ML18011A853

Proposed Tech Specs Re Relocation of Cycle Specific Setpoint Parameters to COLR

ML18011A853
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Site:	Harris
Issue date:	03/20/1995
From:	CAROLINA POWER & LIGHT CO.
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ENCLOSURE 5 SHEARON HARRIS NUCLEAR POVER PLANT NRC DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINT RELOCATION TO THE COLR TECHNICAL SPECIFICATION PAGES 9503230066 9'50320 PDR 'ADQCK 05000400 P

TABLE 2.2-1 Continued TABLE NOTATIONS NOTE 1: OVERTEHPERATURE hT (1 + r,S) (1 + r4S) hT D.To K] Kz + Ka(P P ) f>(hI)

(1 + r,S) 1+ AS (1 + r,S) 1+ r,S Where: hT Heasured hT by RTD Instrumentation; I +T)S Lead-lag compensator on measured. hT; 1+ r~S T)p Tz Time constants utilized in lead-lag compensator for hT, r> - 8 s, r> = 3 s; 1

Lag compensator on measured hT; 1 + r>S Time constants utilized in the lag compensator for hT, ra = 0 s; Indicated hT at RATED THERHAL POWER; K) Jptclf/E J lN 7 k COPE OPEiQ72~ M82fTD EZMR7 (COl/g pheon'pzocrddw pLp-40$ q Kz ~~ >pc'i',~Wk. Coal OP~R~g~~ings gp~~ ggzgg,

~~<4 i Aocf4 exp P~P-ZW~

1+ r4S The function generated by the lead-lag compensator for T, dynamic compensation; 1+ r5S

• T4, T5 Time constants utilized in the lead-lag compensator for Tpp T4 20 s, r5 = 4 s;

TABLE 2.2-1 Continued TABLE NOTATIONS NOTE 1: (Continued)

Average temperature, 'F~

Lag compensator on measured T,,;

1 +reS Time constant utilized in the measured T., lag compensator, r, = 0 s; 580.8'F (Nominal T., at RATED THERHAL POWER);

K3 p $g ye gift COEF OPFRATSP6 WAG~ REIVE@/CO~

i'4e+Pm~J~ NP-406 pressure, psig; 'ressurizer Pl 2235 psig (Nominal RCS operating pressure);

Laplace transform operator, s';

and f, (41) is a function of the indicated difference between top and bottom detectors of the power-c5 range neutron ion chambers; with gains to be selected based on measured instrument response during u'4 plant startup tests such that:

~p5 (1) For, - q, betwe n -21.6% and +12 %, f, (41) 0 where q, and qb ar percent RATED ERHAL PO R in the to and bottom halv s of the core r spectively, and + q, is total T RHAL POWER i percent of TED THERHAL PO R; (2) For each pe cent that the m nitude of q,

- , exceeds -21.6% the 4T Trip Se oint shall be automatica y reduced by 2 6% of its val at RATED THERHA POWER; and

) For eac percent that t magnitude of ,

- q, exceeds + 2.0%, the 4T T p Setpoint shall be autom ically reduced y 1.57% of its value at RATED T RMAL POWER.

NOTE 2: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2.1% 4T span.

INSERT 1 on Pa e 2-8 (1) For q, - q, between the "positive" and "negative" f,(hl) breakpoints as presented in the CORE OPERATING LIMITS REPORT (COLR), Plant Procedure PLP-106, f,(61) 0, where q, and q, are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and q, + q, is the total THERMAL POWER in percent of RATED THERMAL POWER.

(2) For each percent EI that the magnitude of q, - q, is more negative than the f,(BI) "negative" breakpoint presented in the CORE OPERATING LIMITS REPORT (COLR), Plant Procedure PLP-106, the AT trip setpoint shall be automatically reduced by the f,(AI) "negative" slope presented in the CORE OPERATING LIMITS REPORT (COLR), Plant Procedure PLP-106; and (3) For each percent 61 that the magnitude of q, - q, is more positive than the f,(AI) "positive" breakpoint presented in the CORE OPERATING LIMITS REPORT (COLR), Plant Procedure PLP-106, the AT trip setpoint shall be automatically reduced by the f,(QI) "positive" slope presented in the CORE OPERATING LIMITS REPORT (COLR), Plant Procedure PLP-106.

TABLE 2.2-1 (Continued)

O TABLE NOTATIONS Q

M I

NOTE 3o OVERPOMER T

()

(1 Mhere:

+

hT

+ v S) (

'tsS) (1 +

AT i3S)

<ST (K hs defined

-K ~

(1 + ~>S) in Note 1, (1 + ~6S)

T-K T

(

(1 +

1

~6S),- f2(AI))

) T" P

~1+ t S hs defined in Note 1, 1 + v2S Tls T2 hs defined in Note 1, 1 hs defined in Note 1, 1+ v3s T3 hs defined in Note 1,

>> +r Z J, 'ZX COZSOPewee u~S gfp08f (COl.RJ~ p4,af lao~g~~q pZP-404 hs defined in Note 1, Kg Kg 0.02/'F for increasing average temperature and 0 for decreasing average temperature,

~ts+

The function generated by the rate-lag compensator for Tavg dynamic 1

TIES compensation, Time constants utilized in the rate-lag compensator for Tavg' r = l0 s As defined in Note 1, 1 + t6S o

As defined in Note 1,

TABLE 2.2-1 Continued TABLE NOTATIONS NOTE 3: (Continued) gs sptcifisA~ f>< <<io Oprmfieg Jiiil7~

As defined in Note 1, R~p'<7'C<+~)~ p4 7 p~~/cner pM Indicated T, at RATED THERMAL POWER (Calibration temperature for hT instrumentatIon, ~ 580.8'F),

S As defined in Note 1, and f,(51) = 0 for all b,l.

NOTE 4: The channel's maximum Trip Setpoint" shall not exceed its computed Trip Setpoint by more than 2.3% hT span.

NOTE 5: The sensor error for temperature is 1.9 and 1. 1 for pressure.

NOTE 6: The sensor error for steam flow is 0.9, for feed flow is 1.5, and for steam pressure is 0.75.

NOTE 7: This value is associated with measured RCS flow ~ [293,540 gpm x (1.0 + C,)]. Technical Specification 3/4.2.3 requires this setpoint to be reduced at the rate of 1.5% of RTP for each 1%

that measured RCS flow is below [293,540 gpm x (1.0 + C,)].

ADMINISTRATIVE C LS 6.9.1.8 CORE OPERATING LIMITS REPORT 6.9. 1.6. 1 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT (COLR), plant procedure PLP-106, prior to each reload cycle, or prior to any remaining portion of a reload cycle, for the following:

a. Moderator Temperature Coefficient Positive and Negative Limits and 300 ppm surveillance limit for Specification 3/4. 1. 1.3,

b. Shutdown Bank Insertion Limits for Specification 3/4. 1.3.5,

c. Control Bank Insertion Limits for Specification 3/4. 1.3.6,

d. Axial Flux Difference Limits for Specification 3/4.2. 1,

e. Heat Flux Hot Channel Factor, F,~, K(Z), and V(Z) for Specification 3/4.2.2, Enthalpy Rise Hot Channel Factor, F~", and Power Factor Multiplier, PF~ for Specification 3/4.2.3.

g. Boron Concentration for .Specification 3/4.9. 1.

P 6.9. 1.6.2 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC at the time the reload analyses are performed, and the approved revision number shall be identified. in the COLR.

a. XN-75-27(A), latest Revision and Supplements, "Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors," Exxon Nuclear Company, Richland WA 99352.

(Methodology for Specification 3. 1. 1.3 - Moderator Temperature

-

Coefficient, 3.1.3.5 Shutdown Bank Insertion Limits, 3. 1.3.6-Control Bank Insertion Limits, 3.2. 1 - Axial Flux Difference, 3.2.2-Heat Flux Hot Channel Factor, 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor, and 3.9. 1 - Boron Concentration).

b. ANF-89-151(A), latest Revision, "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events,"

Advanced Nuclear Fuels Corporation, Richland WA 99352.

(Methodology for Specification 3. 1.1.3 - Moderator Temperature Coefficient, 3. 1.3.5 - Shutdown Bank Insertion Limits, 3. 1.3.6-Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2-Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

c. XN-NF-82-21(A), latest Revision, "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations,"

Exxon Nuclear Company, Richland WA 99352.

(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

SHEARON HARRIS - UNIT 1 6-24 Amendment No. 44

INSERT 2 on Pa e 6-24 Overtemperature and Overpower Delta T setpoint parameter values for Specification 2.2.1.

INSERT 3 on Pa e 6-24b 2.2.1 - Reactor Trip System Instrumentation Setpoints,

ADMINISTRATIVE OLS

.6.9.1.6 CORE OPERATING LIMITS REPORT (Continued)

h. ANF-88-054(A), latest Revision, "PDC-3: Advanced Nuclear Fuels Corporation Power Distribution Control for Pressurized Water Reactors and Application of PDC-3 to H. B. Robinson Unit 2," Advanced Nuclear Fuels Corporation, Richland WA 99352.

(Methodology for Specification 3.2. 1 - Axial Flux Difference, and 3.2.2 - Heat Flux Hot Channel Factor).

WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY",

July 1985 (W Proprietary).

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

WCAP-10266-P-A, Rev. 2, "The 1981 Version of the WESTINGHOUSE ECCS EVALUATION MODEL USING THE BASH CODE", March 1987 (W Proprietary).

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor).

k. WCAP-11837-P-A, "EXTENSION OF METHODOLOGY FOR CALCULATING TRANSITION CORE DNBR PENALTIES", January 1990 (W Proprietary).

(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

EMF-92-081(A), latest Revision and Supplements, "Statistical Setpoint/Transient Methodology for Westinghouse Type Reactors,"

Siemens Power Corporation, Richland WA 99352.

(Methodology for Specification 3. 1. 1.3 - Moderator Temperature Coefficient, 3. 1.3.5 - Shutdown Bank Insertion Limits, 3. 1.3.6-Control Bank Insertion Limits, 3.2. 1 - Axial Flux Difference, 3.2.2-Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

EMF-92-153(A), latest Revision and Supplements, "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel,"

Siemens Nuclear Power Corporation, Richland WA 99352.

(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

n. XN-NF-82-49(A), latest Revision and Supplements, "Exxon Nuclear Company Evaluation Model EXEM PWR Small Break Model," Exxon Nuclear Company, Richland WA 99352.

(Methodology for Specification 3.2. 1 - Axial Flux Difference, 3.2.2-Heat Flux Hot, Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

SHEARON HARRIS - UNIT 1 6-24b Amendment No. 44

TABLE 2.2-1 Continued TABLE NOTATIONS NOTE 1: OVERTEMPERATURE hT (1 + T,S) ('

1+

K-K 1 2 T 1

(1 + T,S) T3S (1 S) 1 + T,S Where: hT Measured hT by RTD Instrumentation; 1+ T,S Lead-lag compensator on measured hT:

1 + T2S

= Time constants utilized in T,, T, 1 ead-1 ag compensator for hT, T, = 8 s, T, = 3 s; 1

Lag compensator on measured hT; 1+ T3S Time constants utilized in the lag compensator for ET. T, = 0 s:

Indicated hT at RATED THERMAL POWER; K, As specified in the CORE OPERATING LIMITS REPORT (COLR). Plant Procedure PLP-106; K2 As specified in the CORE OPERATING LIMITS REPORT (COLR), Plant Procedure PLP-106; 1 +T4S4 = The function generated by the lead-lag compensator for T,, dynamic compensation; 1 +TSS

= Time constants utilized in the lead-lag = 20 T4, T, compensator for T,,, T4 s, T, = 4 s;

TABLE 2.2-1 Continued TABLE NOTATIONS NOTE 1: (Continued)

Average temperature, 'F; Lag compensator on measured T,,;

1+ r6S constant utilized in the measured T,, lag compensator, r, = 0 s; P'ime K3 580.8'F (Nominal T,, at As specified in the PLP-106; RATED THERMAL POWER):

CORE OPERATING LIMITS REPORT (COLR), Plant Procedure Pressurizer pressure, psig; 2235 psig (Nominal RCS operating pressure);

Laplace transform operator. s';

and f, (BI) is a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startu

~

tests such that:

For q, - q, between the "positive" and "negative" f, (bI) breakpoints as presented in the CORE OPERATING LIMITS REPORT (COLR) Plant Procedure PLP-106, f, (bI) = 0, where are percent RATED THERMA POWER in the top and bottom halves of the q, and q, core respectively, and q, + q, is the total THERMAL POWER in percent of RATED THERMAL POWER.

(2) For each percent EI that the magnitude of q, - q, is more negative than the f, (BI) "negative" breakpoint presented in the CORE OPERATING LIMITS REPORT (COLR). Plant Procedure PLP-106. the hT Trip Setpoint shall be automatically reduced by the f, (bI) "negative" slope presented in the CORE OPERATING LIMITS REPORT (COLR) Plant Procedure PLP-106; and

TABLE 2.2-1 Continued TABLE NOTATIONS NOTE 1: (Continued)

(3) For each percent BI that the magnitude of q, - q, is more positive than the f, (QI) "positive" breakpoint presented in the CORE OPERATING LIHITS REPORT (COLR) Plant Procedure PLP-106, the hT Trip Setpoint shall be automatically reduced by the f, (hI) "positive" slope presented in the CORE OPERATING LIMITS REPORT (COLR), Plant Procedure PLP-106.

NOTE 2 The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2.1X ZiT span.

TABLE 2.2-1 Continued TABLE NOTATIONS NOTE 3: OVERPOWER BIT hT~ 'j ~j ~hT K K ~'j ~j T K T - T" f $ 1j (1 + r~S) (1 + r3S) (1 + r,S) (1 + r6S) (1 + r,S)

Where: hT As defined in Note 1, 1+r,S As defined in Note 1, 1 + r2S As defined in Note 1, As defined in Note 1, 1+ r,S As defined in Note 1, As defined in Note 1, K4 As specified in the CORE OPERATING LIMITS REPORT (COLR), Plant Procedure PLP-106; K5 0.02/'F for increasing average temperature and 0 for decreasing average temperature, r,S The function generated by the rate-lag compensator for T,, dynamic compensation, 1+ r7S Time constants utilized in the rate-lag compensator for T,,, r, = 10 s, As defined in Note 1, 1 + r,S As defined in Note 1,

TABLE 2.2-1 Continued TABLE NOTATIONS NOTE 3: (Continued)

K6 As As specified in the defined in Note 1, Indicated T,at CORE OPERATING RATED THERMAL POWER instrumentat~ion, ~ 580.8'F).

LIMITS REPORT (COLR),

(Calibration temperature for ET I

Plant Procedure PLP-106 [

S As defined in Note 1, and f,(EI) = 0 for all BI.

NOTE 4: The channel's.maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2.3X hT span.

NOTE 5: The sensor error for temperature is 1.9 and 1.1 for pressure.

NOTE 6: The sensor error for steam flow is 0.9, for feed flow is 1.5, and for st'earn pressure is 0.75.

NOTE 7: This value is associated with measured RCS flow a [293,540 gpm x (1.0 + C,)]. Technical Specification 3/4.2.3 requires this setpoint to be reduced at the rate of 1.5X of RTP for each 1X that measured RCS flow is below [293.540 gpm x (1.0 + C,)].

ADMINISTRATIVE C,OLS

.6.9.1.6 CORE OPERATING LIMITS REPORT 6.9. 1.6. 1 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT (COLR), plant procedure PLP-106, prior to each reload cycle, or prior to any remaining portion of a reload cycle, for the following:

a. Moderator Temperature Coefficient Positive and Negative Limits and 300 ppm surveillance limit for Specification 3/4. 1. 1.3,

b. Shutdown Bank Insertion Limits. for Specification 3/4. 1.3.5,

c. Control Bank Insertion Limits for Specification 3/4. 1.3.6.

d. Axial Flux Difference Limits for Specification 3/4.2. 1,

e. Heat Flux Hot Channel Factor. F~" , K(Z). and V(Z) for Specification 3/4.2.2, Enthalpy Rise Hot Channel Factor, F~

' and Power Factor Multiplier, PF~ for Specification 3/4.2.3.

g. Boron Concentration for Specification 3/4.9. 1.

h. Overtemperature and Overpower Delta T setpoint parameter values for Specification 2.2. 1.

6.9.1.6.2 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC at the time the reload analyses are performed'nd the approved revision number shall be identified in the COLR.

XN-75-27(A), latest Revision and Supplements, "Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors," Exxon Nuclear Company, Richland WA 99352.

(Methodology for Specification 3.1. 1.3 - Moderator Temperature Coefficient, 3. 1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6-Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor. and 3.9. 1 - Boron Concentration).

ANF-89-151(A), latest Revision, "ANF-RELAP Methodology f'r Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events," Advanced Nuclear Fuels Corporation, Richland WA 99352.

(Methodology for Specification 3. 1. 1.3 - Moderator Temperature Coefficient. 3. 1.3.5 - Shutdown Bank Insertion Limits, 3. 1.3.6-Control Bank Insertion Limits. 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

SHEARON HARRIS - UNIT 1 6-24 ~

Amendment No.

ADMINISTRATIVE C OLS 6.9.1:6 CORE OPERATING LIMITS REPORT (Continued)

XN-NF-82-21(A), latest Revision, "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations," Exxon Nuclear Company. Richland WA 99352.

(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

XN-75-32(A). Supplements 1. 2, 3, and 4 "Computational Procedure

~

for Evaluating Fuel Rod Bowing," Exxon Nuclear Company, Richland WA 99352.

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

XN-NF-84-93(A), latest Revision and Supplements. "Steamline Break Methodology for PWRs," Exxon Nuclear Company. Richland WA 99352.

(Methodology for Specification 3. 1. 1.3 - Moderator Temperature Coefficient, 3. 1.3.5 - Shutdown Bank Insertion Limits, 3. 1.3.6-Control Bank Insertion Limits, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

EXEM PWR Large Break LOCA Evaluation Model as defined by:

XN-NF-82-20(A), latest Revision and Supplements'Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Updates," Exxon Nuclear Company, Richland WA 99352.

XN-NF-82-07(A), latest Revision, "Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model," Exxon Nuclear Company, Richland WA 99352.

XN-NF-81-58(A), latest Revision and Supplements, "RODEX2 Fuel Rod Thermal Response Evaluation Model," Exxon Nuclear Company, Richland WA 99352.

XN-NF-85-16(A), Volume 1 and Supplements. Volume 2. latest Revision and Supplements, "PWR 17x17 Fuel Cooling Test Program,"

Exxon Nuclear Company. Richland WA 99352.

XN-NF-85-105(A), and Supplements. "Scaling of FCTF Based Reflood Heat Transfer Correlation for Other Bundle Designs," Exxon Nuclear Company, Richland WA 99352.

(Methodology for Specification 3.2. 1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

XN-NF-78-44(A), latest Revision. "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors,"

Exxon Nuclear Company, Richland WA 99352.

(Methodology for Specification 3. 1.3.5 - Shutdown Bank Insertion Limits. 3. 1.3.6 - Control Bank Insertion Limits, and 3.2.2 - Heat Flux Hot Channel Factor).

SHEARON HARRIS - UNIT 1 6-24a Amendment No.

ADMINISTRATIVE C 'OLS

.9.1.6 CORE OPERATING LIMITS REPORT (Continued)

ANF-88-054(A), latest Revision, "PDC-3: Advanced Nuclear Fuels Corporation Power Distribution Control for Pressurized Water Reactors and Application of PDC-3 to H. B. Robinson Unit 2,"

Nuclear Fuels Corporation, Richland WA 99352. 'dvanced (Methodology for Specification 3.2. 1 - Axial Flux Difference, and 3.2.2 - Heat Flux Hot Channel Factor ).

WCAP-9272-P-A. "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY", July 1985 (W Proprietary).

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

WCAP-10266-P-A, Rev. 2, "The 1981 Version of the WESTINGHOUSE ECCS EVALUATION MODEL USING THE BASH CODE", March 1987 (W Proprietary).

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor).

WCAP-11837-P-A. "EXTENSION OF METHODOLOGY FOR CALCULATING TRANSITION CORE DNBR PENALTIES". January 1990 (W Proprietary).

(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

EMF-92-081(A), latest Revision and Supplements. "Statistical Setpoint/Transient Methodology for Westinghouse Type Reactors,"

Siemens Power Corporation, Richland WA 99352.

(Methodology for Specification 2.Z. 1 - Reactor Trip System

-

Instrumentation Setpoints, 3.1. 1.3 Moderator Temperature Coefficient, 3. 1.3.5 - Shutdown Bank Insertion Limits, 3. 1.3.6-Control Bank Insertion Limits'.2. 1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

EMF-92-153(A), latest Revision and Supplements, "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," Siemens Nuclear Power Corporation. Richland WA 99352.

(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

XN-NF-82-49(A), latest Revision and Supplements, "Exxon Nuclear Company Evaluation Model EXEM PWR Small Break Model," Exxon Nuclear Company. Richland WA 99352.

(Methodology for Specification 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

SHEARON HARRIS - UNIT 1 6-24b Amendment No.