ML103120708

From kanterella
Revision as of 06:27, 13 November 2019 by StriderTol (talk | contribs) (Created page by program invented by StriderTol)
Jump to navigation Jump to search
Cycle 22 Core Operating Limits Report, Revision 1
ML103120708
Person / Time
Site: North Anna Dominion icon.png
Issue date: 11/08/2010
From: Funderburk C
Dominion, Dominion Resources Services, Virginia Electric & Power Co (VEPCO)
To:
Office of Nuclear Reactor Regulation, Document Control Desk
References
10-662
Download: ML103120708 (23)
v • d • e


Text

Dominion Resonrces Services, Inc.

Innsbrook Technical Center 5000 Do min ion Boulevard , 25E, Glen Allen, VA 23060 November 8, 2010 U. S. Nuclear Regulatory Commission Serial No.10-662 Attention : Document Control Desk NLOS/ETS One White Flint North Docket No. 50-338 11555 Rockville Pike License No. NPF-4 Rockville, MD 20852-2738 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

NORTH ANNA POWER STATION UNIT 1 CYCLE 22 CORE OPERATING LIMITS REPORT, REVISION 1 Pursuant to North Anna Technical Specification 5.6.5.d , attached is a copy of the Dominion Core Operating Limits Report for North Anna Unit 1 Cycle 22, Pattern PEZ, Revision 1.

If you have any questions regarding this submittal, please contact Mr. Thomas Shaub at (804) 273-2763.

Sincerely, C. L. Funderburk, Director

-

Nuclear Licensing and Operations Support Dominion Resources Services , Inc.

for Virginia Electric and Power Company

Attachment:

CORE OPERATING LIMITS REPORT, North Anna 1 Cycle 22 Pattern PEZ, Revision 1 Commitments made in this letter: None

Serial No.10-662 Docket No. 50-338 Cycle 22 Pattern PEZ COLR Page 2 of 2 cc: U.S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Ave., NE, Suite 1200 Atlanta, Georgia 30303-1257 Mr. J. E. Reasor , Jr.

Old Dominion Electric Cooperative Innsbrook Corporate Center 4201 Dominion Blvd.

Suite 300 Glen Allen, Virginia 23060 NRC Senior Resident Inspector North Anna Power Station Dr. V. Sreenivas NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 G-9A 11555 Rockville Pike Rockville , Maryland 20852-2738 MS.K. R.Cotton NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 G-9A 11555 Rockville Pike Rockville, Maryland 20852-2738

Serial No.10-662 Docket No. 50-338 ATTACHMENT CORE OPERATING LIMITS REPORT FOR NORTH ANNA UNIT 1 CYCLE 22 PATTERN PEZ, REVISION 1 NORTH ANNA POWER STATION VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

Page 1 of 21

Serial No.10-662 Docket No. 50-338 N1C22 CORE OPERATING LIMITS REPORT INTRODUCTION The Core Operating Limits Report (COLR) for North Anna Unit 1 Cycle 22 has been prepared in accordance with North Anna Technical Specification 5.6.5. The Technical Specifications affected by this report are listed below:

TS 2.1.1 Reactor Core Safety Limits TS 3.1.1 Shutdown Margin (SDM)

TS 3.1.3 Moderator Temperature Coefficient (MTC)

TS 3.1.4 Rod Group Alignment Limits TS 3.1.5 Shutdown Bank: Insertion Limit TS 3.1.6 Control Bank Insertion Limits TS 3.1.9 Physics Test Exceptions-Mode 2 TS 3.2.1 Heat Flux Hot Channel Factor TS 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (FN,ill)

TS 3.2.3 Axial Flux Difference (AFD)

TS 3.3.1 Reactor Trip System (RTS) Instrumentation TS 3.4.1 RCS Pressure, Temperature, and Flow DNB Limits TS 3.5.6 Boron Injection Tank (BIT)

TS 3.9.1 Boron Concentration In addition, a technical requirement (TR) in the NAPS Technical Requirements Manual (TRM) refers to the COLR: '

TR 3.1.1 Boration Flow Paths - Operating The analytical methods used to determine the core operating limits are those previously approved by the NRC and discussed in the documents listed in the References Section.

Cycle-specific values are presented in bold. Text in italics is provided for information only.

Page 2 of21

Serial No.10-662 Docket No. 50-338 REFERENCES

1. VEP-FRD-42 Rev 2.1-A, Reload Nuclear Design Methodology, August 2003.

(Methodology for TS 3.1.1- Shutdown Margin, TS 3.1.3 - Moderator Temperature Coefficient, TS 3.1.5 - Shutdown Bank Insertion Limit, TS 3.1.4 - Rod Group Alignment Limits, TS 3.1.6 - Control Bank Insertion Limits, TS 3.1.9 - Physics Test Exceptions-Mode 2, TS 3.2.1 - Heat Flux Hot Channel Factor, TS 3.2.2 - Nuclear Enthalpy Rise Hot Channel Factor, TS 3.5.6 - Boron Injection Tank (BIT) and TS 3.9.1-Boron Concentration)

2. VEP-NE-2-A, Rev. 0, Statistical DNBR Evaluation Methodology, June 1987.

(Methodology for TS 3.2.2 - Nuclear Enthalpy Rise Hot Channel Factor and TS 3.4.1-RCS Pressure, Temperature and Flow DNB Limits)

3. VEP-NE-l- Rev. O.I-A, Relaxed Power Distribution Control Methodology and Associated FQ Surveillance Technical Specifications, August 2003.

(Methodology for TS 3.2.1 - Heat Flux Hot Channel Factor and TS 3.2.3 - Axial Flux Difference)

4. WCAP-8745-P-A, Design Bases for the Thermal Overpower ~T and Thermal Overtemperature ~T Trip Functions, September 1986.

(Methodology for TS 2.1.1 - Reactor Core Safety Limits and TS 3.3.1 - Reactor Trip System Instrumentation)

5. WCAP-I4483-A, Generic Methodology for Expanded Core Operating Limits Report, January 1999.

(Methodology for TS 2.1.1 - Reactor Core Safety Limits, TS 3.1.1 - Shutdown Margin, TS 3.1.4 - Rod Group Alignment Limits, TS 3.1.9 - Physics Test Exceptions-Mode 2, TS 3.3.1 - Reactor Trip System Instrumentation, TS 3.4.1 - RCS Pressure, Temperature, and Flow DNB Limits, TS 3.5.6 - Boron Injection Tank (BIT) and TS 3.9.1- Boron Concentration)

6. BAW-I0227P-A, Rev. 0, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel."

(Methodology for TS 2.1.1- Reactor Core Safety Limits, TS 3.2.1 - Heat Flux Hot Channel Factor)

7. EMF-2103 (P) (A), Rev. 0, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," April 2003.

Page 3 of21

- - - - - -- - - - - - -- - - - - - - - - - - - - - -- - -- - - - - -- - -- - -

Serial No.10-662 Docket No. 50-338 (Methodology for TS 3.2.1 - Heat Flux Hot Channel Factor)

8. EMF-96-029 (P) (A), Rev. 0 "Reactor Analysis System for PWRs," January 1997.

(Methodology for TS 3.2.1 - Heat Flux Hot Channel Factor)

9. BAW-10168P-A, Rev. 3, "RSG LOCA - BWNT Loss-of-Coolant Accident Evaluation Model for Recirculating Stearn Generator Plant s," December 1996.

Volume II only (SBLOCA models).

(Methodology for TS 3.2.1 - Heat Flux Hot Channel Factor)

10. DOM-NAF-2, Rev. 0.2-P-A , "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," including Appendix A, "Qualification of the F-ANP BWU CHF Correlations in the VIPRE-D Computer Code," August 2010.

(Methodology for TS 3.2.2 - Nuclear Enthalpy Rise Hot Channel Factor and TS 3.4.1 -

RCS Pressure, Temperature and Flow DNB Limits)

Page 4 of21

Serial No.10-662 Docket No. 50-338 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in COLR Figure 2.1-1; and the following SLs shall not be exceeded.

2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained greater than or equal to the 95/95 DNBR criterion for the DNB correlations and methodologies specified in the References Section.

2.1.1.2 The peak: fuel centerline temperature shall be maintained < 5173°F, decre asing by 65°F per 10,000 MWD/MTU of burnup.

Page 5 of2 1

Serial No.10-662 Docket No. 50-338 COLR Figure 2.1-1 NORTH ANNA REACTOR CORE SAFETY LIMITS 665 660 ~ __.............._

-""

._---_._..._--_.__........ .....

~

655 .........

650

............... ~ 1000..

............

~

psia 645 640

"" ~

r-;

............. ~o psia

...............

i"- ....

<, ......

635 i:i:'

............... ~

'"""" <,

"

~

~ 630 .........

.......

~ <,...... ....._-_........ _.........__.. ... ...-.....*.---- ~ \

I1l 625

<,

"

.............

OJ Co E 620 '- .... """"""" < ; 2000 psia \.\

~

OJ 615

<, --.......... ........ \' ~

~ .........

\

"

....

I1l OJ

'"""" < , <, 1860 psia <,

> 610

<<

G:i 605 .....

............... --; 1\ '

VI VI OJ

> 600

............

<,

~

r-, \

<, \\.

595 " ..........

590 ~ \ .._-

585 \

580 \,

575 570 o 10 20 30 40 50 60 70 80 90 100 110 120 Percent of RATED THERMAL POWER Page 6 of 21

- -- - - - - - - - - - - - -- - - -- ~

Serial No.10-662 Docket No. 50-338 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM)

LCO 3.1.1 SDM shall be ~ 1.77 % Llk/k.

3.1.3 Moderator Temperature Coefficient (MTC)

LCO 3.1.3 The MTC shall be maintained within the limits specified below. The upper limit of MTC is +0.6 x 10-4 ~k/k/oF, when < 70% RTP, and 0.0

~oF when ~ 70% RTP .

The BOC/ARO-MTC shall be ~ +0.6 x 10-4 ~oF (upper limit), when <

70% RTP, and ~ 0.0 ~oF when ~ 70% RTP.

The EOC/ARO/RTP-MTC shall be less negative than - 5.0 x 10-4 ~oF (lower limit).

The MTC surveillance limits are:

The 300 ppm/ARO/RTP-MTC should be less negative than or equal to

-4.0 x 10-4 ~oF [Note 2].

The 60 ppm/ARO/RTP-MTC should be less negative than or equal to

-4.7 x 10-4 ~oF [Note 3].

SR 3.1.3.2 Verify MTC is within -5.0 x 10-4 ~oF (lower limit).

Note 2: If the MTC is more negative than -4.0 x 10-4 ~k/k/oF, SR 3.1.3.2 shall be repeated once per 14 EFPD during the remainder of the fuel cycle.

Note 3: SR 3.1.3.2 need not be repeated if the MTC measured at the equivalent of equilibrium RTP-ARO boron concentration of ~ 60 ppm is less negative than -4.7 x 10-4 ~oF.

3.1.4 Rod Group Alignment Limits Required Action ALI Verify SDM to be ~ 1.77 % Llk/k.

Required Action B.1.1 Verify SDM to be ~ 1.77 % Llk/k.

Required Action D.1.1 Verify SDM to be ~ 1.77 % Llk/k.

Page 7 of21

Serial No.10-662 Docket No. 50-338 3.1.5 Shutdown Bank Insertion Limits LCO 3.1.5 Each shutdown bank shall be withdrawn to at least 228 steps.

Required Action A.1.1 Verify SDM to be ~ 1.77 % Ak/k.

Required Action B.1 Verify SDM to be ~ 1.77 % Ak/k.

SR 3.1.5.1 Verify each shutdown bank is withdrawn to at least 228 steps.

3.1.6 Control Bank Insertion Limits LCO 3.1.6 Control banks shall be limited in physical insertion as shown in eOLR Figure 3.1-1. Sequence of withdrawal shall be A, B, C and D, in that order; and the overlap limit during withdrawal shall be 100 steps.

Required Action A.1.1 Verify SDM to be ~ 1.77 % Ak/k.

I Required Action B.1.1 Verify SDM to be ~ 1.77 % Ak/k.

Required Action C.1 Verify SDM to be ~ 1.77 % Ak/k.

SR 3.1.6.1 Verify estimated critical control bank position is within the insertion limits specified in eOLR Figure 3.1-1.

SR 3.1.6.2 Verify each control bank is within the insertion limits specified in eOLR Figure 3.1-1.

SR 3.1.6.3 Verify each control bank not fully withdrawn from the core is within the sequence and overlap limits specified in LeO 3.1.6 above.

3.1.9 PHYSICS TESTS Exceptions - MODE 2 LCO 3.1.9.b SDM is ~ 1.77 % Ak/k.

SR 3.1.9.4 Verify SDM to be ~ 1.77 % Ak/k.

Page 8 of21

Serial No.10-662 Docket No. 50-338 COLR Figure 3.1-1 North Anna 1 Cycle 22 Control Rod Bank Insertion Limits 230

/ (0.5 ~9 , 228) 220 210 /

200 /

190 / (1. ,194) ;

Fully w d positi nn ::: 228 180

(-BANK

/ /

"'0

"'-

3:170 /  ;/

a.

~160

/ /

III

150 / ..V r
::.~

...

,2140 / /

'Vi / /

~130

§l20 / /

o

....

~110 (0, liE) /

"'0

~100 / D-BANK 90

/

80 /

70 V

60 /

50 V

40 /

30 /

20 /

10 /

o / (0.048,0) 0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 Fraction of Rated Thermal Power Page 9 of21

Serial No.10-662 Docket No. 50-338 3.2 POWER DISTRIBUTION LIMITS 3.2.1 Heat Flux Hot Channel Factor (FQ(Z))

LCO 3.2.1 FQ(Z), as approximated by FQM(Z), shall be within the limits specified below.

CFQ = 2.32 The Measured Heat Flux Hot Channel Factor, FQM(Z), shall be limited by the following relationships:

CFQ K(Z)

F M(Z):::;---- for P>0.5 Q P N(Z)

CFQ K(Z)

FM(Z):::;---- for P:::;0.5 Q 0.5 N(Z)

THERMAL POWER where: P-  ; md

- RATED THERMAL POWER K(Z) is provided in COLR Figure 3.2-1, N(Z) is a cycle-specific non-equilibrium multiplier on FQM(Z) to account for power distribution transients during normal operation, provided in COLR Table 3.2-1.

The discussion in the Bases Section B 3.2.1 for this LCO requires the application of a cycle dependent non-equilibrium multiplier, N(Z), to the measured peaking fa ctor, FQM(Z), befor e comparing it to the limit. N(Z) accounts for power distribution transients encountered during normal operation. As fun ction N(Z) is dependent on the predicted equilibrium FQ(Z) and is sensitive to the axial power distribution, it is typically generated from the actual EOC burnup distribution that can only be obtained after the shutdown of the previous cycle. The cycle-specific N(Z) fun ction is presented in COLRTable 3.2-1.

Page 10 of21

Serial No.10-662 Docket No. 50-338 COLR Table 3.2-1 NIC22 Normal Operation N(Z)

NODE HEIGHT o to 1000 1000 to 3000 3000 to 5000 5000 to 7000 7000 to 9000 9000 to 11000 (FEET) MWDIMTU MWDIMTU MWDIMTU MWDIMTU MWDIMTU MWDIMTU 10 10.2 1.109 1.116 1.115 1.139 1.139 1.123 11 10.0 1.116 1.117 1.114 1.138 1.138 1.121 12 9.8 1.122 1.121 1.116 1.136 1.136 1.122 13 9.6 1.127 1.126 1.121 1.135 1.135 1.128 14 9.4 1.128 1.127 1.122 1.132 1.132 1.129 15 9.2 1.129 1.129 1.126 1.134 1.135 1.135 16 9.0 1.138 1.140 1.138 1.142 1.150 1.151 17 8.8 1.146 1.150 1.150 1.154 1.167 1.167 18 8.6 1.150 1.154 1.155 1.159 1.173 1.173 19 8.4 1.150 1.155 1.155 1.162 1.176 1.176 20 8.2 1.151 1.156 1.156 1.167 1.182 1.182 21 8.0 1.150 1.155 1.155 1.169 1.185 1.185 22 7.8 1.149 1.154 1.154 1.169 1.186 1.186 23 7.6 1.145 1.154 1.154 1.170 1.184 1.184 24 7.4 1.139 1.156 1.156 1.174 1.182 1.182 25 7.2 1.135 1.156 1.156 1.175 1.179 1.179 26 7.0 1.134 1.155 1.155 1.175 1.176 1.176 27 6.8 1.133 1.156 1.156 1.175 1.175 1.174 28 6.6 1.130 1.154 1.154 1.174 1.174 1.170 29 6.4 1.125 1.152 1.152 1.171 1.171 1.163 30 6.2 1.119 1.146 1.146 1.164 1.164 1.155 31 6.0 1.114 1.143 1.143 1.163 1.163 1.155 32 5.8 1.110 1.136 1.136 1.156 1.156 1.154 33 5.6 1.103 1.119 1.120 1.139 1.139 1.144 34 5.4 1.098 1.105 1.108 1.122 1.123 1.133 35 5.2 1.094 1.097 1.106 1.115 1.117 1.129 36 5.0 1.095 1.097 1.105 1.111 1.113 1.122 37 4.8 1.098 1.098 1.100 1.107 1.108 1.111 38 4.6 1.104 1.104 1.099 1.107 1.107 1.106 39 4.4 1.110 1.109 1.102 1.111 1.111 1.109 40 4.2 1.118 1.118 1.108 1.117 1.117 1.113 41 4.0 1.127 1.127 1.114 1.123 1.123 1.116 42 3.8 1.135 1.136 1.119 1.126 1.126 1.120 43 3.6 1.141 1.142 1.124 1.127 1.128 1.124 44 3.4 1.146 1.146 1.128 1.129 1.128 1.127 45 3.2 1.151 1.149 1.133 1.133 1.130 1.129 46 3.0 1.157 1.151 1.142 1.141 1.130 1.136 47 2.8 1.165 1.157 1.152 1.152 1.133 1.142 48 2.6 1.174 1.166 1.158 1.158 1.136 1.143 49 2.4 1.187 1.180 1.168 1.168 1.144 1.148 50 2.2 1.204 1.196 1.185 1.184 1.157 1.161 51 2.0 1.215 1.207 1.196 1.196 1.166 1.169 52 1.8 1.218 1.209 1.198 1.198 1.168 1.170 These decks are generated for normal operation flux maps that are typically taken at full power ARO. Additional N(z) decks may be generated for the specific plant conditions at the time of the flux map, if necessary, consistent with the methodology described in the RPDC topical (Reference 3). EOR is defined as Hot Full Power End of Reactivity.

Page 11 of 21

Serial No.10-662 Docket No. 50-338 COLR Table 3.2-1 (continued)

N1C22 Normal Operation N(Z)

NODE HEIGHT 11000 to 13000 13000 to 15000 15000 to 17000 17000 to 19000 19000 to EOR (FEET) MWDIMTU MWDIMTU MWDIMTU MWDIMTU MWDIMTU 10 10.2 1.126 1.133 1.133 1.112 1.113 11 10.0 1.124 1.131 1.131 1.112 1.113 12 9.8 1.122 1.129 1.129 1.111 1.112 13 9.6 1.122 1.129 1.129 1.110 1.111 14 9.4 1.121 1.124 1.125 1.104 1.106 15 9.2 1.125 1.125 1.126 1.107 1.108 16 9.0 1.139 1.138 1.133 1.125 1.126 17 8.8 1.155 1.155 1.144 1.147 1.147 18 8.6 1.161 1.161 1.149 1.153 1.153 19 8.4 1.167 1.167 1.156 1.158 1.158 20 8.2 1.177 1.177 1.168 1.172 1.172 21 8.0 1.183 1.183 1.177 1.181 1.181 22 7.8 1.185 1.185 1.179 1.183 1.183 23 7.6 1.187 1.187 1.181 1.185 1.185 24 7.4 1.189 1.188 1.185 1.190 1.190 25 7.2 1.189 1.189 1.188 1.192 1.192 26 7.0 1.188 1.188 1.187 1.191 1.191 27 6.8 1.187 1.187 1.187 1.191 1.191 28 6.6 1.184 1.184 1.185 1.189 1.189 29 6.4 1.177 1.179 1.181 1.184 1.185 30 6.2 1.167 1.171 1.177 1.178 1.179 31 6.0 1.161 1.167 1.179 1.178 1.175 32 5.8 1.154 1.162 1.177 1.177 1.17 1 33 5.6 1.143 1.154 1.170 1.170 1.167 34 5.4 1.135 1.148 1.162 1.161 1.163 35 5.2 1.131 1.145 1.159 1.158 1.162 36 5.0 1.126 1.139 1.153 1.15 4 1.158 37 4.8 1.118 1.129 1.143 1.148 1.151 38 4.6 1.113 1.120 1.130 1.140 1.142 39 4.4 1.113 .1.114 1.119 1.133 1.133 40 4.2 1.115 1.114 1.114 1.126 1.129 41 4.0 1.119 1.119 1.118 1.122 1.131 42 3.8 1.127 1.127 1.122 1.126 1.133 43 3.6 1.136 1.136 1.128 1.136 1.137 44 3.4 1.143 1.143 1.133 1.145 1.144 45 3.2 1.149 1.149 1.138 1.153 1.153 46 3.0 1.154 1.154 1.142 1.160 1.159 47 2.8 1.157 1.157 1.146 1.164 1.164 48 2.6 1.158 1.158 1.146 1.165 1.165 49 2.4 1.160 1.160 1.151 1.168 1.168 50 2.2 1.166 1.166 1.161 1.176 1.177 51 2.0 1.170 1.170 1.170 1.18 4 1.187 52 1.8 1.171 1.171 1.173 1.188 1.192 These decks are genera ted for normal operation flux maps that are typically taken at full power ARO . Additio nal N(z) decks may be generated for the specific plant conditio ns at the time of the flux map, if necessary, consistent with the methodo logy describ ed in the RPDC topical (Reference 3). EOR is defined as Hot Full Po wer End of Reactivity.

Page 12 of 21

Serial No.10-662 Docket No. 50-338 COLR Figure 3.2-1 K(Z) - Normalized FQ as a Function of Core Height 1.2 1.1 1.0 0.9

-- -- --

6,1.0 r---

r--

~

(12 .925 )

0.8

-

N "LL o

w N

0.7

J

<C 0.6

E a::

oz if 0.5

~

0.4 0.3 0.2 0.1 0.0 o 1 2 3 4 5 6 7 8 9 10 11 12 13 CORE HEIGHT (FT)

Page 13 of2 1

Serial No.10-662 Docket No. 50-338 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (FNMI)

LCO 3.2.2 FNMI shall be within the limits specified below.

~m ::; 1.587{1 + 0.3(1- PH THERMAL POWER where:

P= RATED THERMAL POWER SR 3.2.2.1 Verify FNMI is within limits specified above.

3.2.3 AXIAL FLUX DIFFERENCE (AFD)

LCO 3.2.3 The AFD in % flux difference units shall be maintained within the limits specified in COLR Figure 3.2-2.

Page 140f21

Serial No.10-662 Docket No. 50-338 COLR Figure 3.2-2 North Anna 1 Cycle 22 Axial Flux Difference Limits 120 110

(-12 (+6 100 100 V \

U aeeeptab l Oper ation Ur aeeeptabl ~

90 Ooer ation

\

.

Ql 80 / 1\

1/

==

0 a.

.

"iii E 70

/

A eept able ()perction

\

-

Ql

.s::

I-

"C Ql 60

/

/ 1\

\

I'll a:

'5 C /

.

Ql U

Ql a.

50

(-27, 50) (+20 40 30 20 10 o

-30 -20 -10 o 10 20 30 Percent Flux Difference (Delta-I)

Page 15 of 21

Serial No.10-662 Docket No. 50-338 3.3 INSTRUMENTATION 3.3.1 Reactor Trip System (RTS) Instrumentation TS Table 3.3. 1-1 Note1: Overtemperature ~T The Overtemperature ~T Function Allowable Value shall not exceed the following nominal trip setpoint by more than 2% of ~T span, with the numerical value s of the parameters as specified below.

~T ::;

(I +1'1s))[T-T]+K (P-P')-f (M) }

st; {K 1-K 21+1'

(

2 s 3 1 where: ~T is measured RCS ~T , oF.

~To is the indicated ~T at RTP, "F, s is the Laplace transform operator, sec".

T is the measured RCS average temperature, OF.

T ' is the nominal T avg at RTP , ::; 586.8 OF.

P is the measured pressurizer pressure, psig.

P ' is the nominal RCS operating pressure , ~ 2235 psig.

K3 ~ 0.001144 /psig 1'], 1':2 = time constants utilized in the lead-lag controller fo r Tavg

'tJ ~ 23.75 sec 't 2 s 4.4 sec (1 + 1']s)/( 1 + 1'2S) =function generated by the lead-lag controllerfor Tavg dynamic compensation f J(~I) ~ 0.0165 { (qt - qb) } when (qt - qb) < -35 % RTP o when -35 % RTP::; (qt - qb) ::; +3 % RTP 0.0198{ (qt - qb) - 3} when (qt - qb) > +3 % RTP Where qt and qb are percent RTP in the upper and lower halves of the core, respectively, and qt + qb is the total THERMAL POWER in percent RTP.

Page 16 of2 1

Serial No.10-662 Docket No. 50-338 TS Table 3.3.1-1 Note 2: Overpower .L1T The Overpower Af Function Allowable Value shall not exceed the following nominal trip setpoint by more than 2% of .L1T span, with the numerical values of the parameters as specified below.

where: .L1T is measured RCS .L1T, OF.

.L1To is the indicated .L1T at RTP, OF.

s is the Laplace transform operator, sec".

T is the measured RCS average temperature, "F.

T' is the nominal T avg at RTP, s 586.8 OF.

x, s 1.0865 K, ~ 0.0197;oF for increasing T avg  !<<) ~ 0.00162/oF when T > T' o / oF for decreasing T avg o / oF when T ~ T' T3 = time constant utilized in the rate lag controller for Tavg 1:3 ~ 9.5 sec T3S/( 1+ T3S) = function generated by the rate lag controller for Tavg dynamic compensation f2(.L1I) =0, for all M .

Page 17 of21

Serial No.10-662 Docket No. 50-338 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)

Limits LCO 3.4.1 RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified below:

a. Pressurizer pressure is greater than or equal to 2205 psig;
b. RCS average temperature is less than or equal to 591°F; and
c. RCS total flow rate is greater than or equal to 295,000 gpm.

SR 3.4.1.1 Verify pressurizer pressure is greater than or equal to 2205 psig.

SR 3.4.1.2 Verify RCS average temperature is less than or equal to 591°F.

SR 3.4.1.3 Verify RCS total flow rate is greater than or equal to 295,000 gpm.

SR 3.4.1.4 ------------------------------NOTE-------------------------------------------

Not required to be performed until 30 days after z 90% RTP.

Verify by precision heat balance that RCS total flow rate is ;:::

295,000 gpm.

Page 18 of21

- -------------~ . - - . __ ._ .._ - - --- --_.....

Serial No.10-662 Docket No. 50-338 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.6 Boron Injection Tank (BIT)

Required Action B.2 Borate to an SDM ~ 1.77 % Ak/k at 200 of.

Page 19 of21

Serial No.10-662 Docket No. 50-338 3.9 REFUELING OPERATIONS 3.9.1 Boron Concentration LCO 3.9.1 Boron concentrations of the Reactor Coolant System (RCS), the refueling canal, and the refueling cavity shall be maintained ~ 2600 ppm.

SR 3.9.1.1 Verify boron concentration is within the limit specified above.

Page 20 of21

Serial No.10-662 Docket No. 50-338 NAPS TECHNICAL REQUIREMENTS MANUAL TRM 3.1 REACTIVITY CONTROL SYSTEMS TR 3.1.1 Boration Flow Paths - Operating Required Action 0.2 Borate to a SHUTDOWN MARGIN ~ 1.77 % M<Ik at 200 OF, after xenon decay.

Page 21 of21

Retrieved from "https://kanterella.com/w/index.php?title=ML103120708&oldid=2124932"