ML20238C996

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Proposed Tech Specs,Incoporating Operating Limits for All Fuel Types for Cycle 8 Operation of Unit
ML20238C996
Person / Time
Site: Brunswick Duke Energy icon.png
Issue date: 09/04/1987
From:
CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML20238C895 List:
References
87TSB10, NLS-87-184, NUDOCS 8709100488
Download: ML20238C996 (61)


Text

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i ENCLOSURE 1 TO SERIAL: NLS-87-184 i PROPOSED TECHNICAL SPECIFICATION PAGES BRUNSWICK PLANT, UNIT 2 TSC 87TSB10 i

8709100488 870904 (5273WRM/lah)

PDR ADOCK 05000324

p. PDR A: .

SUMMARY

LIST OF REVISIONS PAGE NO. DESCRIPTION OF CHANGES IV Revise to correct the pagination of Section 3/4.2.

1-2 Revise definition of CRITICAL POWER RATIO to refer to "an NRC-approved CPR correlation" instead of the GEXL correlation.

1-5 Revise the definition for PHYSICS TESTS to reference Chapter 14 of the Updated Final Safety Analysis Report (FSAR)instead of Chapter 13 of the original FSAR.

3/4 2-2 (old) Delete figure of 8x8R fuel type MAPLHGR limits. .

3/4 2-2 (new) Repaginate. Revise figure number (previously Figure 3.2.1-3).

3/4 2-3 (old) Delete figure of 8x8R fuel type MAPLHGR limits.

3/4 2-3 (new) Repaginate. Revise figure number (previously Figure 3.2.1-4) 3/42-4 Repaginate. Revise figure number (previously Figure 3.2.1-5) 3/42-5 Incorporate new MAPLHGR curve for GE8 fuel type.

3/42-6 Incorporate new MAPLHGR curve for GE8 fuel type.

I 3/42-7 Delete the design total peaking factor for 8x8R fuel type and add design total peaking factor value for GE8 fuel type.

I' Remove references to the 8x8R fuel type MCPRs, add MCPR 3/42-8 values for the GE8 fuel type, and revise MCPR values for existing fuel types.

3/42-10 Revise the values of mu and sigma. Add reference to notch 36 in the Limiting Condition for Operation.

3/4 2-12 Remove references to the 8x8R fuel type. Add MCPR values for the GE8 fuel type. Revise MCPR values for remaining existing fuel.

types.

3/4 2-14 Add Linear Heat Generation Rate (LHGR) limit of 14.4 kw/f t for GE8 fuel type.

3/4 3-42 Revise reference to the factor "T" to be consistent with the "T" factors in TS 3.2.2.

B3/4 2-2 Revise table to incorporate new GE8 fuel type.

B3/4 2-3 Revise Section 3/4.2.2 to delete total peaking factor reference to 8x8R fuel type and add reference to GE8 fuel type.

(5273WRN/lah)

(BSEP-2-119) l INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS' SECTION PACE 3/4.0 APPLICABILITY.............................................. 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN HARCIN.......................................... 3/4 1-1 3/4.1.2 REACTIVITY AN0HALIES..................................... 3/4 1-2 3/4.1.3 CONTROL RODS Control Rod Operability.................................. 3/4 1-3 Control Rod Maximum Scram Insertion Times................ 3/4 1-5 Control' Rod-Average Scram Insertion Times................ 3/4 1-6 Four Control Rod Group Insertion Times................... 3/4 1-7 Control Rod Scram Accumulators........................... 3/4 1-8 Control Rod Drive Coupling............................... 3/4 1-9 Control Rod Position Indication.......................... 3/4'l-11 Control Rod' Drive Housing Support........................ 3/4 1-13 3/4.1.4 CONTROL ROD PROGRAM CONTROLS Rod Worth Minimizer...................................... 3/4 1-14 Rod Sequence Control System.............................. 3/4 1-15 Rod Block Monitor........................................ 3/4 1-17 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM............................ 3/4 1-18 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERACE PLANAR LINEAR HEAT CENERATION RATE............... 3/4 2-1 3/4.2.2 APRM SETP0INTS........................................... 3/4 2-7 3/4.2.3 MINIMUM CRITICAL POWER RATI0............ ................ 3/4 2-8 3/4.2.4 LINEAR HEAT CENERATION RATE.............................. 3/4 2-14 .

I 1'

i l

BRUNSWICK - UNIT 2 IV Amendment No.

(BSEP-2-119)

DEFINITIONS CHANNEL FUNCTIONAL TEST (Continued)

b. Bistable channels - the injection of a simulated signal into the channel sensor'to verify OPERABILITY including alarm and/or trip functions.

CORE ALTERATION-CORE ALTERATION shall be the addition, removal, relocation, or movement of fuel, sources, incore instruments, or reactivity controls in the reactor core with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component.to a safe, conservative location.

CRITICAL POWER RATIO The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in an assembly which is calculated, by application of an NRC approved CPR correlation, to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131.shall be concentration of I-131, p Ci/ gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The following is defined equivalent to 1 pCi of I-131 as determined from Table III of-TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites": I-132, 28 uCi; I-133, 3.7 pCi; I-134, 59 uCi; I-135, 12 uCi.

E -AVERACE DISINTEGRATION ENERGY E shall be the average, weighted in proportion to the concentration of each radionuclides in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes with half lives greater than 15 minutes making up at least 95% of the total non-iodine activity in the coolant.

EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME l- The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation setpoint at the channel sensor-until the ECCS equipment is capable of performing its safety function (i.e., the valves travel to their required positions,' pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

I BRUNSWICK - UNIT 2 1-2 Amendment No.

- _ = - _

(BSEP-2-119)

DEFINITIONS OFFSITE DOSE CALCULATION HANUAL (ODCM)

The OFFSITE DOSE CALCULATIONAL MANUAL (ODCM) is a manual which contains the current methodology and parameters to be used to calculate offsite doses resulting from the release of radioactive gaseous and liquid effluents; the methodology to calculate gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints; and, the requirements of the environmental radiological monitoring program.

OPERABLE - OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s).

Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electric power sources, cooling or seal wat r, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their related support function (s).

OPERATIONAL CONDITION An OPERATIONAL CONDITION shall be any one inclusive combination of mode switch position and average reactor coolant temperature as indicated in Table 1.2.

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and are 1) described in Section 14 of the Updated FSAR, 2) authorized under the l provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE i PRESSURE BOUNDARY LEAKAGE shall be leakage through a non-isolable fault in a reactor coolant system component body, pipe wall, or vessel wall.

PRIMARY CONTAINMENT INTECRITY PRIMARY CONTAINMENT INTECRITY shall exist when:

a. All penetrations required to be closed during accident conditions are either:
1. Capable of being closed by an OPERABLE containment automatic isolation valve system, or
2. Closed by at least one manual valve, blit.d flange, or deactivated automatic valve secured in its closed position, except as provided in Table 3.6.3-1 of Specification 3.6.3.1, or BRUNSWICK - UNIT 2 1-5 Amendment No.

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POWER DISTRIBUTION LIMITS- 1

~/4.2.2 3 APRM SETPOINTS LIMITING CONDITION FOR OPERATION

'3.2.2 The flow-biased-APRM scram trip setpoint (S) and rod block trip set point (SRB) shall be established according to the following relationship:

l lS $-(0.66W + 54%) T SRB $}(0.66W+42%)T where: S and SRB are in percent of RATED THERMAL POWER.

W = Loop recirculation flow in percent of rated flow, T_= Lowest value of the ratio of design TPF divided by the MTPF obtained for any class of fuel in the core (T s 1.0), and Design TPF for P8 X 8R, fuel = 2.39 l BP8 x 8R fuel = 2.39 CE8 fuel = 2.48 l APPLICABILITY: OPERATIONAL' CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

ACTION:

With S or SRB exceeding the allowable value, initiate corrective action within 15 minuten and con inue corrective action so that S and SRB are within the required limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2.2 The HTPF for each class of fuel shall be determined, the value of T calculated,'and the flow biased APRM trip setpoint adjusted, as required:

a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,
b. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and
c. Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating.with a LIMITING CONTROL ROD PATTERN for MTPF.

BRUNSWICK'- UNIT 2 3/4 2-7 Amendment No.

(BSEP-2-119)

POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3'.1[ The MINIMUM CRITICAL POWER RATIO (MCPR), as a function of core flow, shall be equal to or greater than the MCPR limit times the Kg shown in Figure 3.2.3-1 with the following MCPR limit adjustments:

a. Beginning-of-cycle (BOC) to end-of-cycle (EOC) minus 2000 MWD /t with ODYN OPTION A analyses in effect'and the end-of-cycle recirculation pump trip system inoperable, the MCPR limits are listed below:
1. MCPR for P8 x 8R fuel = 1.34 2.. MCPR for BP8 x 8R fuel = 1.34
3. MCPR for CE8 fuel = 1.34
b. EOC minus 2000 MWD /t to EOC with ODYN OPTION A analyses in effect and the end-of-cycle recirculation pump trip system inoperable, the MCPR limits are listed below:
1. MCPR for P8 x 8R fuel = 1.35
2. MCPR for BP8 x BR fuel = 1.35
3. MCPR for CE8 fuel = 1.35
c. BOC to EOC minus 2000 MWD /t with ODYN OPTION B analyses in effect and the end-of-cycle recirculation pump trip system inoperable, the MCPR limits'are listed below:
1. MCPR for P8 x 8R fuel = 1.27
2. MCPR for BP8 x BR fuel = 1.27
3. MCPR for CE8 fuel = 1.27
d. EOC minus 2000 MWD /t to EOC with ODYN OPTION B analyses in effect and the end-of-cycle recirculation puup trip system inoperable, the MCPR limits are listed below:
1. MCPR for P8 x SR fuel = 1.31
2. MCPR for BP8 x 8R fuel = 1.31
3. MCPR for CE8 fuel = 1.31 APPLICABILITY: OPERATIONAL CONDITION 1 when THERMAL POWER is greater than or equal to 25% RATED THERMAL POWER BRUNSWICK - UNIT 2 3/4 2-8 Amendment. No.

______.______________._____j

i (BSEP-2-119) i POWER' DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO (ODYN OPTION B)

LIMITING CONDITION FOR OPERATION 3.2.3.2 For the OPTION B MCPR limits listed in specification 3.2.3.1 to be used, the cycle average 20% (notch 36) scram time (t * ) shall be less.than or l equal to the Option'B scram time limit (tg), where t, and T B are determined as follows: n E

i=1 Nt y t__

t , where ave " n.-

N.

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i=1

.i = Surveillance test number, n = Number of surveillance tests performed to date in the cycle (including BOC),

th surveillance test, and Ni = Number of rods tested in the i t; = Average scram time to notch 36 for surveillance test i N

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r B = u + 1.65 ( n N) g E

i=1

-i = Surveillance test number n = Number of surveillance tests performed to date in the cycle (including BOC),

th surveillance test Ni = Number of rods. tested in the i Ng = Number'of rods tested at BOC, p = 0.813 seconds l (mean value for statistical scram time distribution from de-energization of scram pilot valve solenoid to pickup on notch 36),

o = 0.018 seconds l (standard deviation of the above statistical distribution).

APPLICABILITY:- OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% RATED THERMAL POWER.

BRUNSWICK - UNIT 2 3/4 2-10 Amendment No.

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POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT CENERATION RATE n

LIMITING CONDITION FOR OPERATION 3.2.4 The LINEAR HEAT CENERATION RATE (LHCR) shall not exceed 13.4 kw/ft for P8x8R and BP8x8R fuel assemblies and 14.4 kw/ft for CE8 fuel assemblies.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than'or-equal to 25% of RATED THERMAL POWER.

ACTION:

With the LHCR of any fuel rod exceeding the above limit, initiate corrective action within 15 minutes and continue corrective action so that the LHCR is within the limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2.4 LHGRs shall be determined to be equal to or less than the limit:

a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,
b. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and
c. Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating on a LIMITING CONTROL ROD PATTERN for LHCR.

1 BRUNSWICK - UNIT 2 3/4 2-14 Amendment No.

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(BSEP-2-119) l Bases Table B 3.2.1-1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS FOR BRUNSWICK - UNIT 2 Plant Parameters; Core Thermal Power 2531 Mwt which corresponds to 105% of rated steam flow Vessel Steam Output 10.96 x 106Lbm/h which corresponds to 105% of rated steam flow Vessel Steam Dome Pressure 1055 psia Recirculation Line Break Area for Large Breaks

a. Discharge 2.4 ft2 (DBA); 1.9 ft2 (80% DBA)
b. Suction 4.2 ft 2 Number of Drilled Bundles 520 Fuel Parameters:

PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAT AXIAL CRITICAL FUEL BUNDLE CENERATION RATE PEAKING POWER **

FUEL TYPES CEOMETRY (kw/ft) FACTOR RATIO Reload Core BP/P8x8R 13.4 1.4 1.20 GE8x8EB 14.4 1.4 1.20 A more detailed list of input to each model and its source is presented in Section II of Reference 1.

'* This power level meets the Appendix K requirement of 102%.

    • To account for the 2% uncertainty in bundle power required by Appendix K, the SCAT calculation is performed with an MCPR of 1.18 (i.e., 1.2 divided by 1.02) for a bundle with an initial MCPR of 1.20.

o BRUNSWICK - UNIT 2 B 3/4 2-2 Amendment No.

(BSEP-2-119)

POWER. DISTRIBUTION LIMITE BASES 1 3/4.2.2 APRM SETPOINTS The fuel cladding integrity Safety Limits of Specification 2.1 were based on a TCTAL PEAKING FACTOR of 2.39 for P8x8R and BP8x8R fuel and 2.48 for CE8 fuel. The scram setting and rod block functions of the APRM instruments must be adjusted to ensure that the MCPR does not become less than 1.0 in the i degraded situation. The scram settings and rod block settings are adjusted in accordance with the formula in this specification when the combination of THERMAL FGWER and peak flux indicates a TOTAL PEAKING FACTOR greater than 2.39 for P8x8R and BP8x8R fuel and 2.48 for CE8 fuel. This adjustment may be l accomplished by increasing the APRM gain and thus reducing the slope and intercept point of the flow referenced APRM high flux scram curve by the recipsocal of the APRM gain change. The method used to determine the design TPF shall be consistent with the method used to determine the MTPF.

3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as speeliied in Specification 3.2.3 are derived from the established fuel claeJingintegritySafeg{)LimitMCPRof1.07,andananalysisofabnormal For any abnormal operating transient analysis operational transients.

evaluation with the initial condition c: the reactor being at the steady state operating limit, it is required that the resulting MCPR does not deccesse below the Safety Limit MCPR at any time during the transient, assuming an instrument trip setting as given in Specification 2.2.1.

To r..sure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and J coolant temperature decrease.

I Unless.otherwise stated in cycle specific reload analyses, the limiting transient which determines the required steady state MCPR limit is the turbine trip with failure of the turbine bypass. This transient yields the largest 6 MCPR. Prior to the analysis of abnormal operational transients an initial fuel bundle MCPR was determined. This parameter is based on the bundle flow calculatedbyaCEmultichannelsteadygjateflowdistributionmodelas and on core parameters shown in described in Section 4.4 of NEDO-20360 Reference 3, response to Items 2 and 9.

1 1

BRUNSWICK - UNIT 2 B 3/4 2-3 Amendment No.

I-i ENCLOSURE 2 TO SERIAL: NLS-87-184

" SUPPLEMENTAL RELOAD LICENSING REPORT FOR BRUNSWICK STEAM ELECTRIC PLANT UNIT 2, RELOAD 7, CYCLE 8" TSC 87TSB10 i

(5273WRM/lah)

23A3855 Revision 0 Class I August 1987 (23A5855,'Rev. 0)

-l SUPPLEMENTAL RELOAD LICENSING REPORT FOR BRUNSWICK STEAM ELECTRIC PLANT .. .

~

UNIT 2, RELOAD 7, CTCLE 8 n~ >.~ 2 m . .-

i Prepared:

P. A. Lambett Fuel Licensing i

Verified: - . 8 P. E. Elliott Fuel Licensing {

l i

Approved:

J.S./haynley, Manager )

j Fuel L.iceAsing

)

)

NUCLEAR ENERGY BUSINESS OPERATIONS . GENERAL ELECTRIC COMPANY

- SAN JOSE, CALIFORNIA 95125 G EN ER AL h) ELECTRIC n

1/2

-- __________-________________a

.. 23A5855 'Rav. 0 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for Carolina Power

]

and' Light Company (CP&L) for CP&L's use with the U.S. Nuclear Regulatory Com-mission (USNRC) to amend CP&L's operating license of the Brunswick Steam Electric Plant Unit 2. The information contained in this report is believed by General Electric to'be an accurate and true representation of the facts

~

known, obtained or provided'to General Elcetric at the time this report was prepared.

The only undertakings of the General Electric Company respecting informa- ~

tion in this document are contained in the Supplemental Agreement to the Con-

' ~ tract between Carolina Power and Light Company and General Electric Company for Reload Fuel Supply and Related Services for Brunswick Steam Electric Plant i.

Unit 2, and nothing contained in this document shall be construed as changing said contract. The use of this information except as defined by said con- l

~

tract, or for any purpose other than that for which it is intended, is not authorized; and with respect to any such unauthorized use, neither General Electric Company nor any of the contributors to thir, document makes any repre-santation or warranty (express or implied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of such information may not infringe privately owned rights; nor do they assume l

any responsibility for liability or damage of any kind which may result from I such use of such information. i 3

1 3/4 l(

__ __ __D

. . 23A5855 Rev. 0-ACKNOWLEDGMENT The engineering and reload licensing' analyses, which form the technical basis ~of this Supplemental Reload Licensing Report, were performed by T. P.

Lung and R. E. Polomik'of the Nuclear Fuel and Engineering Services Department. '

  • 4 l

1 i

1 1

.. j i

L I

1 .

5/6 t _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _

'23A3855 Rev. 0

-1. PLANT-UNIQUE ITEMS'-(1.0)*

Appendix A: _ Limiting Conditions for Operation

-Appendix B:- Bases for Limiting Conditions for Operation .

Appendix C: Safety Relief Valve.Out-of-Service

' Appendix D: Transient Operating Parameters-Appendix E: Turbine Control Valve Configuration Appendix F: Use of GEMINI Methods for Cycle 8

2. RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1 AND 4.0) -

l Cycle Loaded Number 1,

I Irradiated

- P8DRB265H 5 44 BP8DRB299 6 184 l , BP8DRB299 7 148 New' l f l

Bb317A' 8' 92

]

BD323A 8 92 Total 560

3. REFERENCE CORE LOADING PATTERN (3.3.1)

Nominal previous cycle core' average exposure at end of cycles 20,,449 mwd /MT Minimum previous cycle core average exposure at end of cycle from cold shutdown considerations: 20,00e Mud /Mr 3 Assumed reload cycle core average exposure at end of-cycle: 20,814 mwd /MT Core loading pattern: Figure 1 l

l

  • (- ) Refers to area of discussion in " General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A-8, dated May 1986. A letter "S" preceding the number refers to the appropriate section in the United States Supplement, NEDE-24011-P-A-8-US, May 1986.

7

23A5855 Rev. 0 4 CALCULATNDCOREEFFECTIVEMULTIPLICATJONANDCONTROLSYSTEMWORTH-NO VOIDS, 20*C (3.3.2.1.1 AND 3.3.2.1.2) e Beginning of Cycle,-Keff Uncontrolled 1.114 Fully Controlled 0.968 Strongest Control Rod Out 0.988 R, Maximum Increase in Cold Core Reactivity with 0.0 Exposure into Cycle, AK .-

5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)

Shutdown Margin (aK) pppg (20*C, Xenon Free)

- - - '600 0.031

6. RELOAD-UNIQUE TRANSIENT ANALYSIS INPUT (3.3.2.1.5 AND S.2.2) -l l

(Cold Water Injection Events Only)

Void Fraction (%) 41.71 Average Fuel Temperature (*F) 1104 Void Coefficient N/A* (d/% Rg) -7.34/-9.18 Doppler Coefficient M/A* (d/*F) -0.205/-0.195 Scram Worth N/A* ($)

l.

i

  • N
  • Nuclear Input Data, A = Used in Transient Analysis
    • Generic exposure independent values are used as given in " General Electric i Standard Application for Reactor Fuel," NEDE-24011-P-A-8, dated May 1986.

8

23A5855 Rev. 0

7. RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (S.2.2) en ng Factors Fuel Bundle Power Bundle Flow Initial Design Local Radial Axial R-Factor (MWt) (1000 lb/hr) MCPR Exposure: BOC8 to EOC8-2000 mwd /ST BP/P8x8R 1.20 1.55 1.40 1.051 6.574 111.5 1.22 GE8x8EB 1.20 1.56 1.40 1.051 6.602 113.6 1.23 Exposure: E0C8-2000 mwd /ST to EOC8

~

BP/P8x8R 1.20 1.49 1.40 1.051 6.320 113.0 1.28 GECx8EB 1.20 1.50 1.40 1.051 6.382 114.8 1.27

8. SELECTED MARGIN IMPROVEMENT OPTIONS (S.2.2.2)

^

Transient Recategorization: No Recirculation Pump Trip: .No Rod Withdrawal Limiter: No Thermal Power Monitor: Yes  !

Improved Scram Time: No Exposure Dependent Limits: Yes Exposure Points Analyzed: EOC8 and E0C8-2000 mwd /ST

9. OPERATING FLEXIBILITY OPTIONS (S.2.2.3) l Single-Loop Operation: Yes Load Line Limit: Yes Extended Load Line Limit: No Increased Core Flow: No Flow Point Analyzed: N/A I Feedwater' Temperature Reduction: No ARTS Program: No ,

Maximum Extended Operating Domain: No l 4

l 9 fl l

I

i 23A3855 Rev. O j 1

1

10. CORE-WIDE' TRANSIENT ANALYSIS RESULTS (S.2.2.1)

Hethods Used: GEMINI Flux ACPR Q/A Transient (% NBR) (% NBR) BP/P8x8R GE8x8EB Figure Exposure Range: BOC8 to EOC8 Inadvertent HPCI 123 119 0.15 0.15 2 Exposure Range: BOC8 to EOC8-2000 mwd /ST Load Rejection Without 379 118 0.15 0.15 3 Bypass Feedwater Controller 107 105 0.04 0.04 4 Failure .

Exposure Range: EOC8-2000 mwd /ST to EOC8 Load Rejection Without 381 122 0.20 0.20 5 Bypass i Feedwater Controller 153 107 0.05 0.05 6 Failure

11. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT

SUMMARY

(S.2.2.1)

Limiting Rod Pattern: Figure 7 Rod Block Rod Position aCPR MLHGR (kW/ft)

Reading (feet withdrawn) BP/P8x8R GE8x8EB BP/P878R GE8x8EB 104 4.5 0.15 0.15 14.08 15.08

~

j 105 5.0 0.16 0.16 14.52 15.52 106 5.5 0.18 0.18 14.90 15.90 '

107 5.5 0.18 0.18 14.90 15.90 108 6.0 0.19 0.19 15.24 16.24 109 8.5 0.24 0.24 16.28 17.28 110 9.5 0.2* 0.24 16.28 17.28 Setpoint Selected: 107 i'

10

23A5855 Rev. O v

i

12. CYCLE MCPR VALUES (S.2.2)

Non-Pressurized Events BP/P8x8R GE8x8EB Exposure Range: BOC8 to EOCS Inadvertent HPCI 1.22 1.22 Fuel Loading Error --

1.20 Rod Withdrawal Error 1.25 1.25 Pressurization Events Option A Option B BP/P8x8R GE8x8EB BP/P8x8R GE8x8EB

~

Exposure Rang'e: BOC8 to EOC8-2000 mwd /ST Load Rejection Without Bypass 1.32 1.32 1.25 1.25 Feedwater Controller Failure 1.16 1.16 1.14 1.14 Exposure Range: EOC8-2000 mwd /ST to EOC8 Load Rejection Without Bypass 1.33 1.33 1.29 1.29 Feedwater Controller Failure 1.17 1.17 1.14 1.14

13. OVDlPRESSURIZATION ANALYSIS SUNY (S.2.3) s1 y Transient (psig) (osig) Plant Response MSIV Closure 1213 1251 Figure 8 (Flux Scra'm) 11 I

l 23A5855 Rev. 0 i

14. LOADING ERROR RESULTS (S.2.5.4)

Variable Water Gap Disoriented Bundle Analysis Yes*

i Event ACPR Disoriented 0.11 1

15. CONTROL ROD DROP ANALYSIS RESULTS (S.2.5.1)

Bounding Analysis Results Doppler Reactivity Coefficient: Figure 9 Accident Reactivity Shape Functions: Figures 10 and 11 Scram Reactivity Functions: Figures 12 and 13

- Plant-Specific Analysis Results:

Resultant Peak Enthalpy, Cold: 139.4 Resultant Peak Enthalpy, HSB: 192.7 ,

16. STABILITY ANALYSIS RESULTS (S.2.4)

Rod Line Analyzed: Extrapolated Decay Ratio: Figure 14 Reactor Core Stability Decay Ratio, x2 /*0: 0.80 Channel Hydrodynamic Performance "reay Ratio, x2 /*0 Channel Type -

BP/P8x8R ~ -

0.31 GE8x8EB 0.28

  • aCPR penalty'of 0.02 for the tilted disoriented bundle is applied to the cycle MCPR value reported in Section 12.

i 12

23A5855 Rev. 0

17. LOSS-OF-C'00LANT ACCIDENT RESULT (S.2.5.2) i See " Loss-of-Coolant Analysis Report for Brunswick Steam Electric Plant Unit No. 2," NEDO-24053, September 1977 (as amended), and NEDE-24053-P, August 1987.

S 9

i, t

1 l l l

i!

13

? ______ a

\

t 23A5855 Rev. O I

~

l

.1 e

@MMMMM@

e a s @ e @m@ e e @ @ @,@

4s @@im@l@@+0m'@@iscirri@@+@@T@@

HMMMMMMMMMM 4
@MMMMMMMMMMME l
MMMMMMMMMMMMM .
MMMMMMMMMMMMM 23- Et@ @ @ Ci@ @i@ @t@ @ E @ @ @ @ @i@ @ @ @ @ @ @ T @  ;

26- @T@ @ @ @T@ @T@ @T@ @ @ @ @ @ @ @T[ @ @ @ @ @ @

24-Es@Ym@,c@,s@Ym@,@@YmmY@mY@@Y@@@Y 22'-@Y@ @ O @TO @TD @ @ @TE @ @ @ @ @ % , @ @TE

MMMMMMMMMMMMM ~

ll "MMMMMMMMMMM" 8

MEMMMMMMMMM

@ @,@ @ @ @ @ @i@ @ @ @g@ @ @ @ @ @g@ @

)

G @T@ @ @ @ $ @T@ @ @ @l@ @ @ @ @ @ @

l "MMMMM" l l l l l l l l l l 1 3 57 9111315171921232527293133353739414345474951 FUEL TYPE A = BD323A D- 8DRE299 (Cycle 6) '

B = P8DRB265H E

  • E SDRB299 (Cycle 7)

C.- BD317A Figure 1. Reference Core Loading Pattern 14

._-_--__A

23A5855 Rev. 0 1 NEUTRDN FLUX 1 VESSEL PRESS RISE (psi)

, 4'2 AVE SU9 FACE HEAT FLUX 4-4, 2 REUEF VALVE FLOW 4 3 CORE INLET FLOW 3 BYPASS VALVE FLOW 150 150 _

4 CORE INLET SUB 4 HPCI FLOW (% of fw)

S 4

_2 M 2 -14 w2-1---o-t 2 y 3 2

a::

o 100 - 3 -b 3 :3 33 100 o

5 k

50 50 ,

4_4 4- 4 4- 4-4-4-4-4 g- 1--- _1 -1 1 0' O 4~3 2-3-2 3-2--3 2-3-2 3-2-3 2 0 50 100 0 50 100 TIME (seconds) TIME (seconds) 1 LEV'EL (inch-REF-SEP SKRT) 1 VOID REACTIVITY 2 VESSEL STEAMFLOW 2 DOPPLER REACTIVITY 150 3 TURBINE STEAMFLOW .

1 3 SCRAM REACTIVITY 4 FEEDWATER FLOW 4 TOTI._ REACTIVITY

% I

.- 1

- - 3 3 -32-3 2 - 32 5 100 . O y[da c--3 -3 4 4 4-34

\ 4 4_444-4-4-4-4-4-4 s 2 2- 2- 2 --- 2 o

O 50 E _1 7'1 1-~1 1 1 E

O -2 0 50 100 0 50 1CO 1

, TIME (seconds) TIME (seconas)

Figure 2. Plant Response to Inadvertent Startup of HPCI 15 l _ _ _ - _ _ _ _ _ _

4

?

23A5855 Rev. 0

.1 NEUTRON FLUX 1 VESSEL PRESS RISE (psi) 2 AVE SURFACE HEAT FLUX 2 SAFETY VALVE FLOW 3 CORE INLET FLOW 3 RELIEF VALVE FLOW 150 300 4 BYPASS VALVE FLOW 6

w q l 1 .'

a: 2 a

v s

'w 100,, ,-

3

's 200 1

c 2 3%

'E 3%3 I 50 N2

'2 100 3 k c.

r, 1

1 1 0 Oi r4 2 4 2- /2- 4-2 4 4-4 0 2 '4 '6 0 2 4 6 TIME (seconds) TIME (seconds) 1 LEVEL (inch-REF-SEP-SKRT) fV 1 VOID REACTIVITY

2. DOPPLER REACTIVITY 2 VESSEL STEAMFLOW  ?

j ' 3 SCRAM REACTIVITY 200 3 TURBINE STEAMFLOW 4 FEEDWATER FLOW 4 TOTAL REACTIVITY 4

O m }) 2

/

E 100 -

4 w

2 o i

1 '2./

/

q ' 2

^%-24 d.- 2.

2 2 '

2' O

2 l

\ #

3 1

1 h

h O

3-3 3 3 3-3 -1 N .~..

~0

~

4 6 2 4 6 o 2 TIME (seconds) TIME (secor>Js)

Figure 3. Plant Response to Generator Load Rejection Without Bypass  !

(ZOC8-2000 mwd /ST) i 16

23A5855. Rev. 0 l

1 NEUTRON FLUX -

1 VESSEL PRESS RISE (psil 2 AVE SURFACE HEAT FLUX 2 SAFETY VALVE FLOW 3 CORE INLET FLOW  !

150 _ 3 RELIEF VALVE FLOW 4 CORE INLET SUB 4 SYPASS VALVE FLOW

'5 4 . 100 o' a3- 3-3 s 100 = 2- 3 g-- g D' 2 '"

g .

4

.gt 50

{. 4 2

2 0,M4- 3 4 2-3 4 4 32-3-23 1 0 1 1

0 50 100 0 50 100 TIME (seconds) TIME (seconds)

)

1 LEVEL (inch-REF SEP SKRT) 1 VOID REACTIVITY 2 VESSEL STEAMFLOW 2 OOPPLER REACTIVITY 150 3 TURBINE STEAMFLOW . _j 3 SCRAM REACTIVITY' 4 4- 4 FEEDWATER FLOW

  • 4 TOTAL REACTIVITY 1

a j 2 2 f b 1 100 3 p O' r.2 h 2 2-j 2 2

[>

1 50 'l' '

-1 g .

\ I 4 3 0 3 34-3 -2 0 -

50 100 0 50 100 TIME (seconds) TIME (seconds) a l

i l

)

Figure 4. Plant Response to Feedwater Controller Failure (EOC8-2000 mwd /ST) 11 17

h 23A5855 Rev. 0 l

1 NEUTRON RUE 1 YESSEL PREshi R!$E(P$1) 2 AVE $URF 4CE M AT FLUX 2 SAFETY VALVE FLCW 150 300 i

i.-100 g

y -

200 z

s N ^

50 100 0  :

0 -- - ~

0-2 4 6 0 2 4 6 O

TIME (seconds) TIME (seconds) )

! LEVEL (INCH-PEF.SEP.SKRT) I votD REACTINITY 2 00PPLER RE AtitVtTY I 2 VESSEL STEANFLOW

~

200 1 V

l l

100-z 0 0

" ' - [ e

[

u N r

--i 3 _.

O {fg-  :  : _

@ _1 l

\

5 m

-100 -2 O 2 S 6 0 2 4 6 TIME (seconds) TIME .asconds) l Figure 5. Plant Response to Generator Load Rejection Without Bypass (EOC8) 1; 13

23A5855 Rev.'0 1 VE5 EEL PRESS RISECPSI)

I MEurRON FLUX 2 $AF ETY YALVE FLOW 2 AVE SURFACE M AT FLUX ,

.3 REL (EF VALYE FLOW 3 EORE [P4.ET FLOW 4 ernss VALVE FLOW 150

- I c-100

{m A o 2 1[ k r 100 o 5

O E 50 1

50 -

3 ,

0;4 4  ; - :: I::: : ::  ;

0 10 20 0 10 20 0 TIME (seconds) TIME (seconds) t t VOID REACTI O ITY ILEv L(tNCH REF.$EP.$KRf) 2 OOPPLER RE ACT!v!TY 2 WE5 EL STEAMFLCW 3 ScapM REACT vtTY 3 FUR $1NE STEAMFLOW ' " ' " " ' ' ~ ' ' ' ' '

1' -

150 ' "P ' - " " 1 ~

.~- >

$ l

. f 3 7

. d 100- ,

04 .

. , g 2

O U '

l

  1. -1 50 v  ! l g ,

a:

1

\ k

-2 <-

0-  : . .,: 3 20 10 20 0 10 l O

TIME isecondst t: TIME (seconds) h

.{

)'

I Figure 6. Plant Response to Teedwater Controller Failure (EOC8) l 19 l

__--___--______a

23A5855 Rev. O l-2 6 10 14 18 22 26 30 34 38 42 46 50 52 18 18 48 36 44 44 44 36 44 14 2 10 10 2 14 .

40 44 44 44 44 -

44 i

36 .2 .10 2 2 10 2

~

32 44 44 44 44 28 -

10 2 0 0 2 10 24 44 44 44 44 l 20 2 10 2 2 10 2 16 44 44 44 44 44 12 14 2 10 10 2 14 8 36 44 44 44 36 4 18 18 NOTES: 1. NUMBER INDICATES NUMBER OF NOTCHES WITHDRAWN OUT OF 48. BLANK IS A WITHDRAWN ROD.

2. ERROR R00 IS (22,28).

Figure 7. Limiting Rod Pattern 20 ,

l i

23A5855 ,

Rey, 0 v.

1 NEUTRON F UX 1 VESSEL PPESS RISE (PSI)

, 2 AVE SURFAZ MEAT FLUX 2 SAFETY YALVE FLOW 3 CORE IPLET FLOW 3 RELIEF VdVE FLOW

{ 300 '""'""""E~

150 8

E e ^

200

  • 100 -

w.

a '

e_

  • 50 100 , . , g

{  ;

O 0; ..:=  :  :  : _J 0 5 0 5

  • . TIME (seconds) TIME (seconds) 1 LEVEL (INde.REF.$EP.$KAT) 1 VQ10 REAC!T!Y!TV 2 3DPPLER RE ACTIV!TT

-200 2 VESSEL STEAMFlow

? ta 1 W V.'

5 100 ,

A g O. .- ..-- 7

's  :

s 8

0 - - - -

{ -1 5

x

-100 -2 TIME (seconds) TIME (seconds)

I Figure 8. Plant Response to MSIV Closure - Flux Scram i

21 .

2 4

23A3855 Rev. 0

0. 0 l

l 1

-5. 0 l h A

-10.0

-l

\ Y -,5. o /r

/

3-

/

~

3 '

] -20.0 IJ 3

)

-25.0 3

J

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)

3

-30.0

-35.O A C ALCQLA s tu VALut -LULU B CALCULATED VALUE -HSB C BOUNQ VAL 290 C/ L/G COLD D BOUNC' VAL 280 C/ L/G HS8

-40.0

0. 0 500.0 1000.0 1500.0 2000.0 2500.0 3000.0 FUEL TEMPERATURE DEG C.

Figure 9. Fuel Doppler Coefficient in 1/A*C 0

22

23A3855 Rev. O l

20.0 i

17.5 15.0 ,

I 4

m o

JD C

~

D i, 12.5 z

F-

.J. .

w 10.0 ,,

, f, o

E

> 7.5

~

u w

c:

5. O c 2.5  ;

A ACCIDENT i: UNCTION B BOUNDING '/ALUE 280 CAL /G

0. O n
0. 0 5.0 10.0 15.0 20.0 ROD POSITION, FEET DUT l Figure 10. Accident Reactivity Shape Function - Cold Startup 23

23A5855 Rev. 0 20.0 17.5 t5.0 ,.

m o - mm n .

g "

w 12.5 I .#"

7 e z

0--

J - .

w 10.0 o /

C==

7.5 e u

w QC

5. 0 p 2.5 3 A ACCIDENT FUNCTION 8 BOUNDING '/ALUE 280 CAL /G
0. 0 ,,
0. 0 5.0 10.0 15, 9 20.0 ROD POSITION, FEET OUT Figure 11. Accident Reactivity Shape Function - HSS l..

'4 24

23A3855 Rev. 0 30.0 A SCRAM FLNCTION G BOUNDING VALUE 280 CAL /G 25.0 m

o 1 -

W

  • 20.0 m H

_J w

CD

~ '

^ 15.0 o

w Z

s W

10.0 H

U w

cc 5.0 /

0 . 0 ,; , , ,

J.

0. 0 1. 0 2. 0 3. 0 4.0 5. 0 6. 0 ELAPSED TIME, SECONDS Figure 12. Scram Reactivity Function - Cold Startup 25

23A5855'- Rcy. 0 1 I

)

i 50.0 J A. SCRAM FLNCTION I

B BOUNDING VALUE 280 CAL /G i i

40.0

/

l m 1 O ,, '

I W

Z

< 30.0 l

w l J 4 W

o * ,

i

- i n

co I w i 1

z w

m

- 20.0 W

W U

w CC 10.0 0.0 ~

.0- . 1. 0 2.0 3.0 4.0 5.0 6.0 ELAPSED TIME, SECONDS Figure 13. Scram Reactivity Function -- HSB .

26

23A5855' Rev.'0 1.25 A NATURAL C!RCULATION B 105 PERCENT ROD LINE 1.00

,A

'O.75 ,

o. AB 5

p . >

.g E

. =

k 8

o 0.50 3

{

l A

0

\

0.25 9-B 0 .

O 20 40 60 80 100 120 POWER (%)

Figure 14. Reactor Core Decay Ratio Versus Power 27/28 -

23A5855 Rev. O APPENDIX A LIMITING CONDITIONS FOR OPERATION This appendix provides the limiting condition for operation (LCO) for each of the power distribution limits identified below:

(1) Average Planar Linear Heat Generation Rate (APLHGR)

(2) Operating Limit MCPR (3) APRM Setpoints Surveillance requirements and required actions are specified in the Tech-nical Specifications. The power distribution limit bases are given in Appen-dix B.

A.1 APLHGR .

During steady-state power operation, the APLHGR for each type of fuel as l

a function of axial location and average planar exposure shall not exceed limits based on applicable APLHGR limit values which have been approved for the respective fuel and lattice types determined by the approved methodology described in GESTAR-II (NEDE-24011-P-A). When hand calculations are required, the APLHGR for each type of fuel as a function of average planar exposure i shall not exceed the limiting value for the most limiting lattice (excluding natural uranium) shown in Figures A-1 through A-4, during two recirculation loop operation.

A.2 OPERATING LIMIT MCPR The fuel cladding integrity safety limit MCPR is 1.07. During steady-state power operation, the MCPR for each type of fuel shall not be less than the limiting value (shown in Table A-1) times the Kf (shown in Figure A-5),

for two recirculation loop operation.

29

_..7_..

__; _7 __ 7_ - .____

_ _j _ _ _ _

1 i

l 23A3855 Rev. 0 l

l In reference to Technical Specification 3.2.3.2, the OLMCPR for T,y, less than or equal to T is the greater of the non pressurization transient or 3

the Option B OLMCPR (Table A-1), where T,y,and TD are given by:

i n  !

}] Ng i=1 T

g j ave n

}[ N g i=1 where:

l i = Surveillance test number.

I n a Number of surveillance tests performed to date in the cycle (includ-ing BOC).

Ng =

Number of rods tested in the i th surveillance test. I

=

T g Average scram time to notch 36 for surveillance test 1.

and 1

T3 = a + 1.65 [N-- \

y a

g i/

where:

Ng =

Number of rods tested at BOC.

u = 0.813 seconds (mean value for statistical scram time distribution from de-energi:ation of scram pilot valve solenoid to pickup on notch 36).

o = 0.018 seconds (standard deviation of the above statistical distribution).

l 30

23A5855 Rev. O

{

in reference to Technical Specification 3.2.3.2, the OLMCPR for r,y, j greater than T shall be either:

B

a. The greater of the non pressurization transient (Table A-1) or the adjusted pressurization transient MCPR (MCPRadj) where:

r T i Option B Option A

~

Option B adj g -r B 1 TA = 1.05 seconds (control rod average scram insertion-time limit to notch 36),

and MCPR as given in Table A-1 0ption A MCPR as ghen in Table A-1 0ption B or s l

b. MCPR as ghen in Table A-1.

0ption A A.3 APRM SETPOINTS The flow-biased APRM scram trip setpoint (S) and rod block trip setpoint (SRB) shall be:

S 1 (0.66% + 54%)T, and i Sg 1 (0.66W + 42%) T; 1

where S and S RB are in percent of rated thermal power; W s loop recirculation flow in percent of rated flow.

T is the ratio of Fraction of Rated Thermal Power (FRTP) divided by Core Maxi-mum Average Planar Linear Heat Generation Rate Ratio (CMAPEAT):

l T=C I where T 1 1.

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l' 36

23A5855 Rev. O Table A-1 hCPRs Fuel Type: P8x8R. BP8x8R, and GE8x8EB Non-Pressurized Transient MCPR = 1.25 Pressurization Transients Exposure Range MCPR 0ption A Option L

-)1 BOC8 to E0C8-2000 mwd /ST 1.32 1.25

~

E0C8-2000 mwd /ST to EOC8 1.33 1.29 l

l i

l l 1

I l

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37/38

l l

23A3855 Rev. Of APPENDIX B BASES FOR LIMITING CONDITIONS FOR OPERATION This appendix provides the bases for each of the power distribution limits identified in Appendix A.

B.1 APLHGR i

This specification assures that the peak cladding temperature (PCT) fol-loving the postulated design basis loss-of-coolant accident (LOCA) will not ,I exceed the limits specified in 10CFR50.46 and that the fuel mechanical design analysis limits specified in Reference B-1 will not be exceeded.

Thermal Mechanical Design Analysis: NRC approved methods (specified in

. Reference B-1,) are used to demonstrate that all fuel rods in a lattice oper-ating at the bounding power history meet the fuel design limits specified in Reference B-1. No single fuel rod follows, or is capable of following, this bounding power history. This bounding power history is used as the basis for the fuel design analysis APLHGR limit.

LOCA Analysis: A LOCA analysis is performed in accordance with 10CFR50, Appendix K to demonstrate that the permissible planar power (maximum APLHGR) limits comply with the ECCS limits specified in 10CFR50.46. The analysis is performed for the most limiting break size, break location, and single failure combination for the plant. The methods used are discussed in Reference B-2.

The APLEGR limit is the most limiting composite of the fuel design analy-sis APLHGR limit and the ECCS APLHGR limit.

B.2 0FLP.ATING LIMIT MCPR The required operating limit MCPRs at steady state operating conditions as specified in Appendix A are derived from the established fuel cladding integrity safety limit MCPR specified in Appendix A and an analysis of s 1

1 39

w 23A5855 Rev. 0 abnormal operational transients. For any abnormal operating transient analy-sis evaluation with the initial condition of the reactor being at the steady-state operating limit, it is required that the resulting MCPR does not decrease below the safety limit MCPR at any time during the transient, assum-ing inscrument trip setting as given in Specification 2.2.1 of the Technical Specifications.

To assure that the fuel cladding integrity safety limit is not exceeded during any anticipated abnormal operational transient, the most limiting tran-sients have been analyzed to determine which ones result in the largest reduc- --l tion in Critical Power Ratio (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insert' ion, and coolant temperature decrease. I I

. The codes used to perform the transient analyses that serve as the basis for the operating limit MCPR are described in Reference B-1. Conditions at -

limiting exposures are used for nuclear data to provide conservatism relative i to core exposure aspects. Plant-unique initial conditions and system param- l eters are used as inputs to the transient codes. The ACPR' calculated by the transient codes is adjusted using NRC approved adjustment factors to account for code uncertainties and to provide a 95/95 licensing basis.

The limiting transient yields the largest ACPR. The ACPR for the limiting transient is added to the fuel cladding integrity safety limit to MCPR to establish the minimum operating limit MCPR. ."

The purpose of the gK factor is to define operating limits at other than rated flow conditions. At less than 100:: flow, the required MCPR is the product of the operating limit MCPR and the Kg factor. Specifically, the Kf factor provides the required thermal margin to protect against a flow increase transient. The most limiting transient initiated from less than rated flow conditions is the recirculation pump speedup caused by a motor generator speed control failure.

40

23A5855 Rev. O I

I j

For opera' tion in the automatic flow control mode, the gK factors assure j that the operating limit MCPR in Appendix A will not be violated should the l most limiting transient occur at less than rated flow. In the manual flow control mode, the K factors assure that the safety limit MCPR will not be f

violated should the most limiting transient occur at less than rated flow. I Ihe K factor values are generically developed as described in f

Reference B-1. ,

l The K factors are conservative for the General Electric plant opera-f ,

tion because the operating limit MCPRs in Appendix A are greater than the original 1.20 operating limit MCPR used for the generic derivation of K .

f At core thermal power levels less than or equal to 25%, the reactor will l ,

be operating at minimum recirculation pump speed and the moderator void con-1 I tent will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience indicated that the result-ing MCPR value is in excess of requirements by a considerable margin. With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR. During initial startup testing of the plant, a MCPR evaluation will be made at 25% initia]

l power level with minimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary. The daily requirement for calculating MCPR above 25: rated thermal power is sufficient, since power distribution shifts are very slow when there have not been significant power or control rod changes. I 1

l The requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in power or power shape, regardless of magnitude that could place operation at a thernal limit.

41 l

23A3855 Rev. O s

B.3 APRM SETPOINTS The flow-biased thermal power upscale scram setting and flow-biased neu-tron flux upscale control rod block functions of the APRM instruments are adjusted to ensure that fuel design and safety limits are not exceeded. The scram setting and rod block setting are adjusted in accordance with the for-mula in Appendix A when the combination of Thermal Power and CMAPRAT indicate a highly peaked power distribution. This adjustment may be accomplished by increasing the APRM gain and thus reducing the slope and intercept point of the flow referenced APRM high flux scram curve by the reciprocal of the APRM .

gain change.

t B.4 REFERENCES

1. " General, Electric Standard Application for Reactor Fuel", NEDE-24011-P-A (latest approved revision).
C
2. " General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50, Appendix K", NEDO-20566, January 1970.

l 1

l 42

23A5855 Rev. 0

(

~

APPENDIX C SAFETY-RELIEF VALVE OUT-OF-SERVICE The analysis was performed for safety-relief valve out-of-service and there was no change in the ACPR. The change in pressure for MSIV with flux scram is shown below.

P,7 P l

y (psig) (psig) Plant Response MSIV Closure 1225 1261 Figure C-1 .!

(Flux Scram) ,

1 I

l l

0 43

23A5855 Rev.'O 1

1 NEUTRON FLUX 1 VESSEL P'RESS RISE (psi)-

2 AVE SURFACE HEAT FLUX 2 SAFETY VALVE FLOW 15 0. 0

{ 3 CORE INLET FLOW 3 RELIEF VALVE FLOW 300.0 4 BYPASS VALVE FLOW

\

F 2

100.0 2 y2 3 200.0 'I 'l 3

E 3\ 2 s Q 3

i 30.0 \ %3 2 300.0 3-3 3 3- -3

'*" 1- 0'8

- 1

0. 0 43 324-242 24--4-2-4 4 5.0 0. 0 5. 0 TIME (SECONDS)

, TIME (SECONOS1 Z

1 LEVEL (inch-REF-SEP.SKRT) ( 1 VOlO REACTIVITY 2 VESSEL STEAMFLOW 2 DOPPLER REACTIVITY 2: 1 3 TURBINE STEAMFLOW ~"~

g,0 3 SCRAM REACTIVITY 4 FEEDWATER FLOW / 4 TOTAL REACTIVITY

/ -

4 4 4

. .4 m42 .,444-44-4 0.0 4 4 4 _ 43. 3 2

'3 h 2-4 2 ** 4 c::::-- 2 { 2'3 [

\ 4

  • 3 3 - 'QV 3 a . ,) 2> 3s 3 -- 3 g ~. 9

\3 4

101.0 '

-2.0

0. 0 5.0 0. 0 5, o fiME (SECONOS)

TIME (SECONOS) l

)

l 4 I l

Figure C-1. Plant Response to MSIV Closure - Flux Scram (SRV005) 1 44 l l

l j

23A5855 Rev. O i l

I J

AhPENDIXD j PIANT PARAMETER DIFFERENCES l

i GETAB and Transient Analysis Initial Conditions l The values used in the GETAB and Transient Analysis are given in Table D-1. The following values differ from the values reported in j Tables S.2-4.1 and S.2-6 in NEDE-24011-P-A-8-US, May 1986.

i l

l Table D-1 ~!

PLANT PARAMETER i

Parameter Analysis Value NEDE-24011 Value l

Dome Pre'esure 1005 1020 1 2 psi Rated Steam Flow 10.47 10.96 1 0.2% )

Turbine Pre.sure 950 960 1 2 psi Non-Fuel Power Fraction 0.039 0.04

,l I

i i

l

. e  !

45/46 l

23A5855 Rev. O APPENDIA E TURBINE CONTROL VALVE CONFIGURATI0N The transient GETAB analyses presented in the body of this report are based on turbine control valves in a full-arc configuration and on the power supply to the recirculation Motor-Generator Sets from offsite power.

0 l

l l

l l

l l

J -

l 47/48 l 1

I I

23A5855 Rev. 0 APPENDIX F USE OF GEMINI METHODS FOR CYCLE 8 The analyses required for this cycle vere performed with GE's advanced reload licensing methods, known as GEMINI. Any differences between this reload and the previous one are due not only to cycle differences, but also to the difference in the methods. Therefore, making direct comparisons between the two cycles will be inconclusive.

l l

l I

J J

s 49/50 (FINAL)