ML042640026

License Amendment Request, Technical Specification Changes for Single Recirculation Loop Operation

ML042640026
Person / Time
Site:	Pilgrim
Issue date:	09/02/2004
From:	Balduzzi M Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
2.04.074
Download: ML042640026 (32)

Text

SEntergy Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360 September 2, 2004 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001 Michael A. Balduzzi Site Vice President

SUBJECT:

Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station Docket No. 50-283 License Number. DPR-35 License Amendment Request Technical Specification Changes for Single Recirculation Loop Operation LETTER NUMBER:

2.04.074

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Nuclear Operations, Inc. (Entergy) requests approval of the enclosed changes to the Pilgrim Nuclear Station (PNPS) Operating License and Technical Specifications to allow continued plant operation with a single recirculation loop in service.

The proposed changes have been evaluated in accordance with TO CFR 50.91 (a)(1) using criteria in 10 CFR 50.92(c) and it has been determined that these changes involve no significant hazard considerations. The bases for these determinations are provided in the enclosure.

ENO requests NRC approval by May 10, 2005., GE Report GE-NE-0000-0027-5301, is proprietary and an affidavit signed by an authorized representative of GE is provided pursuant to 10 CFR 2.790. It is requested that this proprietary information be withheld from public disclosure. Attachment 3 provides a non-proprietary version of the report.

Commitments made by the Licensee in this letter are listed in Attachment 4.

If you have any questions or require additional information, please contact Bryan Ford at (508) 830-8403.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 001A day of s

, 2004.

Sincerely, Michael A. Balduzzi Site Vice President

Enclosure:

Attachments:

Evaluation of Proposed Changes, (14 pages)

1. License and Technical Specification Mark-ups with Bases (16 pages)

2. General Electric Report GE-NE-0000-0027-5301 (proprietary)

3. General Electric Report GE-NE-0000-0027-5301 (non-proprietary)

4. List of Regulatory Commitments

$,01 204074

Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station cc:

Mr. Lee Licata, Project Manager Office of Nuclear Reactor Regulation Mail Stop: 0-8B-1 U.S. Nuclear Regulatory Commission 1 White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. Robert Walker, Director Massachusetts Department of Public Health Radiation Control Program 90 Washington Street, 2nd Floor Dorchester, MA 02121 Letter no. 2.04.074 Page 2 Ms. Cristine McCombs, Director Mass. Emergency Management Agency 400 Worcester Road Framingham, MA 01702 Senior Resident Inspector Pilgrim Nuclear Power Station U.S. Nuclear Regulatory Commission Region 1 475 Allendale Road King of Prussia, PA 19408

ENCLOSURE to Letter 2.04.074 Evaluation of Proposed Changes

Subject:

License Amendment to allow Single Loop Operation

1.

DESCRIPTION

2.

PROPOSED CHANGES

3.

BACKGROUND

4.

TECHNICAL ANALYSIS

5.

REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Evaluation 5.2 Environmental Consideration

6.

COORDINATION WITH OTHER PENDING TS CHANGES

7.

PRECEDENTS

8.

REFERENCES

Letter 2.04.074 Enclosure Application for License Amendment to Allow Single Loop Operation

1. Description This letter is a request to amend Operating License DPR-35 and applicable Technical Specifications (TS) for Pilgrim Nuclear Power Station (PNPS). The proposed license amendment requests changes to include provisions for continued reactor operation in single recirculation loop operation (SLO) mode under certain specified conditions. The current PNPS Operating License (OL) allows for only 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of SLO.

The proposed changes are based on NUREG 1433, Rev. 3 (Reference 1), Standard Technical Specifications (STS) for General Electric Boiling Water Reactors (BWR/4),

which includes provisions for SLO. provides a mark-up of the proposed changes to the existing OL and associated TS. Mark-ups of the associated technical specification bases are also enclosed for the purpose of adding clarity and completeness to the submittal. and 3 are proprietary and non-proprietary copies of General Electric Report GE-NE-0000-0027-5301, "Pilgrim Nuclear Power Station Single Loop Operation." This report identifies the operational restrictions that must be implemented in order to justify single loop operation without undue risk to public health and safety. Attachment 4 is a list of the regulatory commitments applicable to this license amendment request.

SLO is a recognized operating practice for Boiling Water Reactor (BWR) operation and many BWRs have TS that allow indefinite operation with a single recirculation pump in service. The ability to operate in single-loop mode provides operational flexibility and allows continued power operation in the event of an inoperable recirculation loop.

2. Proposed Changes The following changes are proposed as part of the OL and TS amendment.

A. License Condition 3.E.

This license condition requires that the reactor not be operated for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if one recirculation loop is out of service. This change deletes the license condition.

B. Safety Limit for MINIMUM CRITICAL POWER RATIO (SLMCPR)

TS 2.1.2 identifies that the SLMCPR for two-loop operation (TLO) is 2 1.06.

Based on increased uncertainties in total core flow and Traversing In-core Probe (TIP) readings, the TLO Safety Limit MCPR must be increased by 0.02 for SLO.

TS 2.1.2 is revised to identify a SLMCPR limit of 21.08 for SLO.

C. Jet Pumps Existing TS 3/4.6.E - Jet Pumps, established the LCO and surveillance requirements that ensure each jet pump remains operable whenever the reactor is in the Startup or Run Mode. Based on removal of license condition 3E, existing surveillance requirement TS 4.6.E.1 cannot be performed when the plant is operating in SLO. The proposed surveillance revision provides for three 2

Letter 2.04.074 Enclosure F. Table of Contents The Table of Contents is updated to reflect the title revision to TS 3/4.6.F and to remove reference to two previously deleted TS sections, TS 3/4.6.H and TS 3/4.6.1. Reference to TS 3/4.6.H was also deleted from page 3/4.6.8.

3.

Background

The reactor coolant recirculation system provides forced coolant flow through the reactor core and, in combination with control rods, provides a means to control and change reactor power over a broad range. The recirculation system consists of two recirculation pump loops and drive units, each with a separate variable speed motor generator (MG) set, recirculation pump, and piping loop. The individual recirculation pumps are located in the drywell and provide drive flow to two separate jet pump headers inside the reactor vessel, which in turn provide core recirculation flow. For a detailed discussion of the system components and operating characteristics, refer to the description in Chapter 4.3 in the Updated Final Safety Analysis Report (UFSAR).

During normal power operation both recirculation pumps are operated at near-matched speeds to provide forced recirculation flow. Recirculation pump speed and flow can be changed using the variable speed recirculation system MG set and, thus, be used to change core power. Chapter 4.3 of the FSAR states that operation with a single recirculation loop is possible at reduced power. Power generation with a single recirculation loop in service is also a recognized mode of operation for Boiling Water Reactors (BWRs) and many BWRs have TS that allow for SLO.

The most obvious benefit of adding an allowance for SLO is the ability to continue power operation in the event of the loss of a recirculation loop due to component malfunction. Several active components in the recirculation system are located in the reactor building and are readily accessible during power operation. These include the MG set drive motors, fluid couplers, generators, and associated oil coolers. Also accessible are the recirculation system controllers, logic relaying, and system electrical panels and breakers. Typically, most of these components can be repaired with the reactor in service with no impact on power operation other than the unavailability of the affected loop itself.

While the reactor recirculation system is a very reliable system, temporary unavailability of a recirculating loop is occasionally experienced. Currently, Operating License, Section 3.E restricts reactor operation with one recirculation loop out of service to a period of less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. As might be expected, it is not always possible to diagnose and effect repairs to the system within these time frames. PNPS most recently had a recirculation system TS required shutdown in February 2002 due to a MG set generator field wiring failure because the repair could not be diagnosed, planned, and implemented within the LCO time frame currently allowed.

Many BWRs (Hatch, Peach Bottom, Vermont Yankee, Fitzpatrick, Cooper, Duane Arnold, and Grand Gulf) have previously provided justification for SLO and have obtained TS changes allowing operation in single-loop. Thus, there is ample industry experience demonstrating the viability of operating in single-loop and precedent in analyzing the safety considerations of SLO.

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Letter 2.04.074 Enclosure The NRC has previously determined SLO is generically acceptable as set forth in Generic Letter 86-09, "Technical Resolution of Generic Issue B-59 (N-1) Loop Operation in BWRs and PWRs," (Reference 3). SLO is also recognized as a standard mode of operation addressed in NUREG-1433, BWR/4 ISTS.

In this submittal, Entergy is providing the technical analyses that form the basis for changing the license and associated technical specifications to allow indefinite SLO operation. This license change and associated TS changes would allow continued operation in the event of loss of a single recirculation loop. As noted above, it is expected that, in most cases, a malfunctioning loop could be repaired with the reactor in service, and normal operations resumed. Thus, for these circumstances, approval of these TS changes would have a positive economic impact in allowing continued power generation in SLO. Similarly, the avoidance of shutdowns due to recirculation system failures would eliminate risk factors associated with unnecessary reactor shutdowns and restarts.

4.

Technical Analysis With two recirculation loops in service, the reactor recirculation system, in combination with the control rods, is designed to support the full range of power operations. Power operation with a single recirculation loop in service is also a viable mode of operation although with a single loop in service, full core flow and, hence, full core power cannot be achieved.

Reactor control and operation in single-loop is very similar to two-loop recirculation operation. The primary difference is that the maximum achievable power level is reduced because of the reduction in total core flow. Drive flow on the operating pump is increased and part of the total flow from the active jet pump loop will backflow through the inactive jet pumps. This affects the normal relationship between drive flow and core flow as compared to two-loop operation.

The primary analysis of the safety considerations in support of PNPS proposed SLO TS is presented in Attachment 2, General Electric Report GE-NE-0000-0027-5301, "Pilgrim Nuclear Power Station Single Loop Operation," July 2004. This report was prepared for the purpose of evaluating the effects of SLO on the plant transient and accident analyses. The basic conclusion of this report is that the plant can be operated in SLO mode safely for an unrestricted period of time provided that operation is controlled in accordance with the identified safety settings established for SLO operation.

Although it is not a currently permitted mode of operation, SLO has been included as an operating flexibility option in the cycle-specific core reload analyses that are performed for each fuel cycle. These analyses are performed in accordance with the latest approved version of GE Licensing Topical Report NEDE-2401 1-P-A,

'General Electric Standard Application for Reactor Fuel" (Reference 4).

Safety Limit Minimum Critical Power Ratio (SLMCPR)

The SLMCPR is established such that no fuel damage is calculated to occur if the limit is not violated during reactor transients. The limit is calculated using a statistical model that includes considerations for uncertainties. For SLO as discussed in GE-NE-0000-0027-5301, the SLMCPR increases by 0.02 due to increased uncertainties in the total core flow and Traversing In-Core Probe (TIP) readings compared to two-loop operation. Except for these readings, the 5

Letter 2.04.074 Enclosure uncertainties used in the statistical analysis for SLMCPR are not dependent on whether core flow is provided by one or two recirculation pumps.

The net effect of the revised core flow and TIP uncertainties is an increase in SLMCPR of 0.02 for SLO operation. This is the value included in the proposed change to TS 2.1.2. The SLO SLMCPR limit will be revised if necessary based on the evaluations performed for each subsequent core reload.

The plant process computer will be updated with revised input data to allow for calculation of MCPR and fuel specific MCPR operating limits (OLMCPR) applicable to SLO. Existing TS 3.11.C provides the requirements for monitoring MCPR and relies on the Core Operating Limits Report (COLR) to define the OLMCPR limits. By maintaining MCPR 2 OLMCPR, the SLMCPR specified in TS 2.1.2 will not be challenged in the event of the most limiting transient.

Core Operating Transients Core wide transients potentially impacted by single loop operation were evaluated in Section 4 to GE-NE-0000-0027-5301. The evaluation concludes that the consequences of the abnormal operating transients (pressurization, flow increase, cold water injection events) were less severe than the same events analyzed for two-loop mode, and thus bounded by the two-loop UFSAR analyses.

The cycle-specific reload analyses performed by Global Nuclear Fuels Inc. (GNF) considers SLO an operating flexibility option that is evaluated in accordance with the methods described in NEDE-2401 1-P-A, "General Electric Standard Application for Reactor Fuel." The OLMCPR with an SLO MCPR adder, as specified in the COLR, was evaluated and provides adequate protection for transients initiated during SLO.

The Rod Withdrawal Error (RWE) evaluations are independent of the source of core flow (i.e. one recirculation loop or two) and consequently, these evaluations are valid for both two-loop operation and single loop operation. RWE is evaluated each fuel cycle as part of the cycle-specific reload analysis. Cycle specific flow biased APRM rod block and scram setpoints are calculated to define the scram trip and rod block limits. The cycle specific APRM scram trip and rod block limits are defined on the power flow map provided in the COLR. For single loop operation, the flow-biased APRM scram trip and rod block set points must be adjusted to account for the change in the relationship between drive flow and core flow due to reverse flow in the inactive loop jet pumps and lower core resistance. GE-NE-0000-0027-5301, Section 4.1.6 identifies the equations used to correct the flow biased APRM scram trip and rod block set points for long-term SLO operation.

PNPS is currently in operating Cycle 15. The Operating Cycle-16 COLR will identify the revised TLO and SLO APRM setpoints and existing TS 3.11.D ensures that the plant will be operated in accordance with the power-flow map, which will be updated to reflect APRM setpoint revisions. In the plant, flow control trip reference (FCTR) cards will be updated to enforce TLO and SLO APRM scram trip and rod block setpoint limits. After loss of a recirculation loop, operator action will be required to manipulate a FCTR card toggle switch within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to adjust for SLO operation. The time period to make the SLO adjustments is consistent with standard TS 3.4.1.

6

Letter 2.04.074 Enclosure Accident Evaluation An evaluation of the effects of SLO on applicable accident analyses is provided in GE-NE-0000-0027-5301. The LOCA and Recirculation Pump Seizure events were evaluated since the remaining Chapter 14 UFSAR accidents are not significantly affected by recirculation mode or are bounded by two-loop accident analysis.

The Recirculation Pump Seizure accident is evaluated in GE-NE-0000-0027-5301, Section 4.1.5 and is characterized by a near instantaneous stoppage of the pump and the associated pump flow. With sudden stagnation of the drive flow, the active loop flow rapidly decreases, and the resultant core flow decreases causing the core void fraction to increase which in turn causes a rapid decrease in core power.

The purpose of the analysis is to ensure that the radiological consequences of the event are acceptable. The analyses referenced in the GE Report indicate that the MCPR will remain greater than the SLMCPR, and that the event will terminate with the reactor continuing to operate in natural circulation. A cycle independent OLMCPR was calculated for a recirculation pump seizure event when operating in SLO. This value was verified for Cycle 15 and will be reevaluated for each subsequent core reload in the cycle specific Supplemental Reload Licensing Report. Based on the analyses performed, recirculation pump seizure is not a limiting event and will not result in fuel entering boiling transition. Consequently, radiological release is avoided and there is no challenge to 10 CFR 100 radiation release limits.

A summary of Emergency Core Cooling System (ECCS) performance during a Loss of Coolant Accident (LOCA) when operating in SLO is provided in Section 6 of GE-NE-0000-0027-5301. This report references NEDC-31852 Rev 2, which describes the SAFER/GESTR-LOCA analysis performed for Pilgrim Station. The DBA-LOCA analysis that was performed for PNPS relied on SAFER/GESTR codes and methodology for performing LOCA analyses. With breaks smaller than the DBA, there is a longer period of nucleate and/or film boiling prior to fuel uncovery to remove the fuel's stored energy. This analysis report provides the results of LOCA analysis for SLO and concluded that the design basis accident (large breaks) are more severe than small break sequences and, therefore, the large break results are bounding for SLO. The analysis resulted in an SLO PLHGR/MAPLHGR multiplier of 0.8 for both GE-1l and GE-14 fuel. The calculated Appendix K PCT's using these PLHGR/MAPLHGR multipliers are below the 10 CFR 50.46 limit of 22000F. Therefore, the calculated SLO PLHGR/MAPLHGR multipliers are conservative and assure that the SLO results satisfy the acceptance criteria of 10 CFR 50.46 and NRC Safety Evaluation Report (SER) requirements for the SAFER/GESTR application methodology.

Following approval of the proposed SLO TS, the PLHGR/MAPLHGR SLO multiplier will be documented in the COLR report and the plant process computer will be programmed to apply the SLO PLHGR/MAPLHGR multiplier when calculating core thermal limits when in single loop operating mode to support the analysis assumptions.

Stability Requirements The primary contributing factors to the stability performance with one recirculation loop not in service are the power-flow ratio and the recirculation loop characteristics. For low core flows occurring at minimum pump speed, the jet pumps for both recirculation loops will exhibit forward flow. At higher pump speeds 7

Letter 2.04.074 Enclosure the core flow is increased in SLO and the inactive jet pump forward flow decreases with increasing core flow. The reduced flow in the inactive loop reduces the resistance that the recirculation loops impose on the reactor water flow perturbations, thereby adding a destabilizing effect. At the same time, the increased flow results in a lower power-flow ratio, which has a stabilizing effect.

These countering effects result in a slightly decreased stability margin (higher decay ratio) initially as core flow is increased (from minimum) in SLO, and then an increase in stability margin (lower decay ratio) as the core flow is increased further and reverse flow in the inactive loop is established.

As core flow is increased beyond 40% of rated during SLO, substantial reverse flow is established in the inactive loop. A cross flow is established in the annular downcomer region near the jet pump suction entrance caused by the reverse flow of the inactive recirculation loop. At higher flow, with substantial reverse flow in the inactive recirculation loop, the effect of cross flow results in an increase in system noise which increases the total core flow noise which tends to increase observed neutron flux noise.

GE has evaluated the SLO effects on stability, including increased noise, and determined that stability characteristics are not significantly different from TLO conditions. At low core flow, SLO may be slightly less stable than TLO, but as core flow increased and reverse flow is established, the stability performance of TLO and SLO is similar.

Because of generic stability concerns and the experience of some BWR reactors regarding the potential for power oscillations at low flow/high power operating map conditions, the NRC issued Bulletin No. 88-07 which was supplemented in December 1988.Bulletin 88-07, Supplement 1 (Reference 5), requested Licensees to adopt the Boiling Water Reactor Owners Group (BWROG) stability monitoring guidelines similar to those originally issued by GE in Services Information Letter (SIL) 380 Revision 1 (Reference 6). The NRC later requested licensees to address long-term solutions for thermal-hydraulic instabilities in Generic Letter (GL) 94-02 dated July 11, 1994 (Reference 7). In the GL, the NRC requested utilities to update operating procedures and provide operator training to address detection and suppression of power oscillations based on more recent operating experiences and revised BWROG Guidelines.

In response to the above referenced NRC operating guidance, the BWROG developed licensing strategies and methodology to provide a long-term resolution for thermal-hydraulic instability. Initially, NEDO-31960 and NEDO-31960, Supplement 1 were submitted to and approved by the NRC in 1993 (Reference 8).

The BWROG subsequently developed specific reports to address selected options for addressing thermal-hydraulic stability. NEDO-32339-A (Reference 9) addresses Stability Option ElA, and NEDO-32465-A (Reference 10) addresses Stability Option 1 D. The NRC has specifically reviewed these reports and has approved the referenced stability monitoring and protection methodologies.

PNPS currently relies on administrative controls and plant hardware to enforce the protection methods described in Stability Option ElA. However, PNPS is pursuing a plant modification to adopt the protection features defined in Stability Option 1 D.

Since thermal-hydraulic stability protection is not predicated on the number of recirculation loops in service, either Stability Option ElA or 1 D will provide 8

Letter 2.04.074 Enclosure adequate core protection from thermal-hydraulic instabilities when operating in the SLO operating mode.

Reactor Internal Vibration In single-loop recirculation mode, increases in APRM noise and core delta p fluctuations have been observed in some plants while operating at high drive flows which may be associated with increased vibration of the active jet pumps. The impact of jet pump vibration during SLO was evaluated to ensure reactor internal vibration levels are maintained at acceptable levels.

GE Report GE-NE-0000-0027-5301, Section 7, identifies that SLO operation will be restricted to operation below the 100% Current Licensed Thermal Power (CLTP) rod line with a maximum power limit of 65% of rated power (2028 Mwt) and a maximum core flow limit of 52% of rated flow (69 Mlb/hr) to ensure vibration levels are maintained at acceptable levels. These SLO operating limits will be identified in the COLR. TS 3.11.D addresses the power-flow map and ensures that the limiting values identified in the COLR are not exceeded. The SLO pump operating restrictions defined in the COLR will be reviewed for adequacy for each subsequent reload cycle. SLO pump operating limits will be incorporated into operating procedures after approval of the amendment request.

In addition to the GE evaluation, a separate evaluation of the effects of SLO on the structural integrity of the reactor vessel and associated internal components was performed and documented in Structural Integrity Associates (SIA) Inc., Report SIR-04-026 Rev. 2 - "Evaluation of Single Loop Operation at Pilgrim Nuclear Power Station," June 2004. The SIA report addresses structural integrity via the evaluation of TLO and SLO vibration data. This data was obtained from the plant start-up program, thirty (30) years of successful operation at 100% power, operating data obtained with the plant operating in SLO mode, and detailed jet pump inspections performed during recent Refueling Outages (RFOs). The report concludes that SLO operation is acceptable from a structural analysis standpoint since the vibration levels observed in the core for a single operating recirculation pump are enveloped by the vibration levels observed for TLO at 100% power.

The SIA Report also identifies that the jet pumps are typically the component of most concern with respect to vibration because of high internal flow and RPV annulus downcomer flow rates, which can lead to significant vibratory response.

The report documents that jet pump inspection history since RFO-1 1 (February 1996), has been extensive and has not detected the presence of any cracking in the jet pump welds. Detailed evaluation of the jet pump swing gate gaps was also performed and found to be structurally acceptable based on vibration frequency and harmonic analyses, which revealed that the jet pumps are nominally stressed to acceptable levels.

Recirculation Loop Operating Restrictions TS 3/4.6.F - Jet Pump Flow Mismatch and TS 3.6.A.6 - Thermal-Hydraulic Stability contain the LCO conditions applicable to controlling recirculation loops during power operation. Review of the NRC approved STS revealed that STS 3.4.1 - Recirculation Loops Operation, identifies the necessary LCO requirements for controlling TLO, SLO and no recirculation loop operation. Based on the removal of License Condition 3.E, TS 3/4.6.F was modified to incorporate the STS recirculation loop operating requirements identified in STS 3.4.1.

9

Letter 2.04.074 Enclosure The title of TS 3/4.6.F was changed to Recirculation Loops Operating and a mode switch applicability restriction was also incorporated consistent with STS 3.4.1.

The LCO applies when the reactor mode switch is in either Run or Startup, which are the operating modes where control of the recirculation pumps is required.

TS 3/4.6.F.1 was not revised. This TS ensures that recirculation loop TLO operation is conforming to the accident analyses previously evaluated for LOCA and LPCI Loop Select Logic. Maintaining matched recirculation pump speeds (and drive flows) within limits is consistent with the restrictions defined in STS 3.4.1 for TLO operation.

New TS limits applicable to SLO operation were identified based on the LCO limits defined in STS 3.4.1. These new SLO limits reference reactor fuel LCO limits established for APLHGR (TS 3/4.11.A) and MCPR (TS 3/4.11.C), and the APRM High Flux Trip Setting (as defined in TS Table 3.1.1). In accordance with the proposed TS, calculated values for SLO will be identified in the COLR.

TS 3/4.6.F.3 identifies the required actions and time limits for restoration of compliance to TS 3/4.6.F.1 and TS 3/4.6.F.2. A 24-hour completion time limit to restore compliance is identified. If compliance cannot be achieved or no recirculation pumps are operating, action to be in Hot Shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is required. The required action and completion times are consistent with STS 3.4.1.

The 24-hour limit to achieve compliance is justified based on engineering judgment of the likelihood of a transient or DBA occurring simultaneously with the limits out of specification while providing sufficient time for deliberate controlled operator action. The 12-hour shutdown requirement is acceptable based on operating experience related to actions necessary to shutdown the plant in an orderly manner without challenging plant systems.

TS 3.6.A.6 - Thermal-Hydraulic Stability included the requirement to ensure forced recirculation when core thermal power is 2 25%. This TS did not identify a required action or required surveillance. Revised TS 3/4.6.F includes the requirement for ensuring core flow during power operation as well as a required shutdown action and associated completion time limit. As previously discussed above, a 12-hour shutdown action and associated completion time is justified based on operating experience related to completing shutdown activities.

Jet Pumn ODerabilitv A revision to TS 3/4.6.E - Jet Pumps is identified to ensure that jet pump surveillance requirements can be performed under either SLO or TLO operating conditions. The revisions also ensure consistency with STS 3.4.2 requirements.

Based on the removal of License Condition 3.E, existing surveillance requirement TS 4.6.E.1 cannot be performed when the plant is operating in SLO. The proposed surveillance revision provides for three separate, loop-specific options to verify jet pump Operability.

The recirculation pump speed operating characteristics (pump flow and loop flow versus pump speed) are determined by the flow resistance from the loop suction through the jet pump nozzles. A change in the relationship indicates a plug, flow restriction, loss in pump hydraulic performance, leakage, or new flow path between the recirculation pump discharge and jet pump nozzle. For this criterion, the pump flow and loop flow versus pump speed relationship must be verified.

10

Letter 2.04.074 Enclosure Individual jet pumps in a recirculation loop normally do not have the same flow.

The unequal flow is due to the drive flow manifold, which does not distribute flow equally to all risers. The flow (or jet pump diffuser to lower plenum differential pressure) pattern or the relationship of one jet pump to the loop average is repeatable. An appreciable change in this relationship is an indication that increased (or reduced) resistance has occurred in one of the jet pumps. This may be indicated by an increase in the relative flow for a jet pump that has experienced beam cracks.

The deviations from normal are considered indicative of a potential problem in the recirculation drive flow or jet pump system. Normal flow ranges and established jet pump flow and differential pressure patterns are established by plotting historical data as discussed in GE SIL No. 330, June 9,1980 (Reference 2).

The required TS 3.6.E.1 LCO action and the associated action time limit was also revised from "be in Cold Shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />" to "be in Hot Shutdown in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />." The completion time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is reasonable, based on operating experience to complete shutdown actions in an orderly manner without challenging plant systems. These revisions are consistent with STS 3.4.2 actions and required time limits.

An editorial change was also implemented to capitalize Startup and Run Modes and Operable in TS 3.6.E.1.

TS 3/4.11-Reactor Fuel Assembly Revisions Revisions to TS 3/4.11.A, TS 3/4.11.B, and TS 3/4.11.C are identified to clarify TS LCO applicability and to improve consistency with STS 3.2.1, 3.2.2, and 3.2.3. TS 4.11.C was revised to make an editorial correction to specified power level.

The revised specifications identify that when the reactor is operated below 25%

rated thermal power, compliance with the reactor fuel limit LCOs is not required.

APLHGR, MCPR and LHGR limits are derived from fuel design evaluations and LOCA and transient analyses that are assumed to occur at high power levels.

Design calculations and operating experience have shown that when power is reduced, the margin to the required fuel limits increases. Below 25% rated thermal power the reactor is operating with substantial margin and as a result, the LCO is not required. This LCO applicability restriction is consistent with STS and their associated bases.

The revised specifications identify that if the reactor fuel limits cannot be restored, prompt action is required. The two (2) hour limit to restore compliance is sufficient to perform the necessary actions and is acceptable due to the reduced likelihood of a transient or DBA occurring simultaneously with the limits out of specification.

If compliance can not be achieved, a four (4) hour action to bring the plant to a safe configuration where power levels are reduced, fuel margins are improved, and the LCO does not apply. These LCO actions and time limits are consistent with STS and their associated bases.

Editorial TS Changes A revision to the Table of Contents was identified to remove reference to two previously deleted TS sections, TS 3/4.6.H and TS 3/4.6.1. These TS sections were deleted and the TS section number is not required to maintain the TS 3/4.6 number sequence. Reference to TS 3/4.6.H was also deleted from page 3/4.6.8.

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Letter 2.04.074 Enclosure

5.

Regulatory Safety Analvsis 5.1 No Significant Hazards Consideration Entergy Nuclear Operations, Inc. (Entergy) is proposing to modify the Pilgrim Station Technical Specifications to allow plant operation with only one recirculation in service. Entergy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

i)

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

Response: No.

The proposed license and technical specification changes will allow the plant to be operated with one recirculation pump for longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> provided that appropriate limits are instituted. Extended single recirculation loop operation has been evaluated and methodologies have been established for determining appropriate operating limits. Implementation of the single recirculation loop operating limits ensures that system operation is in conformance with the conditions established to minimize the probability of accidents and the associated consequences. Required completion times for implementing the system operating limits and restoring out of specification limits minimize the probability that an accident occurs when out of specification condition's exist while allowing for deliberate operator action.

Therefore, this proposed amendment does not involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated.

ii)

Does the proposed change create the possibility'of a new or different kind of accident from any previously evaluated?

Response: No.

The proposed license and technical specification changes will allow plant operation with a single recirculation loop for longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The proposed changes introduce an additional recirculation system-operating mode, however, existing system component operating equipment or operating characteristics will not change. The Pilgrim Station Single Loop Analysis Report identifies required operating limits that apply when the system will be operated in the single'loop operation mode. Implementation of these operating limits will ensure that the system is operated in accordance with design. Additionally, revised jet pump surveillance ensures that loop specific surveillance is performed as'required to validate the bounding assumptions of existing accident analyses. As such, no new failure mechanisms are created and existing design evaluations bound system operation. '

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

12

Letter 2.04.074 Enclosure iii)

Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed license and technical specification changes identify the operating limits that apply to single recirculation loop operation. These proposed recirculation system limits were identified to ensure that system operation would be in conformance to the conditions evaluated in applicable accident and transient analyses. Implementation of the proposed limits for single recirculation loop operation ensures that safety margins are maintained. Required completion times for implementing the system operating limits minimizes the possibility that an accident occurs when out of specification conditions exist.

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

Based on the above, Entergy concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Environmental Consideration The proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 1 0 CFR 51.22(b), no environmental assessment needs to be prepared in connection with the proposed amendment.

6.

Coordination With Other Pending TS Changes There are no other pending TS changes that would be affected by this proposed license amendment.

7.

Precedents This proposed license amendment request is similar those submitted the by Tennessee Valley Authority on June 2, 1997 and approved December 23, 1998 (TAC Nos. M98885, M98886, and M8887). This proposed license amendment request is not identical to those requests due to differences in the Technical Specifications between the plants and the exact approach each licensee took to amend their Technical Specifications, however, the overall intent of all the requests are believed to be the same with very similar justifications.

8.

References

1. NUREG 1433, Standard Technical Specifications for General Electric Plants, BWR/4, Revision 3, June 2004.

2. GE Services Information Letter (SIL) No. 330, "Jet Pump Beam Cracks, "June 9, 1980

3. NRC Generic Letter 86-09, 'Technical Resolution of Generic Issue B-59 (N-1)

Loop Operation in BWRs and PWRs," March 1986.

13

Letter 2.04.074 Enclosure

4. General Electric Licensing Topical Report NEDE-24011-P-A, June 2000; and the US Supplement, NEDE-24011-P-A-14-US, "General Electric Standard Application for Reactor Fuel," June 2000.

5. NRC Bulletin No. 88-07 and Bulletin 88-07, Supplement 1, 'Power Oscillations in Boiling Water Reactors," December 1988.

6. GE Services Information Letter (SIL) 380 and SIL 380, Rev. 1, " BWR Thermal-Hydraulic Stability," February 1984.

7. NRC Generic Letter (GL) 94-02, "Long-Term Solutions and Upgrade of Interim Operating Recommendations for Thermal-Hydraulic Instabilities in Boiling Water Reactors," July 1994.

8. NEDO-31960-A, and NEDO-31960-A, Supplement 1, November 1995, "BWR Owners Group Long-Term Stability Solutions Licensing Methodology," November 1995.

9. NEDO-32339-A, "Reactor Stability Long-Term Solution: Enhanced Option 1-A,"

and associated Supplements 1, 2, 3 and 4, April 1998.

10. NEDO-32465-A, " Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications (Options 1 D and 1II)," August 1996.

11. GE Services Information Letter (SIL) No. 517, "Single Loop Operation-GE BWR/

3, 4, 5 and 6 Plants," July 1990.

14 to 2.04.074 Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Plant Proposed Amendment to the Technical Specifications Mark-Up of License and Technical Specifications (with Bases)

(16 total pages)

S B.

Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 205, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

C.

Records ENO shall keep facility operating records in accordance with the requirements of the Technical Specifications.

D.

Ecualizer Valve Restriction - DELE-TED E.

Recirculation Loool nooerable-t£ LETi 19 1

reaitor shall not b operated wth one rcirculation loop t of servic for more thary?4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. )th the r ctor opeting, o

recir lation loop iyout of se e, the t shall b plac in hot sh down conditin within hours u ess the l26p is r turned to service.-

F.

Fire Protection ENO shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility and as approved in the SER dated Decerriber 21, 1978 as supplemented subject to the following provision:

ENO may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

G.

Phvsical Protection TXhe licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10CFR73.55 (51 FR27817 and 27822) and to the authority of 10CFR50.90 and 10CFR50.54(p). The plans, which contain Safeguards Information protected under iOCFR73.21, are entitled: "Pilarim Nuclear Power Station Physical Security Plan," with revisions submitted through September 18,1967;

'Pilarim Nuclear Power Station Guard Training and Qualification Plan,"

with revisions submitted through September 24,1984; and "Pilarim Nuclear Power Station Safeguards Contingency Plan," with revisions submitted through February 15,1984. Changes made in accordance with 10CFR73.55 shall be implemented in accordance with the schedule set forth therein.

Amendment C-, 4-8, I

TABLE OF CONTENT$

'1 LIMITING CONDImONS FOR OPERATION SURVEILLANCE REQUIREMENTS'.

3.6 B.

C.

D.

PRIMARY SYSTEM BOUNDARY Thermal and Pressuri~ation Limitations Coolant Chemistry Coolant Leakage Safety and Relief Valves Jet Pumps

.1 1

K c:c I

,t et e Structund IntearirN 4.6 A

B C.

D 3/4.6-1 3/4.6-1 3/4.6-3 3/4.6-4 314.6-6 32/4.6-7 3/4.8-3 3I4.56 B3/4.6-1

/O' 5 0f,4 ~,

.

Z...

- I F--;

I

-BASES

-I 3.7 CONTAINMENT SYSTEMS A.

Primary Containment B.

Standby Gas Treatment System and Control Room High Efficiency Air Filtration System C..

Secondary Containment BASES 3.8 PLANT SYSTEMS 3.8.1 Main Condenser Offaas 3.8.2 Mechanical Vacuum Pump BASES 3.9 AUXILIARY ELECTRICAL SYSTEM A.

Auxiliary Electrical Equipment E.

Operation with Inoperable Equipment BASES 4.7 A

B 3/4.7-1 3/4.7-1 C

4.8 4.8.1 4.8.2 4.9 A.

B 3/4.7-11 3/4.7-16 B3/4.7-1 3/4.8-1 3/4.8-1 3/4.8-2 B3/4.8-1 3/4.8-1 3/4.9-1 3/4.9-4 B3/4.9-1 3.10 CORE ALTERATIONS A.

Refueling Interlocks B.

Core Monitoring C.

Spent Fuel Pool Water Level D.

Multiple Control Rod Removal BASES 3.11 REACTOR FUEL ASSEMBLY A.

Average Planar Linear Heat Generation Rate (AFLHGR)

B.

Linear Heat Generation Rate (LHGnR)

C.

Minimum Critical Fower Ratio (MCPR)

D.

Power/Flow Relatcnship During Power Operation EASES 4.10 AB C

D 3/4.10-1 3/4.10-1 3/4.10-1 3/4.10-2 3/4.10-2 B3/4.1 0-1 4.11 A.

3/l4.11-1 E

D 3/4.1 1-1 3/4.11-2 3/4.11-2 3/4.11-4 E3/4.11 -1 RX h,.iti J^..

I Amendment No. P l FNPS ii

• i J*

2.0 SAFETY LIMITS 2.1 Safety ULmits 2.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10% of rated core flow:

THERMAL POWER shall be < 25% of RATED THERMAL POWER.

2.1.2 With the reactor steam dome pressure > 785 psig and core flow> 10o of rated core'flow:

MlMUM CRITICAL POWER RATiIO shall be> -aim6k 4-v re-e IR I' e v ir oiateOrwe the reactor vessel, the water level shall not be less than 12 inches above the top of the normal active fuel zone.

2.1.4 Reactor steam dome pressure shall be < 1325 psig at any time when irradiated fuel is present in the reactor vessel.

2.2 Safetv Limit Violation With any Safety Limit not met the following actions shall be met:

2.2.1 Within one hour notify the NRC Operations Center in accordance with 1 OCFR50.72.

2.2.2 Within two hours:

A. Restore compliance with all Safety Limits, and B. Insert all insertable control rods.

2.2.3 The Station Director and Senior Vice President - Nuclear and the Nuclear Safety Review and Audit Committee (NSRAC) shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2.2.4 A Licensee Event Report shall be prepared pursuant to 1 0CFR50.73. The Licensee Event ReDort shall be submitted to the Commission, the Operations Review Committee (ORC), the NSRAC and the Station Director and Senior Vice President - Nuclear within 30 days of the violation.

2.2.5 Critical operation of the unit shall not be resumed until authorized by the Commission.

Amendment <

.'=,

0 2, 72, '13, 8E,

.7 2-1

BASES:

3.1 REACTOR PROTECTIONASYSTEM (Cont) could exceed the safety limit.

The 15% APRM scram remains active until the mode switch is placed in the RUN position.

The analysis to support operation at various.power and flow relationships has considered opQ-rt&zz a.ih two recirculation vpe r boor Oe-A 4T;M anc Si 1e o

The IRM system consists of 8 chambers, 4 in each of the reactor protection system logic channels.

The IRM is a 5-decade instrument which covers the range of power level between that covered by the SRM and the APRM.

The 5 decades are covered by the IRM by means of a range switch and the 5 decades are broken down into 10 ranges, each being one-half of a decade in size.

The IRM scram setting of 120/125 of full scale is active in each range of the IRM.

For example, if the instrument were on Range 1, the scram setting would be a 120/125 of full scale for that range; likewise, if the instrument were on Range 5, the scram would be 120/125 of full scale on that range.

Thus, as the IRM is ranged up to accommodate the increase in power level, the scram setting is awlso ranged up.

The most significant sources of reactivity change during the power increase are due to control rod withdrawal. For in-sequence control rod withdrawal, the rate of change of power is slow enough due to the physical limitation of withdrawing control rods that heat flux is in equilibrium with the neutron flux, and an IRM scram would result in a reactor shutdown well before any safety limit is exceeded.

In order to ensure that she IRM provided adequate protection against the single rod withdrawal error, a range of rod withdrawal accidents was analyzed.

This analysis included starting the accident at various power levels.

The most severe case involves an initial condition in which the reactor is just subcritical and the IRM system is not yet on scale.

This condition exists at quarter rod density. Additional conservatism was taken in this analysis by assuming that the IRM channel closest to the withdrawn rod is bypassed.

The results of this analysis show that the reactor is scrammed and peak core power limited to one percent of rated power, thus maintaining MCPR above the safety limit MCPR.

Based on the above analysis, the IRM provides protection against local control rod withdrawal errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.

Reactor Low Water Level The setpoint for-low level scram is above the bottom of she separator skirt.

This level has been used ir.

transient anaivses dealing with coolant invento~y decrease.

The results show that scram at this level properly protects the -uel and the pressure barrier, because MCPR remains well above the safety limit MCPR in all cases, and system pressure does not reach the safety valve settings.

The scram setting is approximately 15 inches below the normal operating range and is thus sufficient to avoid spurious scrams.

Revision.

5h 79;-1337-147.1 9-=_

B3/4.1-4

LIMITING CONDITION FOR OPERATION 3.6 PRIMARY SYSTEM BOUNDARY (Cont)

SURVEILLANCE REQUIREMENTS A.

Thermal and Pressurization I

Limitations (Cont)

In the event this requirement is not met, achieve stable reactor conditions with reactor vessel temperature above that defined by the appropriate curve and obtain an engineering evaluation to determine the appropriate course of action to take.

3. The reactor vessel head bolting studs shall not be under tension unless the temperature of the vessel head flange and the head is greater than 55'F.

4. The pump in an idle recirculation loop shall not be started unless the temperatures of the coolant within the idle and operating recirculation loops are within 50OF of each other.

4.6 PRIMARY SYSTEM BOUNDARY (Cont)

A.

Thermal and Pressurization Limitations (Cont)

Test specimens of the reactor vessel base, weld and heat affected zone metal subjected to the highest fluence of greater than 1 Mev neutrons shall be installed in the reactor vessel adjacent to the vessel wall at the core midplane level. The specimens and sample program shall conform to the requirements of ASTM E 185-66.

Selected neutron flux specimens shall be removed at the frequency required by Table 4.6-3 and tested to experimentally verify adjustments to Figures 3.6-1, 3.6-2, and 3.6-3 for predicted NDT temperature irradiation shifts.

3. When the reactor vessel head bolting studs are tensioned and the reactor is in a Cold Condition, the reactor vessel shell temperature immediately below the head flange shall be permanently recorded.

4. Prior to and during startup of an idle recirculation loop, the temperature of the reactor coolant in the operating and idle loops shall be permanently logged.

5. Prior to starting a recirculation pump, the reactor coolant temperatures in the dome and in the bottom head drain shall be compared and permanently logged.

.Vt

5. The reactor recirculation pumps shall not be started unless the coolant-temperatures between the dome and the bottom head drain are within 145'F.

1-_

Amendment No. 2 7 T-2 8 T-4 2 T 827-143 3/4.6 -2

MO O

SC R

Q LIMFlPING CONDITONS FOR OPERAMON' SURVEILLAN&E REQUIREMENTS.'

3.6 PRIMARY SYSTEM BOUNDARY (Cont) 4.6 PRIMARY SYSTEM BOUNDARY (Cont)

Safetv Relief Valves (Con't)

4. Any safety relief valve whose discharge pipe temperature exceeds 212OF for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or more shall be removed at the next cold shutdown of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or more, tested in the as-found condition, and recalibrated as necessary prior to reinstallation. Power ooeration shall nct continue beyond 90 days from the initial discovery of discharge pipe

-temperatures in excess of 2120F for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without prior NRC approval of the engineering evaluation delineated in 3.6.D.3.

5. The limiting conditions of operation for the instrumentation that monitors tail pipe temperature are given in Table 3.2-F.

E. Jet PumDs E. Jet PumDs

1.

• henever the reactor is in th e/tartup or mps shll beNOTES

,iernbodes, all jet pumps shall be

1. Not required to be performed until 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> ePrable. If it is determ after the associated recirculation loop is in Spumio is 1noperable ntropsuttion.

Hall b nitiated d the ctor Sor n

(in a5 Shut n Cort on

2. Not required to be performed until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after >25% Rated Thermal Power.

(A7V1 e<A Wheever there owwith the 7.,.

a Ireactor in the startup or 'run modes, jet pump 1770 era;+ 4.no operability shal e checked daily by verifying lZ Acr S

/that no two offe following conditions occur

_simultane sly./

1. two recirculatio ops have a flow balance of 1 aT r more when the

/

pumps are ope ed at the same speed.

2. The indica d value of core flow rate varies fr the value derived fro Co flow m surements by more It 10%.

3. The diffuser to lower pi m differential pressure reading on individual jet pump varies from estabF ed jet pump delta F characteristics by more than 10%.

Amendment No. 4, 56, 74,,2, 123;205-.

3/4.6-7 I

Insert for SR 4.6.E - Jet Pumps Whenever there is recirculation flow with the reactor in the Startup or Run Modes, jet pump operability shall be checked daily by verifying at least one of the following criteria (1, 2, or 3) is satisfied for each operating recirculation loop:

1. Recirculation pump flow to speed ratio differs by < 5% from established patterns, and jet pump loop flow to recirculation pump speed ratio differs by

< 5% from established patterns.

2. Each jet pump diffuser to lower plenum differential pressure differs by < 20%

from established patterns.

3. Each jet pump flow differs by < 10%

from established patterns.

LI1ITING CONDITIONS FOR:6PERATION SUJRVEIILLANCE REQUIMENS -

3.6 PRIMARY SYSTEM BOUNDARY (Cont) 4.6 PRIMARY SY F.

Jet Pump Flow Mismatch F.

Jaz -ut -

1. Whenever both recirculation Recirculat pumps are in operation, ump checked an' speeds shall be mainta ed' per day.

within 10% of each oaer when power level is great than 8 %

and within 15% of e ch other c i when Dower level i less than or ecual to 89%. /

X 1

2. If Specificaton 3.6.F.1 is exceeded imm iate corrective action shal be taken. If recircula on pump speed mismatch s not corrected within 0 minutes, an orderly G.

Structural shutd shall be initiated and the actor shall be in the Inservice Col Shutdown condition within shall be p 24 ours unless the with the P:

re irculation pump speed Program.

mismatch is brought within from compl ssooner.

will be ev, completion G.

Structural Integritv interval.

evaluation

1. The structural integritv of the STEM BOUNDARY (Cont) ion pump speeds shall be d logged at least once

,r.o> Im L "ps Inteeritv inspection of components erformed in accordance NPS Inservice Inspection The resultfs obtained iance with this program aluated at the of each ten year The conclusions of this will be reviewed with

.J primary system boundary shall be maintained at the level required by the ASME Boiler and Pressure Vessel Code,Section XI "Rules for Inservice Inspection of Nuclear Power Plant Components," Articles IWA, IWB, IWC, IWD and IWF and mandatory appendices as required by 10CFR50.55a(g),

except where specific relief has been granted by the NRC pursuant to 10CFR50.55a(g)(6)(i).

tne NKG.

i;.

Deleted Ij ARmenient o.

7 Amendment No. 197T-9 3 T -1; 3i 3/4.6-8

Insert for TS 3.6.F - Recirculation Loops Operating F. Recirculation Loops Operating During operation in the Run and Startup Modes, at least one recirculation pump shall be operating.

1. Whenever both recirculation pumps are in operation, pump speeds shall be maintained within 10% of each other when power level is greater than 80%

and within 15% of each other when power level is less than or equal to 80%.

2. Whenever a single recirculation loop is operating, the following limits are applied when the associated LCO is applicable:

a) LCO 3.11.A, "Average Planar Linear Heat Generation Rate (APLHGR)," single loop operation limits specified in the COLR, b) LCO 3.11.C, "Minimum Critical Power Ratio (MCPR)," single loop operation limits specified in the COLR, and c) LCO 3.1, "Reactor Protection System," Average Power Range Monitor High Flux function, trip level setting is reset for single loop operation per Table 3.1.1.

3. If the requirements of Specification 3.6.F.1 or 3.6.F.2 are not met, restore compliance within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If compliance is not restored or with no recirculation pumps in operation the reactor shall be in Hot Shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

'.'I7 i

BASES:

3/4.6 PRIMARY SYSTEM BOUNDARY (Cont) e E.

Jet Pumos El jet upnzl assembly hold down mnechanism, nozze assembly and/or nser, would increase the cross-sectional flow area for blowdown following the design basis double-ended recirculation line break. Therefore, if a failure occurred, repairs must be made.

A nozzle riser failure coul se the coincident failure a jet pump body; however, becaus of the lack of any substan stress in the jet pump body e converse is not possible. Therefo e, failure of a jet pump y cannot occur without the f re of the nozzle riser.

The following fa rs form the basis for the surv ance requirements:

A br in a jet pump decreases the w resistance characteristi f the external pipin lo causing the recirculation pum a operate at a higher flow nditon when compar revious operation.

The change in flow rate of t ailed jet pump produces a ange in the indicated flow rate of that pump relative tohe other jet pumps in that I p. Comparison of the data with a normal relationship r pattern provides the indi ion necessary to detect a failed jet pump.

The jet pump flow d ation pattern derived from Z diffuser to lower plenum differential pressure reading ill also be used to evaluate' pump operability.

The surveillance is m fed by two Notes. Note 1 ows the surveillance not to be pdrformed until 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after th ssociated recirculation lo s in operation, since these che s can only be performed duri he jet pump operation. T 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is an acceptable time t establish conditions appro ate for data collection and aluation. Note 2 allows the su Hillance not to be performed n thermal power is <25% Rated Thermal Power. Duringw flow conditions, jet mp noise approaches the Ireshold response of the associaed flow instrumentatgo and precludes the coile& n of repeatable and meaningful ata. The 24-hour frequency has-been shown by operatin experience to be timely for dete ng jet pump degradation and is consistent with the urveillance frequency for recirc ation loop operability verification.

The surveillance requires verificai n of jet pump operability on a ily basis by verifying two of three of the following conditions o not occur simultaneously: 1, he recirculation loops have a flow imbalance of less than 10 when the pumps are operat at the same speed, 2) the variation in the indicated car (ow rate and the loop flow ra is less than 10%, and 3) the diffuser to lower plenum diff ential pressure reading on individual jet pump is less an 100/.

The jet pump operability d ection technique is as follos. With the two recirculatia pumps balanced in speed to witlrp l 5%, the flow rates in b recirculation loops will be v rified by Control Room monitoring instruments. If the two rate values do not differ by /ore than 10%, riser and nozzle assembly integrity has be verified. If they do differ by

% or more after correction for the difference in pump spews, the diffuser to lower plenu differential pressure of all jet pumps will be compared toHstablished jet pump ALP cha teristics. In the event of a failed jet pump nozzle (or riser),, e affected jet pump diffuser erential pressure signal would be reduced because the backflow would be less than the n al forward flow. If the jet pump AP indications are within 10% of established jet pump characteristics, jet pump nozzle and riser integrity have been established. If the indica det pump AP varies from the established jet pump characteristics by more than 10%, indi

'd core flow will be compared to the core flow derived from loop flow measurements. If the difference between measured and Revision 24-w'

N 1

-7,

t.,"

I " !",

I

- ;- I

- I ! i I

1.-.,

BASES:

3/4.6 PRIMARY SYSTEM BOUNDARY (Cont) olse-4ri-

..N, E E. Jet Pumps (Cont) deried core flow rate is10% or more, afailed je pupnz or riser i.s indicated and te plant shall be shut do for repairs. If the otential blowdown flow area is increased, the system resistance to e recirculation p is also reduced; hence e affected drive mp will

'run out' to a sub ntially higher flow e (approximately 115% 120% for a single ozzle failure). If the loops are balanc in flow at the same pum peed, the resistan e characteristi annot have chan d.

Any imb ncs between drive op flow rates wculd be I cated bythe plant pr cess instrume tation. In additio ihe affected jet pump wo provide a leakage p h past the core thus reducing the core flo rate. The reverse flow t ugh the inactive Jet p

,p would still be indicated by a positive rferential pressure but the effect would be a sli ht decrease (3% to 6%) in the total core f 'w measured. This decrease, together with the loop flow increase, would result in a lack of c-of-elation dg~e~eq r

and derived core flow F.

Jokt tum p F OpA,2 The LPCI loop selection logic has been previously described in the Pilgrim Nuclear Power Station FSAR. For some limited low probability accidents with the recirculation loop operating with large speed differences, it is possible for the logic to select the wrong loop for injection. For these limited conditions the core spray itself is adequate to prevent fuel temperatures from exceeding allowable limits. However, to limit the probability even further, a procedural limitation has been placed on the allowable variation in speed between the recirculation pumps.

The licensee's analyses indicate that above 80% power the loop select logic could not be expected to function at a speed differential of 15%. At or below 80% power the loop select logic would not be expected-to function at a speed differential of 20%. This specification provides a margin of 5% in pump speed differential before a problem could arise. If the reactor is operating on one pump, the loop select logic trips that pump before making the loop selection.

The flow mismatch restriction also derives from the 'Core Flow Coastdown" concern. This concern postulates that if the recirculation loop with the higher flow is broken, the 'effective core flow' is determined by the loop with the lower flow. Compared to a matched flow condition, this would start pump coastdown from a lower flow/speed with the reactor power effectively above the rated rod line. Therefore, boiling transition may occur earlier during a postulated LOCA event, which could result in higher calculated peak cladding temperatures (PCTs). Therefore, the purpose of the "Core Flow Coastdown' flow mismatch restriction is to maintain Pilgrim within its analyzed conditions.

'=pecification 3.6.F allows 30 minutes to correct amismatch in recirculation pump s lorder to take manual control of the recirculation pump MG set scoop tube positioner in the even t hould fail.

jt Revision IL4&~

3461 B3/4.6-1 0

Insert for TS Bases Section B3/4.6.E - Jet Pumps The jet pumps are part of the Reactor Coolant Recirculation System and are designed to provide forced circulation through the core to remove heat from the fuel. Jet pump OPERABILITY is an explicit assumption in the design basis loss of coolant accident (LOCA) analysis., The structural failure of any of the jet pumps could cause significant degradation in the ability of the jet pumps to allow reflooding to two-thirds-core height during a LOCA. OPERABILITY of all jet pumps is required to ensure that operation of the Reactor Coolant Recirculation System will be consistent with the assumptions used in the licensing basis analysis.

The jet pumps are required to be OPERABLE when the reactor is operating in the Run or Startup Modes since there is a large amount of energy in the reactor core and since the limiting DBAs are assumed to occur in these MODES. This is consistent with the requirements for operation of the Reactor Coolant Recirculation System.

An inoperable jet pump can increase the blowdown area and reduce the capability of reflooding during a design basis LOCA. If one or more of the jet pumps are inoperable, the plant must be brought to a Mode in which the LCO does not apply. To achieve this status, the plant must be brought to Hot Shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is reasonable, based on operating experience, to reach Hot Shutdown from full power conditions in an orderly manner and without challenging plant systems.

The surveillance is designed to detect significant degradation in jet pump performance that precedes jet pump failure. The surveillance is required to be performed only when the loop has forced recirculation flow since surveillance checks and measurements can only be performed during jet pump operation. The jet pump failure of concern is a complete mixer displacement due to jet pump beam failure. Jet pump plugging is also of concern since it adds flow resistance to the recirculation loop. Significant degradation is indicated if the specified criteria confirm unacceptable deviations from established patterns or relationships. The allowable deviations from the established patterns have been developed based on the variations experienced at plants during normal operation and with jet pump assembly failures. Each recirculation loop must satisfy one of the performance criteria provided. Since refueling activities (fuel assembly replacement or shuffle, as well as any modifications to fuel support orifice size or core plate bypass flow) can affect the relationship between core flow, jet pump flow, and recirculation loop flow, these relationships may need to be re-established each cycle. Similarly, initial entry into extended single loop operation, off-normal operating conditions, and changes in monitoring equipment may also require establishment of these relationships. During the initial weeks of operation under such conditions, while base-lining new established patterns, engineering judgment of the daily surveillance results is used to detect significant abnormalities which could indicate a jet pump failure.

The recirculation pump speed operating characteristics (pump flow and loop flow versus pump speed) are determined by the flow resistance from the loop suction through the jet pump nozzles. A change in the relationship indicates a plug, flow restriction, loss in pump hydraulic performance, leakage, or new flow path between the recirculation pump discharge and jet pump nozzle. For this criterion, the pump flow and loop flow versus pump speed relationship must be verified.

Insert forTS Bases Section B3/4.6.E -Jet Pumps (continued)

Individual jet pumps in a recirculation loop normally do not have the same flow. The unequal flow is due to the drive flow manifold, which does not distribute flow equally to all risers. The flow (or jet pump diffuser to lower plenum differential pressure) pattern or relationship of one jet pump to the loop average is repeatable. An appreciable change in this relationship is an indication that increased (or reduced) resistance has occurred in one of the jet pumps. This may be indicated by an increase in the relative flow for a jet pump that has experienced beam cracks.

The deviations from normal are considered indicative of a potential problem in the

-.recirculation drive flow or jet pump system. Normal flow ranges and established jet pump flow and differential pressure patterns are established by plotting historical data as discussed in GE SIL No. 330, June 9, 1980.

The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> frequency has been shown by operating experience to be timely for detecting jet pump degradation and is consistent with the Surveillance Frequency for recirculation loop OPERABILITY verification.

The surveillance is modified by two Notes. Note 1 allows this Surveillance not to be performed until 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the associated recirculation loop is in operation, since these checks can only be performed during jet pump operation. The 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is an acceptable time to establish conditions appropriate for data collection and evaluation. Note 2 allows the surveillance not to be performed when THERMAL POWER is m 25% of RTP. During low flow conditions, jet pump noise approaches the threshold response of the associated flow instrumentation and precludes the collection of repeatable and meaningful data.

Insert for TS Bases Section B3/4.6. F - Recirculation Loops Overatinc A plant specific LOCA analysis has been performed assuming only one operating recirculating loop. This analysis has demonstrated that, in the event of a LOCA caused by pipe break in the operating recirculation loop, the Emergency Core Cooling System response will provide adequate core cooling, provided the APL-IGR requirements are modified accordingly.

The transient analyses of Chapter 14 of the FSAR have also been performed for single loop operation and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed provided the MCPR requirements are modified. During single loop operation modification of the Reactor Protection System (RPS) average power range monitor (APRM) instrument setpoints is also required to account for the different relationships between recirculation drive flow and the reactor core flow.

The APLHGR, MCPR, and APRM High Flux limits for single loop operation are specified in the COLR.

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.11 REACTOR FUEL ASSEMBLY Aoplicabilitv: -

The Limiting Conditions for Operation associated with fuel rods apply to those parameters which monitor the fuel rod operating conditions.

Objective:

4.11 REACTOR FUEL ASSEMBLY Avolicabilitv:

The surveillance requirements apply to the parameters which monitor the fuel rod operating conditions.

Obiective:

The Objective of the Surveillance Requirements is to specify the type and frequency of surveillance to be applied to the fuel rods.

The Objective of Limiting Conditions for Operation is to assure the performance of the fuel rods.

Svecifications:

/

Ž 5

_ri 4

Soecifications:

A.

Average Planar Linear Heat/

Generation Rate (APLHGR)

A.

Averaze Planar'Linear Heat During-p.xeo operation both ro-}-hi-Pi-nvPsnoPar-;=I 4 =

the APLHGR for each type of fuel as a function of average planar exposure shall not exceedthe applicable limiting value specified in the CORE OPERATING Generation Rate (APLHGR)

The APLHGR for each type of fuel as a function of average planar exposure shall be determined daily during reactor operation at> 25% -

rated thermal power.

LIMITS REPORT.

ŽZ /

rt et A-je5-If at any time during operation zIt.t the limiting value for APLHGR is being exceeded, action shall be.

initiated within 15 minutes to restore operation to within the prescribed limits.

If the APLHGR is not returned to within the prescribed limits within two (2) hours*. e reactor snar Drought t the Col Shu down ondiioe ithin96 hou.s.

Su ve Cat n cor-4-pon, ng

-eac_-- ope-i--cr. i4 -_-

hz eascribed M-:-

k el~

r e-n o(H)

Couf J-c v4'~S' U-;

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-G -

Amendment No. l5;-24T-27;-49>1Q5-.4.;-

3/4.11-1

.Il LIMMNG CONDMONS FOR OPERATION SURVEINCE REQUIREMENTS 3.11 REACTOR FUEL ASSEMBLY (Cont) 4.11 REACTOR FUEL ASSEMBLY (Cant)

B.

Linear Heat Generation Rate (LHGR)

B.

Linear Heat Generation Rate (LHGR)

Duringgr operatic KeLLHGR The LHGR as a function of core height shall not exceed the limits specified in the shall be checked daily during reactor CORE OPERATING UMITS REPORT.

operation at >25% rated theffna power.

If at any time during operatico limiting value for LHGR. is being exceeded, c

~:A action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the LHGR is not retumed to within th rescibed Ii a

o within two (2) hours e

tor shl be TA C

/

fprougn t e utdo

)vcondiion

.W hOWS within 30hours.

arveiIlaoe and/

=

)

11-------;W_--9M----MW

Z /-%

4 C.

Minimum Critfiri Pnwer Ratio (MCPRI'-

Ilv i 1111 8

16 i S9 -PJ lt i

t SJVVrr [

1. During-pewer operatior iCPR shall be

' the MCPR operating limit specified in the Core Operating Limits Report. If at any time.during operationio a-l-the limiting value for MCPR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the steady-state MCPR is not returned to within the prescribed lirriits within two 1.- MCPR shall be deteftmin uring reactor power operation

'at 25%ated thermal power and followod an change in power level or di&reibition that would cause operation with a limiting control rod pattern as specified

Žh

,2Sc°). reed Outfr~it 1p°v¢ r

2. The value of X in Specification 3.11.C.2.

shall be equal to 1.0 unless determined from the result of surveillance testing of Specification 4.3.C as follows:-

I a) X is defined as Tave - -B I2=

1.252 - T5 II Amendment No. 15, 27, 29, 42, 5'4, 105, 433, 165, -29 3/4.11-2