Text

License Amendment Request - Revise Technical Specifications to Apply TSTF-542, Reactor Pressure Vessel Water Inventory Control, Revision 2
	ML17349A027
Person / Time
Site:	Nine Mile Point
Issue date:	12/15/2017
From:	Jim Barstow; Exelon Generation Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NMP1L3142
	Download: ML17349A027 (105)
	v • d • e

200 Exelon Way Exelon Generation Kennett Square. PA 19348 www.exeloncorp.com 10 CFR 50.90 NMP1L3142 December 15, 2017 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Nine Mile Point Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-63 NRG Docket No. 50-220

Subject:

License Amendment Request - Revise Technical Specifications to Apply TSTF-542, "Reactor Pressure Vessel Water Inventory Control," Revision 2 In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (Exelon) requests an amendment to the Technical Specifications, Appendix A, of Renewed Facility Operating License No. DPR-63 for Nine Mile Point Nuclear Station, Unit 1 (NMP1 ). The proposed amendment is consistent with NRG-approved Technical Specification Task Force (TSTF) 542, Revision 2, "Reactor Pressure Vessel Water Inventory Control."

The proposed amendment would revise the NMP1 Technical Specification (TS) by replacing the existing specifications related to Operation with a Potential for Draining the Reactor Vessel (OPDRV) with revised requirements for Reactor Pressure Vessel Water Inventory Control (RPV WIG).

Attachment 1 provides the Evaluation of Proposed Changes. Attachment 2 provides the Proposed TS Marked-Up Pages. Attachment 3 provides revised (clean) TS pages.

Attachment 4 provides the Bases Marked-Up Pages.

The proposed changes have been reviewed by the NMP Plant Operations Review Committee in accordance with the requirements of the Exelon Quality Assurance Program.

Exelon requests approval of the proposed amendment by January 31, 2019. Once approved, the amendment shall be implemented no later than the start of the NMP1 Spring 2019 refueling outage.

U.S. Nuclear Regulatory Commission License Amendment Request Apply TSTF-542, Revision 2 Docket No. 50-220 December 15, 2017 Page2 There are no regulatory commitments contained in this request.

Exelon has concluded that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"

paragraph (b), Exelon is transmitting a copy of this application and its attachments to the designated State Officials.

Should you have any questions concerning this submittal, please contact Ron Reynolds at (610) 765-5247.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 15th day of December 2017.

R;j;,cJ /. ~ Jr-=i-'-

James Barstow Director - Licensing & Regulatory Affairs Exelon Generation Company, LLC Attachments:

1) Evaluation of Proposed Changes

2) Proposed Technical Specification Marked-Up Pages

3) Revised Technical Specification Pages

4) Proposed Technical Specification Bases Marked-Up Pages cc: USNRC Region I, Regional Administrator w/attachments USNRC Senior Resident Inspector, NMP w/attachments USNRC Project Manager, NMP w/attachments A. L. Peterson, NYSERDA w/attachments

ATTACHMENT 1 License Amendment Request Nine Mile Point Nuclear Station, Unit 1 Docket No. 50-220 EVALUATION OF PROPOSED CHANGES CONTENTS

SUBJECT:

Revise Technical Specifications to Apply TSTF-542, "Reactor Pressure Vessel Water Inventory Control," Revision 2

1.0 DESCRIPTION

2.0 ASSESSMENT 2.1 Applicability of Published Safety Evaluation 2.2 Variations

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration

4.0 ENVIRONMENTAL CONSIDERATION

5.0 REFERENCES

License Amendment Request Attachment 1 Adoption of TSTF-542, Revision 2 Page 1 of 17 Docket No. 50-220 Evaluation of Proposed Changes

1.0 DESCRIPTION

Pursuant to 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (Exelon) requests an amendment to the Technical Specifications (TS), Appendix A, of Renewed Facility Operating License No. DPR-63 for Nine Mile Point Nuclear Station, Unit 1 (NMP1 ).

The proposed changes would replace the existing specifications related to Operation with a Potential for Draining the Reactor Vessel (OPDRV) with revised specifications for Reactor Pressure Vessel Water Inventory Control (RPV WIG).

2.0 ASSESSMENT 2.1 Applicability of Published Safety Evaluation Exelon has reviewed the Safety Evaluation provided to the Technical Specifications Task Force on December 20, 2016 (Reference 1), as well as the information provided in TSTF-542, Revision 2, (Reference 2). Exelon has concluded that the justifications presented in TSTF-542, Revision 2, and the Safety Evaluation prepared by the NRG staff is applicable to NMP1, and justify this amendment for the incorporation of the changes to the NMP1 TS.

The following NMP1 TS are affected by the proposed changes:

TOG 1.0 Definitions 3.1.4 Core Spray System 3.1.9/4.1.9 Reactor Pressure Vessel (RPV) Water Inventory Control (New section added by this License Amendment Request) 3.2.7 Reactor Coolant System Isolation Valves 3.4.0 Reactor Building 3.4.1 Leak Rate 3.4.2 Reactor Building Integrity -Isolation Valves 3.4.3 Access Control 3.4.4 Emergency Ventilation System 3.4.5 Control Room Air Treatment System 3.6.2/4.6.2 Protective Instrumentation 2.2 Variations Exelon is proposing the following variations from the TS changes described in TSTF-542, Revision 2. These variations do not affect the applicability of TSTF-542, Revision 2, or the NRG staff's safety evaluation to the proposed license amendment. Similar to Cooper Nuclear Station (CNS) submittal to implement TSTF-542, Revision 2 (Reference 3), there are STS requirements on which TSTF-542, Revision 2, is based that do not appear in the NMP1 TS. The variations due to the NMP1 Custom Technical Specifications are described below.

License Amendment Request Attachment 1 Adoption of TSTF-542, Revision 2 Page 2 of 17 Docket No. 50-220 Evaluation of Proposed Changes 2.2.1 NMP1 Pre-GDC Assessment The TSTF-542, Revision 2, Traveler and Safety Evaluation discuss the applicable regulatory requirements and guidance, including the 10 CFR 50, Appendix A, General Design Criteria (GDC). NMP1 was not licensed to the 10 CFR 50, Appendix A, GDC. The NMP1 Updated Final Safety Analysis Report (UFSAR) provides an assessment against the GDC in Table 1-1.

This UFSAR table refers to the NMP1 Technical Supplement to Petition for Conversion from Provisional Operating License to Full-Term Operating License, July 1972, for the details of the assessment against the GDC current at that time. A review has determined that the plant-specific requirements for NMP1 are sufficiently similar to the Appendix A GDC as related to the proposed change. The proposed license change is consistent with the NMP1 current licensing basis in that the design requirements for instrumentation, reactor coolant leak detection, the reactor coolant pressure boundary, and reactor coolant makeup are unaffected. See below for excerpts (in italics) from the July 1972 NMP1 Technical Supplement to Petition for Conversion from Provisional Operating License to Full-Term Operating License describing how the NMP1 design aligns to the applicable GDC.

Criterion 13 - Instrumentation and Control The fission process is monitored and controlled for all conditions from the source range through the power range. The neutron monitoring system detects core conditions that could potentially threaten the overall integrity of the fuel barrier due to excess power generation and provides a corresponding signal to the reactor protection system.

Fission chambers, located in the core, are used to sense neutron flux from the source range through the power range. The detectors are located to provide maximum sensitivity to control-rod movement during startup and to provide optimum monitoring in the intermediate and power ranges.

Also, the reactor protection system is provided to initiate automatically appropriate action whenever specific Station conditions reach established limits. The protective system functions are tabulated in the Technical Specifications. The system is designed to mitigate the consequences of Station normal and accident transients, and operator errors, to ensure that core safety limits are not exceeded and to ensure the integrity of the reactor coolant boundary, containment and associated systems.

Instrumentation and control features of systems which can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems are described in detail in the FSAR. Also described there are instrumentation and control features of other systems associated with the reactor.

The detailed information referenced above substantiates compliance with this criterion.

Criterion 14 - Reactor Coolant Pressure Boundary This criterion is met since these systems were designed and analyzed for all possible adverse conditions and the resultant stresses are very conservative when compared to the American National Standards Institute code B31. 1 allowables. All code required tests, inspections and material certifications were performed.

Criterion 30 - Quality of Reactor Coolant Pressure Boundary As discussed in the evaluation for Criterion 14 above, components which are part of the reactor coolant pressure boundary are designed and constructed to the highest quality

License Amendment Request Attachment 1 Adoption of TSTF-542, Revision 2 Page 3 of 17 Docket No. 50-220 Evaluation of Proposed Changes standards practical. Drywell equipment and floor drains are monitored and sampled.

From analyses of the samples, the location of the source of reactor coolant leakage will be determined. In addition, a continuous air monitor is employed to monitor rises in drywell atmosphere radioactivity which can be related to leakage.

Criterion 33 - Reactor Coolant Makeup As discussed in the accident analysis presented in the FSAR, leaks from smaller coolant lines are detected by increased drywell temperature or pressure, or by drywell sump level buildup. For line breaks with an equivalent area less than 0. 15 square feet, feedwater flow provides adequate core cooling. A control-rod-drive pump is continuously available for high-pressure inventory makeup, even if all offsite power is lost. The control-rod-drive pump would deliver 50 gpm of makeup to the reactor at operating pressure.

For breaks above the capacity of these systems, or in the event that neither system is available, a redundant core spray system is provided. This system is discussed in the FSAR. Core spray coolant does not enter the reactor vessel until reactor pressure has dropped to 365 psig. For break areas less than 0. 15 square feet, the reactor is depressurized by the redundant auto-relief system. This allows core spray operation and prevents significant clad damage.

As an additional means of accommodating small breaks a feedwater high pressure coolant injection system will be provided. The system consists of two redundant sets of motor-driven pumps. The system will be capable of delivering 7,600 gpm into the reactor vessel at reactor pressure. Condensate and feedwater-booster pumps are normally operating when the primary system is at pressure. The two motor-driven feedwater pumps would be automatically started on either a turbine trip or low reactor water level. The system will use off-site power for operation and on-site power for the lubrication pump. Installation of final wiring and electrical controls will make the system operable by the end of the first major refueling outage.

Based on the analysis performed in the Technical Supplement, NMP1 believes that the plant-specific requirements for NMP1 are sufficiently similar to the Appendix A GDC and represent an adequate technical basis for adopting TSTF-542, Revision 2.

2.2.2 NMP1 Custom TS Summary The NMP1 TS are Custom TS (CTS) and are a different format, different numbering and titles than the NRC Improved Standard Technical Specifications in NUREG-1433 (STS) on which TSTF-542, revision 2, was based. As examples, the NMP1 CTS format does not capitalize definitions and uses the term Parameter instead of Function. These variations and other similar variations are discussed in the variation details that follow. These variations are administrative and do not affect the applicability of TSTF-542, Revision 2, to the NMP1 TS.

In keeping the CTS format, NMP1 includes a new Section 3.1.9 and 4.1.9 created for the Reactor Pressure Vessel (RPV) Water Inventory Control (WIC), Limiting Condition of Operations (LCOs) and associated Surveillance Requirements. Section 3.1.0 is for systems to protect the fuel cladding. Section 3.5.0 of the NMP1 TS is for Shutdown and Refueling.

Locating the new TS for RPV WIC in Section 3.1.0 is the logical location to be consistent with TSTF-542, Revision 2, Proposed Safety Basis (Section 3.1.2), and therefore justified.

License Amendment Request Attachment 1 Adoption of TSTF-542, Revision 2 Page 4 of 17 Docket No. 50-220 Evaluation of Proposed Changes Additionally, the wording for the new LCO Condition, Applicability, Required Action, Completion Time, and Surveillance Requirements have been modified from the STS format in TSTF-542, Revision 2, to align to the NMP1 TS format.

The NMP1 Reactor Operating Conditions are different than the STS. The STS terminology of Modes is not used in the NMP1 TS. The table below describes the differences.

STS Mode NMP1 TS Reactor Operating Condition 1 - Power Operation Power Operating Condition 2 - Startup 3 - Hot Shutdown (>200°F) Shutdown Condition - Hot (>212°F) 4 - Cold Shutdown {S200°F) Shutdown Condition - Cold (S212°F) 5 - Refuelina Refuelina Condition No Mode Mai or Maintenance Condition (defueled)

The NMP1 instrumentation uses the position of the Reactor Mode Switch to enable functions.

The tables in TS 3.6.2/4.6.2 Protective Instrumentation include columns titled "Reactor Mode Switch Position in Which Function Must be Operable." The table below correlates NMP1 Reactor Operating Condition to Mode Switch Position.

NMP1 TS Reactor Operating Condition NMP1 Reactor Mode Switch Position Run Power Operating Condition Startup Shutdown Condition - Hot (>212°F) Shutdown Shutdown Condition - Cold (S212°F) Shutdown or Refuel Refueling Condition Refuel Maior Maintenance Condition (defueled) No Required Position The TSTF-542, Revision 2, traveler is written to support a Safety Limit (SL) that requires RPV water level to remain above the Top of Active Fuel {TAF). The reactor vessel water level of -10 inches indicator scale used by NMP1 as its SL is a reflection of the limitations of the vessel level instrumentation available to the Operations staff and is not indicative of a difference from, or change to, the analysis presented in TSTF-542, Revision 2. The SL is a long-standing TS value and is not changed as part of the proposed changes to apply TSTF-542, Revision 2, to NMP1.

NMP1 TS 2.1.1.d specifies the SL for water level as -10 inches indicator scale. The TAF at NMP1 is defined in the UFSAR as -84 inches below vessel zero. All the trip settings for instruments that are relied upon for accident or transient mitigation are referenced to vessel zero. The setpoint of -10 inches indicator scale corresponds Low-Low-Low Reactor Water Level, which with a simultaneous High Drywell pressure condition, are the initiation signals for the Automatic Depressurization System (ADS). The ADS is used to rapidly depressurize the reactor when the line break is smaller than 0.30 sq. ft, so that the Low-Pressure Permissive signal can be sent to the valves and the Core Spray water admitted to the reactor.

Note also that the minimum level assumed for a Design Basis Accident Loss Of Coolant Accident (LOCA) is below the TAF, and the degree of cladding damage is a function of core uncovery (reflooded) time, and not minimum vessel level. The SL, therefore, does not directly impact reactor safety.

License Amendment Request Attachment 1 Adoption of TSTF-542, Revision 2 Page 5 of 17 Docket No. 50-220 Evaluation of Proposed Changes This SL is a variation to TSTF-542, Revision 2. NMP1 proposes to maintain the wording of protecting the SL, which will maintain RPV water level above -10 inches indicator scale, as opposed to referencing TAF. This use of "above -10 inches indicator scale" is a variation to several changes described in the proposed changes to apply TSTF-542, Revision 2, including the definition of drain time and throughout the new LCO for the RPV WIC. The variation maintains consistency within the NMP1 TS and does not affect the applicability of TSTF-542, Revision 2, to the NMP1 TS.

TSTF-542, Revision 2, identifies both a Low Pressure Core Injection (LPCI) system and a Low Pressure Core Spray (LPCS) system for RCP WIC. NMP1 does not have a LPCI system or LPCS system. NMP1 uses a Core Spray system that operates at low pressure. This is a variation to TSTF-542, Revision 2. The Core Spray system consists of two separate and independent Core Spray loops as described in NMP1 Updated Final Safety Analysis Report (UFSAR)Section VII, Engineered Safeguards. For NMP1, the Core Spray system performance is evaluated for the entire spectrum of break sizes for a postulated LOCA. Both systems (at least one subsystem in each system) are required to operate to limit peak clad temperatures below 2200°F (1 O CFR 50.46(a)(1 )(i) model) for the worst case line break (recirculation discharge line break). The NMP1 UFSAR, Chapter XV, Section 2.4.4, discusses a Core Spray system break, during a Design Basis Accident LOCA, reducing the Core Spray system to one sparger fed by one Core Spray pump and one topping pump. The resultant analysis demonstrated that the 10 CFR 50.46 limits are met. It is reasonable to assume, based on engineering judgment, that while the reactor coolant temperature is less than or equal to 212°F, one Core Spray subsystem can maintain adequate reactor vessel water level. Operation of one pump set and blocking valve is sufficient to establish required delivery rate and flow path. This evaluation represents an adequate technical basis for this variation to TSTF-542, Revision 2.

The Instrumentation specifications are Section 3.6.2 for NMP1. Section 4.6.2 contains the corresponding Surveillance Requirements. Section 3.6.2 is a series of Tables for the various instruments. Each instrument table uses the term Parameter, which is equivalent to the term Function in the STS. For the instrumentation changes to conform to TSTF-542, Revision 2, a new Table 3.6.2m/4.6.2m is created for RPV WIC instrumentation Parameters and associated Surveillance Requirements required for the Shutdown Condition - Cold and Refueling Conditions. The Parameters used in new Table 3.6.2m/4.6.2m come from the existing instrumentation Table 3.6.2b/4.6.2b and Table 3.6.2d/4.6.2d. The applicable notes are also retained in the new Table 3.6.2m/4.6.2m. Due to relocating some Parameters to Table 3.6.2m/4.6.2m, the applicability of those Parameters retained in the existing Table3.6.2b/4.6.2b and Table 3.6.2d/4.6.2d is also modified to align to the requirements of TSTF-542, Revision 2, for retaining applicability in the Shutdown Condition - Hot Operating Condition. This results in changing some of the existing Table 3.6.2 notes. These changes are summarized in the following variation sections where the Table changes are discussed.

TSTF-542, Revision 2, uses the term "Manual Initiation" for Functions in Table 3.3.5.2-1, RPV Water Inventory Control Instrumentation. To remain consistent with the NMP1 TS, the term "Manual" is used to identify specific Parameters in Tables 3.6.2m and 4.6.2m. This is a variation to TSTF-542, Revision 2, that is administrative in nature and do not affect the applicability of TSTF-542, Revision 2, to the NMP1 TS.

License Amendment Request Attachment 1 Adoption of TSTF-542, Revision 2 Page 6 of 17 Docket No. 50-220 Evaluation of Proposed Changes The table below summarizes TSTF-542, Revision 2, TS with OPDRV related requirements using NUREG-1433, BWR/4 STS, and the corresponding NMP1 TS where changes are proposed. The NMP1 TS do not use a Setpoint Control Program. The NMP1 CTS Equivalent column includes reference to the variation discussion sections in bold font. The sections identified in bold font in the table are described in detail following the table.

NUREG-1433, BWR/4 ISTS NMP1 CTS Equivalent Modified for TSTF-542, Rev. 2 3.3.5.1, Emergency Core Cooling System 3.6.2/4.6.2, Protective Instrumentation (ECCS) Instrumentation (Section 2.2.4 and 2.2.5 below)

Table 3.6.2d/4.6.2d Instrumentation That Initiates Core Spray

Table 3.6.2f/4.6.2f Instrumentation that Initiates Auto Depressurization 3.3.5.2 RPV WIC Instrumentation (added by 3.6.2/4.6.2 Protective Instrumentation TSTF-542, Revision 2) (Section 2.2.5 below)

Table 3.6.2b/4.6.2b Instrumentation that Initiates Primary Coolant System and Containment Isolation (Revised)

Table 3.6.2m/4.6.2m RPV WIC Instrumentation (New table added to NMP1 CTS for TSTF-542, Revision 2) 3.3.5.3 RCIC System Instrumentation 3.6.2/4.6.2 Protective Instrumentation (Renumbered from 3.3.5.2) (Section 2.2.6 below)

Table 3.6.2c/4.6.2c Instrumentation That Initiates or Isolates Emergency Coolina.

3.3.6.1, Primary Containment Isolation 3.6.2/4.6.2, Protective Instrumentation Instrumentation (Section 2.2. 7 below)

Table 3.6.2b/4.6.2b Instrumentation that Initiates Primary Coolant System or Containment Isolation

Table 3.6.2c/4.6.2c Instrumentation That Initiates or Isolates Emergency Coolinq.

3.3.6.2, Secondary Containment Isolation 3.6.2/4.6.2, Protective Instrumentation

Table 3.6.2j/4.6.2j, Emergency Ventilation Initiation 3.4.2 Reactor Building Integrity - Isolation Valves (Section 2.2.8 below) 3.3.7.1, [Main Control Room Environmental 3.4.5, Control Room Air Treatment System Control (MCREC)] System Instrumentation 3.6.2/4.6.2, Protective Instrumentation

Table 3.6.21/4.6.21, Control Room Air Treatment System Initiation (Section 2.2.9 below)

License Amendment Request Attachment 1 Adoption of TSTF-542, Revision 2 Page 7 of 17 Docket No. 50-220 Evaluation of Proposed Changes NUREG-1433, BWR/4 ISTS NMP1 CTS Equivalent Modified for TSTF-542, Rev. 2 3.5.2, ECCS Shutdown 3.1.4, Core Spray System 3.1.5, Solenoid-Actuated Pressure Relief Valves (Automatic Depressurization System).

There are no changes to Section 3.1.5.

(Section 2.2.3 below) 3.6.1.3, Primary Containment Isolation 3.2.7, Reactor Coolant System Isolation Valves (PCIVs) Valves 3.3.4, Primary Containment Isolation Valves (Section 2.2.1 O below) 3.6.4.1, [Secondary] Containment 3.4.0, Reactor Building 3.4.1, Leakage Rate 3.4.3, Access Control (Section 2.2.11 below) 3.6.4.2, Secondary Containment Isolation 3.4.2, Reactor Building Integrity - Isolation Valves (SCIVs) Valves (Section 2.2.12 below) 3.6.4.3, Standby Gas Treatment (SGT) 3.4.4, Emergency Ventilation System System (Section 2.2.13 below) 3.7.4, [Main Control Room Environmental 3.4.5, Control Room Air Treatment System Control (MCREC)] System (Section 2.2.14 below) 3.7.5, [Control Room Air Conditioning (AC)] NMP1 does not have an equivalent System standalone TS that requires changes described in the traveler.

(Section 2.2.15 below) 3.8.2, AC Sources - Shutdown The equivalent NMP1 TS do not have 3.8.5, DC Sources - Shutdown OPDRV requirements in these sections and 3.8.8, Inverters - Shutdown are not included.

3.8.10, Distribution Systems- Shutdown 2.2.3- Core Spray System Technical Specification Changes The Applicability of TS 3.1.4, Core Spray System is modified to Power Operating Condition or Shutdown Condition - Hot, to align with the change in applicability in the STS of MODES 1 , 2 or 3, as shown in the above table. The NMP1 design includes a Core Spray System to ensure adequate core cooling. The NMP1 Core Spray system consists of four Core Spray and Core Spray topping pumps. All eight pumps start on an initiation signal. Each Core spray topping pump discharge line is equipped with a PSV (relief valve) that relieves to the Torus (Suppression Chamber) to provide minimum flow protection for the Core Spray system. This water injection system is a low pressure system and requires the ADS to lower reactor vessel pressure to allow water injection. As described in the NMP1 UFSAR,Section VII, the ADS and Core Spray systems are the Engineered Safety Features (ESF) to either prevent or mitigate the consequences of major accidents. The ADS and Core Spray systems are considered the ECCS systems at NMP1. This is consistent with TSTF-542, Revision 2, Proposed Safety Basis (Section 3.1.2), and is therefore justified.

License Amendment Request Attachment 1 Adoption of TSTF-542, Revision 2 Page 8 of 17 Docket No. 50-220 Evaluation of Proposed Changes There are plant-specific TS requirements related to OPDRVs, and Required Actions to "suspend OPDRVs" that do not appear in the STS. Changes to these TS controls are justified by the discussion in TSTF-542, Revision 2, justification for Section 3.4.2, Other Proposed Changes -

Containment, Containment Isolation Valves, and Standby Gas Treatment System Requirements. Affected NMP1 TS are listed in Section 2.1 above.

In alignment with TSTF-542, Revision 2, Proposed Safety Basis (Section 3.1.2}, the existing NMP1 TS 3.1.4.h requirement to suspend core alterations as an action for Core Spray subsystem inoperability is no longer warranted since there are no postulated events associated with core alterations that are prevented or mitigated by the proposed RPV WIC requirements.

In addition, loss of RPV inventory events are not initiated by core alteration operations.

Refueling Limiting Conditions for Operation (LCOs} provide requirements to ensure safe operation during core alterations. Therefore, NMP1 proposes to delete TS 3.1.4.h.

The SRs in TSTF-542, Revision 2, for LCO 3.5.2 are included in the new SR 4.1.9, except for an equivalent for SR 3.5.2.4 to verify the required ECCS injection/spray subsystem, the piping is filled with water from the pump discharge valve to the injection valve. The current SR 4.1.9.g is limited in applicability to greater than 212°F; therefore, this surveillance is currently not required when in the Shutdown Condition - Cold or Refuel and is not being added. The system design and configuration is not susceptible to voiding and would not result in a system inoperability, as reported in the Generic Letter 2008-01 responses provided (References 4 and 5).

The following summary is the alignment from TSTF-542, Revision 2, TS 3.5.2 SRs to the applicable SRs in the new SR 4.1.9.

TSTF-542 LCO 3.5.2 SR number Eauivalent LCO 4.1.9 SR number 3.5.2.1 4.1.9.a 3.5.2.2 Not applicable. NMP1 does not have a LPCI subsvstem.

3.5.2.3 4.1.9.b, modified for plant specific confiauration 3.5.2.4 Not applicable 3.5.2.5 4.1.9.c 3.5.2.6 NMP1 does not have a recirculation line for the Core Sprav Svstem.

3.5.2.7 4.1.9.d 3.5.2.8 4.1.9.e 2.2.4 Instrumentation Technical Specification Variations (Table 3.3.5.1-1)

TSTF-542, Revision 2, Table 3.3.5.1-1, "Emergency Core Cooling System Instrumentation,"

corresponds to the NMP1 Table 3.6.2/4.6.2, "Protective Instrumentation."

2.2.4.1 TSTF-542, Revision 2, Table 3.3.5.1-1, "Emergency Core Cooling System Instrumentation," Function 1 for Core Spray system uses five different functions. NMP1 has three Parameters for Core Spray Initiation as described in paragraph 2.2.5.1 below.

The following Functions in TSTF-542, Revision 2, Table 3.3.5.1-1, Function 1, do not have NMP1 equivalent Parameters:

License Amendment Request Attachment 1 Adoption of TSTF-542, Revision 2 Page 9 of 17 Docket No. 50-220 Evaluation of Proposed Changes 1.c, Reactor Steam Dome Pressure - Low (Injection Permissive) 1.d, Core Spray Pump Discharge Flow- Low (Bypass)

The NMP1 Core Spray system utilizes the Parameter Low-Low Reactor Water Level, instead of TSTF-542, Revision 2, Function 1.a, Reactor Vessel Water Level - Low, Low, Low, Level 1.

2.2.4.2 TSTF-542, Revision 2, Table 3.3.5.1-1, "Emergency Core Cooling System Instrumentation," Function 2 for Low Pressure Coolant Injection (LPCI) system, does not have a direct correlation for NMP1. The NMP1 Core Spray System described in Section 2.2.3 above performs the low pressure injection Parameter.

2.2.4.3 TSTF-542, Revision 2, Table 3.3.5.1-1, "Emergency Core Cooling System Instrumentation," Function 3 for High Pressure Coolant Injection (HPCI) System, has seven functions. The NMP1 HPCI system is a mode of operation of the existing Feed and Condensate system and is not an ECCS system for NMP1, as discussed in the NMP1 UFSAR,Section VII, Engineered Safeguards. (See Section 2.2.3 above.)

Therefore, no changes are required to Table 3.6.2k for Parameters to initiate HPCI.

2.2.4.4 TSTF-542, Revision 2, Table 3.3.5.1-1, "Emergency Core Cooling System Instrumentation," Functions 4 and 5 for ADS, Trip System A and Trip System B, respectively, have eight functions. The NMP1 ADS Initiation Parameters are in Table 3.6.2f. The NMP1 ADS system requires two Parameters to initiate, (1 )a. Low-Low-Low Reactor Water Level and (1 )b. High Drywell Pressure. The following Functions in TSTF-542, Revision 2, Table 3.3.5.1-1, Functions 4 and 5, do not have NMP1 equivalents:

5.c, Automatic Depressurization System Initiation Timer 5.d, Reactor Vessel Water Level - Low, Level 3 (Confirmatory) 5.e. Core Spray Pump Discharge Pressure - High 5.f. Low Pressure Coolant Injection Pump Discharge Pressure - High 5.g. Automatic Depressurization System Low Water Level Actuation Timer 5.h. Manual Initiation There are no changes proposed to Table 3.6.2f. The ADS is only required to lower reactor coolant system pressure to allow Core Spray injection. The existing note (b) allows for the Parameter to be bypassed when reactor pressure is less than 11 O psig and reactor coolant temperature is less than the corresponding saturation temperature.

2.2.5 Instrumentation Technical Specification Variations (Table 3.3.5.2-1) 2.2.5.1 TSTF-542, Revision 2, Table 3.3.5.2-1, "RPV Water Inventory Control Instrumentation,"

Function 1 for the Core Spray system uses Reactor Steam Dome Pressure- Low (Injection Permissive), Core Spray Pump Discharge Flow- Low (Bypass) and Manual Initiation. For NMP1, the Core Spray system Initiation Parameters are identified in Table 3.6.2d. To start the Core Spray pumps, either Parameter (1) High Drywell Pressure or (2) Low-Low Reactor Water Level, is required. Additionally, Table 3.6.2d includes Parameter (3) Reactor Pressure to open the Core Spray discharge valves, in conjunction

License Amendment Request Attachment 1 Adoption of TSTF-542, Revision 2 Page 10 of 17 Docket No. 50-220 Evaluation of Proposed Changes with one of either Core Spray pump start Parameters. In alignment with TSTF-542, Revision 2, Proposed Safety Basis (Section 3.3.1 ), the High Drywall Pressure Parameter is not realistic in the Shutdown Condition - Cold and is not transferred to the new Table 3.6.2m/4.6.2m.

The current Table 3.6.2d is modified with Note (g) for the Shutdown position of the mode switch. Note (g) states, "The Parameters for Start Core Spray Pumps and Open Core Spray Discharge Valves in Table 3.6.2d are only applicable in the Shutdown Condition -

Hot. See Table 3.6.2m for Parameter applicability in the Shutdown Condition - Cold."

The applicability in the Refuel position of the mode switch is removed from Table 3.6.2d, for the Parameters Start Core Spray Pumps and Open Core Spray Discharge Valves.

This Note is added to provide clarification and to direct the reader to TS Table 3.6.2m for the corresponding Parameter applicability in the Shutdown Condition - Cold. This is a variation to TSTF-542, Revision 2, and is administrative in nature and does not affect the applicability of TSTF-542, Revision 2, to the NMP1 TS.

New Table 3.6.2m is organized consistent with TSTF-542, Revision 2, Table 3.3.5.2-1.

The Parameters necessary to inject water into the core are listed first, followed by Primary Coolant Isolation Parameters. The Parameters from Table 3.6.2d/4.6.2d are for opening of the discharge valves and are on new TS page 247b. The automatic start surveillance requirement in new Table 4.6.2m is replaced with "Manual" Initiation to align with TSTF-542, Revision 2, Table 3.3.5.2-1. Table 3.6.2m/4.6.2m retains the Note (f) content, as relabeled Note (d) because it is applicable in all conditions. Table 3.6.2m/4.6.2m also retains the Note (e) content, as relabeled Note (c) because it is applicable in the Refuel Condition. Note (b) is also retained in new Table 3.6.2m/4.6.2m, and is not relabeled. These variations align with TSTF-542, Revision 2, Proposed Safety Basis (Section 3.3.1 ), for instrumentation required to support automatic isolation of penetration flow paths that may be credited in a calculation of drain time.

New Table 4.6.2m contains the applicable surveillance requirements from Table 4.6.2d, and is modified with Note 1 to indicate the surveillance intervals are specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.6.2m.

Note (d) from Table 4.6.2d is retained in the new Table 4.6.2m and relabeled as Note (g).

2.2.5.2 TSTF-542, Revision 2, Table 3.3.5.2-1, "RPV Water Inventory Control Instrumentation,"

Function 2 for Low Pressure Coolant Injection (LPCI) System, does not have a direct correlation for NMP1. The NMP1 Core Spray system described in Section 2.2.3 above performs the low pressure injection function.

2.2.5.3 TSTF-542, Revision 2, Table 3.3.5.2-1, "RPV Water Inventory Control Instrumentation,"

Function 3 for RHR System Isolation, corresponds to the Shutdown Cooling System Isolation Parameter in NMP1 Section 3.6.2, Protective Instrumentation, Table 3.6.2b, Instrumentation that Initiates Primary Coolant System or Containment Isolation. The Parameters in Table 3.6.2b for Shutdown Cooling System Isolation are (1 )(b) Low-Low Reactor Water Level and (2) Manual. These parameters for the Shutdown Condition -

Cold are added to the new Table 3.6.2m.

License Amendment Request Attachment 1 Adoption of TSTF-542, Revision 2 Page 11of17 Docket No. 50-220 Evaluation of Proposed Changes The current Table 3.6.2b is modified with Note (k) for the Shutdown position of the mode switch. Note (k) states, "The Primary Coolant Isolation Parameters for Cleanup and Shutdown Cooling in Table 3.6.2b are only applicable in the Shutdown Condition - Hot.

See Table 3.6.2m for Parameter applicability in the Shutdown Condition - Cold." This Note is added to provide clarification and to direct the reader to TS Table 3.6.2m for the corresponding Parameter applicability in the Shutdown Condition - Cold. This is a variation to TSTF-542, Revision 2, and is administrative in nature and does not affect the applicability of TSTF-542, Revision 2, to the NMP1 TS. The applicability in the Refuel position of the mode switch is removed from Table 3.6.2b, for the Primary Coolant Isolation Parameters of (1 )(b) Shutdown Cooling, and (2) Manual. Additionally, Table 3.6.2b/4.6.2b Note (i) is retained in the Notes section since it is used for other instruments not transferring to Table 3.6.2m. Note U) is deleted from the Notes section because it only applies to the Parameters moving to Table 3.6.2m/4.6.2m in the Shutdown Condition - Cold or Refuel Condition. Table 3.6.2m/4.6.2m retains the Note U) content, as Note (e) because it is only applicable in the Shutdown Condition -

Cold. These variations align with TSTF-542, Revision 2, Proposed Safety Basis (Section 3.3.1 ), for instrumentation required to support automatic isolation of penetration flow paths that may be credited in a calculation of drain time.

New Table 3.6.2m is organized consistent to TSTF-542, Revision 2, Table 3.3.5.2-1.

The Parameters necessary to inject water into the core are first, followed by Primary Coolant Isolation Parameters. The Parameters from Table 3.6.2b/4.6.2b are for Primary Coolant Isolation and are on new TS page 247c. Parameter (1 )(a) automatically isolates Cleanup on Low-Low Reactor Water Level. The applicability is annotated with Note (a) which states, "The Parameters in this table are only applicable in the Shutdown Condition - Cold and Refuel. See Table 3.6.2b or Table 3.6.2d for Parameter applicability in the Shutdown Condition - Hot." The Note (f) is the same Note (f) from Table 3.6.2b/4.6.2b. Table 3.6.2m/4.6.2m Note (e) is the Note U) content from Table 3.6.2b/4.6.2b because it is only applicable in the Shutdown Condition - Cold.

2.2.5.4 TSTF-542, Revision 2, Table 3.3.5.2-1, "RPV Water Inventory Control Instrumentation,"

Function 4 for Reactor Water Cleanup (RWCU) System Isolation, is called Cleanup at NMP1, and is in Section 3.6.2, Protective Instrumentation, Table 3.6.2b, Instrumentation that Initiates Primary Coolant System or Containment Isolation. The Parameters in Table 3.6.2b for Cleanup System Isolation are (1 )(a) Low-Low Reactor Water Level and (2) Manual. These Parameters for the shutdown condition are being removed from table 3.6.2b and added to the new Table 3.6.2m. This variation aligns with TSTF-542, Revision 2, Proposed Safety Basis (Section 3.3.1 ), for instrumentation required to support automatic isolation of penetration flow paths that may be credited in a calculation of drain time.

2.2.5.5 The corresponding surveillances are for the Primary Coolant Isolation for the Cleanup and Shutdown Cooling Isolation Parameters for Low-Low Reactor Water Level (Table 4.6.2b(1 )) and Manual (Table 4.6.2b(2)). New Table 4.6.2m contains the applicable surveillance requirements from Table 4.6.2b, and is modified with Note 1 to indicate the surveillance intervals are specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.6.2m. Note (c) from Table 4.6.2b is retained in the new Table 4.6.2m and relabeled as Note (g). These variations align with TSTF-542,

License Amendment Request Attachment 1 Adoption of TSTF-542, Revision 2 Page 12 of 17 Docket No. 50-220 Evaluation of Proposed Changes Revision 2, Proposed Safety Basis (Section 3.3.1 ), for instrumentation required to support automatic isolation of penetration flow paths that may be credited in a calculation of drain time.

2.2.6 RCIC System Instrumentation (3.3.5.3)

TSTF-542, Revision 2, renumbered the existing RCIC system instrumentation section to support insertion of the new 3.3.5.2 for RPV WIC Instrumentation section. NMP1 has chosen to insert the new instrumentation section at the end of the existing tables. Therefore, renumbering of the previous tables is not required. A new Table 3.6.2m/4.6.2m for the RPV WIC Instrumentation is created.

The Emergency Cooling system at NMP1 has two instrumentation Parameters for initiation of Emergency Cooling that are applicable in the Shutdown Condition - Hot. This is not addressed for the equivalent system (RCIC) in TSTF-542, Revision 2, because RCIC is not applicable in modes 4 or 5. For NMP1, Table 3.6.2c, "Instrumentation That Initiates or Isolates Emergency Cooling," the Initiation Parameters (1) and (2) for the Shutdown position of the Reactor Mode Switch are retained in Table 3.6.2c. They are not applicable to the RPV WIC Instrumentation because they do not have an applicability in the Shutdown Condition- Cold or in Refuel. This is consistent with TSTF-542, Revision 2, Proposed Safety Basis (Section 3.1 .2) and therefore justified.

2.2.7 Primary Containment Isolation Instrumentation (3.3.6.1)

TSTF-542, Revision 2, deletes Required Action J.2 from 3.3.6.1. This is discussed above in Section 2.2.5.3 as part of the variation for the Shutdown Cooling Isolation instrumentation. No additional changes required.

2.2.8 Secondary Containment Isolation Instrumentation (3.3.6.2)

TSTF-542, Revision 2, deletes the OPDRV related requirement for the instrumentation isolation signals for Secondary Containment. Instrumentation in Table 3.6.2j/4.6.2j, "Emergency Ventilation Initiation," is revised to remove the OPDRV related requirement in Note (a).

Secondary containment isolation requirements associated with OPDRVs are also contained in Section 3.4.2, "Reactor Building Integrity - Isolation Valves." The TS markups provided in have deleted the OPDRV requirement from these sections, consistent with TSTF-542, Revision 2.

2.2.9 Main Control Room Environmental Control (MCREC) System Instrumentation (3.3.7.1)

TSTF-542, Revision 2, deletes the OPDRV related requirement for the MCREC system instrumentation. NMP1 TS Section 3.6.2, Protective Instrumentation, Table 3.6.21/4.6.21, Control Room Air Treatment System Initiation, contains the Initiation Parameters (1) and (4) for the Shutdown position of the Reactor Mode Switch. These Parameters are only applicable in the Shutdown Condition - Hot and are not relocated to Table 3.6.2m. The Control Room Air Treatment System isolation requirements associate with OPDRVs is contained in Section 3.4.5.

The TS markups provided in Attachment 2 have deleted the OPDRV requirement from this section, consistent with TSTF-542, Revision 2. This is consistent with TSTF-542, Revision 2, Proposed Safety Basis (Section 3.1.2) and therefore justified.

License Amendment Request Attachment 1 Adoption of TSTF-542, Revision 2 Page 13 of 17 Docket No. 50-220 Evaluation of Proposed Changes 2.2.10 Primary Containment Isolation Valves (PCIVs) (3.6.1.3)

TSTF-542, Revision 2, revises LCO 3.6.1.3, Action H to remove the OPDRV related Condition, Required Action H.1 and associated Completion Time. NMP1 TS 3.2.7, "Reactor Coolant System Isolation Valves," LCO 3.2.7.f is the NMP1 equivalent section. The TS markups provided in Attachment 2 have deleted the OPDRV requirement from this section, consistent with TSTF-542, Revision 2. NMP1 TS 3.3.4, "Primary Containment Isolation Valves," does not include reference to OPDRVs. Therefore, no changes to NMP1 TS 3.3.4 are required.

2.2.11 Secondary Containment (3.6.4.1)

TSTF-542, Revision 2, deletes the OPDRV related requirement for Secondary Containment.

The corresponding sections for NMP1 TS for Secondary Containment Isolation requirements associate with OPDRVs are contained in Section 3.4.0, "Reactor Building" and 3.4.1, "Leakage Rate," and 3.4.3, "Access Control." The TS markups provided in Attachment 2 have deleted the OPDRV requirement from these sections, consistent with TSTF-542, Revision 2.

2.2.12 Secondary Containment Isolation Valves (3.6.4.2)

TSTF-542, Revision 2, deletes the OPDRV related Applicability statement from LCO 3.6.4.2, and revises Action D to remove the OPDRV related Condition, Required Action and associated Completion Time. Secondary Containment Isolation requirements associate with OPDRVs are contained in Section 3.4.2, "Reactor Building Integrity - Isolation Valves." The TS markups provided in Attachment 2 have deleted the OPDRV requirement from this section, consistent with TSTF-542, Revision 2.

2.2.13 Standby Gas Treatment (SGT) System (3.6.4.3)

TSTF-542, Revision 2, deletes the OPDRV related Applicability statement from LCO 3.6.4.3, and revises Actions C and E to remove the OPDRV related Condition, Required Action and associated Completion Time. The equivalent NMP1 requirements associate with OPDRVs are contained in 3.4.4, "Emergency Ventilation System," LCO 3.4.4.e. The TS markups provided in Attachment 2 have deleted the OPDRV requirement from this section, consistent with TSTF-542, Revision 2.

2.2.14 Main Control Room Environmental Control (MCREC) System (3.7.4)

TSTF-542, Revision 2, deletes the OPDRV related Applicability statement from LCO 3.7.4.

requirement and revises Actions D and F to remove the OPDRV related Condition, Required Action and associated Completion Time. The NMP1 TS equivalent is the Control Room Air Treatment System Isolation requirements associate with OPDRVs is contained in Section 3.4.5.a and 3.4.5.h. The TS markups provided in Attachment 2 have deleted the OPDRV requirement from this section, consistent with TSTF-542, Revision 2.

License Amendment Request Attachment 1 Adoption of TSTF-542, Revision 2 Page 14 of 17 Docket No. 50-220 Evaluation of Proposed Changes 2.2.15 Control Room Air Conditioning (AC) System (3.7.5)

The NMP1 TS do not include a separate LCO for Control Room Air Conditioning. Therefore, the changes proposed in the traveler to delete the references to OPDRVs is not implemented for NMP1.

2.2.16 Surveillance Frequency Control Program The NMP1 TS contain a Surveillance Frequency Control Program. Therefore, the Surveillance Requirement Frequencies for Specifications 4.1.9 and 4.6.2 are "In accordance with the Surveillance Frequency Control Program."

3.0 REGULATORY ANAL VSIS 3.1 No Significant Hazards Consideration Exelon requests adoption of TSTF-542, Revision 2, "Reactor Pressure Vessel Water Inventory Control," which is an approved change to the Standard Technical Specifications (STS}, into the NMP1 Technical Specifications (TS}. The proposed amendment replaces the existing requirements in the TS related to "Operations with a Potential for Draining the Reactor Vessel" (OPDRV) with new requirements on Reactor Pressure Vessel Water Inventory Control (RPV WIC) to ensure RPV water level remains above -10 inches indicator scale (74 inches above the Top of Active Fuel).

Exelon has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes replace existing TS requirements related to OPDRVs with new requirements on RPV WIC that will ensure RPV water level remains above -10 inches indicator scale. Draining of RPV water inventory in the cold shutdown and refueling conditions is not an accident previously evaluated; therefore, replacing the existing TS controls to prevent or mitigate such an event with a new set of controls has no effect on any accident previously evaluated. RPV water inventory control in the cold shutdown or refueling condition is not an initiator of any accident previously evaluated. The existing OPDRV controls or the proposed RPV WIC controls are not mitigating actions assumed in any accident previously evaluated.

The proposed changes reduce the probability of an unexpected draining event (which is not a previously evaluated accident) by imposing new requirements on the limiting time in which an unexpected draining event could result in the reactor vessel water level dropping to -1 O inches indicator scale. These controls require cognizance of the plant configuration and control of configurations with unacceptably short drain times. These requirements reduce the probability of an unexpected draining event. The current TS

License Amendment Request Attachment 1 Adoption of TSTF-542, Revision 2 Page 15 of 17 Docket No. 50-220 Evaluation of Proposed Changes requirements are only mitigating actions and impose no requirements that reduce the probability of an unexpected draining event.

The proposed changes reduce the consequences of an unexpected draining event (which is not a previously evaluated accident) by requiring a Core Spray subsystem to be operable at all times in the cold shutdown and refueling conditions. The change in requirement from two Core Spray subsystems to one Core Spray subsystem in the cold shutdown or refueling conditions does not significantly affect the consequences of an unexpected draining event because the proposed Actions ensure equipment is available within the limiting drain time that is as capable of mitigating the event as the current requirements. The proposed controls provide escalating compensatory measures to be established as calculated drain times decrease, such as verification of a second method of water injection and additional confirmations that containment and/or filtration would be available if needed.

The proposed changes reduce or eliminate some requirements that were determined to be unnecessary to manage the consequences of an unexpected draining event, such as automatic initiation of a Core Spray subsystem and control room ventilation. These changes do not affect the consequences of any accident previously evaluated since a draining event in the cold shutdown or refueling condition is not a previously evaluated accident and the requirements are not needed to adequately respond to a draining event.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed changes replace existing TS requirements related to OPDRVs with new requirements on RPV WIC that will maintain RPV water level above -10 inches indicator scale. The proposed changes will not alter the design function of the equipment involved. Under the proposed changes, some systems that are currently required to be operable during OPDRVs would be required to be available within the limiting drain time or to be in service depending on the limiting drain time. Should those systems be unable to be placed into service, the consequences are no different than if those systems were unable to perform their function under the current TS requirements.

The event of concern under the current requirements and the proposed change is an unexpected draining event. The proposed changes do not create new failure mechanisms, malfunctions, or accident initiators that would cause a draining event or a new or different kind of accident not previously evaluated or included in the design and licensing bases.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

License Amendment Request Attachment 1 Adoption of TSTF-542, Revision 2 Page 16 of 17 Docket No. 50-220 Evaluation of Proposed Changes

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed changes replace existing TS requirements related to OPDRVs with new requirements on RPV WIC. The current requirements do not have a stated safety basis and no margin of safety is established in the licensing basis. The safety basis for the new requirements is to maintain RPV water level above -10 inches indicator scale. New requirements are added to determine the limiting time in which the RPV water inventory could drain to the top of the fuel in the reactor vessel should an unexpected draining event occur. Plant configurations that could result in lowering the RPV water level to -10 inches indicator scale within one hour are now prohibited. New escalating compensatory measures based on the limiting drain time replace the current controls.

The proposed TS establish a safety margin by providing defense-in-depth to maintain RPV water level above -10 inches indicator scale to protect the public health and safety.

While some less restrictive requirements are proposed for plant configurations with long calculated drain times, the overall effect of the change is to improve plant safety and to add safety margin.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above, Exelon concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c) and, accordingly, a finding of "no significant hazards consideration" is justified.

4.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

License Amendment Request Attachment 1 Adoption of TSTF-542, Revision 2 Page 17 of 17 Docket No. 50-220 Evaluation of Proposed Changes

5.0 REFERENCES

1. Final Safety Evaluation of Technical Specifications Task Force Traveler TSTF-542, Revision 2, "Reactor Pressure Vessel Water Inventory Control" (TAC No. MF3487) dated December 20, 2016, ADAMS Accession No. ML163439008

2. TSTF-542, Revision 2, "Reactor Pressure Vessel Water Inventory Control," dated December 20, 2016

3. Letter from Nebraska Public Power District (NPPD) to US NRC, "Application to Revise Technical Specifications to Adopt TSTF-542, Revision 2, "Reactor Pressure Vessel Water Inventory Control" Cooper Nuclear Station, Docket No. 50-298, License No. DPR-46," dated August 7, 2017, ADAMS Accession No. ML17228A042

4. Letter from Constellation Energy, Nine Mile Point Nuclear Station to US NRC, "Three Month Response to NRC Generic Letter 2008-01, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems," dated April 10, 2008

5. Letter from Constellation Energy, Nine Mile Point Nuclear Station to US NRC, "Supplemental Response to NRC Generic Letter 2008-01, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems," dated July 6, 2009

ATTACHMENT2 License Amendment Request Nine Mile Point Nuclear Station, Unit 1 Docket No. 50-220 Revise Technical Specifications to Adopt TSTF-542, "Reactor Pressure Vessel Water Inventory Control," Revision 2 Proposed Technical Specification Marked-Up Pages TS Pages TOC 8

54 55 56 79a thru 79c 109 (note page 108 is not changed, but included for continuity) 164 165 168 171 (note page 170 is not changed, but included for continuity) 174 (note page 173 is not changed, but included for continuity) 175 178 179 194 196 (note page 195 is not changed, but included for continuity) 205 213 (note page 212 is not changed, but included for continuity) 217 219 242 247b thru 247h

NINE MILE POINT NUCLEAR STATION UNIT 1 - TECHNICAL SPECIFICATIONS CONTENTS SECTION DESCRIPTION PAGE 1.0 Definitions 1 2.0 Safety Limits and Limiting Safety System Setting Safety Limits Limiting Safety System Setting 2.1.1 Fuel Cladding Integrity 2.1.2 Fuel Cladding Integrity 9 2.2.1 Reactor Coolant System 2.2.2 Reactor Coolant System 23 3.0 Limiting Condition for Operation (LCO) and Surveillance Requirement (SR) Applicability 27 3.1.0 Fuel Cladding 28 Limiting Condition for Operation Surveillance Requirements 3.1.1 Control Rod System 4.1.1 Control Rod System 29 3.1.2 Liquid Poison System 4.1.2 Liquid Poison System 44 3.1.3 Emergency Cooling System 4.1.3 Emergency Cooling System 50 3.1.4 Core Spray System 4.1.4 Core Spray System 54 3.1.5 Solenoid-Actuated Pressure Relief Valve 4.1.5 Solenoid-Actuated Pressure Relief Valve 60 3.1.6 Control Rod Drive Coolant Injection 4.1.6 Control Rod Drive Coolant Injection 62 3.1.7 Fuel Rods 4.1.7 Fuel Rods 65 3.1.8 High Pressure Coolant Injection 4.1.8 High Pressure Coolant Injection 76 3.1.9 Reactor Pressure Vessel (RPV) Water Inventory 4.1.9 Reactor Pressure Vessel (RPV) Water Inventory Control Control 79a AMENDMENT NO. 142, 182 i

1.28 (Deleted) 1.29 (Deleted) 1.30 Reactor Coolant Leakage

a. Identified Leakage (1) Leakage into closed systems, such as pump seal or valve packing leaks that are captured, flow metered and conducted to a sump or collecting tank, or (2) Leakage into the primary containment atmosphere from sources that are both specifically located and known not to be from a through-wall crack in the piping within the reactor coolant pressure boundary.

b. Unidentified Leakage All other leakage of reactor coolant into the primary containment area.

1.31 Core Operating Limits Report The CORE OPERATING LIMITS REPORT is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.6.5. Plant operation within these operating limits is addressed in individual specifications.

1.32 Shutdown Margin (SDM)

SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical throughout the operating cycle assuming that:

a. The reactor is xenon free,

b. The moderator temperature is 68 F, corresponding to the most reactive state, and

c. All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn. With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.

1.33 INSERVICE TESTING PROGRAM The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).

AMENDMENT NO. 142, 176, 180, 181, 216, 227 Insert A 8

Insert A 1.34 Drain Time The drain time is the time it would take for the water inventory in and above the Reactor Pressure Vessel (RPV) to drain to -10 inches indicator scale (74 inches above the top of the active fuel seated in the RPV) assuming:

a. The water inventory above -10 inches indicator scale is divided by the limiting drain rate;

b. The limiting drain rate is the larger of the drain rate through a single penetration flow path with the highest flow rate, or the sum of the drain rates through multiple penetration flow paths susceptible to a common Mode failure (e.g.,

seismic event, loss of normal power, single human error), for all penetration flow paths below -10 inches indicated scale except:

1. Penetration flow paths connected to an intact closed system, or isolated by manual or automatic valves that are locked, sealed, or otherwise secured in the closed position, blank flanges, or other devices that prevent flow of reactor coolant through the penetration flow paths;

2. Penetration flow paths capable of being isolated by valves that will close automatically without offsite power prior to the RPV water level being equal to -10 inches indicator scale when actuated by RPV water level isolation instrumentation; or

3. Penetration flow paths with isolation devices that can be closed prior to the RPV water level being equal to

-10 inches indicated scale by a dedicated operator trained in the task, who is in continuous communication with the control room, is stationed at the controls, and is capable of closing the penetration flow path isolation device without offsite power.

c. The penetration flow paths required to be evaluated per paragraph b) are assumed to open instantaneously and are not subsequently isolated, and no water is assumed to be subsequently added to the RPV water inventory;

d. No additional draining events occur; and

e. Realistic cross-sectional areas and drain rates are used.

A bounding drain time may be used in lieu of a calculated value.

AMENDMENT NO. 8a

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.1.4 CORE SPRAY SYSTEM 4.1.4 CORE SPRAY SYSTEM Applicability: Applicability:

Applies to the operating status of the core spray Applies to the periodic testing requirements for the systems when in the Power Operating Condition or core spray systems.

Shutdown Condition - Hot.

Objective:

To verify the operability of the core spray systems.

To assure the capability of the core spray systems to cool reactor fuel in the event of a loss-of-coolant accident.

Specification:

The core spray system surveillance shall be

a. Whenever irradiated fuel is in the reactor vessel performed as indicated below.

and the reactor coolant temperature is greater than 212F, each of the two core spray systems a. In accordance with the Surveillance Frequency shall be operable except as specified in Control Program automatic actuation of each Specifications b and c below. subsystem in each core spray system shall be demonstrated.

b. If a redundant component of a core spray system becomes inoperable, that system shall be b. In accordance with the Surveillance Frequency considered operable provided that the component Control Program pump operability shall be is returned to an operable condition within 7 days checked.

and the additional surveillance required is performed. c. In accordance with the Surveillance Frequency Control Program the operability of power-

c. If a redundant component in each of the core operated valves required for proper system spray systems becomes inoperable, both systems operation shall be checked.

shall be considered operable provided that the component is returned to an operable condition within 7 days and the additional surveillance required is performed.

AMENDMENT NO. 142, 222, 54

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

d. If Specifications a, b and c are not met, a normal d. (Deleted) orderly shutdown shall be initiated within one hour and the reactor shall be in the cold shutdown condition within ten hours.

e. During reactor operation, except during core e. Surveillance with Inoperable Components spray system surveillance testing, core spray isolation valves 40-02 and 40-12 shall be in the When a component becomes inoperable its open position and the associated valve motor redundant component or system shall be verified starter circuit breakers for these valves shall be to be operable immediately and in accordance with locked in the off position. In addition, redundant the Surveillance Frequency Control Program valve position indication shall be available in the thereafter.

control room.

f. Whenever irradiated fuel is in the reactor vessel f. With a core spray subsystem suction from the and the reactor coolant temperature is less than CST, CST level shall be checked in accordance or equal to 212F, two core spray subsystems with the Surveillance Frequency Control Program.

shall be operable except as specified in g and h below. (Deleted)

g. If one of the above required subsystems g. In accordance with the Surveillance Frequency becomes inoperable, restore at least two Control Program when the reactor coolant subsystems to an operable status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months temperature is greater than 212F, verify that or suspend all operations that have a potential the piping system between valves 40-03, 13 and for draining the reactor vessel. (Deleted) 40-01, 09, 10, 11 is filled with water.

AMENDMENT NO. 142, 192, 222, 55

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

h. If both of the above required subsystems become inoperable, suspend core alterations and all operations that have a potential for draining the reactor vessel. Restore at least one subsystem to operable status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or establish secondary containment integrity within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . (Deleted)

i. With the downcomers in the suppression chamber having less than three and one half foot submergence, two core spray subsystems and the associated raw water pumps shall be operable with the core spray suction from the condensate storage tanks (CST), and the CST inventory shall not be less than 300,000 gallons.

AMENDMENT NO. 142, 56

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.1.9 Reactor Pressure Vessel (RPV) Water Inventory 4.1.9 Reactor Pressure Vessel (RPV) Water Inventory Control Control Applicability: Applicability Applies to the operating status of the core spray Applies to the periodic testing requirements for the systems and Reactor Water Inventory Control when core spray system and RPV water inventory.

the reactor coolant temperature is less than or equal to 212F.

Objective: Objective:

To assure the RPV water inventory is To verify the operability of the core spray system maintained -10 inches indicator scale. and RPV water inventory.

Specification: Specification:

a. Whenever irradiated fuel is in the reactor vessel a. Verify drain time 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months in accordance and the reactor coolant temperature is less than or with the Surveillance Frequency Control equal to 212°F, drain time of RPV water inventory Program.

to the TAF shall be 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months and one core spray subsystem shall be operable except as specified in b. Verify, for a required core spray subsystem, the Specifications b through f below. downcomers in the suppression chamber have greater than or equal to three and one half foot

b. If the required core spray subsystem becomes of submergence or the condensate storage inoperable, the component shall be returned to an tank inventory is not less than 300,000 gallons, operable condition within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . in accordance with the Surveillance Frequency Control Program.

c. If Specifications a and b are not met, then c. Verify for the required core spray subsystem, immediately initiate action to establish a method of each manual power operated and automatic water injection capable of operating without offsite valve in the flow path, that is not locked, sealed electrical power. or otherwise secured in position, is in the correct position, in accordance with the Surveillance Frequency Control Program.

AMENDMENT NO. 79a

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

d. If drain time < 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months and 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , within 4 d. Verify each valve credited for automatically hours perform the following actions: isolating a penetration flow path actuates to the isolation position on an actual or simulated (1) Verify secondary containment boundary is isolation signal, in accordance with the capable of being established in less than the Surveillance Frequency Control Program.

drain time.

and e. Verify the required core spray subsystem actuates (2) Verify each secondary containment penetration on a manual initiation signal, in accordance with flow path is capable of being isolated in less the Surveillance Frequency Control Program.

than the drain time, Vessel spray may be excluded.

and (3) Verify one RBEVS is capable of being placed in operation in less than the drain time.

e. If drain time < 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , immediately perform the following actions:

(1) Initiate action to establish an additional method of water injection with water sources capable of maintaining RPV water level above -10 inches indicator scale for 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

and (2) Initiate action to establish secondary containment boundary, and (3) Initiate action to isolate each secondary containment penetration flow path or verify it can be manually isolated from the control room. and (4) Initiate action to verify one RBEVS is capable of being placed in operation.

AMENDMENT NO. 79b

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

f. Specifications d and e not met, or drain time is

< 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , immediately initiate action to restore drain time to 36 hour4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

AMENDMENT NO. 79c

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.2.7 REACTOR COOLANT SYSTEM ISOLATION VALVES 4.2.7 REACTOR COOLANT SYSTEM ISOLATION VALVES Applicability: Applicability:

Applies to the operating status of the system of Applies to the periodic testing requirement for the isolation valves on lines connected to the reactor reactor coolant system isolation valves.

coolant system.

Objective: Objective:

To assure the capability of the reactor coolant system To assure the capability of the reactor coolant system isolation valves to minimize reactor coolant loss in the isolation valves to minimize reactor coolant loss in the event of a rupture of a line connected to the nuclear event of a rupture of a line connected to the nuclear steam supply system, and to minimize potential leakage steam supply system, and to limit potential leakage paths from the primary containment in the event of a loss- paths from the primary containment in the event of a of-coolant accident. loss-of-coolant accident.

Specification: Specification:

a. Whenever fuel is in the reactor vessel and the reactor The reactor coolant system isolation valves coolant temperature is greater than 212°F, all reactor surveillance shall be performed as indicated below.

coolant system isolation valves on lines connected to the reactor coolant system shall be operable except a. In accordance with the Surveillance as specified in Specification 3.2.7.b below. Frequency Control Program the operable automatically initiated power-operated

b. In the event any isolation valve becomes isolation valves shall be tested for automatic inoperable whenever fuel is in the reactor vessel and initiation and closure times.

the reactor coolant temperature is greater than 212°F, the system shall be considered operable b. Additional surveillances shall be performed as provided that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months at least one valve in required by the INSERVICE TESTING PROGRAM.

each line having an inoperable valve is in the mode corresponding to the isolated condition, except as noted in Specification 3.1.1.e.

AMENDMENT N0.142, 145, 173, 181, 197, 222, 227 108

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

c. If Specifications 3.2.7a and b above are not met, c. In accordance with the Surveillance Frequency initiate normal orderly shutdown within one hour Control Program the feedwater and main-steam and have reactor in the cold shutdown condition line power-operated isolation valves shall be within ten hours. exercised by partial closure and subsequent reopening.

d. Whenever fuel is in the reactor vessel and the reactor coolant temperature is less than or equal to 212°F, d. At least once per plant cold shutdown the the isolation valves on the shutdown cooling system feedwater and main steam line power-operated lines connected to the reactor coolant system shall be isolation valves shall be fully closed and operable except as specified in Specification 3.2.7.e reopened, unless this test has been performed below. within the previous 92 days.

e. In the event any shutdown cooling system isolation valve becomes inoperable whenever fuel is in the reactor vessel and the reactor coolant temperature is less than or equal to 212°F, the system shall be considered operable provided that, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , at least one valve in each line having an inoperable valve is in the mode corresponding to the isolated condition.

f. If Specifications 3.2.7.d and 3.2.7.e above are not met, immediately initiate action to restore the valve(s) to operable status. either:

(1) Immediately initiate action to suspend operations with a potential for draining the reactor vessel (OPDRVs); or (2) Immediately initiate action to restore the valve(s) to operable status.

AMENDMENT NO. 142, 197, 222, 109

3.4.0 REACTOR BUILDING APPLICABILITY Applies to the operating status of the reactor building (secondary containment).

OBJECTIVE To assure the integrity of the reactor building.

SPECIFICATION Reactor building integrity must be in effect for the following conditions:

a. Power operating condition,

b. When the reactor water temperature is above 215°F, or

c. Whenever recently irradiated fuel or an irradiated fuel cask is being handled in the Reactor Building, and.

d. During operations with a potential for draining the reactor vessel (OPDRVs).

AMENDMENT NO. 142, 170, 194, 164

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.4.1 LEAKAGE RATE 4.4.1 LEAKAGE RATE Applicability: Applicability:

Applies to the leakage rate of the secondary Applies to the periodic testing requirements of the containment. secondary containment leakage rate.

Objective: Objective:

To specify the requirements necessary to limit exfiltration of To assure the capability of the secondary containment fission products released to the to maintain leakage within allowable limits.

secondary containment as a result of an accident.

Specification:

In accordance with the Surveillance Frequency Control At all times when secondary containment integrity is Program - isolate the reactor building and start required, the reactor building leakage rate as determined by emergency ventilation system fan to demonstrate Specification 4.4.1 shall not exceed negative pressure in the building relative to external 1600 cfm. If this cannot be met after a routine surveillance static pressure. The fan flow rate shall be varied so check, then the actions listed below shall be taken: that the building internal differential pressure is at least as negative as that on Figure 3.4.1 for the wind speed

a. Suspend any of the following activities: at which the test is conducted. The fan flow rate represents the reactor building leakage referenced to

1. Handling of recently irradiated fuel in the reactor zero mph with building internal pressure at least 0.25 building. inch of water less than atmospheric pressure. The test shall be done at wind speeds less than 20 miles per

2. Irradiated fuel cask operations in the reactor hour.

building.

3. Operations with a potential for draining the reactor vessel (OPDRVs).

b. Restore the reactor building leakage rates to within specified limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or initiate normal orderly shutdown and be in a cold shutdown condition within 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months .

AMENDMENT NO. 142, 156, 170, 194, 222, 165

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.4.2 REACTOR BUILDING INTEGRITY - ISOLATION 4.4.2 REACTOR BUILDING INTEGRITY - ISOLATION VALVES VALVES Applicability: Applicability:

Applies to the operational status of the reactor Applies to the periodic testing requirements of the building isolation valves. reactor building isolation valves.

Objective: Objective:

To assure that fission products released to the secondary To assure the operability of the reactor building containment are discharged to the environment in a isolation valves.

controlled manner using the emergency ventilation system.

Specification: Specification:

a. The normal Ventilation System isolation valves In accordance with the Surveillance Frequency Control shall be operable at all times when secondary Program, automatic initiation of valves shall be checked.

containment integrity is required.

b. If Specification 3.4.2.a is not met, then the actions listed below shall be taken:

1. The reactor shall be in the cold shutdown condition within ten hours.

2. Suspend any of the following activities:

a. Handling of recently irradiated fuel in the reactor building,

b. Irradiated fuel cask handling operations in the reactor building,

c. Operations with a potential for draining the reactor vessel (OPDRVs).

AMENDMENT NO. 142, 170, 194, 222, 168

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.4.3 ACCESS CONTROL 4.4.3 ACCESS CONTROL Applicability: Applicability:

Applies to the access control to the reactor building. Applies to the periodic checking of the condition of portions of the reactor building.

Objective: Objective:

To specify the requirements necessary to assure the To assure that pump compartments are properly integrity of the secondary containment system. closed at all times and to assure the integrity of the secondary containment system by verifying that Specification: reactor building access doors are closed, as required by Specifications 3.4.3.a.1 and 3.4.3.a.2.

a. At all times when secondary containment integrity is required, the following conditions will be met: Specification:

1. Only one door in each of the double-doored a. The core and containment spray pump access ways shall be opened at one time, compartments shall be checked in accordance except when the access opening is being used with the Surveillance Frequency Control Program for entry and exit. and after each entry.

2. Only one door or closeup of the railroad bay shall be opened at one time.

3. The core spray and containment spray pump compartments' doors shall be closed at all times except during passage in order to consider the core spray system and the containment spray system operable.

AMENDMENT NO. 142, 170, 194, 222, 223 170

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

b. If these conditions cannot be met, then the actions b. Verify in accordance with the Surveillance listed below shall be taken: Frequency Control Program that:

1. If in the power operating condition, restore 1. At least one door in each access to the reactor building integrity within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be secondary containment is closed, except when in at least the hot shutdown condition within the access opening is being used for entry and the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in the cold shutdown exit.

condition within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2. At least one door or closeup of the railroad OR bay is closed.

If the reactor coolant system temperature is above 215°F, restore reactor building integrity within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in cold shutdown within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2. Suspend any of the following activities:

a. Handling of recently irradiated fuel in the reactor building,

b. Irradiated fuel cask handling operations in the reactor building,

c. Operations with a potential for draining the reactor vessel (OPDRVs).

AMENDMENT NO. 142, 170, 194, 222, 223 171

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.4.4 EMERGENCY VENTILATION SYSTEM 4.4.4 EMERGENCY VENTILATION SYSTEM Applicability: Applicability:

Applies to the operating status of the emergency Applies to the testing of the emergency ventilation ventilation system. system.

Objective: Objective:

To assure the capability of the emergency ventilation To assure the operability of the emergency ventilation system to minimize the release of radioactivity to the system.

environment in the event of an incident within the primary containment or reactor building.

Specification: Specification:

a. Except as specified in Specification 3.4.4e Emergency ventilation system surveillance shall be below, both circuits of the emergency ventilation performed as indicated below:

system shall be operable at all times when secondary containment integrity is required. a. In accordance with the Surveillance Frequency Control Program, the following conditions shall be

b. The results of the in-place cold DOP and halo- demonstrated:

genated hydrocarbon tests at design flows on HEPA filters and charcoal adsorber banks shall (1) Pressure drop across the combined HEPA show > 99% DOP removal and > 99% halogen- filters and charcoal adsorber banks is less ated hydrocarbon removal when tested in than 6 inches of water at the system rated accordance with ANSI N.510-1980. flow rate ( 10%).

(2) Deleted AMENDMENT NO. 142, 179, 201, 222 173

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

c. The results of laboratory carbon sample analysis b. The tests and sample analysis of Specification shall show 95% radioactive methyl iodide removal 3.4.4b, c and d shall be performed in when tested in accordance with ASTM D3803-1989 accordance with the Surveillance Frequency at 30°C and 95% R.H. Control Program, or after 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation, whichever occurs first or following

d. Fans shall be shown to operate within +/-10% significant painting, fire or chemical release in design flow. any ventilation zone communicating with the system.

e. During reactor operation, including when the reactor coolant system temperature is above c. Cold DOP testing shall be performed after each 215°F, from and after the date that one circuit complete or partial replacement of the HEPA of the emergency ventilation system is made or filter bank or after any structural maintenance on found to be inoperable for any reason, reactor the system housing.

operation is permissible only during the succeeding seven days unless such circuit is d. Halogenated hydrocarbon testing shall be sooner made operable, provided that during such performed after each complete or partial seven days all active components of the other replacement of the charcoal adsorber bank or emergency ventilation circuit shall be operable. after any structural maintenance on the system housing.

During handling of recently irradiated fuel in the reactor building, or handling of an irradiated fuel e. Each circuit shall be operated at least 15 minutes cask in the reactor building, and operations with a in accordance with the Surveillance Frequency potential for draining the reactor vessel Control Program.

(OPDRVs), from and after the date that one circuit of the emergency ventilation system f. Test sealing of gaskets for housing doors is made or found to be inoperable for any reason, downstream of the HEPA filters and charcoal recently irradiated fuel handling in the reactor adsorbers shall be performed at and in building, or irradiated fuel cask handling in the conformance with each test performed for reactor building, or OPDRVs are is permissible compliance with Specification 4.4.4b and during the succeeding seven days unless such Specification 3.4.4b.

circuit is sooner made operable, provided that AMENDMENT NO. 142, 146, 170, 171, 194, 201, 222, 174

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT during such seven days all active components of g. In accordance with the Surveillance Frequency the other emergency ventilation circuit shall be Control Program, automatic initiation of each operable. Recently irradiated fuel handling in the branch of the emergency ventilation system reactor building, or irradiated fuel cask handling in shall be demonstrated.

the reactor building, or OPDRVs may continue beyond seven days provided the operable h. In accordance with the Surveillance Frequency emergency ventilation circuit is in operation. Control Program, manual operability of the bypass valve for filter cooling shall be demonstrated.

f. If these conditions cannot be met, within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months , the reactor shall be placed in a condition i. When one circuit of the emergency ventilation for which the emergency ventilation system is system becomes inoperable all active compon-not required. ents in the other emergency ventilation circuit shall be verified to be operable within two hours and in accordance with the Surveillance Frequency Control Program thereafter.

AMENDMENT NO. 142, 194, 222, 175

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.4.5 CONTROL ROOM AIR TREATMENT SYSTEM 4.4.5 CONTROL ROOM AIR TREATMENT SYSTEM Applicability: Applicability:

Applies to the operating status of the control room air treatment Applies to the testing of the control room air treatment system and Control Room Envelope (CRE) boundary. system and CRE boundary.

NOTE----------------------------

The CRE boundary may be opened intermittently under administrative control.

Objective: Objective:

To assure the capability of the control room air treatment To assure the operability of the control room air treatment system to minimize the amount of radioactivity or other system.

gases entering the control room in the event of an incident.

Specification:

a. In accordance with the Surveillance Frequency Control

a. Except as specified below, the control room air treatment Program, the pressure drop across the combined system shall be operable for the following conditions: HEPA filters and charcoal adsorber banks shall be demonstrated to be less than 1.5 inches of water at

1. Power operating condition, and whenever the system design flow rate ( 10%).

reactor coolant system temperature is greater than 212°F, . b. The tests and sample analysis of Specification 3.4.5b, c and d shall be performed in accordance with

2. Whenever recently irradiated fuel or an irradiated the Surveillance Frequency Control Program, or after fuel cask is being handled in the reactor building, . 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation, whichever occurs first and or following significant painting, fire or chemical release in any ventilation zone communicating with the system.

3. During operations with a potential for draining the reactor vessel (OPDRVs).

b. The results of the in-place cold DOP and halogenated hydrocarbon tests at design flows on HEPA filters and charcoal adsorber banks shall show > 99% DOP removal and > 99% halogenated hydrocarbon removal when tested in accordance with ANSI N.510-1980.

AMENDMENT NO. 142, 170, 171, 179, 194, 195, 222, 178

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

c. The results of laboratory carbon sample analysis shall c. Cold DOP testing shall be performed after each show 95% radioactive methyl iodine removal when complete or partial replacement of the HEPA filter tested in accordance with ASTM D3803-1989 at 30°C bank or after any structural maintenance on the and 95% R.H. system housing.

d. Fans shall be shown to operate within +/-10% design flow. d. Halogenated hydrocarbon testing shall be performed after each complete or partial

e. From and after the date that the control room air treatment replacement of the charcoal absorber bank or system is made or found to be inoperable for any reason, after any structural maintenance on the system except for an inoperable CRE boundary during the power housing.

operating condition, restore the system to operable within the succeeding seven days. e. The system shall be operated at least 15 minutes in accordance with the Surveillance Frequency Control

f. If the control room air treatment system is made or found to Program.

be inoperable due to an inoperable CRE boundary during the power operating condition: immediately initiate action to f. In accordance with the Surveillance Frequency implement mitigating actions; within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , verify Control Program, automatic initiation of the mitigating actions ensure CRE occupant exposures to control room air treatment system shall be radiological, chemical, and smoke hazards will not exceed demonstrated.

limits; and within 90 days, restore the CRE boundary to operable status. g. In accordance with the frequency and specifications of the Control Room Envelope

g. If Specifications 3.4.5.e or 3.4.5.f cannot be met during the Habitability Program, perform required CRE power operating condition, or when reactor coolant system unfiltered air inleakage testing.

temperature is greater than 212°F, reactor shutdown shall be initiated and the reactor shall be in cold shutdown within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

h. If Specification 3.4.5.e cannot be met whenever recently irradiated fuel or an irradiated fuel cask is being handled in the reactor building, or during OPDRVs: immediately suspend handling of recently irradiated fuel or the irradiated fuel cask in the reactor building; and immediately initiate action to suspend OPDRVs.

AMENDMENT NO. 142, 171, 194, 195, 201, 222, 179

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.2 PROTECTIVE INSTRUMENTATION 4.6.2 PROTECTIVE INSTRUMENTATION Applicability: Applicability:

Applies to the operability of the plant instrumentation Applies to the surveillance of the instrumentation that that performs a safety function. performs a safety function.

Objective: Objective:

To assure the operability of the instrumentation To verify the operability of protective required for safe operation. instrumentation.

Specification: Specification:

a. The set points, minimum number of trip systems, a. Sensors and instrument channels shall be and minimum number of instrument channels that checked, tested and calibrated at the frequency must be operable for each position of the reactor specified in the Surveillance Frequency Control mode switch shall be as given in Tables 3.6.2a to Program unless otherwise noted in Tables 4.6.2a 3.6.2l2m. to 4.6.2l2m.

If the requirements of a table are not met, the actions listed below for the respective type of instrumentation shall be taken.

(1) Instrumentation that initiates scram -

control rods shall be inserted, unless there is no fuel in the reactor vessel.

AMENDMENT NO. 142, 222, 194

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT (2) Primary Coolant and Containment b. Each trip system shall be tested each time the Isolation - Isolation valves shall be closed respective instrument channel is tested.

or the valves shall be considered inoperable and Specifications 3.2.7 and 3.3.4 shall be applied.

(3) Emergency Cooling Initiation or Isolation -

The emergency cooling system shall be considered inoperable and Specification 3.1.3 shall be applied.

(4) Core Spray Initiation - The core spray system shall be considered inoperable and Specification 3.1.4 shall be applied.

(5) Containment Spray Initiation - The containment spray system shall be considered inoperable and Specification 3.3.7 shall be applied.

(6) Auto Depressurization Initiation - The auto depressurization system shall be considered inoperable and Specification 3.1.5 shall be applied.

(7) Control Rod Withdrawal Block - No control rods shall be withdrawn.

AMENDMENT NO. 142, 143 195

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT (8) Mechanical Vacuum Pump Isolation -

The mechanical vacuum pump shall be isolated or the instrument channel shall be considered inoperable and Specification 3.6.1 shall be applied.

(9) Diesel Generator Initiation - The diesel generator shall be considered inoperable and Specification 3.6.3 shall be applied.

(10) Emergency Ventilation Initiation - The emergency ventilation system shall be considered inoperable and Specification 3.4.4 shall be applied.

(11) High Pressure Coolant Injection Initiation - The high pressure coolant injection system shall be considered inoperable and Specifi-cation 3.1.8.c shall be applied.

(12) Control Room Ventilation - The control room ventilation system shall be considered inoperable and Specification 3.4.5 shall be applied.

(13) Reactor Pressure Vessel Water Inventory Control - Specification 3.1.9 shall be applied for the applicable drain time.

AMENDMENT NO. 142, 143, 176, 196

TABLE 3.6.2b INSTRUMENTATION THAT INITIATES PRIMARY COOLANT SYSTEM OR CONTAINMENT ISOLATION Limiting Condition for Operation Minimum No. of Operable Instrument Reactor Mode Switch Minimum No. Channels per Position in Which of Tripped or Operable Function Must Be Parameter Operable Trip Systems Trip System Set Point Operable Shutdown Refuel Startup Run PRIMARY COOLANT ISOLATION (Main Steam, Cleanup, and Shutdown Cooling)

(1) Low-Low Reactor Water Level (a) Main Steam and Cleanup 2 2(f) 5 inches (i)(k) x x (Indicator Scale)

(b) Shutdown Cooling 2(j) 2(f)(j) 5 inches X(k) x x x (Indicator Scale)

(2) Manual 2 1 --- X(k) x x x MAIN-STEAM-LINE ISOLATION (3) High Steam Flow Main-Steam Line 2 2(f) 105 psid x x AMENDMENT NO. 142, 197, 205

NOTES FOR TABLES 3.6.2b and 4.6.2b (a) May be bypassed in the refuel and startup positions of the reactor mode switch when reactor pressure is less than 600 psi.

(b) May be bypassed when necessary for containment inerting.

(c) May be bypassed in the shutdown mode whenever the reactor coolant system temperature is less than 215F.

(d) The trip circuit will be calibrated and tested in accordance with the Surveillance Frequency Control Program, the primary sensor will be calibrated and tested in accordance with the Surveillance Frequency Control Program.

(e) Deleted.

(f) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for required surveillances without placing the Trip System in the tripped condition provided at least one Operable Instrument Channel in the same Trip System is monitoring that Parameter.

With the number of Operable Channels one less than required by the Minimum Number of Operable Instrument Channels per Operable Trip System requirement for one trip system, either

1. Place the inoperable channel(s) in the tripped condition within

a. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for Parameters common to SCRAM Instrumentation, and

b. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for Parameters not common to SCRAM Instrumentation.

or

2. Take the ACTION required by Specification 3.6.2a for that Parameter.

With the number of Operable Channels one less than required by the Minimum Number of Operable Instrument Channels per Operable Trip System requirement for both trip systems,

1. Place the inoperable channel(s) in one trip system in the tripped condition within one hour.

and

2. a. Place the inoperable channel(s) in the remaining trip system in the tripped condition within (1) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for Parameters common to SCRAM Instrumentation, and (2) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for Parameters not common to SCRAM Instrumentation.

or

b. take the ACTION required by Specification 3.6.2a for that Parameter.

AMENDMENT NO. 142, 222 212

NOTES FOR TABLES 3.6.2b and 4.6.2b (g) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for required surveillances without placing the Trip System in tripped condition provided at least one Operable Instrument Channel in the same Trip System is monitoring that Parameter.

With the number of Operable channels one less than required by the Minimum Number of Operable Instrument Channels for the Operable Trip System, either

1. Place the inoperable channel(s) in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

or

2. Take the ACTION required by Specification 3.6.2a for that Parameter.

(h) Only applicable during startup mode while operating in IRM range 10.

(i) May be bypassed in the cold shutdown condition.

(j) In the cold shutdown and refueling conditions, only one Operable Trip System is required provided shutdown cooling system integrity is maintained. With one of the two required Operable Channels in the required Trip System not operable, place the inoperable channel in the tripped condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Otherwise, either:

1. Immediately initiate action to restore the channel to operable status.

or Immediately initiate action to isolate the shutdown cooling system. Deleted (k) The Primary Coolant Isolation Parameters for Cleanup and Shutdown Cooling in Table 3.6.2b are only applicable in the Shutdown Condition - Hot. See Table 3.6.2m for Parameter applicability in the Shutdown Condition - Cold.

Note 1: Surveillance intervals are specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.6.2b.

AMENDMENT NO. 142, 153, 197, 222 213

TABLE 3.6.2d INSTRUMENTATION THAT INITIATES CORE SPRAY(e)

Limiting Condition for Operation Minimum No. of Operable Instrument Reactor Mode Switch Minimum No. Channels per Position in Which of Tripped or Operable Function Must Be Parameter Operable Trip Systems Trip System (f) Set Point Operable Shutdown Refuel Startup Run START CORE SPRAY PUMPS (1) High Drywell Pressure 2 2 3.5 psig (d)(g) x (a) (a)

(2) Low-Low Reactor Water Level 2 2 5 inches (b)(g) x x x (Indicator Scale)

OPEN CORE SPRAY DISCHARGE VALVES (3) Reactor Pressure and either (1) or 2 2 365 psig x(g) x x x (2) above.

AMENDMENT NO. 142, 217

NOTES FOR TABLES 3.6.2d AND 4.6.2d (a) May be bypassed when necessary for containment inerting.

(b) May be bypassed when necessary for performing major maintenance as specified in Specification 2.1.1.e.

(c) The trip circuit will be calibrated and tested in accordance with the Surveillance Frequency Control Program, the primary sensor will be calibrated and tested in accordance with the Surveillance Frequency Control Program.

(d) May be bypassed when necessary for integrated leak rate testing.

(e) The instrumentation that initiates the Core Spray System is not required to be operable, if there is no fuel in the reactor vessel.

(f) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for required surveillances without placing the Trip System in the tripped condition provided at least one Operable Instrument Channel in the same Trip System is monitoring that parameter.

With the number of Operable channels less than required by the Minimum Number of Operable Instrument Channels per Operable Trip System requirement:

1. With one channel inoperable, place the inoperable channel in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or take the ACTION required by Specification 3.6.2a for that Parameter.

2. With more than one channel inoperable, take the ACTION required by Specification 3.6.2a for that Parameter.

(g) The Parameters for Start Core Spray Pumps and Open Core Spray Discharge Valves in Table 3.6.2d are only applicable in the Shutdown Condition - Hot. See Table 3.6.2m for Parameter applicability in the Shutdown Condition - Cold.

Note 1: Surveillance intervals are specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.6.2d.

AMENDMENT NO. 142, 222, 219

NOTES FOR TABLES 3.6.2j AND 4.6.2j (a) This function shall be operable whenever recently irradiated fuel or an irradiated fuel cask is being handled in the reactor building, and during operations with a potential for draining the reactor vessel (OPDRVs).

(b) Deleted.

(c) Immediately prior to when function is required and in accordance with the Surveillance Frequency Control Program thereafter until function is no longer required.

(d) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for required surveillances without placing the Trip System in the tripped condition provided at least one Operable Instrument Channel in the same Trip System is monitoring that parameter.

With the number of Operable channels one less than required by the Minimum Number of Operable Instrument Channels for the Operable Trip System, either

1) Place the inoperable channel(s) in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

or

2) Take the ACTION required by Specification 3.6.2a for that Parameter.

Note 1: Surveillance intervals are specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.6.2j.

AMENDMENT NO. 142, 194, 222, 242

TABLE 3.6.2m RPV WATER INVENTORY CONTROL INSTRUMENTATION Limiting Condition for Operation Minimum No. of Operable Instrument Reactor Mode Switch Minimum No. Channels per Position in Which of Tripped or Operable Function Must Be Parameter Operable Trip Systems Trip System Set Point Operable Shutdown Refuel Startup Run START CORE SPRAY PUMPS (c)

(1) Manual 1 1 (d) --- (a)(b) (a)(b)

OPEN CORE SPRAY DISCHARGE VALVES (c)

(2) Reactor Pressure and (1) above. 1 per 1 per pump (d) 365 psig (a) (a) pump AMENDMENT NO. 247b

TABLE 3.6.2m RPV WATER INVENTORY CONTROL INSTRUMENTATION Limiting Condition for Operation Minimum No. of Operable Instrument Reactor Mode Switch Minimum No. Channels per Position in Which of Tripped or Operable Function Must Be Parameter Operable Trip Systems Trip System Set Point Operable Shutdown Refuel Startup Run PRIMARY COOLANT ISOLATION (1) Low-Low Reactor Water Level (a) Cleanup 2 2(f) 5 inches (a) (a)

(Indicator Scale)

(b) Shutdown Cooling 2(e) 2(e)(f) 5 inches (a) (a)

(Indicator Scale)

(2) Manual 2 1 --- (a) (a)

AMENDMENT NO. 247c

TABLE 4.6.2m RPV WATER INVENTORY CONTROL INSTRUMENTATION Surveillance Requirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration START CORE SPRAY PUMPS (1) Manual Note 1 Note 1(g) Note 1(g)

OPEN CORE SPRAY DISCHARGE VALVES (2) Reactor Pressure and Note 1 Note 1(g) Note 1(g)

(1) above AMENDMENT NO. 247d

TABLE 4.6.2m RPV WATER INVENTORY CONTROL INSTRUMENTATION Surveillance Requirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration PRIMARY COOLANT ISOLATION (Cleanup and Shutdown Cooling)

(1) Low-Low Reactor Water Note 1 Note 1(g) Note 1(g)

Level (2) Manual --- Note 1 ---

AMENDMENT NO. 247e

NOTES FOR TABLES 3.6.2m AND 4.6.2m (a) The Parameters in this table are only applicable in the Shutdown Condition - Cold and Refuel. See Table 3.6.2b or Table 3.6.2d for Parameter applicability in the Shutdown Condition - Hot.

(b) May be bypassed when necessary for performing major maintenance as specified in Specification 2.1.1.e.

(c) The instrumentation that initiates the Core Spray System is not required to be operable if there is no fuel in the reactor vessel.

(d) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for required surveillances without placing the Trip System in the tripped condition provided at least one Operable Instrument Channel in the same Trip System is monitoring that parameter.

With the number of Operable channels less than required by the Minimum Number of Operable Instrument Channels per Operable Trip System requirement:

1. With one channel inoperable, place the inoperable channel in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or take the ACTION required by Specification 3.6.2a for that Parameter.

2. With more than one channel inoperable, take the ACTION required by Specification 3.6.2a for that Parameter.

(e) In the cold shutdown and refueling conditions, only one Operable Trip System is required provided shutdown cooling system integrity is maintained.

With one of the two required Operable Channels in the required Trip System not operable, place the inoperable channel in the tripped condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , otherwise

1. Declare associated penetration flow path(s) incapable of automatic isolation, and

2. Calculate drain time, and

3. a. Immediately initiate action to restore the channel to operable status, or b Immediately initiate action to isolate the shutdown cooling system.

AMENDMENT NO. 247f

NOTES FOR TABLES 3.6.2m AND 4.6.2m (f) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for required surveillances without placing the Trip System in the tripped condition provided at least one Operable Instrument Channel in the same Trip System is monitoring that Parameter.

With the number of Operable Channels one less than required by the Minimum Number of Operable Instrument Channels per Operable Trip System requirement for one trip system:

1. Declare associated penetration flow path(s) incapable of automatic isolation, and

2. Calculate drain time, and

3. Place the inoperable channel(s) in the tripped condition within,

a. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for Parameters common to SCRAM Instrumentation, and

b. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for Parameters not common to SCRAM Instrumentation, or

4. Take the ACTION required by Specification 3.6.2a for that Parameter.

With the number of Operable Channels one less than required by the Minimum Number of Operable Instrument Channels per Operable Trip System requirement for both trip systems:

1. Declare associated penetration flow path(s) incapable of automatic isolation, and

2. Calculate drain time, and

3. Place the inoperable channel(s) in one trip system in the tripped condition within one hour, and

4. Place the inoperable channel(s) in the remaining trip system in the tripped condition within,

a. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for Parameters common to SCRAM Instrumentation, and

b. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for Parameters not common to SCRAM Instrumentation, or

5. Take the ACTION required by Specification 3.6.2a for that Parameter.

AMENDMENT NO. 247g

NOTES FOR TABLES 3.6.2m AND 4.6.2m (g) The trip circuit will be calibrated and tested in accordance with the Surveillance Frequency Control Program, the primary sensor will be calibrated and tested in accordance with the Surveillance Frequency Control Program.

Note 1: Surveillance intervals are specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.6.2m.

AMENDMENT NO. 247h

ATTACHMENT 3 License Amendment Request Nine Mile Point Nuclear Station, Unit 1 Docket No. 50-220 Revise Technical Specifications to Adopt TSTF-542, "Reactor Pressure Vessel Water Inventory Control," Revision 2 Revised Technical Specification Pages TS Clean Pages TOG 175 Ba 178 54 179 55 194 56 196 79a thru 79c 205 109 213 164 217 165 219 168 242 171 247b thru 247h 174

NINE MILE POINT NUCLEAR STATION UNIT 1 - TECHNICAL SPECIFICATIONS CONTENTS SECTION DESCRIPTION PAGE 1.0 Definitions 1 2.0 Safety Limits and Limiting Safety System Setting Safety Limits Limiting Safety System Setting 2.1.1 Fuel Cladding Integrity 2.1.2 Fuel Cladding Integrity 9 2.2.1 Reactor Coolant System 2.2.2 Reactor Coolant System 23 3.0 Limiting Condition for Operation (LCO) and Surveillance Requirement (SR) Applicability 27 3.1.0 Fuel Cladding 28 Limiting Condition for Operation Surveillance Requirements 3.1.1 Control Rod System 4.1.1 Control Rod System 29 3.1.2 Liquid Poison System 4.1.2 Liquid Poison System 44 3.1.3 Emergency Cooling System 4.1.3 Emergency Cooling System 50

3. 1.4 Core Spray System 4.1.4 Core Spray System 54 3.1.5 Solenoid-Actuated Pressure Relief Valve 4.1.5 Solenoid-Actuated Pressure Relief Valve 60 3.1.6 Control Rod Drive Coolant Injection 4.1.6 Control Rod Drive Coolant Injection 62 3.1. 7 Fuel Rods 4.1.7 Fuel Rods 65 3.1.8 High Pressure Coolant Injection 4.1.8 High Pressure Coolant Injection 76 3.1.9 Reactor Pressure Vessel (RPV) Water Inventory 4.1.9 Reactor Pressure Vessel (RPV) Water Inventory Control Control 79a AMENDMENT NO. 142. 182,

1.34 Drain Time The drain time is the time it would take for the water inventory in and above the Reactor Pressure Vessel (RPV} to drain to

-10 inches indicator scale (74 inches above the top of the active fuel seated in the RPV} assuming:

a. The water inventory above -1 O inches indicator scale is divided by the limiting drain rate,

b. The limiting drain rate is the larger of the drain rate through a single penetration flow path with the highest flow rate, or the sum of the drain rates through multiple penetration flow paths susceptible to a common Mode failure (e.g.,

seismic event, loss of normal power, single human error}, for all penetration flow paths below -1 O inches indicator scale except:

1. Penetration flow paths connected to an intact closed system, or isolate by manual or automatic valves that are locked, sealed, or otherwise secured in the closed position, blank flanges or other devices that prevent flow of reactor coolant through the penetration flow paths;

2. Penetration flow paths capable of being isolated by valves that will close automatically without offsite power prior to the RPV water level being equal to -10 inches indicator scale when actuated by RPV water level isolation instrumentation; or

3. Penetration flow paths with isolation devices that can be closed prior to the RPV water level being equal to the

-1 O inches indicator scale by a dedicated operator trained in the task, who is in continuous communication with the control room, is stationed at the controls, and is capable of closing the penetration flow path isolation device without offsite power.

c. The penetration flow paths required to be evaluated per paragraph b are assumed to open instantaneously and are not subsequently isolated, and no water is assumed to be subsequently added to the RPV water inventory;

d. No additional draining events occur; and

e. Realistic cross-sectional areas and drain rates are used.

A bounding drain time may be used in lieu of a calculated value.

AMENDMENT NO. Ba

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.1.4 CORE SPRAY SYSTEM 4.1.4 CORE SPRAY SYSTEM Applicability: Applicability:

Applies to the operating status of the core spray Applies to the periodic testing requirements for the systems when in the Power Operating Condition or core spray systems.

Shutdown Condition - Hot.

Objective:

To verify the operability of the core spray systems.

To assure the capability of the core spray systems to cool reactor fuel in the event of a loss-of-coolant accident.

Specification:

The core spray system surveillance shall be

a. Whenever irradiated fuel is in the reactor vessel performed as indicated below.

and the reactor coolant temperature is greater than 212°F, each of the two core spray systems a. In accordance with the Surveillance Frequency shall be operable except as specified in Control Program automatic actuation of each Specifications b and c below. subsystem in each core spray system shall be demonstrated.

b. If a redundant component of a core spray system becomes inoperable, that system shall be b. In accordance with the Surveillance Frequency considered operable provided that the component Control Program pump operability shall be is returned to an operable condition within 7 days checked.

and the additional surveillance required is performed. c. In accordance with the Surveillance Frequency Control Program the operability of power-

c. If a redundant component in each of the core operated valves required for proper system spray systems becomes inoperable, both systems operation shall be checked.

shall be considered operable provided that the component is returned to an operable condition within 7 days and the additional surveillance required is performed.

AMENDMENT NO. 142. 222 54

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

d. If Specifications a, b and c are not met, a normal d. {Deleted) orderly shutdown shall be initiated within one hour and the reactor shall be in the cold shutdown condition within ten hours.

e. During reactor operation, except during core e. Surveillance with Inoperable Components spray system surveillance testing, core spray isolation valves 40-02 and 40-12 shall be in the When a component becomes inoperable its open position and the associated valve motor redundant component or system shall be verified starter circuit breakers for these valves shall be to be operable immediately and in accordance with locked in the off position. In addition, redundant the Surveillance Frequency Control Program valve position indication shall be available in the thereafter.

control room.

f. (Deleted) f. With a core spray subsystem suction from the CST, CST level shall be checked in accordance with the Surveillance Frequency Control Program.

g. (Deleted) g. In accordance with the Surveillance Frequency Control Program when the reactor coolant temperature is greater than 212°F, verify that the piping system between valves 40-03, 13 and 40-01 , 09, 10, 11 is filled with water.

AMENDMENT NO. 142. 192. 222, 55

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

h. (Deleted)

i. With the downcomers in the suppression chamber having less than three and one half foot submergence, two core spray subsystems and the associated raw water pumps shall be operable with the core spray suction from the condensate storage tanks (CST), and the CST inventory shall not be less than 300,000 gallons.

AMENDMENT NO. ~. 56

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.1.9 Reactor Pressure Vessel (RPV) Water Inventory 4.1.9 Reactor Pressure Vessel (RPV) Water Inventory Control Control Applicability: Applicability:

Applies to the operating status of the core spray systems Applies to the periodic testing requirements for the and Reactor Water Inventory Control when the reactor core spray system and RPV water inventory.

coolant temperature is less than or equal to 212°F.

Objective:

To assure the RPV water inventory is maintained -1 O To verify the operability of the core spray system inches indicator scale. and RPV water inventory.

Specification: Specification:

a. Whenever irradiated fuel is in the reactor vessel a. Verify drain time <::36 hours in accordance and the reactor coolant temperature is less than or with the Surveillance Frequency Control equal to 212°F, drain time of RPV water inventory to Program.

-1 O inches indicator scale shall be <::36 hours and one core spray subsystem shall be operable except b. Verify, for a required core spray subsystem, the as specified in Specifications b through f below. downcomers in the suppression chamber have greater than or equal to three and one half foot

b. If the required core spray subsystem becomes of submergence or the condensate storage inoperable, the component shall be returned to an tank inventory is not less than 300,000 gallons, operable condition within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . in accordance with the Surveillance Frequency Control Program.

c. If Specifications a and b are not met, then immediately initiate action to establish a method of c. Verify for the required core spray subsystem, water injection capable of operating without offsite each manual power operated and automatic electrical power. valve in the flow path, that is not locked, sealed or otherwise secured in position, is in the correct position, in accordance with the Surveillance Frequency Control Program.

AMENDMENT NO, 79a

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

d. If drain time <36 hours and ~8 hours, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months d. Verify each valve credited for automatically perform the following actions: isolating a penetration flow path actuates to the isolation position on an actual or simulated (1) Verify secondary containment boundary is isolation signal, in accordance with the capable of being established in less than the Surveillance Frequency Control Program.

drain time,

e. Verify the required core spray subsystem actuates and on a manual initiation signal, in accordance with the (2) Verify each secondary containment penetration flow Surveillance Frequency Control Program. Vessel path is capable of being isolated in less than the spray may be excluded.

drain time, and (3) Verify one RBEVS is capable of being placed in operation in less than the drain time.

e. If drain time <8 hours, immediately perform the following actions:

(1) Initiate action to establish an additional method of water injection with water sources capable of maintaining RPV water level above -10 inches indicator scale for ~36 hours, and (2) Initiate action to establish secondary containment boundary, and (3) Initiate action to isolate each secondary containment penetration flow path or verify it can be manually isolated from the control room, and (4) Initiate action to verify one RBEVS is capable of being placed in operation.

AMENDMENT NO. 79b

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

f. Specifications d and e not met, or drain time is

<1 hour, immediately initiate action to restore drain time to ~36 hours.

AMENDMENT NO. 79c

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

c. If Specifications 3.2.7a and b above are not met, c. In accordance with the Surveillance Frequency initiate normal orderly shutdown within one hour Control Program the feedwater and main-steam and have reactor in the cold shutdown condition line power-operated isolation valves shall be within ten hours. exercised by partial closure and subsequent reopening.

d. Whenever fuel is in the reactor vessel and the reactor coolant temperature is less than or equal to 212°F, d. At least once per plant cold shutdown the the isolation valves on the shutdown cooling system feedwater and main steam line power-operated lines connected to the reactor coolant system shall be isolation valves shall be fully closed and operable except as specified in Specification 3.2.7.e reopened, unless this test has been performed below. within the previous 92 days.

e. In the event any shutdown cooling system isolation valve becomes inoperable whenever fuel is in the reactor vessel and the reactor coolant temperature is less than or equal to 212°F, the system shall be considered operable provided that, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , at least one valve in each line having an inoperable valve is in the mode corresponding to the isolated condition.

f. If Specifications 3.2.7.d and 3.2.7.e above are not met, immediately initiate action to restore the valve(s) to operable status.

AMENDMENT NO. 142. 197. 222, 109

3.4.0 REACTOR BUILDING APPLICABILITY Applies to the operating status of the reactor building (secondary containment).

OBJECTIVE To assure the integrity of the reactor building.

SPECIFICATION Reactor building integrity must be in effect for the following conditions:

a. Power operating condition,

b. When the reactor water temperature is above 215°F, or

c. Whenever recently irradiated fuel or an irradiated fuel cask is being handled in the Reactor Building.

AMENDMENT NO. 142. 170. 194, 164

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.4.1 LEAKAGE RATE 4.4.1 LEAKAGE RATE Applicability: Applicability:

Applies to the leakage rate of the secondary Applies to the periodic testing requirements of the containment. secondary containment leakage rate.

Objective: Objective:

To specify the requirements necessary to limit exfiltration of To assure the capability of the secondary containment fission products released to the to maintain leakage within allowable limits.

secondary containment as a result of an accident.

Specification:

In accordance with the Surveillance Frequency Control At all times when secondary containment integrity is Program - isolate the reactor building and start required, the reactor building leakage rate as determined by emergency ventilation system fan to demonstrate Specification 4.4.1 shall not exceed negative pressure in the building relative to external 1600 cfm. If this cannot be met after a routine surveillance static pressure. The fan flow rate shall be varied so check, then the actions listed below shall be taken: that the building internal differential pressure is at least as negative as that on Figure 3.4.1 tor the wind speed

a. Suspend any of the following activities: at which the test is conducted. The fan flow rate represents the reactor building leakage referenced to

1. Handling of recently irradiated fuel in the reactor zero mph with building internal pressure at least 0.25 building. inch of water less than atmospheric pressure. The test shall be done at wind speeds less than 20 miles

2. Irradiated fuel cask operations in the reactor per hour.

building.

b. Restore the reactor building leakage rates to within specified limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or initiate normal orderly shutdown and be in a cold shutdown condition within 1O hours.

AMENDMENT NO. 142. 156. 170. 194. 222, 165

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.4.2 REACTOR BUILDING INTEGRITY - ISOLATION 4.4.2 REACTOR BUILDING INTEGRITY - ISOLATION VALVES VALVES Applicability: Applicability:

Applies to the operational status of the reactor Applies to the periodic testing requirements of the building isolation valves. reactor building isolation valves.

Objective: Objective:

To assure that fission products released to the secondary To assure the operability of the reactor building containment are discharged to the environment in a isolation valves.

controlled manner using the emergency ventilation system.

Specification: Specification:

a. The normal Ventilation System isolation valves In accordance with the Surveillance Frequency Control shall be operable at all times when secondary Program, automatic initiation of valves shall be checked.

containment integrity is required.

b. If Specification 3.4.2.a is not met, then the actions listed below shall be taken:

1. The reactor shall be in the cold shutdown condition within ten hours.

2. Suspend any of the following activities:

a. Handling of recently irradiated fuel in the reactor building,

b. Irradiated fuel cask handling operations in the reactor building.

AMENDMENT NO. 142. 170. 1 Q4. 222, 168

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

b. If these conditions cannot be met, then the actions b. Verify in accordance with the Surveillance listed below shall be taken: Frequency Control Program that

1. If in the power operating condition, restore 1. At least one door in each access to the reactor building integrity within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be secondary containment is closed, except when in at least the hot shutdown condition within the access opening is being used for entry and the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in the cold shutdown exit.

condition within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2. At least one door or closeup of the railroad OR bay is closed.

If the reactor coolant system temperature is above 215°F, restore reactor building integrity within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in cold shutdown within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2. Suspend any of the following activities:

a. Handling of recently irradiated fuel in the reactor building,

b. Irradiated fuel cask handling operations in the reactor building.

AMENDMENT NO. 142. 170. 1Q4. 222. 223, 171

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

c. The results of laboratory carbon sample analysis b. The tests and sample analysis of Specification shall show ~95% radioactive methyl iodide removal 3.4.4b, c and d shall be performed in when tested in accordance with ASTM D3803-1989 accordance with the Surveillance Frequency at 30°C and 95% R.H. Control Program, or after 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation, whichever occurs first or following

d. Fans shall be shown to operate within +/- 10% significant painting, fire or chemical release in design flow. any ventilation zone communicating with the system.

e. During reactor operation, including when the reactor coolant system temperature is above c. Cold DOP testing shall be performed after each 215°F, from and after the date that one circuit complete or partial replacement of the HEPA of the emergency ventilation system is made or filter bank or after any structural maintenance found to be inoperable for any reason, reactor on the system housing.

operation is permissible only during the succeeding seven days unless such circuit is d. Halogenated hydrocarbon testing shall be sooner made operable, provided that during such performed after each complete or partial seven days all active components of the other replacement of the charcoal adsorber bank or emergency ventilation circuit shall be operable. after any structural maintenance on the system housing.

During handling of recently irradiated fuel in the reactor building or handling of an irradiated fuel e. Each circuit shall be operated at least 15 minutes cask in the reactor building from and after the in accordance with the Surveillance Frequency date that one circuit of the emergency ventilation Control Program.

system is made or found to be inoperable for any reason, recently irradiated fuel handling in the f. Test sealing of gaskets for housing doors reactor building or irradiated fuel cask handling in downstream of the HEPA filters and charcoal the reactor building is permissible during the adsorbers shall be performed at and in succeeding seven days unless such circuit is conformance with each test performed for sooner made operable, provided that compliance with Specification 4.4.4b and Specification 3.4.4b.

AMENDMENT NO. 142. 140. 170. 171. 194. 201. 222, 174

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT during such seven days all active components of g. In accordance with the Surveillance Frequency the other emergency ventilation circuit shall be Control Program, automatic initiation of each operable. Recently irradiated fuel handling in the branch of the emergency ventilation system reactor building or irradiated fuel cask handling in shall be demonstrated.

the reactor building may continue beyond seven days provided the operable emergency ventilation h. In accordance with the Surveillance Frequency circuit is in operation. Control Program, manual operability of the bypass valve for filter cooling shall be demonstrated.

f. If these conditions cannot be met, within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months , the reactor shall be placed in a condition i. When one circuit of the emergency ventilation for which the emergency ventilation system is system becomes inoperable all active compon-not required. ents in the other emergency ventilation circuit shall be verified to be operable within two hours and in accordance with the Surveillance Frequency Control Program thereafter.

AMENDMENT NO. 142. 194. 222, 175

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.4.5 CONTROL ROOM AIR TREATMENT SYSTEM 4.4.5 CONTROL ROOM AIR TREATMENT SYSTEM Applicability: Applicability:

Applies to the operating status of the control room air treatment Applies to the testing of the control room air treatment system and Control Room Envelope (CRE) boundary. system and GAE boundary.

NOTE----------------------------

The GAE boundary may be opened intermittently under administrative control.

Objective: Objective:

To assure the capability of the control room air treatment To assure the operability of the control room air treatment system to minimize the amount of radioactivity or other system.

gases entering the control room in the event of an incident.

Specification:

a. In accordance with the Surveillance Frequency Control

a. Except as specified below, the control room air treatment Program, the pressure drop across the combined system shall be operable for the following conditions: HEPA filters and charcoal adsorber banks shall be demonstrated to be less than 1.5 inches of water at

1. Power operating condition, and whenever the system design flow rate {+/- 10%).

reactor coolant system temperature is greater than 212°F. b. The tests and sample analysis of Specification 3.4.5b, c and d shall be performed in accordance with

2. Whenever recently irradiated fuel or an irradiated the Surveillance Frequency Control Program, or after fuel cask is being handled in the reactor building. 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation, whichever occurs first or following significant painting, fire or chemical release any ventilation zone communicating with the system.

b. The results of the in-place cold DOP and halogenated hydrocarbon tests at design flows on HEPA filters and charcoal adsorber banks shall show ::: 99% DOP removal and ::: 99% halogenated hydrocarbon removal when tested in accordance with ANSI N.510-1980.

AMENDMENT NO. 142. 170. 171. 179. 194. 195. 222, 178

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

c. The results of laboratory carbon sample analysis shall c. Cold DOP testing shall be performed after each show ;;::95% radioactive methyl iodine removal when complete or partial replacement of the HEPA filter tested in accordance with ASTM 03803-1989 at 30°C bank or after any structural maintenance on the and 95% R.H. system housing.

d. Fans shall be shown to operate within +/-10% design flow. d. Halogenated hydrocarbon testing shall be performed after each complete or partial

e. From and after the date that the control room air treatment replacement of the charcoal absorber bank or system is made or found to be inoperable for any reason, after any structural maintenance on the system except for an inoperable CRE boundary during the power housing.

operating condition, restore the system to operable within the succeeding seven days. e. The system shall be operated at least 15 minutes in accordance with the Surveillance Frequency Control

f. If the control room air treatment system is made or found Program.

to be inoperable due to an inoperable CRE boundary during the power operating condition: immediately initiate f. In accordance with the Surveillance Frequency action to implement mitigating actions; within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , Control Program, automatic initiation of the verify mitigating actions ensure CRE occupant exposures control room air treatment system shall be to radiological, chemical, and smoke hazards will not demonstrated.

exceed limits; and within 90 days, restore the CRE boundary to operable status. g. In accordance with the frequency and specifications of the Control Room Envelope

g. If Specifications 3.4.5.e or 3.4.5.f cannot be met during the Habitability Program, perform required CRE power operating condition, or when reactor coolant system unfiltered air inleakage testing.

temperature is greater than 212°F, reactor shutdown shall be initiated and the reactor shall be in cold shutdown within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

h. If Specification 3.4.5.e cannot be met whenever recently irradiated fuel or an irradiated fuel cask is being handled in the reactor building, immediately suspend handling of recently irradiated fuel or the irradiated fuel cask in the reactor building.

AMENDMENT NO. 142. 171. 1Q4. 1Q5. 201. 222, 179

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.2 PROTECTIVE INSTRUMENTATION 4.6.2 PROTECTIVE INSTRUMENTATION Applicability: Applicability:

Applies to the operability of the plant instrumentation Applies to the surveillance of the instrumentation that that performs a safety function. performs a safety function.

Objective: Objective:

To assure the operability of the instrumentation To verify the operability of protective required for safe operation. instrumentation.

Specification: Specification:

a. The set points, minimum number of trip systems, a. Sensors and instrument channels shall be and minimum number of instrument channels that checked, tested and calibrated at the frequency must be operable for each position of the reactor specified in the Surveillance Frequency Control mode switch shall be as given in Tables 3.6.2a to Program unless otherwise noted in Tables 4.6.2a 3.6.2m. to 4.6.2m.

If the requirements of a table are not met, the actions listed below for the respective type of instrumentation shall be taken.

(1) Instrumentation that initiates scram -

control rods shall be inserted, unless there is no fuel in the reactor vessel.

AMENDMENT NO. 142. 222. 194

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT (8) Mechanical Vacuum Pump Isolation -

The mechanical vacuum pump shall be isolated or the instrument channel shall be considered inoperable and Specification 3.6.1 shall be applied.

(9) Diesel Generator Initiation - The diesel generator shall be considered inoperable and Specification 3.6.3 shall be applied.

(10) Emergency Ventilation Initiation - The emergency ventilation system shall be considered inoperable and Specification 3.4.4 shall be applied.

(11) High Pressure Coolant Injection Initiation -

The high pressure coolant injection system shall be considered inoperable and Specification 3.1.8.c shall be applied.

(12) Control Room Ventilation - The control room ventilation system shall be considered inoperable and Specification 3.4.5 shall be applied.

(13) Reactor Pressure Vessel Water Inventory Control - Specification 3.1.9 shall be applied for the applicable drain time.

AMENDMENT NO. 142. 143. 176, 196

TABLE 3.6.2b INSTRUMENTATION THAT INITIATES PRIMARY COOLANT SYSTEM OR CONTAINMENT ISOLATION Limiting Condition for Operation Minimum No. of Operable Instrument Reactor Mode Switch Minimum No. Channels per Position in Which of Tripped or Operable Function Must Be Parameter Operable Trip Systems Trip System Set Point Operable Shutdown Refuel Startup Run PRIMARY COOLANT ISOLATION (Main Steam, Cleanup, and Shutdown Cooling)

(1) Low-Low Reactor Water Level (a) Main Steam and Cleanup 2 2(f) 5 inches (k) x x (Indicator Scale)

(b) Shutdown Cooling 2 2(f) 5 inches (k) x x (Indicator Scale)

(2) Manual 2 1 --- (k) x x MAIN-STEAM-LINE ISOLATION (3) High Steam Flow Main-Steam Line 2 2(f) 105 psid x x AMENDMENT NO. 142, 197, 205

NOTES FOR TABLES 3.6.2b and 4.6.2b (g) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for required surveillances without placing the Trip System in tripped condition provided at least one Operable Instrument Channel in the same Trip System is monitoring that Parameter.

With the number of Operable channels one less than required by the Minimum Number of Operable Instrument Channels for the Operable Trip System, either

1. Place the inoperable channel(s) in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

or

2. Take the ACTION required by Specification 3.6.2a for that Parameter.

(h) Only applicable during startup mode while operating in IRM range 10.

(i) May be bypassed in the cold shutdown condition.

G) Deleted.

(k) The Primary Coolant Isolation Parameters for Cleanup and Shutdown Cooling in Table 3.6.2b are only applicable in the Shutdown Condition - Hot. See Table 3.6.2m for Parameter applicability in the Shutdown Condition - Cold.

Note 1: Surveillance intervals are specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.6.2b.

AMENDMENT NO. 142. 15a. 1Q7. 222, 213

TABLE 3.6.2d INSTRUMENTATION THAT INITIATES CORE SPRAY<El Limiting Condition for Operation Minimum No. of Operable Instrument Reactor Mode Switch Minimum No. Channels per Position in Which of Tripped or Operable Function Must Be Parameter Operable Trip Systems Trip System (f) Set Point Operable c

-0 "C

s

.c en Ci)

-::::s Q) cc Q.

s

...ent:m c

s cc START CORE SPRAY PUMPS (1) High Drywell Pressure 2 2 :S 3.5 psig (d)(g) x (a) (a)

(2) Low-Low Reactor Water Level 2 2 ~ 5 inches (b)(g) x x x (Indicator Scale)

OPEN CORE SPRAY DISCHARGE VALVES (3) Reactor Pressure and either (1) or 2 2 ~ 365 psig (g) x x x (2) above.

AMENDMENT NO. +42, 217

NOTES FOR TABLES 3.6.2d AND 4.6.2d (a) May be bypassed when necessary for containment inerting.

(b) May be bypassed when necessary for performing major maintenance as specified in Specification 2.1.1.e.

(c) The trip circuit will be calibrated and tested in accordance with the Surveillance Frequency Control Program, the primary sensor will be calibrated and tested in accordance with the Surveillance Frequency Control Program.

{d) May be bypassed when necessary for integrated leak rate testing.

(e) The instrumentation that initiates the Core Spray System is not required to be operable, if there is no fuel in the reactor vessel.

(f) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for required surveillances without placing the Trip System in the tripped condition provided at least one Operable Instrument Channel in the same Trip System is monitoring that parameter.

With the number of Operable channels less than required by the Minimum Number of Operable Instrument Channels per Operable Trip System requirement:

1. With one channel inoperable, place the inoperable channel in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or take the ACTION required by Specification 3.6.2a for that Parameter.

2. With more than one channel inoperable, take the ACTION required by Specification 3.6.2a for that Parameter.

(g) The Parameters for Start Core Spray Pumps and Open Core Spray Discharge Valves in Table 3.6.2d are only applicable in the Shutdown Condition - Hot. See Table 3.6.2m for Parameter applicability in the Shutdown Condition - Cold.

Note 1: Surveillance intervals are specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.6.2d.

AMENDMENT NO. 142. 222, 219

NOTES FOR TABLES 3.6.2j AND 4.6.2j (a) This function shall be operable whenever recently irradiated fuel or an irradiated fuel cask is being handled in the reactor building.

(b) Deleted.

(c) Immediately prior to when function is required and in accordance with the Surveillance Frequency Control Program thereafter until function is no longer required.

(d) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for required surveillances without placing the Trip System in the tripped condition provided at least one Operable Instrument Channel in the same Trip System is monitoring that parameter.

With the number of Operable channels one less than required by the Minimum Number of Operable Instrument Channels tor the Operable Trip System, either

1) Place the inoperable channel(s) in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

or

2) Take the ACTION required by Specification 3.6.2a tor that Parameter.

Note 1: Surveillance intervals are specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.6.2j.

AMENDMENT NO. 142. 194. 222, 242

TABLE 3.6.2m RPV WATER INVENTORY CONTROL INSTRUMENTATION Limiting Condition for Operation Minimum No. of Operable Instrument Reactor Mode Switch Minimum No. Channels per Position in Which of Tripped or Operable Function Must Be Parameter Operable Trip Systems Trip System Set Point Operable c

0 "C

J

.c en a;

J Cl) a: -

c.

J t:

m en c

J a:

START CORE SPRAY PUMPS (cl (1) Manual 1 1 (d) (a)(b) (a)(b)

OPEN CORE SPRAY DISCHARGE VALVES (cl (2) Reactor Pressure and (1) above. 1 per 1 per pump (d) 2: 365 psig (a) (a) pump AMENDMENT NO. 247b

TABLE 3.6.2m RPV WATER INVENTORY CONTROL INSTRUMENTATION Limiting Condition for Operation Minimum No. of Operable Instrument Reactor Mode Switch Minimum No. Channels per Position in Which of Tripped or Operable Function Must Be Parameter Operable Trip Systems Trip System Set Point Operable c

=0

I

.c en "iii

I G>

a:

~

a.

I (ii c

I a:

PRIMARY COOLANT ISOLATION (1) Low-Low Reactor Water Level (a) Cleanup 2(f) 2(f) 2:5inches (a) (a)

(Indicator Scale)

(b) Shutdown Cooling 2(e) 2(e)(f) 2:5inches (a) (a)

(Indicator Scale)

(2) Manual 2 1 --- (a) (a)

AMENDMENT NO. 247c

TABLE 4.6.2m RPV WATER INVENTORY CONTROL INSTRUMENTATION Surveillance Requirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration START CORE SPRAY PUMPS (1) Manual Note 1 Note 1(9) Note 1(9)

OPEN CORE SPRAY DISCHARGE VALVES (2) Reactor Pressure and Note 1 Note 1(9) Note 1(9)

(1) above AMENDMENT NO. 247d

TABLE 4.6.2m RPV WATER INVENTORY CONTROL INSTRUMENTATION Surveillance Requirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration PRIMARY COOLANT ISOLATION (Cleanup and Shutdown Cooling)

(1) Low-Low Reactor Water Note 1 Note 1!gJ Note 1!gJ Level (2) Manual Note 1 AMENDMENT NO. 247e