ML20151P167

Annual Rept of Changes,Tests & Experiments for 1987

ML20151P167
Person / Time
Site:	Prairie Island
Issue date:	07/07/1988
From:	NORTHERN STATES POWER CO.
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ML20151P151	List: ML20151P148 ML20151P167
References
NUDOCS 8808090238
Download: ML20151P167 (7)

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f-1 EXHIB1T A PRAIRIE ISLAND NUCLEAR GENERATING PIANT ANNUAL REPORT OF CHANGES, TERTS AND EXPERIMENTS January 1, 1987 to December 31, 1987 The following sections include a brief description and a summary of the safety evaluation for those changes, tests and experiments which were carried out without prior NRC approval,' pursuant to the requirements of 10CFR 50.59(b). I

1. REMOVAL OF CVCS MONITOR TANK FLOATING DIAPHRAGM Descriotion of Chance Safety Evaluation (SE) No. 192 reviewed the removal of the floating diaphragm from the CVCS monitor tank.

Summary of Safety Evaluation The CVCS monitor tanks are used only to hold water prior to discharge.

Original design allowed for maintaining this water deoxygenated using a floating diaphragm. Because the water is only to be held for release, deoxygenation is not required. Therefore the diaphragm is not required, and was removed.

2. INSTALL NEW STEAM GENERATOR WIDE RANGE LEVEL CHANNELS (80Y143)

Descriotion of Chance Installed new safety grade steam generator wide range level channel post accident monitoring instrumentation. This change resulted from )

the fire hazards review and Three Mile Island Lessons Learned.

This design change addressed the installation of the transmitter, l manifold, all associated tubing in containment and the wiring from the ,

transmitter to the containment penetration cabinet, l

Summary of Safety Evaluation l l

l The new tubing taps into the existing tubing. It was installed in accordance with the same requirements as the existing tubing. A break in the new tubing will result in the same effective hole size as that already analyzed for. The channel is required to operate in a i post accident environment for at least 30 days. The entire l installation is designed to meet the required seismic and environmental l qualification requirements. In the. case of a single active failure only the affected channel will be lost. In each unit the individual l l

channels are powered from different trains, routed separately for both cable and wiring, and sense different steam generators. Failure of the new instrument tubing may also fail the existing wide range instrument 8808090238 880707 PDR A-1 K ADOCK 05000282 PDC

for the affected steam generator. This failure would not effect the other two existing instrument channels. Approved installation and I testing procedures were utilized. The new channels have been added to j the surveillance procedures to allo.i for early detection of any '

problems.

The conclusion of the Safety Evaluation states that there are no l

conflicts with the Technical Specifications, unreviewed scfety ques- j tions, nor effects on the probability of occurrence and consequences of )

an accident or malfunction of equipment previously analyzed in the US.'R l or amendments. l

3. UPGRADE EXISTING WIDE RANGE RCS TEMPERATURE CHANNELS (80Y144)

Descriotion of Change i

Upgraded the existing wide range RCS temperature channels to post accident monitoring instrumentation requirements. This change resulted l from the fire hazards review and Three Mile Island Lessons Learned.  !

This design change addressed the rerouting of the cables from the existing sensors to safety grade penetrations via a safety grade raceway. This dasign change also addressed the interface module used to condition the signal from the new equipment so that existing l

instruments will read accurately without further modification.

Summary of Safety Evaluation The consequences of a single failure, such as the loss of an RTD or a single instrument, would cause the loss ot' that instrument channel.

This failure could be detected by comparison with other plant instrumentation, such as core thermocouples. Approved installation and testing procedures were utilized. The new channels have been added to the surveillance procedures to allow for early detection of any problems. The modifications add two more instruments to the loop. The loop error following the modification increased by only 0.367% or  ;

2.2*F. Tbis additional error is not significant when compared to the j scale of the recorder. The existing RTD's are of the type used by the l protection system for inputs to analog protection. They are '

environmentally qualified as safety related Class 1E process instrumentation. New wire is qualified for use in QA I installations for the environment in which it is installed. The new E/I replacing the R/I in the existing loop is of the same style of equipment used in the protection racks and was procured for that use. I The conclusion of the Safety Evaluation states that there are no conflicts with the Technical Specifications, unreviewed safety ques-tions, nor effects on the probability of occurrence and consequences of l an accident or malfunction of equipment previously analyzed in the USAR ]

or amendments. i A2

4. STEAM CENERATOR BLOWDOWN '82Y315)

Descriotion of Chance Steam Generator Blowdown (SGB) flow rate from the Steam Generators was increased from a design rate of approximately 40 gpm to 100 gpm. In order to accomplish this increased flow rate, the inlet valves to the Flash Tank were replaced with valves of higher capacity and of improved suitability for flashing service.

Enmmary of Safety Evaluation Closure of the SGB control valves upon receipt of an Auxiliary Feedwater Pump start signal or High Radiation signal are verified monthly. The SGB motor operated Containment Isolation valves will close in an amergency. The Emergency Operating Procedures require verification of SGB isolation when required. The rupture of the blowdown line has been analyzed as a small steam line break. The results of this analysis conclude that a valve failure would be less severe than a line rupture. In addition, a recent PRA indicates that the quantitative impact of the SGB failing to isolate is expected to be insignificant because several independent failures of blowdown components are required to impact Auxiliary Feedwater failure prebability.

The conclusion of the Safety Evaluation states that there are no conflicts with the Technical Specifications, unreviewed safety ques-tions, nor effects on the probability of occurrence and consequences of an accident or malfunction of equipment previously analyzed in the USAR or amendments.

5. DC CAPACITY AND DISTRIBUTION ADDITION (82N505)

Description of Change Parts I & II Parts I & II of this Design Change provided for installation of double-throw DC transfer switches in selected Unit 1 & 2 DC c'icuits so that certain Safeguard DC loads necessary for both units can be supplied by the same battery train from either unit. This al)aws a Safeguard Battery to be removed from service for maintenante and/or testing during a respective unit outage.

Part III Part III of this Design Change provided for installation of new non-safeguard Station tatteries 31 & 42 and the associated main fuses, 125 VDC distribution panels, battery chargers, and connection equipment for a spare portable charger. It also provided for removal of certain non-safeguard loads from the Safeguard Batteries to these new non safeguard ,

batteries. 1 A3

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l' Summary of Safety EvaluatioD Parts I & II  ;

i The two safety evaluations prepared for Parts I & II (corresponding to i Units 2 & 1 respectively) addressed USAR requirements, operation i (switching) considerations, equipment qualification, required codes and j standards and power considerations. The evaluations also identified any special procedural requirements necessary for switching.

The conclusion of the Safety Evaluations states that there are no conflicts with the Technical Specifications, unreviewed safety ques-tions, nor effects on the probability of occurrence and consequences of an accident or malfunction of equipeent previously analyzed in the USAR or amendments.

Another non-design Safety Evaluation No. 116 was written under Part II to address the operability of equipment during transfer of DC power supply. It war concluded that transfer operation can be performed without affecting the operability of supplied loads.

Part III A safety evaluation was prepared to address removal of non-safeguard loads from the Safegucid Entteries as part of the original Part III packago. However this was superseded by another safety evaluation under Addendum No. 2 because of various scope changes during the project and the effect of other concurrent projects affecting the Safeguard Batteries.

Addendum No. 1 This addendum included a safety evaluation which addresses routing the feeder for 21 Emergency Turbine Oil Pump through Safeguard Busrooms 16

& 25. Although this circuit is non safety related, it was supported off of the north walls of those busrooms which are seismic category I filled masonry walls previously analyzed by Flcur in 1980. Support details were provided for use in construction. This safety evaluation indicates no degradation of the masonry walls due to this new load.

Addendum No. 2 )

l This addendum was prepared to provide a revised safety evaluation for 1 the Part III work as mention above. It addressed battery capacities and operating times for all Safeguard Batteries and 31 & 42 Batteries  ;

in the event of a loss of AC battery charging power (loss of all AC l scenario). This analysis was based on the following criteria: I 1

1) Actual non-safeguard loads removed from the Safeguard )

Batteries, l l

2) Calculations based on present Safeguard Battery specifications )

after replacement and actual installed 31 & 42 Battery specifications, i

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3) More accurate DC load assumptions including some actual test measurements, and

4) Included loads for new RVLIS Inverters.

This evaluation concluded that there is greater than one hour of capacity in the Safeguard Batteries under the loss of AC scenario.

The conclusion of the Safety Evaluations states that there are no conflicts with the Technical Specifications, unroviewed safety ques-tions, nor effects on the probability of occurrence and consequences of an accident or malfunction of equipment previously analyzed in the USAR or amendments.

6. AUXILIARY FEEDWATER PUMP NOS. 11 & 12 TURBINE STEAM SUPPLY VALVE RELOCATIONS (84L838)

Description of Change This modification provided for the relocation of the steam supply valves for the Auxiliary Feedwater Pump 11 and 12 turbines from the Auxiliary Feedwater Pump Room in the Turbine Building Class 1 corridor to a location outside the Class 1 corridor.

Summary of Safety Evaluation Relocation of the Auxiliary Feedwater Pump Steam Supply valves from the Auxiliary Feedwater Pump Rooms was necessitated by the requirement of the Prairie Island Environmental Qualification program that the Auxiliary Feedwater Pump Rooms be maintained as mild environments.

The following safety concerns associated with this modification were considered as part of this evaluation:

1) Environmental Qualiftcation of safety related electrical equipment,

2) Seismic design and protection of safety related piping and components,

3) Postulation of and effects of high energy line breaks,

4) Fire protection of safe shutdown systems and components (10 CFR Part 50, Appendix R),

5) Fire barrier penetration seal design, >

6) Turbine drain system, and

7) Security.

The conclusion of the Safety Evaluations states that there are no conflicts with the Technical Specifications, unreviewed safety ques-tions, nor effects on the probability of occurrence and consequences of an accident or malfunction of equipment previously analyzed in the USAR or amendments, A5 I

7. UNIT 2 CYCLE 12 RELOAD (87L018)

Description of Chanc.e Prairie Island Unit 2 Cycle 12 began operation in February 1988 and is i expected to shutdown in March 1989. Cycle 12 is projected to reach and  ;

and of hot full power exposure of 13,710 MWD /MTU and to shutdown with l an exposure of 14,460 MWD /MTU. This will result in a coastdown of 22 days to 82% of full power at shutdown.

During the Unit 2 Cycle 11/12 refueling outage, 44 Exxon TOPROD assemblies were replaced. The reload consisted of 44 new Westinghouse ,

OFA assemblies, 40 once burned Westinghouse OFA assemblies, and 37 twice burned Exxon TOPROD assemblies. The core uses gadolinium as a burnable poison to control the temperature coefficient and power peaking. The fuel also has approximately 6 inches of natural uranium at the top and bottom of the fuel to enhance uranium utilization.

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Summary of Safety Evaluation I

l The analyses performed in the design and licensing of Unit 2 Cycle 12 l operation were done by NSP's Nuclear Analysis Department (NAD) and are l summarized in the "Prairie Island Unit 2 Cycle 12 Final Reload Design Report (Reload Safety Evaluation)", September 1987. The analyses  ;

indicates that the core can be operated within Technical Specification  !

and USAR limits. )

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8. FILTERED BACKWASH WASTE WATER PIPING (87L969)

Description of Change l This modification installed piping which allows filtered backwash waste l water to be directed to the Backwash Water Storage Tanks (BWST) or to the Unit 2 Turbine Building Sump (TBS).

i l Summary of Safety Evaluation I The original design directed all backwash waste water to either the #11 or #21 BWST where it was used in various operations throughout the i Filter /Pemin System. However, it was determined that resin fines could i l be introduced into the BWST by the backwash waste water and contribute to plugging of the optimized septume now being used in the Filter /Demin System, i Piping was added to the #11 BWST inlet piping that allows the backwash l waste water to be directed to the Unit 2 TBS. This alternate flow path serves to improve the Filter /Demin System performance by allowing the #11 BWST to remain as a source of clean water for use on Unit 1 l

Condensate Polishers during precoating operations, and thereby preventing resin fines from being introduced into the system. The TBS l was reviewed and determined to be satisfactory for storage of the l bachwash waste water.

l All piping was installed and tested under QA 3 specifications and had no adverse effects on the existing plant design, i

1 j A-6 l

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9. CHANGE TO OPERATIONAL OUALITY ASSURANCE PLAN APPENDIX C Revision 12 to the NSP Operational Quality Assurance Plan was internally reviewed and approved on May-26, 1988. We have concluded that this revision does not reduce the commitments of NSP's Operational Quality Assurance Program and does not adversely impact the safe operation of the nucimar plants. Specific changes with the reason for the change and the basis for concluding no reduction in commitments

[per 10 CFR 50.54(a)(3)] aro presented in Appendix D to the plan. The Operational Quality Assurance Plan, Revision 12, is Appendix C of the USAR.

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