ML20058A246
| ML20058A246 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 11/01/1978 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20058A248 | List: |
| References | |
| NUDOCS 7811170259 | |
| Download: ML20058A246 (18) | |
Text
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D UNITED STATES
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NUCLEAR REGULATORY COMMISSION n
h WASHING TON. D. C. 20555
% 'tkb
%, '..... s#
NORTHERN STATES POWER CCMPANY DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amencnent No. 32 License No. OPR-42 1.
The Nuclear Regulatory Connission (the Conmission) has found that:
A.
The applications for amendment by Northern States Power Company (the licensee) datec December 9,1577 and September 8,1978, comply with the standards and recuirements of the Atomic Energy Act of 1954, as amended (the Act), and the Connission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Connission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conductec without endangering the health and safety of the public, and (ii) tnat such activities will be conducted in compliance with the Connission's regulations; D.
The issuance of this anendment will no: be inimical to the cccree defense and security or to the health and safety of the auolic; and E.
Tne issuance of this anenement is in accordance witn 10 CFR Part 51 of tne Connission's regulations and all applicable requirements have been satisfied.
7811170269'
V m 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license anendment, and paragraph 2.C(2) of Facility License No. DPR-42 is hereby amended to read as follows:
(2) Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 32, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendnent is effective as of the cate of its issuance.
FCR THE NUCLCAR REGULATCRY C0f MISSION i
,s A (UW $&
- /_
A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: November 1,1973
/
4 UNITED STATES D~/.'t
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NUCLEAR REGULATORY COMMISSION i
WASHINGTON, D. C. 20555
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NCRTHERN STATES P0',4ER COMPANY DOCKET 110. 50-306 PRAIRIE ISLAND HUCLEAD GENERATING PLANT UNIT NO. 2 AldEH0f1ENT TO FACILITY OPERATING LICENSE Amendment No. 26 License No. DPR-60 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The applications for amendment by Northern States Power Company (the licensee) dated December 9,1977 and September 8,1978, comply with the standards and requirements of the Atoric Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the Sublic, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defente and security or to the health and safety of the public; 590 E.
The issuance of t9t s amencment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
J
_ 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amenenent, anc paragraph 2.C(2) of Facility License No. DPR-60 is hereby amended to read as follows:
(2) Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amencment No. 26, are hereby incorporated in the license. The licensee shall operate the f3cility in accordance with the Technical Specifications.
3.
Tnis license amencment is effective as of the date of its issuance.
FCR THE NUCLEAR REGULATORY COMMISSION
,b,
&N$Q A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: November 1, 1978
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ATTACHMENT TO LICENSE AMENDMENT HOS. 32 AND 26 FACILITY OPERATING LICENSE NOS. DPR 42 AND DPR-60 DOCKET NOS. 50-282 AND 50-306 Replace the folicwing pages of the Technical Specifications contained in Appendix A of the acove-indicated if censes with the attached pages bearing the same nunbers, except as otherwise indicated. The changed areas on the revised pages are reflected by a marginal line.
Remove Insert ii 11 iii iii iv iv TABLE TS.3.1-1 TABLE TS.3.1-1 TABLE TS.3.1-2 TABLE TS. 3.1-2 FIGURE TS.3.1-1 FIGURE TS.3.1-1 FIGURE TS.3.1-2 FIGURE TS.3.1-2 FIGURE TS.3.1-3 FIGURE TS.3.1-3 FIGURE TS.3.1-4 FIGURE TS.3.1-4 TS 3.10-4 TS 3.10-4 TS 3.10-5 TS 3.10-5 TS 5.3-1 TS 5.3-1
TS-il APPENDIX A TECHNICAL SPECIFICATIONS TA3LE OF CONTENTS (Continued)
TS SECTION TITLE PAGE 5.0 Desir_n Features TS.5.1-1 5.1 Site TS.S.1-1 5.2 containment System TS.5.2-1 5.3 Reactor TS.5.3-1 5.4 Engineered Safety Features TS.5.4-1 5.5 Radioactive Waste System TS.5.5-1 5.6 Fuel Handling IS.5.6-1 6.0 Adninistrative Controls TS.6.1-1 6.1 Organization TS.6.1-1 6.2 Review and Audit IS.6.2-1 63 Special Inspections and Audits TS.6.3-1 6.4 Safety Limit Violation IS.6.4-1 l
6.5 Plant Operating Procedures TS.6.5-1 6.6 Pla nt Operating Records TS.6.6-1 6.7 Reporting Requirements TS.6.7-1 Amendment No.
32 1 26
v TS-iii APPENDIX A TECHNICAL SPECIFICATIONS LIST OF TABLES TS TABLE TITLE 3 1-1 Unit 1 Reactor Vessel Toughness Data 3 1-2 Unit 2 Reactor Vessel Toughness Data T.5-1 Engineered Safety Features Initiation Instrucent Limiting Set Points 3.5-2 Instru=ent Operating Conditions for Reactor Trip 3.5-3 Instru=ent Operating Conditions for Emergency Cooling System 3.5
- Instrument Operacing Conditions for Isolation Functions 3.5-5 Ins trucent operating Conditions for Ventilation Systems 3.9-1 Radioactive Liquid Was te Sampling and Analysis 3.9-2 Radioactive Gaseous Waste Sampling and Analysis 3 12-1 Safety Related Shock Suppressors (Snubbe rs )
3.14-1 Safety Related Fire Detection Instruments 4.1-1 Minimum Frequencies for Checks, Calibrations and Test of Ins trume nt Channels 4.1-2A Minicus Frequencies for Equipment Tests 4.1-23 Minimum Frequencies for Sampling Tests 4.2-1 Reactor Coolant System In-Service Inspection Schedule Section 1.0 - Reactor Vessel Section 2.0 - Pressurizer Section 3.0 - Steam Generators and Class A Heat Exchangers Section 4.0 - Piping Systems Section 5.0 - Reactor Coolant Pumps Section 6.0 - Valves 4.2-2 System Boundaries for Piping Requiring Volumetric Inspection Under Exa=ination Category IS-251 J-l 4.2-3 System Soundaries Extending Beyond Those of Table TS.4.2-2 for Piping Requiring Surface Inspection Under Examination Category 15-251 J-1 4.2-4 System Boundaries Extending Beyond Those of Tables TS.4.2-2 and -3 for Piping Excluded f rco Examination under IS-251 but Requiring Visual Inspection (Which need not Require Re= oval of Insulation) of all Welds during System Hydrostatic Test 4.4-1 Unit 1 and Unit 2 Fenetration Designation for Leakage Tests 4.10-1 Prairie Island Nuclear Generating Plant-Radiation Envirencental Monitoring Program Sa=ple Collection and Analysis Environmental Monitoring Program l
4.12-1 Steam Generator Tube Inspection l
5.5-1 Anticipa.ed Annual Release of Radioactive Material in Liquid Ef fluents Free Prairie Island Nuclear Generating Plant (Per Unit) 5.5-2 Anticipated Annual Release of Radioactive Nuclides in Gasecus Effluent Frec ?rairie Island Nuclear Generating Plant (Per Unit) 6.1-1 Minimus Shif t Crew Composition 6.7-1 Special Reports Amendment
.*10. 32 & 26
TS-iv APPENDIX A TECHNICAL SPECIFICATIONS LIST OF FIGURES TS FIGURE TITLI 2.1-1 Saf ety Limits, Reactor Core, Thermal and Hydraulic Two Loop Operation 3 1-1 Unit I and Unit 2 Reactor Coolant System Heatup Limitations 3 1-2 Unit I and Unit 2 Reactor Coolant Sys tem Cooldown Limitations 3.1-3 Ef fect of Fluence and Copper Content og Shift of RT Reactor Vessel Steels Exposed to 550 Tempe ratur e' DT N
3.1-4 Fast Neutron Fluence (E ) 1 !!eV) as a Function of Full Power Service Life 3 10-1 Required Shutdown Reactivity Vs Reactor Boron Concentration 3.10-2 Control Bank Insertion Limits 3.10-3 Insertion Limits 100 Step overlap with One Bottomed Rod 3 10-4 Inse rtion Limits 100 Step ove rlap with One Inoperable Rod 3.10-5 Hot Channel Factor Normalized Operating Envelope For F = 2.32 3.10-6 Fluk Dif ference Control Schematic 3.10-7 Rod Bow Penalty (R3P) Fraction Versus Region Average 3urnup 4.4-1 Shield Building Design In-Laakage Rate 4 10-1 Prairie Island Nuclear Generating Plant Radiation Environmental Monitoring Program (Sample Location Map) 4.10-2 Prairie Island Nuclear Generating Plant Radiation Environmental Monitoring Program (Sample Location Map) 6.1-1 NSF Corporate Organizational Relationship to On-site Opera ting Organization 6.1-2 Prairie Island Nuclear Generating Plant Functional Organization for On-site Operating Group Amendment No. 32 & 26
TS.3.1-4 B.
HEATUP AND COOLDOWN Specification:
1.
The Unit 1 end Unit 2 reactor coolant te:::penture cad pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures TS.3.1-1 and TS.3.1-2 for the first full power service period.
a.
Allowable ecmbinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown.
Limit lines for cooldown rates between those presented may be obtained by interpolation.
b.
Figures TS.3.1-1 and TS.3.1-2 define limits to assure prevention of non-ductile failure only.
For normal operation other inherent plant characteristics, e.g.,
pump heat addition and pressurizer heater capacity may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.
2.
The limit lines shown in Figures IS.3.1-1, TS.3.1-2 shall be recalculated periodically using methods discussed in the Basis section.
3.
The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the vessel is below 700F.
4.
Se p--esuri:er heatup rate shall not exceed ICCCF/hr and the pressurizer cocidevn rste shall not ex eed 2CCCF/hr.
The spray shall not be used if the tenpersture difference betweed the pressurizer and the spray fluid is greater than 32CcF.
Basis Se reactor coolant systen heatup and cocidevn rates in Figuras 3 31-1 and
~S.3 1-2 are applicable to both Unit 1 and Unit 2.
Ite cu: tes are based on Unit 1 to-@ss data ani are cense: tative for the Unit 2 vessel. Toughness data is included in Tables TS.3 1-1 and :s.3 1-2.
k en e.ent No. 32 1 26
Table TS 3.1-1 UtlIT N0. 1 REACTOR VESSEL TOUGilNESS DATA (UNIRRADIATED) i Transe ese *I I
50 ft Ib/35 nuls Material Cu P
NDTT Lateral E mpennon NDT Aserage Transverse *I HT I
Component Type
(%)
(%)
(*F)
Temp. (*F)
(*F)
Upper Shelf (ft Ib)
M 4NI 75 I#l Closure llead Donne A533 Gr. D. Cl. I 4
64 Ill Nl bI llead Flange A508 Q. 3 4
12 A
84 d
ICI fd Vessel Flange A508 O. 3 4
41 A
77.5 injection N.;t.
A508 Cl. 3 22 114hl 22 ("}
gy f2]
EI 5 EI 92 bl Inlet 4eul Outlet Nossie A SOH Cl. 3
+5 39 III 4 I'}
85 NI Upper Shell A508 Cl. 3 4
39 lbl inter. Shell A SOS Cl. 3 0 06 0 013
+14 14 14 143 lbl Lower Sheil A508 O. 3 0 07 0 014 4
45 4
134 bI 5
79 kl Trans. Ring A508 Q. 3
+5 63 Y
Uottom itead A533 Gr. B, Cl. I 4
57 1
-3 68.5 <I M
I2 I
lbl h
Welitment Wetil 0.13 0 017 0
10 0
78.5 IhI flAZ ilAZ 0
<.100 0
211 eH e.
54 women orienied noenwl to it.e manos wuebng duection ta liased on actuel teensweese data thsough she su.we.alance proy om j
c I metic e uung Pectswee.leneceasu e turwti." Section 5.12 of Sreederd Renew Pfan. NUREG.75/OST.
I'. h. f on tonfiuJenal dass Amendment NO. 32&2p
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4
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4
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i IABLE 'IS.3.1-2 i
UNrf NO. 2 REACTOR VESSEL TOUGliNESS DATA (UNIRRADIATED)
Transvesse 'I I
50 ft Ib/35 mits Mateeial Cu P
NDTT Lateral Empanseon NDT Average Transeerse *I RT I
Cornpoewn Type
(%)
(%)
("F )
Temp. (*F)
(*F)
Upper Shelf (f t Ib) 5 52ICI 5
G4 (Cl Cosuse flead Donw A533 Gr. B. O 1 liced Flange A508 Q. 3
-1 18
-31 edcl Vessel Flange abo 8 C. 3 22 18
-22 et1ICI M
loject on Nonles ASou Q. 3
-22 114 22 9dCI ICI Irdet and Out!ct Nonle A508 Cl. 3
-13 50 ICI 10 89 l
ICI Upser Stielt A508 C. 3 13 41
-13 85ICI inter. Sheillbl A508 Q. 3 0.075 0 010
-4 SG
- 4 112 lbi Lower Sheti A508 c. 3 0.085 0.011 13 54 6
108 l
Trans. Hing A508 O. 3 10 50 10 76ICI Hottorn licad AS33 Gr. B. Cl.1
-13 56 4
6">ICI weidnienilb) weid 0.032 0 019
-31 G
-31 103 Q
lh) llAZ ItAZ h
31 35 31 117 w
- e. Speumen ossensed nunneet so t'he mesue wwuehmg dieect on
- p
- b. Besed on scives sese.sveese deve etwovgh the sueveesteme proviem b
c t.t.== trJ owy 'Psessuee-Temsweasues Lemets." Section 5.3.2 of Srsaders Aenew Men, NUREG-75/08F.
i l',M. f oes sontivJenal data.
Amendment No. 32 & 26 i
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(DISd)330SS38d031Ml0NI P
hendment No. 32 & 26
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(91Sd) 330SS3Hd 03J.nl0Ni Mendment No.32 & 26
FIGU?.E TS.3 1-3 400 0.30'. COPPE4 8ASE. 0.251
100 C
80 o[
60 E
g 0.15% COPPER SA$E. 0.10% WELD 40 o io, c0PPEt sASE. 0.05% wEl.3 20 I
I I
l I
I I
10 20 10'8 2
4 6
8 10
2 4
6 8 10 2
FLUENCE (N/CH > I NEV)
EFFECT OF FLUENCE AND COPPER CONTENT ON SHIFT OF RT FOR RUCTOR VESSEL NOT STEELS EXFOSED TO 550 F TEMPERATURE FIGURES.3.1-3 Arendment No. 32 & 25
[IGUF2TS.31-4 3
2 t/ar 10
8 6
37,7
'4 E
4 3
g 2
3 5
m x
88 2
10 5
a 2
6 4
7 2
87 10 0
5 10 15 20 25 30 SERVICE LIFE (EFFECTIVE FULL POWER YEARS)
FAST NEUTRON FLUENCE (E>1MeV) AS A FUNCTION OF FULL F0WER SERVICE LIFE FIGURE TS.3.1-4 Amendment No. 32 & 26
TS.3.10-4 2.
If tne percentage quandrant power tilt exceeds 2% but is less than 7 for a sustained period of more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or if such a tilt re curs intermittently, the reactor shall be brought to the hot shutdown condition. Subsequent operation below 50% of rating, for tes ting, shall be pe rmi t ted.
3.
Except for physics tes ts if the quadrant power tilt ratio exceeds L.0 7, tne reactor shall be brought to the hot shutdown condition.
Subsequent operation below 50% of rating, for tes ting, shall be permitted.
4.
If tne core is operating above 85 power with one excore nuclear cnanncl out of se rvice, then the core quadrant power balance shall be dete rmined daily and af ter a 10% power change using either 2 movable detectors or 4 core the rmocouples ' pe r q uad rant, per Specification 3 11.
D.
Rod Insertion Limits 1.
The shutdown rods shall be fully withdrawn when the reactor is critical or approaching criticality.
2.
Excep t during low power physics tes ting, operation with part length ro ds shall be restricted such that the part length rod bank is nog inserted in the reactor core at any time the reactor is critical.'
3.
When the reactor is critical or approaching criticality, the control banks shall be limited in physical insertion; insertion limits are shown in Figure TS.3.10-2, 3 and 4 for normal and abnormal operating conditions.
4.
Control bank insertion may be furthe r res tricted by Specification 3.10.A if, (1) the =easured control rod worth of all rods, less the worth of the worst stuck rod, is less than 5.52% reactivity at the ceginning of the firs t cycle or the equivalent value if measured at any othe r ti=e, or ( 2) if a rod is inoperable (Specification 3.10.G).
5.
Insertion limits do not apply during physics tests or during periodic exe rcise of individual rods. The shutdown =argin shown in Figure TS.3.10-1 must be maintained except for the icw pcwer =argin.
Fo r this te s t the reactor =ay be critical with all but one high worth full-length control rod inserted and all part-length rods fully withdrawn for a period not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per year provided a rod drop test is run on the high worch full-length red prior to this 3
particular low power physics test This limiting condition for operation as pertains to the part le ng t h rods is no longe r applicable for a particular Pral:1e Island unit f o llowing removal of the part length rods from that unit.
3 See footnote 2.
Amendment No. 32 3 26
TS.3.10-5 E.
Rod Misalignment Limitations 1.
If a f ull-length or part-length rod cluster control (RCC) assembly is misaligned f rom its bank by more than 15 inches, the rod will be realigned or the core power peaking f actors shall be determined within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and Specification 3 10.3 applied.
If peaking f actors a re no t determined within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the high neutron flux trip setpoint shall be reduced to 85 percent of rating.'
l 2.
If the misaligned rod cluster control is not rt.ligned within a total of B hours, the rod shall be declared inoperable.
F.
Inoperable Rod Position Indicator Channels 1.
If a rod position indicator (RPI) channel is out of service then For operation between 50% and 100% of rating, the position of a.
the RCC shall be checked directly by core ins truce ntation (excore detector and/or ther= occupies and/or covable incore de tec to rs) eve ry shif t or subsequent to rod motion exceeding a total of 24 steps, whichever occurs firs t.
b.
During operation below 50% of rating, no special monitoring is required.
2.
The plant shall be brought to the hot shutdown condition should core tnan one RPI channel per sub-bank or more than two RPI channels per bank be found to be incperable during power operation.
3.
If a full length or part length rod having a rod position it.dicator channel out of service is found to be misali d from 1.a. above, then Specification 3.10 E. will be applied.' gne l
G.
Inope rable Rod Lisitations 1.
An inoperable rod is a rod vnich (a) does not trip, (b) is decla red inoperable under Specification 3 10 E. or 3.10 H. or (c) cannot be moved by its drive =echaniss and cannot be corrected within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
' See f ootnot e 2, page TS.3.10-4.
5 See footnote 2, page TS.3.10-4.
I Amendment No. 32 & 25 i
i
TS.5.3-1 53 REACTOR
~
A.
Reactor Core m.
1 The reactor core contains approximately 48 metric tons of uranium in the form of sligntly enriched uranium dioxide pellets. The pellets are encapsulated in Zircaloy-4 tubing to form fuel rods.
The reactor core is made up of lgf uel assemblies. Each fuel
.ssembly contains 179 fuel rods.
2.
The ave rage enrichment of the initial core is a nominal 2 90 weight per cent of U-235.
Three fuel enrichments are used in the enrichment is a nominal 3.40 weight per initial core.
of U-2 3!. ge highes t cent 3.
Reload f uel will be similar in design to the initial core.
4.
Bu,rnable poison rods are incorporated in the initial core.
There are 704 poison rods in the form of 8, 12, and 16-rod cluste g whica are located in vacant rod cluster control guide tubes.
The burnable po clad with stainless steel.gn rods consist of borosilicate glass 5.
In the reactor core, there are 29 full-length RCC assemblies that 142-inglength of silver-indium-cadmium alloy clad with centain a stainless steel 6.
Up to 10 grams of enriched fissionable material may either be used in the core or be available on the plant site in the focu of fabricated neutron flux detectors for the purposes of monitoring core neutron flux.
S.
Reactor Coolant System The design of the reactor coo system complies with all 1
applicablecoderequirements.{ ant Amendment No. 32 & 26