ML20142A528

4 to Updated Final Safety Analysis Report, Appendix 15A, Purpose

ML20142A528
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Issue date:	05/21/2020
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APPENDIX 15A 15A.l PURPOSE The purpose of this appendix is to present assumptions and computer codes used in the analysis of certain accidents treated in Section 15.

15A. 2 FORMAT Each accident discussed will be presented in the order used in Section 15.

Within each section, the following topics will be discussed: mathematical model, transport assumptions, computer codes used, and dose assumptions.

Table 15A-l presents the isotopic core inventory for various decay times.

Table 15A-2 presents the radionuclide concentrations used in the accident analyses.

15A.3 CONTROL ROD DROP ACCIDENT (Section 15.4.9)

This Section has been deleted.

Equations 15A-1 through 15A-24 have been deleted.

Pages lSA-2 through lSA-10 have been deleted.

15A.4 INSTRUMENT LINE FAILURE ACCIDENT (Section 15.6.2) 15A.4.1 Mathematical Model 15A.4.1.1 Design Basis Analysis The design basis analysis is presented in Section 15.6.2.5.2.

15A.4.2 Transport Assumptions 15A.4.2.1 Design Basis Analysis The transport assumptions are discussed in Section 15.6.2.5.2.2.

15A.4.2.2 Deleted HCGS-UFSAR 15A-l Revision 16 May 15, 2008

I 15A.4.3 Computer Codes Used The computer code used to model transport in the Reactor Building and to the environment is the RADTRAD 3.02 computer code (NUREG/CR-6604}.

15A.4.4 Dose Assumptions The breathing rates used in this Regulatory Guide 1.183, Revision 0.

0 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> analysis of this accident are taken from 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Specifically, they are:

3.50E-4 m3 /s

1. SOE-4 m3/s

2. 30E-4 m3/s The iodine dose conversion factors used for the thyroid inhalation and whole body doses for an adult were taken from Federal Guidance Reports (FGR}

11 and 12 respectively.

15A.5 STEAM LINE BREAK ACCIDENT (Section 15.6.4) 15A.5.1 Deleted 15A.5.2 Transport Assumptions The transport assumptions are described in Sections 15.6.4.5.2.

15A.5.3 Computer Codes Used The computer code used to model transport to the environment is the RADTRAD 3.02 computer code (NUREG/CR-6604).

15A.5.4 Dose Assumptions The dose assumptions described in Section 15A. 4. 4 apply to this accident as well.

15A.6 LOSS-OF-COOLANT ACCIDENT (Section 15.6.5) 15A-2 HCGS-UFSAR Revision 16 May 15, 2008

15A.6.1 Transport Assumptions The reactor building exhaust rate is modeled as a step function.

The value assumed for each step is the value the function would have at the beginning of the time period.

The approximation continues until time t = 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, at which time the exhaust rate is assumed to be a constant value of 3324 cfm.

The exhaust rate for each time step is doubled to account for 50% mixing in the Reactor Building and 10% is added to account for flow variation.

Initially, the FRVS exhaust rate is 19,800 cfm.

The ESF leakage is modeled assuming a constant leakage rate of 2 gpm.

assumed that 10% (that is, 0.2 gpm) of the leakage becomes airborne.

15A.6.2 Computer Code Used It is The RADTRAD 3.02 computer code (NUREG/CR 6604) was used to calculate the off-site dose consequences of primary containment leakage, ESF leakage outside the primary containment, and MSIV leakage.

15A.6.3 Dose Assumptions The breathing rates used in this Regulatory Guide 1.183, Revision 0, analysis of this accident Specifically, they are:

are taken from 0 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 3.5E-4 m3/s

1. 8E-4 m3/s

2. 3E-4 m3 /s 15A.7 FEEDWATER LINE BREAK ACCIDENT (Section 15.6.6) 15A.7.1 Mathematical Model 15A.7.1.1 Design Basis Analysis The design basis analysis is presented in Section 15.6.6.

15A.7.2 Transport Assumptions 15A.7.2.1 Design Basis Analysis All assumptions are described in Section 15.6.6.5.

15A-3 HCGS-UFSAR Revision 16 May 15, 2008 I

I 15A.7.3 Computer Codes Used The computer code used to model the activity transport to the environment is RADTRAD 3.02 (NUREG/CR 6604).

15A.7.4 Dose Assumptions The dose assumptions described in Section 15A. 4. 4 apply to this accident as well.

15A.8 WASTE GAS SYSTEM FAILURE ACCIDENT {SECTION 15.7.1)

All information is presented in Section 15.7.1.

15A. 9 LIQUID RADWAS'rE TANK FAILURE ACCIDENT (SECTION 15.7. 3)

All information is presented in Section 15.7.3.

15A.l0 FUEL HANDLING ACCIDENT (SECTION 15.7.4) 15A.10.1 Mathematical Model The model describing Section 15.7.4.9.2.

the 15A.10.2 Transport Assumptions transport of activity is described The transport assumptions are described in Sections 15.7.4.9.1 and 15.7.4.9.2.

15A.l0.3 Computer Codes Used in The RADTRAD computer code {Version 3. 02) was used to calculate the off-site radiological consequences of a fuel handling accident.

15A.l0.4 Dose Assumptions The dose assumptions described in Section 15A. 4. 4 apply to this accident as well.

lSA-4 HCGS-UFSAR Revision 16 May 15, 2008

Isotope HCGS-UFSAR TABLE 15A-1 CORE INVENTORIES FOLLOWING SHUTDOWN, Ci 0 Min 30 Min 1 of 1 1440 Min Revision 0 April 11, 1988 Security Related Information withheld Under 10 CFR 2.390

Isotope Noble Gases HCGS-UFSAR TABLE 15A-2 RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT AND MAIN STEAM(l), ~C/g (FROM GALE CODE)

Reactor Coolant Design Basis(2) 1 of 2 Reactor Steam Design Basis(2)

Revision 0 April 11, 1988 Security Related Information withheld Under 10 CFR 2.390

TABLE lSA-2 (Contd)

(1) The reactor coolant concentration is specified at the nozzle where reactor water leaves the reactor vessel.

Similarly, the reactor steam concentration is specified at time 0 at the nozzle.

(2) *Design basis concentrations correspond to 350,000, p.Ci/s 30 min.

(3)

All iodine concentrations have been adjusted lower to account for the reduced I -131 source term, which was reported in Revision 1 of NUREG-0016.

2 of 2 HCGS-UFSAR Revision 0 April 11, 1988

APPENDIX 15B SPECIAL ANALYSIS 158.1. INTRODUCTION An analysis of the transient caused by continuous control rod withdrawal in the startup range (Section 15.4.1.2) was performed to demonstrate that the licensing~basis criterion for fuel failure will not be exceeded when an out of sequence control rod is withdrawn at the maximum allowable normal drive speed.

The sequence and timing assumed in this special analysis is shown in Table 158-1..

The rod worth minimizer (RWM) constraints on rod sequence will prevent the continuous withdrawal of an out of sequence rod.

This analysis was performed to demonstrate that, even for the unlikely event where the RWM fails to block the.continuous withdrawal of an out of sequence rod, the licensing basis criterion for fuel £ailura is still satisfied.

The methods and design basis used for performing the detailed analysis for this event are similar to those previously approved for the control rod drop accident (CRDA)

(References.lSB-1/ 15B-2, and 15B-3). Additional, simplified point.model kinetics calculations were p.erformed to evaluate the dependence of peak fuel en.thalpy on the control blade worth.

The licensing basis criterion for fuel failure is that the contained energy of a £uel pellet located in the peak power region of the core shall not exceed 170 cal/g-uo2..

15B-1 HCGS-UFSAR Revision 15 October 27, 2006 I

15B.2 METHODS OF ANALYSIS Since the rod worth calculations using the approved design-basis methods (References 158-1, 158-2, and 158-3) use three dimensional geometry, it is not practical to do a detailed analysis of this event by parameterizing control rod worths.

Therefore, the methods of analysis employed were to perform a detaiJed evaluation of this event for a

typical BWR and control rod worth (1.6 percent ~K) and to use a point model kinetics calculation to evaluate the results over the expected ranges of out of sequence control rod worths.

The detailed calculations are performed to demonstrate 1) the consequences of this event over the expected power operating range and 2) the validity of the approximate point-model kinetics calculation.

The point model kinetics calculation demonstrates that the licensing criterion for fuel failure is easily satisfied over the range of expected out of sequence control rod worths.

These methods are described in more detail below.

The methods used to perform the detailed calculation are identical to those used to perform the design basis CRDA with the following exceptions:

1.

The rod withdrawal rate is 3.6 ips (0.3 fps} rather than the blade drop velocity of 3.11 fps.

Although faster withdrawal rates are possible, it would require the failure of the associated control rod drive mechanism or hydraulic control unit

{as described in Section 4.6.2) in addition to the assumed failure of the RWM.

If the associated control rod drive mechanism or hydraulic control unit were assumed to be the worst single failure, then the RWM would terminate the event prior to the full rod withdrawal, or even prior to control rod movement.

2.

Scram is initiated either by the intermediate range monitor {IRM) or by a 15 percent power scram initiated by the average power range monitor (APRM} in the startup range.

The IRM system is assumed to be in the worst bypass condition allowed by technical specifications.

3.

The blade being withdrawn is inserted along with remaining drives at technical specification insertion rates upon initiation of the scram signal.

15B-2 HCGS-UFSAR Revision 15 October 27, 2006

Examination of a number of rod withdrawal transients in the low power startup range using a two-dimensional R/Z model has shown clearly that a higher fuel enthalpy addition would result from transients starting at the 1 percent power level rather than from lower power levels.

The analysis further shows that for continuous rod withdrawal from these initial power levels { 1 percent range),

the APRM 15 percent power-level scram is likely to be reached as soon as the degraded (worst bypass condition} IRM scram.

Consequently, credit is taken for either the IRM or APRM 15 percent power scram in meeting the consequences of this event.

The transients for this response were initiated at 1 percent of power and were performed using the APRM 15 percent power scram.

An initial point kinetics calculation was run to determined the time required to scram based on an APRM scram setpoint of 15 percent power and an initial power level of 1 percent.

From this time and the maximum allowable rod withdrawal speed, itis possible to show the degree of rod withdrawal before reinsertion due to the scram.

From this information, Figure 15B-l, showing the modified effective reactivity shape, was constructed.

The point model kinetics calculations use the same equations employed in the adiabatic approximation described on Page 4-1 of Reference 15B-1.

The rod reactivity characteristics and scram reactivity functions are input identically to the adiabatic calculations, and the Doppler reactivity is input as a

function of core average fuel enthalpy.

The Doppler reactivity feedback function used in the point model kinetics calculations was derived from the detailed analysis of the 1.6 percent rod worth case described above.

This is a conservative assumption for higher rod worths since the power peaking and hence spatial Doppler feedback will be larger for higher rod worths.

As will be seen in the results section, maximum enthalpies resulted from cases initiated at 1 percent of rated power.

In this power range, the APRM will initiate scram at 15 percent of power; hence, the APRM 15 percent power scram was used for these calculations thereby eliminating the 15B-3 HCGS-UFSAR Revision 0 April 11, 1988

need to perform the spatial analysis required for the IRM scram.

inputs are consistent with the detailed transient calculation.

All other The point model kinetics calculations result in core-average enthalpies.

The peak enthalpies were calculated using the following equation:

where:

h h

h 0

h f

(P/A)

T h

+ (P/A)

(h o

T f

h ) f 0

final peak fuel enthalpy, initial fuel enthalpy, final core average fuel enthalpy, and total peaking factor (radial peaking) x (axial peaking) x (local fuel pin peaking)Q.

For these calculations, the radial and axial peaking factors were obtained as a function of rod worth from the calculations performed in Section 3.6 of Reference 15B-2 and are shown in Figure 15B-2.

It was conservatively assumed that no power flattening due to Doppler feedback occurred during the course of the transient.

I 15B. 3 RESULTS The reactivity insertion resulting from moving the control rod is shown in Figure 158-1 for the point model kinetics calculations.

The core-average power versus time and the global peaking factors from Section 3.6 of Reference 158-2 are shown in Figures 158-3 and 158-2, respectively.

The results of the point model kinetics calculation are summarized in Table 15B-2 along with the results of the detailed analysis.

158-4 HCGS-UFSAR Revision 15 October 27, 2006

From Figure 158-3 and Table 158-2, it is shown that the core average energy deposition is insensitive to control rod worth; therefore, the only change in peak enthalpy as a function of rod worth will result from differences in the global peaking, which increases with rod worth.

Comparison of the global peaking factors shown in Figure 15B-2 with the values used in the detailed calculation demonstrates that the Reference 15B-2 values are reasonable for their application in this study.

For all cases 1 the peak fuel enthalpy is well below the licensing basis criterion of 170 cal/g.

Cases 4 and 5 of Table 15B-2 show that the point model kinet'i'CS calculations give conservative results relative to the detailed evaluations.

The primary di:fference is, that the global peaking will flatten during the t:t;'ansient due to Doppler _feedback.

This.is accounted _for in the detailed calculation, but the point.mode1.- kinetics

• calculations conservatively assumed t-hat the peaking remains constant.at its initial value.

The dif~ference-s, in core-average and peak enthalpy between cases 1 and 5 are due to the fact that _for case 1 the scram was. initiated by the APRM 15 percent power scram setpoint; whereas, in case 5 the scram was initiated -bY the IRMs.

As can be seen by Figure 15B-4, this would occur at a core-average pow.er of 21 percent.

Since the APRM trip point will be reached first, it is reasonable to take credit for the APRM scram.

1'5B. 4 This Section Deleted HCGS-UFSAR 15B-5 Revision 15 October 27, 2006

1

58.5 CONCLUSION

S The above evaluations of continuous withdrawal of a control rod in the startup range indicate that the peak fuel enthalpies due to the continuous withdrawal of an out of sequence rod in the startup range will be much less than the licensing basis criterion of 170 cal/gm.

In light of the conservative nature of these evaluations and the markedly different fuel designs and vendor methodologies, the substantial margins to 170 cal/gm limit support a generic conclusion that the peak fuel enthalpy associated with continuous withdrawal of a control rod in the startup range in the HCGS core will remain below 170 cal/gm.

15B.5 15B-1 15B-2 15B-3 15B-4 15B-5 HCGS-UFSAR REFERENCES

c. J. Paone, et al 1 "Rod Drop Accident Analysis For Large Boiling Water Reactors", NED0-10527, March 1972.

R. c. Stirn, et al, "Rod Drop Accident Analysis For Large Boiling Water Reactorsn, NED0-10527, Supplement 1, July 1972.

R. C. Stirn, "Rod Drop Accident Analysis F'or Large Boiling Water Reactors, Addendum No. 2, Exposed Cores", NED0-10527, Supplement 2, January 1973.

R.

C.. Stirn, J. F. Klapproth, "Continuous Rod Withdrawal Transient in the Startup Rangen, NED0-23842, April 1978.

Deleted.

l5B-6 Revision 15 October 27, 2006

Time 1&..

0

>0 4

4-8 S-9 10

1.

2..

3..

4.

5.

6.

HCGS-UFSAR TABLE 15B-1 SEQUENCE OF EVENTS FOR CONTINUOUS ROD WITHDRAWAL DURING REACTOR STARTUP Event The reactor is critical and operating in the startup range.

The operator selects and withdraws an out-of-sequence control rod at the maximum normal drive speed of 3.6 ips.

Either the RWM or the second qualified verifier fail to block the selection (selection error) and continuous withdrawal {withdraw error) of the out-of-sequence rod.

The reactor scram is initiated by the IRM system or the APRM system.

The prompt power burst is terminated by a combination of Doppler and/or scram feedback.

The transient is finally terminated by the scram of all rods, including the control rod being withdrawn. Scram insertion times are assumed to be 5 seconds to 90 percent insertion.

1 of 1 Revision 9 June 13, 1998

SUMMARY

OF RESULTS FOR DETAILED AND POINT MODEL KINETICS CALCULATIONS OF CONTINUOUS ROD WITHDRAWAL IN THE STARTUP RANGE Control Rod h (cal/ g)

(P/A) (2)

~

Yorth ($AK) f G

h (call&)

1 1.6 17.3 24.2 42.7 2

2.0 17.3 30.9 50.0 3

2.5 17.2 46.0 58.5 4

1.6(l) 18.3 19.7(3) 56.2 5

1.6<4>

18.3 19.7 59.6 (1) Detailed transient calculation. All other data reported are for point model kinetics calculations.

(2)

(P/A) -global peaking factor (radial x axial).

G (3)

The (P/A)

- 19.7 is the initial value.

For the detailed analysis, this value will decrease during the course of the transient since the power shape will flatten due to Doppler feedback.

(4)

Point model kinetics calculation with an IRK-initiated scram and 3-D simulator global peaking.

1 of 1 HCGS-UFSAR Revision 0 April 11, 1988

.026"

.024

.022

.020

.018

.016

.014

<]

.012

.010

.008

.006

.004

.002 4

I I

I ~SCRAM INSERTS

' 'I CONTROL ROD l II l II

\\ l I 8

12 16 20 2.5% ROD WORTH 2.0% ROD WORTH 1.6% ROD WORTH CONTROL ROD BEING PULLED 24 28 32 36 40 TIME (SECONDS)

REVISION 0 APRIL 11, 1988 PUBLIC SERVICE ELECTRIC AND GAS COMPANY HOPE CREEK NUCLEAR GENERATING STATION POir~T KINETICS CONTROL ROD REACTIVITY INSERTION (4)

UPDATED FSAR FIGURE 158-1

6~------------------------------------------------------

X

<(

)(

50 40

...1 30

<(

0

<(

a:

~

Q.

20 10

,a~

P/A FROM DETAILED ANALYSIS 0--------------------------~------------------------~

1.0 2.0 3.0 CONTROL ROD WORTH (%.6K)

REVISION 0 APRIL 11. 1988 PUBLIC SERVICE ELECTRIC AND GAS COMPANY HOPE CREEK NUCLEAR GENERATING STATION PIA.vs ROD WORTH NED0-10527 SUPPLEMENT 1 (2) AND DETAILED ANALYSIS (4)

UPDATED FSAR FIGURE 158~2

a:

w

~

~

w

~

c{

-l w

a:

w CJ c{

a:

w >

c{

w a:

0 CJ 2.5% RODWORTH

-*-*-*- 2.0% ROD WORTH 1.6% ROD WORTH A

1\\ i' I \\

I I

I

\\

I

\\

I

\\ *'

\\

1 v

• I I\\

I I \\

I

\\

I

.I

\\

I

.I I

I I

I I

.I

/ *'

/

I I

/

./

//./

,o-2

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

2.0 4.0 TIME (SECONDS) 6.0 8.0 REVISION 0 APRIL 11. 1988 PUBLIC SERVICE ELECTRIC AND GAS COMPANY HOPE CREEK NUCLEAR GENERATING STATION CONTINUOUS RWE IN THE STARTUP RANGE CORE AVERAGE POWER vs TIME FOR 1.6%, 2.0% AND 2.5%

WPRTH'S (POINT MODEL KINETICS)

(4)

~

UPDATED FSAR FIGURE 158-3

a:

w

~

0 a..

w c,

<(

a:

w >

<(

w a:

0

(,)

w >

~

<(

...I w

a:

-2~------~--------~------~~-------L ________._ ______ _. ______ ~

10 2

3 4

5 6

7 8

9 TIME (SECONDS)

ASSUMPTtONS:

1.

1.6% ak ROD

2. 0.3 fps WITHDRAWAL VELOCITY

3.

IRM SCRAM FOR WORST BYPASS CONDITION

4.

P0 = 10-2 OF RATED

5.

1967 PRODUCT UNE TECH SPEC SCRAM RATE

6.

EXPOSURE= 0.0 GWDIT REVISION 0 APRIL 11, 1988 PUBLIC SERVICE ELECTRIC AND GAS COMPANY HOPE CREEK NUCLEAR GENERATING STATION CONTINUOUS CONTROL ROD WITHDRAWAL FROM HOT STARTUP (4)

UPDATED FSAR FIGURE 158-4

HCGS-UFSAR APPENDIX 15C HOPE CREEK SINGLE LOOP OPERATION ANALYSIS FEBRUARY 1986 Prepared for PUBLIC SERVICE ELECTRIC AND GAS COMPANY HOPE CREEK GENE.RATING STATION Prepared by GENERAL ELECTRIC COMPANY fWCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE, CALIFORNIA 95125 Hcvjsion 0 Apri 1 11, 1 qgg

15.C 15.C.l 15.C.2 15.C.2.1 15.C.2.1.1 15.C.2.1.2 15.C.2.2 15.C.3 15.C.3.1 15.C.3.1.1 15.C.3.1.2 15.C.3.1.4 15.C.3.2 15.C.3.3 15.C.4 15.C.4.1 l5.C.4.2 lS.C.S 15.C.5.1 15.C.S.2 15.C.5.3 HCGS-UFSAR APPENDIX 15.C TABLE OF CONTENTS RECIRCUTATION SYSTEM SINGLE-LOOP OPERATION INTRODUCTION AND

SUMMARY

MCPR FUEL CLADDING INTEGRITY SAFETY LIMIT

• core Flow Uncertainty lS.C.l-1 lS.C.l-1 15.C.2-l 15.C.2-1 Core Flow Measurement During Single-Loop Operation 15.C.2-1 Core Flow Uncertainty Analysis TIP Reading Uncertainty MCPR OPERATING LIMIT Abnormal Operational Transients Feedwater Controller Failure - Maximum Demand Generator Load Rejection With Bypass Failure Summary and Conclusions Rod Withdrawal Error Operating MCPR Limit STABILITY ANALYSIS Phenomena Compliance to Stability Criteria LOSS-OF-COOLANT ACCIDENT ANALYSIS Break Spectrum Analysis Single-Loop MAPLHGR Determination Small Break Peak Cladding Temperature 15.C-i 15.C.2-2 15.C.2-4 15.C.3-1 15.C.3-1 15.C.3-2 15.C.3-3 15.C.3-5 15.C.3-6 15.C.3-7 15.C.4-1 15.C.4-l 15.C.4-2 lS.C.S-1 lS.C.S-2 lS.C.S-2 lS.C.S-3 Revision 14 July 26, 200 5

TABLE OF CONTENTS (Continued) 1S.C.6 CONTAINMENT ANALYSIS 1S.C.7 MISCE~LANEOUS IMPACT EVALUATION 1S.C.. 7.1 Anticipated Transient* Without Scram Impact Analysis 15.C.7.2 Fuel Mechanical Performance 1S.. C.7.3 Vessel Internal Vibration 15.C.8 REFERENCES lS.C-ti HCGS-UFSAR Page 15.C.6-l 1S.C.7-l 1S.C.7-l 1S.C.7-l 1S.C.7-Z IS.C.S-1 Revision 0 April 11, 1988

NUMBER 15.C.3-1 15.C.3-2 15.C.3-3 15.C.3-4 HCGS-UFSAR LIST OF TABLES TITLE Input Parameters and Initial Conditions Sequence of Events for Figure 15.C.3-1, Feedwater Controller Failure, Maximum Demand Sequence of Events for Figure 15.C.3-2, Generator Load Rejection with Bypass Failure Summary of Transient Peak Value and CPR Results lS.C-iii PAGE 15.C.3-9 1S.C.3-11 15.C.3-12 15.C.3-13 Revision 0 April 11, 1988

NUMBER 1S.C.2-1 1S.C.3-1 15.C.3.2 lS.C.S-1 HCGS-UFSAR LIST OF FIGURES TITLE Illustration of Single Recirculation Loop Operation Flows Feedwater Controller Failure - Maximum Demand, 75\\ Power/60\\ Core Flow Generator Load Rejection with Bypass Failure, 75\\ Power/60\\ Core Flow Deleted lS.C-iv 15.C.2-5 15.C.3-14,15, 16,17 15.C.3-18,19 20,21 Revision 11 November 24, 2000

15.C RECIRCULATION SYSTEMS SINGLE-LOOP OPERATION The information presented in Appendix 15C is historical in nature.

The single-loop operation (SLO) required operating limits are confirmed or determined on a reload basis in accordance with the requirements in Reference 15. C. 8-6.

In

addition, SLO has been determined to be acceptable for CPPU operating conditions as described in Reference 15.C.8-12.

15.C.1 INTRODUCTION AND

SUMMARY

Single-loop operation (SLO) at reduced power is highly desirable in the event recirculation pump or other component maintenance renders one loop inoperative.

To justify single-loop operation, accidents and abnormal operational transients associated with power operations, as presented in Sections 6.2 and 6.3 and the main text of Chapter 15.0, were reviewed for the single-loop case with only one pump in operation.

This appendix presents the results of the safety evaluation for the operation of the Hope Creek Generating Station (HCGS) with single recirculation loop inoperable.

This safety evaluation was performed for GE and ABB fuel in Hope Creek. The analysis shows that the transient consequences for SLO {ACPR) are bounded by the full power analysis results given in the FSAR.

The conclusion drawn from the transient analysis results presented in this report is applicable to reload cycle operation.

Increased uncertainties in the core total flow and Traversing In-Core Probe (TIP) readings result in an incremental increase in the Minimum Critical Power Ratio (MCPR) fuel-cladding integrity safety limit during single-loop operation.

No increase in rated MCPR operating limit and no change in the power or flow dependent MCPR limit is required because all abnormal operational transients analyzed for single-loop operation indicated that there is more than enough MCPR margin to compensate for this increase in MCPR safety limit.

The recirculation flow rate dependent rod block and scram setpoint equation given in Chapter 16 (Technical Specifications) are adjusted for one-pump operation.

15.C.l-1 HCGS-UFSAR Revision 17 June 23, 2009

I.

Thermal-hydraulic was evaluated for its adequacy with respect to General Design Criteria 12

( 10CFR50, Appendix A). It is shown that this criterion is satisfied during SLO. It is further shown that the increase in neutron noise observed during SLO is of system stability margin.

To prevent potential control oscillations from occurring in the recirculation flow control the operation mode of the recirculation flow control system must be restricted to operation in the manual control mode for single-loop operation.

The Maximum Average Planar Linear Heat Generation Rate loop operation is reduced to accommodate the impact

s.

for s of SLO on the LOCA The impact of loop operation on the FSAR specifications for containment response including the containment dynamic loads was evaluated.

It was confirmed that the containment response under SLO is within the present design values.

The impact of single-loop operation on the Anticipated Transient Without Scram (ATWS) analysis was evaluated. It is found that all ATWS acceptance criteria are met puring SLO.

The fuel thermal and mechanical duty for transient events occurring during SLO is found to be bounded by the fuel bases. The Power Range Monitor (APRM) fluctuation* should not exceed a flux amplitude of +/-15% of rated and the core plate-differential pressure fluctuation should not exceed 3.2 psi peak to peak to be consistent with the fuel rod and assembly design bases.

A recirculation pump drive flow limit is imposed for SLO. The highest drive flow that meets acceptable vessel internal vibration criteria is the drive flow limit for SLO. The pump speed at Hope Creek Generating Station should be limited to 90% of rated under 15.C.1-2 HCGS-UFSAR conditions.

Revision 18 May 10, 2011

l5.C.2.MCPR PUBL CLADDING INTEGRITY SAPBTY LIMIT Except for eore total flow and TIP reading, the Wlcartaintiea used in the statistical analysis to determine the MCPR fuel cladding integrity safety limit are not dependent on whether coolant flow is provided by one or two recirculation pumps.

A 6t core flow measurement uncert*inty has been established for aingle-loop operation (compared to

2. st for two-loop operation). At shown below, this value conservatively reflects the one standard deviation (one sigma) acaura~ of the core flow measurement system doCUmented in Reference lS.C.8-l.

The random noise conwonent of the 'riP reading uncertainty waa revised. for single recirculation loop operation to reflect the operating plant test results given in Subsection 1s.c.2.2. This revision resulted in a single-loop operation process computer effective TIP uncertainty of 6.8t of initial cores and 9.lt for reload cores. Comparable two-loop process ~uter uncertainty values are

6. 3t for initial cores and 8. 7t for reload

~orea. This represents a 4. 6t increase in process computer determination relative assembly power.

The net effect of these two revised uncertainties is an incremental increase in the required MCPR fuel cladding integrity safety limit.

15.C.~.l Core Plow Uncertainty lS.C.2.l.l Core Flow Measurement During Single-LoDe Qperation The jet pump core flow measurement system is calibrated to meaaure core flow when both sets of jet pumps are in forward flow1 total core flow is the sum of the indicated loop flows. For single-loop operation, however, some inactive jet pumps will be baokflowing (at active pump speeds above approximately 40t).

Therefore, the measured flow in the backflowing jet pumps must be subtracted from the measured flow in the active loop to obtain the total core flow. In addition, the jet pump coefficient is different for reverse flow than for forward flow, and the measurement of reverse flow must be modified to account for this difference.

1S.C.2-l HCGS.. tmSAR Revision 11 November 24, 2000 I

I I

I In formula:

operation, the total core flow is derived by the following Total Core Flow

=

ActiveLoop Indicated Flow InactiveLoop

-C Indicated Flow The coefficient C (=0.95) is defined as the ratio of "Inactive True Flow" to "Inactive Loop Indicated Flow".

"Loop Indicated Flow" is the flow measured by the jet pump "single-tap" loop flow summers and indicators, which are set to read forward flow correctly.

The 0. 95 factor was the result of a conservative to the single-tap flow coefficient for reverse flow.* If a more exact, less conservative, core flow is in-reactor calibration tests can be made. Such calibration tests would involve: calibrating core support plate

~p versus core flow during one-pump and two-pump operation along with 100% flow control line and calculating the correct value of C based on the core

~P and the loop flow indicator readings.

15.C.2.1.2 The uncertainty analysis procedure used to establish the core flow uncertainty for one-pump is the same as for two-pump operation with some exceptions. The core flow uncertainty analysis is described in Reference 15.C.8-l. The of one-pump core flow uncertainty is suromarized below.

• The analytical expected value of the "C" coefficient for HCGS is 0.84.

15.C.2-2 HCGS-UFSAR Revision 18 May 10, 2011

For single-loop operation, the total core flow can be expressed as follows (refer to Figure 15.C.2-1):

where:

total core flow,_.,_

.. \\.

WA active loop flow, and w1 inactive loop (true) flow.

By applying the "propagation of errors" method to the variance of the total can be

+(~r * (rr2 WI rand +

where:

unc~rtainty* of total core flow; to both crW*

of loop only; A rand above by:

ere) random uncertainty of inactive loop only; uncertainty-of "C" co~fficient; a6d

equation, the a

ratio of iriacti ve loop trow (WI) t*o active l-oop flow (WA)'~

From an uncertainty analysis, the conservative, crw A rand crw I rand and are 1. 6%,

values of and 2.

respectively.

Based on the above uncertainties and a bounding value of 0. 36 for "a", the variance of the total flow uncertainty is approximately:

• This flow split ratio varies from about 0.13 to 0. 36. The 0. 36 value is a bounding value. The expected value of the flow split ratio for HCGS is - 0.33.

HCGS-UFSAR 15.C.2-3 Revision 18 May 10, 2011

(1. 6 ) 2

+

(2.6 )2

+ (

0.36 )2 * ((3.5 )2 + (2.8 )2) 1 -

0. 36 (5. 0% )2 When the effect of 4.1% core bypass flow split uncertainty at 12%

(bounding case) flow fraction is added to the total core flow the active-coolant flow a-2 active coolant (5.0%)2

+

which is less than the 6% flow In summary, core flow way and its 15.C.2.2 To ascertain the TIP noise a test was level 59.3% of rated with a 46.3%

rated). A the test.

is:

( 4. 1% )2 (5.1% }2 assumed in the statistical one-pump is measured in a conservative evaluated.

for recirculation BWR.

The test was recirculation pump in symmetric control rod at a power (core flow existed during Five con~ecutive traverses were made with each of five TIP machines, a

total of 25 traverses.

of this data resulted in a nodal TIP noise of 2.85%. Use of this TIP noise value as a of the process computer total results in a

value for for reload cores.

HCGS-UFSAR process total effect TIP of 6.8% for initial cores and 9.1%

15.C.2-4 18 May 10, 2011

PSE&G HCGS-rFSAR We

• Ttl11 COPI r1o*

w,

• Act1vt Leo* **

W1

• laac,tve Loop r1..

ILLUSTRATION OF SINGLE RECIRCULATION LOOP OPERATION FLOWS 1S.C.2*5 FIGURE

, s. c. 2-,

Revision 0 April ll, 1988

15.C.3 MCPR OPERATING LIMIT 1S.C.3.1 Abnormal Operational Transients Operating with one recirculation loop results in a maximum power output which is about 30' below that which is attainable for two-pump operation.

Therefore, the consequences of abnormal operational transients from one-loop operation will be considerably less severe than those analyzed for two-loop operation.

For pressurization, flow increase, flow decrease, and cold water injection transients, the results presented in Chapter 15 bound both the thermal and overpressure consequences of one-loop operation.

The consequences of flow decrease transients are bounded by the full power analysis. A single pump trip from one-loop operation is less severe than a two-pump trip from full power because of the reduced initial power level.

The worst flow increase transient results from a recirculation flow controller failure, and the worst cold water injection transient results from the loss of J

feedwater heating. For the former event, the impact on CPR is derived assuming both recirculation loop controllers fail. This condition produces the maximum I

possible power increase and hence maximum &CPR for transients initiated from less than rated power and flow. During operation with only one recirculation loop, the flow and power increase associated with this failure with only one loop will be less than that associated with both loops; therefore, the impact on CPR of the worst flow increase event derived with the two-pump assumption is conservative for single-loop operation.

The latter event, loss of feedwater heating/ is generally the most severe cold water event with respect to increase in core power. This power increase is caused by positive reactivity insertion from increased core inlet subcooling and it is relatively insensitive to initial power level. A generic statistical loss of feedwater heater analysis using different 15.C.3-1 HCGS-UFSAR Revision 11 November 24, 2000

initial power levels and other core design parameters concluded one-pump operation with lower initial power level is conservatively bounded by the full power two-pump analysis.

The conclusions regarding the consequences of the inadvertent restart of the idle recirculation pump in Chapter 15.4.4 are still applicable for single-loop operation.

Assessments of the relative impact on the limiting pressurization transients for single-loop and two-loop conditions show that the consequences for single-loop conditions are bounded by the two-loop results.

The following sections provide examples of these assessments and confirm the generic nature of the conclusions.

15.C.3.1.1 Feedwater Controller Failure -Maximum Demand {Cycle 1l This event is postulated on the basis of a single failure of a master feedwater control device, specifically one which can directly cause an increase in coolant inventory by increasing the total feedwater flow.

The most severe applicable event is a feedwater controller failure during maximum flow demand.

The feedwater controller is assumed to fail to its upper limit at the beginning of the event.

A feedwater controller failure during maximum flow demand at 75% power and 60%

flow during single recirculation loop operation produces the sequence of events listed in Table 15.C.3-2.

Figure 1S.C.3-1 shows the changes in important variables during this transient.

References to percent power, percent of rated, etc., contained in the text, figures, and tables describing this event are relative to the Cycle 1 licensed power level of 3293 MWth*

The computer model described in Reference 15. C. 8-2 was used to simulate this event.

The analysis has been performed with the plant conditions tabulated in Table 15.C.3-l. with the initial vessel water level at Level 4 (instead of normal water level) for conservatism. By lowering the initial water level, more cold feedwater will be injected before Level 8 is reached resulting in higher heat fluxes.

15.C.3-2 HCGS-UFSAR Revision 12 May 3, 2002

The cbndi t*ion is* at 7 5% rated thermal-power and 60% rated core flow, which represents

-recirculation loop operation *at 100% pump speed on the 105% rod line: End of. **cycle (all rod out*) scram characteristics are assumed. The safety..:.relief *valve* act-ion is -eonse1::vati vely assumed to- *occur with higher th 1an nominal setpoints. The transient is *.simulated by ptogrartiining an upper limit failure in the feedwater system such that 159% of rated feedwater flow at the reacto~ d'ome pressure 973 and of *rated

'flow would occur fit of psig.

The high feedwater controller transient is shown in 15.C.3-l. The water level turbine and feedwater approximately 6.1 seconds~ Scram occurs pump trip are

~*..

initiated at valve closure, and fuel thermal transient. The turbine bypass and limits the neutron flux system opens to lim~t peak_

in the steam supp~y

. Events caused by low water level functions are

.will follow initiation of HPCS and RCIC core cooling incl~ded in the simulation. Should these events occur, the and to be less severe than those have by The limita~G~R,of.l.

so

~- -

~.....

.~

. ' ~

no

-pea{'.

,y~_ssel of 1375 15.C.3.1.2 Fast closure of the turbine control* valves (TCV) is initiated wherever di"sturhances occur- ;which*

result*~, ih fi*cant:

• loss of electrical load on the The turbine control valves are

. t'o close as as to prevent (T-G) rotor. Closure of the turbine 15.C.3-3 HCGS-UFSP~R the turbine-will increase Revision 18 May 10, 2011 I

A loss of generator electrical load with bypass failure at 75% power and 60%

flow during recirculation loop operation produces the sequence of events listed in Table 15.C.3-3.

Figure 15.C.3-2 shows the changes in important variables during this transient.

References to percent power, of rated, etc., contained in the text, figures; and tables describing this event are relative to the Cycle 1 licensed power level of 3293 MWth*

Generator load ection causes turbine control valve (TCV) fast closure which initiates a scram trip signal for power levels greater than 40% NB rated. In addition, recirculation pump trip is initiated. Both of these trip signals satisfy single failure criterion and credit is taken for these protection features.

The pressure relief system which operates the relief valves independently when system pressure exceeds relief valve instrumentation setpoints is assumed to analyzed.

function normally during the time All plant control systems maintain normal operation unless to the The computer model described in Reference 15. C. 8-2 was used to simulate this event.

The has been performed with the plant conditions tabulated in Table 15.C.3-1, except that the turbine bypass function is assumed to fail.

The safety condition is at 75% rated thermal power and 60% rated core flow, which recirculation loop operation at 100% pump speed on the 105% rod line.

The turbine electro-hydraulic control system (EHC) power/load unbalance device detects load rejection before a measurable speed change takes The closure characteristics of the turbine control valves are assumed such that the valves operate in the full arc (FA) mode and have a full stroke closure time, from open to fully closed, of 0.15 second.

15.C.3-4 HCGS-UFSAR Revision 18 May 10, 2011

Auxiliary Power would normally be independent of any turbine-generator over-speed effects and continuously be supplied at rated frequency as automatic fast transfer to auxiliary power supplies occurs.

The simulated generator load rejection with bypass failure is shown in Figure 15.C.3-2.

Events caused by low water level trips, including initiation of HPCI and RCIC core cooling system functions are not included in this simulation.

If these events occur, they will follow sometime after the primary concerns of fuel margin and overpressure effects have passed, and will result in effects less severe than those already experienced by the reactor system, and will provide long-term reactor inventory control.

Table 15.C.3-4 summarizes the transient analysis results. The peak neutron flux reaches about 120% of rated and average surface heat flux peaks at about 104%

of its initial value. The peak vessel pressure predicted is 1162 psig and is well below the ASME limit of 1375 psig. The calculated MCPR is 1.16 which is considerably above the cycle 1 safety limit MCPR of 1.07.

l5C.3.1.3 Evaluation for ABB Fuel The impact of pressurization transients for single-loop operation (SLO) conditions relative to two-loop conditions has also been evaluated for the limiting pressurization events in a mixed SXB-4 and SVEA-96+ core.

These calculations were performed with the ABB licensing analysis methodology in Reference lS.C.B-10.

The calculations show that MCPR operating limits established by the limiting two loop transients are conservatively applicable to transients initiated from SLO conditions.

This conclusion accommodates the fact that the SVEA-96+ and BxB-4 SLMCPR for SLO is increased by an increment appropriate to accommodate the increased SLO uncertainties discussed in Section 15.C.2.

These results provide further confirmation that MCPR operating limits establishes by the limiting pressurization events based on the two loop evaluations will conservatively protect the fuel during postulated limiting pressurization transients initiated from SLO conditions.

15.C.3-5 HCGS-UFSAR Revision 11 November 24, 2000

Appendix 150 provides more information on the SLO analysis that is performed during the reload.

15.C.3.1.4 Summary and Conclusions The discussion in section 15C. 3. 1.1 through

15. C. 3.1. 2 ill'ustrates the conclusion that the operating limit MCPRs is established by pressurization transients for two-pump operation are also applicable to single-loop operation conditions.

For pressurization, Table 15. C. 3-4 indicates that the peak pressures are well below the ASME code value of 1375 psig. Hence, it is concluded that the pressure barrier integrity is maintained under single-loop operation.

15.C~3.2 The rod withdrawal error at rated power is given *in ~the FSAR. These analyses are performed to demonstrate, even if the operator ignores all instrument indications and the alarm which could occur -during the*aourse of the transient, the rod block system will stop rod withdrawal at a minimum critical power ratio

{MCPR) which is* higher than the fuel cladding inte'grity safety limit.

For ARTS/MELLLA analyses, the RWE is conservatively performed without a rod block and ensures the MCPR is higher* *t:han the fuel cladding integrity safety limit.

Modification of the rod block equation {below) and lower power assures the MCPR safety limit is not violated.

One-pump operation results in backflow through 10 of the 20 jet pumps while the flow is being supp.lied

• into the lower plenum from *the 10

• active jet pumps.

Because of the backflow through the inactive jet pumps, the..-pre.Setlt rod.block equation was conservatively modified for use during one-pump operation because the direct active-loop *flow measurement may not indicat-e actual flow above about 40% core flow without correction.

A procedure has been established for correcting the rod block equat-ion to account for the discrepancy between actual flow and indicated flow in the active loop. This preserves the original relationship between rod block and actual effective drive flow when operating with a single-loop.

15. C.. 3-6 HCGS-UFSAR Revision 15 October 27, 2006 I

The two-pump rod block equation is:

RB = mW + RB100 - m(lOO}

The one-pump equation becomes:

RB mW + RB100 - m{lOO} -

mAW where difference between two-loop and single-loop effective drive flow at the same core flow.

This value is expected to be 8\\ of rated (to be determined by PSE&G).

RB power at rod block in \\;

m flow reference slope for the rod block monitor (RBM) w drive flow in \\ of rated.

RB100

=

top level rod block at 100\\ flow.

If the rod block setpoint (RB100 ) is changed, the equation must be recal-culated using the new value.

The APRM trip settings are flow biased in the same manner as the rod block monitor trip setting. Therefore, the APRM rod block and scram trip settings are subject to the same procedural changes as the rod block monitor trip settings discussed above.

15.C.3-7 HCGS-UFSAR Revision 11 November 24, 2000

1S.C.3.3 Operating MCPR Limit For single-loop operation, the operating, MCPR limit remains unchanged from the normal two-loop operation limit. Although the increased uncertainties in core flow and TIP readings resulted in an incremental increase in MCPR fuel cladding integrity safety limit during single-loop operation (Section 15.C.2),

the results in Section 15.C.3 indicate that there is more than enough MCPR margin during single-loop operation to compensate for this increase in safety limit.

For single-loop operation at lower flows, the steady-state operating MCPR limit is established by reduced flow operating MCPRs.

This ensures the 99.9%

statistical limit requirement is always satisfied for any postulated abnormal operational occurrence.

Since the maximum core flow runout during single loop operation is only about 60\\ of rated, the current reduced flow MCPRs which are generated based on the flow runout up to rated core flow are also adequate to protect the flow runout events during single-loop operation.

15.C.3-8 HCGS-UFSAR Revision 11 November 24, 2000 I

I I

I I

TABLE 15.C.3-1 INPUT PARMETERS AND INITIAL CONDITIONS

1.

Thermal Power Level, MWt

2.

Steam Flow, lb per hr

3.

Core Flow, lb per hr

4.

Feedwater Flow Rate, lb per sec

5.

Feedwater Temperature, °F

6.

Vessel Dome Pressure, psig

7.

Vessel Core Pressure, psig

8.

Turbine Bypass Capacity, \\ NBR

9.

Core Coolant Inlet Enthalpy, Btu per lb

10.

Turbine Inlet Pressure, psig

11.

Fuel Lattice

12.

Core Average Gap Conductance, 2

Btu/sec-ft

- °F

13.

Core Bypass Flow, \\

14.

Required Initial MCPR

15.

MCPR Safety Limit

16.

Doppler Coefficient, ¢/°F

17.

Void Coefficient, ¢/\\Rated Voids

18.

Core Average Rated Fraction, \\

19.

Scram Reactivity, $&K

20.

Control Rod Drive Speed Position versus Time 2470 10.17 X 106 60.00 X 106 2824 390 973 978 25 512.1 944 C(P8x8R) 0.1744 11.27 1.28**

1.07 45.1 Figure 15.0-1 This value is calculated within the computer code (Reference 15.C.8-2) for end of Cycle 1 conditions based on input from the CRUNCH file.

Kf times the Rated Operating Limit MCPR 15.C.3-9 HCGS-UFSAR Revision 8 September 25, 1996

TABLE 15.C.3-1 (Cont.)

21.

Fuel

22.

Jet Pump Ratio, M I

23.

Valve Capacity, % NBR

@ 1121 psig Manufacturer End of Cycle 1

'3. 56 85.8 14

24.

Relief Function Delay, seconds 0.4

25.

Relief Function Time Constant, seconds 0.15

26.

for Valves

27.

Number of Valve Simulated Function

28.

29.

30.

• Flux

31.

Vessel Level Skirt Bottom Level 8 Valve Reclosure

, Feet Above

( L8*), feet Level 4-(L4), feet Level 3-(L~), feet Level 2-(L2), feet (121 X 1.043)

32.

APRM Simulated Thermal Power Trip Setpoint,

% NBR (117 x 1.043)

33.

34.

Inertia

35.

36.

Total Steamline Volume, ft 3 Pressure Pump HCGS-UFSAR

• of ATWS Recirculation psig 15.C.3-10 112lr 1131, 1141 3

71 10971 1107 126.2....

1071. ;

, *6.*cQ42 *:

• 3

1. 7>92

-3.708.

122.0 0.175, 4.5 6619 1101 Revision 18 May 10, 2011

Time-sec 0

6.1 6.1 6.. 1 6.Z 6.3 HCGS-UFSAR TABLE lS.C.l-2 SEQUENCE OF EVENTS FOR FIGURE 15.C.3*1 FEEDWATER CONTROLLER FAILURE, MAXIMUM DEMAND Initiate simulated failure to the upper 11m1t on feedwater flow.

LS vessel level setpoint trips mafn turbine and feedwater pumps. Turbine bypass operation initiated.

Reactor scram trip actuated from main turbine stop valve pos i tfon switches.*

Recirculation pump trip (RPT) actuated by stop valve position switches.

Main turbine stop valves closed and turbine bypass valves start to open.

Recirculation pump motor circuit breaker opens causing decrease in core flow.

15.C.. 3*ll Revision 0 April 11, 1988

Time-sec

(-)0.015 (approx.)

0 0

0 TABLE lS.C.l-3 SEQUENCE OF EVENTS FOR FIGURE 15.C.3-2 GENERATOR LOAO REJECTION WITH BYPASS FAILURE Turbine-generator detects loss of electrical load.

Turbine-generator load rejection sensing devices trip to initfatf turbine control valve fast closure.

Turbine bypass valves fail to operate.

Fast control valve closure (FCV) initiates scram trip and recirculation pump trip (RPT}.

0.07 Turbine control valves closed.

0.175 Recirculation pump motor circuit breaker opens causing decrease in core flow.

2.2 2.6 2.8 4.4 9.5 HCGS-UFSAR Group 1 relief valves actuated.

Group 2 relief valves actuated.

Group 3 relief valves actuated.

G~oup 3 relief valves start to close.

All relief valves are closed.

15.C.3*12 Revision 0 April 11, 1988

TABLE 15.C.3*4

SUMMARY

OF TRANSIENT PEAl VALUE AND CPR RESULTS

~

Initial Power/Flow CS Rated) 75/60 Peak Neutron Flux (S Rated) 91.2 Peak Heat Flux (S Initial) 103.3 Peak Dome Pressure {psfg) 1107 Peak Vessel Bottoa Pressure (psi g) 1121 1.28 Required Initial MCPR Transient MCPR **

1.17 Safety Limit MCPR (for SLO) 1.07 Margin to Safety Limit

o. 10

• Kf times the Rated Operating Limit MCPR.

• '* Includes Optton A adder.

15.C.3-13 HCGS-UFSAR

~

75/60 119.7 103.9 1148 1162 l.ZB 1.16 1.07 0.09 Revision 0 April 11, 1988

c n

0 Cll I c::

1-rj Cll g;

o*

~

1-1*

Cll 1-1*

~ 0

~::I 0

~

I U1.

n.

w I -

~,

150.

100.

50.

1 LEVELIIN H-REF-SEP-SKIRT 2 W A SENS 0 LEVELliNCHESI

• 3 N A SENS 0 LEVEL (INCHES I 4 CORE INL T FLOW lPCTJ 5 DRIVE FL W 2 lPCTJ 0. [ 1 '

I I I I I

I I I I

I I

0

• 5.

1 0.

1 5 *

20.

TIME lSECJ PSE&G FEE~~ATER CONTROLLER FAILURE - MAXIMUM DEMAND FIGURE lS.C.J-1

X:

n G1

(/)

I c tTJ

(/)

t>

Q

>::t:J "d (I) ti <

~~

I-' {II.....

1-'0 I-' :::.'I n.

w ' -"'

0

\\0 CD CD Cl I.&J 1-a:

0:

lL 0

I-z w

u a:

w Q_

PSEIG I NEUTRON LUX 2 PERK FUE CENTER TEHP 150 1 1 \\

1 t3 AVE SURF CE HEAT FLUX q FEEOWATE FLOH 5 VESSEL S EAH FLOW

• 100.

so.

I

0. I ** '

• I I I

I I I I

10.

15.

0.

5.

TIME ISECJ

. FEEDWAT£R CONTROLLER FAILURE - MAXHIJM OEIMND 751 POI<<R/601 CORE FLOW

(

I 2 1

-r

20.

FIGURE lS.C.J-1 CONT'D.

(

r:

()

GJ C/:1 I

c::::

l"lj C/:1

~

1-'

_.I U'l.

p w * -

0\\ I 0

200.

\\00.

0.

1 VESSEL P ES RISE lPSIJ 2 STM LINE PRES RISE lPSIJ 3 TURBINE RES RISE. lPSIJ 4 BYPASS 5 EAM FLOWCPCTJ 5 AELIEF V LVE FLOWlPCTJ 6 TURB STE M FLQW IPCTJ

• 1

15.

20.

-1 00 * [ I I

I I I I I 5

• 10 i I ME ( SEc I

0.

PSE&G fEEOWATER CONTROLLER FAILURE - MAXIHUH DEMAND 75~ PO\\IER/601 CORE FLOW FIGURE 15.C.J-1

X: n Ci)

Cfl I

c:::::

~

Cfl

~

?;:!

(I) <

!-.1--

(j')

!-.1--

t-"0 1-"::l 0

I-"

I Uti.

n.

~I I

I

(

)-

t-

~

t-u a:

w a:

PSE&G I VOID REA TIVITY 2 DOPPLER

[ACTIVITY 1 1 1

1 13 SCRAM ~~ CTIVITl 4 TOTn

~

tTY

o.

-1.

-2. 0.

5.

tO.

15.

TIME ISECJ FEEOUATER CONTROllER FAILURE - MAXIMUM OEMAHD 751 POWER/601 CORE FlOW

(

20.

-1 FIGURE 15.C.3-1 COHT*o.

(

r::

CJ 0

(!)

I c::

forj

(!)

0

0

~

1-'*

Cfl 1-'*

1--'0 1--'::l 00 co 0

I U1.

n.

w t -

(X) I 150.

100.

50.

1 LEVELliN H-REF-SEP-SKIRT 2 W R SENS 0 LEVEL(INCHESJ 3 N A SENS 0 LEVELtiNCHESJ 4 CORE INL T FLOW lPCTJ 5 DRIVE FL H 2 lPCTJ I

• 7 t

I

0. [!! II I I I I I I

4.

6.

8.

0.

2*

TIME ISECl PSE&G.

GENERATOR LOAD REJECTION I~ITH BYPASS FAILURE, 75l POWER/60S CORE FLO~

fiGURE 15.C.J.2

r:

(")

0 C/)

I c

"'rj C/)

r,;

d (I) <

1-'*

en 1-'--

1-' 0

_:-::s 0

1-'

~I w

I \\DI I

I

(_

0 w a:

a:

u_

0 t-z UJ u a:

w Q_ -

PSE&G 150.

1 NEUTRON 2 PEAK FUE CENTER TEMP 3 AVE SURF CE HEAT FLUX 4 FEEOWATE FLOH 5 VESSEL S EAM FLOW 100.~~~--+-------4-------~-------r-------

50. I I h \\J I...............,.. II P

._,.~tiH

>st<::::::

I

~ t II 1

11 V

I

.n Q,t.\\AIIa*aaf

4.

0.

2.

TIME GENERATOR LOAD REJECTION WITH BYPASS FAILURE, 75S POWER/60S CORE fLOW FIGltHE 15.C.l.2 COHT'O.

(

r::

n G')

(I)

I c:::

~

Ul

~

iAl

~

1-1*

(ll 1-1*

t-'0 t-':;:1 0

t-'

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~

n.

w

• N C)

I 200.

100.

0.

1 VESSEL PES RISE lPSIJ 2 STM LINE PRES RISE lPSIJ 3 TURBINE RES RISE lPSIJ 4 RELIEF V LVE FLOwr-lPCTJ

-lOO.Ww~~~W-------~------~~------~----~

2.

Lj.

6.

8.

0.

PSE&G TIME l SEC l GE~EP~TOR LOAD REJECTION WITH BYPASS FAILURE. 75% POWER/60S CORE fLOl.

fiGURE 15.C.J.l CONT' 0.

X::

CJ G1 C/)

I c::

t-:j Ul

~

d (I) <

I-'*

00 1-'*

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t-t-

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15.C.4 STABILITY ANALYSIS 15.C.4.1 The primary contributing factors to the stability performance with one recirculation loop not in service are the power/flow ratio and the recirculation loop characteristics.

As forced circulation with only one recirculation loop in operation, the reactor core stability is influenced by the inactive recirculation loop. As core flow increases in SLO, the inactive jet pump forward flow decreases because the driving head across the inactive jet pumps decreases with core flow. The reduced flow in the inactive loop reduces the resistance that the recirculation on reactor core flow perturbations thereby a destabili effect. At the same time the increased core flow results in a lower power/flow ratio which is a stabilizing effect. These two countering effects result in decreased margin ratio) as core flow is increased (from minimum) in SLO and then an increase in margin (lower ratio) as core flow is increased further and reverse flow in the inactive loop is established.

As core flow is increased further during SLO and substantial reverse flow is established in the inactive loop an increase in jet pump flow, core flow and neutron noise is observed. A cross flow is established in the annular downcomer near the jet *pump suction entrance caused by the reverse flow of the inactive recirculation loop.

This cross flow interacts with the jet pump suction flow of the active recirculation loop and in-creases the jet pump flow noise. This effect increases the total core flow noise which tends to drive the neutron flux noise.

I To determine if the increased noise is being caused by reduced margin as SLO core flow was increased, an evaluation was which phenomenologically accounts for single-loop operation effects on stability, as summarized in Reference 15.C.8-3. The model 15.C.4-1 HCGS-UFSAR were initially Revision 18 May 10, 2011

compared with test data and showed very good agreement for both two-loop and test conditions.

An evaluation was performed to determine the effect of reverse flow on stability during SLO. With increasing reverse flow, SLO exhibited slightly lower decay ratios than two-loop operation. However, at core flow conditions with no reverse flow, SLO was less stable. This is consistent with observed behavior in tests at operating BWRs I (Reference 15.C.8-4).

In addition to the above analyses, the cross flow established during reverse flow conditions was simulated analytically and shown to cause an increase in the individual and total jet pump flow noise, which is consistent with test data (Reference 15. C. 8-3).

The results of these analyses and tests indicate that the characteristics are not significantly different from two-loop operation. At low core flow, SLO may be slightly less stable than operation but as core flow is increased and reverse flow is established the is similar. At higher core flow with substantial reverse flow in the inactive recirculation loop, the effect of cross flow on the flow noise results in an increase in noise (jet pump, core flow and neutron flux noise).

15.C.4.2 Compliance with the criteria set forth in 10CFR50 Appendix A,

General Design Criterion (GDC-12),

is achieved by either stability-related neutron flux oscillations or and suppres the oscillations prior to exceeding Acceptable Fuel Limits.

The BWR Owners' Group (BWROG) has developed solutions, which incorporate either prevention or detection and suppression features, or a combination of both features, to ensure compliance with GDC-12.

Methodologies have been to support the of these long-term solutions.

The BWROG has also developed guidelines (Reactor Stabil Interim Corrective Actions) for the licensee to use prior to the licensee's successful implementation of a Term Stability Solution. These guidelines expand the interim corrective actions identified in NRC Bulletin 88-07, Supplement 1.

15.C.4-2 HCGS-UFSAR Revision 18 May 10, 2011 I

The expanded guidelinea primarily accommodate the experience gained from plant stability events as well as conclusions based on recent analytical st\\ldies supporting' the Long Term Stability solution.

Based on the COlmlunications between the NRC and the BWROG, these guidelines fully aatiafy the Bulletin 88-07, supplement 1, requirements.

HCGS has implemented Reactor Stability Interim Corrective Actions baaed on the BWROG's recommendations to reduce the potential for unaccepta~le oscillations associated with the single-loop operation prior to the implementation of the Long Term Stability Solution.

lS.C.4.. 3 HCGS-UFSAR Revision ll November 24, 2000

I I

LOSS-OF-COOLANT ACCIDENT ANALYSES If two recirculation loops are operating and a pipe break occurs in one of the two recirculation loops, the pump in the unbroken loop is assumed to immediately trip and.begin to coast down. The decaying core flow clue to the pump coastdown results in very effective heat transfer (nucleate boiling) during the initial phase of the blowdown. Typically, nucleate boiling will be sustained during the first 5 to 9 seconds after the accident, for the design basis accident (DBA).

If only one recirculation loop is operating, and the break ooeurs in the operating loop, continued core flow ill provided only by natural circulation because the veaael is blowing down to the reactor containment through both sections of the broken loop. The core flow decreases more rapidly than in the two-loop operating case, and the departure fro. nucleate boiling for the high power node might occur 1 or 2 seconds after the postulated accident, resulting in more severe cladding heatup for the one-loop operating case.

In addition to changing the blowdown heat transfer characteristics, losing recirculation pwnp coastdown flow can alao affect the system inventory and reflooding phenomena.

Of particular interest are the changes in the high-power node uncovery and reflooding times, the system pressure and the time of rated core spray for different break sizes.

One-loop operation results in small changes in the high-power node uncovery times and times of rated spray.

The effect of the reflooding times for various break sizes is also generally small.

Analyses single recirculation loop operation using the modela and assumptions documented in References lS.C.S-9 or lS.C.S-10~ as appropriate, are performed for HCGS.

Using the appropriate methods, limiting pipe breaks are identified.

The single loop LOCA evaluation results in maximum planar linear heat generation rate (MAPLHGR) curves specific to single loop operation which aeeume that LOCA acceptance criteria in 10CFR50.46 are satisfied.

l.S.C.S-1 HCGS-UFSAR Revision 11 November 24, 2000

FIGURE lS~C~S-1 HAS BEEN DELETED lS.C.S-2 HCGS*UFSAR Revision ll November 24, 2000

lS.C. 6 CONTAINMENT ANALYSIS The range of power/flow conditions which are included in the SLO operating domain for Hope creek were investigated to determine if there would be any impact on the FSAR specifications for containment response, including the containment dynamic load.s. The SLO operating conditions were confirmed to be within the range of operating conditions which have previously been considered in defining the containment pres~e and temperature response and containment dynamic loads for two-loop operation. Therefore, the containment response for Hope Creek with single.. loop operation has been confirmed to be within the present design values.

1.s.c.6.. J.

HCGS.. UFSAR Revision 0 April 11., 1988

lS.C.7 MISCELLANEOUS IMPACT EVALUATION lS.C.7.l Anticipated Transient Without Scram (ATWS) Impact Evaluation The principal difference between single-loop operation (SLO} and normal two-loop operation (TLO) affecting Anticipated Transient Without Scram (ATWS) performance is that of initial reactor conditions. Since the SLO initial power flow condition is less than the rated condition used for TLO ATWS analysis, the transient response is less severe and therefore bounded by the TLO analyses.

It is concluded that if an ATWS event were initiated at HCGS from the SLO conditions, the results would be less severe than if it were initiated from rated conditions.

1S.C.7.2 Fuel Mechanical Performance Component pressure differential and fuel rod overpower values for anticipated operational occurrences initiated from SLO conditions have been found to be bounded by those applied in the fuel rod and assembly design bases.

It is observed that due to the substantial reverse flow established during SLO both the Average Power Range Monitor

{APRM) noise and core plate differential pressure noise are slightly increased.

An analysis has been carried out to determine that the APRM fluctuation should not exceed a flux amplitude of +/-15\\ of rated and the core plate differential pressure fluc-tuation should not exceed 3.2 psi peak to peak to be consistent with the fuel rod and assembly design bases.

1S.C.7-1 HCGS-UFSAR Revision 11 November 24, 2000

15.C.7.3 Vessel Internal Vibration Vibration tests for SLO were performed during the startup of two BWR 4-251 plants. An extensive vibration test was conducted at a prototype BWR 4-251

plant, Browns Ferry 1, the results of which are used as a standard for comparison.

A confirmatory vibration test was performed at the Peach Bottom 2

& 3 plants.

The Browns Ferry 1

test data demonstrates that all instrumented vessel internals components vibrations are within the allowable criteria. The highest measured vibration in terms of percent criteria for single-loop operation was 70%. This was measured at a jet pump riser brace during cold flow conditions at 100% of rated pump speed.

The Peach Bottom vibration test data shows that vessel internals vibration levels are within the allowable criteria for all test conditions. The highest measured vibration in terms of percent criteria for single-loop operation was 96%. This was measured at a jet pump elbow location during 68% power condition at 92% of rated pump speed. This vibration amplitude is the highest, in terms of percent criteria, experienced in vessel internals for the BWR 4-251 plants studied.

The conclusion is that under all operating conditions, the vibration level is acceptable.

However, due to the high vibration levels

recorded, it is recommended that Hope Creek not perform single-loop operation with pump speed exceeding 90% of rated pump speed. The same recommendation has been accepted by the Browns Ferry and Peach Bottom plants.

This analysis is conservative because the criteria are developed by assuming that the plant operates on a steady state single loop operations througho~t the plant life.

15-C.?-2 HCGS-UFSAR Revision 14 July 26, 2005

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i

15.C.8 15.C.8-1 IS.C.B-2 15.C.8-3 15.C.8-4 15-C-8-5 15.C.8-6 15.C.8-7 15.C.8-8 HCGS-UFSAR REFERENCES 11General Electric BWR Thermal Analysis Basis

{GETAB);

Data, Correlation, and Design Application", NED0-10958-A, January 1977.

11Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors 11, NED0-24154-A, August 1986.

Letter 1 H.C. Pfefferlen (GE) to c.o. Thomas {NRC), "Submittal of Response to Stability Action Item from NRC Concerning Single-Loop Operation," September 1983.

s. F. Chen and R. o. Niemi, "Vermont Yankee Cycle 8 Stability and Recirculation Pump Trip Test Report", General Electric Company, August 1982 (NEDE-25445, Proprietary Information).

G.A. Watford, "Compliance of the General Electric Boiling Water Reactor Fuel Designs to Stability Licensing Criteria", General Electric

Company, October 1984 Information).

(NEDE-22277-P-1, Proprietary 11General Electric Standard Application for Reactor Fuel 11,

NEDE-24011-P-A, and "General Electric Standard Application for Reactor Fuel (Supplement for United States}," NEDE-24011-P-A-US, latest revision.

11BWR Core Thermal Hydraulic Stability". General Electric Company/

February 10, 1984 (Service Information Letter-380, Revision 1).

Letter, C.O. Thomas (NRC) to H.C. Pfefferlen (GE) 1 "Acceptance for Referencing of Licensing Topical Report NEDE-24011, Rev. 6, Amendment B, Thermal Hydraulic Stability Amendment to GESTAR II, 11 April 24, 1985.

15.C.8-1 Revision 14 July 26, 2005

15.C.8-9 I

15.C.8-10 15.C.8-ll I

15.C.8-12 HCGS-UFSAR 15.C.8 REFERENCES(Cont'd)

"SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis for Hope Creek Generating Station at Power Uprate, '

1 NEDC-33172P, March 2005.

ABB Combustion *Engineering Nuclear

Power, "Reference Safety Report for Boiling Water Reactor Reload Fuel,"

ABB Report CENPD-300-P-A (proprietary), July 1996.

Latest BWROG recommendations for "interim Stability Solution 11 NEDC-33076P, Rev.

2, "Safety Analysis Report for Hope Creek Constant Pressure Power Uprate 11

, August 2006.

15.C.8-2 Revision 17 June 23, 2009

APPENDIX 15D CYCLE 23 RELOAD ANALYSIS RESULTS TABLE OF CONTENTS 15D.1 INTRODUCTION AND PURPOSE 15D-1 15D.2 RELOAD METHODOLOGY 15D-1 15D.3 RELOAD ANALYSIS RESULTS 15D-2 15D.3.1 Loss of Feedwater Heating 15D-2 15D.3.1.1 Initial Conditions 15D-2 15D.3.1.2 Sequence of Events 15D-3 15D.3.1.3 Results 15D-3 15D.3.2 Feedwater Controller Failure - Maximum Demand 15D-3 15D.3.2.1 Initial Conditions 15D-3 15D.3.2.2 Sequence of Events 15D-3 15D.3.2.3 Results 15D-3 15D.3.3 Generator Load Rejection, No Bypass 15D-4 15D.3.3.1 Initial Conditions 15D-4 15D.3.3.2 Sequence of Events 15D-4 15D.3.3.3 Results 15D-4 15D.3.4 Turbine Trip, No Bypass 15D-4 15D.3.4.1 Initial Conditions 15D-4 15D.3.4.2 Sequence of Events 15D-5 15D.3.4.3 Results 15D-5 15D.3.5 Rod Withdrawal Error 15D-5 15D.3.5.1 Initial Conditions 15D-5 15D.3.5.2 Sequence of Events 15D-5 15D.3.5.3 Results 15D-5 15D-i HCGS-UFSAR Revision 24 May 21, 2020

TABLE OF CONTENTS (cont) 15D.3.6 Inadvertent High Pressure Injection Startup 15D-6 15D.3.6.1 Initial Conditions 15D-6 15D.3.6.2 Sequence of Events 15D-6 15D.3.6.3 Results 15D-6 15D.3.7 Loss of Coolant Accident 15D-6 15D.3.8 Misloaded Fuel Bundle Accident 15D-7 15D.3.8.1 Mislocated Bundle 15D-7 15D.3.8.2 Misoriented Bundle 15D-7 15D.3.9 Control Rod Drop Accident 15D-7 15D.3.10 Fuel Handling Accident 15D-8 15D.3.11 Shutdown Without Control Rods 15D-8 15D.3.12 Core Thermal-Hydraulic Stability 15D-8 15D.3.13 ASME Over-Pressurization 15D-8 15D.3.14 Deleted 15D-9 15D.4 Single Loop Operation 15D-9 15D.5 References 15D-9 15D-ii HCGS-UFSAR Revision 24 May 21, 2020

LIST OF TABLES Table Title 15D-1 INPUT PARAMETERS AND INITIAL CONDITIONS FOR RELOAD LICENSING ANALYSIS 15D-2 RESULT

SUMMARY

FOR LOSS OF FEEDWATER HEATING MANUAL CONTROL 15D-3 SEQUENCE OF EVENTS FOR FEEDWATER CONTROLLER FAILURE MAXIMUM DEMAND 15D-4 SEQUENCE OF EVENTS FOR GENERATOR LOAD REJECTION WITHOUT BYPASS OPERATION 15D-5 SEQUENCE OF EVENTS FOR INADVERTENT HIGH PRESSURE COOLANT INJECTION STARTUP 15D-6 DELETED 15D-7 SEQUENCE OF EVENTS FOR MISLOADED FUEL BUNDLE ACCIDENT 15D-8 MISLOCATED FUEL ASSEMBLY RESULTS 15D-9 MISORIENTED FUEL ASSEMBLY RESULTS 15D-iii HCGS-UFSAR Revision 24 May 21, 2020

LIST OF FIGURES Figure Title 15D-1 Plant Response to FW Controller Failure (EOC ICF & FWTR (UB))

15D-2 Plant Response to Load Rejection w/o Bypass (EOC ICF (UB))

15D-3 Plant Response to Turbine Trip w/o Bypass (EOC ICF (UB))

15D-4 Plant Response to Inadvertent High Pressure Coolant Injection Startup (EOC ICF (UB))

15D-5 Deleted 15D-6 Deleted 15D-7 Deleted 15D-8 Deleted 15D-9 Deleted 15D-10 Deleted 15D-11 Deleted 15D-12 Deleted 15D-iv HCGS-UFSAR Revision 24 May 21, 2020

LIST OF FIGURES (cont)

Figure Title 15D-13 Deleted 15D-14 Deleted 15D-15 Deleted 15D-16 Deleted 15D-17 Deleted 15D-18 Deleted 15D-19 Deleted 15D-20 Deleted 15D-21 Deleted 15D-22 Deleted 15D-23 Deleted 15D-24 Deleted 15D-25 Deleted 15D-v HCGS-UFSAR Revision 18 May 10, 2011

Appendix 15D Cycle 23 Reload Analysis Results 15D.1 INTRODUCTION AND PURPOSE During each reload, fresh fuel assemblies are loaded into the core. A change in fuel design and core configuration has the potential to affect the results of the Section 15 events. Therefore an analysis of the potentially limiting events is performed on a cycle-to-cycle basis. This analysis is known as the reload licensing analysis. This appendix to Section 15 represents the results of cycle specific reload licensing analysis.

The purpose of this appendix is to summarize the cycle specific reload licensing analysis. This appendix is referenced throughout Section 15 for the results of the appropriate events. It is also referenced in Section 5.2.2.

15D.2 RELOAD METHODOLOGY The NRC-approved reload methodology is documented in GESTAR II (Reference 15D.5-1).

The reload methodology is used to perform an evaluation of the potentially limiting events. The potentially limiting events can be divided into three groups: Anticipated Operational Occurrences (AOOs), Design Basis Accidents (DBAs) and Special Events. The AOOs are:

Loss of Feedwater Heating (LOFH):

See Section 15.1.1 Feedwater Controller Failure Maximum Demand (FWCF):

See Section 15.1.2 Generator Load Rejection, No Bypass(GLRNB):

See Section 15.2.2 Turbine Trip, No Bypass (TTNB):

See Section 15.2.3 Rod Withdrawal Error at Power (RWE):

See Section 15.4.2 Inadvertent High Pressure Coolant Injection See Section 15.5.1 Startup (IHPCIS) 15D-1 HCGS-UFSAR Revision 24 May 21, 2020

The DBAs are:

Loss of Coolant Accident (LOCA):

See Section 15.6.5 Misloaded Fuel Bundle Accident (Mislocated or Misoriented):

See Section 15.4.7 Control Rod Drop Accident (CRDA):

See Section 15.4.9 Fuel Handling Accident:

See Section 15.7.4 The special events are:

Shutdown without Control Rods: (none identified)

Core Thermal-Hydraulic Stability: (none identified)

ASME Over-Pressurization:

See Section 5.2.2 Anticipated Transient Without Scram (ATWS):

See Section 15.8 In addition to the aforementioned events, an assessment is made to re-confirm that the results of the events evaluated for two recirculation loop operation bounds the single recirculation loop configuration, or specific single loop operation limits are established.

15D.3 RELOAD ANALYSIS RESULTS The results of the Cycle 23 reload analysis are presented within this section, or can be found in Reference 15D.5-2.

15D.3.1 Loss of Feedwater Heating The description of the Loss of Feedwater Heating (LOFH) is found in Section 15.1.1.

The results presented in this section assume that the plant is operating in manual flow control mode.

15D.3.1.1 Initial Conditions The analysis has been performed with the conditions tabulated in Table 15D-1.

15D-2 HCGS-UFSAR Revision 24 May 21, 2020

15D.3.1.2 Sequence of Events The LOFH event is analyzed with a three-dimensional core simulator (Reference 15D.5-1). This is a conservative steady state analysis for the determination of the appropriate power distribution limits during the event. Since it is not a dynamic simulation, no sequence of events is available.

The event can be initiated by closure of an extraction line to a feedwater heater or by bypassing one or more feedwater heaters. No subsequent operator action to mitigate plant response to the loss of feedwater heating is assumed.

15D.3.1.3 Results The initiation of the LOFH event is an assumed 110F reduction in feedwater temperature. The analysis results for the LOFH in the manual flow control mode are summarized in Table 15D-2 and in Reference 15D.5-2.

15D.3.2 Feedwater Controller Failure - Maximum Demand The description of the Feedwater Controller Failure - Maximum Demand (FWCF) is found in Section 15.1.2.

15D.3.2.1 Initial Conditions The analysis has been performed with the conditions tabulated in Table 15D-1.

The FWCF event has the potential to be the limiting event.

15D.3.2.2 Sequence of Events The sequences of events for the FWCF analysis are listed in Table 15D-3.

15D.3.2.3 Results Analysis results for the FWCF events are presented in Figure 15D-1. This figure presents the transient variation of various important system parameters (Reference 15D.5-2).

15D-3 HCGS-UFSAR Revision 17 June 23, 2009

15D.3.3 Generator Load Rejection, No Bypass The description of the Generator Load Rejection, No Bypass (GLRNB) is found in Section 15.2.2.

15D.3.3.1 Initial Conditions The analysis has been performed with the conditions tabulated in Table 15D-1.

The values tabulated in Table 15D-1 represent analysis assumptions, which were established as design input for this event as described in GESTAR (Reference 15D.5-1).

15D.3.3.2 Sequence of Events The sequence of events for the GLRNB analysis is listed in Table 15D-4.

15D.3.3.3 Results The analysis results for the GLRNB are presented in Figure 15D-2. This figure presents the transient variation of various important system parameters (Reference 15D.5-2).

15D.3.4 Turbine Trip, No Bypass The description of the Turbine Trip, No Bypass (TTNB) is found in Section 15.2.3.

The TTNB event is similar to the GLRNB event. Although the two events have different initiating faults, the TTNB event parameter responses follow the same trend as the GLRNB event response. The TTNB event is analyzed for each fuel cycle.

15D.3.4.1 Initial Conditions The analysis has been performed with the conditions tabulated in Table 15D-1.

The values tabulated in Table 15D-1 represent analysis assumptions, which were established as design input for this event as described in GESTAR (Reference 15D.5-1). These analysis assumptions are the same as for the GLRNB event.

15D-4 HCGS-UFSAR Revision 24 May 21, 2020

15D.3.4.2 Sequence of Events The sequence of events for the TTNB analysis is similar to the GLRNB in Table 15D-4.

15D.3.4.3 Results The analysis results for the TTNB are presented in Figure 15D-3. This figure presents the transient variation of various important system parameters (Reference 15D.5-2).

15D.3.5 Rod Withdrawal Error The description of the Rod Withdrawal Error (RWE) is found in section 15.4.2.

15D.3.5.1 Initial Conditions The analysis has been performed with the conditions tabulated in Table 15D-1.

The values tabulated in Table 15D-1 represent analysis assumptions, which were established as conservative design input for this event as described in GESTAR (Reference 15D.5-1).

15D.3.5.2 Sequence of Events The RWE event is analyzed with a three-dimensional core simulator (see Reference 15D.5-1). This is a conservative steady state analysis for the determination of the appropriate power distribution limits during the event.

Since, it is not a dynamic simulation, no sequence of events is available.

An operator is assumed to erroneously select and continuously withdraw a control rod at its maximum withdrawal rate at rated conditions until rod withdrawal is terminated by the Rod Block Monitor system.

15D.3.5.3 Results The ARTS based rod withdrawal error is evaluated for each fuel cycle and the results are provided in Reference 15D.5-2.

15D-5 HCGS-UFSAR Revision 24 May 21, 2020

15D.3.6 Inadvertent High Pressure Coolant Injection Startup The description of the Inadvertent High Pressure Coolant Injection Startup (IHPCIS) event is found in section 15.5.1. The results presented in this section conservatively assume that the feedwater control system will not prevent a Level 8 turbine trip and 100% of the HPCI flow enters the vessel through the feedwater sparger.

15D.3.6.1 Initial Conditions The analysis has been performed with the conditions tabulated in Table 15D-1.

The values tabulated in Table 15D-1 represent analysis assumptions, which were established as design input for this event as described in GESTAR (Reference 15D.5-1). These analysis assumptions are the same as for the GLRNB and TTNB events.

15D.3.6.2 Sequence of Events The sequence of events for the IHPCIS analysis is listed in Table 15D-5.

15D.3.6.3 Results The analysis results for the IHPCIS are presented in Figure 15D-4. This figure presents the transient variation of various important parameters (Reference 15D.5-2).

15D.3.7 Loss of Coolant Accident The description of the loss of coolant accident (LOCA) is found in Section 15.6.5.

The LOCA is a design bases accident. The GE14 fuel was analyzed for the LOCA and the results are summarized in Reference 15D.5-5. The GNF2 fuel was analyzed for the LOCA and the results are summarized in Reference 15D.5-14.

The consequences of a design basis LOCA are evaluated for each unique reload fuel design to support the establishment of core operating limits for that fuel design. This evaluation establishes appropriate Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits for the reload fuel (Reference 15D.5-2).

The operation of the core within these established MAPLHGR limits ensures that the ECCS LOCA requirements are met. The MAPLHGR operating limits for reload fuel are presented and controlled in the cycles Core Operating Limit Report (COLR).

15D-6 HCGS-UFSAR Revision 24 May 21, 2020

15D.3.8 Misloaded Fuel Bundle Accident The description of the Misloaded Fuel Bundle Accident is found in Section 15.4.7.

The reload licensing methodology analyzes two events in this category: the mislocated fuel bundle event and the misoriented fuel bundle event (Reference 15D.5-1). Although both events are classified as accidents, each is analyzed as an operating transient (AOO) in accordance with GESTAR (Reference 15D.5-1).

15D.3.8.1 Mislocated Fuel Bundle This design basis accident involves the mislocation of a fuel assembly into the wrong core location and the subsequent operation of the reactor with the mislocated assembly.

The sequence of events for the mislocated fuel bundle accident is presented in Table 15D-7.

The results of the mislocated fuel bundle accident are presented in terms of cycle Operating Limit MCPR in Table 15D-8. (Reference 15D.5-2) 15D.3.8.2 Misoriented Fuel Bundle This design basis event involves the misorientation (rotation) of a fuel assembly relative to the orientation assumed in the reference core design.

The sequence of events for the misloaded fuel bundle accident is presented in Table 15D-7.

The results of the misoriented fuel bundle accident are presented in terms of cycle Operating Limit MCPR in Table 15D-9. (Reference 15D.5-2) 15D.3.9 Control Rod Drop Accident The description of the Control Rod Drop Accident is described in Section 15.4.9.

15D-7 HCGS-UFSAR Revision 24 May 21, 2020

HCGS is a Banked Position Withdrawal Sequence (BPWS) plant, and therefore, in accordance with GESTAR II (Reference 15D.5-1), does not need to analyze the control rod drop accident (CRDA) each reload. The results of a generic analysis of the event are provided in Section 15.4.9.

15D.3.10 Fuel Handling Accident The current HCGS licensing analysis bounds the consequences of any fuel handling accident (see Section 15.7.4).

15D.3.11 Shutdown Without Control Rods The Standby Liquid Control System (SLCS) shutdown capability has been evaluated at a moderator temperature of 181°C as a function of exposure for a core Boron concentration equivalent to 660 ppm at 20°C. The minimum shutdown margin for the Hope Creek Cycle 23 core is 0.030 k (Reference 15D.5-2).

15D.3.12 Core Thermal-Hydraulic Stability Hope Creek has implemented the Detect and Suppress Solution - Confirmation Density (DSS-CD) as described in Reference 15D.5-6. The reload validation for the DSS-CD solution is performed for every reload cycle in accordance with Reference 15D.5-1.

Hope Creek uses the backup stability protection (BSP) methodology in the event that the OPRM system is declared inoperable according to Reference 15D.5-6.

15D.3.13 ASME Over-Pressurization The ASME over-pressurization analysis is performed to evaluate margin to the vessel pressure safety limit. The basis for this event is described in Section 5.2.2.

MSIV closure with flux scram was found to be the most limiting event in terms of vessel pressure. The results are summarized as follows:

15D-8 HCGS-UFSAR Revision 24 May 21, 2020

Maximum Vessel Pressure 1294 psig Maximum Steam Dome Pressure 1270 psig Maximum Steam Line Pressure 1264 psig The scram on MSIV position is not credited for this event. The maximum pressures during the event are below the ASME upset code limit of 1375 psig, which is 110% of the reactor vessel design pressure. Furthermore the maximum steam dome pressure predicted during the event is below the Technical Specification steam dome pressure safety limit of 1325 psig. (Reference 15D.5-

2) 15D.3.14 Section Deleted 15D.4 Single Loop Operation GNF has confirmed that the basis for single loop operation (SLO) presented in Appendix 15C remains valid for the current cycle. The confirmation involves an analysis at core power consistent with, or bounding the Technical Specification limit. Reference 15D.5-2 identifies that for Cycle 23, for single loop operation, the safety limit MCPR will be 1.13 and the LHGR and MAPLHGR multiplier will be 0.80. The DSS-CD solution supports SLO per Reference 15D.5-

6. The reload validation for the DSS-CD solution is performed for every reload cycle in accordance with Reference 15D.5-1.

15D.5 References 15D.5-1 Global Nuclear Fuel, General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A, and General Electric Standard Application for Reactor Fuel (Supplement for United States), NEDE-24011-P-A-US, latest revision.

15D.5-2 Global Nuclear Fuel - Americas, Supplemental Reload Licensing Report for Hope Creek Reload 22 Cycle 23, 004N8343, Revision 0, September 2019.

15D.5-3 Deleted 15D.5-4 Deleted 15D-9 HCGS-UFSAR Revision 24 May 21, 2020

15D.5-5 SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis for Hope Creek Generating Station at Power Uprate, NEDC-33172P, March 2005.

15D.5-6 GE Hitachi Nuclear Energy, GE Hitachi Boiling Water Reactor Detect and Suppress Solution - Confirmation Density, NEDC-33075P-A, Revision 8, November 2013.

15D.5-7 Deleted.

15D.5-8 Deleted.

15D.5-9 Deleted.

15D.5-10 Deleted.

15D.5-11 Deleted.

15D.5-12 Deleted.

15D.5-13 Deleted.

15D.5-14 Hope Creek Generating Station GNF2 ECCS-LOCA Evaluation, 002N5176-R0, Revision 0, August 2016 15D-10 HCGS-UFSAR Revision 23 November 12, 2018

TABLE 15D-1 INPUT PARAMETERS AND INITIAL CONDITIONS FOR RELOAD LICENSING ANALYSIS

1. Loss of Feedwater Heating:

Power (% of Rated) 100 Core Flow (% of Rated) 97.1 Feedwater Temperature (F) minus 110F 323.5 Core Mid-Plane Pressure (psia) 1034.6 Core Coolant Inlet Enthalpy (BTU/lbm) 524.4 Core Average Void Fraction (%)

50.6 Cycle Exposure MOC (5,500 MWd/ST)

2. Feedwater Controller Failure - Maximum Demand Power (% of Rated) 100 Core Flow (% of Rated) 105 Steam Flow (Mlbm/Hr) 15.0 Feedwater Flow Rate (Mlbm/Hr) 15.0 Feedwater Temperature (F) 331.5 Steam Dome Pressure (psig) 986.0 Core Exit Pressure (psig) 998.4 Core Coolant Inlet Enthalpy (BTU/lbm) 510.9 Core Average Void Fraction (%)

44.5 Cycle Exposure EOC with ICF EOC-RPT Operable Yes 1 of 2 HCGS-UFSAR Revision 24 May 21, 2020

TABLE 15D-1 (Cont)

3. Generator Load Rejection, No Bypass, Turbine Trip, No Bypass; and Inadvertent High Pressure Coolant Injection Startup Power (% of Rated) 100 Core Flow (% of Rated) 105 Steam Flow (Mlbm/Hr) 17.1 Feedwater Flow Rate (Mlbm/Hr) 17.1 Feedwater Temperature (F) 433.6 Steam Dome Pressure (psig) 1005.3 Core Exit Pressure (psig) 1018.5 Core Coolant Inlet Enthalpy (BTU/lbm) 526.2 Core Average Void Fraction (%)

49.9 Cycle Exposure EOC with ICF EOC-RPT Operable Yes

4. Rod Withdrawal Error Power (% of Rated) 100 Core Flow (% of Rated) 100 Steam Flow (Mlbm/Hr) 17.1 Feedwater Flow Rate (Mlbm/Hr) 17.1 Feedwater Temperature (F) 433.5 Core Mid-Plane Pressure (psig) 1020.3 Core Coolant Inlet Enthalpy (BTU/lbm) 525.1 Core Average Void Fraction (%)

51.4 High Worth Nominal Rod Pattern (For MCPR Yes Determination)

Cycle Exposure 2,500 MWd/ST 2 of 2 HCGS-UFSAR Revision 24 May 21, 2020

TABLE 15D-2 RESULT

SUMMARY

FOR LOSS OF FEEDWATER HEATING MANUAL CONTROL Final Conditions:

Power (% of Rated) 119.5 Core Flow (% of Rated) 99.2 Feedwater Temperature (F) 323.5 Core Mid-Plane Pressure (psig) 1027.4 Core Exit Pressure (psig)

N/A Core Coolant Inlet Enthalpy (BTU/lbm) 503.9 Core Average Void Fraction (%)

49.2 Peak Neutron Flux (% of Rated)(1)

N/A (1) The APRM simulated thermal Flux scram is not credited.

1 of 1 HCGS-UFSAR Revision 24 May 21, 2020

TABLE 15D-3 SEQUENCE OF EVENTS FOR FEEDWATER CONTROLLER FAILURE MAXIMUM DEMAND Time (Seconds)

Event Descriptions 0

A simulated failure to the Feedwater pump runout flow 12.4 L8 vessel level setpoint trips main turbine and feedwater pumps.

Turbine bypass operation is initiated.

12.4 Reactor scram actuated from TSV position switches.

16.0 Relief valves start to open.

1 of 1 HCGS-UFSAR Revision 24 May 21, 2020

TABLE 15D-4 SEQUENCE OF EVENTS FOR GENERATOR LOAD REJECTION WITHOUT BYPASS OPERATION Time (Seconds)

Event Descriptions

< 0.0 Loss of electrical load detected by the turbine generator.

0.0 Turbine generator load rejection sensing devices trip to initiate turbine control valve fast closure.

0.0 Turbine bypass valves fail to operate.

Fast closure of TCVs initiates scram and RPT.

0.07 TCVs are closed.

1.8 Group 1 SRVs are actuated in relief mode.

1 of 1 HCGS-UFSAR Revision 24 May 21, 2020

TABLE 15D-5 SEQUENCE OF EVENTS FOR INADVERTENT HIGH PRESSURE COOLANT INJECTION STARTUP Time (Seconds)

Event Descriptions 0.0 Inadvertent High Pressure Coolant Injection Startup.

17.0 Maximum Narrow Range Water Level / L8 Setpoint Level.

L8 vessel level setpoint trips main turbine, feedwater pumps, and HPCI. Turbine bypass operation is initiated.

17.0 Reactor scram actuated from TSV position switches.

19.4 Relief valves start to open 1 of 1 HCGS-UFSAR Revision 24 May 21, 2020

HCGS-UFSAR TABLE 15D-6 THIS INFORMATION HAS BEEN DELETED 1 of 1 Revision 14 July 26, 2005

TABLE 15D-7 SEQUENCE OF EVENTS FOR MISLOADED FUEL BUNDLE ACCIDENT

1.

During core loading operation, a fuel bundle is placed in the wrong location (or misoriented in the proper core location).

2.

Subsequently, the bundle intended for this location is placed in the assigned location of the previously misplaced bundle (Not Applicable for misoriented bundle accident)

3.

During core verification procedure, these errors are not observed

4.

The plant is brought to full power operation without detecting misplaced bundles

5.

The plant continues to operate HCGS-UFSAR 1 of 1 Revision 14 July 26, 2005

TABLE 15D-8 Mislocated Fuel Assembly Results Burnup Range OLMCPR GNF2/GE14 BOC13-EOC13 1.21 BOC23-EOC23 Non-Limiting The Mislocated Fuel Loading Error was determined to be non-limiting for Cycle 23 based on Cycle 13 results and the experience and procedural basis for the limiting fuel type, GNF2. The Cycle 23 Non-Limiting disposition for GNF2 also applies for the GE14 fuel.

1 of 1 HCGS-UFSAR Revision 24 May 21, 2020

TABLE 15D-9 Misoriented Fuel Assembly Results Burnup Range OLMCPR GNF2 GE14 BOC23-EOC23 1.23 1 of 1 HCGS-UFSAR Revision 24 May 21, 2020

C 2000 PSEG Nuclear, LLC. All Rights Reserved.

Updated FSAR PSEG Nuclear, LLC Hope Creek Nuclear Generating Station HOPE CREEK NUCLEAR GENERATING STATION Figure 15D-1 PLANT RESPONSE TO FW CONTROLLER FAILURE

% RATED (EOC ICF & FWTR (UB) )

% RATED LEVEL (INCHES ABOVE SEPARATOR SKIRT)

% RATED NEUTRON FLUX (% RATED)

TIME (SEC)

TIME (SEC)

TIME (SEC)

TIME (SEC) 0 2

4 6

8 10 12 14 16 0

2 4

6 8

10 12 14 16 0

2 4

6 8

10 12 14 16 0

2 4

6 8

10 12 14 16 0

25 100 125 150 0

100 150 200 250 300 0

100 125 25 50 20 30 40 50 60 70 80 0

10 20 30 40 50 1250 1100 1150 1200

-40

-30

-20

-10 REACTIVITY ($)

10 0

90 150 75

-25 1000 1050 50 50 75 PRESSURE (PSIA)

REVISION 24 MAY 21, 2020

C 2000 PSEG Nuclear, LLC. All Rights Reserved.

Updated FSAR PSEG Nuclear, LLC Hope Creek Nuclear Generating Station HOPE CREEK NUCLEAR GENERATING STATION

% RATED

% RATED TIME (SEC.)

TIME (SEC.)

TIME (SEC.)

PLANT RESPONSE TO LOAD REJECTION W/O BYPASS

% RATED 2

3 4

5 6

0 1

2 3

4 5

6 0

1 2

3 4

5 6

NEUTRON FLUX (% RATED)

-25 LEVEL (INCHES ABOVE SEPARATOR SKIRT) 0

-40 DOME PRESSURE RISE (PSI) 80

-30

-30

-10 10 0

100 125 150 75 50 25 0

0 10 20 30 40 50 60 70 TIME (SEC.)

0 1

2 3

4 5

6 1000 1050 1100 1200 1250 1300 1350 1400 1150 70 60 50 40 30 20 10 0

400 350 300 250 200 150 100 50 0

200 175 150 125 100 25 75 50 REACTIVITY ($)

Figure 15D-2 (EOC ICF (UB) )

REVISION 24 MAY 21, 2020

C 2000 PSEG Nuclear, LLC. All Rights Reserved.

Updated FSAR PSEG Nuclear, LLC Hope Creek Nuclear Generating Station HOPE CREEK NUCLEAR GENERATING STATION Figure 15D-3 PLANT RESPONSE TO TURBINE TRIP W/O BYPASS

% RATED

% RATED

% RATED TIME (SEC.)

TIME (SEC.)

TIME (SEC.)

(EOC ICF (UB) )

REACTIVITY ($)

LEVEL (INCHES ABOVE SEPERATOR SKIRT)

NEUTRON FLUX (% RATED)

PRESSURE (PSIA) 0 1

2 3

4 5

6 0

1 2

3 4

5 6

TIME (SEC.)

1000 1050 1100 1150 1200 1250 1300 1350 1400

-25 0

10 20 30 40 50 60 70 0

50 80 70 60 50 60 30 20 10 0

10 0

-40

-30

-10

-20 100 125 150 0

25 50 75 100 125 150 175 75 50 25 0

350 300 250 200 150 100 0

1 2

3 4

5 6

0 1

2 3

4 5

6 REVISION 24 MAY 21, 2020

C 2000 PSEG Nuclear, LLC. All Rights Reserved.

Updated FSAR PSEG Nuclear, LLC Hope Creek Nuclear Generating Station HOPE CREEK NUCLEAR GENERATING STATION

% RATED

% RATED LEVEL (INCHES ABOVE SEPARATOR SKIRT)

% RATED NEUTRON FLUX (% RATED)

TIME (SEC)

TIME (SEC)

TIME (SEC)

TIME (SEC) 0 2

4 6

8 10 12 14 16 0

2 4

6 8

10 12 14 16 0

2 4

6 8

10 12 14 16 0

2 4

6 8

10 12 14 16 0

25 100 125 150 0

100 150 200 250 0

100 125 25 50 0

10 20 30 40 50 60 0

10 20 30 40 50 1250 1100 1150 1200

-40

-30

-20

-10 REACTIVITY ($)

10 0

70 150 75

-25 1000 1050 50 50 75 PRESSURE (PSIA)

Figure 15D-4 (EOC ICF (UB))

COOLANT INJECTION STARTUP PLANT RESPONSE TO INADVERTENT HIGH PRESSURE 18 20 22 18 20 22 18 20 22 60 1300 18 20 22 REVISION 24 MAY 21, 2020

THIS FIGURE HAS BEEN DELETED PSEG NUCLEAR L.L.C.

HOPE CREEK GENERATING STATION HOPE CREEK UFSAR -REV 14 SHEET 1 OF 1 July 26, 2005 F15D-5

THIS FIGURE HAS BEEN DELETED PSEG NUCLEAR L.L.C.

HOPE CREEK GENERATING STATION HOPE CREEK UFSAR -REV 14 SHEET 1 OF 1 July 26, 2005 F15D-6

THIS FIGURE HAS BEEN DELETED PSEG NUCLEAR L.L.C.

HOPE CREEK GENERATING STATION

-HOPE CREEK UFSAR -REV 14 SHEET 1 OF 1 July 26,2005 F15D-7

THIS FIGURE HAS BEEN DELETED PSEG NUCLEAR L.L.C.

HOPE CREEK GENERATING STATION HOPE CREEK UFSAR -REV 14 SHEET 1 OF 1*

July 26,2005 F15D-8

THIS FIGURE HAS BEEN DELETED PSEG NUCLEAR L.L.C.

HOPE CREEK GENERATING STATION HOPE CREEK UFSAR -REV 14 SHEET 1 OF 1 July 26,2005 F15D-9

THIS FIGURE HAS BEEN DELETED PSEG NUCLEAR L.l.C.

HOPE CREEK GENERATING STATION HOPE CREEK UFSAR -REV 14 SHEET 1 OF 1 July 26,2005 F15D-10

THIS FIGURE HAS BEEN DELETED PSEG NUCLEAR L.L.C.

HOPE CREEK GENERATING STATION HOPE CREEK UFSAR -REV 14 SHEET 1 OF 1 July 26, 2005 F15D-11

THIS FIGURE HAS BEEN DELETED PSEG NUCLEAR L.L.C.

HOPE CREEK GENERATING STATION HOPE CREEK UFSAR -REV 14 July 26,_2005 SHEET1 OF 1 F15D-12

THIS FIGURE HAS BEEN DELETED PSEG NUCLEAR LL.C.

HOPE CREEK GENERATING STATION HOPE CREEK UFSAR -REV 14 SHEET1 OF1 J~l)' ~~. ~~~----* *-*

F15D-13

THIS FIGURE HAS BEEN DELETED PSEG NUCLEAR L.L.C.

HOPE CREEK GENERATING STATION HOPE CREEK UFSAR - REV 14 SHEET 1 OF 1 July 26,2005 F15D-14

THIS FIGURE HAS BEEN DELETED PSEG NUCLEAR L.L.C.

HOPE CREEK GENERATING STATION HOPE CREEK UFSAR -REV 14 SHEET 1 OF 1 July 26,2005 F15D.;15

THIS FIGURE HAS BEEN DELETED PSEG NUCLEAR L.L.C.

HOPE CREEK GENERATING STATION HOPE CREEK UFSAR -REV 14 SHEET 1 OF 1 July 26, 2005 F15D-1-6

THIS FIGURE HAS BEEN DELETED PSEG NUCLEAR L.L.C HOPE CREEK GENERATING STATION HOPE CREEK UFSAR -REV 14 SHEET 1 OF 1 July 26,2005 F15D-17

THIS FIGURE HAS BEEN DELETED.

PSEG NUCLEAR L.L.C.

HOPE CREEK GENERATING STATION HOPE CREEK UFSAR -REV 14 SHEET 1 OF 1 July 26,2005 F15D-18

THIS FIGURE HAS BEEN DELETED PSEG NUCLEAR L.L.C.

HOPE CREEK GENERATING STATION HOPE CREEK UFSAR -REV 14 SHEET 1 OF 1 July 26.2005 F15D-19