Text

Issuance of Amendment 187 Relocation of Specific Surveillance Frequencies to a Licensee - Controlled Program Based on Technical Specification Task Force (TSTF) Change TSTF-425
	ML103410243
Person / Time
Site:	Hope Creek
Issue date:	02/25/2011
From:	Richard Ennis; Plant Licensing Branch 1
To:	Joyce T; Public Service Enterprise Group
	Ennis R, NRR/DORL, 415-1420
References
	TAC ME3545
	Download: ML103410243 (145)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 25, 2011 Mr. Thomas Joyce President and Chief Nuclear Officer PSEG Nuclear LLC P.O. Box 236, N09 Hancocks Bridge, NJ 08038

SUBJECT:

HOPE CREEK GENERATING STATION -ISSUANCE OF AMENDMENT RE:

RELOCATION OF SPECIFIC SURVEILLANCE FREQUENCIES TO A LICENSEE-CONTROLLED PROGRAM BASED ON TECHNICAL SPECIFICATION TASK FORCE (TSTF) CHANGE TSTF-425 (TAC NO. ME3545)

Dear Mr. Joyce:

The Commission has issued the enclosed Amendment No. 187 to Facility Operating License No. NPF-57 for the Hope Creek Generating Station (HCGS). This amendment consists of changes to the Technical Specifications (TSs) and Facility Operating License in response to your application dated March 19, 2010, as supplemented by letters dated July 28, 2010, and January 10, 2011.

The amendment modifies the TSs by relocating specific surveillance frequencies to a licensee controlled program. The changes are based on Nuclear Regulatory Commission-approved TS Task Force (TSTF) change TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control-RITSTF [Risk-Informed TSTF] Initiative 5b."

A copy of our safety evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, Richard B. Ennis, Senior Project Manager Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket 1\\10. 50-354

Enclosures:

1. Amendment No. 187 to License No. NPF-57

2. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 PSEG NUCLEAR LLC DOCKET NO. 50-354 HOPE CREEK GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 187 License No. NPF-57

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment filed by PSEG Nuclear LLC dated March 19,2010, as supplemented by letters dated July 28,2010, and January 10, 2011, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-57 is hereby amended to read as follows:

-2 (2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 187, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into the license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

The license amendment is effective as of its date of issuance and shall be implemented within 120 days.

FOR THE NUCLEAR REGULATORY COMMISSION

~::~

Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: February 25, 2011

ATTACHMENT TO LICENSE AMENDMENT NO. 187 FACILITY OPERATING LICENSE NO. NPF-57 DOCKET NO. 50-354 Replace the following page of the Facility Operating License with the revised pqge. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Remove Insert 3

3 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert Remove Insert Remove Insert Remove Insert 3/4 1-2 3/41-2 3/43-66 3/43-66 3/46-10 3/46-10 3/4 8-4 3/4 8-4 3/4 1-4 3/4 1-4 3/43-67 3/43-67 3/46-13 3/46-13 3/4 8-5 3/4 8-5 3/4 1-5 3/4 1-5 3/43-74 3/43-74 3/46-14 3/46-14 3/4 8-6 3/48-6 3/41-7 3/4 1-7 3/43-82 3/43-82 3/46-15 3/46-15 3/4 8-8 3/4 8-8 3/41-10 3/4 1-10 3/43-83 3/43-83 3/46-16 3/46-16 3/4 8-9 3/4 8-9 3/41-14 3/41-14 3/43-87 3/43-87 3/46-18 3/46-18 3/48-13 3/48-13 3/4 1-19 3/41-19 3/43-88 3/43-88 3/46-44 3/46-44 3/48-14 3/48-14 3/41-20 3/4 1-20 3/43-105 3/43-105 3/46-45 3/46-45 3/48-20 3/48-20 3/42-1 3/42-1 3/4 3-108 3/43-108 3/46-46 3/46-46 3/48-23 3/48-23 3/42-3 3/42-3 3/43-109 3/43-109 3/46-47 3/46-47 3/48-24 3/48-24 3/42-5 3/42-5 3/43-110 3/43-110 3/46-48 3/46-48 3/48-25 3/48-25 3/43-1 3/43-1 3/44-2a 3/44-2a 3/46-49 3/46-49 3/48-30 3/48-30 3/43-7 3/4 3-7 3/4 4-4 3/4 4-4 314 6-51 3/46-51 3/48-38 3/48-38 3/4 3-8 3/4 3-8 3/4 4-5 3/44-5 3/46-51 a 3/46-51 a 3/48-40 3/48-40 3/43-10 3/43-10 3/44-8 3/44-8 3/46-52 3/46-52 3/48-41 3/48-41 3/43-28 3/43-28 3/44-9 3/44-9 3/46-52a 3/46-52a 3/48-44 3/48-44 3/43-29 3/43-29 3/44-10a 3/44-10a 3/46-53 3/46-53 3/4 9-2 3/49-2 3/43-30 3/43-30 3/44-12 3/44-12 3/46-53a 3/46-53a 3/4 9-3 3/4 9-3 3/43-31 3/43-31 3/44-20 3/44-20 3/46-55 3/46-55 3/4 9-4 3/4 9-4 3/43-32 3/43-32 3/44-21 3/44-21 3/4 7-2 3/4 7-2 3/4 9-5 3/4 9-5 3/43-39 3/43-39 3/44-22 3/44-22 3/4 7-4 3/47-4 3/4 9-11 3/49-11 3/43-40 3/43-40 3/44-25 3/44-25 3/47-5 3/47-5 3/49-12 3/49-12 3/43-41 3/43-41 3/44-28 3/44-28 3/47-6a 3/47-6a 3/49-14 3/49-14 3/43-44 3/43-44 3/44-29 3/44-29 3/4 7-7 3/47-7 3/49-16 3/49-16 3/43-46 3/43-46 3/45-4 3/45A 3/47-8 3/4 7-8 3/49-17 3/49-17 3/43-50 3/43-50 3/4 5-5 3/45-5 3/4 7-9 3/4 7-9 3/49-18 3/49-18 3/43-51 3/43-51 3/4 5-7 3/4 5-7 3/4 7-11 3/47-11 3/4 10-1 3/4 10-1 3/43-55 3/43-55 3/4 5-9 3/4 5-9 3/47-12 3/47-12 3/410-3 3/4 10-3 3/43-60 3/43-60 3/46-1 3/46-1 3/47-19 3/47-19 3/4 10-4 3/4 10-4 3/43-61 3/43-61 3/4 6-6 3/46-6 3/47-21 3/47-21 3/4 10-6 3/4 10-6 3/43-62 3/43-62 3/4 6-9 3/4 6-9 3/4 11-2 3/411-2 3/411-17 3/4 11-17 6-16e 6-/16e

- 3 (4)

PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5)

PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6)

PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility. Mechanical disassembly of the GE14i isotope test assemblies containing Cobalt-60 is not considered separation.

(7)

PSEG Nuclear LLC, pursuant to the Act and 10 CFR Part 30, to intentionally produce, possess, receive, transfer, and use Cobalt-60.

C.

This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level PSEG Nuclear LLC is authorized to operate the facility at reactor core power levels not in excess of 3840 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 187, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into the license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

Inservice Testing of Pumps and Valves (Section 3.9.6, SSER No. 4)*

This License Condition was satisfied as documented in the letter from W. R. Butler (NRC) to C. A. McNeill, Jr. (PSE&G) dated December 7, 1987. Accordingly, this condition has been deleted.

The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.

Amendment No. 187

REACTIVITY CONTROL SYSTEMS 3/4.1.2 REACTIVITY ANOMALIES LIMITING CONDITION FOR OPERATION 3.1.2 The reactivity equivalence of the difference between the actual ROD DENSITY and the predicted ROD DENSITY shall not exceed 1 % delta klk.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With the reactivity equivalence difference exceeding 1 % delta klk:

a.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months perform an analysis to determine and explain the cause of the reactivity difference; operation may continue if the difference is explained and corrected.

b.

Otherwise. be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.1.2 The reactivity equivalence of the difference between the actual ROD DENSITY and the predicted ROD DENSITY shall be verified to be less than or equal to 1 % delta klk:

a.

During the first startup following CORE ALTERATIONS. and

b.

In accordance with the Surveillance Frequency Control Program during POWER OPERATION.

HOPE CREEK 3/41-2 Amendment No. 187

REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

ACTION (Continued)

d.

One or more BPWS groups with four or more inoperable control rods*****, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , restore control rod(s) to OPERABLE status.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

e.

With more than 8 control rods inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

f.

With one or more scram discharge volume (SDV) vent or drain lines*** with one valve inoperable, isolate the associated line within 7 days or be in at least HOT SHUTDOWN within the next 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s:***

g.

With one or more SDV vent or drain lines"'** with both valves inoperable, isolate the associated line within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . "'***

SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The scram discharge volume drain and vent valves shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by:

a.

Verifying each valve to be open,* and

b.

Cycling each valve through at least one complete cycle of full travel.

These valves may be closed intermittently for testing under administrative controls.

May be rearmed intermittently, under administrative control, to permit testing associated with restoring the control rod to OPERABLE status.

Separate Action entry is allowed for each SDV vent and drain line.

An isolated line may be unisolated under administrative control to allow draining and venting of the SDV.

Not applicable when THERMAL POWER is greater than 8.6% RATED THERMAL POWER.

HOPE CREEK 3/4 1-4 Amendment No. 187

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.1.3.1.2 When above the low power setpoint of the RWM, all withdrawn control rods not required to have their directional control valves disarmed electrically or hydraulically shall be demonstrated OPERABLE by moving each control rod at least one notch:

a.

In accordance with the Surveillance Frequency Control Program, and

b.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when any control rod is immovable as a result of excessive friction or mechanical interference.

4.1.3.1.3 All control rods shall be demonstrated OPERABLE by performance of Surveillance Requirements 4.1.3.2, 4.1.3.3, 4.1.3.5, 4.1.3.6 and 4.1.3.7.

4.1.3.1.4 The scram discharge volume shall be determined OPERABLE by demonstrating:

a.

The scram discharge volume drain and vent valves OPERABLE in accordance with the Surveillance Frequency Control Program, by verifying that the drain and vent valves:

1.

Close within 30 seconds after receipt of a signal for control rods to scram, and

2.

Open when the scram signal is reset.

HOPE CREEK 3/4 1-5 Amendment No. 187

REACTIVITY CONTROL SYSTEIVIS CONTROL ROD SCRAM INSERTION TIMES 3.1.3.3 No more than 13 OPERABLE control rods shall be "slow," in accordance with Table 3.1.3.3-1, and no more than 2 OPERABLE control rods that are "slow" shall occupy adjacent locations.

Table 3.1.3.3-1

NOTES ---------------------------------------------------------------

1.

OPERABLE control rods with scram times not within the limits of this Table are considered "slow."

2.

Enter applicable Conditions and Required Actions of LCO 3.1.3.2, "Control Rod Maximum Scram Insertion Times," for control rods with scram times> 7.0 seconds to notch position 05. These control rods are inoperable in accordance with SR 4.1.3.2 and are not considered "slow."

Scram Times(a)(b) (Seconds) When Reactor Notch Position Steam Dome Pressure;::: 800 psig 45 0.52 39 0.86 25 1.91 05 3.44 (a)

Maximum scram time from fully withdrawn position, based on de-energization of scram pilot valve solenoids at time zero.

(b)

Scram times as a function of reactor steam dome pressure, when < 800 psig are within established limits.

APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With more than 13 OPERABLE control rods exceeding any of the above limits or more than 2 OPERABLE control rods that are "slow" occupy adjacent locations, be in at least HOT SH UTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.1.3.3 During single control rod scram time surveillances with the control rod drive pumps isolated from the accumulators:

a.

Verify each control rod scram time is within the limits of Table 3.1.3.3-1 with reactor steam dome pressure;::: 800 psig prior to THERMAL POWER exceeding 40% RATED THERMAL POWER after each reactor shutdown;::: 120 days.

b.

Verify for a representative sample, each tested control rod scram time is within the limits of Table 3.1.3.3-1 with reactor steam dome pressure ~ 800 psig in accordance with the Surveillance Frequency Control Program.

c.

Verify each affected control rod scram time is within the limits of Table 3.1.3.3-1 with any reactor steam dome pressure prior to declaring control rod OPERABLE after work on control rod or CRD System that could affect scram time.

d.

Verify each affected control rod scram time is within the limits of Table 3.1.3.3-1 with reactor steam dome pressure;::: 800 psig prior to THERMAL POWER exceeding 40%

RATED THERMAL POWER after fuel movement within the affected core cell AND prior to exceeding 40% RTP after work on control rod or CRD System that could affect scram time.

HOPE CREEK 3/41-7 Amendment No. 187

REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

ACTION (Continued)

3.

With one or more control rod scram accumulators inoperable and reactor pressure < 900 psig, a)

Immediately upon discovery of charging water header pressure <

940 pSig, verify all control rods associated with inoperable accumulators are fully inserted otherwise place the mode switch in the shutdown position**, and b)

Within one hour insert the associated control rod(s), declare the associated control rod(s) inoperable and disarm the associated control valves either electrically or hydraulically by closing the drive water and exhaust water isolation valves.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

In OPERATIONAL CONDITION 5*:

1.

With one or more withdrawn control rods inoperable, upon discovery immediately initiate action to fully insert inoperable withdrawn control rods.

SURVEILLANCE REQUI REMENT§ 4.1.3.5 Each control rod scram accumulator shall be determined OPERABLE:

a.

In accordance with the Surveillance Frequency Control Program by verifying that the indicated pressure is greater than or equal to 940 psig unless the control rod is inserted and disarmed or scrammed.

At least the accumulator associated with each withdrawn control rod. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

Not applicable if all inoperable control rod scram accumulators are associated with fully inserted control rods.

HOPE CREEK 3/4 1-10 Amendment No. 187

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.3.7 The control rod position indication system shall be determined OPERABLE by verifying:

a.

In accordance with the Surveillance Frequency Control Program that the position of each control rod is indicated,

b.

That the indicated control rod position changes during the movement of the control rod drive when performing Surveillance Requirement 4.1.3.1.2, and

c.

That the control rod position indicator corresponds to the control rod position indicated by the "Full Out" position indicator when performing Surveillance Requirement 4.1.3.6.b.

HOPE CREEK 3/41-14 Amendment No. 187

REACTIVITY CONTROL SYSTEMS 3/4.1 5 STANDBY LIQUID CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.1.5 The standby liquid control system consists of two redundant subsystems and shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1, and 2 ACTION:

a.

In OPERATIONAL CONDITION 1 or 2:

1.

With one system subsystem inoperable, restore the subsystem to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2.

With both system subsystems inoperable, restore at least one subsystem to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.1.5 The standby liquid control system shall be demonstrated OPERABLE:

a.

In accordance with the Surveillance Frequency Control Program by verifying that;

1.

The temperature of the sodium penta borate solution in the storage tank is greater than or equal to 70°F.

2.

The available volume of sodium penta borate solution is within the limits of Figure 3.1.5-1.

3.

The heat tracing circuit is OPERABLE by determining the temperature of the pump suction piping to be greater than or equal to 70°F.

HOPE CREEK 3/4 1-19 Amendment No. 187

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b.

In accordance with the Surveillance Frequency Control Program by:

1.

Verifying the continuity of the explosive charge.

2.

Determining that the available weight of sodium pentaborate is greater than or equal to 5,776 Ibs and the concentration of boron in solution is within the limits of Figure 3.1.5-1 by chemical analysis.*

3.

Verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

c.

Demonstrating that, when tested pursuant to the 1ST Program, the minimum flow requirement of 41.2 gpm, per pump, at a pressure of greater than or equal to 1255 psig is met.

d.

In accordance with the Surveillance Frequency Control Program by:

1.

Initiating one of the standby liquid control system subsystem, including an explosive valve, and verifying that a flow path from the pumps to the reactor pressure vessel is available by pumping demineralized water into the reactor vessel and verifying that the relief valve does not actuate.

The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch which has been certified by having one of that batch successfully fired. Both injection subsystems shall be tested in accordance with the Surveillance Frequency Control Program.

2.

- Demonstrating that all heat traced piping between the storage tank and the injection pumps is unblocked and then draining and flushing the piping with demineralized water.

3.

Demonstrating that the storage tank heaters are OPERABLE by verifying the expected temperature rise of the sodium pentaborate solution in the storage tank after the heaters are energized.

This test shall also be performed anytime water or boron is added to the solution or when the solution temperature drops below 70°F.

This test shall also be performed whenever both heat tracing circuits have been found to be inoperable and may be performed by any series of sequential, overlapping or total flow path steps such that the entire flow path is included.

HOPE CREEK 3/41-20 Amendment No. 187

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITIONEOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) shall be less than or equal to the limits specified in the CORE OPERATING LIMITS REPORT.

APPLICABI L1TY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 24% of RATED THERMAL POWER.

ACTION:

With an APLHGR exceeding the limits specified in the CORE OPERATING LIMITS REPORT, initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 24% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE;QUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the limits specified in the CORE OPERATING LIMITS REPORT:

a.

Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER is greater than or equal to 24% of RATED THERMAL POWER and in accordance with the Surveillance Frequency Control Program.

b.

Initially and in accordance with the Surveillance Frequency Control Program when the reactor is operating with a LIMITING CONTROL ROD PATTERN for APLHGR.

HOPE CREEK 3/42-1 Amendment No. 187

POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITIONFOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater than the MCPR limit specified in the CORE OPERATING LIMITS REPORT.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 24% of RATED THERMAL POWER.

ACTION:

a.

With the end-of-cycle recirculation pump trip system inoperable per Specification 3.3.4.2, operation may continue provided that, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , MCPR is determined to be greater than or equal to the EOC-RPT inoperable limit specified in the CORE OPERATING LIMITS REPORT.

b.

With MCPR less than the applicable MCPR limit specified in the CORE OPERATING LIMITS REPORT, initiate corrective action within 15 minutes and restore MCPR to within the required limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 24% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEilLANCE REQlJlREMENTS 4.2.3 MCPR, shall be determined to be equal to or greater than the applicable MCPR limit specified in the CORE OPERATING LIMITS REPORT:

a.

Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER is greater than or equal to 24% of RATED THERMAL POWER and in accordance with the Surveillance Frequency Control Program thereafter.

b.

Initially and in accordance with the Surveillance Frequency Control Program when the reactor is operating with a LIMITING CONTROL ROD PATTERN for MCPR.

Amendment No. 187 HOPE CREEK

POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall not exceed the limit specified in the CORE OPERATING LIMITS REPORT.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 24% of RATED THERMAL POWER.

ACTION:

With the LHGR of any fuel rod exceeding the limit specified in the CORE OPERATING LIMITS REPORT, initiate corrective action within 15 minutes and restore the LHGR to within the limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 24% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.2.4 LHGR's shall be determined to be equal to or less than the limit specified in the CORE OPERATING LIMITS REPORT:

a.

Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER is greater than or equal to 24% of RATED THERMAL POWER and in accordance with the Surveillance Frequency Control Program thereafter.

b.

Initially and in accordance with the Surveillance Frequency Control Program when the reactor is operating on a LIMITING CONTROL ROD PATTERN for LHGR.

HOPE CREEK 3/4 2-5 Amendment No. 187

3/4.3 INSTRUMENTATION 3/44.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protection system instrumentation channels shown in Table 3.3.1-1 shall be OPERABLE.

APPLICABILITY:

As shown in Table 3.3.1-1.

ACTION:

a.

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip system, place the inoperable channel(s) and/or that trip system in the tripped condition* within twelve hours.

b.

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for both trip systems, place at least one trip system** in the tripped condition within one hour and take the ACTION required by Table 3.3.1-1.

SURVEILLANCE REQUIREMENTS 4.3.1.1 Each reactor protection system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.1-1.

4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of aU channels shall be performed in accordance with the Surveillance Frequency Control Program.

4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip functional unit shall be demonstrated to be within its limit in accordance with the Surveillance Frequency Control Program. Neutron detectors are exempt from response time testing. For the Reactor Vessel Steam Dome Pressure - High Functional Unit and the Reactor Vessel Water Level - Low, Level 3 Functional Unit, the sensor is eliminated from response time testing for RPS circuits.

4.3.1.4 The provisions of Specification 4.0.4 are not applicable for entry into OPERATIONAL CONDITION 2 or 3 from OPERATIONAL CONDITION 1 for the Inter-mediate Range Monitors.

An inoperable channel need not be placed in the tripped condition where this would cause the Trip Function to occur. In these cases, the inoperable channel shall be restored to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or the ACTION required by Table 3.3.1-1 for that Trip Function shall be taken.

If more channels are inoperable in one trip system than in the other, place the trip system with more inoperable channels in the tripped condition, except when this would cause the Trip Function to occur.

HOPE CREEK 3/43-1 Amendment No. 187

TABLE 4.3.1.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS FUNCTIONAL UNIT

1.

Intermediate Range Monitors:

a.

Neutron Flux - High

b.

Inoperative

2.

Average Power Range Monitor{f):

a.

Neutron Flux - Upscale, Setdown

b.

Flow Biased Simulated Thermal Power-Upscale

c.

Fixed Neutron Flux - Upscale

d.

Inoperative

3.

Reactor Vessel Steam Dome Pressure - High

4.

Reactor Vessel Water Level - Low, Level 3

5.

Main Steam Line Isolation Valve Closure

6.

This item intentionally blank

7.

Drywell Pressure - High HOPE CREEK CHANNEL CHANNEL FUNCTIONAL CHANNEL CHECK (m)

TEST (m)

CALI BRA TION (a) (m)

(b)

NA NA (b)

(I)

(9)

(d) (e) (h)

(d)

NA NA (k)

(k)

NA (k) 3/4 3-7 OPERATIONAL CONDITIONS FOR WHICH SURVEILLANCE REQUIRED 2

3,4,5 2,3,4,5 2

3,4,5 1

1 1,2,3,4,5 1,2 1,2 1

1,2 Amendment No. 187

TABLE 4.3.1.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH FUNCTIONAL UNIT CHECK (m)

TEST(m)

CALI BRA TION (m)

SURVEILLANCE REQUIRED

8.

Scram Discharge Volume Water Level

- High:

a.

Float Switch NA 1,2,50l

b.

Level TransmitterfTrip Unit (k) 1,2,5m

9.

Turbine Stop Valve - Closure NA 1

10. Turbine Control Valve Fast Closure Valve Trip System Oil Pressure - Low NA 1

11. Reactor Mode Switch Shutdown Position NA NA 1,2,3,4,5

12. Manual Scram NA NA 1,2,3,4,5 (a)

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(b)

The IRM and SRM channels shall be determined to overlap for at least 1/2 decades during each startup after entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be determined to overlap for at least 1/2 decades during each controlled shutdown, if not performed within the previous 7 days.

(c)

DELETED This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER 2: 24% of RATED THERMAL POWER. Adjust the APRM channel if the absolute difference is greater than 2% of RATED THERMAL POWER.

(e)

This calibration shall consist of the adjustment of the APRM fiow biased channel to conform to a calibrated flow signal.

(f)

The LPRMs shall be calibrated in accordance with the Surveillance Frequency Control Program.

(g)

Verify measured core flow (total core fiow) to be greater than or equal to established core flow at the existing recirculation loop flow (APRM % flow).

This calibration shall consist of verifying the 6 +/- 0.6 second simulated thermal power time constant.

(i)

This item intentionally blank With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

(k)

Verify the tripset point of the trip unit in accordance with the Surveillance Frequency Control Program.

(i)

Not required to be performed when entering OPERATIONAL CONDITION 2 from OPERATIONAL CONDITION 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering OPERATIONAL CONDiTION 2.

(m)

Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

HOPE CREEK 3/4 3-8 Amendment No. 187

INSTRUMENTATION 4.3.2.1 Each isolation actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.2.1-1.

4.3.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed in accordance with the Surveillance Frequency Control Program.

4.3.2.3 The ISOLATION SYSTEM RESPONSE TIME of each isolation trip function shall be demonstrated to be within its limit in accordance with the Surveillance Frequency Control Program. Radiation detectors are exempt from response time testing. The sensor is eliminated from response time testing for MSIV isolation logic circuits of the following trip functions:

Reactor Vessel Water Level - Low Low Low, Level 1; Main Steam Line Pressure - Low; Main Steam Line Flow - High.

HOPE CREEK 3/43-10 Amendment No. 187

TABLE 4.3.2.1-1 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL FUNCTIONAL TRIP FUNCTION CHECK (e)

TEST (el

1.

PRIMARY CONTAINMENT ISOLATION

a.

Reactor Vessel Water Level

1) Low Low, Level 2

2) Low Low Low, Level 1

b.

Drywell Pressure - High

c.

Reactor Building Exhaust Radiation High

d.

Manual Initiation NA (al

2.

SECONDARY CONTAINMENT ISOLATION

a.

Reactor Vessel Water Level Low Low, Level 2

b.

Drywell Pressure - High

c.

Refueling Floor Exhaust Radiation - High

d.

Reactor Building Exhaust Radiation High

e.

Manual Initiation NA (a)

3.

MAIN STEAM LINE ISOLATION

a.

Reactor Vessel Water Level Low Low Low, Level 1

b.

Main Steam Line Radiation - High, High

c.

Main Steam Line Pressure - Low

d.

Main Steam Line Flow - High

e.

Condenser Vacuum - Low

f.

Main Steam Line Tunnel Temperature-High NA

g.

Manual Initiation NA (a)

CHANNEL CALIBRATION (e)

OPERATIONAL CONDITIONS FOR WHICH SURVEILLANCE REQUIRED 1,2,3 1,2,3 1,2,3 NA 1,2,3 1,2,3 1,2,3 and

1,2,3 1,2,3 and

NA 1,2,3 and

1,2,3 and

1,2,3 1,2,3 1

1,2,3 1, 2**, 3**

NA 1,2,3 1,2,3 HOPE CREEK 3/43-28 Amendment No. 187

TABLE 4.3.2.1-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP FUNCTION CHANNEL CHECK (e)

CHANNEL FUNCTIONAL TEST (e)

CHANNEL CALIBRATION (e)

4.

REACTOR WATER CLEANUP SYSTEM ISOLATION

a.

RWCU Ll Flow - High

b.

RWCU Ll Flow - High, Timer

c.

RWCU Area Temperature - High

d.

RWCU Area Ventilation Ll Temperature - High

e.

SLCS Initiation

f.

Reactor Vessel Water Level - Low Low, Level 2 NA NA NA NA (0)

NA

g.

Manual Initiation NA (a)

NA

5.

REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION

a.

b.

RCIC Steam Line Ll Pressure (Flow) - High RCIC Steam Line Ll Pressure (Flow) - High, Timer NA NA

c.

d.

RCIC Steam Supply Pressure - Low RCIC Turbine Exhaust Diaphragm Pressure - High NA NA

e.

f RCIC Pump Room Temperature - High RCIC Pump Room Ventilation Ducts Ll Temperature - High NA NA

g.

h.

RCIC Pipe Routing Area Temperature - High RCIC Torus Compartment Temperature -High NA NA

i.

j.

Drywell Pressure - High Manual Initiation NA NA OPERATIONAL CONDITIONS FOR WHICH SURVEILLANCE REQUIRED 1,2,3 1,2,3 1,2,3 1,2,3 1,2 1,2,3 1,2,3 1, 2, 3 1,2,3 1,2,3 1, 2, 3 1,2,3 1,2,3 1, 2, 3 1,2,3 1,2,3 1,2,3 HOPE CREEK 3/43-29 Amendment No. 187

TABLE 4.3.2.1-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP FUNCTION

6.

HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION

a.

HPCI Steam Line!':!. Pressure (Flow) - High

b.

HPCI Steam Line!':!. Pressure (Flow) - High, Timer

c.

HPCI Steam Supply Pressure - Low

d.

HPCI Turbine Exhaust Diaphragm Pressure-High

e.

HPCI Pump Room Temperature - High f

HPCI Pump Room Ventilation Ducts !':!.

Temperature - High

g.

HPCI Pipe Routing Area Temperature - High

h.

HPCI Torus Compartment Temperature -High

i.

Drywell Pressure - High

j.

Manual Initiation

7.

RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION

a.

Reactor Vessel Water Level - Low, Level 3

b.

Reactor Vessel (RHR Cut-in Permissive)

Pressure - High

c.

Manual Initiation HOPE CREEK CHANNEL CHECK (el CHANNEL FUNCTIONAL TEST (e)

CHANNEL CALIBRATION (e)

OPERATIONAL CONDITIONS FOR WHICH SURVEILLANCE REQUIRED NA NA NA NA NA NA NA NA NA NA NA 1,2,3 1,2,3 1,2,3 1,2,3 1,2,3 1,2,3 1,2,3 1,2,3 1,2,3 1,2,3 NA NA (a)

NA 1,2,3 1,2,3 1,2,3 3/43-30 Amendment No. 187

TABLE 4.3.2.1-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS (a)

(b)

(c)

When handling recently irradiated fuel in the secondary containment and during operations with a potential for draining the reactor vessel.

When any turbine stop valve is greater than 90% open and/or when the key-locked bypass switch is in the Norm position.

Manual initiation switches shall be tested in accordance with the Surveillance Frequency Control Program. All other circuitry associated with manual initiation shall receive a CHANNEL FUNCTIONAL TEST in accordance with the Surveillance Frequency Control Program as part of circuitry required to be tested for automatic system isolation.

Each train or logic channel shall be tested in accordance with the Surveillance Frequency Control Program.

Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

HOPE CREEK 3/4 3-31 Amendment No. 187

INSTRUMENTATION 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3 The emergency core cooling system (ECCS) actuation instrumentation channels shown in Table 3.3.3-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.3-2.

APPLICABILITY:

As shown in Table 3.3.3-1.

ACTION:

a.

With an ECCS actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.3-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

b.

With one or more ECCS actuation instrumentation channels inoperable, take the ACTION required by Table 3.3.3-1.

SURVEILLANCE REQUIREMENTS 4.3.3.1 Each ECCS actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.3.1-1.

4.3.3.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed in accordance with the Surveillance Frequency Control Program.

4.3.3.3 The ECCS RESPONSE TIME of each ECCS trip function shall be demonstrated to be within the limit in accordance with the Surveillance Frequency Control Program. ECCS actuation instrumentation is eliminated from response time testing.

HOPE CREEK 3/43-32 Amendment No. 187

TABLE 4.3.3.1-1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP FUNCTION CHANNEL CHECK (a)

1.

CORE SPRAY SYSTEM

a.

Reactor Vessel Water Level - Low Low Low, Level 1

b.

c.

d.

Drywell Pressure - High Reactor Vessel Pressure - Low Core Spray Pump Discharge Flow - Low (Bypass)

e.

f.

g.

Core Spray Pump Start Time Delay - Normal Power Core Spray Pump Start Time Delay - Emergency Power Manual Initiation NA NA NA

2.

LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM

a.

Reactor Vessel Water Level - Low Low Low, Level 1

b.

c.

d.

e.

f.

Drywell Pressure - High Reactor Vessel Pressure - Low (Permissive)

LPCI Pump Discharge Flow - Low (Bypass)

LPCI Pump Start Time Delay - Normal Power Manual Initiation NA NA

3.

HIGH PRESSURE COOLANT INJECTION SYSTEM#

a.

Reactor Vessel Water Level - Low Low, Level 2

b.

c.

d.

e.

f.

g.

Drywell Pressure - High Condensate Storage Tank Level-Low Suppression Pool Water Level - High Reactor Vessel Water Level - High, Level 8 HPCI Pump Discharge Flow - Low (Bypass)

Manual Initiation NA CHANNEL OPERATIONAL CONDITIONS FUNCTIONAL CHANNEL FOR WHICH SURVEILLANCE (a)

CALIBRATION (a)

REQUIRED NA NA 1,2,3,4*,5*

1,2,3 1,2,3,4*,5*

1,2,3,4*,5*

1,2,3 1,2,3,4*,5*

1, 2, 3, 4*, 5*

1,2,3,4*,5*

NA 1,2,3 1,2,3 1,2,3 1,2,3 1,2,3 1,2,3 1,2,3 HOPE CREEK 3/43-39 Amendment No. 187

TABLE 4.3.3.1-1 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CONDITIONS CHANNEL FUNCTIONAL CHANNEL FOR WHICH SURVEILLANCE TRIP FUNCTION CHECK (a)

TEST (a)

CALIEIRATION (al REQUIRED

4.

AUTOMATIC DEPRESSURIZATION SYSTEM##

a.

Reactor Vessel Water Level - Low Low Low, Level 1 1,2,3

b.

Drywell Pressure - High 1,2,3

c.

ADS Timer NA 1,2,3

d.

Core Spray Pump Discharge Pressure - High 1,2,3

e.

RHR LPCI Mode Pump Discharge Pressure 1,2,3 High

f.

Reactor Vessel Water Level - Low, Level 3 1,2,3

g.

ADS Drywell Pressure Bypass Timer NA 1,2,3

h.

ADS Manual Inhibit Switch NA NA 1,2,3

i.

Manual initiation NA NA 1,2,3

5.

LOSS OF POWER

a.

4.16 kv Emergency Bus Under-voltage (Loss NA NA of Voltage) 1, 2, 3, 4**, 5**

b.

4.16 kv Emergency Bus Under-voltage (Degraded Voltage) 1, 2, 3, 4**, 5**

(a)

Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

When the system is required to be OPERABLE per Specification 3.5.2.

Required OPERABLE when ESF equipment is required to be OPERABLE.

Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 200 psig.

Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 100 psig.

HOPE CREEK 3/43-40 Amendment No. 187

3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.4.1 The anticipated transient without scram recirculation pump trip (ATWS-RPT) system instrumentation channels shown in Table 3.3.4.1-1 shall be OPERABLE with their trip setpoints set consistent with values shown in the Trip Setpoint column of Table 3.3.4.1-2.

APPLICABILITY:

OPERATIONAL CONDITION 1.

ACTION:

a.

With an A TWS recirculation pump trip system instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.4.1-2, declare the channel inoperable until the channel is restored to OPERABLE status with the channel trip setpoint adjusted consistent with the Trip Setpoint value.

b.

With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels per Trip System requirement for one or both trip systems, place the inoperable channel(s) in the tripped condition within one hour.

c.

With the number of OPERABLE channels two or more less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip system, and:

1.

If the inoperable channels consist of one reactor vessel water level channel and one reactor vessel pressure channel, place both inoperable channels in the tripped condition within one hour, or if this action will initiate a pump trip, declare the trip system inoperable.

2.

If the inoperable channels include two reactor vessel water level channels or two reactor vessel pressure channels, declare the trip system inoperable.

d.

With one trip system inoperable, restore the inoperable trip system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least STARTUP within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

e.

With both trip systems inoperable, restore at least one trip system to OPERABLE status within one hour or be in at least STARTUP within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.3.4.1.1. Each ATWS recirculation pump trip system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies in accordance with the Surveillance Frequency Control Program.

4.3.4.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed in accordance with the Surveillance Frequency Control Program.

HOPE CREEK 3/43-41 Amendment No. 187

TABLE 4.3.4.1-1 INFORMATION ON THIS PAGE HAS BEEN DELETED HOPE CREEK 3/43-44 Amendment No. 187

INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.4.2.1 Each end-of-cycle recirculation pump trip system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies in accordance with the Surveillance Frequency Control Program.

4.3.4.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed in accordance with the Surveillance Frequency Control Program.

4.3.4.2.3 The END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME of each trip function shown in Table 3.3.4.2-3 shall be demonstrated to be within its limit in accordance with the Surveillance Frequency Control Program. Each test shall include at least the logic of one type of channel input, turbine control valve fast closure or turbine stop valve closure, such that both types of channel inputs are tested in accordance with the Surveillance Frequency Control Program.

4.3.4.2.4 The time interval necessary for breaker arc suppression from energization of the recirculation pump circuit breaker trip coil shall be measured in accordance with the Surveillance Frequency Control Program.

HOPE CREEK 3/43-46 Amendment No. 187

TABLE 4.3.4.2.1-1 INFORMATION ON THIS PAGE HAS BEEN DELETED HOPE CREEK 3/43-50 Amendment No. 187

INSTRUMENTATION 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.5 The reactor core isolation cooling (RCIC) system actuation instrumentation channels shown in Table 3.3.5-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.5-2.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2 and 3 with reactor steam dome pressure greater than 150 psig.

ACTION:

a.

With a RCIC system actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.5-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

b.

With one or more RCIC system actuation instrumentation channels inoperable, take the ACTION required by Table 3.3.5-1.

SURVEILLANCE REQUIREMENTS 4.3.5.1 Each RCIC system actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.5.1-1.

4.3.5.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed in accordance with the Surveillance Frequency Control Program.

HOPE CREEK 3/43-51 Amendment No. 187

TABLE 4.3.5.1-1 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL CHANNEL FUNCTIONAL UNITS CHECK (b)

FUNCTIONAL TEST (b)

CALIBRATION (b)

a.

Reactor Vessel Water Level - Low Low, Level 2

b.

Reactor Vessel Water Level - High, Level 8

c.

Condensate Storage Tank Level-Low NA

d.

Manual Initiation NA (a)

NA (a)

Manual initiation switches shall be tested in accordance with the Surveillance Frequency Control Program. All other circuitry associated with manual initiation shall receive a CHANNEL FUNCTIONAL TEST in accordance with the Surveillance Frequency Control Program as part of circuitry required to be tested for automatic system actuation.

(b)

Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

HOPE CREEK 3/43-55 Amendment No. 187

TABLE 4.3.6-1 CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK (I)

TEST (I)

CALlBRATION(a) (I)

SURVEILLANCE REQUIRED

1.

ROD BLOCK MONITOR

a.

Upscale NA (e) 1*

b.

Inoperative NA (e)

NA 1*

(e)

c.

Downscale NA 1*

2.

APRM

a.

Flow Biased Neutron Flux - Upscale NA 1

b.

Inoperative NA NA 1,2,5

c.

Downscale NA 1

d.

Neutron Flux - Upscale, Startup NA 2,5

3.

SOURCE RANGE MONITORS

a.

Detector not full in NA NA 2, 5

b.

Upscale NA 2, 5

c.

Inoperative NA NA 2, 5

d.

Downscale NA 2,5

4.

INTERMEDIATE RANGE MONITORS

a.

Detector not full in NA NA 2,5

b.

Upscale NA 2, 5

c.

Inoperative NA NA 2,5

d.

Downscale NA 2,5

5.

SCRAM DISCHARGE VOLUME

a.

Water Level-High (Float Switch)

NA 1,2, 5**

6.

REACTOR COOLANT SYSTEM RECIRCULATION FLOW

a.

Upscale NA 1

b.

Inoperative NA NA 1

c.

Comparator NA 1

7.

REACTOR MODE SWITCH SHUTDOWN POSITION NA (e)

NA 3,4 HOPE CREEK 3/43-60 Amendment No. 187

TABLE 4.3.6-1 (Continued)

CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS NOTES:

a.

Neutron detectors may be excluded from CHANNEL CALIBRATION.

b.

DELETED

c.

Includes reactor manual control multiplexing system input.

d.

DELETED

e.

Not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after reactor mode switch is in the shutdown position.

f.

Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

With THERMAL POWER ~ 30% of RATED THERMAL POWER.

With more than one control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

HOPE CREEK 3/43-61 Amendment No. 187

INSTRUMENTATION 3/4.3.7 MONITORING INSTRUMENTATION RADIATION MONITORING INSTRUMENTATION 3.3.7.1 The radiation monitoring instrumentation channels shown in Table 3.3.7.1-1 shall be OPERABLE with their alarm/trip setpoints within the specified limits.

APPLICABILITY:

As shown in Table 3.3.7.1-1.

ACTION:

a.

With a radiation monitoring instrumentation channel alarm/trip setpoint exceeding the value shown in Table 3.3.7.1-1, adjust the setpoint to within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel inoperable.

b.

With one or more radiation monitoring channels inoperable, take the ACTION required by Table 3.3.7.1-1.

c.

The provisions of Specifications 3.0.3 are not applicable.

4.3.7.1 Each of the above required radiation monitoring instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the conditions and at the frequencies in accordance with the Surveillance Frequency Control Program.

HOPE CREEK 3/43-62 Amendment No. 187

TABLE 4.3.7.1-1 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS INSTRUMENTATION CHANNEL CHECK (a)

CHANNEL FUNCTIONAL TEST (a)

CHANNEL CALIBRATION (a)

OPERATIONAL CONDITIONS FOR WHICH SURVEILLANCE REQUIRED

1.

Control Room Ventilation Radiation Monitor 1,2,3, and *

2.

Area Monitors

a. Criticality Monitors

1) New Fuel Storage Vault

2) Spent Fuel Storage Pool

b. Control Room Direct Radiation Monitor At all times

3.

Reactor Auxiliaries Cooling Radiation Monitor At all times

4.

Safety Auxiliaries Cooling Radiation Monitor At all times

5.

Offgas Pre-treatment Radiation Monitor HOPE CREEK 3/43-66 Amendment No 187.

TABLE 4.3.7.1-1 (Continued)

RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE NOTATION (a) Frequencies are specified in the Surveillance Frequency Control Program unless othelWise noted in the table.

With fuel in the new fuel storage vault.

1. With fuel in the spent fuel storage pool.

When recently irradiated fuel is being handled in the secondary containment and during operations with the potential for draining the reactor vessel.

- When the offgas treatment system is operating.

HOPE CREEK 3/43-67 Amendment No. 187

INSTRUMENTATION REMOTE SHUTDOWN SYSTEM INSTRUMENTATION AND CONTROLS LIMITING CONDITION FOR OPERATION 3.3.7.4 The remote shutdown system instrumentation and controls shown in Table 3.3.7.4-1 and Table 3.3.7.4-2 shall be OPERABLE.

APPLICABILITY OPERATIONAL CONDITIONS 1 and 2.

ACTION:

a.

With the number of OPERABLE remote shutdown monitoring instrumentation channels less than required by Table 3.3.7.4-1, restore the inoperable channel(s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

With the number of OPERABLE remote shutdown system controls less than required in Table 3.3.7.4-2, restore the inoperable control(s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.3.7.4.1 Each of the above required remote shutdown monitoring instrumentation channels shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies specified in the Surveillance Frequency Control Program unless otherwise noted by Table 4.3.7.4-1.

4.3.7.4.2 At least one of each of the above remote shutdown control switch(es) and control circuits shall be demonstrated OPERABLE by verifying its capability to perform its intended function(s) in accordance with the Surveillance Frequency Control Program.

HOPE CREEK 3/43-74 Amendment No. 187

11 TABLE 4.3.7.4-1 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS INSTRUMENT CHANNEL CHANNEL CHECK (a)

CALIBRATION (a)

1.

Reactor Vessel Pressure

2.

Reactor Vessel Water Level

3.

Safety/Relief Valve Position (Energization)

NA

4.

Suppression Chamber Water Level

5.

Suppression Chamber Water Temperature

6.

RHR System Flow

7.

Safety Auxiliaries Cooling System Flow

8.

Safety Auxiliaries Cooling System Temperature

9.

RCIC System Flow

10.

RCIC Turbine Speed RCIC Turbine Bearing Oil Pressure Low Indication

12.

RCIC High Pressure/Low Pressure Turbine Bearing Temperature High Indication HOPE CREEK 3/43-82 Amendment No. 187

TABLE 4.3.7.4-1 (Continued)

REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS INSTRUMENT CHANNEL CHANNEL CHECK (a)

CALIBRATION (a)

13.

Condensate Storage Tank Level Low-Low Indication

14.

Standby Diesel Generator 1AG400 Breaker Indication NA

15.

Standby Diesel Generator 1 BG400 Breaker Indication NA

16.

Standby Diesel Generator 1 CG400 Breaker Indication NA

17.

Standby Diesel Generator 1DG400 Breaker Indication NA

18.

SWitchgear Room Cooler 1AVH401 Status Indication NA

19.

Switchgear Room Cooler 1 BVH401 Status Indication NA

20.

Switchgear Room Cooler 1CVH401 Status Indication NA

21.

Switchgear Room Cooler 1DVH401 Status Indication NA (a)

Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

HOPE CREEK 3/43-83 Amendment No. 187

TABLE 4.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS APPLICABLE CHANNEL CHANNEL OPERATIONAL INSTRUMENT CHECK (a)

CALIBRATION (a)

CONDITIONS

1.

Reactor Vessel Pressure

2.

Reactor Vessel Water Level

3.

Suppression Chamber Water Level

4.

Suppression Chamber Water Temperature

5.

Suppression Chamber Pressure

6.

Drywell Pressure

7.

Drywell Air Temperature

8.

Deleted

9.

Safety/Relief Valve Position Indicators

10.

Drywell Atmosphere Post-Accident Radiation Monitor

11.

North Plant Vent Radiation Monitor'l

12.

South Plant Vent Radiation Monitor'l

13.

FRVS Vent Radiation Monitor#

14.

Primary Containment Isolation Valve Position Indication 1,2,3 1,2,3 1,2,3 1,2,3 1,2,3 1,2,3 1,2,3 1,2,3 1,2,3 1,2,3 1,2,3 1,2,3 1,2,3 (a)

Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

CHANNEL CALIBRATION shall consist of an electronic calibration of the channel, not including the detector, for range decades above 10 Rlhr and a one point calibration check of the detector below 10 R/hr with an installed or portable gamma source.

High range noble gas monitors.

HOPE CREEK 3/43-87 Amendment No. 187

INSTRUMENTATION SOURCE RANGE MONITORS LIMITING CONDITION FOR OPERATION 3.3.7.6 At least the following source range monitor channels shall be OPERABLE:

a.

In OPERATIONAL CONDITION 2*, three.

b.

In OPERATIONAL CONDITION 3 and 4, two.

APPLICABI LlTY:

OPERATIONAL CONDITIONS 2*,3 and 4.

ACTION:

a.

In OPERATIONAL COl\\IDITION 2* with one of the above required source range monitor channels inoperable, restore at least 3 source range monitor channels to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

In OPERATIONAL CONDITION 3 or 4 with one or more of the above required source range monitor channels inoperable, verify all insertable control rods to be inserted in the core and lock the reactor mode switch in the Shutdown position within one hour.

SUB\\IEILLANCE REQUIREMENTS 4.3.7.6 Each of the above required source range monitor channels shall be demonstrated OPERABLE by:

a.

Performance of a:

1.

CHANNEL CHECK:

a)

In accordance with the Surveillance Frequency Control Program in CONDITION 2*, and b)

In accordance with the Surveillance Frequency Control Program in CONDITION 3 or 4.

2.

CHANNEL CALIBRATION** in accordance with the Surveillance Frequency Control Program.

b.

Performance of a CHANNEL FUNCTIONAL TEST in accordance with the Surveillance Frequency Control Program.

c.

Verifying, prior to withdrawal of control rods, that the SRM count rate is at least 3 cps with the detector fully inserted.

d.

The provisions of Specification 4.0.4 are not applicable for entry into OPERATIONAL CONDITION 2* or 3 from OPERATIONAL CONDITION 1.

With IRM's on range 2 or below.

Neutron detectors may be excluded from CHANNEL CALIBRATION.

HOPE CREEK 3/43-88 Amendment No. 187

INSTRUMENTATION 3/4.3.9 FEEDWATER/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.9 The feedwater/main turbine trip system actuation instrumentation channels shown in Table 3.3.9-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.9-2.

APPLICABILITY:

As shown in Table 3.3.9-1.

ACTION:

a.

With a feedwater/main turbine trip system actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.9-2, declare the channel inoperable and either place the inoperable channel in the tripped condition until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value, or declare the associated system inoperable.

b.

With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels requirement, restore the inoperable channel to OPERABLE status within 7 days or be in at least STARTUP within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c.

With the number of OPERABLE channels two less than required by the Minimum OPERABLE Channels requirement, restore at least one of the inoperable channels to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least STARTUP within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.3.9.1 Each feedwater/main turbine trip system actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies specified in the Surveillance Frequency Control Program.

4.3.9.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed in accordance with the Surveillance Frequency Control Program.

HOPE CREEK 3/43-105 Amendment No. 187

TABLE 4.3.9.1-1 INFORMATION ON THIS PAGE HAS BEEN DELETED HOPE CREEK 3/43-108 Amendment No. 187

INSTRUMENTATION 3/4.3.10 MECHANICAL VACUUM PUMP TRIP INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.10 Two channels of the Main Steam Line Radiation - High, High function for the mechanical vacuum pump trip shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2 with mechanical vacuum pump in service and any main steam line not isolated.

ACTION:

a.

With one channel of the Main Steam Line Radiation - High, High function for the mechanical vacuum pump trip inoperable, restore the channel to OPERABLE status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Otherwise, trip the mechanical vacuum pumps, or isolate the main steam lines or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

With mechanical vacuum pump trip capability not maintained:

1.

Trip the mechanical vacuum pumps within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; or

2.

Isolate the main steam lines within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; or

3.

Be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c.

When a channel is placed in an inoperable status solely for the performance of required Surveillances, entry into the associated ACTIONS may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the mechanical vacuum pump trip capability is maintained.

SURVEILLANCE REQUIREMENTS 4.3.10 Each channel of the Main Steam Line Radiation - High, High function for the mechanical vacuum pump trip shall be demonstrated OPERABLE by:

a.

Performance of a CHANNEL CHECK in accordance with the Surveillance Frequency Control Program;

b.

Performance of a CHANNEL FUNCTIONAL TEST in accordance with the Surveillance Frequency Control Program;

c.

Performance of a CHANNEL CALIBRATION in accordance with the Surveillance Frequency Control Program. The Allowable Value shall be s 3.6 x normal background; and

d.

Performance of a LOGIC SYSTEM FUNCTIONAL TEST, including mechanical vacuum pump trip breaker actuation, in accordance with the Surveillance Frequency Control Program.

HOPE CREEK 3/43-109 Amendment No. 187

3/4.3 INSTRUMENTATION 3/4.3.11 OSCILLATION POWER RANGE MONITOR LIMITING CONDIIION FOR OPERATION 3.3.11 Four channels of the OPRM instrumentation shall be OPERABLE*. Each OPRM channel period based algorithm amplitude trip setpoint (Sp) shall be less than or equal to the Allowable Value as specified in the CORE OPERATING LIMITS REPORT.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 24% of RATED THERMAL POWER.

ACTIONS

a.

With one or more required channels inoperable:

1.

Place the inoperable channels in trip within 30 days, or

2.

Place associated RPS trip system in trip within 30 days, or

3.

Initiate an alternate method to detect and suppress thermal hydraulic instability oscillations within 30 days.

b.

With OPRM trip capability not maintained:

1.

Initiate alternate method to detect and suppress thermal hydraulic instability oscillations within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and

2.

Restore OPRM trip capability within 120 days.

c.

Otherwise, reduce THERMAL POWER to less than 24% RTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.3.11.1 Perform CHANNEL FUNCTIONAL TEST in accordance with the Surveillance Frequency Control Program.

4.3.11.2 Calibrate the local power range monitor in accordance with the Surveillance Frequency Control Program in accordance with Note f, Table 4.3.1.1-1 of TS 3/4.3.1.

4.3.11.3 Perform CHANNEL CALIBRATION in accordance with the Surveillance Frequency Control Program. Neutron detectors are excluded.

4.3.11.4 Perform LOGIC SYSTEM FUNCTIONAL TEST in accordance with the Surveillance Frequency Control Program.

4.3.11.5 Verify OPRM is enabled when THERMAL POWER is ;;:: 26.1 % RTP and recirculation drive flow s the value corresponding to 60% of rated core flow in accordance with the Surveillance Frequency Control Program.

4.3.11.6 Verify the RPS RESPONSE TIME is within limits in accordance with the Surveillance Frequency Control Program. Neutron detectors are excluded.

When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated ACTIONS may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , provided the OPRM maintains trip capability.

HOPE CREEK 3/43-110 Amendment No. 187

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.1.1.1 With one reactor coolant system recirculation loop not in operation in accordance with the Surveillance Frequency Control Program verify that:

a.

Reactor THERMAL POWER is ~ 60.86% of RATED THERMAL POWER, and

b.

The recirculation flow control system is in the Local Manual mode, and

c.

The speed of the operating recirculation pump is less than or equal to 90% of rated pump speed.

4.4.1.1.2 With one reactor coolant system recirculation loop not in operation, within no more than 15 minutes prior to either THERMAL POWER increase or recirculation loop flow increase, verify that the following differential temperature requirements are met if THERMAL POWER is

~ 38% of RATED THERMAL POWER or the recirculation loop flow in the operating recirculation loop is ~ 50% of rated loop flow:

a.

~ 145°F between reactor vessel steam space coolant and bottom head drain line coolant, and

b.

~ 50°F between the reactor coolant within the loop not in operation and the coolant in the reactor pressure vessel, and

c.

~ 50°F between the reactor coolant within the loop not in operation and the operating loop.

The differential temperature requirements or Specifications 4.4.1.1.2b and 4.4.1.1.2c do not apply when the loop not in operation is isolated from the reactor pressure vessel.

4.4.1.1.3 DELETED.

HOPE CREEK 3/44-2a Amendment No. 187

REACTOR COOLANT SYSTEM JET PUMPS LIMITING CONDITION FOR OPERATION 3.4.1.2 All jet pumps shall be OPERABLE.

APPLICABI LlTY:

OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With one or more jet pumps inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS*

4.4.1.2 All jet pumps shall be demonstrated OPERABLE as follows:

a. Each of the above required jet pumps shall be demonstrated OPERABLE prior to THERMAL POWER exceeding 24% of RATED THERMAL POWER and in accordance with the Surveillance Frequency Control Program by determining recirculation loop flow, total core flow and diffuser-to-Iower plenum differential pressure for each jet pump and verifying that no two of the following conditions occur when the recirculation pumps are operating in accordance with Specification 3.4.1.3.

1. The indicated recirculation loop flow differs by more than 10% from the established pump speed-loop flow characteristics.

2. The indicated total core flow differs by more than 10% from the established total core flow value derived from recirculation loop flow measurements.

3. The indicated diffuser-to-Iower plenum differential pressure of any individual jet pump differs from the established patterns by more than 20%.

b. During single recirculation loop operation, each of the above required jet pumps in the operating loop shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by verifying that no two of the following conditions occur:

1. The indicated recirculation loop flow in the operating loop differs by more than 10%

from the established* pump speed-loop flow characteristics.

2. The indicated total core flow differs by more than 10% from the established* total core flow value derived from single recirculation loop flow measurements.

3. The indicated diffuser-to-Iower plenum differential pressure of any individual jet pump differs from established* single recirculation loop pattern by more than 20%.

c.

The provisions of Specification 4.0.4 are not applicable provided that this surveillance is performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding 24% of RATED THERMAL POWER.

During startup following any refueling outage, baseline data shall be recorded for the parameters listed to provide a basis for establishing the specified relationships. Comparisons of the actual data in accordance with the criteria listed shall commence upon conclusion of the baseline data analysis. Single loop baseline data shall be recorded the first time the unit enters single loop operation during an operating cycle.

HOPE CREEK 3/44-4 Amendment No. 187

REACTOR COOLANT SYSTEM RECIRCULATION LOOP FLOW LIMITING CONDITION FOR OPERATION 3.4.1.3 Recirculation loop flow mismatch shall be maintained within:

a.

5% of rated core flow with effective core flow** greater than or equal to 70% of rated core flow.

b.

10% of rated core flow with effective core flow** less than 70% of rated core flow.

APPLICABILITY:

OPERATIONAL CONDITIONS 1

and 2* during two recirculation loop operation.

ACTION:

Witli the recirculation loop flows different by more than the specified limits. either:

a.

Restore the recirculation loop flows to within the specified limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . or

b.

Declare the recirculation loop of the pump with the slower flow not in operation and take the ACTION required by Specification 3.4.1.1.

SURVEILLANCE REQUIREMENTS 4.4.1.3 Recirculation loop flow mismatch shall be verified to be within the limits in accordance with the Surveillance Frequency Control Program.

See Special Test Exception 3.10.4.

Effective core flow shall be the core flow that would result if both recirculation loop flows were assumed to be at the smaller value of the two loop flows.

HOPE CREEK 3/44-5 Amendment No. 187

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.2.1 The acoustic monitor for each safety/relief valve shall be demonstrated OPERABLE with the setpoint verified to be :s; 30% of full open noise level by performance of a:

a.

CHANNEL FUNCTIONAL TEST in accordance with the Surveillance Frequency Control Program, and a

b.

CHANNEL CALIBRATION in accordance with the Surveillance Frequency Control Program.

4.4.2.2 At least 1/2 of the safety relief valve pilot stage assemblies shall be removed, set pressure tested and reinstalled or replaced with spares that have been previously set pressure tested and stored in accordance with manufacturer's recommendations in accordance with the Surveillance Frequency Control Program, and they shall be rotated such that all 14 safety relief valve pilot stage assemblies are removed, set pressure tested and reinstalled or replaced with spares that have been previously set pressure tested and stored in accordance with manufacturer's recommendations in accordance with the Surveillance Frequency Control Program. All safety relief valves will be re-certified to meet a +/-1 % tolerance prior to returning the valves to service after setpoint testing.

4.4.2.3 The safety relief valve main (mechanical) stage assemblies shall be set pressure tested, reinstalled or replaced with spares that have been previously set pressure tested and stored in accordance with manufacturer's recommendations in accordance with the Surveillance Frequency Control Program.

HOPE CREEK 3/4 4-8 Amendment No. 187

REACTOR COOLANT SYSTEM SAFETY/RELIEF VALVES LOW-LOW SET FUNCTION LIMITING CONDITION FOR OPERATION 3.4.2.2 The relief valve function and the low-low set function of the following reactor coolant system safety/relief valves shall be OPERABLE with the following settings:

Low-Low Set Function Setpoint* (psig) +/-2%

Valve No.

Open Close F013H 1017 905 F013P 1047 935 APPLICABI LlTY:

OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

a.

With the relief valve function and/or the low-low set function of one of the above required reactor coolant system safety/relief valves inoperable, restore the inoperable relief valve function and low-low set function to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

With the relief valve function and/or the low-low set function of both of the above required reactor coolant system safety/relief valves inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.2.2.1 The relief valve function and the low-low set function pressure actuation instrumentation shall be demonstrated OPERABLE by performance of a:

a.

CHANNEL FUNCTIONAL TEST in accordance with the Surveillance Frequency Control Program.

b.

CHANNEL CALIBRATION, LOGIC SYSTEM FUNCTIONAL TEST and simulated automatic operation of the entire system (excluding actual valve actuation) in accordance with the Surveillance Frequency Control Program.

The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.

HOPE CREEK 3/4 4-9 Amendment No. 187

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.3.1 The reactor coolant system leakage detection systems shall be demonstrated OPERABLE by:

a.

Drywell atmosphere gaseous radioactivity monitoring system-performance of a CHANNEL CHECK, a CHANNEL FUNCTIONAL TEST and a CHANNEL CALIBRATION in accordance with the Surveillance Frequency Control Program.

b.

The drywell pressure shall be monitored in accordance with the Surveillance Frequency Control Program and the drywell temperature shall be monitored in accordance with the Surveillance Frequency Control Program.

c.

Drywell floor and equipment drain sump monitoring system-performance of a CHANNEL FUNCTIONAL TEST and a CHANNEL CALIBRATION TEST in accordance with the Surveillance Frequency Control Program.

d.

Drywell air coolers condensate flow rate monitoring system-performance of a CHANNEL FUNCTIONAL TEST and a CHANNEL CALIBRATION in accordance with the Surveillance Frequency Control Program.

HOPE CREEK 3/44-10a Amendment No. 187

REACTOR COOLANT SYSTEM 4.4.3.2.1 The reactor coolant system leakage shall be demonstrated to be within each of the above limits by:

a.

Monitoring the drywell atmospheric gaseous radioactivity in accordance with the Surveillance Frequency Control Program (not a means of quantifying leakage),

b.

Monitoring the drywell floor and equipment drain sump flow rate in accordance with the Surveillance Frequency Control Program, and

c.

Monitoring the drywell air coolers condensate flow rate in accordance with the Surveillance Frequency Control Program, and

d.

Monitoring the drywell pressure in accordance with the Surveillance Frequency Control Program (not a means of quantifying leakage), and

e.

Monitoring the reactor vessel head flange leak detection system in accordance with the Surveillance Frequency Control Program (not a means of quantifying leakage), and

f.

Monitoring the drywell temperature in accordance with the Surveillance Frequency Control Program (not a means of quantifying leakage).

4.4.3.2.2 Each reactor coolant system pressure isolation valve specified in Table 3.4.3.2-1 shall be demonstrated OPERABLE by leak testing pursuant to the 1ST Program and verifying the leakage of each valve to be within the specified limit:

a.

In accordance with the Surveillance Frequency Control Program,** and

b.

Prior to returning the valve to service following maintenance, repair or replacement work on the valve which could affect its leakage rate.

The provisions of Specification 4.0.4 are not applicable for entry into OPERATIONAL CONDITION 3.

4.4.3.2.3 The high/low pressure interface valve leakage pressure monitors shall be demonstrated OPERABLE with alarm setpoints per Table 3.4.3.2-2 by performance of a CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION at the frequencies specified in the Surveillance Frequency Control Program.

P.I.V. leak test extension to the first refueling outage is permissible for each RCS P.I.V.

listed in Table 3.4.3.2-1, that is identified in Public Service Electric & Gas Company's letter to the NRC (letter No. NLR-N87047), dated April 3, 1987, as needing a plant outage to test. For this one time test interval, the requirements of Section 4.0.2 are not applicable.

HOPE CREEK 3/44-12 Amendment No. 187

TABLE 4.4.5-1 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT AND ANALYSIS

1.

Gross Beta and Gamma Activity Determination

2.

Isotopic Analysis for DOSE EQUIVALENT 1-131 Concentration

3.

Radiochemical for E Determination

4.

Isotopic Analysis for Iodine

5.

Isotopic Analysis of an Off-gas Sample Including Quantitative Measurements for at least Xe-133, Xe-135 and Kr-88 SAMPLE AND ANALYSIS FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program

a) At least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , whenever the specific activity exceeds a limit, as required by ACTION b.

b)

At least one sample, between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following the change in THERMAL POWER or off-gas level, as required by ACTION c.

In accordance with the Surveillance Frequency Control Program OPERATIONAL CONDITIONS IN WHICH SAMPLE AND ANALYSIS REQUIRED 1,2,3 1

1 1#,2#,3#,4#

1,2 1

Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer.

Until the specific activity of the primary coolant system is restored to within its limits.

HOPE CREEK 3/44-20 Amendment No. 187

REACTOR COOLANT SYSTEM 3/4.4.6 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FORQPERATION 3.4.6.1 The reactor coolant system temperature and pressure shall be limited in accordance with the limit lines shown on Figure 3.4.6.1-1 (hydrostatic or leak testing), and Figure 3.4.6.1-2 (heatup by non-nuclear means, cooldown following a nuclear shutdown and low power PHYSICS TESTS), and Figure 3.4.6.1-3 (operations with a critical core other than low power PHYSICS TESTS), with:

a.

A maximum heatup of 100°F in anyone hour period,

b.

A maximum cooldown of 100°F in anyone hour period,

c.

A maximum temperature change of less than or equal to 20°F in anyone hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves, and

d.

The reactor vessel flange and head flange metal temperature shall be maintained greater than or equal to 79°F when reactor vessel head bolting studs are under tension.

APPLICABILITY At all times.

ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out of-limit condition on the structural integrity of the reactor coolant system; determine that the reactor coolant system remains acceptable for continued operations or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURYEIL~NCE REQUIREMENTS 4.4.6.1.1 During system heatup, cooldown and inservice leak and hydrostatic testing operations, the reactor coolant system temperature and pressure shall be determined to be within the above required heatup and cooldown limits and to the right of the limit lines of Figures 3.4.6.1-1,3.4.6.1-2, and 3.4.6.1-3 as applicable, in accordance with the Surveillance Frequency Control Program.

HOPE CREEK 3/44-21 Amendment No. 187

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (continued) 4.4.6.1.2 The reactor coolant system temperature and pressure shall be determined to be to the right of the criticality limit line of Figure 3.4.6.1-3 within 15 minutes prior to the withdrawal of control rods to bring the reactor to criticality and in accordance with the Surveillance Frequency Control Program during system heatup.

4.4.6.1.3 The reactor vessel material surveillance specimens shall be removed and examined, to determine changes in reactor pressure vessel material properties, as required by 10 CFR 50, Appendix H. The results of these examinations shall be used to update the curves of Figures 3.4.6.1-1,3.4.6.1-2, and 3.4.6.1-3.

4.4.6.1.4 The reactor vessel flange and head flange temperature shall be verified to be greater than or equal to the limit specified in 3.4.6.1.d.

a.

In OPERATIONAL CONDITION 4 when reactor coolant system temperature is:

1.

2.

s 110°F, in accordance with the Surveillance Frequency Control Program.

s gO°F, in accordance with the Surveillance Frequency Control Program.

b.

Within 30 minutes prior to and in accordance with the Surveillance Frequency Control Program during tenSioning of the reactor vessel head bolting studs.

HOPE CREEK 3/44-22 Amendment No. 187

REACTOR COOLANT SYSTEM REACTOR STEAM DOME LIMITING CONDITION FOR OPERATION 3.4.6.2 The pressure in the reactor steam dome shall be less than 1020 psig.

APPLICABILITY:

OPERATIONAL CONDITION 1

and 2*.

ACTION:

With the reactor steam dome pressure exceeding 1020 psig, reduce the pressure to less than 1020 psig within 15 minutes or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.6.2 The reactor steam dome pressure shall be verified to be less than 1020 psig in accordance with the Surveillance Frequency Control Program.

Not applicable during anticipated transients.

HOPE CREEK 3/44-25 Amendment No. 187

REACTOR COOLANT SYSTEM 3/4.4.9 RESIDUAL HEAT REMOVAL HOT SHUTDOWN LIMITING CONDITIOO FOR OPERATION 3.4.9.1 Two# shutdown cooling mode loops of the residual heat removal (RHR) system shall be OPERABLE and, unless at least one recirculation pump is in operation, at least one shutdown cooling mode loop shall be in operation*'##, with each loop consisting of:

a.

One OPERABLE RHR pump, and

b.

One OPERABLE RHR heat exchanger.

APPLICABILITY:

OPERATIONAL CONDITION 3, with reactor vessel pressure less than the RHR cut-in permissive setpoint.

ACTION:

a.

With less than the above required RHR shutdown cooling mode loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible. Within one hour and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, demonstrate the operability of at least one alternate method capable of decay heat removal for each inoperable RHR shutdown cooling mode loop. Be in at least COLD SHUTDOWN within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .**

b.

With no RHR shutdown cooling mode loop or recirculation pump in operation, immediately initiate corrective action to return at least one loop to operation as soon as possible. Within one hour establish reactor coolant circulation by an alternate method and monitor reactor coolant temperature and pressure at least once per hour.

SURVEILLANCE REQUIREMENTS 4.4.9.1 At least one shutdown cooling mode loop of the residual heat removal system, one recirculation pump, or alternate method shall be determined to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program.

One RHR shutdown cooling mode loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided the other loop is OPERABLE and in operation or at least one recirculation pump is in operation.

The shutdown cooling pump may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided the other loop is OPERABLE.

The RHR shutdown cooling mode loop may be removed from operation during hydrostatic testing.

Whenever two or more RHR subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.

HOPE CREEK 3/44-28 Amendment No. 187

REACTOR COOLANT SYSTEM COLD SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.9.2 Two# shutdown cooling mode loops of the residual heat removal (RHR) system shall be OPERABLE and, unless at least one recirculation pump is in operation, at least one shutdown cooling mode loop shall be in operation',## with each loop consisting of:

a.

One OPERABLE RHR pump, and

b.

One OPERABLE RHR heat exchanger.

APPLICABILITY:

OPERATIONAL CONDITION 4 and heat losses to ambienC are not sufficient to maintain OPERATIONAL CONDITION 4.

ACTION:

a.

With less than the above required RHR shutdown cooling mode loops OPERABLE, within one hour and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, demonstrate the operability of at least one alternate method capable of decay heat removal for each inoperable RHR shutdown cooling mode loop.

b.

With no RHR shutdown cooling mode loop or recirculation pump in operation, within one hour establish reactor coolant circulation by an alternate method and monitor reactor coolant temperature and pressure at least once per hour.

4.4.9.2 At least one shutdown cooling mode loop of the residual heat removal system, recirculation pump or alternate method shall be determined to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program.

One RHR shutdown cooling mode loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided the other loop is OPERABLE and in operation or at least one recirculation pump is in operation.

The shutdown cooling pump may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided the other loop is OPERABLE.

The shutdown cooling mode loop may be removed from operation during hydrostatic testing.

Ambient losses must be such that no increase in reactor vessel water temperature will occur (even though COLD SHUTDOWN conditions are being maintained).

HOPE CREEK 3/44-29 Amendment No. 187

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.1 The emergency core cooling systems shall be demonstrated OPERABLE by:

a.

In accordance with the Surveillance Frequency Control Program:

1.

For the core spray system, the LPCI system, and the HPCI system:

a)

Verifying by venting at the high point vents that the system piping from the pump discharge valve to the system isolation valve is filled with water.

b)

Verifying that each valve, manual, power operated or automatic, in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct* position.

c)

Verify the RHR System cross tie valves on the discharge side of the pumps are closed and power, if any, is removed from the valve operators.

2.

For the HPCI system, verifying that the HPGI pump flow controller is in the correct position.

b.

Verifying that, when tested pursuant to the 1ST Program:

1.

The two core spray system pumps in each subsystem together develop a flow of at least 6150 gpm against a test line pressure corresponding to a reactor vessel pressure of ~105 psi above suppression pool pressure.

2.

Each LPGI pump in each subsystem develops a flow of at least 10,000 gpm against a test line pressure corresponding to a reactor vessel to primary containment differential pressure of ~20 psid.

3.

The HPGI pump develops a flow of at least 5600 gpm against a test line pressure corresponding to a reactor vessel pressure of 1000 psig when steam is being supplied to the turbine at 1000, +20, -80 psig.**

c.

In accordance with the Surveillance Frequency Gontrol Program:

1.

For the core spray system, the LPGI system, and the HPGI system, performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence and verifying that each automatic valve in the flow path actuates to its correct position. Actual injection of coolant into the reactor vessel may be excluded from this test Except that an automatic valve capable of automatic return to its EGGS position when an EGGS signal is present may be in position for another mode of operation.

The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is adequate to perform the test.

HOPE CREEK 3/4 5-4 Amendment No. 187

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENT~m(Continued)

2.

For the HPCI system, verifying that:

a)

The system develops a flow of at least 5600 gpm against a test line pressure corresponding to a reactor vessel pressure of

200 psig, when steam is being supplied to the turbine at 200 + 15, -0 psig.**

b)

The suction is automatically transferred from the condensate storage tank to the suppression chamber on a condensate storage tank water level - low signal and on a suppression chamber water level high signal.

3.

Performing a CHANNEL CALIBRATION of the CSS, and LPCI system discharge line "keep filled" alarm instrumentation.

4.

Performing a CHANNEL CALIBRATION of the CSS header ~P instrumentation and verifying the setpoint to be s the allowable value of 4.4 psid.

5.

Performing a CHANNEL CALIBRATION of the LPCI header ~P instrumentation and verifying the setpoint to be s the allowable value of 1.0 psid.

d.

For the ADS:

1.

In accordance with the Surveillance Frequency Control Program, performing a CHANNEL FUNCTIONAL TEST of the Primary Containment Instrument Gas System lOW-low pressure alarm system.

2.

In accordance with the SUrveillance Frequency Control Program:

a)

Performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence, but excluding actual valve actuation.

b)

Verify that when tested pursuant to the 1ST Program, that each ADS valve is capable of being opened.

c)

Performing a CHANNEL CALIBRATION of the Primary Containment I nstrument Gas System lOW-low pressure alarm system and verifying an alarm setpoint of 85 +/- 2 psig on decreasing pressure.

The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is adequate to perform the test.

HOPE CREEK 3/4 5-5 Amendment No. 187

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.2.1 At least the above required ECCS shall be demonstrated OPERABLE per Surveillance Requirement 4.5.1.

4.5.2.2 The core spray system shall be determined OPERABLE in accordance with the Surveillance Frequency Control Program by verifying the condensate storage tank required volume when the condensate storage tank is required to be OPERABLE per Specification 3.5.2.a.2.b.

HOPE CREEK 3/4 5~7 Amendment No. 187

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.3.1 The suppression chamber shall be determined OPERABLE by verifying the water level to be greater than or equal to:

a.

74.5" in accordance with the Surveillance Frequency Control Program in OPERATIONAL CONDITIONS 1, 2, and 3.

b.

5.0" in accordance with the Surveillance Frequency Control Program in OPERATIONAL CONDITIONS 4 and 5*.

4.5.3.2 With the suppression chamber level less than the above limit or drained in OPERATIONAL CONDITION 4 or 5*, in accordance with the Surveillance Frequency Control Program:

a.

Verify the required conditions of Specification 3.5.3.b to be satisfied, or

b.

Verify footnote conditions

to be satisfied.

HOPE CREEK 3/4 5-9 Amendment No. 187

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT PRIMARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2* and 3.

ACTION:

Without PRIMARY CONTAINMENT INTEGRITY, restore PRIMARY CONTAINMENT INTEGRITY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be demonstrated:

a.

After each closing of each penetration subject to Type B testing, except the primary containment air locks, if opened following Type A or B test, by leak rate testing in accordance with the Primary Containment Leakage Rate Testing Program.

b.

In accordance with the Surveillance Frequency Control Program by verifying that all primary containment penetrations** not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in position, except for valves that are opened under administrative control as permitted by Specification 3.6.3.

c.

By verifying each primary containment air lock is in compliance with the requirements of Specification 3.6.1.3.

d.

By verifying the suppression chamber is in compliance with the requirements of Specification 3.6.2.1.

See Special Test Exception 3.10.1 Except valves, blind flanges, and deactivated automatic valves which are located inside the primary containment, and are locked, sealed or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except such verification need not be performed when the primary containment has not been de-inerted since the last verification or more often than once per 92 days.

HOPE CREEK 3/46-1 Amendment No. 187

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6.1.3 Each primary containment air lock shall be demonstrated OPERABLE:

a.

By verifying seal leakage rate in accordance with the Primary Containment Leakage Rate Testing Program.

b.

By conducting an overall air lock leakage test in accordance with the Primary Containment Leakage Rate Testing Program.

c.

In accordance with the Surveillance Frequency Control Program by verifying that only one door in each air lock can be opened at a time.**

Except that the inner door need not be opened to verify interlock OPERABILITY when the primary containment is inerted, provided that the inner door interlock is tested within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after the primary containment has been de-inerted.

HOPE CREEK 3/4 6-6 Amendment No. 187

CONTAINMENT SYSTEMS DRYWELL AND SUPPRESSION CHAMBER INTERNAL PRESSURE LIMITING CONDITION FOR OPERATION 3.6.1.6 Drywell and suppression chamber internal pressure shall be maintained between -0.5 and +1.5 psig.

APPLICABI LlTY:

OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

With the drywell and/or suppression chamber internal pressure outside of the specified limits, restore the internal pressure to within the limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.6.1.6 The drywell and suppression chamber internal pressure shall be determined to be within the limits in accordance with the Surveillance Frequency Control Program.

HOPE CREEK 3/4 6-9 Amendment No. 187

CONTAINMENT SYSTEMS DRYWELL AVERAGE AIR TEMPERATURE LIMITING CONDITION FOR OPERATION 3.6.1.7 Drywell average air temperature shall not exceed 135°F.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

With the drywell average air temperature greater than 135°F, reduce the average air temperature to within the limit within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4.6.1.7 The drywell average air temperature shall be the volumetric average of the temperatures at the following locations and shall be determined to be within the limit in accordance with the Surveillance Frequency Control Program:

Elevation Zone Approximate Azimuth*

a.

86'11"-112'8" (under vessel)

b.

86'11"-111'10" (outside of pedestal)

c.

111'10"-139'2" 55°,240°, 155°, 315°

d.

139'2"-168'0" 45°,215°,0°,90°,180°,270

e.

168'0"-192'7" At least one reading from each elevation zone is required for a volumetric average calculation.

HOPE CREEK 3/46-10 Amendment No. 187

CONTAINMENT SYSTEMS LIMITING CONDITION FOR OPERATION (continued)

ACTION: (Continued)

3.

With the suppression chamber average water temperature greater than 120°F, depressurize the reactor pressure vessel to less than 200 psig within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c.

With one drywell-to-suppression chamber bypass leakage in excess of the limit, restore the bypass leakage to within the limit prior to increasing reactor coolant temperature above 200°F.

4.6.2.1 The suppression chamber shall be demonstrated OPERABLE:

a.

By verifying the suppression chamber water volume to be within the limits in accordance with the Surveillance Frequency Control Program.

b.

In accordance with the Surveillance Frequency Control Program in OPERATIONAL CONDITION 1 or 2 by verifying the suppression chamber average water temperature to be less than or equal to 95°F, except:

1.

At least once per 5 minutes during testing which adds heat to the suppression chamber, by verifying the suppression chamber average water temperature less than or equal to 105°F.

2.

At least once per hour when suppression chamber average water temperature is greater than 95°F, by verifying:

a)

SuppreSSion chamber average water temperature to be less than or equal to 110°F.

c.

At least once per 30 minutes in OPERATIONAL CONDITION 3 following a scram with suppression chamber average water temperature greater than 95°F, by verifying suppression chamber average water temperature less than or equal to 120°F.

d.

By an external visual examination of the suppression chamber after safety/relief valve operation with the suppression chamber average water temperature greater than or equal to 17rF and reactor coolant system pressure greater than 100 psig.

e.

In accordance with the Surveillance Frequency Control Program by a visual inspection of the accessible interior and exterior of the suppression chamber.

HOPE CREEK 3/46-13 Amendment No. 187

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (continued)

f.

In accordance with the Surveillance Frequency Control Program by conducting a drywell-to-suppression chamber bypass leak test at an initial differential pressure of 0.80 psi and verifying that the differential pressure does not decrease by more than 0.24 inch of water per minute for a period of 10 minutes. If any drywell-to suppression chamber bypass leak test fails to meet the specified limit, the test schedule for subsequent tests shall be reviewed and approved by the Commission. If two consecutive tests fail to meet the specified limit, a test shall be performed at least every 9 months until two consecutive tests meet the specified limit, at which time the Surveillance Frequency Control Program schedule may be resumed.

HOPE CREEK 3/46-14 Amendment No. 187

CONTAINMENT SYSTEMS SUPPRESSION POOL SPRAY LIMITING CONDITION FOR OPERATION 3.6.2.2 The suppression pool spray mode of the residual heat removal (RHR) system shall be OPERABLE with two independent loops, each loop consisting of:

a.

One OPERABLE RHR pump, and

b.

An OPERABLE flow path capable of recirculating water from the suppression chamber through an RHR heat exchanger and the suppression pool spray sparger.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

a.

With one suppression pool spray loop inoperable, restore the inoperable loop to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

With both suppression pool spray loops inoperable, restore at least one loop to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN* within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.6.2.2 The suppression pool spray mode of the RHR system shall be demonstrated OPERABLE:

a.

In accordance with the Surveillance Frequency Control Program by verifying that each valve, manual, power operated or automatic, in the flow path that is not locked, sealed or otherwise secured in position, is in its correct position.

b.

By verifying that each of the required RHR pumps develops a flow of at least 540 gpm on recirculation flow through the RHR heat exchanger (after consideration of flow through the closed bypass valve) and suppression pool spray sparger when tested pursuant to the 1ST Program.

Whenever both RHR subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.

HOPE CREEK 3/46-15 Amendment No. 187

CONTAINMENT SYSTEMS SUPPRESSION POOL COOLING LIMITING CONDITION FOR OPERATION 3.6.2.3 The suppression pool cooling mode of the residual heat removal (RHR) system shall be OPERABLE with two independent loops, each loop consisting of:

a.

One OPERABLE RHR pump, and

b.

An OPERABLE flow path capable of recirculating water from the suppression chamber through an RHR heat exchanger.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

a.

With one suppression pool cooling loop inoperable, restore the inoperable loop to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

With both suppression pool cooling loops inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN* within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.6.2.3 The suppression pool cooling mode of the RHR system shall be demonstrated OPERABLE:

a.

In accordance with the Surveillance Frequency Control Program by verifying that each valve, manual, power operated or automatic, in the flow path that is not locked, sealed or otherwise secured in position, is in its correct position.

b.

By verifying that each of the required RHR pumps develops a flow of at least 10,160 gpm on recirculation flow through the RHR heat exchanger (after consideration of flow through the closed bypass valve) and the suppression pool when tested pursuant to the 1ST Program.

Whenever both RHR subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.

HOPE CREEK 3/46-16 Amendment No. 187

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6.3.1 Each primary containment isolation valve shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by cycling the valve through at least one complete cycle of full travel and verifying the specified isolation time.

4.6.3.2 Each primary containment automatic isolation valve shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by verifying that on a containment isolation test signal each automatic isolation valve actuates to its isolation position.

4.6.3.3 The isolation time of each primary containment power operated or automatic valve shall be determined to be within its limit when tested pursuant to the 1ST Program.

4.6.3.4 In accordance with the Surveillance Frequency Control Program, verify that a representative sample of reactor instrumentation line excess flow check valves# actuates to the isolation position on a simulated instrument line break signal.

4.6.3.5 Each traversing in-core probe system explosive isolation valve shall be demonstrated OPERABLE*:

a.

In accordance with the Surveillance Frequency Control Program by verifying the continuity of the explosive charge.

b.

In accordance with the Surveillance Frequency Control Program by removing the explosive squib from at least one explosive valve, and initiating the explosive squib. The replacement charge for the exploded squib shall be from the same manufactured batch as the one fired or from another batch which has been certified by having at least one of that batch successfully fired. No squib shall remain in use beyond the expiration of its shelf-life or operating life, as applicable.

Exemption to Appendix J of 10 CFR Part 50.

The reactor vessel head seal leak detection line (penetration J5C) is not required to be tested pursuant to this requirement.

HOPE CREEK 3/46-18 Amendment No. 187

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6.4.1 Each suppression chamber - drywell vacuum breaker shall be:

a.

Verified closed in accordance with the Surveillance Frequency Control Program*.

b.

Demonstrated OPERABLE:

1.

In accordance with the Surveillance Frequency Control Program and within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after any discharge of steam to the suppression chamber from the safety-relief valves, by performing a functional test of each vacuum breaker.

2.

In accordance with the Surveillance Frequency Control Program by verifying the opening setpoint of each vacuum breaker to be less than or equal to 0.20 psid.

Not required to be met for vacuum breaker assembly valves that are open during surveillances or that are open when performing their intended functions.

HOPE CREEK 3/46-44 Amendment No. 187

CONTAINMENT SYSTEMS REACTOR BUILDING - SUPPRESSION CHAMBER VACUUM BREAKERS J.J.MlTING CONDITION FOR OPERATION 3.6.4.2 Each reactor building - suppression chamber vacuum breaker assembly shall be OPERABLE APPLICABI LlTY:

OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

a.

With one reactor building - suppression chamber vacuum breaker assembly, with one or two valves inoperable for opening, restore the vacuum breaker assembly to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

With two reactor building - suppression chamber vacuum breaker assemblies with one or two valves inoperable for opening, restore both valves in one vacuum breaker assembly to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c.

With one or two reactor building - suppression chamber vacuum breaker assemblies, with one valve not closed, close the open vacuum breaker assembly valve(s) within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

d.

With two valves in one or two reactor building - suppression chamber vacuum breaker assemblies not closed, close one open vacuum breaker assembly valve in each affected assembly within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.6.4.2 Each reactor building - suppression chamber vacuum breaker assembly shall be:

a.

Verified closed in accordance with the Surveillance Frequency Control Program*.

b.

Demonstrated OPERABLE:

1.

In accordance with the Surveillance Frequency Control Program by:

a)

Performing a functional test of each vacuum breaker assembly valve.

2.

In accordance with the Surveillance Frequency Control Program by:

a)

Verifying the opening setpoint of each vacuum breaker assembly valve to be less than or equal to 0.25 psid.

Not required to be met for vacuum breaker assembly valves that are open during surveillances or that are open when performing their intended functions.

HOPE CREEK 3/46-45 Amendment No. 187

This Page Intentionally Blank HOPE CREEK 3/46-46 Amendment No. 187

CONTAINMENT SYSTEMS 3/4.6.5 SECONDARY CONTAINMENT SECONDARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR QPERATION 3.6.5.1 SECONDARY CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY:

OPERATIONAL CONDITIONS 1,2,3 and *.

ACTION:

Without SECONDARY CONTAINMENT INTEGRITY:

a.

In OPERATIONAL CONDITION 1,2 or 3, restore SECONDARY CONTAINMENT INTEGRITY within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

In Operational Condition *, suspend handling of recently irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.

4.6.5.1 SECONDARY CONTAINMENT INTEGRITY shall be demonstrated by:

a.

Verifying in accordance with the Surveillance Frequency Control Program that the reactor building is at a negative pressure.

b.

Verifying in accordance with the Surveillance Frequency Control Program that:

1.

All secondary containment equipment hatches and blowout panels are closed and sealed.

2.

a.

For double door arrangements, at least one door in each access to the secondary containment is closed.

b.

For single door arrangements, the door in each access to the secondary containment is closed except for routine entry and exit.

3.

All secondary containment penetrations not capable of being closed by OPERABLE secondary containment automatic isolation dampers/valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic dampers/valves secured in position.

When recently irradiated fuel is being handled in the secondary containment and during operations with a potential for draining the reactor vessel.

HOPE CREEK 3/46-47 Amendment No. 187

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

c.

In accordance with the Surveillance Frequency Control Program:

1.

Verifying that four filtration recirculation and ventilation system (FRVS) recirculation units and one ventilation unit of the filtration recirculation and ventilation system will draw down the secondary containment to greater than or equal to 0.25 inches of vacuum water gauge in less than or equal to 375 seconds, and

2.

Operating four filtration recirculation and ventilation system (FRVS) recirculation units and one ventilation unit of the filtration recirculation and ventilation system for four hours and maintaining greater than or equal to 0.25 inches of vacuum water gauge in the secondary containment at a flow rate not exceeding 3324 CFM.

HOPE CREEK 3/46-48 Amendment No. 187

CONTAINMENT SYSTEMS SECONDARY CONTAI NM ENT AUTOMATIC ISOLATION DAMPERS LIMITING CONDITION FOR OPERATION 3.6.5.2 The secondary containment ventilation system (RBVS) automatic isolation dampers shown in Table 3.6.5.2-1 shall be OPERABLE with isolation times less than or equal to the times shown in Table 3.6.5.2-1.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3 and *.

ACTION:

With one or more of the secondary containment ventilation system automatic isolation dampers shown in Table 3.6.5.2-1 inoperable, maintain at least one isolation damper OPERABLE in each affected penetration that is open and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months either:

a.

Restore the inoperable dampers to OPERABLE status, or

b.

Isolate each affected penetration by use of at least one deactivated damper secured in the isolation position, or

c.

Isolate each affected penetration by use of at least one closed manual valve or blind flange.

Otherwise, in OPERATIONAL CONDITION 1, 2 or 3, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Otherwise, in Operational Condition *, suspend handling of recently irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.6.5.2 Each secondary containment ventilation system automatic isolation damper shown in Table 3.6.5.2-1 shall be demonstrated OPERABLE:

a Prior to returning the damper to service after maintenance, repair or replacement work is performed on the damper or its associated actuator, control or power circuit by cycling the damper through at least one complete cycle of full travel and verifying the specified isolation time.

b.

In accordance with the Surveillance Frequency Control Program by verifying that on a containment isolation test signal each isolation damper actuates to its isolation pOSition.

c.

By verifying the isolation time to be within its limit in accordance with the Surveillance Frequency Control Program.

When recently irradiated fuel is being handled in the secondary containment and during operations with a potential for draining the reactor vessel.

HOPE CREEK 3/46-49 Amendment No. 187

CONTAINMENT SYSTEMS 3.6.5.3 FILTRATION, RECIRCULATION AND VENTILATION SYSTEM (FRVS)

FRVS VENTILATION SUBSYSTEM LIMITING CONDITIONFQR OPERATION 3.6.5.3.1 Two FRVS ventilation units shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3 and *.

ACTION:

a.

With one of the above required FRVS ventilation units inoperable, restore the inoperable unit to OPERABLE status within 7 days, or:

1.

In OPERATIONAL CONDITION 1, 2 or 3, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

In Operational Condition *, place the OPERABLE FRVS ventilation unit in operation or suspend handling of recently irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.

b.

With both ventilation units inoperable in Operational Condition *, suspend handling of recently irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3. are not applicable.

SURVEILLANCE REQUIREMENTS 4.6.5.3.1 Each of the two ventilation units shall be demonstrated OPERABLE:

a.

In accordance with the Surveillance Frequency Control Program by verifying that the water seal bucket traps have a water seal and making up any evaporative losses by filling the traps to the overflow.

b.

In accordance with the Surveillance Frequency Control Program by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the subsystem operates for at least 15 minutes.

When recently irradiated fuel is being handled in the secondary containment and during operations with a potential for draining the reactor vessel.

HOPE CREEK 3/46-51 Amendment No. 187

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (continued)

c.

In accordance with the Surveillance Frequency Control Program or upon determination** that the HEPA filters or charcoal adsorbent could have been damaged by structural maintenance or adversely affected by any chemicals, fumes or foreign materials (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the subsystem by:

1.

Verifying that the subsystem satisfies the in-place penetration testing acceptance criteria of less than 0.05% and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rates are 9,000 cfm +/- 10% for each FRVS ventilation unit.

2.

Verifying within 31 days after removal from the FRVS ventilation units, that a laboratory test of a sample of the charcoal adsorber, when obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C and a relative humidity 95%.

3.

Verifying a subsystem flow rate of 9,000 cfm +/- 10% for each FRVS ventilation unit during system operation when tested in accordance with ANSI N510-1980.

d.

After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying within 31 days after removal from the FRVS ventilation units, that a laboratory analysis of a representative carbon sample, when obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows a methyl iodide penetration less than 5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C and a relative humidity of 95%.

This determination shall consider the maintenance performed and/or the type, quantity, length of contact time, known effects and previous accumulation history for all contaminants which could reduce the system performance to less than that verified by the acceptance criteria in items c.1 through c.3 below.

HOPE CREEK 3/46-51 a Amendment No. 187

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (continued)

e.

In accordance with the Surveillance Frequency Control Program by:

1.

Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 5 inches Water Gauge in the ventilation unit while operating the filter train at a flow rate of 9,000 cfm +/-

10% for each FRVS ventilation unit.

2.

Verifying that the filter train starts and isolation dampers open on each of the following test signals:

a.

Manual initiation from the control room, and

b.

Simulated automatic initiation signal.

f.

After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter bank satisfies the inplace penetration testing acceptance criteria of less than 0.05% in accordance with Regulatory Position C.5.a and C.5.c of Regulatory Guide 1.52, Revision 2 March 1978, while operating the system at a flow rate of 9,000 cfm +/- 10% for each FRVS ventilation unit.

g.

After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorber bank satisfies the inplace penetration testing acceptance criteria of less than 0.05% in accordance with Regulatory Position C.5.a and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 9,000 cfm +/- 10% for each FRVS ventilation unit.

HOPE CREEK 3/46-52 Amendment No. 187

CONTAI NIVIENT SYSTEMS 3.6.5.3 FILTRATION, RECIRCULATION AND VENTILATION SYSTEM (FRVS)

FRVS RECIRCULATION SUBSYSTEM LIMITING CONDITION FOR OPERATION 3.6.5.3.2 Six FRVS recirculation units shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and *.

ACTION:

a.

With one or two of the above required FRVS recirculation units inoperable, restore all the inoperable unit(s) to OPERABLE status within 7 days, or:

1.

In OPERATIONAL CONDITION 1,2, or 3, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

In Operational Condition*, suspend handling of recently irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.

b.

With three or more of the above required FRVS recirculation units inoperable in Operational Condition *, suspend handling of recently irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.

c.

With three or more of the above required FRVS recirculation units inoperable in OPERATIONAL CONDITION 1,2, or 3, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.6.5.3.2 Each of the six FRVS recirculation units shall be demonstrated OPERABLE:

a.

In accordance with the Surveillance Frequency Control Program by verifying that the water seal bucket traps have a water seal and making up any evaporative losses by filling the traps to the overflow.

b.

In accordance with the Surveillance Frequency Control Program by initiating, from the control room, flow through the HEPA filters and verifying that the subsystem operates for at least 15 minutes.

When recently irradiated fuel is being handled in the secondary containment and during operations with a potential for draining the reactor vessel.

HOPE CREEK 3/46-52a Amendment No. 187

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (continued)

c.

In accordance with the Surveillance Frequency Control Program or upon determination** that the HEPA filters could have been damaged by structural maintenance or adversely affected by any foreign materials (1) after any structural maintenance on the HEPA filters or housings by:

1.

Verifying that the subsystem satisfies the in-place penetration testing acceptance criteria of less than 0.05% and uses the test procedure guidance in Regulatory Positions C.5.a and C.5.c of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rates are 30,000 cfm +/-

10% for each FRVS recirculation unit.

2.

Verifying a subsystem flow rate of 30,000 cfm +/- 10% for each FRVS recirculation unit during system operation when tested in accordance with ANSI N510-1980.

d.

not used

e.

In accordance with the Surveillance Frequency Control Program by:

1.

Verifying that the pressure drop across the exhaust duct is less than 8 inches Water Gauge in the recirculation filter train while operating the filter train at a flow rate of 30,000 cfm +/- 10% for each FRVS recirculation unit.

2.

Verifying that the filter train starts and isolation dampers open on each of the following test signals:

a.

Manual initiation from the control room, and

b.

Simulated automatic initiation signal.

f.

After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter bank satisfies the inplace penetration testing acceptance criteria of less than 0.05% in accordance with Regulatory Position C.5.a and C.5.c of Regulatory Guide 1.52, Revision 2 March 1978, while operating the system at a flow rate of 30,000 cfm +/- 10% for each FRVS recirculation unit.

This determination shall consider the maintenance performed and/or the type, quantity, length of contact time, known effects and previous accumulation history for all contaminants which could reduce the system performance to less than that verified by the acceptance criteria in items c.1 and c.2 below.

HOPE CREEK 3/46-53 Amendment No. 187

CONTAINMENT SYSTEMS This Page Intentionally Blank HOPE CREEK 3/46-53a Amendment No. 187

CONTAINIVIENT SYSTEMS DRYWELL AND SUPPRESSION CHAMBER OXYGEN CONCENTRATION LIMITING CONDITION FOR OPERATION 3.6.6.2 The drywell and suppression chamber atmosphere oxygen concentration shall be less than 4% by volume.

APPLICABI LlTY:

OPERATIONAL CONDITION 1*, during the time period:

a.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER is greater than 15% of RATED THERMAL POWER, following startup, to

b.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to reducing THERMAL POWER to less than 15% of RATED THERMAL POWER preliminary to a scheduled reactor shutdown.

ACTION:

With the drywell and/or suppression chamber oxygen concentration exceeding the limit, restore the oxygen concentration to within the limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least STARTUP within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

SURVEILLANCE REQUIREMENTS 4.6.6.2 The drywell and suppression chamber oxygen concentration shall be verified to be within the limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER is greater than 15% of RATED THERMAL POWER and in accordance with the Surveillance Frequency Control Program thereafter.

See Special Test Exception 3.10.5.

HOPE CREEK 3/46-55 Amendment No. 187

PLANT SYSTEMS LIMITING CONDITION FOR OPERATION (continued)

ACTION: (Continued)

c.

In OPERATIONAL CONDITION 4 or 5 with the SACS subsystem, which is associated with safety related equipment required OPERABLE by Specification 3.5.2, having two SACS pumps or one heat exchanger inoperable, declare the associated safety related equipment inoperable and take the ACTION required by Specification 3.5.2.

d.

In OPERATIONAL CONDITION 5 with the SACS subsystem, which is associated with an RHR loop required OPERABLE by Specification 3.9.11.1 or 3.9.11.2, having two SACS pumps or one heat exchanger inoperable, declare the associated RHR system inoperable and take the ACTION required by Specification 3.9.11.1 or 3.9.11.2, as applicable.

e.

In OPERATIONAL CONDITION 4, 5, or **, with one SACS subsystem, which is associated with safety related equipment required OPERABLE by Specification 3.8.1.2, inoperable, realign the associated diesel generators within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to the OPERABLE SACS subsystem, or declare the associated diesel generators inoperable and take the ACTION required by Specification 3.8.1.2. The provisions of Specification 3.0.3 are not applicable.

f.

In OPERATIONAL CONDITION 4, 5, or **, with only one SACS pump and heat exchanger and its associated flowpath OPERABLE, restore at least two pumps and two heat exchangers and associated flowpaths to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or, declare the associated safety related equipment inoperable and take the associated ACTION requirements.

SURVEILLANCE REQUIREIVIENTS 4.7.1.1 At least the above required safety auxiliaries cooling system subsystems shall be demonstrated OPERABLE:

a.

In accordance with the Surveillance Frequency Control Program by verifying that each valve in the 'flow path that is not locked, sealed or otherwise secured in position, is in its correct position.

b.

In accordance with the Surveillance Frequency Control Program by verifying that:

1) Each automatic valve servicing safety-related equipment actuates to its correct position on the appropriate test signal(s), and 2) Each pump starts automatically when its associated diesel generator automatically starts.

HOPE CREEK 3/4 7-2 Amendment No. 187

PLANT SYSTEMS LIMITING CONDITION FOR OPERATION (continued)

ACTION: (Continued)

b.

In OPERATIONAL CONDITION 4 or 5:

With only one station service water pump and its associated flowpath OPERABLE, restore at least two pumps with at least one flow path to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or declare the associated SACS subsystem inoperable and take the ACTION required by Specification 3.7.1.1.

c.

In OPERATIONAL CONDITION *:

With only one station service water pump and its associated flowpath OPERABLE, restore at least two pumps with at least one flow path to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or declare the associated SACS subsystem inoperable and take the ACTION required by Specification 3.7.1.1. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.1.2 At least the above required station service water system loops shall be demonstrated OPERABLE:

a.

In accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power operated or automatic), serviCing safety related equipment that is not locked, sealed or otherwise secured in position, is in its correct position.

b.

In accordance with the Surveillance Frequency Control Program, by verifying that:

1.

Each automatic valve servicing non-safety related equipment actuates to its isolation position on an isolation test signal.

2.

Each pump starts automatically when its associated diesel generator automatically starts.

When handling recently irradiated fuel in the secondary containment.

HOPE CREEK 3/4 7-4 Amendment No. 187

PLANT SYSTEMS ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION 3.7.1.3 The ultimate heat sink (Delaware River) shall be OPERABLE with:

a.

A minimum river water level at or above elevation -9'0 Mean Sea Level, USGS datum (80'0 PSE&G datum), and

b.

An average river water temperature of less than or equal to 85.0°F.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3, 4, 5 and *.

ACTION:

With the river water temperature in excess of 85.0°F, continued plant operation is permitted provided that both emergency discharge valves are open and emergency discharge pathways are available. With the river water temperature in excess of 88.0°F, continued plant operation is permitted provided that all of the following additional conditions are satisfied: all SSWS pumps are OPERABLE, all SACS pumps are OPERABLE, all EDGs are OPERABLE and the SACS loops have no cross-connected loads (unless they are automatically isolated during a LOP and/or LOCA); with ultimate heat sink temperature greater than 89°F and less than or equal to 91.4°F, verify once per hour that water temperature of the ultimate heat sink is less than or equal to 89°F averaged over the previous 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period; otherwise, with the requirements of the above specification not satisfied:

a.

In OPERATIONAL CONDITIONS 1, 2 or 3, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

In OPERATIONAL CONDITIONS 4 or 5, declare the SACS system and the station service water system inoperable and take the ACTION required by Specification 3.7.1.1 and 3.7.1.2.

c.

In Operational Condition *, declare the plant service water system inoperable and take the ACTION required by Specification 3.7.1.2. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.1.3 The ultimate heat sink shall be determined OPERABLE:

a.

By verifying the river water level to be greater than or equal to the minimum limit in accordance with the Surveillance Frequency Control Program.

b.

By verifying river water temperature to be within its limit:

1) in accordance with the Surveillance Frequency Control Program when the river water temperature is less than or equal to 82°F.

2) in accordance with the Surveillance Frequency Control Program when the river water temperature is greater than 82°F.

When handling recently irradiated fuel in the secondary containment.

HOPE CREEK 3/4 7-5 Amendment No. 187

PLANT SYSTEMS 3/4.7.2 CONTROL ROOM EMERGENCY FILTRATION SYSTEM LIMITING CONDITION FOR OPERATION (continued)

2.

With both control room emergency filtration subsystems inoperable for reasons other than Condition b.3, suspend handling of recently irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel.

3.

With one or more control room emergency filtration subsystems inoperable due to an inoperable CRE boundary##, immediately suspend handling of recently irradiated fuel and operations with a potential for draining the vessel.

c.

The provisions of Specification 3.0.3 are not applicable in Operational Condition*.

SURVEILLANCE REQUIREMENTS 4.7.2.1 Each control room emergency filtration subsystem shall be demonstrated OPERABLE:

a.

In accordance with the Surveillance Frequency Control Program by verifying that the control room air temperature is less than or equal to 85°F#.

b.

In accordance with the Surveillance Frequency Control Program by initiating, from the control room, the control area chilled water pump, flow through the HEPA filters and charcoal adsorbers and verifying that the subsystem operates for at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months with the heaters on in order to reduce the buildup of moisture on the carbon adsorbers and HEPA filters.

When recently irradiated fuel is being handled in the secondary containment and during operations with a potential for draining the reactor vessel.

This does not require starting the non-running control emergency filtration subsystem.

The main control room envelope (CRE) boundary may be opened intermittently under administrative control.

HOPE CREEK 3/47-6a Amendment No. 187

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (continued)

c.

In accordance with the Surveillance Frequency Control Program or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the subsystem filter train by:

1.

Verifying that the subsystem satisfies the in-place penetration testing acceptance criteria of less than 0.05% and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system filter train flow rate is 4000 cfm ~ 10%.

2.

Verifying within 31 days after removal, that a laboratory test of a sample of the charcoal adsorber, when obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 0.5% when tested in accordance with ASTM 03803-1989 at a temperature of 30°C and a relative humidity 70%.

3.

Verifying a subsystem filter train flow rate of 4000 cfm ~10% during subsystem operation when tested in accordance with ANSI N510-1980.

d.

After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying within 31 days after removal from the Control Room Emergency Filtration units that a laboratory analysis of a representative carbon sample, when obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows a methyl iodide penetration less than 0.5% when tested in accordance with ATSM 03803 -1989 at a temperature of 30°C and a relative humidity of 70%.

e.

In accordance with the Surveillance Frequency Control Program by:

1.

Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 7.5 inches Water Gauge while operating the filter train subsystem at a flow rate of 4000 cfm ~ 10%.

2.

Verifying with the control room hand switch in the recirculation mode that on each of the below recirculation mode actuation test signals, the subsystem automatically switches to the isolation mode of operation and the isolation dampers close within 5 seconds:

a)

High Orywell Pressure b)

Reactor Vessel Water Level Low Low Low, Level 1 c)

Control room ventilation radiation monitors high.

HOPE CREEK 3/4 7-7 Amendment No. 187

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (continued)

3. Verifying with the control room hand switch in the outside air mode that on each of the below pressurization mode actuation test signals, the subsystem automatically switches to the pressurization mode of operation:

a)

High Drywell Pressure

'b)

Reactor Vessel Water Level Low Low Low, Level 1 c)

Control room ventilation radiation monitors high.

4.

Verifying that the heaters dissipate 13 +/-1.3 Kw when tested in accordance with ANSI N51 0-1980 and verifying humidity is maintained less than or equal to 70% humidity through the carbon adsorbers by performance of a channel calibration of the humidity control instrumentation.

f.

After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter bank satisfies the inplace penetration testing acceptance criteria of less than 0.05% in accordance with Regulatory Positions C.5.a and C.5.c of Regulatory Guide 1.52, Revision 2, March 1978, while operating the system at a flow rate of 4000 cfm +/- 10%.

g.

After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorber bank satisfies the inplace penetration testing acceptance criteria of less than 0.05% in accordance with Regulatory Positions C.5.a and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 4000 cfm +/- 10%.

4.7.2.2 The control room envelope boundary shall be demonstrated OPERABLE:

a.

At a frequency in accordance with the Control Room Envelope Habitability Program by performance of control room envelope unfiltered air inleakage testing in accordance with the Control Room Envelope Habitability Program.

HOPE CREEK 3/4 7-8 Amendment No. 187

PLANT SYSTEMS 3/4.7.3 FLOOD PROTECTION LIMITING CONDITION FOR OPERATION 3.7.3 Flood protection shall be provided for all safety related systems, components and structures when the water level of the Delaware River reaches 6.0 feet Mean Sea Level (MSL) USGS datum (95.0 feet PSE&G datum) at the Service Water Intake Structure.

APPLICABILITY:

At all times.

ACTION:

a.

With severe storm warnings from the National Weather Service which may impact Artificial Island in effect or with the water level at the service water intake structure above elevation 6.0 feet MSL USGS datum (95.0 feet PSE&G datum), initiate and complete:

1.

The closing of all service water intake structure watertight perimeter flood doors identified in Table 3.7.3-1 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , or declare affected service water system components inoperable and take the actions required by LCO 3.7.1.2;

- and

2.

The closing of all power block watertight perimeter flood doors identified in Table 3.7.3-1 within 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Once closed, all access through the doors shall be administratively controlled.

b.

With the water level at the service water intake structure above elevation 10.5 feet MSL USGS datum (99.5 feet PSE&G datum), be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.7.3 The water level at the service water intake structure shall be determined to be within the limit by:

a.

Measurement in accordance with the Surveillance Frequency Control Program when the water level is below elevation 6.0 MSL USGS datum (95.0 feet PSE&G datum), and

b.

Measurement at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months when severe storm warnings from the National Weather Service which may impact Artificial Island are in effect.

c.

Measurement in accordance with the Surveillance Frequency Control Program when the water level is equal to or above elevation 6.0 MSL USGS datum (95.0 feet PSE&G datum).

HOPE CREEK 3/47-9 Amendment No. 187

PLANT SYSTEMS 3/4.7.4 REACTOR CORE ISOLATION COOLING SYSTEM LIMITING CONDITION FOR OPERATION 3.7.4 The reactor core isolation cooling (RCIC) system shall be OPERABLE with an OPERABLE flow path capable of automatically taking suction from the suppression pool and transferring the water to the reactor pressure vessel.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3 with reactor steam dome pressure greater than 150 psig.

ACTION:

Note: LCO 3.0.4.b is not applicable to RCIC.

With the RCIC system inoperable, operation may continue provided the HPCI system is OPERABLE; restore the RCIC system to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to less than or equal to 150 psig within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.7.4 The RCIC system shall be demonstrated OPERABLE:

a.

In accordance with the Surveillance Frequency Control Program by:

1.

Verifying by venting at the high point vents that the system piping from the pump discharge valve to the system isolation valve is filled with water.

2.

Verifying that each valve, manual, power operated or automatic in the flow path that is not locked, sealed or otherwise secured in position, is in its correct position.

3.

Verifying that the pump flow controller is in the correct position.

b.

When tested pursuant to the 1ST Program by verifying that the RCIC pump develops a flow of greater than or equal to 600 gpm in the test flow path with a system head corresponding to reactor vessel operating pressure when steam is being supplied to the turbine at 1000 + 20, - 80 psig.*

The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is adequate to perform the test.

HOPE CREEK 3/47-11 Amendment No. 187

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (continued)

c.

In accordance with the Surveillance Frequency Control Program by:

1.

Performing a system functional test which includes simulated automatic actuation and restart# and verifying that each automatic valve in the flow path actuates to its correct position. Actual injection of coolant into the reactor vessel may be excluded.

2.

Verifying that the system will develop a flow of greater than or equal to 600 gpm in the test flow path when steam is supplied to the turbine at a pressure of 150 + 15, - 0 psig. *

3.

Verifying that the suction for the RCIC system is automatically transferred from the condensate storage tank to the suppression pool on a condensate storage tank water level-low signal.

The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is adequate to perform the tests.

Automatic restart on a low water level signal which is subsequent to a high water level trip.

HOPE CREEK 3/47-12 Amendment No. 187

PLANT SYSTEMS 3/4.7.6 SEALED SOURCE CONTAMINATION 3.7.6 Each sealed source containing radioactive material either in excess of 100 microcuries of beta and/or gamma emitting material or 5 microcuries of alpha emitting material shall be free of greater than or equal to 0.005 microcuries of removable contamination.

APPLICABILITY:

At all times.

ACTION:

a.

With a sealed source having removable contamination in excess of the above limit, withdraw the sealed source from use and either:

1.

Decontaminate and repair the sealed source, or

2.

Dispose of the sealed source in accordance with Commission Regulations.

b.

The provisions of Specification 3.0.3 are not applicable.

4.7.6.1 Test Requirements - Each sealed source shall be tested for leakage and/or contamination by:

a.

The licensee, or

b.

Other persons specifically authorized by the Commission or an Agreement State.

The test method shall have a detection sensitivity of at least 0.005 microcuries per test sample.

4.7.6.2 Test Frequencies - Each category of sealed sources, excluding startup sources and fission detectors previously subjected to core flux, shall be tested at the frequency described below.

a.

Sources in use - In accordance with the Surveillance Frequency Control Program for all sealed sources containing radioactive material:

1.

With a half-life greater than 30 days, excluding Hydrogen 3, and

2.

In any form other than gas.

HOPE CREEK 3/47-19 Amendment No. 187

PLANT SYSTEMS 3/4.7.7 MAIN TURBINE BYPASS SYSTEM LIMITING CONDITION FOR OPERATION 3.7.7 The main turbine bypass system shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDITION 1 when THERMAL POWER is greater than or equal to 24% of RATED THERMAL POWER ACTION: With the main turbine bypass system inoperable, restore the system to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than or equal to 24% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.7.7 The main turbine bypass system shall be demonstrated OPERABLE:

a.

In accordance with the Surveillance Frequency Control Program by cycling each turbine bypass valve through at least one complete cycle of full travel, and

b.

In accordance with the Surveillance Frequency Control Program by:

1.

Performing a system functional test which includes simulated automatic actuation and verifying that each automatic valve actuates to its correct position.

2.

Demonstrating TURBINE BYPASS SYSTEM RESPONSE TIME meets the following requirements when measured from the initial movement of the main turbine stop or control valve:

a) 80% of turbine bypass system capacity shall be established in less than or equal to 0.3 second.

b)

Bypass valve opening shall start in less than or equal to 0.1 second.

HOPE CREEK 3/47-21 Amendment No. 187

ELECTRICAL POWER SYSTEMS 4.8.1.1.1 Each of the above required independent circuits between the offsite transmission network and the onsite Class 1 E distribution system shall be:

a.

Determined OPERABLE in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignments and indicated power availability, and

b.

Demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program during shutdown by transferring, manually and automatically, unit power supply from the normal circuit to the alternate circuit.

4.8.1.1.2 Each of the above required diesel generators shall be demonstrated OPERABLE: *

a.

In accordance with the Surveillance Frequency Control Program by:

1.

Verifying the fuel level in the fuel oil day tank.

2.

Verifying the fuel level in the fuel oil storc~ge tank.

3.

Verifying the fuel transfer pump starts and transfers fuel from the storage system to the fuel oil day tank.

4.

Verifying each diesel generator starts** from standby conditions and achieves steady state voltage;::: 3828 and :c;; 4580 volts and frequency of 60 :!: 1.2 Hz.

5.

Verifying the diesel generator is synchronized, loaded to between 4000 and 4400*** kw and operates with this load for at least 60 minutes.

All engine starts and loading for the purpose of this surveillance testing may be preceded by an engine prelube period and/or other warmup procedures recommended by the manufacturer so that mechanical stress and wear on the diesel engine is minimized.

A modified diesel generator start involving idling and gradual acceleration to synchronous speed may be used for this surveillance. When modified start procedures are not used, the time, voltage, and frequency tolerances of Surveillance Requirement 4.8.1.1.2.g must be met.

Momentary transients outside the load range do not invalidate this test.

HOPE CREEK 3/4 8-4 Amendment No. 187

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

6.

Verifying the diesel generator is aligned to provide standby power to the associated emergency busses.

7.

Verifying the pressure in all diesel generator air start receivers to be greater than or equal to 325 psig.

8.

Verifying the lube oil pressure, temperature and differential pressure across the lube oil filters to be within manufacturer's specifications.

b.

In accordance with the Surveillance Frequency Control Program by visually examining a sample of lube oil from the diesel engine to verify absence of water.

c.

In accordance with the Surveillance Frequency Control Program and after each operation of the diesel where the period of operation was greater than or equal to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months by checking for and removing accumulated water from the fuel oil day tank.

d.

In accordance with the Surveillance Frequency Control Program by removing accumulated water from the fuel oil storage tanks.

e.

In accordance with the Surveillance Frequency Control Program by performing a functional test on the emergency load sequencer to verify operability.

f.

In accordance with the surveillance interval specified in the Diesel Fuel Oil Testing Program and prior to the addition of new fuel oil to the storage tank, samples shall be taken to verify fuel oil quality. Sampling and testing of new and stored fuel oil shall be in accordance with the Diesel Fuel Oil Testing Program contained in Specification 6.8.4.e.

HOPE CREEK 3/4 8-5 Amendment No. 187

ELECTRICAL POWER SYSTEMS

g.

In accordance with the Surveillance Frequency Control Program by verifying each diesel generator starts from standby conditions and achieves 2':: 3950 volts and 2':: 58.8 Hz in ::;; 10 seconds after receipt of the start signal, and subsequently achieves steady state voltage 2':: 3828 and::;; 4580 volts and frequency of 60 +/- 1.2 Hz.

h.

In accordance with the Surveillance Frequency Control Program#, during shutdown, by:

1.

Deleted.

2.

Verifying the diesel generator capability to reject a load of greater than or equal to that of the RHR pump motor for each diesel generator while maintaining voltage 2':: 3828 and::;; 4580 volts and frequency at 60 +/- 1.2 Hz.

3.

Verifying the diesel generator capability to reject a load of 4430 kW without tripping. The generator voltage shall not exceed 4785 volts during and following the load rejection.

4.

Simulating a loss of offsite power by itself, and:

a)

Verifying loss of power is detected and deenergization of the emergency busses and load shedding from the emergency busses.

b)

Verifying the diesel generator starts on the auto-start signal, energizes the emergency busses with permanently connected loads within 10 seconds after receipt of the start signal, energizes the autoconnected shutdown loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads. After energization, the steady state voltage and frequency of the emergency busses shall be maintained 2':: 3828 and::;; 4580 volts and 60 +/- 1.2 Hz during this test.

For any start of a diesel generator, the diesel may be loaded in accordance with the manufacturer's recommendations.

HOPE CREEK 3/4 8-6 Amendment No. 187

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

10.

Verifying the diesel generator's capability to:

a)

Synchronize with the offsite power source while the generator is loaded with its emergency loads upon a simulated restoration of offsite power, b)

Transfer its loads to the offsite power source, c)

Be restored to its standby status, and d)

Diesel generator circuit breaker is open.

11.

Verifying that with the diesel generator operating in a test mode and connected to its bus, a simulated ECCS actuation signal overrides the test mode by (1) returning the diesel generator to standby operation, and (2) automatically energizes the emergency loads with offsite power.

12.

Verifying that the fuel oil transfer pump transfers fuel oil from each fuel storage tank to the day tank of each diesel via the installed cross connection lines.

13.

Verifying that the automatic load sequence timer is OPERABLE with the interval between each load block within +/- 10% of its design interval.

14.

Deleted.

i.

In accordance with the Surveillance Frequency Control Program or after any modifications which could affect diesel generator interdependence by starting all diesel generators simultaneously, during shutdown, and verifying that all diesel generators accelerate to at least 514 rpm in less than or equal to 10 seconds.

j.

In accordance with the Surveillance Frequency Control Program by:

1.

Draining each fuel oil storage tank, removing the accumulated sediment and cleaning the tank using a sodium hypochlorite solution or equivalent, and HOPE CREEK 3/4 8-8 Amendment No. 187

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

2.

Performing a pressure test of those portions of the diesel fuel oil system designed to Section III, subsection NO of the ASME Code in accordance with ASME Code Section XI Article IWD-5000.

k.

In accordance with the Surveillance Frequency Control Program# by:

1.

Verifying the diesel generator operates for at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . During the first 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months of this test, the diesel generator shall be loaded to between 4000 and 4400 kW## and during the remaining 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of this test, the diesel generator shall be loaded to between 4652 and 4873 kW. The diesel generator shall achieve ~ 3950 volts and ~ 58.8 Hz in ~ 10 seconds following receipt of the start signal and subsequently achieve steady state voltage ~ 3828 and ~ 4580 volts and frequency of 60 +/- 1.2 Hz.

2.

Within 5 minutes after completing 4.8.1.1.2.k.1, verify each diesel generator starts and achieves ~ 3950 volts and ~ 58.8 Hz in ~ 10 seconds after receipt of the start signal, and subsequently achieves steady state voltage ~ 3828 and ~ 4580 volts and frequency of 60 +/- 1.2 Hz.

-OR-Operate the diesel generator between 4000 kW and 4400 kW for two hours. Within 5 minutes of shutting down the diesel generator, verify each diesel generator starts and achieves ~ 3950 volts and ~ 58.8 Hz in

~ 10 seconds after receipt of the start signal, and subsequently achieves steady state voltage ~ 3828 and::;; 4580 volts and frequency of 60 +/- 1.2 Hz. This test shall continue for at least five minutes.

4.8.1.1.3 Reports - Not used.

4.8.1.1.4 The buried fuel oil transfer piping's cathodic protection system shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by subjecting the cathodic protection system to a performance test.

For any start of a diesel generator, the diesel may be loaded in accordance with manufacturer's recommendations.

Momentary transients outside the load range do not invalidate this test.

HOPE CREEK 3/4 8-9 Amendment No. 187

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS 4.8.2.1 Each of the above required batteries and chargers shall be demonstrated OPERABLE:

a.

In accordance with the Surveillance Frequency Control Program by verifying that:

1.

The parameters in Table 4.8.2.1-1 meet the Category A limits, and

2.

Total battery terminal voltage for each 125-volt battery is greater than or equal to 129 volts on float charge and for each 250-volt battery the terminal voltage is greater than or equal to 258 volts on float charge.

b.

In accordance with the Surveillance Frequency Control Program and within 7 days after a battery discharge with battery terminal voltage below 108 volts for a 125-volt battery or 210 volts for a 250-volt battery, or battery overcharge with battery terminal voltage above 140 volts for a 125-volt battery or 280 volts for a 250-volt battery, by verifying that:

1.

The parameters in Table 4.8.2.1-1 meet the Category B limits,

2.

There is no visible corrosion at either terminals or connectors, or the connection resistance of these items is less than 150 x 10-6 ohms, excluding cable intercell connections, and

3.

The average electrolyte temperature of each sixth cell of connected cells is above 72°F.

c.

In accordance with the Surveillance Frequency Control Program by verifying that:

1.

The cells, cell plates and battery racks show no visual indication of physical damage or abnormal deterioration,

2.

The cell-to-cell and terminal connections are clean, tight, free of corrosion and coated with anti-corrosion material,

3.

The resistance of each cell-to-cell and terminal connection is less than or equal to 150 x 10-6 ohms, excluding cable intercell connections, and

4.

The battery charger will supply the current listed below at the voltage listed below for at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

CHARGER Minimum Voltage CURRENT (AMPERES) 1A0413, 1A0414 129 200 1 B0413, 1 B0414 1C0413,1C0414 1C0444,100414 100444, 100413 100423, 100433 258 50 HOPE CREEK 3/48-13 Amendment No. 187

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (continued)

d.

In accordance with the Surveillance Frequency Control Program, during shutdown, by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status all of the actual or simulated emergency loads for the design duty cycle when the battery is subjected to a battery service test.

e.

In accordance with the Surveillance Frequency Control Program, during shutdown, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. This performance discharge test may be performed in lieu of the battery service test.

f.

At least once per 18 months, during shutdown, performance discharge tests of battery capacity shall be given to any battery that shows signs of degradation or has reached 85% of the service life expected for the application. Degradation is indicated when the battery capacity drops more than 10% of rated capacity from its average on previous performance tests, or is below 90% of the manufacturer's rating. At this once per 18 months interval, this performance discharge test may be performed in lieu of the battery service test.

HOPE CREEK 3/48-14 Amendment No. 187

ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

APPLICABI LlTY:

OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

a.

With one of the above required A.C. distribution system channels not energized, re-energize the channel within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

With one ofthe above required 125 volt D.C. distribution system channels not energized, re-energize the division within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c.

With anyone of the above required 250 volt D.C. distribution systems not energized, declare the associated HPCI or RCIC system inoperable and apply the appropriate ACTION required by the applicable Specifications.

d.

With one or both inverters in one channel inoperable, energize the associated 120 volt A.C. distribution panel(s) within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and restore the inverter(s) to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4.8.3.1 Each of the above required power distribution system channels shall be determined energized in accordance with the Surveillance Frequency Control Program by verifying correct breaker/switch alignment and voltage on the busses/MCCs/panels.

HOPE CREEK 3/48-20 Amendment No. 187

ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

APPLICABILITY:

OPERATIONAL CONDITIONS 4, 5 and *.

ACTION:

a.

With less than two channels of the above required A.C. distribution system energized, suspend CORE ALTERATIONS, handling of recently irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel.

b.

With less than two channels of the above required D.C. distribution system energized, suspend CORE ALTERATIONS, handling of recently irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel.

c.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.8.3.2 At least the above required power distribution system channels shall be determined energized in accordance with the Surveillance Frequency Control Program by verifying correct breaker/switch alignment and voltage on the busses/MCCs/panels.

When handling recently irradiated fuel in the secondary containment.

HOPE CREEK 3/48-23 Amendment No. 187

ELECTRICAL POWER SYSTEMS PRIMARY CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES LIMITING CONDITION FOR OPERATION 3.8.4.1 All primary containment penetration conductor overcurrent protective devices shown in Table 3.8.4.1-1 shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

a.

With one or more of the primary containment penetration conductor over current protective devices shown in Table 3.8.4.1-1 inoperable, declare the affected system or component inoperable and apply the appropriate ACTION statement for the affected system, and

1.

For 4.16 kV circuit breakers, de-energize the 4.16 kV circuit(s) by tripping the associated redundant circuit breaker(s) within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and verify the redundant circuit breaker to be tripped at least once per 7 days thereafter.

2.

For 480 volt circuit breakers, remove the inoperable circuit breaker(s) from service by disconnecting* the breaker within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and verify the inoperable breaker(s) to be disconnected at least once per 7 days thereafter.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4.8.4.1 Each of the primary containment penetration conductor overcurrent protective devices shown in Table 3.8.4.1-1 shall be demonstrated OPERABLE:

a.

In accordance with the Surveillance Frequency Control Program:

1.

By verifying that each of the medium voltage 4.16 kV circuit breakers are OPERABLE by performing:

a)

A CHANNEL CALIBRATION of the associated protective relays, and b)

An integrated system functional test which includes simulated automatic actuation of the system and verifying that each relay and associated circuit breakers and overcurrent control circuits function as designed.

After being disconnected, these breakers shall be maintained disconnected under administrative control.

HOPE CREEK 3/48-24 Amendment No. 187

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

2.

By selecting and functionally testing a representative sample of at least 10% of each type of lower voltage circuit breakers. Circuit breakers selected for functional testing shall be selected on a rotating basis.

Testing of these circuit breakers shall consist of injecting a current with a value between 150% and 300% of the pickup of the long time delay trip element and verifying that the circuit breaker operates within the time delay bandwidth for that current specified by the manufacturer. The instantaneous element shall be tested by injecting a current in excess of 120% the pickup value of the element and verifying that the circuit breaker trips instantaneously with no intentional time delay. Molded case circuit breaker testing shall also follow this procedure except that generally no more than two trip elements, time delay and instantaneous, will be involved. Circuit breakers found inoperable during functional testing shall be restored to OPERABLE status prior to resuming operation. For each circuit breaker found inoperable during these functional tests, an additional representative sample of at least 10% of all the circuit breakers of the inoperable type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested.

b.

In accordance with the Surveillance Frequency Control Program by subjecting each circuit breaker to an inspection and preventive maintenance in accordance with procedures prepared in conjunction with its manufacturer's recommendations.

HOPE CREEK 3/48-25 Amendment No. 187

ELECTRICAL POWER SYSTEMS MOTOR OPERATED VALVES - THERMAL OVERLOAD PROTECTION (BYPASSED)

LIMITING CONDITION FOR OPERATION 3.8.4.2 The thermal overload protection bypass circuit of each motor operated valve (MOV) required to have thermal overload protection shall be OPERABLE.

APPLICABILITY:

Whenever the MOV is required to be OPERABLE.

ACTION:

With the thermal overload protection bypass circuit for one or more of the above required MOVs inoperable, restore the inoperable thermal overload protection bypass circuit(s) to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or declare the affected MOV(s) inoperable and apply the appropriate ACTION statement(s) for the affected system(s).

SURVEILLANCE REQUIREMENTS 4.8.4.2.1 The thermal overload protection bypass circuit for each of the above required MOVs shall be demonstrated OPERABLE:

a.

In accordance with the Surveillance Frequency Control Program by the performance of a CHANNEL FUNCTIONAL TEST for:

1.

Those thermal overload protection devices which are normally in force during plant operation and bypassed only under accident conditions.

2.

A representative sample of at least 25% of those thermal overload protection devices which are bypassed continuously and temporarily placed in force only when the MOVs are undergoing periodic or maintenance testing.

3.

A representative sample of at least 25% of those thermal overload protection devices which are in force during normal manual (momentary push button contact) MOV operation and bypassed during remote manual (push button held depressed) MOV operation.

b.

Following maintenance on the motor starter.

4.8.4.2.2 The thermal overload protection for the above required MOVs which are continuously bypassed and temporarily placed in force only when the MOV is undergoing periodic or maintenance testing shall be verified to be continuously bypassed following such testing.

HOPE CREEK 3/48-30 Amendment No. 187

ELECTRICAL POWER SYSTEMS MOTOR OPERATED VALVES - THERMAL OVERLOAD PROTECTION (NOT BYPASSED)

LIMITING CONDITION FOR OPERATION 3.8.4.3 The thermal overload protection of each motor operated valve (MOV) shown in Table 3.8.4.3-1 shall be OPERABLE.

APPLICABILITY:

Whenever the MOV is required to be OPERABLE.

ACTION:

With the thermal overload protection for one or more of the above required MOVs inoperable, restore the inoperable thermal overload{s) to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or declare the affected MOV(s) inoperable and apply the appropriate ACTION statement(s) for the affected system{s).

SURVEILLANCE REQUIREMENTS 4.8.4.3 The thermal overload protection for each of the above required MOVs shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program and following maintenance on the motor starter by the performance of a CHANNEL CALIBRATION.

HOPE CREEK 3/48-38 Amendment No. 187

ELECTRICAL POWER SYSTEMS REACTOR PROTECTION SYSTEM ELECTRICAL POWER MONITORING LIMITING CONDITION FOR OPERATION 3.8.4.4 Two RPS electric power monitoring channels for each inservice RPS MG set or altemate power supply shall be OPERABLE.

APPLICABI LlTY:

At all times.

ACTION:

a.

With one RPS electric power monitoring channel for an inservice RPS MG set or alternate power supply inoperable, restore the inoperable power monitoring channel to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or remove the associated RPS MG set or alternate power supply from service.

b.

With both RPS electric power monitoring channels for an inservice RPS MG set or alternate power supply inoperable, restore at least one electric power monitoring channel to OPERABLE status within 30 minutes or remove the associated RPS MG set or alternate power supply from service.

4.8.4.4 The above specified RPS electric power monitoring channels shall be determined OPERABLE:

a.

By performance of a CHANNEL FUNCTIONAL TEST each time the plant is in COLD SHUTDOWN for a period of more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , unless performed in the previolJs 6 months.

b.

In accordance with the Surveillance Frequency Control Program by demonstrating the OPERABILITY of over-voltage, under-voltage, and under frequency protective instrumentation by performance of a CHANNEL CALIBRATION including simulated automatic actuation of the protective relays, tripping logic and output circuit breakers and verifying the following setpoints.

1.

Over-voltage s 132 VAC, (Bus A), 132 VAC (Bus B)

2.

Under-voltage ~ 108 VAC, (Bus A), 108 VAC (Bus B)

3.

Under-frequency ~ 57 Hz. (Bus A and Bus B)

HOPE CREEK 3/48-40 Amendment No. 187

ELECTRICAL POWER SYSTEMS CLASS 1 E ISOLATION BREAKER OVERCURRENT PROTECTIVE DEVICES LIMITING CONDITION FOR OPERATION 3.8.4.5 All Class 1 E isolation breaker (tripped by a LOCA signal) overcurrent protective devices shown in Table 3.8.4.5-1 shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

a.

With one or more of the overcurrent protective devices shown in Table 3.8.4.5-1 inoperable, declare the affected isolation breaker inoperable and remove the inoperable circuit breaker(s) from service within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and verify the inoperable breaker(s) to be disconnected at least once per 7 days thereafter.

SURVEILLANCE REQUIREMENTS 4.8.4.5 Each of the Class 1 E isolation breaker overcurrent protective devices shown in Table 3.8.4.5-1 shall be demonstrated OPERABLE:

a.

In accordance with the Surveillance Frequency Control Program:

By selecting and functionally testing a representative sample of at least 10% of each type of lower voltage circuit breakers. Circuit breakers selected for functional testing shall be selected on a rotating basis. Testing of these circuit breakers shall consist of injecting a current with a value between 150% and 300% of the pickup of the long time delay trip element and a value between 150% and 250% of the pickup of the short time delay, and verifying that the circuit breaker operates within the time delay band width for that current specified by the manufacturer. The instantaneous element shall be tested by injecting a current in excess of 120% of the pickup value of the element and verifying that the circuit breaker trips instantaneously with no intentional time delay. Molded case circuit breaker testing shall also follow this procedure except that generally no more than two trip elements, time delay and instantaneous, will be involved.

For circuit breakers equipped with solid state trip devices, the functional testing may be performed with use of portable instruments designed to verify the time current characteristics and pickup calibration of the trip elements. Circuit breakers found inoperable during functional testing shall be restored to OPERABLE status prior to resuming operation. For each circuit breaker found inoperable during these functional tests, an additional representative sample of at least 10% of all the circuit breakers of the inoperable type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested.

b.

In accordance with the Surveillance Frequency Control Program by subjecting each circuit breaker to an inspection and preventive maintenance in accordance with procedures prepared in conjunction with its manufacturer's recommendations.

HOPE CREEK 3/48-41 Amendment No. 187

ELECTRICAL POWER SYSTEM POWER RANGE NEUTRON MONITORING SYSTEM ELECTRICAL POWER MONITORING LIMITING CONDITION FOR OPERATION 3.8.4.6 The power range neutron monitoring system (NMS) electric power monitoring channels for each inservice power range NMS power supply shall be OPERABLE.

APPLICABI LlTY:

At all times.

ACTION:

a.

With one power range NMS electric power monitoring channel for an inservice power range NMS power supply inoperable, restore the inoperable power monitoring channel to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or deenergize the associated power range NMS power supply feeder circuit.

b.

With both power range NMS electric power monitoring channels for an inservice power range NMS power supply inoperable, restore at least one electric power monitoring channel to OPERABLE status within 30 minutes or deenergize the associated power range NMS power supply feeder circuit.

4.8.4.6 The above specified power range NMS electric power monitoring channels shall be determined OPERABLE:

a.

By performance of a CHANNEL FUNCTIONAL TEST each time the plant is in COLD SHUTDOWN for a period of more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , unless performed in the previous 6 months.

b.

In accordance with the Surveillance Frequency Control Program by demonstrating the OPERABILITY of over-voltage, under-voltage, and under frequency protective instrumentation by performance of a CHANNEL CALIBRATION including simulated automatic actuation of the protective relays, tripping logic and output circuit breakers and verifying the following setpoints.

1.

Over-voltage:s; 132 VAC (BUS A), 132 VAC (BUS B)

2.

Under-voltage ~ 108 VAC (BUS A), 108 VAC (BUS B)

3.

Under-frequency ~ 57 Hz. -0, +2%

HOPE CREEK 3/48-44 Amendment No. 187

REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS 4.9.1.1 The reactor mode switch shall be verified to be locked in the Shutdown or Refuel position as specified:

a.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to:

1.

Beginning CORE ALTERATIONS, and

2.

Resuming CORE ALTERATIONS when the reactor mode switch has been unlocked.

b.

In accordance with the Surveillance Frequency Control Program.

4.9.1.2 Each of the above required reactor mode switch Refuel position interlocks* shall be demonstrated OPERABLE by performance of a CHANNEL FUNCTIONAL TEST within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to the start of and in accordance with the Surveillance Frequency Control Program during control rod withdrawal or CORE ALTERATIONS, as applicable.

4.9.1.3 Each of the above required reactor mode switch Refuel position interlocks* that is affected shall be demonstrated OPERABLE by performance of a CHANNEL FUNCTIONAL TEST prior to resuming control rod withdrawal or CORE ALTERATIONS, as applicable, following repair, maintenance or replacement of any component that could affect the Refuel position interlock.

The reactor mode switch may be placed in the Run or Startup/Hot Standby position to test the switch interlock functions provided that all control rods are verified to remain fully inserted by a second licensed operator or other technically qualified member of the unit technical staff.

HOPE CREEK 3/4 9-2 Amendment No. 187

REFUELING OPERATIONS 3/4.9.2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 At least 2 source range monitor* (SRM) channels shall be OPERABLE and inserted to the normal operating level with:##

a.

Annunciation and continuous visual indication in the control room,

b.

One of the required SRM detectors located in the quadrant where CORE ALTERATIONS are being performed and the other required SRM detector located in an adjacent quadrant, and

c.

Unless adequate shutdown margin has been demonstrated per Specification 3.1.1, the "shorting links" removed from the RPS circuitry prior to and during the time any control rod is withdrawn.#

d.

During a SPIRAL UNLOAD, the count rate may drop below 3 cps when the number of assemblies remaining in the core drops to sixteen or less.

e.

During a SPIRAL RELOAD, up to four fuel assemblies may be loaded in the four bundle locations immediately surrounding each of the four SRMs prior to obtaining 3 cps. Until these assemblies have been loaded, the 3 cps count rate is not required.

APPLICABILITY: OPERATIONAL CONDITION 5.

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS and fully insert all insertable control rods.

4.9.2 Each of the above required SRM channels shall be demonstrated OPERABLE by:

a.

In accordance with the Surveillance Frequency Control Program:

1.

Performance of a CHANNEL CHECK, The use of special movable detectors during CORE ALTERATIONS in place of the normal SRM nuclear detectors is permissible as long as these special detectors are connected to the normal SRM circuits.

Not required for control rods removed per Specification 3.9.10.1 and 3.9.10.2.

Three SRM channels shall be OPERABLE for critical shutdown margin demonstrations.

An SRM detector may be retracted provided a channel indication of at least 100 cps is maintained.

HOPE CREEK 3/4 9-3 Amendment No. 187

REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued)

2.

Verifying the detectors are inserted to the normal operating level, and

3.

During CORE ALTERATIONS, verifying that the detector of an OPERABLE SRM channel is located in the core quadrant where CORE ALTERATIONS are being performed and another is located in an adjacent quadrant.

b.

Performance of a CHANNEL FUNCTIONAL TEST in accordance with the Surveillance Frequency Control Program.

c.

Verifying that the channel count rate is at least 3 cps.

1.

Prior to control rod withdrawal,

2.

Prior to and in accordance with the Surveillance Frequency Control Program during CORE ALTERATIONS***, and

3.

In accordance with the Surveillance Frequency Control Program***.

d.

Unless adequate shutdown margin has been demonstrated per Specification 3.1.1, verifying that the RPS circuitry "shorting links" have been removed, within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to and in accordance with the Surveillance Frequency Control Program during the time any control rod is withdrawn.**

Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.

Except as noted in Specifications 3.9.2.d and 3.9.2.e.

HOPE CREEK 3/4 9-4 Amendment No. 187

REFUELING OPERATIONS 3/4.9.3 CONTROL ROD POSITION LIMITING CONDITION FOR OPERATION 3.9.3 All control rods shall be inserted.*

APPLICABILITY:

OPERATIONAL CONDITION 5, during CORE ALTERATIONS. **

ACTION:

With all control rods not inserted, suspend all other CORE ALTERATIONS, except that one control rod may be withdrawn under control of the reactor mode switch Refuel position one-rod out interlock.

SURVEILLANCE REQUIREMENTS 4.9.3 All control rods shall be verified to be inserted, except as above specified:

a.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to:

1.

The start of CORE ALTERATIONS.

2.

The withdrawal of one control rod under the control of the reactor mode switch Refuel position one-rod-out interlock.

b.

In accordance with the Surveillance Frequency Control Program.

Except control rods removed per Specification 3.9.10.1 or 3.9.10.2.

See Special Test Exception 3.10.3.

HOPE CREEK 3/4 9-5 Amendment No. 187

REFUELING OPERATIONS 3/4.9.8 WATER lEVEL - REACTOR VESSEL LIMITING CONDITION FOR OPERATION 3.9.8 At least 22 feet 2 inches of water shall be maintained over the top of the reactor pressure vessel flange.

APPUCABI LlTY:

During handling of fuel assemblies or control rods within the reactor pressure vessel while in OPERATIONAL CONDITION 5 when the fuel assemblies being handled are irradiated or the fuel assemblies seated within the reactor vessel are irradiated.

ACTION:

With the requirements of the above specification not satisfied, suspend all operations involving handling of fuel assemblies or control rods within the reactor pressure vessel after placing all fuel assemblies and control rods in a safe condition.

4.9.8 The reactor vessel water level shall be determined to be at least its minimum required depth within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to the start of and in accordance with the Surveillance Frequency Control Program during handling of fuel assemblies or control rods within the reactor pressure vessel.

HOPE CREEK 3/4 9-11 Amendment No. 187

REFUELING OPERATIONS 3/4.9.9 WATER LEVEL - SPENT FUEL STORAGE POOL LIMITING CONDITION FOR OPERATION 3.9.9 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the spent fuel storage pool racks.

APPLICABILITY:

Whenever irradiated fuel assemblies are in the spent fuel storage pool.

ACTION:

With the requirements of the above specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the spent fuel storage pool area after placing the fuel assemblies and crane load in a safe condition. The provisions of Specification 3.0.3 are not applicable.

4.9.9 The water level in the spent fuel storage pool shall be determined to be at least at its minimum required depth in accordance with the Surveillance Frequency Control Program.

HOPE CREEK 3/49-12 Amendment No. 187

REFUELING OPERATIONS SU..BVEILLANCE REQUIREMENTS 4.9.10.1 Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to the start of removal of a control rod and/or the associated control rod drive mechanism from the core and/or reactor pressure vessel and in accordance with the Surveillance Frequency Control Program thereafter until a control rod and associated control rod drive mechanism are reinstalled and the control rod is inserted in the core, verify that:

a.

The reactor mode switch is OPERABLE per Surveillance Requirement 4.3.1.1 or 4.9.1.2, as applicable, and locked in the Shutdown position or in the Refuel position with the "one rod out" Refuel position interlock OPERABLE per Specification 3.9.1.

b.

The SRM channels are OPERABLE per Specification 3.9.2.

c.

The SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied per Specification 3.9.1 0.1.c.

d.

All other control rods in a five-by-five array centered on the control rod being removed are inserted and electrically or hydraulically disarmed or the four fuel assemblies surrounding the control rod or control rod drive mechanism to be removed from the core and/or reactor vessel are removed from the core cell.

e.

All other control rods are inserted.

f.

All fuel loading operations are suspended.

HOPE CREEK 3/49-14 Amendment No. 187

REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS 4.9.10.2.1 Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to the start of removal of control rods and/or control rod drive mechanisms from the core and/or reactor pressure vessel and in accordance with the Surveillance Frequency Control Program thereafter until all control rods and control rod drive mechanisms are reinstalled and all control rods are inserted in the core, verify that:

a.

The reactor mode switch is OPERABLE per Surveillance Requirement 4.3.1.1 or 4.9.1.2, as applicable, and locked in the Shutdown position or in the Refuel position per Specification 3.9.1.

b.

The SRM channels are OPERABLE per Specification 3.9.2.

c.

The SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied.

d.

All other control rods are either inserted or have the surrounding four fuel assemblies removed from the core cell.

e.

The four fuel assemblies surrounding each control rod and/or control rod drive mechanism to be removed from the core and/or reactor vessel are removed from the core cell.

f.

All fuel loading operations are suspended.

4.9.10.2.2 Following replacement of all control rods and/or control rod drive mechanisms removed in accordance with this specification, perform a functional test of the "one-rod-out" Refuel position interlock, if this function had been bypassed.

HOPE CREEK 3/49-16 Amendment No. 187

REFUELING OPERATIONS 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION HIGH WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.11.1 At least one shutdown cooling mode loop of the residual heat removal (RHR) system shall be OPERABLE and in operation* with:

a.

One OPERABLE RHR pump, and

b.

One OPERABLE RHR heat exchanger.

APPLICABILITY: OPERATIONAL CONDITION 5, when irradiated fuel is in the reactor vessel and the water level is greater than or equal to 22 feet 2 inches above the top of the reactor pressure vessel flange and heat losses to ambient** are not sufficient to maintain OPERATIONAL CONDITION 5.

ACTION:

a.

With no RHR shutdown cooling mode loop OPERABLE, within one hour and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, demonstrate the operability of at least one alternate method capable of decay heat removal. Otherwise, suspend all operations involving an increase in the reactor decay heat load and establish SECONDARY CONTAINMENT INTEGRITY within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b.

With no RHR shutdown cooling mode loop in operation, within one hour establish reactor coolant circulation by an alternate method and monitor reactor coolant temperature at least once per hour.

SURVEILLANCE REQUIREMENTS 4.9.11.1 At least one shutdown cooling mode loop of the residual heat removal system or alternate method shall be verified to be in operation and Circulating reactor coolant in accordance with the Surveillance Frequency Control Program.

The shutdown cooling pump may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8-hour period.

Ambient losses must be such that no increase in reactor vessel water temperature will occur (even though REFUELING conditions are being maintained).

HOPE CREEK 3/49-17 Amendment No. 187

REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.11.2 Two shutdown cooling mode loops of the residual heat removal (RHR) system shall be OPERABLE and at least one loop shall be in operation,

with each loop consisting of:

a.

One OPERABLE RHR pump, and

b.

One OPERABLE RHR heat exchanger.

APPLICABILITY:

OPERATIONAL CONDITION 5, when irradiated fuel is in the reactor vessel and the water level is less than 22 feet 2 inches above the top of the reactor pressure vessel flange and heat losses to ambient** are not sufficient to maintain OPERATIONAL CONDITION 5.

ACTION:

a.

With less than the above required shutdown cooling mode loops of the RHR system OPERABLE, within one hour and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, demonstrate the OPERABILITY of at least one alternate method capable of decay heat removal for each inoperable RHR shutdown cooling mode loop.

b.

With no RHR shutdown cooling mode loop in operation, within one hour establish reactor coolant circulation by an alternate method and monitor reactor coolant temperature at least once per hour.

SURVEILLANCE REQUIREMENTS 4.9.11.2 At least one shutdown cooling mode loop of the residual heat removal system or alternate method shall be verified to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program.

The shutdown cooling pump may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8-hour period.

Ambient losses must be such that no increase in reactor vessel water temperature will occur (even though REFUELING conditions are being maintained).

HOPE CREEK 3/49-18 Amendment No. 187

3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY 3.10.1 The provisions of Specifications 3.6.1.1, 3.6.1.3 and 3.9.1 and Table 1.2 may be suspended to permit the reactor pressure vessel closure head and the drywell head to be removed and the primary containment air lock doors to be open when the reactor mode switch is in the Startup position during low power PHYSICS TESTS with THERMAL POWER less than 1% of RATED THERMAL POWER and reactor coolant temperature less than 200°F.

APPLICABI LlTY:

OPERATIONAL CONDITION 2, during low power PHYSICS TESTS.

ACTION:

With THERMAL POWER greater than or equal to 1 % of RATED THERMAL POWER or with the reactor coolant temperature greater than or equal to 200°F, immediately place the reactor mode switch in the Shutdown position.

4.10.1 The THERMAL POWER and reactor coolant temperature shall be verified to be within the limits in accordance with the Surveillance Frequency Control Program during low power PHYSICS TESTS.

HOPE CREEK 3/4 10-1 Amendment No. 187

SPECIAL TEST EXCEPTIONS 3/410.3 SHUTDOWN MARGIN DEMONSTRATIONS LIMITING CONDITION FOR OPERATION 3.10.3 The provisions of Specification 3.9.1, Specification 3.9.3 and Table 1.2 may be suspended to permit the reactor mode switch to be in the Startup position and to allow more than one control rod to be withdrawn for shutdown margin demonstration, provided that at least the following requirements are satisfied.

a.

The source range monitors are OPERABLE with the RPS circuitry "shorting links" removed per Specification 3.9.2.

b.

The rod worth minimizer is OPERABLE per Specification 3.1.4.1 and is programmed for the shutdown margin demonstration, or conformance with the shutdown margin demonstration procedure is verified by a second licensed operator or other technically qualified member of the unit technical staff.

c.

The "rod-out-notch-override" control shall not be used during out-of-sequence movement of the control rods.

d.

No other CORE ALTERATIONS are in progress.

APPLICABILITY:

OPERATIONAL CONDITION 5, during shutdown margin demonstrations.

ACTION:

With the requirements of the above specification not satisfied, immediately place the reactor mode switch in the Shutdown or Refuel position.

SURVEILLANCE REQUIREMENTS 4.10.3 Within 30 minutes prior to and in accordance with the Surveillance Frequency Control Program during the performance of a shutdown margin demonstration, verify that;

a.

The source range monitors are OPERABLE per SpeCification 3.9.2,

b.

The rod worth minimizer is OPERABLE with the required program per Specification 3.1.4.1 or a second licensed operator or other technically qualified member of the unit technical staff is present and verifies compliance with the shutdown demonstration procedures, and

c.

No other CORE ALTERATIONS are in progress.

HOPE CREEK 3/4 10-3 Amendment No. 187

SPECIAL TEST EXCEPTIONS 3/4.10.4 RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION 3.10.4 The requirements of Specifications 3.4.1.1 and 3.4.1.3 that recirculation loops be in operation with matched pump speed may be suspended for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for the performance of:

a.

PHYSICS TESTS, provided that THERMAL POWER does not exceed 5% of RATED THERMAL POWER.

APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2, during PHYSICS TESTS.

ACTION:

a.

With the above specified time limit exceeded, insert all control rods.

b.

With the above specified THERMAL POWER limit exceeded during PHYSICS TESTS, immediately place the reactor mode switch in the Shutdown position.

SURVEILLANCE REQUIREMENTS 4.10.4.1 The time during which the above specified requirement has been suspended shall be verified to be less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in accordance with the Surveillance Frequency Control Program during PHYSICS TESTS.

4.10.4.2 THERMAL POWER shall be determined to be less than 5% of RATED THERMAL POWER in accordance with the Surveillance Frequency Control Program during PHYSICS TESTS.

HOPE CREEK Amendment No. 187

SPECIAL TEST EXCEPTIONS 3/4.10.6 TRAINING STARTUPS 3.10.6 The provisions of Specification 3.5.1 may be suspended to permit one RHR subsystem to be aligned in the shutdown cooling mode during training startups provided that the reactor vessel is not pressurized, THERMAL POWER is less than or equal to 1% of RATED THERMAL POWER and reactor coolant temperature is less than 200°F.

APPLICABILITY:

OPERATIONAL CONDITION 2, during training startups.

ACTION:

With the requirements of the above specification not satisfied, immediately place the reactor mode switch in the Shutdown position.

4.10.6 The reactor vessel shall be verified to be unpressurized and the THERMAL POWER and reactor coolant temperature shall be verified to be within the limits in accordance with the Surveillance Frequency Control Program during training startups.

HOPE CREEK 3/4 10-6 Amendment No. 187

RADIOACTIVE EFFLUENTS LIQUID HOLDUP TANKS LlMITINGgONDITION FOR OPERATION 3.11.1 A The quantity of radioactive material contained in any outside temporary tank shall be limited to less than or equal to 10 curies, excluding tritium and dissolved or entrained noble gases.

APPLICABILITY:

At all times.

ACTION:

a.

With the quantity of radioactive material in any of the above tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank, within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit, and describe the events leading to this condition in the next Radioactive Effluent Release Report, pursuant to Specification 6.9.1.7.

b.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1 A The quantity of radioactive material contained in each of the above tanks shall be determined to be within the above limit by analyzing a representative sample of the tank's contents in accordance with the Surveillance Frequency Control Program when radioactive materials are being added to the tank.

HOPE CREEK 3/4 11-2 Amendment No. 187

RADIOACTIVE EFFLUENTS MAIN CONDENSER LIMITING CONDITION FOR OPERATION 3.11.2.7 The radioactivity rate of noble gases measured at the recombiner after-condenser discharge shall be limited to less than or equal to 3.30 E+5 microcurieslsec after 30 minute decay.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2* and 3*.

ACTION:

With the radioactivity rate of noble gases at the recombiner after-condenser discharge exceeding 3.30 E+5 microcurieslsec after 30 minute decay, restore the radioactivity rate to within its limit within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.11.2.7.1 The radioactivity rate of noble gases at the recombiner after-condenser discharge shall be continuously monitored in accordance with Specification 3.3.7.1.

4.11.2.7.2 The radioactivity rate of noble gases from the recombiner after-condenser discharge shall be determined to be within the limits of Specification 3.11.2.7 at the following frequencies by performing an isotopic analysis of a representative sample of gases taken near the discharge of the main condenser air ejector:

a.

In accordance with the Surveillance Frequency Control Program.

b.

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following an increase, as indicated by the Offgas Pretreatment Radiation Monitor, of greater than 50%, after factoring out increases due to changes in THERMAL POWER level, in the nominal steady-state 'fission gas release from the primary coolant.

c.

The provisions of Specification 4.0.4 are not applicable.

When the main condenser air ejector is in operation.

HOPE CREEK 3/4 11-17 Amendment No. 187

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) 6.8.4.i INSERVICE TESTING PROGRAM This Program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program shall include the following:

a.

Testing frequencies applicable to the ASME Code for Operations and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as follows:

ASME OM Code and applicable Required Frequencies for Addenda terminology for performing inservice inservice testing activities testing activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

b.

The provisions of Specification 4,0.2 are applicable to the above required frequencies and to other normal and accelerated frequencies specified as 2 years or less in the Inservice Testing Program for performing inservice testing activities,

c.

The provisions of Specification 4.0.3 are applicable to inservice testing activities, and d,

Nothing in the ASME OM Code shall be construed to supersede the requirements of any Technical Specification.

6.8.4,j Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a.

The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.

b.

Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.

c.

The provisions of Surveillance Requirements 4.0.2 and 4.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.

HOPE CREEK 6-16e Amendment No. 187

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 187 TO FACILITY OPERATING LICENSE NO. NPF-57 PSEG NUCLEAR LLC HOPE CREEK GENERATING STATION DOCKET NO. 50-354

1.0 INTRODUCTION

By letter dated March 19, 2010, as supplemented by letters dated July 28,2010, and January 10, 2011, (Agencywide Documents Access and Management System (ADAMS)

Accession Nos. ML100900224, ML102230417, and ML110200059, respectively), PSEG Nuclear LLC (PSEG or the licensee) submitted a request for changes to the Hope Creek Generating Station (HCGS) Technical Specifications (TSs). The supplements dated July 28,2010, and January 10, 2011, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC or the Commission) staff's original proposed no significant hazards consideration determination as published in the Federal Register (FR) on June 15, 2010 (75 FR 33842),

The requested change is the adoption of NRC-approved TS Task Force (TSTF) change TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control - RITSTF

[Risk-Informed TSTF] Initiative 5b" (Reference 1). When implemented, TSTF-425 relocates most periodic frequencies of TS surveillances to a licensee-controlled program, the Surveillance Frequency Control Program (SFCP), and provides requirements for the new program in the Administrative Controls section of the TSs. All surveillance frequencies can be relocated except:

Frequencies that reference other approved programs for the specific interval (such as the Inservice Testing Program or the Primary Containment Leakage Rate Testing Program);

Frequencies that are purely event-driven (e.g" "each time the control rod is withdrawn to the

'full out' position");

Frequencies that are event-driven, but have a time component for performing the surveillance on a one-time basis once the event occurs (e.g., "within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after thermal power reaching ~ 95% RTP [rated thermal power]"); and

Frequencies that are related to specific conditions (e.g., battery degradation, age and capacity) or conditions for the performance of a surveillance requirement (e.g., "drywell to suppression chamber differential pressure decrease").

A new program would be added to the Administrative Controls in TS Section 6.0 as TS 6.8A.j.

The new program is called the SFCP and describes the requirements for the program to control Enclosure

- 2 changes to the relocated surveillance frequencies. The proposed licensee changes to the Administrative Controls of the TSs to incorporate the SFCP include a specific reference to Nuclear Energy Institute (NEI) 04-10, "Risk-Informed Technical Specifications Initiative 58, Risk Informed Method for Control of Surveillance Frequencies," Revision 1 (Reference 2), as the basis for making any changes to the surveillance frequencies once they are relocated out of the TSs.

In a letter dated September 19, 2007 (Reference 9), the NRC staff approved NEI 04-10, Revision 1, as acceptable for referencing by licensees proposing to amend their TSs to establish a SFCP, to the extent specified and under the limitations delineated in NEI 04-10, and in the NRC staff's safety evaluation (SE) providing the basis for its acceptance of NEI 04-10.

The NRC staff issued a Notice of Availability for TSTF-425, Revision 3, in the Federal Register on July 6,2009 (74 FR 31996). The notice included a model SE. In its application dated March 19, 2010, the licensee stated that "PSEG has concluded that the justifications presented in the TSTF proposal and the safety evaluation prepared by the NRC staff are applicable to HCGS." The SE that follows is based, in large part, on the model SE for TSTF-425.

2.0 REGULATORY EVALUATION

In the "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors," published in the Federal Register on July 22,1993 (58 FR 39132), the NRC addressed the use of Probabilistic Safety Analysis (PSA, currently referred to as Probabilistic Risk Assessment (PRA)) in determining the content of the TSs. On page 39135 of this FR publication the Commission states, in part, that:

The Commission believes that it would be inappropriate at this time to allow requirements which meet one or more of the first three criteria [in Title 10 of the Code of Federal Regulations (10 CFR), Section 50.36(c)(2)(ii)] to be deleted from Technical Specifications based solely on PSA (Criterion 4). However, if the results of PSA indicate that Technical Specifications can be relaxed or removed, a deterministic review will be performed.....

The Commission Policy in this regard is consistent with its Policy Statement on "Safety Goals for the Operation of Nuclear Power Plants," 51 FR 30028, published on August 21, 1986. The Policy Statement on Safety Goals states in part, "* *

probabilistic results should also be reasonably balanced and supported through use of deterministic arguments. In this way, judgments can be made * *

about the degree of confidence to be given these [probabilistic]

estimates and assumptions. This is a key part of the process of determining the degree of regulatory conservatism that may be warranted for particular decisions.

This defense-in-depth approach is expected to continue to ensure the protection of public health and safety."....

The Commission will continue to use PSA, consistent with its policy on Safety Goals, as a tool in evaluating specific line-item improvements to Technical SpeCifications, new requirements, and industry proposals for risk-based Technical Specification changes.

- 3 Approximately two years later, the NRC provided additional detail concerning the use of PRA in the "Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities; Final Policy Statement," published in the Federal Register on August 16,1995 (60 FR 42622). On page 42627 of this FR publication, the Commission states, in part, that:

PRA addresses a broad spectrum of initiating events by assessing the event frequency. Mitigating system reliability is then assessed, including the potential for multiple and common-cause failures. The treatment therefore goes beyond the single failure requirements in the deterministic approach. The probabilistic approach to regulation is, therefore, considered an extension and enhancement of traditional regulation by considering risk in a more coherent and complete manner.

On pages 42628 and 42629 of this FR publication, the Commission provided its policy on use of PRA which states:

Although PRA methods and information have thus far been used successfully in nuclear regulatory activities, there have been concerns that PRA methods are not consistently applied throughout the agency, that sufficient agency PRAIstatistics expertise is not available, and that the Commission is not deriving full benefit from the large agency and industry investment in the developed risk assessment methods. Therefore, the Commission believes that an overall policy on the use of PRA in nuclear regulatory activities should be established so that the many potential applications of PRA can be implemented in a consistent and predictable manner that promotes regulatory stability and efficiency. This policy statement sets forth the Commission's intention to encourage the use of PRA and to expand the scope of PRA applications in all nuclear regulatory matters to the extent supported by the state-of-the-art in terms of methods and data. Implementation of the policy statement will improve the regulatory process in three areas:

Foremost, through safety decision making enhanced by the use of PRA insights; through more efficient use of agency resources; and through a reduction in unnecessary burdens on licensees.

Therefore, the Commission adopts the following policy statement regarding the expanded NRC use of PRA:

(1) The use of PRA technology should be increased in all regulatory matters to the extent supported by the state-of-the-art in PRA methods and data and in a manner that complements the NRC's deterministic approach and supports the NRC's traditional defense-in-depth philosophy.

(2) PRA and associated analyses (e.g., sensitivity studies, uncertainty analyses, and importance measures) should be used in regulatory matters, where practical within the bounds of the state-of-the-art, to reduce unnecessary conservatism associated with current regulatory requirements, regulatory guides, license commitments, and staff practices. Where appropriate, PRA should be used to support the proposal for additional regulatory requirements in accordance with 10 CFR 50.109 (8ackfit Rule). Appropriate procedures for

- 4 including PRA in the process for changing regulatory requirements should be developed and followed. It is, of course, understood that the intent of this policy is that existing rules and regulations shall be complied with unless these rules and regulations are revised.

(3) PRA evaluations in support of regulatory decisions should be as realistic as practicable and appropriate supporting data should be publicly available for review.

(4) The Commission's safety goals for nuclear power plants and subsidiary numerical objectives are to be used with appropriate consideration of uncertainties in making regulatory judgments on the need for proposing and backfitting new generic requirements on nuclear power plant licensees' The Commission's regulatory requirements related to the content of the TSs are set forth in 10 CFR 50.36, "Technical specifications." This regulation requires that the TSs include items in the following five specific categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements (SRs);

(4) design features; and (5) administrative controls. The regulation does not specify the particular requirements to be included in a plant's TSs.

As stated in 10 CFR 50.36(c)(3), "Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met." To meet this requirement, the SR must specify an adequate test, calibration, or inspection and an appropriate frequency of performance. The licensee has proposed to implement changes to surveillance frequencies in the SFCP using the methodology in NEI 04-10, which includes qualitative considerations, results of risk analyses, sensitivity studies and any bounding analyses, recommended monitoring of structures, systems, and components (SSCs), and documentation of the evaluation. Furthermore, changes to frequencies are subject to regulatory review and oversight of the SFCP implementation through the rigorous NRC review of safety-related SSC performance provided by the reactor oversight process (ROP).

Licensees are required by the TSs to perform surveillance test, calibration, or inspection on specific safety-related system equipment (e.g., reactivity control, power distribution, electrical, and instrumentation) to verify system operability. Surveillance frequencies, currently identified in the TSs, are based primarily upon deterministic methods such as engineering jUdgment, operating experience, and manufacturer's recommendations. The licensee's use of NRC approved methodologies identified in NEI 04-10 provides a way to establish risk-informed surveillance frequencies that complement the deterministic approach and support the NRC's traditional defense-in-depth philosophy.

The licensee's SFCP is intended to ensure that SRs specified in the TSs are performed at intervals sufficient to assure the above regulatory requirements are met. Existing regulatory requirements, such as 10 CFR 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants," and Appendix B to 10 CFR Part 50 require licensee monitoring of surveillance test failures and implementation of corrective actions to address such

- 5 failures. One of these actions may be to consider increasing the frequency at which a surveillance test is performed. In addition, the SFCP implementation guidance in NEI 04-10 requires monitoring the performance of SSCs for which surveillance frequencies are decreased to assure reduced testing does not adversely impact the SSCs. These requirements, and the monitoring required by NEI 04-10, are intended to ensure that surveillance frequencies are sufficient to assure that the requirements of 10 CFR 50.36 are satisfied and that any performance deficiencies will be identified and appropriate corrective actions taken.

Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk Informed Decisions on Plant-Specific Changes to the Licensing Basis" (Reference 5), describes a risk-informed approach, acceptable to the NRC, for assessing the nature and impact of proposed licensing-basis changes by considering engineering issues and applying risk insights.

This regulatory guide also provides risk acceptance guidelines for evaluating the results of such evaluations.

RG 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications" (Reference 3), describes an acceptable risk-informed approach specifically for assessing proposed TS changes.

RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities" (Reference 4), describes an acceptable approach for determining whether the quality of the PRA, in total or the parts that are used to support an application, is sufficient to provide confidence in the results, such that the PRA can be used in regulatory decision making for light-water reactors.

3.0 TECHNICAL EVALUATION

The licensee's adoption of TSTF-425 for HCGS provides for administrative relocation of applicable surveillance frequencies, and provides for the addition of the SFCP to the Administrative Controls section of the TSs. TSTF-425 also requires the application of NEI 04-10 for any changes to surveillance frequencies within the SFCP. The licensee's application for the changes proposed in TSTF-425 included documentation regarding the PRA technical adequacy consistent with the requirements of RG 1.200, Revision 1 (Reference 4). In accordance with NEI 04-10, PRA methods are used, in combination with plant performance data and other considerations, to identify and justify modifications to the surveillance frequencies of equipment at nuclear power plants. Th is is consistent with the guidance provided in RG 1.174 (Reference 5) and RG 1.177 (Reference 3).

3.1 RG 1.177 Five Key Safety Principles RG 1.177 identifies five key safety principles required for risk-informed changes to the TSs.

Each of these principles is addressed by the industry methodology document, NEI 04-10, and is evaluated below in SE Sections 3.1.1 through 3.1.5 with respect to the proposed amendment.

- 6 3.1.1 The Proposed Change Meets Current Regulations The first key safety principle in RG 1.177 is that the proposed change meets the current regulations.

Paragraph (c)(3) in 10 CFR 50.36 requires that TSs will include SRs which are "requirements relating to test, calibration, or inspection to assure that necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met."

The proposed amendment would relocate most periodic SR frequencies, currently shown in the HCGS TSs, to a licensee-controlled program (Le., the SFCP). The SRs themselves would remain in the TSs, as required by 10 CFR 50.36(c)(3). The requirements for the SFCP would be added to new TS 6.8.4.j. In accordance with TS 6.8.4.j, any changes to the SR frequencies would be made in accordance with NEI 04-10, Revision 1. By letter dated September 19, 2007 (Reference 9), the NRC staff found that the methodology in NEI 04-10, Revision 1, met NRC regulations, specifically 10 CFR 50.36(c)(3), and was an acceptable program for controlling changes to surveillance frequencies.

Based on the above considerations, the NRC staff concludes that the proposed change is consistent with the requirements in 10 CFR 50.36(c)(3). Therefore, the proposed change meets the first key safety principle of RG 1.177.

3.1.2 The Proposed Change Is Consistent With the Defense-in-Depth Philosophy Consistency with the defense-in-depth philosophy, the second key safety principle of RG 1.177, is met if:

A reasonable balance is preserved among prevention of core damage, prevention of containment failure, and consequence mitigation.

Over-reliance on programmatic activities to compensate for weaknesses in plant design is avoided.

System redundancy, independence, and diversity are preserved commensurate with the expected frequency, consequences of challenges to the system, and uncertainties (e.g., no risk outliers).

Defenses against potential common cause failures are preserved, and the potential for the introduction of new common cause failure mechanisms is assessed.

Independence of barriers is not degraded.

Defenses against human errors are preserved.

The intent of the General Design Criteria in 10 CFR Part 50, Appendix A, is maintained.

TSTF-425 requires the application of NEI 04-10 for any changes to surveillance frequencies within the SFCP. NEI 04-10 uses both the core damage frequency (CDF) and the large early release frequency (LERF) metrics to evaluate the impact of proposed changes to surveillance frequencies. The guidance of RG 1.174 and RG 1.177 for changes to CDF and LERF is achieved by evaluation using a comprehensive risk analysis, which assesses the impact of proposed changes including contributions from human errors and common cause failures.

- 7 Defense-in-depth is also included in the methodology explicitly as a qualitative consideration outside of the risk analysis, as is the potential impact on detection of component degradation that could lead to an increased likelihood of common cause failures. The NRC staff concludes that both the quantitative risk analysis and the qualitative considerations assure that a reasonable balance of defense-in-depth is maintained. Therefore, the proposed change meets the second key safety principle of RG 1.177.

3.1.3 The Proposed Change Maintains Sufficient Safety Margins The engineering evaluation that will be conducted by the licensee under the SFCP, when frequencies are revised, will assess the impact of the proposed frequency change in accordance with the principle that sufficient safety margins are maintained. The guidelines used for making that assessment will include ensuring the proposed surveillance test frequency change is not in conflict with approved industry codes and standards or adversely affects any assumptions or inputs to the safety analysis, or, if such inputs are affected, justification is provided to ensure sufficient safety margin will continue to exist.

The design, operation, testing methods, and acceptance criteria for SSCs, specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the Updated Final Safety Analysis Report and bases to the TSs), since these are not affected by changes to the surveillance frequencies.

Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis.

Based on the above considerations, the NRC staff concludes that there is reasonable assurance that safety margins will be maintained through use of the SFCP methodology. Therefore, the proposed change meets the third key safety principle of RG 1.177.

3.1.4 When Proposed Changes Result in an Increase in Core Damage Frequency or Risk. the Increases Should Be Small and Consistent With the Intent of the Commission's Safety Goal Policy Statement RG 1.177 provides a framework for evaluating the risk impact of proposed changes to surveillance frequencies. This requires the identification of the risk contribution from impacted surveillances, determination of the risk impact from the change to the proposed surveillance frequency, and performance of sensitivity and uncertainty evaluations. TSTF-425 requires application of NEI 04-10 in the SFCP. NEI 04-10 satisfies the intent of RG 1.177 requirements for evaluating the change in risk, and for assuring that such changes are small.

3.1.4.1 Quality of the PRA The quality of the HCGS PRA is compatible with the safety implications of the proposed TS change and the role the PRA plays in justifying the change. That is, the more the potential change in risk or the greater the uncertainty in that risk from the requested TS change, or both, the more rigor that must go into ensuring the quality of the PRA.

The licensee used RG 1.200 to address the technical adequacy of the HCGS PRA. RG 1.200 is NRC's developed regulatory guidance which, in Revision 1, endorsed with comments and qualifications the use of the American Society of Mechanical Engineers (ASME) RA-Sb-2005,

- 8 "Addenda to ASME RA-S-2002 Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," (Reference 6), NEI 00-02, "PRA Peer Review Process Guidelines,"

(Reference 7) and NEI 05-04, "Process for Performing Follow-On PRA Peer Reviews Using the ASME PRA Standard" (Reference 8). The licensee has performed an assessment of the PRA models used to support the SFCP against the requirements of RG 1.200 to assure that the PRA models are capable of determining the change in risk due to changes to surveillance frequencies of SSCs, using plant-specific data and models. Capability category II of ASME RA-Sb-2005 was applied as the standard, and any identified deficiencies to those requirements are assessed further to determine any impacts to proposed decreases to surveillance frequencies, including by the use of sensitivity studies where appropriate. The NRC staff notes that in Revision 2, RG 1.200 endorsed with comments and qualifications an updated combined standard which includes requirements for fire, seismic, and other external events PRA models.

The existing internal events standard was subsumed into the combined standard, but the technical requirements are essentially unchanged. Since NEI 04-10 Rev. 1 specifically identified the use of RG 1.200 Revision 1 to assess the internal events standard, the licensee's approach is reasonable and consistent with the approved methodology.

The licensee identified that there are no open significant facts and observations (F&Os) remaining from its 1999 Industry PRA Peer Review. In October 2008, a peer review of the HCGS internal events PRA model was conducted using the ASME PRA Standard (Reference 6), which identified a small number of supporting requirements which were not met at capability category II. The NRC staff reviewed the licensee's assessment of these identified deficiencies (Table 2.2-1 of Attachment 2 of the license amendment request). The staff's assessment of these open "gaps," to assure that they may be addressed and dispositioned for each surveillance frequency evaluation per the NEI 04-10 methodology, is provided below.

DA-D1: Plant-specific data was not used in the most recent update of reliability data, except for a subset of systems. The licensee identified that the systems updated represent the majority of high importance systems, and that sensitivity studies will be conducted as required to address the reliability data of other systems, and therefore this deficiency can be addressed per the methodology of NEI 04-10.

QU-E4: The most recent industry guidance for identifying and addressing modeling uncertainties has not been completely addressed. The licensee identified that uncertainty assessments for each surveillance frequency evaluation are required by the SFCP methodology and therefore this deficiency can be addressed per the methodology of NEI 04-10.

SY-A6 and SY-C2: System boundaries are not defined in the system notebooks, and the peer review identified other recommendations for improving the system notebook documentation.

These deficiencies in system notebook documentation do not affect the technical adequacy of the PRA model, and therefore can be addressed per the methodology of NEI 04-10.

Based on the licensee's assessment using the applicable PRA standard and RG 1.200, the level of PRA quality, combined with the proposed evaluation and disposition of the identified deficiencies, is sufficient to support the evaluation of changes proposed to surveillance frequencies within the SFCP, and is consistent with Regulatory Position 2.3.1 of RG 1.177.

- 9 3.1.4.2 Scope of the PRA The licensee is required to evaluate each proposed change to a relocated surveillance frequency using the guidance contained in NEI 04-10 to determine its potential impact on risk, due to impacts from internal events, fires, seismic, other external events, and from shutdown conditions. Consideration is made of both CDF and LERF metrics. In cases where a PRA of sufficient scope or where quantitative risk models were unavailable, the licensee uses bounding analyses, or other conservative quantitative evaluations. A qualitative screening analysis may be used when the surveillance frequency impact on plant risk is shown to be negligible or zero.

The licensee does not maintain a fire PRA model, but uses a combination of the Fire Induced Vulnerability Evaluation (FIVE) methodology and fire PRA techniques, which were completed for the Individual Plant Examination of External Events (IPEEE). Similarly, the IPEEE seismic PRA study will be used to provide seismic insights. Other external hazards were assessed as insignificant during the IPEEE assessment. Therefore, the risk contribution from these sources will be assessed either qualitatively or by bounding analyses for evaluation of surveillance frequency changes.

The licensee's evaluation methodology is sufficient to ensure the scope of the risk contribution of each surveillance frequency change is properly identified for evaluation, and is consistent with Regulatory Position 2.3.2 of RG 1.177.

3.1.4.3 PRA Modeling The licensee will determine whether the SSCs affected by a proposed change to a surveillance frequency are modeled in the PRA. Where the SSC is directly or implicitly modeled, a quantitative evaluation of the risk impact may be carried out. The methodology adjusts the failure probability of the impacted SSCs, including any impacted common cause failure modes, based on the proposed change to the surveillance frequency. Where the SSC is not modeled in the PRA, bounding analyses are performed to characterize the impact of the proposed change to the surveillance frequency. Potential impacts on the risk analyses due to screening criteria and truncation levels are addressed by the requirements for PRA technical adequacy consistent with guidance contained in RG 1.200, and by sensitivity studies identified in NEI 04-10.

The licensee will perform quantitative evaluations of the impact of selected testing strategy (Le., staggered testing or sequential testing) consistent with the guidance of NUREG/CR-6141 and NUREG/CR-5497, as discussed in NEI 04-10.

Thus, through the application of NEI 04-10, the HCGS PRA modeling is sufficient to ensure an acceptable evaluation of risk for the proposed changes in surveillance frequency, and is consistent with Regulatory Position 2.3.3 of RG 1.177.

3.1.4.4 Assumptions for Time-Related Failure Contributions The failure probabilities of SSCs modeled in the HCGS PRA include a standby time-related contribution and a cyclic demand-related contribution. NEI 04-10 criteria adjust the time-related failure contribution of SSCs affected by the proposed change to surveillance frequency. This is consistent with RG 1.177 Section 2.3.3 which permits separation of the failure rate contributions

- 10 into demand and standby for evaluation of surveillance requirements. If the available data do not support distinguishing between the time-related failures and demand failures, then the change to surveillance frequency is conservatively assumed to impact the total failure probability of the SSC, including both standby and demand contributions. The SSC failure rate (per unit time) is assumed to be unaffected by the change in test frequency, and will be confirmed by the required monitoring and feedback implemented after the change in surveillance frequency is implemented. The process requires consideration of qualitative sources of information with regard to potential impacts of test frequency on SSC performance, including industry and plant specific operating experience, vendor recommendations, industry standards, and code-specified test intervals. Thus the process is not reliant upon risk analyses as the sole basis for the proposed changes.

The potential beneficial risk impacts of reduced surveillance frequency, including reduced downtime, lesser potential for restoration errors, reduction of potential for test caused transients, and reduced test-caused wear of equipment, are identified qualitatively, but are conservatively not required to be quantitatively assessed. Thus, through the application of NEI 04-10, the licensee has employed reasonable assumptions with regard to extensions of surveillance test intervals, and its approach is consistent with Regulatory Position 2.3.4 of RG 1.177.

3.1.4.5 Sensitivity and Uncertainty Analyses NEI 04-10 requires sensitivity studies to assess the impact of uncertainties from key assumptions of the PRA, uncertainty in the failure probabilities of the affected SSCs, impact to the frequency of initiating events, and of any identified deviations from capability category II of ASME PRA Standard ASME RA-Sb-2005 (Reference 6). Where the sensitivity analyses identify a potential impact on the proposed change, revised surveillance frequencies are considered, along with any qualitative considerations that may bear on the results of such sensitivity studies.

Required monitoring and feedback of SSC performance once the revised surveillance frequencies are implemented will also be performed. Thus, through the application of NEI 04-10, the licensee has appropriately considered the possible impact of PRA model uncertainty and sensitivity to key assumptions and model limitations, and is consistent with Regulatory Position 2.3.5 of RG 1.177.

3.1.4.6 Acceptance Guidelines The licensee will quantitatively evaluate the change in total risk (including internal and external events contributions) in terms of CDF and LERF for both the individual risk impact of a proposed change in surveillance frequency and the cumulative impact from all individual changes to surveillance frequencies using the guidance contained in NRC-approved NEI 04-10 in accordance with the TS SFCP. Each individual change to surveillance frequency must show a risk impact below 1 E-6 per year for change to CDF, and below 1 E-7 per year for change to LERF. These criteria are consistent with the limits of RG 1.174 for very small changes in risk.

Where the RG 1.174 limits are not met, the process either considers revised surveillance frequencies which are consistent with RG 1.174 or the process terminates without permitting the proposed changes. Where quantitative results are unavailable to permit comparison to acceptance guidelines, appropriate qualitative analyses are required to demonstrate that the associated risk impact of a proposed change to surveillance frequency is negligible or zero.

Otherwise, bounding quantitative analyses are required which demonstrate the risk impact is at least one order of magnitude lower than the RG 1.174 acceptance guidelines for very small

- 11 changes in risk. In addition to assessing each individual SSC surveillance frequency change, the cumulative impact of all changes must result in a risk impact below 1 E-5 per year for change to CDF, and below 1 E-6 per year for change to LERF, and the total CDF and total LERF must be reasonably shown to be less than 1 E-4 per year and 1 E-5 per year, respectively. These are consistent with the limits of RG 1.174 for acceptable changes in risk, as referenced by RG 1.177 for changes to surveillance frequencies. The NRC staff interprets this assessment of cumulative risk as a requirement to calculate the change in risk from a baseline model utilizing failure probabilities based on the surveillance frequencies prior to implementation of the SFCP, compared to a revised model with failure probabilities based on changed surveillance frequencies. The NRC staff further notes that the licensee includes a provision to exclude the contribution to cumulative risk from individual changes to surveillance frequencies associated with insignificant risk increases (less than 5E-8 CDF and 5E-9 LERF) once the baseline PRA models are updated to include the effects of the revised surveillance frequencies.

The quantitative acceptance guidance of RG 1.174 is supplemented by qualitative information to evaluate the proposed changes to surveillance frequencies, including industry and plant-specific operating experience, vendor recommendations, industry standards, the results of sensitivity studies, and SSC performance data and test history.

The final acceptability of the proposed change is based on all of these considerations and not solely on the PRA results compared to numerical acceptance guidelines. Post implementation performance monitoring and feedback are also required to assure continued reliability of the components. The NRC staff concludes that the licensee's application of NEI 04-10 provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies, consistent with Regulatory Position 2.4 of RG 1.177.

Therefore, the proposed change satisfies the fourth key safety principle of RG 1.177 by assuring that any increase in risk is small and consistent with the intent of the Commission's Safety Goal Policy Statement.

3.1.5 The Impact of the Proposed Change Should Be Monitored Using Performance Measurement Strateg ies The licensee's adoption of TSTF-425 requires application of NEI 04-10 in the SFCP. NEI 04-10 requires performance monitoring of SSCs whose surveillance frequency has been revised as part of a feedback process to assure that the change in test frequency has not resulted in degradation of equipment performance and operational safety. The monitoring and feedback includes consideration of maintenance rule monitoring of equipment performance. In the event of degradation of SSC performance, the surveillance frequency will be reassessed in accordance with the methodology, in addition to any corrective actions which may apply as part of the maintenance rule requirements. The NRC staff concludes that the performance monitoring and feedback specified in NEI 04-10 is sufficient to reasonably assure acceptable SSC performance and is consistent with Regulatory Position 3.2 of RG 1.177. Therefore, the proposed change meets the fifth key safety principle of RG 1.177.

- 12 3.2 Addition of Surveillance Frequency Control Program to Administrative Controls The proposed amendment would add the SFCP to the Administrative Controls section of the HCGS TSs. Specifically, new TS 6.8.4.j, "Surveillance Frequency Control Program," would read as follows:

This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure that the associated Limiting Conditions for Operation are met.

a. The Surveillance Frequency Control Program shall contain a list of Frequencies of the Surveillance Requirements for which the Frequency is controlled by the program.

b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.

c. The provisions of Surveillance Requirements 4.0.2 and 4.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.

The NRC staff concludes that the proposed addition to the Administrative Controls section of the TSs adequately identifies the scope of the SFCP and defines the methodology to be used in a revision of SR frequencies. Therefore, the proposed TS change is acceptable.

3.3 TS Bases Changes PSEG's application dated March 19, 2010, provided proposed changes to the TS Bases to be implemented with the associated TS changes. These pages were provide for information only and will be revised in accordance with the HCGS TS Bases Control Program.

3.4 Technical Evaluation Conclusion

The NRC staff has reviewed the licensee's proposed relocation of some surveillance frequencies to a new licensee-controlled program, the SFCP, and its proposal to control changes to surveillance frequencies in accordance with the new program. Based on the above considerations, the NRC staff concludes that the proposed amendment is acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the New Jersey State Official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes SRs.

- 13 The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (75 FR 33842). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: A. Howe R. Ennis Date: February 25, 2011

- 14

7.0 REFERENCES

1. TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control-RITSTF Initiative 5b," March 18,2009 (ADAMS Accession Number: ML090850642).

2. NEI 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 58, Risk-Informed Method for Control of Surveillance Frequencies," April 2007 (ADAMS Accession Number:

ML071360456).

3. Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking:

Technical Specifications," August 1998 (ADAMS Accession Number: ML003740176).

4. Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 1, January 2007 (ADAMS Accession Number: ML070240001).

5. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk Informed Decisions on Plant-Specific Changes to the Licensing 8asis," NRC, Revision 1, November 2002 (ADAMS Accession Number: ML023240437).

6. ASME PRA Standard ASME RA-Sb-2005, Addenda to ASME RA-S-2002, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Application."

7. NEI 00-02, Revision 1 "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance," Revision 1! May 2006 (ADAMS Accession Number: ML061510621).

8. NEI 05-04, "Process for Performing Follow-On PRA Peer Reviews Using the ASME PRA Standard," Revision 0, August 2006.

9. Letter from Ho. K. Nieh (NRC) to 8iff Bradley (NEI), "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) 04-10, Revision 1, Risk-Informed Technical Specification Initiative 5-8, Risk-Informed Method for Control of Surveillance Frequencies,"

dated September 19, 2007 (ADAMS Accession No. ML072570267).

February 25, 2011 Mr. Thomas Joyce President and Chief Nuclear Officer PSEG Nuclear LLC P.O. Box 236, N09 Hancocks Bridge, NJ 08038

SUBJECT:

HOPE CREEK GENERATING STATION -ISSUANCE OF AMENDMENT RE:

RELOCATION OF SPECIFIC SURVEILLANCE FREQUENCIES TO A LICENSEE-CONTROLLED PROGRAM BASED ON TECHNICAL SPECIFICATION TASK FORCE (TSTF) CHANGE TSTF-425 (TAC NO. ME3545)

Dear Mr. Joyce:

The Commission has issued the enclosed Amendment No. 187 to Facility Operating License No. NPF-57 for the Hope Creek Generating Station (HCGS). This amendment consists of changes to the Technical Specifications (TSs) and Facility Operating License in response to your application dated March 19,2010, as supplemented by letters dated July 28,2010, and January 10, 2011.

The amendment modifies the TSs by relocating specific surveillance frequencies to a licensee controlled program. The changes are based on Nuclear Regulatory Commission-approved TS Task Force (TSTF) change TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control-RITSTF [Risk-Informed TSTF] Initiative 5b."

A copy of our safety evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Docket No. 50-354

Enclosures:

1. Amendment No. 187 to License No. NPF-57

2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION PUBLIC LPL 1-2 RtF RidsNrrDorlLpl1-2 Resource RidsNrrLAABaxter Resource RidsNrrPMHopeCreek Resource RidsNrrDorlDpr Resource RidsNrrDraApla Resource ADAMS Accession No: ML103410243 OFFICE LPL 1-2/PM NAME REnnis DATE 2/3/11 Sincerely, Ira!

Richard B. Ennis, Senior Project Manager Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation RidsOgcRp Resource RidsAcrsAcnw_MailCTR Resource RidsNrrDirsltsb Resource RidsRgn 1 MailCenter Resource GHiII,OIS AHowe, APLA GWaig,lTSB

SE dated 8/5/10 HChernoff 2/25/11

v t e Letter Sequence Approval
TAC:ME3545, Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5B (Approved, Closed)
Initiation Request, Request Acceptance Supplement Administration Questions 2 July 2010 Answers 28 July 2010 Results Approval Other: ML110980151
MONTHYEAR2010-07-28 [Table View]

ML103410243

Text

SUBJECT:

Dear Mr. Joyce:

Enclosures:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

3.4 Technical Evaluation Conclusion

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

7.0 REFERENCES

SUBJECT:

Dear Mr. Joyce:

Enclosures:

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