iAda- of 3 974o Attachment 1 HADDAM NECK PLANT Annual Report Summary of Changes Mark the Appropriate Choice:

Design Change X Setpoint Change Test Tech Requirements Manual Change Experiment Procedure Change Jumper Bypass

1. Change Number: Revision Number:

Title:

System cateceW Determinatien in a Decem4 ssioned plant -

Domestic Water Mystem

2. Description of Change:

This Safety Evaluation addresses the system category determination for the Domestic Water System, as show on P&ID No. 16103-26067, Sheets 1 through 6. The entire system will remain "Available" to supply water for plant personnel.

3. Reason for the Change:

The Domestic Water System will remain "Available" to supply water for plant personnel. This system supplies all domestic water needs such as showers, toilets, and sinks.

4. Safety Evaluation:

a. This change was safe for the following Tensons:

Per this 10CFR50.59 Safety Evaluation Review, the cha.nge deat., not affect the probability or consequences of previously evaluated accidents or malfunctions of equipment important to safety.

There is no reduction in the margin of plant safety,

b. This change does not constitute AN UNREVIEWED SAFETY QUESTION hecause 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 fdr this statement is:

peg pcy 00 007-97 gy . av 00Z+

Pc$e 2 4 3 CY is in a defueled condition and has notified the NRC of its decision to permanently cease power operation. Fuel has been permanently removed from the reactor vessel. As such, much of the equipment which previously had a safety related and/or operational function during normal plant operation, may no longer be important to safety or required to be operable. This includes the Domestic Water System. The operation and required function of the system components which will remain 'Available" is unchanged from previous operation and function. Therefore, the probability of occurrence of a previously evaluated malfunction is also unchanged. In addition, the Domestic Water System does not directly communicate in any way with any system containing radioactivity, the Spent Fuel Pool or the Spent Fuel Handling System. Therefore, the probabilic/ of occurrence or consequences of a fuel handling accident or a radwaste system failure, which are the only. accidents applicable to the defueled condition, is unaffected.

THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFEYENT TYPE THAN ANY EVALUATED PREVIOUSLY IN Th3 SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:

The operation and required function of the Domestic Water System components which will remain "Available" is unchanged from previous operation and function. Therefore, there is no possibility of a different type of malfunction. Also, there is no possibility of an accident of a different type created because the Domestic Water System does not directly communicate in any way with any system containing radioactivity, the Spent Fuel Pool or the Spent Fuel Handling System.

THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:

"Available" systems are those which no longer have Technical i specification requirements associated with them. The Domestic Water System never had any Technical Specification requirements associated with it for any mode of plant operation. Therefore, I there is no impact on en, margin of safety as defined in the Bases of any Technical Specification.

c. This change did not require a change to the Technical Specifications.

In addition, a 10CFR50.82 Decommissioning Review was also performed that concluded that this change does not constitute an unreviewed decommissioning question for the following reasons:

' peg pcy . oo. os2 -97

- sy . sv . 97 - 002+

Pogs 3 of 3 Placing the Domestic Water System in the "Available" category does not foreclose (preclude) release of the site for possible unrestricted use. There is no potential failure mode which could create a major spill or rupture, causing a significant radioactive contamination of the soil or buildings which could not easily be decontaminated. It also does not permanent,/ install or retain major structures or equipment on site, or render it difficult to remove systems, structures or components from the site.

Placing the Domestic Water System in the "Available" category does not result in a significant environmental impact not previously reviewed. It does not result in the dischaige of radioactivity or chemicals, either by a controlled release, or due to a malfunction or failure of equipment. It also does not affect the site environment (e.g., terrain, noise, visual appearance, transmission lines) in any way.

Placing the Domestic Water System in the 'Available" category does not remove reasonable ascurar.ce that adequare funds will be available for decommissioning. Tr.e cost of procedure revisions, UFSAR revisions, etc., taken in total, are not likely to jeopardize decommissioning funds.

There is no applicability to the PSDAR of placing the Comestic Water System in the "Available" category since that document has not yet been developed for CY. Rather, the system categorization process will be used as input to the PSDAR development.

1

-mmii-m. -s

" 5 y'- sy- 974/ O ocu va-w 083-9

/ Pqs f d 3 Attachment 1

• EADDAM NECK iLANT .

Annual Report Summary of Changes Mark the Appropriate Choice:

Design Change Y Setpoint Change Test Tech Requirements Manual Change Experiment Procedure Change Jumper Bypass

1. Change Number: Revision F Jer:

Title:

System categerv Deteminati on in a Decoxmi mmioned Plant -

Hydrarine Feed System

2. Deweription of Chant This Safety Euluation addresses the system category determination for the Hydrazine Feed System, as shown on P&ID No. 16103-26068. The entire system is being ' Abandoned".

3. Reason for the Change:

The Hydrazine Feed System is used for corrosion centrol in the Feedwater & Condensate Systems. Since the Feedwater &' Condensate Systems will no longer be used, the Hydra:ine Feed System is no longer required and is being 'Abandvned".

4. Safety Evaluation:

a. This change was safe for the following reasons:

Per this 10CFR50.S9 Safety Evaluation Review, the change does not affect the probability or consequences Uf previously evaluated accidents or malfunctions of equipment important so safety.

There is no reduction in the margin of plant 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:

DCl4 DC'i- C* * *

• U sy . o v o$ 10 Page 1 d 3 CY is $n a defueled condition and has notified the NRC of its decision to permanently cease power operation. Fuel has been permanently removed from the reactor vessel. As such, much of the equipment which previously had a safety related and/or operational function during normal plant operation, may no longer be important to safety or required to be operable. This includes the Hydrazine Feed System. Previously, during normal plant operation the Hydrazine Feed System was used for corrosion control in the Feedwater & Condensate Systems. With the plant in a defueled condition, this function is no longer required and a malfunction of any of the subject equipment is of no consequence. In addition, the Hydrazine Feed System does not directly communicate in any way with any system containing radioactivity, the Spent Fuel Pool or the Spent Puel Handling System. Therefore, the probability of occurrence or consequences of a fuel handling accident or a radwaste system failure, which are the only accidents applicable to the defueled condition, is unaffected.

THE POSSIBILITY FOR AN ACCIDENT Ok 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 ' Abandoned" Hydra:ine Feed System components no lenger have a required function in the plant and, therefore, cannot malfunction.

Also, there is no possibility of an accident of a dtfferent type created because the Hydra:ine Feed System does not directly l

communicate in any way with any system containing radioactivity, the Spent Fuel Pool or the Spent Fuel Handling System.

THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:

• Abandoned" systems are those which no longer have Technical

.pecification requirements associated with them. The Hydrazine Feed System never had any Technical Specification requirements asso ' ted with it for any mode of plant operation. Therefore, the as no impact on the margin of safety as defined in the Bases of any Technical Specification,

c. This change did not require a change to the Technical Specifications.

In addition, a 10CFR50.82 Decommissioning Review was also performed that concluded that this change does not constitute an unreviewed decommissioning question for the following reasons:

Placing the Hydrazine Feed System in the ' Abandoned" category does not foreclose (preclude) release of the site for possible unrestricted use. There is no potential failure mode which could l

_ _ _ _ -. I

gg peg. to- 083 - 9'I g . C;V - 97 ~ '" #

Page 3d3 create a major spill er rupture, causing a significant radioactiv'e contamination of the soil or buildings which could not easily be decontaminated. It also does not permanently install or retain major structures or equipment on site, or render it difficult to remove systems, structures'or components from the site.

Placing the Hydrazine Feed System in the ' Abandoned" category does not result in a significant environmental impact not previously reviewed. It does not result in the discharge of radioactivity or chemicals, either by a controlled release, or due to a malfunction or failure of equipment. It also does not affect the site environment (e.g. , terrain, noise, visual appearance, transmission lines) in any way.

Placing the Hydrazine Feed System in the ' Abandoned" category does not remove reasonable assurance that adequate funds will be available for decommissioning. The cost of procedure revisiens, UFSAR revisions, etc., taken in total, are not likely to jeopardize decommissioning funds.

There is no applicability to the PSDAR of placing the Hydrazine Feed System in the " Abandoned" category since that document has not yet been developed for CY. Rather, the system categorization process will be used as input to the PSDAR development.

b

f#^ L 'fJ-o3

/ keens 40 [ct A!L-24e(?7 CONNECTICUT YANKEE Docutnec+ Uhange Manual Summnry of Changes (conducted under 10CFR50.59) PPcR 1436 A

Proc Change N/A Jumper Bypass N/A

• • As part of the design change process, the Project Engineer is to fill out one of these forms. These summary forms are to be forwarded to Nuclear Licensing once a year for those CHANGES actually implemented since the last report. Nuclear Licensing will prepare a transmittal package for submittal to the NRC Staff in accordance with 10CFR50.59

1. Change Numbert O_ Revision Number: O_

Title:

PDCR 1435 "Renlacement of Battery Charr r BC-1-1 A" Revised Safety Evaluation

2. Description r hange:

PDCR M45 "Ihe Replacement of Battery Charger BC-1-1A" replaced a W=Wa:Nuse 200 amp Banery Charger with a Solidstate Controls, Inc. 300 amp Battery Charger. 'Ibe "A" Banery Charger Rep 1=~. was implemented during Refueling Outage RFO 18 and released for operation on March 1,1995 Upon placing the charger into service it was noted that the ammeter for the output current of the charger was fluedag= Prior to installation, the charger was tested satisfactorily at the factory and at Coaa-dcut Yankee with a load ,

bank with no fluctuations noted. Significant time was spent by plant perwunc! during the outage to determme the cause of the noted ammeter fluctuations. The vendors for the inverters (Cyberex) and battery charBer (SCI) have both evaluated the system. SCI noted that the DC voltage has a ripple between 131.94 and 132.26 vohs (320 mv) and Cyberex has noted a DC ripple of 300 my with random spikes as high as 400 my occurring. This is a ripple of 0.3%

and is an acceptable level for the associated system components. Each vendor starsi their equipment (i.e. battery charger and inverters) was operating correctly, and no conclusions regarding the cause of the fluctuations were reached.

3. Reason for the Change:

NRC Inspection Report 50-213/96 201 considered the cr:ginal safety evaluation (SE-1435, Rev. 0) deficient because a battery charger of different design and circuitry (solid state) wss installed, and the evaluation did not adequately assess the possibility of a malfunction of a different type than previously evaluated. Specifically, the safety evaluation did not consider the potential new failure modes and degrading effects on the 125-Vdc system.

1

A PDG A M SE % 009 Pase 2 of 5 Revised Battery Chater Safety Evaluation Swwury Review

4. Safety Evaluationt

a. This change was safe for the following reasonst ne effects (voltage spikes and current fluctuations) that have been ssen on the 125 Vdc system are typical and do not adversely affect any other system components. Dese conclusions are based on evaluations by plant personnel and rea-ai=d exnnts in the field of EMI and DC Systems ne 125 Vdc system is considered operable and able to perform its intended safety functions.

b. This change does not cor4stitute AN UNREVIEWED SAFETY QUESTION because His evaluation has shown that this modi 6 cation had no negative effects on the probability of occurrence of a previously evaluated erWat or =%= Mon of equipment important to safety. This enange had no negative effect on the consequences of a previously evaluated accident or malfunction of equipmer.t important to safety. This change also did not create the possibility of an accident or malfunction of a different type than any previously evaluated nor did it reduce the margm of safety.

De effects (voltage spikes and curret fluctuations) that have been see . on the 125 Vdc system are typical and do not adwnely affect any other system components. These conclusions are based on evaluations by plant personnel and reaaai-d experts in the field of EMI and pC Systems. De 125 Vdc symm is considered operable and ab!s to y. form its intended safety functions.

Based on the above, this change did not increase the risk to the public and was therefore not considered an unreviewed safety question (USQ).

THERE IS NO INCREASE IN THE .DROBABILITY 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:

Here was no increase in the probability of occurrence or the consequences of an accident or malfunction of equipment during the implementation or as a result of this change since this modification replaced edsting equipment with upgraded equipment to perform identical functions. Prior to and after this change a loss of offsite power combined with a subsequent failure of Diesel Generator "A" to start will render Battery Charger "A" inoperable. Prior to and after this change a Station Blackout event will render Battery Charger "A" inoperable.

Page 2 of 3

./

/ P0cR 144 2 % -00 9 Die 5 cf 3 Revised Battay Charger Safety Evaluation

/s THFPOSSIBILITY FOR AN ACCIDENT OR MALFUNCTf0N OF A DIFFERENT summary Review TYPE THAN ANY EVALUATED PREVIOUSLY IN THC SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED.

The basis for this statement is:

Due to the effects that have been not ? on the 125 Vdc system the possibility of a malfur.ction or failure mode of a different type th r previously evaluated is A" below, Upon placing the new battery charger into service it was noted that the ammeter for the outrut current of the charger was fluctuating Significant time was spent by plant personnel e determine the cause of the noted ammster benatiana. %e vedors for the inverters (Cyberex) and battery charger (SCl) have both evaluated the system. Each vendor stated their equipment (i.e. battery charger and inverters) was operating correctly. He alarm and protective devices within the battery charger and inverters would operate before any damage occurmd tothese ,- ,

CHAR Services and Wa*iag6m Power Systems Faai-ing both have concumd .<sth the l previous evaluations performed by the battery charger manufacturer a a Inverter manufacturer. Dese evaluations have concluded that the voltage spikes and current flucmntions seen on the DC system are typical and are caused by the switching of power transistors in the inverters and do not adversely affect any other system components.

Addrtinwly, the voltage spikes though high in --a'** have little energy n i='~l with them aa.i nmunta due to circuit i'aa~l== as they propagate along the system winng s

Based on the results from tbc art =<ive troubl-6*ia: by plant personnel and outside experts it is concluded that the possibility of a =h* ion or failure mode of a different type than previously evaluated does not exist. Hence, the =~i='~' 125 Vdc system is operable and able to perform its intW safety function.

THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED.

The basis for this statement is:

His change did not negatively impact the basis of the Technical Specifications. Le margin of safety was not reduced, since the safety limits and the parameters of the protective boundaries, as described in the Technical Specifications, were unchanged. His change replaced existing equipment with new QA Category 1, Class 1E equipment.

De replacement of Battery Charger BC-1.'A was performed during Mode 6. Technical Specification limiting conditions for operatior. regardmg operability of DC electrical power sources was followed. Derefore, there was no impact on the margm of safety,

c. Did this change require a chanc' to the Technical Specifications: Yes / Eq Page 3 of 3

• PDCR 145 CON"ECTICUT YANKEE 9 604 10CFR50.59 Summ ry tY- 5EI' f IA$8 This summary applies to:

Safety Evaluation CY-SE-97-004

'"PDCR lill Set Point Change Test Procedure Tech Requirements Manual Change Experiment Tech Spec Basis Change Only Procedure Change FSAR Changes _E_

Jumper Bypass 1.

Title:

CY-SE-97-004, Remove SW-V-849 and SW-V-850

2. Description of Changes:

This Safety Evaluation addressed the removal of two Service Water valves from the upper supply lines to a CAR Fan Cooling coil. The valves had previously been installed to isolate the upper cooling coil, thereby allowing inspection of the coil.

This change returned the piping to a configuration equivalent to it's original configuration.

3. Reason for the Changes

(

i I

The CAR Fan performance monitoring program had been completed and inspection capabilities were no longer required.

Components installed to support tb4 inspection program were l therefore removed.

4. Safety Evaluation:

a. The changes are safe for the following reasons:

The changes will not contribute to any new or previously analyzed accident or malfunction of equipment and do not reduce the margin of safety of the systems involved.

b. The changes do not constitute AN UNREVIEWED SAFETY QUESTION because:

THERE 3 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:

t

PDCE 3 cf- SE l- 00+

CONNECTICUT YANKEE 10CFR50.59 Summary cont'd Safety Evaluaticn CY-SE-97-004

, None of the propoi.ed changes have any effect on the malfunctions evaluated and therefore have no effect on the probability of their occurrence. None of these changes affect the operation or performance of equipment involved.

This change returned the piping to a configuration equivalent to it's original configuration.

THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HA5 NOT BEEN CREATED. The basis for this

))} statement is:

() The proposed changes do not introduce any new accident initiators. This change returned the piping to a configuration equivalent to it's original configuration.

The proposed changes do not introduce any new failure

[) modes or malfunctions. None of the changes involve a system or component operating beyond its design basis.

This change returned the piping to a configuration equivalent to it's original configuration.

THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:

It was concluded that the proposed changes have no impact on previously evaluated accidents and do not introduce the potential for a new unanalyzed event. The proposed changes do not result in the operation of any system or component beyond its design basis. Therefore, it is also concluded they have no impact on the margin of safety. This change returned the piping to a configuration equivalent to it's original configuration,

c. These changes do not require a change to the Technical Specifications.

CY SE k$$'3 Rev.0 2/2/96 FIGURE 72 - SAFETY EVALUATION FORMAT (Use Attrohment 8.A for Guidance)

Safety Evaluation Number: CY-SE 1553 Revision No. O Plant Change Number: 1553 Revision No. O Plant Change

Title:

Service Water MIC Chemical iniection

1.

SUMMARY

INFORMATION 1.1 Saferv Evaluation Concluli.cns ,

3 This safety evaluation concludes that the proposed design change is safe and is not an Unreviewed Safety Question (USQ).

1.2 Descriction of the Change Install a new chemical addition system for the injection of a bio-penetrant and a bio-dispersant into the service water system. The new system consists of bulk storage of chemicals in the Hypochlorite Room, a pump skid with manual and automatic controls, pressure indication and flow indication, with the necessary isolatir and testing capabilities. The bio-penetrant shall be BULAB 8007 and the bio-oncersant shall be BULAB 7005.

1.3 Aseects of the Change Evaluated This safety evaluation addresses the ability of the service water system to perform it's safety related function of providing cooling to QA Cat i equipment, and transmitting the heat to the ultimate heat sink (CT River).

This safety evaluation also addresses the personnel and equipment safety aspects from venting the BULAB 8007 and BULAB 7005 directly into the Hypochlorite Room.

1.4 Malfunctions Evalunted The failure modes associated with the proposed design change are:

Failure of the new Chemical System piping / tubing, causing a loss of service water pressure / flow.

A cr_d burst and subsequent cloggi,g of safety related heat exchangers, caused by the sloughing off of massive quantities of existing tuberculation, after exposure to the BULAB chemicals.

NGP 3.12 Rev.9 Page 1 of (

tocE2 t687

^

SE- 1jf>74

. f,t et t FIGURE 7.2 - S AFETY EVALUATION FORMAT (Use Attachment 8.A for Guidance)

Safety Evaluation Number: SE-1587a Revision No. O Plant Change Number: 1587 Revision No. O Plant Change

Title:

Connecticut Yankee Control Room Modifications

1.

SUMMARY

fNFORMATION ,

] 1.1 Saferv Evaluation Conclusion _s The moditication associated with PDCR 1587 is :;afe and is not a USQ.

1.2 Descriotion of the Change Modifications to tne Connecticut Yankee Control Room are proposed to better serve the needs of the Operations department. Changes will include remodeling of office space, carpeting and installation of a new cont ol room lighting system, ne annunciator silence switch will be relocated to the Control Operators' desks. This change also provides for new operator's consoles, a new shift tnanager console and a new drawing table, ne present control room configuration requires modifications to enhance Control Room Operator's response to plant operations. The proposed modifications provide for a general refurbishment of the cottrol room area. The changes will provide better viewing of the control panels, increased console and office space and reduced noise levels. The new layout will serve to better restrict personnel traffic in the " red line" area resulting in fewer distractions to plant operators.

A new lighting system (Circadian Lighting) will be installed in order to enhance operators ability to biologically adjust to rotational shiftwork. Further, the lighting system has been designed to minimize glare on the control panels.

1.3 Aspects of the Chang Evaluated This safety evaluation addresses the electrical and seismic aspects of the design change modification, as well as the impact of the m 4'ication on the control room operators.

1.4 Malfunctions Evaluated This safety evaluation addresses the fo!!owing malfuretions:

1) Failure of the new lighting system.

2) Failure of the new lighting system support structure due to a seismic event.

1.5 References

1) Electrical and I&C Programs Review per SP-ST-EE-299 (Attached to PDCR).

---mm-- -==

PbcE iG87: -

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f. I c 2.

Safety Evaluation SE~l5FG h RevisionNo. 0 Plant Change Number 1587 Revision No. O Plant Change

Title:

Connecticut Yankee Control Room Modificatim

1.

SUMMARY

INFORMATION ',

1,1 Safety Evaluation Conc!usion Per NGP 3.12, Rev. 09, and in accordance with the rules of 10CFR50.59, the l

proposed PDCR changes are SAFE and h not constitute an Unreviewed Safety Question in that the changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of fire. The changes do constitute a degradation of the fire protection level of the control room due to the increase in the combustible loading and the corresponding fire exposure potential. However, active fire protection in the room and the availability of manual fire fighting capability by l the operators in attendance provides compensation for the added fire load.

j l.2 Description of the Change This PDCR will provide human factors and aesthetic improvements to the CY Control Room. These renovations will include the installation of new work stations for the operators, installation of floor carpeting in the entire room and the installation of a new lighting system in the area in front of the main control board.

Each of these improvements involves the installation of materials which are

' combustible. From a Fire Hazards Analysis standpoint, this is a degradation of the fire protection of this area. As a result of this modification, combustible loading of the room will b,e more than doubled from the present level and the fire duration increased from the current 11 minutes to over 24 minutes. As such, this modification degrades the existing fire protection of the Control Room and does not follow the basic fire protection " defense in depth" philosophy' espoused in the Haddam Neck fire protection program.

1.3 Asnects of the' Change Evaluated 1.3.1 Ability of the plant to be safely shutdown in the event of a fire in the Control Room. ,

I m -__u

. TPCR1581 S E~s51b P' 3 4 A .

1.4 Malfunctions Evaluated 1.4.1 Fire which originates in the Control Room.

i

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[PDCE 159846- f7W sy sv.97 00c5 ihse t 4 +

CONNECTICUT YANKEE Document Change Manual Summary of Changes (conducted under 10CFR50.59)

PDCR: 1123 Proc Change: N/A Jumper Bypasst ElA Safety Evaluation No.: SY-EV-97-0005 "As part of the design change process, the Project Engineer is to fill out one of these forms. These summary forms are to be forwarded to Nr lear Licensing once a year for those CHANGES actually implemented since the last .e,> ort. Nuclear Licensing will prepare a transmittal package for submittal to the NRC Staff in accordance with 10CFR50.59

1. Change Number:._PDCR 15o8 Revision Number :B

Title:

_Ninin Feedwnter Pumn Nfotor Renlacement- Early Release B

2. Descrintion of Change:

PDCR 1598, " Main Feedwater Pump Motor Replacement," intended to replace the existing Westinghouse 4500 hp Main Feedwater Pump motors with new Asea Brown-Boveri (ABB) 5000 hp motors. However, the completion of the design change was stopped on October 9,1996, due to the permanent shutdown of the plant. As a result, the PDCR has never been submitted to PORC for overall review and approval.

Prior to the permanent shutdown, however, two Early Work Releases had already been appr;ved by the Unit Director to support the PDCR implementation schedule. The implementation work performed under the Early Releases inc!'t te preparation for installation and testing of the new equipment stored in the Ware- the removal of the existing MFP motors and their associated lube oil systems, and une construction of new foundations for the new MFP motors. In parallel with these field activities, the pumps were also removed for refurbishment under a separate task.

10CFR59. doc -

Page 2 of 4 Consequently, part of the implementation work required by this design change nas been f

either completed, staned, or never started, i

As of this date, the actual status of the equipment is as follows:

Old 4500 hp MFP motors are removed and stored in the Boneyard.

New 5000 hp MFP motors, accessories, and couplings are stored in the Warehouse.

MF pumps are refurbished and stored in the Turbine Building.

] -

a The 4.16 kV breakers for the MFP motors are tagged, racked out, provided with workman's grounds, and have their de control power fuse disconnects open.

. Lube Oil Pumps P 82-1 A and IB breakers are tagged open.

,A -

All power, control, annunciation, and computer cables for process monitoring are

] disconnected at the load side, insulated, and secured.

Discharge Isolation Valves FW MOV-15 and 16 art tagged out.

The MFP lube oil systems have been removed. All associated oil and closed cooling water lines are isolated from the plant systems by manually operated valves tagged in closed position. The lines are not capped.

The new foundations for the MFP motors are in place, however, no equipment has been mounted on them.

All equipment is in stable condition and secured in a safe manner.

3. Reason for the Change:

The actual status of the Main Feedwater Pumps and associated equipment has no adverse impact on the operation of the Spent Fuel Pool and/or the plant in permanent shutdown and defueled condition, and does not create safety hazards for plant personnel. However, since the decommissioning process will extend over a long period of time, it is prudent to take additional actions in order to enhance the long-term plant and personnel safety.

These additional actions will consist of the de-termination of power, control, and annunciation cables at the source sir'e, and the capping of closed cooling water lines serving the pump and motor oil lubrication system.

10CFR59. doc

h %v'.$ 0o06 Page 3 of 4

4. Safety Evaluntion:

a. His change is safe for the following reasons:

The proposed work outlined above will be implemented on non safety equipment. The equipment is neither required nor used to support the operation of the Spent Fuel Pool (SFP), and/or the plant in permanent shutdown and defueled condition.

b. This change does not constitute AN UNREVIEWED SAFETY QUESTION because:

] His evaluation has shown that this modification will not increase the probability of n occurrence or the consequences of an accident or malfunction of equipment important to L.3 safety previously evaluated in the safety analysis report. This change will not create the possibility of an accident or malfunction of equipment of a different type previously evaluated in the safety analysis report, and will not reduce the margin of safety as defined in the basis of any Technical Specifications.

THERE IS NO INCREASE IN THE PROBAB4LITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO S AFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT.

The basis for this statement is:

This change will affect only non safety related equipment neither required nor used for the operation of the Spent Fuel Fool (SFP) and/or the plant in pumanent shutdown and condition. Except for the annunciation cables, the equipment involved in this scope of work is presently deenergized, and electrically and mechanically isolated from the plant systems. The low voltage, ow energy cnnunciation cables are still energized but used for indication only.

THE POSSIBILITY OR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED.

He basis for this statement is:

The equipment affected by this modification, except the annunciation cables, is presently deenergized, and electrically and mechanically isolated from the plant systems. The low voltage and energy annunciation cables are still energized but used for indication only.

10CFR59. doc .-

6 WA.W?

Page 4 of 4 THE MAROIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED.

The basis for this statement is:

This change will not affect the basis of the Technical Specifications as the Main Feedwater Pumps are neither required nor used to support the operation of the SFP and/or the plant in permanent shutdown and defueled condition. The margin of safuy was not reduced since the safety limits and the parameters of the protective boundaries for the present plant mode

! of operation will not be affected by this change.

l c. Does this change require a change to the Technical Specifications: Yes No.4 10CFR59. doc -

/ f?o/tt 97+7 DCR CT- 97001 ST- EV- 97 " 0002 Pge i 4 3 Attachment 1 i

HADDAM NECK PLANT Annual Report Summary of Changes Mark the Appropriate Choice:

Design Change L Setpoint Change Test _ _ Tech Requirements Manual Change Experiment Procedure Change Jumper Bypass l

1. Change Number:J cn cY-97001 Revision Number _A

Title:

Removal of Main Feedwater Isolation Valves

2. Description of Change:

The CY main feedwater isolation valves (FW-MOV-11 to 14) were f purchased by Millstone Unit 1. These valves and motor operators were removed from the CY feedwater system via DCR CY-97001, Rev A.

3. Reason for the Change:

CY is currently in a defueled condition and has notified the NRC at its decision to permanently cease power operation and remove fuel from the reactor vessel. The feedwater system is not required for decommissioning. The removal of FW-MOV-11 to 14 was evaluated for the plant in it's current defueled condition and found to be safe, both electrically and structurally.

4. Safety Evaluation:

a. This change was safe for the following reasons:

The change did not contribute to any previously analyzed or new accident or malfunction of equipment and did not reduce the margin of safety of the systems involved.

t

gg cY 97ool SV- EW 0001 Page 2 d 3

b. This change does not constitute AN UNREVIEWED SAFETY QUESTION because it does not affect the probability or ,

consequences of previously evaluated accidents or l malfunctions of equipment important to safety. There is no reduction in the margin of plant safety. ,

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:

The probability of occurrence of a previously evaluated g malfanction of equipment important to safety was unaffected gl by this change because there is no longer any feedwater equipment which has a safety related function with the plant

, in its permanently defueled condition. It was concluded that this change would not adversely impact the structural 3 integrity of the system. The feedwater system remained well supported and in a structurally acceptable condition.

Electrically, other than the removal of the valve motors and actuators, the change only involred lifted leads and label changeouts. No equipment was removed, i.e., the integrity of the Main Control Board and McCs was unaffected.

The only accidents addressed were the Fuel Handling Accident and the Radioactive Waste System Failure. No other accidents were credible accidents with the plant in it's permanent defueled condition and with fuel removed from the reactor vessel. The probability of occurrence of a fuel handling accident or a radioactive waste system failure was unaffected by this change. Operation of the feedwater system has no affect on the probability or occurrence of an accident resulting from the dropping of a spent fuel assembly onto another fuel assembly and cannot have any affect on the probability of a "adioactive waste system failure.

THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT RAS NOT BEEN CREATED. The basis for this statement is:

There was no possibility of a malfunction of a different type than previously evaluated as a result of this propoced k

peg cY. 97001

$1. sV- 9]- 0002 Pa$e 3 er 3

1. change. Th6 malfunction of a service water line presently providing flow to the spent fuel pool was considered as a reault of this proposed change. The only possibility of this malfunction occurring was if the feedwater MOVs were dropped on a service water line in the Turbine Building during the removal process. However, since the MOVs were not transported over any portion of the service water system presently providing flow to the spent fuel pool during removal, this was not a credible malfunction and does not need to be evaluated.

III There was no possibility of an accident of a different type than previously evaluated as a result of this change. A loss of service water accident was considered as a result of this proposed change. The only possibility of this accident occurring was if the feedwater MOVs were dropped on a service water line in the Turbine Building during the removal process, which was discussed above.

THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION RAS NOT BEEN REDUCED. The basis for this statement is:

The margin of safety as defined in Technical Specification 3/4.7.9 was not affected by the proposed change since the consequences or probability of occurrence of an accident were not affected. Also, the change did not affect the consequences or probability of occurrence of a previously evaluate ( malfunction of equipment important to safety and did not create the possibility of a new malfunction. The referenced Tech Spec was applicable to Modes 1 through 4, i.e., it had no applicability for a defueled reactor,

c. Did this change require a change to the Technical Specifications: Yes / No Tech Spec 3/4.7.9, "Feedwater Isolation Valves", needed to be deleted since these valves ~were being removed. This change was encompassed by PTSCR C-19-96, which was already in process, to delete those requirements unnecessary for a defueled reactor from the Tech Specs.

OcR cf. M 002

$, EV- QJ - 000)

Ne 3 i of 2 FIGURE 7 2 SAFEW EVA'_UATION FORJAT Safety Evaluation Number SV EV-97-0003 Revision No. Q Plant Change Number CY 97002 Revision No. Q Plant Change

Title:

Check Valve Mftion to Service Water Supply to Spent Fuel Heat Exchangers

1.

SUMMARY

INFORMATION

] 1.1 Descriotion of the Chance This DCR will install a 6" carbon steel 150 # Anchor Darling swing check alve in the Service Water supply line to the Spent Fuel Heat Exchangers. A 3/4" vent line and valve will also be installed into an existing plunged threadolet located upstream of the new check valve. This vent connection will tacilitate in-service testing of the new check valve.

The check valve is being installed to prevent water from draining (backflowing) from the piping following a loss of normal power (LNP) and associated loss of Service Water, as identified by ACR 97-0119. Draining of this pipe has the potential to result in Service Water column separation in the high areas of piping in the Spent Fuel Building (Creare Technical Memorandum TM 1788a). Repressurization of the piping by a restarted Service Water Pump after column separation has occurred cc;ld result in a severe water hammer which has the potential to damage Service Water piping. This modification will prevent the piping from draining, thereby precluding the water hammer (refer to Creare 3/17/97 latter.

1.2 Aspects of the Chance Evabated This Safety Evaluation will discuss the effects of the installation of the new c..eck valve and test connection on system design and malfunctions and potential accidents associated with these components.

1.3 Safety Evaluation Summarv 1.3.1 Description of Change:

This change adds a new check valva and test connection to the Service Water supply line to the Spent Fuel Heat Exchangers.

1.3.2 Reason for Change:

The check valve will prevent rapid draining and column separation in the Service Water supply line after a loss of normal power / loss of Service Water flow. Draining of the line could result in water hammer when a Service Water pump restarts. The test connection is being installed to facilitate in-service testing.

Rev.10 3.1210NGP NGP 3.12 Page 7.2-1 of 7

pct, c $ 97003 sv .el 003 6ge a d 2.

1.3.1 Safety Evaluadon:

This change is safe for the following reasons:

. The installation of the check valve and test connection will not adversely affect any design basis accident applicable to the current mode (defuelod).

. The installation of the check valve and test connection will not introduce any malfunction which could adversely affect safety related equipment.

e The installation of the check valve and test connection will not adversely affect the .

ability to cool the Spent Fuel Pool. j The change has been determined to be safe and not to constitute an Unreviewed Safety Question because: ,

This changw 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 l

' TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:

De modification affects Service Water Spent Fuel cooling supply which can not affect any design j basis accidents in the defueled modes. It also did not increase the probability of ary malfun'.tlons.

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 can not create a new accident since the loss of the SW can not initiate a fuel handling accident, a ment fuel- pool draindown or a radioactive materials re' ease, The potential malfunctions associated with this change have been evaluated and determined to be bounded by existing procedures and analysis.

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 safety is not affected by this change since the consequences or probability of occurrence of an accident is not affected. Also this change does not effect the consequences or probability of occurrence of a previously evaluated malfunction important to safety and does not create the possibility of a new malfunction.

Did this change require a change to the Technical Specifications: No Rev.10 3.12-10NGP NGP 3.12 Page 7.2-2 of 7

Oc& CY- 97003 CONNECTICUT YANKEE CY. SE - *)7 - 00 8 10CFRSO.59 Summary FAge i of f This summary applier to:

PDCR Set Point Change Test Procedure Tech Requirements Manual Change _

Experiment Tech Spec Basis Change Only Procedure Change FSAR Changes Jumper Bypass Other: Removal of Reactor-Related Accidents from CY Licetisina Bases 1.

Title:

CY-SE-97-008. Deletion of Reactor " elated Accidents from the Licensina and Desien Basis of CX

2. Description of Change:

The proposed change removes the reactor-related accidents from the licensing and design basis of CY.

3. Reason for the Change:

The proposed change is the result of the permanent cessation of operations of CY and the permanent removal of fuel assemblies from the reactor vessel into the spent fuel pool as certified in CYAPCO's letter to the NRC, dated December 5, 1996.

4. Safety Evaluation:

a. This change was safe for the following reasons:

s With the fuel assemblies permanently removed from the reactor vessel, the change causes no increase in the risk to public health and safety.

b. This change does not constitute AN UNREVIEWED SAFETY QUESTION because u -

pcR cY- 97003 cY- SE 97- 008 2 page 2J4 10CFR50.59 Summary Continued .

'bEhtEk'SNOINCREASEINTHEPROBABILITYOF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFCTY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT.

The bas.s for this statement is:

The reactor-related transients and ae:idents __

previously evaluated in the UFSAR are only postulated to occur with fual assemblies in the reactor vessel. There are no longer any fuel assemblies in the reactor vessel. Without fuel assemblies in the reactor vessel, and with the permanent cessation of operations, the accidents Cannot occur.

"he proposal does not af t'ect any fission product arrier. Due to the significant decay of noble asea and iodines since permanont reactor shutdown, any release of radioactivity currently in the reactor coolant is bounded by the existing accident analyses in UFSAR Section 15.5. Therefore, there is no increase in radiation dose consequences to the public.

All of the malfunctions listed in the UFSAR for the reactor-related accidents are concerned with the power operation of the plant. Since power operation has been permanently terminated, these malfunctions are no longer applicable and therefore cheir probability of occurrence cannot be increased. The previously evaluated malfunctions of other equipment important to safety used to mitigate the remaining accidents in UFSAR Section 15.5 are not impacted by this change.

This proposed change does not increase the consequences of a previously evaluated malfunction of equipment important to safety, because the malfunctions are no longer applicable, and therefore their malfunction consequences cannot be increased.

' NK CY- 97063 3 ey.e.;7-008 I*t' 3 Y 10CFR50.59 Summary Continued '

t 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 accidents evaluated in other UFSAh sections have p uno physical or procedural connection with the reactor-related accidents analyzed in the UFSAR.

Therefore, there is no possibility of undesirable system interactions, bypasses of safety functions,.

or new failure modes which would result in an accident of a different type. Accidener of a different type the.a previously evaluated are due to decommissioning activities. Decommissioning activity accidents will be addressed and justified by a separate 10 CFR 50.59 safety evaluation.

Malfunctions of a different type than previously evaluated are due to decommissioning activities. , .ryps The malfunctions of equipment during decommissioning activities will be addressed and juctified by a separate 10 CFR 50.59 safety evaluation. -

Additionally, removal of the reactor-related accidents does not create an interface with other systems which are required to perform a safety function for other accidents. Therefore, there is no possibility of undesirable system . interactions, bypasses of safety functions, or new failure modes for the other accidents, which would result-in a malfunction of a different. type.

THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:

With no fuel assemblies in the reactor vessel, no accepr.ance limits which are applicable to reactor-related accidents can be exceeded, and no-failure point or system limitation is reduced. The fission product barriers associated with this proposal are

pct ef- 97009 008 4 cY- SE

• 97 4 10CFR50.59 Summary Emi" 4 'I Continued .

4 the fuel cladding, reactor coolant system boundary and containment. These barriers are no longer required as defense-in-depth protsetive features for the reactor-related transients and accidents analyzed in the UFSAR.

c. This change did not require a change to the Technical Specifications 8

ACP 1.2 2.42 Rev. ? MAJOR Foim 3 10 CFR 50.59 (b)(2) Report Page1of1 Safety Evaluation Number: EY EV 97 0009 Revision: 0 Document Number:,DCR CY 97004 Revision: 0 Document

Title:

__ Domestic Water oli Treatment system Proside a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Ilrief Description of Change and Safety Evaluation Summary:

This safety evaluation addresses the addition of a chemical injection skid and associated chm deal tank to control pil of the Domestic Water system, it also addresses the installation of a backDow preventer in the dunestk ter supply line to the closed cooling uter system.

This Safety Evaluation is required because this activity constitutes a change in the facility as described in the Safety Analysis Report. The potable water system is described brie 0y in UFSAR Section 9.2.5 and is shown on Figure 9.2 4, which is P&lD 16103 26003, Sht 1. The scope of this design change is shown on this P&lD, as w ell as on P&lD 16103 26067, Sht 1.

Als change dou not constitute an unreview ed safety question benuse:

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 that the domestic water system never had any equipment with a safety related function. UFSAR Section 9.2.5.3 states,"The potable water system has no safety related function and does not transport radioactive Dulds. Failure of the system does not compromise any safety related system or coinponent or prevent a safe shutdown of the plant."

The possibilky N an accident or malfunction of a different type than any evaluated previously in the safety analysis report be

- men created. The basis for this statement is that the domestic water system never had any equipment with l a safety related function. UFSAR Section 9.2.5.3 states,"The potable water system has no safety-related function and does not transport radioactive fluids. Failure of the system does not compromise any safety related system or component or prevent a safe shutdown of the plant."

The margin of safety as defined in the basis for any technical specification has not been reduced. The basis for this statement is that the Domestic Water System does not have any Technical Specification requirements associated with it.

2. Reason for the Change:

The Connecticut DEP considers the CY domestic water system to be a Non-Transient, Non-Community water supply.

As such, orgoing sampling and analysis is required. If results are not within speci0 cation for two consecutive 6 month periods, a plan to condition the water supply must be submitted to the State. This has occurred in that samples taken between July 1,1993 and June 30,1994 showed unacceptable copper and lead levels in the water, in addition, the State has identified the need to install a backDow preventer h the domestic water piping between the hydropneumatic (llP) tank and the closed cooling w ater system to meet the requirements of the Connecticut Public Health Code Section 1913-B38g.

The preferred method of reducing copper and lead corrosion is to increase the pil level of the water from the existing level of approximately 6.2 to 8.2 using sodium hydroxide. In brief, the scope of the modification included in DCR CY-97004 will be the addition of a chemical injection skid and associated chemical tank which will continuously recirculate water from the discharge of the IIP tank back to the tank inlet. Chemicals to control pil will be injectM into the recirculated stream. The requir?d bacL0ow preventer will also be installed.

Preparer _ Scott penlev - SI Date i Rkt) DOC

/

r f ACP 1.2 2.42 Rev.1 MAJOR l Form 3 10 CFR 50.59 (b)(2) Report Page1of1 Safety Evaluation Number: SY EV 97 0011 Revision: 0 Document Number: _DCR CY 97003 Revision: 0 Document Titje: REPLACEMENT OF THE ABOVE GROUND DfESEL STORAGE TANK Provide a brief description of the change and a summary of the Safety Evaluation in the format below,

l. Brief Description of Change and Safety Evaluation Surnmary:

nis Safety Evaluation addresses the design and replacement of the existing 42,000 gallon above ground diesel storage tank with a new hortrontal 20,000 gallon horizontal tank. De new tank will be installed adjacent to and due East of the location of the esisting tank. De new tank is made of carbon steel and will have its own self contained dike w/ tain shields. It will be supplied with a remote overfill alarm and a remote gauge that will read in gallons. De tank will also be prt wired with non-explonive lighting so that inspections within the dike area can be performed. The new tank will be connected to the existing fill line and supply line. The tank is anchored to a reinforced concrete foundation to resist a seismic effect and buoyancy due to flooding, and is located at an elevation to allow the tank to gravity feed to the existmg underground Emergency Diesel stooge tanks, he operation and required function of the new tank is unchanged from previous operation and function. However the size of the tank hu been reduced from a 42000 gallon to a 20,000 gallon capas!*y, Since CY is in the ctefueled mode the fuel consumption rate of the Emergency Diesel Generators have been also reduced from 205 gph to 70 gph. ne smaller tank will be able to provide the sufficient fuel to the Emergency Diesel Generators so that there will be no reduction in available Ernergency Diesel running time. The Diesel's available amou .! of back up power to the spent fuel cooling and the sersice water purnps will also be maintained. The New Above Ground Diesel Tank does not affect nor does it directly communicate in any other way with the Spent Fuel Pool or the Spent Fuel Handling System. Therefore, the probability of occurrence of a fuel handling accident or a radwaste system failure, which are the only accidents applicable to the defueled state, is unaffected and there is no possibility for an accident of a different type, nis change does not foreciose (preclude) release of the site for possible unrestricted use. it does not result in a significant environmental impact rm previously reviewed. It does not remove reuonable assurance that adequate funds will be available for decomu.tssioning.

This change is safe and does not involve an unreviewed safety question. De change does not affect the probability or consequences of previously esaluated accidents or malfunctions of equipment important to safety, here is no reduction m the margm of plant safety. In addition, this change does not involve an unresolved decommissioning question.

2. Reason for the Change:

De existing carbon steel above ground diesel storage tank,(TE13 I A), is a vertical tank and has a capacity of 42.000 gallons, ne tank is classified as Non-QA, Non Seismic and has no safety significance other than giving credit to the lesel requirements according to Appendix R", The tank's main function is to provide diesel fuel to the auxiliary boilers for Plant heat. It is also used to fill the under ground diesel storage tanks when their level approaches 3,250 gallons. Over the years the tank bottom hu show1n signs of corrosion. In 1993 the tank wu emptied and an ultrasonic grid test wu performed to determine the condition of the tank Door and the shell,12" up. Inspection of the tank bottom revea'ed several areu of pitting significant enough to require repair. Portions of the tank floor were repaired by welding plates to the floor. This change is being initiated to decrease the risk of polluting the environment caused by failure of an antiquated tank. This change will also reduce the quantity of diesel fuel stored on site and will also provide a remote overfill alarm and a more accurate gauge that will read in gallons.

D Preparer a Date M wrmo poc W ~

t Wegne mm-

ACP 1.2 2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page 1 of 2 Safety Evaluation Nurnber: SY-EV 97 00.18 Revision: 0 Document Number: DCR CW97006 Revision: 0 Document

Title:

Installation of Test lloles in Sill. Pall & NSGD Ventila' ion Doctwork Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Iirief Description of Change and Safety Evaluation Summary:

1his DCR installs eighteen test holes in the ductwork of the Spent fuel fluilding (Sill), Primary Auxillaries 11uilding (PAD) ond New Switchgear fluilding 3rd Iloor (NSGil) Ventilation Systems. The test holes will be manufactured by Ventfabrics, Inc.1he ports will facilitate data acquisition which will allow sytem performance to be evaluated. The test holes will be permanent and will be threaded capped when not in use to present leakage. This change will not affect the operation of any active components or the flow rates of the systems. The test holes will be a part of the pressure boundary of the ductwork and will be installed to the standards of the ductwork.

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 previoui!y evaluated in the safety analysis report. The basis for this statement is:

The instrument test holes are gasketed components with tight Otting screw caps. They are the type originally speciGed for use in the system. Any leakage will be no greater than that expected for seams and Ottings in the existing ductwork. Even if the leakage were greater than expected, the consequences would not result in consequences greater than expected in the SAR because of their localon. Test holes located on positively pressurized ductwork are located such that any leakage from the ductwork ends up in the area into which the ductwork ultimately discharges, in the case of test holes located on negathcly pressurized ductwork, any leakage will be into the ductwork and will not result in a release to the surrounding area.

The possibility for an accident or malfunction of a different type than any evaluated previously in the safet < analysis report has not been created. The basis for this statement is:

The instrument test holes are installed to the speciGcations of the original ductwork.

Therefore they will not result in accidents or failures of the ductwork that have not been considered.

The margin of safety as denned in the basis for any technical specincation has not been reduced. The basis for this statement is:

1he instrument test holes are irstalled to the speci0 cations of the original ductwork.

Therefore they will not reduce the margin of safety of the ductwork. They also will not affect the design Dow rates of sne system.

$059f nT IxX'

ACP 1.21.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (h)(2) Report Page 2 cf 2 Safety Evaluation Number: SY EV 97 00.18 Revision: 0 Document Number: DCR CY 97006 Revision: 0 Document

Title:

Installation of Test lloles in $FD. Pall & NSGil Ventilation Duetwork

2. Reason for the Change:

'the change allows data to be collected on the performance of the ventilation systems. Once the test holes are installed, the caps on a test hole can be removed, thereby allowing instruments to be placed in the duct airstream. Air velocity, temperature, and humidity are parameters that can be measured. The data will be used to benchmark system performance which is necessary as part of the closure to ACRs 97 0025,97 0170 and AR 95-002917.

f

\

Preparer Carlo Darton Date

$039FET lxE

ACP l.24A2 Form 3 10 CFR 50.59 (b)(2) Report Page l of 2 Safety Evaluation Number: SYav.97 0044 Revision: 0 Document Number: D"R CY.97007 Revision: 0 Document

Title:

Revhion to CYAPCO H Adam Neck Undated Final Enfety Analvth Renort Seedan 13.5 Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Brief Description of Change and Safety Evaluaticn Summaryt This change changes the licensing basis and design basis for the CY cecidents. The current bounding radwaste release accident is the waste gas decay tank rupture. In this change, the resin container cceident becomes the bounding radwaste release accident for CY. This DCR revises UFSAR Section 15.5 by: 1) revising the waste gas incident for the trace amount of noble gases remaining in the waste gas decay tank, 2) revising the fuel handling accident for the longer decayed radioactivity in the fuel assemblies, and 3) adding a resia container occident and other decommissioning activity accidents.

The site boundary doses for the waste gas incident and fuel nandliMg occident in this change are well within 10 CFR 100 guidelines. The cite boundary dose for the resin container accident is within the EPA PAG limits.

This change does not constitute an unreviewed safety question because there is no increase in the probability of occurrence or consequences of an ace.ident previously evaluated in the SAR. This is because no changes are being made to the design, operation or. testing of structures, systems, and components, and the site boundary doses remain within acceptable limits.

The probability of occurrence or consequences of a malfunction of '

equipment previously evaluated in the SAR is not increased because this change does not make changes to the equipment, the procedure, or the way the procedure is used to uewater resins.

There is no possibility of creating an accident or malfunction of a different type than any previously evaluated in the SAR because all the revised accidents are still within the design basis and the resin container accident being added was not previously outside the design basis (since it is-an-airborne--release with a= site boundary dose-less than the previous bounding waste gas accident).

$M970044 NEW

ACP 1.2 2.42 Rev. I MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report l

Page 2 of 2 Safety Evaluation Number. SY.EV.97 0044 Revision: 0 Document Number: DCR CY.97007 Revision: 0 DO:ument

Title:

. Revision to CYAPCO 11addam Neck Undsted Final Safew Annivsis Renort Section 1S 5 The margin of safety as defined in the bases for any Technical Specification is not reduced because the margin of cafety for the physical protective boundary for CY (the fuel cladding) is not affected by this change. =

l

2. Reason for the Change: l The reason for this change is to update the UFSAR to the accidents =

chat are applicable to CY in the permanently defueled condition and -

4 h ing decommissioning activ.ities. The expected effect of this change is to allow engineering, des'lgn and operating activities to proceed in ,

acccrdance with the new licensing basis and design basis for accidents.

W 4

]

Preparer

)- Date 8 5 97

$M970064 NTw

ACP :.2 2.42 Rev.1 MAJOR Form 3 10 CFR 50,59 (b)(2) Report Page1of1 Safety Evaluation Number: SY EV 97 0054 Revision: 0 ,

Do ument Number: _DCR CY 97 008 Revision: 0 Document

Title:

- Auxiliary Particulate and lodine samnle system Provide a brief descripti.>n of the change and a summary of the Safety Evaluation in the format below.

1.11rief Description of Change and Safety Evaluation Summary:

The safety evaluation overall was writter, to technically address the installat:sn of the auxiliary partiet. late and iodine sample system. This system is being installed to replace the R 14A sample system as backup to R 14D for particulate and lodine sampling. nis modification, performed via DCR CY 97008, installed valves, tubing and a sample pump in the same sample lines as R 140. When R 14D is out of service this auxiliary system may be put into service to perform the required Technical Specification sampling. This system has no effect on the normal operating characteristics on the R 14D sample system and adequately perfonns the subject sampling at the same sample rate as R 14D, This change does not constitute an unreviewed safety question because no impact on any accident analysis in the defueled condition is affected by this modification. This modi 0 cation does not alter the normal operation of any plant equipment and provides an auxiliary method of sampling for paniculate and iodine from the main stack when the Wide liange Monitor is out of service.

There is no increase in the pro? ability of occurrence or the consequences of an accident or malfunction of equipment important to sa.'ety previously evaluated in the safety analysis report. The basis for this statement is that no important to safety equipment is afTected by this change. All changes performed by this modification are isolated to equipment that is not imponant to safety. The changes to the R 14D, Wide Rage Oas Monitor, do not impact an evaluated ma% -tn of any important to .afety equipment and this system is not used to mitigate the consequences of any accident.

The possibility for an accident or malfunction of a different type than evaluated previously in the safety analysit report has not been created. The basis for this statement is that the equipmert modified by this change cannot be an initiator of an accident. The R 14D, Wide Range Gas Monitor, cannot bitiste an event and malfunction of this equipment does not change the existing analysis. The modi 0 cation performed used similar parts as the existing equipment and does not alter the normal operation or plant con 0guratior. of R 14D. No new malfunctions or accidents result from 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 that the change to R 1411, Wide A ange Gas Monitor, does not alter the normal plant operation of any equipment nor does this modincat* ion change the ability of the plant to comply with Technical Speci0 cations.

The new auxiliary system will only be in service when R 14D is out of service so no impact on Dow characteristics of R 14D will occur due to the operation of the new auxiliary system. The new auxiliary system will be isolated and downstream of the particulate and iodine filters woen R 14D is in service so no impact on the normal operation of R 14D will occur.

2. Reason for the Change:

R 14A was considered to be an improper backep to R 14D for particulate and iodine sampling. The R 14 A sample system was incapable of drawing,a proper isokinetic sample. This auxiliary system draws a proper sampli - sough existing Isakinetic noules to provide an adequate backup sample system to R 140. If R 14B is out of service this auxiliary sample system may be utilized. During normal operation of R 14D this auxiliary system does not impact R 14D.

Preparer Date 5059r1.01 txx'

,p -wi.

ACP 1.2 2.42 Rev.1 MAJOR Form 3 10 CFR 50.59 (b)(2) Report Pa ge 1 of 1 Safety Evaluation Number: fiY LV 97 0061 Revision: ,, O Document Number: DCR CY 07000 Revision: 0 Document

Title:

_. Af1tfas12emoval for Decontaminatlon Testine Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

l 1. Ilrief Description of Change and Safety Evaluation Summary:

This Safety Evaluation addresses the DCR that removes and replaces a 6 foot section of 3" piping in CVCS letdown line 3" Cll 2$0lR 170 %e removed pipe is required for testing the proposed fluids and processes for decontamination of primary plant systems.

1he removed section of piping is being replaced with a new section of piping which will have an isolation valve and a Danged spool piece with an associatect drah: The section of piping will be reclaoWd to a lower design pressure and temperature. The insulation will also be permanently removed.

The replac.cment piping and valves require re rating or the associated line.

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 i important to safety previously evaluated in the safety analysis report. The basis for this statement is that CY is in a i

defueled condition and has permanently ceased power op rations. The only accidents that are applicable in the condition are related to fuel handling and radw aste system failure. The system that is being modined does not communicate or interface with Spent Fuel Pool or Radwaste systems.

The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report has not been created 1he basis for this statement is that the system is currently classi0ed as " Lay up" and is not functioning, the remos al and reinstallation of a short se-tion of piping will not create the possibility of an accident or malfunction of a different type.

r The margin o safety as denned in the basis for any technical specincation has not been reduced. The Sasis for this statement is that the only Technical Speci0 cation requirements for the RCS and adjoining sections of CVCS which must be met in the cunen; defueled condition are the performance of the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> chemistry sampling of Tech Spec 3/4.4,7.

These requirements will be met using the hot leg sample line w hich will remain "Available". The only other RCS requirements which apply at all times are the heatup and cooldown limits of Tech Specs 3/4.4.9.1 and 3/4.4.9.2. In the current plant con 0guration, it is impossible not to meet these requirements.

2. Reason for the Change:

In order to determine the effectiveness of the proposed decontamination Guids and processes, tests must be performed on a representative sample of pipe. The section ofletdown piping that will be removed has been determined to be representative of th9 majority of the RCS system piping that is being considered for decontamination.

Preparer _ Scott Penlev Date sM IRAD EXT

<w v y N

ACP 1,2 2A2 Rew I MAJOR Form 3 10 CFR 50.59 (h)(2) Report Page 1 of 2 Safety Evaluation Number: s%EV 97 0076 Revision: 0 Document Number: DCR C%97013 Revision: 0 Document

Title:

_ sodium Ilvpachlorite TanLRcpAccment Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

l. lirief Description of Change and Safety Evaluation Summary:

DCR CY.97013, which this SE supports will replace the two existing 1500 gallon Sodium flypochlorite (NaOCl) tanks, TK 44.l A & 111(Drawings 16103 26014 sht I and 16103 20279), with new 305 gallon tanks. The tanks are part of the NaOCl system and reside in the llypochlorite room of the Screenhouse Inu ilding.

As documented in section 10A6.3 of the CY l'SAR, the NaOCl system is not safety related and failure ofit or any ofits components will not damage any safety related system or component. lurther, pDCR 1335, ' Sodium flypochlorite System Upgrade',1993 and, NUSCO calculation No.90-036 749 GM, Rev. 0 'CY Sodium flypochlorite System Modifications: NaOCl Tank Rupture Analysis' address catastrophic failure of both esisting 1500 gallon tanks and demonstrate that the existing dike wall contiguration of the hypochlorite room is adequate to retain the spillage of these tanks and, therefore the new 305 gnllon tanks; thereby preventing any environmental concem or accident.

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 ;ireviously evaluated in the safety analysis report. The basis for this statement is th:a the tanks are not credited for mitigating any design basis accidents and have no failure mode that could impact DilA probability, that the tanks are loc 6ed in the intake structure and cannot interact with either the radaaste system or the spent fuel such that the consequences of a radioactive waste system failure or a fuel handling accident are increased, that the flypochlorite system is not safety related and failure of this system or any of its components will not damage any safety related system or corr.ponent, and finally that the decrease in tank site still provides an adequate supply of NaOCl to prevent bio-fouling and the tanks are isolated away from any safety related equipment.

The possibility for an accident or malfunction of a difTerent type than any evabated previously in the safety analysis report has not been created.1he basis for this statement is that the new tanks will be placed in the same location as the old tanks, be manufanured of the same material as the old tanks, contain the same chemical solution and be exposed to the same extemal environment as the old tanks.

1he margin of safety as defined in the basis for any technical specification has not been reduced.1hc basis for this statement is that the NaOCl system is not a system w hich is addressed ir, the Technical Specifications.

2. Reason for the Change:

Tanks TK 441 A & Ill are being replaced because both existing tanks have developed leaks which are not directly repairable. Tank capacities will be reduced from 1500 gallons each to 305 gal lons each because system demand has been ,

drastically reduced due to plant shutdown. CY TS 97 0315, 'llypochlor;te System Decommissioning Action plan',

U23/97 and ' Connecticut Yankee decommissionin;; Action plan for the Sodium llypochlorite System', Rev. O,6!!'P97 indicate that the Circulating Water systern no longer requires NaOCl and CY.TS-97 0278, 'llypochlorite Tank Assessment and Corrective Action Recommendation',6!4'97, indicates that usage for the Service Water system at the 225 mtutoc

ACP 1.2 2.42 llev,1 MAJOR Form 3 10 CFR 50.59 (h)(2) Report Page20f2 Safety Evaluation Number: SY.EV.97 0076 Revision: 0 Document Number: ncn.cv.97013 Revision: 0 Document

Title:

Sodium livnochlorite Tant Reclacement beginning of June 1997 was 35 40 gallons per week. At this rate the new tanks would provide approximately 16 wecks of NaOCl before refillis tequired.

Preparer _ Vincent A. Ursitti Date 9/29'97 s

2281 Rh0A IXE

ACP 1.2 2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page1of2 Safety Evaluation Number:. SY TV 97 0076 Revision: 1 Document Number: DCR.CY 97013 Revision: 0 Document 'litle: __ Sodium livnochlorite Tank Replacement and System Uncrades Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Drief Description of Change and Safety Evaluation Summary:

Revision 0 of SY EV 97 0076 addressed the replacement of the two 1500 gallon Sodium flypochlorite tanks, under DCR CY 97013 and DCN DCY 00-0228 97. Revision 1 of SY EV 97-0076 has been generated to incorporate additional scope work to DCR CY 97013. DCN DCY 00-0363 97, ' Relief Valve Recirculation Lne for ilP RV.157, SW liypo Metering Pump P 178 l A', details the added scope as the installation of a short section of Dexible tub;ng between the NaOCl Service Water injection metering pump relief valve and the metering pump suction line. He tubing will be about 2' of 1/4" OD Dexible tenon tubing and a 3/8" x 3/8" x 1/4" tenon reducing union Tee. De metering pump and relief valve are shown in section G'll 3 of P&lD 16103 26014 Sh.1, Rev 23. This tubing will be part of the NaOCl system and res!Je in the flypochlorite Room of the Screenhouse Building. Here is an individual Form 1 for each DCN mentioned.

During plart operation, the NaOCl system provided a means of injecting measured amounts of Sodium flypochlorite into the Circulating Water and Service Water systems. Since the plant has been permanently shut down, the Circulating Water system is no longer provided with NaOCl solution. This is documented in CY TS-97 0315, 'llypochlorite System Decommissioning Action Plan', 6/23/97 and ' Connecticut Yankee Decommissioning Action Plan for the Sodium llypochlorite System', Rev 0,6/18/97. The Service Water System remains active and currently has a reduced NaOCl demand. Further, as documented in section 10.4.6.3 of the CY FSAR, the NaOCl system is not safety related and failure ofit or any of its components will not damage any safety related system or component.

His changed does not constitute an unrevb ed safety question because:

here is no increase in the probabili:y of occurrence or the consequences of an accident or malfur.ction of equipment important to safety previously evaluated in the safety analysis report. The basis for this statement is that neither the storage tarks nor the service water NaOCl metering pump relief valve, the metering pump suction line nor the NaOCl system itself are credited for mitigating any design basis accidents and base no failure mode that cot.ld impact DBA probability, that the components involved are located in the intake structure and cannot interact with either the radwaste system or the spent fuel such that the consequences of a radioactive waste system failure or a fuel handling accident are increased, thrt the flypochlorite system is not safety related and failure of this system or any of its components will not damage any safety related system or component, that the decrease in the tank size still provides an adequate supply of NaOCl to prevent bio fouling and the tanks are isolated away from any safety related equipment and adding and Ocally that adding a 2' long piece of tubing between the metering pump relief valve and the metering pump suction line serves no safety function and does not afTect any safety related equipment or the consequences of a malfunction of equipment important to safety.

En 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 that the new tanks will be placed in the same location as the old tanks, be manufactured of the same material as the old tar ks, contain the same chemical solution and be exposed to the same external environment as the old tanks and adding about 2' of tubing between the service water metering pump relief valve and suction line does not change the NaOCl system in any way so as to create the possibility of an accident, malfunction or failure that result in a need to be included in the SAR, sinct public risk is not compromised, l' rat.tRVI A IXV' s.

l l

1 ACP 1.2 2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (h)(2) Report Page l of 2 Safety Evaluation Number: SY EV 97 0076 Revision: 1 Document Number: DCR CY 97013 Revision: 0 Document

Title:

Sodium flypochlorite Tank Renlacement and System Untrades The marsi n of safety as defined in the basis for any technical specification has not been reduced. De basis for this statement is that the NaOCl sptem is not a system which is addressed in the Technical Specifications.

2. Reason for the Change:

De request to add a recirculating line between the NaOCl system injection, to Service Water System, metering pump relief s alve and the metering pump suction line was made so as to prevent a malfunction of the relief valve and to prevent the solution, vented through the valve, frorn dripping on the side of the pump and the floor it makes good sense to have this work done at the same time the NaOCl storage tank replacement work will be done in the Hypochlorite Room.

Preparer Vincent A Ursitti Date 12/4'97 IRM3RVI A DOC

/ pcR cY- 9799Z f sv . EV 0093 CONNECTICUT YANKEE Page t of 2 Document Change Manual Summary of Changes (conducted under 10CFR50.59)

DCR CY 97992 Proc Change Jumper Bypass

" As part of the design change process, the Project Engineer is to fill out one of these fonns. These summary forms are to be forwarded to Nuclear Licensing once a year for those CIIANGES actually implemented since the last report. Nuclear Licensing will prepare a transmittal package for submittal to the NRC Staffin accordance with 10CFR50.59

1. Change Number: SV-EV-97-0003

Title:

Check Valve Addition to Service Water Surolv to Spent Fuel Hea.L Exchancers

2. Description of Change:

This DCR will installed a 6" carbon steel 150 # Anchor Darling swing check valve in the Service Water supply line to the Spent Fuel Heat Exchangers. A 3/4" vent line and valve will also be installed into an existing plugged threadolet located upstream of the new check valve to facilitate in service testing of the new check valve.

I I

3. Reason for the Change:

The check valve was installed to prevent water from draining (backflowing) from the piping following a loss of normal power (LNP) and associated loss of Service Water, as identified by ACR 97 0119.

Draining of this pipe had the potential to result in Service Water column separation in the high areas of piping in the Spent Fuel Building (Creare Technical Memorandum TM 1788a). Repressurization of the piping by a restarted Service Water Pump after column separation could have resulted in a water hammer which has the potential to damage Service Water piping. This modification prevents the piping from draining, thereby precluding the water hammer m e

SV - EV o00 Po $e 2 cf 2

4. Safety Evaluation:

a. This change was safe for the following reasons:

. The installation of the check valve and test connection will not adversely affect any design basis accident applicable to the current mode (defueled).

. De installation of the check valve and test connection will not introduce any malfunction which could adversely affect safety related equipment. i e The installation of the check valve and test connection will not I adversely affect the ability to cool the Spent Fuel Pool.

i 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 modification affects Service Water Spent Fuel cooling supply which can not affect any design basis accidents in the defueled modes. It also did not increase the probability of any 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. De basis for this statement is:

This change can not create a new accident since the loss of the SW can not initiate a fuel handling accident, a spent fuel pool draindown or a radioactive rnaterials release. The potential rnalfunctions associated with this change have been evaluated and detennined to be bounded by existing procedures and analysis.

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 safety is not affected by this change since the consequences or probability of occurrence of an accident is not affected. Also this change does not effect the consequences or probability of occurrence of a previously evaluated malfunction important to safety and does not create the possibility of a new malfunction.

c. Did this change require a change to the Technical Specifications: No

HADDAM NECK PLANT SECTION ll Procedure Changes (Page 1 of 3)

Safety Evaluation Procedure Number Number Title ACP 1.21.4 Rev.0 SY EV 97-0114 CY Nuclear Safety Assessment Board ACP 1.2 3.11 Rev. O SY-EV 97 0107 Performance of Control Panel Design Reviews AOP 3.219 Rev.12 SY EV 97-0084 Less of Service Water AOP 3.2 57 Rev.15 SY EV 97 0142 Station Fire i AOP 3.2 59 Rev. 3 SY EV 97 0001 Rev. 0 Emergency Service Water Cooling for the "A" Spent Fuel Pool Cooling Heat Exchangers (TPC 97-54)

AOP 3.2 59 Rev. 3 SY EV 97 0001 Rev.1 Emergency Service Water Cooling for the "A" Spent Fuel Pool Cooling Heat Exchangers AOP 3.2 59 Rev. 3 SY EV-97 0001 Rev. 2 Emergency Service Water Cooling for the "A" Spent Fuel Pool Cooling Heat Exchangers AOP 3.2 59 Rev. 3 SY-EV 97-0103 Loss of Spent Fuel Pool Cooling (TPC 97-281)

ENG.1.7-102 Rev. 2 SY EV 97-00?O Spent Fuel Pool Cooling Heat Exchanger Performance Test Revision 4

1MQDAM NECK PLANT SECTION 11 <

Procedure Changes  :

(Page 2 of 3)

Safety Evaluation Procedure Number Number Ijile ENG.1.7-102 Rev. 3 SY EV 97-0057 Spent Fuel Pool Cooling Heat Exchanger Performance Test EOP 3.1-10 Rev.18 SY EV-07-0006 Temporary Procedure Change (TPC 97-68) .

EOP 3.1-48 None Loss of Refueling Cavity Inventory NGP 2.14 SY EV 97-0078 CY Fire Protection Program Manual NOP 2.101 Rev.14 SY EV 97-0070 Spent Fuel Pit Cooling System Operation (TPC 97-188)

NOP 2.14-7 Rev.13 SY-EV-97-0065 Waste Gas Degasifier Operation NOP 2.20-4 None Hypochlorite System Operation (TPC 97-3)

NOP 2.20-7 Rev.1 SY EV 97 0086 Traveling Water Screen Operation NOP 2.23-1 SY EV 97-0028 Water Treatment Plant Operation (TPC 97-113)

NOP 2.24 3 SE-ODM 96-53 Rev 1 Filtered SWS and Adams Filter Operation, and Cancellation of TPC 4

w+rP- 3 -

m qwg,-mm=-- + . -& w~ .-,v---- - sp~ m --n m, - - - - -wy-----,4-4,

BAQDAM NECK PLANT l SECTION ll j Procedure Changes (Page 3 of 3)

Safety Evaluation Procedure Number Number lille NOP 2.24 3 Rev.17 SY EV 97 001 Rev. 4 Emergency Service Water Cooling for the "A" Spent Fuel Pool Cooling Heat Exchanger (TPC 97 222)

NOP 2.261 Rev.11 SY EV 97 0082 Operation of Control Air System (TPC)

PMP 9.1-61 Rev. O SY EV 97-0073 R-18 Sample Pump Strainer Cleanhg RPM 1.6 8 Rev. O SY EV-97 0075 Connecticut Yankee Health Physics Organization RPM 2.7-1 SY EV 97-0144 Cancellation of Protective -

Clothing Procedure SUR 5.2 81.3 SE-PDCR1550 Rev. O Radiation Monitoring System Calibration (R14A Main Stack)

SUR 5.7-217 Rev. 3 SY-EV-97-001, Rev. 3 Inservice Testing of SW-CV-963 (TPC-97 221)

ACP 1.2 2.42

/ '

Rev,1 MAJOR Form 3 10 CFR 50.59 (b)(2) Report Page1of1 Safety Evaluation Number: EYJ%97 0114 Revision: 0 Document Number: _ ACP 13.l 4 Revision: Orininal Document

Title:

cY Nuclear Rafarv At**timent Board Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Brief Description of Change and Safety Evaluation Summary:

This safety Evaluation addresses the deletion of Note 1 in Attachment $ (NSAB Audit ltr nsiof the subject procedure which states: Over a five year period, all items contained in the Technical Specification: will be audited". The reason for this change is to eliminate this unnecessary requirement, which is inconsistent with the current audit program as described in Section 18.2.1 of the CYQAP. Rev. 01, which uns a performance oriented approach to auditing. The requirement to perform Technical Specification Audits at a frequency of I per 24 months will remain as specified in the subject procedure and Technical Specification 6.5.2.7. nis revised approach at the same audit frequency will hase no adverse effects, but rather will enhance the process by focusing on items of importance and concent only.

This change relates only to the approach to auditing for conformance to the Technical Specifications. The probability of occunence or the consequences of an accident or malfunction of equipment important to safety previously identified in the SAR are unaffected. l ikewise, the change does not create the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the SAIL This change does not foreclose (preclude) release of the site for possible unrestricted use. it does not result in any environmental impact, it does not remove reasonable assurance that adequate funds will be available for decommissioning, nis change is safe and does not involve an unreviewed safety question. There is no reduction in the margin of plant safety. In addition, this change does not involve an unresolved decommissioning question.

2. Reason for the Change:

This change is proposed to permit auditing for confonnance to the Technical Specifications to be conducted on a perfonnance oriented basis, consistent with the approach to auditing contained in the CYQAP, Rev 01. Industry data,-

appropriate to specific Technical Specifications for both operating and decommissioning plants, as well as identified issues, will form the basis for these perfonnance oriented audits. In addition, as decommissioning progresses, Technical Specification volume will be reduced accordingly, resulting in the specific item audited becoming a higher percentage of the total Technical Specifications for the plant. The net result will an enhanced auditing program, focusing on items of importance and concern.

Preparer T. F. Turheo WE^$I 2 Date I!5If!I7 I

J d

audit 3 dot.

1 ACP 1.2 2.42 Rev,1 MAJOR Form 3 10 CFR 50.59 (b)(2) Report Page1ofI Safety Evaluation Number: W.Ev.97 0107 Revision: 0 Document Number: ACP 1.2 3.11 (NGP S.23 Rev. 21 Revision: OrIntnal Document

Title:

Perforenance of Cnntrol Panel Detlan Reviews Provide a brief description of the change and a summary of the Safety Evaluation in the format below,

l. firief Description of Change and Safety Evaluation Summary:

CancetWlon of ACP 1.2 3.11 (NOP $.25 Rev. 2) w hlch pertains to the control room local control panels.

This change does not constitute an unreviewed safety question because:

Dere 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. ne basis for this statement is:

ne 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:

In the defueled state, the cancellation of this procedure has no effect on the revised accidents analyzed since 1) good lluman Factors Engineering principles will be applied to control roont' control panel modifications taking into consideration the guidance provided in SP.EE.261

• Design Standards for the Modification of Control Panels at Connecticut Yankee, and Millstone Units 1,2 and 3." and 2) the remaining postulated accidents require very few parameters to be monitored and the control room staff has ample time to take corrective actions.

De margin of safety as defined in the basis for any technical specification has not been reduced. De basis for this statement is:

ne cancellation of this procedure does not reduce the margin of safety since this procedure is not related to or associated with any Technical Specification.

2. Reason for the Change:

Dit procedure is being canceled because it is no longer warranted in the defueled state. His procedure, the CY CRDR and the NRC criteria specified in NUREO 0737 Supplement 1, item 1.D.I and NUREO 0700 " Guidelines for Control Room Design Reviews" were developed for an operating plant, where following an accident Control room personnel are bombarded with large amounts of complex and sometimes conflicting information which they must assimilate, use to diagnose the event and take corrective actions in a relatively short period of time. At CY, this is not the case. De remaining postulated events require very few parameters to be monitored and the control room stalThas ample time to take torrective actions. How ever, to ensure consistent application of good lluman Factors Engineering principles, DCM 1.3 1 " Design Changes" requires that design changes made to the control room or local control panels take into consideration the guidance provided in SP-EE 261

• Design Standards for the Modification of Control Panels at Connecticut Yankee and Millstone Units 1,2 and 3" Therefore, this procedure is no longer needed.

Preparcr Date 51970107 DOC

ACP 1,2 2.42 Rev.1 MAJOR Form 3 10 CFR 50.59 (h)(2) Report Page 1 of 2 Safety Evaluation Number: SY EV 97 0084 Revision: 0 Document Number: AOP 12-19 Revision: 12 Document

Title:

1 oss of Scndte Water Provide a brief description of the change and a summary of the Safety Evaluation in the format below,

l. Drief Descripuon of Change and Safety Evalcation Summary:

1 The proposed major changes to AOP 3.219 are:(1) the placement of the SWP's in TPO and stopping the EDG's if a pipe rupture is suspected;(2) the addition of a step to clean the SWP strainers;(3) the addition of attachments for isolating various portions of the SWS as necessary to isolate leaks;(4) the addition of a TRM action statement entry for the FP system when aligned to the EDG's for emergency cooling; (5) the addition of steps to isolate SW to the SFP heat exchangers if a break occurs;(6) deletion of the step to open SW.V 281 to crosstic FPS to SWS; (7) deletion of steps referring to letdown and RCP seal supply;(8) addition of steps necessary to align normal SW to both EDG's by opening crosstie valve SW V.147 and (9) deletion of reference to enter E-0.

This change does not constitute an uttreviewed 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 changes to AOP 3.219 only afTect the way in which the plant responds to a loss of SW cooling (i.e. shutdown of equipment to prevent damage, isolation of failed components, shutdown of SWP's to reduce flooding, etc.). This change improves the chances of maintaining vital equipment available by taking prompt preventative measures to isolate SW breaks and provide backup cooling. There are no effects on the consequences of the Design Basis Accidents currently listed in Chapter 15 due to this procedure change since the changes do not effect the handling of fuel or the w aste liquid' gas systems, nor do they adversely effect the actions necessary to respond to these accidents.

The changes to AOP 3.219 will decrease the probability of equipment malfunctions by providirig appropriate acti ns after a partial or total loss of Service Water. The potential for EDO damage will be diminished by promptly shutting down the component. Ruptured piping will be isolated to reduce flooding and critical components will be provided with altemate cooling were possible. The Fire Protection System will be slightly degraded when cross tied to the EDG's, however a loss of SW concurrent with a station fire is outside the design basis and licensing commitments and considered a non-credible event. The probability of occurrence of a previously evaluated raalfunction of equipment (i.e. partial or total loss of SFP cooling) important to maintain the spent fuel pit bulk temperature below 150*F will not be effected by these changes since adequate steps are being provided to maintain SFP laventory should a loss of SFP senice water supply occur by directing the operator to implement AOP 3.2 59 for loss of SFP cooling. The changes to AOf' 3.219 actually reduces the consequences of postulated failures of equipment important to safety by providing guidance necessary to shutdown or isolate important components and piping prior to any damage occurring. Backup sources of cooling w ater are also provided to restore important equipment as soon as possible.

The possibility for an accident or malfunction of a different type than any evaluate previously in the safety analysis report has not been created. The basis for this statement is:

The changes to AOP 3.219 provide guidance to decrease the possibility of component failures after a loss of Service Water and steps necessary to restore the sjstem.1hc possible of another accident otbr those previously evaluated concurrent with the loss of SW is not considered credible and would be outside the current design and f ut3ns4 Doc

ACP 1.2 2.42 Rev. I MAJOR Form 3 10 CFR 50.59 (h)(2) Repott Page 2 of 2 Safety Evaluation Number: s%E%97 0084 Revision: 0 Document Number: AOP 3.219 Revision: 12 Document

Title:

Lou.pf Service Water licensing basis. The AOP being revised would only be entered as a result of a malfunction or abnormal condition in the SWS. The major changes simply provide guidance and actions to prevent damage to plant equipment and restoration of cooling now. The only potential malfunctions caused from the changes in this procedure affect the FPS. Ilowever, concurrent SWS failures and station Dres are outside the licensing and design basis. Even if a fire did occur after aligning the FPS to the LDG's, the amount of How diverted from the FPS (less than 1000 rpm based on previous testing) would still allow the system to respond to most anticipated st.ition fires now that the turbine is shutdown.

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 change serves to enhance the ability to respond to a loss of Service Water by shutting down important equipment prior to damage .nd providing backup cooling to vital equipment. With the exception of the potential for slightly degrading the FPS, the changes improve the reliability and availability of vital plant equipment. Technical Specincatica 3/4.7.3 was reviewed although not applicable to the defueled mode or this AOP. No other Technical Speci0 cations appear applicable.

2. Reason for the Change:

The reasons for the changes are:(1) if pipe rupture is suspected flow must be terminated to prevent Hooding and EDG's must be stopped to prevent damage;(2) strainer differatial pretsure could be excessive and causing low header pressure; (3) the NRC identified the need for more detailed guidance when isolating SW subsystems;(4) aligning FPS to SW degrades FPS, requiring TRM entry;($) no previous guidance existed; (6) opening main crosstic would cause the FPS to be total inoperable and potentially feed a pipe break in the SWS;(7) these systems are no longer operable and have been recategorized;(8) both EDG's can be maintained operable with normal SWS flow with SW.W147 open and (9) the plant is defueled, therefore E 0 is no longer applicable.

Preparer _ R. W. Kasugn Date 1/1498 4

i Rhtl$1 s4 DOC

1 I

ACP 1.2 2.42 l Rev.1 MAJOR t Form 3 - 10 CFR 50.59 (h)(2) Report Page1of2 Safety Evaluation Number: SY EV 97 0112 Revision: 0 Document Number: AOp 3.2 57 Revision: 15 Document

Title:

Station Fire Provide a brief denription of the change and a summary of the Safety Evaluation in the format below.

1. Brief Description of Charqe and Safety Evalu tlon Summary:

Procedure was revised to incorporated improved strategies for the Control Room to assist the Fire Brigade when fighting fires. The steps that were previously required for safe plant shutdown we e removed from the procedure since they are no longer necessary. Actions that were previously assigned to specific operaters are no longer included, since the positions no longer exist.

The most significant change to the procedure is the isolation of pow er to Spent Tucl Cooling in the event of a fire in that area. With the plant in a defueled condition Spent fuel Pool Cooling is the most cri icalt plant system. Safety evaluation reviewed the consequences of this action as a fire fighting strategy, in many cases electrical faults can be the cause or a contributing factor to a fire. In other cases the damage to electrical equipment can be greatly reduced by deenergizing that equipment prior to fire damage. Without having to consider consequences and actions required for safe plant shutdown the choice to deenergize portions of the electrical systeros becomes much simpler.

Plant parameters and systems that are affected by this procedure are first alTected by the fire. It is the fire that has the greatest alTect on plant parameters and systems. Iloweve6 based on the procedural response of the Control Room power will be isolated to various systems in different scenarios. When power is isolated operating system will be lost.

Of greatest concem will be isolation of pow er to Spent Tuel Pool Cooling in the event of a fire in that area. However, it is imperative to remember that isolation of electrical equipment in a fire is a temporary action. Temporary isolation of Spent fuel Pool Cooling is an acceptable action in the event of a fire di.e to the current average spent fuel temperatures, spent fuel heat load and maximum pool temperature. Since it would take more than 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> for the pool to boil from an initial temperature of 150' F, and the maximum calculated fire duration is ler. :han 10 minutes, there is a considerable margin during which time Spent Tuel Pool Cooling can be restored.

Based on review of the references and completion of the 50:59 Safety Evaluation the procedure revision is considered a safe change.

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 orquipment impoaant to safety previously evaluated in the safety analysis report. The basis for this statement is:

There is no change to the plant or how the plant is operated the proposed procedure change provides responses to a casualty.

In the event that a fire in the Spent fuel Building necessitated power isolation to Spent Fuel Pool Cooling it would not be an increase in the consequences, but an attempt to mitigate damage.

The possibility for an accidera 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:

Control Room strategies outlined in AOP 3.2 57 do not create any permanent changes to the plant. Intentional power isolation to systems affected by the fire is no different than pow er isolation due to equipment malfunction, which has aiready been considered.

AOPDR t .tXX'

l ACP 1.2 2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page 2 of 2 Safety Evaluation Number: SY EV 97-0142 '

Revision: 0 Document Number: AOP12 57 Revision: 13 Document

Title:

J'ation Fire The margin of safety as defined in the basis for any technical specification has not been reduced. The basis for this statement is:

The fire itself is the challeny to the margin of safety. Enhancements and improvements to the strategies employed by the Control Room which allow prompt Fire Brigade response will preserve the margin of safety. In cases where power is isolated, it is likely to limit the damage of fire thus allowing a quicker recovery from the abnormal event.

2. Reason for the Change:

%e primary reason for the procedure change is the transition from ath Operating Nuclear Power Plant to a Decommissioning Plant.

Preparer Date

{

AOPf3RI IXX'

l l

TPC 47-74 A0r 'S.2 - 59

",Ej~$7-flb.0 FIGURE 7.2. sAFrTY EVALUATION FORMAT (Use Aanchment s.A for Guidance)

Safety Evaluation Number: sY-EV 97-0001 Revision No: J _

Procedure Number: AOP3259 Revision No. _ 1 Safety Evaluation

Title:

F==.aev hevien wat., cnolin. ror th, "A" M W Fuel Paol cantine Heat Frehaneer' I' st IMMARY INFORMATION I O. O DA 3 v s-y 7 1.1 De<crintiaa of the De'l?" Chaa?e .' -

POKC CMAIRM6~

The proposed Abnormal Operating Procedure (AOP) change will allow installation of tc...,~i-i cooling to the "A" Spem Fuel Pool Heat exchanger ..

(SFPHx) in the unlikely event of a Service Water (SW) supply or return piping rupture. The AOP will provide guidance on using a 21/2 or 3" firehose supplied from either or both "A" and "B" Adams Filter druin line (SW.v.216/244) in the PAB l':e heat exchanger return now path will be throu6h a 21/2 or 3" hose which will connect to the service water retum header drain valve (SW.y.6 43A) located in the east end of the PAB lower level.

1.2 Annects of the chanee Evsluateri This safety evaluation will address the consequ 'cas on Spent Fuel Cooling and the SW system when temporary cooling supply and return hoses are used to supply the "A" SFPHx. aAer the normal Spent Fuel Pool (SFP) SW supply and/or retum line has failed.

1.3 Saferv Evaluation s-mmy 1.3.1 Reston for chanee . The use of temporary SW supply hoses for the "A" SFPHx would provide an attemate method of responding to a loss of SW Dow to the SFP cooling system.

1.3.2 Conclusions - The propbsed changes are safe for the following rease.ns:

1.3.2.1 The change will not adversely affect any design basis accident in the current defueled mode.

1.3.2.2 The change will not result in any malfunctions which will adversely affect safety related equipment.

l.3.2.3 The change will allow sufficient Dow to the SFP cooling system.

1. --m orJ.12 = n,,, i,

, 1

]

W S.t-%f sf.sy. 97 -0001 R:v.0 Pasje 2 ef 2.

l.3.2.4 There is no increase la the probabiitty oroccunence or consequences of an accident or malfunedon of equipment Important to safety or affect any design basis accidents in the current defueled mode.

1.3.2.5 The change does not constitute an Unreviewed Safety Question (USQ) because it does not create a new accident or malfunction, or increase the probability of occurrence or se consequences of an accident or malfuncdon of equipment pre iously evaluated or reduce the margin of safety as defmed in the basis for any Technical Specification.

i

1.3.2.6 This change will prohibit the inovement of fuel and loads over the spent fuel pool while the firehoses sre insts!!ed.

1 1

5

. .. s -

AOP 3.2.-59 W-EV- 9*l- 0 001 R4 Page I of 2.

FlCURE 7.2 - SAFETY EVALUATION FORMAT (Use Attachment 8.A for Guidsace) .

Safety Evaluation Number: SV-EV-97-0001 Revision No: 1 Procedure NuuN OP 3.2-59 Revision No. _ 3

]

Safety Evaluation

Title:

Fmercenev Service Water Cooling for the "A" Soer t Fuel Pool Cooling Heat Exchanger PORC REVIEWED M T G. # 9 7'S I" I

1.

SUMMARY

INFORMATION o

1.1 Descrinticant he t Desien Change RC'CRAIRMAN The proposed Abnormal Operating Procedure (AOP) change will allow installation of temporary cooling to the "A" Spent Fuel Pool Heat exchanger (SFPHx) in the unlikely event of a Service Water (SW) supply or return piping mpture. The AOP will provide guidance on using a 21/2 or 3" firehose supplied from either or both "A" and "B" Adams Filter drain line (SW-V-216/244) in the PAB. The heat exchanger retum flow path will be through a 21/2 or 3" hose which will connect to the service wtter retum becder drain valve (SW V-643 A) located in the east end of the PAB lower level.

1.2 Ascects of the Chr we Evaluat+d T5ds safety evaluation will address the consequences on Spent Fuel Cooling and the SW system when temporary cooling supply and return hoses are u2ed to supply the "A" SFPHx, eAer the normal Spent Fuel Pool (SFP) SW supply and/or return line has failed.

1.3 Safetv Evaluation Summarv 1.3.1 Reason for change - The use of temporary SW supply hoses for the "A" SFPHx would provide an alternate method of responding to a loss of SW flow to the SFP cooling system.

1.3.2 Conclusim - The pmposed changes are safe for the following rea: ans:

1.3.2.1 The change will not adversely affect any design basis accident in the current defueled mode.

1.3.2.2 The change will not result in any malfunctions which '

will adversely affect safety relaad equipment.

. 1.3.2.3 The change will allow mafficient flow to the SFP cooling system.

NGP 3.12 _Rev.10 Page 7.2.1 of 9

AOP 1.1-59 ST. EN 0003 Rt kge 2 4 2 1.3.2.4 Tnere is no increase in the probability of occurrence or consequences of an accident or malfunction of equipment importapt to safety or affect any design basis accidents in the current defueled mode. ,.,

1.3.2.5 The, change does not constitute an Unreviewed Safety Question (USQ) because it does not create a new accident or malfunction, or increase the probability of occurrence or the consequences of an accident or malfunction of equipment previously evaluated or reduce the margin of safety as defined in the basis for any Technical Specification.

1.3.2.6 His change will prohibit the movement of fuel and loads over the spent fuel pool while the firehoses are installed.

i l

I NGP3.12 Rev.10

' Page 7.2.2 of 9

l A0P '4.1-59 sY-DV-97-oo01 b FIGURE 71 - SAFETY EVALUATION FORM AT .

"U

) (Use Auschment s.A for Guidance)

SY-EV-97 0001 Revision No: 2 Safety Evaluation Number:

AOP 3.2-59 Revision No. 3 Procedure Number:

Safety Evaluation

Title:

Fmemenev Service Water Can11no for the ME ' bl Pool CoolIno Heat Frehanoer [~g*['g7 7 ,

DATE * * -f 7

1.

SUMMARY

INFORMATION PORC CHAIRM/Ji 1.1 Decerintion of the necien chance The proposed Abnormal Operating Procedure (AOP) change will allow installation of temporary cooling to the "A" Spent Fuel Pool Heat exchanger (SFPHx) in the unlikely event of a Service Water (SW) supply or retum piping rupture. The AOP will provide guidance rg asing a 21/2 or 3" firebose supplied from either or both "A" and "B" Adams Filter drain line (SW V-216/244) in the PAB. The heat exchanger retum flow path will be through a 21/2 or 3" hose which will connect to the senice water retum header drain valve (SW-V-643 A) located in the east end of the PAB lower level.

h 1.2 Annects of the Chance Evminated This safety evaluation will address the consequenc-s on Spent Fuel Cooling and the SW system when temporary cooling supply anc etum hoses are used to supply the "A"SFPHx, after the normal Spent Fue, Jool (SFP) SW supply and/or return line has failed. Also, the safety evaluation addresses the maintenance and 2 testing when supply a.nd/or renirn hose are in place.

1.3 Rafety Evaluation Summarv 1.3.1 R**=on for Chance - The use of temporary SW supply hoses for the "A" SFPHx would provide an altemate method of responding to a loss of SW flow to the SFP cooling system.

1.3.2 Conclusions - The proposed changes are safe for the following reasons:

1.3.2.1 The change will not adversely affect any design basis accident in the current defueled mode.

1.3.2.2 The change will not result in any malfunctions which will adversely affect safety related equipment.

NGP 3.12 Rev.10 Page 7.2.1 of 9

-'N=- .iem.

-~

A0P 3.2 -5 9 4

r/.EV- 91-000( Rev.1

,,- Paje 2 of 2 1.3.2.3 The change will allow sufficient flow to the SFP cooling system.

1.3.2.4 There is no increase in the probability of occurrence or consequences of an accident or malfunction of equipment

., important to safety or affect any design basis accidents in the current defueled r.1:xie.

1.3.2.5 The change does not constitute an Unreviewed Safety Question (USQ) because it does not create a new accident or malfunction, or increase the probability of occurrence or the consequences of an accident or malfunction of equipment previously evaluated or reduce the margin of safety as dermeel in th: basis for any Technical Specification.

1.3.2.6 This change will prohibit the movement of fuel and loads over the spent fuel pool while the firehoses are installed.

l l

NGP 3.12 1 Page 7.2.2 of 9 s

j ,

ACP 1.2-2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page 1 of 3 Safety Evaluation Number: SY rV.97 0103 Revision: 0 Document Number: AOP 3.2 59. TPC 97-281 Revision: 3 Document

Title:

Lon of soent ruel Pool cooling Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Brief Description of Change and Safety Evaluation Summary:

he proposed Abnormal Operating Procedure is being changed to designate the DWST as the emergency deionized water source for the Spent Fuel Pool. This is a substitution due to the slightly degraded condition of the RWST. Sufficient deionized water will be maintained in the DWST and portable hoses and a pump will be available to transfer water from the DWST to the Spent Fuel Pool.

This change does not constitute an unreview ed safety question because:

Dere 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 Pool Cooling (SFC) system removes decay heat transmitted to the spent fuel pool water and releases it to the Connecticut River via the Service Water System (SWS). If the heat sink is removed (failure of SFC, or failure of SWS) the y fuel pool temperature will slowly rise. The maximum possible temperature is 212 degrees. A recent calculation indicates the highest actual temperature will be approximately 170 degrees.

As specified in the SER far amendment 188, the safe storage of spent fuel is assured even if the pool reaches the boiling point (212 degrees) as long as makeup to the pool is provided. The three sources of makeup specified are: the PWST, the RWST and Fire Water System. The RWST was speciGed since it is a seismically qualified source of high quality water.

The use of DWST for makeup to the Spent Fuel Pool does not change the power being drawn from the grid. If power is lost the gas driven pump will independently provide makeup water to the Spent Fuel Pool. This proposed change has no impact on the probability that the loss of power will occur.

De proposed AOP revision w hich designates the DWST as the emergency deionized water source for make-up tt . .e SFP will not increase the probability of an accident previously evaluated in the SAR. This change will provide a reliable source of make-up water to the spent fuel pool in the event normal makeup methods are unavailable.

The proposed AOP revision designates the DWST as the emergency deionized water source for the SFP. This is a replacement (seismically qualified) for the RWST which is slightly degraded. He DWST meets the same seismic standard as the RWST. The DWST has sufficient volume to act as a makeup source to the SFP, capable of supplying 6 gpm of deionized water which provides the capability to maintain minimum water depth with deionized water due to loss by evapo.stion.

De boron concentration should remain the same due to makeup for evaporation. His procedure will not require the addition of boric acid to the SFP.

During a loss of offsite power the fuel pool heat sink will be lost and will have to be manually restored. The ability to pump from the PWST or RWST will be lost. Manual breaker alienments (no prcmedure) and or temporary fire hoses will have to be used to transfer water to the Spent Fuel Pool. This effect will be unchanged by the proposed change and therefore will not increase the consequences of a loss of offsite power.

SE970103 DOC

ACP 1.2 2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page20f3 Safety Evaluation Number: SY EV-97-0103 Revision: 0 Document Number: AOP 3 2-59. TPC 97-281 Revision: 3 Document

Title:

Loss of Soent Fuel Pool Cooline m

The 1 1/2 inch fire hoses are designed for 300 lb. well above the operation of the maximum pressure of the engine driven pump. The fire hoses are tested to 200 psig. The unlikely failure of the 300 lb. fire hose would temporarily affect makeup, however, if a hose w cre to sever or rupt. ire it could be easily replaced to reestablish SFP makeup.

The spent fuel pool door to the first floor will be kept closed while hoses are in the staircase to protect MCC 2 n the event of a hose rupture. Except for 115 KV switchyard the hoses will not be routed near equipment important to safety.

The fence will provide a barrier to keep fire hoses out of the 115 KV switchyard. If the 115 KV switchyard was sprayed with water this would be no different than rain. De piping from the RWST is not all seismically qualified and touk fait during a seismic event in which hoses would be required. The 1 1/2 inch fire hoses will provik th same level of assurance as the 150 lb piping used for the RWST to supply water to the SFP. Fire hoses have been evaluated previously for emergency operation.

De self priming centrifugal gasoline engine driven pump has more than the capacity and head required to provide emergency makeup to the SFP.

The proposed A OP revision replaces the R' /ST with a similarly qualified passive component, therefore, the cons.equet:ces remain within the bounds of analyzed malfunctions.

The possibility for an accident or malfunction of a different ty},e than any evaluated previously in the safety analysis report has not been created. The basis for this statement is:

The DWST is a fimetional equivalent to the RWST for makeup to the spent fuel pool and does not create the possibility of an accident or malfunction of a different type as use of the RWST requires non qualified piping and pumps as does the use of the DWST, Spent fuel building ventilation has been demonstrated capable of maintaining a negative pressure in the SFP building with the roll up door open for hoses, therefore, routing the temporary hoses through the normal personnel access door will not impact this ability to maintain negative pressure.

The margin of safety as defined in the basis for any technical specification has not been reduced. The basis for this statement is:

De margin of safety is maintained as long as the fuel pool level is maintained at least 20 feet above the spent fuel assembly per Technical Specification. This AOP will maintain the level in the pool due to loss of water by evsporation .

The technical specifics. tion does not define or identify the make up source to the spent fuel pocl. De SAR requires that a source of water be from a seismically qualified source or the diesel fire pump. The DWST is the equivalent (seismically qualified) of the RWST as an emergency deionized water sourci to the SFP during a loss of cooling.

51:970103 DOC y) ~ s

ACP 1.2 2.42 Rev. I MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page 3 of 3 Safety Evaluation Number: SY.Ev.97 0103 Revision: 0 Document Number: AOP 3.2 59. TPC 97 231 Revision: 3 Document

Title:

Im of wa ruel Pool Cooline

2. Reason for the Change:

The RWST is slightly degraded and may be adding to the source term for tritium in the ground water. The DWST provides a '

comparable substitute, allowing the RWST to be drained.

i Preparer Date SE970103 DOC n n -

ACP 1.2-2.42 Rev. I MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page 1 of 2 Safety Evaluation Number: SY-EV 97-0020 Revision: 0 Document Number; _ENG 1.7102 Revision:, 2 Document

Title:

n SrP Coo ne llent INehancer Performance Test Revision Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1.13rief Description of Change and Safety Evaluation Summary:

This evaluation addresses an historic revision to ENG 1.7-102, " Spent Fuel Pool Cooling IIcat Exchanger and Pump Performance Test" which was PORC approved on 8/2/96. This revision (2 Majcr) of ENG 1.7-102 increased the spent fuel pool temperature at the onset of the test by 10* and changes the acceptance criteria for the plate heat exchanger (E-101D) from a cahulated Overall lleat Transfer Coefficient to a calculated Fouling Factor among other system line-up changes to ensure NPSil requirements are met as required by PDCR 1592 " Connecticut Yankee Spent Fuel Pool Rerack".

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 a previously evaluated malfunction of equipment (i.e. partial or total loss of SFP cooling)important to maintain the spent fuel pit bulk temperature below 150*F will not be effected by this

' change because at no time is the equipment expected to perform in conditions outside of their design basis or different than what has previously been evaluated to be acceptable. The probability of occurrence of previously evaluated accidents in not effected by this change since the a Fuel Handling Accident is umelated to this performance of this procedure, which simply increases pool temperature test range by 10*F and measures heat exchanger performance by utilizing previously approved system alignments.

The changes to ENG l.7-102 have no effect on the consequences of a postulated failure of a loss of spent fuel pool cooling because during the performance of this test the system is operated in accordance with normal operating procedurer. Thus, no new failure mechanisms are introduced. In the event of a loss of forced cooling at any time,(i.e.

due to a loss of offsite power), power to at least one spent fuel pool cooling pump will be established by implementing EOP 3.1 10. This EOP is implemented in the event of a loss of power and directs the restoration of power to all necessary station loads as soon as possible. These activities can be accomplished to restore power within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> to preclude boiline in the spent fuel pool. Should failure to restore power in sufficient time to prevent boiling to occur, makeup to the pool can be accomplished by several means as outlined in AOP 3.2 59, " Loss of Spent Fuel Cooling" There are also no effects on the consequences of the previously evaluated accidents (i.e. fuel handling accident) due to this procedure change because the changes do not effect the handling of fuel nor do they adversely effect the actions necessary to respond to a fuel handling accident.

The possibility for an accident or malfunction of a different type than ety evaluated previously in the safety analysis report has not been created. The basis for this statement is:

The equipment in the spent fuel pool cooling system was not modified in anyway or operated in a manner outside currently approved procedures, during the performance of ENG 1.7102. In addition, testing the heat exchangers at a higher temperature which is well within their design basis will not adversely effect their ability to perform their safety function. His revision to ENG 1.7-102 increases the spent fuel pool temperature at the onset of the test by 10*F and changes the acceptance criteria for the plate heat exchanger (E-10-IB) from a calculated Overall Heat Transfer Coefficient to a calculated Fouling Factor among other system line-up changes to ensure NPSH requirements are met as FRht35E20 DOC

ACP 1.2 2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page 2 cf 2 Safety Evaluation Number: SY-EV.97-0020 Revision: 0 Document Number: ENO 1.7102 Revision: 2 Document

Title:

,,, SFP Cooling lleat Exchanger Performance Test Revision ,

required by PDCR 1592," Connecticut Yankee Spent Fuel Pool Rcrack". These changes do t.ot in any way create the possibility of a new unanalyzed accident.

E E The margin of safety as defined in the bads for any technical specification has not been reduced. The basis for this statement is-The procedure change simply ssrves to enhance the ability to ensure proper monitor the "B" heat exchanger and therefore the reliability of the SFPCS.

2. Reason for the Change:

(a) Increasing the star'ing temperature of the spent fuel pool yields more accurate test result.' by increasing the measurable delta T across the heat exchanger being tested. Ari evaluation of the Spent Fuel Pool Coolir g System (SFPCS) was performed for CYAPCo License Amendment 188 and determined the cooling system has sufticient capacity to maintain the bulk pool temperature at, or below,150*F (increased from 140*F) for any postulated discharge scenario, including a failure of the most effdent spent fuel pool cooling pump. Therefore, the starting temperature of ENG 1.7-102 was proportionately increased from 125' to 135' to yield more accurate results. As a result of this procedure change, a 15' margin has been maintained between staning tet.'p to maximum temperature. (b) Changing the acceptance criterion from a calculated U, overall heat transfer coefficient to f, a calculated fouling factor, was done for Maintenance Rule Trending purposes. Since system parameters can not be duplicated every time the heat exchanger is tested and are always different than design parameters, trending using the OllTC is not useful. A comparison of the calculated fouling resistance to the design maximum fouling resistance provides indication of the heat removal capability of the heat exchanger (Ref.1.4.2) (c) Changing the System line-up by opening both suction and discharge lines to the operating pump was required by PDCR 1592 to ensure adequate NPSil margin and reduce head losses in the piping. The thermal hydraulic analysis which utilized these new line-ups are discussed in the safety evaluation for the PDCR and bound this procedure change.

Preparer R. W. Kasuga Date 1/14'98 9

FRAT 3SE20 IX)C

ACP 1,2-2.42 Rev. ) MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report ,

Page 1 of 2 Safety Evaluation Number: S%r%97-0057 Redsion: _Q Document Number: ENO 1.7102 Revision: 3  ;

Document

Title:

SFP Cooline llent Euhnneer Performance Trst Provide a brief description of the change and a summary of the Safety Evaluation in the format below,

l. Brief Description of Change and Safety Evaluat on Summary:

His evaluation addresses a revision to ENG 1.7102," Spent Fuel Pool Cooling Heat Exchanger and Pump Performance Test" This revision (Rev. 3 Major) of ENG 1.7102 will change the ame mt of test data acquired and how it is analyzed to determine acceptability of the SFP cooling system. It also adds administrative steps for tracking en!ry and exit to/from Tech Spec 3/4.9.15 due to isolation of the "B" SFP heat exchanger while testing the "A" hea<. exchanger. None of these changes result ia a change to the way testing was previously perfo::ned so they can not adversely affect system operation.

Therefore the change is considered safe.

1 his 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 revises the methodology for analysis of data and monitoring of the SFP pumps. The initiation of a Fuel llandling Accident is unrelated to the performance of this procedure, which measures heat exchanger perforriance by utilizing previously approved system alignments and procedures. There are no effects on the consequences of the Design Basis Accident (i.e. fuel handling accident) due to this procedure change because the changes do not effect the handli < af fuel nor do they adversely effect the actions necessary to respond to a fuel handling accident.

The changes to ENG 1.7-102 will not have an adverse impact on the service ~ater, spent fuel or any other system important to maintain the spent fuel pit bulk temperature below 150'F because at no time is the equipment expected to perform in conditions outside of their design basis or different than what has previously been evaluated to be acceptable. The changes have no effect on the consequences of a postulated failure or loss of spent fuel pool cooling because during the performance of this test the system is operated in accordance with normal operating procedures.

Rus, no new failure mechanisms are introduced in the event of a loss of forced cooling at any time, (i.e. due to a loss of offsite power), power it it least one spent fuel pool cooling pump will be established by implementing EOP 3.1-10, fhese activities can be accomplished to restore power within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> to preclude boiling in the spent fuel pool. Sheuld failure to restore power in sufficient time to prevent boding to occt , makeup to the pol can be accomplished by several means as outlined in AOP 3.2 59," Loss of Spent Fuel Cooling".

The possibility for an accident or malfunction of a different type than any evaluate previously in the safety analysis repor, has not been created. The basis for this statement is:

The revision to ENG 1.7-102 changes the value calculated for determining if the "B" heat exchanger is acceptable, adds a Tech Spec Action statement entry / exit due to a previous Tech Spec change (Amendment 188) and reduces the number of SFP pump flow points necessary to monitor the pump performance. These changes do not in any way create the possibility of a new unanalyzed accident. The eqaipment in the spent fuel pool cooling system will not modified in anyway or operated in a manner outside ct':rently anproved procedures, during the performance of ENG 1.7 102.

rRM3SE57. DOC 2

ACP 1.2 2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (h)(2) Report Page 2 0f 2 Stfety Evaluation Number: SY EV-97-0057 Revision: 0 Document Number: ENO 1.7102 Revision: 3 Documer.t

Title:

srP Cooline llent Exchanoer Performance Test The margin of safety as defined in the basis for any technical specification he= not been reduced. The basis for this statement is:

The procedure change simply serv-s to enhance the ability to ensure proper monitoring of the "B" heat exchanger i.nd therefore the reliability of the SFPCS. No system changes are be implemented or Technical Specification sections, other than the bases, being revised.

2. Reason for the Change:

The aosaton of the steps for Tech Spec 3/4.9.15 Action statement entry and exit are due to the requirement added by Amendment 188 wh:ch requires the "B" SFT heat exchanger to be inservice when the core is offlosded. The "B" heat exchanger must be isolated while testing "A". The acceptance criteria for the "B" heat exchanger is being changed from a value of"U" which varies substantially with temperature and flow, to a very trendable and useful value of heat exchanger overall fouling factor, which is less dependent on test parameters and accurately predicts when the unit must be cleaned. The reduction its the number of Spent Fuel Pool pump flow data points from 7 to 2 brings the trending method in line with IST testing or. other safety related pumps. The previous revision of the ENO had just added 'he requirement to measure pump ficw and it was mainly meant as a one time test to validate the Holtec SFP cooling system model. Retaining the acquisition of pump flow is useful for Maintenance Rule System monitoring.

Preparer R.W. Kamen Date 1/14/98 V

FRM3SE57. DOC

- - _ - - _ _ _ - . _ - . . . .- . _ ~ --

EOP ').1- to sf.EV 17-OM4 foy I 4 3 FIGURE 7 2 - SAFETY EVALUAlON FORMAT Safety Evaluation Number SY EV-97-QQQQ Revision No. Q Procedure Change Number 917&B Revision No. Q Plar.Chango

Title:

Temporary Procedure Change to EOP 3.1 "O Rev.18

1.

SUMMARY

INFORMATION 1.1 Description of the Chance OCR CY 97002 installed a check valve in the Service Water supply line t 7 . rbi

!iest Exchangers. The check valve was installed to prevent water fro - ,

(backflowing) from the piping following a loss of normal power (LNP) ano _.sociated loss of Servk,e Water, as identified by ACR 97-0119. Draining of this pipe has the potential tc result in Service Water column separation in the high areas of piping in the Spent Fuel Building (Creare Technical Memorandum TM 1788a). Repressurization of the piping by a restarted Service Water Pump, after column separation has occurred, could result in a water hammer which has the potential to damage Service Water piping. This modification prevents the piping from draining for loss of power events with a nomial diesel start, thereby precluding the water hammer (refer to SY-EV-97-0003).

The check valve has a maximum permissible leakage of 2 gpm. This leakage is based on a Service Water flow interruption of 48 seconds, which corresponds to the time that is required for a diesel to start and load a Service Water pump after a loss of offsite power.

For extended loss of power events, EOP 3.1-10 is being revised so that the restart of a Service Water pump after a prolonged loss of flow coupled with slight leakage of the new check valve will not result in the sudden pressurization of a strction of voided Service Water piping, with its potential for water hammer.

The EOP change requires that the Emergency Stop buttons for the dieses be pushed and the Service Water pumps be placed in Trip Pull Out (TPO) if the emergency diesels fail to -

start after a loss of offsite power. An operator is then dispatched to locally isolate the Service Water supply line to the Spent Fuel Pnol Cooling heat exchangers. After the supply line is isolated, the Service Water pumps are placed back in Auto and the diesel Emergency Stop is reset. When power and Service Water flow are restored, the supply valves are locally opened to fill the supply piping without introducing a water hammer.

1.2 Aspects of the Chance Evaluated This Safety Evaluation will discuss the effects of the EOP change which resulted from the

. installation of the new check valve and its effect on malfunctions and potential accidents .

E0 f(.P '3.1-tosy. 9'f-00a b Paqs 1 cf I 1.3 Safety Evaluation Summarv 1.3.1 Description of Change:

This EOP change adds steps to isolate the Sevice Water supply line to the Spent Fuel Heat Exchangers during a loss of offsite power with failure of onsite power sources.

1.3.2 Reason for Change:

The check valve prevents rapid draining and column separation in the Service Water supply line after a loss of normal power / loss of Service Water flow, assuming a normal diesel start. Draining of the line could result in water hammer when a Service Water pump restarts. The EOP is being revised to preclude a water hammer during loss of power events of longer duration, during which the piping could drain by means of normal (allowable) leakage through the new check valve.

1.3.1 Safety Evaluation:

This change is safe for the following reasons:

The change to the EOP will not adversely affect any design basis accident applicable to the current mode (defueled).

Tne change to the EOP will not introduce any malfunction which cou'd adversely affect safety related equipment.

The change to the EOP will not adversely affect the ability to cool the Spent Fuel Pool.

The change has been determined to be safe and not to cor.stitute an Unreviewed Safety Question because:

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 to the EOP affects Service Water Spent Fuel cooling supply which can not affect any design basis accidents in the defueled modes. It also did not increase the probability of any malfunctions.

THE P.,SSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATi'D PREVIOUSLY IN THE SAFETY ANALYS!S REPORT HAS NOT BEEN CREATED. The basis for this statement is: '

This change can not create a new accident since the emergency stopping the diesel ecd docLS the stan of the SW pumps can not initiate a fuel handling accident, a spent fuel pool draindown or a radioactive materials release. The potential malfunctions associated with this change have been evaluated and determined to be bounded.

THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:

De margin of safety is not afDeted by this change since the consequences or probability o.'

occurrence of an accident is not affected. .Also this change does not effect the consequences or 4 , L

EoP 'A.I-to 9f EV- 97-000le Fay 3 cf 3 probability of occurrence of a previously evaluated malfunction important to safety and does not create the possibility of a new malfunction.

Did this change require a change to the Technical Specifications?: No l

9

)

W0\/'._ 1 M e O Attschment 7 10CFR50.59 Summary Format [EoP !,.l. '48d[

faaje i of 2.

This summary applies to PDCR -

Setpoint Change Test Procedure Tech Requirements Manual Change Experiment Tech Spec Basis change only Procedure Change E0P 3.1-4 8 TSAR Changes Jumper Bypass

1. Other:

Titl6: EOP 3.1-40, Loss of Refueling Cavity Inventory Preparer: D.J. Parker Date: 11/06/96

2. Description of Change:

The change revises the Loss of Refueling Cavity Inventory procedure. The changes improve the guidance provided to the p operators for handling a loss of Refueling Cavity Inventory.

3. Reason for the Change:

, The procedure was reviewed and areas where improvements in the instructions to the operators could be made wers found.

The charige adds those improvements.

4. Safety Evaluation:

a. This change was safe for the following reasons:

The procedure provides appropriate guidance to mitigate a-loss of Refueling Cavity Inventory. The procedure changes provide additional guidance for rastoring the RHR system if an RHR pump cavitates. The change improved the guidance for loss of inventory with a latched asse261y in the upender and for proper CAR fan operation.

O .

t

,. ,-a nna anno snt:.i n nimuana rns tu 000 Mu une . .

E0P 1,.1- 4&

Pbst .9 2.

iiOV1l'.t,-

b.

O This change does not constitute AN UNREVIEWED SAFETY

. QUESTION becauset THERE IS NO INCREASE IN THE 7BOBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SATETY FREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT. The basis for this statement is The appropriate procedure is used for restoring RHR, which adds water of greater than or equal boron concentration than the water that has been drained from the refueling cavity and therefore dilution accident. does not increase the likelihood of a boron The changes do not effect the refueling equipment and therefore do not increase the likelihood of a fuel handling accident. The changes use equipment within its design capabilities.

THE POSSIBILITY FOR AN ACCIDENT OR 14ALFUNCTION OF A DI FFERSr. .' TYPE THAN ANY FVALUATED PREVIOUSLY IN TdE SAFETY ANALYSIS REPORT HAS NOT BEEN ~ 7:I.TED. The basis for this statement ist The procedure uses equipment within its capabilitien. No hardware modifications were made.

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 enhanced the guidance provided to the operators for a loss of refueling cavity inventory. The modification improved guidance for restoring RHR if an RER pump cavitates. The modified guidance when a latched assembly is in the upender provides grc ter assurance that adequate shielding will be maintained.

C. Did this change require a charge to the Technical Specifications? Vee '/ No f

ACP 1.2-2.42 Rev.1 MAJOR E P 2.14 Form 3 - 10 CFR 50.59 (b)(2) L port Page 1 of 2 Safety Evaluation Number: SY.EV.97 0078 Revision: 0 Document Number: Not Annliemble Revision:

Document

Title:

Cannecticut Yanbe Fire Protectinn ProcragtManual Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Brief De'etiption of Change:

l Lis change implements the Fire Protection Program Manual which replaces Nuclear Group Procedure 2.14 (Connecticut Yankee Procedure 2.5 6), Nuclear Plant F) e Protection Program Requirement. De Fire Protection Program Manual s

addresses the differences in the Fire Protection Program that have arisen from becoming a plant in the decommissioning process. In doing so, all references to Appendix R and safe shutdown have been removed. In accordance with 10CFR50.48 (f), the Fire Protection Program now establishes the fire protection policy for the protection of structures, systems, and components (SSC) from fires that could cause the release or spread of radioactive materials. His includes those SSC important to the operation of the Spent Fuel Pool Systems. Many major parts of the previous prt,i; ram have not changed, la particular, administrative controls on hot work, combustible loading, fire brigade duties and training have not changed. Tests and inspections of fue protection systems and components have not changed. De three major changes associated with the issuance of the new Fire Protection Program Manual are:

A. Fire Protection Program Organization

%e program addresses the new Connecticut Yankee organization, and their responsibilities.

B. Removal of Appendix R and Safe Shutdown criteria.

This removes the references to Appendix R and to the need for safe shutdowm.

C. Incorporation, Reference, and Linkage of Fire Protection Program Documents 2, Reason for the Change:

. This change will incorporate in one document a list of all references, and will provide the linkages necessary to effectively manage the program during the decommissioning of the plant..

3.~ Safety Evaluation Summary -

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 an accident (fire) occurring is affected by either an increase in ignition sources or the addition of a combustible that is easier to ignite. His proposed change does not affect the ignition sources or the type of combustibles in any plant area.

Here are no malfunctions of equipment important to safety that are considered applicable to this proposed change. The malfunction of equipment due to fire could be affected by the duration of the fire, which could cause either more or less damage. Le length of a fire can be affected by the type of combustible. This proposed change does not involve any change of combu tible material, nor does it add any new equipment.

5059FPP3. DOC

ACP 1.2-2.42 Rev.1 MAJOR NGP 2,14 Form 3 - 10 CFR 50.59 (b)(2) Report Page 2 of 2 Safety Evaluation Number: SY.EV-97-007R Revision: 0 Document Number: Not Anplicable Revision: _

Document

Title:

Cnnnecticut Yank ee Fire Protection Program hiannal The consequences of a fire can be affected by the type of combustibles buming, the location of the fire, or the length . : time the fire burns. This proposed change does not reduce the level of fire protection in any plant area containing or adjacent to radioactive waste materials.

This proposed change does not involve an increase in ignition sources, a change in type or quantity of combustible material, or any new equipment.

The possibility for an accident or malfunction of a different type than any previously evaluated in the safety analysis report has not been created. The basis for this statement is:

The type of fire that can cccur in an area is dependent upon the type of combustible in the area. This proposed change does not change the type of combustibic in any plant area. It does not change the fire protection design basis for the plant.

The malfunction of equipment due to fire could be affected by the duration of the fire, which could cause either -

more or less damage. The duration of a fire can be affected by the type of combustible. This proposed change does not involve any change of combustible material, nor does it add any new equipment.

f The margin of safety as defined in the basis for any technical speciti:ation has not been reduced. The basis for this statement is:

This proposed change establishes a new Fire Protection Manual, a high tier document that does not change the .

elements of the Fire Protection Program that now apply to Connecticut Yankee, including those procedures used to implement the program. The Fire Protection Program has been removed from Plant Technical Specifications and placed in the Technical Requirements Manual. There are no fire protection margins of safety def med in the technical specifications.

Preparer Jdward A. Sawver. Fire Protection # BM Date /M/

f '- v q y

5059FP03. DOC

ACP 1.2-2.42 Rev.1 M AJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page 1 of 2 Safety Evaluation Number: SY EV 97 0070 Revision: 0 Document Number: TPC 97-188 to NOP 2.10-1 Revision:

Document

Title:

SFP Cooline System Oneration (Rev 14)

Pros ide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Brief Description of Change and Safety Evaluation Summary:

This evaluation addresses a revision to NOP 2.10-1, " Spent Fuel Pit Cooling System Operation", nis change will improve the method of backwashing the "B" heat exchanger (E 10-IB) by isolating the Service Water How to the "A" heat exchanger, thus forcing more flow through "B" It also adds administrative steps for tracking entry and exit to/from Tech Spec 3'4.9.15 due to isolation of normal SFP cooling, even though "B" will still be removing some heat load.

None of these changes result in a dramatic change to the way backwashing was previously performed so it will not adversely affect system operation. A slight SFP temperature increase may result during the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> duration of backwashing. Therefore the change is considered safe.

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 equipment important to safety previously evaluated in the safety analysis eport. The basis for this statement is:

1 i

This change simply isolates normal SFP cootmg for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> which could result in a slight increase in SFP tempera *ure.

The initiation of a Fuel 1landling Accident is unrelated to this performance of this procedure, since only the se: vice water valve alignments will be affected. There are no effects on the consequences of the Design Basis Accident (i.e.

fuel handling accident) due to this procedure change becattse the changes do not effect the handling of fuel nor do j they adversely effect the actions necessary to respond to a fuel handling accident.

Equipment necessary to main'.ain the spent fuel pit bulk temperatve below 150*F will not be effected by this i.hange because at no time is the equipment expected to perform in conditions outside of their design basis or different than what has previously been evaluated to be acceptable. The proposed change will temporarily realign SW flow in an attempt to clean the secondary side of the "B" heat exchanger. The changes to NOP 2.10-1 have no effect on the consequences of a postulated failure of a loss of spent fuel pool cooling because during the performance of this change the system is operated similar to previously approved normal operating procedures, with the exception ofisolating the "A" heat exchanger. Spent Fuel Cooling has been routinely taken out-of-service in the past for similar durations with no consequences. Thus, no new failure mechanisms are introduced. An EOP would be implemented in the event of a loss of power and directs the restoration of power to all necessary station loads as soon as possible. These activities can be accomplished to restore power in time to preclude boiling in the spent fuel pool. Should failure to restore power in sufticient time to prevent boiling occur, makeup to the pool can be accomplished by several means as outlined in AOP 3.2 59," Loss of Spent Fuel Cooling".

The possibility for ari accident or malfunction of a different type than any evaluate previously in the safety analysis report has not been created. The basis for this statement is:

This TPC simply realigns Service Water to the "B" heat exchanger for a short period of time to flush out fouling. No problems were previously encountered during backwashing activities. Therefore, this change does not create the possibility of a new unanalyzed accident. In addition, the equipment in the spent fuel pool cooling system will not be modified in anyway or operated in an unusual manner as a result of this procedure change.

FRMISE70 DOC

ACP 1.2 2A2 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page 2 0f 2 Safety Evaluation Number: SWEV-97-0070 Revision: 0 Document Number: TPC 97188 to NOP 1101 Revision:

Document

Title:

SFP Coolino System Ooeration (Rev 14)

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 change simply serves to enhance the ability to clean the "B' heat exchanger and therefore improve the reliability of the SFPCS. The temporary isolation of Service Water to the "A" SFP heat exchanger, and entry into TS 3/4.9.15, does not effect the margin of safety or Tech Specs.

2. Reason for the Change:

Isolating the "A" SFP heat exchangu during backwashing "B" will improve the efficiency of the flush by dramatically increasing flow.

The addition of the steps for Tech Spec 3/4.9.15 Action statement entry and exit are due to the requirement added by Amendment 188 which requires the "B" SFP heat exchanger to be insersice when the core is offloaded. Although the "B" heat exchanger can be technically considered to be inservice, it will not be in its normal, l- most efficient valve lineup for SFP heat removal.

Preparer R W. Katuca i~f - Date 1/14/98 v

FRM3SE70. DOC i

ACP 1.2-2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page 1 of 2 Safety Evaluation Number: SY-EV.97 0065 Revision: 1 Document Number: NOP114 7 Revision: 13 Document Titi.: _ waste Gas Degasifier Operation Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Brief Description of Change and Safety Evaluation Summary:

NOP 2.14 7, is an operating procedure used to remove the gaseous waste produced by the fission process when CY was in operation. With the plant shut down and no gaseous fission products being produced, operations is proposing a procedural addition to align the Primary Drains Tank (PDT) to the Boron Waste Storage Tanks (BWST) bypassing the degassification process and alignir.g straight to the Primary Vent Stack. By doing so, several components will be bypassed that are not needed at the present time and will conserve on energy usage, but are available, if needed, for the future decontamination process. The equipment operational capabilities will not be altered in anyway to accomplish the proposed procedural addition.

This change does not constitute an unreviewed safety quenion because:

! There is no increase in the probahility 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 system as well as the equipment being util; zed will be operating within their designed capabilities. There are no modifications that will affect the system's pe:formance. With the plant shut down, theie are no gaseous fission products being produced. He potential for an uncor. trolled release of radioactive gases that will effect the safety of the plant, the personnel, the public, or the environmcat is less likely to occur.

The alignment of the PDT to the BWST, bypassing the degassification process and aligning straight to the Primary Vent Stack, will not increase the probability of equipment malfunctions. The equipment will be operating within its capabilities.

The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report has not been created. De basis for this statement is:

With the plant shut down, there are no gaseous fission products being produced. The contents of the Waste Gas Decay Tanks have been permitted and released. The primary gas in the system is nitrogen which has no adverse effect on the safety of the plant, the personnel, the public, or the environment. The lineup and operation of the involved equipment is all within its designed operating capabilities.

He margin of safety as defined in the basis for any technical specification has not been reduced. The basis for this statement is:

With the plant shut down, there are no gaseous fission products being produced. What fission gases that had been produced, prior to shut down, have been planned, permitted, and released. Technical Specification Bases 3/4.11.2,3/4.11.2.1, 3/4.11.2.2,3/4.11.2.") discuss the permitted release atid dose limits of Gaseous Effluents, as described in 10 CFR 20 and 50.

They were reviewed, and under our present operating condition all of the above mentioned would be well belo.e the allowabk release and dose limits for the safety of the piant, the personnel, the public, and the environment. l

\

SE970065. DOC

ACP 1.2-2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (h)(2) Report Page 2 of 2 Safety Evahmtion Number: SY-EV.97 0065 Revision: 1 Document - 1ber: NOP114 7 Revision: 13 Document

• 4: . Waste Gas Dennairier Oncratinn

2. Reason for the Change:

ne primary reason for the proposed procedural addition is to use the equipment that is actually needed to process the gaseous effluents that are applicable to a non-operating plant. He bypassed or isolated equipment will remain in lay up and can be readily realigned if needed for the decontaminatien process involved in decommissioning. At the present time, the principle gas that is found in the system is nitrogen, which w:ll have no impact on the safety of the plant, the personnel the public, or the environment.

(

1 Preparer _ Date SE970065. DOC

06A6 9 7-7 Attachment 1 NoP 1 20-4 HADDAM NECK PLANT DC' E Annual Report Summary of Changes Mark the Appropriate choice:

Design Change Setpoint Change Test Tech Requirements Manual Change Experiment Procedure Change Z Jumper Bypass

1. Change Number: HA Revision Number: HA

Title:

TPC 97-3 to NOP 2.20-4 "Mvnochlorite System Oreration" Preparer: n.w. Kasuga Date: 1/17/97 l 2. Description of Change:

This change allows isolation of the sodium hypochlorite injection system to the running service water pump (s) whenever the Connecticut River is below 35'F.

t

3. Reason for the Change:

The reason for this change is to allow freeze protection flow to the idle service water and fire protection pumps when the river is below 35'F, without violating the current NPDES permit.

4. Safety Evaluation:

a. This-change was safe for the following reasons:

1. The potential for micro and macrofouling and any resultant corrosion is significantly reduced at river temperatures below 35'F.

2.

The flow requirements of the Service Water system are much lower than at power conditions and therefore, significantly margin exists even if fouling were to occur.

3. The termination of the sodium hypochlorite system can not adversely effect any postulated accidents or safety related equipmen* in the current defueled mode of operation.

y/ .

WOP 2 20-4 Page 2.cf 2

/

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 allow the sodium hypochlorite system to be isolated at river temperatures below 35'F for aafety related equipment freeze protection. The hypochlorite system is not a safety related system and therefore can not adversely effect the probability of occurrence or consequences of an accident or malfunction of important equipment for the reasons stated in (a) above.

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 can not create a new accident or malfunction in the current defueled mode since the loss of the SWS, which would be the worst case scenario if the system were to somehow completely foul, is a previously evaluated accident / malfunction. Procedures currently exist for responding to a loss of service water event.

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 safety is not affected by this change since the consequences or probability of occurrence of an accident is not affected. Also this change does not effect the consequences or probability of occurrence of a previously evaluated malfunction important to safety and does not create the possibility of a new malfunction. Safety margin will actually be increase by providing freeze protection to safety related pumps (SW and FP).

c. Did this change require a change to the Technical Specifications: No

ACP 1.2-2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page 1 of I Safety Evaluation Number:_CY FV-97-0086 Revision: 0 Document Number: NOP 2 20-7 , Revision: 1 Document

Title:

Traveline Water Screen Oneration Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Brief Description of Change and Safety Evaluation Summary:

The procedure change allows two circulating water pumps to be run with two non-corresponding traveling water screens (e.g. A&C pumps with D&D screens).

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. De basis for this statement is that NPSH for all service water and fire pumps will be maintained and thus the pumps will be operable. REF Haddam Neck Plant UFSAR 10.4.6.3.

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 that there are no other possible failure modes that are introduced by this procedure change. The only possible failure mode is addressed in UFSAR section 10.4.6.3.

The margin of safety as defined in the basis for any technical specification has not been reduced. The basis for this statement is that there will be no reduction in the cooling capacity of the service water system per T/S 3/4.7.3. There will also be no l reduction in the ability of the fire water system to provide water for fire protection purposes.

l

2. Reason for the Change:

The change is required because two of the screens are currently out of service and two non corresponding circulating water pumps sre also out of service.

Preparer Stephen Willard Date 01/20'98 CIR.FRM3 DOC

ACP 1.2 2.42 Rev.1 MAJOR Form 3 10 CFR 50.59 (b)(2) Report (rev.1)

Page 1 of 2 Safety Evaluation Number; SY EV-97-0028 Revision: 0 Document Number: TPC 07113 Rev8 sion: 0 Document

Title:

NOP 2.23 l: Water Treatment Plant Operation (TPC 97-113 allows the water treatment plant to be coerated with an alternate flowpath tnat bvoasses the dearator and forwardine numes)

Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Drief Description of Change and Safety Evaluation Summary:

Temporary Procedure Change 97113 describes a valve line up that creates a different Oow path than normally seen in the water treatment system. This alternate flow is such that the outlet flow of water treatment pre-filter and pressure regulating valve flows directly to the Ecolochem truck without passing through the Vacuum Deaerator and forwarding pumps as described in the Haddam Neck FSAR.

De net effect of this change is that the water which is produced by the water treatment system follows a different flow path and is not deacrated. All the piping that is in water treatment is rated above the maximum pressure it will see and the oulet i

ficw of the system will be directed via its normal route, so the change in flow path has no effect on safety at all. Regarding the increased dissolved gas levels: during power operation an increase in dissolved oxygen levels of make-up water can increase crosion / corrosion rates and potentially increase the likelihood of a pipe rupture in either the secondary sHe or the RCS. Ilowever lladdam Neck Plant has formally declared its intent to cease power operation, in the pennanently shutdown and defueled state. the systems that are normally prone to erosion / corrosion are open to atmosphere and below the temperature w here elevated oxygen levels are a concern. Therefore the elevated level of dissolved gasses in the output of water treatment is not an unreviewed safety question and the TPC is considered safe.

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 that this pro:cdure change cannot cause or in any way affect any design basis accidents in a decommissioning state.

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 that the only effect of this procedure change is to produce water that has slightly elevated levels of dissolved gasses. This will have no effect on any accidents or equipment malfunctions.

The margin of safety as defined in the basis for any technical specification has not been reduced. De basis for this statement is that this procedure change has no effect on any technical specification.

{ WWS-FM3B DOC 1

_l

ACP 1.2 2.42 Rev.1 MAJOR 1

Form 3 - 10 CFR 50.59 (b)(2) Report (rev.1) 1 Page 2 cf 2 Safety Evaluation Number: SY EV 97 0028 Revision: 0 Document Number: TPC 97 Il3 Revision: 0 Document

Title:

NOP 2.23 1: Water Treatment Plant Operation (TPC 97 113 allows the water treatment olant to be

.,ggerated with an attemate flownath that bvoasses the dearator and forwardinn numos)

2. .. ason for the Change:

The reason for the change is that ths inlet pipe to the vacuum deserstor developed a leak. The plant has a contract for the Ecolochem mobile deminerallier for a set duration citime, with fina .cial penalties for exceeding this time period. The leak in the pipe could not be repaired immediately, so a TPC was written to bypass the leaky pipe which also required l

circumventing the dearator and forwarding pumps as well. The leaking pipe in the water treatment system will be repaired and the TPC canceled after the current Ecolochem trailer has been exhausted and the water treatment system has been removed from service.

Preparer Steohen Willard Date 01/2098 wwsnt3n DOC w

Noe 2.24-3 SE- opM.%-53 Rev.)

Attachment 1 HADDAM NECK PLANT Annual Report Summary of Changes Mark the Appropriate Choice:

Denign Change , Setpoint Change Test Tech Requirements Manual Change Experiment Procedure Chance X Jumper Bypass

1. Change Number: NA Revision Number: 1

Title:

Cancellation of TPC to NOP 2.24-3 Preparer: R.W. Kasuca Date: 1/2/97

2. Description of Change:

This change eliminates the current restriction on parallel operation of the Primary Adams Filters.

i 3. Reason for the Change:

The reason for this change is to allow parallel Adams Filter operation when river conditions do not permit effective filtration with only one filter inservice.

4. Safety Evaluation:

a. This change was safe for the following reasons:

1. The hydraulic performance-of the Service Water System (SWS) will no' be affected by parallel operation of the filters since they are automatically bypass during a DBA.

2. Short term periods of low flow to the Containment Air  !

Recirculation (CAR) fans will not affect their ability to cool containment at power or post-accident.

3. Short term loss of CAR fan flow indication will not result in the inability of the piping to perform its safety function, nor will it cause a complete loss of the ability of Operations to recognize a SWS break in

NOP 2.14-3

$E- COM- 96-53 Rtv.I containment. 'S'

b. This change does not constitute A'1 UNREVIEWED SAFETY QUESTIO!1 because:

THERE IS NO INCREASE IN THE PROBABILITY OF OCCURREllCE OR i THE CO!1 SEQUENCES OF All ACCIDE!1T OR MALTUtvCTIO!1 OF EQUIPME!1T IMPORTAt1T TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS PEPORT. The basis for this statement is:

This change allow both Adams Filters to be placed in service. Both filters are automatically bypassed during a DBA ensuring design basis flow. If both filters were to clog while on line they would be cleaned expeditiously which is the current practice.

THE POSSIBILITY FOR AN ACCIDENT OR MALP')NCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement ist This change can not create a new accident since the loss of the SWS can not initiate a DBA. The potential malfunctions associated with this change are actually reduced due to reduced notential for plugging from lower flow rates through ea,.. filter, THE MARCIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHt1IQhL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:

The margin of safety is not affected by this change since the consequences or probability of occurrence of an accident is not affected. Also this change does not effect the consequences or probability of occurrence of a previously evaluated malfunction importart to safety and does not create the possibility of a new Malfunction.

c. Did this change require a change to the Technical Specifications: No

NoP 214-3 SY-av-97. col Rev 4 f.ge n of 'l FIGURE 7.2 RFETY EVALUATION FORMAT (Use Anachment 8.A for Ovidance)

Safety Evaluation Number: SY FV.97 001 Revision No: - 4 Procedure Number:

TPC 97.??? to NOP 124 3 Revision No: 17 Safeiy Evaluation

Title:

Emercenev service Water Cooline for the "A" Knent Fuel Pool Cooline Heat Exchaneer

1. E' fMARY INFORMATION 1.1 Descrintion of the Desien Chance Operation of the spent fuel cooling system with the A heat exchanger isolated with a hose manifold and the B heat exchanger in service. The A heat exchar.ger has been isolated to allow testing of the senice water check valve SW CV 963.

l l

The Abnormal Operating Procedure (AOP) will allow installation of temporary cooling to tne "A" Spent Fuel Pool Heat exchanger (SFPHx) in the unlikely ever.:

of a Service Water (SW) supply or retum piping rupture. The AOP will provide guidance on using a 21/2 ter 3" firehose supplied from either or both "A" and "B" Adams Filter drain line (SW V 216/244)in the PAB. The heat exchanger retum flow path will be through a 21/2 or 3" hose which will connect to the senice the water PAB lower retum level. header drain velve (SW V 643A) located in the east end 1.2 his of the Change Evaluated The safety evaluation will address the consequences of operating the B heat exchanger with the A heat exchanger isolated with a hose manifold, in case of a rupture this safety evaluation will also address the cons equences on Spent Fuel Cooling and the SW system when temporary cooling supply and retum hoses are used to supply the "A" SFPHx, after the normal Spent Fuel Pool (SFP)

SW supply and/or return line has failed. Also, the safety evaluation addresses tha.

ntaintenance and testing when supply and/or return hose are in place.

1.3 Safety Evaluation Summan 1.3,1 Reason for Change - To allow the operation of the spent fuel cooling system with the A heat exchanger in operation and the B heat exchanger isolated with a hose maniloid. Also, the use of temporary'SW supply hoses for the "A" SFPHx would provide an ahemate method of responding to a loss of SW flow to the SFP cooling system.

1.3.2 f

Gonciusions The proposed changes are safe for the following reasons:

NGP 3.12 Rev.10 Page 7.2.1 of 11 w

W t.14 -> 1 Ref a 4 SY-CV*91-00 Pas 1 of &

1.3.2 Conclusions The proposed changes are safe for the following reasons:

1.3.2.1 The change will not adversely affect any design basis accident in the current defueled mode.

1.3.2.2 The change will not result in any malfunctions which will adversely affect safety related equipment.

I.3.2.3 The change will allow sufficient flow to the SFP cooling system.

1.3.2.4 There is no increase in the probability of occurrence or consequences of an accident or malfunction of equipment imponant to safety or affect any design basis accidents in the current defueled mode.

1.3.2.5 The change does not constitute an Unreviewed Safety Question (USQ) because it does not create a new accident or malfunction, or increase the probability of occurrence or the consequences of an accident or malfunction of equipment previously evaluated or reduce the margin of safety as defined in the basis for any ---

Technical Specification.

1.3.2.6 This change will prohibit the movement of fuel and loads over the spent fuel pool while the firehoses are installed.

4

ACP 1.2 2.42  !

Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page 1 of 2 Safety Evalution Number: SY.Ev.97 0082 Revision: 0 Document Number: TPC NOP 2 261 Revision: 11 Document

Title:

,,, w iration of Control Air System Provide a brief description of the change and a summary of the Safety Evaluat ion in the format below.

1. Brief Description of Change and Safety Evaluation Summary:

A Temporary Procedure Change (TPC)is being proposed for NOP 2.261, Rev. I1," Operation of Control Air System". The proposed TPC will add a Step (6.3.7)i Sction 6.3 giving directions to perfenn a new Attachment (7)" Service Air to Control Air". The purpose of the TPC n.o cross connect the Service and Control Air Systems. The Service Air Compretsors will be the primary source for control air, while using the Control Air Compressors as a system bac) up.

Both systems have been reclassined as AVAILABLE for the decommissioning process and are not required for malr.taining the integrity of the Spent Fuel Pool. At the present time, both system compressors are cycling in rapid succession. The load and unload cycle time for the Service Air Compressors is approximately 10 to 15 seconds, while the load and unload cycle for the Control Air Compressors is approximately 20 to 30 seconds. Rapid cycling can be detrimental to equipment operability and reliability. Both systems produce compressed air at the same pressure,100 psig. The capacity of one Service Air Compressor is rated at 647 standard cubic feet per minute (scfm) while the rated capacity of two Control Air Compressors is a combined 336 (168 each) standard cubic feet per minute (scfm).

'IS.is change does not constitute an unreviewed safety question because:

'Ihere is no increase in the probability of occurrence or the consequences of an accident or malfunction of equipment important to safer / peviously evaluated in the safety analysis report. The basis for this statement is:

There are ,o accidents described or evaluated in the technical speci0 cations, the design basis documentation, or the FSAR associated with the Control and'or Service Air Systems. TM5e systems are not classined as having a Safety Related function in the operation of the plant.

There are no equipment malfunctions associated with the Control and'or Service Air Systems that are described or evaluated in the technical specifications, the design basis documentation, or the FSAR. These systems are not classified a: having a Safety Related function in the operation of the plant. Both systems have been reclassi0ed as AVAILABLE for the decommissioning process and are not required for maintaining the integrity of the Spent Fuel Pool.

The possibility for an accident or malfunction cf a different type than any evaluated previously in the safety analysis report has not been created. The basis for this statement is:

These systems are not classined as having a Safety Related function in the operation of the plant. Both systems have teen reclassified as AVAILABLE for the decommissioning process and are not req' aired for maintaining the integrity of the Spent Fuel Pool.

The margin of safety as defined in the basis for any technical specification has not been reduced.11e basis for t' is statement is:

The margin of safcty will not be reduced by the proposed cross connection of the Control and Service Air Systems. The documentation reviewed for this determination were as follows: Technical Specification 3/4.7 " Plant Systems," 3/4.0 Bases

• Lhaiting Conditions for Operation and Surveillance Requirements," FSAR Chapter 9.3 " Process Auxiliaries," 9.3.1 st:9700s2 Doc

ACP 1.2 2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page 2 of 2 Safety Evaluation Number: SY.rN 97 0082 Revision: 0 Document Number: TPC NOP 2.261 _ Revision: 11 Document

Title:

Operation of Control Air System

" Compressed Air Systems," 9.3.1.1 " Design Bases". 9.3.1.2," Control Air System" and " Service Air System". ENG 1.7 156, Rev. 2 " System Category Determination in a Decommissioned Plant", Attachment 12.2 " System Category Determination Document" Service Air System and Contro' Air System.

2. Reason for the Change:

Now that the plant is shutdown and being decommissioned the oemand for compressed air is down. Dy cross connecting the Control and Service Air we will develop a compressed alt system that will serve a dual function, w hile meeting the required air demands of the plant and at the same time accomplish r.* other goals. l'irst, we will help increase equipment operability and reliability by increasing compressor load cycle times. arid second, we can shutdown the equipment not needed for maintaining system pressure.

1 1

Preparer Date

$0970082 IX1C

ACP 1.2 2.42 3 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (h)(2) Report Page 1 of 2 Safety Evaluation Number: W rV 97 0073 Revision: O Document Number: PMP 9 l 61 Revision: _ original Document

Title:

R 18 Samnle Pomn Strainer CleA!dng Provide a brief description of the charge and a summary of the Safety Evaluation in the format below,

l. Brief Description of Change and Safety Evaluation Summary:

This safety evaluation is being written to evaluate PMp 9.161 R 18 Sample Pump Stralner Cleaning. His is a new procedure to provide guidance on how to clean the R 18 sample pump strainer. Bis procedure was toutinely performed in the past by appropriate personnel but without procedoral guidance. This procedure is being written to previde that guidance and document the performance of the strainer cleaning. During performance of this procedure it is required to remove RMS R 18 from service. In order to do this a Technical Specification Action Statement must be entered. This procedure provides the guldance to perform 151$ entry and apply the appropriate compensatory actions. Beyond this, the procedure is routine and normal and rninor in affect to other plsnt systems.

Performanct of this procedure is safe and does not impact any previous analyses nor will it create the requirement for any new analyses, his change also does not impact any decommissioning funding or goal or environmental regulations, nis change does not constitute an unreviewed safety question because no effect on the existing accident analyses result from

• his change. De procedure being evaluated performs routine plant maintenance and does not alter the ability of the plant to operate in a manner consistent with Techmcal Specifications, nere is no increa:e 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 that the procedure being ovaluated decs not effect any important to safety equipment, ne subject procedure isolates the R 18, Service Water Effluent Monitor (and associated sampling equipment), for a shoit period of time for routine maintenance. No other equipment is effected by this procedure. His equipment is not required to mitigate any accident and its malfunction cannot cause or worsen an accident.

De possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report has not been created. He basis for this statement is that this procedure is not performing any task outside routine maintenance. No Technical Specifications are being violated and the subject equipment is not being put in a condition outside its design basis. De R 18, Service Water Effluent Monitor (and associated sampling equipment) cannot initiate an accident and this procedure is not introducing any new equipment or operation th.! can cause a different failure than previously analyzed. His procedure performs routine maintenance only, ne margin of safety as defined in the basis for any Technical Specification has not been reduced. De basis for this statement is that equipment is not being operated outside the guidelines of Technical Specifications, ne procure being evaluated performs routine maintenance of plant equipment and is controlled in accordance with Technical Specifications.

No new equipment or mode of operation is being introduced by this procedure.

3STRNR. DOC

ACP 1.2 2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (h)(2) Report Page20f2 Safety Evaluation Number: SY.EV.97 0073 Revision: 0 Document Number: PMP9161 Revision: Oricinni Document

Title:

_ R.18 Samele l'enn Straint r cleanine

3. Renon for the Change:

This procedure is being written due to the lack of procedural guidance for implementation of this work. This procedure ensures that proper inethodology and compensatory actions are performed for perforir.ance of this work.

i i

Preparer Date 35TRNR. DOC

~ ~ -

ACP 1,2 2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (h)(2) Report Page 1 of 1 Safety Evaluation Number W.EV-97-0075 Revision: _00 Document Number:, RPM 1.6 8 Revision: _00 Document

Title:

Connecticut Yankee 11galth Physics Organintion Provide a brief description of the change and a summary of the Safety Evaluation in the format below, l Brief Description of Change and Safety Evaluation Summary:

Major revision to the procedure to include functional descriptions and organizational responsibilities and expectations.

The change reflects the organizational realignment or rediological engineering functions and radioactive material l handling and waste shipments to the llealth Physics Manager. The safety evaluation indicates that the functional l responsibility is not materially diminished by the realignment alreponing responsibility and that this is only an i adminhtrative change that does not affect the safety of the plant.

This change does not constitute an unreviewed safety question because:

There is no increase in the probability of the occurrence or the consequences of an accident or malfunction of equipment imponant to safety previously evaluated in the safety analysis repon. The basis for this statement is:

The organization of the llealth Physics Department is an administrative change which does not affect any initiating events associated with the consequences or malfunction of equipment of an accident.

The possibility for an accident or malfunction of a different type than any evaluated previously is the safety analysis repon has not been created. The basis for this statement is:

The organization of the llealth Physics Department does not affect any initiating events, therefore there is no assoclated impact due to organizational changes that can result in an accident or malfunction of a different type than previously evaluated..

The response of health physics personnel to limit the dose consequences of an event is pan of the Emergency Plan.

The qualification of personnel available to perform emergency functions is not being reduced as a result of this change.

The margin of safety as defined in the basis for ny technical specification has not been reduced. The basis for this statement is:

Technical Specification 6.0 was reviewed. No changes to the Technical Specifications are required as a result of this change. The organization of the Health Physics Department is not a basis for system parameters. Therefore the margin of safety is not reduced.

3. Reason for the Change:

Organizational restructuring collects similar functions and responsibilities under a common management structure to enable focus and prioritization on all radiological issues.

Preparer Date - - --

org3 DOC

ACP 1.2 2.42 Rev.1 M.UOR Form 3 a 10 CFR 50.59 (b)(2) Report Page l of t i Safety Evaluation Number: sv.Ev.970tu Revision: fi Document Number: FSARER 97.cY.*!7 Revision: 0 Document

Title:

Paarallatina d RPM '!.7 1 "Pratective cinehine" Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Brief Description of Change and Safety Evaluation tummary:

4 Change:

Deleted the requirement in the UFSAR 12.3.2 to require that protective clothing donning and removal be reflected in plant procedures.

Safety Evaluation Summary:

Personnel are trained in the appropriate use of protective clothing and the requirements for protective clothing are described 4

in work control documents such as the RWP. De change does not affect the use or availability of the protective clothing only the procedurai guidance for donning and removal.

nis is an administrative change not affecting the function or performance of any safety systems, structures or components or

affecting the availability of that equipment to perform iu safety function. Similarly, no equipment important to safety is affected by this change ne possibility or consequences of an accident or malfunction either previously evaluated or not l previously evaluated is not affected because the change is not related to any initiating events. The change does not affect the i decommissioning funds available for the plant or the environmentalimpact or the aval

ab!!ity of the site for unrestricted use.

j nerefore, there is no safety significance to this change and the proposed change is considered safe plant change.

3

2. Reason for the Change:

J The existence of the procedure is redundant to the training normally provided ano provides a negligable benefit.

Prepar i1 h

- 7, , w

- Date M /7 3 b e

$E PCM. DOC

@~.a? W 56  %~ % b SE- PDcR M0 Rev.0 Fesje i ef 2. 1

, SAFETY EVALUA110N FORM Safety Evaluation Number: SE PDCR 1510 Revision No. 0 PDCR-

Title:

Radiation Monitoring System Upgrsde

1. S u m m s rv In fn em s tin n 1.1 S a fe tv Ev31ustion Cenelusient . Electriesi It has been concluded that these modifientions are safe and do not invohe s.: unreviewed safety question, l.; Deverintinn of Chunve PDCR 1550 implements certain recommended plant improvements resultant from ongoing majatenance and operability problems

• with the subject Radiation Monitoring System (RMS) channels.

P ll (Containmtnt Particulate). R.12 (Containment Air). R. I .t A (Main Stackh R 15 (Air Ejector). R.16A/B (Steam Generator Blowdown). R.13 (Service Water Dischstge). and R 22 (Waste Test Tank Dischstge) are all being replaced with new open frame akid mounted sample systems manufactured by Victorcen. Inc. Each channel will be comprised of a locally mounted Universst Digital Ratemeter (UDR) for local rendout and control, which will comme.iicate with a control room mounted ScanRad computer s; v..s The ScanRad computer will be used by the operstors to perform trending, setpoint changes, channel indics:lon, check source operation, and annunciation. Control board annunciation will abo be provided exclusive of the ScanRad computer alarms.

R 16A. R 15 A. R 22. R.13, R.ll and R 12 will provide plant process computer indication of channel readings. Recorder

, inputs will not be altered from the existing strangement.

Automatic trip functions unique to R.I.lA. R 16A/B. RIS and R.22 will still be performed. Each new channel will consist of a new detector and associated input / output piping and valves, pumps as required, local UDR controls as reqaired, and input to the Scan ,

Rad computer system and recorders / plant computer. All mounting of new equipment will be evaluated for seismic 2 over I concerns. All new equipment is non-QA.

Deliberate entrance into Technical Specification Actiort Statements will be performed during this modifiention. Required contingency plans will be performed during these entrances. It is not considered that these entrances will increase the probability of occurrence of a previousiy evaluated accident, increase the probability of occurence of a previously evaluated malfunction of equipment impossnt to safety, increase the -

consequences of the previously evaluated accidents, increase the '

consequences of a previously evaluated malfunction of equipment important to safety, increase the possibility of an

, :ccident o. a different type than previously analyzed, or increase th,e possibility of a malfunction of a different type than

SUR. 0 2- 01.3 Su- PPcR 15.30 Rev.0 Ptge 141

• previously analyzed. Entrance into these action statements will not decrease the marigin of safety as all requ! red contingency plans will be enforced to ensure compliance with Technical Specifications.

l.) A r nee tt af the Chante Evslusted This safety evaluation considers the instrumentation. control.

power and functional operability aspects of the proposed modification.

i 1.4 Msifunetlene Evslusted The following credible failure modes are associated with the RMS Upgrade modification:

a) component allu re s ,

b)

• loss of power

ACP 1.2 2.42 Rev.1 MAJOR Form 3 10 CFR 50.59 (b)(2) Report Page 1 of 2 Safety Evaluation Number: S%EV 97 0001 Revision: 3 Document Number: .IPC-97 221 to SUR 5.7 217 Revision: 3 Document

Title:

Inic vice Testing of SW cV 963 Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

I. firief Description of Change and Safety Evaluation Summary:

Operation of the spent fuel cooling system with the A heat exchanger isolated with a hose manifold and the B heat exchanger in service. The Abnormal Operating Procedure (AOP) will allow installation of temporary cooling to the "A" Spent Fuel Pool licat exchanger (SFPlix) in the unlikely event of a Service Water (SW) supply or return piping rupture.

The AOP will provide guidance on using a 21/2 or 3" firehose supplied from either or both "A" and "B" Adams Filter drain line (SW V 216/244) in the PAB. The heat exchanger return flow path will be through a 21/2 or 3" hose which wih .onnect to the service water return header drain valve (SW V-643A) located in the east end of the Pall lower level.

This change does not constitute an unreviewed safety question because:

'Ihere 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:

Operation with only the B heat exchanger in service has no change on the probabiiity of occurrence of evaluated accidents since the normal operational alignment has the A heat exchanger isolated. The difference is the A heat exchanger is normally valved out and available with a valving change. In this case the A heat exchanger is isolated with a hose manifold and can only be placed back into serv'ce with hoses. The temporary cooling firehoses will be installed from the North or South SW header, vb. the Adams Filter drains, to the "A" SFPilx inlet with the retum hose connecting to the SW retu n header. The SW System (and thus the temporary hoses) do not directly interface with SFP, therefore, the response to and consequences of a fuel handling accident would be unchanged.

The installation of alternate emergency cooling in the event of a SW line rupture to the SFP will provide a4quate flow from a difTerent supply point ("A" and'or "B" Adams Filter drain). Although the alternate supply will not be filtered, the SW pump strainers should minimira the SFPilx fouling to a satisfactory level.

The unlikely failure of the hose will not adversely affect the SW system due to low plant heat loads, equipment out of-service and significant flow margins. Adequate flow would still be mairGd to the Emergency Diesel Generator llent Exchangers. in addition, if a hose were to sever or rupture it could be easily replaced to reestablish cooling and the supply isolation valves closed to regain service water integrity.

Previously installed hoses for alternate cooling were evaluated for seismicity. The firehose is flexible and will be run along the ground which results n the fircho

• being essentially continuously supported. Deadweight thermal and seismic losdings are not a concern.

The probability of sn LNP 4 also uaaffected by the proposed procedure change since the hoses do not interface with the station electncal distribution systerr nor do the hoses run in an area w hich could mitiate an LNP in the PAB.

Flooding issues in the PAB resulting from a hose failure are bounded in the PAB by the current large diameter service water header line break flood analysis.

The use of temporary hoses may increase the probability of a temporary loss of SFP cooling due to a supply hose rupturing.- llowever, a new hose can be run (already tested and staged) to replace a ruptured hose (s) well within the period of time it takes to exceed SFP maximum allowed temperature.

FRM35001 DOC

d ACP 1.2 2.42

,. Rev.1 MAJOR Form 3 10 CFR 50.59 (b)(2) Report Page 2 of 2 Safety Evaluation Number: SY EV 97 0001 Revision: 3 Document Number: TPC 97 221 to SUR 5.7 217 . Revision: 3 Dor *Jment

Title:

Intervice T.uine orSW cv 963 The possibility for an accident or malfunction of a differcat type than any evaluated previously in the safety analysis report has not been created. The basis for this statement is:

This proposed procedure change simply supplies a temporary source of SW cooling to the "A" CFPilx via firehoses. In the current defueled mode of operation there are no credible failures of these hoses which could create the possibility of anat.. .nt of a different type than previously evaluated.

1 The failure of the temporary SFP cooling firehose would cause a temporary loss of SFP cooling. Gooding in either the upper level PAB, RCA yard of Lower Level SFB and a loss of service water integrity in the North or South SW header.

Loss of SFP cooling would be handled by cut.cnt p actic:s (i.e., reestablish pipinghose integrity or provide makeup 4

water per AOP 3.2 59). Flooding of the PAB due to hose rupture is bounded by the Dooding analysis of record done on a j larger bore SW line break. Ruptu.e of a hose ',ould be tenninated by isolating the Adams Filter drain or filter inlet j isolation. Since there are no other credible faih res or new malfunctions created, the possibility of a malfunction of a different type than previously evaluated is not tretted.

I The margin of safety as defined in the basis for any technical specification has not been reduced. The basis for this statement is:

Operation with only the B heat exchanger in service has no change on the margin of safety as denned in the Technical Specincations since the normal operational alignment has the A heat exchanger isolated. The procedure change serves to enhance the ability to meet current SFP Technical Specl0 cations by providing emergency cooling after a total failure of the normal service water supply and'or rerum piping. The temporary hoses will provide sumcient water to maintain the spent fuel pool below its technical speci0 cation limit of 150'F with the current and future levels of decay heat.

i

2. Reason for the Change

The A heat exchanger has been isolated to allow testing of the service water check valve SW CV 963.

Preparer R. W. Katuna 'M Date 1/14'98 e

FRM3SE01 DOC

HADDAM NECK PLANT ,

S E C fl O N lli Jumoer-Lifted Lead and Bvoans (J LL-B) Chanaes (Page 1 of 1) l Safetv Evaluation J LL-B Number Number Ilile 95-41Rev. O SY EV 97-0021 Removal of SF-CV-846 96-63 None "A" Spent Fuel Pool Cooling i

= = = = = = = = _ _ = ==_ = =

I

ACP 1.2 2A2 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page1ofI Safety Evaluation l' umber; sv.Ev.97 0021 _ _ _ Revision: 0 Document Number: .flypaiss Jumpet 95-4 i Revision: 0 __,,

Documerit

Title:

• temoval of sr-cv.846 provide a brief dest -iption of the change and a summary of the Safety Evaluation in the format below.

l. Brief Description of Change and Safety Evaluation Summary:

Dypass Jumper P$ 41 removed check valve SF.CV 846 on 1/23/96 to eliminate a leaking weld betw een the check valve and the vent isolation valve SF V 845. This check velve is located in a vent line downstream of a normally closed isolation volve (SF V 846)directly off the $F inlet line to the "A" heat exchanger (E 101 A). De removal of this valve has no efTect on the operation of the spent fuel pool cooling system.

His change does not constitute an unreviewed safny questic.1 because.

There is tio increase in the probability Joccurrence of the consequences of an accident or malfunction of equipment impertant to safety previously evaluated e the safety analysis report. De basis for this statement is:

j De probability of occurrenet and conseq" nces of previously evaluated accidents is not efrected by this change.

De initiation cf a Fuck llandling Accident is unrelated to this procedure change.

Ilypassjumper 95 41, which removes check valve SF CV 846, will have no impact on the spcM fuel pool cooling system. The prob.:bility of occurrence of a previously evaluated malfunction of equipment (i.e. partial or 10 2 loss of SFp cooling)imponant to maintain the spent fui i pit bulk temperature below 150'F is unaffected by this change. De system lineup remains unchanged since this check valve was located downsmam of a normally closed isolation valve.

The removal of check valve FF CV 846 has no effect on the postulated failure of the spent fuel pool coming system.

The possibilhy for an accident or melfimetion of a different type than any evaluated previously in the afety analysis repon has not been creaud. The basis for this statement is:

Removal of check valve (SF CV 846) does not in any way create the possibility of a new unanalyzed accident, ne opeiation of the spent fuel pool coo!1ng system was not modified in anyway by the removal of this check valve.

, 2. Kenson for the Change:

It w as determined by engineering and operations to be more c1st effective to remcve this valve and eliminate the historical leaking weld then to do an ASME Section XI Code weld repair when clearly there is no need for this check valve. The leaking w cid was radiologically and aesthetically undesirable.

(

Prepr.ar Date -.__

SE970021 DOC

T'~fdN k(p-M (NcNtb E h h M Attachment 1 r 94-G HADDAM NECK PLANT Annual Report Summary of Changes Mark the Appropriate Choices Design Change t

Test _

setpeint Change Experiment Tech Requirements Manual Change _

l Procedure Chnnge l Jumper Bypac: 1.

1. Change Number E Revision Heinbert n

Title:

Jumne* Dvemar 96-61 to "A*

RFP coM 43 Preparer: RA resuca Date: n /s/96

{ 2. Description of Change:

This enange provides temporary cooling to the "A" SFP Hx using firuhoses from the North-SW header to the inlet of E-10-1A.

3. Reason for the Change:

The reason for this change is to replacement of a leaking valve in = the SFp supply header (SW-V-23 9) .

4. Safety Evaluations

a. This change was safe for the following reasons:

1.

The jumper will not adversely affect any design basis occidents-in the current mode of use (5 or 61

2. The jumper will-not result in any malfunctions which will adversely e.ffect sa#ety related equipment.

3. The jumper will provide cooling system.

sufficient flow to the SFP I

~

m .

m

- . - - . .__ _ _ __ ' "' A m Ak my.3 4 _

Jumper Bypsp 4Q fa$e 1 cT l-a

b. This change does not constitute AN UNREVIEWED SAFETY QUESTION

< becauses THERE IS NO INCREASE IN THE PROBABI!ITY OF OCCURRENCE O CONSEQUENCES OF AN ACCIDENTEVALUATED OR MALFUNCTION IN THE SAFETYOF EQUIPME IMPORTANT TO SAFETY PREVIOUSLY 4 ANALYSIS REPORT. The basis for this statement is:

The jumper only affects Service Water Spent Fuel cooling supply which can not affect any design basis accidents in the current modes of use.

OR MALFUNCTION OF A POSSIBILITY FOR AN ACCIDENT

- THE DIFFERENT TYPE THAN ANY EVALUATED ThePREVIOUSLY basis for this IN THE l 4 ANALYSIS REPORT HAS NOT BEEN CREATED.

dtAtement is:

This change can not create a new sccident since the loss of the SWS can not initiate a DBA. The potential malfunctions and associated with this change have been evaluated and determined to be bounded by existing procedures analysis.

i 4

IN THE BASIS FOR ANY THE MARGIN OF SAFETY AS DEFINED The basis for TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED.

this statement is:

The margin of safety iJ not affected by this change since the consequences or probability of occurrence of an accident Also this change does not effect the is not affected.

consequences or probability of occurrence of a previously safety and does not to evaluated malfunction important create the possibility of a new malfunction.

require a change to the Technical

c. Did this change Specifications: No

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1 l.

I HADDAM NECK PLANT SECTION IV 1

1 Tests 4

l (Page 1 of 1)

The following tests were performed under the provision of Title 10, Code of j Federal Regulations, Section 50.59 during 1997.

i Safety Evaluation j Procedure Number Number lille

ENG 1.7-182 SY EV 97-0090 Spent Fuel Building ',entilation Air

! Flow Test, Revision No.1 -

{ ST 11.7-168 None SWS Acute Treatnw. r MIC, PDCR 1553 Functional Test, , avision No. 2 i

4 s

t 4

4

'A 4

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ACP 1.2 2.42 Rev.1 MAJOR Form 3 10 CFR 50.59 (h)(2) Report Page I of I Safety Evaluation Number: SY EV 97 0090 Revision: 0 Document Number: ENO l.7162 . Revision: 1 Document

Title:

Soent Fuel Buildine Ventilation Air Flow Test Provide a brief descrip'hn of the change and a summary of the Safety Evaluation in the format below.

l. Brief Description of Change and Safety Evaluation Summary:

The Spent Fuel Building (SFB) and Primary Auxiliary Building (PAB) ventilation systems will be in service in order to collect data that will be used in calculations to support the spent fuel cooling analysis.

Per UFSAR section 9.4.2, the design flow of the exhaust fan exceeds supply capacity by 1000 cfm maintaining a small continuous innitration into the SFB to prevent re!:ase ofcontamination from the building. Based on initial test data, the SFB ventilation system is not operating as designed. A positive pressure was developed in the Spent Fuel Buildirt; following startup of the SFB supply fan.

This evaluation documented the controls in place to ensure adequate monitoring of any release from the SFB. The change was determined to be safe.

This change does not constitute an unreviewed safety question because:

1here 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 system will be operated as described in the UFSAR but will no: maintain the SFB at a negative pressure w hile the SFB supply faa is aperating. Operations insolving movement of fuel or crane operation with loads over the storage pool will not be performed during testing. The building exhaust air will be monitored by the stack radiation monitor.

Chemistry and ilP will take additional samples in the SFB to ensure adequate monitoring.

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 system will be aligned as described in the SAR. Any releases will be monitored as described.

The margin of safety as denned in the basis for any technical specification has not been reduced. The basis for this statement is:

Operations involving movement of fuel or crane operation with loads over the storage pool will nat be performed during testing. Chemistry and llP will take additional samples in the SFB to ensure adequate monitoring,

2. Reason for the Change:

The change allows data to b: collected on the performance of the ventilation systems.

Preparer D. Carnesi Date 9'2'97 50591 utttxx'

/ )%M 97-l ST M.7- 16S Attachment 1 Pase 6 ef t HADDAM NECK PLANT Annual Report i Summary of Changes Mark the Appropriate Choice:

Design Change Setpoint Change Te s t ,,,,x,,,, Tech Requirements Manual Change  ;

Experiment '

Procedure Change Jumper Bypass *

1. Change Number ST 11.7-168 l

Revision Number: A

Title:

SWS Acute treatment for MIO. PDER 1553 functional test Preparer: R.W. Manugri Date: 1/2/97 2, Description of Change:

This change injects new chemicals into the Service Water System

, (SWS) to combat Microbiological 1y Influenced corrosion (MIC) and aligns flow to several standby components which places them in an unusual alignment.

3. Reason for the Change:

=

-This- -change is being performed- to provide functional verification of PDCR 1553 and to continuously treat _the SWS for 45 days (called Acute Treatment Period).

4. Saf ety Evaluatiota

a. This change cas safe for the following reasons:

-1. Continous flow to the EDG's does_ not adversely affect engine lube oil or room ambient temperature.

2. .The hydraulic performance of the SWS will not be-f 3 adversely affected by placing standby components in service V since important system valves will be maintained or placed in their required post-accident positions.

ST11.7-[M Pay 2 4 2.

4, 3. The EDG fouling levels will be closely monitored and the "o flow control valves closed prior to adversely affecting the

, EDG's post-accident response r*, quire ~nts.

4. No Technical Specification regt aments will be challenged.

5. No EDG alarm functions or indica ions, other than the raw water admission valves,will be adversely affected or lost.

] .

b. This change does not constitute AN UNREVIEWED SAFETY QUESTION because:

I THERE IS NO INCREASE IN THE PROBABILITY OF OCCURRENCE OR THE ,

i 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 puts safety-related valves and components in their fail-safe position or condition for responding to a j DBA and the EDG will not be adversely impacted.

I THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A l DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY  ;

j ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this ,

statement is:

This change simply aligns flow to standby components, wich if fouled can be cleaned online. Tne test only. effects SWS cooling and . the loss of SW in not an accident initiating event.

THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:

See above.

c. Did this change require a change to the Technical Specifications: No i

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HADDAM NECK PLANT SECTION V Erneriments (Pace 1of1)

There were no experiments performed under the provisions of Title 10, Code of Federal Regulations, Section 50,59 during 1997,  !

. - - - - - . -= . - . . - . - - - _ __- _- -__ - .--- . _. - -- - -

HADDAM NECK PLANI SECTION V1 Final Safety Analysis Reoort Changes (Page 1 of 2)

SA[etv Evaluation FSARCR Number Number Iille 96-CY 24 CY-SE 97-001 Various UFSAR Section Revisions for the Service Water System 96 CY 24 CY-SE-97-002 Revise Service Water System Design Basis to Reflect Normal inlet Operating lemperatures 96-CY-24 CY-SE-97-007 Revise Section 9.2.1.2 to Address 97-CY-17 Freeze Protection for the Fire and Service Water Pumps and Service Water System Discharge Points 97 CY 1 S%EV 97-0004 Diesel Generator Floor Response Spectra Safety Evaluation 97-CY-2 SY EV-97-0066 Service Air System 97 CY-3 SY EV-97-0039 Fiber Optic (Synchronous Optical Network - SONET) Communication Equipment in CY PBX Room 97 CY-09 SY-EV 97-0007 Non-CYAPCO Structures on Site 97-CY-15 SY-EV 97-0035 UFSAR Fire Protection Organization Change 97 CY-17 SY-EV-97-0123 Mir silaneous UFSAR Changes 97-CY-24 Aff: : ting Description of Water Supply in FSAR Chapters 2 and 9 97-CY-17 SY-EV-97-0127 Redefinition of the CY Restricted Area 97-CY-27 m+mr - - ---

4 HADDAM NECK PLANT EECTION VI Final Safety Analysis Reoort Channs (Page 2 of 2)

Safety Evaluation FSARCR Number Number .Tlila l 97-CY-18 SY EV 97128 Appendix R Safe Shutdown Program 97-CY-21 97-CY-22 97-CY-23 97-CY-24 97-CY-19 SY EV 97-0080 Revision to UFSAR Chapter 4 to Reflect Defueled 97-CY-21 SY EV 97-0072 Dose to Control Room Operators (UFSAR Section 6.4) 97-CY-2'4 SY EV-97-0085 Control Air System 97-CY-27 SY-EV-97-0134 Intemal Dos' metry Program Enhancements 97-CY-36 SY-EV 97-0047 Configuration of Power Supply to Semi-Vital Panels 1 and 2 97-CY-36 SY-EV-97-0058 Battery Load Calculation Revision 97-CY-39 SY EV 97-0113 Incorporation of PDCRs 69 and 84 into

UFSAR 1

97-CY-44 SY EV-97-0081 Addition to CY UFSAR on Snow Roof Load 97-CY-45 SY-EV-97-0088 UFSAR Figure 2,1-4, CY Site Plan Update

- . . . . . - . - . . .. . - - - _ - - - - - = - - - - . - - . . . . . .

1

. CONNECTICUT YANKEE i 10CFR50.59 Summary , g ,,og This summary applies toi ga i *

• 4

PDCR Set Point Change Test Procedure Tech Requirements Manual Change I

Experiment Tech Spec Basis Change Only l Procedure Change FSAR Changes _X_

Jumper Bypass 1.

Title:

CY-SE-97-001. Revise Various UFSAR Sections for 1 the Service Water System to correct Known Text 4

Discrecancies and to make reauired revisions, j I

2. Description of Changes:

! The topics included within this safety Evaluation

! include raising the maximum SWS inlet temperature to 4

90'F, increasing the storage capacity of the Spent Fuel 4

Pool, adding discussion of various instrumentation connections points to SWS components, adding details on chemical addition systems, clarifying the automatic closure of valves SW-Mov-1 and 2, adding clarifications j regarding the as-built conditions of the plant, and

correcting identified typographical errors.

3. Reason for the Changes:

Safety Evaluation CY-SE-97-001 was developed to support

-various changes included as part of FSARCR 96-CY-24.

The Safety Evaluation supports UFSAR changes related to the Service Water System (SWS) and the Spent Fuel Pool '

(SFP).

The proposed changes are the result of the following.

}

4 4

. , , . - . . . ,.._--,._,..-__.-__,,m .. _ . , _ _ . _ . , ~ . _

OCFR50.59 Summary f(* ,5E+, $]= Dol Continued

1. Connecticut Yankee Service Water System, NRC Generic Letter 89-13, Item No. IV, Design Basis Summary Report, Revision 0, July 15, 1994.

2. NRC Letter, dated January 22, 1996, " Issuance of Amendment No. 188"

3. PDCE CY-89-061. " Addition of Instrumentation for SFP System"

4. PDCE CY-89-063, " Addition of Instrumentation Ports for RHR System"

5. PDCE CY-89-064, " Addition of Instrumentation for EJG System"

6. PDCE CY-89-065, " Addition of SW Instrumentation for SW-V-133/134"

7. PDCR 955, " Installation of Emergency Automatic Closure Circuitry for the Turbine Bldg. Service Water Header Isolation Valves, SW-MOV-1 and 2"

8. PDCR 1354, " CAR Fan Service Water Flow Instrumentation Rsplacement"

9. PDCR 1391, " Service Water Pump Flow Indicator Replacement"

10. PDCR 1447, "AFW Modifications, Electric Pump Addition"

11. PDCR 1553, " Service Water System Chemical Addition for MIC Control"

12. QAS Audit A25099, Recommendation R-06, "Need to Correct FSAR Typographical Errors"

13. WCAP-12196, " Service Water System Design Basis Temperature Increase to 95'F for the Connecticut Yanke,e, Haddam Neck Plant," Westinghouse Electric Corp., July 1989

. 3 10CFR5n.59 Summary ty 5E 9[-Coi -

Continued Page 3 g 4

4. Safety Evaluation

a. The changes are safe for the following reasons:

The changes will not contribute to any new or

, previously analyzed accident or malfunction of equipment and do not reduce the margin of safety of the oystems involved.

1

b. The changes do 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 DAFETY ANALYSIS REPORT.

. The basis for this statement is:

The only accident applicable to the changes being proposed is the fuel handling accident. The a

proposed changes do not increase the probabili*y that a fuel assembly is dropped since none of the changes involve a change to any system or component involved in the dropped fuel assembly scenario.

. None of the proposed changes have any effect on the malfunctions evaluated and therefore have no effect on the probability of their occurrence. The proposed changes applicable to equipment important to safety are the changes to the SFP maximum temperature and the proposed changes to the SFP ,

itself. None of these changes affect the operation or performance of equipment involved.

In addition changes being made to reflect plant <

conditions (i.e., as-built or clarify the system description) have no impact on any malfunctions because they do not involve any plant change.

, .-- .-- , n, -,.-~- . , --- , , r- .-g , , - - - - - -

I

, e i 10CFRSO.59 Summ ry cy. 5E- 97 00I Continued Page 4 of f 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:

Since CY is permanently defueled and all of the fuel has been moved to the SFP, the only credible accident is the fuel handling accident which has been previously evaluated. The proposed changes do not introduce any new accident initiators.

The proposed changes do not introduce any new failure modes or malfunctions that could lead to inadequate SFP cooling. None of the changes involve a system or component operating beyond its design basis.

THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:

It was concluded that the proposed changes have no impact on previously evaluated accidents and'do not introduce the potential for a new unanalyzed event.

The proposed changes do not result in the operation of any syetem or component beyond its design basis.

Therefore, it is also concluded they have no impact on the margin of safety.

c. These changes do not require a change to the Technical Specifications

/

cy-n-97 22 1

-- Attachment 1 py, t ,t s HADDAM NECK PLANT Annual Report Summary of Changes Mark the Appropriate Choice:

Design Change _ Setpoint Change Test Tech Requirements Manual Change Experiment FSAR Change Y Procedure Change Jumper Bypass

1. Change Number: Resvision Number

Title:

cY- sE - 9 *1 - c o 2 . Revise sewice water svstem Des!-m Basis te Reflect Neimal Tnlet Oceratina Temperatures

2. Description of Change:

Revise the Service Water (SW) system description in the UFSAR to reflect the normal inlet operating temperatures of the SW system and upda'te the necessary pipe stress calculations to reflect this temperature. The UFSAR wording is being revised to reflect a conservative minimum operating temperature of 28'F, which is the freezing point of sea water.

3. Reason for the Change:

In 02/96 (ACR 96-0128) and again in 01/97 (ACR 97-022), it was determined that the plant had been operating outside of the analyzed SW system supply temperature range. CY UFSAR Section 9.2.1.2 references 35'F as the minimum SW supply temperature.

The SW supply piping stress analysis also utilized a minimum supply temperature of 35'F. However, the average SW supply temperature hr.s been recorded below 35'F, as documented in the above referenced ACRs. The corrective action for this condition conoists of revising the SW system description in the UFSAR to reflect the normal inlet operating temperatures of the SW system and performing the necessary design basis calculation changes.

/

2

• y-S 97 001 Po9e 2 of 3

4. Safety Evaluation:

a. This change was safe for the following reasons:

The change will not contribute to an'y previously analyzed or new accident or malfunction of equipment and does not reduce the margin of safety of the systems involved. It is concluded that the proposed change is not an unreviewed safety question and 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 FREVIOUSLY EVALUAlaD IN THE SAFETY ANALYSIS REPORT. The basis for this statement is:

For a LOCA with and without an LNP, the only instance identified as a potential negative impact of lower supply temperature was to increase the maximum CAR fan heat removal rate. However, the consequences of the increased CAR fan heat removal rate have been analyzed and found not to adversely impact the performance of the ECCS.

There is no negative impact of lowar SW supply temperature on the consequences of a MSLB. Increasing the efficiency of the heat removal by lowering the SW temperature to the CAR fan units will decrease the containment peak pressure and temperature.

There is no negative impact of lower SW supply temperature on the consequences of a fuel handling accident. This accident would typically involve spent fuel assembly failure and increased radiological levels in the spent fuel building, which is not impacted by a lower SW temperature.

The SW system structural integrity is not degraded by operating at the lower temperature. An evaluation was made of the effect of the lower temperature on che structural integrity of the SW piping, pipe supports, valves and equipment. It was found that the material allowable values will not change appreciably a- a result of lowering the temperature 7'F f rom 35'F t, '?.

/

I C,Y,- ,.

SE-97 >001 7 THf POSSIBILITY FOR AN ACC1 DENT OR MALFUNCTIGU 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 possibility of an accident of a different type than previously evaluated as a result of this proposed change because the SW syntem is not an accident initiator for any UFSAR Chapter 15 accident.

There is no possibility of a malfunction of a different type than previously evaluated as a result of this proposed ,

change because the structural integrity and heat removal capability of the SW system components is not adversely

! affected. There are no other system malfunctions that could occur as a result of the lower ' temperatures.

THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for thic statemert is:

The margin of safety is not affected by the proposed change since the consequences or probability of. occurrence of an accident is not affected. Also, the change does not affect the consequences or probability of occurrence of a previously evaluated malfunction important to safety and

-does not create the possibility of a new malfunction.

c. This change did not require a change to the Technical

' Specifications.

M

QA-O CONNECTICUT YANKEE /*f9-jpg 10CFR50.59 Summtry This summary applies to: cy. 5E 007 Pac 3e t et 3 PDCR Set Point Change ,

Test Procedure Tech Requirements Kanual Change Experiment Tech Spec Basis Change Only Procedure Change FSAR Changes X Jumper Bypass 1.

Title:

CY-SE-97-007, Revise section 9.2.1.2 to addr_r g i

l fraere Drotection for the Fire and Service Water Dumos and Service Water System discharoe coints l 2. Description of Changes:

The propoced change provides clarification that SW provides freeze protection to the Fire and SW pumps and that their discharge is within the Screenwell Fouse.

The inclusion of the proposed change therefore details the location of all of the SWS discharge points.

3. Reason for the Changes:

The proposed changes is the result of ACR 97-0087, dated 2/13/97 which noted a discrepancy between the UFSAR and NOP 2.24-6, " Service Water System Normal Operation".

t 9

4

2 10CFR50.59 Summary tY 5E 007 Continued Ph c 2 of 3

4. Safety Rvaluation:

e

a. The changes are safe for the following reasons:

The changes will not contribute to any new or previously a.alyzed accident or malfunction of equipment and do not reduce the margin of safety of the systems involved.

b. The changes do 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 EV).i.UATED IN THE SAFETY A!!ALYSIS REPORT.

The basis for this statement is:

The only accident applicable to the change being proposed is the fuel handling accident. The proposed change do not increase the probability that a fuel assembly is dropped since it does not involve a change to any system or component involved in the dropped fuel assembly scenario.

The proposed change does not have any effect on the malfunctions evaluated and therefore has no effect on the probability of its occurrence. The proposed change has no detrimental affect on equipment import, ant to safety, nor does the chpnge affect the operation or performance of equipment invo2ved.

In addition the proposed change is being made to reflect plant conditions (i.e., as-built or clarify the system description) and has no impac -

any malfunctions because it does not involvt. j plant change.

m s

/ 3 30CFR50.59 Summary cy.5E 007 Continued Page 3 'f 3 e

THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE TRAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT HAS NOT BEEN CREATED.* The basis for this statement is:

-Since CY is permane.stly defueled and all of the fuel has been moved to the SFP, the only credible l accident is the fuel handling accident which has l been previously evaluated. The proposed change does not introduce any new accident initiators.

The proposed change does not introduce any new failure modes or malfunctions that could lead to inadequate SFP cooling. The gnanges does not involve a system or component operating beyond its design basie.

THE MARGIN OF SAFETY AS DEFINEL IN THE BASIS FOR ANY TECHNICAL SPECIFICATION HAS NOT BEEN REDUCED. The basis for this statement is:

It was concluded that the proposed change has no impact on previously evaluated accidents and does not introduce the potential for a new unanalyzed event. The proposed change does not result in the ,

operation of any system or component beyond its design basis. Therefore, it is also concluded it has no impact on the' margin of safety.

c. These changes do not require a change to the Technical Specificatione.

e

ACP 1.2 2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page1of1 Safety Evaluation Number: SY.CV 97 0004 Revision: 0 Document Number: FSARCR 97-CY-1 Revision 0 Document

Title:

Cannecticut Yankaa - Diesel cienerator Floor reenante Snectra Safety Evahindon Provide a brief descriptior, of tha change and a summary of the Safety Evaluation in the format below, l

l 1. Brief Description of Change and Lfety Evaluation Summary:

The floor response spectra for the Emergency Diesel Generator Building was revised to account for the effects of soil I

structure interaction. The building structure and the associated safety related equipment was reviewed utilizing the new rusponse spectra to asses the resultant impact on structural integrity under SSE loading.

The safety evaluation considered both mechanical and electrical safety related equipment and reviewed the previous conclusion reached under SEP Topic !!I 6 and the USI A 46 efforts.

The safety evaluation concluded that structural integrity and functionality of the building structure and the associated safety related meclenical and electrical equipment remains acceptable for the new seismic demand. Therefore the SEP and USI A-46 conclusion 1 remain unchanged, his change does not constitute a unreviewed safety question because:

i Dere 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. De basis for this statement is:

The analysis using the newly developed seismic spectra shows that the diesel building and associated equipment remain eismically qualified for the increased seismic demand.

The possibilie for an accident or malfunction of a different type than any evahtated previoudy in the safety anslysis report has not been created. The basis for this statement is:

There are no changes to plant components or systems. The existing diesel generator building and equipment has been found to be acceptable for the higher sei:mic demand. Therefore there are no new seismic failures that create the possibility of an accident or malfunction of equipment different than previously evaluated.

He margin of safety as defined in tne basis for any techni<:al specification has not been reduced. The basis for this statement is:

The Technical Specification does not define any seismic allowable limit, ne maximum stress of the co.uponents and piping systems are tr.4intained when applying the new FRS.

- 2. Reason for the Change:

It was unclear if the effects of soil suucture interaction wen adequately considered in the development of the previous '

building floor response spectra utilized for the SEP and USI-A 46 resolution efforts.

Preparer __

. _. Date _

SE970004. DOC

ACP 1,2-2.42 Rev. I MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page 1 of 2 Safety Evalu'ition Number: SY-EV.97-0066 Revision: 0 Document Number: FS ARCR 97-CY-2 Revision: WA Document

Title:

Service Air System Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

I. Brief Description of Change and Safety Evaluation Summary:

At the present time the FSAR describes the Service Air System cross tie to Control Air System as follows:

"The service air system sup; sies the control air system on extreme low p cssure in the latter through a tie provided with a regulating valve set at 70 psig gage and a low pressure alarm signal to the Control Room set at 75 psi gage."

He proposed change to the FSAR will read:

"The service air system supplies the control air system on extreme low pressure in the latter through a tie provided with a check valve and a low pressure alarm signal to the Control Rcom set at 75 psi gage."

He change to the system took place with the implementatior. of PDCR 595, dated 3/31/84. Ths regulating valve was removed from the system, the P&lD updated, but the FSAR was never altered to reflect the change. The safety to the plant ,

the public, or the environment is not efketed by this clerical change. The Service Air System is not a safety related s; stem and has no efrect on the operability of tiv plant under power or decommissioned.

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 descriptive change will not increase the probability or consequences of an accident. The Service Air System has no safety related function to the operation of the plent.

This is a descriptive change that will have no effect on the operability or increase the probability of occurrence or a malfunction of the equipment involved in the system. There are no equipmerc malfenctions associated with the Service Air System that are described or evaluated in the technical speciScations, the design basis documentation or the FSAR.

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 b:: sis for this statement is:

ne descriptive change to the system will not create the possibility of an accident or malfunction of different type. There are no evaluated accidents associated with the Service Air System that are described in the technical specifications, the design basis documentation or the FSAR. The Service Air System has no srfety related function to the operation orthe plant.

The margin of safety as defined in the basis for any technical specification has not been reduced. He basis for this statement is:

The plant safety margins are notjeopardized by the Service Air System. The Service Air System does not have a safety related function associated with the safe operation or safe shut down of the plant. The proposed change to the Service Air System is a descriptive change to the FSAR only. The syste.n change was accomplished over ten (10) years ago, but the FSAR was never updated. He documentation reviewed for this determination was as follows: Technical Specification 3/4.7 SE970066 DOC i

ACP 1.2-2,42 Rev 1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page 2 0f 2 Safety Evaluation Number: SY-EV 97-0066 Revision: 0 _

Document Number:JSARCR 97-CY-2 Revision: N/A Document

Title:

Service Air Svuem l

" Plant Systems",3/4.0 Bases " Limiting Conditions for Operation and Surveillance Requirements", FSAR Chapter 9.3

" Process Auxiliaries",9.3.1 " Compressed Air System", 9.3.1.2," Service Air System"

2. Reason for the Change:

To update the FSAR so it will accurately reflect the physical depiction of the Service Air System and how it works e

9 Preparer Date SE970066 DOC

ACP 1,2-2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page1of1 Safety Evaluation Number: SY-EV-97-0039 Revision: 0 Document Number: FSARCR 97-CY 3 Revision: 0 Document

Title:

Fiber Optic (Synchronous Optical NetworbSONET) Communication Eouipment in CY PBX P,oom Provide a brief description of the change and a summary of the Saf-ty Evaluation in the format below.

1. Brief Descr!ption of Change and Safety Evaluation Summary:

His change adds a fiber optic communication system and adds a description to the UFSAR. The Synchronous Optical Network (SONET) is a fiber optic communication system linking Berlin to Wethersfield to CY to Millstone and supplements the existing microwave system. The three postulated Design Base Accidents (DBAs) evaluated in the UFSAR for the plant in a decommissioning status are the Waste Evapor; tor Failure, Waste Gas incident and the Fuelllandling Accident. These postulated DBAs are not considered applicable to this change since this rhange only adds an additional independent off-site intra-company communications path to the site independent of the microwave system and other safety related systems and does not remove any existing off-site communication parameters. The systems associated with these postulated DBAs are not impacted by this change. This system is not an accident initiator or mitigator nor does this system mteract with systems that perform mitigation functions. The system does not perform any safety function or direct mitigst on function. This l change does not prevent any actions assumed in the accident analysis. The system is not a radioactive system and does not have any anect on any radiological sources. The fiber optic system is a non-safety, light transmitting system that does not

• interface with any class IE plant electrical systems. The equipment is located in the front room of the PBX room. This room and the equipment in this room are not safety related. The only credible failure modes associated with this change would be a complete failure of the SONET fioer optic communication system. The SONET Network element is deployed in a ring j

topology which is referred to as "self healing" because ung topologies can survive a single point of failure. In the unlikely

) event that this system should fail the existing microwave system would still be availaMe. This change does not affect any I

other safety related equipment nor degrade the performance of any system imponant to safety. There are no other systems or parameters affected by this change. The change had no adverse affect on the design function of any equipment or systems.

It is therefore concluded that a failure of this system will not have a safety impact on the plant or increase the risk to the public health or safety. This is a safe change and not an unreviewed safety question.

2. Reason for the Change:

This safety evaluation justifies a change to the plant with Maintenance Suppan Engineering Evaluation: MSEE 95-048 R/0, dated 2/26/96 and a change u UFSAR Section 9.5.2.2.2. Off-site Communications, for conformance to the as-built condition of the plant. The change is an enhancement of the existing microwave system and provides a high bandwidth transport communications system supporting information exchaage on the Northeast Utilities computer network.

Preparer Eric Urban Date 5/29/,92 97CY35E3 DOC li

• ACP 1.2-2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page 1 of 1 Safety Evaluation Number: SY EV-97-0007 Revision: 0 Document Number: FSARCR 97-CY 09 Revision: 0 Document

Title:

Non-CYAPCO Stmetures on Site Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Drief Description of Change and Safety Evaluation Summary:

The UFSAR stated "no structure shall be located on the site except structures owned by the Connecticut Yankee Power Company and used in conjunction with normal utility operations".

No plant equipment was affected by thi:: change. This change is considered safe since the ownership of on-site structures does not affect the risk to the health and safety of the public. CYAPCO retains control of all on-site activities regardless of the ownership of any on-site structures. Neither the probability nor the consequences of any analyzed malfunction was affected by this change. This change did not introduce a new or different malfunction than previously evaluated. Neither the probability nor the consequences of any previausly analyzed accident was affected by this change. This change did not introduce a new or diff erent a:cident than previously evaluated. The margin of safety is not affected by this change and did not affect any protective boundaries, nor did it affect any Technical Specification equipment or requirements.

2. Reason for the Change:

Since the use of non-CYAPCO owned structures is occasionally necessary, this change allows the use of leased structures on the Haddam Neck site.

1 Preparer Date sE97 007. DOC

FSARCR 97.CY-$

ACP 1.2-2,42 -

Rev,1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page 1 (J 2 Safety Evaluation Number: SY-EV-97-0035 Revision: 0 Document Number: UFSAR Sectinn 9.5.1.51 and UFSAR Finure 9.5-1 Revision:

Document

Title:

UFSAR Fire Protection OronninHan C%noe Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Brief Description of Change This proposed change will revise UFSAR Section 9.5.1.5.1, Fire Protection Organization, by removing the descripdon of the fire protection organization and organizational responsibilities from the UFSAR. It will also remove the Fire Protection Organizational Chart, Figure 9.5-1, from the UFSAR. All of this information, including the organizatiot chart, is shown in the Connecticut Yankee Nuclear Plant Fire Protection Program (NGP 2.14, WCM 2.5 6); which. under the broad dermition of the SAR, remains a part of the SAR. The actual organization and responsibilities are not affected by this removal of the information from the UFSAR.

The postulated design basis accident previously evaluated in the SAR that is considered applicsble to this preposed change is a fire in the plant. His is in accordance with the guidelines for performing a Safety Evaluation as documented in Generic Letter 8610, April 24,1986, P.ragraph F., Addition of Fire Protection Program into FSAR. This paragraph states, in pen that "the provisions of 10 CFR 50.59 would then apply directly for changes the licensee desires to make in the fire protection program that would not adversely affect the ability to achieve and maintain safe shutdown. In this centext, the determination of the involvement of an unreviewed safety question defined in $0.59 (a)(2) would be based on the ' accident....pr:viously cvaluated' being the postulated fire in the fire hazard analysis for the fire area affected by the change". For that reason, for fire protection program changes, the fire is the accident.

2. Reason for the Change:

ne reason for this change is to (1) reduce the potential for confusion and/or conflict caused by having identical information in two documents; and (2) allow CY to revise the organization and division of responsibilities for the Fire Protection Program without having to revise the UFSAR as the plant proceeds through the different phases of decommissioning.. Any changes to the program will still require a 50.59 Safety Evaluation, as required in 10CFR50.48 (f).

3. Safety Evaluation Sununary This change does not constitute an unreviewed safety question because:

Here is no increase in the probability of occurrence or the consequences of an accident or malfunction of equipomewnt important to safety previously evaluated in the safety analysis report. He basis for this statement is:

The probability of a fire occurring is affected by either an increase in ignition sources or the addition of a combustible that is easier to ignite. This proposed change does not affect the ignition sources or the type of combustibles in any plant area.

- He probability of occurrence of a malfunction of equipment can be affected by the probability of occurrence of a fire. The probability of occurrence of a fire is affected by an increase in ignition sources or a change in the type of combustible material. His proposed change does not involve an increase in ignition sources or a change in combustible material; nor does it involve any new equipment.

The consequences of a fire can be affected by the type of combustibles burning, the location of *.he fire, or the length of time the fire burns. This proposed change does not affect any of these things.

50590RO3. DOC

O TsARcit 97-CY-5 ACP 1.2-2.42 Rev.1 MAJOR Form 3 - 10 CFR $0.59 (b)(2) Rep,rt Pr- 2 of 2 Safety Evaluation Number: Sv.Ev.97.0035 Revision: 0 Document Nember: UFSAR Section 9.5.1.5.1 and UFS AR Figure 9.5.1 Revision:

Document

Title:

UFSAR Fire Protection Oronnivatinn r'hnnue His proposed change does not involve an increase in ignition sources, a change in type or quantity of combustible material, Or any new equipment. He dose due to a malfunction will not increase.

The pos0ility for an accident or malfunction of a different type than any previously evaluated in the safety analysis report lits not been created. The basis for this statement is:

he type of fire that can occur in an area is dependent upon the type of combustible in the area. His proposed change does not address the type af combustible in any plant area.

No new equipment is involved in this proposed change, here is no change in the type or amount of combustibles, or the number ofignition sources. Therefore, the fire in any plant area is not changed by this proposed change. If the fire is not changed, its effect on equipment is not changed, and there is no malfunction of a different type than that previously evaluated The margin of safety a* defined in the basis for any technical specification has not been reduced. He basis for this statement is:

i This proposed change does not change the organization, nor does .: change the responsibilities of the organizatic n ,

given in Technical Specification 6.0, Administrative Controls. l

/ /

preparer Edward A. Sawyer M . MI Dete  !, 7#'

yyye ,

00590RG3. DOC

FS.'RCR 97.CY 17 Aap 97 CY 24 ACP 1,2 2,42 Rev,1 MAJOR Form 3 - 10 CFR 50,59 (b)(2) Report Page1ofI l

Safety Evaluation Number: SY.EV 97 0123 Revision: ,_.0 Document Number: None Revision: 11/A Document

Title:

Miscellaneous UFSAR Channen Affectino Descrintion of Water Sunniv in FSAR Chanters 2 and 9 Provide a brief description of the change and a summary of the Safety Evaluation in the format below,

l. Brief Description of Change and Safety Evaluation Summary:

nis safety evaluation addresses changes being made to the UFSAR in Chapter 2," Site Characteristics," Section 2.4.1, "llydrologic Description," and Section 2.4.13 " Groundwater," as well as Chapter 9," Auxiliary Systems," Section 9.2.5,

" Potable Water System." The changes are being made to correct the descriptions cf public water supplies in the area of the plant and chemical treatment of potable water, ne changes affect the description of the Well Water and Domestic Water System in the UFSAR but have no affect on these systems in a physical sense.

Specifically, the proposed revisions to the UFSAR change the description of public water supplies in the vicinity of the lladdam Neck Plant and removes a reference to hypochlorination equipment in the Domestic Water System that does not exist.

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 that there will be no physical work or alterations made to plant systems as a result of this change. Also, the change affects the potable water system which has no safety related function, and does not interface with any 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 that the proposed revision to the UFSAR does not modify any equipment or the methods of system operation.

The margh of safety as defined in the basis for any technical specification has not been reduced. The basis for this statement is that there will be no physical work or alterations made to plant systems as a result of this change.

2. Reason for the Change:

These chatiges are being made to provide a more accurate description of water supplies in the vicinity of the plant and to correct a reference to hypochlorination equipment that does not exist.

Preparer Scott Penlev , M Date FRM3. DOC

l ACP 1.2 2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page l of 2 Safety Evaluation Number: S%EV 97-0127 Revision: 00 Document Number: 97-CV-17 and 97 CY-27 _ Revision: .,00 i)ocument

Title:

Redefinit;on of the Cr Restricted Arca ,_

Provide a brief description of the change and a summary of the Safety Evaluation in the format below, f Brief Descriptmn of Change and Safety Evaluation Summary:

l Title 10 Part 20 (10CFR20) of the code of federal regulations defines three distinct areas used for controlling radiation and radioactive material at NRC licensed sites. ihe Restricted Area is defined as an area " access to which is limbed by the licensee for the nurpose of protecting individuals against undue risks from aposure to radiation and radioactrve I

' materials." He Controlkd Area is dermed as an area "outside of a Restricted Area but inside the site boundary, access to which can be limited by the licensee for any reason." The Unrestricted.1rea is defined ar, an area where l " access to which neither limited nor controlled by the licensee".

l At Connecticut Yarsee, the Restricted Area has been dermed as " generally corresponding to the RCA" in the UFSAR I

Chapter 12 and Health Pitysics prscedures until mid 1996. At that time, changes were made to Health Physics procedures and tcchnical documents to rederme the 10CFR20 Restricted Area as being generally the plant Proteced a Area. A UFSAR Change Request was also submitted to modify the defmition in the UFSAR in March 1997, a QA surrillance identified a discrepancy hi the definition of the Restricted Area in the UFSAR an Health Physics piocedures because the UFSAR change request submitted was not prwessed. An ACR was subsequent!y submitted describing this concern, in addition, Chapter 2 of the UFSAR contains a definition of RestrictedArea that is not consistent with the Chapter 12 definition.

This safety evaluation supports the redsfmition of the CY Restricted Area from " generally the RCA" to " generally the Protected Area".

This change does not constitute an unreviewed safety question because:

There is no increase in the probability of the occurrence or tne consequences of an accident or malfunction of equipment impartant to safety previously evduated in the safety analysis report. The be;is for this statement is:

De change in defimtion of the Restricted Area is an administrati e program wh;ch does not affect any initiating events associated with the consequences or malfunction of equipment of an accident. Herefore, there is no increase in the probability of the occurrences or the consequences of an accident or malfunction previously evaluated.

He possibility for an accident or malfunction of a different type than any evaluated previously is the safety analysis report has not been created. The bwis for this statement is:

The redefinition of the CY Restricted Area does not involve plant equipment or impact operability.

Radioactive material used and stored in all plant locations is controlled by the CY Health Physics program. No physical changes that could create the possibility of an accident of a diffetnt type than previously evaluated will be made.

• 0$9RA3. DOC m J

ACP 1.2-2.42 -

Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page 2 0f 2 Safety Evaluation Number: SY.rv.97 0127 Revision: 00 Documeld Number: _97.CY.17 and 97.CY 27 Revision: 00 __

Document

Title:

Redefinition of the CF Restricted Arca The margin of safety as defined in the b:: sis for any technical specificulon has not been reduced. The basis for this statement is:

Technical Specification 6.0 wes reviewed. No changes to the Technical Specif: cations are required as a result of this change. The definition of Restrictsd Arca is not a bas!s for system parameters. Therefore the twgin of safety is not reduced.

2. Reason for the Change:

l This :hange is an administrative UFSAR change that etTects how radioactive material is controlled on site.

Maintaining the definition of the Restricted Area as the RCA is not consistent with we control of radioactive material located in systems are area outside of the RCA. The Pre'ected Area boundary is currently a boundary that is used by llealth Physics to restrict access to the plant for radiation protection purposes. Defining the Restricted Area as the Protected Area in the UFSAR will provide consistent documentation of a current practice that must be maintained for the security o(radioactive material on site.

Preparer Date t ..

I 5059RA3.DGC A  % _

ACP 1.2-2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page1ofI rafety Evaluation Number: SY-EV 97-128 Revision: 0 Docurrent Number:-. FSA%It. Nos. 97 cY- 68, 21,21,2b 14 gevisiom Document

Title:

Annendix R Safe Shtdown Procram Provide 2 bnef description of the change and a summary of the Safety Evaluation in the format below,

l. Brief Description of Change and Safety Evaluation Summary:

The Appendix R Safe Shutdown Program is canceled.

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:

In the defueled state, the cancellation of the Appendix R Safe Shutdown Program has no effect on the revised ,

accidents analyzed, since the fuel has been permanently removed from the vessel and safe shutdown is no longer applicable.

Them are no malfunctions of equipment important to safety that are considered applicable to this proposed change.

The Appendtx R Program was to ensure safe shut down of the plant when it was operating. Amndix R no longer applies.

The possibility of 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:

In the defueled state, the cancellation of the Appendix R Safe Shutdown Program can not create the possibility of an accident of a different type than previously evaluated in the SAR since the fuel has been permaaently remove from the reactor vessel and safe shutdown is no longer applicable.

The margin of ssfety as defined in the basis for any technical specification has not been reduced. The basis for this statement is:

The Technical Specifications do not address Appendi: R safe shutdown requirements.

2. Reason for the Change:

This program is being cancel, d because it no longer applies following submittal of the 10CFR50.82 certifications of permanent cessation of power operations and permanent removal of fuel from the reactor vessel.

Preparer _

Date -

EV97118. DOC l

__Y

m FsAR cg 47 cY-19

/ ACP 1.2 2.42 p Rev.1 MAJOR

+,v Form 3 - 10 CFR 50.59 (b)(2) Report Page1of 1 Safety Evaluation Number: SY EV 97-@0 Revision: 0 Document Number : VES AR Chaeter 4 Rev.: June 1997 Document

Title:

Revision to Chaeter 4 to refleet Defueled Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Brief Deteription of Ch.nge and Safety Evaluation Summary:

The proposed changes are being made to revise Chapter 4, Reactor, to reflect the permanent removal of the fue in preparation for tbc decommissioning of the Haddam Neck Plant.

% yys vuse%

These changes do not constitute a unreviewed safety question because there is no increase in the probability of occurrence or consequences of any accident previously evaluated in the SAR. This is because all reactor related accidents heve been removed from the HNP licensing and design basis and the fuel is no longer in the reactor vessel.

He probability of occurrence or consequences of a malfuaction of equipment previously evaluated in the SAR is not increased because these changes do not make changes to the fuel, only its location; or the equipment required to assure the l cladding's integrity.

Dere is no possibility of creating an accident or malfunction of a different type than any previously evaluated in the S AR G because all the fuel has been permanently removed from the reactor vessel. All applicable accidents are still within the V design basis and site boundsry dose limits.

The margin ofiafety as defined in the Bases for any Technical Specification is not reduced because the margin of safety for the physical protective boundary for the fuel cladding, while in the spent fuel pool, is not affected by this change.

2. Reason for the Change:

Since HNP has pe:Tnanently ceased power operations, moved all fuel to the spent fuel pool and does not or will not have any new fuel on site, Chapter 4 does not reflect the current condition of the p! ant. These changes will result it a UFSAR chapter that represents the plant as it prepares fer decommissioning.

F\

U Preparer ! h -- Jb /4 bh Date 8 ' 2

, /

SE9700:0. DOC N

ACP 1.2 2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page1 ofl' Safety Evaluation Number: SY.EV.97 0072 Revision: 0-Document Number: FS ARCR 97.cY.21 llevision: 0 _

- Document Tttle: CYAPco Haddam Neck UaA**d Final Saferv Analvain Pm Provies a brief description of the change and a summary of the Safety Evaluation in the format below.

I. Brief Description of Change and Safety Evaluation Summary:

This proposal is to revise the dose to comrol room operators in UFSAR Section 6.4. De revised cortrol room dows in Section_6.4 are based on the defueled condition of the reactor and decommissioning activities. These accide added to Section 15.5 of the UFSAR under DCR # CY.97007. The two accidents considered are:

1) Resin Container Accident

2) Fuel Handling Accident The doses from direct shine, immersion from outside the control room, and sky shine we no longer considered signi As a result, the calculated dose to the control room operators will change based on the new accident parameters. The references to the "J" size air boules in Section 6 have been reinoved.

This change does not constitute an unreviewed safety question because there is no increase in the probability of occur or consequences of an accident previously evaluated in the SAR. This is because the control room doses are based on the accidents stready in UFSAR Section 15.5, and the dose to the control room operators is less than previously evaluated.

The probability of occurrence or consequences of a malfbaction of equipment previously evaluated in the SAR is not increased because isolation of the control room is accomplished in the same manner as before. Mitigating actions from inside the control room are not required. There is no possibility of creating an accident or malfhnetion of a differen than any previously evaluated because the new DBAs upon which control room habitability is based cannot lead to a different initiating event, different failure mode, or differen: response than the types previously evaluated in the SAR.

The margin of safety is not reduced because the dose to control room operators is less than previously calculated.

} 2. Reuon for the Change:

The previous design basis accident (DBA) used in the design evaluation of the control rocm habitabilir/ systems was a LOCA. However, the roscoor related accidents in UFSAR Section 15 were removed by DCR # CY.97003. -Be new DB As were added to UFSAR Section 15.5 by DCR 8 CY.97007. The effectiveness of the control room habitability systems was re.

- evaluated using the preameters of the new accident scenarios.

The methou alogy used to calculate the doses has changed since the significant pathways have changed with the new DBAs.

Since the use of the "J" size air bottles is not credited in the control room dose calculation, and the doses to operators are less i

than the limits of GDC 19, the references to the "J' size air bottles in the cos. trol room have been removed.

PreparerMcH6M 3 MibDki b ,

4 f h1

/ - _ Dats 9 2d , ,

7 ,_,

50!*FRM3 DOC

~ -

f . ACP 1.2-2.42 Rev.1 MAJOR Form 3 10 CFR 50.59 (b)(2) Report Page1of1 Safety Evaluation Number: SY EV 97 0085 Revision: 0 Document Number: FSARCR No 97 C%24 Revision: N/A Document

Title:

Control Air Svstem UFS AR Chance Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Brief Description of Change and Safety Evaluation Summary:

This Safety Evaluation addresses modifications to the UFSAR text for the Control Air System to reflect changes made via PDCR 164, which was implemented in 1974. The text for UFSAR Section 9.3.1.2 and Table 7.4 5 still makes reference in some places to only two control air compressors even though a third compressor was added via the subject PDCR. The third air compressor was added due to additional control air demand in the plant. It should be noted that the Control Air System

- Description in UFSAR Section 9.3.L2 describes three control air compressors but is not consistent in the wording throughout the text, in addition, UFSAR T' *le 7.4 5 contains a typographical error for the power supply for the control air compressors.

The Probability of occurrence or consequences of a previously evaluated malfunction in the Control Air System or EDGs is unchanged and there is no possibility for a malfunction of a different type. The control air supply to systems containing radioactivity and to the service water / spent fuel pool heat exchanger interface is unchanged by this change. The probability of occurrence of a fuel handling accident or a radwaste system failure or a loss of spent fuel pool cooling is unaffected and there is no possibility for an accident of a different type.

This change does not foreclose (preclude) release of the site for possible unrestricted use. It does not result in a significant environmentalimpact not previously reviewed. It does not remove reasonable assurance that adequate funds will be available for decommissioning.

This change is safe and does not involve an unreviewed safety quest;on. The change does not affect the probability or consequences of previously evaluated accidents or malfunctions of equipment important to safety. There is no reduction in the margin of plant safety. In addition, this change does riot involve an unresolved decommissioning question.

2. Reason for the Chrnge:

The proposed change to the Control Air System was ac:omplished over twenty (20) years ago, but the UFSAR was never completely updated to reflect the change.

Preparer NePbfl 3. IO

. N_/_A_ 1 Date h22 9 8 '

s.- -

97CY241. DOC

ACP 1.2-2,42 Rev.1 MAJOR l Form 3 - 10 CFR 50.59 (b)(2) Report Page1ofI Safety Evaluation Number

• SY-EV 97-01M Revision: 0 Document Number: 8'3ARCft 97*CT* 27 Revision:

Document

Title:

Intamal Dmimetry Pronram Enhancemenn Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Brief Description of Change and Safety Evaluation Summary:

The intemal dose action level of 4 DAC-hrs is being replaced with new criteria to take additional / follow up actions based on general area grap samples, lapel air samples, and random checks associated with air sample and respiratory l_

cfTectiveness evaluations, ne expected affect of the change is an overall improvement to the internal program because the controls will include all airborne radioactive hazards, ne change does not result in any USQs or UDQs.

His change doas not constitute an unreviewed safety question because:

Dere is no increase in the probability of the occurrence or the consequences of an accident or malfunction of .

equipment important to safety previously evaluated in the safety analysis report, ne basis for this statement is:

'fhe Internal Dosimetry program of the Health Physics i.epartment is an administrative program which does not affect any initiating events associated with the consequences or malfunction of equipment of an accident.

Herefore, there is no increase in the probaMlity of the occurences er the consequences of an accident or malfunction previously evaluated.

The possibility for an accident or malfunction of a different type than any evaluated previously is the safety analysis

. report has not been created. De basis for this statement is:

The intemal dosimetry program of the Health Physics Department does not affect any mitiating events, therefore there is no associated impact due to changes in action levels, or program structure that can result in an acc%t or malfunction of a different type than previously evaluated.

The margin of safety as defined in the basis for any technical ,pecification has not been reduced. De basis for this statement is:

Technical Specification 6.0 was reviewed. No changes to the Technical Specifications are required as a result of this change. The internal dosimetry program of the Health Physics Depa.tment is not a basis for system parameters.

Herefore the margin of safety is not reduced.

2. Reason for the Change:

he rease 1 for the change is to address the accounting for h ed-to-measure nuclides (i.e. transumnics) which previously was not performed.

Preparer ____ Date int _ne do e

ACP 1.2-2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page l of 1 Safety Evaluation Number: SY EV-97-0047 Revisien: 0 Document Number: FSARCR 97-CY 36 / UIR 052-195 Revision: 0 Document

Title:

CONFIGURATION OF POWER SUPPLY TO SEMIVITAL PANELS 1 AND 2 Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Brief Description of Change and Safety Evaluation Summary:

This change altered the configuration of the 125 Vac Semivital Panels I and 2 performed via PDCR I195. However, PDCR 1195 did not incorporate a 10 CFR 50.59 (b)(2) safety evaluation, and did not revise the FSAR to reflect the as-built condition of the plant. This change modified the in-series connection of SV Panels I and 2 into in-parallel connection. As a result, this change assured the independency of SV Panel 2 in that it precluded the loss of both Pa,els 1 and 2 if Panel 1 main breaker would trip inadvertently, or as a result of a fault in SV Panel 1. Prior to this charge the feeder for SV Puel 2 had been connected downstream of the 100 A main breaker for SV Panel 1. Therefore, this change reduced substantially the totalload of Panel 1. As part of this %e, the # 6 AWG cable connection from Panel 2 to Panel I was replaced with a # 2 AWG cable with adequate ampacNfor supplying Panel 2 loads. All connections were made in the MCB using the existing raceways. The weight aference between the # 6 AWG cable and # 2 was negligible and therefore no structural modifications were required.

The three postulated Design Base Accidents (DBAs) evaluated in FSAR for the plant in the decommissioning status are the Waste Evaporator Failure, Waste Gas Incident, and Fut. riandling Accident. These postulated DBAs are not affected by this change since the change did not alter the intended function and operation of equipment powered from SV Panels I and 2. This change can not initiate or contribute to the possibility of occurrence of an accident not previously evaluated. This change will not prevent any actions assumed in the accident analysis. The total or partial loss of SV panels I and 2 will have the same effects on the plant safety as before the change implementation. Existing Emergency Operation Procedure outlines the operator corrective actions to be taken in response to a loss of SV Panels.

The chanh did not affect any oth:r safety related equipment, and did not degrade any system important to safety. It is therefore concluded that this change did not have an adverse impact on the plant safety or increase the risk to the public heahh, This was and still is a safe change, and did not create an unreviewed safety question.

The possibility for an accident or malfunction of a different type than any evamated previously in the safety analysis report has not been created because this modification does not change the intended function or operation of the equipment powered from either Panel 1 or 2.

The margin of safety as defined in the basis for any technical specification has not been reduced because this change enhanced the independence of SV Panel 2 from Panel 1, decreased Panel 1 loading, and did not alter any other safety related systems.

2. Reason for the Change:

This safety evaluationjustifies the change to the configuration of the 125 Vac Semivital Panels 1 and 2 performed via PDCR 1195, and the associated change to FSAR Section 5.3.1.1.3,120 V System Description, to reflect as-built condition of the plant. The change was an enhancement of the existing SV Panels I and 2 configuration, as',uring the independency of SV Panel 2.

Prept.rer: S. Rosenberc Da.c 195FRM3 DOC

ACP 1.2-2.42 Rev. I MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page 1 of 2 Safety Evaluation Number: SY-EV-97-005R Revision: 0 Document Number: FSARCR No. 97-CY-36 / UIR 014 Revision: 0 .

Document

Title:

BATTERY LOAD CALCULATION REVISION Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Brief Description of Change and Safety Evaluation Summary:

Calculation ABBATRRY 1409 EY, Rev. 0," Class 1E Batteries I A and 1B Size Verification for LNP/LOCA and SBO Conditions" determined the duty cycle of station batteries. However, based on the Unresolved item Report (UIR)

No. 014, the calculation did not take into account all loads required to support the EDGs starting. Derefore, a new Revision I of the calculation was issued in August 1996 when the plant was in operation. Subsequent to the issuing of Revision 1, however, UFSAR Table 8.3 2, Battery Duty Cycle, had not been updated to reflect the corrections made by Revision 1. In October 1996 the plant was permanently shutdown, and at the present time the duty cycles of Batteries I A and IB are substantially lower, ne calculation determined the batteries duty cycles under LNP/LOCA conditions for two hours, and SBO conditions for four hours, and assumed a loss of AC power coincident with a SIS signal. Furthermore, the calculation has been performed based on the IEEE 485 1983 methodology. De calculation results indicate that under LNP/LOCA Batteries 1 A and IB have spare capacities of 20.5 %, and 9.13 %, respectively, of the batteries manufacturer's ratings. Under SBO conditions Batteries 1 A and i B have spare capacities of 24.25 % and 9.13 %, respectively, of the manufacturer's ratings. Therefore, this change does not increase the probability of an accident previously evaluated in the SAR for both tie plant normal and decommissioning modes of operation.

De postolated Design Base Ae:idents (DBAs) evaluated in UFSAR for the plant in normal operation are the Loss of coolant accident and Loss ofuTsite power. The postulated Design Bases Accidents (DBAs) evaluated in UFSAR for the plant in the decommissioning status are the Waste Evaporator Failure, Waste Gas incident, and Fuel Handling Accident.

The above DBAs are not affected by this change since it did not alter the intended function and operation of equipmed powered from Batteries I A and IB.

This change can not initiate or contribute to the possibility of occurrence of an accident not previously evaluated. This change will not prevent any actions assumed in the accident analysis. The total or partial loss of Batteries 1 A or IB will have the same effects on the plant safety as before the change implementation.

The change did not affect any other safety related equipment, and did not degrade any system important to safety. It is therefore concluded that this change did not have an adverse impact on the plant safety or increase the risk to the public health. This was and still is a safe change, and did not create an unreviewed sefety question.

There is no increase in the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the safety analysis report because the batteries have adequate capacity to supply the load profile at the required voltage for both the normal and decommissioning modes of operation for the entire cycles. In the very unlikely event that one battery is lost, the other redundant battery would still be available to support the operation of the safety systems.

The possibility for an accident or malfunction of a different ty; a than any evaluated previously in the safety analys:s report has not been created because: (1) the calculation results indicate the batteries have adequate spare capacity to supply the load profile at the required voltage for both the normal and decommissioning modes of operation for the entire cycles, and (2) no physical modifications were made to the de system 14FRM3 DOC

ACP 1.2-2.42 Rev.1 M AJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page 2 of 2 Safety Evaluation Number: SY EV-97 0058 Revision: 0 Document Number: FSARCR No. 97-CY 36 / UIR 014 Revision: 0 Document

Title:

BATTERY LOAD CALCULATION REVISION The margin of safey as defined in the basis for any tehnical specification has not been reduced because this change does not affect any TS requirements.

~ 2. Reason for the Change:

This sa:cj evaluation supports the change described in Calculation ABBATRRY 1409-EY, Rev.1," Class IE Batteries l A and IB Size Verification for LNP/LOCA and SBO Conditions," and justifies a change to the UFSAR. Table 8.3 2, Battery Duty Cycle, to reficct the as-built condition of the plant.

Preparer S . Rosenberg Date ..

1 *?RM3. DOC

ACP 1.2-2,42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page 1 of 2 Safety Evaluation Number: SY-EV-97-0113 Revision: 0 Document Number: FSARCR 47 CY-39 Revision: 0 Document

Title:

Incornoration of PDCRs 69 and 84 into the UFSAR: resolution of ACR 97-0297 Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Brief Description of Change and Safety Evaluation Summary:

The change involves the incorporation ofinformation from previously completed PDCRs (69 and 84) into the UFSAR The interconnections installed in the PDCRs introduce the potential for increasing or decreasing the fuel pool level. This is addressed by providing high and low level alarms for the fuel pool well within the limits specified in the Technical Specifications, if the operators do not respond to the alarm (s), the change in level is slow enough that the level change would be observed by operators taking rounds long before any impact on the Spent Fuel Cooling System er N occur (would require a decrease in the pool level approximately 6 feet below the low level alarm). If the pool overflowed (nigh level alarm ignored) water would be directed to a sump that would transfer the water to the Aerated Drain Tank. This tank has a high level alarm, in either case no USQ was identified His 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:

Piping and valves installed by the PDCRs satisfy the code requirements of the originally installed ion exchange piping.

Herefore, the probability of a rupture of the ion exchange (SFC) piping is not increased.

l Since the new equipment is not electric powered there can be no impact from a loss of offsite power De latest accident analysis assumes no decontamination of the released gases by the fuel pool water. Since the des antamination factor for the water 6 the spent fuel pool used in the current fuel handling accident is not affected by water temperature (assumes a DF of 1.0), a decrease in pool level or an increase in pool temperature can have no impact on the activity of fission products released from the spent fuel during/following the Fuel Handling Accident The impact of mlsuse of the installed modifications (valve lineup errors that divert water to or from the fuel pool)is a high or low fuel pool level. A decrease in pool level in the order of seven feet would be required before a loss of SFC could occur. Since operators take fuel pool level readings each eight how shift, any significant fuci pool level change would require an operator error in the valve lineup followed by a fain.a of the operator to respond to alarm response procedures and a failure of the operator making shiftly rounds to fail to see the change in pool. A consideration of two operator errors is beyond the design brsis.

Wit 1 pool level maintained at least 13 feet above the spent fuel the maximum temperature that the fuel pool water can reach during any credible failure is approximately 170* F. Fuel cladding temperatures at this water temperature are significantly below temperatures that can cause mechanical failure and much lower than during normal full power operation.

Ollitb*

N __

_,-m,,--,..- -- .-,-----.--- '

I ACP 1.2-2.42 Rev.1 MAJOR Form 3 10 CFR 50.59 (b)(2) Report Page 2 of 2 Safety Evaluation Number: SY EV-97-0113 Revision: 0 Document Number: FSARCR 97-CY 39 Revision: 0 Document

Title:

Incorrioration of PDCRs 69 and 84 into the UFS AR: resolution of ACR 97-0297 The possibility for an accider.t 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:

Equipment inste!!ed by the referenced PDCRs consisted of stainless steel manual valves and piping. These all have failure modes consistent with comparable equipment currently installed in the fuel building cnd currently evaluated in the SAR. That is, no different equipment was installed by the referenced PDCRs.

l Loss of pool level and loss of cooling (excessive temperature in the pool) could be created by misuse of the new connections, but this has been evaluated in the SAR and bounds the che.nges initiated by the referenced PDCRs l

He margin of safety as defined in the basis for any technical specification has not been reduced. He basis for this statement is:

Although not specifically defined in the basis for any Technical Specification, the margin of safety potentially affected by this change is the maximum analyzed concrete differential temperature and a time-dependent margin above the maximum pool temperature specified in the TS (150' F) and the concrete degradation point or temperature at which concrete exceeds Code allowables of the pool structure.

Considering the extended time to reach 150* F (approximately 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> from a normal temperature of 110' F), the lower heat reject capability required to reduce pool temperature below 150' F, and the contingency actions available to restore cooling , the credible failures associated with the installed interconnections on the SFC system will not result in more time that the pool water (and thus the concrete) remains above 150' F.

2. Reason for the Change:

This change will resolve the issue identified in the ACR and ensure the FSAR reflects the r.s-built system.

Preparer G A. Johnson

> Date 18 September 1997

(  !

Oll3. DOC I

1 1

_A

ACP 1.2 2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page t of 1 Safety Evaluation Number: SY-EV-97-0081 Revision: 0 Document Number: FSARCR No. 97-CY-44 Revision: 0 Document tTi le: FSARCR No. 97-CY-44 Addition to CY UFSAR on Snow Roof Load Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Brief Description of Change and Safety Evaluation Summary:

The recorrmended changes to the UFSAR reflect additional documented evaluations performed to show the design adequacy of the Haddam Neck roofs to withstand the extreme loads currently described in the UFSAR and to reduce further confusion.

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 i

of equipment important to safety previously evaluated in the safety analysis report. The basis for this statement is that the UFSAR change only provides additional detail with regard to the design loads for the roofs of structures at CY. The design of the roofs are acceptable and remain unchanged, therefore, the probability of and/or consequences of an accident or malfunction of equipment important to safety remains unchanged.

The possibility of an accident or malfunction of a different type than any evaluated previously in the safety analysis report (SAR) has not been created. The basis for this statement is that the roofs remain acceptable for the required loads, and there will be no adverse effects which could create an accident or malfunction of a different type than previously evaluated in the SAR.

The margin of safety as defined in the basis for any technical specification has not been reduced. The basis for this statement is that the design of the roofs remain acceptable for the required loads and there are no pertinent Technical Specification Sections applicable to this UFSAR change.

2. Reason for the Change:

The UIR 274 assignment was initiated in order to evaluate snow loading impact and recommend UFSAR revisions to reflect the current information. The present UFSAR Sections 2.3.1.2.6 and 3.8 appear to be inconsistent with each other regarding snow loadiags for Haddam Neck buildings. The descriptions in these sections are also inconsistent with other docketed correspondence (CYAPCo and USNRC), and past descriptions in the UFSAR and it's predecessor the FSDA. The recommended changes to the UFSAR reflect additional docum.ated evaluations performed to show the design adequacy of the Haddam Neck roofs to withstand the extreme loads currently described in the UFSAR and to reduce further confusion.

Preparer 8- -

Date 8/8/97 c

ACP242sN. doc

ACP 1.2-2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (h)(2) Report Page 1 of i Safety Evalaation Number: SY E\ 97-0088 Revision: 0 Document Number: FSARCR No 97 CY-45 Revision: 0 Document

Title:

UFSAR Fie. 21-4 "CY Site Plan Update Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Brief Description of Change and Safety Evaluation Summary:

The change is to update the UFSAR, Fig 2,1-4 " Connecticut Yankee Site Plan". The location and the orien6ation of principal plant structures within the site area are shown on this figure. This activity is being done in support to the ongoing UFSAR Update. He primary plant structures that will added in the updated CY Site l Plan are as follow:

Emergency Operations Facility.

Condensate Storage Tank lon Exchange Structure Spent Resin Storage i Spent Resin Facility Waste Disposal Bldg.

Containment Equipment Access Terry Turbine Building 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 activity does not physically add structures in the plant. It is mainly a documentation update to identify principal plant structures that are currently not shown in the CY Site Plan. This activity does not impact the spent fuel pool ofits support systems. Nor does it affect any system or equipment in the plant.

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 rtatement is: the proposed change is an update to the CY site plan, Figure 2.1-4 of UFSAR. It does not make physical change to the plant.

The margin of safety as dermed in the basis for any technical specification has not been reduced. The basis for this statement is: the proposed change is an update to the CY Site Plan for existing structures, Figure 2.1-4 of UFSAR. It does not make a physical change to the plant. There are no sections in the Technical Specification that are affected by this change.

2. Reason for the Change:

The Update of the UFSAR, Fig 2.1-4 " Connecticut Yankee Site Plan" is necessary to show the principal plant structures that are currently not shown or identified in the CY Site Plan. The updated CY Site Plan will also show the plant's current protected area boundary.

I Preparer Charlie L Nara JMH C <>. & Date January 20.1998 5059FRht3-1

HADDAM NECK PLANT SECTION_W1 Technical Reaulraments Manual Changes (Page 1 of 1)

Safety Evaluation 12*dCR Number Number Iltle C 97-1 SY-EV 97-0067 Deletion of Containment Requirements and Tech Spec Clarifications C 97-4 SY-EV 97-0087 Appendix R Shutdown Related Con ,,onents 4

v

ACP 1.2 2.D.

Rev. I MAJOR Form 3 - 10 CFR 50.59 (b)(2'> Heport Page 1 of 2 Safety Evahation Number:,,,,,,D'.EV 97 0067 Revision: 0 Document Number: TRMCR No. C 971 Revision: 0 Document

Title:

Deletion Of Containment Reauirements And L:h Soee Clarl0 cations Provide a brief descri; lon of the change and a summary of the Safety Evaluation in the format below.

1.11rief Description of Change and Safety Evaluation Summary:

The Technical Requirements Manual Change Request (TRMCR) deletes several CY Tech Spec Clarincations (C-TSCs) and

) deletes a Technical Requirement for the containment.

l He deletions discussed in the TRMCR affect how surveillances are impleniented. The affected surveillances are for systems that are no longer needed for facility safety in its permanantly defueled condition. His change has no effect on the plant l physical condguration.

2. Reason far the Change:

On December 5,1996 CYAPCO submitteu certifications pursuant to 10CFR50.82(af t)(1) and 10CFR50.82(a)(!)(ii) stating that the IINP has permanently ceased operations and that the fuel has been permanently removed from the reactor. As a l

result of submitting these certific*.tions,10CFR50.54to) states that 10CFR50, Appendix J is no longer applicable. Also, PTSCR C 19.a6 (proposed Defueled Tech Spec submittal) was submitted to the NRC on May 30,1997.

In addition, several active C-TSCs defer those Tech Specs that are not applicable to a permanently defueled facility. He Tech Specs referencM by these C TSCs will remain deferr i until the referenced Tech Specs are deleted in a future License Amendment, ne deletion of Technical Requirement 3/4.6.1," Primary Containment" is based on 10CFRC3.54(o).

3 Safety Esaluation Summary:

nis change does nct constitute an careviewed safety question because:

nere 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 aaalysh report. He basis for this statement is:

rhe deletions discussed in the TRMCR affect how surveillances are implemented. The affected surveillances are for systems that are no longer needed for facility safety in its permanently defueled condition. Therefore, the probability of occurrence of a previously evaluated malfunctiori of equipment irr.portant to safety failure is unaffected by these deletions.

De deletions discussed in the TRMCR ufTect how surveillances art implementeJ. The affected surveillances are for systems that are a longer needed for facility safety in its permanen'ly defueled condition. The consequences of a previously evaluated malfunction of equipment important to safety failure is unaffected by these deletiens.

herefore, the ratological dose limits speci0ed in 10CFR100 are not challenged.

The deletion of Technical Requirement 3/4.6.1," Primary Containment" is band on 10CFR50.54(o).

97015I3A DOC

l - i 4

4 ACP 1.2 2.42

Rev.1 MAJOR ,

): Form 3 10 CFR 50.59 (b)(2) Report Page 2 of 2 i

j Safety Evaluation Number: s%EW9') 0067 Revision: 0 +

j Document Number: TRMCR No. C 971 Revision:,,, O l

! Document

Title:

- Deletion Of Conulnment Reautrements And Tech Spec Clarificat'ons ,

)'

i ne possibility for an accident or malfunction of a different type than any previously evaloated in the safety analysis report .

has not been created, ne basis for this statement is

'

l i ne deletions discussed in the TRMCR affect how surveillances are implemented. De affected surveillances are l for systems that are no longer needed for facility safety in its penanently defueled condition. Therefore, the  ;

j possibility of an accident of a different type is not created by these deletions.

b ne deletions discussed in the TRMCR affect how surveillances are implemented. The affected surveillances are

for systems that are no longer needed for facility safety in its permanently defueled condition. Derefore, the possibility of a malfunction of a different type is not created by these deletions.

The deletion of Technical Requirement 3/4.6.1,

• Primary Containment" is besed on 10CFR50.$4(o).

l l

i ne margin of safoty as defined in the basis for any technical specification has not been reduced. De basis for this statement is:

1 De deletions discussed in the TRMCR affect how surveillances are implemented. The affected surveillances are i for systems that are no longer needed for facility safety in its permanently defueled condition. These deletions are j for surveillances that are not applicable to the cunent plant configuration. Therefore, these deletions do not create d

any reduction in the margin of safety as defined in the basis for any Technical Specification.

He deletion of Technical Requirement 3/4.6.1. " Primary Containment" is based on 10CFR50.54(o).

l l

i i

Preparer O. O. Hofer Date 1/1498 i

l 4

i p

970lSDA. DOC i

l

- --~- ~------ - -

C 97 ACP 1.2 2.42 Rey,1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page 1 of 2 Safety Evaluation Number: SY.sv.97.0087 Revision: 0 J Document Number: Technien1 Requirements Manum 114 F~ition: 0 Document

Title:

_ APPENDIX R S1IUTDOWN REl ATED COMPONENTS ,

l Provide a brief description of the change and a summary of the Safety Evaluation in the format behw. l

1. Drief Description of Change and Safety Evaluation Summary:

This proposed ci..mge del:tes the present Section !!.2, APPENDIX R Sl!UTDOWN RELATED COMPONEN1S of the Fire Protection Technical Requirements. He ventilation for the "A" and *D" Switchgear Rooms is addressed in a new TRM Section developed to address those plant systems which are still needed to support the Spent Fuel Pool.

At the present time, the Spent Fuel Pool lluilding and Systems still uses electric power supplied through the "A" and the "B" Switchgear Rooms. The ventilation equipment for these rooms in Ested in the TRM Section to be deleted. Therefore,'he l

need for ventilation in the switchgear rooms must be addressed in a new sectior

2. Reason for the Change:

The reason for this change is that, with Appendix h no longer applying to the plant, the change will allow us to (1) remove unnecessary compensatory measures and surveillance requirements from plant equipment that is no longer needed for Appendix R safe shutdown, (2) remove unnecessary compensatory measures and the requirement to perform an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> discharge test on emergency lighting that was once necessary for safe shutdown, and (3) semove the requirement to keep equipment that was necessary for use in manual actions required to bring the plant to a safe shutdown condition. In the case of the emergency lights, they will still remain in service, and will be tested and maintained in accordance with the plant procedures used to test and maintain all non Appendix R plant er .gency lights. The manual equipment will be removed from the plant.

3. Safety Evalu'ition Sununary f

his change does not constitute an unreviewed safety question because:

here is no increase in the probability of occurrence or the consequences of an accident or malfunction of equipment impe: tant to safety previously evaluated in the safety analysis report. The basis for this statement is:

ne probability of a nre occurring is afTected by either an increase in ignition tources or the addition of a combustible that is easier to ignite. His proposed change does not change any fue protection equipment or procedures, and does not change fue brigade actions. It addresses only the removal of compensatory measures and surveillance requirements from non fire protection equipment that was necessary to bring the plant from operation to a safe shutoc,wn condition in case of Gre. It does not affect the ignition sources or the type of combustibles in any plant area.

De malfunction of equipment due to fue could be affected by the duration of the nre, which could cause either more or less damage, ne duration of a fire can be affected by the type of combustible, the addition of combustible, lack of detection, lack of suppression, or la,k of nre brigade action. This proposed change does not involve any change to any of these.

ne consequences of a fue can be affected by the type of combustibles burning, the location of the Ere, or the length of time the nre burns before activation of detection and suppiession systems. This proposed change does not affect APPRTRM3. DOC,10/15/97 l

c.974 ACP 1.2 2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (h)(2) Report Page 2 of 2 Safety Evaluation Number: SY SV.97 00R7 Revision: 0 Document Number: Technlent Reautremente Mantin1112 Revision: 0 Document

Title:

APPENDIX R SilUTDOWN Rr1 ATEf),CQMPONENTS any of these things. It does not change any fire protection equipment or procedures, and does not change Are brigade actions.

His proposed change does not involve an increase in ignition sources, a change in type or quantity of combustible material, of any new equip'ncnt. De dose due to a malfunction will not increase.

De possibility for an accident or malfunction of a different type than any previously evaluated in the safety ani.ysis report has not been created. ne basis for this statement is:

He type of nre that can occur in an area is dependent upon the type of combustible in the area. His proposed change does not address the type of combustible in any plant area.

He malfunction of equipment due to fire could be affected by the duration of the fire. A longer fire could cause more damage to nsore equipment. The duration of a fire can be affected by the type of combustible, the addition of comb"stible, lack of detection, lack of suppression, or lack of fire brigade action. His proposed change does not affect any of then items.

The margin of safety as defined in the basis for any technical specification has not been reduced. The basis for this statement is:

The following Technical Specifications were reviewed in making the determination that the margin of safety as defined in the basis for any Technical Specification :

TS Section 6, Administrative Controls nere are no margins of safety discussed in the fire protection technical specifications.

M N Preparer Edward A. Sawyer '( Date - /

./

l [{ f '

APPRTRM3.DUC,10/15/97 vmem s -

HADDAM NECK PLANT ,

SECTION Vill Technical Soecification Bases Chano_ es (Page 1 of 1)

Csfetv Evaluation PTSCR Number Number lille C 5 97 SY EV 97-0125 Bases Change 3/4.9.15 Spent Fuel Pool Cooling System m

ACP 1.2 2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (h)(2) Report Page 1 of 2 Safety Evaluation Number: _SY.Ev.97 0125 Revision: O Document Number: .EISCR C.$.9 and FSARCR 9 734 Revision:

Document

Title:

PTSCR for Section 3/4.9.15."SFP Cooling Bases" and UFSAR Section 9131 Provide a brief description of the change and a summary of the Safety Evaluation in the format below,

l. Brief Description of Change and Safety Evaluation Summary: i This evaluation addresses a revision to the Bases section o' Technical Specification 3/4.9.15 " Spent Fuel Pool cooling System". This change will reduce the maximum required heat removal capability of the SFP cooling system due to the decommissioning of CY. The previous heat removal capability of 22.4 million BTUhr was based on the emergency core ollload at the end of the previously licensed life (2007). The revised bases utilites the lloitec Decay llent and lleatup rate Analysis number ofless than 3 million BTU /hr (as of 10/1/97). This is an administrative chutge only. Therefore the ch age is considered safe.

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 of the bases section is due to the decommissioning of CY, dramatic drop in the SFP pool heat load and the elimination of any future emergency core ofTloads or spent fuel assembly additions. The probability of occurrence of ,

previously evaluated fuel handling accident, seismic event or lots of offsite power is not effected by this change which simply changes the current Spent Fuel Pool Cooling system Technical Specification Bases maximum heat removal requirement from 22.4 million BTU /hr to less than 3 million BTUthr. The consequences of the pieviously evaluated accidents are not affected by the reduction in the design bases heat load since the heat load is not a parameter used in determining the consequences of the current design basis accidents.

The redaction in heat load in the SFP does not directly influence the probability of any previously evaluated malfunctions. The reduction in the SFP heat load will have a positive influence on the previously evaluated malfunctions since more time will be available to restore any of the failed equipment important in maintaining the SFP bulk temperature less than 150'F, thus providing greatly increased margins. Since significantly more margin exists, the consequences of any previously evaluated malfunctions is greatly reduced.

The possibility for an accident or malfunction of a difTerent type than any evaluate previously in the safety analysis report has not been created. The basis for this statement is:

The change to tim ba:es section of the Spent Fuel Pool cooling system Tech Spec does :o, adversely affect ar'y plant equipment, systems or analyses. The equipment in the spent fuel pool cooling system will not be modified .n anrway or opeinted in an unusual manner as a result of this procedure change.

1he 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 safety is improved by the proposed change since the consequences or probability of occurrence of an accident or malfunction are decreased. The bases change to 3/4.9.15 reflects the signifbantly diminished Spent Fuel Pool nose 125. Doc

ACP 1.2 2.42 Rev.1 MAJOR Form 3 10 CFR 50.59 (b)(2) Report Page 2 of 2 Safety Evaluation Number: S%rv.97 0123 Revision: 0 Document Number: PTSCR C.S.97 and TSARCR 97 24 Revision:

Documen'

Title:

PTSCR for Sectinn 3/4.9.15. "SFP Cooline naten" and UFSAR Sectinn 9.111 i

heat load as a result of the decommissioning of Connecticut Yankee. The current calculated time to boiling after a total loss of SFP cooling is greater than 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />, from a starting temperature of iS0'F, excluding the effects of evaporative cooling. Since the change does not affect any other plant systems (bcluding SFP or SWS flow rates), n) other margins of '

safety are adversely affected.

3. Ikuon for the Change:

The current bases requirement of 22.4 million B1U/hr is out dated and requires the SFP heat exchangers to t>e maintained in an unreasonably clean condition. The calculated heat load in the SFP as of October I,1997 is less than 3 million BTU /hr.

Preparer M Date / 98 V '

nossinnoc

HADDAM NECK PLANT SEQ. TION IX General (Page 1 of 1)

Safetv Evaluation i Number Iltle l SY-EV 97-0011 CYAPCO Dry Active Waste (DAW) On Site Storage in Ses/ Land Containers i

SY EV-97-0117 Use of Portable Radiation Monitor on the Spent Fuel Bridge During Fuel Shuffling in the Spent Fuel Pool i SY EV 97-0069 Refueling Operations Crane Travel- Spent Fuel Storage Be'iding - Tech Spec. C;arification C-TSC-099 SY-EV 97-0101 ACR 97-0694, Discovery of Radiological Contamination in Closed Cooling Water System SY EV 97-0112 ACR 97-0729, Radiological Contamination in House Heating System SY-EV 97 0120 Electrical Equipment Qualification Program SY EV 97-0130 Continued Use of the Yard Storm Drainage System as a Contaminated System SY-EV 97-0131 Continued Use of the Component Cooling Water Systern as a Contaminated System b

. SYo EM 00ll Attachment 1 HADDAM NECK PLANT ,

Annual Report Summary of Changes Mark the Appropriate Choicus Design Change Setpoint Change Test Change _ X Tech Requirements Manua)

Experiment Procedure Ctyange Jumper Bypass

1. Change Number:

Revision Number:

Title:

_ Dry Active (DAW) Waste On-Site Storage in_ Sea / Land Containers

2. Description of Change:

Section 11.4.1 (5) of the CYAPCo UFSAR indicates the Radwaste Reduction Facility (RRF) in the Solid Waste Management System is used to " Hold" (store) DAW. In addition to stonge of DAW in the RRF, a supplementary practice at CYAPCo is to utilize on-site Sea / Land Containers to collect and temporarily store DAW prior to shipment to off-site processors.

3. Reason for the Change:

As a result of a recent NRC audit at the Seabrook Nuclear Station, it has been requested that the practice of storing DAW in on-site Sea / Land Containers at CYAPCo be evaluated in acccrdance with 10 CFR 50.59. This document evaluates the safety significance of storing low-level contaminated materials in sea / land containers in the CYAPCo protected area.

4. Safety Evaluation:

a. This change was safe for the following reasons:

There are no credible " malfunctions" of radiological significance. A radiological technical evaluation was performed for an accidental release, within the protected area, of 1% of the stored radioactive material via the airborne pathway and 100% released via the liquid pathway from a single Sea / Land Container as a result of a fire.

Title:

Dry Active (daw) waste on-site storage in sea / Land containers page l of 3

ST. EV 00 t f The accidental release of the material contents of the Sea / Land container is an accident bounded by the previously evaluated Radioactive Release from Gaseous Radwaste failure licensing basis accident listed in NGP 3.12, Figure A.3.

The evaluated condition will not increase the probability of en accident er malfunction affect accident mitigation or increase the consequences o,f an accident for the arguments given above. The changes are therefore safe and do not raise an unteviewed safety question.

b. This changs 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 ANALYSIS TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY REPORT.

i i

The basis for this statement is:

There are no credible " malfunctions" of radiological significance.

Sea / Land Containers are self contained storage units with no " equipment important to safety" to malfunction. The dispersion of the contaminated contents of the on-site Sea / Land Container by fire is classified as an

" accident" not a " malfunction".

THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFLTY ANALYSIS REPORT HAS NOT BEEN CREATED. The basis for this statement is:

The accidental release of the material contents of the Sea / Land container is an accident bounded by the previously evaluated Radioactive Release from Gaseous Radwaste failure licensing basis accident listed in NGP 3.12 Figure A.3.

Sea / Land Containers are self contained storage unite with no

" equipment important to safety" to 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 site radiological limits were evaluated and the accident analysis consequences due to a fire are bounded by the previously evaluated Radioactive Release from a-gaseous

Title:

,_ Dry Active (DAW)

Waste on-Site Storage in Sea / Land Containers page 2 of 3 eu i

'. SY- EV- 97-00f I radwaste failure licensing basis accident listed in NGP 3.12 (ref. 3) figure A.3.

c. Did this change require a change to the Technical Specifications: No

Title:

Dry Active (daw) waste on-site storage in sea / Land containers page 3of3 l

l

4

' ACP 1.3o3.43 Rev. I MAJOR 4' Foran 3 - 10 CFR 50.59 (b)(2) Report Psge1ofI Safety Evaluation Number: SY . EV . 97 0117 .- Revision: 0 Document Number:.lfA Revision:

Dowment

Title:

U= ' ' Fla=1 Enfatv Aaalvals *==i Tkd n-- r 38. Portable 'd-'lan Manitar On Snent Fuel Brid==

provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Brief Descnption of Change and Safety Evaluation Summary:

A portable radiation monitor is being und on the spent fuel bridge in place of the fixed R34 radiation monitor, which is inoperable, during fuel shuffling of fuel assemblies in the spent fuel pool.

a The change does not constitute an unreviewed safety question because there is no increase in the probability of occurrence et consequences of an accident previously evaluated in the SAR. This is because the spent fuel bridge radiation .nonitor is not credited for any mitigation actions for a fuel handling accident.

De probability of occurrence of a malfunction of equipment previously evaluated in the SAR is not increased because the portable radiation monho. will provide adequate range and sensitivity to detect abnormal exposure rates. There is no possibility of creating an sccident or malfunction of a different tj* pe than any previously evaluated because the change does not make any accident or malfunction previously considered outside the design basis now within the cesign basis.-

The margin of safety is not reduced because the offsite and onsite doses are not increased.

2. Reason for the Change:

The proposed activity is to allow fuel shuffling in the spent fuel pool to be undertaken with the R34 radiation monitoring thannel inoperable. -

Preparer Stenhen 1 Milioti

, ,b -

Date 10/897 _

1 ,_

n l

U

ACP 1.2 2.42 Rev. I MAJOR Form 3 - 10 CFR 50.59 (h)(2) Report Page1of1 Safety Evaluation Number: __ SWEV-97 0069 Revision: 0 Document Number:.C:TSC-099 Revision: WA Document

Title:

Refuelinn Operationt Crane Travel . snent ruel storace Hulk line l

Provide a brief description of the change and a summary of the Safety Evaluation in the format below.

I, Dricf Description of Change and Safety Eva'i ..on Summary:

The TSC clarines the existing limit on loads over spent fuel.

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. This basis for this statement is:

The change reduces the probability of an accident by enforcing the limit of one fuel bundle or less than the 2300 pour 3 load used in the assumptions. This limits the censequences of an accident by maintaining operation within the ar%mption of the existing analysis. The proposed TSC does not affect the calculated consequences of a fuel handling accident. This accident involves the rupture of one fuel assembly (221 day decay), and therefore the tource term is not affected by the proposed TSC.

The fuel assemblies are being handled with the same tooling which has been designed for the loads analyzed. The loads indicated in the TSC are such that binding of the fuel assembly will not cause damage to the bundle structure in excess of the analyzed accident.

Equinment reliability is not affected. The consequences of c dropped load on the spent fuel storage rack have been analyzed for w eights as high as 2300 lb. v ithout damage to the active region stored fuel. The analyzed fuel handling accident, dropping a spent fuel assembly directly on top of another spent fuel assembly, assumes the tupture of all pins h one fuel assembly,221 days after pow er operation, as an unnitered gr >und level release. The 1SC does not allow movement of heavy loads over spent fuel, nor allow lifting more than one spent fuel assembly, therefore the consequences remain within the beund of analyzed 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:

By maintaining the limits analyzed for power operation, accidents more lhiting than the ones previously analyzed will n.t be induced.

This ESC does not add equipment nor complexity to any equipment or systrms.

The margin of safety as denned in the basis for any technhal speci0 cation has not been redaced. The basis for this ,tatement is:

The TSC serves to enha ce the ability to meet current SFP Technical Specifications by reemphasizing the technical specincation load limit. The proposed TSC supports the analyzed basis of the technical speci0 cation by limiting the amount of spent fuel that can be lifted to one assembly and maintaining the loads lified over spent fuel to less than the limiting impact load which would damage active regions c'the f. pent fuel.

2. Reason for the Change:

To ensure the Technical $pecincation load limit will not be exceeded during fuel handling.

Preparer Date

$l970069 IXX'

ACP 1.2 2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page 1 of 2 ,

Safety Evaluation Number: s%Ev.97 0101 Revision: O Document Number: AcR 97-0694 Revision:

Doct ment

Title:

Discoverv of Radiologient contamination in closed cooling Water system Provide a brief description of the change and a ,ummary of the Safety Evaluation in the fonnat below.

1. Baef Description of Change and Safety Evaluation Summary: '

i Radiological Contamination has been found in the Closed Cooling Water System. His is supposed te be a l

his change addresses continued operation of Closed Cooling Water with slight amounts of radioactive contamina water. NRC Bulletin 8010 requires that a safety Evaluation be performed whenever an assumed clean system is fou contam radiological contamination. The Safety Evaluation has concluded that is event is not a USQ. "he basis for this dete mination is that the dose to the general public would be extremely small and well below any regulatoiy limit content of the Clased Cooling Water System u ere instantaneously discharged to the environment in an undiluted form.

His change does not constitute an unreviewed safety question because:

Here is no increase in the probability of occurrence or the consequences of an accident or malfunction of equ important to safety previously evaluated in the safety analysis report. De basis for this statement is:

C '.C does not interface with fuel handling activities or radioactive waste systems in any way, it therefore has no effect on th( probability or consequences of any accidents evaluateiin the SAR.

11 e small amount of radioactivity present in CLC water has no effect on the system's ability to function, nor w ould it be es xcted to cause failure of sny CLC components or piping. Continued use of the system will not increase %t occurrence of a malfunction of CLC equipment. Analysis shows that the CLC water in an undiluted state is below Maximum Permissible Concentration (MPC) limits for water released from the plant and is only a small fraction of the perm radioactive discharge from the plant. A dose analysis to the public shows that the dose to the general public from this would be bounded by the design basis accident and is a very small fraction of the limits established in 10 CFR 20 40 C 190 and 10 CFR 50 Appendix 1.

He possibility of an accident or malfunction of a different type than any previously evaluated in the safety ana not created: ne basis for this statement is:

ne presence of small amounts of radioactivity in the CLC System will not result in an equipment failure which could ca an accident of a dilTerent type than previously evaluated in SAR. He CLC System does not interface with any s associated with accidents evaluated in the SAR. ne possibility of a pipe rupture in the CLC System has always exi the presence of a small amount of radioactivity does not change that possibility. He consequences of this break has be addressed in the dose analysis.

He is:

margin of safety as defined in the basis of any technical specification has not been reduced, ne basis for this statem ne Basis for Technical Specification 3/4.11.1.2 states that the purpose of the Specification is to implement 10 CFR 5 Appendix 1. As stated above, the dose from an instantaneous, t.ndiluted release of all of the contents of the C wo sid result in only a small fraction of Appendix I limits.

5059FRNt3. DOC b

. i

c... _ _.__ _ _ _ _ _ . . _ _ _ _ _ - . _ _ _ _ _ _ . _ -

1

. l ACP 1.2 2.42 i

Rev.1 MAJOR Form 3 - 10 CFR $0.59 (b)(2) Report i Page 2 of 2 2

j Safety Evaluation Numkr: sv.Ev.97-0101 Revision: o i Document Number: Ack 97-0694 Revision:

. Document

Title:

Dinoverv of Radiala ical cane minnelu in clawa caniina water sven,m 1

- 2. Reason for the Change

Rt Jiological Cont mination has been found in the Closed Cooling Water System. NRC Bu!!ctin 8010 requires that a Safe .

Evaluation be performed whenever an assumed clean system is found to contain radiological contamination.

8 s

i 1-1

.i

?

- Preparer

{ Date i

i e

5 1

d 5059FIU43 DOC

, . . . . . - . _ . - - . . - . - - ~ . _ - . - . . - - - . . - - . - - - . . . . -

ACP 1.2-2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page 1 of 2 Safety Evaluatir,n Number: SY.EV.97 0112 Revision: 0 m.

Document Nurnber: ACR 97 020 Revision:

Document

Title:

__ Radiologiral contmination in closeJ Cooline Water System Provide a brirf description of the change and a summary of the Safety Evaluetion in the format below.

1. Brief Description of Change and Safety Evaluation Summary:

Radiological Contamination has been found in the llouse Heating System. This system was designed to be a non.

contaminated, system. His change addresses continued operation of House llenting Systetn with slight amounts of radioactive contamination in the system. NRC Bulletin 8010 requires that a Safety Evaluation be performed whenever a system w hich w as designed to be a clean system is found to contain radiological contamination. The Safety Evaluation h concluded that is event is not a USQ. ne basis for this determination is that the dose to the general public would be within regulatory limit if the entire content of the House lleating System were instantaneously discharged to the environment.

l This change does not constitute an unreviewed safety question occause:

There is no increase in the probability of occurrence or the consequences of an accident or malfunttlon of equipment important to safety previously evaluated in the safety analysis reputt. The basis for this statement is:

The llenting Steam System dom not interface with fuel handling activities other than providmg heat to the Spent Fuel Building. De presence of small amounts of radioactivity in the system cannot cause a fuel handling accident. Steam is also supplied to sorn radioactive waste processing equipment. The presence of small amounts of rm.lioactivity in the steam system will not cause these pieces of equipment to fall it therefore has no effect on the probability of any accidents (valuated pm.

in the SAR. The llenting Steam System is not used to mitigate fuel handling or radioactive waste accidents The nce of small amounts of radionetivity in the system therefore have no effect on the consequences of any accidents evaluated in the SAR.

The small amount of radioactivity present in llcating Steam System has no effect on tue system's ability to function, nor would it be expected to cause failure of any lleating Steam components or pi%ng. Continued use of the system will not increase the probability of occurrence of a malfunction of Heating Steam equipment Dose analysis

. shows that if th e activity in the lleating Steam System is released as described above, the release would be below Maximum Permissible Concentration (MPC) limits for water emuent from the plant and is only a fraction of the permissible radioactive disch from the plant. A dose analysis to the shows that the dose to the general public from this event would be 3 lE(-) 2 m whole body dose and 6.52E(-) 2 mrlyr, maximum organ dose (refer to memo HP 97 335), This dose is bounded by the design basis accident and is within the limits established in 10 CFR 20,40 CFR 190 and 10 CFR 50 Appendix 1.

We possi'oility of an accident or malfunction of a dif ferent type than any previous'y evaluated in the safety analysis not created: The basis for this statement it:

The presence of small amour.ts of radioactivity in the lleating Steam System will not cause an equipment failure which coulo cause an accident of a different type than previously evaluated in SAR. The possibility of a pip: ruptures in the Heating Steam System has always existed and the presence of a small amount of radioactivity does not change that possibi change in consequen:es of this break has been addressed in the cose analysis 5059rRht1 DOC

, l

ACP 1.2 2.42 Rev.1 MAJOR Foria 3 - 10 CFR 50.$9 (b)(2) Report Page 2 of 2 Safety Evaluation Number: EY EV.97 0112 _ Revision: 0 Document Number: ACR 97-0729 Revision:

Document

Title:

n.Atata.ic.t cani ntanelan in clated c<,ntine water svitem r

The margin of safety as defined in the basis of any technical specification has not been reduced. The basis for this statement is:

The Basis for Technical Specification 3/4.11.1 states that the purpose nf the Specification is to irr.plement 10 CFR $0 Appendix 1. As stated above, the dose from an instantaneous, release of all of the contents of the Heating Steam System would result in only a small fraction of Appendix I limits.

2. Reason for the Change:

Radiological Contarnination has been found in the House lleating System. NRC Bulletin 8010 requires that a Safety Evaluation be perfortned a system which was designed to be a clean system is found to contain radiological contamination.

Preparer - Date N

$0$9FRhD IE

_- _ . _ _ _ _ . . . ___._ -_,-._ _ i_. . _ - , _ . . - _ . _ __ _ , . . - . , _ _ _ _ _ - . . - _ - _ . . - - - , _ - , - . . - - . . - _ _ _ . - - ,

i I

ACP 1.2 2A2 Rev.1 MAJOR

/

Forat 3 10 CFR 50.59 (b)(2) Report Page1ofI Safety Evaluation Number: _ SY.EV 97 Cl20 Revision: 0 l Document Number: _See Anelleability Review Revision:

Document

Title:

Electrical Eaulement Oualineation Pronram l

Provide a brief desenption of the change and a summary of the Safety Evaluation in the format below.

1. Bilef Description of Change and Safety Evaluation Summary:

CFR50.82 certifications of permanent cessation of power opera in the defueled state, the cancellation of thk Prostam has no ef"ect on the reviced accidents ana the SAR and does not reduce the margin of safety and barause credit the fue is ne taken for a..e environmental qualification of electrical equipment.

2. Reason for the Change:

CFR$0.82 certi0 cations of permanent cessation of power operat vessel. In the defueled mode, credit is not taken for environmental qualification of elect-ical equipm also no longer applicable Therefore it is recommended that this program b '

Preparer -

• Date ?

ACP 1.2-2.42 Rev. I MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page 1 of 2 Safety Evaluation Number: SY.Ev.97 0130 Revision:

Document Number: Reinonst 1Mgmo CH.97 203 " System Review in Resnonse to IE BulletinEh10 Revision:

Document

Title:

Continued Use of the Yard Storm Drainace System as a Contaminated System provide a brief description of the change and a summary of the Safety Evaluation in the format below.

l. Briel Description of Char ge and Safety Evaluation Summary:

Radiological contamination is present in portions of the Yu Storm Drainage System. His system was designed as clean l

system, but over the years of service radioactive contamimilon has been introduced. De most significant is the low levels of radiological contamination found in the drains adjacent to or leaving the Radiological Control Area (RCA) Yard. nis l

change addresses the continued use of the Yard Storm Drainage System with radiological contamination present ' in the I system. NRC IE Dullet!n 8010 requires that a 10 CFR 50.59 Review and Safety Evaluation be performed, whenever a system which is designed as a clean system is found to contain radiological contamination and its continued operation with contaminneon is desired. De Safety Evaluation has concluded that an Unresolved Safety Question (USQ) is not involved.

The basis for this determination is that the dose to the general public would be extremely low and well below any re limits if the entire sediment contents of the Yard Storm Drainage System Catch Basins were released to the environment, his change does not constitute an unreviewed safety question because:

I f

here is no increase in the probability ofoccurrence or the consequences of an accident or malfunction of equipment importe.nl to safety previously evaluated in the safety analysis report. He basis for this statement is:

De Yard Storm Drainage System does not interface with fuel handling activities or radioactive waste systems as described in the UFSAR. It, therefore, has no etTcet on the probability of any accident evaluated in the UFSAR. The amour.. of radioactivity present in the Yard Stotm Drainage System has no effect on the system's ability to function, not would it ba expected to increase the probability of an accident or consequences of an accident.

Building and Equipment Drains that had a potential for the release of radiological contaminated liquids to the environment have been scaled.

De routine discharge of contaminated liquid effluents via the normal Dow paths, along with the small amounts of contamination present, will has e no effect of the systems ability to function. Continued u,e of the notsysten, increase j the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the uSAR and will '

have minimal impact on the safety of the plant, the personnel, the public, or the environment he possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis rep has not been created. De basis for this statement is:

The contamination levels present in the Yard Storm Drainage System will not result in an equipment failure which could cause an accident of a diDerent type than previously evaluated in the UFSAR. The Yard Storm Drainage System does not interface with any systenis associated with accidents evaluated in the UFSAR.

De possibility of a total dislodginent of the sediment contained in the Yard Storm Drainage System catch basins has existed and the presence of contaminated sediment does not change that possibility he consequences of sediment dislodgment have been addressed and was found to be acceptable.

$t970D0 DOC 9 -

ACP 1.2 2.42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page 2 of 2 Safety Evaluation Number: tv Ev.97 0110 . Revision:

Documen* Number: .Regior.u to M*=a Cll 97 203 " System Review in Retnnna* to IE Bulletin 1010 Revision: 4 Document

Title:

Caralnued Use of the Yard Carm Draina=e Syste*,mCnnemminne.A Svet.m De margin of safety as defined in the basis for any technical specification has not been reduced. The basis for this statement is:

ne Basis for Technical1f *cification 3/4.11.1.2 ttat s that the purpose of the Specification is to implement 10 CFR 50 Appendix 1. As stated above, the dose from the complete discharge o 'the sediment contained in the Yard Storm Drainage System c tch basins would result in only a small fraction of the estabdished limits of 10 CFR 50, Appendix 1,

2. Reason for the Change:

NRC IE Bulletin 8010 requires t)at a 10 CFR 50.59 Review and hafety Evaluation be performed, whenever a system which is designed as a clean system is found to contain radiological contamination and its continued operation with contamination is desire J.

4 Preparer Date

. $E970130. DOC .

n a

ACP 1.2 2.42 Rev.1 MAJOR Form 3 - 10 CFR 50,59 (b)(2) Report Page 1 of'2 Safety Evaluation Number: SY.EV.97 0131 Revision:

Document Number: Response to Memo cil.97 203 " system Review 5 Resnonne to ir Bulletin 8010" Revision:

Document

Title:

continued Use orcomnonent cooline Water system as a contaminated system provide a brief description of the change and a summary of the Safety Evaluation in the format below.

1. Drief Description of Change and Safety Evaluation Summary:

Radiological contamination is present in portions of the Yard Storm Drainage System. His system was designed as l

system, but over the yetrs of service radioactive contamination has been introduced. The most significant is the low level radiologica, contamination found in the drains adjacent to or leaving the Radiological Control Area (RCA) Yaid. His change addresses the continued use of the Yard Storm Drainage System with radiological contamination present in system. NRC IE Bulletin 8010 req'aires that a 10 CFR 50.59 P,eview and Safety Evaluation be performed, whenever a system which is designed as a clean system is found to contain radiological contamination and its continued operatk,n contamination is desired. De Safety Evaluation has concluded that an Unresolved Safety Question (USQ) is not involved.

De basis for this determination is that the dose to the general public would be extremely low and well below limits if the entire sediment contents of the Yard Storm Drainage System Catch Basins were released to the environment His change does m,t constitute an unreviewed safety question because:

Dere is no increase in the probability of occurrence or the consequences of an accident or malfunction of equipme traportant to safety previously evaluated in the safety analysis report. De basis for this statemei;t is:

De CCW System does not interface with I"uel Handling Activities. De CCW System does interface with the comp the Waste Gas System. His interface is for cooling of the Waste Gas Compressors, Compressor Oil Coolers, and After Coolers, ne low activity levels will not increase the pr0 ability of any accident previously evaluated in the UFSAR for the Radioactive Waste Systems.

De small amount of radioactivity present in the CCW has no effect on the system's ability to function, nor wc'ild it be expected to cause a failure of any CCW System components or piping. Contbued use of the system will not increase the probability of an occurrence of a malfunction of CCW equipment.

As a result of the levels of radiological contamination present in the CCW System, a malfunction. such as a leak into the Service Water System or a piping leak outside the PAB into the Yard Storm Drainage System cculd result in an unmonh and uncontrolled release of radioactivity. He water contained in the CCW System was sampled and analyzed on Oc 20,1997 for radionuclides. An analys.is of the water was performed to determine the Gross Alpha activity present in system resulted in less than Minimum Detectable Activity (MDA).

A total release of the CCW System contents would pose no threat to the safety of the plant, the personnel, the publi environment.

The possibility for an accident or malfunction of a difTerent type than any evaluated previously in the safety an has not been created. He basis for this statement is:

De amount of radiological contamination present in the CCW System will not result in an equipment failure which could cause an accident of a different type than previously evaluated in the UFSAR.

SI M 0131.D O C i,.

l ACP 1.2 2,42 Rev.1 MAJOR Form 3 - 10 CFR 50.59 (b)(2) Report Page 2 of 2 Safety Evaluation Number: SY.EV.97.0131 Revision:

Document Number: R==aaaa* to M*=a c14.97.201 "svai = Review in nz:=,

to ir noni ela a0.10" Revision:

Document

Title:

cnattaa d Une of cn===r-: can11a= Water Evan = an a caae -!reA svai-=

/

De ponibilities for CCW System leakage into the ! er Ice Water System or Yard Storm Drainage System h and the amounts of radioactivity present in the system does not change that possibility, he sma'l amo the CCW System will not cause any physical degradation of the CCW System or components that would b create a differeht type of malfunction previously evaluated in the UFSAR.

De is: margin of safety as defined in the basis for any technical specification has not been toducei The basis for th We Basis for Technical Specification 3/4.11.1.2, states that the purpose of the Specification is to implement 10 CI Appendix 1. As stated above, the dose from the complete discharge of the CCW System would result in only of the established limits of 10 CFR 50, Appendix ! and would not pose any undue risk to the safety of the plant, personnel, the public, or the environment,

2. Reason for the Change:

Radiological contamination is present in the Component Cooling Water System. NRC IE Bulletin 8010 require Safety Evaluation be performed w heneve an assumed clean *ystem is found to contain radiological contamination.

Preparer Date

$E970131. DOC