ML20090D223
| ML20090D223 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 06/29/1984 |
| From: | Musolf D NORTHERN STATES POWER CO. |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20090D233 | List: |
| References | |
| NUDOCS 8407180142 | |
| Download: ML20090D223 (33) | |
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g 'T Northem States Power Company 414 Neollet Mall Minneapoks, Minnesota 55401 Telephone (612) 330-5500 June 29, 1984 Director Office of Nuclear Reactor Regulation U S Nuclear Regulatory Commission Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANI Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Submittal of Revision No. 2 to the Updated Safety Analysis Report (USAR)
Purusant to 10CFR.50.71(e) we are submitting 13 copies of Revision No. 2 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, 1983 through December 31, 1983.
Exhibit A contains a description and a summary of the sa#ety evaluation i
for changes, tests and experiments made under the provisions of 10CFR 50.59 during this period.
Exhibit B contains the USAR page changes and instructions for entering the pages. Included in Exhibit B is Levision 9 to the Northern States -
Power Company Operational Quality Assurance Plan in compliance with 10CFR 50.54(a). Changes included in Revision 9 to the plan are described in Exhibit A (Item 18, page 10) of this letter, ws David Musoif Maneger - Nuclear S . ort Services Dre!/ TAP /bd c: Regional Administrator-III, NRC NRR Project Manager (w/o enclosure)
Resident Inspector, NRC G Charnoff Attachments b
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i EXHIBIT A ;
PRAIRIE ISLAND NUCLEAR GENERATING PLANT !
ANNUAL REPORT OF CHANGES, TEST AND EXPERIMENTS l January 1, 1983 to December 31, 1983 ,
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. ITEMS WHICH CONTRIBUTE TO HYOROGEN GENERATION INSIDE CONTAINMENT (SE #97)
Description of Change
. Portions of galvanized ductwork inside containment which were not necessary were removed so as not to contribute to hydrogen generation.
' Summary of Safety Evaluation Galvanized ductwork was removed from containment resulting in less hydrogen generaton. Hydrogen build-up is thus, slower.
- 2. ANALYSIS OF ADEQUACY OF STATION ELECTRICAL DISTRIBUTION SYSTEM VOLTAGES (SE# 114)
Description'of Change An " Analysis of Adequacy of Station Electric Distribution System Voltages" was performed and submitted to the NRC as an attachment to a letter dated July 17, 1981. This had been implemented by Reference'in A re-analysis was performed which considered normal-
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USAR Section 8.
grid conditions with abnormal 1R and 2R transformer configurations and worst case plant loads.
Summary of Safety Evaluation This analysis does not alter prior acceptance criteria for safeguards equipment electrical support. It establishes bounding conditions that are within the capabilities of the NSP system that assure support in any forseeable event.
- 3. CIRCULATING WATER INTAKE AND DISCHARGE MODIFICATIONS (78YO73)
Description of Change This design change provided for the modification of the Circulating Water Intake and Discharge.
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~ Summary of Safety Evaluation This design change is non-safety related, QA Type III. Portions of
, the USAR describing the circulating water system are affected and have been updated in Revision 2. Interfaces between plant systems.
and the circulatfng water intake and discharge modifications were reviewed and no safety concerns were identified.
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- 4. REACTOR COOLANT VENT SYSTEM (80Y117)
Description of Change The purpose of this design change is to install the capability of remotely venting noncondensable gases from the reactor coolant system high points, in particular the reactor vessel head and the pressurzier steam space. Noncondensable gases in the reactor coolant system can degrade the core cooling effectiveness of the ECCS. The design bases for.the reactor coolant gas vent system was specified in NUREG-0578 (short term lessons learned) and the appropriate sections of the NRC letter to all operating power plants of October 30, 1979.
Summary of Safety Evaluation The Reactor Coolant Vent System is designed to safely and reliably vent noncondensible gases from either the reactor vessel head or the pressurizer steam space in order to maintain an adequate mode of core cooling. The system design has taken into account those design considerations as contained in the detailed design change / safety evaluation package, which will guarantee that the installation and operation of the system will meet the criteria as defined by NUREG-0578 and,related followup letters from the NRC.
- 5. INCORE THERMO COUPLE MODIFICATIONS (80Y120)
Description of Change This design change addresses the installation of the inside of containment penetration adaptors and mineral insulated, metal sheathed thermocouple extension cable from the refueling pool-side disconnect to the penetration.
This design change modified the incore thermocouple system by: 1)
Replacing the.Thermo-Electric type K head connector and mate with qualified mating connector. 2) Replace the thermocouple extension wire in containment with a qualified mineral insulated-metal sheathed cable which extended to the containment penetration. 3) The reference temperature junction boxes were replaced and their location changed to outside of containment. 4) Thermoc uple extension wire was added from the containment penetrations to the reference junction boxes and copper field wiring was run from the reference junction boxes to the incore panel and computer.
Summary of Safety Evaluation The objectives of the design change was to upgrade the Incore Thermo-couple System to a qualified IE system. This allows qualified input of Core Exit Thermocouples to both the subcooled margin monitors and the Tech Support Center B&W Recall Computer. This upgrade does not eliminate any of the present functions of the thermocouple system but only upgrades their reliability.
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- 6. HOT SHUTDOWN PANEL MODIFICATIONS (80Y140)
Description of Change This design change fulfills part of an-NSP commitment to the NRC l to isolate the indication of selected parameters from the effects of a control room / relay room fire. The indicator portion of the Hot
> l Shutdown Panels was.reconfigured to display more parameters using dual l channel indicators.' When this design change was completed, each Hot '
, Steam Shutdown panel displayed the following protected outputs:
Generator Level (2/ unit), RC Loop Pressure (2/ unit), RC Loop Temperature
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Th (2/ unit),'RC Loop Temperature Tc (2/ unit), and Pressurizer Level - j
- Cold Cal. Each Hot shutdown Panel will continue to display the ;
i following nonisolated signals: Steam Header Pressure (2/ unit) and j Pressurizer Pressure. !
Summary of Safety Evaluation l 3
This design change involved fabrication of subpanels, installation of !
new instrumentation, bracing the existing Hot Shutdown Panels, cutting !
away portions of the Hot Shutdown Panel fronts, and mounting the new subpanels in the Hot Shutdown Panel enclosures. The display of plant t
. parameters at the Hot Shutdown Panel will not be degraded by this j design change. The indicators selected for this QA III function l are of the same type and manufacturer as those selected for some QA I l
applications at Prairie Island.
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- 7. T0XIC GAS MONITORS (80Y154) i Description of Change l' This project provided for the installation of gas analyzers to monitor r
for. e:blorine, ammonia, formaldehyde, and hydrogen chloride and upon ,
actuation place the Control, Relay, and Computer Rooms ventilation :
system into the recirculation mode through the PAC filter and close t the outside air intake. dampers. This is in compliance with Regulatory Guides 1.78 and 1.95 covering operator exposure to toxic gases. !
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! -Summary of Safety Evaluation !
4 This design change provides for the control system design modification, j procurement, and installation of gas analyzers for the Control, Relay, The regulatory requirements l and Computer Room Ventilation System.
pertaining to Chlorine Detection are contained in the Regulatory Guide 1.95. The Control, Relay and Computer Room Ventilation System consists ;
of two trains of ventilation equipment. When chlorine is detected, t the outside air dampers and steam exclusion dampers are closed, and Particulate, Absolute, and Charcoal-(PAC) filter dampers are opened.
Control Room isolation is accomplished with redundant dampers, Train A j and Train B, in each external air path. All equipment necessary for j the operation of the chlorine detectors and system The system actuation are is designed !
Quality Assurance Type 1 and Design Class 1.It is concluded by this evaluation l in accordance with IEEE 344-1975.
that this change provides additional protection for the control room cperation.
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- 8. AUXILIARY _ FEEDWATER p0MP MODIFICATIONS (80Y240, part J)
- t Description of Change -
This design change is _to insure that the #21 Auxiliary Feedwater Pump
( AFW) is operable in tha event of a relay room or Appendix control'Rroom fire.
, .Certain requirements.
relays were relocated and circuits modified forThese modific
\ to prevent.a breaker trip from a fire-induced sh' ort, the f l v of aQemote trarisfer switch to insure tots 1 electrical iso 1'ation by local operation, and where relays cannot be guaranteed to return to
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.' normal positions after loss of off-site power and a relay room fire, 1
bypasses while in local ~ operation or remote transfer switches areIn a
being installed. eliminated and a timer to the Aux.The Lubsame
.011modifi- Pump installed which w lubricate the #21 AFW for 15 minutes ev'ery 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
sation will apply to the Aux Lube Oil P, ump for #12 AFW.
g Sumbiry of Safety Evaluation
.\ s This designtchange does not alter the AFW system mech
-s i oil ~regulrements.
aking relay T03/826 out of local op'eration will not affect the FSAR Load restoration will remain the same section on load sequencing.While the AFW pump is in local operation, for remote operation.
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af.tivation of relay LRY/826 would still trip the pump since the remote switch in series with'this contact will be from the diesel generator
}' controls. For Appendix R reouirements, the remote diesel generator controls, a.d 9 the load rejection and restoration schemes for "A" Train, are assumed to be unreliable after a relay room fire.
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- 9. INTERNAL HYOROGEN RECOMBINERS (82Y255)
Description of Change
- 1 This design change provides for the installation of two redundant of'this system include redundant Westinghouse type re each containment, redundant power supply and control panels in the Auxiliary Building, new electrical penetration assemblies reference junction boxas anE control wiring and power cables from new
'- MCC to power sus;' lies and recombiners.
s" ^rv ' s ''" " he operation of the hydrogen recombiners
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All equipment necessary for t The system is qualified in accordance are QA Type I and Ot: sign Class 1.to the plant environmental specifications as with IEEE-323, 1974 detailed in our rer,ponse to IE Bulletin 79-01B and IEEE-
'N The recombiners will be powered from new safeguards MCCs located in
\ Train A and B safeguards chiller rpoms.
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- 10. REACTOR COOLANT PUMP DISCHARGE CROSS-CONNECT TO CONTROL SYSTEM (82L702)
Description of Change This design change installed a permanent cross-connect from the reactor coolant drain pump discharge to the chemical and volume control system. The cross-connect is used for cleaning up the RCS and refueling pool during refueling shutdowns, when The reactor thetank drain RCSpumps is are depressurized and in cold shutdown.
needed to provide head for flow through the CVCS for cleaning.
Summary of Safety Evaluatin All new piping installation and fabrication were in accordance with Proper quality specifications. Piping was seismically analyzed. Chances
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assurance documentation and qualifications was provided.
of operational and installation errors are significantly reduced by this design change due to the elimination of a temporary jumper installation for every refueling outage and a simplified operating procedure for the permanent jumper.
- 11. MANIPULATOR CRANE " INCHING" AND SPLIT SCREEN UNDERW SYSTEM (83Y410)
Description of Change This design change is to upgrade the unit 1 and 2 manipulator cranes to include the " inching" crane movement system and the Split Screen Underwater TV Indexing systems.
This will give a finer, more precise indexing system for fuel handling in addition to a viewing system for inserting fuel into the core.
Summary of Safety Evaluation
- This design change was implemented to aid the operator in the placementBo of fuel assembites into the reactor core. The designed for use on the Prairie Island Manipulator Cranes.
inching modification allows the operator to make very precise bridge and trolley movements while centering the fuel assembly over its An " inching" permissive must be switched on intended core location. This permissive locks out prior to the use of the joy stick control.all the other bridge and trolley inching joy stick. The inching modification is being supplemented by This gives the operator a 360' an underwater Split-Screen TV system. The view of the fuel assembly as it is entering or leaving the core.
operator can use the inching control and the TV system to center the fuel assembly to ensure that it can be lowered Overall,into thethe with core usewithout of contacting the surrounding assemblies.
these two systems, the movement of fuel will be performed in a more controlled, precise and sc.fe manner than was done in the past.
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- 12[ UNINTE1RUTABLE POWER SUPplIEh.(83Y415) m y n.
Description o/ Change /
3 This design change replaces' the existing four (4) 75 KVA' inverters '
with four (4) Uninterru7 table Power Supplies which each Jnclude a 5.0 -
KVA inverter,-sutomatic transfer switch, and a manual' bypass switch.
The new' power supplies-allows for automatic transfer of power to Also the manual ,
an alternate AC source in-case of inverter failure. When t
byp' ass' switch. tallows fpr inverter maintenance under power. ,
any one of tfe< inverters is bypassed it is annunciated separately !
in the Control Room.,.,,
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. Summary of Safety Evaluation d A1.1 variations between the old inverters and the new Uninterruptable l Power Supplies have been evaluated and addressed in the detailed !
destpn description and safety evaluation. Included were an evaluation '
a icf the 5.0 KVA rating vs 7.5 KVA (the maximum loading is 4.0 (KVA), ,
the[a'ddition of an automatic static transfer switc!),which is designed l to transfer the instrument bus load from the inverter, if it fails, to the ACLsource through a bypass breaker (this transfer switch adds to :
7- f the reliability of the 120 VAC<saurce being supplied to the instrument l
.4 bus). addition of ~a manual' bypart iswitch allowing for inverter repair
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,') or maintenance without ngehto rely on panel 217 to supply the load,
^ seismic considerations and electrical variations.
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^ With the addition of ode morelahdenate source of power to the instrument m power, buses there will be in$rease in the reliability of the buses and ,'
- therefore an. increase in the margin i, of plant safety.
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13.- INSTALLATION OF A MANUAL'ISOLAT,l_0N VALVE DOWNSTREAM OF CV31420 (83L742)
Description of Change .
3 This design change 51s for the installation of a manual isolation valve downstream of CV31420. The valve is normally open and was 7~ t installed in the line so CV31420 may be isolated from RCS pressure and repaired under. power.
i Summary of Safety Evaluation ;
The installation of a manual valve in the charging line to the RCS does not create a possibility for an. accident or malfunction of a different. type than evaluated previously. The valve utilized has
'been use( in many places in the RCS and is compatible with the piping specification. The line has been stress analyzed with the additional valve installed,,and the load increases appear to be insignificant.
The stress levels are within acceptable limits. -I t
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- 14. CAUSTIC ADDITION SYSTEM (83L745)
Description of Change i
This design change performs modification to the Caustic Addition System which assures that the pH limit to which the electrical equip-ment in containment is qualified will not be exceeded at any time.
This design change is a result of an evaluation which was performed in accordance with IE Bulletin 79-Olb on the qualification of safety >
related electrical equipment when exposed to various harsh environ-mental conditions. The evaluation showed that some accident scenarios resulted in a containment spray pH greater than 13 during the injection phase. -The pH would not decrease to 10.5 until the initiation of the recirculation mode. Safety related electrical equipment has been evaluated to withstand a chemical environment consisting of 2100 ppm boric acid with the pH adjusted to a maximun of 10.5 through the cddition of sodium hydroxide.
Summary of Safety Evaluation i This design change resulted in the containment spray injection pH ;
varying from 10.45 to 9.1 depending uoon pumping combinations (RHR, '
SIP and CSPs) which is acceptable. Recirculation phase pH could r:ach a minimum value of 8.2 which is less than the criteria in the Standard Review Plan but still consistent with the Prairie Island ,
FSAR. A pH less than 8.5 but greater than 7 is considered acceptable !
because; a) the tank volumes and concentrations for this pH value were taken in the worst cast situation, not the normal condition, b) the ,
n:rmal tank levels and solution concentrations would give a sump pH (R:cirewater phase pH) of 8.4, c) the changes in pH at this level do n:t affect stress corrosion cracking at Prairie Island d) at the pH of
~ 8.2 the partition factor is above 2000, consequently, most if not all, of the iodine absorbed during the spray should remain in solution.
e) the FSAR analysis takes credit for a reduction of iodine concentration . 1 by a factor of only 100 (DE=100). The FSAR also shows that 10CFR 100 , -
guidelines are NOT exceeded if the Caustic Addition System were not installed, f) the recirculation phase water pH can be monitored by '
sampling to ensure the pH levels are greater than 7. Additives can be used during the recirculation phase to maintain the pH at an acceptable j level.
Given the proceeding considerations, th9 modifications meet the limiting ;
Conditions of Operation, the FSAR and-10CFR 100,-
Piping and structural analyses were perfermed and were still within ,
the acceptable limits.
.15. RADI0 ACTIVE WASTE LIQUID TREATMENT SYSTEM (83L754) [
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Description of Change A CHEM-Nuclear 24 inch diameter pressure demineralizer vessel containing activated carbon was. installed as a filter media in Ifeu !
of the ADT Evaporator in the Radioactive Waste Liquid Treatment _
System. The filter unit is located in the old Waste Concentrates L
The feed line to the 5 Tank Pump Room of the RAD Waste Building.
G.M9 ADT Evaporator was modified to allow flow to the new filterA If unit or to the evaporator.
filter unit back into the 5 GPM Evaporator RoomThis and tie into the distillate discharge line to the Condensate Receiver Tanks.
allowed flow from the ADT Collection Tanks through the ADT Filters to the Activated Charcoal filter and on to the ADT Condensate Receiver Tanks.
Ion Exchangers prior to release from the ADT Monitor Tanks.
Summary of Safety Evaluation A11 piping was QA Type III and the No tie-ins were made into existin new penetrations Type III lines as defined by plant data files.
were Boundary walls.
made in any structual, fire barrier, or Special Ventilati The USAR has been updated to reflect the Installation Procedures. change in the description of the Waste Liquid Trea 8,__ CORE RELOAD (83L7621
- 16. UNIT 2. CYCLE Description of Change This design change addresses the Core Reload for Unit 2 Cycle 8.
During the Prairie Island Unit 2 Cycle 7/8 refueling outage, 40 spen
& Exxon STANDARD region 7(G) fuel assemblies and 1 Westinghous 6(F) assembly was replaced with 20 fresh Exxon TOPROD region discharged from Unit 1 at the The end of cycle remainder of blies region 7 after o operation, and 1 Westinghouse region 2(B) assembly.
the' 9(I), core 36 twice will consist burned of 40 once Exxon STANDARD burned Three Exxon TOPROD assemblies of the region assem 8(H twice burned Exxon lead TOPROD assemblies region SL(H). Cycle 8 are th assemblies to be placed in the core for Unit 2 The. failed pins .
assemblies which were reconstituted in April, 1983.
(1 per assembly) were replaced with pins of similar burnups and histories.
Cycle 8 is projected to last 11,700 MWD /MTU (262 EFPD including a coast down of 600 MWD /MTU. 4 w/o gadolinia blies.
the loading pattern is a low leakage pattern with 64, Summary of Safety Evaluation The analyses performed in the design and licensing D) and of Unit 2 C operation were done by NSP's Nuclear Analysis l dDepartment (NA are summarized in the " Prairie Island Unit 2, Cycle 8 Final Re oa Design Report (Reload Safety Evaluation)" Sept.,1983 and th The analyses indicate that theking corefactor will operat Specification limits with respect to shutdown reactivity worth, temperature coefficient restrictions, and hot channel pea F g.
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During cycle 7 the primary coolant activity indicated the presence of :
at least one failed fuel pin. The decision was made to reconstitute ;
3 the failed assembly (ies) during the shuffle. NSP's Nuclear Analysis Department has analyzed the effects of replacing the failed pin (s) ,
with an inert zire rod (s) and has concluded that there would be no unresolved safety ouestions as long as guidelines for the reconstitution provided by NAD to the plant were followed.
No additional analysis was required for the three fuei assemblies l
I (I-23, 37.and 40) that were reconstituted in April,1983 since the replacement pins were from Fuel Assembly H-90 and had similar burnups. l In summary, sufficient analyses have been-performed to show that the f Unit 2 Cycle 8 reload core can be safely loaded into Unit 2 Reactor. .
- 17. UNIT 1.~ CYCLE 9 CORE RELOAD j Description of Change '
This design change addresses the Core Reload for Unit 1 Cycle 9.
During the Prairie Island Unit 2 Cycle 8/9 refueling outage, 40 Exxon i region 8(H) assemblies and 1 Westinghouse region 2(B) assembly will l be replaced with 20 fresh Exxon TOPROD region 11(L) assemblies, 20 once burned Exxon TOPROD region 9(I) from Unit 1 Cycle 7 and 1 twice l
burned Westinghouse region 2(B) assembly. The remainder of the core 3 will consist of 40 once burned Exxon TOPR00 region 10(J) assemblies !
' and 40 once burned Exxon TOPROD region 10A(K) assemblies. Due to }
outage schedule considerations, the low leakage core design was not !
I used and all new assemblies will be loaded on the outside of the core. This loading pattern allows for a cycle 9 hot full power !
lifetime of 11, 400 MWO/MTU (approximately 305 EFPDs) assuming an EOL l exposure of cycle 8 of 12, 738 MWO/MTV.
6 Summary of Safety Evaluation The " Prairie Island Unit 1, Cycle 9, Final Reload Design Report ;
-(Reload Safety Evaluation)", October, 1983 validated a safe fuel loading and subsequent operation using the specified loading pattern provided the end of Cycle 8 exposure was 12,738 i 500 MWD /MTV. The ,
transient analysis and reactor physics sections were completed by NSP's Nuclear. Analysis Department (NAD) while the LOCA/ECCS analysis L-was completed by Exxon Nuclear using data supplied by NSP through the l L
Reload Safety' Evaluation.- However, Unit 1 was shut down early at an '
exposure of 11,940 MWD /MTU which is not within the bounds of the
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. Reload Safety Evaluation. The Reload Safety Evaluation was re-analyzed ,
' ' -by NSP and verified to be valid. The updated data was also supplied i i-to Exxon and the LOCA/ECCS analysis was also verified to be valid !
considering this new.EOC exposure. The Unit 1 Cycle 9 physics character- l 1stics have been analyzed by NSP's NAD and are contained in " Prairie i M -Island Unit 1 Cycle 9, Startup and Operations Report" December, 1983. ;
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- 18. CHANGE TO THE OPERATIONAL QUALITY ASSURANCE PLAN APPENDIX 138 Revision 9 to the NSP Operation Quality Assurance Plan was internally ,
reviewed and approved on April 12, 1984. This revision does not i reduce the commitments in NSP's Operational Quality Assurance Program ,
'and does not adversely impact the safe operation of the nuclear plants. A summary of significant changes in Revision 9 is presented below. . Specific changes with the reason for the change and th.2 basis for concluding no reduction in commitments per 10CFR 50.54(a)(3)(iii), i are presented in Appendix D to the plan. The Operatioral Quality Assurance Plan, Revision 9, is includad in Section.13 to the USAR. !
- 1. Change Security _and Radiation Environmental Monitoring Description Exceptions to sections in ANSI N18.7-1976 concerning -
security and radiation envrionmental monitoring have been ,
deleted. Management positions in these two areas have !
been added to.the organizational description. These changes place security and radiation environmental I monitoring under the quality assurance program. ,
- 2. Change Inspections and Test Control Description Exceptions to sections in ANSI N18.7-1976 concerning I inspections and test control have been deleted. NSP's ;
alternate wording within the plan remains essentially unchanged. NSP's program implements the N18.7 sections and the plan sections.
- 3. Change Vendor Evaluation and Verification and Vendor Inspections Description The differences between vendor evaluation and verification and vendor inspection have been clarified.
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, .. 4 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Revision 2 to the Updated Safety Analysis Report
'The. attached instructions should be followed when making this revision to the Updated Safety Analysis Report. If you have any questions concerviing this revision call:
Terry Pickens (612) 330-5671 5
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- 2. Plant Site and 2-v 2-v 2 Environs 2-vi 2-vi O 2.9-4 2.9-4 2 2.9-5 2.9-5 0 j TBL 2.9-5 TBL 2.9-5 2
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I 3. Reactor
- FIG 3.3-1 FIG 3.3-1 FIG 3.3-2 FIG 3.3-2 2 FIG 3.3-3 FIG 3.3-3 2 FIG 3.3-4 FIG 3.3-4 2 FIG 3.3-5 FIG 3.3-5 2 FIG 3.3-6 FIG 3.3-6 2 FIG 3.6-5 2
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TBL 4.7-1 T P~ 4.7-1 2 (2 of 3) (2 of 3)
TBL 4.7-1 TBL 4.7-1 2 (3 of 3) (3 of 3)
FIG 4.1-1 -- FIG 4.1-1 2 FIG 4.1-la 2 FIG 4.5-1 -
II 5. Plant Ccntainment Systems 5-i 5-i 2 5-ii ,5-ii 2 5-iii 5-iii 2 5-iv 5-iv 0 r s 5-v 5-v 0 5-vi 5-vi 2 5-vii 5-vii 2 5-viii 5-viii O j; 5-ix 5-ix 2 5-x (BLANK PAGE) NA 5.1-1 5.1-1 2 5.1-2 5.1-2 0 5.2-2 5.2-2 2 5.2-3 5.2-3 O 5.2-4 5.2-4 2 5.2-5 5.2-5 2 5.2-6 5.2-6 0 5.2-7 5.2-7 2 5.2-8 5.2-8 2 5.2-9 5.2-9 2
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II 5. Plant Containment Systems -
5.2-9a 2 (BLANK PAGE) NA 5.2-12 5.2-12 0 5.2-13 5.2-13 2 5.2-16 5.2-16 2 5.2-17 5.2-17 2 5.2-18 5.2-18 2 5.2-18a 5.2-18a 2 5.2-18b 5.2-18b 1 5.2-19 5.2-19 2 5.2-20 5.2-20 2 5.2-21 5.2-21 0 5.2-22 5.2-22 2 5.2-23 5.2-23 2 5.2-24 5.2-24 2 5.2-25 5.2-25 2 5.2-26 5.2-26 2 5.2-27 5.2-27 2 5.2-28 5.2-28 2 5.2-29 5.2-29 2 Y'- .
5.2-30 5.2-30 2 5.2-31 5.2-31 2 5.2-32 5.2-32 2 5.2-33 5.2-33 2 5.2-34 5.2-34 2 5.3-1 5.3-1 2 5.3-2 5.3-2 2 5.3-3 5.3-3 0 5.3-4 5.3-4 2 5.3-5 5.3-5 2 5.3-6 5.3-6 2 5.3-7 5.3-7 0 6
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II 5. Plant Containment Systems 5.3-10 5.3-10 2 5.3-11 5.3-11 2 l
1 5.3-12 5.3-12 2 l 5.3-13 5.3-13 0 i 5.4-2 5.4-2 O l 5.4-3 5.4-3 2 I i
5.4-4 5.4-4 2 5.4-5 5.4-5 2 i
5.4-5a 2 !
5.4-5b 2 l 5.4-10 5.4-10 2 5.4-11 5.4-11 O i i
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TBL 5.2-1 TBL 5.2-1 1 (6 of 17) (6 of 17)
TBL 5.2-1 TBL 5.2-1 1 (7 of 17) (7 of 17) i TBL 5.2-1 TBL 5.2-1 2 (8 of 17) (8 of 17) l TBL 5.2-1 TBL 5.2-1 1 l
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II 5. Plant Containment Systems TBL 5.3-3 TBL 5.3-3 O (3 of 3) (3 of 3)
TBL 5.3-4 TBL 5.3-4 2 TBL 5.4-1 TBL 5.4-1 2 (1 of 2) (1 of 2)
TBL 5.4-1 TBL 5.4-1 O (2 of 2) (2 of 2)
FIG 5.2-2 FIG 5.2-2 2 FIG 5.2-4 FIG 5.2-4 2 (1 of 2) (1 of 3)
FIG 5.2-4 FIG 5.2-4 2 (2 of 2) (2 of 3)
FIG 5.2-4 2 (3 of 3)
FIG 5.2-10 FIG 5.2-10 2 FIG 5.4-1 FIG 5.4-1 2
- 6. Engineered Safeguards 6-1 6-i O 6-ii 6-ii 2 6-iii 6-iii 2 6-iv 6-iv .O 6.2-23 6.2-23 0 6.2-24 6.2-24 2 6.2-25 6.2-25 0 6.2-26 6.2-26 2 6.2-26a 2 (BLANK PAGE) NA 6.2-27 6.2-27 2 6.2-28 6,2-28 0 6.2-29 6.2-29 2
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r II 6. Engineered Safeguards 6.4-11 6.4-11 2 6.4-12 6.4-12 O FIG 6.3-1 FIG 6.3-1 2
- 7. Plant Instrumentation and Control Systems l 7.8-6 7.8-6 0 7.8-7 7.8-7 2 FIG 7.6-1 FIG 7.6-1 2 III 8. P,lant Electrical Systems 8.3-4 8.3-4 2 8.3-5 8.3-5 1 8.5-1 8.5-1 0 8.6-1 8.6-1 2 8.9-1 8.9-1 2 8.10-1 8.10-1 2 FIG 8.2-2 FIG 8.2-2 2 FIG 8.5-2 FIG 8.5-2 2 i
- 9. Plant Radioactive Waste Control Systems 9.2-2 9.2-2 0 9.2-3 9.2-3 2 9.3-4 9.3-4 0 9.4-1 9.4-1 2 FIG 9.1-6 FIG 9.1-6 2 L
FIG 9.1-8 FIG 9.1-8 2 t
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III 10. Plant Auxiliary 10-1 10-1 2 r Systems 10-11 10-11 O l.
10-iii 10-iii 2 l 10-iv 10-iv 2 10-v 10-v 2 10-vi 10-vi O i 10.2-7 10.2-7 2 10.2-8 10.2-8 O 4 l
10.2-17 10.2-17 2 10.2-18 10.2-17a 2 (BLANK PAGE) NA 10.2-18 0 10.3-2 10.3-2 2 ;
10.3-3 10.3-2a 2 ;
I (BLANK PAGE) NA 10.3-3 0 .
l 10.3-10 10.3-10 0 10.3-11 10.3-11 2 10.3-11a 2 (BLANK PAGE) NA 10.3-26 O i 10.3-26 .
10.3-27 10.3-27 2 (BLANK PAGE) NA f 10.4-1 O l l i
10.6-1 10.6-1 2 l 10.6-2 10.6-2 2 ;
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l FIG 10.2-7 FIG 10.2-7 2 i
2 l FIG 10.3-6 FIG 10.3-6 IV 11. Plant Power Conversion i Systems 11.4-4 11.4-4 0 11.5-1 11.5-1 2 11.5-2 11.5-2 2 i 11.6-1 11.6-1 0 10 l
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IV 11. Plant Power Conversion Systems 11.9-2 11.9-2 0 11.9-3 11.9-3 2 11.9-4 11.9-4 2 !
11.9-5 11.9-5 1 !
FIG 11.1-17 FIG 11.1-17 2 FIG 11.1-18 FIG 11.1-18 2
- 12. Plint Structures 12-ix 12-ix 2 and Shielding 12-x 12-x 2 i e
12-xi 12-xi O i 12-xii 12-xii 2 !
12.2-104 12.2-104 1 12.2-105 12.2-105 2 12.2-106 12.2-106 1 l 12.2-107 12.2-107 , 2 12.2-108 12.2-108 2 12.2-109 12.2-109 2 12.2-110 2 12.2-111 2 12.2-112 . 2 .
(BLANK PAGE) NA i
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TBL 12.2-40 2 (2 of 2)
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! 13. Plant Operations 13-i 13-i 1 13-11 13-ii 2 ,
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13.2-4 13.2-4 2 l 13.3-1 13.3-1 2 i
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Volume No. Tab Page No. Page No. No. 1 IV 13. Plant Operations 13.4-1 13.4-1 2 13.4-2 13.4-2 2 13.4-3 13.4-3 2 13.5-1 13.5-1 O l 13.6-1 13.6-1 0 ;
13.7-1 2 13.8-1 2 13A. Operational QA - Remove the Operational QA Plan Plan Revision 8 (cover sheet, Table of Contents and 78 pages).
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and 91 pages). l V 14. Safety and Accident Analysis 14-1 14-1 2 14-11 14-11 2 14-iii 14-iii 2 14-iv 14-iv 2 14-v 14-v 2 14-vi 14-vi .O 14-vii 14-vii 2 14-viii 14-viii 2 14-ix 14-ix 2 14-x 14-x 2 14-xi 14-xi 2 14-xii 14-xii 2 ,
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V 14. Safety and Accident l Analysis 14.4-1 14.4-1 2 14.4-2 14.4-la 2 ;
14.4-1b ?
14.4-ic 2 14.4-1d 2 !
14.4-2 2 14.4-3 14.4-3 2 14.4-4 14.4-4 2 1 14.4-5 14.4-5 0 14.4-6 14.4-6 2 :
14.4-7 14.4-7 2 14.4-0 14.4-8 2 14.4-9 14.4-9 0 14.4-10 14.4-10 2 14.4-11 14.4-11 2 14.4-12 14 3 4-12 0 14.4-13 14.4-13 2 14.4-14 14.4-14 2 :
14.4-15 14.4-15 2 14.4-16 14.4-16 2 14.4-17 14.4-17 2 14.4-18 14.4-18 2 14.4-19 14.4-19 2 14.4-20 14.4-20 2 14.4-21 14.4-21 2 14.4-22 14,4-22 2 14.4-23 14.4-23 0 14.4-24 14.4-24 2 l
14.4-25 14.4-25 2 '
14.4-26 14.4-26 2 i
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V 14. Safety and Accident Analysis 14.5-1 14.5-1 2 14.5-2 14.5-2 0 14.5-5 14.5-5 2 14.5-6 14.5-6 2 14.5-7 14.5-7 2 14.5-8 j 14.5-8 2 14.5-9 14.5-9 2 14.5-10 14.5-10 2 14.5-11 14.5-11 2 14.5-12 14.5-12 0 14.5-21 14.5-21 2 14.5-22 14.5-22 2 14.5-23 14.5-23 2 14.5-24 14.5-24 2 14.5-25 14.5-25 2 14.5-26 14.5-26 2 14.5-27 14.5-27 2 14.5-28 (BLANK PAGE) NA 1
14.6-1 14.6-1 0 14.6-2 14.6-2 2 14.6-3 14.6-3 2 ,
14.6-4 14.6-4 2 14.6-5 14.6-5 2 14.6-6 14.7-1 2 14.6-7 -
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14.9-15 14.9-15 0 14.9-16 14.9-16 2 14.10-1 14.10-1 2 14.10-2 14.10-2 2 6
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V 14.' Safety.and' Accident Analysis TBL 14.3-1 TBL 14.3-1 2 (1 of 2)
TBL 14.3-1 TBL 14.5-2 2 (2 of 2)
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V 14. Safety and Accident Analysis FIG 14.4-1 FIG 14.4-1 2 FIG 14.4-2 FIG 14.4-2 2 FIG 14.4-3 FIG 14.4-3 2 FIG 14.4-4 FIG 14.4-4 2 FIG 14.4-5 FIG 14.4-5 2 FIG 14.4-6 FIG 14.4-6 2 FIG 14.4-7 FIG 14.4-7 2 FIG 14.4-8 FIG 14.4-8 2 FIG 14.4-9 FIG 14.4-9 2 FIG 14.4-10 FIG 14.4-10 2 FIG 14.4-11 FIG 14.4-11 2 FIG 14.4-12 FIG 14.4-12 2 FIG 14.4-13 FIG 14.4-13 2 FIG 14.4-14 FIG 14.4-14 2 FIG 14.4-15 FIG 14.4-15 2 FIG 14.4-16 FIG 14.4-16 2 FIG 14.4-17 FIG 14.4-17 2 t
FIG 14.4-18 FIG 14.4-18 2 FIG 14.4-19 FIG 14.4-19a 2 ,
i FIG 14.4-19b 2 FIG 14.4-20 FIG 14.4-20 2 FIG 14.4-21 FIG 14.4-21 2 i FIG 14.4-22 FIG 14.4-22 2 16
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V 14. Safety and Accident Analysis FIG 14.4-23 FIG 14.4-23 2 FIG 14.4-24 FIG 14.4-24 2 FIG 14.4-25 FIG 14.4-25 2 FIG 14.4-26 FIG 14.4-26 2 FIG 14.4-27a FIG 14.4-27 2 FIG 14.4-27b -
FIG 14.4-28 FIG 14.4-28 2 FIG 14.4-29 FIG 14.4-29 2 FIG 14.4-30 FIG 14.4-30 2 FIG 14.4-31 FIG 14.4-31 2 FIG 14.4-32 FIG 14.4-32 2 FIG 14.4-33 FIG 14.4-33 2 FIG 14.4-34 FIG 14.4-34 2 FIG 14.4-35 FIG 14.4-35 2 FIG 14.4-36 FIG 14.4-36 2 FIG 14.4-37 FIG 14.4-37 2 FIG 14.4-38 FIG 14.4-38 2 FIG 14.4-39 FIG 14.4-09 2 FIG 14.4-40 FIG 14.4-40 2 FIG 14.4-41 FIG 14.4-41 2 FIG 14.4-42 FIG 14.4-42 2 FIG 14.4-43 FIG 14.4-43 2 FIG 14.4-44 FIG 14.4-44 2 17
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V 14. Safety and Accident l Analysis FIG 14.4-45 FIG 14.4-45 2 FIG 14.4-46 FIG 14.4-46 2 FIG 14.4-47 FIG 14.4-47 2 FIG 14.4-48 FIG 14.4-48a 2 [
FIG 14.4-48b 2 FIG 14.4-48c 2 FIG 14.4-49 -
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FIG 14.4-51 -
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FIG 14.4-53 -
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FIG 14.4-55 -
FIG 14.4-56 -
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FIG 14.4-58 -
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FIG 14.4-60 -
FIG 14.4-61 -
FIG 14.4-62 -
FIG 14.4-63 -
FIG 14.4-64 -
FIG 14.4-65 -
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V 14. Safety and Accident Analysis FIG 14.4-66 -
FIG 14.4-67 -
FIG 14.4-68 -
FIG 14.4-69 -
FIG 14.4-70 -
FIG 14.4-71 -
FIG 14.4-72 -
FIG 14.4-73 -
FIG 14.4-74a -
FIG 14.4-74b -
FIG 14.4-74c -
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V 14. Safety and Accident Analysis FIG 14.5-14 -
FIG 14.5-15 -
FIG 14.5-16 -
FIG 14.5-17 -
FIG 14.5-18 -
FIG 14.5-19 -
FIG 14.5-20 -
FIG 14.5-21 -
FIG 14.5-22 -
FIG 14.5-23 -
FIG 14.5-24 -
FIG 14.5-2b -
FIG 14.5-26 -
FIG 14.5-27 -
FIG 14.6-1 FIG 14.6-1 2 FIG 14.6-2 FIG 14.6-2 2 FIG 14.6-3 FIG 14.6-3 2 FIG 14.6-4 FIG 14.6-4 2 FIG 14.6-5 FIG 14.6-5 2 FIG 14.6-6 FIG 14.6-6 2 FIG 14.6-7 FIG 14.6-7 2 '
FIG 14.6-8 FIG 14.6-8 2 FIG 14.6-9 FIG 14.6-9 2 20
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V 14 . Safety and Accident Analysis FIG 14.6-10 FIG 14.6-10 2 FIG 14.6-11 FIG 14.6-11 2 FIG 14.6-12 FIG 14.6-12 2 FIG 14.6-13 FIG 14.6-13 2 FIG 14.6-14 FIG 14.6-14 2 FIG 14.6-15 FIG 14.6-15 2 FIG 14.6-16 FIG 14.6-16 2 FIG 14.6-17 FIG 14.6-17 2 FIG 14.6-18 FIG 14.6-18 2 FIG 14.6-19 FIG 14.6-19 2 FIG 14.6-20 2 FIG 14.6-21 FIG 14.6-21 2 FIG 14.6-22 FIG 14.6-22 2 FIG 14.6-23 FIG 14.6-23 2 FIG 14.6-24 FIG 14.6-24 2 FIG 14.6-25 FIG 14.6-25 2 FIG 14.6-26 FIG 14.6-26 2 FIG 14.6-27 FIG 14.6-27 2 FIG 14.6-28 FIG 14.6-28 2 FIG 14.6-29 FIG 14.6-29 2 FIG 14.6-30 FIG 14.6-30 2 FIG 14.6-31 FIG 14.6-31 2 FIG 14.6-32 FIG 14.6-32 2 t
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V 14. Saf ety and Accident-
^ t 'Analysin 3 FIG 14.6-33 -
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14A. Offsite Dose Calculations for- .
High Burnup Fuel
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22