NLS2005051, Response to Request for Additional Information License Amendment Request to Revise the Required Channels Per Trip System for Primary & Secondary Containment Isolation and Control Room Emergency Filter System Instrumentation

Response to Request for Additional Information License Amendment Request to Revise the Required Channels Per Trip System for Primary & Secondary Containment Isolation and Control Room Emergency Filter System Instrumentation

ML052170167
Person / Time
Site:	Cooper
Issue date:	08/01/2005
From:	Fehrman W Nebraska Public Power District (NPPD)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NLS2005051
Download: ML052170167 (36)

Text

N Nebraska Public Power District Always there when you need us NLS2005051 August 1, 2005 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001

Subject:

Response to Request for Additional Informnation Re: License Amendment Request to Revise the Required Channels per Trip System for Primary and Secondary Containment Isolation and Control Room Emergency Filter System Instrumentation Cooper Nuclear Station, Docket No. 50-298, DPR-46

Reference:

1. Letter to R. Edington (Nebraska Public Power District) from U.S. Nuclear Regulatory Commission dated June 7, 2005, "Request for Additional Information Re: License Amendment Request to Revise the Required Channels per Trip System for Primary and Secondary Containment Isolation and Control Room Emergency Filter System Instrumentation (TAC No.

MC503 1)."

2. Letter to U. S. Nuclear Regulatory Commission from R. Edington (Nebraska Public Power District) dated October 25, 2004, "License Amendment Request to Revise the Required Channels per Trip System for Primary and Secondary Containment Isolation and Control Room Emergency Filter System Instrumentation" (NLS2004122).

The purpose of this letter is for the Nebraska Public Power District (NPPD) to respond to the Request for Additional Information (RAI) provided in Reference 1 by the Nuclear Regulatory Commission (NRC) regarding the previously submitted License Amendment Request of Reference 2. Please find the RAI responses in Attachment 1. This letter also provides requested supplementary information associated with a teleconference held with the NRC Staff on June 23, 2005.

Should you have any questions concerning this matter, please contact Mr. Paul Fleming, Licensing Manager, at (402) 825-2774.

ACI COOPERNUCLEARSTATION P.O. Box 98/ Brownville, NE 68321-0098 Telephone: (402) 825-3811/ Fax: (402) 825-5211 www.nppd .com

NLS200505 1 Page 2 of 2 I declare under penalty of perjury that the foregoing is true and correct.

Executed on______

(Date)

< Iiam J. Fehrman President and Chief Executive Officer

/'vv Attachment Enclosures cc: Regional Administrator w/ attachment, enclosures USNRC - Region IV Senior Project Manager Nv/ attachment, enclosures USNRC - NRR Project Directorate IV-I Senior Resident Inspector w/ attachment, enclosures USNRC NPG Distribution wv/o attachment, enclosures Records w/attachment, enclosures

NLS200505 1 Attachment I Page 1 of4 Attachment 1 Response to Request for Additional Information Re:

License Amendment Request to Revise the Required Channels per Trip System for Primary and Secondary Containment Isolation and Control Room Emergency Filter System Instrumentation Question 1: In the October 25, 2004 submittal, the nmark-up of the technical specification (TS)

Table 3.3.6.1-1 showved Functions2a, 5d, and 6b, Table 3.3.6.2-1 Function 1, and Table 3.3.7.1-1 Function 1 as Reactor Vessel Water Level - Low (Level 3).

However, in Amendment 209 Table 3.3.6.1-1 Function 5d, Table 3.3.6.2-1 Function 1, and Table 3.3.7.1-1 Function 1 are shown to be Reactor Vessel Water Level - Louw Lowv (Level 2).

Pleaseexplain this discrepancy between Amendinent 209 and the October 25, 2004 submittal.

Response: Reference 2 was submitted to the NRC prior to the issuance of License Amendment 209. That license amendment pertained to a different subject than Reference 2. However, both of these documents affect Technical Specifications Pages 3.3-52, 3.3-57, and 3.3-63; and the license amendment affected certain descriptive information contained in Reference 2. This was recognized by NPPD and the need for providing supplemental information was dialogued with the NRC Project Manager. Enclosure I provides the replacement pages to the Reference 2 submittal which reflect the necessary changes due to License Amendment 209.

Question 2: The October 25, 2004 submittalstated, "As a result of the ITS fimproved technicalspecification] conversion the number of Required Channels Per Trip System was changedfrom 2 to 4. This change was based on ani ITS convention that defined each divisional logic to be one trip system. Since allfour instrument channels provided input signals in that divisional/tripsystem logic, the result was that there be 4 Required ChannelsPer Trip System. However, the CNS [Cooper Nuclear Station] design basis defines each divisional logic as having tvo trip systems, thus 2 Required Channels Per Trip System. The discussion of changes subnitted with the ITS revisions acknowledged the change in practicethat was introduced. -

(a) Pleaseprovide more details concerninghow this change was introduced duringthe ITS conversion.

NLS2005051 Attachment I Page 2 of 4 (b) Please identiby the pages of the ITS submittal where the discussion of changes, that acknowvledged the change in practice,appeared.

(c) Wxat was the constnrction of the TSfor these instruments priorto the conversion to the ITSfornat? Howv many channels, what wvere the action statements, etc.

Response: Enclosure 2 provides the salient ITS submittal pages, including the markup of the Custom Technical Specification pages (including action statements) and the Discussion of Changes pages. Note A.9 to ITS 3.3.6.1 explains that the change from 2 channels per trip system to 4 channels per trip system for the Reactor Low Water Level and High Drywell Pressure functions are departures from CNS administrative practice. Notes A.3 and M.3 to ITS 3.3.6.2 pertain to the Reactor Building Ventilation Radiation function. Note A.3 changed the column header from "Number of Sensor Channels provided by Design" to "Required Number of Sensor Channels per Trip System." Note M.3 acknowledged the CNS convention of having two trip systems for each divisional logic, with two channels per trip system, and the ITS deviation from that convention in establishing each divisional logic as being one trip system with four channels. NPPD has been unable to find any additional internal documentation as to wvhy this convention was used in the ITS submittal.

Question 3: Il the October 25, 2004 submittal, the nmarkup ofpage B 3.3.157, Function 6.b, secondparagraph,second sentence, reads, "ight Four channels (four channels per trip system of the Reactor Vessel [Water Level - Low (Level 3) Function are availableand are required to be OPERABLE to ensure that no single instnrment failure can preclude the isolationfinction.

If there arefour channels total, is "(four channelsper trip systemn)" correct?

Response: The parenthetical "(four channels per trip system)" should read "(two channels per trip system)." A replacement page is provided in Enclosure 1 for this oversight in Reference 2.

Question 4: Some of the instninzents involved in this request also appearin other TS tables with the Required Number of ChannelsPer Trip System being differentfrom the numnber of channels in your request. Reactor Vessel Water Level - Low Low (Level 2) appearsin Table 3.3.5.1-1, "Emergency Core Cooling System Instnruentation,"Function 3a and Table 3.3.5.2-1, "Reactor Core Isolation Cooling System Instntentation Function 1, with 4 as the Required Channels

NLS200505 1 Page 3 of 4 PerTrip Systeni. Drywell Pressure- High appearsin Table 3.3.5.1-1 Function lb, 2b, and 3b, with 4 as the Required Channels Per Trip System.

Please explain whyyour request does not agree with the Required Channels Per Trip System in TS Tables 3.3.5.1-1 and 3.3.5.2-1 from the same variables.

Response: The headers for TS Tables 3.3.5.1-1 and 3.3.5.2-1 are "Required Channels Per Function," not "Required Channels Per Trip System." Thus, for an othervise equivalent one-out-of-two-taken-twice logic, twice the number of channels would be expected in TS Tables 3.3.5.1-1 and 3.3.5.2-1 than in TS Tables 3.3.6.1-1, 3.3.6.2-1, and 3.3.7.1-1. Moreover, the Emergency Core Cooling System (ECCS) actuation instrumentation contains different sensors, channels, and trip system configurations than the Primary and Secondary Containment Isolation and Control Room Emergency Filter System Initiation Instrumentation. Therefore, a correlation should not be drawn between the Reactor Vessel Water Level - Low and Drywvell Pressure - High functions that are common to all these tables.

In a teleconference held between NPPD and the NRC Staff on June 23, 2005, additional questions were asked, for which supplementary information in this response was requested:

1. Note A9 in the ITS conversion for TS 3.3.6.1 states, in part: "Each divisional logic is one-out-of-two taken twice, and provides a signal to close both the inboard and outboard dampers and start both SGT subsystems." This wording seems to state that each logic actuates both isolation devices. Is this true?

Response: The wording was not well-written. A single divisional logic will actuate an inboard or outboard isolation device, not both.

2. Are TS 3.3.6.1 Instrument Functions 5d (Reactor Water Cleanup isolation on Reactor Vessel low level) and 6b (Residual Heat Removal Shutdown Cooling isolation on Reactor Vessel low level) covered by Figures 1, 2, and 3?

Response: Instrument Function 5d is a Primary Containment Isolation System (PCIS) Group 3 isolation. The overall logic arrangement for a Group 3 isolation is the same as for a Group 2 (i.e., the channels are combined into two trip systems that combine into a single divisional logic that controls an inboard or outboard isolation valve).

Figure 4 has been provided in Enclosure 3 to describe the Group 3 logic arrangement.

NLS200505 1 Page 4 of 4 Instrument Function 6b is a Group 2 isolation with a logic that is similar to that shown on Figure 1.

3. A markup of TS Bases Page B 3.3-139 was not provided in the License Amendment Request. The first paragraph states:

[The outputs from these channels are] arranged into two one-out-of two taken twice logic trip systems. One trip system initiates isolations of all inboard primary containment isolation valves, while the other trip system initiates isolation of all outboard primary containment isolation valves. Each logic closes one of the two valves on each penetration, so that operation of either logic isolates the penetration.

Do you plan to revise this page?

Response: The BACKGROUND sections to TS Bases 3.3.6.1 and 3.3.6.2 do not clearly distinguish between use of the terms "trip system," "logic trip system," and "logic," including certain instrument functions that are not being changed by this License Amendment Request. The fundamental PCIS logic arrangement is that there are two divisional trip logics, with the Division I logic controlling an inboard isolation valve, and the Division II logic controlling the outboard isolation valve. As shown on enclosed Figures 1, 2, 3, and 4, each divisional logic is made up from the output contacts of two Trip Systems (Trip System A and B). Within each Trip System are two Trip Subsystems which include the individual channels for a given instrument function. Thus, there are typically two channels per Trip System. The Trip System output contacts are arranged as one-out-of-two-taken-twice within each divisional logic.

Since the clarification of the Bases BACKGROUND wording to the above discussion extended beyond the specific instrument functions that were the subject of this License Amendment Request, it was felt that providing comprehensive Bases markups would not have been beneficial to the NRC's review. However, revision of the Bases pages to provide terminology that is consistent with the above discussion is designated as an implementing action of this License Amendment Request. These Bases changes will be made under the provisions of IOCFR50.59 and TS 5.5.10 (Technical Specifications Bases Control Program). A markup of Bases Page B 3.3-139 is provided in Enclosure 4 to illustrate the types of changes that are anticipated.

NLS2005051 Enclosure I Enclosure I Listing of Replacement Pages to NLS2004122 The following pages are replacements to the Attachments provided in NLS2004122: : Page 2 of 8 : Technical Specifications Page 3.3-52 Technical Specifications Page 3.3-57 Technical Specifications Page 3.3-63 : Technical Specifications Page 3.3-52 Technical Specifications Page 3.3-57 Technical Specifications Page 3.3-63 : Technical Specifications Bases Page B 3.3-156 Technical Specifications Bases Page B 3.3-157 Technical Specifications Bases Page B 3.3-170 Technical Specifications Bases Page B 3.3-187 : Figure 1 Figure 2 Figure 3

NLS2004122 Page 2 of 8 LICENSE AMENDMENT REQUEST TO REVISE THE REQUIRED CHANNELS PER TRIP SYSTEM FOR PRIMARY AND SECONDARY CONTAINMENT ISOLATION AND CONTROL ROOM EMERGENCY FILTER SYSTEM INSTRUMENTATION 1.0 Description This letter requests an amendment to the Cooper Nuclear Station (CNS) Technical Specifications (TS) to revise the Required Channels Per Trip System for several instrumentation functions described on TS Table 3.3.6.1-1 (Primary Containment Isolation Instrumentation), Table 3.3.6.2-1 (Secondary Containment Isolation Instrumentation), and Table 3.3.7.1-1 (Control Room Emergency Filter System Instrumentation). The changes address inconsistencies with the CNS design bases regarding the definitions of "Instrument Channel" and "Trip System." These inconsistencies were introduced with the conversion to Improved Standard Technical Specifications (ITS) in License Amendment 178 (Tables 3.3.6.1-1 and 3.3.6.2-1) (Reference 7.1),

as carried forxvard in License Amendment 187 (Table 3.3.7.1-1) (Reference 7.2). The revisions are administrative in nature, as they have no impact on facility configuration, operation, or testing.

2.0 Proposed Change Attachments 2 and 3 describe the proposed changes. These changes revise the number of Required Channels Per Trip System from 4 to 2 for the designated instrumentation functions. provides the applicable TS Bases changes, which revise the total number of channels from eight to four.

3.0 Background High Drywvell Pressure or Low Reactor Water Level will provide a Group 2 Primary Containment Isolation System (PCIS) signal. A Group 6 PCIS signal is actuated by a High Drywell Pressure, Low-Low Reactor Water Level, or High Reactor Building Ventilation Exhaust Plenum Radiation signal. These instrumentation functions are reflected in TS Table 3.3.6.1-1, Functions 2.a, b, c, 5.d, 6.b, and Table 3.3.6.2-1, Functions 1, 2, 3. As a result of the ITS conversion, the number of Required Channels Per Trip System was changed from 2 to 4. This change was based on an ITS convention that defined each divisional logic to be one trip system. Since all four instrument channels provided input signals in that divisional/trip system logic, the result was that there be 4 Required Channels Per Trip System. However, the CNS design basis defines each divisional logic as having two trip systems, thus 2 Required Channels Per Trip System. The discussion of changes submitted with the ITS revisions acknowledged the change in practice that was being introduced. However, since then, this has

Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 2 of 3)

Primary Containment Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SURVEILLANCE SPECIFIED PER TRIP REQUIRED REQUIREMENT ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION C.1 S VALUE

3. High Pressure Coolant Injection (HPCI) System Isolation

a. HPCI Steam Line 1.2.3 1 F SR 3.3.6.1.2 <250% rated Flow- High SR 3.3.6.1.4 steam flow SR 3.3.6.1.6

b. HPCI Steam Line 1,2,3 1 F SR 3.3.6.1.2 <6 seconds Flow-Time Delay Relays SR 3.3.6.1.4 SR 3.3.6.1.6

c. HPCI Steam Supply 1,2,3 2 F SR 3.3.6.1.2 > 107 psig Line Pressure - Low SR 3.3.6.1.4 SR 3.3.6.1.6

d. HPCI Steam Line Space 1,2,3 2 per F SR 3.3.6.1.2 <195F Temperature - High location SR 3.3.6.1.4 SR 3.3.6.1.6

4. Reactor Core Isolation Cooling (RCIC) System Isolation

a. RCIC Steam Line 1,2.3 1 F SR 3.3.6.1.2 <288% rated Flow - High SR 3.3.6.1.4 steam flow SR 3.3.6.1.6

b. RCIC Steam Line 1,2,3 1 F SR 3.3.6.1.2 <6 seconds Flow-Time Delay Relays SR 3.3.6.1.4 SR 3.3.6.1.6

c. RCIC Steam Supply 1,2,3 2 F SR 3.3.6.1.2 > 61 psig Line Pressure - Low SR 3.3.6.1.4 SR 3.3.6.1.6

d. RCIC Steam Line Space 1,2,3 2 per F SR 3.3.6.1.2 <195-F Temperature - High location SR 3.3.6.1.4 SR 3.3.6.1.6

5. Reactor Water Cleanup (RWCU) System Isolation

a. RWCU Flow-High 1,2,3 1 F SR 3.3.6.1.2 <191% of SR 3.3.6.1.4 Rated SR 3.3.6.1.6

b. RWCU System Space 1,2,3 2 per F SR 3.3.6.1.2 <195F Temperature - High location SR 3.3.6.1.4 SR 3.3.6.1.6

c. SLC System Initiation 1,2 1 H SR 3.3.6.1.6 NA

d. Reactor Vessel Water 1,2,3 42 F SR 3.3.6.1.1 >-42 inches Level - Low Low SR 3.3.6.1.2 (Level 2) SR 3.3.6.1.4 SR 3.3.6.1.6 Amendment 209 3.3-52 2/4one5

Secondary Containment Isolation Instrumentation 3.3.6.2 Table 3.3.6.2-1 (page 1 of 1)

Secondary Containment Isolation Instrumentation APPLICABLE MODES OR REQUIRED OTHER CHANNELS SPECIFIED PER SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS TRIP SYSTEM REQUIREMENTS VALUE

1. Reactor Vessel Water Level - Low Low (Level 2) 1,2,3, (a) 42 SR SR 3.3.6.2.1 3.3.6.2.2

> -42 Inches I SR 3.3.6.2.3 SR 3.3.6.2.4

2. Drywell Pressure - High 1,2,3 42 SR 3.3.6.2.2 <1.84 psig SR 3.3.6.2.3 I SR 3.3.6.2.4

3. Reactor Building Ventilation 1,2.3, -42 SR 3.3.6.2.1 <49 mR/hr Exhaust Plenum (a),(b) SR 3.3.6.2.2 I Radiation - High SR 3.3.6.2.3 SR 3.3.6.2.4 (a) During operations with a potential for draining the reactor vessel.

(b) During CORE ALTERATIONS and during movement of Irradiated fuel assemblies In secondary containment.

Amendment 2G9 3.3-57 2/1 0/05

CREF System Instrumentation 3.3.7.1 Table 3.3.7.1-1 (page 1 of 1)

Control Room Emergency Filter System Instrumentation APPLICABLE MODES OR REQUIRED OTHER CHANNELS SPECIFIED PER TRIP SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM REQUIREMENTS VALUE

1. Reactor Vessel Water 1.2.3, -4 2 SR 3.3.7.1.1 > -42 inches I Level - Low Low (Level 2) (a) SR 3.3.7.1.2 SR 3.3.7.1.3 SR 3.3.7.1.4

2. Drywell Pressure - High 1,2.3 -42 SR 3.3.7.1.2 SR 3.3.7.1.3

<1.84 psig I SR 3.3.7.1.4

3. Reactor Building Ventilation 1.2.3. -42 SR 3.3.7.1.1 <49 mR/hr I Exhaust Plenum (a).(b) SR 3.3.7.1.2 Radiation - High SR 3.3.7.1.3 SR 3.3.7.1.4 (a) During operations with a potential for draining the reactor vessel.

(b) During CORE ALTERATIONS and during movement of Irradiated fuel assemblies in the secondary containment.

Amendment 2G9 3.3-63 2/4iWa

Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 2 of 3)

Primary Containment Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SURVEILLANCE SPECIFIED PER TRIP REQUIRED REQUIREMENT ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION C.1 S VALUE

3. High Pressure Coolant Injection (HPCI) System Isolation

a. HPCI Steam Line 1,2,3 1 F SR 3.3.6.1.2 <250% rated Flow - High SR 3.3.6.1.4 steam flow SR 3.3.6.1.6

b. HPCI Steam Line 1,2,3 1 F SR 3.3.6.1.2 < 6 seconds Flow-Time Delay Relays SR 3.3.6.1.4 SR 3.3.6.1.6

c. HPCI Steam Supply 1,2,3 2 F SR 3.3.6.1.2 > 107 psig Line Pressure - Low SR 3.3.6.1.4 SR 3.3.6.1.6

d. HPCI Steam Line Space 1,2,3 2 per F SR 3.3.6.1.2 < 195'F Temperature - High location SR 3.3.6.1.4 SR 3.3.6.1.6

4. Reactor Core Isolation Cooling (RCIC) System Isolation

a. RCIC Steam Line 1,2,3 1 F SR 3.3.6.1.2 < 288% rated Flow - High SR 3.3.6.1.4 steam flow SR 3.3.6.1.6

b. RCIC Steam Line 1,2,3 1 F SR 3.3.6.1.2 <6seconds Flow-Time Delay Relays SR 3.3.6.1.4 SR 3.3.6.1.6

c. RCIC Steam Supply 1,2.3 2 F SR 3.3.6.1.2 > 61 psig Line Pressure - Low SR 3.3.6.1.4 SR 3.3.6.1.6

d. RCIC Steam Line Space 1,2,3 2 per F SR 3.3.6.1.2 < 195-F Temperature - High location SR 3.3.6.1.4 SR 3.3.6.1.6

5. Reactor Water Cleanup (RWCU) System Isolation

a. RWCU Flow - High 1,2,3 1 F SR 3.3.6.1.2 < 191% of SR 3.3.6.1.4 Rated SR 3.3.6.1.6

b. RWCU System Space 1,2,3 2 per F SR 3.3.6.1.2 c 195-F Temperature - High location SR 3.3.6.1.4 SR 3.3.6.1.6

c. SLC System Initiation 1.2 1 H SR 3.3.6.1.6 NA

d. Reactor Vessel Water 1,2,3 2 F SR 3.3.6.1.1 >-42Inches Level - Low Low SR 3.3.6.1.2 (Level 2) SR 3.3.6.1.4 SR 3.3.6.1.6 Amendment 3.3-52

Secondary Containment Isolation Instrumentation 3.3.6.2 Table 3.3.6.2-1 (page 1 of 1)

Secondary Containment Isolation Instrumentation APPLICABLE MODES OR REQUIRED OTHER CHANNELS SPECIFIED PER SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS TRIP SYSTEM REQUIREMENTS VALUE

1. Reactor Vessel Water 1,2,3, 2 SR 3.3.6.2.1 >-42 inches Level - Low Low (Level 2) (a) SR 3.3.6.2.2 I SR 3.3.6.2.3 SR 3.3.6.2.4

2. Drywell Pressure - High 1,2,3 2 SR 3.3.6.2.2 < 1.84 psig SR 3.3.6.2.3 I SR 3.3.6.2.4

3. Reactor Building Ventilation 1,2,3, 2 SR 3.3.6.2.1 < 49 mRlhr Exhaust Plenum (a),(b) SR 3.3.6.2.2 I Radiation - High SR 3.3.6.2.3 SR 3.3.6.2.4 (a) During operations with a potential for draining the reactor vessel.

(b) During CORE ALTERATIONS and during movement of irradiated fuel assemblies in secondary containment.

Amendment 3.3-57

CREF System Instrumentation 3.3.7.1 Table 3.3.7.1-1 (page 1 of 1)

Control Room Emergency Filter System Instrumentation APPLICABLE MODES OR REQUIRED OTHER CHANNELS SPECIFIED PER TRIP SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM REQUIREMENTS VALUE

1. Reactor Vessel Water Level - Low Low (Level 2) 1,2,3, (a) 2 SR SR 3.3.7.1.1 3.3.7.1.2

> -42 inches I SR 3.3.7.1.3 SR 3.3.7.1.4

2. Drywell Pressure - High 1,2.3 2 SR 3.3.7.1.2 < 1.84 psig SR 3.3.7.1.3 I SR 3.3.7.1.4

3. Reactor Building Ventilation 1,2,3, 2 SR 3.3.7.1.1 <49 mRlhr Exhaust Plenum (a),(b) SR 3.3.7.1.2 I Radiation - High SR 3.3.7.1.3 SR 3.3.7.1.4 (a) During operations with a potential for draining the reactor vessel.

(b) During CORE ALTERATIONS and during movement of irradiated fuel assemblies in the secondary containment.

Amendment 3.3-63

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE SAFETY ANALYSES, LCO, AND APPLICABILTY 5.d Reactor Vessel Water Level - Low Low (Level 2) (continued) peak cladding temperature remains below the limits of 10 CFR 50.46.

The Reactor Vessel Water Level - Low Low (Level 2) Function associated with RWCU isolation is not directly assumed in the USAR safety analyses because the RWCU System line break is bounded by breaks of larger systems (recirculation and MSL breaks are more limiting).

Reactor Vessel Water Level - Low Low (Level 2) signals are initiated from four level switches that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Eight Four channels of Reactor Vessel Water Level - Low Low (Level 2) Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.

The Reactor Vessel Water Level - Low Low (Level 2) Allowable Value was chosen to be the same as the High Pressure Coolant Injection/Reactor Core Isolation Cooling (HPCI/RCIC) Reactor Vessel Water Level - Low Low (Level 2) Allowable Value (LCO 3.3.5.1and LCO 3.3.5.2), since this could indicate that the capability to cool the fuel may be threatened.

This Function isolates the Group 3 valves, as listed in Reference 1.

Shutdown Cooling System Isolation 6.a. Reactor Pressure - High The Reactor Pressure - High Function is provided to isolate the shutdown cooling portion of the Residual Heat Removal (RHR) System.

This Function is provided only for equipment protection to prevent an intersystem LOCA scenario, and credit for the interlock is not assumed in the accident or transient analysis in the USAR.

The Reactor Pressure - High signals are initiated from two pressure switches that are connected to different taps on a recirculation pump suction line. Two channels of Reactor Pressure - High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Function is only required to be OPERABLE in MODES 1, 2, and 3, since these Cooper B 3.3-156 2/1 0/05

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY 6.a. Reactor Pressure - High (continued) are the only MODES in which the reactor can be pressurized; thus, equipment protection is needed. The Allowable Value was chosen to be low enough to protect the system equipment from overpressurization.

This Function isolates both RHR shutdown cooling pump suction valves.

6.b. Reactor Vessel Water Level - Low (Level 3)

Low RPV water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, isolation of some reactor vessel interfaces occurs to begin isolating the potential sources of a break. The Reactor Vessel Water Level - Low (Level 3) Function associated with RHR Shutdown Cooling System isolation is not directly assumed in safety analyses because a break of the RHR Shutdown Cooling System is bounded by breaks of the recirculation and MSL. The RHR Shutdown Cooling System isolation on Level 3 supports actions to ensure that the RPV water level does not drop below fuel zone zero during a vessel draindown event caused by a leak (e.g., pipe break or inadvertent valve opening) in the RHR Shutdown Cooling System.

Reactor Vessel Water Level - Low (Level 3) signals are initiated from four level switches that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Ekjht Four channels (fotf two channels per trip system) of the Reactor Vessel Water Level - Low (Level 3) Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. As noted (footnote (b) to Table 3.3.6.1-1), only one trip system of the Reactor Vessel Water Level - Low (Level 3) Function is required to be OPERABLE in MODES 4 and 5, provided the RHR Shutdown Cooling System integrity is maintained. System integrity is maintained provided the piping is intact and no maintenance is being performed that has the potential for draining the reactor vessel through the system.

The Reactor Vessel Water Level - Low (Level 3) Allowable Value was chosen to be the same as the RPS Reactor Vessel Cooper B 3.3-1 57 Cunle-+, Ic999

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY

1. Reactor Vessel Water Level-Low Low (Level 2) (continued) the actual water level (variable leg) in the vessel. Eght Four channels of Reactor Vessel Water Level - Low Low (Level 2) Function are available and are required to be OPERABLE in MODES 1, 2, and 3 to ensure that no single instrument failure can preclude the isolation function.

The Reactor Vessel Water Level - Low Low (Level 2) Allowable Value was chosen to be the same as the High Pressure Coolant Injection/

Reactor Core Isolation Cooling (HPCI/RCIC) Reactor Vessel Water Level Low Low (Level 2) Allowable Value (LCO 3.3.5.1 and LCO 3.3.5.2) since this could indicate that the capability to cool the fuel is being threatened).

The Reactor Vessel Water Level - Low Low (Level 2) Function is required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists in the Reactor Coolant System (RCS); thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. In MODES 4 and 5, the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES; thus, this Function is not required. In addition, the Function is also required to be OPERABLE during operations with a potential for draining the reactor vessel (OPDRVs) because the capability of isolating potential sources of leakage must be provided to ensure that offsite dose limits are not exceeded if core damage occurs.

2. DrDwell Pressure-High High drywell pressure can indicate a break in the reactor coolant pressure boundary (RCPB). An isolation of the secondary containment and actuation of the SGT System are initiated in order to minimize the potential of an offsite dose release. The isolation on high drywell pressure supports actions to ensure that any offsite releases are within the limits calculated in the safety analysis. The Drywell Pressure - High Function associated with isolation is not assumed in any USAR accident or transient analyses, but will provide an isolation and initiation signal. It is retained for the overall redundancy and diversity of the secondary containment isolation instrumentation as required by the NRC approved licensing basis.

Cooper B 3.3-170 2/^nfn5

CREF System Instrumentation B 3.3.7.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY

1. Reactor Vessel Water Level - Low Low (Level 2)

Low reactor pressure vessel (RPV) water level indicates that the capability of cooling the fuel may be threatened. A low reactor vessel water level could indicate a LOCA and will automatically initiate the CREF System, since this could be a precursor to a potential radiation release and subsequent radiation exposure to control room personnel.

Reactor Vessel Water Level - Low Low (Level 2) signals are initiated from level switches that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Ekjht Four channels of Reactor Vessel Water Level - Low Low (Level 2) Function are available and are required to be OPERABLE in MODES 1, 2, and 3 to ensure that no single instrument failure can preclude CREF System initiation.

The Reactor Vessel Water Level - Low Low (Level 2) Allowable Value was chosen to be the same as the Secondary Containment Isolation Allowable Value (LCO 3.3.6.2) to enable initiation of the CREF System at the earliest indication of a breach in the nuclear system process barrier, yet far enough below normal operational levels to avoid spurious initiation.

The Reactor Vessel Water Level - Low Low (Level 2) Function is required to be OPERABLE in MODES 1, 2, and 3, and during operations with a potential for draining the reactor vessel (OPDRVs) to ensure that the Control Room personnel are protected during a LOCA. In MODES 4 and 5 at times other than OPDRVs, the probability of a vessel draindown event resulting in the release of radioactive material to the environment is minimal. Therefore, this Function is not required in other MODES and specified conditions.

2. Drywell Pressure - High High drywell pressure can indicate a break in the reactor coolant pressure boundary. A high drywell pressure signal could indicate a LOCA and will automatically initiate the CREF System, since this could be a precursor to a potential radiation release and subsequent radiation exposure to control room personnel.

Cooper B 3.3-187 2110105

TRIP SYSTEMA TT. ClIR

-1 -

• CHANNEL

~ ~ ~ 5A-K6A LOGIC IS16A-K5A CHANNEL TRIP SI_

StA S A-K4A l tTRI' -SIGNAL TRIP SUBSYSTEM A2 CHANNEL

[5~c~ A -K6C LOGIC 16A-K5C CHANNEL R TRIP SIGNLC l StC§t 5A-K4C -iTIP -- SIGNAL

___ I I I ACTUATION DEVICE TRIP SYSTEMP TRIP SUBSYSTEUB1 PROTECTNVE CHANNEL ACTON RW-AOV-AO95 OUTBOARO DRYWELL

[5A,-MK SB LOGIC EQUIPMENTDRAIN ACTUATED SUMP ISOLATION VALVE DEVICE (IS TABLE3.3.6.1-I I I FUNCTIONS2a & 2b)

I CHAN[NEL >TRIP SIGNL*4 B16A-K5 I - - - I I I LEGEND NSO2 5M-K4B 1 t RP SIGNAL I ELECTRICALRELAYSOLENOID I MECHANICALCONNECTION I -,-. ELECTRICALCONNECTION I

CHANNEL TRIP SUBSYSTEM82 5A-KBD 1 LOGIC I

I I

I ir

REFERENCES:

ELECTRICALCONTACTS(SHOWNCLOSED)

VALVE It6A-K5D CHANNEL >TRIP SIGNAL 1 SIN " - - - - - I DWG 791E256 SH 9 & 10 DWG 79tE266 SH 5. 6. & 7 5A K4 J <-RISIG FIGURE 1 - PCIS GROUP 2 REACTOR WATER LEVEL/DRYWELL PRESSURE INSTRUMENTATION b

CM-l

TRIP SYSTEM TRIP SUBSYSTEM Al CHANNEL LOGIC SENO 16A-K81A CHANNEL nTRIP SIGNAL 1 - 1

-1 A A-K4A N

_ __ - FROu RMAX SHEET 2 TRIP SUBSYSTEM A2

_ - - FROM RMBX SHEET2 CHANNEL TRIP SIGNAL

> 1> ACTUATIONDEVICE 2AX I

CHANNEL >TRIP SIc w

NAL*.

< 16A-K81C I I 1"

PS1C A-K4C -tTRIP SIGNAL I

II 16A- K79 TRIP SYSTEMB I TIRIP SIGNAL CHANNEL

--- j TRIP SUBSYSTEMB1 LOGIC 1

I.

PROTECTIVE N HV-MUOV-M0258 REACTORBUILDING EXHAUSTISOLATIONVALVE ACTUATED (TS TABLE3.3.6.2-1 FUNCTIONS1.2.3)

I SENSOR I I DEVICE (HOP)

CHANNEL I

g rT 1.

TRIP SiGNA

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- WR

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I I IEGEN

-18 5A-K4B t RP SCI ELECTRICALRELAYSOLENOID

- - - - MECHANICAL CONNECTION ELECTRICAL CONNECTION TRIP SUBSYSTEM 82 ELECTRICALCONTACTS(SHOWNCLOSED)

CHANNEL VALVE IFu s S I SENSORl LOGIC

REFERENCES:

DIW 791E256 SH 9 & 10 CHANNEL TRIP SIGNAL ;

DWO 791E268 SH 5. 8 & 8 EN SO__

-120 5A-K4D TRIP SIGNAL DWG 3065 FIGURE 2 - PCIS GROUP 6 REACTOR WATER LEVEL/DRYWELL PRESSURE INSTRUMENTATION e

P

TRIP SYSTEM A

- - -I1 LOGIC

--- I -1 LOGIC I I

I ad/\ r (O LIC Ir lI _

RMAX

- g? LOGIC f t VJ tTRIP SIGNAL U SHEED 1)

TRIP SYSTEME tRIP SIM TRISHEET 1) i _I I

__ _ _ I i

REFERENCES:

DWG 791E257 SH 6 DWG 3085. SH 17A FIGURE 3 - PCIS GROUP 6 REACTOR BUILDING VENTILATION EXHAUST PLENUM RADIATION INSTRUMENTATION t1w a cob3ml

NLS200505 1 Enclosure 2 Excerpts From Improved Technical Specifications Submittal'

1. Letter from G. Horn (NPPD) to U.S. Nuclear Regulatory Commission, dated March 27, 1997, "Proposed Change to CNS Technical Specifications Conversion to Improved Standard Technical Specifications" (NLS970002).

wT-COOPE NUCLEAR STATION I Adl 11DES9/eJ S TABLE (Page 1)

3. preopscd hpp itc.e.ii PRIMARY CONTAINMENT AND REACTOR VESSEL ISOLATION INSTRUMENTATION l 60q4VfOd R fM( f u14 AJOn)6. 6c P / ra S/YO6 - f ,7 e,t

>Main Steam Line Hi 4~-r Radiation f Wa te Reactor Low Water Level  ?#4.5 in. Indicated Level Reactor Low Low Low Water 2-145.5 in. Indicated Level 2 Level L~.A~J(PO ose4 \

(1) > Main Steam Detection s 200°F (IjhD I/"I'% s 150% of Rated Steam 2L3 2(3)

.Acro C10 FtI 01 &

4Main Steam Line High Flow Flow G > Main Steam Line Low 2 825 psig Pressure 2 1 High Drywell Pressure s 2 p3ig

@ High Reactor Pressure

! cTcJ &

_ rD...

Main Condenser Low kj~)a Vacuum

~ AC7-.r N Reactor Water Cleanup 1

-I C tSystem High Flow I" A C To g 't?I 0;

.1.

Lr--

Uj 9-."

COOPER CLEAR STATION TABLE 3. .A (Page 2)

PRIMARY CONTAINMENT AND REACTOR VESSEL ISOLATION INSTRUMENTATION

-ns owbeVaq) inimum Num b Action Required When Intumn Insruen S >pf Operablel oponents Component Operability Intrmet ID.No /+ _Per Trip System (1) i9 Not Assured (2)

RWCU System Hi Space RWCU-TS-15 0 151 1 00 2 6 Temp. 152, 153/ 14, 155, 156 . ,T 157, 1J, 159 RW U

T TS-81

-6S1 IL-A,B,E,F, C,D,G, AJ /9) 0 m co n .

O1

1 oecJFov 3 asa6 LE Cfis 3-2-A vd 77 h .S,'

rq u1 /H sh atl 4 4 Whenever Primary Containment in egri y is required a be two o erable o

-ere CTJ tripped trip s stems for each tunction.

f e minimum number of operable ent stem requirement cannot be met by a trip system, (that trip ssem hall be ripped' If the

/ W' le requirements cannot be met b bat tri sY tems, e appropriate ac isted below shall be taken.

AC I ide shut and(have the reactor in a cold shutdown condition;>

aeductionand have the Main Steam Isolation Valves hut within Whours. < rb; o~

Isolate the Reactor Water Cleanup Sy 12 A014 t5 0/'d ; L Isolate the Shutdown Cooling mode of the RHR Sys0e OF4 (

Isolate the Reactor Water Sample Valves Two required for each steam line. -Ae i 4. These signals also start the Standby Gas Treatment System and initiate Secondary)

Containment isolation.

Not required in the refuel, shutdown, and startup/hot standby modes Iinterl < ed with [

I X tlRequi one nnwl eeaal tch trip system.

Low vacuum isolation is bypassed when the turbine stop is not full open,X a A) j.d,- / itches are in~fypass and mode switch is not in RUN. i

8. The instruen a le produce primary containmen system isolati ns. The following listing groups the system signals and the system isolate I .1- Isolation Signals:

1. Reactor Low Low Low Water Level in.)

2. Main Steam Line Low Pressure 25 psig in the RUN mode)

3. Main Steam Line Leak Det on (s2OO0F)

4. Condenser Low Vacuum " Hg vacuum)

5. Main Steam Line Flow (l50% of rated flow)

Isolations:

2. an Steam Line Drains Amendment No. 158 03/02/93 Poadr FT 2E

DISCUSSION OF CHANGES ITS: 3.3.6.1 - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION ADMINISTRATIVE A.5 Details contained in Note 1 concerning the use of CTS Table 3.2.B are proposed to be deleted. NUREG-1433 provides a new format for proposed Table 3.3.6.1-1 that eliminates the need for instructions on its use.

The actual requirements (ACTIONS) have not changed except where justified in another Discussion of Change in this section. Therefore, this change is considered to be administrative.

A.6 The conversion to ITS precludes the need for extensive referencing within the instrumentation tables. Appropriate instruments are listed in each section to which they apply. Therefore, CTS Tables 3.2.D, Note 1.E, Tables 4.2.A through 4.2.F, Notes 5, 11, and 12, and the Note in Table 4.2.D to "see Table 4.2.A" are not necessary and are proposed to be deleted. This change is considered to be administrative.

A.7 The column title in CTS Table 3.2.D is proposed to be changed to the required number of channels per trip system rather than the current number of channels provided by design. Thus, except as otherwise noted, the number of channels in the proposed columns is halved for Functions having two trip systems. This change is considered administrative.

A.8 The Frequency "once/Operating Cycle" in CTS Table 4.2.A for the Main Steam Line Leak Detection and RWCU High Space Temperature Functions, in CTS Table 4.2.B for the HPCI Steam Line Space High Temperature, HPCI Steam Line High delta P Actuation Timers, RCIC Steam Line Space Excess Temperature, and RCIC Steam Line High delta P Actuation Timer Functions, and CTS Table 4.2.D for the Mechanical Vacuum Pump Isolation Function (high steam line radiation) is proposed to be changed to "18 months".

This change is administrative since 18 months is a normal operating cycle.

A.9 CTS Table 3.2.A requires two channels per trip system for the Reactor Low Water Level and High Drywell Pressure Functions. Each of these sensors provides input into two divisional logics. Each divisional logic is one-out-of-two taken twice, and provides a signal to close both the inboard and outboard dampers and start both SGT subsystems. Current administrative practice specifies that each divisional logic consists of two trip systems, with two channels (arranged in a one-out-of two logic) in each trip system. The ITS will specify that each divisional logic consists of one trip system; a one-out-of-two taken twice trip system logic. Therefore, the number of channels in the ITS for these Functions (proposed Functions 2.a, 5.d, and 6.b for Reactor Low Water Level and Function 2.b for High Drywell Pressure) have been specified as four.

Since this change does not change the actual requirements, it is considered administrative.

CNS 2 Revision 0

Lzj

2 r?

I 1I/IOvJ z System L1Z ,5-H J LzInstpg1ent 6 No. I \. Vabe

/ Steam Jet Air Ejector Off-G Gas RMP-RM-150 A & B (3)

N . System _._._ . . _ - _- _ _ . _ _ ...................... _

> Reactor Building Isolation ZZ-13M~ ~A , Bl, (3f r/hrmr. v and Standby Gas Treatment EO-0 or Initiation

,eerT. Liquid Radwaste Discharge RMP-RM-1 (2) -

1 C)

Isolation -

Control Room Emergency Filter RMV-RM-1 4xlO CPM 1 D 4_ ,._ ...

. .Vcuu . .... R- .A,.B .Pum-3t

/ Mchnial acumPup RP-M-51A, B. 3 T im, es normal full power 4 E Isolation (4) C & D backg]round. Alarm at 1.5 t imes normal full w power -bacgon

, / __ __

33@,

0 W

t'j 4- W Z),

(,A

LL5pe7;kod3/4Q; 3.3.6.2

'-7 O' 0~ --

1. Action required when component operability is not assured. f A. (1) If radiation level exceeds 1.0 ci/sec (prior to 30 min. delay line) for a period greater than 15 consecutive minutes, the off-gas isolation valve shall close and reactor shutdown shall be initiated immediately andth /

reactor placed in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Refer Spto Specification 3.21.

u nut crabe shallnelo per trip sms m shall be operabl mov n Za3aus -nsiae secondarv contain-^ hclste oetao

' \ T Ed'damagvi . ~~~rnAegdftlonE-hsrqirement ctth p u ,i ACTS toP~~~~i~~iS~~t}5iEF ~~- 1 If psse this salb rpe requirmn a i oing actions jtj_ irradiated fuel inside secondary containme eaehnlof n dremov7e /uJ endmen No. 1580

.During release of radioactive wastes, the effluent control monitor shal he SS- - 1 be set to alarm and automatically close the waste discharge valve priorz.vS 3 to exceeding the limits of Specification 3.21.B-1. I al

[OD. Refer to Section entitled 'Additional SafetyRelated Plant Capabilities, E. Refer to Section 3.2.D.5 and the requirements for Primary Containment

/Isolation on high main steam line radiation, Table 3.2.A.

A.Trip settings to correspond to Specification 3.21.B.1.

3.Trip settings to correspond to Specification 3.21.C.6.a.

4. Minimum number of channels operable shall be one during mechanical vacuum pum operation.

Amendment No. 158 -63a- 03/02/93 Pqge Xof l

DISCUSSION OF CHANGES ITS: 3.3.6.2 - SECONDARY CONTAINMENT ISOLATION INSTRUMENTATION ADMINISTRATIVE A.1 In the conversion of the Cooper Nuclear Station current Technical Specifications (CTS) to the proposed plant specific Improved Technical Specifications (ITS), certain wording preferences or conventions are adopted that do not result in technical changes (either actual or interpretational). Editorial changes, reformatting, and revised numbering are adopted to make the ITS consistent with the BWR Standard Technical Specifications, NUREG-1433, Rev. I (i.e., Improved Standard Technical Specifications (ISTS)).

A.2 A Note at the start of the Actions Table ("Separate Condition entry is allowed for each channel.") is proposed to be added to provide more explicit instructions for proper application for the new Actions for Technical Specification compliance. In conjunction with the proposed Specification 1.3 "Completion Times," this Note provides direction consistent with the intent of the Required Actions for inoperable Secondary Containment Isolation instrumentation channels, functions, or trip systems. It is intended that each Required Action be applied regardless of it having been applied previously for other inoperable Secondary Containment Isolation instrumentation channels, functions, or trip systems.

A.3 The column title in CTS Table 3.2.D is proposed to be changed to a per Trip System basis rather than the current "provided by design" basis.

The Reactor Building Ventilation Exhaust Plenum Radiation-High Function (proposed Function 3) has two trip systems per design; therefore, this is considered to be an administrative change.

A.4 CTS Table 3.2.D, Note 1.B(1) requires suspension of irradiated fuel handling inside the secondary containment when Reactor Building Ventilation Exhaust Plenum Radiation - High channels are inoperable and not tripped in both trip systems. This action is proposed to be replaced with proposed Required Actions C.1.2 and C.2.2, which require the associated secondary containment isolation valves (SCIVs) and Standby Gas Treatment (SGT) subsystem to be declared inoperable. With channels in both trip systems inoperable, all automatic SCIVs and both SGT subsystems are affected (i.e., they are associated with the inoperable channels). The ACTIONS of proposed LCO 3.6.4.2, "Secondary Containment Isolation Valves (SCIVs)" will require suspension of irradiated fuel handling in the secondary containment 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after declaring the SCIVs inoperable. The ACTIONS of proposed LCO 3.6.4.3, "Standby Gas Treatment (SGT) System" will require immediate suspension of irradiated fuel handling in the secondary containment after declaring the SGT System inoperable. Therefore, since the actions are the same (irradiated fuel handling in the secondary containment continues to be required to be suspended immediately), this change is considered administrative.

CNS 1 Revision 0

DISCUSSION OF CHANGES ITS: 3.3.6.2 - SECONDARY CONTAINMENT ISOLATION INSTRUMENTATION TECHNICAL CHANGES - MORE RESTRICTIVE M.1 Currently, no LSFT is required for the Drywell Pressure High Function in CTS Table 4.1.1. While this CTS Table is of RPS Instrumentation, the Drywell Pressure High Sensor is common to both the RPS and Secondary Containment Isolation Function. Therefore, the CTS only provides the Surveillances in the RPS Specification. A specific SR for performance of a Logic System Functional Test is proposed to be added to the High Drywell Pressure Function. This additional SR will ensure Operability of the secondary containment isolation instrumentation ensuring it can fulfill its safety function. The addition of a new SR constitutes a more restrictive change; however, it is consistent with current administrative practices.

M.2 A new Applicability is proposed to be added (proposed Note a to Table 3.3.6.2-1) for the Reactor Vessel Water Level-Low (Level 3) Function (proposed Function 1). This Function will be required to be Operable during operations with a potential for draining the reactor vessel (OPDRVs). OPDRVs could result in a vessel draindown event and subsequent release of radioactivity, such that these instruments would be needed to isolate the secondary containment and start the SGT System.

This is an additional restriction on plant operation. In addition, the Applicability for the CTS Table 3.2.0 Reactor Building Ventilation Exhaust Plenum Function (proposed Function 3) has been modified. CTS Table 3.2.D, Note 1.B specifies an Applicability of "when handling irradiated fuel inside secondary containment." The new Applicability for this Function includes not only irradiated fuel handling in the secondary containment (proposed Note (b) to Table 3.3.6.2-1), but also MODES 1, 2, and 3 (proposed Applicability Column in Table 3.3.6.2-1),

Core Alterations (proposed Note (b) to Table 3.3.6.2-1), and operations with a potential for draining the reactor vessel (proposed Note (a) to Table 3.3.6.2-1). These new Applicabilities will ensure the Function is Operable to mitigate accidents in the MODES and other specif-ed conditions assumed in the accident analysis. This is also a, additional restriction on plant operation.

M.3 The number of channels per trip system in CTS Table 3.2.D, Note 1.8 for proposed Function 3, Reactor Building Ventilation Exhaust Psenum Radiation-High, is proposed to be increased from 1 to 4. This is being done for consistency with the design of the instruments, at described in CTS Table 3.2.D (the Number of Sensor Channels provided by Design columns), and with the USAR description of these instruments in USAR Section VII-12.4.3 and to properly accommodate the potential for single failure. Each of these sensors provides input into two divisional logics. Each divisional logic is one-out-of-two taken twice, and provides a signal to close both the inboard and outboard dampers and start both SGT subsystems. Current administrative practice specifies that each divisional logic consists of two trip systems, with two channels (arranged in a one-out-of-two logic) in each trip system. Due CNS 3 Revision 0

DISCUSSION OF CHANGES ITS: 3.3.6.2 - SECONDARY CONTAINMENT ISOLATION INSTRUMENTATION TECHNICAL CHANGES - MORE RESTRICTIVE M.3 (continued) to the one-out-of-two taken twice logic of these instruments, if only one channel per trip system is required to be maintained Operable, the single failure criterion is not met. An increase in the number of channels per trip system required to be Operable constitutes a more restrictive change. In addition, the ITS will specify that each divisional logic consists of one trip system; a one-out-of-two taken twice trip system logic. Therefore, the number of channels in the ITS for this Function (proposed Function 3) has been specified as four.

This portion of the change does not change the actual requirements and is considered administrative.

M.4 A finite Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in CTS Table 3.2.D, Note 1.B(2)

(proposed Required Actions C.1.1 and C.2.1) is proposed to be provided to isolate the secondary containment and start the Standby Gas Treatment subsystem. Currently, no Completion Time is provided. This ensures the appropriate actions are completed in a timely manner. This change is consistent with the BWR Standard Technical Specifications, NUREG 1433 and is considered more restrictive on plant operation.

M.5 The Frequencies for performance of Channel Checks are proposed to be changed from once per day to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for the Reactor Low Water Level Function in CTS Table 4.2.A and from once per day during releases via the pathway (which is essentially all the time) to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for the Reactor Building Exhaust Plenum Radiation Monitor in CTS Table 4.21.A.2.

The Channel Check ensures once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that a gross failure of instrumentation has not occurred. This Frequency is based on operating experience that demonstrates that channel failure is rare and the Frequency coincides with performance each shift. This change is consistent with NUREG-1433, which requires the SR to be performed every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

M.6 The Frequency for performance of the Channel Functional Test for CTS Table 4.21.A.2, item 3.f (Reactor Building Ventilation Exhaust Plenum Radiation-High Function, proposed Function 3) is proposed to be reduced from a refueling outage Frequency (18 months) to once every 92 days.

The Frequency of 92 days is based on the reliability analyses of NEDC-31677P-A, "Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation," July 1990, and will ensure the Function is maintained Operable. This is being done for consistency with the Frequencies of other Channel Functional Tests and to conform with the recommendations of NUREG-1433, Rev. 1.

CNS 4 Revision 0

NLS2005051 Enclosure 3 PCIS GROUP 3 INSTRUMENTATION CONFIGURATION FIGURE Figure 4- PCIS Group 3 Reactor Water Cleanup Instrumentation

TRIP SYSTEM A TRIP SUBSYSTEMAl CHANNEL LOGIC NOTESi lLS-57A SN16A-KLA A - STANDBY LIQUID CONTROL PUMP START SWITCH.

PUMP 'A' CONTACT IN INBOARD LOGIC AND PUMP "B CONTACT IN OUTBOARD LOGIC.

'-TRIP SIGNAL B - HIGH NON-REGENERATIVE HEAT EXCHANGER TEMPERATURE TRIP SIGNAL. ONE SENSOR AND RELAY WITH CONTACTS IN BOTH THE INBOARD AND OUTBOARD LOGICS.

A2 TRIP SUBSYSTEM I C - HIGH AREA TEMPERATURE/HIGH SYSTEM FLOW I TRIP SIGNAL. ONE SET OF TEMPERATURE SWITCHES AND ONE FLOW SWITCH IN THE INBOARD LOGIC, I AND ONE SET OF TEMPERATURE SWITCHES AND ONE FLOW SWITCH IN THE OUTBOARD LOGIC.

CHANNEL 1 - ----------- lI TRIP SIGNAL ACTUATION 1 .r .l DEVICE TRIP SYSTEM SEE SEE TRIP SUBSYSTEM81 NOTE A NOTE C . t RIP ~SIGALl6A 2 PROTECTIVE ACTION i~b 4ib RWCU-MOV-MO 15 INBOARDREACTOR 1 l ACTUATED WATERCLEANUP ISOLATIONVALVE CHANNEL I I DEVICE (TS TABLE3.3.6.1-1 FUNCTION 5) 1IS- --- - - -' _ - - ____I -____-__-_ J I TLRI SN LEGEND ELECTRICALRELAY SOLENOID

_ _ \__

MECHANICALCONNECTION ELECTRICALCONNECTION TRIP SUBSYSTEM82 irb ELECTRICALCONTACTS(SHOWN CLOSED)

VALVE CHANNEL

REFERENCES:

lLIS-583 SENSOR 16A-K60 SIGNA - - - - - - - -i 0WG 791E266 SH 5, 6. 12 &13 ILTRP S2NA FIGURE 4 - PCIS GROUP 3 REACTOR WATER CLEANUP INSTRUMENTATION

NLS2005051 Enclosure 4 Markup of TS Bases Page B 3.3-139

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES BACKGROUND 2. Primary Containment Isolation (continued) arranged into two one-out-of-two taken twice fogie trip systems logics.

Each logic receives input from two trig systems. One trip-system logic initiates isolation of all inboard primary containment isolation valves, while the other trip-system logic initiates isolation of all outboard primary containment isolation valves. Each logic closes one of the two valves on each penetration, so that operation of either logic isolates the penetration.

The exception to this arrangement is the Main Steam Line Radiation - High Function. This Function has four channels, whose outputs are arranged in two, two-out-of-two logie trip systems logics for the recirculation sample valves, and in two one, one-out-of-two taken twice Woie trip systems iociics for the mechanical vacuum pumps and associated isolation valves. Each of the recirculation sample valve trip systems logics isolates one of the two valves. Both of the The single mechanical vacuum pump trip-systems logic must actuate to trip the both mechanical vacuum pumps and isolate the associated valves.

The valves isolated by each of the Primary Containment Isolation Functions are listed in Reference 1.

3. 4. High Pressure Coolant Iniection Svstem Isolation and Reactor Core Isolation Cooling System Isolation The Steam Line Flow-High Functions that isolate HPCI and RCIC receive input from two channels, with each channel comprising one trip system using a one-out-of-one logic. Each of the two trip systems in each isolation group (HPCI and RCIC) is connected to one of the two valves on each associated penetration. Each HPCI and RCIC steam Line Flow-High Channel has a time delay relay to prevent isolation due to flow transients during startup.

The HPCI and RCIC Isolation Functions for Steam Supply Pressure-Low receive inputs from four channels. The outputs from these channels are combined in two trip systems, each with two-out-of-two logic to initiate isolation of the associated valves. One tripsystem logic isolates the inboard valve and the other trip-system logic isolates the outboard valve.

Cooper B 3.3-139 Revision 0

ATTACHMENT 3 LIST OF REGULATORY COMMITMENTSl Correspondence Number: NLS2005051 The following table identifies those actions committed to by Nebraska Public Power District (NPPD) in this document. Any other actions discussed in the submittal represent intended or planned actions by NPPD. They are described for information only and are not regulatory commitments. Please notify the Licensing Manager at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.

I COMMITMENT COMMITTED DATE I COMMITMENT NUMBER OR OUTAGE The TS 3.3.6.1 and 3.3.6.2 Bases pages that do not clearly distinguish between "trip Within 90 days after system," "logic trip system," and "logic" are designated for revision as an implementing NLS2005051-01 issuance of the action of this License Amendment Request, and will be changed under the provisions of 10CFR50.59 and TS 5.5.10.

4 4 4 4 4 4 4 4 4 4 4

4 4 4 4 4 4 PROCEDURE 0.42 l REVISION 17 l PAGE 20 OF27