ML042640029

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Proposed Amendment to the Technical Specifications, General Electric Report GE-NE-0000-0027-5301, Pilgrim Nuclear Power Station, Single Loop Operation (Non-proprietary)
ML042640029
Person / Time
Site: 05000283
Issue date: 07/31/2004
From:
General Electric Co
To:
Office of Nuclear Reactor Regulation
References
-RFPFR, 2.04.074, DRF 0000-0027-4472 GE-NE-0000-0027-5301a, Rev 1
Download: ML042640029 (43)
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Attachment 3 to 2.04.074 Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Plant ProDosed Amendment to the Technical SDecifications General Electric Report GE-NE-0000-0027-5301 Pilgrim Nuclear Power Station Single Loop Operation (Non-proprietary)

GE Nuclear Energy Non-proprietary verision GE-NE-0000-0027-5301 a Revision 1 DRF 0000-0027-4472 Class I July 2004 Pilgrim Nuclear Power Station Single Loop Operation

GE-NE-0000-0027-5301 a Revision 1 DRF 0000-0027-4472 Class I July 2004 Non-ProprietaryVersion Pilgrim Nuclear Power Station Single Loop Operation

GE-NE-0000-0027-5301a, Revision I NON-PROPRIETARY VERSION IMPORTANT NOTICE REGARDING THE CONTENTS OF THIS REPORT Please Read Carefully A. Disclaimer The only undertakings of the General Electric Company (GE) respecting information in this document are contained in the contract between Entergy Nuclear Operations, Inc (Entergy) and GE. Nothing contained in this document shall be construed as changing the applicable contract. The use of this information by anyone other than Entergy as authorized by GE to have this document, or for any purpose other than that for which it is intended, is not authorized. With respect to any unauthorized use, GE makes no representation or warranty, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document, or that its use may not infringe privately owned rights.

B. Information Notice This is a non-proprietary version of the document GE-NE-0000-0027-5301, which has the proprietary information removed. Portions of the document that have been removed are indicated by an open and closed bracket as shown here ((

ini

GE-NE-0000-0027-5301a, Revision I NON-PROPRIETARY VERSION TABLE OF CONTENTS Paae

1. INTRODUCTION AND

SUMMARY

................................... I-I 1.1 Introduction ................................... 1-1 1.2 Summary ................................... 1-1

2. OVERALL ANALYSIS APPROACH 2-1
3. FUEL CLADDING SAFETY LIMIT MCPR ................................... 3-1 3.1 Core Flow Uncertainty ................................... 3-1 3.1.1 Core Flow Measurement During 3-1 Single-Loop Operation 3.1.2 Core Flow Uncertainty Analysis 3-2 3.2 TIP Reading Uncertainty ................................... 3-4 3.3 Calculation Options for SLMCPR . .................................. 3-4
4. TRANSIENT EVALUATIONS ................................... 4-1 4.1 Anticipated Operational Occurrences ............................... 4-1 4.1.1 Core Flow Decrease Events 4-2 4.1.2 Core Flow Increase Events 4-2 4.1.3 Cold Water Injection Events 4-2 4.1.4 Pressurization Events 4-3 4.1.5 Recirculation Pump Seizure Accident 4-3 4.1.6 Rod Withdrawal Error 4-5 4.2 Plant Operating Limit MCPR ............................. 4-6 4.3 Overpressurization Protection 4-8
5. STABILITY ANALYSIS .. ................................. 5-1 5.1 Phenomena ................................... 5-1 5.2 Compliance to Stability Criteria ................................... 5-2
6. ECCS PERFORMANCE ANALYSIS ................................... 6-1
7. REACTOR INTERNAL VIBRATIONS ................................... 7-1
8. PROPOSED DOCUMENTATION CHANGES ................................... 8-1
9. CONCLUSIONS ................................... 9-1
10. REFERENCES ................................... 10-1 iv

GE-NE-0000-0027-5301a, Revision I NON-PROPRIETARY VERSION LIST OF TABLES Page 2-1 EVALUATIONS PRESENTED IN THIS REPORT 2-2 8-1 RECOMMENDED DOCUMENTATION CHANGES 8-1 LIST OF FIGURES Pane 1-1 PNPS. SLO POWER/FLOW MAP .................................... 1-3 31 ILLUSTRATION of SINGLE RECIRCULATION ............................... 3-5 LOOP OPERATION FLOWS 4.2-1 Power Dependent MCPR Limits (Turbine Bypass In-Service) 4-9 (FFWTR included in limits) 4.2-2 Power Dependent MCPR Limits (Turbine Bypass Out-of-Service) .4-10 (FFWTR included in limits) 4.2-3 Power Dependent LHGR Factors with Turbine Bypass .4-11 In or Out of Service (FFWTR included in limits) 4.2-4 Power Dependent MAPLHGR Factors with Turbine Bypass . 4-12 In or Out of Service (FFWTR included in limits) 4.2-S Flow Dependent MCPR Limits (All Operating Flexibility Options)..4-13 4.2-6 Flow Dependent LHGR Factors (All Operating Flexibility Options) 4-14 4.2-7 Flow Dependent MAPLHGR Factors .- 15 V

GE-NE-0000-0027-5301a, Revision I NON-PROPRIETARY VERSION ACRONYMS Term Definition AOO Anticipated Operational Occurrence APRM Average Power Range Monitor ARTS APRM/RBM/Technical Specification ATWS Anticipated Transient Without Scram BWR Boiling Water Reactor CPR Critical Power Ratio DBA Design Basis Accident ECCS Emergency Core Cooling System ELLLA Extended Line Load Limit Analysis EOOS Equipment Out-Of-Service FFWTR Final Feedwater Temperature Reduction FIV Flow-Induced Vibration FSAR Final Safety Analysis Report FWCF Feedwater Controller Failure FWHOOS Feedwater Heater Out-of-Service GE General Electric ICGT In-Core Guide Tube ICF Increased Core Flow LHGR Linear Heat Generation Rate LOCA Loss-Of-Coolant Accident LRNBP Load Rejection with No Bypass MAPLHGR Maximum Average Planar Linear Heat Generation Rate MCPR Minimum Critical Power Ratio MELLLA Maximum Extended Load Line Limit Analysis MOP Mechanical Over-Power MSIV Main Steamline Isolation Valve vi

GE-NE-0000O27-53Oa, Revision I NON-PROPRIETARY VERSION ACROTNYMS (Continued)

Term Definition OLMCPR Operating Limit Minimum Critical Power Ratio OOS Out-of-Service PCT Peak Clad Temperature PNPS Pilgrim Nuclear Power Station RBM Rod-Block Monitor RWE Rod Withdrawal Error SLMCPR Safety Limit Minimum Critical Power Ratio SLO Single Loop Operation SRLR Supplemental Reload Licensing Report SRV Safety/Relief Valve TIP Traversing In-Core Probe TLO Two Loop Operation TOP Thermal Over-Power TTNBP Turbine Trip with No Bypass USNRC US Nuclear Regulatory Commission vii

GE-NE-0000-0027-5301a, Revision 1 NON-PROPRIETARY VERSION I. INTRODUCTION AND

SUMMARY

1.1 Introduction The capability to operate in the single-loop operation (SLO) mode is useful if a recirculation pump or other component malfunction renders one recirculation loop inoperative for longer time than allowed by standard plant technical specifications. To support the use of long term SLO for Pilgrim Nuclear Power Station (PNPS), the accidents, abnormal operational transients and other events associated with power operation were reviewed for the case with only one recirculation pump in operation. The results of this evaluation are presented in this report.

1.2 Summary In this report, SLO operation for PNPS is justified based on the results of the evaluations documented here. The major conclusions of this report are summarized as follows:

  • Increased uncertainties in the total core flow and Traversing In-Core Probe (TIP) readings are conservatively accounted for by an incremental increase of -0.01 to 0.02 in the Safety Limit Minimum Critical Power Ratio (SLMCPR) during single-loop operation. The two loop and single loop SLMCPR are calculated each cycle, so the exact incremental increase may vary. This increase in SLMCPR must be included in the determination of the Operating Limit MCPR (OLMCPR) for SLO (Sections 3 and 4).
  • The reduced core flow coastdown for recirculation line breaks requires a Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) and Linear Heat Generation Rate (LHGR) SLO multiplier of 0.80 (Section 6).

1-1

GE-NE-0000-0027-5301a, Revision I NON-PROPRIETARY VERSION

  • The flow-biased Average Power Range Monitor (APRM) rod block and scram setpoints are adjusted for SLO condition (Sections 4 and 8).
  • The maximum rod line allowed for SLO operation is the rod line with a maximum power of 65% and a core flow of 52% of rated (Reference 1) due to SLO vessel internal vibration evaluation (Section 7).

Based on the evaluations presented herein, it is concluded that SLO for PNPS is acceptable within the region shown on Figure 1-1. The low power interlock is shown to be adequate to protect against jet pump cavitation for SLO (Reference 1). The PNPS recirculation flow control system should be operated in the manual flow control mode during SLO to prevent potential control oscillations from occurring in the recirculation flow control system. A summary of the plant Technical Specifications changes is shown in Section 8.

1-2

GE-NE-0000-0027-5301a, Revision I NON-PROPRIETARY VERSION Core Flow (Mll)/hr) 0 10 20 30 40 S0 6O 70 R0 120

  • 2400 A! Natural Circulation 110 - at 26% Minimum Pump Speed 2200 Ci 63.2% Power/ 36.2% Flow D: 100.0% Power/ 76.7% Flow E F 202%MWI 100 .- D'i 98.52% Power/ 75.0% Flow RIW: 2(X)()(

Es 100.0% Power/ 100.0% Flow E's 98.S2% Power/ 100.0% Flow EF' F: 100.0% Power/ 107.5 Flow 90 t F't 99.5% Power/ 107.5% Flow IF00 Gi 78.e8 Power/ 112.5% Flow Hi 49.3% Power/ 112.51 Flow 1: 49.3% Power/ 100.0% Flow *1600 ci 0_

70 O 1CF 1400 ;;

L..

I.Zt 2..

0) 60 -. 110, 1200 0 0 .1~

50 *1 6 -0 II L.

40 -

30 (.00 Maximum SLO Bound 20 100t Licensed Core Power . 2028 MWt *400 100t Core Flow - 69.0 Mlb/hr

. '.  ; / SLO Jet Pump 10 -f 200 II Cavitation Line

. . . .- I - - ... ,.

l i5* -- I - I - - - - I - - - - - - I - - - - - - - -0 0 10 20 30 40 50 60 70 80 90 100 110 120 Core Flow (%)

Fipire 1-1 PNPS SLO POWER/FLOW M1AP 1-3

GE-NE-0000-0027-53Oa, Revision I NON-PROPRIETARY VERSION_

2. OVERALL ANALYSIS APPROACH The Reference I report documents a previous SLO evaluation performed for the PNPS.

This section describes the various analyses that are being updated for the PNPS for SLO.

The following safety and regulatory concerns are identified as potentially being affected as a result of SLO:

  • Thermal limits during abnormal operational occurrences.
  • Thermal-hydraulic stability performance.
  • ECCS/loss of coolant accident (LOCA) performance.
  • Reactor vessel internals vibration responses.
  • Effects on plant Technical Specifications.

Each applicable safety and regulatory concern implied in the above listed items was reviewed to determine the acceptability of plant operation in the single loop mode. Other analyses, such as Containment Response and Anticipated Transients Without Scram are not addressed specifically here or in previous SLO for PNPS (Reference 1). The analyses not addressed here are assumed to be both: (a) not sensitive to the SLO plant state, and (b) valid in the power to flow operating map for SLO. Additionally, the reactor vessel internals vibration response to SLO from Reference I is applied to this evaluation and is summarized in Section 7. Discussions are presented in the sections indicated in Table 2-1.

The SLO analysis may be combined with other PNPS non-standard plant operating conditions and equipment out-of-service (EOOS), such as final feedwater temperature reduction (FFWTR), Feedwater Heater out-of-service (FWHOOS) or Main Steam Isolation Valve (MSIV) out-of-service (OOS), because the analyses for SLO that also affect these options conclude that the analysis for two loop is applicable to SLO.

However, SLO is not compatible with the Maximum Extended Load Line Limit Analysis (MELLLA) or Increased Core Flow (ICF) options as these operating regions are excluded in SLO.

2-1

GE-NE-0000-0027-5301a, Revision 1 NON-PROPRIETARY VERSION Table 2-1 EVALUATIONS PRESENTED IN THIS REPORT Item Section Result Fuel Cladding Integrity Safety Limit 3.0 Acceptable with correction MCPR Fuel Transient Thermal Limits and 4.0 Acceptable with correction and Over-Pressure Limit setpoint adjustment Thermal-Hydraulic Stability 5.0 Acceptable with setpoint adjustment for Option ID ECCS/LOCA Performance 6.0 Acceptable with multiplier Reactor Vessel Internals Vibration 7.0 Previous analysis and conclusions applied Effects on Technical Specifications 8.0 Acceptable Changes required 2-2

GE-NE-0000-0027-5301a, Revision 1 NON-PROPRIETARY VERSION

3. FUEL CLADDING INTEGRITY SAFETY LIMIT MCPR Except for the total core flow and TIP reading, the uncertainties used in the statistical analysis to determine the SLMCPR are not dependent on whether coolant flow is provided by one or two recirculation pumps. Uncertainties used in the two-loop operation analysis are documented in Reference 2. A 6% core flow measurement uncertainty has been established for SLO (compared to 2.5% for two-loop operation). As shown in Subsection 3.1, this value conservatively reflects the one standard deviation (one sigma) accuracy of the core flow measurement system. The random noise component of the TIP reading uncertainty was revised for single recirculation loop operation to reflect the operating plant test results discussed in Subsection 3.2. This revision results in a 3DMonicore effective TIP uncertainty of 9.1% for reload cores. The comparable two-loop 3DMonicore uncertainty value is 8.7%. The net effect of revised core flow and TIP uncertainties is an increase in the required SLMCPR of approximately 0.01 to 0.02. The actual increase may vary from cycle to cycle, since the two loop and single loop SLMCPR values are calculated each fuel cycle.

3.1 Core Flow Uncertainty 3.1.1 Core Flow Measurement During Single-Loop Operation The jet pump core flow measurement system is calibrated to measure core flow when both sets of jet pumps are in forward flow, the total core flow being the sum of the indicated loop flows. For SLO, however, the inactive jet pumps will be operating in reverse flow at active pump flows above approximately 40%. Therefore, the measured flow in the inactive jet pumps must be subtracted from the measured flow in the active loop to obtain the total core flow. The jet pump flow coefficient is different for reverse flow than for forward flow, and the measurement of reverse flow must be modified to account for this difference.

3-1

GE-NE-000-0027-5301a, Revision 1 NON-PROPRIETARY VERSION In SLO, the total core flow is derived from the following formula:

"(active coret = loop rC inactive loop, (3-1) indicated indicated flow) flow )

The coefficient "C" is a correction factor and "loop indicated flow" means the flow indicated by the jet pump "single-tap" loop flow summers and indicators, which are set up to indicate forward flow correctly. A value of 0.95 is used for the "C" coefficient, as the result of a conservative evaluation. ((

)) If more exact, less conservative core flow is required, special in-reactor calibration tests could be made. Such calibration tests would involve (I) calibrating core support plate AP versus core flow during one-pump and two-pump operation along a constant flow control line and (2) calculating the correct value of "C" based on the core support plate AP and the loop flow indicator readings. Until such tests are made, the 0.95 correction factor should be used for the core flow calculations during single-loop operation.

3.1.2 Core Flow Uncertainty Analysis The uncertainty analysis procedure used to establish the core flow uncertainty for SLO is essentially the same as for two-pump operation. For SLO, the total core flow can be expressed as follows (Figure 3-1):

WC = WA - WI (3-2) where:

Wc = total core flow 3-2

GE-NE-0000-0027-5301a, Revision I NON-PROPRIETARY VERSION WA= active loop flow WI = inactive loop true flow 3-3

GE-NE-0000-0027-5301a, Revision I NON-PROPRIETARY VERSION 3.2 TIP Reading Uncertainty Tests were performed at an operating BWR to establish the TIP noise uncertainty for single recirculation loop operation. The test was performed at a power level of 59.3% of rated with a single recirculation pump in operation (core flow of 46.3% of rated). A rotationally symmetric control rod pattern existed during the test. Results of the test are directly applicable to PNPS because the data collected are typical of random neutron, electronic and boiling noise during SLO for a BWR.

3.3 Calculational Options for SLMCPR All SLMCPR evaluations for both TLO and SLO are based on the original GETAB model for licensing applications that use absolute adaption regardless of what adaptive (or non-adaptive) methods are used for core monitoring. For core monitoring systems that utilize shape adaptive option, the original GETAB uncertainties and models related to the power distribution have been shown to yield conservative SLMCPR values.

However, core monitoring systems that utilize shape adaptive option may credit REVISED plant uncertainties and REDUCED power distribution uncertainties for both TLO and SLO. PNPS uses 3DMonicore (shape adaptive option) with REVISED plant uncertainties and REDUCED power distribution uncertainties. For 3DMonicore (shape adaptive option) the REVISED power distribution model (Reference 3) and REVISED 3-4

GE-NE-0000-0027-5301a, Revision 1 NON-PROPRIETARY VERSION plant uncertainties and REDUCED power distribution uncertainties (References 3 and 4) using the shape adaptive option have been reviewed and approved by the NRC (Reference 5).

ACTrVT Loop WWJG FLOW VmCnomziON WC = Total Core Flow, WA = Active Loop Flow, WI = Inactive Loop Flow Figure 3-1 ILLUSTRATION of SINGLE RECIRCULATION LOOP OPERATION FLOWS 3-5

GE-NE-0000-0027-5301a, Revision I NON-PROPRIETARY VERSION

4. TRANSIENT EVALUATIONS 4.1 Anticipated Operational Occurrences The reactor response to abnormal operational transients (AOOs) is for the most part independent of the source of core flow (i.e. whether the flow is provided in one or both recirculation loops). Thus, the consequences of an event initiated from SLO will be the same as the consequences of an event initiated from two-loop operation given the same initial power/flow conditions. The highest allowable core flow, with one active pump, is 52%. For A0Os, the maximum power that can be justified is 78.1% power corresponding to the MELLLA boundary, however, power is restricted to a maximum of 65% because of vibration limitations discussed in Section 7. ((

)) The results of the full power analyses in the cycle specific Supplemental Reload Licensing Report (SRLR) for Pilgrim bound the thermal and over-pressure consequences of these events during SLO. Cycle specific analysis verifies or establishes reactor operating limits as reported in the SRLR (Reference 10) that cover ICF, MELLLA, and Final Feedwater Temperature Reduction (FFWTR) operating domains, both FWH-OOS and end-of-cycle (EOC) FFWTR extension, in combination with one MSIV out-of-service (MSIV-OOS). Cycle specific analysis verifies or establishes ARTS curves as reported in the SRLR and are used in the determination of off-rated (power and flow dependent MCPR and LHGR) limits for equipment in-service and for supported EOOS options. However, because the OLMCPR and the Thermal and Mechanical Over-Power (TOP/MOP) are influenced by relative change (compared to initial values), to justify SLO, safety analyses must be reviewed for one pump operation. The fuel cladding integrity safety limit MCPR, flow decrease, flow increase, cold water injection, and pressurization transients are evaluated.

4-1

GE-NE-0000-0027-5301a, Revision I NON-PROPRIETARY VERSION Increased uncertainties in the total core flow and traversing in-core probe tip readings result in a small increase in the fuel cladding integrity SLMCPR for SLO. Therefore, when operating in SLO the plant SLMCPR is typically higher than when operating in TLO.

4.1.1 Core Flomw Decrease Events 1]

4.1.2 Core Flow Jncrease Events 4.1.3 Cold Water Injection Events 4-2

GE-N-OOO-002-531a, evison GE-NE-0000-0027-53Ola, Revision I NON-PROPRIETARY VERSION

))

4.1.4 Pressurization Events

[1

)) This is because at high core flows the transport time required for the feedwater to make its way to the core inlet is reduced, thereby increasing the core inlet subcooling more rapidly. The more rapid reactivity increase produced by the more rapid core inlet subcooling increase makes the transient more severe. In SLO, the maximum allowable core flow is 52.0% of rated core flow. ((

]i 4.1.5 Recirculation Pump Seizure Accident When operating with only one recirculation pump in operation, this active pump provides drive flow for the jet pumps in that loop. This bank of jet pumps is referred to as the active jet pumps. The other recirculation pump is out of service and the jet pumps in that loop are referred to as the inactive jet pumps. The inactive jet pumps have reverse flow.

In this condition of forward flow in the active loop and reverse flow in the inactive loop, the maximum core flow is limited to 52% of rated for Pilgrim. This reduced flow 4-3

GE-NE-0000-0027-5301a, Revision 1 NON-PROPRIETARY VERSION capability limits the maximum achievable core power to 78.1% as governed by the MELLLA boundary line.

The Recirculation Pump Seizure is characterized by a near instantaneous stoppage of the pump and associated pump flow. With this sudden stagnation of the drive flow, the active loop flow rapidly decreases. The resultant core flow decrease causes the core void fraction to increase which in turn causes a rapid decrease in core power. Because of the extremely low probability of occurrence for this event, a pump seizure is classified as an accident rather than an AOO. The purpose of evaluating a Pump Seizure event is to assure that the radiological consequences of the event are acceptable. The analysis was performed using the reload licensing methodology documented in References 2 and 6 and is based on the PNPS Cycle 14 core configuration..

Although the Cycle 14 core configuration was used, the analysis is considered cycle independent. The cycle-independent OLMCPR calculated for a recirculation pump seizure event when operating in Single Loop Operation (SLO) is 1.46. When adjusted for the off-rated power/flow conditions for SLO, (i.e., Kp of 1.082), this limit corresponds to a rated power/flow OLMCPR of 1.35, assuming a SLO SLMCPR

  • 1.12.

4-4

GE-NE-0000-0027-5301 a, Revision I NON-PROPRIETARY VERSION 4.1.6 Rod Withdrawal Error The implementation of the ARTS Program eliminated the flow-biased rod block monitor (RBM) trips and replaced them with power-dependent trips (Reference 7). The Rod Withdrawal Error (RWE) evaluations are independent of the source of the core flow (i.e.,

one recirculation loop versus two). Consequently, these evaluations are valid for both two-loop and single-loop operation.

The Flow Biased APRM rod block and scram system provides additional alarms, rod blocks and scram when power levels are grossly exceeded. Modification of the Flow Biased APRM function is therefore required to maintain the TLO rod block and scram versus power relationship when in SLO.

SLO results in backflow through 10 of the 20 jet pumps while flow is being supplied into the lower plenum from the 10 active jet pumps. The present rod block equation must be modified for use during SLO because of the backflow. Without correction the relationship between drive flow and core flow measurements may be non-conservative with respect to two-loop operation above approximately 35% core flow.

A procedure has been established for correcting the Flow Biased APRM rod block and scram equations to account for the discrepancy between actual flow and indicated flow in the active loop.

For normal TLO, the Flow Biased APRM Rod Block function is defined as:

RB=mWd +c For SLO, the above equation is revised to include the flow correction term:

RB=mWd + c -m AWd Or equal to RB = mWd + c' 4-5

GE-NE-0000-0027-5301a, Revision 1 NON-PROPRIETARY VERSION For example, a nominal trip setpoint of the TLO rod block equation given by:

RB = mWd + c = 0.66 *Wd + 58.5% (may include several similar flow segments)

Following the process described above, using a value of 10% to represent the SLO to TLO flow correction AWd, the SLO equation becomes:

RB= mWd+ c-mAWd= 0.66*Wd +51.9%

Where:

RB = Power at rod block in % of rated m = Flow reference slope, assumed 0.66 in example A Wd = Difference in drive flow between single and two loop operation at the same core flow. A value of 10% is verified to be bounding for PNPS based on plant specific recirculation system calculations. A more direct value can be derived using the procedure described in section 3.1.1.

'Al = Drive flow in % of rated c = Constant in flow-biased rod block equation, assumed.58.5 in example The Flow Biased APRM scram trip settings are corrected in the same manner as the Flow Biased APRM rod block setting. Therefore, the Flow Biased APRM flux scram settings are subject to the same procedural changes as the rod block setting discussed above.

4.2 Plant Operating Limit M1CPR The uncertainties used in the statistical analysis to determine the fuel cladding integrity SLMCPR are independent of whether coolant flow is provided by one or two recirculation pumps. However, the SLMCPR is impacted by uncertainties in core total flow and TIP readings. The effects of the increase in core flow and TIP uncertainties would be bounded by a 0.02 incremental increase in the required SLMCPR as shown in Reference 10. This increment will be confirmed or revised on a plant/cycle specific basis. For SLO the OLMCPR is corrected by this increment and the adjustment is described in References 7 and 10.

4-6

GE-NE-0000-0027-5301a, Revision I NON-PROPRIETARY VERSION With the implementation of ARTS, power- and flow-dependent MCPR limits are determined from the rated power and flow OLMCPR through the use of MCPR adjustment factors. The power and flow dependencies are based on different transient events. For any given power and flow condition, both the power- and flow-dependent MCPRs must be determined. The OLMCPR is, then, the more restrictive of these two MCPRs, as follows:

OLMCPR = Maximum[ MCPR(P), MCPR(F)]

The PNPS power-dependent MCPR adjustment factors discussed in Reference 7 are based on analyses of the limiting pressurization transients ((

)) initiated from various rated and off-rated initial power conditions. The power-dependent MCPR multiplier, Kp, was established based on plant specific analyses. The applicability of these multipliers was then confirmed by performing the PNPS Cycle 15 specific calculations. MCPR(P) is calculated as follows:

MCPR(P) = Kp

  • OLMCPR at 1OOP/IOOF However, prior to application of SLO operation the OLMCPR should be adjusted as described in 4.1.5 and in References 7 and 10.

PNPS flow dependent MCPR curves are based on generic recirculation slow flow run-out evaluations (Reference 7). This flow dependent limit, MCPR(F), assures that the SLMCPR will not be violated during a slow flow run-out event which does not terminate in a scram. As discussed in Section 4.1.2 these limits, as established for two-loop operation are conservatively applicable for SLO even with the SLMCPR increase of 0.02 (Reference 10) because of the significantly reduced recirculation run-out capability with only one recirculation pump in operation.

4-7

GE-NE-0000-0027-5301a, Revision 1 NON-PROPRIETARY VERSION Also, part of the ARTS improvement program is the power and flow dependent MAPLHGR/LHGR limits. These limits are developed to maintain the fuel thermal and mechanical design criteria under postulated transient events at off rated power/flow conditions. From this transient overpower protection viewpoint, the power and flow dependent MAPLHGR/LHGR limits are not impacted by SLO because their evaluations are independent of the source of the core flow.

The power and flow dependent ARTS Limits for Pilgrim Cycle 15 are provided in Figures 4.2-1 through 4.2-7. The power dependent limits below P-bypass are based on Cycle 15 specific analyses with a two-loop safety limit of 1.06. The flow dependent ARTS limits are based on Figure 3-4 and Figure 3-5 of Reference 7.

4.3 Overpressurization Protection The limiting event, the MSIV closure with flux scram in SLO, will yield peak pressures well below those calculated for rated power operation due to the reduced initial core power/flow condition during SLO.

4-8

GE-NE-0000-0027-5301a, Revision I NON-PROPRIETARY VERSION 3.25

> 50% Flow 3.00 The 32.5% power limits may be

. applied at powers >32.5% and S 45%

for conditions where the direct scram 2.75 is bypassed in this region 2.50

\l I , I ________S 2.25 El 5 50% Flow Operating Limit MCPR (P) = K(P) x Operating Limit MCPR (100)

-I 0 .For P c 25%: No Thermal Limits Required 10 go For 25% S P < 32.5% , > 50% Flow:

-_ 1.75 -. . -. . OLMCPR(P) = 2.95 + 0.0333(32.5% - P) a.

For 25% S P c 32.5% s 50% Flow:

OLMCPR(P) = 2.34 For 32.5% sP c 100%: K(P) a 1.0 0.00375(100% - P) 1.50 For Pk100%: K(P)= 1.00

.-0 ~ - -

1.25 . . I 1.00 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 Power (% Rated)

Figure 4.2-1 Power Dependent MCPR Limits (Turbine Bypass in Service)

(FFWTR included in limits) 4-9

GE-NE-0000-0027-5301a, Revision I NON-PROPRIETARY VERSION 4.25 4.00 \ _ _ _ _ _ _ _ _ _ I

)> 50% Flow 3.75

_ _ _ _ I f ! _ __._ __'

3.50 I

3.25 The 32.5% power limits may be applied at powers >32.5% and 1.

3.00 45% for conditions where the direct scram Is bypassed In this region 2.75

,, i l I I i I I,

\ >..............

El 2.5 Operating Limit MCPR (P) = K(P) K Operating Limit MCPR (100)

-J C- S 50% Flow For P c 25%: No Thermal Limits Required O 2.25 For25% S P < 32.5% , > 50% Flow:

' OLMCPR(P)

  • 3.25 + 0.125(32.5% - P) y 2.00 For 25% S P c 32.5% s S 50% Flow:
  • 2.51
  • 0.0453(32.5% - P)

For 32.5% s P < 100%: K(P) = 1.0 + 0.00375(100% - P) 1.75 i For P 2 100%: K(P) a 1.00 1.50 1.25 -* I 1.00 225 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 Power (% Rated)

Figure 4.2-2. Power Dependent MCPR Limits (Turbine Bypass Out-of-Service)

(FFWTR included in limits) 4-10

GE-NE-0000-0027-5SO3a, Revision I NON-PROPRIETARY VERSION 1.10 1.00 1

0.90 0.80 0.70 i The 32.5% power limits

. may be applied at powers >32.5% and < LHGR(P) = LHGRFAC(P)

  • LHGRstd I.- 4 55%for conditions LHGRstd=Rated LHGR limits 0.60 . _ where the direct scram is bypassed in this For Pc 25%: No Thermal Limits Required U.

region I For 25% s P < 32.5% , > 50% Flow:

LHGRFAC(P) = 0.44 + 0.008(P - 32.5%)

-J 0.50 For 25% s P < 32.5% , s 50% Flow:

S 50% Flow LHGRFAC(P) = 0.49 + 0.00533(P -

32.5%)

0.40 For 32.5% s P < 90%:

LHGRFAC(P) =0.90 + 0.004(P - 90%)

For 90% s P < 100%:

7> 50No Flow - LHGRFAC(P) =1.0 + 0.01(P - 100%)

0.30 For P k 100%: LHGRFAC(P) 1.00 I I I I I I I 0.20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 Power (% Rated)

Figure 4.2-3. Power Dependent LHGR Factors with Turbine Bypass In Service or Out of Service (FFWTR included in limits) 4-11

GE-NE-0000-0027-5301a, Revision I NON-PROPRIETARY VERSION 1.10 1.00 0.90 0.80

_____ I _____

0.70 I I I I 0-- he 32.5% power limits I  : may be applied at powers I

>32.5% and 1-45% for conditions where the direct /11_ MAPLHGR(P) =

0.60 scram is bypassed In this MAPFAC(P)xMAPLHGRstd region MAPLHGRstd=Rated MAPLHGR limits 0.

<:~

For P< 25%: No Thermal Limits Required

  • 0.50  ; *w-j- -- ,....'. I I I For 25% s P < 32.5% , > 50% Flow:

- 50%j Flow MAPFAC(P) = 0.44 + 0.008(P - 32.5%)

0.40 7 i For 25% S P < 32.5% , s 50% Flow:

MAPFAC(P) = 0.49 + 0.00533(P - 32.5%)

For32.5% s P < 90%:

F I MAPFAC(P) =0.90 + 0.004(P - 90%)

> 50% Flow I 0.30 . , .

For 90% s P < 100%:

i MAPFAC(P) =1.0 + 0.01(P - 100%)

I For P 2 100%: MAPFAC(P) = 1.00,*

I 0.20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 Power (% Rated)

Figure 4.2-4. Power Dependent MAPLHGR Factors with Turbine Bypass In Service or Out-Of-Service (FFWTR included in limits) 4-12

GE-NE-0000-0027-5301a, Revision I NON-PROPRIETARY VERSION 1.8 1.7 1.6 S.

0.

.8 C.,

zIL 1.5 t-1.

1.)

la 1.4 C_

EC, 0

1.3 1.2 1.1 20 30 40 50 60 70 80 90 100 110 120 Core Flow (%Rated)

Figure 4.2-5. Flow Dependent MCPR Limits (All Operating Flexibility Options) 4-13

GE-NE-0000-0027-5301a, Revision I NON-PROPRIETARY VERSION 1.1 1

Max Flow = 102.5% - lo i 107%

112% 1 0.9 1170%/

S-W 0.8 I W,

I LHGR(F) =LHGRFAC(F) X LHGRstd LHGRstd = STANDARD LHGR LIMITS 0:

I- 0.7 X7 (7

//Iii.=

LHGRFAC(F)-102.5% =The Minimum of EITHER 1.0 OR( 0.6784 x (WTI1 00) + 0.4861 )

0t-J LHGRFAC(F)-107% = The Minimum of EITHER 1.0 OR { 0.6758 x (WTI100) + OA574) 0C, 0.6 LHGRFAC(F)-112% =The Minimum of EITHER 1.0

=L OR { 0.6807 x (WT/1 00) + 0.42141

-J0.

LHGRFA C(F)-117% = The Minimum of EITHER 1.0 OR { 0.6886 x (WT/1 00) + 0.38281 WT = % Rated Core Flow 0.5 0.4 0.3 4 - .1. 4 30 40 50 60 70 80 90 100 110 Core Flow (% Rated)

Figure 4.2-6. Flow Dependent LHGR Factors (All Operating Flexibility Options) 4-14

GE-NE-0000-0027-5301a,. Revision 1 NON-PROPRIETARY VERSION 1.1 1

Max Flow= 102.5V /

107%

112%

0.9 o.

IL a- 0.8 2

0 MAPLHGR(F) MAPFAC(F) X MAPLHGRstd I-U MAPLHGRstd t STANDARD MAPLHGR LIMITS u0 0.7 MAPFAC(F)-102.5% The Minimum of EITHER 1.0 0 OR ( 0.6784 x (WT/100) + 0.4861 1

-j MAPFAC(F)-107% = The Minimum of EITHER 1.0 C, OR( 0.6758 x (WT/100) + 0.45741

_ MAPFAC(F)-112% - The Minimum of EITHER 1.0 cc a- OR ( 0.6807 x (WT/100) + 0.4214)

MAPFAC(F)-1 17% - The Minimum of EITHER 1.0 OR ( 0.6886 x (WT/1 00). 0.3828)

WT r % Rated Core Fbw 0.5 0.4 i I 0.3 3W o 40 50 60 70 80 90 100 110 Core Flow (%Rated)

Figure 4.2-7. Flow Dependent MAPLHGR Factors 4-15

,: & ,-.t , '"'

GE-NE-0000-0027-530la, Revision I NON-PROPRIETARY VERSION

5. STABILITY ANALYSIS 5.1 Phenomena 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. At forced circulation with only one recirculation loop in operation, the reactor core stability is influenced by the inactive recirculation loop characteristics. For low core flows occurring at minimum pump speed, the jet pumps on both recirculation loops will exhibit forward flow. At higher pump speeds the core flow is increased in SLO and the inactive jet pump forward flow decreases because the driving head across the inactive jet pumps decreases with increasing core flow. The reduced flow in the inactive loop reduces the resistance that the recirculation loops impose on reactor core flow perturbations, thereby adding a destabilizing effect. At the same time, the increased core flow results in a lower power-flow ratio, which is a stabilizing effect. These two countering effects result in 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 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, an increase in jet pump flow, core flow and neutron noise is observed. A cross flow is established in the annular downcomer region near the jet pump suction entrance caused by reverse flow of the inactive recirculation loop. This cross flow interacts with the jet pump suction flow of the active recirculation loop and increases the jet pump flow noise. This effect increases the total core flow noise, which tends to increase the neutron flux noise.

To determine if the increased noise condition represents reduced stability margin as SLO core flow was increased, an evaluation was performed which phenomenologically accounts for SLO effects on stability (Reference 11). The model predictions were 5-1

GE-NE-0000-0027-5301a, Revision I NON-PROPRIETARY VERSION initially compared with test data and showed very good agreement for both two-loop and single-loop test conditions. An evaluation was performed to determine the effect of reverse flow on stability during SLO. With increasing reverse flow, SLO exhibited similar decay ratios as two-loop operation. However, at core flow conditions with no reverse flow, SLO was slightly less stable. This is consistent with observed behavior in stability tests at operating BWRs (Reference 11).

In addition to the above analyses, the cross flow established during reverse flow conditions was simulated analytically and shown to cause an increase in the individual and total jet pump flow noise, which is consistent with test data (Reference I1). The results of these analyses and tests indicate that the stability characteristics are not significantly different from two-loop operation. At low core flow, SLO may be slightly less stable than two-loop operation, but as core flow is increased and reverse flow is established, the stability performance is similar. At higher core flow, with substantial reverse flow in the inactive recirculation loop, the effect of cross flow on the flow noise results in an increase in system noise (jet pump, core flow and neutron flux noise).

5.2 Compliance to Stability Criteria The stability compliance of GE fuel designs is documented in Reference 2.

The USNRC acceptance of thermal-hydraulic stability includes the condition that the plant has systems and procedures in place, supported by Technical Specifications, as appropriate, which provide adequate instability protection. The USNRC has accepted stability Option ElA (Reference 12), which is consistent with the operating practices at PNPS, and has accepted stability long-term solution Option ID (Reference 13), which may replace Option ElA at PNPS in the future.

5-2

GE-NE-0000-0027-5301aa, Revision I NON-PROPRIETARY VERSION Stability Option ElA Stability Option EIA provides Safety Limit MCPR protection by defining regions in the power-flow map where instabilities are credible (Reference 12). To prevent the violation of specified acceptable fuel design limits, the plant protection system provides an automatic scram upon entry into a defined exclusion region and a rod block upon entry into a defined restricted region, as well as additional Option EIA protection features. Since the stability performance for SLO is comparable to two-loop operation in the EIA regions, and only slightly decreased for the short duration that the inactive loop flow is low, the reactor scram and rod block stability setpoints are applicable to SLO without pump flow to core flow alignment (Reference 12, Appendix G).

Stability Option ID Stability Option ID provides MCPR Safety Limit protection by using the NRC approved methodology (References 13 and 14). To achieve this purpose the analysis demonstrates 1) an Exclusion Region on the power flow map where instabilities are expected to occur and 2) Detection and Suppression of power oscillations using the Flow-Biases APRM Neutron Flux scram. A core decay ratio < 0.80 at the Exclusion Region Boundary and assurance that core wide is the predominant mode of reactor instability are required for the Option I-D solution to be applicable (Reference 13). Since the stability performance for SLO is comparable to two-loop operation, and only slightly decreased for the short duration that the inactive loop flow is low, the Exclusion Region Boundary and Flow-Biased APRM scram setpoints are applicable to SLO with pump flow to core flow adjustment.

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GE-NE-0000-0027-5301a, Revision I NON-PROPRIETARY VERSION

6. ECCS PERFORMANCE ANALYSIS The Emergency Core Cooling System (ECCS) performance for single-loop operation (SLO) has been evaluated for PNPS in Reference 17, using the SAFER/GESTR-LOCA methodology. SLO analysis is performed for Design Basis Accident (DBA), since SLO wvill affect the DBA results more than the smaller breaks. 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 stored energy. ((

)) These assumptions are conservative and provide bounding results for the Design Basis Accident (DBA) under SLO.

)) Using the 0.8 PLHGR/MAPLHGR multipliers with the Appendix K assumptions yields PCT values well below the 10CFR50.46 PCT limit of 2200 "F.

Therefore, the calculated SLO PLHGR/MAPLHGR multipliers are conservative and assure that the SLO results satisfy the acceptance criteria of 10CFR50.46 and the NRC SER requirements for the SAFER application methodology. The ECCS LOCA analysis in Reference 17 for SLO is based on an initial power level of 77.6% of rated for nominal cases and 79% of rated for Appendix K cases. The initial core flow was 52% of rated.

These initial conditions are consistent with limits established for SLO in this report.))

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GE-NE-0000-0027-5301a, Revision I NON-PROPRIETARY VERSION

7. REACTOR INTERNAL VIBRATIONS The Reference I report documents the conclusions of the SLO internals vibration review for PNPS based on plant start-up data. The conclusion of that review was that the maximum core flow should be limited to 52% of rated core flow on the rated rod line at 65% core power, which corresponds approximately to 94% of rated pump speed. This pump speed and flow correspond approximately to the rated core power and flow for two pump operation conditions.

The limitation in core power and flow does not change with newer fuel designs as the thermal hydraulic environment of the internal components remain unchanged with new fuels.

The 65% core power limit is scaled to reflect the subsequent change in the Pilgrim licensing bases due to the increase in thermal power from the thermal power uncertainty reduction.

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GE-NE-0000-0027-5301a, Revision I NON-PROPRIETARY VERSION

8. PROPOSED DOCUMENTATION CHANGES The implementation of the SLO will require the changes to specific items in the PNPS technical specifications, core operating limits report and associated bases. The APRM flow biased rod block and scram values must be modified to reflect the difference in core flow to power relationship for SLO. The recommended changes to these items are shown below.

Table 8-1 RECOMMENDED DOCUMENTATION CHANGES Tech Specs Item Description of Change Safety Limit MCPR Modify any LCO and Action statements regarding the Safety Limit MCPR to include SLO values. Furthermore, the basis for the SLMCPR may need revision and the cycle specific values may be included in the Core Operating Limits Report.

Operating Limit MCPR Modify any LCO and Action statements regarding specific values of the Operating Limit MCPR to include the SLO values. Furthermore, the basis for the OLMCPR may need revision and the cycle specific OLMCPR for SLO may be included in the COLR.

MAPLHGR Modify specific values of the MAPLHGR to include the SLO multiplier. Furthermore, the basis for the MAPLHGR may need revision and the cycle specific MAPLHGR values for SLO may be included in the COLR.

LHGR Modify specific values of the LHGR to include the SLO multiplier.

Furthermore, the basis for the LHGR may need revision and the cycle specific LHGR values for SLO may be included in the COLR.

Flow Biased APRM Modify any scram trip equation to include SLO values.

Scram Flow Biased APRM Rod Modify any rod block trip equation to include SLO values.

Block SLO Condition Modify any reference to operation in SLO to include the applicable power and flow restrictions. Furthermore, incorporate the SLO operating map in applicable plant documents and procedures.

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GE-NE-0000-0027-5301ia, Revision I NON-PROPRIETARY VERSION

9. CONCLUSIONS The following paragraphs summarize the conclusions resulting from the SLO safety evaluation for PNPS.

The transient safety evaluation shows that the transient consequences for SLO are bounded by the full power TLO analyses results.

Increased uncertainties in the core total flow and TIP readings results in increased fuel cladding integrity SLMCPR value during SLO. An increase in rated and off-rated OLMCPR is required to account for the SLMCPR increase. The flow-biased APRM rod block and flux scram setpoints are adjusted for SLO.

Thermal-hydraulic stability was evaluated for adequacy with respect to General Design Criterion 12 (IOCFR50, Appendix A). It is shown that this stability criterion is satisfied during SLO. It is further shown that the increase in neutron noise associated with SLO is independent of system stability margin. The flow-biased APRM rod block and flux scram setpoints are adjusted for SLO.

To prevent potential control oscillations from occurring in the recirculation flow control system, the flow control should be in master manual for the duration of SLO.

The limiting MAPLHGR and LHGR reduction factor for SLO is calculated to be 0.80.

A limit on the maximum core thermal power and flow is imposed for SLO based on consideration of the reactor internals vibration.

Based on the above, it is concluded that the PNPS can operate in single loop mode without undue risk to the public health and safety.

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GE-NE-0000-0027-5301a, Revision I NON-PROPRIETARY VERSION

10. REFERENCES
1. GE-NE-187-10-0591, General Electric Company, "Pilgrim Nuclear Power Station Single Loop Operation", October 1991.
2. NEDE-2401 I-P-A-14, General Electric Company, "GESTAR II General Electric Standard Application for Reactor Fuel", June 2000.
3. NEDC-32601P-A, Methodology and Uncertainties for Safety Limit MCPR Evaluations, August 1999.
4. NEDC-32694P-A, Power Distribution Uncertainties for Safety Limit MCPR Evaluations, August 1999.
5. Letter F. Akstulewicz (NCR) to G. A. Watford (GE), Acceptance for Referencing of Licensing Topical Reports NEDC-32601P, "Methodology and Uncertainties for Safety Limit MCPR Evaluations"; NEDC-32694P, "Power Distribution Uncertainties for Safety Limit MCPR Evaluation"; and Amendment 25 to NEDE-2401 1-P-A on Cycle-Specific Safety Limit MCPR (TAC NOS. M97490, M99069 and M97491),

March 11, 1999.

6. NEDC-24154P-A, "Qualification of the One-Dimensional Core Transient Model (ODYN) for Boiling water Reactors", Volume I through 3 and Supplement 1, Revision 1 to Volume 4, February 2000.
7. NEDC-31312P, General Electric Company, "ARTS Improvement ProgramAnalysis for PilgrimNuclear Power Station (PNPS)" September 1987.
8. Not Used
9. NEDC-32306P, General Electric Company, "Maximum ExtendedLoad Line Limitfor PilgrimNuclearPower Station Reload 9 Cycle 10," March 1994.
10. General Electric Company, Supplemental Reload Licensing Report for Pilgrim NuclearPowerStation Reload 14 Cyclel5, 0000-0008-6613SRLR, Revision 1, March 2003.

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GE-NE-0000-0027-5301a, Revision I NON-PROPRIETARY VERSION

11. Letter, H. C. Pfefferlen (GE) to C.O. Thomas (NRC), "Submittal of Response to Stability Action Item from NRC Concerning Single-Loop Operation," September 1983.
12. NEDO-32339-A, "Reactor Stability Long-Term Solution: Enhanced Option I-A,"

April 1998.

13. NEDO-32465-A, General Electric Company, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," August 1996.
14. NEDO-31960-A and NEDO-31960-A Supplement 1, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.
15. Not used
16. Not used
17. "SAFER/GESTR-LOCA Loss of Coolant Accident Analysis for Pilgrim Nuclear Power Station," NEDC-31852-P, Rev. 2, Jan. 2003.

10-2

Hi. , I I :; .

Attachment 4 to 2.04.074 Entergy Nuclear Operations, In&.

Pilgrim Nuclear Power Plant Proposed Amendment to the Technical Specifications List of Regulatory Commitments:

The following table identified those actions committed to by Pilgrim in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

I1

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