ML20094L820

Forwards Rept of Changes,Tests & Experiments Per 10CFR50.59 for Oct 1994 - Sept 1995

ML20094L820
Person / Time
Site:	Calvert Cliffs
Issue date:	11/16/1995
From:	Denton R BALTIMORE GAS & ELECTRIC CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9511210140
Download: ML20094L820 (155)

Text

Ro:ERT E. DENTON Baltimore Gas tnd Electric Company Vice President Calvert Cliffs Nuclear Power Plant Nuclear Energy 1650 Calvert Cliffs Parkway L.usby, Maryland 20657 410 586-2200 Ext.4455 local 410 260-4455 Baltimore November 16,1995 U. S. Nuclear Regulatory Commission Washington,DC 20555 ATTENTION:

Document Control Desk

SUBJECT:

Calvert Cliffs Nuclear Power Plant Unit Nos.1 & 2; Docket Nos. 50-317 & 50-318 Report of Changes. Tests and Exocriments - 10 CFR 50.59 In accordance with 10 CFR 50.59(b)(2), Baltimore Gas and Electric Company hereby submits a report containing brief descriptions of changes, tests, and experiments approved under the provisions of 10 CFR 50.59 Attachment (1) of this report includes 50.59 evaluations approved during the 12-month period from October 1,1994 to September 30,1995. Attachment (2) includes a number of 10 CFR 50.59 summaries which were recently identified as missing from our database and not previously provided to NRC. This deficiency is being addressed under our corrective action program.

Items in the report are sorted by 50.59 identification number.

Should you have questions regarding this matter, we will be pleased to discuss them with you.

Very truly yours, 4n&

/

7 for R. E. Denton Vice President -Nuclear Energy RED /MRC/dlm Attachments:

(1)

Calvert Cliffs Nuclear Power Plant Report of Changes, Tests, and Experiments

[10 CFR 50.59(b)(2)]

(2) 10 CFR 50.59 Summaries Not Previously Provided to NRC 2

9511210140 951116 4

PDR ADOCK 05000317 R

PDR l

vv Document Control Desk r

November 161995 t

Page 2 i

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(

cc:

(Without Attachments)

D. A. Brune, Esquire J. E. Silberg, Esquire L. B. Marsh, NRC D. G. Mcdonald, Jr., NRC

[

T. T. Martm, NRC l

Resident Inspector,NRC R. I. McLean, DNR J. H. Walter, PSC i

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ATTACHMENT (1)

CALVERT CLIFFS NUCLEAR POWER PLANT REPORT OF CHANGES, TESTS, AND EXPERIMENTS

[10 CFR 50.59(b)(2)]

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i Baltimore Gas and Electric Company Docket Nos. 50-317 & 50-318 November 00,1995

NWRB018 NUCLEIS 10/15/1995 Search Proces2 A& oc Report 1

STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

Document ID Revision Status

as==w=======================

==

94-1-061-015-R01 62 54 ject:

INSTALL PERMANENT COVERS OVER TSP BASKETS IAW 94 052 012 00 Alias:

POSRC #:

94-141 Assoc Doc ID: 94-052-012-00 Revision To: 0000 Assoc Stat: O Assoc Type: MCR Ref Doc ID:

Rev:

Refer Type.

Sender Xmtl #

Xmt! Date

==================================================================================================_

Other refs:

Pers Rzfs:

i Equipment:

Org/Div:

System Code: 061 CONTAINMENT SPRAY Text:

NRC SUfetARY:

THE PROPOSED ACTIVITY 15 THE INSTALLATION OF MCR 94 052 012 00. THE MCR INSTALLS PERMANENT COVERS OVER THE TRISODIUM PHOSPHATES D0DECANYDRATE (TSP) BASKETS.

THE COVERS WILL ALLOW THE CONTAINMENT DECONTAMINATION EFFORT TO BEGIN AS EACH UNIT ENTERS ITS SHUTDOWN PROCESS.

THIS 50.59 IS BEING PREPARED BECAUSE THE MODIFICATION AFFECTS THE UFSAR DESCRIPTIONS OF THE TSP BASKETS.

THE COVER ASSEMBLIES MET ALL THE APPLICABLE DESIGN REQUIREMENTS FOR USE IN THIS APPLICATION. THE COVERS WILL NOT PREVENT THE TSP BASKETS FROM PERFORMING THEIR DESIGN BASIS FUNCTION. FURTHERMORE, THE COVERS ARE RIGIDLY MOUNTED SUCH THAT THEY WILL NOT MOVE FREELY AND BLOCK THE CONTAINMENT SUMP.

i THIS ACTIVITY WILL NOT DEGRADE THE RELIABILITY OF ITS EQUIPMENT / SSC.

THE ACTIVITY DOES NOT INVOLVE AN UNREVIEWED SAFETY QUESTION, NOR DOES IT REDUCE THE MARGIN OF SAFETY AS DESCRIBED IN THE TECHNICAL SPECIFICATION i

BASES.

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NWRB018 NUCLEIS 10/15/1995 Search Procesa A& oc Report 2

STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

Document ID Revision Status

e=======================

==

94-B-064-101-R00 62 Stbject:

ADD TWO ISOLATION VALVES IN THE COMON DISCHARGE LINE DOWNSTREAM OF THE REACTOR VALVE Alias:

94-8-064-101-R00 POSRC #:

94-148 Assoc Doc ID: 94-064-011-00 Revision To: 0000 Assoc Stat: C Assoc Type: MCR Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xmtl Date

===================================================================================================

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Equipment:

Org/Div:

System Code: 064 REACTOR COOLANT Text:

PROPOSED ACTIVITY:

THIS ACTIVITY ADDS TWO ISOLATION VALVES IN THE COMON DISCHARGE LINE DOWNSTREAM OF THE REACTOR VESSEL AND PRESSURIZER VENT LINE SOLENOID VALVES ( 1 / 2 - SV - 103, - 104, - 105, - 106).

Document ID Revision Status

================================

94-8-036-105-R00 62

Subject:

ALLOW USE OF ALTERNATE MATERIALS FOR STEAM DRIVEN AFW PUNPS IAW MCR 94 036 005 00 Atlas:

POSRC #:

94-150

m..

NWR9018 NUCLEIS 10/15/1995 Search Process A& oc Report-3 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

Assoc Doc ID: 94-036-005-00 Revision To: 0000 Assoc Stat: 0 Assoc Type: MCR Ref Doc ID:

Rev:

Refer Type:

I Sender Xmtl #

Xmtl Date

ne===============================================================================

============

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Equipment:

Org/Div:

system Code: 036 AUXILIARY FEEDWATER Text:

NRC SIAMARY:

THIS MCR PROVIDES THE DESIGN ENGINEERING REQUIRED TO ALLOW THE USE OF ALTERNATE MATERIALS FOR THE STEAM DRIVEN AUXILIARY FEEDWATER (AFW) PUMPS. THE MCR Alt 0WS THE USE OF MATERIAL CONFORMING TO THE REQUIREMENTS OF ASTM A 479 FOR THE PLMP SHAFT AND ASTM A 743 FOR THE PUMP IMPELLER.

THIS IS AN EQUIVALENT MATERIAL CHANGE WHICH USES A DIFFERENT ASTM SPECIFICATION FOR THE SAME MATERIAL TYPE / GRADE AS THE EXISTING TYPE / GRADE NO PROTECTIVE OR SAFETY FEATURES OF THE AFW SYSTEM ARE ALTERED. THE ALTERNATE ASTM SPECIFICATION WILL NOT DEGRADE OR PREVENT ACTIONS DESCRIBED OR ASSLMED IN THE SAR. THIS ACTIVITY IS CONSISTENT WITH THE REQUIREMENTS OF THE ORIGINAL DESIGN CODES AND STANDARDS.

THEREFORE, THIS ACTIVITY DOES NOT INVOLVE AN UNREVIEWED SAFETY QUESTION, NOR DOES IT REDUCE THE MARGIN OF SAFETY AS DESCRIBED IN THE TECHNICAL SPECIFICATION BASES.

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NWR8018 NUCLEIS 10/15/1995 Search frocess Adioc Report 4

STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

Document ID Revision Status

================================

94-8-045-100-R00 62

Subject:

REVISE UFSAR TO LOWER THE MAIN FEEDWATER FLOW AFTER A TURBINE TRIP.

Alias:

POSRC #:

94-150 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: C Assoc Type: ESP Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xmtl Date

============

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Equipment:

Org/Div:

System Code: 045 FEEDWATER Text:

SupW4ARY: (FOR NRC REPORT)

THIS ACTIVITY ADJUSTS THE POST TRIP MAIN FEEDWATER FLOW TO 3.8%. ANALYSIS HAS SHOWN THAT 3.8% IS ADEQUATE TO REMOVE DECAY HEAT AND PREVENT ACTUATION OF AUXILIARY FEEDWATER. THE 3.8% POST TRIP MAIN FEEDWATER FLOW ALSO MAXIMIZES THE TIME AVAILABLE FOR OPERATOR ACTION TO PREVENT RCS COOL DOWN BELOW THE NORMAL TEMPERATURE CONTROL BAND. THE POST TRIP MAIN FEEDWATER FLOW FUNCTION IS PERFORMED BY NON SAFETY RELATED COMPONENTS AND IS NOT CREDITED IN RESPONSE TO ACCIDENTS.

NO SYSTEM DESIGN CHANGES ARE MADE OTHER THAN A NEW BYPASS FEED VALVE POST TRIP SETPOINT. THEREFORE, THIS ACTIVITY DOES NOT INVOLVE AN UNREVIEWED SAFETY QUESTION.

.m

NNR8018 NUCLEIS 10/15/1995 Search Process A&oc Report 5

STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995) t Doctment 10 Revision Status

===============================

94-B-060-084-R00 62

Subject:

ALLOWS FOR REPLACEMENT OF EXISTING FIBERGLASS MIST ELIMINATION PADS WITH STAINLESS STEEL WIRE PADS.

Alias:

POSRC #:

94-150 t

Assoc Doc ID: 93-060-003-00 Revision To: 0000 Assoc Stat:

C Assoc Type: MCR Ref Doc 10:

Rev:

Refer Type:

Sender Xmtt #

Xstl Date

===================================================================================================

Other refs:

PIrs RIfs:

Equipment:

Org/Div:

System Code: 060 PRIMARY CONTAINMENT HEAT AND VENT Text:

SUPMARY: (FOR NRC REPORT)

THIS MCR ALLOWS REPLACING CONTAINMENT IODINE REMOVAL FILTERS' DISPOSABLE FIBERGLASS MIST ELIMINATION PADS WITH WASHABLE STAINLESS STEEL WIRE PADS.

THE EXISTING (FIBERGLASS) AND REPLACEMENT (STAINLESS STEEL) MIST ELIMINATION PADS ADEQUATELY PERFORM THE SAME FUNCTION OF REMOVING WATER DROPLETS FROM THE AIR PRIOR TO REACHING THE HEPA FILTERS AND CHARCOAL BEDS IN THE CONTAINMENT i

FILTER UNITS. HOWEVER, THE STAINLESS STEEL PADS PROVIDE SUPERIOR DURABILITY.

i ThE MODIFIED FILTER UNITS WILL REMAIN STRUCTURALLY AND SEISMICALLY ADEQUATE.

THIS ACTIVITY DOES NOT ALTER THE DESIGN CHARACTERISTICS OF ANY EQUIPMENT IMPORTANT TO SAFETY THAT PERFORM AN ACCIDENT MITIGATION FUNCTIONS. IN ADDITION, ACCIDENT INITIATORS ARE NOT AFFECTED BY THIS ACTIVITY. THEREFORE, THE PROBASILITY AND CONSEQUENCES OF PREVIOUSLY EVALUATED MALFUNCTIONS AND ACCIDENTS HAS NOT BEEN INCREASED, AND THE POSSIBILITY OF ACCIDENTS AND MALFUNCTIONS OF A DIFFERENT TYPE OTHER THAN PREVIOUSLY ANALYZED HAVE NOT BEEN CREATED. THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL i

SPECIFICATIONS HAS NOT BEEN RECUCED. THUS, THERE IS NO UNREVIEWED SAFETY QUESTIONS ASSOCIATED WITH THIS ACTIVITY.

t Document 10 Revision Status

==

94-B-058-097-R00 62 I

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NNRB018 NUCLEIS 10/15/TC Search Proces3 A<ftoc Report 6

STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

Subject:

MODIFY UFSAR TO STATE THAT THE THERMAL MARGIN / LOW PRESSURE (TM/LP) PRETRIP SETPolNT OCCURS AT 50 PSI ABOVE THE VARIABLE TM/LP TRIP SETPOINT.

Alias:

POSRC #:

94-153 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat:

C Assoc Type: ESP Ref Doc ID:

IRO-033-301 Rev:

0000 Refer Type:

IR ISSUE REPORT Sender Xmtl #

Xmt! Date

===================================================================================================

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Pars Rafs:

Equipment:

Org/Div:

System Code: 058 REACTOR PROTECTIVE Text:

Supe 4ARY: (FOR NRC REPORT)

THE PROPOSED ACTIVITY IS TO MODIFY UFSAR TABLE 7-1 AND UFSAR FIGURE 7-6 A TO DOCUMENT THAT THE THERMAL MARGIN / LOW PRESSURE (TM/LP) PRETRIP SETPOINT WAS REDUCED FROM 100 PSI TO ONLY 50 PSI ABOVE THE VARIABLE TM/LP TPIP SETPOINT BACK IN 1975. ISSUE REPORT IRO - 033 - 301 DOCUMENTS THIS UFSAR DEFICIENCY.

THEREFORE, THIS 50.59 SAFETY EVALUATION SUPPORTS THE "AS - BUILDING" 0F THE UFSAR.

THIS 50.59 ONLY AFFECTS THE TM/LP PRETRIP SETPOINT AND DOES NOT IN ANY WAY AFFECT THE CALVERT CLIFFS TECHNICAL SPECIFICATION COLOR DEFINED LIMITS /

REQUIREMENTS FOR THE TM/LP TRIP SETPOINT.

THE TM/LP PRETRIP SETPOINT OF 50 PSI ABOVE THE VARIABLE TM/LP TRIP PROVIDES THE CONTROL ROLN OPERATORS WITH VISUAL AND ALDIBLE NOTIFICATION OF PLAFT PARAMETERS APPROACHING TRIP CONDITIONS AND IT INITIATES A CEA WITHDRAWAL PROHIBIT. THE ACCIDENT ANALYSES DOES NOT CREDIT THE TM/LP PRETRIP OR CEA WITHDRAWAL PROHIBIT TO LIMIT OR MITIGATE THE CONSEQUENCES OF AN ACCIDENT.

THE TM/LP TRIP LIMITS THE CONSEQUENCES OF AN ACCIDENT.

THE PROPOSED ACTIVITY DOES NOT INCREASE THE PROBABILITY OR CONSEQUENCES OF A MALFUNCTION OR AN ACCIDENT PREVIOUSLY ANALYZED IN THE SAR, IT DOES NOT CREATE THE POS$1BILITY OF A NEW MALFUNCTION OR NEW ACCIDENT, AND IT DOES NOT AFFECT THE MARGIN OF SAFETY AS DEFINED IN THE TECHNICAL SPECIFICATIONS BASES.

THEREFORE, THE PROPOSED ACTIVITY IS NOT AN UNREVIEWED SAFETY QUESTION.

NNRB018 NUCLEIS 10/15/1995 i

Setrch Proces2 A & oc Report 7

STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

L i

Document ID Revision Status

==

94-B-102-053-R00 62

Subject:

REPLACE EXISTING CONTROL ROOM FURNITURE AT THE SUPERVISOR'S DESK AND THE TWO UNIT OPERATOR'S DESKS WITH NEW i

FURNITURE.

f Alias:

POSRC #:

94-153

[

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Assoc Doc ID: 89-0079-55 Revision To: 0000 Assoc Stat: C Assoc Type:

FCRSUP Ref Doc 10:

Rev:

Refer Type:

Sender Xmtl #

Xmtl Date '

===================================================================================================

Cther refs:

Pers Rifs:

Equipment:

Org/Div:

l System Code: 102 PLANT AREAS Text:

SUMARY: (FOR NRC REPORT)

THIS ACTIVITY INVOLVES THE REPLACEMENT OF THE EXISTING CONTROL ROOM FURNITURE I

AT THE SUPERVISOR'S DESK AND THE TWO UNIT OPERATOR'S DESKS WITH NEW i

FURNITURE. THE NEW WORK SPACES ALLOW THE EXISTING OPERATOR UTILITY CRT'S AND t

SUPERVISOR SPOS CRT'S TO BE RELOCATED FROM THE TOP OF THE WORK SPACES TO UNDERNEATH GLASS COUNTER TOPS, BOOK SHELVES ARE ALSO INCORPORATED INTO THE NEW WORK SPACE DESIGN. BOOK SHELVES AND COUNTER TOPS ARE ADDED TO THE NEW COMUNICATIONS CONSOLE LOCATED BEMIND THE SUPERVISOR'S DESK. CONTROL ROOM l

CARPETING WILL BE REPLACED AS PART OF THIS ACTIVITY. THE NEW FURNITURE WILL BE SEISMICALLY MOUNTED (SEISMIC II/I REQUIREMENTS) TO ENSURE NO DAMAGE RESULTS TO ANY SAFETY RELATED EQUIPMENT LOCATED WITHIN THE CONTROL R0(NI. MFE AND FIRE PROTECTION EVALUATIONS WERE ALSO FAVORABLE.

THIS ACTIVITY WILL IfPROVE THE AESTNETIC AND FUNCTIONALITY OF THE CONTROL ROOM AS WELL AS PROVIDE ADDED CONVENIENCE AND CONFORT TO THE CONTROL ROOM i

OPERATORS DURING THE CONDUCT OF CONTROL ROOM ACTIVITIES. -THE NEW WORK SPACES I

WILL ENHANCE THE OPERATOR'S FIELD OF VISION WHEN VIEWING THE MAIN CONTROL

[

ROOM PANELS FROM THE WORK SPACES, ENHANCES THE OVERALL ORGANIZATION OF THE r

CONTROL ROOM, AND PROVIDES ADDED WORK SPACE FOR THE OPERATIONS PERSONNEL.

THIS ACTIVITY HAS NO IMPACT ON THE OPERATION, PERFORMANCE, OR STRUCTURAL-i INTEGRITY OF ANY $$C IMPORTANT TO SAFETY. THIS ACTIVITY MAS NO IMPACT ON THE i

PROBABILITY OF ANY ACCIDENT OR MALFUNCTION OF EQUIPMENT Ifr0RTANT TO SAFETY.

AS PREVIOUSLY EVALUATED IN THE SAR. TMIS ACTIVITY HAS NO IMPACT ON THE CONSEQUENCES OF AN ACCIDENT OF MALFUNCTION OF EQUIPMENT IMPORTANT TO l

SAFETY AS PREVIOUSLY EVALUATED IN THE SAR. THIS ACTIVITY DOES NOT CREATE THE POSSIBILITY OF ANY NEW ACCIDENT OR MALFUNCTION. THIS ACTIVITY DOES NOT INVOLVE AN UNREVIEED SAFETY QUESTION.

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i WNR3018 NUCLEIS 10/15/1995 Search Procesa A & oc Ceport 9

STATUS 62 OR 64 50.59S (10/01/1994 inRU 09/30/1995)

Assoc Doc ID: 89-0176 Revision To: 0000 Assoc Stat: C Assoc Type:

FCR Ref Doc ID:

Rev:

Refer Type:

Sender Xmtt #

Xatt Date

...............a.................................................................========================

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l Drg/Div:

System Code: 102 PLANT AREAS Text:

SUMMARY

(FOR NRC REPORT)

THE ACTIVITY COVERED BY THIS 50.59 EVALUATION IS THE UPGRADE OF THE EXISTING SECURITY SYSTEM, WNICH INCLLA)ES THE REPLACEMENT OF THE SOUTH PROCESSING BUILDING (FORMERLY KNOWN AS THE SECURITY PROCESSING BUILDING) WITH A NEW NUCLEAR SECURITY FACILITY (NSF). THE NSF WILL MOUSE THE NEW SECURITY CIMPUTER SYSTEM, VIDEO SWITCHER, SECONDARY ALARM STATION (SAS),

ADMINISTRATIVE OFFICES, STAFF SUPPORT FUNCTION (E.G., MANAGERS' OFFICES, WEIGHT ROOM, REST ROOMS), AND SECURITY SCREENING. THE CURRENT PERIMETER MICROWAVE / INFRARED / VIBRATION INTRUSTION DETECTION SYSTEMS AND THE CLOSED CIRCUIT TELEVISION SYSTEM WILL BE REPLACED WITH NEW EQUIPMENT. THE PROTECTED AREA BOUNDARY WILL BE EXPANDED AND UPGRADED. AN OFFSITE POWER SYSTEM FOR THE NEW SECURITY SYSTEM WILL BE PROVIDED AND WILL INCLUDE AN UNINTERRUPTISLE POWER SOURCE (UPS) AND A STANDBY DIESEL GENERATOR. UFSAR FIGURES 1-2, 9-22 AND 9-22 E WILL BE REVISED TO REFLECT CHANGES FROM THE RELOCATION OF THE PROTECTED AREA BOUNDARY, CONSTRUCTION OF THE NSF, AND THE ADDITION OF A LINE AND ISOLATION VALVE ON THE TIE-IN 04 THE OUTSIDE FIRE PROTECTION SYSTEM FOR FUTURE PLANT OFFICE FACILITY (PDF) SPRINKLER SYSTEM.

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.m NNRB018 NUCLEIS 10/15/1995 Search Procesa A & oc Report 11 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

Pers RIfs:

Equipment:

Org/Div:

System Code: 024 EMERGENCY DIESEL GENERATOR Text:

SupMARY: (FOR NRC REPORT)

THIS MCR ADDRESSES THE REMOVAL OF THE MANUAL VALVES IN THE JACKET COOLING SUPPLY LINES TO THE EMERGENCY DIESEL GENERATOR TURBOCHARGERS AND THE REMOVAL OF THE RELIEF VALVES (RV) IN THE OUTLET PIPING.

STANDARD PIPING SYSTEM FITTINGS (E.G.: UNION) WILL BE INSTALLED IN PLACE OF THE MANUAL VALVES AND A PLUG WILL BE INSTALLED IN PLACE OF THE RELIEF VALVES.

THIS MCR APPLIES TO 11, 12, AND 21 EDG.

RECENT INVESTIGATION INTO THE DESIGN BASES FOR THE RV'S INDICATED THAT THE RV'S WERE INSTALLED FOR PROTECTION GF THE TURBOCHARGERS CASINGS DURING TESTING OF THE JACKET COOLING SYSTEM. WITH THE MANUAL VALVES LOCKED OPEN OR REMOVED (SUPPLY OR RETURN LINES), THERE IS A DIRECT PATH To THE SYSTEM VENT AND THERE ARE NO MECHANISM WHICH COULD RESULT IN OVERPRESSURIZING THE TURBOCHARGER CASINGI HENCE, THE RV'S ARE NO LONGER NEEDED. HOWEVER, WITH THE NEW RV'S REMOVED, THE MANUAL VALVES SHOULD ALSO BE REMOVED TO ELIMINATE THE POSSIBILITY OF INADVERTENTLY ISOLATING JACKET COOLING TO THE TURBO-CHARGERS AND OVERPRESSURIZING THE PIPING AND THE TURBOCHARGER CASING.

NO PROTECTIVE OR SAFETY FEATURES OF THE EDG'S ARE ALTERED. THE MODIFICAITON WILL NOT DEGRADE OR PREVENT ACTIONS DESCRIBED OR ASSUMED IN THE SAR. THIS ACTIVITY l$ CONSISTENT WITH THE REQUIREMENTS OF THE ORIGINAL DESIGN CODES AND STANDARDS.

i THEREFORE, THIS ACTIVITY DOES NOT INVOLVE AN UNREVIEWED SAFETY QUESTION, NOR DOES IT REDUCE THE MARGIN OF SAFETY AS DESCRIBED IN THE TECHNICAL SPECIFICATION BASES.

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NNR8018 NUCLEIS 10/15/1995 Search Process Adioc Report 12 STATUS 62 OR 64 50.595 (10/01/1994 TNRU 09/30/1995) r t

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f l

Document ID Revision Status f

======================== =_.___i=

93-B-064-035-R02 62 Stbject:

REPLACE THE PRESSURIZER LEVEL CONTROLLERS.

[

Alias:

POSRC #:

94-157 l

Assoc Doc ID: 93-064-007-00 Revision To: 0000 Assoc Stat: C Assoc Type: MCR Cef Doc ID:

Rev:

Refer Type:

Sender Xmt! #

Xmtl Date t

i

===================================================================================================

Other refs:

Pers RIfs:

Equipment:

Org/Div:

System Code: 064 REACTOR COOLANT Text:

SUIMARY: (FOR NRC REPORT)

__..- - _ _-._-~, -._____.,..__~ - - -.--- - -,... -. - -.. _ -.,

NNR8018 NUCLEIS 10/15/1995 Search Procesa A & oc Report 13 STATUS 62 OR 64 50.595 (10/01/1996 THRU 09/30/1995)

THIS SAFETY EVALUATION CONCLUDES THAT THIS ACTIVITY IS NOT AN UNREVIEWED SAFETY QUESTION.

NCR 93-064-007-00 REPLACES THE PRESSURIZER LEVEL CONTROLLERS, 1(2)-LIC-110x AND 110Y. THIE OLD CONTROLLERS ARE OBSOLETE AND STOCK IS DEPLETED.

THESE CONTROLLERS ARE NON SAFETY RELATED AND ARE ELECTRICALLY ISOLATED FROM THE SAFETY RELATED PORTION OF THE PRESSURIZER LEVEL INSTRUMENTATION LOOPS BY QUALIFIED ISOLATION DEVICES. THE NEW CONTROLLERS ARE MOUNTED TO SEISMIC II/I CRITERIA FOR THE CONTROL ROOM PANELS.

THE NEW CONTROLLERS ARE MICROPROCESSOR BASED DEVICES WHICH WILL ACCEPT THE SAME INPUT SIGNALS AND PROVIDE THE SAME OUTPUT AS THE EXISTING CONTROLLER.

l

NWRB018 NUCLEIS 10/15/1995 Search Proces2 A e oc Report 14 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

Doctment ID Revision Status

================================

94-B-029-001-R00 62

Subject:

RETIRE THE CONTAINMENT SUBSYSTEM OF THE PLANT HEATING SYSTEM.

Atlas:

POSRC #:

94-159 Assoc Doc ID: 93-029-005-00 Revision To: 0000 Assoc Stat: C Assoc Type: MCR Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xmt! Date

================_________====================_===================================================

Other refs:

PErs R;fs:

Equipment:

Org/Div:

System Code: 029 PLANT HEATING Text:

SUMMARY

(FOR NRC REPORT)

MCR 93-029-005-00 WILL RETIRE THE CONTAINMENT SUBSYSTEM OF THE PLANT HEATING SYSTEM BECAUSE IT PERFORMS NO FUNCTIONS INSIDE CONTAINMENT. THE PLANT HEATING INLET TO CONTAINMENT IS AT PENETRATION 64 AND THE OUTLET IS AT PENETRATION 62. THE PENETRATION PIPING WILL BE CUT AND CAPPED INSIDE CONTAINMENT, AND THE METHOD OF PERFORMING THE CONTAINMENT ISOLATION FUNCTION AT PENETRATION 62 AND 64 WILL CHANGE AS A RESULT. ADDITIONALLY, FOUR NON SAFETY RELATED 480V DISCONNECT SWITCHES WILL BE INSTALLED IN EACH CONTAINMENT USING THE ABANDONED POWER SUPPLIES TO THE UNIT HEATERS' FANS.

IT DOES NOT RESULT IN AN UNREVIEWED SAFETY OUESTION BECAUSE IT IS CONSISTENT ASSEMBLIES WILL BE PROVIDED FOR WHICH NO SINGLE, CREDIBLE FAILURE OR MALFUNCTION OF AN ACTIVE COMPONENT CAN RESULT IN LOSS OF ISOLATION OR IN TOLERABLE LEAKAGE. ADDITIONALLY: (1) THE PENETRATION PIPING WILL BE CUT AND CAPPED INSIDE CONTAINMENT CLOSE ENOUGH TO THE STRUCTURE TO PRECLUDE EXCESSIVE VIBRATION AND STRESS ON THE PENETRATION ASSEMBLY FROM CANTILEVER PIPES, (2) UNSUPPORTED PORTIONS OF ABANDONED PIPING INSIDE CONTAINMENT WILL BE REMOVED, (3) THE ENDS OF THE REMAINING ABANDONED PORTIONS OF THE PLANT HEATING SYSTEM WILL BE CAPPED, AND (4) THE NON SAFETY RELATED 480V DISCONNECT SWITCHES WILL BE POWERED FROM NON SAFETY RELATED MOTOR CONTROL CENTERS.

i

NNRB018 NUCLEIS 10/15/1995 Search Proceso Adioc Report STATUS 62 OR 64 50.59s (10/01/1994 THRU 09/30/1995)

n......

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NNR9018-NUCLEIS 10/15/1995 Search Procesa A & oc Report 16 STATUS 62 OR 64 50.59S (10/01/1996 THRU 09/30/1995)

I i

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F E

i i

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i Document ID Revision Status

====================================== ______

94-8-062-092-R00 62 S4 ject:

MODIFY EQUIPEMNT ASSOCIATED WITH THE EMERGENCY POWER SOURCES AND THE 13.8 KV, 4.16 KV AND 480 V DISTRIBUTION SYSTEM.

Alias:

POSRC #:

94-159 Assoc Doc !D: 89-0079 Revision To: 0000 Assoc Stat: C Assoc Type:

FCR Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xmtl Date

========================================.....=========..... - _=====================================

Other refs:

Pers Rifs:

Equipment:

Org/Div:

i System Code: 062 CONTROL BOARDS Text:

SUMMARY

(FOR NRC REPORT) t i

1 5

4

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b NNRB018 NUCLEIS 10/15/1995 Search Procesa A& oc Report 17 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

THIS ACTIVITY MODI 51ED EQUIPMENT ASSOCIATED WITH THE EMERGENCY POWER SOURCES AND THE 13.8 KV, 4.16 KV AND 480V DISTRUBUTION SYSTEMS. THE ELECTRICAL AUXILIARY CONTROL PANEL (EACP) FOR UNIT 1 (IC17,1C18 AND 1C19) WILL BE ALTERED TO ENHANCE THE PRESENTATION OF INFORMATION TO THE OPERATOR FOR MONITORING THE ELECTRICAL POWER SYSTEMS. INSTRUMENTATION AND CONTROLS ARE REARRANGED TO CORRECT DISCREPANCIES IDENTIFIED BY A DETAILED CONTROL ROOM DESIGN REVIEW (DCRDR) IN THE 1980'S. THIS ACTIVITY RELOCATES EXISTING METERS ON THE METER SECTION OF THE PANELS IN ORDER FOR THE METERS TO PROPERLY ALIGN WITH THE ASSOCIATED CONTROLS ON THE BENCH SECTION OF THE PANELS. THE MODIFICATIONS TO THE EACP REMOVE NONFUNCTIONAL CONTROLS, STATUS INOICATION AND METERS ASSOCI ATED WITH DG 11, DG 12 AND EMERGENCY BUSES 11,14 AND 21 AS A PART OF DEDICATING EACH EMERGENCY DIESEL GENERATOR TO A SINGLE ENGINEERED SAFETY FEATURED BUS.

THE STRUCTURAL ADEQUACY AND SEISMIC QUALIFICATION OF NEW AND EXISTING SSC'S, OPERABILITY OF PLANT ELECTRICAL DISTRIBUTION SYSTEMS AND CONTROL PANEL REQUIREMENTS WERE EVALUTED TO ENSURE THE PROBABILITY AND CONSEQUENCES OF A PREVIOUSLY EVALUATED ACCIDENTS AND MALFUNCTIONS HAVE NOT BEEN INCREASED BY THIS ACTIVITY. PRECAUTIONS ARE OBSERVED IN ORDER TO PREVENT INSTALLATION ACTIVITIES FROM INTRODUCING A NEW MALFUNCTION OR ACCIDENT DURING MODIFICATION OF THE EACP. THIS ACTIVITY DOES NOT AFFECT THE OPERABILITY OF ELECTRICAL

tSTRIBUTION SYSTEMS. THUS, THE MARGIN OF SAFETY AS DEFINED IN THE TECHNICAL SPtC'FICATIONS IS NOT REDUCED.

THEREF0k2, THERE ARE NO UNREVIEWED SAFETY QUESTIONS ASSOCIATED WITH THIS ACTIVITY.

l Docanent ID Revision Status

================================

94-8-035-068-R00 62

Subject:

ADDITION OF MONOETHANOLAMINE (ETA), ALSO REFERRED TO AS ETHAN0LAMINE, AS AN ACCEPTABLE CHEMICAL ADDITIVE TO

+

THE CONDENSATE AND FEEDWATER SYSTEM FOR PH CONTROL IN ORDER TO REDUCE FORMATION OF CORROSION PRCDUCTS WHICH RESULT IN SLUDGE FORMATION.

Alias:

POSRC #:

94-160 Assoc Doc ID: 93-035-001-00 Revision To: 0000 Assoc Stat: C Assoc Type: MCR Ref Doc ID:

Rev:

Refer Type:

Sender Xmtt #

Xmtl Date

.____================================================================================================

Other refs:

Pers Rifs:

Equipment:

Org/Div:

System Code: 035 CHEMICAL ADDITIONS - TURBINE b

~

m

NNRB018 Search Process A & oc Report 18 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995) i Text:

SUMMARY

(FOR NRC REPORT)

THE SCOPE OF THE SUBJECT MCR ACTIVITY AND THIS SAFETY EVALUATION IS LIMITED TO AN EVALUATION / JUSTIFICATION OF THE USE OF THE MON 0ETHAN0LAMINE (ETA)

CHEMICAL AS A SUPERIOR CHEMICAL ADDITIVE FOR SECONDARY PLANT SYSTEM PH CONTROL. CURRENTLY, HYDRAZINE AND AMMONIA AND/OR MORPHOLINE ARE ADDED TO THE CCNPP CONDENSATE SYSTEM FOR PH CONTROL. THE NEGATIVE CHARACTERISTICS OF AMMONIA AND MORPHOLINE PROMPTED EPRI TO INITIATE A NUMBER OF STUDIES TO IDENTIFY AMINE COMPOUNDS WHICH HAVE PROPERTIES WICH ARE DESIRABLE FOR SECONDARY SYSTEMS PM CONTROL. EPRI IDENTIFIED SEVERAL AMINES AS HAVING DESIRABLE CHARACTERISTICS SUITABLE FOR CONSIDERATION AS A REPLACEMENT CHEMICAL ADDITIVE FOR THIS APPLIATION. OF THESE, ETA (C 22 H7 NO) WAS CHOSEN AS A SUPERIOR SECONDARY CYCLE PH ADDITIVE. THE EXISTING CHEMICAL ADDITION EQUIPMENT (CHEMICAL ADDITION SYSTEM MORPHOLINE TANKS, PUMPS, PIPING AND VALVES) CURRENTLY USED, AS WELL AS EQUIPMENT WITHIN THE SCOPE OF FCR 91-254, FOR THE STORAGE, HANDLING, AND INJECTION OF MORPHOLINE INTO THE CONDENSATE AND FEEDWATER SYSTEM WILL BE SUITABLE FOR ETA ADDITION. THEREFORE, NO PLANT HARDWARE CHANGES ARE INCLUDED IN THE SCOPE OF THIS SAFETY EVALUATION.

THE PURPOSE OF THIS MCR ACTIVITY 15 TO ALLOW THE USE OF ETA AS AN ACCEPTABLE ALTERNATIVE CHEMICAL ADDITIVE FOR PH CONTROL OF THE CONDENSATE AND FEEDWATER.

CURRENTLY MORPHOLINE OR AMMONIA ARE USED FOR THIS PURPOSE. ETA IS LESS VOLATILE AND HAS BETTER TRANSPORT CHARACTERISTICS WHEN COMPARE TO EITHER AMMONIA OR MORPHOLINE, SUCH THAT IT IC EXPECTED TO PROVIDE BETTER OVERALL CORROSION PROTECTION, PARTICULARLY IN THE WET STEAM AREAS OF THE CONDENSATE AND FEEDWATER SYSTEM. ETA HAS BEEN SUCCESSFULLY USED FOR SECONDARY PLANT SYSTEM PM CONTROL IN OTHER PWR FACILITIES WITH RECIRCULATING STEAM GENERATORS. BASED ON IN PLANT TESTING AND LABORATORY TESTING CONDUCTED BY EPRI ON ETA AND A VARIETY OF OTHER ADVANCED AMINES, ETA HAS BEEN INCLUDED IN REVISION 3 OF THE EPRI PWR SECONDARY WATER CHEMISTRY GUIDELINES AS A SUITABLE CHEMICAL ADDITIVE FOR THIS APPLICATION. ALL AMINES USED FOR ALL VOLATILE TREATMENT (A V T) FOR CONDENSATE PM CONTROL HAVE CERTAIN LIMITATIONS HOWEVER, ETA HAS BEEN FOUND TO PROVIDE A NUMBER OF ADVANTAGES WITHOUT MANY OF THE INHERENT PROBLEMS ASSOCI ATED WITH MORPHOLINE AND AMONI A INCLUDING BETTER CORROSION PROTECTION IN WET STEAM AREAS OF THE PLANT, LESS IMPACT ON THE CONDENSATE POLISHER OPERATION THROUGH REDUCED CATIONIC LOADING, AND ECONOMIC ADVANTAGES DUE TO SMALLER AMINE USAGE (QUANTITIES) AND REDUCED CONDENSATE DEMINERALIZER REGENERANT COSTS.

CURRENTLY, THE CCNPP UFSAR STATES THAT CHEMICALS (HYDRAZINE AND AMONIA OR MORPHOLINE) ARE ADDED TO THE CONDENSATE FLOW FOR OXYGEN SCAVENGING AND PH CONTROL. SINCE THIS ACTIVITY PROPOSES TO ALLOW THE USE OF A NEW ALTERNATIVE CHEMICAL ADDITIVE (ETA) FOR THIS APPLICATION, A UFSAR CHANGE IS NECESSARY AND IS INCLLDED AS PART OF THE SCOPE OF THIS ACTIVITY.

THERE ARE NO NEW SYSTEM INTERACTIONS ASSOCIATED WITH ALLOWING THE USE OF ETA SECONDARY PLANT SYSTEM PH CONTROL. THERE ARE NO NEW MALFUNCTIONS OR ACCIDENTS CREATED AS A RESULT OF THE SUBJECT MCR ACTIVITY. ALSO THERE IS No

NNRB018 NUCLEIS 10/15/1995 Search Process Adioc Report 19 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

EFFECT ON OFFSITE DOSE CONSEQUENCES AS A RESULT OF IMPLEMENTING THE SUBJECT MCR ACTIVITY. SINCE THE PROBABILITY AND CONSEQUENCES OF A PREVIOUSLY EVALUATED ACCIDENT OR MALFUNCTION HAVE NOT BEEN INCREASED BY THIS ACTIVITY, SINCE NO NEW MALFUNCTIONS OR ACCIDENTS MAVE BEEN CREATED BY THIS ACTIVITY, AND SINCE THE MARGIN OF SAFETY DEFINED BY THE TECNNICAL SPECIFICATIONS BASES IS NOT REDUCED, THERE ARE NO UNREVIEWED SAFETY QUESTIONS ASSOCIATED WITH THIS ACTIVITY.

k k

i h

Docunent ID Revision Status

================================

94-8-999-076-R00 62

Subject:

MODIFIY EXISTING ELECTRICAL DISTRIBUTION SYSTEM TO DEDICATE EMERGENCY DIESEL GENERATOR 21 To UNIT 2 AND TO DISCONNECT DG 12 FROM AN ENGINEERED SAFETY FEATURES BUS AT UNIT 1.

Alias:

POSRC #:

94-161 i

Assoc Doc ID: 89-0079 Revision To: 0000 Assoc Stat:

C Assoc Type:

FCR Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xmtl Date

===============_____======______===================================================================

Other refs:

Pers Rifs:

Equipment:

i

NNRB018 NUCLEIS 10/15/1995 Search Proces2 A & oc Report 20 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

Org/Div:

System Code:

Text:

SUMMARY

(FOR NRC REPORT)

THIS ACTIVITY MODIFIES THE EXISTING ELECTRICAL DISTRIBUTION SYSTEM TO DEDICATE EMERGENCY DIESEL GENERATOR 21 (DG 21) TO UNIT 2 AND DISCONNECT DG 12 FROM A DIFFERENT ENGINEERED SAFETY FEATURES BUS AT UNIT 1.

IN ADDITION, THIS ACTIVITY ENSURES PROCEDURAL CHANGES ARE IMPLEMENTED FOR ALIGNMENT OF THE SERVICE WATER SYSTEM TO DG 21 AND DG 12. THIS ACTIVITY WILL DISCONNECT DG 12 FROM EMERGENCY BUS 11 CONCURRENT WITH MODIFICATIONS TO CONNECT EMERGENCY BUS 21 INDICATIONS TO THE DIESEL GENERATOR CONTROL CONSOLE (DGCC) AND TO DISCONNECT DG 12 FROM, AND CONNECT DG OC TO EMERGENCY BUS 24. WHILE PERFORMING MODIFICATIONS TO DG 12, ACTIVITIES EVALUATED IN SAFETY EVALUATION LOG NO. 94 8 999 045 R00 REQUIRED AN EXTENSION OF ACTION STATEMENT B 0F TECHNICAL SPECIFICATION 3.8.2.2 TO ALLOW DG 12 TO REMAIN OUT OF SERVICE FOR UP TO 14 DAYS. DURING THIS TIME, DG 21 WILL BE CAPABLE OF BEING ALIGNED TO SUPPLY POWER TO EMERGENCY BUS 14 AND SUPPLY POWER FOR SELECTED LOADS ON EMERGENCY BUS 24.

ONCE DG 12 IS RETURNED TO SERVICE, THIS ACTIVITY WILL DEDICATE A UNIT 2 SERVICE WATER TRAIN To DG 21 AND IMPLEMENT MODIFICATIONS TO DISCONNECT DG 21 FROM EMERGENCY BUS 14 CONCURRENT WITH PREVIOUSLY EVALUATED MODIFICATIONS (SAFETY EVLAUATION LOG NO. 94 B 999 045 ROO) TO TRANSFER CONTROL OF DG 21 TO THE DIESEL GENERATOR CONTROL CONSOLE.

THIS ACTIVITY WILL BE PERFORMED DURING A UNIT 2 PLANT OUTAGE IN MODES 5 OR 6.

IT IS EXPECTED THAT UNIT 1 WILL OPERATE IN MODES 1, 2, 3, 4, 5, OR 6.

IN MODE 5 WITH DG 12 IN OPERABLE, REDUCED INVENTORY CONDITIONS ARE NOT ALLOWED.

IN MODE 6, AT LEAST 23 FEET OF WATER WILL BE MAINTAINED DVER IRRADIATED FUEL ASSEMBLIES SEATED WITHIN THE REACTOR PRESSURE VESSEL.

MODIFICATIONS IMPLEMENTED BY THIS ACTIVITY WERE EVALUATED TO ENSURE THEY 00 NOT INCREASE THE PROBABILITY OF A MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY (E.G., EMERGENCY DIESEL GENERATORS (EDGS), ELECTRICAL DISTRIBUTION SYSTEMS, SERVICE WATER SYSTEM AND MAIN CONTROL ROOM CONTROL PANELS).

EQUIPMENT IDENTIFIED AS INITIATORS OF ACCIDENTS ARE NOT AFFECTED BY THIS ACTIVITY. THEREFORE, THE PROBABILITY OF PREVIOUSLY EVALUATED MALFUNCTIONS AND ACCIDENTS HAS NOT BEEN INCREASED.

THE CONSEQUENCES OF PREVIOUSLY EVALUATED MALFUNCTIONS AND ACCIDENTS HAVE NOT BEEN INCREASED BY THIS ACTIVITY BECAUSE EQUIPMENT REQUIRED To SERVE MITIGATION FUNCTIONS UNDER THESE CONDITIONS HAVE NOT BEEN ADVERSELY AFFECTED AND CONTROL ROOM AND OFFSITE DOSES PREVIOUSLY CALCULATED REMAIN UNCHANGED AND WITHIN THE PREVIOUSLY STATED LIMITS. AN ADEQUATE NUMBER OF DIESEL GENERATORS WILL BE AVAILABLE TO SUPPORT POWER OPERATION OF UNIT 1 AND SHUT DOWN OPERATION OF UNIT 2.

PROCEDURAL CHANGES TO THE EDG COOLING WATER SYSTEMS WILL NOT AFFECT THE FLOW OF SERVICE WATER TO CTHER SSCS WHICH FUNCTION TO MITIGATE THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION.

NNRB018 NUCLEIS 10/15/1995 Search Proces3 A4oc Report 21 STATUS 62 OR 64 50.59s (10/01/1994 THRU 09/30/1995)

ONE EDG WILL REMAIN AVAILABLE FOR A SHUTDOWN UNIT AND TWO EDGS WILL BE AVAILABLE FOR A UNIT OPERATING AT POWER. IN ADDITION, WHEN OPERATING TWO UNITS AT POWER, TWO EDGS WILL BE AVAILABLE (ONE OF WHICH WOULD BE A SWING DIESEL GENERATOR CAPABLE OF SERVING EITHER UNIT) WILL BE AVAILABLE FOR EACH UNIT. ADMINISTRATIVE CONTROLS ON COOLING WATER SUBSYSTEMS ENSURES THAT A FAILURE OF A SERVICE WATER SUBSYSTEM ASSOCIATED WITH AN ENGINEERED SAFETY FEATURES BUS FORMERLY CONNECTED DG 12 OR DG 21 WILL NOT AFFECT THE OPERABILITY OF THE EDGS. NO NEW SYSTEMS INTERACTIONS ARE BEING CREATED BY THIS ACTIVITY. THEREFORE, THE POSSIBLITY OF A NEW MALFUNCTION OR ACCIDENT CREATED BY THIS ACTIVITY.

THE MARGIN OF SAFETY EXPRESSED IN THE BASES OF THE TECHNICAL SPECIFICATIONS IS NOT REDUCED BECAUSE SEQUENCING OF INSTALLATION ACTIVITIES ENSURES THAT EITHER AN EDG WILL BE AVAILABLE TO SUPPLY EMERGENCY POWER TO AN ENGINEERED SAFETY FEATURES BUS AT UNIT 2 AT ALL TIMES, OR A TEMPORARY DIESEL GENERATOR WILL BE CONNECTED IN ACCORDANCE WITH THE TECHNICAL SPECIFICATIONS.

UPON COMPLETION OF THIS ACTIVITY, TWO OPERATIONAL EDGS (OhE OF WHICH WOULD BE A SWING CIESEL GENERATOR CAPABLE OF SERVING EITHER UNIT) WILL BE AVAILABLE TO SUPPLY EMERGENCY POWER TO ENGINEERED SAFETY FEATURES BUSES AT EACH UNIT.

THEREFORE, THERE ARE NO UNREVIEWED SAFETY QUESTIONS ASSOCIATED WITH THIS ACTIVITY.

Document ID Revision Status

================================

94-2-062-050-R01 62

Subject:

THIS ACTIVITY MODIFIES EQUIPMENT ASSOCIATED WITH THE EMERGENCY POWER SOURCES AND THE 13.8KV, 4.16 KV AND 480V DISTRIBUTION SYSTEMS.

Alias:

POSRC #:

94-162 Assoc Doc ID: 89-0079 Revision To: 0000 Assoc Stat: O Assoc Type:

FCR Cef Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xmtl Date

= = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = - = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = === =========== ======

Oth2r refs:

P rs Rafs:

Equipment:

Org/Div:

System Code: 062 CONTROL BOARDS Text:

THIS ACTIVITY MODIFIES EQUIPMENT ASSOCIATED WITH THE EMERGENCY POWER SOURCES AND THE 13.8KV, 4.16 KV AND 480V DISTRIBUTION SYSTEMS. THE ELECTRICAL AUXILIARY CONTROL PANEL (EACP) FOR UNIT 2 (1C19,1C20 AND 2C17) WILL BE ALTERED TO ENHANCE THE PRESENTATION OF INFORMATION TO THE OPERATOR FOR

NNRB018 NUCLEIS 10/15/1995 Search Proces) Adioc Report 22 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

MONITORING THE ELECTRICAL POWER SYSTEMS. INSTRUMENTATION AND CONTROLS ARE REAG ANGED TO CORRECT DISCREPANCIES IDENTIFIED BY A DETAILED CONTROL ROOM DESIGN REVIEW (DCRDR) IN THE 1980'S. THIS ACTIVITY RELOCATES EXISTING METERS ON THE METER SECTION OF THE PANELS IN ORDER FOR THE METERS TO PROPERLY ALIGN WITH THE ASSOCIATED CONTROLS ON THE BENCH SECTION OF THE PANELS. INDICATION AND CONTROL SYSTEM RELOCATION INVOLVES REMOVING NONFUNCTIONAL INSTRUPENTATION ON THE BENCH SECTIONS OF THE EACP FOR DG 21 (PANELS 1C19 AND 1C20) AND PROVIDING PATCHING FOR THE EXISTING CUTOUTS.

THE STRUCTURAL ADEQUACY AND SEISMIC QUALIFICATION OF NEW AND EXISTING SSC'S OPERABILITY OF PLANT ELECTRICAL DISTRIBUTION SYSTEMS AND CONTROL PANEL REQUIREMENTS WERE EVALUATED TO ENSURE THE PROBA8ILITY AND CONSEQUENCES OF A PREVIOUSLY EVALUATED ACCIDENTS AND MALFUNCTIONS MAVE NOT BEEN INCREASED BY THIS ACTIVITY. PRECAUTIONS ARE OBSERVED IN ORDER TO PREVENT INSTALLATION ACTIVITIES FROM INTRODUCING A NEW MALFUNCTION OR ACCIDENT DURING MODIFICATION OF THE EACP. THIS ACTIVITY DOES NOT AFFECT THE OPERABILITY OF ELECTRICAL DISTRIBUTION SYSTENS. THUS, THE MARGIN OF SAFETY AS DEFINED IN THE TECHNICAL SPECIFICATIONS IS NOT REDUCED. THEREFORE, THERE ARE NO UNREVIEWED SAFETY QUESTIONS ASSOCIATED WITH THIS ACTIVITY.

Document ID Revision Status

============________===========

94-8-041-091-R00 62

Subject:

REPLACE THE DEMINERALIZED WATER FLOW CONTROLLERS AND THE BORIC ACID FLOW CONTROLERS.

I i

1 m.

. m -

-.,,e,

NWRB018 NUCLEIS 10/15/1995 Search Process A& oc Report 23 STATUS 62 OR 64 50.595 (10/01/1996 THRU 09/30/1995)

Alias:

POSRC #:

94-163 Assoc Doc ID: 94-041-009-00 Revision To: 0000 Assoc Stat:

C Assoc Type: MCR Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xmt! Date

===================================================================================================

Other rsfs:

Pers Rifs:

Equipment:

1FIC210X 1 CVC M/U WTR FLO INDIC C 1FIC210Y BORIC ACD VOL CONT TNK 11 2FIC210X 2 CVC M/U WTR FLO INDIC C 2FIC210Y BORC ACD VOL CONT TANK 21 Org/Div:

System Code: 041 CHEMICAL & VOLUME CONTROL SYSTEM (CVCS)

Text:

SLMMARY: (FOR NRC REPORT)

THIS SAFETY EVALUATION CONCLtDES THAT THIS ACTIVITY IS NOT AN UNREVIEWED SAFETY QUESTION.

THIS MODIFICATION REPLACES THE UNIT 1 AND UNIT 2 DEMINERALIZED WATER AND BORIC ACID FLOW CONTROLLERS 1 / 2 FIC 210 X / Y WITH FISCHER & PORTER (F&P)

MICR0 DCI 53 MC 5000 SERIES FLOW INDICATING CONTROLLERS. DUE TO INCREASED CAPABILITIES OF THESE CONTROLLERS, OTHER INSTRUMENTS IN THE LOOP ARE NO LONGER REQUIRED AND ARE BEING REMOVED. THESE CONTROL LOOPS ARE NON SAFETY RELATED. THE F&P MICR DCI 53 MC 5000 SERIES CONTROLLER IS A MICROPROCESSOR BASED CONTROLLER THAT INTERFACES WITH THE FLOW TRANSMITTER AND PROVIDES THE SAME OUTPUT AS THE EXISTING CONTROL SCHEME.

THIS ACTIVITY IS NOT AN UNREVIEWED SAFETY QUESTION SINCE IT DOES NOT INCREASE THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF ANY SAR ACCIDENT OR MALFUNCTION NOR DOES IT CREATE THE POSSIBILITY OF A DIFFERENT TYPE OF ACCIDENT OR MALFUNCTION PREVIOUSLY EVALUATED IN THE SAR.

NNRB018 NUCLEIS 10/15/1995 Search Process A& oc Report 24 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

Document ID Revision Status mas=========================================

94-5-052-113-R00 62 Stbject:

CHANGE IN THE NORMAL POSITION OF THE LOW PRESSURE SAFETY INJECTION MOTOR OPERATED ISOLATION VALVES TO ALLOW MAINTENANCE ON THE MOTOR ACTUATORS WHILE THE PLANT IS IN " NORMAL OPERATING MODE".

Aliae:

POSRC #:

94-168 l

Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: C Assoc Type: ESP Ref Doc ID:

2-94-0059 Rev:

0000 Refer Type:

TMOD TEMPORARY MODIFICATIONS 2-94-0060 0000 TMOD TEMPORARY MODIFICATIONS Sender Xmtl #

Xmt! Date

========= ========================================================================================

Other refs:

Pers RIfs:

Equipment:

Org/Div:

System Code: 052 SAFETY INJECTION SYSTEM Text:

SUMMARY

(FOR NRC REPORT)

THE ACTIVITY EVALUATED IS THE TEMPORARY CHANGE IN POSITION OF THE LOW PRESSURE SAFETY INJECTION ISOLATION MOTOR OPERATED VALVES (MOV). THESE MOV'S ARE NORMALLY CLOSED. THEY OPEN ON SAFETY INJECTION ACTUATION SIGNAL (SIAS).

THESE VALVES MAY NEED MAINTENANCE WHILE THE UNIT IS AT POWER. TO PERFORM THIS MAINTENANCE, ONE VALVE AT A TIME WOULD BE OPENED AND GAGC-ED WHILE THE ACTUATOR IS SERVICED. IN THIS POSITION, THE MOV WOULD ALLOW FOR NORMAL ECCS DELIVERY ON SIAS. THE COMPANION CHECK VALVE WOULD ISOLATE LPSI ON RAS. WITH THE MOV OPEN THE POTENTIAL FOR LEAKAGE PAST THE TWO RCS CHECK VALVES AND THE LPSI CHECK VALVE EXISTS. THE CONSEQUENCES OF THIS CHANGE HAVE BEEN EVALUATED AND DETERMINED TO BE NOT SIGNIFICANT. THEREFORE, THERE IS NO CHANGE TO THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAR. NO NEW MALFUNCTIONS OR ACCIDENTS ARE CREATED AND THERE IS NO CHANGE TO THE MARGIN OF i

SAFETY.

1 im

NNR8018 NUCLEIS 10/15/1995 Search Proces2 A & oc Report 25 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

Document ID Revision Status semew==========================

==

94-2-024-088-R01 62 Stbject:

INSTALLATION OF AN UPGRADE KIT PROVIDED BY THE EDG SUPPLIER ON EDG'S NO. 11 AND 21.

Alias:

POSRC #:

94-170 Assoc Doc ID: 93-0203 Revision To: 0000 Assoc Stat: C Assoc Type: FCR Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Matt Date

===================================================================================================

Othrr rsfs:

Pars Rafs:

Equipment:

i Org/Div:

System Code: 024 EMERGENCY DIFSEL GENERATOR Text:

SupW4ARY: (FOR NRC REPORT)

THIS ACTIVITY ADDRESSES THE INSTALLATION OF AN UPGRADE KIT ON EMERGENCY DIESEL GENERATOR (EDG) NO. 11 & 21, PROVIDED BY THE EDG SUPPLIER, THAT INCLtBES REPLACING THE CYLINDER LINERS, PISTONS, SCAVENGING AIR SYSTEM, AND FUEL INJECTORS WITH COMPONENTS OF AN IMPROVED DESIGN. ALSO MODIFICATIONS OF SOME SUPPORTING SYSTEMS ARE INCLUDED. THE INSTALLATION OF THE UPGRADE KIT FACILITATES AN INCREASE TO THE ELECTRICAL CAPA8ILITY OF EDG'S NO. 11 & 21 IN THE FUTURE. WITH THE INCORPORATION OF FUTURE LOADS (PLANNED AND POTENTIAL),

THE MARGINS WITHOUT THIS UPGRADE WOULD BE UNACCEPTABLE. HOWEVER, WITH THE CAPACITY UPGRADE (FOR EDG'S No. 11 & 21) AND WITH THE FUTURE LOADS ADDED, POSITIVE LOAD MARGINS WILL BE MAINTAINED ON ALL BUSES.

THE PROBASILITY AND CONSEQUENCES OF MALFUNCTIONS AND ACCIDENTS PREVIOUSLY EVALUATED IN THE SAR ARE NOT IMPACTED BY THIS ACTIVITY BECAUSE THE RELIABILITY OF THE EDG IS NOT IMPACTED BY THIS ACTIVITY.

NWRB018 NUCLEIS 10/15/1995 Search Procesa A & oc Ceport 26 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

NO ADDITIONAL FAILURE MODES OF THE EDG ENGINE ARE BEING CREATED BY THIS ACTIVITY, AND NO NEW INTERACTIONS BETWEEN SYSTEMS ARE CREATED BY CHANGES COVERED UNDER THIS SAFETY EVALUATION. FURTHERMORE, THE EDGS ARE ACCIDENT MITIGATORS AND CANNOT BECOME AN INITIATOR OF A NEW ACCIDENT. THEREFORE, THE POSSI81LITY OF A NEW MALFUNCTION OR ACCIDENT HAS NOT BEEN CREATED BY THIS ACTIVITY.

THE QUANTITY OF FUEL OIL REQUIRED TO BE STORED (IN DAY TANKS AND FUEL DIL STORAGE TANKS) BY THE TECHNICAL SPECIFICATIONS IS SUFFICIENT SINCE THE EDG WILL CONSUME LESS FUEL AT THE EXISTING RATINGS THAN PRIOR TO THE UPGRADE.

THE OPERABILITY AND CAPABILITY OF THE SERVICE WATER SYSTEM ARE NOT ADVERSELY AFFECTED BY THIS ACTIVITY. THUS, THE MARGIN OF SAFETY AS DEFINED IN THE TECHNICAL SPECIFICATIONS IS NOT REDUCED BY THIS ACTIVITY, AND THERE ARE NO UNREVIEWED SAFETY QUESTIONS ASSOCIATED WITH THIS ACTIVITY.

THE ACTIVITIES COVERED BY THIS SAFETY EVALUATION ARE THE NECESSARY PLANT MODIFICATIONS REQUIRED TO INCREASE THE ELECTRICAL CUTPUT CAPACITY OF THE EDGS. NO INCREASE IN THE EDG'S ELECTRICAL OUTPUT CAPACITY 15 REALIZED UPON IMPLEMENTATION OF THIS ACTIVITY UNTIL DES APPROVES THE PLANT TEST RESULTS AND THE TECHNICAL SPECIFICATIONS ARE REVISED.

I L

i

NMR3018 NUCLEIS 10/15/1995 Search 1*rocesa As9HM: Report

' 27-STATUS 62 OR 66 50.595 (10/01/1996 THRU 09/30/1995)

Document ID Revision Status

================================

96-B-011-119-R00 62 S4]ect:

MCR 96 011 007 00: INSTALL SLEEVES IN INLET ENDS OF TME TUBES FOR SRWMX.

Alles:

POSRC #:

94-170 Assoc Doc ID: 96-011-007-00 Revision To: 0000 Assoc Stat: 0 Assoc Type: MCR Ref Doc ID:

Rev:

Refer Type:

Sender Matl #

Natt Date

= = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = =======-================

Other rsfs:

Pers R2fs:

Equipment:

Org/Div:

System Code: 011 SERVICE WATER COOLING Text:

NRC SUIBIARY:

THIS ACTIVITY PROVIDES THE DESIGN ENGINEERING REGUIRED TO INSTALL SLEEVES

(- 8" LONG X 0. 028' W LL) IN THE INLET ENDS OF THE TUBES FOR THE SERVICE WATER (SRW) MEAT EMCMANGERS (MM) MX'S, TME SLEEVES WILL BE ROLLED AIS EXPANDED IN PLACE WITM A FLARED INLET EDGE FLUSN WITM THE MX TUSE SMEET.

PRIOR TO INSTALLING THE SLEEVE, APPROVED ADMESIVE WILL DE APPLIED SETW EN THE OUTER DFAMETER OF TME SLEEVE Als TME IMMER DIAETER OF TME TtBE TO FORM A LEAK TIGHT JOINT.

CURRENTLY THE SRWMX'S ARE EXPERIENCING TUBE INLET (SALTWATER) E W EROSION /

CORROSION. THE INSTALLATION OF TMESE SLEEVES WILL ENTElm TME LIFE OF THE MX'S TUDES AND ALLOWS FOR LONGER PERIODS BETWEEN RETUBING.

REASON FOR 50.59 SAFETY EVALUATION: A NOTE IS BEING ADDED TO UFSAR TABLE 917 TO INDICATE TMAT St 111 70 / 30 CUNI SLEEVES MAY BE INSTALLED.

NO PROTECTIVE OR SAFETY FEATURE OF TME SRWMX'S IS ALTERED. TME MODIFICATION WILL NOT DEGRADE OR PREVENT ACTIONS DESCRIBED OR ASSLNED IN THE SAR. TMIS ACTIVITY IS CONSISTENT WITH TME REGUIREENT SOF THE ORIGINAL DESIGN CISES AND STAISARDS.

TMEREFORE, THIS ACTIVITY DOES NOT INVOLVE AN UNREVIEWED SAFETY GUESTION, NOR DOES IT REDUCE THE MARGIN OF SAFETY AS DESCRISED IN TME TECINIICAL SPECIFICATION BASES.

(CMM)

NMR9018 NUCLEIS 10/15/1995 Search Process Ae oc Report 28 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

E

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- - ~.

NINtB018 NUCLEIS 10/15/1995 Search Procesa A& oc Report 29 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995) i L

Document ID Revision Status

================================

94-B-012-118-R00 62

Subject:

INSTALL BYPASS LINE AROUND 11 & 21 ECCS PP RM AIR CLR SW INLET ISO VLVS IAW MCR 94 012 016 00 Atlas:

POSRC #:

94-170 Assoc Doc ID: 94-012-016-00 Revision To: 0000 Assoc Stat: 0 Assoc Type: MCR Ref Doc ID:

Rev:

Refer Type:

t Sender Xmtl #

Xmtl Date

============

Other rsfs:

Pers Rafs:

Equipment:

Org/Div:

[

System Code: 012 SALT WATER COOLING Text:

NRC

SUMMARY

THIS MCR PROVIDES THE DESIGN FOR INSTALLING A BYPASS LINE AROUND 11 & 21 ECCS PUMP ROOM AIR COOLER SALT WATER (SW) INLET ISOLATION VALVES 1 (2) CV 5170.

CURRENTLY THE NORMAL POSITION FOR THE INLET AND OUTLET CONTROL VALVES (CV) 0F 11 AND 21 ECCS PUMP ROOM AIR COOLERS IS SHUT. THIS SYSTEM LINE UP HAS RESULTED IN LIFTING RELIEF VALVES 1 (2) RV 5205, WHICH ARE INSTALLED BETWEEN THE CV'S AFTER CERTAIN SYSTEM TESTS.

SPECIFICALLY, RELATIVELY COLD BAY WATER IS TRAPPED BETWEEN THE CONTROL VALVES SHORTLY AFTER THE PERFORMANCE OF A SYSTEM TEST. AS THE WATER IN THE PIPING IS HEATED TO ROOM AMBIENT TEMPERATURE, THE RELIEF VALVES (RV'S) LIFT TO DISSIPATE THE ASSOCIATED PRESSURE INCREASE.

i THE BYPASS LINE, WHICH IS CONSTRUCTED MAINLY OF 1/2" TUBING, WILL ALLOW L

[

NNRB018 NUCLEIS 10/15/1995 Search Procesa Adioc Report 30 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

EXPANDING WATER TO RETURN TO THE SW SYSTEM HEADER WITMOUT HAVING TO CHALLENGE THE SYSTEM RV'S.

REASON FOR A 50.59 SAFETY EVALUATION: TA8tE 9 16A STATES THAT CV 5170 IS NORMALLY CLOSED AND AUTOMATICALLY OPENS, ALONG WITH THE COOLER'S SW DISCHARGE CV, IN ORDER TO REGULATE THE ECCS PLMP ROOM AMBIENT TE9FERATURE.

1 THIS ACTIVITY WILL NOT RESULT IN A CHANGE TO THE UFSAR'S CURRENT DESCRIPTION OF THE DESIGN, FUNCTION OR METHOD OF PERFORMING THE FUNCTION OF THE SW SYSTEM OR THE ECCS PUMP ROOM AIR COOLERS. HOWEVER, INSTALLING THE BYPASS LINE WILL ALLOW AN OPEN PATH AROUND THE INLET CV'S. THEREFORE, A NOTE IS BElhG ADDED TO TABLE 9 16A TO STATE THE PURPOSE OF THE BYPASS LINE.

NO PROTECTIVE OR SAFETY FEATURES OF TME SW SYSTEM OR THE ECCS PUMP ROOM AIR COOLERS ARE ALTERED. THE MODIFICATION WILL NOT DEGRADE OR PREVENT ACTIONS DESCRIBED OR ASSUMED IN THE SAR. THIS ACTIVTY IS CONSISTENT WITH THE REQUIREMENTS OF THE ORIGINAL DESIGN CEDES AND STANDARDS.

A THEREFORE, THIS ACTIVTY DOES NOT INVOLVE AN UNREVIEWED SAFETY QUESTION, NOR DOES IT REDUCE THE MARGIN OF SAFETY AS DESCRIBED IN THE TECHNICAL SPECIFICATION BASES.

(CMH) i l

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NNRB018 NUCLEls 10/15/1995 Fearch Procesa Adtoc Report 31 STATUS 62 OR 64 50.59s (10/01/1994 THRU 09/30/1995) 5 1

I i

NWR8018 NUCLEIS 10/15/1995 Search Proces3 Acitoc Report 32 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

Document ID Revision Status

x=================================

94-8-052-117-R00 62

Subject:

REPLACE THE SHUTDOWN COOLING FLOW CONTROLLERS.

Alias:

POSRC #:

94-170 Assoc Doc ID: 91-052-029-00 Revision To: 0000 Assoc Stat: C Assoc Type: MCR Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xmtl Date

============

Other refs:

Pers Rafs:

Equipment:

1FIC306 S/D CLG FLOW CONTROLLER 2FIC306 S/D CLG FLOW CONTROLLER Org/Div:

System Code: 052 SAFETY INJECTION SYSTEM Text:

SUMMARY

(FOR NRC REPORT)

THIS MODIFICATION REPLACES THE UNIT 1 AND UNIT 2 SHUTDOWN COOLING FLOW L

CONTROLLERS 1 / 2 FIC 306 WITH FISCHER & PORTER (F&P) MICR0 DCI 53 NC 5000 SERIES FLOW INDICATING CONTROLLERS. THESE CONTROL LOOPS ARE NON SAFETY RELATED FOR THEIR CONTROL FUNCTION BUT SAFETY RELATED FOR INDICATION ONLY.

l THE F&P MICR DCI 53 MC 5000 SERIES CONTROLLER IS A MICROPROCESSOR RASED CONTROLLER THAT INTERFACES WITH THE FLOW TRANSMITTER AND PROVIDES THE SAME OUTPUT AS THE EXISTING CONTROL SCHEME.

THIS ACTIVITY IS NOT AN UNREVIEWED SAFETY QUESTION SINCE IT DOES NOT INCREASE THE PROBASILTIY OF OCCURENCE OR THE CONSEQUENCES OF ANY SAR ACCIDENT OR MALFUNCTION NOR DOES IT CREATE THE POSSIBILITY OF A DIFFERENT TYPE OF ACCIDENT OR MALFUNCTION PREVIOUSLY EVALUATED IN THE SAR.

1 i

i t

NWRB018 NUCLEIS 10/15/1995 Search Process Adtoc Report

-33 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

Document ID Revision Status

================================

94-B-064-110-R00 62

Subject:

ESTABLISH A NEW SUBC00 LING MARGIN MONITOR TEMPERATURE MARGIN SETPOINT.

Alias:

POSRC #:

94-172 Assoc Doc ID: 94-064-013-00 Revision To: 0000 Assoc Stat: C Assoc Type: MCR Ref Doc ID:

Rev:

Refer Type:

Serder Xmtt #

Xmtl Date

=================================_____=====______=================================================

Other rifs:

Pers RIfs:

Equipment:

Org/Div:

System Code: 064 REACTOR COOLANT Text:

SumARY: (FOR NRC REPORT)

MCR 94-064-013-00 ESTABLISHES A NEW SUBC00 LING MARGIN MONITOR (S M ) ALARM SETPOINT VALUE TO ELIMINATE NUISANCE ALARMS. THESE NUISANCE ALARMS RESULT FROM NORMAL FULL POWER PLANT OPERATION TO THE RANGE OF 50 - 58 DEGREES F SUSC00 LED WHICH DVERLAPS THE ALARM SETPolNT. UFSAR SECTION 7.5.9.1 (INADEQUATE CORE C00LIN INSTRUMENTATION - SUBC00 LED MARGIN MONITOR) DEFINES A TEMPERATURE MARGIN SETPOINT OF 50 DEGREES F. MCR 94-064-013-00 DOCUMENTS THE SETPOINT BASIS FOR A REVISED ALARM SETPOINT. A UFSAR CHANGE REQUEST HAS BEEN INITIATED TO DELETE THE EXPLICIT SETPOINT VALUE REFERENCE FROM THE UFSAR.

THE SETPOINT WILL BE CONTROLLED BY THE ENGINEERING CHANGE PROCESS.

THE PROBABILITY OR CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION ARE NOT INCREASED BY THIS MODIFICATION. THIS MODIFICAITON DOES NOT CREATE THE POSSIBILITY OF AN ACCIDENT OR MALFUNCTION NOT PREVIOUSLY EVALUATFD. THIS j

• 0DIFICATION DOES NOT AFFECT ANY MARGIN OF SAFETY AS DEFINED IN THE TECHNICAL SPECIFICATION. THEREFORE, THERE ARE NO UNREVIEWED SAFETY QUESTIONS ASSOCIATED l

t 1

l

NNRB018 NUCLEIS 10/15/1995 Search Process A & oc Report 34 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

WITH THIS NODIFICATION.

Document 10 Revision Status

====================================== ______

94-2-058-061-R00 62

Subject:

EXCHANGE CABLES AT THE CONTAINNENT PENETRATION FOR EXCORE DETECTORS.

Atlas:

POSRC #:

95-003 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: O Assoc Type:

ESP Ref Doc ID:

2-94-0091 Rev:

0000 Refer Type:

TNOD TEMPORARY NODIFICATIONS Sender Xmtl #

Xmtl Date

============

OthIr refs:

P:rs RIfs:

Equipment:

Org/Div:

System Code: 058 REACTOR PROTECTIVE Text:

SupstARY: (FOR NRC REPORT)

THIS ACTIVITY EXCHANGES THE CABLES AT THE CONTAINNENT PENETRATION FOR LINEAR RANGE NUCLEAR INSTRUNENTS 2 NE 007 (SAFETY CHANNEL "C") AND 2 NE 009 (REACTOR

(

l l

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NMRB018 NUCLEIS 10/15/1995 Search Proces) A& oc Report 35 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

REGULATING CHANNEL "X").

THIS WILL EFFECTIVELY EXCHANGE THE OPERATIONAL FUNCTIONS OF THESE TWO DETECTORS WITHOUT MODIFYING THEM PHYSICALLY. THE CHANNEL "C" DETECTOR IS EXPERIENCING NOISE DEGRADATION. EXCHANGING THE TWO DETECTORS WILL ALLOW THE FULLY OPERABLE DETECTOR TO BE USED AS A SAFETY CHANNEL INPUT, AND THE DEGRADED DETECTOR TO BE USED IN THE LESS CRITICAL ROLE OF THE REACTOR REGULATING CHANNEL. THIS ACTIVITY ALSO INSTALLS JUMPERS AT THE REACTOR REGULATING CHANNEL "X" AND "Y" INPUTS TO THE REACTOR PROTECTION SYSTEM POWER RATION CALCULATOR. INSTALLATION OF THE JUMPERS WILL REMOVE THE CHANNEL "X" INPUT (DEGRADED CHANNEL "C"), AND INPUT THE CHANNEL "Y" SIGNAL TWICE TO THE POWER RATIO CALCULATOR.

THIS ACTIVITY DOES NOT CONSTITUTE AN UNREVIEWED SAFETY QUESTION, OR REQUIRE A CHANGE TO PLANT TECHNICAL SPECIFICATIONS. THE RPS AND RR$ WILL CONTINUE TO FUNCTION AS CURRENTLY DESCRIBED IN THE SAR. APD COEFFICIENTS WILL BE RECALIBRATED TO ACCOUNT FOR THE CHANGE IN DETECTOR GEOMETRY WITHIN FUNCTIONAL GROUPS. THIS ACTIVITY MAINTAINS THE DESIGN REQUIREMENTS FOR BOTH S~'TEMS AND INTRODUCES NO NEW FAILURE MODES THAT CC'JLD ADVERSELY AFFECT EQUIPMLui IMPORTANT TO SAFETY.

Docunent ID Revision Status

======

95-0003 0000 62

Subject:

TEMPORARY ALTERATION PROVIDES THE DESIGN TO INSTALL A BLIND FLANGE ON THE DISCHARGE OF 2-RV-105.

Alias:

POSRC #:

95-005 Assoc Doc 10: ES9300001 Revision To: 0000 Assoc Stat: 0 Assoc Type: ESP Ref Doc ID:

2-94-0057 Rev:

0000 Refer Type:

TMOD TEMPORARY MODIFICATIONS Sender Xmtl #

Xmtl Date

===================================================================================================

m u

m

tv NMR9018 NUCLEIS 10/15/1995 Search Proces2 A&oc Report 36 STATUS b2 OR a4 50.59S (10/01/1994 THRU 09/30/1995)

Other rsfs:

Pers RIfs:

E@ipment:

2RV105 VCT RV i

i Org/Div:

System Code: 041 CHEMICAL & VOLUME CONTROL SYSTEM (CVCS)

Text:

SupptARY: (FOR NRC REPORT)

THIS TEMPORARY ALTERATION (TA) PROVIDES THE DESIGN TO INSTALL A BLIND FLANGE ON THE DISCHARGE OF 2-RV-105.

THE RV IS BEIN REMOVED FOA MAINTENANCE DURING A PERIOD WHEN THE VCT IS NOT NEEDED FOR SYSTEM OPERATION; HOWEVER, THE RV DISCHARGE PIPING MUST REMAIN IN OPERATION AND QUALIFIED TO THE REQUIREMENTS OF THE ORIGINAL DESIGN CODE.

A BLIND FLANGE SERVES AS ISOLATION FROM TME PLANT VENT HEADER.

A 50.59 SAFETY EVALUATION IS NEEDED SINCE THIS ACTIVITY WILL RESULT IN A CHANGE TO THE SYSTEM AS DESCRIBED IN UFSAR FIGURES 9-24 AND 11-2.

THE TA WILL NOT DEGRADE OR PREVENT ACTIONS DESCRIBED OR ASSUMED IN THE SAR.

THIS ACTIVITY IS CONSISTENT WIT't THE REQUIREMENTS OF THE ORIGINAL DESIGN CODES AND STANDARDS.

THEREFORE, THIS ACTIVITY DOES NOT INVOLVE AN UNREVIEWED SAFETY QUESTION, NOR DOES IT REDUCE THE MARGIN OF SAFETY AS DESCRIBED IN THE TECHNICAL SPECIFICATION BASES.

NNRB018 NUCLEIS search Froces2 A& oc Report STATUS 62 OR 64 50.59s (10/01/1994 THRU 09/30/1995)

NNRt018 NUCLEIS 10/15/1995 Search Proces3 Ae oc Report 38 STATUS 62 OR 64 50.595 (10/01/1996 THRU 09/30/1995)

Document ID Revision Status

================================

95-0004 0000 62 Sthject:

REMOVAL OF SERVICE WATER PUMP ROOM FLOOR DRAIN CHECK VALVES DURING THE SERVICE WATER HEAT EXCHANGER TUBE i

CLEANING.

Alias:

POSRC #:

95-0G5 Assoc Doc ID: SRWHX-02 Revision To: 0000 Assoc Stat: C Assoc Type: MP Ref Doc 10:

Rev:

Refer Type Sender Xmtl #

Xmtl Date

============

Othir refs:

Pers hfs:

Equipment:

Org/Div:

System Code:

Text:

SUMMARY

(FOR NRC REPORT)

THIS EVALUATION IS TO JUSTIFY THE REMOVAL OF SERVICE WATER PUMP ROOM FLOOR DRAIN CHECK VALVES DUR*G THE SERVICE WATER HEAT EXCHANGER TUBE CLEANING IN l

ACCORDANCE WITH TECH W AL PROCEDURE SRWHX-02, SERVICE WATER HEAT EXCHANGER TUBE CLEANING. THIS 2?ALUATION IS TO PROVIDE JUSTIFICATION FOR THE REINSTALLATION OF THE CHECK VALVES WHEN FLOODING OCCURS IN THE TUR81NE BUILDING DURING THE PERFORMANCE OF TECHNICAL PROCEDURE SRWHX-02.

l THERE ARE NO U4 REVIEWED

• SAFETY QUESTIONS ASSOCIATED WITH THIS ACTIVITY.

I

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NNRB018 NUCLEIS 10/15/1995 Search croces3 A&oc Report 39 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

Document ID Revision Status

================================

94-8-052-109-R00 62

Subject:

THIS ACTIVITY INVOLVES A NON MODIFICATION CHANGE TO SECTIONS 6 3 2 1 AND 6 3 2 2 0F THE UFSAR AS A RESULT OF REVIEWS DONE UNDER ISSUE REPORT IRO 035 352.

Alias:

POSRC #:

95-012 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: C Assoc Type: ESP Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xmtl Date

========================================================================================_==========

Other refs:

Ptra R fs:

Equipment:

Org/Div:

System Code: 052 SAFETY IWJELTION SYSTEM Text:

SUPMARY:

THIS ACTIVITY IS DONE IN SUPPORT OF ISSUE REPORT IRO 035 352, AND ADDRESSES NON MODIFICATION CHANGES TO SECTIONS 6.3.2.1 AND 6.3.2.2 OF THE OFSAR. THESE CHANGES CLARIFY THE EXTERNAL COOLING REQUIREMENTS FOR THE HPSI AND LPSI PUMPS AND BEARINGS. THE PRESENT UFSAR DESCRIPTION IMPLIES THAT THE PUMP SEALS CAN OPERATE FOR EXTENDED PERIODS AT PUMP FLOW TEMPERATURES OF 300 DEGREES F WITH-OUT COOLING FROM THE CCW SYSTEM. TO THE CONTRARY, DISCUSSIONS WITH THE PUMP VENDORS INDICATE THE SEALS CAN ONLY OPERATE FOR TWO HOURS WITHOUT THE SEAL FLOW BEING COOLED BY EXTERNAL MEANS (I.E., COMPONENT COOLING WATER) FOR PUMP FLOW TEMPERATURES OF 250 DEGREES F AND 300 DEGREES F FOR THE HPSI AND LPSI PUMPS RESPECTIVELY.

A REVIEW OF THE ORIGINAL PUMP SPECIFICATION SHOWS THAT IT WAS NEVER INTENDED FOR THESE PUMPS TO OPERATE FOR MORE THAN TWO HOURS WITHOUT EXTERNAL COOLING.

ALSO, IN TABLE 9-17 0F THE UFSAR IT IS SEEN THAT THE M*'SI PUMP IS ONLY

NNRB018 NUCLEIS 10/15/1995 Search Procesa A & oc Report 40 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

REQUIRED TO OPERATE FOR TWO HOURS AFTER A POST RAS PASSIVE FAILURE OF THE COMPONENT COOLING WATER SYSTEM. TWO HOURS ALLOWS SUFFICIENT TIME FOR THE OPERATORS TO ALIGN THE AIR COOLED CONTAINMENT SPRAY PUMPS FOR CORE INJECTION OPERATION. THE LPSI PUMPS ARE SECURED AFTER A RAS.

THE RADIATION LEVELS TWO HUURS AFTER A LBLOCA WITH A LOSS OF CCW HAVE BEEN COMPUTED AND ARE ACCEPTABLE TO ALLOW AN OPERATOR TO ACCESS THE REQUIRED VALVE TO ALIGN THE CONTAINMENT SPRAY PUMP FOR CORE COOLING OPERATION. THEREFORE, THE PROBAFILTIY OF MALFUNCTION IS NOT INCREASED. SINCE NO FIELD OR PROCEDURE CHANGES ARE OCCURRING THE CONSEQUENCES OF MALFUNCTION IS NOT INCREASED AND THE POSSIBLITY OF A NEW MALFUNCTION IS NOT CREATED. SINCE THE HPSI AND LPSI PUMPS ARE USED ONLY AFTER AN ACCIDENT HAS BEEN INITIATED THEY DO NOT CONTRIBUTE TO THE PROBABILITY OF AN ACCIDENT OR THE POSSIBLITY OF A NEW TYPE OF ACCIDENT. FINALLY, SINCE, THE ABOVE REVIEW HAS SHOWN THAT IT IS NOT REQUIRED THAT THESE PUMPS OPERATE FOR EXTENDED PERIODS WITHOUT EXTERNAL COOLING THE CONSEQUENCES OF A LOCA ARE NOT AFFECTED AND THEREFORE THIS ACTIVITY DOES NOT INVOLVE AN UNREVIEWED SAFETY QUESTION.

(CMH) k b

NWRB018 NUCLEIS 10/15/1995.

Search Proces) A & oc Report 41 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995).

Document ID Revision Status

================================

94-B-999-082-R00 62

Subject:

MODIFIES EXISTING ELECTRICAL DISTRIBUTION SYSTEM IN ORDER TO CONNECT A NEW SR DG.

Atlas:

POSRC #:

95-013 Assoc Doc ID: 89-0079 Revision To: 0000 Assoc Stat:

Assoc Type:

FCR Ref Doc ID:

Rev:

Refer Type:

Sender Xmt! #

Xmtl D4te

=========_=========================================================================================

Other rsfs:

Pers RIfs:

Equipment:

Org/Div:

System Code:

Text:

SUMMARY

(FOR NRC REPORTS)

IN ORDER TO CONNECT DG 1A TO EMERGENCY BUS 11, THIS ACTIVITY DISCONNECTS EMERGENCY DIESEL GENERATOR 11 (DG 11) FROM EMERGENCY BUS 11 FROM ITS NORMALLY CLOSED DISCONNECT SWITCH (DISCONNECT SWITCH 1103) AND FROM THE CIRCUIT BREAKER CUBICLE AT THE BUS. POWER CABLING FOR DG 1A WILL THEN BE CONNECTED TO THE CIRCUIT BREAKER AT EMERGENCY BUS 11. DG 1A WILL NOT BECOME OPERATIONAL UNTIL TESTING IS COMPLETED. IN SUPPORT OF THE DG 1A TIE-IN THIS ACTIVITY ALSO ADDS RACEWAYS IN THE UNIT 1 ELECTRICAL SWITCHGEAR ROCM TO CONNECT DG 1A TO EMERGENCY BUS 11 AND INSTALLS / CONNECTS WIRING BETWEEN THE ELECTRICAL AUXILIARY CONTROL PANEL (EACP) AND THE DG CONTROL / CONSOLE (DGCC) FOR DG 1A INSTRUMENTATION, ANNUNCIATION AND CONTROLS. IN AD0! TION, THIS ACTIVITY RE DESIGNATES EMERGENCY DG 11 AS EMERGENCY DIESEL GENERATOR 2A, THEREAFTER REFERRED TO AS DG 2A, AND MODIFIES THE EXISTING ELECTRICAL DISTRIBUTION SYSTEM IN ORDER TO COMPLETE THE DEDICATION OF DG 2A TO AN ENGINEERED SAFETY FEATURES BUS IN UNIT 2 ANC TRANSFER THE INDICATIONS, ANNUNCIATION AND CONTROLS FOR DG 2A AND BREAKER CONTROLS FOR EMERGENCY BUS 21 FROM THE EACP TO THE DGCC. THE INTERNAL WIRING IN THE EACP FOR DG 2A WILL BE DISCONNECTED.

IN ORDER TO CONNECT DG 1A S AUTOMATIC START AND LOADING CIRCUITS TO THE i

PLANT, THIS ACTIVITY WILL REMOVE THE UNIT 2 AUTOMATIC START SIGNALS (SIAS AND BUS UNDERVOLTAGE) FROM DG2A. THESE SIGNALS WILL BE CONNECTED TO DG 1A TO AUTOMATICALLY START DG 1A UPON RECEIPT OF A SIAS OR, START AND LOAD DG 1A ON RECIEPT OF A BUS UNDERVOLTAGE ESFAS SIGNAL.

I IN ORDER TO DEDICATE DG 2A TO UNIT 2, THIS ACTIVITY DEDICATES SERVICE WATER COOLING FOR DG 2A To UNIT 2 SERVICE WATER SUBSYSTEM.

i THIS ACTIVITY WILL BE PERFORMED DURING A UNIT 1 PLANT OUTAGE IN MODE 5 OR 6 OR DEFUELED. THE DESIGN INSTRUCTIONS IDENTIFY PORTIONS OF THIS ACTIVITY

+

NWRB018 NUCLEIS 10/15/1995 Search ProcesO A& oc Report C2 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

WHICH MAY BE PERFORMED DURING NON-0UTAGE CONDITIONS. IT IS EXPECTED THAT UNIT 2 WILL OPERATE IN MODES 1, 2, 3, 4, 5, OR 6 OR DEFUELED. IN MODE 6, AT LEAST 23 FEET OF WATER WILL BE MAINTAINED OVER IRRADIATED FUEL ASSEMBLIES SEATED WITHIN THE REATOR PRESSURE VESSEL. WORK WITHIN THE EACP WILL NOT BE PERFORMED WHEN THE PLANT IS IN A TECHNICAL SPECFICATION LCO ACTION STATEMENT FOR ANY OF THE EDGS, THEIR ASSOCIATED EMERGENCY BUSES OR THE OFFSITE POWER SOURCES (I.E. TECHNICAL SPECIFICATION 4.8.1.1 AND 4.8.1.2).

NEW SSC'S ADDED SY THIS ACTIVITY HAVE BEEN EVAULATED TO ENSURE THE EFFECT OF THEIR INSTALLATION (E.G., SEISMIC ADEQUACY OF EXISTING STRUCTURES, HEAT LOADS, CABLE SEPARATION) DO NOT INCREASE THE PROBABILITY OF PREVIOUSLY EVALUATED MALFUNCTIONS. SSC'S ADDED BY THIS ACTIVITY WILL NOT BECOME OPERATIONAL UNTIL TESTING OF DG 1A IS COMPLETE. EQUIPMENT IDENTIFIED AS INITIATORS OF ACCIDENTS ARE NOT AFFECTED BY THIS ACTIVITY. THEREFORE, THE PROBABILITY OF PREVIOUSLY EVALUATED MALFUNCTIONS AND ACCIDENTS HAS NOT BEEN INCREASED.

THE CONSEQUENCES OF PREVIOUSLY EVALUATED MALFUNCTIONS AND ACCIDENTS HAVE NOT BEEN INCREASED BY THIS ACTIVITY BECUASE EQUIPMENT REQUIRED TO SERVE MITIGATION FUNCTIONS UNDER THESE CONDITIONS HAVE NOT BEEN AFFECTED, AND CCMTROL ROOM AND OFF SITE DOSES PREVIOUSLY CALCULATED REMAIN WITHIN THE PREVIOUSLY STATED LIMITS.

AN EVALUATION WAS PERFORMED TO ASSESS THE POSSIBILITY OF AN INSTALLATION ERROR IN THE EACP WHICH COULD RESULT IN LOSS OF AN EDG OR ENGINEERED SAFETY FEATURE BUS OR THAT COULD CAUSE A PLANT TRIP IN THE OPERATING UNIT. NO NEW SYSTEMS INTERACTIONS ARE BEING CREATED BY THIS ACTIVITY. THERFORE, THE POSSIBILITY OF A NEW MALFUNCTION OR ACCIDENT IS NOT CREATED BY THIS ACTIVITY.

THE MARGIN OF SAFETY EXPRESSED IN THE BASES OF THE TECHNICAL SPECIFICATIONS IS NOT REDUCED BECAUSE COMPLETION OF THIS ACTIVITY WILL RESULT IN TWO OPERATIONAL EDGS FOR EACH UNIT. PRIOR TO IMPLEMENTING THIS ACTIVITY, AN NRC APPROVED EXTENSION OF THE SEVEN DAY LIMITATION OF ACTION STATEMENTS AND OF TECHNICAL SPECIFICATION 3.7.6.1 WILL BE REQUIRED.

THEREFORE, THERE ARE NO UNREVIEWED SAFETY QUESTIONS ASSOCIATED WITH THIS ACTIVITY.

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NWRB018 NUCLEIS 10/15/1995 Search Procesa Advoc Report 43 STATUS 62 OR 64 50.59S (10/01/1994 TNRU 09/30/1995) f 3

Y k

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Doctanent 10 Revision Status 3

================================

95-0018 0000 62

Subject:

INSTALLATION OF LEAK REPAIR DEVICES ON NSR MAIN STEAM PIPING.

1 Atlas:

POSRC #:

95-014

{

Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: O Assoc Type: ESP Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xartl Date

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NNRB018 NUCLEIS 10/15/1995' Search Proces2 A & oc Report 44 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

Other rsfs:

Pers Rifs:

Equipment:

Org/Div:

System Code: 083 MAIN STEAM Text:

SUMMARY

(FOR NRC REPORT)

UFSAR CHAPTER 10.1.3 INDICATES THAT " THE COMPONENTS OF THE MAIN STEAM SYSTEM ARE CONVENTIONAL AND OF THE TYPE THAT HAVE BEEN EXTENSIVELY USED IN FOSSIL FUEL PLANT AND IN OTHER NUCLEAR POWER PLANT." CLEARLY LEAK REPAIR CLAMPS ARE NOT CONVENTIONAL COMPONENTS WHICH ARE USED FOR SYSTEM DESIGN.

THIS ACTIVITY ADDRESSES THE INSTALLATION OF LEAK REPAIR DEVICES ON NSR MAIN STEAM PIPING. THIS EVALUATION IS WRITTEN GENERICALLY TO ADDRESS ALL LEAK REPAIR DEVICES THAT ARE PLACED ON NSR MAIN STEAM PIPING.

THE USE OF LEAK REPAIR CLAMPS ARE USED THROUGHOUT THE FOSSIL AND NUCLEAR INDUSTRY AND THE FOSSIL INDUSTRY. ALTHOUGH THE CLAMPS ARE NOT

" CONVENTIONAL" DESIGN COMPONENTS, THEY ARE DESIGNED IN ACCORDANCE WITH THE RULES AND REQUIREMENTS OF THE ORIGINAL CONSTRUCTION CODE TO ASSURE THAT THERE WILL BE NO IMPACT ON STRUCTURAL INTEGRITY. FURTHERMORE, THE CLAMP IS NOT BEING USED TO SUPPLEMENT OR MAINTAIN STRUCTURAL INTEGRITY OF THE SYSTEM, ITS SOLE PURPOSE IS TO ACT AS A LEAK LIMITING DEVICE AND PREVENT FURTHER DEGRADATION OF STRUCTURAL COMPONENTS BY STEAM CUTTING.

THIS ACTIVITY DOES NOT CREATE AN UNREVIEWED SAFETY QUESTION AS DEFINED BY 10CFR50.59.

Document ID Revision Status

================================

94-B-064-110-R01 62

Subject:

MCR 94 064 013 01 ESTABLISHES A NEW SUBC00 LING MARGIN MONITOR TEMPERATURE MARGIN SETPOINT Alias:

POSRC #:

95-015 Assoc Doc ID: 94-064-013-00 Revision To: 0000 Assoc Stat: C Assoc Type: MCR Ref Doc ID:

Rev:

Refer Type:

Sender Xmtt #

Xmtl Date

ri==

+

Othir refs:

Pws RIfs:

Equipment:

Org/Div:

System Code: 064 REACTOR COOLANT

._m NWRB018 Search Proces3 A& oc Report 45 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

Text:

SUMARY:

MCR 94 064 013 01 ESTABLISHES A NEW SUBC00 LING MARGIN MONITOR (SM) ALARM SETPOINT VALUE TO ELIMINATE NUISANCE ALARMS. THESE NUISANCE ALARMS RESULT FROM NORMAL FULL POWER PLANT OPERATION IN THE RANGE OF $0 - 58 DEGREE F SUB-COOLED WHICH DVERLAPS THE ALARM SETPOINT. UFSAR SECTION 7.5.9.1 (INADEQUATE CORE COOLING INSTRUMENTATION - SUBCOOLED MARGIN MONITOR) DEFINES A TEMPERATURE MARGIN SETPOINT OF 50 DEGREES F. MCR 94 064 013 01 DOCUMENTS THE SETPOINT BASIS FOR A REVISED ALARM SETPOINT. A UFSAR CHANGE REQUEST HAS BEEN INITIATED TO DELETE THE EXPLICIT SETPOINT VALUE REFERENCE FROM THE UFSAR.

THE PROBABILITY OR CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION ARE NOT INCREASED BY THIS MODIFICATION. THIS M00lFICATION DOES NOT CREATE THE POSSIBILITY OF AN ACCIDENT OR MALFUNCTION NOT PREVIOUSLY EVALUATED. THIS MODIFICATION DOES NOT AFFECT ANY MARGIN OF SAFETY AS DEFINED IN THE TECHNICAL SPECIFICATION. THEREFORE, THERE ARE NO UNREVIEWED SAFETY QUESTIONS ASSOCIATED WITH THIS MODIFICATION.

(CMH)

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NNRB018 NUCLEIS 10/15/1995 Search Process A & oc Report 46 Sit"JS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

Document ID Revision Status

w==w=======================

==

95-0007 0000 62

Subject:

THIS MCR PROVIDES THE DESIGN FOR REROUTING THE DISCHARGE PIPING FROM THE UNIT 1 TURBINE BUILDING SAMPLE SINK TO A NEARBY FLOOR DRAIN Alias:

POSRC 8:

95-017 Assoc Doc ID: 94-076-001-00 Revision To: 0000 Assoc Stat:

Assoc Type: MCR

[

Ref Doc ID:

Rev:

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Sender Xmtl #

Xmtl Date

===================================================================================================

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P:;rs RIfs:

Equipment:

Org/Div:

System Code: 076 SECONDARY SAMPLE Text:

SUMMARY

(FOR NCR REPORT)

THIS MCR PROVIDES THE DESIGN FOR REROUTING THE DISCHARGE PIPING FROM THE UNIT 1 TUR8tNE BUILDING SAMPLE SINK TO A NEARBY FLOOR DRAIN.

THE SAMPLE SINK DISCHARGE LINES ARE ROUTED FROM THE 12' ELEVATION UP TO A 22' ELEVATION AND THEN SACK DOWN TO 11 MISCELLANEOUS DRAIN TANK. THE VERTICAL PIPING RUN IS CREATING BACKPRESSURE ON THE SYSTEM WHICH IS HINDERING THE OPERATION OF THE NOTWELL SAMPLING PORTION OF THE SINK AND PLACING UNNECESSARY DEMANDS ON THE HOTWELL PUMPS. CURRENTLY TEMPORARY ALTERATION (TA) 1 94 10 (UNIT 1 ONLY) HAS REROUTED THE HOTWELL SAMPLE DISCHARGE LINES TO A NEARBY TURBINE BUILDING FLOOR DRAIN.

l A 50.59 SAFETY EVALUATION IS NEEDED SINCE THIS ACTIVITY WILL RESULT IN A CHANGE TO THE SYSTEM AS DESCRIBED IN UFSAR FIGURES 9 - 30.

THIS MCR WILL NOT DEGRADE OR PREVENT ACTIONS DESCRIBED OR ASSUMED IN THE SAR.

THIS ACTIVITY IS CONSISTENT WITH THE REQUIREMENTS OF THE ORIGINAL DESIGN CODES AND STANDARDS.

THEREFORE, THIS ACTIVITY DOES NOT INVOLVE AN UNREVIEWED SAFETY QUESTION, NOR DOES IT REDUCE THE MARGIN OF SAFETY AS DESCRIBED IN THE TECHP.ICAL SPECIFICATION 8ASES.

i

i NNRs018 NUCLEIS 10/15/1995 Search Procesa A& oc Report 47 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995) l f

Document ID Revision Status

================================

95-0025 0000 62

Subject:

REPLACENENT PARTS FOR THE PERSONNEL AIRL LOCK (PAL) WHICH ARE NOT STAMED IN ACCORDANCE WITH THE ASME 80!LER AND PRESSURE VESSEL CODE SECTION III.

I Alias:

POSRC #:

95-017 i

+

Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: O Assoc Type:

ESP Ref Doc ID:

Rev:

Refer Type:

Sender Xmtt #

Xmtt Date

======================================================_============================================

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I l

---m

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__-_.-__r_____--_a-_-,._w____m__...__c___-_______-____

NMRB018 NUCLEIS 10/15/1995 Search Proces3 A&oc Report 48 STATUS 62 OR 64 50.59s (10/01/1994 THRU 09/30/1995)

Equipment:

Org/Div:

System Code:

Text:

SUMMARY

(FOR NRC REPORT)

UFSAR CHAPTER 5.1.B.A AND 5.1.D INDICATE THAT THE PAL IS DESIGNED AND REQUIRED TO BE TESTED TO SAME SECTION III CLASS B REQUIREMENTS. THE ORIGINAL DESIGN SPECIFICATION FOR THE PAL INDICATES THAT PAL APPLICABLE YEAR OF THE CODE IS THE 1968 EDITION. THE 1968 EDITION OF ASME SECTION III (AND IN ALL LATER EDITIONS OF SECTION III), ARTICLES 15 AND 8 REQUIRE THAT:

(N - 814) THOSE PARTS OF NUCLEAR VESSELS REQUIREING INSPECTION UNDER THIS SUBSECTION, WHICH ARE FURNISHED BY OTHER THAN THE SHOP OF THE MANUFACTURER RESPONSIBLE FOR THE COMPLETED VESSEL SHALL APPLY THE CODE N - SYMBOL, AS SHOWN IF FIG. N - 811 (B).

THE REPLACEMENT HINGE PIN AND HINGE ARM ASSEMBLY ARE BEING FABRICATED FROM MATERIAL BY BGE AS OPPOSED TO THE ORIGINAL EQUIPMENT MANUFACTURER (CHICAGO BRIDGE IRON). AS SUCH THE ABOVE PARAGRAPH FROM THE CODE WDJLD REQUIRE THAT THE PARTS OR ASSEMBLIES BE STAMPED. THEREFORE, THIS 50.59 SAFETY EVALUATION IS REQUIRED TO ALLOW BGE TO FABRICATE AND INSTALL THESE PARTS WITHOUT MEETING THE STAMPING REQUIREMENTS. IN ADDITION, THE ASME SECTION XI REQUIREMENTS AND POTENTIAL 10CFR50.55A REQUIREMENTS HAVE BEEN REVIEWED WITH NO RULES WHICH ARE APPLICABLE TO THIS ACTIVITY. NOTE THE PAL IS CLASSIFIED AS CLASS NC AND AS SUCH IS NOT REQUIRED TO BE INCLUDED IN THE SECTION XI PROGRAM.

IN THIS INSTANCE, BGE HAS THE DETAILED FABRICATION AND MACHINING DRAWINGS FOR THE HINGE ARM ASSEMBLY AND THE HINGE PIN. IN ADDITION, ALL MATERIALS AND INSPECTION REQUIREMENTS ARE ALSO CALLED OUT IN DETAIL. ALL MATERIALS BEING USED WILL ASSUREDLY COMPLY WITH BGE'S APPENDIX B PROGRAM AND WILL BE MATERIAL WHICH IS APPROVED BY THE CODE AND FABRICATED IN ACCORDANCE WITH ASME CODE SPECIFICATIONS. ALL FABRICATION WILL BE CARRIED OUT UNDER THE QA PROGRAM AND WILL BE INSPECTED BY OUR AUTHORIZED NUCLEAR INSPECTOR.

WITH ALL OF THESE CONTROLS IN PLACE, EQUIVALENT ASSURANCE EXISTS THAT ALL REQUIREMENTS GERMANE TO THIS ACTIVITY WILL BE PERFORMED TO A SUFFICIENT LEVEL OF QUALITY AS TO COMPENSATE FOR THE NEED TO HAVE THE PARTS STAMPED.

THIS ACTIVITY DOES NOT REPRESENT AN UNREVIEWED SAFETY QUESTION.

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NMR9018 NUCLEIS 10/15/1995 Search Proces) A& oc Report 49 L

STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995) i Docunent 10 Revision Status

================================

94-B-999-120-R00 62

Subject:

REVISE FUEL HANDLING INCIDENT ANALYSIS PRESENTED IN CHAPTER 14.18 OF THE UFSAR.

5 Alias:

POSRC #:

95-019 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: C Assoc Type: ESP Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xatt Date l

===================================================================================================

Other rsfs:

Pers Rafs:

Equipment:

Org/Div:

System Code:

Text:

SUMMARY

(FOR NRC REPORT)

THIS ACTIVITY REVISES CHAPTER 14.18 0F THE UFSAR. ACCOMPANYING CHANGES TO THE UFSAR REVISE CHAPTER 14.18 TO DESCRIBE THE RESULTS OF ANALYSIS OF A FUEL HANDLING INCIDENT IN CONTAINMENT WITH BOTH PAL 000RS OPEN. THIS ANALYSIS WAS PREVIOUSLY REVIEWED AND APPROVED BY THE LICENSE AMENDMENT (TECH. SPEC. CHANGE) PROCESS. THIS SAFETY EVALUATION APPLIES THE METHODOLOGY USED FOR THE CONTAINMENT ANALYSIS TO THE CASE OF A FUEL HANDLING INCIDENT IN THE SPENT FUEL POOL. NO CHANGES ARE MADE TO ANY FUEL HANDLING EQUIPMENT, a

NNR8018 NUCLEIS 10/15/199*

Search Proces3 Advoc Report

?J STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

+

PROCEDURES, OR ADMINSTRATIVE CONTROLS OF FUEL HANDLING ACTIVITIES. THE RESULTS OF THIS EVALUATION DEMONSTRATE THAT THE OFF-SITE DOSE CONSEQUENCES OF A FUEL HANDLING ACCIDENT REMAIN BELOW THOSE PREVIOUSLY REVIEWED AND ACCEPTED BY THE NRC.

t I

r Docunent ID Revision Status

================================

95-0009 0000 62

Subject:

ALLOW REMOVAL OF EACH UNIT 2 CVCS CHARGING PUMP SUCTION LINE RELIEF VALVE AND ALLOW INSTALLATION OF BLIND FLANGE ON THE REMAINING RV DISCHARGE PIPING FLANGE OF THE WPS SYSTEM.

Atlas:

i POSRC #:

95-019 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: 0 Assoc Type: ESP Ref Doc ID:

2-95-0002 Rev:

0000 Refer Type:

TMOD TEMPORARY MG)!FICATIONS 2-95-0003 0000 TMOD TEMPORARY MODIFICATIONS 2-95-0004 0000 TMOD TEMPORARY MODIFICATIONS Sender Xmtl #

Xmtl Date

===================================================================================================

m Other refs:

t i

NMR3018 NUCLEIS 10/15/1995 Search Process A&oc Report 51

. STATUS 62 OR 64 50.595 (10/01/1994 TIntu 09/30/1995)

Pers Rifs:

EcPJipment:

2Rv315 21 CMG PP SUCT RV 2RV318 22 CMG PP SUCT RV' 2Rv321 23 CMG PP SUCT RV

.Org/Div:

Systent Code:

Text:

SUW4ARY: (FOR NRC REPORT)

THIS SAFETY EVALUATION ADDRESSES TlutEE TEWORARY ALTERATION ACTIVITIES TO ALLOW REMOWAL OF EACM UNIT 2 CV;;S CHARGING PLNEP SUCTION SIDE RELIEF VALVE (2 - RV - 315, 318, 321) AND ALLOW INSTALLATION OF A BLIND FLANGE AT THE OUTLET PIPE FLANGk FOR THE SUBJECT RV.

THE THREE TAS ARE:

TA 2 95 0002 2-RV-318 822 CVCS CMARGlWG PUMP SUCTION RV TA 2 95 0003 2-RV-321 #23 CVCS CMARGING PUMP SUCTION RV TA 2 95 0006 2-RV-315 821 CVCS CMARGING PUMP SUCTION RV THE SUBJECT RV WILL BE REMOVED FOR MAINTENANCE AND THE BLIND WILL BE INSTALLED TO PREVENT THE RELEASE OF WATER (Alm RADIO GASES) FRGE THE CGeq0N RV OUTLET MEADER DOWIISTREAlt OF SUBJECT RV.

TME ASSOCIATED CMARGING Pt3F.

WILL DE OUT OF SERVICE (SAFETY TAGGED AIS ISOLATED) FOR TME DURATION OF THIS TA, WMILE THE OTHER CMARGING PU N S REIIAIN IN SERVICE.

THE CHARGING PU W SUCTION RV PROVIDES THEllMAL OVERPRESSURE PROTECTION FOR THE PIPING Alm CGEPONENTS AT THE SUCTION SIDE OF THE CHARGING PWIP. THE RV DISCHARGES TO THE WASTE Pft0 CESSING SYSTEM (WPS) VIA A CGOION MEADER TIED TO THE OUTLET OF THE OTHER UIIIT 2 CHARGING Pt#EPS SUCTICII RVS.

THE PIPING AT THE INLET AND OUTLET OF TME RV IS CLASS MC-2 AND IS ANSI B31.7 CLASS 3 DESIGII. THE PIPIIIG AT THE INLET IS SR-PS PER THE Q-LIST AND TME OUTLET PIPIIIG IS AG-WS.

THE OUTLET PIPIIIG IS MSR EMCEPT THAT IT IS DESIGIIED SEISMIC CLASS I.

ALL DESIGIl REQUIREMIITS OF THE WPS SYSTEM PIPING ARE ET, TIIE REMAINING CVCS AND WPS PIPING IS ADEQUATELY SUPPORTED AND MEETS SEISMIC REQUIREIENTS, AND THERE ARE 100 IIIPACTS TO OTHER PLANT SYSTEftS. THER AltE 100 AFFECTS ON AIIALY2ED MALFUIICTICIIS OR ACCIDENTS AND 110 IIEW MALFUIICTICIIS OR ACCIDENTS ARE CREATED. THEREFORE, THIS ACTIVITY DOES Of CONSTITUTE A USG.

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NNRS018 NUCLEIS 10/15/1995' Search Process Ah Report 52 STATUS 62 OR 64 50.59S (10/01/1994 TMau 09/30/1995)

Docunent ID Revision Status

==

95-0010 0000 62

Subject:

WILL INCORPORATE THE RESULTS OF CALCULATION 95 - 0028, EVALUATION OF H0DE 5 CLOSURE HEAD DETENSIONING FOR LMIT 1 & 2 REACTOR VESSELS, REV. O INTO THE UFSAR.

Alias:

POSRC #:

95-019 Assoc Doc ID: 95-0028 Revision To: 0000 Assoc Stat:

Assoc Type: DCALC Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xmtl Date

===========================================================================

=

Other refs:

Pcts Rtfs:

NNRB018

>UCLEIS 10/15/1995 Search Procest A s oc Report 53 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

Equipment:

Org/Div:

System Code:

Text:

SUMMARY

(FOR NRC REPORT)

THIS ACTIVITY WILL INCORPORATE THE RESULTS OF CALCULATION 95 - 0023, EVALUATION OF MODE 5 CLOSURE HEAD DETENSIONING FOR UNIT 1 & 2 REACTOR VESSELS, REV. O DATED 2/15/95, INTO THE UFSAR, WHICH ANALYTICALLY DEMONSTRATES THE ABILITY TO SAFELY DETENSION AND REMOVE 2/3 'S OF THE REACTOR VESSEL HEAD STUDS WHILE IN MODE 5 AND ALLOW FOR FUTURE REDUCTIONS IN OVERALL REFUELING OUTAGE DURATION. EIGHTEEN OF THE TOTAL 54 STUDS REMAIN IN PLACE (EVERY THIRD STUD), TENSIONED TO A PRE-LOAD EQUAL TO 757,OF DESIGN PRELOAD, WITH RCS PRESSURE EQUAL TO 500 PSIA. TWO LOADING CONDITIONS ARE CONSIDERED: (1) RCS TEMPERATURE IS SET EQUAL TO THE SATURATION TEMPERATURE 467 DEGRESS FARENHEIT, SIMULATING THE WORST CASE (HIGHEST) TEMPERATURE, WHICH COULD OCCUR AS A RESULT OF LOSS OF SHUTDOWN COOLING, AND (2) RCS TEMPERATURE EQUAL TO 200 DEGREES FARENHEIT, WHICH IS THE NORMAL MAXIMUM TEMPERATURE FOR MODE 5 OPERATION. THE ANALYSIS INDICATES THAT THE REACTOR VESSEL 0-RINGS REMAIN IN COMPRESSION DURING BOTH CONDITIONS (NO LEAKAGE), AND THE STUD STRESSES DUE TO DETENSIONING ARE IN MODE 5 ARE LESS THAN ALLOWABLE. CALCULATED STRESSES

,t WITHIN THE FLANGES WHICH ARISE FROM MODE 5 DETENSIONING ARE BOUNDED BY THE STRESSES FROM THE ORIGINAL DESIGN. IT WAS ALSO DETERMINED THAT THE MINIMUM NUMBER OF STUDS REQUIRED TO PROPERLY SEAT THE VESSEL HEAD IN MODE 5 IS 12 BUT 18 IS USED FOR CONSERVATISM.

THIS ACTIVITY DOES NOT RESULT IN AN UNREVIEWED SAFETY QUESTION BECAUSE:

1) REACTOR VESSEL PRESSURE WILL BE MAINTAINED EQUAL TO OR LESS THAN 500 PSIA BY ADMINSTRATIVELY CONTROLLING RCS VENT PATHS IN THE MOST LIMITING CONDITION SPECIFIED BY THE TECHNICAL SPECIFICATIONS;

2) THE DESIGN BASIS EVENTS IN THE LOW TEMPERATURE REGION ARE UNAFFECTED BY THIS CHANGE;

3) ALL CREDIBLE PASSIVE MECHANICAL FAILURE MODES AND ASSOCIATED EVENTS ARE ALREADY ANALYZED IN THE SAR; AND s

4) THE MARGINS OF SAFETY ASSOCIATED WITH LTOP AS DEFINED IN THE TECHNICAL SPECIFICATIONS ARE MAINTAINED.

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NNRB018 NUCLEIS 10/15/1995 Search Proces3 A& oc Report 54 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

Doctment ID Revision Status

================================

95-0026 0000 62

Subject:

INSTALLATION OF AN UPGRADE KIT PROVIDED BY THE EMERGENCY DIESEL GENERATOR EDG SUPPLIER (FAIRBANKS MORSE [FM] /

COLTEC) ON EDG NO. 12.

Atlas:

POSRC #:

95-020 Assoc Doc ID: 93-0203 Revision To: 0000 Assoc Stat: C Assoc Type:

FCR Ref Doc ID:

Rev:

Refer Type:

Sender Matl #

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============

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Text:

SupmARY: (FOR NRC REPORT)

THIS fCTIVITY ADDRESSES THE INSTALLATION OF AN UPGRADE KIT ON EMERGENCY DIESCL GENERATOR (EDG) NO. 12, PPOVIDED BY THE EDG SUPPLIER, THAT INCLUDES k

NMRB018 NUCLEIS 10/15/1995 Search Proces3 Advoc Report 55 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

REPLACING THE CYLINDER LINERS, PISTONS, SCAVENGING AIR SYSTEM, AND FUEL INJECTORS WITH COMPONENTS OF AN IMPROVED DESIGN. ALSO MODIFICATIONS OF SOME SUPPORTING SYSTEMS ARE INCLUDED. THE INSTALLATION OF THE UPGRADE KIT FACILITATES AN INCREASE TO THE ELECTRICAL CAPABILITY OF EDG NO. 12 IN THE FUTURE. WITH THE INCORPORATION OF FUTURE LOADS (PLANNED AND POTENTIAL), THE MARGINS WITHOUT THIS UPGRADE WOULD BE UNACCEPTABLE. HOWEVER, WITH THE CAPACITY UPGRADE (FOR EDG NO.12) AND WITH THE FUTURE LOADS ADDED, POSITIVE LOAD MARGINS WILL BE MAINTAINED ON ALL BUSES.

THE PROBABILITY AND CONSEQUENCES OF MALFUNCTIONS AND ACCIDENTS PREVIOUSLY I

EVALUATED IN TMS SAR ARE NOT IMPACTED BY THIS ACTIVITY BECAUSE THE RELIABILITY OF THE EDG IS NOT IMPACTED BY THIS ACTIVITY.

NO ADDITIONAL FAILURE MODES OF THE EDG ENGINE ARE BEING CREATED BY THIS ACTIVITY, AND NO NEW INTERACTIONS BETWEEN SYSTEMS ARE CREATED BY CHANGES COVERED UNDER THIS SAFETY EVALUATION. FURTHERMORE, THE EDGS ARE ACCIDENT MITIGATORS AND CANNOT BECOME AN INITIATOR OF A NEW ACCIDENT. THEREFORE, i

THE POSSIBILITY OF A NEW MALFUNCTION OR ACCIDENT HAS NOT BEEN CREATED BY THIS ACTIVITY.

THE QUANTITY OF FUEL OIL REQUIRED TO BE STORED (IN DAY TANKS AND FUEL OIL STORAGE TANKS) BY THE TECHNICAL SPECIFICATIONS IS SUFFICIENT. THE OPERABILITY AND CAPABILITY OF THE SERVICE WATER SYSTEM ARE NOT ADVERSELY AFFECTED BY THIS ACTIVITY, AND THERE ARE NO UNREVIEWED SAFETY QUESTIONS ASSOCIATED WITH THIS ACTIVITY.

P

NNRB018 NUCLEIS 10/15/1995 Search Procesa Adioc Report 56 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995) i l

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L a

Document ID Revision Status

================================

92-B-042-066-R02 62 i

Subject:

ALLOW TEMPORARY OPENING OF SEALED INTAKE STRUCTURE PENETRATIONS Alias:

POSRC #:

95-021 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: 0 Assoc Type: ESP Ref Doc ID:

Rey:

Refer Type:

t Sender Xmtl #

Xmtl Oate

==_m

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Pers R2fs:

Equipment:

Org/Div:

System Code: 042 CIRCULATING WATER Text:

NRC SLSWtARY:

THIS ACTIVITY ALLOWS THE TEMPORARY OPENING OF SEALED INTAKE STRUCTURE F

l PENETRATIONS (I.E. REMOVAL OF BLIND FLANGES, MANWAYS, PUNPS) WHICH PREVENT WATER FROM ENTERING THE BUILDING AND THEREBY, PREVENT FLOODING OF THE INTAKE STRUCTURE FROM OUTSIDE SOURCES. THE ONLY EXCEPTION TO THIS ARE THOSE PENETRATIONS COVERED BY TECHNICAL SPECIFICATIONS.

BOTH UNITS ARE AFFECTED BY AN OPEN PENETRATION ON EITHER UNIT. THE INTAKE STRUCTURE HOUSES THE CIRCULATING WATER PUMPS AND SALTWATER PLMPS FOR BOTH UNITS.

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NMR8018 NUCLEIS 10/15/1995 Search Process Adioc Report 57 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

THIS ACTIVITY FACILITATES MAINTENANCE ACTIVITIES THAT OCCUR IN THE INTAKE STRUCTURE. THIS ACTIVITY RESULTS IN A TEMPORARY CHANGE TO THE SAR DESCRIPTION OF THE DESIGN FUNCTION OF THE INTAKE STRUCTURE, WHICH IS THE REASON FOR THIS SAFETY EVALUATION.

THE OPENING OF THE INTAKE STRUCTURE PENETRATIONS WILL NOT ADVERSELY AFFECT ANY SAFETY RELATED EQUIPMENT IN THE INTAKE STRUCTURE. THEREFORE, THE A8!LITT OF THE OPERATING UNIT (S) TO SAFELY SHUTDOWN AND REMAIN SHUTDOWN IS NOT AFFECTED BY TMIS ACTIVITY. IN ADDITION, THE INTAKE STRUCTURE IS MAINTAINED AS Q LIST CLASSIFICATION SR CLASS 1.

THIS EVALUATION DOES NOT ADDRESS THE SECURITY REQUIREMENTS OF AN OPEN PENETRATION. T"E SECURITY REQUIREMENTS DIFFER DEPENDING ON THE SIZE OF THE OPENING. IN GEhiRAL, A PENETRATION WITH AN OPENING OF % IN (2) OR GREATER CREATES AN AD0!*.IONAL SECURITY CONCERN.

THIS ACTIVITY D(Es NOT CREATE AN UNREVIEWED SAFETY QUESTION, NOR DOES IT REDUCE THE MARGIN OF SAFETY AS DESCRIBED IN THE TECHNICAL SPECIFICATION 8ASES.

(CMH)

Docunant ID Revision Status

================================

95-0013 0000 62

Subject:

INSTALLATION OF CLAMPS TO RESTRAIN 218 RCP SHAFT IAW TEMP ALT 2 95 0012

NWRB018 NUCLEIS 10/15/1995 Search Process Ae oc i;eport 58 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

Alias:

POSRC #:

95-021 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: 0 Assoc Type: ESP Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xmtl Date

ss=====

==========3m===================================================

==========z3

==

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Equipment:

Org/Div:

System Code: 064 REACTOR COOLANT Text:

NRC SupWERY:

TEMPORARY ALTERATION 2 95 0012 INSTALLS CLAMPS TO RESTRAIN 21B RCP SHAFT IN l

ORDER TO FACILITATE REMOVAL OF THE RCP MOTOR IN MODES 5 & 6. THE ACTIVITY IS NECESSARY IN ORDER TO RESTRAIN THE SHAFT AGAINST AXIAL MOTION THAT WILL OTHERWISE OCCUR ON REMOVAL OF THE MOTOR WITH THE RCS PRESSURIZED AXIAL MOTION WILL RESULT IN SEAL DAMAGE AND COULD RESULT IN A BREACH OF THE RCS PRESSURE BOUNDARY.

THE CLAMPS HAVE BEEN ANALYZED FOR A THRUST LOAD CORRESPONDING TO AN RCS PRESSURE OF 500 PSI CONCURRENT WITH THE DESIGN BASIS SEISMIC EVENT. THE i

l ANALYSIS PROVIDES QUALIFICATION OF THE CLAMP COMPONENTS AS WELL AS THE PUMP COMPONENTS TO WHICH THE CLAMPS ARE ATTACHED. CLAMP MATERIALS ARE IN ACCORDANCE WITH THE CODE OF RECORD, ASME B & PV SECTION III SUBSECTION NF, AND STRESS INTENSITIES ARE IN ACCORDANCE WITH DIVISION 1 OF THE SAME CODE.

STRESSES ARE CONSERVATIVELY LIMITED TO S M VALUES RATHER THAN AISC BASED 3

ALLOWABLES TO RECOGNIZE THAT FIALURE OF THE CLAMPS REPRESENTS A DIRECT BREACH i

OF THE RCS PRESSURE BOUNDARY RATHER THAN AN INDIRECT BREACH IN THE CASE OF NORMAL SUPPORTS FOR CLASS 1 SYSTEMS.

SAFETY EVALUATION 95 0013 CONCLUDES THAT THE ACTIVITY IS NOT AN UNREVIEWED SAFETY OUESTION (USQ). EXISTING ACCIDENTS ANALYZED IN THE SAR ARE NOT i

AFFECTED AND NO NEW ACCIDENT CONDITIONS ARE INTRODUCED. THE ACTIVITY DOES NOT VALIDATE TECHNICAL SPECIFICATION REQUIREMENTS.

(CMH) l i

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Nuts 018 NUCLEIS 10/15/1995 Search Procesa A & oc Report 59 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995) i i

Document ID Revision Status a

================================

95-0017 0000 64 SLbject:

ISOLATE CHEMICAL AND VOLUME CONTROL SYSTEM MANUAL VALVE, 2 CVC 397, FROM THE LETDOWN LINE WHICH TAPS OFF THE COLD LEG OF REATOR COOLANT LOOP 22A.

Atlas:

POSRC #:

95-021 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: C Assoc Type: ESP Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xatt Date-

===================================================================================================

Other refs:

Pers R.tfs:

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Org/Div:

System Ccde:

Text:

SUMMARY

THE PURPOSE OF TEMPORARY ALTERATION 2 95 0014 IS TO INSTALL A FREEZE SEAL IN THE 2" 22A LETDOWN LINE BETWEEN THE 22A REACTOR COOLANT LOOP AND 2 CVC 397 l

s

NNRB018 NUCLEIS 10/15/1995 Search Proces3 A & oc Report 60 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

THIS MANUAL ISOLATION VALVE MAS EXHIBITED SEAT LEAKAGE PROBLEMS AND REQUIRES ~

COMPLETE REPLACEMENT. NO ISOLATION VALVES EXIST IN THE 22A LETDOWN LINE BETWEEN THE 22A REACTOR COOLANT LOOP AND 2 CVC 397. THE FREEZE SEAL WILL ALLOW 2 CVC 397 TO BE REPLACED DURING MODE 6 WITH THE REACTOR VESSEL HEAD REMOVED.

THE FREEZE SEAL HAS BEEN EVALUATED AS EQUIVALENT TO A SYSTEM BOUNDARY ISOLATION VALVE. DESIGN REQUIREMENTS HAVE BEEN CONSIDERED, THAT ARE EQUIVALENT TO SUCH A VALVE, AND WERE DETERMINED TO BE ACCEPTABLE. THE FREEZE SEAL WILL HAVE NO EFFECT ON ANY INSTRUMENTATION USED BY THE OPERATORS DURING MODE 6 SINCE THE PLANT IS NOT OPERATING. IF NITROGEN SUPPLY TO THE FREEZE WERE LOST INTEGRITY OF THE FREEZE SEAL WILL BE MAINTAINED FOR 2 HOURS. DURING THIS 2 HOUR PERIOD A MECHANICAL PIPE PLUG WOULD BE INSTALLED IN THE OPEN END OF THE P!PE. THEREFORE, IF THE FREEZE SEAL FAILS, WATER WILL FILL THE PREVIOUSLY EMPTY PIPE BETWEEN THE FREEZE SEAL AND THE NORMAL LOCATION OF 2 CVC 397, WHERE THE PIPE PLUG IS INSTALLED. THIS WILL CAUSE WATER TO DRAIN FROM THE REFUELING POOL. THE DROP IN THE WATER LVEL IN THE REFUELING POOL WILL gr ""IGNIFICANT (LESS THAN AN INCH) DUE TO THE RELATIVE SIZE OF THE LETD N L.. LOMPARED TO THE REFUELING POOL. THERE IS NO THREAT OF EXPOSING THE F8' 0d VIOLATING MINIUMUM POOL LEVELS.

THE TEMPORARY ALTERATION TEMPORARILY AFFECTS UFSAR FIGURE 4 17. THIS ACTIVITY IS NOT A USQ, NOR DOES IT REDUCE THE MARGIN OF SAFETY DESCRIBED IN THE TECHNICAL SPECIFICATION BASES.

(CMH)

Document ID Revision Status

================================

SE00006 0000 62 i

Subject:

THE MCR ALLOWS THE INSTALLATION OF AN INTERIM MECHANICAL PLUG INSIDE THE LEAK OFF PORT OF THE REACTOR VESSEL I

MEAD FLANGE LEAKAGE DETECTION SYSTEM j

Alias:

NWRB018 NUCLEIS 10/15/1995 Search Proces2 Adioc Report 61 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

POSRC #:

95-021 Assoc Doc ID: 94-084-001-03 Revision To: 0003 Assoc Stat: 0 Assoc Type: MCR Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xett Date

me================================================================================

============

6 Other rafs:

Pers RIfs:

Equipment:

Org/Div:

System Code: 064 REACTOR COOLANT Text:

SUMARY:

THE MCR ALLOWS THE INSTALLATION OF AN INTERIM MECHANICAL PLUG INSIDE THE LEAK OFF PORT OF THE REACTOR VESSEL HEAD FLANGE LEAKAGE DETECTION SYSTEM.

THE PLUG SUPPORTS ONGOING MAINTENANCE OF THE DETECTION PIPING AND WILL ONLY BE INSTALLED IN MODES 5, 6 OR DEFUELED. THE PLUG IS INSTALLED TO PREVENT LEAKAGE PAST THE RV HEAD FLANGE.

THE 50.59 SAFETY EVALUATION IS BEING WRITTEN SINCE THE SYSTEM'S DESCRIPTION AS DEFINED IN FIGURE 4-17 0F THE UFSAR IS TEMPORARILY ALTERED.

THIS ACTIVITY WILL NOT DEGRADE THE RELIABILITY OF ITS EQUIPMENT / SSC. THE MCR DOES NOT INVOLVE AN UNREVIEWED SAFETY QUESTION, NOR DOES IT REDUCE THE MARGIN OF SAFETY AS DESCRIBED IN THE TECHNICAL SPECIFICATION BASES.

(CMH)

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NNRB018 NUCLEIS 10/15/1995 Search Procesa Achoc Report 62 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995) i i

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NNRB018 NUCLEIS

.10/15/1995 Search Procesa Ac9 toc Report (3

STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

Docunent ID Revision Status

================================

95-0017 0000 62

Subject:

S/G BLOWDOWN SYS Alias:

POSRC #:

95-022 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: O Assoc Type: ESP Ref Doc ID:

Rev:

Refer Type:

Sender Xmti #

Xmtt Date

===================================================================================================

Other refs:

P;rs RIfs:

Equipment:

Org/Div:

System Code: 083 MAIN STEAM Text:

NRC SUPMARY:

THE PURPOSE OF TEMPORARY ALTERATIONS 2 95 0015 AND 2 95 0020 IS TO INSTALL FREEZE SEALS IN THE 2" BOTTOM BLOWDOWN LINES 2" EB 06 2007 AND 2" EB 06 2008 RESPECTIVELY, TO ALLOW REPLACEMENT OF DOWNSTREAM SECTIONS OF THE PIPING SYSTEM. THE STEAM GENERATORS WILL BE IN WET LAY UP AND THE N0ZZLE DAMS WILL BE INSTALLED WHILE THE FREEZE SEAL IS ACTING AS THE GENERATOR PRESSURE BOUNDARY.

THE FREEZE SEAL HAS BEEN EVALUATED AS EQUIVALENT TO A SYSTEM BOUNDARY ISOLATION VALVE. DESIGN REQUIREMENT HAVE BEEN CONSIDERED THAT ARE EQUIVALENT TO SUCH A VALVE AND HAVE BEEN DETERMINED TO BE ACCEPTABLE. IF NITROGEN SUPPLY TO THE FREEZE SEAL IS LOST INTERGRITY OF THE FREEZE SEAL WILL BE MAINTAINED FOR 2 HOURS. DURING THIS 2 HOUR PERIOD A MECHANICAL PIPE PLUG WILL BE INSTALLED IN THE OPEN END OF THE PIPE. THEREFORE, IF THE FREEZE SEAL FAILS COMPLETELY, WATER WILL FILL THE PREVIOUSLY EMPTY PIPE BETWEEN THE FREEZE SEAL AND THE PIPE PLUG. THIS WILL CAUSE WATER TO DRAIN FROM THE STEAM GENERATOR. THE DROP IN THE WATER LEVEL IN THE GENERATOR WILL BE INSIGNIFICANT (LESS THAN IN INCH) DUE TO THE RELATIVE S!ZE OF THE BLOWOWN LINE COMPARED TO THE STEAM GENERATOR.

THE TEMPORARY ALTERATIONS 2 95 0015 / 0020 TEMPORARILY AFFECT UFSAR FIGURE 10 9. THIS ACTIVITY IS NOT A USO, NOR DOES IT REDUCE THE MARGIN OF SAFETY DESCRIBED IN THE TECHNICAL SPECIFICATION BASIS.

(CMH)

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NMRB018 NUCLEIS 10/15/1995 Search Procesa Ac9 toc Report 65 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

Other reefs:

Pers Rtfs:

f Equipment:

Org/Div:

System Code: 084 REACTOR VESSEL INTERNAL Text:

SupmARY: (FOR NRC REPORT)

THE MCR ALLOWS THE INSTALLATION OF AN INTERIM MEC?ANICAL PLUG INSIDE THE LEAK OFF PORT OF THE REACTOR VESSEL HEAD FLANGE LEAKAGE DETECTION SYSTEM.

THE PLUG SUPPORT ONGOING MAINTENANCE OF THE DETECTION PIPING AND WILL ONLY BE INSTALLED IN MODES 5, 6, OR DEFUELED. THE PLUG IS INSTALLED TO PREVENT LEAKAGE PAST THE RV HEAD FLANGE.

THE 50.59 SAFETY EVALUATION IS BEING WRITTEN SINCE THE SYSTEM'S-DESCRIPTION AS DEFINED IN FIGURE 4-17 0F THE UFSAR IS TEMPORARILY ALTERED.

THIS ACTIVITY WILL NOT DEGRADE THE RELIA 8ILITY OF ITS EQUIPMENT / SSC. THE MCR DOES NOT INVOLVE AN UNREVIEWED SAFETY QUESTION, NOR DOES IT REDUCE THE MARGIN OF SAFETY AS DESCRIBED IN THE TECHNICAL SPECIFICATION BASES.

Docunent ID Revision Status

me========================================

92-B-012-075-R01 62

Subject:

ACC0pm0DATE MAINTENANCE OR CLEANING OF THE U1 OR U2 CCWHX, ECCS PP RM AIR CLRS AND THE ECCS PP RM CLR BASKET STNRS, THE COMPONENTS MUST BE DRAINED BACK INTO THE SALT WATER SYSTEM VERSUS INTO THE EXISTIN AUXILIARY BUILDING FLOOR DRAINS.

Alias:

POSRC #:

95-024 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat:

C Assoc Type: ESP Ref Doc ID:

Rev:

Refer Type:

Sender Xmt! #

Xmtl Date

,==========================================_____=========================================================

k Other rsfs:

1 I

.m

NNROO18 NUCLEIS 10/15/1995 Search Proces3 Acftoc Report 66 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

Pers Rifs:

Equipment:

Org/Div:

System Code: 012 SALT WATER COOLING Text:

THIS EVALUATION ADDRESSES DRAINING OF THE SALTWATER FROM THE UNIT ONE OR UNIT TWO COMPONENT COOLING WATER HEAT EXCHANGERS, ECCS PUMP ROOM AIR COOLERS AND THE ECCS PUMP ROOM AIR COOLER BASKET STRAINERS TO AN OPERATING UNIT ONE OR UNIT TWO COMPONENT COOLING BASKET STRAINER. THE COMPONENTS ARE CURRENTLY DRAINED BY OPERATIONS PER PROCEDURE UI-29 TC ALLOW MAINTENANCE OR CLEANING OF THE DRAINED COMPONENT. THIS EVALUATION IS BEING PERFORED BECAUSE THE PROCEDURE REQUIRES NSR HOSE TO BE CONNECTED TO THE SAFETY RELATED DRAIN LINES OF THE COMPONENTS RECEIVING THE WATER. THESE ACTIVITIES ARE ACCEPTABLE FOR THE FOLLOWING REASONS:

1. ONLY ONE OF TWO CCWHX AND NEITHER OF THE TWO ECCS PUMP ROOM COOLERS ARE NEEDED FOR NORMAL OPERATION (REF. UFSAR SECTION 9.5.5).

2. A MINIMUM OF ONE CCWHX AND ONE ECCS PUMP ROOM COOLER IS NEEDED FOLLOWING A LOCI. (REF. UFSAR SECTION 9.5.5).

3. WHEN DRAINING TO A SALT WATER LOOP DN A UNIT, THE SALT WATER FLOW MARGIN WILL NOT BE OVERRUN BECAUSE THE OIL WILL REQUIRE OPERATIONS TO ENSURE THAT ADEQUATE FLOW MARGIN EXISTS.

4. FLOODING OF THE CCWHX ROOMS AND THE ECCS PUMP ROOMS FROM THE SALT WATER SYSTEM HAS BEEN ANALYZED FOR AND FOUND TO BE ACCEPTABLE. (REF. UFSAR SECTION 9.5.5 AND THE BGE FLOODING DESIGN GUIDELINES MANUAL).

AS A RESULT OF THE PROPOSED ACTIVITIES THERE ARE NO UNREVIEWED SAFETY QUESTIONS AND NO CHANGES TO THE TECHNICAL SPECIFICATIONS OR BASES.

NMR8018 NUCLEIS 10/15/1995 Search Process Actioc Report 67 '

STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995) i i

t k

I i

Doctment ID Revision Status

================================

92-B-012-075-R1 0000 62

Subject:

ACCopWOODATE MAINTENANCE OR CLEANING OF THE U1 OR U2 CCw X, ECCS PP RM AIR CLRS AND THE ECCS PP RM CLR BASKET STNRS, THE COMPONENTS MUST BE DRAINED BACK INTO THE SALT WTER SYSTEM VERSUS INTO THE EXISTIN AUKILIARY BUILDING FLOOR DRAINS.

Atlas:

POSRC #:

95-024 Assoc Doc ID: ES9300001 Revision Tv: 0000 Assoc Stat: C Assoc Type: ESP Ref Doc ID:

Rev:

Refer Type

• 1 l

I

~

.. +.

=-

m

NNRB018 NUCLEIS 10/15/1995 Search Process A& oc Report 68 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

Sender Xmtl #

Xmtl Date n======================================================================================

============

Other rsfs:

Pers R;fs:

Equipment:

Org/Div:

System Code: 012 SALT WATER COOLING Text:

THIS EVALUATION ADDRESSES DRAINING OF THE SALTWATER FROM THE UNIT ONE OR UNIT TWO COMPONENT COOLING WATER HEAT EXCHANGERS, ECCS PLMP ROOM AIR COOLERS AND THE ECCS PUMP ROOM AIR COOLER BASKET STRAINERS TO AN OPERATING UNIT ONE OR UNIT TWO COMPONENT COOLING BASKET STRAINER. THE COMPONENTS ARE CURRENTLY DRAINED BY OPERATIONS PER PROCEDURE 01-29 TO ALLOW MAINTENANCE OR CLEANING OF THE DRAINED COMPONENT. THIS EVALUATION IS BEING PERFORMED BECAUSE THE PROCEDURE REQUIRES NSR HOSE TO BE CONNECTED TO THE SAFETY RELATED DRAIN LINES OF THE COMPONENTS RECEIVING THE WATER. THESE ACTIVITIES ARE ACCEPTABLE FOR THE FOLLOWING REASONS:

1. ONLY ONE OF TWO CCWHX AND NEITHER OF THE TWO ECCS PUMP ROOM COOLERS ARE NEEDED FOR NORMAL OPERATION (REF. UFSAR SECTION 9.5.5).

2. A MINIMUM OF ONE CCWHX AND ONE ECCS PUMP ROOM COOLER IS NEEDED FOLLOWING A LOCI. (REF. UFSAR SECTION 9.5.5).

3. WHEN DRAINING TO A SALT WATER LOOP ON A UNIT, THE SALT WATER FLOW MARGIN WILL NOT BE OVERRUN BECAUSE THE DIL WILL REQUIRE OPERATIONS TO ENSURE THAT ADEQUATE FLOW MARGIN EXISTS.

4. FLOODING OF THE CCWHX ROOMS AND THE ECCS PUMP ROOMS FROM THE SALT WATER STSTEM HAS BEEN ANALYZED FOR AND FOUND TO BE ACCEPTABLE. (REF. UFSAR SECTION 9.5.5 AND THE BGE FLOODING DESIGN GUIDELINES MANUAL).

AS A RESULT OF THE PROPOSED ACTIVITIES, THERE ARE NO UNREVIEWED SAFETY QUESTIONS AND NO CHANGES TO THE TECHNICAL SPECIFICATIONS OR BASES.

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b NWRs018 NUCLEIS 10/15/1995 Search Procesa AcSHM: Report 69 STATUS 62 OR 64 50.59s (10/01/1994 THRU 09/30/1995) i I

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Document ID Revision Status

================================

95-0005 0000 62 i

Subject:

THIS IS AN EVALUATION OF FOUR LEAD FUEL ASSEMBLIES (LFA'S) MANUFACTURED BY ABS CCMOUSTION ENGINEERING TO BE INSERTED INTO THE CALVERT CLIFFS UNIT 2 CORE, BEGINNING WITH CYCLE 11.

Alles:

POSRC #:

95-024 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: 0 Assoc Type: ESP Ref Doc ID:

Rev:

Refer Type:

i L

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NMR5018 NUCLEIS 10/15/1995 Search Process Adioc Report 70 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

Sender Xmtl #

Xmtl Date semanc=== ======================= ==================================== ==================== === ==========

Other rsfs:

Pers RIfs:

Equipment:

Org/Div:

System Code: 080 NEW FUEL STORAGE AND ELEVATOR Text:

SUPMARY:

THIS SAFETY EVALUATION CONSIDERS THE USE OF FOUR AB8 COMBUSTION ENGINEERING LEAD FUEL ASSEMBLIES (LFA'S). THE LFA'S WILL RESIDE IN NON-LIMITING LOCATIONS IN THE CALVERT CLIFFS UNIT 2 CORE DURING CYCLES 11, 12 AND 13. PERFORMANCE OF THE LFA'S WILL BE EVALUATED AFTER CYCLES 11 AND 12 TO ENSURE SATISFACTORY PERFORMANCE IN CYCLE 13. DATA FROM THE LFA'S IS INTENDED TO SUPPORT THE DEVELOPMENT OF NEW AND IMPROVED FUEL DESIGNS AND FUEL EVALUATION METHOD-OLOGIES TO ACHIEVE HIGHER BURNUPS, AND ATTAIN OVERALL BETTER FUEL CYCLE ECONOMICS. DESIGN FEATURES CHANGES INCLUDE A SHORTER FUEL ROD LENGTH, A THINNER CLAD, AND A LARGER AND HEAVIER FUEL PELLET. THE LFA'S ARE ALSO HOST TO A DEMONSTRATION OF AN ADVANCED CLADDING MATERIAL, ZIRCALOY 2P. THE RELOAD ANALYSIS PERFORMED FOR THE CORE BOUNDS THE LFA'S. NO CHANGES TO TECHNICAL SPECIFICATIONS ARE REQUIRED, CONTINGENT UPON NRC ISSUANCE OF AMENDMENT FOR CHAPTER 5 " DESIGN FEATURES" SECTION DESCRIBING THE FUEL ASSEMBLY. USE OF THE LFA'S IN CYCLE 11 WAS FOUND TO NOT CONSTITUTE AN UNREVIEWED SAFETY QUESTION.

(CMH)

NNR9018 NUCLEIS 10/15/1995 Search Procesa Ae oc Report 71 STATUS 62 OR 64 50.59s (10/01/1994 THRU 09/30/1995)

Document ID Revision Status

o=a0:o======================

==

95-0028 0000 62

Subject:

TEMP ALT 2 95 033 INSTALLS TEMPORARY AIR COMPRESSOR Alias:

POSRC #:

95-025 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: 0 Assoc Type: ESP Ref Doc ID:

Rev:

Refer Type:

Sender Xat1 #

Xmtt Date

===========================================================_

-=================================

Other rofs:

Pero R2fs:

Equipment:

Org/Div:

System Code:

Text:

NRC SUP9tARY:

IN SUPPORT OF THE 1995 UNIT 2 OUTAGE, TA 2 95 033 INSTALLS A TEMPORARY AIR COMPRESSOR TO SUPPLY AIR, VIA THE INTEGRATED LEAK RATE TEST (ILRT) PIPING, FOR TOOLS AND MISCELLANEOUS LOADS IN THE UNIT 2 REFUELING POOL AREA.

THE TA IS NEEDED TO SUPPLY AIR TO PNEUMATIC TOOLS INSIDE CONTAINMENT DURING THE UPCOMING OUTAGE (UNIT 2 SPRING 1995).

THE 50.59 SAFETY EVALUATION IS BEING WRITTEN SINCE THE ILRT SYSTEM'S DESCRIPTION AS DEFINED IN FIGURE 9 20A 0F THE OFSAR IS TEMPORARILY ALTERED.

THE TEMPORARY AIR SUPPLY WILL NOT EXCEED THE DESIGN PRESSURE OF THE ILRT PIPING. AN ASME SECTION VIII RELIEF VALVE IS SET TO PROVIDE OVERPRESSURE PROTECTION. THE TEMPORARY EQUIPMENT (HOSES / CONNECTION FITTINGS) ARE QUALIFIED FOR THIS APPLICATION. THE TEMPORARY AIR HOSES AND FITTINGS MEET THE SEISMIC II / I DESIGN REQUIREMENTS. CONTAINMENT CLOSURE IS NOT COMPROMISED BY THIS ACTIVITY.

THIS ACTIVITY WILL NOT DEGRADE THE RELIABILITY OF ITS EQUIPMENT / SSC. THE TA DOES NOT INVOLVE AN UNREVIEWED SAFETY QUESTION, NOR DOES IT REDUCE THE MARGIN OF SAFETY AS DESCRIBED IN THE TECHNICAL SPECIFICATION BASES.

(CMH)

NWR3018 NUCLEIS 10/15/1995 Search Process Adioc Report 72 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

Document ID Revision Status

================================

95-0030 0000 62 l

l S4]ect:

DISABLE WIDE RANGE GASEOUS EFFLUENT MONITORING SYSTEM (WRGM) MID-RANGE DETECTOR, 1 RE 5417.

Atlas:

POSRC #:

95-028 Assoc Doc ID: ES?300001 Revision To: 0000 Assoc Stat: 0 Assoc Type: ESP Ref Doc ID:

Rev:

Refer Type:

Sender Matt #

Xatl Date

===================================================================================================

Other refs:

PTrs Rafs:

Equipment:

Org/Div:

1rNRB018 NUCLEIS 10/15/1995 Search Proces3 M oc Report 73 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

System Code: 079 PROCESS RADIATION MONITORING Text:

SUMMARY

THIS TEMPORARY ALTERATION DISABLES THE WIDE RANGE GASEOUS EFFLUENT MONITORING SYSTEM MID-RANGE DETECTOR DUE TO REPEATED FAILURES WITH THE CHECK SOURCE ASSEMBLY AND POTENTIAL LEAKS. THE DETECTOR WILL BE DISABLED IN SUCH A MANNER THAT THE WRGM SYSTEM RECOGN!ZES ONLY THE LOW AIS HIGH RANGE DETECTORS AS VALID INPUTS. THE SYSTEM IS DESIGNED TO OPERATE IN THIS M(BE, AND WILL CONTINUE TO BE FULLY Fl?tCTIONAL. THE LOW AND HIGH RANGE DETECTORS HAVE ADEQUATE RANGES AND OVERLAP TO COVER THE ENTIRE SYSTEM OPERATING RANGE.

THEREFORE, THERE IS NO LOSS OF INDICATION AS A RESULT OF THIS ACTIVITY.

THIS EVALUATION HAS BEEN PREPARED BECAUSE THE SAR DESCRIPTION OF THE THREE OVERLAPPING DETECTORS IN THE WRGM SYSTEM IS TEMPORARILY ALTERED. THE ACTIVITY HAS BEEN EVALUATED AND DETERMINED NOT TO CONSTITUTE AN UNREVIEWED SAFETY QUESTION, BASED ON THE ABILITY OF THE SYSTEM TO CONTINUE TO PERFORM ITS DESIGN FUNCTION WITH NO LOSS OF INDICATION TO THE OPERATOR.

(CMH)

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NWR8018 NUCLEIS 10/15/1995 Search Proces3 A & oc Report 7'a I

STATUS 62 OR 64 50.59S (10/01/1996 THRU 09/30/1995) i Document ID Revision Status

=u==============================

95-0029 0000 62 ALLOW REDUCTION IN THE N' MBER OF 80LTS REQUIRED ON THE FUEL TRANSFER TUBE BLIND FLANGE TO ONLY 22.

S4 ject:

J Alias:

POSRC #:

95-030 Assoc Doc ID: 95-070-001-00 Revision To: 0000 Assoc Stat: 0 Assoc Type: MCR Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xmtl Date

===================================================================================================

Other rsfs:

Pers Rifs:

Equipnent:

Org/Div:

System Code: 070 REFUELING POOL Text:

MINOR MODIFICATION 95 070 001 ALLOWS REDUCTION IN THE TOTAL NUMBER OF BOLTS INSTALLED IN THE FUEL TRANSFER TUBE BLIND FLANGE FROM 44 TO 22. THE FUNCTION AND ABILITY TO PERFORM THE FUNCTION OF THE FUEL TRANSFER TUBE ASSEMBLY IS NOT CHANGED FROM THE ORIGINAL ASSEMBLY. THEREFORE, THE PROPOSED ACTIVITY IS NOT AN UNREVIEWED SAFETY QUESTION, WILL NOT INCREASE THE PROBA81LTIY OR CONSEQUENCES OF A MALFUNCTION OR ACCIDENT, WILL NOT CREATE THE POSSIBLITY OF A DIFFERENT TYPE OF MALFUNCTION OR ACCIDENT, WILL NOT REQUIRE A CHANGE TO THE TECHNICAL SPECIFICATIONS, AND WILL NOT AFFECT THE MARGIN OF SAFETY AS DEFINED IN THE TECHNICAL SPECIFICATIONS.

F (CMH) t f

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NMRB018 NUCLEIS 10/15/1995 Search Proces3 Adioc Report 75 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995) i r

Doctanent ID Revision Status

================================

t 95-0005 0001 62

Subject:

THIS IS AN EVALUATION OF FOUR LEAD FUEL ASSEMBLIES MANUFACTURED BY ABB COMBUSTION ENGINEERING TO BE INSERTED INTO THE CALVERT CLIFFS UNIT 2 CORE, BEGINNING WITH CYCLE 11.

,=

Alias:

POSRC #:

95-032 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat:

0-Assoc Type: ESP Ref Doc ID:

Rev:

Refer Type:

i Sender Xmtl #

Xmtl Date

ww===================================================================================

===============_.

I Other rsfs:

P rs R2fs:

Equipment:

Org/Div:

System Code: 080 NEW FUEL STORAGE AND ELEVATOR Text:

SUN 4ARY:

THIS SAFETY EVALUATION CONSIDERS THE USE OF FOUR ABB COMBUSTION ENGINEERING LEAD FUEL ASSEMBLIES (LFA'S). THE LFA'S WILL RESIDE IN NON-LIMITING LOCATIONS IN THE CALVERT CLIFFS UNIT CORE DURING CYCLES 11, 12 AND 13. PERFORMANCE OF THE LFA'S WILL BE EVALUATED AFTER CYCLES 11 AND 12 TO ENSURE SATISFACTORY PERFORMANCE IN CYCLE 13. DATA FROM THE LFA'S IS INTENDED TO SUPPORT THE DEVELOPMENT OF NEW AND IMPROVED FUEL DESIGNS AND FUEL EVALUATION

NNRs018 NUCLEIS 10/15/1995 Search Procesa A&oc Report 76 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

METHODOLOGIES TO ACHIEVE HIGHER BURNUPS, AND ATTAIN OVERALL BETTER FUEL CYCLE ECONOMICS. DESIGN FEATURES CHANGES INCLUDE A SHORTER FUEL ROD LENGTH, A THINNER CLAD, AND A LARGER AND MEAVIER FUEL PELLET. THE LFA'S ARE ALSO HOST TO A DEMONSTRATION OF AN ADVANCED CLADDING MATERIAL, ZIRCALOY 2P.

THE RELOAD ANALYSIS PERFORMED FOR THE CORE BOUNDS THE LFA'S. NO CHANGES TO TECHNICAL SPECIFICATIONS ARE REQUIRED, CONTINGENT UPON NRC ISSUANCE OF AMENDMENT FOR CHAPTER 5 " DESIGN FEATURES" SECTION DESCRIBING THE FUEL ASSEMBLY. USE OF THE LFA'S IN CYCLE 11 WAS FOUND TO NOT CONSTITUTE AN UNREVIEWED SAFETY QUESTION.

(CMH)

I t

Docunent ID Revision Status

================================

95-0030 0001 62

Subject:

DISA8LE WIDE RANGE GASEOUS EFFLUENT MONITORING SYSTEM (WRGM) MID-RANGE DETECTOR,1 RE 5417.

f Alias:

POSRC #:

95-032 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: O Assoc Type: ESP Ref Doc ID:

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NMRB018 NUCLEIS 10/15/1995 Search Proces3 A e oc Report 77 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

Org/Div:

System Code: 079 PROCESS RADIATION MONITORING Text:

SUMARY:

THIS TEMPORARY ALTERATION DISABLES THE WIDE RANGE GASEOUS EFFtUENT MONITORING SYSTEM MID-RANGE DETECTOR DUE TO REPEATED FAILURES WITH THE CHECK SOURCE ASSEMBLY AND POTENTIAL LEAKS. THE DETECTOR WILL BE DISABLED IN SUCH A MANNER THAT THE WRGM SYSTEM RECOGNIZES ONLY THE LOW AND HIGH RANGE DETECTORS AS VALID INPUTS. THE SYSTEM IS DESIGNED TO OPERATE IN THIS MODE, AND WILL CONTINUE TO BE FULLY FUNCTIONAL. THE LOW AND HIGH RANGE DETECTORS HAVE ADEQUATE RANGES AND DVERLAP TO COVER THE ENTIRE SYSTEM OPERATING RANGE.

THEREFORE, THERE IS NO LOSS OF INDICATION AS A RESULT OF THIS ACTIVITY.

THIS EVALUATION HAS BEEN PREPARED BECAUSE THE SAR DESCRIPTION OF THE THREE OVERLAPPING DETECTORS IN THE WRGM SYSTEM IS TEMPORARILY ALTERED. THE ACTIVITY HAS BEEN EVALUATED AND DETERMINED NOT TO CONSTITUTE AN UNREVIEWED SAFETY QUESTION, BASED ON THE ABILITY OF THE SYSTEM TO CONTINUE TO PERFORM ITS DESIGN FUNCTION WITH NO LOSS OF INDICATION TO THE OPERATOR.

(CMH)

NNRB018 NUCLEIS 10/15/1995 Search Process Actioc Report 78 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

Document ID Revision Status aw===_========================

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95-0017 0001 64

Subject:

ISOLATE CHEMICAL AND VOLUME CONTROL SYSTEM MANUAL VALVE, 2 CVC 397, FROM THE LETDOWN LINE WHICH TAPS OFF THE COLD LEG OF REACTOR COOLANT LOOP 22A.

Atlas:

POSRC #:

95-033 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: 0 Assoc Type: ESP Ref Doc ID:

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THE PURPOSE OF TEMPORARY ALTERATION 2 95 0014 IS TO INSTALL A FREEZE SEAL IN THE 2" 22A LETDOWN LINE BETWEEN THE 22A REACTOR COOLANT LOOP AND 2 CVC 397.

THIS MANUAL ISOLATION VALVE HAS EXHIBITED SEAT LEAKAGE PROBLEMS AND REQUIRES COMPLETE REPLACEMENT. NO ISOLATION VALVES EXIST IN THE 22A LETDOWN LINE F

BETWEEN THE 22A REACTOR COOLANT LOOP AND 2 Cyc 397. THE FREEZE SEAL WILL ALLOW 2 CVC 397 TO BE REPLACED DURING MODE 6 WITH THE REACTOR VESSEL HEAD REMOVED.

THE FREEZE SEAL HAS BEEN EVALUATED AS EQUIVALENT TO A SYSTEM BOUNDARY ISOLATION VALVE. DESIGN REQUIREMENTS HAVE BEEN CONSIDERED, THAT ARE EQUIVALENT TO SUCH A VALVE, AND WERE DETERMINED TO BE ACCEPTABLE. THE FREEZE SEAL WILL HAVE NO EFFECT ON ANY INSTRUMENTATION USED BY THE OPERATORS DURING MODE 6 SINCE THE PLANT IS NOT OPERATING. IF NITROGEN SUPPLY TO THE FREEZE WERE LOST INTEGRITY OF THE FREEZE SEAL WILL BE MAINTAINED FOR 2 HOURS.

DURING THIS 2 HOUR PERIOD A MECHANICAL PIPE PLUG WOULD BE INSTALLED IN THE OPEN END OF PIPE. THEREFORE, IF THE FREEZE SEAL FAILS, WATER WILL FILL THE PREVIOUSLY EMPTY PIPE BETWEEN THE FREEZE SEAL AIN) THE NORMAL LOCATION OF 2 CVC 397, WHERE THE PIPE PLUG IS INSTALLED. THIS WILL CAUSE WATER TO DRAIN FROM THE REFUELING POOL. THE DROP IN THE WATER LEVEL IN THE REFUELING POOL

NNR9018 NUCLEls 10/15/1995 Search Procesa Adioc Report 79 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

WILL BE INSIGNIFICANT (LESS THAN AN INCH) DUE TO THE RELATIVE SIZE OF THE LETDOWN LINE COMPARED TO THE REFUELING POOL. THERE IS NO THREAT OF EXPOSING THE FUEL OR VIOLATING MINIMUM POOL LEVELS.

THE TEMPORARY ALTERATION TEMPORARILY AFFECTS UFSAR FIGURE 417. THIS ACTIVITY IS NOT A USQ, NOR DOES IT REDUCE THE MARGIN OF SAFETY DESCRIBED IN THE TECHNICAL SPECIFICATION BASES.

(CMH)

Document ID Revision Status

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95-0031 0000 62

Subject:

THIS 50.59 REVIEW IS WRITTEN TO EVALUATE THE USE OF SPARE KEYS TO DEFEAT THE KIRK KEY INTERLOCKS ON THE 4KV DISCONNECTS ALLOWING BOTH DISCONNECTS FEEDING A SINGLE SAFETY BUS TO BE CLOSED AT THE SAME TIME.

Alias:

POSRC #:

95-035 Assoc Doc ID: ES9300001 Revisfort To: 0000 -Assoc Stat: 0 Assoc Type: ESP Ref Doc ID:

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System Code: 024 EMERGENCY DIESEL GENERATOR Text:

SupWIARY:

THIS 50.59 EVALUATED DEFEATING THE KIRK KEY INTERLOCK ASSOCIATED WITH THE DIESEL GENERATOR FEEDER BREAKER DISCONNECT SWITCH. THE INTERLOCK WAS DEFEATED BY OBTAINING A SPARE KEY. THIS PROVIDES THE POTENTIAL FOR ATTEMPTING TO ALIGN TWO ELECTRICAL POWER SOURCES TO THE SAM BUS AT THE SAE TIME.

DEFEATING THE KEY INTERLOCK IS NECESSARY TO PROVE THAT THE ESFAS SIGNALS ARE MODIFIED CORRECTLY. THE BUSES, BREAKERS AND DISCONNECT ASSOCIATIONS ARE LISTED TO BELOW. THE XX 03 AND XX 06 DISCONNECTS ARE INTERLOCKED TO PREVENT AN ATTEMPT TO ALIGN BOTH DIESEL GENERATORS TO A SINGLE BUS.

NWRB018 NUCLEIS 10/15/1995 Search Proces3 A& oc Report 80 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

(CMH)

Doctanent ID Revision Status

=

95-0002 0000 62

Subject:

PROPOSED ACTIVITY: TO FORMALLY ALLOW FOR THE USE OF UNJACKETED FIBERGLASS BLANKET INSULATION WITHIN THE UNIT 2 CONTAINMENT.

Atlas:

POSRC #:

95-036 Assoc Doc ID: 94-169-003-00 Revision To: 0000 Assoc Stat: O Assoc Type: MCR Ref Doc ID:

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System Code: 169 INSULATION, PIPE Text:

SUMMARY

THIS ACTIVITY APPROVES THE USE OF UNJACKETED (I.E., NO METAL JACKETING)

FIBERGLASS BLANKET INSULATION 04 VARIOUS PIPING AND COMPONENTS WITHIN CONTAINMENT. THE FUNCTION OF METAL JACKETING GERMANE TO THIS 10CFR50.59 EVALUATION IS TO SHED WATER UPON ACTUATION OF CONTAIMMENT SPRAY OR DURING SYSTEM LEAKING. THE ABILITY TO SHED WATER ENSURES THAT 1) THERMAL SHOCKING OF THE COVERED EQUIPMENT / PIPING WILL NOT OCCUR, 2) THE INCREASE IN INSULATION WEIGHT DUE TO WATER ABSORPTION DOES NOT ADVERSELY AFFECT PIPING STRESS OR SUPPORT LOADS, AND 3) UNACCEPTABLE CONCENTRATIONS OF BORIC ACID SOLUTIONS DO NOT COME INTO CONTACT WITH INSULATED PIPING AND EQUIPMENT. IN

N4RB018 NUCLEIS 10/15/1995 Search Procesa Ac9 toc Report 81 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

ADDITION, THE UNJACKETED INSULATION WAS EVALUATED WITH REGARD TO BLOCKAGE OF THE CONTAINMENT SUMP; IT WAS DETERMINED THAT BLOCKAGE OF THE CONTAINMENT SUMP WILL NOT OCCbR DUE TO THE DISLODGING OF THE INSULATION CAUSED BY A HIGH ENERGY LINE BREAK. BASED UPON MEETING THE CRITERIA FOR THE FUNCTION OF THE JACKETING GERMANE TO SAFETY RELATED PERFORMANCE, FIBERGLASS BLANKET INSULATION INSTALLED WITH A WDVEN FIBERGLASS CLOTH COVER BUT NO METAL JACKET-ING IS AN ACCEPTABLE CONFIGURATION FOR THE EVALUATED LOCATIONS. AS NOTED ABOVE, THE UNJACKETED FIBERGLASS INSULATION WILL PROVIDE ADEQUATE PROTECTION AGAINST THERMAL SHOCKING OF THE PIPING / COMPONENTS. SECTION 6.4.4 0F THE UFSAR INDICATES THAT THE INSULATION INSIDE OF CONTAINMENT MUST PROVIDE PROTECTION AGAINST THERMAL SHOCKING OF THE PIPING / COMPONENTS FOLLOWING THE INADVERTENT ACTUATION OF CONTAINMENT SPRAY. DURING AN INADVERTENT ACTUATION OF CONTAINMENT SPRAY, THE UNJACKETED FIBERGLASS INSULATION WILL PROVIDE AN ADEQUATE BARRIER TO PREVENT THE COOL SPRAY FROM COMING IN DIRECT CONTACT WITH THE HOT EQUIPMENT / PIPING. A UFSAR CHANGE REQUEST FORM MAS BEEN INITIATED UNDER THIS MCR TO AMEND SECTION 6.4.4.

(CMH) i t

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NNRB018 NUCLEIS 10/15/1995 Search Procesa A & oc Report 82 STATUS 62 OR M 50.595 (10/01/1994 THRU 09/30/1995)

Document ID Revision Status

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93-B-0MA-152-R03 62

Subject:

THIS ACTIVITY WILL PERMIT THE USE OF NUCLEAR ENGINEERING SERVICES (NES) N0ZZLE DAMS AS AN ALTERNATIVE TO THE CE N0ZZLE DAMS.

Alias:

POSRC #:

95-038 Assoc Doc ID: 93-0202-0003 Revision To: 0000 Assoc Stat: 0 Assoc Type:

FEC Ref Doc ID:

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System Code: 064 REACTOR COOLANT Text:

SUMMARY

THE NES N0ZZLE DAM CAN BE INSTALLED IN LESS TIME THAN THE CE N0ZZLE DAMS, WHICH IS IN KEEPING WITH THE GOAL OF MAINTAINING PERSONNEL EXPOSURE AS LOW AS REASONABLY ACHIEVABLE (ALARA). THE NES N0ZZLE DAM WILL ALSO ALLOW A HIGHER REFUELING POOL / REACTOR VESSEL WATER LEVEL TO BE MAINTAINED SINCE THEY WOULD BE INSTALLED AT A HIGHER NGZZLE ELEVATION THAN THE CURRENT PROCEDURALIZED LEVEL. THIS WILL INCREASE THE TIME TO REACH SATURATION CONDITIONS (I.E., B0! LING IN THE CORE) DURING A LOSS OF SHUTDOWN COOLING ACCIDENT AND PROVIDE A GREATER SHUTDOWN SAFETY MARGIN. THE NES N0ZZLE DAM WAS SELECTED SINCE IT IS A PROVEN DESIGN WHICH IS CURRENTLY IN USE AT OTHER CE PLANTS, INCLUDING WATERFORD, ARKANSAS NUCLEAR ONE - 2, ST. LUCIE 1 & 2, AND PALO VERDE 1, 2, AND 3.

Document ID Revision Status

= = = = = = = = = = _ _ _ _ _ _ = = = = = = = = = = _ = = = = = - = = = = =

95-00 %

0000 64 Stbject:

THIS IS AN EVALUATION OF THE DESIGN AND PERFORMANCE OF THE CALVERT CLIFFS UNIT 2 REACTOR FOR THE OPERATION OF CYCLE 11 AT THE FULL RATED POWER OF 2700 MW.

Alias:

Pouc #:

95-038 l

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NMR5018 WUCLEIS 10/15/1995 Search Procesa Ac9mc Report 83 j

STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: O Assoc Type: ESP

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System Code: 080 NEW FUEL STORAGE AND ELEVATOR Text:

SUMARY:

THIS SAFETY EVALUATION CONSIDERED THE OPERATION OF UNIT 2 CYCLE 11.

MWIFICATIONS TO THE FUEL RCD AND THE FUEL ASSE95LY AIS TME RELOAD CORE DESIGN WERE CONSIDERED. THE USE OF ERSIUM FOR UNIT 2 AS A BURNABLE ASSORSER WAS CONSIDERED. THE UNIT 2 CYCLE 11 ANALYSIS ASSUED AN ALLOWED PEAK LINEAR MEAT RATE OF 14.5 KW / FT AND AN EOC SMUTDOWN MARGIN OF 4.5% DELTA RMO. THESE

[

CHANGES WILL BE IMPLEMENTED IN THE UNIT 2 CYCLE 11 CORE OPERATING LIMITS

{

REPORT (COLR). THE PRE-TRIP STEAM LINE BREAK EVENT WAS REANALYZED USING METHODOLOGY DESCRIBED IN THE SAR TO PREDICT THE PERCENTAGE OF FUEL FAILURES.

THE RE-ANALYSIS ASSUMED MORE RESTRICTIVE FUEL FLUX LIMITS IN ORDER TO.

RESTRICT THE PREDICTED NLDIRER OF FUEL FAILURES TO A LESS LIMITING PERCENTAGE 2

THAN THAT PREVIOUSLY REPORTED. THE FUEL FLUX LIMITS ARE EQUAL TO 1.64 & WILL BE IMPLEMENTED IN THE UNIT 2 CYCLE 11 COLR. THE UNIT 2 CYCLE 11 SAFETY ANALYSES ACCOUNTED FOR ALL RELOAD CORE DIFFERENCES. YtfE RESULTS OF ALL REFERENCE SAFETY ANALYSES CONSERVATIVELY APPLY TO UNIT 2 CYCLE 11. IT IS CONCLUDED THAT OPERATION OF UNIT 2 CYCLE 11 DOES NOT INVOLVE AN UNREVIEWED SAFETY QUESTION.

(CMH)

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Document ID Revision Status

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95-0033 0000 64

Subject:

PROPOSED ACTIVITY: BY-PASS THE RCS PRESSURE INTERLOCK IN SAFETY INJECTION TANK (SIT) ISOLATION VALVE 2 MOV 624 CONTROL CIRCUIT TO ALLOW TESTING.

Allas:

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NWRB018 NUCLEIS 10/15/1995 Search Proces2 A& oc Report 84 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

POSRC #:

95-043 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: 0 Assoc Type: ESP Ref Doc ID:

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System Code: 064 REACTOR COOLANT Text:

SUMMARY

THIS TEMPORARY ALTERATION BY PASSES THE RCS PRESSURE INTERLOCK FOR SIT ISOLATION VALVE 2 MOV 624 DURING MODES 4, 5, 6 AND DEFUELED. THE INTERLOCK IS BY PASSED TO ALLOW TESTING (CYCLING) 0F THE MOV CONCURRENT WITH MODIFICATIONS TO THE PRESSURIZER PRESSURE INSTRUMENTATION. MODIFICATIONS CURRENTLY UNDERWAY HAVE DISABLED THIS INSTRUMENT LOOP, CAUSING A SIMULATED HIGH PRESSURE INPUT TO 2 MOV 624 CONTROL CIRCUIT. AS A RESULT, THE VALVE HAS A LOCKED IN "0 PEN" $1GNAL, AND CAN NOT BE CLOSED ELECTRICALLY. THIS TEMPORARY ALTERATION BY PASSES THE RCS PRESSURE INPUT TO ALLOW CYCLING OF 2 MOV 624 WHILE IN MODES 4, 5, 6 AND DEFUELED BY PASSING THE INTERLOCK WILL BE PERFORMED BY LIFTING LEADS AND INSTALLING A JUMPER IN THE VALVE CONTROL CIRCUIT.

THE ACTIVITY HAS BEEN EVALUATED AND DETERMINED NOT TO CONSTITUTE AN UNREVIEWED SAFETY QUESTION, BASED PRIMARILY ON ITS APPLICABILITY TO MODES 4, 5, 6 AND DEFUELED ONLY. IN THESE MODES, OPERABILITY OF THE SIT'S (AND THEIR ASSOCIATED ISOLATION VALVES) IS NOT REQURED. IN ADDITION, THIS ACTIVITY ONLY BY PASSES THE INTERLOCK FOR 2 MOV 624. NO OTHER SSC'S ARE IMPACTED BY THIS TEMPORARY ALTERATION.

(CMH)

NWR8018 NUCLEIS 10/15/1995 Search Procesa A&oc Report 85 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

Document ID Revision Status

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95-0006 0001 64 Stbject:

THIS IS AN EVALUATION OF THE DESIGN AND PERFORMANCE OF THE CALVERT CLIFFS UNIT 2 REACTOR FOR THE OPERATION OF CYCLE 11 AT THE FULL RATED POWER OF 2700 m.

Alias:

POSRC #:

95-046 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: 0 Assoc Type: ESP Ref Doc ID:

Rev:

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Sender Matl #

Matt Date

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NWRB018 NUCLEls 10/15/1995 Search Proces2 Adtoc Report 86 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

THIS SAFETY EVALUATION CONSIDERED THE OPERATION OF U;IT 2 CYCLE 11.

MODIFICATIONS TO THE FUEL ROD AND THE FUEL ASSEMBLY AIO THE RELOAD CORE DESIGN WERE CCMSIDERED. THE USE OF ERBIUM FOR UNIT 2 AS A BURNABLE ABSORBER WAS CONSIDERED. THE UNIT 2 CYCLE 11 ANALYSIS ASSUMED AN ALLOWED PEAK LINEAR HEAT RATE OF 14.5 KW / FT AND AN EOC SHUTDOWN MARGIN OF 4.5% DELTA RHO. THESE CHANGES WILL BE IMPLEMENTED IN THE UNIT 2 CYCLE 11 CORE OPERATING LIMITS REPORT (COLR). THE PRE-TRIP STEAM LINE BREAK EVENT WAS REANALYZED USING METHODOLOGY DESCRIBED IN THE SAR TO PREDICT THE PERCENTAGE OF FUEL FAILURES.

THE RE-ANALYSIS ASSLMED MORE RESTRICTIVE... LIMITS IN ORDER TO RESTRICT THE PREDICTED NUMBER OF FUEL FAILURES TO A LESS LIMITING PERCENTAGE THAN THAT PREVIOUSLY REPORTED. THE... LIMITS ARE EQUAL TO 1.64 AND WILL BE IMPLEMENTED IN THE UNIT 2 CYCLE 11 COLR. THE UNIT 2 CYCLE 11 SAFETY ANALYSES ACCOUNTED FOR ALL RELOAD CORE DIFFERENCES. THE RESULTS OF ALL REFERENCE SAFETY ANALYSES CONSERVATIVELY APPLY TO UNIT 2 CYCLE 11. IT IS CONCLLDED THAT OPERATION OF UNIT 2 CYCLE 11 DOES NOT INVOLVE AN UNREVIEWED SAFETY QUESTION.

(CMH)

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NNRB018 NUCLEIS 10/15/1995 Seerch ProcesO A& oc Report 87

' STATUS 62 OR 64 50.595 (10/01/1996 THRU 09/30/1995)

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96-B-999-065-R01 64

Subject:

THIS ACTIVITY MODIFIES THE EXISTING ELECTRICAL DISTRIBUTION SYSTEM IN ORDER TO CONNECT THE STATION BLACKOUT l

DIESEL GENERATOR, DC OC, TO AN ENGINEERED SAFETY FEATURED BUS IN UNIT 2 Alias:

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NWRB018 NUCLEIS 10/15/1995 Search frocesa A& oc Report 88 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

POSRC #:

95-047 Assoc Doc ID: 89-0079 Revision To: 0000 Assoc Stat: 0 Assoc Type:

FCR Ref Doc ID:

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Xmtt Date

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SLBMARY:

THIS ACTIVITY MODIFIES THE EXISTING ELECTRICAL DISTRIBUTION SYSTEM IN ORDER TO CONNECT THE STATION BLACKOUT (SBO) DIESEL GENERATOR, DC OC, TO AN ENGINEERED SAFETY FEATURES BUS IN UNIT 2 (ENERGENCY BUS 24). THEREFORE, THIS ACTIVITY DISCONNECTS EMERGENCY DIESEL GENERATOR 12 (DG 12) FROM EMERGENCY BUS 24 AND CONNECTS DG OC TO EMERGENCY BUS 24. THIS ACTIVITY ALSO ADDS THE RACEWAY AND CABLES NECESSARY TO COMPLETE THIS PART OF THE PHASE-IN OF DG OC.

ENGINE CONTROLS FOR DG 12 WILL BE MODIFIED TO DELETE START SIGNALS ASSOCIATED WITH EMERGENCY BUS 24.

IN ADDITION, THIS ACTIVITY INSTALLS AND TERMINATES WIRING BETWEEN THE ELECTRICAL AUXILIARY CONTROL PANEL (EACP) AND THE DIESEL GENERATOR CONTROL r

CONSOLE (DGCC) FOR DG 21 INSTRUMENTATION AND CONTROLS. CONTROL OF DG 21 AND EMERGENCY BUS 24 WILL BE TRANSFERRED TO THE DGCC AND THE INTERNAL WIRING IN THE EACP FOR DG 21 WILL BE DISCONNECTED. THIS ACTIVITY ALSO RELOCATES A NUMBER OF ANNUNCIATOR WINDOWS ASSOCIATED WITH DG 21, EMERGENCY BUS 24 AND FUEL OIL STORAGE TANK NO. 21.

IN ORDER TO DISCONNECT DG 12 FROM EMERGENCY BUS 24, THIS ACTIVITY COULD RESULT IN A PLANT CONFIGURATION WHERE NO EMERGENCY DIESEL GENERATORS WOULD BE i

AVAILABLE TO UNIT 2 FOR UP TO 14 DAYS. ACTION STATEMENT B 0F TECHNICAL SPECIFICATION 3.8.2.2 ALLOW UNIT 2 TO CONTINUE SHUTDOWN OPERATIONS WITHOUT EDG'S ALIGNED TO IT FOR SEVEN DAYS DURING THE PERFORMANCE OF STP M 20 INSPECTION ON DG 12 (SURVEILLANCE REQUIREMENT 4.8.1.1.2.D.1). IMPLEMENTATION OF THIS ACTIVITY WILL REQUIRE AN NRC APPROVED EXTENSION OF THE SEVEN DAY LIMITATION OF ACTION STATEMENT B 0F TECHNICAL SPECIFICATION 3.8.1.2 AND 4

3.8.2.2.

CONTROL A/C UNIT NO. 12 IS POWERED FROM EMERGENCY BUS 24. IF AN EDG IS l

UNAVAILABLE FOR EMERGENCY BUS 24, THE LACK OF AN EMERGENCY POWER SOURCE WILL ALSO MAKE CONTROL ROOM A/C UNIT NO. 12 INOPERABLE. IF ONE TRAIN OF CREVS IS INOPERABLE FOR SEVEN DAYS, THE TECHNICAL SPECIFICATIONS REQUIRE OPERATING UNITS TO BE SHUT DOWN. THIS ACTIVITY WILL MAINTAIN THE OPERABILITY OF CONTROL ROOM A/C UNIT No. 12 BY ALLOWING DG 21 TO POWER EMERGENCY BUS 14 AND SELECTED i

LOADS ON EMERGENCY BUS 24. THIS WOULD ALLOW RESTORATION OF CONTROL ROOM A/C j

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NNRB018 NUCLEIS 10/15/1995' Search Procesa Adtoc Report 89 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

UNIT NO. 12 AFTER LOSS OF CONTROL ROOM A/C UNIT NO. 11 DUE TO A FAILURE OF DG 11 OR THE A/C UNIT AFTER A LOSS OF OFFSITE POWER TO BOTH UNITS. THE MANUAL ACTIONS TO PERFORM THE LOAD SHED AND BUS ALIGNMENTS CAN SE PERFORMED AND THE CREVS RESTORED BEFORE THE MCR TEMPERATURE LIMIT OF 104 DEGREES F IS EXCEEDED.

GENERALLY, THIS ACTIVITY WILL BE PERFORMED DURING A UNIT 2 PLANT GUTAGE. THE DESIGN INSTRUCTIONS IDENTIFY PORTIONS OF THIS ACTIVITY WHICH MAY BE PERFORMED DURING NON-0UTAGE CONDITIONS. WITH ONE EXCEPTION, WORK WITHIN THE EACP WILL NOT BE PERFORMED WHEN THE PLANT IS IN A TECHNICAL SPECIFICATION LCO ACTION STATEMENT FOR ANY OF THE EDG'S, THEIR ASSOCIATED EMERGENCY BUSES OR THE OFF SITE POWER SOURCES (I.E., TECHNICAL SPECIFICATION 3.8.1.1 AND 3.8.1.2). THE EXCEPTION WILL BE FOR PLANT MODIFICATIONS PERFORMED IN CONJUNCTION WITH STP M 20 ON DG 12 (SURVEILLANCE REQUIREMENT 4.8.1.1.2.D.1 AND ACTION STATEMENT B OF TECHNICAL SPECIFICATION 3.8.1.2 AND 3.8.2.2).

NEW SSC'S ADDED BY THIS ACTIVITY HAVE BEEN EVALUATED TO ENSURE THE EFFE'.T OF THEIR INSTALLATION (E.G., SEISMIC ADEQUACY OF EXISTING STRUCTURES, HEAT LOADS, CABLE SEPARATION) DO NOT INCREASE THE PROBABILITY OF PREVIOUSLY EVALUATED MALFUNCTIONS. SSC'S ADDED BY THIS ACTIVITY WILL NOT BECOME OPERATIONAL UNTIL TESTING OF DG OC IS COMPLETE. EQUIPMENT IDENTIFIED AS INITIATORS OF ACCIDENTS ARE NOT AFFECTED BY THIS ACTIVITY. THEREFORE, THE PROBABILITY OF PREVIOUSLY EVALUATED MALFUNCTIONS AND ACCIDENTS HAS NOT BEEN INCREASED.

THE W SEQUENCES OF PREVIOUSLY EVALUATED MALFUNCTIONS AND ACCIDENTS HAVE NOT BEEN INCREASED BY THIS ACTIVITY BECAUSE EQUIPMENT REQUIRED TO SERVE MITIGATION FUNCTIONS UNDER THESE CONDITIONS HAVE NOT BEEN AFFECTED, AND THE CONTROL ROOM AND OFFSITE DOSES PREVIOUSLY CALCULATED REMAIN WITHIN THE PREVIOUSLY STATED LIMITS.

INSTALLATION ACTIVITY IS SEQUENCED SUCH THAT AN EDG WILL BE AVAILABLE TO SUPPLY EMERGENCY POWER TO AN ENGINEERED SAFETY FEATURES BUS, OR ELSE THE WORK WILL BE COORDINATED WITH THE PERFORMANCE OF SURVEILLANCE REQUIREMENT 4.8.1.1.2.D.1 ON DG 12. ADEQUATE ELECTRICAL ISOLATION FOR DG OC WILL BE PROVIDED. NO NEW SYSTEMS INTERACTIONS ARE BEING CREATED BY THIS ACTIVITY.

THEREFORE, THE POSSIBILITY OF A NEW MALFUNCTION OR ACCIDENT IS NOT CREATED BY THIS ACTIVITY.

THE MARGIN OF SAFETY EXPRESSES IN THE BASES OF THE TECHNICAL SPECIFICATIONS IS NOT REDUCED BECAUSE TEMPORARY AND PERMANENT MEASURES ARE IN PLACE TO ASSURE THAT PENETRATIONS MADE DO NOT AFFECT RATED FIRE BARRIERS AND MAIN CONTROL ROOM HVAC. DURING PERIODS WHEN PENETRATIONS IN AREAS PROTECTED BY A HALON SUPPRESSION SYSTEM ARE OPEN, THE HALON SYSTEM WILL BE DECLARED INOPERABLE, AND HOURLY FIRE WATCHES AND BACKUP FIRE SUPPRESSION WILL BE INSTITUTED IN ACCORDANCE WITH THE TECHNICAL SPECIFICATIONS. SEQUENCING OF INSTALLATION ACTIVITIES ENSURES THAT EITHER AN EDG WILL BE AVAILABLE TO SUPPLY EMERGENCY POWER TO AN ENGINEERED SAFETY FEATURES BUS AT UNIT 2 ALL TIMES, OR A TEMPORARY DIESEL GENERATOR WILL BE CONNECTED IN ACCORDANCE WITH r

m__._

NMRS018 NUCLEIS 10/15/1995 Search Process A&oc Report 90 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

THE TECHNICAL SPECIFICATIONS. UPON COMPLETION OF THIS ACTIVITY, TWD-i OPERATIONAL EDG'S WILL BE AVAILABLE TO SUPPLY E9ERGENCY POWER TO ENGINEERED SAFETY FEATURES BUSES AT UNIT 2.

THEREFORE, THERE ARE No UNREVIEWED SAFETY QUESTIONS ASSOCIATED WITH THIS

+

ACTIVITY.

l (CNN)

)

Document ID Revision Status

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7 i

95-002%

0000 64 Sthject:

THIS PROCEDURE REVISION CHANGES THE THROTTLED FLOW THROUGH THE 80RON(NIETER OR PROCESS RADIATION MONITOR (PRM)

FROM 0.5 GPM TO 0.75 GPM, WHEN ONLY ONE OF THE DEVICES IS IN SERVICE.

Alias:

POSRC #:

95-048 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: 0 Assoc Type:

ESP-Ref Doc ID:

95-1016 Rev:

0500 Refer Type: PCRESTP ELECT STP PROCEDURE CHANGE REP

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y Text:

SLNNIARY-DURING NORMAL OPERATION THE PROCESS CADIATION MONITOR AIS 80RON(NETER ARE CONNECTED IN PARALLEL WITH A FLOW RATE OF APPROKIMATELY 0.5 GPM TINtOUGH EACN. A LOW FLOW ALARM EXISTS TO ALERT OPERATORS IF FLOW IN THE C(NOION

?

DISCHARGE DROPS BELOW 0.5 GPM. CURRENTLY, IF ONE OF THE DEVICES IS REM 0WED FROM SERVICE, THE LOW FLOW SETPOINT IS IDENTICAL TO THE FLOW RATE THROUGH THE IN-SERVICE INSTRtNIENT AND NUISANCE ALARMS CAN OCCUR.

THIS SAFETY EVALUATION APPROWES INCREASING THE FLOW RATE THROUGH THE IN-SERVICE INSTRUMENT TO 0.75 GPM FROM 0.5 GPM. THE INCREASED FLOW RATE WILL

[

ELIMINATE THE NUISANCE ALARMS SY MAVING THE FLOW TINT 0 UGH A SINGLE INSTRUMENT TO BE GREATER THAN THE SETPOINT.

THIS PROCEDURE CHANGE DOES NOT INCREASE THE PROBASILITY OF AN ACCIDENT OR i

MALFUNCTION OR INCREASE THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION, OR CREATE THE POSSIBILITY FOR A NEW ACCIDENT OR MALFUNCTION, OR REDUCE THE i

MARGIN SAFETY IN THE TECHNICAL SPECIFICATIONS, THEREFORE THIS IS NOT AN k

I h

i

~

NNRB018 NUCLEIS 10/15/1995 Search Process A&oc Report

1. 1 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

UNREVIEWED SAFETY QUESTION.(CMH)

Doctment ID Revision Status na==a=======================================

95-0021 0000 64 Stbject:

THIS NSR TEMPORARY ALTERATION INSTALLS A CROSS CONNECT HOSE FROM THE PLANT AIR (PA) SYSTEM INSIDE CONTAINMENT (2 PA 1042) TO THE IA SYSTEM INSIDE CONTAINMENT (2 IA 407) TO MAINTAIN OPERATION OF THE AFFECTED COMPONENTS RECEIVING IA.

Alias:

POSRC #:

95-050 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: 0 Assoc Type: ESP Ref Doc ID:

Rev:

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Sender Xmtl #

Xmtl Date

============

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Text:

SUMMARY

THIS NSR TEMPORARY ALTERATION INSTALLS A CROSS CONNECT HOSE FROM THE PLANT AIR (PA) SYSTEM INSIDE CONTAINMENT (2 PA 1042) TO THE IA SYSTEM INSIDE CONTAINMENT (2 IA 407) TO MAINTAIN OPERATION OF THE AFFECTED COMPONENTS RECEIVING IA. A COALESCING FILTER WILL BE INSTALLED IN THE CROSS CONNECT HOSE TO ENSURE INSTRUMENT QUALITY AIR IS MAINTAINED.

ISOLATION OF THE NORMAL INSTRUMENT AIR (IA) SUPPLY TO IA LOADS INSIDE UNIT 2 CONTAINMENT DOWNSTREAM OF VALVE 2 CV 2085 IS REQUIRED TO PERFORM MAINTENANCE ON 2 MOV 2080. THE ORIGINAL PLAN WAS TO WORK ON 2 MOV 2080 AFTER THE STEAM GENERATOR N0ZZLE DAMS WERE REMOVED. THIS WOULD ALLOW THE IA HEADER INSIDE CONTAINMENT TO BE REMOVED FROM SERVICE. HOWEVER, TO ELIMINATE POSSIBLE OUTAGE SCHEDULE CONFLICTS OR EXTENSIONS, 2 MOV 2080 WILL BE WORKED WHILE THE N0ZZLE DAMS ARE INSTALLED. THIS REQUIRES A TEMPORARY AIR SUPPLY FOR THE CONTAINMENT IA HEADER.

THE NORMAL IA SYSTEM IS NON SAFETY RELATED AS IS THE TEMPORARY CONFIGURATION.

SUFFICIENT CAPACITY EXISTS IN THE CROSS CONNECTED SYSTEMS TO MEET COMPONENT REQUIREMENTS. THE AIR FROM THE PA SYSTEM WILL BE PASSES THROUGH COALESCING FILTERS TO REMOVE MOISTURE, OIL AND PARTICULATES. THE USE OF FILTERED PA WILL THEREFORE RESULT IN DETRIMENTAL EFFECTS TO TL2 COMPONENTS. THE HOSE AND FITTINGS TO BE USED MEET THE NECESSARY PRESSURE AND TEMPERATURE REQUIREMENTS OF THE AFFECTED PIPING CLASSES. THE IA AND PA SYSTEMS ARE NOT ACCIDENT INITIATORS NOR ARE THEY USED TO MITIGATE THE CONSEQUENCES OF A MALFUNCTION OR ACCIDENT. THE CHANGE TO A TEMPORARY SOURCE OF AIR WILL NOT CAUSE ANY COMPONENTS TO BECOME ACCIDENT INITIATORS. THEREFORE, THIS ACTIVITY DOES NOT

c%e NNRB018 NUCLEIS 10/15/1995 Search Procesa Acfoc Report 92 STATUS 62 OR 64 50.09S (10N1/1994 THRU 09/30/1995)

CREATE AN UNREVIEWED SAFETY QUESTION Ah DEF'dED BY 10 CFR 50.59. (CMH)

Document 10 Revision Status

we========================

==

94-B-102-059-R01 62 Stbiect:

PERMANENT STORAGE OF ADDITIONAL GALVANIZED SCAFFOLDING MATERIAL IN CONTAINMENT Atlas:

POSRC #:

95-051 Assoc Doc ID: 93-102-011-04 Revision To: 0004 Assoc Stat: 0 Assoc Type: MCR Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xmtl Date

=w============================e==================================================================

l Other r:fs:

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System Code: 102 PLANT AREAS Text:

NRC

SUMMARY

THE PREVIOUS REVISION OF THIS SAFETY ANALYSIS ALLOWED THE STORAGE OF GALVANIZED SCAFFOLDING MATERIALS IM SEISMICALLY QUALIFIED RACKS IN THE CONTAINMENT BUILDING OF UNITS 1 AND 2. THIS REVISION OF THE SAFETY ANALYSIS ALSO ALLOWS AN ADDITIONAL AMOUNT OF GALVANIZED GRATING - IN UNIT 2 ONLY. THE GRATING IS AN INTEGRAL PART OF THE SEISMICALLY QUALIFIED SCAFFOLDING RACK.

BASED ON STORAGE CRITERIA, NO INTERACTIONS WILL OCCUR WITH OTHER EQUIPMENT.

BEFORE THE PREVIOUS REVISION OF THIS SAFETY ANALYSIS, THE UFSAR DID NOT EXPLICITLY STATE THAT GALVANIZED SCAFFOLDING MATERIAL WAS ANALYZED IN THE HYDROGEN ACCUMULATION IN CONTAINMENT DESIGN EVENT. THOUGH GALVANIZED GRATING WAS EXPLICITLY ANALYZED, THIS ACTIVITY ADCS AN ADDITIONAL AMOUNT IN UNIT 2. THE REANALYSIS OF THE EVENT RE ALLOCATES MARGIN FOR THE HYDROGEN PRODUCTION DUE TO CONTAINMENT METAL CORROS!04 TO ACCOUNT FOR THE j

INCLUSION OF THESE MATERIALS. THE TOTAL HYDROGEN PRODUCED REMAINS UNCHANGED.

WITH THE HYDROGEN RECOMBINERS STARTED WIMIN ONE DAY OR HYDROGEN PURGE STARTED AT 3 7 V / 0 OR 9 55 DATS (WHICHEVER IS LATER). THE MAXIMLM HYDROGEN CONCENTRATION WILL NOT EXCEED 4 0 V / O. WITH THE PURGE STARTED AFTER 9 55 DAYS, THE RESULTS OF THE OFF SITE DOSE EVALUATION FOR THIS EVENT REMAIN BOUNDING. THEREFORE, THIS EVALUATION SHOWS THAT THE LICENSING BASIS OF THIS EVENT IS MAINTAINED AND NO UNREVIEWED SAFETY QUESTION IS INVOLVED.

(CMH)

unt9018 NUCLEIS 10/15i1995 i

Search Proces3 A& oc Report 93 i

r STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

L I

f I

Document ID Revision Status

================================

95-0035 0000 64

Subject:

THE PROPOSED ACTIVITY DEFEATS THE AUTOMATIC ACTUATION SIGNAL TO #11 SW PP START CIRCUIT....

l Alias:

POSRC #:

95-053 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: 0 Assoc Type: ESP Ref Doc ID:

Rev:

Refer Type:

Sender Matl #

Matt Date

===================================================================================================

i Other refs:

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Org/Div:

System Code: 012 SALT WATER COOLING l

i t

.1 m.

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-m=

mm m

NWR5018 Search Procesa Adioc Report 94 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

Text:

Stp54ARY:

TEMPORARY ALTERATION TA 1 95 0028 ALIGNS NO 13 SALTWATER (SW) PUMP TO NO.11 SALTWATER HEADER AND DEFEATS THE AUTOMATIC ACTUATION SIGNAL TO THE NO.11 SALTWATER PUMP. ADDITIONALLY, THE ELECTRICAL INTERLOCK BETWEEN 1HE BREAKER FOR NO 11 AND 13 SW PUMP IS JUMPERED OUT. THIS ACTIVITY ALLOWS No 11 SW PUMP TO BE INOPERABLE TO HAVE ITS ASME SECTION XI TESTING PERFORMED WHILE NO.

13 SW PUMP IS ALIGNED TO SALTWATER HEADER NO 11 IN STAND 8Y. HAVING AN OPERABLE SW PUMP IS A REQUIREMENT TO MAINTAIN AN OPERABLE SALTWATER TRAIN.

SHOULD A UV AND A SIAS SIGNAL OCCUR, BOTH NO 11 AND 13 SW PUMPS WILL BE LOAD SHED AND ONLY NO. 13 SW PUMP WILL BE LOADED VIA THE SEQUENCER. IF A SIAS OCCURS WITHOUT A UV, 13 SW WILL START AFTER A ONE SECOND DELAY AND THE INOPERA8LE 11 SW PUMP WILL CONTINUE TO RUN. AT NO TIME DURING THE TA CAN TWO SW PUMPS BE AUTOMATICALLY STARTED AND LOADED ONTO THE EDG. EDG 11 REMAINS OPERA 8LE OURING THIS ACTIVITY. SINCE THERE ARE TWO INDEPENDENT OPERABLE SALTWATER HEADERS CAPABLE OF MEETING THEIR DESIGN REQUIREMENT TO PROVIDE COOLING WATER TO EQUIPMENT IMPORTANT TO SAFETY THIS ACTIVITY DOES NOT CONSTITUTE A USQ OR REQUIRE ANY CHANGES TO THE TECHNICAL SPECIFICATIONS.

(CMH)

~

v NNR8018 NUCLEIS 10/15/1995 Search Process A &oc Report 95 i

y STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

Document ID Revision Status satzussnesszzzzzzzzzzzzzzzzzzzzz saz=====

mzzzzz l

95-0027 0001 64 S4]ect:

TA 2 95 0089 ALLOWS INSTALLATION OF A MECHANICAL GAGGING DEVICE ON THE AIR ACTUATOR STEM OF 2 CV 3828 IN THE COMPONENT COOLING SYSTEM Alles:

POSRC #:

95-056 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: 0 Assoc Type: ESP Ref Doc ID:

Rev:

Refer Type:

s Sender Matl #

Xatt Date

mmz==========m::===========zzzz============zz=======mmazzzzzzzzzzzzz=zzzzzzzzzzzzz mzzzzzzzzzzz========

?

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Text:

St.NMARY:

TEMPORARY ALTERATION 2 95 0089 ALLOWS INSTALLATION OF A MECHANICAL GAGGING DEVICE ON THE 21 SDCMX CCW OUTLET VALVE 2 CV 3828 WHILE THE 21 SDCMX IS TAGGED FOR MAINTENANCE. THE INSTALLATION OF THE GAGGING DEVICE WILL ENSURE i

THE SR P8 FUNCTION OF THIS VALVE, AS IT SERVES AS THE ONLY TAGOUT BOUNDARY ON THE CCW OUTLET LINE, WILL NOT BE LOST FOLLOWING A LOSS OF INSTRUMENT AIR TO THE VALVE, AND TMAT THE GAGGING DEVICE CAN WITHSTAND THE M CHANICAL LOADING IT WILL BE SUBJECTED TO FOLLOWING THIS LOSS OF AIR. THE DETAILS OF THE ENGINEERING EVALUATION OF THE GAGGING DEVICE ARE CONTAINED IN THE ENGINEERING EVALUATION SECTION OF TA 2 95 0089. INSTALLATION OF THE GAGGING DEVICE DOES NOT INCREASE THE PROBASILITY OF A MAL 7UNCT10N, ACCIDENT, NEW MALFUNCTION, OR NEW ACCIDENT NOT PREVIOUSLY ANALYZED IN THE SAR. FURTMERMORE, THE CONSEQUENCE OF THE PREVIOUSLY DISCUSSED MALFUNCTIONS AND ACCIDENTS ARE NOT INCREASED.

THIS ACTIVITY DOES NOT CONSTITUTE AN UNRESOLVED SAFETY QUESTION AND DOES NOT VIOLATE TECHNICAL SPECIFICATIONS.

(CMH) 1 I

5

NNRB018 NUCLEIS 10/15/1995 SearchProcesaAdtocReporj STATUS 62 OR M 50.59S (10/01/1994 THRU 09/30/1995) l' i

Doctment ID Revision Status

================================

95-0036 0000 M

Subject:

TA 1 95 0031 INSTALLS INLET AND OUTLET PRESSURE G'UGES ON BASKET STRAINER 1 BS 5205 TO TEMPORARILY REPLACE THE EXISTING BROKEN DIFFERENTIAL PRESSURE INDICATING SWITCH 1 PDIS 5205.

Alias:

POSRC #:

95-056 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: 0 Assoc Type: ESP Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xmtl Date

===================================================================================================

Oth1r rsfs:

Pers Rzfs:

a Equipment:

i Org/Div:

System Code: 012 SALT WATER COOLING Text:

SUPWIARY:

TEMPORARY ALTERATION 1 95 0031 INSTALLS INLET AND OUTLET PRESSURE GAUGES ON BASKET STRAINER 1 BS 5205 TO TEMPORARILY REPLACE THE EXISTING BROKEN DIFFERENTIAL PRESSURE INDICATING SWITCH 1 PDIS 5205.

THIS ACTIVITY DOES NOT REPRESENT A USQ SINCE THE INSTRUNENTS WILL BE INSTALLED IN ACCCRDANCE WITH APPLICABLE DESIGN CRITERIA TO MAINTAIN THEIR St PB CLASSIFICATION.

THE CONSEQUENCES OR PROBABILITY OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY AND THE POSSIBLITY OF A NEW ACCIDENT OR MALFUNCTION ARE NOT INCREASED OR CREATED, ADDITIONALLY THERE ARE NO CHANGES TO THE TECHNICAL SPECIFICATION BASES AS A RESL*LT OF THIS ACTIVITY.

(CMH)

NMRB018 NUCLEIS 10/15/1995 Search Process Adioc Report 97 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

L Document ID Revision Status

================================

95-0037 0000 64

Subject:

TA 1 95 0032 INSTALLS INLET AND OUTLET PRESSURE GAUGES ON BASKET STRAINER 1 BS 5207 TO TEMPORARILY REPLACE THE EXISTING BROKEN DIFFERENTIAL PRESSURE INDICATING SWITCH 1 PDIS 5207 1

Atlas:

POSRC #:

95-056 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: O Assoc Type: ESP Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xmtl Date

===================================================================================================

Other refs:

P;rs RIfs:

Equipment:

Org/Div:

System Code: 012 SALT WATER COOLING Text:

SupWIARY:

TEMPORARY ALTERATION 1 95 0032 INSTALLS INLET AND OUTLET PRESSURE GAUGES ON BASKET STRAINER 1 BS 5207 TO TEMPORARILY REPLACE THE EXISTING BROKEN DIFFERENTIAL PRESSURE INDICATING SWITCH 1 PDIS 5207.

THIS ACTIVITY DOES NOT REPRESENT USQ SINCE THE INSTRUNENTS WILL BE INSTALLED IN ACCORDANCE WITH APPLICABLE DESIGN CRITERIA TO MAINTAIN THEIR SR PB CLASSIFICATION.

l THE CONSEQUENCES OR PROBABILITY OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY AND THE POSSIBILITY OF A NEW ACCIDENT OR MALFUNCTION ARE NOT INCREASED OR CREATED, ADDITIONALLY THERE ARE NO CHANGES TO THE TECHNICAL SPECIFICATION BASES AS A RESULT OF THIS ACTIVITY.

i m

NWR8018 NUCLEIS 10/15/1995 Search Process Adioc Report 98 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

(CMH)

Docunent ID Revision Status

================================

95-0006 0002 64 S4]ect:

THIS IS AN EVALUATION OF THE DESIGN AND PERFORMANCE OF THE CALVERT CLIFFS UNIT 2 REACTOR FOR THE OPERATION OF CYCLE 11 AT THE FULL RATED POWER OF 2700 MW Alias:

{

POSRC #:

95-061 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: 0 Assoc Type: ESP Ref Doc ID:

Rev:

Refer Type Sender Xmtt #

Xmtt Date

===================================================================================================

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Text:

SUMMARY

THIS SAFETY EVAUETION CONSIDERED THE OPERATION OF UNIT 2 CYCLE 11 MODIFICATIONS TO TPZ FUEL RCD AND THE FUEL ASSEMBLY AND THE RELOAD CORE DESIGN WERE CONSIDfiRED. THE USE OF ERBIUM FOR UNIT 2 AS A BURNABLE ABSORSER WAS CONSIDERED. THE UNIT 2 CYCLE 11 ANALYSIS ASSUMED AN ALLOWED PEAK LINEAR HEAT RATE OF UNIT 2 CYCLE 11 CORE OPERATING LIMITS REPORT (COLR). THE PRE f

. m

-..m.

- - - - ~.. -. ~

- +

NWRB018 NUCLEIS 10/15/1995 Search Procesa AcNioc Report 99 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

TRIP STEAM LINE BREAK EVENT WAS REANALYZED USING METHODOLOGY DESCRIBED IN THE SAR TO PREDICT THE PERCENTAGE OF FUEL FAILURES. THE RE ANALYSIS ASSUMED MORE RESTRICT!YE F AND F LIMITS IN ORDER TO RESTRICT THE PREDICTED NUMBER OF FUEL FAILURES TO A LESS LIMITING PERCENTAGE THAN THAT PREVIOUSLY REPORTED. THE F AND F LIMITS ARE EQUAL TO 1.64 AND WILL BE IMPLEMENTED IN THE UNIT 2 CYCLE 11 COLR. THE THERMAL POWER MEASUREMENT UNCERTAINTY FOR THE INCORE DETECTOR MONITORING SYSTEM FOR UNIT 2 CYCLE 11 COLR. THE THERMAL POWER MEASUREMENT UNCERTAINTY FOR THE INCORE DETECTOR MONITORING SYSTEM FOR UNIT 2 CYCLE 11 IS 1. 035 FOR MEASURED THERMAL POWER LESS THAN OR EQUAL TO 50 PERCENT BUT GREATER THAN 20 PERCENT OF RATED FULL CORE POWER AND 1. 020 FOR MEASURED t

THERMAL POWER GREATER THAN 50 PERCENT OF RATED FULL CORE POWER. THIS CHANGE WILL BE IMPLEMENTED IN THE UNIT 2 CYCLE 11 COLR. THE UNIT 2 CYCLE 11 SAFETY ANALYSES ACCOUNTED FOR ALL RELOAD CORE DIFFERENCES. THE RESULTS OF ALL REFERENCE SAFETY ANALYSES CONSERVATIVELY APPLY To UNIT 2 CYCLE 11. IT IS CONCLUDED THAT OPERATION OF UNIT 2 CYCLE 11 DOES NOT INVOLVE AN UNREVIEWED SAFETY QUESTION. (CMH) l l

l

NMRB018 NUCLEIS 10/15/1995 Search Process A& oc Report 100 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

Document ID Revision Status

mewass==================================

94-8-064-090-R00 64

Subject:

THE SAFETY ANALYSIS OF CHAPTER 14 MUST BE REVISED TO ACCOUNT FOR AN INCREASED PSV TOLERANCE BAND. TECH SPECS 3.4.2.1 MUST BE AMENDED.

Allas:

94-B-064-090-R00 POSRC #:

95-067 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat:

C Assoc Type: ESP Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xmtl Date

===================================================================================================

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System Code: 064 REACTOR COOLANT Text:

SUMMARY

THIS ACTIVITY INCREASES THE PRESSURIZER SAFETY VALVE SETPolNT TOLERANCE ASSUMED IN THE UFSAR FROM +/-1% FOR BOTH VALVES TO +2%/-1% FOR RC-200 AND TO

+/-2% FOR RC-201. THE IMPACT OF THIS CHANGE IS REFLECTED IN RE-ANALYSIS OF THE FOLLOWING DESIGN BASIS EVENTS: LOSS OF LOAD, FEEDWATER LINE BREAK, AND LOSS OF FEEDWATER FLOW. THE REMAINING UFSAR EVENTS WERE EVALUATED AND AND DETERMINED NOT TO BE IMPACTED BY INCREASING THE SETPOINT TOLERANCE ON THESE VALVES. THE RESULTS OF THOSE EVENTS WHICH WERE RE - ANALYZED ARE SHOWN TO BE WITHIN THE ACCEPTANCE CRITERIA PREVIOUSLY REVIEWED AND ACCEPTED BY THE NRC FOR THE APPLICABLE EVENTS. THIS ANALYSIS RESULT WAS ACCOMPLISHED PRIMARILY BY RE - ALLOCATING EXISTING ANALYSIS MARGIN IN THE ANALYSES FOR THESE EVENTS. SPECIFICALLY, THE INPUTS ASSUMED FOR MTC ADN MSSV LIFT SETTINGS (FOR THE LOSS OF LOAD EVENT) WERE REVISED TO BE CONSITENT WITH TECH. SPEC. REQUIREMENTS. THEREFORE, INCORPORATION OF AN INCREASED UPPER SETPOINT TOLERANCE FOR THE PRESSURIZER SAFETY VALVES INTO UFSAR ANALYSIS WAS FOUND NOT TO CONSTITUTE AN UNREVIEWED SAFETY QUESTION. THIS ACTIVITY DOES NOT ALTER CURRENT TECH. SPEC. SURVEILLANCE REQUIREMENTS THAT THESE VALVES LIFT WITHIN +/-1% OF THEIR TS SETPOINTS.

1

NNRB018 NUCLEIS 10/15/1995 Search Proces2 A& oc Report 101 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

Docunent ID Revision Status

mas===================================

95-0034 0000 64

Subject:

THE LOOSE PARTS MONITORING SYSTEM (LPMS) IS A NON SAFETY RELATED (NSR) SYSTEM WITH NO TECHNICAL SPECIFICATION REQUIREMENT.

Alias:

POSRC #:

95-068 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: O Assoc Type: ESP Cef Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xmtl Date

=================================================================================================

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Text:

St.MMARY:

FCR 94 0202 REPLACES THE LOOSE PARTS MONITORING SYSTEM. AN UNREVIEWED SAFETY QUESTION IS NOT CREATED BECAUSE THE REPLACEMENT IN CONTAINMENT COMPONENTS ARE OF THE SAME BASIC DESIGN AS THE EXISTING COMPONENTS AND WILL BE INSTALLED PER APPLICA8LE CODES AND STANDARDS. THE REPLACEMENT CONTROL ROOM COMPONENT RECEIVES THOROUGH TESTING OF THE SYSTEM PERFORMANCE AGAINST ThE SYSTEM REQUIREMENTS AND HAS ACHIEVED A RELIABLE OPERATING HISTORY IN OTHER NUCLEAR POWER PLANTS.

(CMH) i r

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v er-~

~ -, -

e

~

i r

NNR8018 NUCLEIS 10/15/1995 Search Procesa Adioc Report 102 1

STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

~I i

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t 1

i

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i i

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t i

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I i

I L

Document ID Revision Status maatraoocmzzz=sszzzzzz= ssma==== mz=mazza muzzzz

(

95-0039 0000 64 l

S4]ect:

INSTALLATION OF MANUAL ISOLATION VALVES IN THE SRW SUPPLY AND RETURN LINES FOR UNIT 1 AND UNIT 2 PLANT AIR COMPRESSORS AFTERCOOLERS.

i Atlas:

I POSRC #:

95-069 I

f O

h

.. ~

...m---

NNRB018 NUCLEIS 10/15/1995 Search Process A& oc Report 103 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

Assoc Doc ID: 89-0173 Revision To: 0000 Assoc Stat: 0 Assoc Type:

FCR 89-0173-01 0000 0

ESP Ref Doc 10:

Rev:

Refer Type:

Sender Xmtl #

Xmtl Date

_. - _ - - - - - - - - = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = =========================

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Text:

SUMMARY

THIS ACTIVITY INSTALLS MANUAL ISOLATION VALVES IN THE NON SAFETY RELATED SERVICE WATER (SRW) SUPPLY AND RETURN LINES ASSOCIATED WITH THE PLANT AIR COMPRESSORS AND AFTERCOOLERS. THESE VALVES ARE BEING INSTALLED TO FACILITATE AIR COMPRESSOR MAINTENANCE AND MODIFICATION ACTIVITIES. ADDITION OF THESE VALVES DOES NOT IMPACT SRW SYSTEM FLOWRATES NOR CHANGE THE OPERATION OF THE SRW SYSTEM DURING NORMAL AND ACCIDENT CONDITIONS. THIS CHANGE DOES NOT REPPESENT AN UNREVIEWED SAFETY QUESTION (USQ) NOR REDUCE THE MARGIN OF SAFETY AS DEFINED IN THE BASES FOR ANY TECHNICAL SPECIFICATION. NO CHANGES TO THE 6

TECHNICAL SPECIFICATIONS ARE REQUIRED. (CMM) t

?

i Document 10 Revision Status

================================

95-0040 0000 64 l

Subject:

95 013 002 00 REMOVES FIRE PROTECTION WATER SUPPLY PIPING SETWEEN 10B VALVE ROOM AND THE SPRINKLERS IN THE 3 ADJACENT TRAILERS.

Alias:

POSRC #:

95-069 Assoc Doc ID: 95-013-002-00 Revision To: 0000 Assoc Stat: C Assoc Type: MCR Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xmtl Date

===================================================================================================

.NMRe018 NUCLEIS'

'10/15/1995~

Search Procesa A & oc Report 106' STATUS 62 OR 66 50.59S (10/01/1996 THRU 09/30/1995)

-Other rsfs:

Pers RIfs:

Equipment:

Org/Div:

System Code: 013 FIRE PROTECTION Text:

SUMMARY

THE PROPOSED ACTIVitt REMOVES SPRI KLER WATER SUPPLY PIPING WHICH SUPPLIES SPRI K LERS IN THREE TRAILERS ADJACENT TO THE INTERIM OFFICE BUILDING. THE TRAILERS ARE BEING REMOVED, SO THAT THE WATER SUPPLY TO THE TRAILER-SPRINKLERS IS NO LONGER REQUIRED. THIS CHANGE AFFECTS THE UFSAR SINGLE LINE DRAWING MAS 8EEN REVISED TO ELIMINATE THE REFERENCE TO THE TRAILERS.

(CMM)

Document ID Revision Status

================================

91-B-019-131 0001 62

Subject:

FCR 89-016 DELETES THE CONTAINMENT BREATHING AIR PURIFIERS / DRYERS FROM UNIT 1 AND 2 AND PACKAGES THE PURIFIERS SUCH THAT THEY CAN BE INSTALLED AT SOME LATER TIME INTO THE AUXILIARY BUILDING.

Alias:

POSRC #:

95-081 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: C Assoc Type: ESP Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xmtl Date j

===================================================================================================

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Org/Div:

4 System Code:

Text:

SupMARY:

FCR 89 0016 00 DELETES THE CONTAINMENT BREATHING AIR PURIFIERS / DRYERS FROM UNIT 1 AND 2. THE RESULTING PIPING CONFIGURATION WAS ANALYZED AND DETERMINED TO BE WITHIN DESIGN AIQ CCOE REQUIREMENTS. THESE CHAhGES DO NOT CONSTITUTE A USS AS THE BREATHING AIR PURIFIERS ARE NOT REQUIRED TO FUNCTION DURING OR I

I I

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..n.-

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NMRs018 NUCLEIS 10/15/1995 Search Process Adioc Report 105 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

AFTER AN ACCIDENT. ALSO, THEY ARE WITHIN ALL DESIGN AND CODE ALLOWA8LES.

THIS REVISION TO TPE 50.59 SAFETY EVALUATION IS ISSUED TO REFLECT THE AS BUILT PIPING CONFIGURATION. THESE CHANGES DO NOT ALTER THE CONCLUSION OF THE SAFETY EVALUATION.

(CMH)

L Doctment ID Revision Status

================================

91-B-019-131-R01 62

Subject:

DELTE CONTAINMENT BREATHING AIR PURIFIERS / DRYERS IAW FEC 89 0016 0001 i

Atlas:

POSRC #:

95-081 Assoc Doc ID: 89-0016 Revision To: 0000 Assoc Stat: 0 Assoc Type:

FCR 89-0016-0001 0000 0

FEC Ref Doc ID:

Rev:

Refer Type:

Sender Xmtt #

Xmtl Date

==================================================================================================

Other refs:

Pcrs RIfs:

Equipment:

b we

i NNRB018 NUCLEIS 10/15/1995 Search Froces3 Actioc Report 106 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995) org/Div:

l System Code: 019 COMPRESSED AIR Text:

NRC SUMARY:

i FCR 89 0016 DELETES THE CONTAINMENT BREATHING AIR PURIFIERS / DRYERS FROM UNIT 1 AND 2. THE RESULTING PIPING CONFIGURATION WAS ANALYZED AND DETERMINED l

TO BE WITHIN DESIGN AND CODE REQUIREMENTS. THESE CHANGES DO NOT CONSTITUE A USQ AS THE BREATHING AIR PURIFIERS ARE NOT REQUIRED TO FUNCTION DURING OR AFTER AN ACCIDENT. ALSO, THEY ARE WITHIN ALL DESIGN AND CODE ALLOWABLES.

THIS REVISION TO THE 50.59 SAFETY EVALUATION IS ISSUED TO REFLECT THE AS BUILT PIPING CONFIGURATION. THESE CHANGES DO NOT ALTER THE CONCLUSION OF THE SAFETY EVALUATION.

(CMH) l Docunent ID Revision Status

,om=====================

==

95-0044 0000 64

Subject:

THE PROPOSED ACTIVITY INSTALLS A JUMPER IN THE ANNUNCIATOR CABINET, 2K02, TO JUMPER OUT THE ALARM CONTACT OF AN OUT OF SERVICE BATTERY CHARGER.

Alias:

POSRC #:

95-081 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: C Assoc Type: ESP Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xmtl Date

_m======

==

other rsfs:

Pers Rifs:

Equipment:

Org/Div:

System Code:

Text:

SUMARY:

THE PROPOSED ACTIVITY INSTALLS A TEMPORARY J'JMPER IN THE ALARM CIRCUIT OF THE BATTERY CHARGERS. THE JUMPER WILL ONLY BE INSTALLED ON A CHARGER THAT IS REMOVED FROM SERVICE. THE JUMPER WILL ENSURE THAT THE OPERATING CHARGERS HAVE AN ALARM TO ALERT OPERATIONS THAT A PROBLEM EXISTS WITH ONE OF THE OPERATING CHARGERS. THE JUMPER WILL BE REMOVED ONCE THE BATTERY CHARGER HAS BEEN PLACED BACK INTO SERVICE. THIS TEMPORARY ALTERATION PREVENTS A HANGING ALARM IN THE CONTROL ROOM FROM MASKING THE STATUS OF THE OPERATING BATTERY CHARGERS.

THIS ACTIVITY DOES NOT CONiTITUTE AN UNREVIEWED SAFETY QUESTION (USQ).

(CMH)

NNRB018 NUCLEIS 10/15/1995 Search Procesa A& oc Report 107 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

I Doctment ID Revision Status

================================

94-B-999-102-R00 64 Stbject:

THE EXISTING ELECTRICAL DISTRIBUTION SYSTEM IS TO BE MODIFIED TO DEDICATE EDG 12 TO ENGINEERED SAFETY FEATURES BUS 14 IN UNIT 1.

Alias:

POSRC #:

95-084 Assoc Doc ID: 89-0079 Revision To: 0000 Assoc Stat: 0 Assoc Type:

FCR Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xmtl Date

===================================================================================================

OthIr rsfs:

Pers Rafs:

Equipnent:

Org/Div:

System Code: 024 EMERGENCY DIESEL GENERATOR Text:

SUMMARY

THE EXISTING ELECTRICAL DISTRIBUTION SYSTEM IS TO BE MODIFIED TO DEDICATE EMERGENCY DIESEL GENERATOR 12 (DG 12) TO ENGINEERED SAFETY FEATURES BUS 14 IN UNIT 1. DG 12 IS TO BE REDESIGNATED AS DG 1B AND ITS ASSOCIATED SUPPORT SYSTEMS WILL BE REDESIGNATED TO REFLECT THE DIESEL'S DEDICATION TO UNIT 1.

DG 12 (DG 18) AUTOMATIC START AND LOADING CIRCUITS WILL BE M00!FIED TO DELETE THE BUS UNDERVOLTAGE SIGNAL ASSOCIATED WITH EMERGENCY BUS 21 AND UNIT 2 SIAS SIGNALS. THIS MODIFICATION WILL PREVENT AUTOMATIC ALIGNMENT OF DG 12 (DG 1B)

TO EMERGENCY BUS 21. HOWEVER, DG 12 (DG 18) WILL BE AVAILABLE FOR MANUAL CONNECTION TO EMERGENCY BUS 21 TO FUNCTION AS A POWER SOURCE TO SHUTDOWN UNIT 2 IN 1HE EVENT OF A FIRE.

THIS ACTIVITY WILL ALSO REVISE THE TECHNICAL SPECIFICATIONS TO REFLECT THE ELIMINATION OF DG 12 S (DG 1B S) SWING CAPABILITY.

UNIT 1 WILL BE IN MODE 5, 6 OR DEFUELED THROUGHOUT IMPLEMENTATION OF THE PORT!aNS OF THIS ACTIVITY THAT ARE ASSOCIATED WITH CHANGES IN SYSTEM OPERATION. UNIT 2 IS CONSIDERED TO BE IN MODE 1, 2, 3, 4, 5 OR 6. AT LEAST 23 FEET OF WATER WILL BE MAINTAINED OVER THE IRRADIATED FUEL ASSEMBLIES SEATED WITHIN THE REACTOR PRESSURE VESSEL WHILE UNIT 1 IS IN MODE 6 AND A

- - =

NNRB018 NUCLEIS 10/15/1995 Search Proces) A & oc Report 108 STATUS 62 OR 64 50.59s (10/01/1994 THRU 09/30/1995)

UNIT 1 ENGINEERED SAFETY FEATURES BUS IS OUf 0F SERVICE. PORTIONS OF THIS ACTIVITY THAT MAY BE PERFORMED DURING NON OUTAGE CONDITIONS ARE LIMITED TO WORK THAT CAN BE PERFORMED WITHOUT CREATING SYSTEM FUNCTIONAL AND OPERATIONAL l

CHANGES.

MODIFICATIONS IMPLEMENTED BY THIS ACTIVITY WERE EVALUATED TO ENSURE THEY DO NOT INCREASE THE PROBABILITY OF A MALFUNCTION OF EQUIPMENT IMPORTANT TO l

SAFETY. EQUIPMENT IDENTIFIED AS INITIATORS OF ACCIDENTS ARE NOT AFFECTED BY THIS ACTIVITY. THEREFORE, THE PROBABILITY OF PREVIOUSLY EVALUATED MALFUNCTIONS AND ACCIDENTS HAS NOT BEEN INCREASED.

THE CONSEQUENCES OF PREVIOUSLY EVALUATED MALFUNCTIONS AND ACCIDENTS HAVE NOT BEEN INCREASED BY THIS ACTIVITY BECAUSE EQUIPMENT REQUIRED TO SERVE MITIGATION FUNCTIONS UNDER THESE CONDITIONS HAVE NOT BEEN ADVERSELY AFFECTED, AND CONTROL ROOM AND OFFSITE DOSES PREVIOUSLY CALCULATED REMAIN UNCHANGED AND WITHIN THE PREVIOUSLY STATED LIMITS. ONE EDG WILL REMAIN AVAILABLE FOR A SHUTDOWN UNIT AND TWO EDG'S WILL BE AVAILABLE FOR A UNIT OPERATING IN MODES 1 THROUGH 4. IN ADDITION, WHEN OPERATING TWO UNITS, TWO EDG'S WILL BE AVAILABLE FOR EACH UNIT. PROCEDURAL CHANGES TO THE EDG SERVICE WATER SUBSYSTEMS WILL NOT AFFECT THE FLOW OF SERVICE WATER TO OTHER SSC'S WHICH FUNCTION TO MITIGATE THE CONSEQUENCES OF AN ACCIDENT OF MALFUNCTION.

ADMINISTRATIVE CONTROLS PLACED ON THE SERVICE WATER SUBSYSTEMS ENSURE THAT A FAILURE OF A UNIT 2 SERVICE WATER SUBSYSTEM WILL NOT AFFECT THE OPERABILITY OF DG 12 (DG 18), NOW DEDICATED TO UNIT 1. NO NEW SYSTEMS INTERACTIONS ARE BEING CREATED BY THIS ACTIVITY. THEREFORE, THE POSSIBILITY OF A NEW MALFUNCTION OR ACCIDENT IS NOT CREATED BY THIS ACTIVITY.

THE MARGIN OF SAFETY EXPRESSED IN THE BASES OF THE TECHNICAL SPECIFICATIONS IS NOT REDUCED BECAUSE THE REQUIRED NUMBER OF EDG'S WILL BE AVAILABLE TO SUPPLY EMERGENCY POWER TO AN ENGINEERED SAFETY FEATURES BUS IN ACCORDANCE WITH THE TECHNICAL SPECIFICATIONS. UPON COMPLETION OF THIS ACTIVITY, ONE OPERATIONAL EDG WILL BE AVAILABLE TO SUPPLY EMERGENCY POWER TO EACH OF THE TWO ENGINEERED SAFETY FEATURES BUSES AT EACH UNIT. THE REQUIREMENTS OF THE TECHNICAL SPECIFICATIONS WILL BE IMPLEMENTED WHEN FIRE BARRIERS ARE PENETRATED OR SUPPRESSION SYSTEMS ARE DISABLED.

t THEREFORE, THERE ARE NO UNREVIEWED SAFETY QUESTIONS ASSOCIATED WITH THIS ACTIVITY.

(CMH)

4 y

y NMRS018 NUCLEIS

' 10/15/1995 Search Procesa A s oc Report 109 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

I

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f i-Document ID Revision Status

[

================================

. I 95-0032 0000 64 S4 ject:

CHANGE TO UFSAR SECTION 14.15 INVOLVES REANALYSIS OF THE STM GEN TUBE RUPTURE EVENT l

Alias:

POSRC #:

95-066 Assoc Doc ID: ES9300001 Revision To: 0000 Asscc Stat:

C Assoc Type: ESP Ref Doc ID:

Rev:

Refer Type:

j Sender Xmtl 8 Xmtl Date l

===================================================================================================

- l I

1 t

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NMRB018 NUCLEIS 10/15/1995 Search Froces3 AcRioc Report 110 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

Other rsfs:

Pers R;fs:

Equipment:

Org/Div:

System Code:

Text:

SUMMARY

THIS ACTIVITY INVOLVES A CHANGE TO THE UFSAR SECTION 14.15, STEAM GENERATOR TUBE RUPTURE (SGTR). THE CHANGE INVOLVES A RE-ANALYSIS OF THE SGTR EVENT AND A RE4 RITE OF SECTION 14.15, PRIMARILY TO PROVIDE FOR USE OF THE ADVS BEYOND WHAT WAS PREVIOUSLY ASSUMED. IN ADDITION, OTHER CONSERVATIVE ASSLMPTIONS WERE LOSS OF FORCED RCS CIRCULATION 3 SECONDS AFTER PLANT TRIP, NO OPERATOR ACTION FOR li MINUTES FOLLOWING REACTOR TRIP, SIMULATION OF PLANT C00LDOWN i

CONF 4 STENT WITH THE E0P AND USE OF A CONSERVATIVE GENERATED IODINE SPIKE. THE 8'-ANALYSIS RESULTS IN AN OFF SITE DOSE WHICH IS LESS THAN THE VALUE CALCULATED BY THE NRC STAFF FOR UNIT 1, CYCLE 6 AND WITHIN THE ACCEPTANCE LIMITS WH!iH ARE LESS THAN 10% OF THE GUIDELINES OF 10CFR100.

THC CHANGES DO NOT INVOLVE AN UNREVIEWED SAFETY QUESTION SINCE THE RESULTING OFF-SITE DOSE CONSEQUENCES ARE WITHIN THE ACCEPTANCE LIMITS. (CMH)

Document ID Revision Status

================================

95-0047 0000 64

Subject:

THIS ACTIVITY REPLACES THE SR AMERON COATING SYSTEM SPECIFIED WITHIN TRD A 1000 TO BE U'ED ON CAR 8ON STEEL l

SUCH AS THE CTMT LINER, ETC WITH SR VALSPAR COATING SYSTEM WICH IS CURRENTLY USED WITHING CTMT ON PIPING AND ASSOCIATED SUPPORTS Alies:

POSRC #:

95-086 i

NMRB018 NUCLEIS 10/15/1995 Search Process A& oc Report til STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

Assoc Doc ID: ES199501038 Revision To: 0000 Assoc Stat:

C Assoc Type: ESP Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xmtl Date

===================================================================================================

l OthIr rsfs:

P%rs Rifs:

Equipment:

Org/Div:

System Code:

Text:

SUMMARY

THIS ACTIVITY, ESP ES 1995 1038, MAINTAINS THE AVAILABILITY OF SAFETY I

RELATED COATINGS FOR USd WITHIN CONTAINMENT. SECTION 6.4 0F THE UFSAR STATES THAT "ALL VALVES IN THE CONTAINMENT ARE CONSTRUCTED OF STAINLESS STEEL OR DIMETCOTE PAINTED CARBON STEEL, NEITHER OF WHICH CORRODES". AMERON IS DIMETCOTE (PRIMER) ALONG WITH AMERON 66 (EP0XY TOP COAT) ARE NO LONGER AVAILABLE FROM THE MANUFACTURER FOR USE WITHIN CCNPP CONTAINMENTS. THIS ACTIVITY REPLACES THE SAFETY RELATED AMERON COATING SYSTEM SPECIFIED WITHIN TRO A 1000 (CCNPP COATING APPLICATION PERFORMANCE STANDARD) TO BE USED ON VARIOUS CARBON STEEL SURFACES WITH THE SAFETY RELATED VALSPAR COATING SYSTEM (13 F 12 / 89). THIS ACTIVITY ALSO REPLACES THE SAFETY RELATED AMERON COATING SYSTEM (110 AA / 66) SPECIFIED WITHIN TRD A 1000 TO BE USED ON CONCRETE WITH THE SAFETY RELATED CARBOLINE COATING SYSTEM (2011 S / 890).

THE VALSPAR COATING SYSTEM WILL PROVIDE EXCELLENT PROTECTION AGAINST CORROSION. THE VALSPAR AND CARBOLINE COATING SYSTEMS WILL REMAIN INTACT AFTER EXPERIENCING THE EFFECTS OF PROLONGED RADIATION EXPOSURE, HIGH TEMPERATURES AND PRESSURES FROM A DBA SUCH AS A LOCI. THE VALSPAR 13 F 12 / 89 AND CARBOLINE 2011 5 / 890 COATING SYSTEMS WILL ALSO PROVIDE A HIGHLY DECONTAMINABLE SURFACE.

THE VALSPAR 13 F 12 / 89 AND CARBOLINE 2011 S / 890 COATING SYSTEMS HAVE BEEN TESTED TIN ACCORDANCE WITH ANSI N 101 4 & N 101 2 BY FIRST EXP MING THE COATING TO AN ACCUMULATED RADIATION EXPOSURE OF AT LEAST 300 000 000 RADS (40 YEAR LIFE). NEXT THE COATINGS WERE TESTED TO A DBA ENVIRONMENT OF TEMPERATURE AND PRESSURE THAT ENVELOPED THE LOCI PRESSURE AND H MPERATURE CURVES SHOWN ON FIGURES 14 20 18 & 14 20 20 IN THE CCNPP UFSAR. THE RESULTS OF THIS TESTING SHOWED THAT THE VALSPAR 13 F 12 / 89 AND CARBOLINE 2011 S /

890 COATING STSTEMS DID NOT EXHIBIT ANY FLAKING, DELAMINATION OR PEELING, BLISTER OR CRACKING. FROM THIS IT COULD BE CONCLLDED THAT VALSPAR 13 F 12 /

89 AND CARBOLINE 2011 S / 890 COATING SYSTEMS WILL REMAIN INTACT FOLLOWING A DBA. THE OPERATION OF THE CONTAINMENT SUMP WILL NOT BE AFFECTED BY THE USE OF THESE CDATINGS INSIDE THE CONTAINMENT. THEIR USE WILL CAUSE NO DEGRADATION TO THE SAFETY INJECTION SYSTEM (EMERGENCY CORE COOLING SYSTEM) OR THE CONTAINMENT SPRAY SYSTEM.

VALSPAR VENDOR DATA (REFERENCE LABORATORY REPORT DATAED 04 29 70 FOR val ***e

NMR5018 NUCLEIS 10/15/1995 Search Proces2 A & oc Report 112 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

COATING SYSTEMS) INDICATES THAT THE OVERALL THERMAL BTU / HR FT DEGREE F.

[

THIS IS HIGHER THAN THE OVERALL THERMAL CONDUCTIVITY (0 67 BTU / HR FT

+

DEGREE F) WHICH WAS USED IN CCNPP UFSAR CHAPTER 14 ACCIDENT ANALYSIS (THE CONCRETE SURFACES CREDITED IN THE ACCIDENT ANALYSIS AS HEAT SINKS ARE AND HAVE ALWAYS BEEN UNC0ATED). THEREFORE, IT CAN BE CONCLLDED THAT THE LOCI PRESSURE AND TEMPERATURE CURVES AS SHOWN ON FIGURES 14 20 18 & 14 20 20 OF THE UFSAR WILL NOT INCREASE WITH THE USE OF THE VALSPAR AND CAR 90LINE COATING SYSTEMS.

THEREFORE, IT CAN BE CONCLUDED THAN NO IMPACT TO PREVIOUS ANALYSES, IN THE SAR IS CREATED, SINCE THE VALSPAR 13 F 12 / 89 AND CARBOLINE 2011 S / 890 COATING SYSTEMS WILL FUNCTION THE SAME OR BETTER THAN THE ORIGINAL (AMERON)

C0ATING SYSTEM. THIS ACTIVITY DOES NOT CONSTITUTE AN UNREVIEWED SAFETY QUESTION.

(CMH) h 1

i

~--n-------

. ~

NMRB018 NUCLEIS 10/15/1995 Search Procesa Adioc Report 113 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

Doctment ID Revision Status sawwzzzszzzzzzzzzz==zzzmazzzzz mz======

==

94-B-999-103-R00 62

Subject:

THIS ACTIVITY MODIFIES THE EXISTING ELECTRICAL DISTRIBUTION SYSTEM TO CONNECT THE $80 DG OC Alias:

POSRC #:

95-087 Assoc Doc ID: 89-0079 Revision To: 0000 Assoc Stat: 0 Assoc Type:

FCR Eef Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xmtl Date azzwzuzzazzazzzzzzzazz=ammazzzzzzz======mmazzzz=szzzzmazzzz===mazzzzzzzzzzzzzzzzz==zzzz mz=========z mzzsazz=mazz Other rsfs:

Pirs Rafs:

Equipment:

Org/Div:

System Code:

Text:

StBMARY:

THIS ACTIVITY MODIFIES THE EXISTING ELECTRICAL DISTRIBUTION SYSTEM IN ORDER TO CONNECT THE STATION BLACKOUT (SBO) DIESEL GENERATOR, DG OC, TO ThE ENGINEERED SAFETY FEATURES BUSES IN UNIT 1 (EMERGENCY BUS 11 AND 14). THIS ACTIVITY ALSO ADDS THE RACEWAY AND CABLES NECESSARY TO COMPLETE THIS PART OF THE PHASE IN OF DG OC. GENERALLY, THIS ACTIVITY WILL BE PERFORMED DURING A UNIT 1 PLANT CUTAGE.

NEW SSC*S ADDED BY THIS ACTIVITY HAVE BEEN EVALUATED TO ENSURE THE EFFECT OF THEIR INSTALLATION (E.G., SEISMIC ADEQUACY OF EXISTING STRUCTURES, HEAT LOADS, CABLE SEPARATION) DO NOT INCREASE THE PROBABILITY OF PREVIOUSLY EVALUATED MALFUNCTIONS. SSC'S ADDED BY THIS ACTIVITY WILL NOT BECOME OPERATIONAL UNTIL TESTING OF DG OC IS COMPLETE. EQUIPMENT IDENTIFIED AS INITIATORS OF ACCIDENTS ARE NOT AFFECTED BY THIS ACTIVITY. THEREFORE, THE PROBABILITY OF PREVIOUSLY EVALUATED MALFUNCTIONS AND ACCIDENTS HAS NOT BEEN INCREASED.

THE CONSEQUENCES OF PREVIOUSLY EVALUATED MALFUNCTIONS AND ACCIDEKTS HAVE NOT BEEN INCREASED BY THIS ACTIVITY BECAUSE EQUIPMENT REQUIRED TO SERVE MITIGATION FUNCTIONS UNDER THESE CONDITIONS HAVE NOT BEEN AFFECTED, AND CONTROL ROOM AND OFFSITE DOSES PREVIOUSLY CALCULATED REMAIN WITHIN THE PREVIOUSLY STATED LIMITS.

INSTALLATION ACTIVITY IS SEQUENCED SUCH THAT ONE EDG WILL BE AVAILABLE TO SUPPLY EMERGENCY POWER TO AN ENGINEERED SAFETY FEATURES BUS FOR UNIT 1, DURING OUTAGE PORTIONS OF THIS ACTIVITY. ADEQUATE ELECTRICAL ISOLATION FOR DG OC WILL BE PROVIDED. NO NEW TYPES OF SYSTEM INTERACTIONS ARE BEING CREATED BY THIS ACTIVITY. THEREFORE, THE POSSIBILITY OF A NEW MALFUNCTION OR ACCIDENT IS

?

3 NNR8018 NUCLEIS 10/15/1995 Search Process Adioc Report 114 i

i I

STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

NOT CREATED BY THIS ACTIVITY.

l THE MARGIN OF SAFETY EXPRESSED IN THE BASES OF THE TECHNICAL SPECIFICATIONS IS NOT REDUCED, BEC4USE THE ADDITIONAL HEAT LOADS ON THE HCR AND AUXILIARY OUILDING HVAC SYSTEMS HAVE BEEN EVALUATED AND DETERMINED NOT TO EXCEED THE SYSTEMS' DESIGN HEAT REMOVAL CAPACITY. DURING PERIODS WHEN PENETRATIONS IN AREAS PROTECTED BY A HALON SUPPRESSION SYSTEM ARE OPEN, THE HALON SYSTEM WILL BE DECLARED INOPERABLE, AND HOURLY FIRE WATCHES AND BACKUP FIRE SUPPRESSION WILL BE INSTITUTED IN ACCORDANCE WITH THE TECHNICAL SPECIFICATIONS.

THEREFORE, THERE ARE NO UNREVIEWED SAFETY QUESTIONS ASSOCIATED WITH THIS ACTIVITY.

(CMH)

Doctaient ID Revision Status t

wx========================

==

95-0052 0000 62

Subject:

DISA8LE THE RCP CC WATER LOW FLOW ALARM IN THE CONTROL ROOM

[

Alias:

POSRC #:

95-089 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: C Assoc Type: ESP

[

Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xmtl Date

===================================================================================================

s Othsr refs:

Pers Rzfs:

Equipuent:

l I

Org/Div:

System Code: 064 REACTOR COOLANT Text:

SUpmARY:

THIS TEMPORARY ALTERATION DISABLES THE REACTOR COOLANT PLMP (RCP) COMPONENT COOLING WATER FLOW ALARM IN THE CONTROL ROOM. THIS ACTIVITY ALSO BYPASSES THE-i RCP START INTERLOCK BY JtmPERING THE INTERMITTENT FLOW SWITCH. A C(MPLETE FAILURE OF COMPONENT COOLING WATER WOULD BE DETECTED BY THE OTHER RCP LOW FLOW ALARMS. A COMPONENT COOLING WATER FLOW FAILURE TO THE 21A RCP IS UNLIKELY, BUT IF IT DOES OCCUR, IT WILL BE DETECTED BY RCP HIGH BEARING AND SEAL LEAK OFF TEMPERATURE ALARMS.

THE PROBASILITY AND CONSEQUENCES OF A SE! ZED RCP ROTOR EVENT ARE NOT INCREASED, BECAUSE THE BASIC INPUTS AND ASSUMPTIONS OF THIS ANALYSIS IS UNCHANGED BY REMOVING THIS LOW FLOW ALARM.

s i

l

~~ -

NMR8018 NUCLEIS 10/15/1995 Search Proces2 Actioc Report 115 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

THE ACTIVITY HAS SEEN EVALUATED AND DOES NOT INCREASE THE PR08A81LITY OR CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION, NOR DOES IT CREATE THE POSSIBILITY OF A NEW ACCIDENT OR MALFUNCTION, NOR DOES IT REDUCE THE MARGIN OF SAFETY OF ANY TECHNICAL SPECIFICATION BASIS, THEREFORE, THIS TEMPORARY ALTERATION DOES NOT CONSTITUTE AN UNREVIEWED SAFETY QUESTION.

(CMH) t I

r Doctment ID Revision Status i

}

================================

SE00002 0000 62 r

Sthiect:

REVISE SECTION 10A.6 0F THE OFSAR TO STATE CORRECT THE DUTY REQUIREMENTS OF SDC SYS AT "HEL8" CONDITIONS.

Alias:

POSRC #:

95-090 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat:

Assoc Type: ESP l

Fef Doc ID: DE1995-0990 Rev:

0000 Refer Type: DMLS DESIGN MEMO LOGGING SYSTEM DE1995-0996 0000 DMLS DESIGN MEMO LOGGING SYSTEM I

. m m

=

w

-w -

NNRB018 NUCLEIS 10/15/1995 Search Proces2 A& oc Report 116 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

Sender Xmtl #

Xmtt Date u==u=======s===========================================================================

============

Other refs:

Pers R fs:

Equipment:

Org/Div:

System Code: 052 SAFETY INJECTION SYSTEM Text:

SUMMARY

THIS ACTIVITY IS A REVISION TO CHAPTER 10 A.6 OF THE UFSAR. THIS SECTION CURRENTLY CONVEYS THAT SINCE THE SHUTDOWN COOLING SYSTEM IS AT HIGH ENERGY CONDITIONS (I.E., GREATER THAN 275 PSIG AND / OR 200 DEGREES F) FOR ONLY ONE HOUR PER EACH COOLDOWN A BREAK IN THIS SYSTEM AT HIGH ENERGY CONDITIONS IS CONSIDERED NON CREDIBLE. IT HAS BEEN DETERMINED THAT THE SDC SYSTEM CAN BE AT HIGH ENERGY CONDITIONS DURING A HEATUP. THEREFORE, SECTION 10 A.6 IS BEING REVISED TO CONVEY THAT THE SDC SYSTEM IS AT HIGH ENERGY CONDITIONS FOR LESS THAN 2% OF THE TIME THAT IT IS AT NON HIGH ENERGY CONDITIONS, AND THAT THIS MAINTAINS THE POSITION THAT A PIPE BREAK AT HIGH ENERGY CONDITIONS IS WON CREDIBLE. THE BASIS OF THIS COMES FROM NUREG 75 087 (" STANDARD REVIEW PLAN",

SEPT 1975) UNDER BRANCH TECHNICAL POSITION MES 3-1, SECTION B.2.E WHERE IT IS STATED THAT A SHORT OPERATIONAL PERIOD IS DEFINED AS ONE WHERE THE FRACTION OF TIME THAT A SYSTEM IS AT HIGH ENERGY CONDITIONS IS LESS THAN 2% OF THE TIME IT IS AT NON HIGH ENERGY CONDITIONS. THEREFORE, THE ORIGINAL POSITION OF i

NOT ANALYZING THE SDC SYSTEM FOR HEL8 S REMAINS VALID. A REVIEW OF THE AVAILABLE PLANT HEATUP AND CODLDOWN LOGS SHOWS THAT THERE IS APOLE MARGIN TIME BEWTEEN THE ACTUAL TIME THE SDC SYSTEM IS AT HIGH ENERGY CONDITIONS, AND THE ALLOWABLE 2% TIME OPERATING TIME. THIS CHANGE DOES NOT REPRESENT AN UNREVIEWED SAFETY QUESTION (USQ) NOR REDUCE THE MARGIN OF SAFETY AS DEFINED IN THE BASES FOR ANY TECHNICAL SPECIFICATIONS. NO CHANGES TO THE TECHNICAL SPECIFICATIONS ARE REQUIRED.

(CMH) i i

. i

. m m

l l

[

NNRB018 NUCLEIS 10/15/1995 Search Process A&oc Report 117

$1ATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

Document ID Revision Status i

================================

SE00003 0000 62

Subject:

CTMT PURGE LINES MODIFICATION Alias:

94-0204-00 (ESP)

POSRC #:

95-090 l

Assoc Doc ID: 94-0204 Revision To: 0000 Assoc Stat: C Assoc Type:

FCR 94-0205-00 0000 0

ESP Ref Doc ID:

Rev:

Refer Type:

Sender Xatl #

Xatt Date

=========____=======================_=============================================================

Other rsfs:

P;rs Rafs:

Equipment:

Org/Div:

System Code: 060 PRIMARY CONTAINMENT HEAT AND VENT Text:

SUpmARY:

THE PROPOSED ACTIVITY WILL ADD BLIND FLANGES TO THE CONTAINMENT PURGE OUTBOARD ISOLATION VALVES TO PROVIDE THE CONTAINMENT PENETRATION PRESSURE BOUNDARY IN LIEU OF THE VALVES, IN MODES 1 - 4. UTILIZING THE BLIND FLANGES WILL ENHANCE THE CONTAINMENT PURGE PENETRATION BOUNDARY'S SEALING CAPABILITY WITHOUT CHANGING THE DESIGN FUNCTION OF THE SYSTEM. THE PROPOSED BLIND FLANGES AND MODIFICATIONS TO THE EXISTING PURGE VALVES ARE DESIGNED IN ACCORDANCE WITH THE REQUIREMENTS OF THE ORIGINAL DESIGN SPECIFICATION FOR THE PURGE VALVES AND THE APPLICABLE CONSTRUCTION CODE. SUPPORTING BGE CALCULATIONS DEMONSTRATE THE BLIND FLANGE AND CONTAINMENT PURGE VfLVE ASSEMBLY WILL UNDERGO LLRT TESTING. THEREFORE, THIS ACTIVITY WILL NOT REDUCE THE MARGIN OF SAFETY AS DESCRIBED IN ANY TECHNICAL SPECIFICATION BASES AND WILL NOT CREATE AN UNREVIEWED SAFETY QUESTION.

(CMH)

NNR9018 NUCLEIS 10/15/1995 Search Procesa A& oc Report 118 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

Document ID Revision Status

================================

95-0046 0000 62

Subject:

THE MODIFICATION WILL REMOVE THE 1 SECOND TIME DELAY FOR THE SWING SERVICE WATER PUMPS 13 & 23 WHEN THE SWING PUMP IS ALIGNED AS THE DEDICATED COOLING SOURCE FOR THE PLANT Alles:

POSRC #:

95-092 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: C Assoc Type: ESP Ref Doc ID:

Rev:

. Refer Type:

Sender Xmtl #

Xmtl Date

===================================================================================================

Other refs:

Pers Rafs:

Equipment:

Org/Div:

System Code:

Text:

SUMARY:

THE PROPOSED ACTIVITY MODIFIES THE CONTROL CIRCUIT OF THE SERVICE WATER SWING PUMPS TO REMOVE THE TIME DELAY FOR STARTING ON AN ESFAS SIGNAL. THE TIME DELAY WILL ONLY BE REMOVED WHEN THE SWING PLMP IS ALIGNED AS THE DEDICATED COOLING SOURCE FOR THE PLANT (I.E. ONE OF THE NORMALLY ALIGNED PUMPS IS OUT FOR MAINTENANCE AND THE HANDSWITCH IS IN THE " PULL TO LOCK" POSITION). THIS WILL ENSURE THAT THE SWING PUMP WILL REACT LIKE A DEDICATED PUMP TO AN ESFAS SIGNAL.

THIS ACTIVITY DOES NOT CONSTITUTE AN UNREVIEWED SAFETY QUESTION (USQ).

I l

~ - -

NMR8018 NUCLEIS 10/15/1995 119 Search Process A & oc Report STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

(CNN)

Document ID Revision Status

_____=============================

94-8-999-082-R01 62

Subject:

THIS ACTIVITY MODIFIES THE EXISTING ELECTRICAL DISTRIBUTION SYSTEM IN ORDER TO CONNECT A NEW SR DIESEL GENERATOR, DG 1A TO AN ENGINEERED SAFETY FEATURES BUS IN UNIT 1. (EMERG. BUS 11)

Atlas:

POSRC #:

95-095 Assoc Doc ID: 89-0079 Revision To: 0000 Assoc Stat: 0 Assoc Type:

FCR Kef Doc ID:

Rev:

Refer Type:

Sender Xmtt #

Xmtt Date

===============================================================================================================.

Oth r refs:

PIrs RIfs:

Equipment:

Org/Div:

System Code: 024 EMERGENCY DIESEL GENERATOR Text:

NRC SLBetARY:

l IN ORDER TO CONNECT DG 1A TO EMERGENCY BUS 11, THIS ACTIVITY DISCONNECTS EMERGENCY DIESEL GENERATOR 11 (DG 11) FROM EMERGENCY BUS 11 FROM ITS 4

NORMALLY CLOSED DISCONNECT SWITCH (DISCONNECT SWITCH 1103) AND FROM THE l

I m.

NWRB018 NUCLEIS 10/15/1995 Search Procesa A& oc Report 120 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

CIRCUIT BREAKER CUBICLE AT THE BUS. POWER CABLING FOR DG 1A WILL THEN BE CONNECTED TO THE CIRCUlf BREAKER AT EMERGENCY BUS 11. DG 1A WILL NOT BECOME OPERATIONAL UNTIL TESTING IS COMPLETED. IN SUPPORT OF THE DG 1A TIE IN, THIS ACTIVITY ALSO ADDS RACEWAYS IN THE UNIT 1 ELECTRICAL SWITCHGEAR ROOM TO COMECT DG 1A TO EKERGENCY BUS 11 AND INSTALLS / CONNECTS WIRING BETWEEN THE ELECTRICAL AUXILIARY CONTROL PANEL (EACP) AND THE DIESEL GENERATUR CONTROL CONSOLE (DGCC) FOR DG 1A INSTRUMENTATION, ANNUNCIATION AND CONTROLS. IN ADDITION, THIS ACTIVITY RE DESIGNATES EMERGENCY DIESEL GENERATOR 11 AS EMERGENCY DIESEL GENERATOR 2A (HEREAFTER REFERRED TO AS DG 2A) AND MODIFIES THE EXISTING ELECTRICAL DISTRIBUTION SYSTEM IN ORDER TO COMPLETE THE DEDICATION OF DG 2A To AN ENGINEERED SAFETY FEATURES BUS IN UNIT 2 AND TRANSFER THE INDICATIONS, ANNUNCIATION AND CONTROLS FOR DG 2A AND BREAf'ER CONTROLS FOR EMERGEhCY BUS 21 FROM THE ELECTRICAL AUXILIARY CONTROL PANEL (EACP) TO THE DIESEL GENERATOR CONTROL CONSOLE (DGCC). THE INTERNAL WIRING IN THE EACP FOR DG 2A WILL BE DISCONNECTED.

14 ORDER TO CONNECT DG 1A S AUTOMATIC START AND LOADING CIRCUITS TO THE PLANT THIS ACTIVITY WILL REMOVE THE UNIT 1 AUTOMATIC START SIGNALS (SIAS AND BUS UNDERVOLTAGE) FROM DG 2A. THESE SIGNALS WILL BE CONNECTED TO DG 1A TO AUTOMATICALLY START DG 1A UPON RECEIPT OF A SIAS OR, START AND LOAD DG 1A 04 RECEIPT OF A BUS UNDERVOLTAGE ESFAS SIGNAL.

IN ORDER TO DEDICATE DG 2A To UNIT 2, THIS ACTIVITY DEDICATES SERVICE WATER COOLING FOR DG 2A TO A UNIT 2 SERVICE WATER SUBSYSTEM.

UNIT 1 WILL BE IN MODE 5, 6 OR DEFUELED THROUGHOUT IMPLEMENTATION OF THE PORTIONS OF THIS ACTIVITY THAT ARE ASSOCIAliD WITH CHANGES IN SYSTEM OPERATION. UNIT 2 IS CONSIDERED TO BE IN MODE 1, 2, 3, 4, 5 OR 6. AT LEAST 23 FEET OF WATER WILL BE MAINTAINED OVER THE IRRADIATED FUEL ASSEMBLIES SEATED WITHIN THE REACTOR PRESSURE VESSEL WHILE UNIT 1 IS IN MODE 6 AND A UNIT 1 ENGINEERED SAFETY FEATURES BUS IS OUT OF SERVICE. PORTIONS OF THIS ACTIVITY, WHICH MAY BE PERFORMED DURING NON OUTAGE CONDITIONS, ARE LIMITED TO WORK ACTIVITIES THAT CAN BE PERFORMED WITHOUT CREATING CHANGES TO THE FUNCTION OR OPERATION OF EXISTING PLANT SYSTEMS. CHANGES TO BE OUT OF SERVICE PLANT COMPONENTS WILL NOT CHANGE THE FUNCTION OR OPERATION OF THESE PLANT l

COMPONENTS ONCE THEY HAVE BEEN RESTORED TO SERVICE. NO TERMINATION /

DETERMINAtl0NS TO FUNCTIONING ELECTRICAL CIRCUITS WILL BE COMPLETED DURING THE NON OUTAGE PERIOD. WORK WITHIN THE EACP WILL NOT BE PERFORMED WHEN THE PLANT IS IN A TECHNICAL SPECIFICATION LCO ACTION STATEMENT FOR ANY OF THE EDG S, THEIR ASSOCIATED EMERGENCY BUSES OR THE OFFSITE POWER SOURCES (I.E, TECHNICAL SPECIFICATION 3 8 1 1 AND 3 8 1 2).

NEW SSC 5 ADDED BT THIS ACTIVITY HAVE BEEN EVALUATED TO ENSURE THE EFFECT OF THEIR INSTALLATION (E.G., SEISMIC ADEQUACY OF EXISTING STRUCTURES, HEAT LOADS, CABLE SEPARATION ) DO NO INCREASE THE PROBABILITY OF PREVIOUSLY EVALUATED MALFUNCTIONS. SSC S ADDED BY THIS ACTIVITY WILL NOT BECOME OPERATIONAL UNTIL TESTING OF DG 1A IS COMPLETE. EQUIPMENT IDENTIFIED AS INITIATORS OF ACCIDENTS ARE NOT AFFECTED BY THIS ACTIVITY. THEREFORE, THE

NMR9018 NUCLEIS 10/15/1995 Search Proceso A& oc Ceport 121 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

PR08A81LITY OF PREVIOUSLY EVALUATED MALFUNCTIONS AND ACCIDENTS HAS NOT BEEN INCREASED.

THE CONSEQUENCES OF PREVIOUSLY EVALUATED MALFUNCTIONS AND ACCIDENTS HAVE NOT BEEN INCREASED BY THIS ACTIVITY BECAUSE EQUIPIENT REQUIRED TO SERVE MITIGATION FUNCTIONS UNDER THESE CONDITIONS HAVE NOT BEEN AFFECTED AND CONTROL ROOM AND OFFSITE DOSES PREVIOUSLY CALCULATED REMAIN WITHIN THE PREVIOUSLY STATED LIMITS.

AN EVALUATION WAS PERFORMED TO ASSESS THE POSSIBILITY OF AN INSTALLATION ERROR IN THE EACP WHICH COULD RESULT IN A LOSS OF AN EDG OR ENGINEERED t

SAFETY FEATURES BUS OR THAT COULD CAUSE A PLANT TRIP IN THE OPEATING UNIT.

NO NEW SYSTEMS INTERACTIONS ARE BEING CREATED BY THIS ACTIVITY. THEREFORE, THE POSSIBILITY OF A NEW MALFUNCTION OR ACCIDENT IS NOT CREATED BY THIS ACTIVITY.

THE MARGIN OF SAFETY EXPRESSED IN THE BASES OF THE TECHNICAL SPECIFICATIONS

[

IS NOT REDUCED BECAUSE COMPLETION OF THIS ACTIVITY WILL RESULT IN TWO t-OPERATIONAL EDG S FOR EACH UNIT. PRIOR TO IMPLEMENTING THIS ACTIVITY, AN NRC APPROVED EXTENSION OF THE SEVEN DAY LIMITATION OF ACTION STATEMENTS A AND 8 OF TECHNICAL SPECIFICATION 3 7 6 1 WILL BE REQUIRED.

THEREFORE, THERE ARE NO UNREVIEWED SAFETY QUESTIONS ASSOCIATED WITH THIS ACTIVITY.

(CMH)

Doctment ID Revision Status

================================

t SE00004 0000 62 Stbject:

SAFETY EVALUATION FOR MCR 94 041 012 f

Atlas:

POSRC #:

95-095 6

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NNRs018 NUCLEIS 10/15/1995-122 Search ProcesJ A e oc Report STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

Assoc Doc ID: 94-041-012-00 Revision To: 0000 Assoc Stat: 0 Assoc Type: MCR Ref Doc ID:

Rev:

Refer Type:

'Xett 8 Xmtl Date Sender

============

===============================__________=========================-....__-...-......

Other refs:

Pers RIfs:

Equipment:

1 HIC 110 11 RC PRZR HIC 2 HIC 110 LETDOWN THROTTLE VLV CNTRL Org/Div:

System Code: 041 CHEMICAL & VOLUME CONTROL SYSTEM (CVCS)

Text:

SU MARY:

THIS SAFETY EVALUATION ADDRESSES THE ACTIVITY OF MCR 94 041 012 00. THE SCOPE OF THIS ACTIVITY IS TO REPLACE THE CHEMICAL AND VOLUME CONTROL SYSTEM (CVCS)

LETDOWN VALVE AUTO / MANUAL CONTROLLER, 1 (2) HIC 110, WITH FISCHER AND PORTER MICR0 DCI MODULAR CONTROLLERS. THE NEW CONTROLLER IS A VENDOR RECOM ENDED DIGITAL REPLACEMENT SERIES FOR THE EXISTING ANALOG MtBEL. THE NEW CONTROLLERS ARE ALSO DESIGNED TO PERFORM THE FUNCTION OF THE EXISTING CURRENT LIMITERS SO THAT THE CURRENT LIMITERS, 1 (2) LY 110 ARE BEING REMOVED. THESE CONTROLLERS AND CURRENT LIMITERS ARE CLASSIFIED NSR AND ARE INSTALLED SEISMIC II / I.

THIS ACTIVITY DOES NOT INCREASE THE PROBABILITY OF A MALFUNCTION, ACCIDENT AND DOES NOT CREATE A NEW MALFUNCTION, OR A NEW ACCIDENT NOT PREVIOUSLY ANALYZED IN THE SAR. FURTHERMORE, THE CONSEQUENCES OF THE PREVIOUSLY DISCUSSED MALFUNCTIONS AND ACCIDENTS ARE NOT INCREASED. THIS ACTIVITY DOES NOT CONSTITUTE AN UNREVIEWED SAFETY QUESTION AND DOES NOT RESULT IN A CHANGE TO TECHNICAL SPECIFICATIONS.

(CMH)

Doctment ID Revision Status

======_____============= ______==

95-0001 0000 62 Sthject:

UPGRADE SITE'S VEHICLE BARRIER SYSTEMS TO PREVENT ACCESS BY A MALEVOLENT VEHICLE WITHIN THE SAFE STANDOFF DISTANCE FROM SELECTED CCNPP SSC'S Allas:

POSRC 8:

95-097 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: 0 Assoc Type: ESP Ref Doc ID:

2-96-0092 Rev:

0000 Refer Type:

TMOD TEMPORARY MODIFICATIONS

NNRB018 NUCLEIS 10/15/1995 Search Proces3 Adioc Report 123 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

Sender Xmtl #

Xmtl Date

===========================================================-

______w======

============

Other rsfs:

Pers Rifs:

Equipment:

Org/Div:

System Code: 064 REACTOR COOLANT Text:

SUMARY:

THE INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) HAUL ROAD PROVIDES A HARD PAVED SURFACE FOR THE TRACTOR TO TRANSPORT SPENT FUEL IN A NUNOMS 24 P DSC / TC FROM THE CCNPP AUXILIARY BUILDING TO THE ISFSI. THE ISFSI USAR DESCRIPTION OF THE TRANSFER ROUTE WAS CHANGED TO ALLOW THE PRESENCE OF A VEHICLE BARRIER TO BE INSTALLED TO COMPLY WITH 10 CFR 73 55, AS AMENDED IN AUGUST 1994. THE CHANGE ALLOWS THE VEMICLE BARRIER'S SUPPORTING BUTTRESSES TO BE INSTALLED WITHIN THE 28 FOOT WIDE TRANSFER ROUTE. IT HAS BEEN CONFIRM-ED BY CALCULATION THAT A CASK DROP DNTO THE VEHICLE BARRIER BUTTRESSES AND A CRASH BEAM DROP ONTO THE TC ARE ENVELOPED BY THE EXISTING CASK DROP ANALYSIS. THIS CHANGE DOES NOT CONSTITUTE AN UNREVIEWED SAFETY QUESTION, A CHANGE TO THE TECHNICAL SPECIFICATIONS OR BASES, A SIGNIFICANT INCREASE IN OCCUPATIONAL EXPOSURE NOR AN UNREVIEWED ENVIRONMENTAL IMPACT FOR THE ISFSI.

(CMH) l

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10/15/1995 NuRs018 NUCLEls

' Search Process Actioc Report 124 STATUS 62 OR 64 50.59S (10/01/1996 THRU 09/30/1995) i i

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SE00011 0000 62 j

S4)ect:

22 AUX FD PNP FLUSN l

Alles:

TMOD 2-95-0126 t

,t POSRC #:

95-100 r

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NNRB018 NUCLEIS 10/15/1995 Search Proces3 A & oc Ceport 125 STATUS 62 OR 64 50.595 (10/01/1996 THRU 09/30/1995)

Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: 0 Assoc Type: ESP Ref Doc ID:

Rev:

Refer Type:

Sender Matl #

Xatt Date

==============================================================.______==============================

Other refs:

P as R2fs:

Equipment:

Org/Div:

System Code: 036 AUXILIARY FEEDWATER Text:

NRC

SUMMARY

THIS ACTIVITY WILL ALLOW THE REMOVAL OF A SPOOL IN THE 22 AFW PUMP SUCTION PIPING, AND INSTALLATION OF A TEMPORARY FLANGE AND REQUIRED HARDWARE TO ALLOW FLUSHING OF THE SYSTEM. THIS IS BEING DONE AS PART OF A TROUBLE SHOOTING EFFORT TO RESOLVE PROBLEMS WITH THE 22 AFW PUMP. THIS FLUSH WILL OCCUR DURING MODE 1 POWER OPERATION.

THERE IS CURRENTLY AN OPERATIONAL PROBLEM WITH THE 22 AFW PLMP. ONE OF THE POTENTIAL CAUSES IS THOUGHT TO BE FOREIGN MATERIAL INTRUSION INTO THE PUMP VIA THE SUCTION PIPING. THIS ACTIVITY WILL SUPPORT THE TROUBLE SHOOTING EFFORT. THIS SYSTEM IS SHOWN AND DESCRIBED IN THE UFSAR CHAPTER 10. SINCE OPENING THE SUCTION ISOLATION VALVE TO ALLOW THE FLUSH TO OCCUR VIOLATES THE TAGGING BOUNDARY, THIS ACTIVITY IS A TEMPORARY MODIFICATION, AND REQUIRES A SAFETY EVALUATION.

DURING MODE 1 OPERATION, THE AFW SYSTEM IS A STAND 8Y EMERGENCY SYSTEM WHICH IS DESIGNED TO PROVIDE FEEDWATER TO THE STEAM GENERATORS FOR THE REMOVAL OF SENSIBLE AND DECAY HEAT, AND COOL THE RCS To 300 F IN THE EVENT THAT MAIN FEEDWATER IS NO LONGER AVAILABLE. DURING MODE 1 OPERATION THIS SYSTEM IS SOLELY A STAND 8Y SYSTEM AND USED FOR EVEN MITIGATION.

SUFFICIENT REDUNDANCY AND DIVERSITY HAS BEEN BUILT INTO THE SYSTEM SUCH THAT ONLY ONE AFW PUMP IS REQUIRED TO PROVIDE FLOW TO THE STEAM GENERATORS TO ENSURE THE SAFETY FUNCTION OF THE SYSTEM CAN BE ACCOMPLISHED. TO ENSURE THAT THE SYSTEM IS NOT SUSCEPTIBLE TO SINGLE FAILURE, THERE ARE TWO STEAM TURBINE DRIVEN PUMPS AND ONE MOTOR DRIVEN PUMP EACH WITH 1001 CAPACITY.

THIS ACTIVITY IS CONSISTENT WITH THE TECHNICAL SPECIFICATIONS WILL NOT AFFECT THE RESPONSE OF THE AFW SYSTEM OR CAPABILITIES DURING ANY EVENT. AS SUCH THIS ACTIVITY DOES NOT CONSTITUTE A USQ.

(CNN)

P l

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NNR8018 NUCLEIS 10/15/1995 Search Procesa Adioc Report 126 STATUS 62 OR 66 50.59S (10/01/1996 THRU 09/30/1995)

F Document ID Revision Status

s=========================

==

SE00001 0000 62

Subject:

TEMP ALT 1 95 078 Alias:

POSRC #:

95-101 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: O Assoc Type: ESP I

Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xmtl Data

===========================================_____==================================================

Other refs:

PIra R;fs:

Equipment:

Org/Div:

System Code: 015 COMPONENT COOLING Text:

NRC SUMARY:

THIS SAFETY EVALUATION ADDRESSES THE ACTIVITY OF TEMPORARILY PLACING CONPONENT COOLING (CC) CONTROL VALVE 1 CV 3828 IN THE FAILED OPEN POSITION DURING ALL MODES OF OPERATION. THE VALVE IS THE CC OUTLET VALVE FROM 11 SHUT DOWN COOLING (SDC) HEAT EXCHANGER (11 SDCMX). INSTRUMENT AIR (IA) MANUAL i

ISOLATION VALVE 1 IA 663 WILL BE TAGGED AND CLOSED TO ISOLATE THE AIR SUPPLY r

TO THE AIR TO CLOSE / FAIL OPEN ACTUATOR OF 1 CV 3828. THIS WILL FAIL CC i

VALVE 1 CV 3828 IN ITS SAFETY RELATED OPEN POSITION. No PHYSICAL CHANGES ARE l

BEING APPROVED AS A RESULT OF THIS TA. THIS ACTIVITY ESSENTIALLY REPRESENTS AN OPERATIONAL TAGOUT ACTIVITY (ISOLATING 1 IA 663); HOWEVER, IT WILL RENDER THE ACTUATOR FOR 1 CV 3828 NON FUNCTIONAL. THE SAFETY INJECTION ACTUATION SIGNAL (SIAS) WILL CONTINUE TO OPERATE THE ASSOCIATED SOLENOID VALVE 1 SV 3828; HOWEVER, WITH THE AIR SUPPLY ISOLATED, 1 CV 3828 WILL NOT CLOSE BUT REMAIN IN ITS SAFETY RELATED (OPEN) POSITION. CONTROL ROOM HANDSWITCH 1 MS i

i i

,r______-.

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l NNR8018 NUCLEIS 10/15/1995 Search Proces3 A & oc Report 127 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995) 3828 WILL BE TAGGED TO THE OPEN POSITION.

NO PHYSICAL CHANGES ARE BEING MADE TO EOJIPMENT AS A RESULT OF THIS ACTIVITY.

VALVE 1 CV 3828 WILL CONTINUE TO PROVIDE A SAFETY RELATED PRESSURE SOUNDARY FOR THE CC SYSTEM. THE VALVE IS NORMALLY SHUT AND REQUIRED TO OPEN UPON A SIAS. FAILING THE VALVE OPEN WILL ALREADY PLACE THE VALVE IN ITS SAFETY POSITION (OPEN) TO PROVIDE CC WATER TO 11 SDCHX TO MEET POSTULATED SAFE SHUTDOWN AND ACCIDENT CONDITIONS. FAILING 1 CV 3828 IN THE OPEN POSITION REMOVES THE POSSIBILITY OF A SINGLE ACTIVE FAILURE PROMI8ITING THE POSITION-ING OF THIS VALVE TO ITS SAFETY POSITION (OPEN).

EVEN WITH A CC FLOW PATH OPEN TO 11 SDCHX, THE CC SYSTEM WILL CONTINUE TO MEET SYSTEM DEMANDS FOR NORMAL OPERATION. IF NECESSARY, A SECOND CC PUMP &

SECOND CCHX CAN BE PLACED IN SERVICE TO MEET THE DEMANDS. ISOLATING AIR TO THE ACTUATOR TO FAIL 11 SDCHX CC CUTLET VALVE 1 CV 3828 IN THE OPEN POSITION DOES NOT INCREASE THE PROBABILITY OF A MALFUNCTION, ACCIDENT, NEW MALFUNCTION OR NEW ACCIDENT NOT PREVIOUSLY ANALYZED IN THE SAR. FURTHERMORE, THE CONSEQUENCES OF THE PREVIOUSLY DISCUSSED MALFUNCTIONS AND ACCIDENTS ARE NOT INCREASED. THIS ACTIVITY DOES NOT CONSTITUTE AN UNRESOLVED SAFETY QUESTION &

DOES NOT VIOLATE TECHNICAL SPECIFICATIONS.

(CMH)

Doctanent ID Revision Status

- =-===

_m===================s

Subject:

THE EACP WILL BE ALTERED TO ENHANCE THE PRESENTATION OF INFORMATION TO THE OPERATOR IAW 89 0079 Alias:

POSRC #:

95-102 Assoc Doc ID: 89-0079 Revision To: 0000 Assoc Stat: O Assoc Type:

FCR Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xmtl Date

==================================== ============================================ = ==================______

Other rifs:

Pers R:fs:

Equipment:

Org/Div:

System Code: 062 CONTROL BOARDS Text:

NRC SUPMARY:

THIS ACTIVITY MODIFIES EQUIPMENT ASSOCIATED WITH THE EMERGENCY POWER SOURCES AND THE 13 8 KV, 4 16 KV AND 480 V DISTRIBUTION SYSTEMS. THE ELECTRICAL AUXILIARY CONTROL PANEL (EACP) (1C17, 1C18, 1C19 & 2C17) WILL BE ALTERED TO ENHANCE THE PRESENTATION OF INFORMATION TO THE OPERATOR FOR MONITORING THE

I NWR5018 NUCLEIS 10/15/1995 Search ProcesO A & oc Report 128 STATUS 62 OR 64 50.595 (10/01/1094 THRU 09/30/1995)

ELECTRICAL POWER SYSTEMS. INSTRUMENTATION AND CONTROLS ARE REARRANGED TO CORRECT DISCREPANCIES IDENTIFIED BY A DETAILED CONTROL ROOM DESIGN REVIEW '

(DCRDR) IN THE 1980'S. THIS ACTIVITY RELOCATES EXISTING METERS ON THE METER SECTION OF THE PANELS IN ORDER FOR THE METERS TO PROPERLY ALIGN WITH THE ASSOCIATED CONTROLS ON THE BENCH SECTION OF THE PANELS. THE MODIFICATIONS TO THE EACP REMOVE NONFUNCTIONAL CONTROLS, STATUS INDICATION AND METERS ASSOCIATED WITH DG 11, DG 12 AND EMERGENCY BUSES 11, 14 AND 21 AS A PART OF DEDICATING EACH EMERGENCY DIESEL CENERATOR TO A SINGLE ENGINEERED SAFETY FEATURES BUS.

THE STRUCTURAL ADEQUACY AND SEISMIC QUALIFICATION OF NEW AND EXISTING SSC'S, OPERASILITY OF PLANT ELECTRICAL DISTRIBUTION SYSTEMS AND CONTROL PANEL REQUIREMENTS WERE EVALUATED TO ENSURE THE PROBA8ILITY AND CONSEQUENCES OF A PREVIOUSLY EVALUATED ACCIDENTS AND MALFUNCTIONS HAVE NOT BEEN INCREASED BY THIS ACTIVITY. PRECAUTIONS ARE OBSERVED IN ORDER TO PREVENT INSTALLATION ACTIVITIES FROM INTRODUCING A NEW MALFUNCTION OR ACCIDENT DURING MODIFICATION OF THE EACP. THIS ACTIVITY DOES NOT AFFECT THE OPERABILITY OF ELECTRICAL DISTRIBUTION SYSTEMS. THUS, THE MARGIN OF SAFETY AS DEFINED IN THE TECHNICAL SPECIFICATIONS IS NOT REDUCED.

THEREFORE, THERE ARE NO UNREVIEWED SAFETY QUESTIONS ASSOCIATED WITH THIS ACTIVITY.

(CMH) r P

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NMRB018 NUCLEIS 10/15/1995 Search Proces:a A & oc Report 129 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

Document ID Revision Status s=wau=========================

==

SE00009 0000 62 Stbject:

11 INSTRUMENT AIR DRYER REPLACEMENT (FCR 89-173, SUPP. 1)

Atlas:

POSRC #:

95-102 Assoc Doc ID: 89-0173 Revision To: 0000 Assoc Stat: 0 Assoc Type:

FCR 89-0173-01 0000 0

ESP Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xmtl Date

============

Other refs:

P;rs RIfs:

Equipment:

Org/Div:

System Code: 019 COMPRESSED AIR Text:

NRC

SUMMARY

THIS ACTIVITY REPLACES THE EXISTING 11 INSTRUMENT AIR DRYER WITH A NEW DRYER IN ORDER TO IMPROVE THE RELIABILITY OF THE INSTRUMENT AIR SYSTEM. THE REPLACEMENT DRYER IS OF THE SAME TYPE OF THE EXISTING DRYER, BUT WITH A LARGER CAPACITY RATING AND IMPROVED MONITORING CAPABILITY. LOCAL FLOW INDICATOR 1 FI 2081 WILL ALSO BE REMOVED UNDER THIS ACTIVITY. THIS CHANGE I

DOES NOT REPRESENT AN UNREVIEWED SAFETY QUESTION (US0) NOR REDUCE THE MARGIN i

0F SAFETY AS DEFINED IN THE BASES FOR ANY TECHNICAL SPECIFICATIN. NO CHANGES TO THE TECHNICAL SPECIFICATIONS ARE REQUIRED.

(CMH) a

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WNRB018 NUCLEIS 10/15/1995 Search Process A & oc Report 130 STATUS 62 OR 64 50.59s (10/01/1994 THRU 09/30/1995) o k

Document ID Revision Status

================================

94-B-999-102-R01 62 Stbject:

THE EXISTING ELECTRICAL DISTRIBUTION SYSTEM IS TO BE MODIFIED TO DEDICATE EDG 12 To ENGINEERED SAFETY FEATURES BUS 14 IN UNIT 1.

Atlas:

POSRC #:

95-106 Assoc Doc ID: 89-VJ79 Revision To: 0000 Assoc Stat: O Assoc Type:

FCR Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xmtt Date

=================_____============================================================================

Other refs:

Pers Rafs:

Equipment:

Org/Div:

System Code: 024 EMERGENCY DIESEL GENERATOR Text:

SUPMARY:

THE EXISTING ELECTRICAL DISTRIBUTION SYSTEM IS TO BE MODIFIED TO DEDICATE EDG 12 TO ENGINEERED SAFETY FEATURES BUS 14 IN UNIT 1. DG 12 IS TO BE REDESIGNATED AS DG 1B AND ITS ASSOCIATED SUPPORT SYSTEMS WILL BE REDESIGNATED TO REFLECT THE DIESEL'S DEDICATION TO UNIT 1.

DG 12 (DG 18) AUTOMATIC START & LOADING CIRCUITS WILL BE MODIFIED TO DELETE THE BUS UNDERVOLTAGE SIGNAL ASSOCIATED WITH EMERGENCY BUS 21 & UNIT 2 SIAS SIGNALS. THIS MODIFICATION WILL PREVENT AUTOMATIC ALIGNMENT OF DG 12 (DG 18)

TO EMERGENCY BUS 21. HOWEVER, DG 12 (DG 18) WILL BE AVAILABE FOR MANUAL CONNECTION TO EMERGENCY BUS 21 TO FUNCTION AS A POWER SOURCE TO SHUTDOWN UNIT 2 IN THE EVENT OF A FIRE.

THIS ACTIVITY WILL ALSO REVISE THE TECHNICAL SPECIFICATIONS TO REFLECT THE ELIMINATION OF DG 12 S (DG 1B S) SWING CAPABILITY.

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NNRBOIS NUCLEIS 10/15/1995 Search Procesa A& oc Report 131 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

UNIT 1 WILL BE IN MODE 5, 6 OR DEFUELED THROUGHOUT IMPLEMENTATION OF THE PORTIONS OF THIS ACTIVITY THAT ARE ASSOCIATED WITH CHANGES IN SYSTEM OPERATION. UNIT 2 IS CONSIDERED TO BE IN MODE 1, 2, 3, 4, 5 OR 6. AT LEAST 23 FEET OF WATER WILL BE MAINTAINED OVER THE IRRADIATED FUEL ASSEMBLIES SEATED WITHIN THE REACTOR PRESSURE VESSEL WHILE UNIT 1 IS IN MODE 6 AND A UNIT 1 ENGINEERED SAFETY FEATURES BUS IS OUT OF SERVICE. PORTIONS OF THIS ACTIVITY THAT MAY BE PERFORMED DURING NON OUTAGE CONDITIONS ARE LIMITED TO WORK THAT CAN BE PERFORMED WITHOUT CREATING SYSTEM FUNCTIONAL & OPERATIONAL CHANGES.

MODIFICATIONS IMPLEMENTED BY THIS ACTIVITY WERE EVALUATED TO ENSURE THEY DO NOT INCREASE THE PROBABILITY OF A MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY. EQUIPMENT IDENTIFIED AS INITIATORS OF ACCIDENTS ARE NOT AFFECTED BY THIS ACTIVITY. THEREFORE, THE PROBABILITY OF PREVIOUSLY EVALUATED MALFUNCTIONS & ACCIDENTS HAS NOT BEEN INCREASED.

THE CONSEQUENCES OF PREVIOUSLY EVALUATED MALFUNCTIONS & ACCIDENTS HAVE NOT BEEN INCREASED BY THIS ACTIVITY BECAUSE EQUIPMENT REQUIRED TO SERVE MITIGATION FUNCTIONS UNDER THESE CONDITIONS HAVE NOT BEEN ADVERSELY AFFECTED

& CONTROL ROOM & OFFSITE DOSES PREVIOUSLY CALCULATED REMAIN UNCHANGED &

WITHIN THE PREVIOUSLY STATED LIMITS. ONE EDG WILL REMAIN AVAILABLE FOR A SHUT DOWN UNIT & TWO EDG S WILL BE AVAILABLE FOR A UNIT OPERATING IN MODES 1 THROUGH 4. IN ADDITION, WHEN OPERATING TWO UNITS, TWO EDG S WILL BE AVAILABLE FOR EACH UNIT. PROCEDURAL CHANGES TO THE EDG SERVICE WATER SUB SYSTEMS WILL NOT AFFECT THE FLOW 0F SERVICE WATER TO OTHER SSC S WHICH FUNCTION TO MITIGATE THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION.

ADMINISTRATIVE CONTROLS PLACED ON THE SERVICE WATER SUBSYSTEMS ENSURE THAT A FAILURE OF A UNIT 2 SERVICE WATER SUBSYSTEM WILL NOT AFFECT THE OPERABILITY OF DG 12 (DG 18), NOW DEDICATED TO UNIT 1. NO NEW SYSTEMS INTERACTIONS ARE BEING CREATED BY THIS ACTIVITY. THEREFORE, THE POSSIBILITY OF A NEW MALFUNCTION OR ACCIDENT IS NOT CREATED BY THIS ACTIVITY.

THE MARGIN OF SAFETY EXPRESSED IN THE BASES OF THE TECHNICAL SPECIFICATIONS IS NOT REDUCED BECAUSE THE REQUIRED NUMBER OF EDG S WILL BE AVAILABLE TO SUPPLY EMERGENCY POWER TO AN ENGINEERED SAFETY FEATURES BUS IN ACCORDANCE WITH THE TECHNICAL SPECIFICATIONS. UPON COMPLETION OF THIS ACTIVITY, ONE OPERATIONAL EDG WILL BE AVAILABLE TO SUPPLY EMERGENCY POWER TO EACH OF THE TWO ENGINEERED SAFETY FEATURES BUSES AT EACH UNIT. THE REQUIREMENTS OF THE TECHNICAL SPECIFICATIONS WILL BE IMPLEMENTED WHEN FIRE BARRIERS ARE PENETRATED OR SUPPRESSION SYSTEMS ARE DISABLED.

THEREFORE, THERE ARE NO UNREVIEWED SAFETY QUESTIONS ASSOCIATED WITH THIS ACTIVITY.

(CMH)

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NNRB018 NUCLEIS 10/15/1995 Search Procesa A& oc Report 133 STATUS 62 OR 64 50.595 (10/01/1994 THRU 09/30/1995)

Assoc Doc ID: 89-0079 Revision To: 0000 Assoc Stat: 0 Assoc Type:

FCR Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xmtl Date

===================================================================================================

Other rsfs:

Pers Rifs:

Equipment:

Org/Div:

System Code: 024 EMERGENCY DIESEL GENERATOR Text:

NRC

SUMMARY

THIS ACTIVITY RE DESIGNATES EMERGENCY DIESEL GENERATOR 21 AS EMERGENCY DIESEL GENERATOR 28. UFSAR TEXT AND VARIOUS FIGURES WILL BE REVISED TO REFLECT THE NEW NOMENCLATURE, ASSOCIATED SUPPORT SYSTEMS WILL ALSO BE REDESIGNATED TO REFLECT THE DIESEL GENERATOR'S DEDICATION TO UNIT 2.

A CHANGE IN NOMENCLATURE WILL NOT AFFECT THE DIESEL GENERATOR'S DESIGN, FUNCTION OR METHOD OF PERFORMING A FUNCTION. THEREFORE, THE PROBABILITY OF PREVIOUSLY EVALUATED MALFUNCTIONS AND ACCIDENTS HAS NOT BEEN INCREASED.

THE CONSEQUENCES OF PREVIOUSLY EVALUATED MALFUNCTIONS AND ACCIDENTS HAVE NOT BEEN INCREASED BY THIS ACTIVITY BECAUSE EQUIPMENT REQURED TO SERVE MITIGATION FUNCTIONS HAVE NOT BEEN ADVERSELY AFFECTED.

THE ONLY PHYSICAL MODIFICATION REQUIRED BY THIS ACTIVITY IS RETAGGING AFFECTED PLANT SSC'S TO REFLECT THE NEW DESIGNATION OF THE DIESEL GENERATOR.

NO NEW SYSTEMS INTERACTIONS ARE BEING CREATED BY THIS ACTIVITY. THEREFORE, THE POSSIBILITY OF A NEW MALFUNCTION OR ACCIDENT IS NOT CREATED BY THIS ACTIVITY. THEREFORE, THE POSSIBILITY OF A NEW MALFUNCTION OR ACCIDENT IS NOT CREATED BY THIS ACTIVITY.

THEREFORE, THERE ARE NO UNREVIEWED SAFETY QUESTIONS ASSOCIATED WITH THIS ACTIVITY.

(CMH)

6 NMRe018 NUCLEIS 10/15/1995 Search Process A & oc Report 134 STATUS 62 OR 64 50.595 (10/01/1996 THRU 09/30/1995)

I l

E i

i Document ID Revision Status

================================

SE00008 0000 62

Subject:

TA 1 95 058, 059 & 060 l

Alias:

'[

POSRC #:

95-111 Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: 0 Assoc Type: ESP Ref Doc ID:

Rev:

Refer Type:

Sender

)hntt #

Xmtl Date

===================================================================================================

t Other rsfs:

P;rs RIfs:

Equipment:

Org/Div:

System Code: 041 CHEMICAL & VOLUME CONTROL SYSTEM (CVCS)

Text:

NRC SUlstARY:

I THIS SAFETY EVALUATION ADDRESSES THREE TEMPORARY ALTERATION ACTIVITIES TO ALLOW REMOVAL OF EACH UNIT 1 CVCS CHARGING PUMP SUCTION SIDE RELIEF VALVE (1 RV 315, 318, 321) AND ALLOW INSTALLATION OF A BLIND FLANGE AT THE OUTLET PIPE FLANGE FOR THE SUBJECT RV. THE THREE TA*S ARE:

~ ~..

~.... _

. - --~~ -.

NNRB018 NUCLEIS 10/15/1995 Search Proceso Adioc Report 135 STATUS 62 OR 64 50.59S (10/01/1996 THRU 09/30/1995)

TA 1 95 060 1 RV 318

1. 12 CYCS CHARGING PUMP SUCTION RV TA 1 95 058 1 RV 321

- #13 CVCS CHARGING PUMP SUCTION RV TA 1 95 059 1 RV 315

1. 11 CVCS CHARGING PUMP SUCTION RV r

THE SUBJECT RV WILL BE REMOVED FOR MAINTENANCE AW THE BLIND WILL BE I

INSTALLED TO PREVENT THE RELEASE OF WATER (AND RADIO GASES) FR(M THE CCISION RV OUTLET HEADER DOWNSTREAM OF SUBJECT RV. THE ASSOCIATED CHARGING PLMP WILL BE OUT OF SERVICE (SAFETY TAGGED AND ISOLATED) FOR THE DURATION OF THIS TA, WHILE THE OTHER CHARGING PUMPS REMAIN IN SERVICE.

THE CHARGING PUMP SUCTION RV PROVIDES THERMAL OVERPRESSURE PROTECTION FOR THE PIPING AND COMPONENTS AT THE SUCTION SIDE OF THE CHARGING PUNP. THE RV DISCHARGES TO THE WASTE PROCESSING SYSTEM (WPS) VIA A CCBBION HEADER TIED TO THE OUTLET OF THE OTHER UNIT 1 CHARGING PLMPS SUCTION RV'S. THE PIPING AT THE INLET & OUTLET OF THE RV IS CLASS HC 2 AND IS ANSI B31 7 CLASS 3 DESIGN. THE P! PING AT THE INLET IS SR PB PER THE Q LIST Am THE OUTLET PIPING IS AQ WPS.

l THE OUTLET PIPING IS NSR EXCEPT THAT IT IS DESIGNED SEISMIC CLASS 1.

i I

ALL DESIGN REQUIREMENTS OF THE WPS SYSTEM PIPING ARE MET, THE REMAINING CVCS t

AND WPS PIPING IS ADEQUATELY SUPPORTED AND EETS SEISMIC REQUIREENTS AND THERE ARE NO IMPACTS TO OTHER PLANT SYSTEMS. THERE ARE NO AFFECTS ON ANALYZED MALFUNCTIONS OR ACCIDENTS AND NO NEW MALFUNCTIONS OR ACCIDENTS ARE CREATED. THEREFORE, THIS ACTIVITY DOES NOT CONSTITUTE A USQ.

(CMH) f e

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-10/15/1995 Search Procesa A e oc Report 136 STATUS 62 OR 64 50.595 (10/01/1994 inau 09/30/1995)

Document ID Revision Status 1

zzzzzzzzzzz===mz=mazzzzz=smz======zzzz mzzzzz I

SE00016 62 S4]ect:

TEMP ALT 1 95 062 Alles i

POSRC #:

95-113 l

l l

I

93-B-064A-152-ROO STATUS: 64

SUBJECT:

WILL PERMIT THE USE OF NUCLEAR ENGINEERING SERVICES NOZZLE DAMS AS AN ALTERNATIVE TO CE NOZZLE DAMS AUAS:

93-844A-152-R00 POSRC #:

93-134 ASSOC DOC :

FCR 93-0202 OTHER REFS:

EQUIPMENT:

ORG/DIV:

SYSTEM:

064 TEXT: PROPOSED ACTIVITY:

THIS ACTIVITY WILL PERMIT THE USE OF NUCLEAR ENGINEERING SERVICES NOZZLE DAMS AS AN ALTERNATIVE TO THE CE NOZZLE DAMS THE INSTALLATION OF THE NES NOZZLE DAMS REQUIRES NO MODIFICATION TO THE STEAM GENERATOR AS THEY WILL BE FASTENED TO THE EXISTING S I G CLAMP RING LOCATED ON THE INSIDE OF THE SINGLE HOT LEG (42") AND TWO COLD LEG (30") NOZZLES. THE DAMS WILL ONLY BE INSTALLED WHEN STEAM GENERATOR MAINTENANCE INSPECTION ACTIVITIES ARE BEING PERFORMED SIMULTANEOUSLY DUR;NG REFUELING. THIS ACTIVITY DOES NOT CONSTITUTE AN UNREVIEWED SAFETY QUESTION.

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93-B-001 129-R00 STATUS: 62

SUBJECT:

MODIFY EXISTING 500 KV TRANSMISSION UNES TO NEW POLE & TOWER STRUCTURES AUAS:

POSRC#:

93-111 ASSOC DOC :

OTHER REFS:

EQUIPMENT:

ORG/DIV:

SYSTEM:

001 TEXT: NRC

SUMMARY

INSTALLATION OF THE NEW 500KV CHALK POINT UNE REQUIRES THE EXISTING 500KV LINES TO BE MOVED TO NEW PLES AND TOWERS EAST OF THE CURRENT LOCATION FOR ABOUT A THREE MILE SECTION OF THE RUN BETWEEN PRINCE FREDERICK & CALVERT CUFFS. THE NEW STRUCTURES ARE SIMILAR IN DESIGN TO THE EXISTING STRUCTURES AND MEET THE BALTIMORE GAS & ELECTRIC, DESIGN REQUIREMENTS FOR A 500KV TRANSMISSION LINE. THE NEW CONDUCTOR RUNS INCREASE FROM TWO TO THREE CONDUCTORS PER PHASE. THE RIGHT OF WAY WILL BE WIDENED BY 200 FEET TO ACCOMMODATE THE NEW UNE FROM CHALK POINT. THE NEW SUPPORT STRUCTURES ARE SPACED EITHER 150 OR 200 FEET CENTERLINE TO THE EXISTING STRUCTURES. THE FAILURE OF A SINGLE STRUCTURE SUPPORT WILL NOT RESULT IN THE LOSS OF THE REMAINING 500 KV UNE. THE EXISTING UNES WILL BE INDIVIDUALLY ENERGlZED, MOVED AND REENERGlZED.

THE TRANSFER OF THE EXISTING 500KV UNES TO NEW POLES AND TOWERS DOES NOT CHANGE THE METHOD OF INTERCONNECTING THE CALVERT CUFFS SITE WITH THE BALTIMORE GAS AND ELECTRIC GRID.

THEREFORE, THIS MODIFICATION TO THE 500KV TRANSMISSION SYSTEM DOES NOT RESULT IN AN UNREVIEWED SAFETY QUESTION. (CMH) 93-B-999-132-R00 STATUS: 62

SUBJECT:

FCR 831085 ALIAS:

POSRC #:

84-058 ASSOC DOC :

FCR 831085 OTHER REFS:

EQUIPMENT:

ORG/DIV:

SYSTEM:

102 TEXT: PROPOSED CHANGE:

l PROVIDE PRE-APPROVAL TO REPLACE EXISTING MAIN FEEDWATER AND MAIN STEAM PIPING (INCLUDING ALSO l

ALL MISC PIPING SUB SYSTEMS IN PIPE CLASSES DB, EB AND GB) WITH CHROME MOLY PIPE ON A ONE TO

(

ONE BASIS. EXISTING PIPE IS CARBON STEEL THIS ACTIVITY DOES NOT CONSTITUTE AN UNREVIEWED SAFETY QUESTION. (CMH)

93-9-036-083-ROO STATUS: 64

SUBJECT:

FCR 90106 SUPPLEMENT 1 AUAS:

POSRC #:

93-086 ASSOC DOC :

OTHER REFS:

90-0106-01 i

EQUIPMENT:

ORG/DIV:

SYSTEM:

036 TEXT: NRC

SUMMARY

THIS ACTIVITY MODIFIED THE AFW SYSTEM, SUCH THAT NEW LEVEL INDICATING PRESSURE GAGES WILL BE ADDED TO THE PIPING ON THE 27' LEVEL OF THE AUXIUARY BUILDING. THIS MODIFICATION DOES NOT REDUCE THE AFW SYSTEM REUABluTY BECAUSE THE MOOlFICATION WILL BE SEISMICALLY OUAUFIED AND s

BE INSTALLED PER ALL APPUCABLE CODES AND STANDARDS TO MAINTAIN THE FULL QUAUFICATION OF THE SYSTEM. THE ADDITIONAL ACTIVITY OF ADOING AN NSR LEVEL INDICATOR TO THE 1C89 FIRE PUMP HOUSE PANEL IS MA!NLY FOR IMPROVEMENT TO TANK FILUNG PROCEDURESiMONITORING ON CST 12. THIS ACTIVITY DOES NOT CONSTITUTE AN UNREVIEWED SAFETY OUESTION. (CMH) i 93-B-0648-097-ROO STATUS: 64 f

SUBJECT:

REVISE UFSAR ALIAS:

POSRC #:

93-102 ASSOC DOC :

FCR 90-0151 OTHER REFS:

EQUIPMENT:

ORG/DIV:

SYSTEM:

064 f

TEXT: NRC

SUMMARY

THIS 50.59 EVALUATION WAS PERFORMED TO EVALUATE A REVISION TO TWO UFSAR FIGURES. THE FIGURES REQUtRE REVISION DUE TO INSTALLATION OF FCR 90151 REPLACES EACH REACTOR COOLANT PUMP (RCP)

_[

MOTOR BEARING OIL RESERVOIR AP LEVEL INSTRUMENTATION LOOP WITH A CAPACITANCE TYPE LEVEL LOOP EXTENDS THE RCP MOTOR BEARING OIL RESERVOIR FILL UNES TO A LOCATION WHERE PERSONNEL WILL RECEIVE LESS RADIATION EXPOSURE DURING RESERVOIR FILUNG. THIS ACTIVITY DOES NOT CONSTITUTE AN UNREVIEWED SAFETY QUESTION.

...................................SEE ORIGIN AL FOR MORE INFORMATION " " " " " " " " " * * " " " " * ' * "

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93-B-012-074-ROO STATUS: 62

SUBJECT:

REVISE UFSAR AWAS:

POSRC #:

93-079 ASSOC DOC :

OTHER REFS:

EQUIPMENT:

ORG/DIV:

T SYSTEM:

012 1

TEXT: NRC

SUMMARY

THE SALTWATER PUMP UPPER FLOW LIMIT AS SPECIFIED IN FSAR SECTION 9.5.2.3 HAS BEEN INCREASED FROM 22.400 GPM TO 25,000 GPM. THE UPPER FLOW UMIT IS DESIGNED TO ENSURE THAT ADEQUATE NPSH IS AVAILABLE FOR THE SALTWATER PUMP, BGE CALCULATION NO. M-93-074 REV O HAS DEMONSTRATED THAT ADEQUATE NPSH IS AVAILABLE UNDER WORST CASE DESIGN CONDITIONS AND THAT THE SALTWATER PUMPS WILL CONTINUE TO OPERATE WITH BAY WATER LEVELS DOWN TO (-) 6' C*. SINCE THE SALTWATER PUMPS CONTINUE TO OPERATE IN ACCORDANCE WITH THE ORIGINAL DESIGN THE PROPOSED ACTIVITY DOES NOT CONSTITUTE A USO. (CMH) 93-B-015-081-ROO STATUS: 64

SUBJECT:

MCR 93-015-006-00 AUAS:

POSRC #:

93-079 ASSOC DOC :

OTHER REFS:

93-015-006-00 EQUIPMENT:

+

ORG/DIV:

SYSTEM:

015 TEXT: NRC

SUMMARY

THis MCR ALLOWS THE USE OF AN ALTERNATIVE MATERIAL DESIGNATION FOR THE IMPELLER $ IN THE COMPONENT COOLING WATER PUMPS.THE CCW PUMPS ARE SAFETY RELATED. THE IMPELLERS AND WEAR RINGS WERE ORIGINALLY SUPPUED TO THE ASTM SPECIFICATION B.145.52.GR.4A. THE ASTM COMMITTEE REPLACED ASTM DESIGNATION B.145.52 BY ASTM B.584.73, COVERING THE REQUIREMENTS FOR THE COPPER ALLOY SAND CASTINGS FOR GENERAL APPLICATIONS. THE CCW PUMP IMPELLER MATERIAL IS SAND CASTINGS FOR GENERAL APPUCATIONS. THE CCW PUMP IMPELLER MATERIAL IS LISTED IN THE UFSAR, THEREFORE A 50.59 EVALUATION IS REQUIRED TO CHANGE THE UFSAR.

THIS ACTIVITY DOES NOT INCREASE THE PROBABILITY OF AN ACCIDENT OR MALFUNCTION: DOES NOT INCREASE THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OR REDUCE THE MARGIN OF SAFETY AS DESCRIBED IN THE TECHNICAL SPECIFICATION. THEREFORE, ACTIVITY DOES NOT CREATE AN UNREVIEWED SAFETY QUESTION NOR DOES IT REDUCE THE MARGIN OF SAFETY AS DESCRIBED IN THE TECHNICAL SPECIFICATION BASES. (CMHI r

i

92-B-024-087-RO1 STATUS: 62

SUBJECT:

FCR B9@79 ALIAS:

POSRC#:

93-119 ASSOC DOC :

FCR 89-0079 OTHER REFS:

EQUIPMENT-ORG/DIV:

SYSTEM:

024 TEXT: NRC

SUMMARY

THE ACTIVITY COVERED BY THIS SAFETY EVALUATION IS THE CONSTRUCTION OF THE DIESEL GENERATOR (DG) BUILDINGS. ONE THREE STORY DG BUILDING (SAFETY RELATED) WITH A PARTIAL BASEMENT, AND A ONE TWO STORY SBO DG BUILDING WITH A FULL BASEMENT WILL BE LOCATED IN THE CURRENT NORTH PARKING LOT AREA. THE DG BUILDINGS WILL SUPPORT THE INSTALLATION AND OPERATION OF TWO NEW DG'S TO MEET CCNPP STATION BLACKOUT REQUIREMENTS IN COMPLIANCE WITH 10CFR50.63 AND REGULATORY GUIDE 1-155, STATION BLACKOUT, AND TO PROVIDE SPARE EMERGENCY ELECTRICAL CAPACITY FOR FUTURE ADDITIONS AND MODIFICATIONS THIS SAFETY EVALUATION ONLY ADDRESSES THE CONSTRUCTION OF THE TWO DG BUILDINGS AND THE ADJACENT DUCT BANK SEGMENTS, THE IMPACT OF THAT CONSTRUCTION ON THE PLANT, AND THE IMPACT OF THE COMPLETED STRUCTURES ON THE PLANT. THE DESIGN, INSTALLATION, STARTUP AND OPERATION OF THE NEW DG'S AND SUPPORTING SYSTEMS IN THE DG BUILDINGS WILL BE ADDRESSED IN A DESIGN REPORT THAT WILL SUPPORT A TECHNICAL SPECIFICATION CHANGE FOR THE NEW DG'S.

POWER BLOCK MODIFICATIONS INCLUDING, BUT NOT LIMITED TO,1) PANEL MODIFICATIONS IN THE CONTROL ROOM,2) SERVICE WATER SYSTEM MODIFICATIONS ASSOCIATED WITH EDG'S 11,12 AND 21,3) ELECTRICAL DtSTRIBUTION SYSTEM MODIFICATIONS,4) THE RELOCATION OF EXISTING YARD UTILITIES AND 5) THE I

t ADDITION OF NEW UTILITIES AND REMAINING ELECTRICAL DUCT BANKS WILL BE ADDRESSED IN OTHER FACILITY CHANGE REQUEST PACKAGES.

THE DESIGN OF THE BUILDINGS PREVENTS DAMAGE TO OTHER SAFETY RELATED SSC*S FROM THE FAILURE OF THESE BUILDINGS. INSTALLATION OF THE SOLDIER PILES OF THE TIED BACK WALLS INTO HOLES PRECLUDES UNNECESSARY VIBRATIONS IN THE VICINITY OF THE POWER BLOCK. SINCE THE CONSTRUCTION ACTIVITIES WILL NOT ADVERSELY IMPACT ANY SSC REQUIRED TO MITIGATE EVALUATED MALFUNCTIONS, ALL CALCULATED OFFSITE AND CONTROL ROOM DOSES WILL REMAIN WITHIN THE UMITS STATED IN THE UFSAR. THEREFORE THIS ACTIVITY DOES NOT INCREASE THE PROBABILITY OR CONSEQUENCE OF ANY ACCIDENT PREVIOUSLY EVALUATED IN THE SAR.

CRANES USED IN CONSTRUCTION WILL HAVE LIMITED MOVEMENT IN AND AROUND THE CONSTRUCTION AREA AND WILL BE LOCATED AND CONTROLLED SO THAT THEY WILL NOT AFFECT SAFETY RELATED EQUtPMENT OR OFFSITE POWER SUPPLIES. DURING TORNADO AND HURRICANE WARNINGS AND HIGH WINDS, CONSTRUCTION CRANES WILL BE SECURED. NO NEW FLOOD PATHS INTO SAFETY RELATED AREAS HAVE BEEN CREATED.THEREFORE, THE POSSIBILITY OF A NEW MALFUNCTION OR ACCIDENT NOT PREVIOUSLY EVALUATED IN THE SAR HAS NOT BEEN CREATED. THEREFORE, NO UNREVIEWED SAFETY QUESTIONS ARE ASSOCIATED WITH THIS ACTIVITY. ICMH)

92-8476-034-RO2 STATUS: 64 i

SUBJECT:

FCR 90-0130 AllAS:

POSRC #:

92-040 ASSOC DOC :

FCR 90-0130 OTHER REFS.

EQUIPMENT:

ORG/DIV:

SYSTEM:

076 TEXT: NRC

SUMMARY

THIS EVALUATION PROVIDES THE DOCUMENTATION FOR THE ACCEPTABILITY OF THE OPERATION OF THE MAIN VENT RADIATION MONITOR (MV RM) SYSTEM. THE SYSTEM is NON SAFETY RELATED AND CAPABLE OF MEETING THE TECHNICAL SPECIFICATION REQUIREMENTS FOR ANALYZING THE PLANT VENT EFFLUENTS FOR

+

GASEOUS, TRITIUM, IODINE AND PARTICULATE RADIATION DURING NORMAL PLANT OPERATIONS AND LOW CONCENTRATION RANGES DURING AN ACCIDENT.

THIS ACTIVITY WILL MODIFY THE MV RM SYSTEM BY ADDING RIGID TUBING FROM THE EXHAUST OF THE LODINE / PARTICULATE SAMPLE PUMPS TO THE PLANT VENTS. THEREFORE. THE IODINE / PARTICULATE SAMPLE PUMP WILL NO LONGER EXHAUST TO THE MAIN VENT EXHAUST EQUIPMENT ROOM.THE IODINE / PARTICULATE SAMPLE PUMPS AND THE TRITIUM SAMPLING RIGS WILL BE PERMANENTLY MOUNTED.

THE PLASTIC HOSE ROUTED TO THE TRITIUM RIGS WILL BE REPLACED WITH METALLIC TUBING. A HASP AND i

PADLOCK JUNCTION BOX WILL REPLACE THE EXISTING LOCAL JUNCTION BOX FOR THE LODINE / PARTICULATE SAMPLE PUMP POWER TO PREVENT ACCIDENTAL UNPLUGGING OF THE SAMPLE PUMP. THE EXISTING DISCONNECT FITTINGS ON THE CHARCOAL CANISTERS WILL BE MODIFIED TO CONFORM WITH NEW SAMPLE PUMP TUBING ALSO ADDED BY THIS ACTIVITY BETWEEN THE CHARCOAL CANISTERS AND THE SAMPLE PUMPS. BECAUSE THE PROBABILITY AND CONSEQUENCES OF AN ACCIDENT HAVE NOT BEEN INCREASED AND THE POSSIBILITY OF NEW MALFUNCTIONS AND ACCIDENTS HAVE NOT BEEN CREATED, NO UNREVIEWED SAFETY QUESTIONS ARE ASSOCIATED WITH THIS ACTIVITY. (CMH) b m.

m

91-B-081-165-ROI STATUS: 64

SUBJECT:

MCR 91-081-001-01 AUAS:

POSRC #:

ASSOC DOC :

MCR 91-081-001-01 OTHER REFS:

EQUIPMENT:

ORG/DIV:

SYSTEM:

081 TEXT: NRC

SUMMARY

MCR 91-081-001-01 ADDS A REMOVABLE WALL TO THE AREA ADJACENT TO THE SPENT FUEL POOL.THE PURPOSE OF THIS WALL IS TO PROVIDE A CLEAN AREA BOUNDARY AND TO PREVENT THE PASSAGE OF LOOSE MATERIALS AND DEBRIS INTO THE SFP.

AS MENTIONED ABOVE. THE ADDED WALL IS NOT RELIED UPON TO PREVENT OR MITIGATE ANY OF THE POSTULATED ACCIDENT SCENARIOS PRESENTED IN THE UFSAR CHAPTER 14.

THEREFORE. THE WALL AND ASSOCIATED SUPPORTS DO NOT CONSTITUTE AN UNREVIEWED SAFETY QUESTION (USC) OR REQUIRE A CHANGE TO THE TECHNICAL SPECIFICATIONS. (CMH) 92-B-076-034-ROO STATUS: 62

SUBJECT:

FCR 90-0130 ALIAS:

POSRC #:

92-040 ASSOC DOC :

FCR 90-0130 OTHER REFS:

EQUIPMENT:

ORG/DIV:

SYSTEM:

076 TEXT: NRC

SUMMARY

THIS EVALUATION PROVIDES THE DOCUMENTATION FOR THE ACCEf'TABluTY OF THE OPERATION OF THE MAIN VENT RADIATION MONITOR (MV RM) SYSTEM. THE SYSTEM IS NON SAFETY RELATED AND CAPABLE OF MEETING THE TECHNICAL SPECIFICATION REQUtREMENTS FOR ANALYZING THE PLANT VENT EFFLUENTS FOR GASEOUS. TRITIUM IODINE AND PARTICULATE RADIATION DURING PLANT OPERATIONS & LOW CONCENTRATION RANGES DURING AN ACCIDENT. THIS ACTIVITY DOES NOT CONSTIRITE AN UNREVIEWED SAFETY QUESTION. (CMH)

F 91-8-024-049-R00 STATUS: 62

SUBJECT:

85-0083

[

ALIAS:

?

POSRC #:

91-118 l

ASSOC DOC :

FCR 85-0083 OTHER REFS:

[

EQUIPMENT:

ORG/DIV:

SYSTEM:

024 i

t TEXT: PROPOSED CHANGE THIS PROPOSED ACTIVITY WILL MAKE THE FOLLOWING MODIFICATIONS TO THE EMERGENCY DIESEL GENERATORS (EDG*S) 11,12 & 21 LUSRICATING OIL SYSTEM IN ORDER TO PROVIDE DIFFERENTIAL PRESSURE i

INDICATION OF THE LUBE OIL FILTER & STRAINER ASSEMBLIES. PRESENTLY THE FILTER & STRAINER I

CANISTERS ARE EACH PROVIDED WITH A PRESSURE GAUGE (OPl4778 & 79, OPl4786 & 87, OPl4794 & 95 RESPECTIVELY). EACH GAUGE IS CONNECTED TO THE INLET & OUTLET CAVITIES OF THE CANISTER SY MEANS OF TUBING CONNECTED TO A THREE WAY VALVE (11DLO1006,1007,120LO1006,1007 &

21DLO1006 & 1007 RESPECTIVELY). THE VALVE IS POSITIONED TO READ THE INLET PRESSURE & THEN

[

REPOSITIONED TO READ THE OUTLET PRESSURE. EACH GAUGE IS SUPPORTED ONLY SY ITS STEM

{

CONNECTION TO THE THREE WAY VALVE. THIS ACTIVITY DOES NOT CONSTITUTE AN UNREVIEWED SAFETY l

QUESTION. (CMH)

.................................. SEE ORIGINAL 50.59 FOR CONTINUATION " * " " * " " " * " ' * * * * " * " " ' * " '

j 91-8-999-064-R01 STATUS: 64

SUBJECT:

FCR 910259 AllAS:

POSRC #:

93-092 i

ASSOC DOC :

FCR 910259 L

4 OTHER REFS:

EQUIPMENT:

i ORG/DIV:

[

SYSTEM:

102 i

TEXT: NRC

SUMMARY

THIS ACTIVITY, FCR 91-0259, DELETES THE NOTATION IN SECTION 10-A.1.15 OF THE UFSAR WHICH NOTES THAT " WATERTIGHT DOORS ARE LOCKED CLOSED AND ALARMED *. THIS PASSAGE REFERS TO THE WATERTIGHT DOORS TO THE MAIN STEAM PENETRATION ROOMS IN THE AUXILIARY BUILDING. A REVIEW OF l

OUR ORIGINAL LICENSING BASIS AND TECHNICAL SPECIFICATION BASIS HAS DETERMINED THAT OUR I

CURRENT ADMINISTRATIVE CONTROLS FOR THE DOORS MEET THE

• ACCEPTANCE LIMIT

• PER NSAC 125. THIS ACTIVITY IS NOT AN UNREVIEWED SAFETY QUESTION. (CMH)

?

.i t

I

90-8-045-129-ROO STATUS 64 SUSJECT:

EVALUATION OF EROSION OF THERMAL SLEEVES AUAS:

POSRC #:

90-138 ASSOC DOC :

OTHER REFS:

EQUIPMENT:

ORG/DIV:

SYSTEM:

045 TEXT: PROPOSED CHANGE THE PROPOSED ACTNITY IS THE EVALUATION OF THE EROSION OF THE THERMAL SLEEVE OF THE STEAM GENERATOR FEEDWATER NOZZLES FOR THE CALVERT CLIFFS UNIT 1 & 2 STEAM GENERATORS & HOW THIS APPUES TO THE SUSCEPTIBlUTY OF THE NOZZLE TO SUSTAIN A WATER HAMMER EVENT. THIS ACTIVITY ALSO CONSIDERS THE EFFECTS OF THERMAL FATIGUE & LOOSE PARTS DUE TO THERMAL SLEEVE EROSION IN THE STEAM GENERATORS. THERE IS ONE THERMAL SLEEVE FOR EACH GENERATOR. THIS ACTIVITY DOES NOT CONSTITUTE AN UNREVIEWED SAFETY QUESTION. (CMH) 90-B-052-179-ROO STATUS: 64

SUBJECT:

RELOCATE SAFETY RELIEF VALVES IAW 89-0098-01 & 89-0098-02 ALIAS:

POSRC #:

90-196 ASSOC DOC :

FCR 89-0098 OTHER REFS:

s EQUIPMENT:

ORG/DIV:

4 SYSTEM:

052 TEXT: PROPOSED CHANGE THIS PROPOSED ACTIVITY WILL RELOCATE THE SAFETY RELIEF VALVES 1(2)RV211,221,231 & 241 FOR THE SAFETY INJECTION TANKS (SIT *S) 11 A(21 A),11 B(228),12A(22A) & 12B(228) RESPECTIVELY. THE s

REUEF VALVES WILL BE RELOCATED FROM THE SIT'S VENT LINES TO THE NITROGEN FILL UNES UPSTREAM OF THE SfT AT AN ELEVATION OF APPROXIMATELY 69 FEET. THIS ACTIVITY DOES NOT CONSTITUTE AN i

UNREVIEWED SAFETY QUESTIN. (CMH) l l

l l

l I

1

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90 8-052-009-ROO STATUS: 64 i

SUBJECT:

FCR 90-0011 AUAS:

POSRC#:

90-018 '

ASSOC DOC :

FCR 90-0011 OTHER REFS:

EQUIPMENT:

I ORG/DIV:

SYSTEM:

052 TEXT: PROPOSED CHANGE THE COMPONENT COOUNG WATER (CCW) FLOWRATE TO THE SHUTDOWN COOUNG HEAT EXCHANGERS (SOCHX *S) WILL BE REDUCED DURING ALL MODES OF OPERATION.

THE VARIOUS DISCUSSIONS OF THE SHUTDOWN COOUNG (SDC) SYSTEM IN THE FSAR WILL BE CLARIFIED TO SHOW THE COMPONENT COOUNG WATER TEMPERATURES EXPECTED DURING DIFFERENT MODES OF OPERATION. THIS ACTNITY DOES NOT CONSTITUTE AN UNREVIEWED SAFETY QUESTION. (CMH) 90-8-999-033-R00 STATUS: 62

SUBJECT:

FCR 88-0221 AUAS:

POSRC #:

90-079 ASSOC DOC :

FCR 88-0221 OTHER REFS:

EQUIPMENT:

ORG/DIV:

SYSTEM:

102 TEXT: PROPOSED CHANGE MODIFICATIONS FOR UNITS 1 & 2 ARE AS FOLLOWS:

1) PROVIDE LAMP TEST CAPABluTY FOR AC GROUND INDICATORS ON PANEL 1C24A,

2) PROVIDE POWER AVAILABLE INDICATION ON PANELS 1C05 (2005) FOR THE REACTOR VESSEL LEVEL MONITORING SYSTEM (RVLMS) UGHT ARRAYS,

3) MODIFY FEEDWATER HEATERS 11(21) AND 12(22) HIGH LEVEL ALARM CONDITIONS TO BE CONSISTENT WITH OTHER ALARMS OF THE SYSTEM AND MOVE FEEDWATER HEATER 14(24) ALARMS FROM PLANT ANNUNCIATION TO PLANT COMPUTER. AND

4) CHANGE LAMP INDICATION AT 1CO3 (2CO3) FOR PANELS 1C47,1C48 (2C65,2C661 POWER LOSS AND LOSS OF REMOTE AND M ANUAL TRIP INDICATION FROM AMBER TO WHITE, NORMALLY ON INSTEAD OF NORMALLY OFF. THIS ACTIVITY DOES NOT CONSTITUTE AN UNREVIEWED SAFETY OUESTION. (CMH)

89-B-045-OOS-RO1 STATUS: 84

SUBJECT:

FCR 88-0128 ALIAS:

POSRC#:

90-153 ASSOC DOC :

FCR 88-0128 OTHER REFS:

EQUIPMENT:

ORG/DIV:

SYSTEM:

045 TEXT: PROPOSED CHANGE TO PERMIT CONTINUED OPERATION OF THE HIGH PRESSURE FEEDWATER HEATERS 16At8) AND 26A(B) EVEN THOUGH THEY ARE NOT IN STRICT COMPLIANCE WITH THE DESIGN CODE. IN THAT THEY ARE NOT EQUIPPED WITH THE HIGH PRESSURE RELIEF PROTECTION REQUIRED BY THE ASME BOILER AND PRESSURE VESSEL CODE. SECTION Vill. AND TO REVISE SECTION 10.2.3 AND TABLE 10-1 OF THE FSAR.

TO REVISE THE ORIGINAL 50.59 SAFETY EVALUATION FOR FCR 88-0128 SUPPLEMENT O. AND REMOVE THE STATEMENT IN SECTION 10-2.2.1 OF THE FSAR THAT CREDITS THE HIGH PRESSURE TRIP AS THE MEANS TO LIMIT ANSI B.31.1 OCCASIONAL LOADS ON THE FEEDWATER PIPING TO 1% OF THE OPERATING PERIOD.

THIS ACTIVITY DOES NOT CONSTITUTE AN UNREVIEWED SAFETY QUESTION. (CMH)

[

89-B-012-032-ROO STATUS: 62

SUBJECT:

FCR 85-0085 ALIAS:

POSRC #:

90-027 ASSOC DOC :

FCR 85-0085 OTHER REFS:

EQUIPMENT:

ORG/DIV:

SYSTEM:

012 TEXT: PROPOSED CHANGE FCR 85-0085 REQUIRES REPLACEMENT OF THE EXISTING BASKET STRAINERS 18S5205.1BS5207 IN UNIT 1 AND 2BS5205,2BS5207 IN UNIT 2. INSTALLED UPSTREAM OF THE ECCS PUMP ROOM AIR COOLERS.

THIS ACTIVITY DOES h0T CONSTITUTE AN UNREVIEWED SAFETY QUESTION. (CMH)

...............

• SEE ORIGINAL SAFETY EVALUATION FOR MORE INFORMATION ' " " " " * " "

3' 88-B-041-066-Rot 07QTUS: 02

SUBJECT:

FCR 88-0038 AUAS:

POSRC#:

90-180 ASSOC DOC :

FCR 88-0038 OTHER REFS:

EQUIPMENT:

ORG/DIV:

SYSTEM:

041 TEXT: PROPOSED CHANGE MODIFY CONTROL SCHEME FOR THE BORIC ACID AND RC MAKE UP PUMPS IN THE

• BORATE

• AND

• DILUTE

• MODES OF OPERATION. THE SIGNAL FROM FLOW TOTAUZER 2FQlS210Y WILL STOP THE BORIC ACID PUMPS AS WELL AS CLOSE VALVE 2CV210Y. SIMILARLY, THE SIGNAL FROM FLOW TOTAUZER 2FQlS210X WILL STOP THE RC MAKE UP PUMP AT THE SAME TIME AS IT CLOSES VALVE 2CV210X. IN ADDITION, THE WIRING FOR 2HS210 WILL BE MODIFIED TO REVERSE THE POSITION BETWEEN

• DILUTE

• AND

• AUTO

• MODE POSITIONS. ALSO, REPLACE EXISTING SELECTOR SWITCHES 2HS210 AND 2HS226. THIS ACTIVITY DOES NOT CONSTITUTE AN UNREVIEWED SAFETY QUESTION. (CMH) 88-B-041-094-ROO STATUS: 62

SUBJECT:

FCR 88 0020 AUAS:

POSRC#:

88-049 ASSOC DOC :

FCR 88 0020 OTHER REFS:

EQUIPMENT:

ORG/DIV:

SYSTEM:

041 TEXT: PROPOSED CHANGE l

ADD POPOFF RELIEF VALVES TO THE FOLLOWING CONTROL VALVES:

1/2CVe11,621,631,641,661 (SAFETY INJECTION TANK FILL VALVES) 1/2CV618,628,638,648 (SAFETY INJECTION TANK CHECK VALVE LEAKAGE) 1/2CV505,4260 (RCP BLEED OFF CONTAINMENT /RCW DRAIN TANK ISOLATION VALVES) l 1/2CV4010,4011,4012,4013 (STEAM GENERATOR BLOWDOWN ISOLATION VALVES)

THE ADDITION OF THE POPOFF REUEF VALVE IN THE AIR TUBING BETWEEN EACH SOLENOID VALVE AND ACTUATOR PROTECTS THE CONTROL VALVE ACTUATOR AGAINST OVER PRESSURIZATION SHOULD THE ASSOCIATED AIR SUPPLY REGULATOR FAIL WITH HIGH OUTPUT PRESSURE. THIS ACTIVITY DOES NOT CONSTITUTE AN UNREVIEWED SAFETY QUESTION. (CMH)

8 7-8-083-OS 7-RO1 STATUS: 62

SUBJECT:

FCR 87-0076 AUAS:

POSRC #:

88-025 ASSOC DOC :

FCR 87-0076 OTHER REFS:

EQUIPMENT:

ORG/DIV:

SYSTEM:

083 TEXT: PROPOSED CHANGE DOCUMENT THE ACCEPTABIUTY OF HEAVIER REPLACEMENT VALVE AND HEAVIER REPLACEMENT MOTOR OPERATOR FOR IMOV6621 (1-DR-24 ISOU. THis IS A RECTIFICATION OF THE LACK OF ANALYSIS AND DOCUMENTATION WHEN REPLACEMENT WAS MADE IN 1984. THIS ACTIVITY DOES NOT CONSTITUTE AN UNREVIEWED SAFETY QUESTION. (CMH) 8 7-B-041-098-ROO STATUS: 62

SUBJECT:

FCR 87-0113 AUAS:

POSRC#:

88477 i

ASSOC DOC :

FCR 87-0113 OTHER REFS:

EQUIPMENT:

ORG/DIV:

SYSTEM:

041 TEXT: PROPOSED CHANGE

1) REPLACE THE VALVE STEM & SPACER ON 1CV518 (128 CHG UNE STOP). 2CV518 (228 CHG UNE STOP),

1CV519 (11 A CHG UNE STOP) & 2CV519 (21 A CHG UNE STOP) WITH A NEW STEM & SPACER FOR 2 INCH STROKE.

2) REPLACE THE EXISTING FLOW CONTROL NEEDLE VALVES ON 1(2)CV517,1(2)CV518 AND 1(2)CV519 WITH NUPRO FINE CONTROL METERING VALVES. THIS ACTIVITY DOES NOT CONSTITUTE AN UNREVIEWED SAFETY QUESTION. (CMH)

i 8 2-1-064-OO1-ROO STATUS: 64

SUBJECT:

FCR 82-0051 l

AUAS:

POSRC #:

83-151 l

l

' ASSOC DOC :

FCR 82-0051 OTHER REFS:

l EQUIPMENT:

ORG/DIV-SYSTEM:

064 TEXT: PROPOSED CHANGE ALLOW THE USE OF MECHANICAL TUBE PLUGS FOR STEAM GENERATOR REPAIR. (SEE ATTACHED USTED FOR QUANTITY AND LOCATION OF TUBES PLUGGED TO DATE - NOT ALL USING MECHANICAL METHOD).

THIS ACTIVITY DOES NOT CONSTITUTE AN UNREVIEWED SAFETY QUESTION. (CMH) 82-1-064-OO.WJO STATUS: 62

SUBJECT:

FCR 83-1031 AUAS:

POSRC #:

85-017 i

ASSOC DOC :

FCR 83-1031 OTHER REFS:

EQUIPMENT:

ORG/DIV:

I SYSTEM:

064 TEXT: PROPOSED CHANGE REPLACEMENT OF EXISTING ROSEMOUNT MODEL 104-1713-001 RESISTANCE TEMPERATURE DEVCES (RTD)

AT ITE111X AND 1TE111Y WITH CONAX RTD'S AND ASSOCIATED JUNCTION BOXES AND CABLING. THIS ACTIVITY DOES NOT CONSTITUTE AN UNREVIEWED SAFETY QUESTION. (CMH) i

ATTACHMENT (2)

CALVERT CLIFFS NUCLEAR POWER PLANT 10 CFR 50.59 SUMMARIES NOT PREVIOUSLY PROVIDED TO NRC Baltimore Gas and Electric Company Docket Nos. 50-317 & 50-318 November 00,1995

F a

4 NMRB018 NUCLEIS 10/15/1995 Search Process A & oc Report 137 STATUS 62 OR 64 50.59S (10/01/1994 THRU 09/30/1995)

Assoc Doc ID: ES9300001 Revision To: 0000 Assoc Stat: O Assoc Type: ESP Ref Doc ID:

Rev:

Refer Type:

Sender Xmtl #

Xmtt Date

===================================================================================================

Other rsfs:

Pers C2fs:

Equipment:

Org/Div:

System Code: 041 CHEMICAL & VOLUME CONTROL SYSTEM (CVCS)

Text:

NRC SUPN4ARY:

THIS SAFETY EVALUATION ADDRESSES ACTIVITIES ASSOCIATED WITH TEMPORARY ALTERATION 1 95 062. THE SCOPE OF THIS TEMPORARY ALTERATION IS THE INSTALLATION OF A BLIND FLANGE AT THE 1" OUTLET PIPE FLANGE FOR 1 RV 311 (CVCS CHARGING PUMPS SUCTION HEADER RELIEF VALVE). THE RV WILL BE REMOVED FOR MAINTENANCE AND THE BLIND WILL BE INSTALLED TO PREVENT THE RELEASE OF RADIO GASES AND WATER FROM THE RV OUTLET PIPING DISCHARGING TO THE WASTE PROCES$1NG SYSTEM (WPS) HEADER AND THE APPLICABLE DEGASIFIER TANK IN SERVICE.

THE CHARGING PUMPS COP 590N SUCTION HEADER RV PROVIDES OVERPRESSURE PROTECTION FOR THE PIPING AND THE COMPONENTS AT THE SUCTION SIDE OF THE CHARGING PUMPS.

THE RV DISCHARGES TO THE WASTE PROCESSING SYSTEM. THE PIPING AT THE INLET OF THE RV IS M 600 CLASS HC 16 & THE PIPING AT THE OUTLET OF THE RV IS M 600 CLASS HC 2. BOTH LINES ARE DESIGNED IN ACCORDANCE WITH AQ WPS, THE OUTLET PIPING IS FUNCTIONALLY NSR, BUT IS DESIGNED SEISMIC CLASS I.

THE CVCS CHARGING SYSTEM WILL BE OUT OF SERVICE (TAGGED OUT GT SERVICE)

DURING THE PERIOD THAT THE 1 RV 311 IS REMOVED. THE VOLUME CONTRut TANK (VCT)

WILL BE DRAINED BEFORE THE VALVE REMOVAL. THEREFORE, THE VCT, CVCS CHARGING PUMP SUCTION HEADER PIPING AND COMPONENTS WILL NOT REQUIRE OVER-PRESSURE PROTECTION. ADDITIONALLY, THE RV INLET PIPING WILL REMAIN OPEN WHILE THE RV IS REMOVED, I.E., NO BLIND WILL BE INSTALLED.

ALL DESIGN REQUIREMENTS OF THE WPS SYSTEM PIPING ARE MET, THE REMAINING CVCS AND WPS PIPING IS ADEQUATELY SUPPORTED AND MEETS SEISMIC REQUIREMENTS, AND THERE ARE NO IMPACTS TO OTHER PLANT SYSTEMS. THERE ARE NO AFFECTS ON ANALY2ED MALFUNCTIONS OR ACCIDENTS, AND NO NEW MALFUNCTIONS OR ACCIDENTS ARE CREATED. THEREFORE, THIS ACTIVITY DOES NOT CONSTITUTE A USQ.

(CMH)

..