ML20217E387

Forwards Rev 14 to USAR for Prairie Island Nuclear Generating Plant.Rev Updates Info in USAR Up Through 970331

ML20217E387
Person / Time
Site:	Prairie Island
Issue date:	09/30/1997
From:	Wadley M NORTHERN STATES POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20217E391	List: ML20217E387
References
NUDOCS 9710070010
Download: ML20217E387 (49)

Text

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Northern States Power Company Pralrie Island Nuclear Generating Plant 1717 Wakonade Dr. East Welch. Mmnesota $5089 1

September 30,1997 -

10 CFR 50.71(e)

U S Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 PRAIRIE ISLAND NUCLEA.R GENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Submittal of Revision No.14 to the Uodated Safety Analysis Reoort (USAR)

Pursuant to 10 CFR 50.71(e) we are submitting one original and 13 copies of Revision

' No.14 to the Updated Safety Analysis Report (USAR) for the Prairie Island Nuclear

. Generating Plant. This revision updates the information in the USAR up through March 31, _1997(although some information is more recent).

Exhibit A contains descriptions and summaries of the safety evaluation for changes, tests and experiments made under the provisions of 10 CFR 50.59 during the period L

'since the last update. Exhibit A also contains discussions of changes made to l'

regulatory commitments made within our Regulatory Commitment Change Process.

. Exhibit B contains the USAR page changes and instructions for entering the pages. l-f3 In this letter we have made no new Nuclear Regulatory Commission commitments.

/;

Please contact Jack Leveille (612-388-1121, Ext. 4662) if you have any questions related to this letter.

ED NO lll$5 0,$)$ fl]lfll O

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PDR, J

I certify that the information presented herein accurately presents changes made since the last updating submittal of the Prairie Island USAR.

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h Michael D Wadley Vice President-Nuclear Ge ration c: Regional Administrator - Region Ill, NRC Senior Resident inspector, NRC NRR Project Manager, NRC J E Silberg -

Attachments: Exhibit A, summaries of safety evaluations Exhibit B, USAR page changes e

Exhibit A PRAIRIE ISLAND NUCLEAR GENERATING PLANT ANNUAL REPORT OF CHANGES, TESTS AND EXPERIMENTS - SEPTEMBER 1997 The following sections include a brief description and a summary of the safety evaluation for each of those changes, tests, and experiments which were carried out without prior NRC approval, purcuant to the requirements of 10 CFR Part 50, Section 50.59(b). Also included are discussions of changes made to regulatory commitmente made within our Regulatory Commitment Change Process.

Modification 85YS80 - New Administration Building Description of Modification Existing 10" main loop fire protection line 2FP-35-4 was tapped by a new 6" fire line to provide service to the new Administration Beliding wet sprinkler fire protection system.

This tie-in to the plant fire protection lina is Tech Spec related due to the out-of-service condition which arises from the hose station line interruption during the new line cut-in and hydrostatic test. In addition, a valve was added downstream and a new hydrant installed.

Summary of Safety Evaluation The Tech Spec specifies that the two hydrants, HH7 and HH8 of the yard loop, not be removed from service more than 14 days without notification of the NRC; the cut-in to provide the new sprinkler service and additional hydrant was performed in 7 days. In addition, the Work Order specified alternate fire stations in the coverage crea to be used in the event of a fire.

With respect to the USAR and pending USAR submittals, the work performed does not create a possibility of an accident different than those previously analyzed.

Modification 88LO74 - Component Cooling Heat Exchanger Retum Channel Drain Valves Description of Modification Part 1 of the Modification added drain valves on the cooling water side of the Component Cooling Heat Exchangers; this will improve the process of draining the cooling water side of the heat exchangers.

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_ _ _ _ _ _~

Exhibit A Part 2 of the Modification replaces the existing drain valves on the cooling water inlet piping to the Component Cooling Heat Exchangers; this will facilitate draining the cooling water side, typically during a refueling outage. The existing globe valves are being replaced with gate valves.

This Modification was performed during consecutive refueling outages. Component Cooing Heat Exchangers were inspected and cleaned at this time.

. Summary of Safety Evaluation With respect to USAR or pending USAR submittals, the work performed does not create the possibility of an accident different than those previously analyzed. Review of the Safety Evaluation "unreviewed Safety Question Determination" responses indicated a negative response for all questions.

Modification 92L360 - Manipulator Crane Motor Drive Circuitry Upgrade Description of Modification

- This modification upgraded the Unit 2 manipulator crane motor drive circuitry with a variable frequency unit, controlling bridge, trolley and hoist motors. It also rep!sced the selsyn bridge positioning system with a video based bridge positioning system and installed a new faceplate on the operator's console.

Summarv of Safety Evaluation The safety evaluation concludes that the proposed activity is not an unreviewed sLfety question. The basis for this conclusion is that the load bearing portion of the crane is the only safsty-related part of the cane and the new equipment is lighter than the old equipment and fits into the existing nvelope, not affecting the load bearing portion of the crane. The circuitry itself is not safety-related, does not affect the interlock circuitry of the crane and has fault diagnostics which shut it down in case of a problem. In all application of the equipment, the crane still lifts only one assembly at a time and is thus bounded by the dropped fuel assembly analysis in USAR Chapter 14.

Modification 93L395 - Permanent Connecf 4 i for Steam Line Radiation Monitors to ERCS Description of Modification

_ Temporary Modification 90T0007 initially installed jumpers to connect the Steam Line Radiation Monitors to the Safety Parameter Display System (SPDS). This Modification 2

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Exhibit A '

replaces the T-Mod, whose scope replaced the jumpers with permanent wiring. Meither the T-Mod nor the subject Modification changed or will change hardware or hardware

locations.-

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p

=

e Summary of Safety Evaluation i'

i p

7With respect to USAR or pending USAR submittals, the Steam Line Radiation Monitors t

.are mentioned in the USAR but are described as being connected only to the Steam Release Computer System. This modification does not alter that connection.- The 4

i

'affected equipment is considered Post-Accident Monitoring Equipment and does not 3

alter the function of either the monitors or ERCS, Review of the Safety Evaluation L

i"unreviewed Safety Question Determination" responses indicated a negative response for all questions.

h Modification 93L416 - Unit 1 Main Feedwater Pumps Discharge Rollef Valves E

Descriotion of Modification :

1 L

The purpose of the modification is the installatlon of relief valves on the discharge of i-No.11 and 12 Main Feedwater Pumps.- The relief piping for the two pumps is routed

beside the pumps in the turbine buMing at approximate ele /ation 695'?- The discharge :

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' piping from the valves will be increased from 3/4" to 1" to lower the potential

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L backpressure on the relief valves.

Summarv of Safety Evaluation

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The modification is the result of plant response to INPO SER 92-038. The response i

recommends relief valves be installed on the' discharge of No.11 and 12 Main g

Feedwater Pumps. The discharge of the pumps is configured with a volume oetween

the discharge check valve and the discharge motor valve that could be pressurized if the volume were to heat up. Thermal expansion of this volume could result in the need to relieve 2.5 gallo_ns of water.- Since a 3/4" manual relief valve connection exists, this-L manual valve will be replaced by an automatic relief valve. Valve selection indicates-L that a 3/4" relief valve is capable.of relieving over 1000 GPM at piping design -

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- pressure. This capacity far exceeds requirements.

lWith respect to the USAR Chapter 14 or.pending USAR submittals, the work performed does not create the possibility of an accident different that those previously analyzed.

- Review of the Safety Evaluation "unreviewed Safety Question Determination"

. responses indicated a negative response for all questions.

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Exhibit A Modification 94L434 Unit 2 Cycle 17 Core Reload Description of Chance This modification replaced depleted Unit 2 fuel assemblies with a fresh reload of 48 Westinghouse VANTAGE + fuel assemblies allowing another cycle of power operation. The new it el assemblies are enriched to a nominal 4.95 w/o U235. The projected cycle length is 71,t22 MWD /MTU, which includes a 35-day coast to approximately 65% of full power.

This is equivalent to 560 effective full power days.

The Unit 2 Cycle 17 reload was developed by the NSP Nuclear Analysis Department using methodology addressed in NSPNAD-8101-A, Qualifications of Reactor Physics Methods for Application to PI Units. More details on the operational parameters can be found in NSPNAD-95002, Rev. O, Prairie Island Unit 2 Cycle 17 Final Reload Design Report, March 1995 and PI NSPNAD-95003, Rev,0, Prairie Island Unit 2 Cycle 17 Startup and Operations Report, April 1995.

Summarv of Safety Evaluation 4

The following safety concems were addressed in the safety evaluation:

A. Thermal Hydraulic Analysis B. Accident and Transient Analysis C.- Main Steam line break / Containment Response Analysis D. LOCA-ECCS Analysis E. Rod Ejection Analysis F. Fuel Handling Accident G. Refueling Shutdown Margin H. Heatup/Cooldown Curves - Reactor Vessel Radiation Surveillance Program

1. Fuel Rod Design Performance

-J. Spent Fuel Heat Load K New Fuel Rack / Spent Fuel Rack Criticality L. Core Exposure Limits /Off-site Dose Calculations M. Peak Linear Generation Rate N. Fuel Assembly Design Changes O. Fuel Reconstitution Campaign P. Startup and Operations Q. Ve.'idity of Safety Evaluation All results were acceptable and are presented in NSPNAD-95002,- Rev. O, Prairie Island Unit 2 Cycle 16 Final Reload Design Report. The LOCA analysis was performed by Westinghouse and is documented in the Unit 2 Cycle 17 LOCA Confirmation Letter 95NS*-

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i Exhibit A G-0012, February 27,1995.- This letter confirms that the operation of Prairie Island Unit 2 Cycle 17 will continue to conform to the acceptance criteria of 10CFR50.46.

- In conclusion, since all tra,1sient analyses met the acceptance criteria, there are no unreviewed safety questians for the PI Unit 2 Cycle 17 Core Reload Modification.

Modification 94L450 - Main Steam Non Return Check Valves Inspection Port Descriotion of Modification

- ASME Section XI now requires periodic evaluations of operability of the NIain Steam Non-Return Check Valves (2 per unit), This modification provides for inspection of the valves without their disassembly. The modification installs a "T" in the drain line directly below each check valve, installs a new gate valve and pipe cap in the line, and reroutes the upper "T" connection to the existing drain configuration. This new configuration will allow borescope examination of the valve internals during refueling -

outages.

Summarv of Safety Evaluation This modification has no active effect on an operating system. Review of the Safety Evaluation "unreviewed Safety Question Determination" responses indicated a negative response for all questions.

Modification 94L459 - Repowering of the Telephone System Descriotion of Modification The telephone system had two sources of power: the telephone inverter, and an

. alternate 120V source. The inverter has failed and will not be replaced. The modification removes the failed inverter and provides two other sources of uninterruptable power to the telephone system.

Summary of Safetv Evaluation

- This Safety Evaluation answered "no" to each of the seven questions for the unreviewed safety question determination, primarily because the telephone system is not a safety-related system. However, each of the USAR accidents assumes the o

telephone system is functional.

The Failure Modes section of this Safety Evaluation demonstrates that the reliability of the power supplies to the telephone system has been improved. Now there are two E

Exhibit A independent uninterruptable power supplies, each with its own battery backup. Also, each UPS is backed by a non-safety related diesel generator.

Modification 94L473 - Cooling Water supply and return to the fan coll units Descriotion of Chance The design change applies to the cooling water supply and return to the fan-coil units.

The design change removes the internals from the inlet check valve which is inside containment, and relocates the relief valve from inside containment to outside containment.

~ Summary of Safety Evaluation

- The safety-related function of the check valves is to open in order to provide cooling to the fan-coil unit. The check valves are not needcd for containment isolation in the event of a passive failure of the inside containment fan-coil unit cooling water piping in conjunction with a LOCA. Removal of the check valve internals will decrease the probability of failure of these valves to perform their safety-related function.

Relief valve relocation to a point outside containment eliminates a concern for potential boron dilution of the containment sump water during a LOCA due to a leaking relief valve.

Modification 94L475 - Remove DC Transfer Switches Description of Modification in the early 1980's, transfer switches were installed for certain DC loads and panels which allowed transfer from the normal power source to the same train alternate power source on the opposite unit. The installation assured reliable power to certain safeguards loads during outages when the safeguards batteries were out of service for testing. EDSFI concerns eventually prevented use of most of these switches. The addition of DS and D6 diesel generators alleviated the need for sharing D1 and D2 diesel generators between units, as well as related diesel controls. The DS/D6 project reduced the need for most of these transfer switches.

Summarv of Safety Evaluation Answers to all the questions in the Unreviewed Safety Question Determination were negative.

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i Exhibit A Modification 94T042 - Removal of No.122 Control Room Air Handler Discharge -

Damper CD 34144 Description of Modification -

- This temporary modification removed the discharge damper from No.122 control room air handler because it was not fully functional and not repairable. The damper will be either refurbished, replaced, or permanently removed under anather modification.

Summarv of Safety Evaluation Temporary removal of CD-34144 does not negatively impact the control room vent system since the exhaust louver in the ductwork is covered and post-maintenance testing assured ac,.quate air flow. The modified system will function within the design basis as summarized in the USAR and subsequent supporting analyses.

. Modification 94T047 (95CL02) - Cooling Water Strainer Backwash Valves Descriotion of Modification Loss of instrument air header pressure will cause the cooling water strainer backwash valves to fall full open. In this failure mode, water which would normally be directed to f

the cooling water header will instead be directed to the intake bay via the backwash piping. Loss of this flow may be detrimental to the cooling water total flow. Installation of a backup compressed gas system ensures that control of the valves will be possible during loss of instrument air scenarios.

Summarv of Safety Evaluation The new backup compressed gas system provides a redundant means of controlling the strainer back'uash valves. Review of the Safety Evaluation "unreviewed Safety Question Determination" responses indicated a negative response for all questions.

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This temporaiy modification will be made permanent with Mod 95CLO2.

Modification 95CLO4 - Valve Travel Stops to Limit Cooling Water Flow Through Component Cooling Heat Exchangers Descriotion of Modification The purpose of this modification is to install manual stops in the Cooling Water (CL)

- System flow control valves to limit the flow through the CC Heat Exchangers during normal plant operations. Physically limiting the CL flow will be accomplished by positioning the flow stops presently installed, on the valves. The ability to limit flow was 7

Exhibit A

always part of the original design, but use of the flow stops was never formally controlled,' and the stops were not used,

!Summarv of Safety Evaluation Results of the CL System ' hydraulic modeling analysis indicate there will be less than -

design flow to the Fan Coll Units and Emergency Diesel Generators, D1 and D2, during y

the worst case' accident, as described in the safety evaluation. Safety Evaluation No.

422 concluded that the CL System is capable of fulfilling accident mitigation functions with these predicted flows. With these flow rates, there is substantially less heat romoval capability associrted with the FCUs. Therefore, other system operational -

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.: changes or modifications are being evaluated in order to restore margin in _the FCU heat removal. - Physically limiting the CL flow through the CC heat exchangers (except when full flow is required) is considered a desirable operational change.

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' With respect to USAR or pending USAR submittals, the work performed does not L

create the possibility of an accident different than those previously analyzed.

l Review of the Safety Evaluation "unreviewed Safety Question Determination"-

responses indicated a negative response for all ques *lons, Modification 95L615 - Unit 1 Cycle 18 Core Reload Modification -

- Description of Chance This modification replaced depleted Unit 1 fuel assemblies with a fresh reload of 48

' Westinghouse VANTAGE + fuel assemblies allowing another cycle of power operation.

- The new fuel assemblies are enriched to a nominal 4,95 w/o U-235. The cycle length

is projected to be 21860 MWD /MTU, which includes a 21-day coast to approximately _

79% of full power, This is equivalent to 563 effective full power days, h The Unit 1 Cycle 18 reload was developed by the NSP Nuclear Analysis Department

_ using methodology addressed in NSPNAD-8101-A, Qualifications of Reactor Physics Methods for Application to PI Units. More details on the operational parameters can be found in NSPNAD-95009, Rev. O, Prairie Island Unit 1 Cycle 18 Startup and Operation Report,: December 1995.

Summarv of Safety Evaluation

. The following safety concerns were addressed in the safety evaluation:

2 A. Thermal Hydraulic Analysis :

B, Accident and Transient Analysis 8

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

p i-1 Exhibit A

C. Main Steam Line Break / Containment Response Analysis D,lLOCA-ECCS Analysisi E.1 Rod Ejection Analysis F. Fuel Hand!!ng Accident

- G.LRefueling Shutdown Margin H. Heatup/Cooldown Curves - Reactor Vessel Radiation Surveillance Program L Fuel Rod Design Performance.

'J.

Spent Fuel Heat Load K; New Fuel Rack / Spent Fuel Rack Criticality L. C. ore Exposure Limits /Off-site Dose Calculations M. Fuel Assembly Design Changes N. Startup and Operations P.. Validity of Safety Evaluation All results were presented in NSPNAD-95008, Rev. O, Prairie Island Unit 1 Cycle 18 Final Reload Design Report (Reload Safety Evaluation) as acceptable. Subsequent'

analysis revealed that not all of the acceptance criteria fcr the large steam line break had been met. Revisions 1 and 2 to NSPNAD-95008 address the questionable

- parameters and found them to be acceptable. Revisions 1 and 2 to the 95L515 safety evaluation covered cycle operations to 13 GWD/MTU and end of cycle, respectively.

, The LOCA analysis was performed by Westinghouse ~and is documented in the Unit 1 Cycle 18 LOCA Confirmation Letter 95NS*-G-0038, September 1,1995.. This letter

- confirms that the operation of Prairie Island Unit 1 Cycle 18 will continue to conform to

- the acceptance criteria of 10CFR50.46.'

In conclusion, since all transient analyses met the acceptance criteria, there are no unreviewed safety questions for the PI Unit 1 Cycle 18 Core Reload Modification.

Modification 95L518 Unit 2 Cycle 18 Core Reload

- This modification replaced depleted Unit 2 fuel assemblies with a fresh reload of 48.

Westinghouse VANTAGE + fuel assemblies allowing another cycle of power operation..The

new fuel assemblies are enriched to a nominal 4.95 w/o U235. The projected cycle length is 21,530 MWD /MTU.

1The Unit 2 Cycle 18 reload was developed by the NSP Nuclear' Analysis Department using '

methodology addressed in NSPNAD-8101-A, Qualifications of Reactor Physics Methrxis for

- Application to P1 Units. More' details on the operational parameters can be found in

NSPNAD-96007,' Rev,0, Prairie Island Unit 2 Cycle 18 Final Reload Design Report, f January 1997 and PI NSPNAD-96006, Rev. O, Prairie Island Unit 2 Cycle 18 Startup and L Operations Report,- December 1996.

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Exhibit A:-

cThe following safety concerns were addressed in the_ safety evaluation:-

A. Thermal Hydraulic Analysis B. Accident and Transient Analysis _

C. Main Steam line break /Conta!nment Response Ana'ys:s D. LOCA-ECCS Analysis -

E. Rod Ejection Analysis F. Fuel Handling Accident.

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- G. Refueling Shutdown Margin._ _

H. Heatup/Cooldown Curves - Reactor Vessel Radiation Surveillance Program

- 1.- Fuel Rod Design Performance J. Spent Fuel Heat Load K. New Fuel Rack / Spent Fuel Rack Criticality L Core Exposure Limits /Off-site Dose Ca!culations M. Peak Linear Generation Rate N. Fuel Assembly Design Changes O. Fuel Reconstitution Campaign P. Startup and Operations

' Q. Validity of Safety Evaluation,

All results were acceptable and are presented in NSPNAD-96007,L Rev. Of Prairie Island Unit-2LCycle 18 Final Reload Design Report. The LOCA analysis-was performed _ by

Westinghouse and is documented in the Unit 2 Cycle 18 LOCA Confirmation Letter 96NS-G-0033, November 25,1996. This letter confirms that the operation of Prairie Island Unit 2 Cycle 18 will co ttinue to conform to the acceptance criteria of 10CFR50.46. In conclusion,-

F.

since all transient analyses met the acceptance criteria, there are no unreviewed safety-

- questions for the PI Unit 2 Cycle 18 Core Reload Modification l Modification 95SG01 - Changes to Steam Generator Welded Tubesheet Sleeve

_ Descriotion of Chanae This Modification incorporated changes which Combustion Engineering had implemented in sleeve fabrication and installation processes. This modification:

_1) Implemented the use of the " straight tubesheet sleeve" configuration.

2) Implemented an installation sequence such that the upper weld and heat -

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treatment'are done F tior t > making the lower weld.

1 i 3) Specified the use of state of the art eddy current technology for baseline -

examination.-

The fabrication of the original style of sleeve installed a 5 d'egree taper (flare) on the lower end of the sleeve. The fabrication of the new sleeve is without the tapered end. In 10

Exhibit A-l both the original tapered. sleeve ~and in the new straight sleeve, the final configuration is:

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the same.1 With the straight sleeve,- the taper is installed once the sleeve is in place in

the tube.; The purpose of the taper is to insure adequate sleeve to tube end contact for-the tube end weld.

^To reduce the residual stress in the upper weld joint, the lower tube end is not welded and remains free ?.o move until the upper joint has been accepted.

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The original sleeving modification, 86L958, requ. ired baseline eddy current inspection of the sleeve using a dual cross wound probe ~and bobbin probe using the multifrequency eddy current method. Recently improved probes have been developed Twhich surpass the capability of the cross wound and bobbin probes.- In particular, -

- rotating pancake probe technology provides a better examination both for baseline and in-service examinations. Therefore, when appropriately qualified, state of the art coils, fsuch as the current ZETEC + Point coil will be used to examine inservice and new sleeves.

- Summarv of Safetv Evaluation Section 14 of the USAR was reviewed for impact. The function of the steam generator tubing (and sleeves) is to 1) maintain the primary system pressure boundary and to 2).

. transfer heat from the reactor coolant system to the secondary side. These changes to-the sleeve installation process indications do not change the failure modes'or failure impact of the Steam Generator tubing. No impact on the USAR Section 14 accidents; was identified.

Modification 95SG01,- Addendum 1 -Installation of Mechanical Tube Plugs in

- Sleeved Tubes

' Description of Chance -

' This Modification; incorporated Combustion Engineering rolled mechanical plug to be-installed in a welded tubesheet sleeve at Prairie Island. The rolled plug is simpler to

~ install and reduces radiation exposure. This plug would also allow recovery of sleeved tubes in the future, if an acceptable repair is available for sleeve repair. The

- Combustion Engineering mechanical tube plug for sleeves is fabricated from Alloy 690 bar stock!

Prior to the plug insertion, the sleeve is rolled out to hard contact with the parent tube

inside diameter. Then, the plug is inserted into the sleeved tube end and rolled into contact with the inside wall of the sleeve by mechanical rolls. The hard roll contact region is similar to the normal mechanical plug at the original hard roll region;

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Exhibit A

-The plug is held in place as a result of the high contact force between the plug and the sleeve ID created during the rolling operation.

Summarv of Safetv Evaluation Section 14 of the USAR was reviewed for impact. The function of the steam generator tubing (and sleeves) is to 1) maintain the primary system pressure boundary and to 2) transfer heat from the reactor coolant system to the secondary side. Use of the Combustion Engineering rolled mechanical plug in a Combustion Engineering welded tubesheet sleeve does not change the failure modes or failure impact of the Steam Generator tubing ho impact on the USAR Section 14 accidents was identified.

Modification 96AF01 Part 2 - Auxiliary Feedwater Pump Runout Protection Description of Chan. ggt l

This modification will separate the low suction pressure switch and the low discharge pressure switch time delay by the lustallation of a new time delay for the low discharge

. pressure switch. The modification will allow the TDAFWP to start, come up to operating speed and normal operating discharge pressure prior to instating a trip from the low discharge pressure switch. The new time delay relay will prevent a trip of the TDAFWP on low discharge pressure during the short t period following pump start while the pressure is stabilizing. The existing time delay for the low suction pressure switch will not be changed.

The modification also provides a bypass of the low discharge pressure trip of the turbine-driven auxiliary feedwater pumps during ATWS conditions by blocking the trip when the reactor trip breakers are closed. This circuit insures that, following completion of the AMSAC initiation of auxiliary foedwater during an ATWS transient, the turbine-driven auxiliary feedwater pumps continue to run.

Summarv of Safety Evaluation With respect to USAR or pending USAR submittals, the work performed does not create the possibility of an accident different than those previously analyzed.

Review of the Safety Evaluation "unreviewed Safety Question Determination" responses indicated a negative response for all questions.

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.I Exhibit A Modification 96CW01 - Circulating Water intake Canal Thermal Distribution Description of Chance A concrete guidewall was constructed in the center of the circulating water intake canal so that equal quantities of cold river water and warm recycle water reach the circulating water pumps condenser inlet temperature can be uniformly controlled.

Summarv of Safety Evaluation The project had no effect on design basis accidents. USAR Section 11.5.2 describes

. the operation of the circulating water system including the intake canal.

Modification 96FP01 - Addition of Priming Tap to No.121 Jockey Fire Pump Description of Modification l

This modification installed a priming tap between the No.121 Jockey Fire Pump and

the Screenhouse Eductor Header. The Jockey Pump had experienced a loss of prime l

on several occasions. This installation ensures that the Jockey Pump remains primed L

during standby conditions.

Summary of Safety Evaluation j

The function of the Jockey Fire Pump is to maintain fire header pressure by cycling on when header pressure decreases and cycling cff when pressure returns to normal. The Jockey Pump is nonsafety-related and is not required to be operable per the Plant Fire

-Protection Program, F5 Appendix K. A review of possible failure modes for the 1

proposed change found no new types of failures created and no effect on safety-related equipment.

Modification 96FP03 - Removal of Relay Room Fire Protection Supply Valve FR-51 2 Description of Modification

- The modification removed FR-51-2, relay room fire protection supply valve. There is no sprinkler system piping downstream of the valve,' which was abandoned in place during-original construction. The valve had corroded and had begun leaking. Figure 10.3-4 will show the removal of the valve.

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Exhibit A Summary of Safety Evaluation The valve is nonsafety-related and in fact its outist has been blank-flanged since original construction. Since the valve will never be used to fight a Re at Prairie Island, it is better removed than replaced.

Modification 96SG01 - Unit 1 Steam Generator Tube Removal D6scriotion of Chance Circumferential eddy current indications and volumetric eddy current indications were found in Combustion Engineering welded tube sheet sleeve upper welds. These indications were not well understood and could have been from either the welding process or service-induced degradation. In order to determine the root cause of these indications and evaluate the effect on the structural and leak tight integrity of the Combustion Engineering welded tubesheet sleeve, it was necessary to remove sleeve samples from 12 Steam Generator hot leg.

This modification:

1) Evaluated the effects of the removal of sleeved tube samples from 12 Steam Generator including the boring of the tubesheet hole to accommodate the outside diameter of the upper sleeve weld joint.

2) Rernoved 5 tube / sleeve samples

3) Installed Framatome Technologies Remote Welded Plugs in the tubesheet hole left by the removed tube.

In order to remove the sleeved tube, the diameter of the tubesheet hole was enlargad by borir:g. The effect of the enlarged bore is to decrease the ligament thickness (the shortest distance between adjacent tube holes in this square pitched tube bundle) of the four bore holes adjacent to the bored hole. The primary stresses in the thinned ligament and the effect on the tubesheet fatigue usage factor as required by ASME Code Section ill were determined to be acceptable.

The portion of the tube romaining in the steam generator was evaluated for the potential of becoming a vibration hazard to adjacent tubes. Turbulence vibration smplitudes and related forces were shown to be small enough to preciude the generation of a locce part through fretting wear or cyclic fatigue failure at the tube support plate.

Summarv of Safety Evaluati.on The welded plug maintains the same presaure boundary integrity and leak tightness of the original tube by plugging the opened tubesheet bore hole.

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L Exhibit A The increase in bore hole size did not significantly reduce the tubesheet stress and fatigue margins.

Section 14 of the USAR was reviewed for impact. The function of the steam generator tu'oing is to 1) maintain the primary system pressure boundary and to 2) transfer heat from the reactor coolant system to the secondary side. This modification does not change the failure modes or failure impact of the Steam Generator tubing. No impact on the USAR Section 14 accidents was identified.

Modificrtion 96T049 (97WLO1)- Silica removal from Refueling Water Storage Tank using reverse osmosis Description of Modification This temporary modification ?nstalled a portable reverse osmosis system to remove si!!ca from the refueling water storage tank.

Summary of Safetv Evaluation Review of the Safety Evaluation "unreviewed Safety Question Determination" responses indicated a negative response for all questions.

This temporary modification will be made permanent with Mod 97WLO1.

Modification 96WS01 - Spent Resin Tank Video Probe Description of Modification The modification removed existing level indicators and an annunciator assembly, which were not reliable, and installed a stainless steel pipe for insertion of a video probe for monitoring Spent Resin Tank level.

Summarv of Safety Evaluation The Spent Resin Tank is not safety-related and is not described in the USAR, except on Figure 9.1-8.

Safety Evaluation 097, Rev.11 - Hydrogen Generation and Control pescription of Chance 15 I

)

Exhibit A USAR Section 5.4.2 tracks zine and aluminem inventories inside containment. The basis for trackirig these material inv atories is (1) to ensure that the conclusions regarding the time to reach a 4% hydrogen concentration are not affected and (2) that the recombiners will perfo.m their design basis function. Recent calculat ons have shown that there is significant margin available between the results and these bases.

Based on these margins, and for simplification, it is desh able to track these material inventories outside of the USAR. A eMculation and Engineering Manual Section were issued to accomplish this objective. 7he purpose of this rSYision to Cafety Evaluation No.

97 is to provide the justification for a USAR revision removing these material inventories.

Summarv of Safety Evaluation Based on the contribution of zine and aluminum surfeces to hydrogen gaeration during an accident, maintaining an accurate inventory of these materials inside of containment is important. This is maintained in the hydrogen control log. In addition to maintaining accountability of material inventories, the control log updates the time to l

reach a 4% hydrogen concentration in containment without use of the recombiners, and calculates the hydrogen generation rate vs. time for evaluation of recombiner effectiveness.

These material inventories will now be tracked as part of this analysis. This analysis is maintained in the analysis of record. The Engineering Manual provides guidance on the use of the control log and performing updates to the calculation. The acceptance l

criteria in the calculation and the Engineering Mar ~al are as follows:

"The maximum hydrogen build up is limited to 4 volume percent. Hydrogen Recombiners are used to control the hydrogen concentiation. From the time the recombiner(s) are started, there must be sufficient time availab!s for the recombiner(s) to reach peak efficiency before the 4% limit is exceeded, and the recombiner(s) must be able to limit the hydrogen concentration to less than the 4%

limit.

The USAR states that the hydrogen concentration will reach 4% in approximately 16 days with no hydrogen removal from the recombiners. Provided that the days to reach 4% remains greater than 19, no safety evaluation or screening is necesseni.

This provides approximately 20% margin above the safety analysis."

The additional 20% margin is self-imposed and controlled by NSP. The 20% margin above the safety analysis provides a basis for concluding that it is unnecessary to track these material inventories in the USAR. Based on the simplification gained from this

)

USAR thange and the lack of impact on the safety analysis, these changes to the USAR are considered acceptable.

16

Exhibit A Safety Evaluation 261, Addendum 2 - Reclassif~ cation of Charging Pumps Description of Chance The safety evaluation supports change of the safety classifkation of the charging pumps to safety related for pressure boundary only.

Summary of S9fety Evaluation Charging pump functions are either not safety related, or, if safety related. can be performed t,y other equipment. NRC accepted the proposalin a letter dateu.anuary 8, 1996.

Safety Evaluation 353, Addendum 1 - Use of Carboline 890 Inside Containment Description of Chance Presently, Phenoline 305 is used inside containment as a qualified coating. Due to consolidation of some of itt products, Carboline is discontinuing the Phenoline 305 products. The Phenotine 305 products are being replaced with Carboline 890 finish coat and 801 primer. Safety Evaluation 353, Revision 0, previously evaluated the acceptability of using Carooline 801 primer. The purpose of this 10CFR50.59 review is to justify the acceptability of using the Carboline 890 product.

Summary of Safety Evaluation Coatings are used on structural components (concrete and steel), piping systems, etc.

inside of containment. The coatings provide corrosion resistance, ease of decontamination, etc. Since coatings are used inside of containment, additional criteria are invoked to ensure that post-accident (LOCA) mitigation is not aggravated; i.e.,

1. Materials which comprise the coating. For example, the coating should not contain zinc. The presence of zinc could generate hydrogen during the DBA.

2. The ability to withstand post-LOCA environmental conditions without deteriorating. This is important from containment sump grating blockage and RHR ingestion perspectives. Test conditions and acceptance criteria are specified in ANSI N101.2.

The Carboline 890 product was evaluated with regard to these criteria and determined acceptable for in-containment use.

17

Exhibit A Safety Evaluation 364, Rev I - Removal of RHR to CS Pump MOV's from the GL 8910 Program Qgigriotion of Chanoe The safety evaluation justifies removal of the following motor-operated valves (MOV's) from the Prairie Island Generic Letur 89-10 Program. The MOV's evaluated by the safety evaluation are:

MV-32096, No.11 RHR Pump to 11 Containment Spray Pump Suction a

MV-32097, No.12 RHR Pump to 12 Containment Spray Pump Suction MV-32108, No. 21 RHR Pump to 21 Containment Spray Pump Suction MV-32109, No. 22 RHR Pump to 22 Containment Spray Pump Suction Summary of Safety Evaluation Removal of these valves from the Generic Letter 8910 program mandates that the breakers for the valves be in the OFF position so the valves cannot be mispositioned at i

any time. This change will alter the valve classification from active to passive for L

accident mitigation. -With the reclassification, different failure modes considerations allow tne safety function of the containment spray pump suction check valves (CA-11-1, 2CA-11-1, CS-16, CS-17, CS-46, and CS-47) to changed, and this reduces the testing required by Section XI of the ASME Code. The safety evaluation discusses these changes and concludes there are no unreviewed safety concerns.

Safety Evaluation #369 - Cable Tray Fill and Spacing Concerns Descriotion of Charyge The Safety Evaluation provides justification for:

1. Implementation of the National Electric Code 1990, Article 318, as guidance for cable tray fill and spacing criteria in the USAR.

2. Revision of the current USAR cable tray fill limit from "approximately 60%" to a more conservative value of "approximately 50%" consistent with NEC guidance.

Summarv of Safety Evaluation The National Electric Code cable tray fill and cable ampacity criteria for power, control and instrumentation cables provides a consistent, documentec. and widely accepted

. basis for determining the acceptab;;ity of cable installations. Existing power cable 18

Exhibit A ladders meet the NEC ampacity criteria. 50% fill for control and instrument cable tray sections is consistent with the NEC requirements and is acceptable based on existing plant configuration. Prairie Island already uses the NEC as a standard.

Safety Evaluation 390, Addendum 1, Rev.1 > Unit 1 Low Pressure Turbine Disk Inspection Interval Extension Description of Chanae The Safety Evaluation addressed the safety impact of extending the Unit i low pressure turbine disk inspections required by USAR Section 12.2.7.2 from a nominal 5 operating years to a maximum 841/2 operating months. The analysis considered a specific fracture analysis and probabilistic study to verify the extended interval did not exceed allowable crack sizes or the safeguards equipment damage design basis of 4

10 per year.

I Summary of Safetv Evaluation The analysis concluded that the Unit 1 inspection interval could be extended to 841/2 operating months from the last disk inspection in 1990. The crack propagation analysis for disks that are potential missiles resulted in a maximum inspection interval of 217 months before a crack would grow to 1/2 the critical crack size for catastrophic failure (1/2 that time for known existing cracks). This analysis is consistent with the original design basis of allowable crack sizes and acceptable probabilities as defined in USAR Section 12.2.7.

Safety Evaluation 390, Addendum 2 - Unit 2 Low Pressure Turbine Disk Inspection Interval Extension Description of Chance The Safety Evaluation addressed the safety impact of extending the Unit 2 low pressure turbine disk inspections required by USAR Section 12.2.7.2 from a nominal 5 operating years to a maximum 89 operating months. The analysis considered a specific fracture analysis and probabilistic study to verify the extended interval did not exceed allowable crack sizes or the safeguards equipment damage design basis of 4

10 per year.

Summarv of Safety Evaluation The analysis concluded that the Unit 2 inspection interval could be extended to 89 operating months from the last disk inspection in 1990. The crack propagation analysis 19

Exhibit A for disks that are potential missiles resulted in a maximum inspection interval of 232 months before a crack would grow to 1/2 the critical crack size for catastrophic failure (1/2 that time for known existing cracks). This analysis is consistent with the original

. design basis of allowable crack sizes and acceptable probabilities as defined in USA 51 Section 12.2.7, Safety Evaluation 399 - USAR Off-Site Dose Analysis Update Description of Chance t

Safety Evaluation No. 399 reviewed the acceptability of updating the off site and control room dose analyses in the USAR.

Summarv of Safetv Evaluation j

During a review of the dose analysis as presented in the USAR, inconsistencies between i

the analysis methodology and regulatory guidance were noted. It is noted that with these l

inconsistencies, the previous USAR analysis is still considered conservative. New dose i

analysis were performed to incorporate these changes. The results showed higher predicted doses, but still well within the regulatory limits. It is also noted that these revisions involved no physical or operational changes to the plant.

Safety Evaluation 421 - Evaluation of Unit 1 Cycle 18 MSLB - Containment

Response

Summarv The purpose of this document is to provide a justification for continued operation for Prairie Island Unit 1 Cycle 18. This safely evaluation is very similar to other safety evaluations, SE 384 and SE 50-402 (References 1'and 2), for previous cycies.

Northern States Power Company has identified deficiencies and errors in the assumptions and input to the current licensing basis for the Main Steam Line Break (MSLB) event which have required a re-evaluation of the event for containment pressure response. The current licensing basis for the MSLB event is the FSAR (Referenco 3) analysis and the NRC-approved response by Northern States Power Company to IE Bulletin 80-04. (Reference 4) The analysis methodology in the FSAR and the response to IE Bulletin 80-04 utilized methods which are conservative and bounding. An engineering analysis of the MSLB event was performed with the

- deficiencies and errors corrected, and the results of that analysis demonstrate the following:

20 l

i 1

- Exhibit A 4

1): There has been no reduction in the margin of safety to the

containment design pressure limit.-

~

= 2) The current licensing basis for the MSLBfwhile containing some deficiencies and errors in the assumptions and input, still represents a conservative and bounding analysis for Unit 1 Cycle 18.

3) There are no unreviewed safety questions,

- Backaround 1

l

_in SE 384, substantial background and historical information is provided that will not be t

reproduced in this safety evaluation. The following summarizes key background

. information regarding the containment response analysis during a MSLB event:

~

'e

-The FSAR analysis and NRC approved May 1980 NSP IE Bulletin 80-04 submittal had bounded all reloads previous to Unit 2 Cycle 16. During each reload safety-

?

evaluation, Nuclear Analysis and Design (NAD) compared DYNODE MSLB mass -

and energy release to the assumptions in original FSAR and the May *>80 I.

submittal.

F

=. In 1982, NSP submitted to the NRC the report, " Reload Safety Evaluation Methods L

for Application to PI Units." (Reference 5): One part of this dLeument outlined - the I

NSP containment pressure response methods that the NRC did not approve citing c e

time constraints. The NRC stated only the mass and energy release data from the :

DYNODE-P/3 code that serves as input to the CONTEMPT-LT code was not qualified; 4

5e: In 1993 and 1994, NAD _ engineers identified minor input problems with the MSLB -

analysis that, when corrected in current methods, yielded energy and mass releases

. that exceeded tho;;e assumed in the FSAR and May 1980 submittal, fc

* -- Fmm March 1994 to June 1994, NAD analyzed 'the _MSLB containment response

[

using minor methodology changes that resulted in acceptable containment 6

pressures. The changes reflected some of the BAST project development work that L

(was i_n progress. SE 384 was written which covered Unit 1 Cycle 16, Unit 2 Cycle -

16, and Unit 1 Cycle 17.

i F

.- On May 19l 1994, NSP met with the NRC to discuss NSP's revised MSLB analysis '

with respect to containment response. The NRC determined that "the changes -

L

/ discussed at the May 19,1994, presentation were acceptable." (Reference 6)-

Safety Evaluation SE 384 was written to justify continued operation.

I~

21 h'.

-~

,... - -., - ~,, -

Exhibit A During March 1995, SE 50-402 was written to justify continued operation of Unit 2 Cycle 17 -

/

Boundino Analysi_t Although the FSAR and the May 1980 submittal provides conservative results, correcting the deficiencies in the analysis yielded overly conservative and unacceptable results with respect to the containment response acceptance criterion.

Analys3s performed outside the FSAR and May 1980 methodology are described in SE 384 and SE 50-402. Problems in the FSAR and Topical analyses were corrected while keeping the analysis as close to the Topical methods (Reference 5) as possible with two exceptions:

1. Credit was taken for the faster closure time of the steam line non-return check valve on the broken loop.

2. Liquid entrainment wss modeled explicitly.

l l

The basis for the two exceptions is detailed in SE 384. A summary of key assumptions l

utilized in this analysis is given in Table 2.

l These analyses performed for SE 384 and SE 50-402 bound Unit 1 Cycle 18 because the core physics parameters for Unit 1 Cycle 18 are bounded and no plant systems l

have been changed that would affect NAD models. While these analyses were l- -

to deviate as little as possible from the previously approved methodology while performed using unapproved methods, they present a bound:ng analysis that attempts correcting for the problems that were found in the original analyses. The analyses are intended to provide an interim solution, through Unit 1 Cycle 18, until the new methods associated with the BAST Project can be submitted to the NRC for review and approval.

Results The results of the analyses from SE 384 and SE 50-402 are presented in Table 1.

Note that the differences in results between Cycles are relatively small even though the core designs are significantly different. This confirms the statement made in the USAR (Reference 7) that the containment pressure response from a MSLB is relatively -

insensitive to core reloads. Case E has been added since SE 384 and SE 50-402 were written.

22 7

Exhibit A Table 1 Results Comparison e

+

.Pif Cy'cle;16L.

g s~w.

~w PI1 Cycie it Pi 2 Cycle 164

v., m a

=,

/.

e vm LPl 2 Cycle 17e >

Case A A7.4 psig 47.1 psig N/A allerrors resolved

@ 160-200 sec

@ 160-200 sec Case B 46.2 psig 46.2 psig.

N/A credit forcheck valves

@ 170-200 sec

@ 170-200 sec Case C 28.7 psig 28.4 psig 27.9 psig check valves, fullentralnment

@ 12 sec

@ 12 sec

@ 11 sec Case D 46.1 psig 45.9 psig N/A check valves, 63% entrainment

@ 80 sec

@ 80 sec Case E 44.5 psig -+

BASTcase which includes

@ 140 sec 1

check valves, entrainment, and no FCU l-Containment Design Pressure Limit = 46 psig L

(source: Reference 8)

Case A in Table 1 shows the results when all of the problems identified (see SE 384) were corrected. The DYNODE analysis for this case is consistent with the approved methodology outlined in Reference 4. The CONTEMPT analysis uses the unapproved methods. The mass and energy releases from DYNODE were reduced by 15% before being input into CONTEMPT.- The results show that this conservative analysis exceeds

' the containment pressure design limit.

Case B in Table i shows the results when the identified problems are corrected and credit is taken for the steam line non-return check valves. The results show that this analysis slightly exceeds the containment pressure design limit. This also provides a measure of the " worth" of the non-return check valves, which is about 1 psi of containment pressure.

Case C in Table 1 shows the results when the identified problems are corrected, credit is taken for the steam line non-return check valves, and full entrainment (break exit quality) is modoled using the WCAP-8822 data (Reference 9) described in SE 384.

The DYNODE and CONTEMPT analyses use unarproved methods. The resulting 23 l

l 1

i

Exhibit A 1

DYNODE mass and energy releases are then left unchanged and input into CONTEMPT. The results show significant margin, over 17 psi, to the containment pressure design limit. This also provides a measure of the " worth" of the new entrainment methodology, which is about 16 psi of containment pressure.

Case D is a sensitivity study in which the entrainment used in Case C was reduced until the containment design pressure limit was reached. The results show that the entrainment could be reduced by 37% before the containment pressure reached the design limit. WCAP-8822 states that in order to account for uncertainties in calculating the break exit quality all of the data points should be increased by 0.1 in quality. This translates into an average reduction in entrainment of 17% for the WCAP-8822 data.

Therefore, Case D demonstrates confidence in the use of the entrainment data. It shows that the entrainment data would have to have a significantly higher error than the published Westinghouse uncertainty before there is any possibility that the analysis approaches the containment design pressure limit.

Case E was analyzed subsequent to SE 384 and SE 50 402 due to indication that the cooling water flow to the containment FCUs would be reduced in a post-MSLB

- environment. Case E is a sensitivity study on the effect o' the FCU cooling on the peak containment pressure. The case was analyzed using the methodology specified in the l

BAST Project final report (Reference 10). Entrainment data from WCAP-8822 (Reference 9) is included in the model. WCAP-8822 states that in order to account foi uncertainties in calculating the break exit quality all of the data points should be increased by 0.1 in quality. This translates into an average reduction in entrainment of 17% for the WCAP-8822 data. The model also includes the non-return check valves on the main steam linas. Two cases are modeled - one with the FCU cooling capacity from the FSAR and one with no FCU cooling. Removing the FCU cooling increases the peak containment pressure by about 0.5 psi. In the long term, the removal o' the FCU cooling does lead to higher temperatures of the containment atmosphere in the analysis; however, the EQ analysis was reviewed with the results from this case and it was determined that "all equipment that is required to function for this event will not be adversely affected.". (Reference 11) Note that these cases were not run for any specific cycle as suggested by the location of the entry in Table 1; however, the model is conservative for all cycles to date (Pl 1-18). As noted previously, the containment response is very insensitive to the core physic parameters.

Conclusion NSP has identified deficiencies and errors in the assumptions and input to the current licensing basis for the MSLB event which have required a re-evaluation of the event for containment pressure response. The current licensing basis for the MSLB event is the FSAR analysis and the NRC-approved response by NSP to IE Bulletin 80-04. An 24

i Exhibit'A _

. engineering analysis of the_ MSLB event with corrected assumptions was performed, it (was necessary to use unapproved methods to perform this analysis. The methods and ;

l assumptions were demonstrated to be conservative and appropriate for the applicationc The analysis presented in this report provides anl interim solution. through Unit 1 Cycle.

~ 18, until new roothods associated with the BAST project can be submitted to the NRC '

1 for review and approval.-

The results of the study have hown that there has been no reduction in the overall margin'of safety to the containment design pressure limit. The current licensing basis for the MSLB still represents a conservative and bounding analysis for Unit 1 Cycle 18.-

1 L There are no unresolved ^ safety questions.

4

f i

C

-w L25-

A

,y

+

Exhibit A h

Table 2:

I Assumptions Used'in' Cases A-D.

l s

l

. - Double-ended rupture at the exit nozzle of the SG" l

{. - l Hot zero power, end of cycle conditior.s -

. LTech.~ Spec. minimum shut down margin of 2%

. Most reactive RCCA stuck out of the core

'.- LOffsite power is available:

LRCPs at full flew is more conservative.

. : : Main FW on full until isolation at 10 seconds:

'no credit for the FW pump coastdown, conservatively increases the available'SG inventory for blowdown. '

. L SG liquid level at 33% of narrow range:

maximizes the available liquid inventory for blowdown the SG level following a reactor trip goes offscale low and takes over 15 minutes to recover.

. ~One safeguard train out:

c this is the limiting singlejailure.

Si actuates at 6 seconds ~

h l Differences from the FSAR analysis:-

. - : RCS flow at 115% of the tech spec minimum:

increases heat transfer from primary to secondary, and bounds actual RCS flows.

'.' = Aux FW at runout conditions to the broken SG L conservatively.incroases the available SG inventory for blowdown.

Si line volumes corrected to modeI each unit specifically--

- Main FW inlet enthalpy at hot conditions, 405 Btullbm:

c.1-

_ correspunds' to FW temperature of 425 "F, I

. - i Broken loop steam line non-return check valve closes in 5.5 seconds

,4.<: Liquid entrainment out the break modeled using WCAP-8822 ' lata -

5 f

'f 26 x.

4 e

w

+

r.m.s y.w-

,-.v.-

Exhibit A Unreviewed Safetv Question Determinatig.rl 1.

May the proposed activity increase the consequences of an accident previously evaluated in the USAR or in a pending USAR submittal?

No. The relevant accident, Main Steam Line Break, hac bosn re-evaluated. As described in the Results section above, the consequences are less severe than the current licensing basis.

2.-

May the proposed activity increase the probability of occurrence of an accident previously evaluated in the USAR or in a pending USAR submittal?

No. There are et operating procedure changes, or plant equipment modifications, as a result of this re-evaluction of the Main Steam Line Break accident. Therefore, there is no increase in the probability of occurrence of this accident.

3.

May the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated Iri the USAR or in a pending USAR submittal?

No. Thsre are no changes to plant equipment, or plant operating procedures, as a rest.It of this ree-evaluation of the Main Steam Line Break accident. Therefore, the probability of occurrence of a malfunction of equipment has not increased.

4.

May the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the USAR or in a pending USAR submittal?

No. The re-evaluation of the Main Steam Line Break accident does not involve any changes to plant equipmunt, or plant operating procedures. Therefore, the consequences of any malfunction previously evaluated has not increased.

27 f1 tr

Exhibit A 5.

May the proposed activity create the possibility of an accident of a different type than previously evaluated in the USAR or in a pending USAR submittal?

No. The re-evaluation of the Main Steam Line Break accident does not involve any changes to plant equipment, or plant operating procedures. The re-evaluation confirms that the current licensing basis is conservative. Therefore, there is no possibility of creating an accident of a different type.

6.

May the proposed activity create the possibility of different type of malfunction of equipment important to safety than any previously rivaluated in the USAR or in a pending USAR submittol?

No. The re-evaluation of the Main Steam Line Break accident does not involve any changes to plant equipment, or plant operating procedures. Therefore, there are no new failure modes created.

7 Does the proposed activity resduce the margin of safety as defined in the basis for any Technical Specification?

No. The re-evaluation of the Main Steam Line Break accident dernonstrates that the current licensing basis is conservative. Therefore, there is no reduction in the margin of safety.

28

Exhibit A References 1.

Prairie Island Safety Evaluathn (Non-Modification) SE 384, June 1994 2.

Prairie Island Safety Evaluation (Non-Modification) SE 50-402, April 1995 3.

Prairie Island Final Safety Analysis Report, Volume 7, Chapter 14.2.

4.

" Main Steam Line Break Safety Analysis", Northern State': Power Co. Internal report, May 8,1980.

5.

"Pralrle Island Nuclear Power Plant Relcad Safety Evaluation Methods for Application to PI Units", NSPNAD-8102 A.

6,

" Main Steam Line Break Methodology", letter from NRC, May 26,1994.

t 7.

Prairie Island Updated Safety Analysis Report, Volume 5, Chapter 14.5.5.

8.

NSP-NAD Calculation Folders:

Ill.SAS 94.002 IV.SAS.94.001 Ill.SAS.94.004 IV.SAS.94.002 I

Ill.SAS.94.008 IV.SAS.94.004 Ill.SAS.94.014 IV.SAS.95.007 lll.SAS.95.011

}

Ill.SAS.95.012 9.

" Mass and Energy Releases Following a Steam Line Rupture", WCAP-8822, Westinghouse Electric Corporation, September 1976.

10.

" Main Steam Line Break Containment Integrity Analysis and Evaluation of the Removal of the Boric Acid Storage Tanks from the Safety injection System of the NSP Pralrie Island Units", NSPNAD 95004, Rev. O, December 1995.

11.

NSP Internal Correspondence, " Qualification of Equipment for the New MSLB Profiles", dated January 31,1996, from Eric Ballou to Kevin Payette.

Safety Evaluation 433 Fan-Coll Unit Manway Covers Descriotion of Chanae The Containmen. Fan Coll unit enclosure have twenty inch manway covers. The covers are not bolted in place per equipment drawings. The FCU equipment specification states the enclosures should be designed for 2 psi internal pressure. The applicability of the 2 psi differential pressure and the potential coil bypass flow were discussed in this evaluation to leave the covers unbolted.

Summarv of Safetv Evaluation

+

SE# 434 justifies generic use of Leak-before-Break technology for elimination of consideration of dynamic effects _of postulated pipe ruptures in the RCS. The dynamic effects to be excluded include the consideration of vessel cavity or sub-compartment pressurization. Additionally, the coil bypass flow was calculated and manway cover removal does not degrade FCU performance below analyzed limits. The temperature of 29

Exhibit A the air over the exterior of the motor enclosure will be essentially unchanged. Motor operating temperatures will be unchanged. Also, the motors are still protected from direct Containment Spray impingement by the motor enclosures inside the FCU enclosure. The largest vertical seismic accelerations on the FCUs are less than ig.

Therefore the covers will remain in place under seismic accelerations, and not become missiles, it is found the manway covers need not be bolted and may be temporarily removed from the openings to aid maintenance during power operations, without logging the equipment inoperable.

Safety Evaluation 434 Generic Use of Leak Before Break Technology Descriotion of Chanae-This safety evaluation reviewed the acceptability of applying Leak-Before Break (LBB) technology for the purposes of eliminating the consideration of dynamic effects of postulated piping ruptures on the Reactor Coolant System.

Summary of Safety Evaluation The use of LBB technology and the submittal required by GDC-4 for elimination of consideration of the dynamics effects of postulated pipe ruptures has been reviewed by the NRC and determined to not increase the risk or probability of an accident. The design of accident mitigstion systems is unchanged. Leakage from the reactor coolant system cari be detected and the plant shutdown in accordance with the plant Technical Specifications prior to cracks growing to a point that safety is challenged. Use of LBB technology will greatly reduce the analyzed loads on the reactor coolant pump horizontal supports, thus requiring no rework beyond repositioning them to the original alignment. Generic use of LBB on only the reactor coolant system is acceptable within the limitations cited in the NRC Safety Evaluations.

Safety Evaluation 435 Old Anti vibration Bar(AVB) Indications Backaround of Anti-vibration Bars (AVB) Dearadation

-The original anti vibration bars installed in the Prairie Island Westinghouse Model 51 steam generators were a chrome-plated Inconel Alloy 600 square bar. The tubing material in the Prairie Island Westinghouse Model 51 steam generators is inconel Alloy 600. During service conditions, the chrome plate thinned due to fretting wear. When the chromium layer was gone, wear increased due to the higher coefficient of wear for Alloy 600 against Alloy 600. This wear resulted in both the bar and tube becoming thinner. Eddy current testing identified the wear. Tubes were plugged when the wear indication reached 40% through wall degradation. The first plugging was in February 30

~

Exhibit A 1981. Due to increasing numbers of tubes with indications of and being plugged due to wear at the AVB's, the AVD's were replaced in Unit 1 in 1980 and in Unit 2 in 1985. In nearly all cases, the new AVB's were located away from the old AVB locations so that the wear degradation mechanism was stopped at the old AVB locations.

Evaluation Eddy current Indicates that the wear locations of the old AVB's are no longer degrading. Examples of old AVB Indications are shown in Table 1 and Table 2. It is seen from Table i that there is very little change in old AVB indications from 1985 to 1994 for Unit i and Unit 2. In fact the growth from 1985 to 1992 was negative, a decrease in % wall penetration, in 1994 the calibration standard was changed and Table 1 does show the old AVB indications increased in % wall penetration with a change in calibrailon standard. The data for Unit i between 1994 and 1996 using the new calibration standard shows very little change in growth. The growth between 1994 and 1996 was negative, similar to the negative growth from 1985 to 1992. Table 2 i

shows two tubes for 11 Steam Generator and the old AVB indications. The data in table i

2 shows a slight decrease in old AVB indications from 1994 to 1996. Therefore the old AVB indications, as shown in Table 1 and Table 2, are not changing over time.

Table 1 AVG.AVB AVG. CHANGE IN AVG.AVB i

GROWTH AVB 92 to 94 GROWTH FROM l

FROM 85 TO 92 (STANDARD) 94 TO 96 11 S/G

-1.4 5.2

-1.9 12 S/G

-0.8 3.4

-2.1 21 S/G

-1.3 4.1 NA 22 S/G

-2.6 2.9 NA note: growth example 34% (85) to 33%(92)= -1% growth Table 2 ROW COL LOCATION

% IN 9405

% IN 9601 38 58 NV2+2.0 44 40 40 59 NV2+2.0 43 41 40 59 NV2+37.1 48 46 40 59 NV4+3.6 44 39 31 I

l l

Exhibit A in the 1996 outage 11 Steam Generator had two tubes (row 41 col 58 and row 31 col

59) returned to service after deplugging. The two tubes had eddy current testing done the full length and the tube area covered by the old plugs was eddy current tested. The eddy current testing of the fulllength showed no degradation in the tube because of not being in service. The area covered by the old plugs showed no damage by the plug removal process and no damage from not being in service. Table 3 shows no degradation in the tubes for AVB indications when corrected for the change in calibration standard. The two tubes that were returned to service had no known degradation in the tubing because of not being in service. These tubes are no more susceptible to degradation due to being plugged than are tubes that have never been plugged. Table 3 also shows the eddy current repeatability for AVB indications. The data shows that the eddy current indication is repeatable using the same equipment (i.e. probe model).

Table 3 ROW COL LOCATION

% IN

% IN

%+ AVG %

% IN

% IN 8312 8501 CHANGE FOR 9601 A 9601 B STD 41 58 NV2 37 41 46 46 48 31 59 NV1

<20 PLG 25 22 22 31 59 NV2 32 PLG 37 40 40 31 59 NV3 35 PLG 40 46*

47 31 59 NV4 40 PLG 45 43 43

• note: growth occurred from 83 to 86 when AVB's were replaced Technical Specification 4.12 D 1.(f) defines the repair limit for degraded tubes.

Reosir limit means the imperfection depth at or beyond which the tube shall be removed from service by plugging or repaired oy sleeving because it may become unserviceable prior to the next inspection and is equal to 50% of the nominal tube wall thickness. If significant general tube thinning occurs, this criteria will be reduced to 40% wall penetration.

Technical Specification 4.12 D.1.(g) defines the unserviceablo condition.

Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break.

Technical Specification 4.12.D.1.(e) defines defect.

Defect means an imperfection of such severity that it exceeds the repair limit. A tube containing a defect is defective.

32

Exhibit A Tubes with wear scars from anti vibration bars have one or two discrete regions of wear (on one or both sides of the tube). This is not significant general tube thinning.

Therefoie, the repair limit for anti vibration wear scars is 50%. Since there is also no growth, IF.e repair limit for anti vibration wear scars per Technical Specification 4.12 D.1 (f) is 50%,

Primary Pressure Pr=2235 psig Secondary Pressure P =705 psig Differential Pressure dP=1530 psi Average Pressure P., =0.5(Pr+P..e)=1470 psl inner Radius of tube Rs.3875 Minimum tensile strength Su=90 ksi tmio = minimum thickness nominal thickness of tube t =.050 in.

1) From Reg. Guide 1.121, it is required that the margin of safety against tube rupture under normal operation conditions should not be less than 3. The thickness is determined from tmina 3 dP R = 3(1530)( 3875) =.020 in.

Su P.,

90,000-1470

% a'!owable degradation = (,050.020)100 = 60%

.050

2) Also, from Reg. Guide 1.121, it is required that the margin of safety against tube failure under postulated accidents (Service level D) should be consistent with the margin of safety determined by the stress limits specified in ND3225 of Section ill of the ASME B & PV code. Appendix F specifies a limit of 0.7So for level D Service limits P. = pressure differential for accident = 2485 psi tmin=

P. Ri

=

(2485)(.3875)

=.016in.

0.7Su-0.5P.

0.7(90,000)-0.5(2485 )

% allowable degradation = (.050.016)100 = 68%

.050 Eddy current uncertainty is 5% for AVB indications, therefore the Technical Specification 4.12 D.1.(f) covers both the eddy current uncertainty and the rninimum thickness per Reg. Guide 1.121, Section 14 of the USAR was reviewed for impact. The function of the steam generator tubing is to 1) maintain the primary system pressure boundary and to 2) transfer heat from the reactor coolant system to the secondary side. This safety evaluation does not 33

Exhibit A j

change the failure modes or failure impact of the Steam Generator tubing. No impact 3

on the USAR Section 14 accidents was identified.

This Safety Evaluation supports leaving old Anti-vibration Bar (AVB) indications in service up to the Technical Specification 4.12 D.1.(f). limit of 50% wall penetration.

1 Future Inspect!ons Technical Specification 4.12 B.2.(a) requires all tubes that previously had wall penetrations >20% that have not been plugged or sleeve repaired in the affected area to be included in the first sample set of tubes selected for each in service inspection of each steam generator. Tubes with old AVB Indications exceeding 50% wall penetration require plugging.

Determination of Unreviewed Safety Question

1. Does the proposed activity increase the consequences of an accident previously evaluated in the USAR or in a pending USAR submittal?

Those accidents which could be affected by this change are:

RupNre of a Steam Pipe or Feedwater Line Break (USAR Section 14.5.5).

Steam Generator Tube Rupture (USAR Sedion 14.5.4).

Large Break Loss of Coolant Accident (USAR Section 14.6)

No. The repair limit per Technical Specification 4.12 D.1.(f) of 50% has been analyzed for consequences of an accident. The old Anti-vibration Bar (AVB) indications will use the same repair limit of 50% wall penetration. Therefore, this activity will not increase the consequences of an accident previously evaluated in the USAR.

2. May the proposed activity increase the probability of occurrence of an accident previously evaluated in the USAR or in a pending USAR submittal?

No. The old Anti-vibration Bar (AVB) indications are not growing. The old AVB's have been removed and this has removed the source of wear on the tubes. The two tubes returned to service had no kna o degradation because of not being in service. The % allowable degradation per Reg. Guide 1.121 is 60%. The Technical Specification 4,12 D.1.(f) of 50 % including the eddy current uncertainty of 5%

conforms to US NRC Reg. Guide 1.121.. Therefore, this activity will not increase the probability of occurrence of an accident previously evaluated in the USAR.

34

Exhibit A 3.

May the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the USAR or in a pending USAR submittal?

No. The removal of the old Anti-vibration Bars (AVB's) has removed the source of wear on the tubes, in that area. The old AVB indications have shown no growth from outage to outage. Therefore, this activity does not increase the probability of occurrence nf a malfunction of equipment important to safety previously evaluated in the USAR.

4. May the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the USAR or in a pending USAR submittal?

No. If any of the steam generator old Anti-vibration Bar (AVB) indications do fail, the malfunction of the tube has been previously evaluated. Accident leakage is bounded by the current analysis for steam generator tube rupture and main steam pipe rupture. Since the repair limit of 50% for old AVB indications is the same as the Technical Specification 4.12 D.1.(f) this will not result in additional steam generator tube leakage, there will be no affect on the radiological consequences of such failures. Therefore, this activity will not increase the consequences of a malfunction of equipment important to safety previously evaluated in the USAR.

5. May the proposed activity create the possibility of an accident of a different type than previously evaluated in the USAR or in a pending USAR submit;al?

No. The repair limit of 50% for old Anti-vibration Bar (AVB) indications is the same as the Technical Specification 4.12 D.1 (f). Therefore, this activity will not create the possibility of an accident of a different type than previously evaluated in the USAR.

G. May the proposed activity create the possibility of a different type of malfunction of equipment important to safety than any previously evaluated in the USAR or in a pending USAR submittal?

No. The failure mode for a steam generator tube leakage or rupture is bounded by current analysos. Therefore, this activity will not create the possibility of a different type of malfunction of equipment important to cafety than previously evaluated in the USAR.

7. Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specifications?

35

)

Exhibit A No. The repair limit of 50% for old Anti-vibration Bar (AVB) indications is the same as the Technical Specification 4.12 D.1.(f). The old AVB indications are not significant gmeral tube thinning and the old AVB indications are not growing.

Therefore, the repair limit of 50% for old AVB indications does not reduce the margin of safety as defined in the basis for any Technical Specifications.

Since all of the above answers are No, there is no Unreviewed Safety Question.

j Safety Evaluation 436 Eddy Current Indications in Welded Tubesheet Sleeve Upper Wold Region.

Descriotion of Chanoe This Safety Evaluation established the plugging criteria for sleeves with ET indications in the upper sleeve weld region. During the 9601 in service and baseline eddy current examinations of the sleeves in 12 Steam Generator, ET indications were identifled at the upper sleeve weld region. The eddy current indications found in the sleeve welds were either volumetric or circumferential in nature. The circumferential indications appeared crack like in nature. There was controversy as to whether or not they were due to the welding process (e.g.,- weld geometry). Because of the inability to size circumferential indications, especially in the weld joint region, all circumferential Indications shall be plugged.

Summary of Safety Evaluation it is possible to make some assessments on the extent of degradation or welding artifacts present in the sleeve weld for volume. tic indications. Laboratory specimens exhibiting volumetric indications have been metallurgically examined. The source of

- these indications has been found to be small cavities at either edge of the sleeve weld.

When these cavities are at the upper edge of the weld, there is no effect on the pressure boundary integrity of the sleeve weld.

Sleeves left in service with volumetric indications were required to be acceptable by all other examinations done including visual examination, installation UT, bobbin coil ET, physical location of the indication at the upper weld edge by rotating coil ET, and less than 0.012 inches wall loss (< 30%) by UT-360 measurements. In addition, all welds left in service with volumetric indications had an everage weld height of greater than 0 080 inches which meets the structural analysis (by a factor of 4) for the CE welded tubesheet sleeve pull-out forces.- All sleeves with volumetric indications will be inspected at each 12 steam generator tubing inspection.

36

Exhibit A Section 14 of the USAR was reviewed for impact. The function of the steam generator tubing is to 1) maintain the primary system pressure boundary and to 2) transfer heat from the reactor coolant system to the secondary side. These indications do not change the failure modes or failure impact of the Steam Generator tubing. No impact on the USAR Section 14 accidents was identified.

Safety Evaluation 441 Reactor Coolant System Leak Basis Description of Chanqq The purpose of this review is to evaluate the USAR Section 14 discussions regarding the capability of the charging pumps in making up for leaks from the reactor coolant system. This evaluation resolves an apparent discrepancy between loss of coolant through a 0.375" (3/8") diameter hole and the makeup capability of a charging pump in Section 14.7.2 and recommend USAR revisions where appropriate.

Summary of Safety Evaluation l

Based on critical flow calculations, it was determined that the charging system can make up for a 3/8" diameter hole and maintain Reactor Coolant System pressure. For hole sizes larger than 3/8" diameter the RCS would depressurize and result in an automatic reactor trip and safety injection. This is consistent with WCAP 9600 and the Basis for the Emergency Operating Procedures. There is no unroviewed safety question or reduction in margin in the bases for any Technical Specifications.

Safety Evaluation 442 - Containment Spray Solution pH Evaluation Description of Chance USAR Section 14.9.6.3 indicates that the gravity feed arrangement to the containment spray pumps will maintain a pH of 8.6 of the spray solution with refueling water storage tank boron concentration of 3500 ppm. The calculation which formed the basis for this pH value used nominal pump capacities as the design input. Subsequent calculations have been performed, varying pump combinations at maximum and degraded flow capacities at minimum and maximum caustic and acid concentrations in the condensato storage tanks and refueling water storage tank, respectively. The results of these determinations indicate that specific pump / flow combinations may result in spray solution pH below 8.0 during specific phases of the injection phase of post LOCA mitigation. The purpose of this evaluation is to provide justification to revise the USAR discussion pertaining to spray solution pH.

37

Exhibit A Summary of Safetv Evaluation The spray solution pH of 8.6 was based on nominal pump flows and provided the basis for the current revision to the USAR. A calculation was performed for all possible ECCS pump combinations, maximum and degraded pump flows, and maximum and minimum caustic and boric acid concentrations in the condensate stnrage tankr. and refueling water storage tank. The results from this analysis show that, at specific pump operating combinations at distinct flow rates and considering the worst case combination of refueling water storage tank boric acid concentration vs. caustic standpipe sodium hydroxide concentration, it is possible for the spray solution pH to be below 8.6. Note that this is not a change in how the system is configured or operated, but only a change resulting from increased detall in the analysis.

The resultant minimum and maximum spray and sump water pH were evaluated and it was shown that there is no effect on off site or control room habitability doses, and no effect on equipment qualification, hydrogen production from containment coatings, etc.

Based on this review, there is no unreviewed safety question or reduction in margin in the bases for any Technical Specifications.

Safoty Evaluation 447 Flooding in the Auxillary Building Description of Chanae Non-Class.1 piping systems having access to large water volumes and/or potentially large flow rates were considered for flooding effects within the Class l portions of the auxillary building. The roedwater System and the Fire Protection System were chosen for evaluation.

Summarv of Safety Evaluation Equipment required to mitigate the flooding events considered are not rendered inoperable. Therefore, for these events, there are no increased consequences than previously recorded in the USAR, and there are no unreviewed safety questions.

t Safety Evaluation 449 TN-40 Cask Weight Description of Chance Table 12.2-40 provides a table of know loads that are handled over safety related components. For the Auxiliary Building Crrane, the TN-40 cask weight is listed as 240,690. The casks fully loaded and assembled at present have been slightly heavier than that.

38

Exhibit A Summary of Safety Evaluation The Auxillary Building Crane is rated at 250,000 lbs and is designed as a single failure proof system. When used with the cask and associated handling equipment, the crane 1

meets all the design margins of the applicable codes and standards. Table 12.2-40 is i

revised to reflect a maximum load of 250,000 lbs as a bounding case when handling a fully loaded cask.

Safety Evaluation 453, Rev.1 Lack of Fusion Indications In Steam Generator Sleeve Wolds found by Reevaluation of Ultrasonic Examination Data Descriotion of Chanae As a result of the metallurgical examination of five sleeved tubes removed from 12 Steam Generator, new analysis guidelines were developed for evaluating ultrasonic examination data of sleeve upper welds. During reevaluation of ultrasonic data for the weld height of 36 sleeves (34 of which are in service) with eddy current volumetric indications, 4 sleeves (3 of which are in service) were ident fied with possible lack of fusion (primary to secondary leak path) using the new analysis guidelines. The lack of fusion was 25 degrees or less. The lack of fusion is contrary to the Sleeve License Amendment NRC Safety Evaluation Report which describes the Combustion 1

Engineering sleeve as a leak-tight sleeve.

As a result of this discoveiy, the UT data for 237 additional sleeves (which represented the remainder of the sleeves with recorded digitized ultrasonic test data from the 9601 sleeving campaign) were reevaluated. There were no additional sleeves with lack of fusion indications.- However, it was determined that the UT data for three of the 237 t

sleeves was taken at the expansion transition zone instead of the weld, The inadequate ultrasonic examination is also contrary to the Sleeve License Amendment NRC Safety Evaluation Report which describes the Combustion Engineering sleeve ultrasonic examination requirement.

Summarv of Safety Evaluation This Safety Evaluation considered the leak-tightness and structural integrity requirements regarding the existing sleeves that are in service in 12 steam generator with indications of lack of fusion and inadequate ultrasonic examination.

The maximum potential leakage was evaluated as 0.016 gpm per sleeve. The potential leakage ~ calculated from the sleeves in service with lack of fusion (3), inadequate ultrasonic examination (3) and volumetric indications (34, corrected for probability of 39

Exhibit A detection) is 0.63 GPM at Main Steam Line Break conditions. This potential leakage is within the Main Steam Line Break analysis of 5 GPM.

The three sleeves uth inadequate ultrasonic examination data are considered acceptable for use until the end of the current fuel cycle because they had acceptable visual examination, acceptable + Point eddy current examination, and because for the 234 sleeves (237 minus the 3 sleeves with inadequato UT examination) which had no

+ Point eddy current indications, there were no sleeves with lack of fusion indications.

Section 14 of the USAR was reviewed for impact. The function of the steam generator tubing is to 1) maintain the primary system pressure boundary and to 2) transfer neat from the reactor coolant system to the secondary side. These indications do not change the failure modes or failure impact of the Steam Generator tubing. No impact on the USAR Section 14 accidents was identified.

The sleeves with inadequate ultrasonic examination can be left in place during Unit 1 Cycle 18 operation.

Safety Evaluation 460 - Movement of Loads over the Spent Fuel Pool Enclosure Roof Descriotion of Chance USAR Section 12.2.12.1,3, ' Auxiliary Building Crane Evaluation 7 contained a passage that states that a load drop on the roof of the Spent Fuel Pool Enclosure is not possible because of physical restriction associated with the height of the crane hook above the enclosure roof. This statement is not true for the auxiliary hook of the crane, and required a change to clarify that the auxiliary hook can move small loads oves the roof.

Summarv of Safety Evaluation The Safety Evaluation discussed an analysis that was done that showed that the impact energy of a 3.0 ton or less load drop from 6 feet (physical restriction associated with crane and structure) is bounded by the enclosure missile analysis, USAR Section 12.2.1,4.3.1.4, and thus is safe and not an unreviewed safety question.

Safety Evaluation 468 - Use of Methoxypropylamine as a pH Control Chemical in the Secondary System Description of Chanqe 40

Exhibit A The change involves use of methoxypropylamine (MPA) for secondary side pH control.

MPA replaces morpholine and ethanolamine as pH control chemicals. Improved retention of sodium on the steam generator reclaim ion exchangers will result in longer service life of the resin beds.

Summary of Safetv Evaluation MPA has been tested in the laboratory and in plants. It provides similar or improved corrosion protection compared to morpholine or ethanolamine. Use of MPA will not affect structural integrity of steam generator shells, tubes or support plates, nor will it affect feee ;ter or main steam line components. Use of MPA will result in lower overall sodium levels in the steam generators while maintaining similar or improved corrosion product transport rates.

Safety Evaluation 473 - Effects of ABSVS Fan Replacement during Original Construction Description of Chanae The safety evaluation justifies changes to the USAR due to changes in the Auxiliary Building Special Ventilation System (ABSVS) configuration during original construction.

ABSVS fans were replaced to improve building drawdown capability. The safety evaluation also serves to clarify the design basis for the ABSVS filter units.

Summarv of Safetv Evaluation The safety evaluation concluded that the system as presently configured satisfies the original and current design bases and that no unreviewed safety question exista.

Safety Evaluation 474 Main Feedwater Line Break Qgseriotion of Chance The safety evaluation supports changing information in USAR Section 11.9 which indicates that 400 gpm auxiliary feedwater flow is needed to mitigate the main feedwater line break accident. Based on recovered correspondence and further analysis, a value of 200 gpm is required for a main feedwater line break inside containment.

Sumrngrv of Safety Evaluation 41 l

Exhibit A Current USAR information regarding auxiliary feedwater flow requirements to mitigate the main feedwater line break accident is inconsistent with the design basis, it assumes a main feedwater line break outside containment with failure of the check valve to function; this accident is outside the design basis of the plant, and should not be part of the USAR.

Safety Evaluation 475 Spent Fuel Pool Enclosure Personnel Access Doors Descriotion of Chanae The evaluation made a new calculation of thyroid dose at the site boundary as a result of a fuel handling accident in the spent fuel pool. The calculation includes a specific allowance for the Spent Fuel Pool Special Ventilation System to be unable to draw a negative prescure for up to one minute during the accident. This allowance bounds the time that the personnel access doore would be expected to be open during the evacuation following a Fuel Handling Accident.

Summarv of Safety Evaluation The new calculation predicts a dose of 23.6 Rem; as opposed to the present USAR calculation of 22.5 Rem. The new dose is still well within 10CFR100 guidelines, Safety Evaluation 47814 01 USAR Section 14, Safety Analysis Qgiqriotion of Chance Safety Evaluation 478-14-01 evaluates corrections and updates to information in Section 14 of the USAR (Safety Analysis). The majority of these changes are clarifications or editorial in nature. The significant part of this change was the incorporation of an update to the large break loss of coolant accident (LBLOCA) analysis discussion, tables and figures.

,!n December 1996, a new LBLOCA analysis was completed by Westinghouse.- This analysis was performed using NRC-approved methodology and continued to show -

compliance with 10 CFR 50.46.

Summary of Safety Evaluation As discussed above, most of these changes are clarifications or editorial in nature.

The significant change is the incorporation of the update to the LBLOCA analysis, completed in December 1996. This analysis was performed using NRC approved

- methodology and continued to show compliance with 10 CFR 50.46.10 CFR 50.46, Section (a)(3) requires calculating the effect of errors in the model used to calculate the PCT to determine if they are significant. Per 10 CFR 50.46, changes or errors to an 42

Exhibit A acceptable LOCA evaluation model which affect the PCT calculation are reported within 30 days if the change is "significant," or at least annually, Chcnges resulting from this update have previously been reported to the NRC in accordance with 10 CFR 50.46.

License Amendments 13 and 7 SAC Review Responsibilities i

Descriotion of Chanae The amendments consisted of changes to the Technical Specifications, in part, to clarify the organization and role of the Safety Audit Committee. In particular, the amendments removed the requirement for the Sefety Audit Committee to review emergency procedures. USAR wording was changed to align with the revised Technical Specification wording.

l Summary of Safety Evaluation The License Amendments were issued May 10,1976.

License Amendments 22 and 16 Spent Fuel Pool Rorack Descriotion of Chanae The amendments consist of changes in the Technical Specifications that permitted modification of the spent fuel storage pool which increased the storage capacity from 198 fuel elements to 687 Part of that license amendment revised the wording in Technical Specification 5.68, Spent Fuel Storage, to read, "The bottoms of the slote-are above the tops of the active fuel in the fuel assemblies which will be stored vertically in specially constructed racks." This is a correct description of the racks as

- they were changed and as they are today USAR wording was changed to align with the revised Technical Specification wording.

Summarv of Safety Evaluation The License Amendments were issued on August 16,1977, License Amendments 73 and 66 - Refueling Boron Concentration Requirements Description of Chance The amendments revised the Technical Specification on boron concentration requirements to agree with the Standard Technical Specifications during refueling 43

Exhibit A operation. USAR wording was changed to align with the revised Technical Speci'ication wording.

Summary of safety Evaluation The License Amendments were issued June 25,1985.

License Amendments 91 and 84 Technical Specifications Upgrade Description of Chance The amendments contained, in part, a change in a reference in the Bases for Teebical Specification 3.2 to USAR Section 14.5.5 concerning boric acid storage tank r

requirements for the main steam line break accident. For clarification, requiremants l

given in the Bases for Technical Specification 3.2 were added to USAR Section 14.5.5.

I Summarv of Safety Eval al!9D 4

The License Amendments were issued October 27,1989.

License Amendments 108 and 100 Increased Fuel Enrichment Limit Changes Descriotion of Chance The amendments revised the Technical Specifications to increase fuel enrichment from 4.25 weight percent to 5.0 weight percent. This included a revision to the Technical Specifications to allow 5.0 weight percent U 235 fuel to be stored in the new fuel vault and the spent fuel pool and used in the core. In addition, the amendments revised the Technical Specifications to increase the minimum refueling water storage tank boron concentration and incorporate references to natural uranium and ZlRLO clad material into the reactor core design description.

Summarv of Safety Evaluation The License Amendments were issued September 3,1993.

License Amendments 122 and 115 Radiological Effluent Technical Specifications Description of Charice 44

Exhibit A The amendments revised the Radiological Effluent Technical Specifications and other sections relating to radiological controls to conform to NUREG 1431,

• Standard Technical Specifications, Westinghouse Plants,' Revision 1, and Generic Letter 89-01,

• lmplementation of Programmt. tic Controls for Radiological Effluent Technical Specifications in the Administrative Controls Section of the Technical Specifications and the relocation of the Procedural Details of RETS to the Offsite Dose Calculation Manual or to the Process Control Programs.

Summary of Safety Evaluation i

The License Arnendments were issued January 24,1996.

i CHANGES TO REGULATORY COMMITMENTS Regulatory Commitment Changes 96 02 and 96<03 The Prairie Island response to Generic Letter 87-06 listed those valves which would im treated as pressure isolation valves, Further review of that list resulted in the finding that the only valves meeting the criteria for pressure isolation valves are the low head injection check valves and the RHR to RCS loop B cold leg check valve. These are the valves listed in Technical Specification 4.3. The balance of the valves listed in the response to GL 87-06 are not pressure isolation valves, and will not be treated as such.

Regulatory Commitment Change 97-02 As part of the response to Generic Letter 96-04, Pralrie Island committed to' revision of procedures "...by temporary memo...." This commitment was changed to allow the procedure revisions to be made gither by the temporary procedure change process or the permanent procedure change process.

Regulatory Commitment Change 97-03 As part of the response to Generic Letter 96-04, Prairie Island committed to maintaining

• irradiated assemblies...in tha cells adjacent to the Boraflex coupons to insure that they have received at least as much dose as any full length Boraflex panel." This commitment is no lor.ger necessary because the spent fuel pool criticality analysis no longer credits Boraflex for criticality control. The coupons will no longer be used to verity the integrity of the Boraflex in the racks.

l o

Exhibit A Regulatory Commitment Change 97 04 As part nf ths response tc Generic Letter 96-04, Prairie Island committed to testing

" BNafiex coupans...for variotis physical properties on a regular basis to insure the Botaflex is performing as dosic-rzed.' This commitment is no longer necessary because the spent fuel porsi criticality analysis no longer credits Boraflex for criticality control.

The coupons will on lonsor 50 used to verify the Integrity of the Boraflex in the racks.

Regulatory Commitment Ciango 97 05 As part of the responss to Generic Letter 96-04, Prairie Island committed to "...a fuel management strategy...to insure that fresh and other high reactivity assemblies...will only be placed la spent fuel pool cells that are projected to have little or no boron carbide loss...." This commitment has been deleted because the spent fuel pool criticality Jnal;s s no longer credits Boraflex for criticality control. Since Dnraflex is no l

longer crec!ited, the amount of boron carbide is irrelevant.

46

Exhibit B a

PRAIRIE ISLAND NUCLEAR GENERATING PLANT Revision 14 to the Updated Safety Analysis Report Instructions:

1.

Remove and discard individual USAR pages, tables, and figures and replace with the new Revision 14 pages provided in this exhibit.-- Special Instructions, where i

applicable, are included with the replacement pages.

2.

When page removal / replacement is complete, review the USAR List of Effective Pages to ensure your copy of the USAR is current and complete. Contact NSP Nuclear Licensing at 612-3881121, Extension 4662 If you require additional assistance.

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