Text

License Amendment Request to Adopt Generic Changes to Standard Technical Specifications
	ML051090436
Person / Time
Site:	Cooper
Issue date:	04/13/2005
From:	Edington R; Nebraska Public Power District (NPPD)
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NLS2005007
	Download: ML051090436 (48)
	v • d • e

Nebraska Public Power District Always there when you need us 50.90 NLS2005007 April 13, 2005 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001

Subject:

License Amendment Request to Adopt Generic Changes to Standard Technical Specifications Cooper Nuclear Station, Docket No. 50-298, DPR-46 The purpose of this letter is for the Nebraska Public Power District (NPPD) to request an amendment to Facility Operating License DPR-46 in accordance with the provisions of 10 CFR 50.4 and 10 CFR 50.90 to revise the Cooper Nuclear Station (CNS) Technical Specifications (TS). The proposed TS revisions clarify the frequency for performing control rod scram time surveillance following outages, clarify which Emergency Core Cooling System (ECCS) instrumentation must be operable when shutdown, eliminate the need to require automatic start of the emergency diesel generators on an ECCS initiation signal at times when the ECCS does not need to be operable, and allow insertion of a control rod block and verification that all rods are fully inserted in lieu of suspending fuel movement when a refueling equipment interlock is inoperable.

These proposed changes are based on the Nuclear Regulatory Commission (NRC)-approved Technical Specification Task Force (TSTF) Travelers TSTF-222-A, TSTF-275-A, and TSTF-300-A, and the portions of TSTF-225 incorporated in the latest issued version of BWR/4 Standard Technical Specifications (NUREG-1433, Revision 3).

NPPD requests NRC approval of the proposed TS change and issuance of the requested license amendment by April 1, 2006. The amendment will be implemented within 60 days of issuance of the amendment. provides NPPD's evaluation of the proposed change. Attachment 2 provides marked up pages showing the proposed changes to the current CNS TS. Attachment 3 provides the revised TS pages in final typed format. Attachment 4 provides marked up TS Bases pages showing the proposed changes to the current Bases for NRC information.

The proposed TS changes have been reviewed by the necessary safety review committees (Station Operations Review Committee and Safety Review and Audit Board). Amendments to the CNS Facility Operating License through Amendment 210, dated February 1, 2005, COOPERNUCLEARSTATION P.O. Box 98 / Brownville, NE 68321-0098 Telephone: (402) 825-3811 / Fax: (402) 825-5211 wwv.nppd .com

NLS2005007 Page 2 of 2 have been incorporated into this request. This request is submitted under oath pursuant to 10 CFR 50.30(b).

By copy of this letter and its attachments, the appropriate State of Nebraska official is notified in accordance with 10 CFR 50.91(b)(1). Copies to the NRC Region IV office and the CNS Resident Inspector are also being provided in accordance with 10 CFR 50.4(b)(1).

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

I declare under penalty of perjury that the foregoing is true and correct.

Executed on 0 1 )

(date)

Ran all K. Edington Vice President - Nuclear and Chief Nuclear Officer

/rer Attachments cc: Regional Administrator w/ attachments USNRC - Region IV Senior Project Manager w/ attachments USNRC - NRR Project Directorate IV-1 Senior Resident Inspector w/ attachments USNRC Nebraska Health and Human Services w/ attachments Department of Regulation and Licensure NPG Distribution w/o attachments CNS Records Iv/

attachments

NLS2005007 Attachment I Page 1 of 16 ATTACHMENT I LICENSE AMENDMENT REQUEST TO ADOPT GENERIC CHANGES TO STANDARD TECHNICAL SPECIFICATIONS COOPER NUCLEAR STATION NRC DOCKET 50-298, LICENSE DPR-46 Revised Pages Technical Specification Pages 3.1-12 3.1-13 3.3-37 3.3-38 3.3-39 3.8-12 3.9-1 1.0 Description 2.0 Proposed Change 3.0 Background 4.0 Technical Analysis 5.0 Regulatory Safety Analysis 5.1 No Significant Hazards Consideration (NSHC) 5.2 Applicable Regulatory Requirements/Criteria 6.0 Environmental Consideration 7.0 References

NLS2005007 Attachment I Page 2 of 16 1.0 Description The Nebraska Public Power District (NPPD) requests that Operating License No. DPR-46 for Cooper Nuclear Station (CNS) be amended by the following four revisions to the Technical Specifications (TS):

1. In TS Section 3.1.4, Control Rod Scram Times, the specified Frequency for Surveillance Requirement (SR) 3.1.4.1 and SR 3.1.4.4 are revised to clarify that control rod scram time testing is required only for core cells in which work on the control rod or drive has been performed or fuel has been moved or replaced.

2. In Section 3.3.5.1, ECCS Instntmentation, Footnote (a) in Table 3.3.5.1-1 is revised to clarify that the ECCS initiation instrumentation, identified as being required in modes 4 and 5, is required to be operable only when the associated ECCS subsystems are required to be operable as defined in LCO 3.5.2, ECCS- Shutdown.

3. In TS Section 3.8.2, AC Sources - Shutdown, an additional note is added as Note 2 to SR 3.8.2.1 to specify that SR 3.8.1.11, the surveillance that verifies automatic start of the emergency diesel generators and automatic load shedding from the emergency buses, is considered to be met without the ECCS initiation signals operable when ECCS initiation signals are not required to be operable per Table 3.3.5.1-1, ECCS Instrumentation.

4. In TS Section 3.9.1, Refueling Equipment Interlocks, Required Actions A.2. 1 and A.2.2 are added to allow insertion of a control rod withdrawal block and verification that all control rods are fully inserted as alternate actions to suspending in-vessel fuel movement in the event that one or more required refueling equipment interlocks are inoperable.

2.0 Proposed Changes

1. Revise Frequency of Surveillance Requirements (SR) 3.1.4.1 and SR 3.1.4.4 SR 3.1.4.1 requires verifying that each control rod scram time is within the limits of Table 3.1.4-1 with reactor steam dome pressure greater than or equal to 800 psig (1) prior to exceeding 40% rated thermal power (RTP) after each refueling, and (2) prior to exceeding 40% RTP after each reactor shutdown greater than or equal to 120 days.

It is proposed to delete the first part of SR 3.1.4.1 Frequency, "prior to exceeding 40%

RTP after each refueling."

SR 3.1.4.4 requires verifying that each control rod scram time is within the limits of Table 3.1.4-1 with reactor steam dome pressure greater than or equal to 800 psig (1)

"prior to exceeding 40% RTP after work on control rod or CRD [Control Rod Drive]

System that could affect scram time," and (2) "prior to exceeding 40% RTP after fuel movement within the reactor pressure vessel."

NLS2005007 Page 3 of 16 It is proposed to revise part (2) to state "prior to exceeding 40% RTP after fuel movement within the affected core cell." It is also proposed to reverse the order of the two parts of SR 3.1.4.4 Frequency.

2. Revise Footnote (a) in Table 3.3.5.1-1, Enmergency Core Cooling System Instnumentation Footnote (a) clarifies when certain functions of the Core Spray (CS) and Low Pressure Coolant Injection (LPCI) System instrumentation are required to be OPERABLE when the plant is shutdown, (i.e., in modes 4 and 5). The footnote currently states:

"When associated subsystems are required to be OPERABLE."

It is proposed to revise the footnote to state:

"When associated ECCS subsystems are required to be OPERABLE per LCO 3.5.2, ECCS - Shutdown."

3. Add New Note in SR 3.8.2.1 SR 3.8.2.1 specifies the SRs of Specification 3.8.1 for AC Sources that are applicable (i.e., must be met) in modes 4 and 5. SR 3.8.1.11 requires verification that an actual or simulated loss of offsite power in conjunction with an actual or simulated ECCS initiation signal results in de-energizing and load shedding of the emergency buses and automatic start of the diesel generators (DGs).

It is proposed to add the following as new Note 2 to SR 3.8.2. 1:

"SR 3.8.1.11 is considered to be met without the ECCS initiation signals OPERABLE when ECCS initiation signals are not required to be OPERABLE per Table 3.3.5.1-1."

It is also proposed to number the current note as "Note 1," and to make the header plural, i.e., "NOTES".

4. Add Alternate Required Actions A.2.1 and A.2.2 to TS 3.9.1 Condition A.

TS 3.9.1, Condition A, addresses a situation in which one or more required refueling equipment interlocks are inoperable. Current Required Action A.1 is to immediately suspend in-vessel fuel movement.

It is proposed to add alternative Required Actions A.2.1 and A.2.2. Required Action A.2. 1 is to immediately insert a control rod block. Required Action A.2.2 is to immediately verify that all control rods are fully inserted.

NLS2005007 Page 4 of 16 Conforming changes to the TS Bases for the proposed changes to the technical specifications in this amendment request are provided for the NRC's information.

In summary, NPPD requests amendment of the CNS operating license (1) to clarify that scram time testing is needed only on control rods that are in core cells affected by fuel movement and/or control rod maintenance, (2) to clarify that certain ECCS instrumentation is required to be operable in modes 4 and 5 only if the associated ECCS subsystem is required to be operable in modes 4 and 5, (3) to allow AC Sources to be considered operable in modes 4 and 5 without requiring that the DGs be capable of responding to ECCS initiation signals when the ECCS subsystems are not required to be operable, and (4) to allow insertion of a control rod block and verification that all control rods are fully inserted as an alternative to suspending fuel movement when a refueling equipment interlock is inoperable.

3.0 Background The following is background information regarding plant systems involved with each of the proposed changes to TS.

1. Revise Frequency of SR 3.1.4.1 and SR 3.1.4.4 The scram function of the Control Rod Drive (CRD) System controls reactivity changes during abnormal operational transients to ensure that specified acceptable fuel design limits are not exceeded. The control rods are scrammed using hydraulic pressure exerted on the CRD piston.

The Design Basis Accident (DBA) and transient analyses assume that all of the control rods scram at a specified insertion rate. The resulting negative scram reactivity forms the basis for the determination of plant thermal limits (for example, MCPR).

Surveillance of each individual control rod scram time ensures the scram reactivity assumed in the DBA and transient analyses can be met.

SR 3.1.4.1 requires that the scram time of each control rod be verified to be within the limits of Table 3.1.4-1 with reactor steam dome pressure > 800 psig prior to exceeding 40% RTP after each refueling and after each reactor shutdown greater than or equal to 120 days. Maximum scram insertion times occur at a reactor steam dome pressure of approximately 800 psig because of the competing effects of reactor steam dome pressure and stored accumulator energy. Therefore, demonstration of adequate scram times at reactor steam dome pressure > 800 psig ensures that the measured scram times will be within the specified limits at higher pressures. Verifying proper scram times before exceeding 40% RTP ensures that scram time testing is performed within a reasonable time after a shutdown duration of greater than or equal to 120 days.

SR 3.1.4.4 requires that the scram time of each control rod be verified to be within the limits of Table 3.1.4-1 with reactor steam dome pressure > 800 psig prior to exceeding 40% RTP after work on control rods or CRD system and after fuel movement within

NLS2005007 Attachment I Page 5 of 16 the reactor pressure vessel. This testing ensures that, prior to withdrawing the control rod for continued operation, the control rod scram performance is acceptable for operating reactor pressure conditions. When only fuel movement occurs, then only those control rods associated with the core cells affected by the fuel movement are required to be scram time tested.

2. Revise Footnote (a) in Table 3.3.5.1-1, Emergency Core Cooling System Instrumentation The purpose of the ECCS instrumentation is to initiate appropriate responses from the systems to ensure that the fuel is adequately cooled in the event of a DBA or transient.

The ECCS instrumentation actuates Core Spray (CS) System, Low Pressure Coolant Injection (LPCI) System (a mode of Residual Heat Removal), High Pressure Coolant Injection (HPCI) System, Automatic Depressurization System (ADS), and the DGs.

Table 3.3.5.1-1 identifies instrumentation associated with the actuation of the ECCS.

The ECCS whose instrumentation is addressed in Table 3.3.5.1-1 are CS, LPCI, HPCI, and ADS.

Footnote (a) in Table 3.3.5.1-1 reads: "When associated subsystem(s) are required to be OPERABLE." This footnote clarifies what ECCS instrumentation is required to be operable when the plant is in modes 4 and 5. The footnote applies to instrumentation in the CS System (Functions la, Ic, Id, and le) and the LPCI System (Functions 2a, 2c, 2f, and 2g).

3. Add New Note to SR 3.8.2.1 The LCO in TS Section 3.8.2, "AC Sources - Slutdoiwn," identifies the electrical AC systems required to be operable when the plant is in shutdown modes 4 and 5, and during movement of irradiated fuel assemblies in the secondary containment. The operability of the minimum AC sources during these times ensures that (a) CNS can be maintained in shutdown or refueling condition for extended periods; (b) sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and (c) adequate AC electrical power is provided to mitigate events postulated during shutdown, such as an inadvertent draindown of the vessel or a fuel handling accident.

SR 3.8.2.1 requires certain SRs from LCO 3.8.1, "AC Sources - Operating," to be performed and met to demonstrate operability of the AC sources in modes other than 1, 2, and 3.

SR 3.8.1.11 requires verifying that three functions will occur in response to an actual or simulated loss of offsite power (LOOP) signal in conjunction with an actual or simulated ECCS initiation signal. The three functions are:

NLS2005007 Page 6 of 16 (1) Emergency buses will de-energize, (2) Loads on the emergency buses will automatically drop off (load shed), and (3) The diesel generator (DG) will automatically start from standby condition and satisfy five specific criteria.

4. Add Required Actions A.2.1 and A.2.2 to TS 3.9.1 Condition A.

Refueling equipment interlocks restrict the operation of the refueling equipment or the withdrawal of control rods to reinforce unit procedures that prevent the reactor from achieving criticality during refueling. The refueling interlock circuitry senses the conditions of the refueling equipment and the control rods. Depending on the sensed conditions, interlocks are actuated to prevent the operation of the refueling equipment or the withdrawal of control rods (rod block). The refueling equipment interlocks protect against prompt reactivity excursions during mode 5.

Refueling interlocks, in combination with core nuclear design and refueling procedures, limit the probability of an inadvertent criticality. The combination of refueling interlocks for control rods and the refueling platform provide redundant methods of preventing inadvertent criticality, even after procedural violations.

Control rods, when fully inserted, serve as the system capable of maintaining the reactor subcritical in cold conditions during all fuel movement activities and accidents.

Current Required Action A.1 requires that in-vessel fuel movement with equipment associated with the inoperable interlocks be immediately suspended if one or more refueling equipment interlocks is inoperable. This action ensures that fuel movement is not performed with equipment that would potentially not be blocked from unacceptable operations (e.g., loading fuel into a cell with a control rod withdrawn).

4.0 Technical Analysis

1. Revise Frequency of SR 3.1.4.1 and SR 3.1.4.4 These revisions were proposed in Technical Specification Task Force (TSTF) traveler TSTF-222-A, Revision 1 (Reference 2), and approved by the NRC as reflected in letter to the Nuclear Energy Institute (NEI) dated May 12, 1999 (Reference 3). These revisions have been incorporated into the latest approved version of BWR/4 Standard Technical Specifications (STS) issued by the NRC (NUREG-1433, Revision 3.0).

The first sentence of TSTF-222-A states its purpose: "Clarify that post-refueling control rod scram time testing only applies to control rods affected by movement of fuel." The remainder of the TSTF-222-A justification discusses how SR 3.1.4.1 effectively requires that all rods be scram time tested even if only one fuel assembly is moved, and how the changes proposed in the TSTF will resolve this misinterpretation.

CNS converted to STS based on Revision 1 of NUREG-1433 by Amendment No. 178 dated July 31, 1998. As part of that conversion CNS adopted a frequency for SR

NLS2005007 Page 7 of 16 3.1.4.1 slightly different from that reflected in NUREG-1433, Revision 1. The modified frequency was intended to clarify that scram time testing on each control rod is required only following a complete refueling, and not after replacement of a limited number of fuel assemblies.

Current CNS SR 3.1.4.1 requires that the scram time of each control rod be verified to be within Table 3.1.4-1 prior to exceeding 40% RTP after a refueling or after a shutdown of 120 days or greater. As revised by this request SR 3.1.4.4 will require scram time testing of the control rod prior to exceeding 40% RTP after fuel movement within the affected core cell. In a typical, routine refueling outage, all core cells are likely to be affected as a result of some fuel movement, e.g., a spent fuel assembly is replaced with a fresh assembly, a fuel assembly is relocated from one cell to another, or a fuel assembly is reoriented within a core cell. Therefore, scram time testing will continue to be conducted on all control rods following a routine refueling.

If a core cell is not affected by (1) movement of one of the four fuel assemblies in the cell, (2) replacement of the control rod in that cell, or (3) maintenance on the control rod drive system for the rod in that cell, the scram time of the control rod in that core cell is not expected to be impacted. As a result there would be no need to conduct scram time testing on that control rod. Furthermore, it is expected that the periodic scram time testing of a representative sample (10% of the control rods), as required by SR 3.1.4.2, will identify any long term phenomenon that could result in degradation of scram time.

Revising the second Frequency of SR 3.1.4.4 to require scram time testing after fuel movement "within the affected core cells" clarifies that only those control rods in core cells in which fuel was moved or replaced or control rod maintenance was performed are required to be scram time tested. It is expected that all core cells will be affected in this manner during a routine refueling outage, and therefore, that scram time testing will be required on all control rods.

Deleting the first part of SR 3.1.4.1 Frequency, and revising the second part and reversing the order of the two parts of SR 3.1.4.4 Frequency makes the CNS TS consistent with these SRs in the current version of BWR/4 STS (NUREG-1433, Rev.

3.0).

These changes are expected to be of benefit in the conduct of outages in which only a limited number of fuel cells are affected by avoiding the need to perform scram time testing on control rods in core cells that were not affected by fuel moves, control rod replacement, or control rod drive maintenance.

2. Revise Footnote (a) in Table 3.3.5.1-1, Emergency Core Cooling System Instninentation This revision was proposed in TSTF-275-A, Revision 0 (Reference 5), which was approved by the NRC as reflected in letter to NEI dated December 21, 1999

NLS2005007 Attachment I Page 8 of 16 (Reference 6). This revision has been incorporated into the latest approved version of BWR/4 STS issued by the NRC (NUREG-1433, Revision 3.0).

TSTF-275-A makes the following statements:

A. "The proposed change to LCO 3.3.5.1 would clarify which, if any, ECCS instrumentation is required to be operable in Mode 5, with RPV level greater than or equal to [23] feet above the RPV flange to support EDG operability." [For CNS the limit is 21 feet above the pressure vessel flange, as reflected in LCO 3.5.2 Applicability].

B. "Consistent with the operability requirements in LCO 3.5.2, ECCS -

Shutdown, ECCS is not required to be operable when the plant is at high water level. If the ECCS is not required then the instrument whose function it is to initiate ECCS should not be required."

C. "...the current footnote implies that the ECCS instrumentation is required to be operable not only when the associated ECCS is required to be operable but also when the associated ECCS support systems are required to be operable. This is incorrect since these support systems also support other functions that are required at times when the ECCS system and associated initiation instrumentation is not needed (e.g., the DGs are required during fuel handling)."

In accordance with the Applicability of TS 3.5.2, an ECCS subsystem is not required to be operable when the plant is in mode 5 with the spent fuel pool gates removed and water level is greater than or equal to 21 feet above the top of the pressure vessel flange. If the ECCS subsystem is not required to be operable, then the instrumentation associated with that ECCS subsystem also should not be required to be operable. In accordance with TSTF-275-A, footnote (a) has been modified to only require these functions to be operable when the associated ECCS subsystems are required to be operable per LCO 3.5.2.

Clarifying this note in Table 3.3.5.1-1 is expected to be beneficial in that it will avoid requiring the applicable instrumentation functions to be operable when ECCS subsystems are not required to be operable in mode 5.

3. Add Note to SR 3.8.2.1 TSTF-300-A, Revision 0 (Reference 7), proposes addition of the following note to SR 3.8.2.1 inNUREG-1433, Rev.1:

SR 3.8.1.12 and SR 3.8.1.19 are not required to be met when associated ECCS subsystem(s) are not required to be OPERABLE per LCO 3.5.2, "ECCS - Shutdown."

NLS2005007 Attachment I Page 9 of 16 As Justification the TSTF states the following:

"Exceptions are added to the DG SR requirements for LCO 3.8.2, AC Sources - Shutdown. These exceptions will eliminate the requirement that the DG be capable of responding to ECCS initiation signals while the ECCS subsystems are not required to be Operable. During shutdown modes, when the vessel is defueled or when the reactor cavity is flooded, the ECCS Systems are not required to be Operable. Therefore, the ECCS-start functions of the DGs serve no safety significant support function. As such, the SRs that test/prove the DG capability to respond to an ECCS-start signal may be removed from DG operability considerations at these times when the ECCS Systems are not required to be Operable."

The above justification from TSTF-300-A is applicable to CNS. When ECCS subsystems are not required to be operable per LCO 3.5.2, "ECCS- Shutndown," the DG is not required to start in response to ECCS initiation signals. However, the DG is still required to meet the other attributes of SR 3.8.1.11 when associated ECCS initiation signals are not required to be operable per TS Table 3.3.5.1-1.

Letter to NEI dated April 21, 1999 (Reference 8) documented the NRC approval of TSTF-300. The TSTF-300 revision has been incorporated into the latest approved version of BWR/4 STS (NUREG-1433, Revision 3.0).

The wording of the new Note 2 proposed for CNS differs from the note reflected in TSTF-300-A in two respects. The first difference is that the note proposed for CNS refers only to SR 3.8.1.11. CNS has one SR (SR 3.8.1.11) that bounds SRs 3.8.1.11, 3.8.1.12 and 3.8.1.19 from NUREG-1433. CNS did not adopt SRs 3.8.1.11 and 3.8.1.12 from NUREG-1433, Revision 1 as part of the conversion to STS because DG testing practices at CNS did not perform Loss of Offsite Power (LOOP) and Loss of Coolant Accident (LOCA) tests individually. The results of the combined LOOP/LOCA test at CNS bounds the results of the individual LOOP and LOCA tests.

CNS SR 3.8.1.11 is equivalent to SR 3.8.1.19 in NUREG-1433, Revision 3.

The second difference is that the note proposed for CNS states that the SR "is considered to be met" whereas the note in TSTF-300-A states that the SR is "not required to be met." The wording of the note proposed for CNS is consistent with that issued for the Duane Arnold Energy Center by Amendment No. 234 dated October 3, 2000 (Reference 9). CNS considers the proposed wording of the note for SR 3.8.2.1 to be more appropriate than the note in TSTF-300-A.

Adding this note to SR 3.8.2.1 is expected to be beneficial in that it has the potential to facilitate performance of maintenance on a DG when CNS is in mode 4 or 5.

NLS2005007 Page 10 of 16

4. Add Required Actions A.2.1 and A.2.2 to TS 3.9.1 Condition A.

The proposed revisions were included in TSTF-225, Revision 2 (Reference 4).

Although the TSTF was not approved by the NRC, the revisions proposed in this amendment request have been incorporated into the latest approved version of BWR/4 STS (NULREG-1433, Revision 3.0).

As stated in the TSTF, the additional Required Actions provide an alternative action for when the refueling interlocks are inoperable. Required Action A.2.1 requires that a control rod withdrawal block be inserted immediately. Required Action A.2.2 requires immediately verifying that all control rods are fully inserted. These actions will ensure that control rods are not withdrawn, and cannot be withdrawn, as a result of the continuous block to control rod withdrawal in place. These alternate actions provide protection against inadvertent criticality.

The safety objective of the refueling equipment interlocks is to prevent an inadvertent criticality during refueling operations. The safety design basis is (1) all controls rods shall be fully inserted during fuel movements in or over the reactor core, and (2) no more than one rod shall be withdrawn from fully inserted when reactor is in the refuel mode.

The insertion of a control rod block (Required Action A.2.1) will ensure that no control rod can be inadvertently withdrawn. The control rod block does not prohibit insertion of control rods. Verifying that all control rods are fully inserted (Required Action A.2.2) will ensure that the maximum amount of negative reactivity available from the control rods is inserted into the core. The combination of blocking control rod withdrawal and ensuring that the rods are fully inserted assures that the reactor cannot inadvertently be made critical. This assures that the safety objective is met and the safety design basis is satisfied.

The TSTF states that the first refueling equipment interlock safety function is to block control rod withdrawal whenever fuel is being moved over or in the reactor vessel, and that the second safety function is to prevent fuel from being loaded into the reactor vessel when a control rod is withdrawn. As stated in the TSTF, the proposed alternative Required Actions will perform these functions by requiring that a control rod block be placed in effect.

The withdrawal block inserted by Required Action A.2.1 will ensure that no control rod will respond (i.e., the rod will remain inserted) if rod withdrawal is attempted. As noted in TSTF-225, Rev. 2, the verification of all control rods being fully inserted by proposed alternate Required Action A.2.2 is in addition to the periodic verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months required by SR 3.9.3.1 when loading fuel assemblies into the core.

Like Required Action A. 1, Required Actions A.2. 1 and A.2.2 will ensure that unacceptable operations (e.g., loading fuel into a cell with a control rod withdrawn) are blocked.

NLS2005007 Page I Iof 16 This proposed change would be beneficial in that it would allow fuel movement to continue in the event that a refueling equipment interlock is inoperable, while continuing to maintain a sufficient level of protection against inadvertent criticality.

5.0 Regulatory Safety Analysis 5.1 No Significant Hazards Consideration 10 CFR 50.91(a)(1) requires that licensee requests for operating license amendments be accompanied by an evaluation of significant hazard posed by issuance of the amendment. Nebraska Public Power District (NPPD) has evaluated this proposed amendment with respect to the criteria given in 10 CFR 50.92 (c).

NPPD is requesting an amendment of the operating license for the Cooper Nuclear Station (CNS.) The requested amendment involves four specific revisions to the CNS Technical Specifications (TS). Those revisions are as follows:

1. Revision of CNS TS Surveillance Requirement (SR) 3.1.4.1 and SR 3.1.4.4 SR 3.1.4.1 requires that the time to insert a control rod (referred to as "scram time") be verified to be within specific limits prior to exceeding 40% Rated Thermal Power (RTP) (1) after each refueling, and (2) after each shutdown greater than or equal to 120 days. SR 3.1.4.4 requires scram time testing of each control rod prior to exceeding 40% RTP (1)after work on the control rod or control rod drive system that could affect scram time, and (2) after fuel movement within the reactor pressure vessel. The requirement to perform SR 3.1.4.1 following each refueling is being replaced by a revised requirement to perform SR 3.1.4.4 after fuel movement within the affected core cell. (An "affected core cell" is one in which the control rod has been replaced, the drive system worked on, or the fuel has been reoriented, relocated or replaced). Thus, the revised TS will continue to require that the scram time of control rods be verified in core cells in which fuel has been relocated or replaced as part of a routine refueling (typically all core cells).

2. Revision of TS Table 3.3.5.1-1 TS Table 3.3.5.1-1, "Emergency Core Cooling System Instrumentation,"

identifies the instrumentation that must be maintained operable to ensure proper actuation of Emergency Core Cooling System (ECCS) subsystems. A footnote in the table explains what instrumentation must be maintained operable when the plant is in operating mode 4 (cold shutdown) and mode 5 (refueling). The footnote is being revised to clarify that certain ECCS instrumentation is required to be operable in modes 4 and 5 only if the associated ECCS subsystem is required to be operable in modes 4 and 5.

NLS2005007 Page 12 of 16

3. Revision of TS SR 3.8.2.1 TS SR 3.8.2.1 requires testing of Alternating Current (AC) equipment when CNS is shutdown. One of the required tests is to verify that each emergency diesel generator (DG) will start automatically in response to a loss of offsite power signal in conjunction with an ECCS initiation signal. (The purpose of the DGs is to provide AC electrical power to essential components to mitigate an accident if power from offsite is not available.) The ECCS subsystems are not required to be operable when shutdown with the pressure vessel opened and water level in the reactor cavity raised to allow fuel movement. Under these conditions the automatic start of the DGs in response to an ECCS initiation signal serves no safety significant support function. A note is being added to SR 3.8.2.1 to state that SR 3.8.1.11, the SR that, in part, tests the capability of a DG to respond to an ECCS initiation signal, is considered to be met without the ECCS initiation signals being operable at those times when the ECCS initiation signals are not required to be operable per Table 3.3.5.1-1.

4. Revision of TS 3.9.1, Condition A Required Action TS 3.9.1, Condition A Required Action A.1 requires that fuel movement be suspended if an interlock on refueling equipment is discovered to be inoperable. Refueling equipment interlocks prevent movement of refueling equipment or withdrawal of a control rod (the means of controlling the reactivity in the core) under certain conditions. This is a protection against an inadvertent (accidental) nuclear criticality. Alternate Required Actions to insert a control rod block (an electronic control that prevents withdrawal of a control rod) and to verify that all control rods are fully inserted into the reactor core are being added in lieu of stopping fuel movement if a refueling equipment interlock is discovered to be inoperable.

Each of the four proposed changes is evaluated against the three criteria of 10 CFR 50.92(c) in the following evaluation. The evaluation supports a finding of "no significant hazards" for the proposed amendment.

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

No.

1. Revision of CNS TS SR 3.1.4.1 and SR 3.1.4.4. The frequency at which control rod scram time is verified is not a precursor of an accident. A scram time slower than required might result in an increase in the consequences of an accident. However, revising the frequency for verifying the scram time of the control rods does not impact the scram time. Verifying that the scram time is acceptable will continue to be required prior to plant startup following

NLS2005007 Page 13 of 16 fuel movement or work on the control rods or control rod drive system.

Therefore, revising the frequency for verifying insertion time to clarify when it is required does not involve a significant increase in the probability of an accident or an increase in the consequences of an accident.

2. Revision of TS Table 3.3.5.1-1. Clarifying when certain ECCS instrumentation must be operable with the plant shut down will not increase either the probability of an accident or the consequences of the accident. The ECCS instrumentation is required to be operable only when the associated ECCS subsystems are required to be operable. This continues to ensure that the instrumentation will be operable when it is required.

3. Revision of TS SR 3.8.2.1. The frequency of verifying certain actions by surveillances is not a precursor to accidents. Clarifying that the actions required in response to an ECCS initiation signal are not required when the ECCS initiation signals are not required to be operable does not result in increased probability of an accident or increased consequences of an accident.

Not requiring that a DG automatically start in response to the ECCS initiation signal when the ECCS subsystems that are supported by the DG are not required to be operable does not reduce the required ECCS protection.

4. Revision of TS 3.9.1. Condition A Required Action. The actions taken when a refueling equipment interlock is inoperable are not initiators of any accident previously evaluated. The level of protection against withdrawing a control rod during the insertion of a fuel assembly or loading a fuel assembly into the vessel with a control rod withdrawn, provided by the proposed alternate Required Actions, is equivalent to that provided by the current Required Action. The radiological consequences of an accident described in the Updated Safety Analysis Report (USAR) while taking the proposed alternate Required Actions are not different from the consequences of an accident under the current Required Actions.

Based on the above NPPD concludes that the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

No. The proposed changes to the CNS operating license involve revisions to the requirements for when certain surveillances are to be performed (change no. 1 and no. 3), clarification of when ECCS instrumentation is required to be operable (change no. 2), and addition of alternative Required Actions if certain plant components are inoperable (change no. 4). These changes will not result in revision of plant design, physical alteration of a plant structure, system, or component (SSC), or installation of a new or different type of

NLS2005007 Attachment I Page 14 ofl6 equipment. The changes do not involve any revision of how the plant, an SSC, or a refueling equipment interlock, are operated. Based on this, the proposed changes do not create the possibility of a new or different kind of accident.

3. Do the proposed changes involve a significant reduction in a margin of safety?

No.

1. Revision of CNS TS SR 3.1.4.1 and SR 3.1.4.4. Sufficiently rapid insertion of control rods following certain accidents (scram time) will prevent fuel damage, and thereby maintain a margin of safety to fuel damage. No change is being made to the required insertion rate specified in plant technical specifications. Clarifying when control rod insertion times must be verified following movement of fuel assemblies, without actually changing the requirement (verification of insertion times will continue to be required whenever work that might impact the rod insertion time is done), does not reduce the margin of safety related to fuel damage.

2. Revision of TS Table 3.3.5.1-1. Clarifying when certain ECCS instrumentation is required to be operable when CNS is in a shutdown mode does not change the requirement. Not requiring ECCS signals that initiate a DG to be operable when the ECCS subsystems that are supported by the DG are not required to be operable does not result in a reduction of a margin of safety for the safety related equipment that is required to be operable.

3. Revision of TS SR 3.8.2.1. Clarifying that automatic start of the DGs in response to the ECCS initiation signal is not required when the ECCS subsystems that are supported by the DG are not required to be operable does not result in a reduction in a margin of safety.

4. Revision of TS 3.9.1. Condition A Required Action. The proposed alternate Required Actions to be taken when a refueling interlock is inoperable provide a level of protection against inadvertent criticality while inserting or moving fuel in the reactor vessel that is equivalent to the level provided by the current Required Action. As a result, the proposed alternate Required Actions do not result in a significant reduction in a margin of safety related to protection against inadvertent criticality when inserting or moving fuel assemblies.

Based on the above NPPD concludes that the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above evaluation of the three criteria, NPPD concludes that the proposed amendment presents no significant hazards consideration under the

NLS2005007 Page 15 of 16 standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory Requirements/Criteria

1. USAR Appendix F, Criterion 15, EngineeredSafety FeaturesProtection Systems, states:

"Protection systems shall be provided for sensing accident situations and initiating the operation of necessary engineered safety features."

2. USAR Appendix F, Criterion 29, Reactivity Shutdown Capability, states:

"At least one of the reactivity control systems provided shall be capable of making the core subcritical under any condition (including anticipated operational transients) sufficiently fast to prevent exceeding acceptable fuel damage limits. Shutdown margins greater than the maximum worth of the most effective control rod when fully withdrawn shall be provided."

CNS continues to satisfy the above applicable regulatory requirements with the proposed changes.

6.0 Environmental Consideration 10 CFR 51.22(b) allows that an environmental assessment (EA) or an environmental impact statement (EIS) is not required for any action included in the list of categorical exclusions in 10 CFR 51.22(c). 10 CFR 51.22(c)(9) identifies an amendment to an operating license which changes a requirement with respect to installation or use of a facility component located within the restricted area, or which changes an inspection or a surveillance requirement, as a categorical exclusion if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant hazards consideration, (2) result in a significant change in the types or significant increase in the amount of any effluents that may be released off-site, or (3) result in an increase in individual or cumulative occupational radiation exposure.

NPPD has reviewed the proposed license amendment and concludes that it meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(c), no EIS or EA needs to be prepared in connection with issuance of the proposed license changes. The basis for this determination is as follows:

1. The No Significant Hazards Consideration evaluation presented in Section 5.1 concluded that the requested changes do not involve a significant hazards consideration.

NLS2005007 Page 16 of 16

2. The proposed changes to TS do not impact any effluent at CNS. Therefore, the proposed changes will not result in a significant change in the types or significant increase in the amounts of any effluents that may be released off-site.

3. The proposed changes to the TS do not significantly increase individual or cumulative occupational radiation exposure.

7.0 References

1. BWR4 Standard Technical Specifications (NUREG-1433, Revision 3.0).

2. TSTF-222-A, "Control Rod Scram Time Testing," Revision 1.

3. Letter from William D. Beckner, NRC, to Mr. James Davis, NEI, dated May 12, 1999.

4. TSTF-225, "Fuel movement with inoperable refueling equipment interlocks,"

Revision 2.

5. TSTF-275-A, "Clarify requirement for EDG start signal on RPV Level - Low, Low, Low during RPV cavity flood-up," Revision 0.

6. Letter from William D. Beckner, NRC, to Mr. James Davis, NEI, dated December 21, 1999.

7. TSTF-300-A, "Eliminate DG LOCA-Start SRs while in S/D when no ECCS is Required," Revision 0.

S. Letter from William D. Beckner, NRC, to Mr. James Davis, NEI, dated April 21, 1999.

9. Duane Arnold Energy Center (License No. DPR-49), Amendment No. 234 dated October 3, 2000.

NLS2005007 Page 1 of8 ATTACHMENT 2 PROPOSED TECHNICAL SPECIFICATION REVISIONS (MARK-UP)

COOPER NUCLEAR STATION NRC DOCKET 50-298, LICENSE DPR-46 Technical Specification Pages 3.1-12 3.1-13 3.3-37 3.3-38 3.3-39 3.8-12 3.9-1

Control Rod Scram Times 3.1.4

(-rS T-- 2za) 3.1 REACTIVITY CONTROL SYSTEMS 3.1.4 Control Rod Scram Times LCO 3.1.4 a. No more than 10 OPERABLE control rods shall be "slow,"

in accordance with Table 3.1.4-1; and

b. No more than 2 OPERABLE control rods that are "slow" shall occupy adjacent locations.

APPLICABILITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months LCO not met.

SURVEILLANCE REQUIREMENTS

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

During single control rod scram time Surveillances, the control rod drive (CRD) pumps shall be isolated from the associated scram accumulator.

SURVEILLANCE FREQUENCY I

SR 3.1.4.1 Verify each control rod scram time is within the limits of Table 3.1.4-1 with reactor steam dome pressure > 800 psig.

I.

Cooper 3.1-12 Amendment No. 44&-

Control Rod Scram Times 3.1.4 r5TF - 222y)

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.1 (continued) Prior to exceeding 40% RTP after each reactor shutdown

> 120 days SR 3.I.4.2 Verify, for a representative sample, each 120 days tested control rod scram time is within the cumulative limits of Table 3.1.4-1 with reactor steam operation in dome pressure > 800 psig. MODE I SR 3.1.4.3 Verify each affected control rod scram time Prior to is within the limits of Table 3.1.4-1 with declaring any reactor steam dome pressure. control rod OPERABLE after work on control rod or CRD System that could affect scram time SR 3.1.4.4 Verify each affected control rod scram time >Prior to is within the limits of Table 3.1.4-1 with exceeding reactor steam dome pressure > 800 psig. 40% RTP after work on control rod or CRD System that could affect scram time AND Prior to exceeding 40%

RTP after fuel movement within , , )

the xr citj Cooper 3.1-13 Amendment No. .PO-

ECCS Instrumentation 1-I~ -iF 5 3.3.5.1 Table 3.3.5.1-1 (page 1 of 6)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES REQUIRED REFERENCED OR OTHER CHANNELS FROC SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE

1. Core Spray System

a. Reactor Vessel Water 1,2,3, 4 (b) B SR 3.3.5.1.1 > -113 inches Level -Low Low Low SR 3.3.5.1.2 (Level 1) 4 (a), 5 (a) SR 3.3.5.1.4 SR 3.3.5.1.5

b. Drywell 1,2,3 4 (b) B SR 3.3.5.1.2 < 1.84 psig Pressure -High SR 3.3.5.1.4 SR 3.3.5.1.5

c. Reactor Pressure-Low 1,2,3 4 C SR 3.3.5.1.2 ' 291 psig (Injection Permissive) SR 3.3.5.1.4 and SR 3.3.5.1.5 < 436 psig 4(a), 5 (a) 4 B SR 3.3.5.1.2 > 291 psig SR 3.3.5.1.4 and SR 3.3.5.1.5 < 436 psig

d. Core Spray Pump 1,2,3, 1 per pump E SR 3.3.5.1.2 > 1370 gpm Discharge Flow-Low SR 3.3.5.1.4 (Bypass) 4 (a), 5 (a) SR 3.3.5.1.5

e. Core Spray Pump 1,2,3, 1 per pump C SR 3.3.5.1.2 > 9 seconds Start-Time Delay Relay SR 3.3.5.1.4 and 4 (a), 5 (a) SR 3.3.5.1.5 < 11 seconds

2. Low Pressure Coolant Injection (LPCI) System

a. Reactor Vessel Water 1,2,3, 4 B SR 3.3.5.1.1 > -113 inches Level -Low Low Low SR 3.3.5.1.2 (Level 1) 4 (a) 5 (a) SR 3.3.5.1.4 SR 3.3.5.1.5 (continued)

(a) When associated tems) are required to be OPERABLE, ) g LO 3,5.2, E cc5-. hucou (b) Also required to initiate the associated diesel generator (DG).

Cooper 3.3-37 Amendment No. 4*8-

ECCS Instrumentation T'S -1F -Z ) 3.3.5.1 Table 3.3.5.1-1 (page 2 of 6)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES REQUIRED REFERENCED OR OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE

2. LPCI System (continued)

b. DrywelL 1,2,3 4 B SR 3.3.5.1.2 < 1.84 psig Pressure -High SR 3.3.5.1.4 SR 3.3.5.1.5

c. Reactor Pressure -Low 1,2,3 4 C SR 3.3.5.1.2 > 291 psig (Injection Permissive) SR 3.3.5.1.4 and SR 3.3.5.1.5 < 436 psig 4(a), 5 (a) 4 B SR 3.3.5.1.2 > 291 psig SR 3.3.5.1.4 and SR 3.3.5.1.5 < 436 psig

d. Reactor Pressure -Low 1 (c), 2 (c). 4 C SR 3.3.5.1.2 > 199 psig (Recirculation SR 3.3.5.1.4 and Discharge Valve 3 (c) SR 3.3.5.1.5 < 221 psig Permissive)

e. Reactor Vessel Shroud 1,2,3 2 B SR 3.3.5.1.1 > -193.19 Level -Level 0 SR 3.3.5.1.2 inches SR 3.3.5.1.4 SR 3.3.5.1.5

f. Low Pressure Coolant 1,2,3, 1 per pump C SR 3.3.5.1.2 Injection Pump SR 3.3.5.1.4 Start -Time Delay 4 (a), 5 (a) SR 3.3.5.1.5 Relay Pumps B,C > 4.5 seconds and

< 5.5 seconds Pumps A,D < 0.5 second (continued)

(a) Wien associated ~~2107 gpm Coolant Injection Pump subsystem SR 3.3.5.1.4 Discharge Flow -Low 4 (a), 5 (a) SR 3.3.5.1.5 (Bypass)~~

~~3. High Pressure Coolant Injection (HPCI) System~~
a. Reactor Vessel Water 1, 4 8 SR 3.3.5.1.1 > -42 inches Level -Low Low SR 3.3.5.1.2 (Level 2) 2 (d), 3 (d) SR 3.3.5.1.4 SR 3.3.5.1.5
b. Drywell 1, 4 B SR 3.3.5.1.2 < 1.84 psig Pressure -High 2(d) (d) SR 3.3.5.1.4 2(d, 3( SR 3.3.5.1.5
c. Reactor Vessel Water 1, 2 C SR 3.3.5.1.1 < 54 inches Level -High (Level 8) 2(d) SR 3.3.5.1.2 (d,3 (d) SR 3.3.5.1.4 SR 3.3.5.1.5
d. Emergency Condensate 1, 2 D SR 3.3.5.1.2 > 23 inches Storage Tank (ECST) d SR 3.3.5.1.3 Level -Low , 3(d) SR 3.3.5.1.5
e. Suppression Pool Water 1, 2 D SR 3.3.5.1.2 < 4 inches Level -High 2(d) d SR 3.3.5.1.4 zr,3C) SR 3.3.5.1.5 (continued)
(a) When the associatedrsubsystem(s) are required to be OPERABL (d) With reactor steam dome pressure > 150 psig.
Cooper 3.3-39 Amendment No. 4;4+
AC Sources - Shutdown
-- sTSF-3oo 3.8.2 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME B. One required DG B.1 Suspend CORE Immediately inoperable. ALTERATIONS.
AND B.2 Suspend movement of Immediately irradiated fuel assemblies in secondary containment.
AND B.3 Initiate action to Immediately suspend OPDRVs.
AND B.4 Initiate action to Immediately restore required DG to OPERABLE status.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.2.1 ----------- NOTE -------------------
J, The following SRs are not required to be performed: SR 3.8.1.3, and SR 3.8.1.9 through SR 3.8.1.11.
For AC sources required to be OPERABLE the In accordance SRs of Specification 3.8.1, except with applicable SR 3.8.1.8, are applicable. SRs
~~2. SR 3.8.1.11 is considered to be met without the ECCS initiation signals OPERABLE when the ECCS initiation signals are not required to be OPERABLE per Table 3.3.5.1-1.~~
Cooper 3.8 -12 Amendment No. JJ.6
rSTF - a Refueling Equipment Interlocks 3.9.1 3.9 REFUELING OPERATIONS 3.9.1 Refueling Equipment Interlocks LCO 3.9.] The refueling equipment interlocks associated with the reactor mode switch refuel position shall be OPERABLE.
APPLICABILITY: During in-vessel fuel movement with equipment associated with the interlocks when the reactor mode switch is in the refuel position.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.] Suspend in-vessel Immediately refueling equipment fuel movement with interlocks inoperable. equipment associated with the inoperable interlocks(s).
a L(.7.1 Iser-ot cer toI Imroc AeND corftoltoJ ros arm~y Ortseod Cooper 3.9- I Amendment No. A-e
NLS2005007 Page 1 of 8 ATTACHMENT 3 PROPOSED TECHNICAL SPECIFICATION REVISIONS (FINAL TYPED)
COOPER NUCLEAR STATION NRC DOCKET 50-298, LICENSE DPR-46 Technical Specification Pages 3.1-12 3.1-13 3.3-37 3.3-38 3.3-39 3.8-12 3.9-1
Control Rod Scram Times 3.1.4 3.1 REACTIVITY CONTROL SYSTEMS 3.1.4 Control Rod Scram Times LCO 3.1.4 a. No more than 10 OPERABLE control rods shall be "slow," in accordance with Table 3.1.4-1; and
~~b. No more than 2 OPERABLE control rods that are "slow" shall occupy adjacent locations.~~
APPLICABILITY: MODES 1 and 2.
ACTIONS CONDITION REQUIRED ACTION COMPLETION A. Requirements of the LCO A.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months not met.
SURVEILLANCE REQUIREMENTS

~~NOTE- ----------~~

During single control rod scram time Surveillances, the control rod drive (CRD) pumps shall be isolated from the associated scram accumulator.
SURVEILLANCE FREQUENCY SR 3.1.4.1 Verify each control rod scram time is within the limits I of Table 3.1.4-1 with reactor steam dome pressure Prior to exceeding
> 800 psig. 40% RTP after each reactor shutdown
> 120 days SR 3.1.4.2 Verify, for a representative sample, each tested 120 days control rod scram time is within the limits of cumulative Table 3.1.4-1 with reactor steam dome pressure operation in
> 800 psig. MODE 1 Cooper 3.1-12 Amendment No.
Control Rod Scram Times 3.1.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.3 Verify each affected control rod scram time is within Prior to declaring the limits of Table 3.1.4-1 with any reactor steam control rod dome pressure. OPERABLE after work on control rod or CRD System that could affect scram time SR 3.1.4.4 Verify each affected control rod scram time is within Prior to exceeding the limits of Table 3.1.4-1 with reactor steam dome 40% RTP after pressure > 800 psig. fuel movement within the affected core cell.
AND Prior to exceeding 40% RTP after work on control rod or CRD System that could affect scram time Cooper 3.1-13 Amendment No. -
ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 1 of 6)
Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES REQUIRED REFERENCED OR OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE
~~1. Core Spray System~~
a. Reactor Vessel Water 1,2,3. 4 (b) B SR 3.3.5.1.1 >-113 inches Level - Low Low Low SR 3.3.5.1.2 (Level 1) 4 (a), 5 (a) SR 3.3.5.1.4 SR 3.3.5.1.5
~~b. Drywell Pressure- High 1.2,3 4 (b) B SR 3.3.5.1.2 < 1.84 psig SR 3.3.5.1.4 SR 3.3.5.1.5~~
c. Reactor Pressure- Low 1,2.3 4 C SR 3.3.5.1.2 > 291 psig (Injection Permissive) SR 3.3.5.1.4 and SR 3.3.5.1.5 < 436 psig 4 (a), 5 (a) 4 B SR 3.3.5.1.2 > 291 psig SR 3.3.5.1.4 and SR 3.3.5.1.5 ' 436 psig
d. Core Spray Pump 1.2,3. I per pump E SR 3.3.5.1.2 > 1370 gpm Discharge Flow - Low SR 3.3.5.1.4 (Bypass) 4 (a), 5 (a) SR 3.3.5.1.5
e. Core Spray Pump 1,2,3, I per pump C SR 3.3.5.1.2 >9 seconds Start-Time Delay Relay SR 3.3.5.1.4 and 4 (a), 5 (a) SR 3.3.5.1.5 < 11 seconds
~~2. Low Pressure Coolant Injection (LPCI) System~~
a. Reactor Vessel Water 1,2,3, 4 B SR 3.3.5.1.1 >-113 inches Level - Low Low Low SR 3.3.5.1.2 (Level 1) 4 (a) 5 (a) SR 3.3.5.1.4 SR 3.3.5.1.5 (continued)
(a) When associated ECCS subsystem(s) are required to be OPERABLE per LCO 3.5.2, ECCS - Shutdown. I (b) Also required to initiate the associated diesel generator (DG).
Cooper 3.3-37 Amendment
ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 2 of 6)
Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES REQUIRED REFERENCED OR OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE
~~2. LPCI System (continued)~~
~~b. Drywell Pressure - High 1,2,3 4 B SR 3.3.5.1.2 <1.84 psig SR 3.3.5.1.4 SR 3.3.5.1.5~~
c. Reactor Pressure - Low 1,2,3 4 C SR 3.3.5.1.2 > 291 psig (Injection Permissive) SR 3.3.5.1.4 and SR 3.3.5.1.5 < 436 psig 4 (a), 5 (a) 4 B SR 3.3.5.1.2 > 291 psig SR 3.3.5.1.4 and SR 3.3.5.1.5 <436 psig
d. Reactor Pressure - Low 1 (c),2 (c), 4 C SR 3.3.5.1.2 > 199 psig (Recirculation Discharge SR 3.3.5.1.4 and Valve Permissive) 3 (c) SR 3.3.5.1.5 <221 psig
e. Reactor Vessel Shroud 1,2.3 2 B SR 3.3.5.1.1 >-193.19 Level - Level 0 SR 3.3.5.1.2 inches SR 3.3.5.1.4 SR 3.3.5.1.5
f. Low Pressure Coolant 1,2,3, 1 per pump C SR 3.3.5.1.2 Injection Pump Start - SR 3.3.5.1.4 Time Delay Relay 4 (a), 5 (a) SR 3.3.5.1.5 Pumps B.C
> 4.5 seconds and
< 5.5 seconds Pumps AD
< 0.5 second (continued)
(a) When associated ECCS subsystem(s) are required to be OPERABLE per LCO 3.5.2, ECCS - Shutdown. I (c) With associated recirculation pump discharge valve open.
Cooper 3.3-38 Amendment
ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 3 of 6)
Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE
~~2. LPCI System (continued)~~
g. Low Pressure 1.2.3, 1 per E SR 3.3.5.1.2 >2107gpm Coolant Injection Pump subsystem SR 3.3.5.1.4 Discharge Flow- Low 4 (a), 5 (a) SR 3.3.5.1.5 (Bypass)
~~3. High Pressure Coolant Injection (HPCI) System~~
a. ReactorVesselWater 1, 4 B SR 3.3.5.1.1 >-42inches Level - Low Low SR 3.3.5.1.2 (Level 2) 2 (d), 3 (d) SR 3.3.5.1.4 SR 3.3.5.1.5
~~b. Drywell Pressure - High 1, 4 B SR 3.3.5.1.2 < 1.84 psig SR 3.3.5.1.4 2 (d) 3 (d) SR 3.3.5.1.5~~
c. Reactor Vessel Water 1, 2 C SR 3.3.5.1.1 <54inches Level - High (Level 8) SR 3.3.5.1.2 2 (d) 3 (d) SR 3.3.5.1.4 SR 3.3.5.1.5
d. Emergency Condensate 1, 2 D SR 3.3.5.1.2 >23inches Storage Tank (ECST) SR 3.3.5.1.3 Level - Low 2 (d), 3 (d) SR 3.3.5.1.5
e. Suppression Pool Water 1, 2 D SR 3.3.5.1.2 <4 Inches Level - High SR 3.3.5.1.4 2 (d) 3 (d) SR 3.3.5.1.5 (continued)
(a) When the associated ECCS subsystem(s) are required to be OPERABLE per LCO 3.5.2, ECCS - Shutdown. I (d) With reactor steam dome pressure > 150 psig.
Cooper 3.3-39 Amendment
AC Sources - Shutdown 3.8.2 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME B. One required DG B.1 Suspend CORE Immediately inoperable. ALTERATIONS.
AND B.2 Suspend movement of Immediately irradiated fuel assemblies in secondary containment.
AND B.3 Initiate action to suspend OPDRVs.
Immediately AND B.4 Initiate action to restore required DG to OPERABLE status. Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.2.1
~~1. The following SRs are not required to be performed: SR 3.8.1.3, and SR 3.8.1.9 through SR 3.8.1.1 1.~~
~~2. SR 3.8.1.1 1 is considered to be met without the ECCS initiation signals OPERABLE when the ECCS initiation signals are not required to be OPERABLE per Table 3.3.5.1-1.~~
For AC sources required to be OPERABLE the In accordance SRs of Specification 3.8.1, except with applicable SR 3.8.1.8, are applicable. SRs Cooper 3.8-1 2 Amendment No. -
Refueling Equipment Interlocks 3.9.1 3.9 REFUELING OPERATIONS 3.9.1 Refueling Equipment Interlocks LCO 3.9.1 The refueling equipment interlocks associated with the reactor mode switch refuel position shall be OPERABLE.
APPLICABILITY: During in-vessel fuel movement with equipment associated with the interlocks when the reactor mode switch is in the refuel position.
ACTIONS CONDITION REQUIRED ACTION COMPLETION A. One or more required A.1 Suspend in-vessel fuel Immediately refueling equipment movement with equipment interlocks inoperable. associated with the inoperable interlocks(s).
OR A.2.1 Insert a control rod Immediately withdrawal block AND A.2.2 Verify all control rods Immediately are fully inserted Cooper 3.9-1 Amendment No._
NLS2005007 Page 1 of 13 ATTACHMENT 4 PROPOSED TECHNICAL SPECIFICATIONS BASES REVISIONS MARKUP FORMAT COOPER NUCLEAR STATION NRC DOCKET 50-298, LICENSE DPR-46 Technical Specification Bases Pages B3.1-25 B3.1-27 B3.3-96 B3.3-98 B3.3-100 B3.3-101 B3.3-104 B3.8-28 B3.8-31 B3.9-3 B3.9-4 Note: TS Bases pages are provided for information. Following approval of the proposed TS change, Bases changes will be implemented in accordance with TS 5.5.10, "Technical Specification (TS) Bases Control Program."
NLS2005007 Page 2 of 13 TS Bases INSERTS Insert 1 (TSTF-275)
Table 3.3.5.1-1 contains several footnotes. Footnote (a) clarifies that the associated functions are required to be OPERABLE in MODES 4 and 5 only when their supported ECCS are required to be OPERABLE per LCO 3.5.2, ECCS - Shutdown.
Insert 2 (TSTF-275)
Per Footnote (a) to Table 3.3.5.1 -1, this ECCS function is only required to be OPERABLE in MODES 4 and 5 whenever the associated ECCS is required to be OPERABLE per LCO 3.5.2.
Insert 3 (TSTF-275)
Automatic initiation of the required DG during shutdown conditions is specified in LCO 3.3.5.1, ECCS Instrumentation, and LCO 3.3.8.1, LOP Instrumentation.
Insert 4 (TSTF-225)
Alternatively, Required Actions A.2.1 and A.2.2 will permit continued fuel movement with the interlocks inoperable if a control rod withdrawal block is inserted, and all control rods are subsequently verified to be fully inserted. Required Action A.2.1 (rod block) ensures no control rods can be withdrawn. The withdrawal block utilized must ensure that if rod withdrawal is requested, the rod will not respond (i.e., it will remain inserted). Required Action A.2.2 is performed after placing the rod withdrawal block in effect, and provides a verification that all control rods are fully inserted. This verification that all control rods are fully inserted is in addition to the periodic verifications required by SR 3.9.3. 1.
Like Required Action A.1, Required Actions A.2.1 and A.2.2 ensure unacceptable operations are blocked (e.g., loading fuel into a cell with the control rod withdrawn).
One use for the A.2 Required Actions is to permit performance of SR 3.9.1.1 once, prior to fuel movement, without the need for subsequent performance if the fuel movement extends longer than the 7 day Frequency of the SR. This permits continued fuel movement under the protection of the continuous rod block inserted by the Required Actions.
Control Rod Scram Times B 3.1.4 (rSTF zz)
BASES (continued)
SURVEILLANCE The four SRs of this LCO are modified by a Note stating that REQUIREMENTS during a single control rod scram time surveillance, the CRD pumps shall be isolated from the associated scram accumulator. With the CRD pump isolated, (i.e., charging valve closed) the influence of the CRD pump head does not affect the single control rod scram times. During a full core scram, the CRD pump head would be seen by all control rods and would have a negligible effect on the scram insertion times.
SR 3.1.4.1 The scram reactivity used in DBA and transient analyses is based on an assumed control rod scram time. Measurement of the scram times with reactor steam dome pressure Ž 800 psig demonstrates acceptable scram times for the transients analyzed in References 3 and 4.
Maximum scram insertion times occur at a reactor steam dome pressure of approximately 800 psig because of the competing effects of reactor steam dome pressure and stored accumulator energy. Therefore, demonstration of adequate scram times at reactor steam dome pressure 2 800 psig ensures that the measured scram times will be within the specified limits at higher pressures. Limits are specified as a function of reactor pressure to account for the sensitivity of the scram insertion times with pressure and to allow a range of pressures over which scram time testing can be performed. To ensure that scram time testing is performed within a reasonable time following
~~at'clinj A=~~
a4t4er a shutdown duration of > 120 days, control rods are required to be tested before exceeding 40% RTP following the shutdown. This Frequency is acceptable considering the additional surveillances performed for control rod OPERABILITY, the frequent verification of adequate accumulator pressure, and the required testing of control rods affected byVwork on control rods or the CRD System.
t(oud w
, Qvepitf U-hf,4n The.
oC5Soctecdi dCssejICellc{ by (continued)
Cooper B 3.1-25 Rcvizicn 0
Control Rod Scram Times B 3.1.4 (Tr5TF - Z-2-2 BASES SURVEILLANCE SR 3.1.4.3 (continued)
REQUIREMENTS Specific examples of work that could affect the scram times are (but are not limited to) the following: removal of any CRD for maintenance or modification; replacement of a control rod; and maintenance or modification of a scram solenoid pilot valve, scram valve, accumulator, isolation valve or check valve in the piping required for scram.
The Frequency of once prior to declaring the affected control rod OPERABLE is acceptable because of the capability to test the control rod over a range of operating conditions and the more frequent surveillances on other aspects of control rod OPERABILITY.
SR 3.1.4.4 When work that could affect the scram insertion time is performed on a control rod or CRD System, or when fuel movement within the reactor pressure vessel occurs, testing must be done to demonstrate each affected control rod is still within the limits of Table 3.1.4-1 with the reactor steam dome pressure > 800 psig. Where work has been performed at high reactor pressure, the requirements of SR 3.1.4.3 and SR 3.1.4.4 can be satisfied with one test.
For a control rod affected by work performed while shut down, however, a zero pressure and high pressure test may be required. This testing ensures that, prior to withdrawing the control rod for continued operation, the control rod scram performance is acceptable for operating reactor pressure conditions. Alternatively, a control rod scram O~iyniti C- >
/ test during hydrostatic pressure testing could also satisfy reuoacto' e uMrQ both criteria. When -#4y- fuel movementAoccurs, .iubem only those control rods associated with the core cells affected by the fuel movement are required to be scram time tested.
The Frequency of once prior to exceeding 40% RTP is acceptable because of the capability to test the control rod over a range of operating conditions and the more frequent surveillances on other aspects of control rod OPERABILITY.
Duri9 3 rostv e re-ULeJ5 ovut o-l t is expec edl. taJ( covlrol roos will b oFfesAe, ._
(continued)
Cooper B 3.1-27 Cooper
-ik:ir.
B3.1-27 0
ECCS Instrumentation
-- B 3.3.5.1 BASES t7T -215 T BACKGROUND Diesel Generators (continued) in approximately 14 seconds, and will run in standby conditions (rated voltage and speed, with the DG output breaker open). The DGs will only energize their respective Engineered Safety Feature buses if a loss of offsite power occurs. (Refer to Bases for LCO 3.3.8.1.)
APPLICABLE The actions of the ECCS are explicitly assumed in the safety SAFETY ANALYSES, analyses of References 5, 6, and 7. The ECCS is initiated LCO, and to preserve the integrity of the fuel cladding by limiting APPLICABILITY the post LOCA peak cladding temperature to less than the 10 CFR 50.46 limits.
ECCS instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 4)e Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion.
The OPERABILITY of the ECCS instrumentation is dependent upon the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.5.1-1. Each Function must have a required number of OPERABLE channels, with their setpoints within the specified Allowable Values, where appropriate. The actual setpoint is calibrated coinsistentwlitb aDpplicable setpoint methodology assumptions.
0 b-e---3-~--5-f~-JAotnote (b), is added to show that certain ECCS instrumentation Functions are also .r-equir4ed-tobe-4GPEMBIE- to perform DG initiation.
Allowable Values are specified for each ECCS Function specified in the table. Nominal trip setpoints are specified in the setpoint calculations. The setpoint calculations are performed using methodology described in NEDC-31336P-A, "General Electric Instrument Setpoint Methodology," dated September 1996. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setpoint is not within its required Allowable Value. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual (continued)
Cooper B 3.3-96 Revision 0
ECCS Instrumentation
't B 3.3.5.1 J-5 T- Z-5
- BASES APPLICABLE l.a. 2.a. Reactor Vessel Water Level-Low Low Low (Level 1)
SAFETY ANALYSES, (continued)
LCO, and APPLICABILITY Function l.a signals. The Reactor Vessel Water Level -Low Low Low (Level 1) is one of the Functions assumed to be OPERABLE and capable of initiating the ECCS during the transients analyzed in References 5 and 7. In addition, the Reactor Vessel Water Level-Low Low Low (Level 1) Function is directly assumed in the analysis of the recirculation line break (Ref. 6). The core cooling function of the ECCS, along with the scram action of the Reactor Protection System (RPS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
Reactor Vessel Water Level-Low Low Low (Level 1) 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.
The Reactor Vessel Water Level-Low Low Low (Level 1)
Allowable Value is chosen to allow time for the low pressure core flooding systems to activate and provide adequate cooling.
Four channels of Reactor Vessel Water Level-Low Low Low (Level 1) Function are only required to be OPERABLE when the Ax rtCS2g ECCS cD r )are required to be OPERABLE to ensure that no l leln-strument failure can preclude ECCS(! b4 initia ion Refer to LCO 3.5.1 and LCO 3.5.2, ECCS-Shutdown," for Applicability Bases for the low pressure ECCS subsystems; LCO 3.8.1, "AC Sources-Operating"; and LCO 3.8.2, "AC Sources-Shutdown," for Applicability Bases for the DGs.
l.b. 2.b. Drvwell Pressure-High High pressure in the drywell could indicate a break in the reactor coolant pressure boundary (RCPB). The low pressure ECCS and associated DGs are initiated upon receipt of the Drywell Pressure-High Function in order to minimize the possibility of fuel damage. The DGs are initiated from Function I.b signals. The Drywell Pressure-High Function, along with the Reactor Water Level-Low Low Low (Level 1)
Function, is directly assumed in the analysis of the (continued)
Cooper B 3.3-98 Revision 0
ECCS. Instrumentation B 3.3.5.1
-T- SF -2q57 BASES APPLICABLE l.c. 2.c. Reactor Steam Dome Pressure-Low (Injection SAFETY ANALYSES, Permissivl) (continued)
LCO, and APPLICABILITY The Allowable Value is low enough to prevent overpressuring the equipment in the low pressure ECCS, but high enough to ensure that the ECCS injection prevents the fuel peak cladding temperature from exceeding the limits of 10 CFR 50.46.
Four channels of Reactor Pressure-Low Function are only required to be OPERABLE when the ECCS is required to be OPERABLE to ensure that no single instrument failure can t2 pe e ECCfSi Refer to LCO 3.5.1 and LCO 3.5.2
<T 1n for Applicability Bases for the low pressure ECCS subsystems.
1.d. 2.q. Core Spray and Low Pressure Coolant Injection Pump Discharge Flow-Low (Bypass)
The minimum flow instruments are provided to protect the associated low pressure ECCS pump from overheating when the pump is operating and the associated injection valve is not fully open. The minimum flow line valve is opened when low flow is sensed, and the valve is automatically closed when the flow rate is adequate to protect the pump. The LPCI and CS Pump Discharge Flow-Low Functions are assumed to be OPERABLE. The minimum flow valves for CS and LPCI are not required to close to ensure that the low pressure ECCS flows assumed during the transients and accidents analyzed in References 5, 6, and 7 are met. The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
One flow transmitter per CS pump and one differential pressure switch per LPCI subsystem are used to detect the associated subsystems' flow rates. The logic is arranged such that each switch or transmitter causes its associated minimum flow valve to open. The logic will close the minimum flow valve once the closure setpoint is exceeded.
The LPCI minimum flow valves are time delayed such that the valves will not open for approximately 3.5 seconds after the switches detect low flow. The time delay is provided to limit reactor vessel inventory loss during the startup of the RHR shutdown cooling mode. The Pump Discharge Flow-Low Allowable Values are high enough to ensure that the pump (continued)
Cooper B 3.3-100 Revision 0
ECCS Instrumentation B 3.3.5.1 BASES "STF S APPLICABLE SAFETY ANALYSES, LCO, and APLICABILITY 1.d. 2.g. Core Spray and Low Pressure Coolant IniectionPump Discharge Flow-Low (Bvpass) flow rate is sufficient to protect the pump.
Each channel of Pump Discharge Flow- Low Function (two CS channels and four LPCI channels) is only required to be OPERABLE-when the associated ECCS is required to be OPERABLE to en, e that no single instrument failure can preclude the ECCS function ARefer to LCO 3.5.1 and LCO 3.5.2 for Applicability Bases for the low pressure ECCS subsystems.
1.e. Core Spray Pump Start-Time Delay Relay The purpose of this time delay is to delay the start of the CS pumps to enable sequential loading of the appropriate AC source. This Function is necessary when power is being supplied from the offsite sources or the standby power sources (DG). The CS Pump Start-Time Delay Relays are assumed to be OPERABLE in the accident analyses requiring ECCS initiation. That is, the analyses assume that the pumps will initiate when required and excess loading will not cause failure of the power sources.
There are two Core Spray Pump Start-Time Delay Relays, one for each CS pump. Each time delay relay is dedicated to a single pump start logic, such that a single failure of a Core Spray Pump Start-Time Delay Relay will not result in the failure of more than one CS pump. In this condition, one of the two CS pumps will remain OPERABLE; thus, the single failure criterion is met (i.e., loss of one instrument does not preclude ECCS initiation). The Allowable Value for the Core Spray Pump Start-Time Delay Relays is chosen to be long enough so that the power source will not be overloaded and short enough so that ECCS operation is not degraded.
Each channel of Core Spray Pump Start-Time Delay Relay Function is required to be OPERABLE only when the associated subsystem is required to be OPERABLE. Refer to LCO 3.5.1 and LCO\3.5.2 for Applicability Bases for the CS subsystems.
Cooper B 3.3-1 01 Cooper~~f B1 3.-11om/Q/4
ECCS Instrumentation B 3.3.5.1 T F- I 75 BASES APPLICABLE 2.f. Low Pressure Coolant Injection Pump Start-TimeDelay Relay SAFETY ANALYSES, (continued)
LCO, and APPLICABILITY There are four LPCI Pump Start -Time Delay Relays, one in each of the RHR pump start logic circuits. While each time delay relay is dedicated to a single pump start logic, a single failure of a LPCI Pump Start - Time Delay Relay could result in the failure of the two low pressure ECCS pumps, powered for the same ESF bus, to perform their intended function (e.g., as in the case where both ECCS pumps on one ESF bus start simultaneously due to an inoperable time delay relay). This still leaves four of the six low pressure ECCS pumps OPERABLE; thus, the single failure criterion is met (i.e., loss of one instrument does not preclude ECCS initiation). The Allowable Value for the LPCI Pump Start - Time Delay Relays is chosen to be long enough so that most of the starting transient of the first pump is complete before starting the second pump on the same 4.16 kV emergency bus and short enough so that ECCS operation is not degraded.
Each LPCI Pump Start - Time Delay Relay Function is required to be OPERABLE only when the associated LPCI subsystem is required to be y\OPERABLS.Refer to LCO 3.5.1 and LCO 3.5.2 for Applicability Bases for the LPCI subsystems.
High Pressure Coolant Injection (HPCI) System 3.a. Reactor Vessel Water Level-Low Low (Level 2)
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, the HPCI System is initiated at Level 2 to maintain level above fuel zone zero. The Reactor Vessel Water Level - Low Low I (Level 2) is one of the Functions assumed to be OPERABLE and capable of initiating HPCI during the transients analyzed in References 5 and 7.
Additionally, the Reactor Vessel Water Level - Low Low (Level 2)
Function associated with HPCI is directly assumed in the analysis of the recirculation line break (Ref. 6). The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
(continued)
Cooper B 3.3-1 04 June 10,1999 l
AC Sources - Shutdown B 3.8.2 BASES 5 TF - Za7 5 LCO ensures that a diverse power source is available for (continued) providing electrical power support assuming a loss of the offsite circuit. Together, OPERABILITY of the required offsite circuit and DG ensures the availability of sufficient AC sources to operate the plant in a safe manner and to mitigate the consequences of postulated events during shutdown (e.g., fuel handling accidents and reactor vessel draindown). <Jvisert3>
The qualified offsite circuit must be capable of maintaining rated frequency and voltage while connected to its respective critical bus, and of accepting required loads during an accident. Qualified offsite circuits are those that are described in the USAR and are part of the licensing basis for the unit. The offsite circuit consists of incoming breaker and disconnect to the startup or emergency station service transformer, associated startup or emergency station service transformer, and the respective circuit path including feeder breakers to all 4.16 kV critical buses required by LCO 3.8.8.
The required DG must be capable of starting, accelerating to rated speed and voltage, connecting to its respective critical bus on detection of bus undervoltage, and accepting required loads. This sequence must be accomplished within 14 seconds. Each DG must also be capable of accepting required loads within the assumed loading sequence intervals, and must continue to operate until offsite power can be restored to the critical buses. These capabilities are required to be met from a variety of initial conditions such as DG in standby with engine hot and DG in standby with engine at ambient conditions.
Proper sequencing of loads, including tripping of nonessential loads, is a required function for DG OPERABILITY. The necessary portions of the Service Water System and Ultimate Heat Sink are also required to provide appropriate cooling to the required DGs.
It is acceptable during shutdown conditions, for a single offsite power circuit to supply both 4.16 kV critical buses.
No fast transfer capability is required for offsite circuits to be considered OPERABLE.
(continued)
Cooper B 3.8-28 Revision 0
AC Sources - Shutdown
.. B 3.8.2 T:)TF -30o BASES ACTIONS A.2.1, A.2.2, A.2.3, A.2.4, B.1, B.2, B.3 and B.4 (continued) to any required 4.16 kV critical bus, ACTIONS for LCO 3.8.8 must be immediately entered. This Note allows Condition A to provide requirements for the loss of the offsite circuit whether or not a division is de-energized. LCO 3.8.8 provides the appropriate restrictions for the situation involving a de-energized division.
SURVEILLANCE SR 3.8.2.1 REQUIREMENTS SR 3.8.2.1 requires the SRs from LCO 3.8.1 that are necessary for ensuring the OPERABILITY of the AC sources in other than MODES 1, 2, and 3. SR 3.8.1.8 is not required to be met since only one offsite circuit is required to be OPERABLE. Refer to the corresponding Bases for LCO 3.8.1 for a discussion of each SR.
This SR is modified byX Notes The reason for 1th Note is to preclude requiring the OPERABLE DG(s) from being paralleled with the offsite power network or otherwise rendered inoperable during the performance of SRs, and to preclude deenergizing a required 4.16 kV critical bus or disconnecting a required offsite circuit during performance of SRs. With limited AC sources available, a single event could compromise both the required circuit and the DG. It is the intent that these SRs must still be capable of being met, but actual performance is not required during periods when the DG and offsite circuit is required to be OPERABLE.
REFERENCES 1. 10 CFR 50.36(c)(2)(ii).
Note 2 states SR 3.8.1.11is considered to be met without the ECCS initiation signals OPERABLE when the ECCS initiation signals are not required to be OPERABLE per Table 3.3.5.1-1. This SR demonstrates the DG response to an ECCS signal in conjunction with a loss of power signal. When ECCS system(s) are not required to be OPERABLE per LCO 3.5.2, "ECCS - Shutdown," the DG is not required to start in response to ECCS initiation signals. This is consistent with the ECCS instrumentation requirements.
However, the DG is still required to meet the other attributes of SR 3.8.1.11 when ECCS initiation signals are not required to be OPERABLE per Table 3.3.5.1-1.
Cooper B 3.8-31 Revision 0
Refueling Equipment Interlocks B 3.9.1 TST F - Za5 BASES (continued)
LCO To prevent criticality during refueling, the refueling interlocks associated with the reactor mode switch refuel position ensure that fuel assemblies are not loaded into the core with any control rod withdrawn.
To prevent these conditions from developing, the all-rods-in, the refueling platform position, the refueling platform fuel grapple fuel loaded, the refueling platform frame mounted hoist fuel loaded, the refueling platform monorail mounted hoist fuel loaded, the refueling platform fuel grapple not full up position, and the service platform hoist fuel loaded inputs are required to be OPERABLE. These inputs are combined in logic circuits, which provide refueling equipment or control rod blocks to prevent operations that could result in criticality during refueling operations.
APPLICABILITY In MODE 5, a prompt reactivity excursion could cause fuel damage and subsequent release of radioactive material to the environment. The refueling equipment interlocks protect against prompt reactivity excursions during MODE 5. The interlocks are required to be OPERABLE during in-vessel fuel movement with refueling equipment associated with the interlocks when the reactor mode switch is in the refuel position. The interlocks are not required when the reactor mode switch is in the shutdown position because a control rod block (LCO 3.3.2.1, "Control Rod Block Instrumentation")
ensures control rod withdrawal cannot occur simultaneously with in-vessel fuel movements.
In MODES 1, 2, 3, and 4, the reactor pressure vessel head is on, and fuel loading activities are not possible.
Therefore, the refueling interlocks are not required to be OPERABLE in these MODES.
KA Baj 6.14r Or*1 ACTIONS A.l ,A-Z, 2 (ih( ro st.fL With one or more of the required re ueling equipment interlocks inoperable, the unit must be aced in a condition in which the LCO does not apply^. %n-vessel fuel movement with the affected refueling equipment must be Cooper B 3.9-3 Revision 0
Refueling Equipment Interlocks B 3.9.1
/ }sTF-2 22&
BASES ACTIONS A.1 (continued) immediately suspended. This action ensures that operations are not performed with equipment that would potentially not be blocked from unacceptable operations (e.g., loading fuel into a cell with a control rod withdrawn). Suspension of in-vessel fuel movement shall not preclude completion of movement of a component to a safe position.
< Inser- +>
SURVEILLANCE REQUIREMENTS SR 3.9.1.1 Performance of a CHANNEL FUNCTIONAL TEST demonstrates each required refueling equipment interlock will function properly when a simulated or actual signal indicative of a required condition is injected into the logic. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping, or total channel steps so that the entire channel is tested.
The 7 day Frequency is based on engineering judgment and is considered adequate in view of other indications of refueling interlocks and their associated input status that are available to unit operations personnel.
REFERENCES 1. USAR, Appendix F, Section F-2.5. I
~~2. USAR, Section VII-6.~~
~~3. USAR, Section XIV-5.3.3.~~
~~4. USAR, Section XIV-5.3.4.~~
~~5. 10 CFR 50.36(c)(2)(ii).~~
Cooper B 3.9-4 12/18/03
I ATTACHMENT 3 LIST OF REGULATORY COMMITMENTS© I Correspondence Number: NLS2005007 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 & Regulatory Affairs Manager at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.
COMMITTED DATE COMMITMENT OR OUTAGE None 1
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4-PROCEDURE 0.42 l REVISION 15 l PAGE 18 OF 24

NLS2005007, License Amendment Request to Adopt Generic Changes to Standard Technical Specifications

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