ML24142A432

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Enclosure 2: Hope Creek Generating Station Improved Technical Specifications Conversion - Volume 4
ML24142A432
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Site: Hope Creek PSEG icon.png
Issue date: 05/20/2024
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Public Service Enterprise Group
To:
Office of Nuclear Reactor Regulation
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ML24142A428 List:
References
LR-N24-0029, LAR H24-02
Download: ML24142A432 (1)
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Text

ENCLOSURE 2

VOLUME 4

HOPE CREEK GENERATING STATION

IMPROVED TECHNICAL SPECIFICATIONS CONVERSION

ITS CHAPTER 2.0 SAFETY LIMITS

Revision 0

LIST OF ATTACHMENTS

1. ITS Chapter 2.0, Safety Limits

ATTACHMENT 1

ITS Chapter 2.0, Safety Limits

Current Technical Specifications (CTS) Markup and Discussion of Changes ( DOCs)

ITS A01 ITS 2.0

(SLs) 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS (SLs) 2.1 SAFETY LIMITS React S THERMAL POWER, Low Pressure or Low Flow 2.1.1. 1 be R 2.1.1. 1 2.1.1 THERMAL POWER shall not exceed 24% of RATED THERMAL POWER with the reactor A vessel steam dome pressure less than 585 psig or core flow less than 10% of rated flow.

< < core :

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. M01

ACTION:

With THERMAL POWER exceeding 24% of RATED THERMAL POWER and the reactor vessel A02 steam dome pressure less than 585 psig or core flow less than 10% of rated flow, be in at least S HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1. ITS Insert 1 Insert 2 A02 Insert 3 M02 THERMAL POWER, High Pressure and High Flow 2.1.1. 2 t 2.1.1. 2 2.1.2 With reactor steam dome pressure greater than 5 85 psig and core flow greater than 10%

of rated flow: M ce The MINIMUM CRITICAL POWER RATIO ( MCPR) shall be 1.07.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. M01

ACTION:

With reactor st eam dome pressure greater than 585 psig and core flow greater than 10% of A02 rated flow and the MCPR below the value for the fuel stated in LCO 2.1.2, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1. S ITS Insert 1 Insert 2 A02 Insert 3 M02 REACTOR COOLANT SYSTEM PRESSURE 2.1.2 2.1.2 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dom e, A shall not exceed 1325 psig. s e be APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4. M01

ACTION:

With the reactor coolant system pressure, as measured in the reactor vessel steam dome, A02 above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less S than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification ITS 6.7.1. Insert 1 Insert 2 A02 Insert 3 M02

HOPE CREEK 2-1 Amendm ent No. 229 ITS 2.0

A02 INSERT 1

2.2 SL VIOL ATIONS

With any SL violation, the following actions shall be completed

A02 INSERT 2

2.2.1 Restore compliance with all SLs; and

M02 INSERT 3

2.2.2 Insert all insertable control rods.

Insert Page 2 -1 ITS A01 ITS 2.0

(SLs)

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

SAFETY LIMITS (Continued)

REACTOR VESSEL WATER LEVEL 2.1.1. 3 great er t han 2.1.1.3 2.1.4 The reactor vessel water level shall be above the top of the active irradiated fuel.

APPLICABILITY: OPERATIONAL CONDITIONS 3, 4 and 5. M01

ACTION:

With the reactor vessel water level at or below the top of the active irradiated fuel, manually A02 M02 initiate the ECCS to restore the water level, after depressurizing the reactor vessel, if required.

Comply with the requirements of Specification 6.7.1. Insert 1 A02

Insert 2 Insert 3 M02

See ITS 5.6

HOPE CREEK 2-2

ITS 2.0

A02 M02 INSERT 1

2.2 SL VIOL A TIONS

With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

A02 INSERT 2

2.2.1 Restore compliance with all SLs; and

M02 INSERT 3

2.2.2 Insert all insertable control rods.

Insert Page 2 -2 ITS A01 ITS 2.0

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

See 2.2 LIMITING SAFETY SYSTEM SETTINGS ITS 3.3.1.1

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS

2.2.1 The reactor protection system instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2.1-1.

APPLICABILITY: As shown in Table 3.3.1-1.

ACTION:

With a reactor protection system instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2.1-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 un til the channel is restored to OPERABLE status with its setpoint adjusted consistent with the Trip Setpoint value.

HOPE CREEK 2-3 ITS A01 ITS 2.0

TABLE 2.2.1-1 See REACTOR PROTECTION SYST EM INSTRUMENTATION SETPOINTS ITS 3.3.1.1

FUNCTIONAL UNI T TRIP SETPOINT ALLOWABLE VALUES

1. Intermediate Range Monitor, 125 div ofull scale 125 div of Neutron Flux-High scale

vage P Range M:

rlux -Upsc RATbaA RATbaA (Sdown) W W

Sherm Upscale**

low Bias -Two 5 6 w + 5 8 %** (a) wit 5 6 w + 6 0%** wit Reccion Loop m of 113.5% of RATED m of 115.5% of lperion THERMAL POWER RATED THERMAL POWER

2) Flow Biased-Single 5 6 -10.8 5 8 %** (a) wit 5 6 -9%) 6 0%** wit Recirculation Loop m of 113.5% of RATED m of 115.5% of Operation THERMAL POWER RATED THERMAL POWER
c. Neutron Flux - Upscale 3% ofATbaA 3% ofATba W TeMAL PlW

Inopere

2 -l - -4 s

lPcale See Cl bRATING LI Pl

Reactsteam 1037 ps 1057 ps Pressu -

Reactsatel - 12.5 inches above instrent 11.0 inches abov Lev z instrento

ain Steam Ision Valv - los 12% clos e

Se B 3/4 3-

    • Tvage P Scr function vi as a functcion loop dre flow (.

( When tutSPcrpoints arentce with Actf Table 3.1-he Shermal P -Upscalow Biaspointt the Cl bRATING LIMIbPT

Pbbh 2 Ament ITS A01 ITS 2.0

TABLE 2.2.1 -1 See REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS ITS 3.3.1.1 (continued)

ALLOWABLE FUNCTIONAL UNIT TRIP SETPOINT VALUES

6. This item intentionally blank
7. Drywell Pressure - High 1.68 psig 1.88 psig
8. Scram Discharge Volume Water Level - Hig h
a. Float Switch Elevation 110' 10.5" Elevation 111' 0.5"
b. Level Transmitter/Trip Unit Elevation 110' 10.5"* Elevation 111' 4.5"*
9. Turbine Stop Valve - Closure 5% closed 7% closed
10. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low 530 psig 465 psig
11. Reactor Mode Switch Shutdown Position NA NA
12. Manual Scram NA NA
  • 80.5" above instrument zero EL 104' 2" for Level Transmitter/Trip Unit A&B (South Header) 83.25" above instrument zero EL 103' 11.25" for Level Transmitter/Trip Unit C&D (North Header)

HOPE CREEK 2-5 Amendment No. 53

DISCUSSION OF CHANGES ITS CHAPTER 2.0, SAFETY LIMITS (SL s)

ADMINISTRATIVE CHANGES

A01 In the conversion of the Hope Creek Generating Station (HCGS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1433, Rev. 5.0, "Standard Technical Specifications - General E lectric BWR/4 Plants" (ISTS).

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 2.1 includes Actions to be taken with any Safety Limit violation.

  • CTS 2.1.1 Action states, in part, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
  • CTS 2.1.2 Action states, in part, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
  • CTS 2.1.3 Action states, in part, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
  • CTS 2.1.4 Action states, in part, manually initiate the ECCS to restore the water level, after depressurizing the reactor vessel.

ITS 2.2 Action states With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s: 2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.

This changes CTS 2.1.1, 2.1.2, 2.1.3, and 2.1.4 by requiring compliance with all SLs be restored within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. CTS 2.1.1 and 2.1.2 do not include an action to restore the Safety Limit, however, restoring the Safety Limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is an acceptable action. CTS 2.1.3 requires reactor coolant system pressure be reduced to less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to restore compliance with its Safety Limit. CTS 2.1.4 requires the reactor vessel water level be raised above the top of irradiated fuel to restore compliance with its Safety Limit, however, CTS 2.1.4 does not provide a completion time. See DOC M02 for the more restrictive change to CTS 2.1.4 to restore compliance with the Safety Limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. CTS does not state that all insertable control rods be inserted.

Though MODE 3 is achieved by inserting all control rods, this action is not required for CTS 2.1.3 and CTS 2.1.4. See DOC M02 for the more restrictive change discussion related to ITS 2.2.2 requirement to Insert all insertable control rods.

The purpose of CTS 2.1 Actions is to ensure specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs), and to protect the RCS against overpressuriza tion. ITS Action 2.2.1 requires compliance with all Safety Limits be restored within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. This change is acceptable because restoring compliance with all Safety Limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> ensures continued safe operation. These changes are designated as adm inistrative changes and are acceptable because they do not result in technical changes to the CTS.

Hope Creek Page 1 of 3 DISCUSSION OF CHANGES ITS CHAPTER 2.0, SAFETY LIMITS (SL s)

MORE RESTRICTIVE CHANGES

M01 CTS 2.1.1 and CTS 2.1.2 state APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2, CTS 2.1.3 states APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4, and CTS 2.1.4 states APPLICABILITY: OPERATIONAL CONDITIONS 3, 4 and 5. CTS 2.1 Safety Limit requirements specifies the individual Applicability for the applicable Operational Conditions. In ITS 2.0, the Applicability is not specified; therefore, the Safety Limit requirem ents are applicable in all Modes. Although it is not p ossible to violate some Safety Limit s in some Modes, all Safety Limits should be applicable at all times to ensure continued safe operation and to ensure compliance with the requirements of 10 CFR 50.36(c)(1) which apply at all times. This change in Applicability will have no negative impact on safety.

The purpose of CTS 2.1 Applicability is to identify the Operational Conditions (MODES) that apply for Safety Limit s 2.1.1, 2.1.2 and 2.1.4 to ensure specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs),

and Safety Limit 2.1.3 to protect the RCS against overpressurization. Th is change is acceptable because requiring each Safety Limit be applicable at all tim es ensures continued safe operation. These changes are designated as more restrictive because the c hanges impose additional restrictions on plant operation.

M02 CTS 2.1 includes Actions to be taken with any Safety Limit violation.

  • CTS 2.1.1 Action states, in part, With THERMAL POWER exceeding 24% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 585 psig or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
  • CTS 2.1.2 Action states, in part, With reactor steam dom e pressure greater than 585 psig and core flow greater than 10% of rated flow and the MCPR below the value for the fuel stated in LCO 2.1.2, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
  • CTS 2.1.3 Action states, in part, With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
  • CTS 2.1.4 Action states, in part, With the reactor vessel water level at or below the top of the active irradiated fuel, manually initiate the ECCS to restore the water level, after depressurizing the reactor vessel.

ITS 2.2.2 states With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s: 2.2.1 Restore compliance with all SLs, and 2.2.2 Insert all insertable control rods.

This changes the CTS by requiring compliance with any Safety Limit, including CTS 2.1.4 be restored and all insertable control rods be inserted within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> with any Safety Limit violation.

Hope Creek Page 2 of 3 DISCUSSION OF CHANGES ITS CHAPTER 2.0, SAFETY LIMITS (SL s)

The purpose of CTS 2.1 is to provide Actions when a Safety Limit is violated to ensure specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs), and to protect against RCS overpressurization. This change is acceptable because the Actions restore compliance with all Safety Limits and requires all insertable control rods be inserted within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to ensure continued safe operation and to be consistent with the requirements of 10 CFR 50.36(c)(1), which requires the reactor to be shutdown if any safety limit is violated. These changes are designated as more restrictive because the c hanges im pose additional restrictions on plant operation.

M03 CTS 2.1.2 sta te s With reactor st eam dome pressure greater than 585 psig and core flow greater than 10% of rated flow : The MINIMUM CRITICAL POWER RATIO (MCPR) shall be 1.07. ITS 2.1.1.2 states With the reactor steam dome pressure 585 psig and core flow 10% rated core flow: MCPR shall be 1.07.

CTS 2.1.2 does not address the condition when steam dom e pressure and core flow are equal to their limits. ITS 2.1.1.2 limits on steam dome pressure and core flow are specified as "greater than or e qual to." This changes CTS 2.1.2 by adding the ITS 2.1.1.2 requirement of "greater than or eq ual to to the reactor steam dome pressure and core flow limits. This change is designated as more restrictive because it resolves a discontinuity and imposes an additional restriction on plant o peration.

RELOCATED SPECIFICATIONS

None

REMOVED DETAIL CHANGES

None

LESS RESTRICTIVE CHANGES

None

Hope Creek Page 3 of 3 Improved Standard Technical Specifications (ISTS) Mark up and Justification for Deviations (JFD s)

CTS SLs 2.0

2.0 SAFETY LIMITS (SLs)

2.1 SLs

2.1 2.1.1 Reactor Core SLs 585

2.1.1 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10% 3 rated core flow:

24 THERMAL POWER shall be 25% RTP. 3 585 2.1.2 2.1.1.2 With the reactor steam dome pressure 785 psig and core flow 10% 3 rated core flow:

MCPR shall be [1.07] [for two recirculation loop operation or [1.08] 2 for single recirculation loop operation. ]

2.1.4 2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.

2.1 2.1.2 Reactor Coolant System Pressure SL

2.1.3 Reactor steam dome pressure shall be 1325 psig.

M01 2.2 SL VIOLATIONS

2.1.1, 2. 1. 2, 2. 1.3, With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

and 2.1. 4 ACTIONS 2.1.1, 2. 1. 2, 2. 1.3, 2.2.1 Restore compliance with all SLs; and and 2.1. 4 ACTIONS M02 2.2.2 Insert all insertable control rods.

General Electric BWR/4 STS 2.0-1 Rev. 5.0 1 Hope Creek Amendment XXX JUSTIFICATION FOR DEVIATIONS ITS CHAPTER 2.0, SAFETY LIMITS (SLs)

1. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant specific nomenclature, number, reference, system description, analysis, licensing basis, or licensing basis description.
2. The ISTS contains bracketed information and/or values that are generic to all General Electric BWR/4 vintage plan ts. The brackets are removed, and the proper plant specific information/value is provided. This is acceptable since the information/value is changed to reflect the current licensing basis.
3. ITS Safety Limits (SLs), specifically Reactor Core SLs 2.1.1.1 and 2.1.1.2 are changed to provide the proper plant specific safety limit values. This is acceptable since the values are changed to reflect the current licensing basis.

Hope Creek Page 1 of 1 Improved Standard Technical Specifications (ISTS) B ases Mar kup and Justification for Deviations (JFD s)

Reactor Core SLs B 2.1.1

B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs

BASES

BACKGROUND GDC 10 (Ref. 1) requires, and SL s ensure, that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs).

The fuel cladding integrity SL is set such that no significant fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a stepback approach is used to establish an SL, such that the MCPR is not less than the limit specified in Specification 2.1.1.2 for [both General Electric Company (GE) and 2 5 Advanced Nuclear Fuel Corporation (ANF) fuel]. MCPR greater than the specified limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.

The fuel cladding is one of the physical barriers that separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking.

Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses, which occur from reactor operation significantly above design conditions.

REVIEWER'S NOTE ------------------------------

In the Background and Applicable Safety Analysis sections, select the SLMCPR95/95 discussion or the 99.9% of the fuel rods discussion as the 4 applicable SL 2.1.1.2 basis.

While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross, rather than incremental, cladding deterioration. Therefore, the fuel cladding SL is defined with a margin to the conditions that would produce onset of transition boiling (i.e.,

MCPR = 1.00). These conditions represent a significant departure from the condition intended by design for planned operation. [This is 2 4 accomplished by having a Safety Limit Minimum Critical Power Ratio (SLMCPR) design basis, referred to as SLMCPR 95/95, which corresponds to a 95% probability at a 95% confidence level (the 95/95 MCPR criterion) that transition boiling will not occur. ] [The MCPR fuel cladding integrity SL 2 4 ensures that during normal operation and during AOOs, at least 99.9% of the fuel rods in the core are not susceptible to boiling transition.]

General Electric BWR/4 STS B 2.1.1-1 Rev. 5.0 Hope Creek Revision XXX Reactor Core SLs B 2.1.1

BASES

BACKGROUND (continued)

Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of transition boiling and the resultant sharp reduction in heat transfer coefficient.

Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY operation and AOOs. [The Tech Spec SL is set generically on a fuel ANALYSES product MCPR correlation basis as the MCPR which corresponds to a 2 4 95% probability at a 95% confidence level that transition boiling will not occur, referred to as SLMCPR 95/95] [The reactor core SLs are established to preclude violation of the fuel design criterion that a MCPR limit is to be 2 4 established, such that at least 99.9% of the fuel rods in the core would not be susceptible to boiling transition.]

The Reactor Protection System setpoints (LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation"), in combination with the other LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System water level, pressure, and THERMAL POWER level that would result in reaching the MCPR limit.

2.1.1.1a Fuel Cladding Integrity [General Electric Company (GE) Fuel] 2 585 GE critical power correlations are applicable for all critical power calculations at pressures 785 psig and core flows 10% of rated flow. 3 For operation at low pressures or low flows, another basis is used, as follows:

Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be > 4.5 psi. Analyses (Ref. 2) show that with a bundle flow of 28 x 103 lb/hr, bundle pressure drop is nearly independent of Subsequent critical power tests for bundle power and has a value of 3.5 psi. Thus, the bundle flow with newer fuel designs acquired data at a 4.5 psi driving head will be > 28 x 10 3 lb/hr. Full scale ATLAS test extended pressures down to 600 psia. data taken at pressures from 14.7 psia to 800 psia indicate that the The data have been shown in Reference 2 to support the critical fuel assembly critical power at this flow is approximately 3.35 MWt. 1 power correlations established for With the design peaking factors, this corresponds to a THERMAL plant specific fuel designs.

POWER > 50 % RTP. Thus, a THERMAL POWER limit of 25% RTP 3 for reactor pressure < 785 psig is conservative. 24 3 585

General Electric BWR/4 STS B 2.1.1-2 Rev. 5.0 Hope Creek Revision XXX Reactor Core SLs B 2.1.1

BASES

APPLICABLE SAFETY ANALYSES (continued)

2.1.1.1b Fuel Cladding Integrity [Advanced Nuclear Fuel Corporation 2 (ANF) Fuel]

The use of the XN-3 correlation is valid for critical power calculations at pressures > 580 psig and bundle mass fluxes > 0.25 x 10 6 lb/hr-ft2 (Ref. 3). For operation at low pressures or low flows, the fuel cladding integrity SL is established by a limiting condition on core THERMAL POWER, with the following basis:

Provided that the water level in the vessel downcomer is maintained above the top of the active fuel, natural circulation is sufficient to ensure a minimum bundle flow for all fuel assemblies that have a relatively high power and potentially can approach a critical heat flux condition. For the ANF 9x9 fuel design, the minimum bundle flow is > 30 x 10 3 lb/hr. For the ANF 8x8 fuel design, the minimum bundle flow is > 28 x 10 3 lb/hr. For all designs, the coolant minimum bundle flow and maximum flow area are such that the mass flux is always > 0.25 x 10 6 lb/hr-ft2. Full scale critical power tests taken at pressures down to 14.7 psia indicate that the fuel assembly critical power at 0.25 x 10 6 lb/hr-ft2 is approximately 3.35 MWt.

At 25% RTP, a bundle power of approximately 3.35 MWt corresponds to a bundle radial peaking factor of > 3.0, which is significantly higher than the expected peaking factor. Thus, a THERMAL POWER limit of 25% RTP for reactor pressures < 785 psig is conservative.

2.1.1.2a MCPR [GE and Westinghouse Fuel] 2 5

The fuel cladding integrity SL is set such that no significant fuel damage is calculated to occur if the limit is not violated. Since the parameters that result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions that result in the onset of transition boiling have been used to mark the beginning of the region in which fuel damage could occur. Although it is recognized that the onset of transition boiling would not result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. [The Technical Specification SL value is dependent on the fuel product line and the corresponding MCPR correlation, which is cycle independent. The value is based on the Critical Power Ratio (CPR) 2 4 data statistics and a 95% probability with 95% confidence that rods are not susceptible to boiling transition, referred to as MCPR 95/95.]

General Electric BWR/4 STS B 2.1.1-3 Rev. 5.0 Hope Creek Revision XXX Reactor Core SLs B 2.1.1

BASES

APPLICABLE SAFETY ANALYSES (continued)

Reviewer's Note --------------------------------

The MCPR95/95 Values by Vendor and Fuel Product Type:

5 Vendor Fuel Type MCPR95/95 Global GE14 1.06 Nuclear Fuel Global GNF2 1.07 Nuclear Fuel Global GNF3 1.07 Nuclear Fuel Westinghouse Optima2 1.06 GNF2

[The SL is based on [Fuel Type] fuel. For cores with a single fuel product line, the SLMCPR 95/95 is the MCPR95/95 for the fuel type. For cores loaded 2 with a mix of applicable fuel types, the SLMCPR 95/95 is based on the largest (i.e., most limiting) of the MCPR values for the fuel product lines that are fresh or once-burnt at the start of the cycle. ]

[However, the uncertainties in monitoring the core operating state and in 2 4 the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity SL is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition, considering the power distribution within the core and all uncertainties.

The MCPR SL is determined using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the approved General Electric Critical Power correlations. Details of the fuel cladding integrity SL calculation are given in Reference 2. Reference 2 also includes a tabulation of the uncertainties used in the determination of the MCPR SL and of the nominal values of the parameters used in the MCPR SL statistical analysis.] The approved revision number of Reference 2 is identified in the COLR. 1 2.1.1.2b MCPR [ANF Fuel] 2 5

The MCPR SL ensures sufficient conservatism in the operating MCPR limit that, in the event of an AOO from the limiting condition of operation, at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition. The margin between calculated boiling transition (i.e.,

MCPR = 1.00) and the MCPR SL is based on a detailed statistical

General Electric BWR/4 STS B 2.1.1-4 Rev. 5.0 Hope Creek Revision XXX Reactor Core SLs B 2.1.1

BASES

APPLICABLE SAFETY ANALYSES (continued)

procedure that considers the uncertainties in monitoring the core 2 5 operating state. One specific uncertainty included in the SL is the uncertainty inherent in the XN-3 critical power correlation. Reference 3 describes the methodology used in determining the MCPR SL.

The XN-3 critical power correlation is based on a significant body of practical test data, providing a high degree of assurance that the critical power, as evaluated by the correlation, is within a small percentage of the actual critical power being estimated. As long as the core pressure and flow are within the range of validity of the XN-3 correlation, the assumed reactor conditions used in defining the SL introduce conservatism into the limit because bounding high radial power factors and bounding flat local peaking distributions are used to estimate the number of rods in boiling transition. Still further conservatism is induced by the tendency of the XN-3 correlation to overpredict the number of rods in boiling transition.

These conservatisms and the inherent accuracy of the XN-3 correlation provide a reasonable degree of assurance that there would be no transition boiling in the core during sustained operation at the MCPR SL.

If boiling transition were to occur, there is reason to believe that the integrity of the fuel would not be compromised. Significant test data accumulated by the NRC and private organizations indicate that the use of a boiling transition limitation to protect against cladding failure is a very conservative approach. Much of the data indicate that BWR fuel can survive for an extended period of time in an environment of boiling transition.

2.1.1.3 Reactor Vessel Water Level

During MODES 1 and 2 the reactor vessel water level is required to be above the top of the active fuel to provide core cooling capability. With fuel in the reactor vessel during periods when the reactor is shut down, consideration must be given to water level requirements due to the effect of decay heat. If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level becomes < 2/3 of the core height. The reactor vessel water level SL has been established at the top of the active irradiated fuel to provide a point that can be monitored and to also provide adequate margin for effective action.

General Electric BWR/4 STS B 2.1.1-5 Rev. 5.0 Hope Creek Revision XXX Reactor Core SLs B 2.1.1

BASES

SAFETY LIMITS The reactor core SLs are established to protect the integrity of the fuel clad barrier to the release of radioactive materials to the environs.

SL 2.1.1.1 and SL 2.1.1.2 ensure that the core operates within the fuel design criteria. SL 2.1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel in order to prevent elevated clad temperatures and resultant clad perforations.

APPLICABILITY SLs 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES.

3 50.67 SAFETY LIMIT Exceeding an SL may cause fuel damage and create a potential for 1 VIOLATIONS radioactive releases in excess of 10 CFR 100, "Reactor Site Criteria,"

limits (Ref. 4). Therefore, it is required to insert all insertable control rods and restore compliance with the SLs within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10.

2. NEDE-24011-P-A (latest approved revision).
3. XN-NF524(A), Revision 1, November 1983. 2 5
4. 10 3 CFR 100. 50.67 1

General Electric BWR/4 STS B 2.1.1-6 Rev. 5.0 Hope Creek Revision XXX RCS Pressure SL B 2.1.2

B 2.0 SAFETY LIMITS (SLs)

B 2.1.2 Reactor Coolant System (RCS) Pressure SL

BASES

BACKGROUND The SL on reactor steam dome pressure protects the RCS against overpressurization. In the event of fuel cladding failure, fission products are released into the reactor coolant. The RCS then serves as the primary barrier in preventing the release of fission products into the atmosphere. Establishing an upper limit on reactor steam dome pressure ensures continued RCS integrity. According to 10 CFR 50, Appendix A, GDC 14, "Reactor Coolant Pressure Boundary," and GDC 15, "Reactor Coolant System Design" (Ref. 1), the reactor coolant pressure boundary (RCPB) shall be designed with sufficient margin to ensure that the design conditions are not exceeded during normal operation and anticipated operational occurrences (AOOs).

During normal operation and AOOs, RCS pressure is limited from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code (Ref. 2). To ensure system integrity, all RCS components are hydrostatically tested at 125% of design pressure, in accordance with ASME Code requirements, prior to initial operation when there is no fuel in the core. Any further hydrostatic testing with fuel in the core may be done under LCO 3.10.1, "Inservice Leak and Hydrostatic Testing Operation." Following inception of unit operation, RCS components shall be pressure tested in accordance with the requirements of ASME Code,Section XI (Ref. 3).

Overpressurization of the RCS could result in a breach of the RCPB, 50.67 reducing the number of protective barriers designed to prevent radioactive releases from exceeding the limits specified in 10 CFR 100, 1 "Reactor Site Criteria" (Ref. 4). If this occurred in conjunction with a fuel cladding failure, fission products could enter the containment atmosphere.

APPLICABLE The RCS safety/relief valves and the Reactor Protection System Reactor SAFETY Vessel Steam Dome Pressure - High Function have settings established ANALYSES to ensure that the RCS pressure SL will not be exceeded.

The RCS pressure SL has been selected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered. The reactor pressure vessel is designed to Section III of the ASME, Boiler and Pressure Vessel Code, [1971 Edition], including 2 Addenda through the [winter of 1972] (Ref. 5), which permits a maximum 2 pressure transient of 110%, 1375 psig, of design pressure 1250 psig.

The SL of 1325 psig, as measured in the reactor steam dome, is 1969 1968

General Electric BWR/4 STS B 2.1.2-1 Rev. 5.0 Hope Creek Revision XXX RCS Pressure SL B 2.1.2

BASES

APPLICABLE SAFETY ANALYSES (continued)

equivalent to 1375 psig at the lowest elevation of the RCS. The RCS is designed to the USAS Nuclear Power Piping Code, Section B31.1,

[1969 Edition], including Addenda through [July 1, 1970 ] (Ref. 6), for the 2 reactor recirculation piping, which permits a maximum pressure transient of 110% of design pressures of 1250 psig for suction piping and 1500 psig for discharge piping. The RCS pressure SL is selected to be the lowest transient overpressure allowed by the applicable codes.

SAFETY LIMITS The maximum transient pressure allowable in the RCS pressure vessel under the ASME Code,Section III, is 110% of design pressure. The maximum transient pressure allowable in the RCS piping, valves, and fittings is 110% of design pressures of 1250 psig for suction piping and 1500 psig for discharge piping. The most limiting of these allowances is the 110% of the suction piping design pressures; therefore, the SL on maximum allowable RCS pressure is established at 1325 psig as measured at the reactor steam dome.

APPLICABILITY SL 2.1.2 applies in all MODES.

SAFETY LIMIT Exceeding the RCS pressure SL may cause immediate RCS failure and VIOLATIONS create a potential for radioactive releases in excess of 10 CFR 100, 50.67 1 "Reactor Site Criteria," limits (Ref. 4). Therefore, it is required to insert all insertable control rods and restore compliance with the SL within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time ensures that the operators take prompt remedial action and also assures that the probability of an accident occurring during this period is minimal.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 14, GDC 15, and GDC 28. 1

2. ASME, Boiler and Pressure Vessel Code,Section III, Article NB-7000.
3. ASME, Boiler and Pressure Vessel Code,Section XI, Article IW-5000.
4. 10 CFR 100. 50.67 1
5. ASME, Boiler and Pressure Vessel Code,Section III, [1971 Edition], 2 Addenda [winter of 1972]. 1969 1968
6. ASME, USAS, Nuclear Power Piping Code, Section B31.1, [1969 2 Edition], Addenda [July 1, 1970 ].

General Electric BWR/4 STS B 2.1.2-2 Rev. 5.0 Hope Creek Revision XXX JUSTIFICATION FOR DEVIATIONS ITS 2.0 BASES, SAFETY LIMITS (SLs)

1. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant specific nomenclature, number, reference, system description, analysis, licensing basis, or licensing basis description.
2. The ISTS contains bracketed information and/or values that are generic to all General Electric BWR/4 vintage plants. The brackets are removed, and the proper plant specific information/value is provided. This is acceptable since the information/value is changed to reflect the current licensing basis.
3. ITS Safety Limits (SLs), specifically Reactor Core SLs 2.1.1.1 and 2.1.1.2 are changed to provide the proper plant specific safety limit values. This is acceptable since the values are changed to reflect the current licensing basis.
4. The Reviewers Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal. License Amendment 219, dated September 19, 2019 (ADAMS Accession No. ML19218A305) modified the MCPR safety limit based on the use 95/95 criterion (i.e., 95% probability at a 95%

confidence level that the hot rod does not experience transition boiling). The ISTS Bases bracketed text associated with use of the SLMCPR 95/95 value applies to the Hope Creek Generating Station.

5. Hope Creek Generating Station uses fuel type Global Nuclear Fuel GNF2 fuel.

Hope Creek Page 1 of 1 Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS CHAPTER 2.0, SAFETY LIMITS (SLs)

There are no specific No Significant Hazards Considerations for this Specification.

Hope Creek Page 1 of 1

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