ML20212R632
| ML20212R632 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 04/13/1987 |
| From: | NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20212R565 | List: |
| References | |
| TAC-65078, TAC-65079, NUDOCS 8704270287 | |
| Download: ML20212R632 (12) | |
Text
{{#Wiki_filter:3 Exhibit B Prairie Island Nuclear Generating Plant License Amendment Request Dated April 13, 1987 Proposed Changes Marked Up on Existing Technical Specification Pages Exhibit B consists of the existing Technical Specification pages with the proposed changes written on those pages. Existing pages affected by this change are listed below: TS-x TS.2.1-2 TS.3.10-1 TS.3.10-2 Figure TS.3.10-8 9 i F 8704270287 870413 PDR ADOCK 0500 2 P
TS-x REV00 11/l'/00 APPENDIX A TECHNICAL SPECIFICATIONS LIST OF FIGURES TS FIGURE TITLE 2.1-1 Safety 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 System Cooldown Limitations 3.1-3 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit l Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity >1.0 pCi/ gram DOSE EQUIVALENT I-131 3.9-1 Prairie Island Nuclear Generating Plant Site Boundary for Liquid Effluents 3.9-2 Prairie Island Nuclear Generating Plant Site Boundary for Gaseous Effluents ~ 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 Insertion Limits 100 Step overlap with one Inoperable Rod 3.10-5 Hot Channel Factor Normalized Operating Envelope ,g 3.10-6 Deviation from Target Flux Difference as a Function of Thermal Power V(Z) as a Function of Core Height ( 3.10-7 4.4-1 Shield Building Design In-Leakage Rate 6.1-1 NSP Corporation Organization Relationship to On-Site Operating Organizations 6.1-2 Prairie Island Nuclear Generating Plant Functional Organization for on-site Operating Group & ffia) e.,d Fsg ( 6 ) kbl f 3.Io-e Any ,e.
o' ) TS.2.1-2 REV 77 4/3/Sf The solid curves of Figure TS 2.1-1 represent the loci of points of thermal power, coolant pressure, and coolant average temperature for which either the coolant enthalpy at the core exit is limiting or the DNB ratio is lhniting. For the 1685 psig and 1985 psig curves, the coolant average enthalpy at the core exit is equal to saturated water enthalpy below power levels of 91% and 74 respectively..For the 2235 psig and 2385 psis curves, the coolant average temperature at the core exit is equal to 650*F below power levels of 64% and 73 respectively. For all four curves, the DNBR is limiting at higher power l levels. The area of safe operation is below these curves. The plant conditions required to violate the limits in the lower power range are precluded by the self-actuated safety valves on the steam generators. The highest nominal setting of the steam generator safety valves is 1129 psig (saturation temperature'560*F). At zero power the difference between primary coolant and secondary coolant is zero and at full power it is 50*F. The reactor conditions at which steam generator safety valves open is shown as a dashed line on Figure TS.2.1-1. Except for special tests, power operation with only one loop or with natural circulation is not allowed. Safety limits for such special tests will be deter =ined as a part of the test procedure. The curves are conservative for the following nuclear hot channel factors: t I T = -h66- [1 + 0.3(1-?)] ; and F --2r30-g, &/.70 q Use of these factors results in more conservative safety limits than would result from power distribution limits in Specification TS.3.10. This combination of hot channel factors is higher than that calculated full power for the range from all control rods fully withdra'vn to at max 1=um allowable control red insertion. The control rod insertion li=its are covered by Specification 3.10. Adverse power distribution factors could occur at lower power levels because additional control rods are in the core. However, the control rod insertion limits specified by Figure TS.3.10-1 assure that the DNB ratio is always greater at part power than at full power. The Reactor Control and Protective System is designed to prevent any l anticipated combination of transient conditions that would result in j a DNB ratio of less than 1.30 for Exxon Nuclear fuel and less than 1.17 for Westinghouse fuel, i l t i k l
6 4 i TS.3.10-1 .REV " ':/2/"6 I 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS l Applicability Applies to the limits on core fission power distribution and to the limits on control rod operations. Objective To assure 1) core suberiticality af ter reactor trip, 2) acceptable core power distributions during power operation, and 3) limited potential reactivity insertions caused by hypothetical control rod ejection. ? l l Specification A. Shutdown Reactivity-The shutdown margin with allowance for a stuck control rod assembly shall exceed the applicable value shown in Figure TS.3.10-1 under all steady-state operating conditions, except for physics tests, from zero i to full power, including effects of axial power distribution. The .i shutdown margin as used here is defined as the amount by which.the reactor core would be suberitical at hot shutdown conditions if all control rod assemblies were tripped, assuming that the highest worth l control rod assembly remained fully withdrawn, and assuming no changes s.c in xenon or boron concentration. i B. Power Distribution Limits 1. At all times, except dgring lgv power physics testing, measured l hot channel factors, F and FaH, as defined below and in the n bases, shall meet the following limits: [ Fa(FM[E) N( l x 1.03 x 1.05 5(2.00/r)* x "(:) h40-x (1+ 0. 3 (1-P) ] **-- w T^'H x 1.04 ( Fu(F) q where the following definitions apply: j - K(Z) is the axial dependence function shown in Figure TS.3.10-5. i - Z is the core height location. i - P is the fraction of rated power at which the core is N operating. In the r limit determination when P 1.50, 0 ,. osa E Fo (FanV amel En Tope 64.= r. - ( 5, ( FeQ a, a,e, %' p, gG h,.io -{21/{ c.;h[;;id f5
- [
- ;;;de;d;;;;;i11:: f:: ";i; 2
-8 wy uw av. t 1.03 a [1 + 0.2 (1-7)} ; hell b; u;;d f;; ";;;i;;h;;;; Stend;;d ;;;;;ilie; l I fe; ";it 2 0 :1 10. 7 1
~ ~- . -. 1 .e .3 ) TS.3.10-2 REV 77 1/3/00 or [ is defined as the measured 7 or T respectively, l - [kth the smallest margin or greatest ekcass N limit. w 4 4 -1.03istheengineering-hotchannelfactor,[ha,nce.liedtothe app I measured T to account for manufacturing tole - 1.05 is applied to the measured [q to account for measurement I j uncertainty. - 1.04 is applied to the measured N ** *** ""* I # "***"****** I H j uncertainty. Hotchannelfactors,[hned,atequilibriumconditionsaccordingandTfH 2. flux difference determ to the following conditions, whichever occurs first: (a) At least once per 31 effective full-power days in conjunction with the target flux difference determination, or I (b) Upon reaching equilibrium conditions af ter exceeding the reactor power at which target flux differance was last. f deter =ined, by 10% or more of rated power. [hthecore: (equil) shall neet the following limit for the middle axial 80: c N (equil) x V(Z) x 1.03 x 1.05 i'.30/7) x U :) 796a* where V(2) is defined Tigure 3.10-7 and other terms are l 1 defined in 3.10.B.1 above. i ) 3. (a) If either measured hot channel factor exceeds its li=it specified in 3.10.B.1, reduce reactor power and the high i neutron flux trip setpoint by 1% for each percent that the measuredT'kthe3.10.3.1 limit. ' or by 3.33% for each percent that the measured T*,'H exceed Then follow 3.10.B.3(c). 4 j (b) If the measured T' (equil) exceeds the 3.10.B.2 linics but not l the 3.10.3.1 lini., take one of the following actions: I l 1. k'ithin 48 hours place the reactor in an equilibrium configuration for which Specification 3.10.B.2 is satis-fied, or I i 2. Reduce reactor power and the high neutron flux trip j satpoint by 1 for each percent that the asasured T{(equil)x1.03x1.05xV(Z)exceedsthe11=1t. l Y i f l ,----.-.-...--,_-.--..-,,_,.,_-.-.-,----..--,...,.,.-.-n
.6 FIGURE TS.3.10-8 REV f///////# 2.4 0.. ~ ~
- ~
l i l(2.4.1.6) [ Operation NOT J Allow ed 2.35- -~~~~ ~ ~~~ h~~ ~ ~ ~ i i e I r L .i O " ~~ '~~ - ~- "- ~~~~j: ~ LL. 2.30- ~ - ~- l Operation i Allowed t [(2.32.1.86) i i 2.25- -~-~~-~~~~~: ~ ~ - ~ ~ ~ ~ ~ ~ - + ~ ~. . ~ ~ I i i i. i 2.20, ) i 1.55 1.80 1.65 1.70 ^ F3g(F ) o l FIGURE 3.10-8 Acceptable Values of F (Fg) and Fg(F ) q q w i i e
r Exhibit C Prairie Island Nuclear Generating Plant License Amendment Request Dated April 13.-1987 Revised Technical Specification Pages Exhibit C consists of the proposed Technical Specification pages with the changes shown in Exhibit B incorporated. The proposed pages are. listed below: TS-x TS.2.1-2 TS.3.10-1 TS.3.10-2 Figure TS.3.10-8 l
I' I TS-x REV 1 APPENDIX A TECHNICAL SPECIFICATIONS l LIST OF FIGURES-TS FIGURE TITLE 2.1-1 Safety Limits, Reactor Core, Thermal and Hydraulic Two Loop Operation 1 3.1-1 Unit I and Unit 2 Reactor Coolant System Heatup Limitations 3.1-2 Unit I and Unit 2 Reactor Coolant System Cooldown Limitations 1 3.1-3 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant i Specific Activity >l.0 uCi/ gram DOSE EQUIVALENT I-131 j 3.9-1 Prairie Island Nuclear Generating Plant Site Boundary for Liquid j Effluents 3.9-2 Prairie Island Nuclear Generating Plant Site Boundary for Gaseous Effluents ] 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 Insertion Limits 100 Step overlap with One Inoperable Rod i 3.10-5 Hot Channel Factor Normalized Operating Envelope 3.10-6 Deviation from Target Flux Difference as a Function of. Thermal j Power ) 3.10-7 V(Z) as a Function of Core Height j 3.10-8 Acceptable Values of F (F n).and F3p(F ) l q a q 4.4-1 Shield Building Design In-Ieakage Kite 6.1-1 NSP Corporate Organizational Relationship to On-Site Operating i Organizations l 6.1-2 Prairie Island Nuclear Generating Plant Functional Organization j for On-site Operating Group t i l i i i i i I i ~.. -.. -.. -,.. - -
TS.2.1-2 REV The solid curves of Figure TS.2.1-1 represent the loci of points of thermal power, coolant pressure, and coolant average temperature for which either the coolant enthalpy at. the core exit is limiting or the DNB ratio is limiting. For the 1685 psig and 1985 psig curves, the coolant average enthalpy at the core exit is equal to saturated water enthalpy below power levels of 91% and 74% respectively. For the 2235 psig and 2385 psig curves, the coolant average temperature at the core exit is equal to 650*F below power levels of 64% and 73% respectively. For all four curves, the DNBR is limitinglat higher power levels. The area of safe operation is below these curves. The plant conditions required to violate the limits in the lower power range are precluded by the self-actuated safety valves on the steam generators. The highest nominal setting of the steam generator safety valves is 1129 psig (saturation temperature 560*F). At zero power the difference between primary coolant and secondary coolant is zero and at full power it is 50*F. The reactor conditions at which steam generator safety valves open is shown as a dashed line on Figure TS.2.1-1. Except for special tests, power operation with only one loop or with natural circulation is not allowed. Safety limits for such special tests will be determined as a par't of the test procedure. The curves are conservative for the following nuclear hot channel factors: F = 1.70 [1 + 0.3(1-P)] ; and F = 2.50 l aH q Use of these factors results in more conservative safety limits than would result from power distribution limits in Specification TS.3.10. This combination of hot channel factors is higher than that calculated at fu11 power for the range from all control rods fully withdrawn to maximum allowable control rod insertion. The control rod insertion limits are covered by Specification 3.10. Adverse power distribution factors could occur at lower power levels because additional control rods are in the core. However, the control rod insertion limits specified by Figure TS.3.10-1 assure that the DNB ratio is always greater at part power than at full power. The Reactor Control and Protective System is designed to prevent any anticipated combination of transient conditions that would result in a DNB ratio of less than 1.30 for Exxon Nuclear fuel and less than 1.17 for Westinghouse fuel.
TS.3.10-1 REV 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Applicability Applies to the limits on core fission power distribution and to the limits on control rod operations. Objectives t j To assure 1) core suberiticality af ter reactor trip, 2) acceptable core power distributions during power operation, and.3) limited potential reactivity insertions caused by hypothetical control rod ejection. Specification A. Shutdown Reactivity The shutdown margin with allowance for a stuck control rod assembly shall exceed the applicable value shown in Figure TS.3.10-1 under all steady-state operating conditions, except for physics tests, from zero to full power, including effects of axial power distribution. The shutdown margin as used here is defined as the amount by which the reactor core would be subcritical at hot shutdown conditions if all control rod assemblies were tripped, assuming that the highest worth control rod assembly remained fully withdrawn, and assuming no changes in xenon or boron concentration. B. Power Distribution Limits 1. At all times, except dgring log power physics testing, measured i hot channel factors, F and FAH, as defined below and in the bases,shallmeetthe9o11owinglimits: F x 1.03 x 1.05 3[F (F g)/P]K(Z) q H(F ) x [1 + 0.3(1-P)] x 1.04 $F H q where the following definitions apply: - K(Z) is the axial dependence function shown in Figure TS.3.10-5. - Z is the core height location. - P is the fraction of rated power at which the core is N operating. In the F limit determination when P 3 50, set q P = 0.50. - The F function [F FAH)] and FAH unction [FAH( Q m to are show in Fi ure TS.3.10-
o-O. TS.3.10-2 I REV or (H is defined as the measured F or F re8Pectively, -FwkththesmallestmarginorgreatestekcessoYlimit. A - 1.03 is thg engineering hot channel factor, F s 8PP11ed'to the l 0 measured F to account for manufacturing tolerance. q 1.05 is applied to the measured F to account for measurement ~ q uncertainty. 1.04 is applied to the measured F t account for measurement H l uncertainty. 2. Hot channel factors, F and FAH, shall be measured and the target fluxdifferencedetermkned,atequilibriumconditionsaccording to the following conditions, whichever occurs first: (a) At least once per' 31 effective full-power days in conjunction with the target flux difference determination, or (b) Upon reaching equilibrium conditions after exceeding the I reactor power at which target flux difference was last determined, by 10% or more of rated power. F (equil) shall meet the following limit for the middle axial 80% o9thecore: q (equil) x V(Z) x 1.03 x 1.05 $[F (Fq 3g)/P] x K(Z) F where V(Z) is defined Figure 3.10-7 and other ttras are defined in 3.10.B.1 above. 3. (a) If either measured hot channel factor exceeds its limit specified in 3.10.B.1, reduce reactor power and the high neutron flgx trip setpoint by 1% for each percent that the mgasured F or by 3.33% for each percent that the measured F exceed the 3.10.B.1 limit. Then follow 3.10.B.3(c). oH (b) If the measured F the3.10.B.1limi9,(equil)exceedsthe3.10.B.2limitsbutnot take one of the following actions: 1. Within 43 hours place the reactor in an equilibrium configuration for which Specification 3.10.B.2 is satis-fled, or 2. Reduce reactor power and the high neutron flux trip s tpoint by 1% for each percent that the measured F (equil) x 1.03 x 1.05 x V(Z) exceeds the limit.
FIGURE TS.3.10-8 REV f///////# 2.40 ~ l(2.4.1.6) i Operation NOT j Allowed 2.35- ~ ' - ~ " ~ ~ ~ " " + 2- <2 IL w O lL 2.30- - """ + Operation l t Allowed l(2.32.1.88) i 2.25- ~ ~ - ""- " ~ ~ - - - - ~ ~ " " " + " - 2.20, ll 1.55 1.60 1.65 1.70 F3g(F ) o 1 I FIGURE 3.10-8 Acceptable Values of F (Fgg) and F6H(F ) q q t ~,-w.,-.,v.- ..a --}}