ML20245G263
| ML20245G263 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 06/30/1989 |
| From: | Parker T NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20245G268 | List: |
| References | |
| NUDOCS 8908150321 | |
| Download: ML20245G263 (8) | |
Text
. - _
1 Northem States Power Company 414 Nicollet Mall Minneapolis, Minnesota 55401-1927 Telephone (612) 330-5500 June 30, 1989 10 CFR 50.71(e)
U S Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 PRAIRIE ISLAND NUCLEAR CENERATING PINTT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Submittal of Revision No. 7 to the Updated Safety Analysis Report (USAR)
Pursuant to 10 CFR 50.71(c), we are submitting 14 copies of Revision No. 7 to the Updated Safety Analysis Report (USAR) for the Prairie Island Generating Plant. This revision updates the information in the USAR for the period from January 1, 1988 through December 31, 1988.
Exhibit A contains a description and a summary of the safety evaluations for changes, tests and experiments made under the provisions of 10 CFR 50.59 during this period.
Exhibit B contains the USAR page changes and instructions for entering the pages.
Inc~.uded in Exhibit B is Revision 13 to the Northern States Power Compan'j Operational Quality Assurance Plan in compliance with 10 CFR 50.54(a).
Changes included in Revision 13 to the plan are described in exhibit A (Item 10, page A-7) of this letter.
19fh Thomas M Parker Manager - Nuclear Support Services TMP/RJM/rj m c:
Regional Administrator - III, NRC Director IE, NRC (w/o Exhibit B)
NRR Project Manager, NRC (w/o Exhibit B) ik Resident Inspector, NRC (w/o Exhibit B)
G Charnoff (w/o Exhibit B)
Attachments v
8908150321 890630 PDR ADOCK 05000282
)
k PLC j
EXHIBIT A PRAIRIE ISLAND NUCLEAR GENERATING PLANT ANNUAL REPORT OF CHANGES, TESTS AND EXPERIMENTS January 1, 1988 to December 31, 1988 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)..
1.
SFP DEMINERALIZED INLET BYPASS LINE (87L967)
Description of Modification This modification was initiated to reduce silica concentrations in the refueling water storage tanks by processing the water through the-reverse osmosis system. The reverse osmosis system, when in service, normally takes suction directly from the spent fuel pool. This modification installed additional piping between the spent fuel pool purification loop and the Refueling Water Storage Tanks (RUST) which allows RWST water and spent fuel pool water to mix. This enables the silica levels in the spent fuel pool and the selected RWST to equalize and be reduced simultaneously.
Summary of Safety Evaluation By procedure, RWST and SFP boron concentrations are checked prior to mixing, so as to prevent inadvertent boron dilution below technical specifications.
No alterations or modifications were performed on safety related portions of the spent fuel cooling system. Therefore, this modification does not create any flowpath for drain down of the SFP or dilution of the RWST not already analyzed in the USAR.
2.
REVERSE OSMOSIS SYSTEM (86L896)
Description of Modification This modification temporarily installed a reverse osmosis system for the purpose of reducing silica levels to within Westinghouse limits in various tanks throughout the plant. During normal operation the reverse osmosis system takes water directly from the spent fuel pool, removes the silica and returns the clean water back to the spent fuel pool.
Upon completion of the silica reduction effort, the reverse osmosis system is disconnected, disassembled and placed into storage.
l I
Summary of Safety Evaluation All work for this modification was performed under normal QA-III requirements.
Installation was such that the spent fuel pool could not l
A-1 l
i l
l
4 be drained below to hnical specifications.
In addition, a temporary dike was installed around tne reverse osmosis system to direct system leakage and rejected effluent to the liquid radwaste system.
Design
{
analysis and review determined that use of the reverse osmosis system j
would have no adverse impact on plant operation or design.
3.
Addition of Clean out Flannes on the Laundry and Hot Shower Tanks Overflow Line ;87L012)
Description of Modification Modification 87LO12 involves replacing an elbow on the Hot Shower Tanks overflow line (2-WL-21) with a tee and a clean-out flange.
The clean-out flange consists of the flanged connection, a diaphragm valve, and a hose connection for demineralized water.
This will allow for the flushing oi hot spots in the Aerated Drains Monitor Tanks overflow line to the Waste Hold-Up Tank. Figure 9.1-3 of the USAR will be changed to show the addition of the clean-out flange.
Summary of Safety Evaluation All work associated with this modification took place in the Aux.
Building (695' el.).
During installation, a fire watch was established during all cutting and welding. A fire retardant blanke, was laid over j
neighboring cable trays to minimize the fire hazard due to sparks caused from cutting and welding. A temporary containment was also installed around the work area in order to prevent contamination of tihe surrounding area. Additional lead shielding was placed over the nearby
" hot" piping in order to minimize worker exposure.
All materials conform to Fluor Pioneer Inc. Specification 106A-Section I
IV, SS-M 380 (NSP), and were fabricated and ins".alled as per \\NSI-B31.1.
Since the line in cuestion is a4 overflow line (i.e. cannot be isolated), a hydrostatic test was not performed on the new welds as is called for in ANSI-B31.1. A visual inspection combined with a dye-penetrant test and a leak check during flushing of the line with demineralize > <ater were used as acceptance criteria for the new welds.
l The only portions of the Waste Liquid Treatment Systems on the Q-list are those associated with containment penetration.
The elbow is classified as QA-III, and the r.sw parts will also be classified as QA-III. The affected overflow lines (3-WL-21 and 2-WL-21) do not perform a safety related function, do not support any safety related components, nor are they required to bring the plant to safe shutdown. No safety power supply components will be affected.
l.
A-2 l
1.
4.
Removal of Hot Chem Lab Fume Hood and Addition of New Laundry and CVCS_ Monitor Tanks Exhaust (86L960)
Description of Modification In order to' increase the efficient operation of the Hot Chem Lab, Modification 86L960 involved removing the #121 Hot Chem Lab Fume Hood to make room for more bench space. The #121 Hot Chem Lab Exhaust equipment (fan, PAC filter, etc.) was then used for the new laundry dryers that were insc 41ed in the Demineralized Lift Roon (735' el. Aux. Building) as a part of Modification 86L943. To ninimize the possibility of an airborne hazard in the area, the 4" vents on the CVCS Monitor Tanks were also tied into this exhaust line. A particulate filter was installed f n the line near the dryers, in order '
temove th-excess lint due to the laundry, before reaching the PAC filter unit.
The equipment labels were changed as follows:
The #121 Hot Chta Lab Exhaust equipment became the #121 Laundry / Monitor Tanks Exhaust equipment, and the #122 Hot Chem Lab Exhaust equipment became the #121 Hot Chem Lab Exhaust equipment.
Figure 10.3-5 of the USAR will be changed to reflect the changes mentioned above.
The status of the Category 1 Vent Zone Boundaries on the Hot Lab and Sample Room Vent system were also corrected to as found conditions.
Summary of Safety Evaluation All equipment installed or affected by this modification are classified as QA-III.
This equipment does not perform a safety furstion, nor is it directly associated with safety related equipment. No safety related power supply components were affected.
5.
Addition oi Laundry Washer to Demineralized Water System (86L943)
Description of Modificati2D In order to increase the efficient operation of the radioactive laundry process, Modification 86L943 involved combining the wet wash and dry leaning operations by installing a new wet washer and two new dryers in the Demineralized Lift Room (735' el. Aux. Bldg.).
To keep the fire hazard in the area (lint from laundry) to a minimum, the dryers utilize steam ccils to heat the air for drying, as opposed to electric coils which could supply an ignition source.
The steam for these heating coils comes from the heating system.
Since there is no l
potable water source available in the area, demineralized water is used for the washer. This water is heated by injecting steam directly into the water inside the washer cylinder.
A-3
i l.
p 1
L i
Summary of Safety Evaluation j
Presently, the waste water from the original laundry washers is collectea in the aerated sump and is treated as aerated waste. The vaste liquid from the new washer is dumped down a local floor drain, c ore it then goes to the aerated sump. Therefore, no significant change was made to the waste liquid system. However, USAR 9.2-2 states that the laundry waste is collected in the laundry and hot shower tanks.
USAR 9.2-2 will be updated to show the new changes, and Figure 10.2-9 of the USAR will be updated to show the addition of the new washer on the demineralized water system.
The equipment installed in this modification is not safety related, nor is it directly connected to any safety related systems. All components installed or affected by this modification are classified as QA-III. No safety related power supply components were installed or affected.
6.
Reclacengnt V/_LCOR Solenoid Control Valves on the RHR Dmp.la Lines (87L006)
Descrintf oq of ifodification Modification 87 LOO 6 consisted of replacing the Valcor solenoid control valves on the RHR sample lines (SV-33640 for Unit 1 and SV-33668 for Unit 2) with a rection of austenitic stainless steel tubing. The aason for tLeir removs1 was because of the risk that'th y might fail clo;ed during a post accident condition. These risks were just.ified when SV-
?3668 failed to open when tested prior to the Unit 2 refueling outage in January of 1988, and when SV-33640 failed to open during the forced Unit 1 outage in March of 1988. The RHR Sample is now controlled by a manual valve on the Sample Panel in the Sample Room.
Summayv cf Safety Evaluation These solenoid valves (SV-33640 and SV-33668) and the sample lines they are on are classified as QA-III according to Appendix B of 10CFR50.
All sample lines consist of sustenitic stainless steel tubing and are designed for high pressure service (the PHR system pressure is 450 psig,
)
j whereas the design pressure for sample piping is 2500 psig.).
The sample lines are routed in uni-strut and are locsted so as to protect them frem at.cidental dana,e during routine operation and maintenance.
]
3 l
The electrically operated Valcor solenoid valves were DC powered and, therefore, their removal does not affect any safety related power i
sources. These valves were controlled from the Sample Room by an "Open-i Close" switch. They did not close upon a containment isolation signal.
i A-4 l
i i
,.., d 7.
Auxiliary Feedwater Turbine Throttle Valve ' Stem Leak Off (86L332)
Description of Modification Previously the throttle valve'atem leak off lines were routed to floor drains near the auxiliary feedwater turbine exhaust vent stack. This modification rerouted the. stem leak.off lines to the auxiliary feedwater pump turbine exhaust. stack. Stem leak off can now blow unrestricted to the' atmosphere via the turbine exhaust stack during operation of the auxiliary feedwater turbine driven pumps.
Summarv of Safety Evaluation Previously,.the outlets for the auxiliary feedwater turbine throttle valve stem leak off were ruuted to floor drains near the auxiliary feedwater pump turbines. The purpose for the stem leak off lines is to prevent' pressure from building up within the stems of thu valves as a result of leaks through the stem bushings. However, since these lines vera'disigned to discharge into the atmosphere of the auxiliary i
feedwater pump rooms, flow throagh these lines was restricted by partially closing the isolation valves in an attempt to minicize the potential for the development of a steam environment.
Modification 86L932 rerouted the stem' leak off lines to direct their discharge to the,suxiliary feedwater turbine exhaust header. This-modification removed the potential for creating a steam environment in the auxiliary.feedwater pump room during auxiliary feedwater turbine driven pump operation.
all piping installed under this modification was seismically designed and mounted. Piping modification were analyzed and found acceptable in accordance with the requirements of the Prairie Island USAR.
8.
Unit 1 Cvele 13 Reload (88LO411 Description of Chance Prairie Island Unit 1 Cycle 13 began operation in September 1988 and is expected to shutdown in January 1990.
Cycle 13 is projected to reach an end of hot full power exposure of 15,830 MWD /MTU and to shutdown with an exposure of 16,600 MWD /MIV. This will result in a coastdown of 23 days to 81% of full power at shutdown.
Prairie Island Unit 1 Cycle 13 contains 48 fresh Westinghouse Improved Optimized Fuel Assemblics (OFA), 44 once burned Westinghouse OFA assemblies. 28 twice burned westinghouse (OFA) assemblies and one twice burned Exxon TOPROD assembly. The core uses gadolinium as a burnable poison to control the temperature coefficient and power peaking. The fresh fuel also has approximately 6 inches of natural uranium at the top and bottom of the gadolinium pins only.
The natural uranium blankets were not incorporated in the enriched fuel pins. This reduces axial peaking and increases cycle Ivngth.
A-5
)
2 Summary of Safety Evaluation The. analyses performed in the design and licensing of Unit 1 Cycle 13 operation were done by USP's Nuclear Analysis Depsrtment (NAD) and are summarized in the " Prairie Island Unit 1 Cycle 13 Final Reload Design Report (Reload Safety L'raluation)", NSPNAD-8810P, May 1, 1988. The analyses indicate that he core can be operated within Technical Specification and USAR.3mits.
9.
Unit ? Cvele '1_neload (89L080)
Description of Chance Prairie Island Unit. 2 Cycle 13 began operation in April 1989 and is expected to shutdown in August 1990.
Cycle 13 is projected to reach an end of hot full power exposure of 16,530 MWD /MTU and to shutdown with an exposure of 17,570 MUD /MTU. This will result in a coastdown of 32 days to 69% of full power at shutdown.
Prairie Island Unit 2 Cycle 13 contains 48 fresh Westinghouse Improved Optimized Fuel Assemblies (OFA),
44 once burned Westinghouse OFA assemblies, 28 twice burned westinghouse (OFA) assembi.ies and one twice burned Exxon TOPROD assembly. The core uses gadolinium as a burnable p)ison to control the temperature coefficient and power peaking.
The f21sh fuel also has approximately 6 inches of natural uranium at the top and bottom of the gadolinium pins only.
The natural uranium blankets were not incorporated in the enriched fuel pins.
This reduces axial peaking and increases cycle length.
Summary of Safety Evaluation The anslyses performed in the design and licensing of Unit 2 Cycle 13 operation were done by NSP's Nuclear Analysis Department (NAD) and are summarized in the " Prairie Island Unit 2 Cycle 13 Final Reload Design Report (Reload Safety Evaluation)", NSPNAD-8829P, December 1988. The analyses indicate that the core can be operated within Technical Specification and USAR limits.
10.
CHANGE TO OPERATIONAL OUALITY ASSURANCE PLAN APPENDIX C Revision 13 to the NSP Operational Quality Assurance Plan was internally reviewed and approved on May 24, 1989. 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 nuclear 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)] are presented in Appendix D to the plan.
Tho Operational Quality Assurance Plan, Revision 13, is Appendix C of the USAR.
A-C
L PAGE CHANGES TO THE:
PRAIRIE ISLAND USAR 1
NORTHERN STATES POWER COMPANY DESCRIPTION:
Page Revisions for USAR Revision No. 2 (Exhibit B of NSP Letter dated June 30, 1989)
Manifest Date: August 10, 1989 USNRC Fluor Power Services
[#30) 1 DCD.
'[#27, #31-thru 43]
- 14-Attn: A V Setlur Regional Admin-III
[#44]
- 1
+ Plant Nuc Tech Services
[#14) 1
+ Resident Inspector
[#10]
1 Attn: J E Goldsmith Prod Training Quadrex 2
+ T E Amundson
[#11 & #12, #55-60) 8 Attn: Librarian
[#29)
+ EOF (Lee Finholm)
[#54) 1 Attn: R Tomkiewicz
[#46]
+ Prairie Island Plant Mgr
[#2-6,#53) 6 Westinghouse
+ Pwr Supply QA Attn: L Kish
[#50]
1 Attn: T W Bacon
[#64) 1 Attn: R Tomkiewicz (Chpt III Only)
Attn: Library Copy
[#16) 1
+ T N Vogel
[#48]
1
+ M B Sellman
[#67 6 #68) 2 Wisconsin Public Service 2
+ D M Vincent
[#8) 1 Attn: D J Ropson
[#19)
+ F W Hartley
[#9]
1 Attn: Library Copy
[#16) ' 1
+ E'F Eckholt
[#22) 1 MDC
[#61]
+ R 0 Anderson
[#17) 1 Attn: F E Gregor
+ D E Gilberts
[#52]
1 EPM (on loan)
[#20]
1
+ H S Isbin
[#24) 1 Attn:
E Margalejo
+ D M Musolf
[#26]
1 Gasser Assoicates
[#23]
1
+ J A Leveille
[#66) 1 LIS
[#62) 1
+ J A Thie
[#47]
1 Attn:-Lyle Graber
+ Plt Engr & Constr Library
[#15) 1 L J McDonnell
[#51]
1
+ NSS Master Copy
[#1] ** 1 MPCA
[#25]
[#45]
1 Attn: J W Ferman NSS Document Control File (Manifest Only)
American Nuclear Insurers [#21]
1 Attn: Librarian TLG Engineering Attn: Tom LaGuardia
[#65]
1
- Distributed under P-47 manifest
+ Indicates Controlled Copy [
]
Indicates USAR Book No, Insert the attached pages per the attached instructions within 5 working days.
ACkNOWLEDG EMr.
- r all controlled USAR copies, copy hold ign, date and return this page t hern States Power Com uclear Support Services Dept., 414 Nicollet Mall, a olis 55401.
Signature:
Date:
USAR Book No.
Page 1 of 2 1
l l