ML092890336

Report of Facility Changes, Tests and Experiments and Summary of Commitment Changes

ML092890336
Person / Time
Site:	Palisades
Issue date:	10/15/2009
From:	Schwarz C Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML092890336 (17)

Text

Entergy Nuclear Operations, Inc.

Palisades Nuclear Plant

-- Entergy 27780 Blue Star Memorial Highway Covert, MI 49043 Tel 269 764 2000 Christopher J. Schwarz Site Vice President October 15, 2009 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Report of Facility Changes, Tests and Experiments and Summary of Commitment Changes Palisades Nuclear Plant Docket 50-255 License No. DPR-20

Dear Sir or Madam:

Entergy Nuclear Operations, Inc. (ENO) is providing the Palisades Nuclear Plant (PNP)

Report of Facility Changes, Tests and Experiments for the time period of September 30,2007, through September 30,2009. This report is being submitted in accordance with the requirements of 10 CFR 50.59(d)(2).

Attachment 1 contains a description of each change to the facility, and a summary of the evaluation performed for each change, in accordance with 10 CFR 50.59. There were no changes made to the facility in accordance with 10 CFR 72.48 during this period.

Attachment 2 contains summaries of regulatory commitment changes requiring NRC notification, from September 30,2007, through September 30,2009. The summaries include justification for each change per Nuclear Energy Institute (NEI) Guideline NEI 99-04, "Guidelines for Managing NRC Commitment Changes," and Regulatory Issue Summary 2000-17, "Managing Regulatory Commitments Made by Power Reactor Licensees to the NRC Staff."

Document Control Desk Page 2 This letter contains no new commitments.

Sincerely, cjs/jse Attachment(s): 1. Report of Changes, Tests, and Experiments

2. Summary of Commitment Changes CC Administrator, Region III, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades USNRC

ATTACHMENT 1 REPORT OF FACILTY CHANGES, TESTS, AND EXPERIMENTS 8 Pages Follow

Report of Changes, Tests, and Experiments Log Number: 07-0199 Document Number: Engineering Change 10306

Title:

ESS Suction Header Cross-Tie Operation Activity

Description:

This Engineering Change consists of the following activities:

(1) A change to Emergency Operating Procedure (EOP) Supplement 42 that directs operators to open the low pressure safety injection (LPSI) cross-tie valves MO-3190 and 3199 in the event of a loss of coolant accident that has progressed to the recirculation mode of emergency core cooling and in which one or more containment spray pumps and/or one or more high pressure safety injection pumps are not in operation. Opening the cross-tie valves prevents postulated emergency core cooling system back-leakage from entering the safety injection and refueling water (SIRW) tank following the loss of coolant accident. Back-leakage into the SIRW tank, which is located above the control room, affects control room habitability.

(2) A change to EOP Supplement 42 to provide alternate electrical control and power to the low pressure safety injection motor operated valves MO-3190 and MO-3199 such that both valves can be operated in the event of a loss of offsite power and failure of one train of emergency diesel generator power.

Summary of Evaluation:

The acceptability of the change to EOP Supplement 42 that directs operators to perform the cross-tie action is based on a failure modes and effects analysis (FMEA). The FMEA concludes that there is no detrimental design basis effect in performing the LPSI cross-tie or 480 VAC bus cross-tie. The potential exists for a steam bubble to form in the elevated LPSI cross-tie piping and to impact the transition to shutdown cooling. Provisions have been added to EOP Supplement 42 to address this potential situation.

The acceptability of the change to EOP Supplement 42 that provides alternate power for MO-3190 and/or MO-3199 via the cross-tie of load centers #11 and #12 is based on the load centers being designed to be cross-tied, and on existing guidance in System Operating Procedure (SOP) 30, upon which the new EOP Supplement 42 cross-tie steps are based. Operators are routinely trained on SOP-30, and the SOP-30 steps have been demonstrated to produce the desired result. The changes to EOP Supplement 42 include the steps necessary to effectively align the cross-tie and maintain transformer and emergency diesel generator loading within required limits.

None of the systems, structures, or components (SSCs) involved or the methods of controlling these SSCs can cause an accident, introduce the possibility of a change in the consequences of an accident, introduce new failure modes due to their failure, nor introduce any new accident or scenario not already bounded by the safety analysis. No fission product barriers are negatively impacted.

1

Changes, Tests, Experiments Log Number: 07-0213 Document Number: Engineering Change 10816

Title:

Radiological Calculations for Alternate Source Term Implementation Activity

Description:

This Engineering Change issues the radiological design basis calculations necessary to support the alternative source term (AST) radiological analysis methodology. The calculations:

(1) NAI-1149-014, Revision 4, "Palisades Design Basis AST MHAILOCA Radiological Analysis," and (2) NAI-1149-015, Revision 3, "Palisades Design Basis Control Rod Ejection AST Radiological Analysis,"

incorporate revised input assumptions. NAI-1149-014 uses a revised back-leakage scenario that increases the allowable back-leakage through the safety injection and refueling water tank discharge lines from 2.2 gpm to 3.6 gpm, decreases the allowable operator action to cross-tie the low pressure safety injection suction lines from 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 41 minutes after recirculation actuation signal (RAS) to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after RAS, and decreases the post-RAS leakage from 0.025 gpm to 0.00625 gpm. Both calculations use revised atmospheric dispersion factors and increased control room envelope unfiltered in-leakage.

Summary of

Description:

The revised input assumptions do not result in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR. The above design basis radiological calculations were revised to incorporate the changed input parameters, and conclude that the dose limits for the exclusion area boundary, the low population zone, and control room access and occupancy continue to be met. The calculated doses are either unchanged from the previous revisions of these calculations or are changed slightly from the previously calculated doses, but remain within 10% of the difference between the current calculated dose and the regulatory requirement.

Therefore, the proposed activity does not result in more than a minimal increase in the consequences of an accident or malfunction of an system, structure, or component (SSC) important to safety previously evaluated in the UFSAR. This activity also does not increase the probability of any FSAR described accidents or malfunctions, does not create the possibility of an accident of a different type, does not create the possibility of a malfunction of a SSC important to safety with a different result, and does not affect any design basis limits for fission product barriers.

2

Report of Changes, Tests, and Experiments Log Number: 08-001 Document Number: Engineering Change 10306

Title:

ESS Suction Header Cross-Tie Operation (Revision 1)

Activity

Description:

This Engineering Change, in part, changes operating procedures to direct operators to open the low pressure safety injection (LPSI) cross-tie valves MO-3190 and 3199 in the event of a loss of coolant accident that has progressed to the recirculation mode of emergency core cooling and in which one or more containment spray pumps and/or one or more high pressure safety injection pumps are not in operation. This revision of the 50.59 Evaluation conservatively increases the time specified in Emergency Operating Procedure (EOP) Supplement 42 that the operators have to align the LPSI cross-tie from 30 minutes to 45 minutes.

Summary of Evaluation:

The increase in time available for the operators to align the LPSI cross-tie does not affect the conclusions of the 50.59 Evaluation. The acceptability of the change to EOP Supplement 42 that directs operators to perform the cross-tie action continues to be based on a failure modes and effects analysis (FMEA). The FMEA concludes that there is no detrimental design basis effect in performing the LPSI cross-tie.

None of the involved systems, structures, or components (SSCs) or the methods of controlling these SSCs can cause an accident, introduce the possibility of a change in the consequences of an accident, introduce new failure modes due to their failure, nor introduce any new accident or scenario not already bounded by the safety analysis. No fission product barriers are negatively impacted.

3

Report of Changes, Tests, and Experiments Log Number: 08-002 Document Number: Engineering Change 8350

Title:

Replace Containment Spray Isolation Valves per GSI-191 Resolution (Revision 1)

Activity

Description:

This Engineering Change (EC) proposed to recover engineering safeguards system pumps' net positive suction head margin during post-recirculation actuation signal (RAS) operation by throttling containment spray flow to the containment spray headers.

The 50.59 Evaluation for this EC was revised to better align with guidance for performing 50.59 Evaluations contained in NEI 96-07, "Guidelines for 10 CFR 50.59 Implementation," Revision 1. Specifically, the conclusion in the responses to questions 2 and 4 were editorially reworded to align more closely with the wording of the Evaluation question being asked.

Summary of Evaluation:

The changes made in the revision of this 50.59 Evaluation have no impact on any of the conclusions of the Evaluation. None of the information provided in the Evaluation as a basis fornot requiring prior NRC approval is impacted by the editorial changes to the responses to questions 2 and 4.

The 50.59 Evaluation concluded the new post-RAS throttled design function for the valves is not an initiator of any accident and that no new failure modes that could initiate an accident are introduced. Additionally, there was no change in the likelihood of a malfunction since no new failure modes are created and the likelihood of malfunctions that exist for the existing failures will not be increased. The Evaluation also concluded that, at most, the dose impact of reducing spray system flow post-RAS was to only 5% of the containment iodine source term and this portion of the source term could not result in a more than minimal increase in the consequences of an accident. None of these conclusions are impacted by the editorial changes made to the 50.59 Evaluation.

4

of Changes, Tests, and Experiments Log Number: 08-0030 Document Number: Engineering Change 5885

Title:

Emergency Diesel Generator 1-1 and 1-1 Air Start Pressure Control Valve, Solenoid Valve, and Isolation Valve Upgrades Activity

Description:

This Engineering Change (EC) increases both the emergency diesel generator (EDG) starting air system supply volume and supply pressure. These changes will allow the air start motors to produce more horsepower and torque, thereby accelerating the EDG flywheel quicker during start attempts, and lowering the start times of the EDGs. The EC replaces the starting air system pressure control valves, and also changes the setpoints of several pressure switches and the EDG overcrank timers. One aspect of this EC was considered to be adverse in the accompanying 50.59 Screening and required review in a 50.59 Evaluation. This aspect concerned the design basis of the starting air system, which was revised from an air supply capacity adequate for approximately 40 seconds of cranking time to an air supply capacity adequate for 24.3 seconds of cranking time.

Summary of Evaluation:

The FSAR states that there is sufficient air in the air receiver tanks for approximately 40 seconds of EDG engine cranking time. Since the current overcrank timer setpoint is 35 seconds, the current design only allows for one unsuccessful start attempt since the remaining stored air after 35 seconds of cranking is not likely to be successful in starting the engine. Under this EC, the starting air system total available cranking time will be reduced to 24.3 seconds. This decrease in cranking time is accompanied by a higher air flow rate available through the larger pressure control valves being installed under the EC. To counter the decrease in net cranking time, the EC changes the amount of time the overcrank timers allow the engine to crank. Since a normal engine start only requires nominally three seconds of cranking time and the longest time that engine cranking would proceed for a normal start is nominally six seconds, setting the timers to ten seconds will ensure a normal start occurs as it currently does while preserving sufficient air pressure in the receiver tanks to allow a second start attempt. Industry experience indicates that EDGs either start promptly or completely fail to start; therefore allowing the engine to crank longer than ten seconds only wastes stored air.

Providing for an additional EDG start attempt significantly reduces the likelihood that the EDG will not successfully start. The shortened total cranking time, when combined with the significantly shortened overcrank timer setting of ten seconds, does not change the likelihood that the EDG will start successfully since the cranking time will still be more than sufficient for the design normal start ("ready-to-Ioad" status in ten seconds).

Therefore, the revised design basis, which reduces the total available cranking time but shortens the cranking time on a failed start attempt, and adds the potential for a second start attempt, does not result in more than a minimal increase in the likelihood of a malfunction of a SSC important to safety previously evaluated in the UFSAR.

5

Report of Changes, Tests, and Experiments Log Number: 08-0094 Document Number: Engineering Change 669

Title:

Modification Package for Cycle 19 Reload Design, Reload W Assemblies (Revision 2)

Activity

Description:

This Engineering Change documents the Cycle 19 reload. The Cycle 19 reload consists of 52 fresh batch "W" assemblies, four fresh batch "W" shield assemblies, 60 once-burnt "V" assemblies, 64 twice-burnt "U" assemblies, 20 thrice-burnt "T" assemblies, and four batch "U" shield assemblies. Cycle 19 is a low leakage design which will minimize fluence on the critical reactor vessel axial welds and vessel base metal. All assemblies to be loaded into the core have been manufactured by AREV A. Four "N" shield assemblies are being replaced with four fresh low enrichment "w" shield assemblies. Refueling activities will include core shuffle, and irradiated fuel inspections with the possibility of in-mast fuel sipping, ultrasonic inspections, and fuel repair and reconstitution.

This 50.59 Evaluation is revised to add justification for the use by Palisades of an NRC Safety Evaluation issued specifically for the St. Lucie plant concerning the use of three solid stainless steel rods in each of two fuel assemblies. NRC-endorsed guidance in NEI 96-07, "Guidelines for 10 CFR 50.59 Implementation," allows the adoption of Safety Evaluations issued at other sites as long as the terms and conditions for its use are met.

This revision of the 50.59 Evaluation documents that these terms and conditions are met.

Summary of Evaluation:

The 50.59 Evaluation was revised to document that Palisades has reviewed the terms and conditions for the use of the Safety Evaluation and has determined that all terms and conditions are met. The Evaluation compared the restrictions on the use of solid stainless steel rods in the St. Lucie Safety Evaluation with Palisades' core design and concluded that Palisades complies with all terms and conditions.

The revision of the 50.59 Evaluation had no impact on the conclusions in the remainder of the Evaluation.

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Report of Changes, Tests, and Experiments Log Number: 08-0442 Document Number: Engineering Change 10869

Title:

Palisades Nuclear Plant Containment Compartment Analysis Activity

Description:

This Engineering Change calculates the peak containment steam generator compartment pressures in support of Engineering Change (EC) 5000122407. This EC would install permanent grating in several floor opening spaces on the 649' elevation of containment.

Typically, several of the floor openings require solid plugs to be installed during refueling outages to increase the amount of floor space available for lay down areas. To reduce outage duration, permanent grating andlor floor plugs will be installed in several of these floor openings, thereby eliminating the time required to install the temporary floor plugs during refueling outages. In support of this EC, a compartmentalized containment model using GOTHIC 7.2a (Generation of Thermal-Hydraulic Information for Containments) software is used to model the steam generator compartment pressurization effects due to a large break loss of coolant accident with various grating configurations installed on the refueling floor.

Currently, the FSAR describes a methodology for containment compartment pressurization utilizing the COPATTA computer code. Due to the age of the current analysis, and unavailability ofthe COPATTA computer code, it is desirable to replace the COPATTA with a current state-of-the-art GOTHIC analysis. Under 10 CFR 50.59, a method of evaluation may be changed to another without prior NRC approval provided the NRC has approved the method for the intended purpose, and the terms, conditions and limitations concerning the use of the method are met. The NRC has approved the use of the GOTHIC computer code, specifically for containment sub-compartment pressurization analysis, as documented in a Safety Evaluation Report (SER) for River Bend Station, Unit 1 Facility Operating License Amendment No. 139.

Summary of Evaluation The change from COPATTA to the GOTHIC computer code is not considered a departure from the method of evaluation because GOTHIC has been approved for this application via a SER issued to River Bend Station with specific conditions and limitations, and Palisades use of the GOTHIC model complies with the River Bend Station SER terms, conditions, and limitations. Therefore, the use of the GOTHIC model does not result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses.

7

Report of Changes, Tests, and Experiments Log Number: 09-0398 Document Number: Engineering Change 15011

Title:

Revision of FSAR Description of Station Battery Recharge Time Activity

Description:

This Engineering Change generates revision 2 of EA-ELEC-AMP-025, "Battery Chargers ED-IS, 16, 17, 18 Output Current Required to Recharge Station Batteries ED-O 1 &

ED-02." The revised calculation determined that the calculated station battery recharge time using both chargers to charge a fully discharged battery is 11.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for both station batteries ED-Ol and ED-02. The calculated recharge time does not align with the FSAR description of a nine hour recharge time. However, actual test data shows that the fully discharged station batteries can be recharged within nine hours using two chargers, and confirms that the methodology used in the calculation is conservative. The recharge time formula used in the calculation predicts a longer recharge time than actual test data indicates. Although the FSAR description is met by use of the test data, it does not conform to the calculated recharge time of 11.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. This 50.59 Evaluation justifies changing the FSAR description of the station battery recharge time from nine hours to thirteen hours.

Summary of Evaluation:

There are no accidents evaluated in the FSAR that result from or are affected by an increase in the calculated recharge time of the station batteries. The proposed change to the FSAR description of the station battery charger performance characteristics does not involve an actual physical change to either the station batteries or battery chargers. There are no FSAR-described design functions dependent on the recharge time of the station batteries. The recharge time of the station batteries does not affect their ability to perform their credited design functions, the design function of the DC and preferred AC systems, or the design function of the battery chargers. In the event of a station blackout (i.e. the loss of all AC power), the station batteries are relied upon for a four hour coping period, but the battery chargers are not credited in this scenario. Therefore, this change does not increase the likelihood of a malfunction of equipment powered by the station DC system.

There are no accidents evaluated in the FSAR that are altered or affected by an increase in the calculated recharge time of the station batteries 8

ATTACHMENT 2

SUMMARY

OF COMMITMENT CHANGES 5 Pages Follow

ATTACHMENT 2

SUMMARY

OF COMMITMENT CHANGES COMMITMENT SOURCE COMMITMENT DESCRIPTION REVISED JUSTIFICATION NUMBER DOCUMENT/DATE COMMITMENT 1010980 Letter dated September 5, Perform vibration monitoring of Cancel Commitment This commitment was made in support of deletion of 1985 reactor vessel internals on a quarterly Technical Specification 4.13. The bases for deletion of basis per ASME Standard OM-05, Tech Spec 4.13 were as follows:

1981

1. The data analyzer used to perform the monitoring was damaged and deemed un-repairable with no replacement equipment readily available.

2. Surveillance data from the previous 10 years had been predictable and within established limits.

3. A modification that had been installed to increase core barrel clamping force had been proven effective over several years and fuel cycles.

Twenty-four years after deletion of the Tech Spec and commencement of the quarterly monitoring per ASME Standard OM-05, the vibration monitoring has provided no insightful data. Since implementation of the quarterly monitoring, no fuel failures or other component damage or degradation has been predicted or prevented by reactor internals vibration monitoring. Two fuel failures were postulated to have been due to misalignment between the core support barrel and upper guide structure, but in neither case was the vibration data beneficial in predicting the failures or determining the cause of the failures.

1012381 Response to Confirmatory 2-I.F.5 Station Batteries (Cmt #45): If Cancel Commitment Station battery surveillance is performed in accordance Action Letter and technical concurrence on our position with Technical Specifications. Commitment is not needed Information Request regarding these batteries can be to meet the obligation.

Pursuant to obtained from the manufacturer, 10CFR50.54(f) dated necessary surveillance procedures 5/21/86 and 11/20/86- will be changed prior to the next Additional Info Request service or discharge test.

1128187, SSFI Report 12/22/86, Request for Restart

ATTACHMENT :2

SUMMARY

OF COMMITMENT CHANGES COMMITMENT SOURCE COMMITMENT DESCRIPTION REVISED JUSTIFICATION

• NUMBER DOCUMENTIDATE COMMITMENT 1012733 Response to Confirmatory Attach 4 - Snubber inservice dates Cancel Commitment This commitment was completed for the 1988 refueling Action Letter and Info (Cmt #397): Issue procedure outage. Procedure EM-09-07 was replaced by an Request Pursuant to establishing proper test means before Entergy fleet procedure. Snubber lSI is performed in 10CFR50.54(f) dated performing tests during 1988 refueling accordance with ASME Section XI. The commitment is 5/21/86 and 11/20/86- outage. not needed to meet the obligation.

Additional Info Request 1/28/87, SSFI Report 12/22/86, Request for Restart 1013149 Response to NRC Attach 2 SFE - Circulating Water Cancel Commitment The cooling tower pumps (P-39A and P-39B) and 12/23/86 Additional System: Cooling Tower Pumps- circulating water system do not perform a safety function.

Information Request Normal plant operation verifies Trending of the cooling tower pump performance is Regarding 12/1/86 adequate circulating water flow - performed as part of the thermal performance monitoring Response to Confirmatory trend program will monitor program. In addition, cooling tower pump vibration and Action Letter of 5/21/86. performance during power escalation. oil monitoring is performed as part of the predictive Request to Restart maintenance program. The commitment is not necessary 3/26/87 for this non-safety related system performance monitoring.

1013169 Response to NRC Attach 2 SFE - Fire protection Cancel Commitment The fire protection jockey pump (P-13) does not perform 12/23/86 Additional system: Motor driven fire pump - No a safety function. Trending of the fire protection jockey Information Request periodic test to verify capacity of fire pump discharge pressure is performed by the system Regarding 12/1/86 system jockey pump. This will be engineer as part of the fire protection system Response to Confirmatory reviewed for inclusion in the performance monitoring plan. The commitment is not Action Letter of 5/21/86. equipment trend program. necessary for this non-safety related system performance Request to Restart monitoring.

3/26/87 2001221 Generic Letter 91-11 , Evaluate changes made to Sections Cancel Commitment Procedure SOP-30 contains the requirement and is January 21,1992 7.3.1 and 7.3.2 of SOP-30 as to how it maintained through reference to Technical Specification may affect our response to Generic 3.8.9 and 3.8.10. Commitment tracking is not required as Letter 91-11. SOP-30 specifies that modification would require a change to the procedure load centers 11 and 12 are and/or the Technical Specifications in accordance with administratively inoperable if the load normal site processes.

centers are cross-connected and powered from the redundant bus.

2001923 Response to Confirmatory Item 2 - Augmented Test Program- Cancel Commitment The electrohydraulic control (EHC) pumps do not perform Action Letter and Info Recommended periodic testing (Table a safety function. Trending of the EHC pump flow Request Pursuant to 4-1 XIV. Turbine generator-1) EHC performance is performed as part of the turbine generator 10CFR50.54(f), dated hydraulic. Pre-refueling. 2) Pump system performance monitoring plan. In addition, EHC May 21, 1986 and inspection/performance monitoring. pump vibration monitoring is performed as part of the November 20, 1986- predictive maintenance program. The commitment is not Additional Information necessary for this non-safety related system performance Request monitoring.

2

ATTACHMENT 2

SUMMARY

OF COMMITMENT CHANGES COMMITMENT SOURCE COMMITMENT DESCRIPTION REVISED JUSTIFICATION NUMBER DOCUMENT/DATE COMMITMENT 1015752 Response to GL 89-10 Complete all design basis reviews, Cancel Commitment Statement in MOV program document EM-28-04, "With June 28, 1989 analyses, verifications, tests and regard to dynamic testing, the closure letter to Generic inspections that have been instituted Letter 96-05 indicates that, as dynamic test data is in order to comply with items A obtained and evaluated, it is possible that there will be through H within five years or three sufficient justification to adjust the dynamic test refueling outages of the date of this periodicity. Plant and industry data support the one-time letter, whichever is later. verification (no periodic) dynamic testing of all safety related globe motor operated valves (MOVs) at Palisades In response to item D of Generic to confirm design basis MOV calculations. Specifically, Letter 89-10, the motor operated valve any calculations that reference values of valve factor (V F)

(MOV) program scheduled dynamic or rate of loading (ROL) can be validated with only one testing of all testable safety-related dynamic testing of an unbalanced globe MOV. All safety-MOVs to be done on a periodicity of related unbalanced globe MOVs have each been rd every 3 refueling outage, unless dynamically tested once. All test data for these MOVs 25% margin could be evaluated have shown that MOV calculation assumptions for ROL through testing. If 25% margin was and VF are conservative.

evaluated for the MOV, then dynamic th testing was deferred to every 5 Recent industry data, reviewed and approved by the refueling outage. NRC, has shown that for all globe valves meeting JOG screening criteria for Category A Status:

1) There is no age-related degradation in required thrust.

2) There is no service-related degradation in required thrust.

All Palisades safety related globe MOVs are identified as unbalanced discs, with no special characteristics, in a water system, with a system temp below 150 degrees F.

No capability to detect MOV degradation is lost in eliminating the periodic dynamic testing of these MOVs.

Periodic dynamic testing for all safety-related gate valve MOVs will also cease. PMs to dynamically test gate MOVs will remain active, as they will be required for a period of time following any internal valve repairs until a time when the valve's internal coefficient of friction (COF) has leveled off and has shown to remain constant.

Justification is as for globe MOVs above. All safety-related gate MOVs at Palisades have been sufficiently dynamically tested to validate MOV calculation assumptions. All calculation assumptions have been shown to be conservative.

3

ATTACHMENT 2

SUMMARY

OF COMMITMENT CHANGES COMMITMENT SOURCE COMMITMENT DESCRIPTION REVISED JUSTIFICATION NUMBER DOCUMENTIDATE COMMITMENT 2010247 Letter dated June 7,1995, Alloy 600 butt welds on both ends of Revise Commitment as The three welds were volumetrically examined before "Additional Information to the pressurizer surge line and the hot follows: Entergy will inspect they were returned to service in 1995. The two surge line Support Continued leg shutdown COOling line were the Alloy 600/82/182 welds were performance demonstration initiative (POI)

Operation for the Repaired subjected to MSIP stress pressurizer butt-welds examined in 2006, and the shutdown cooling weld was Pressurizer instrument improvement in 1995. Palisades subjected to MSIP in 1995 POI examined in 2007, comprising 100% of the treated Nozzle - PWSCC committed to volumetrically examine per MRP-139, Section 6.3.2 welds. No cracks were found. Therefore, MRP-139 Inspection of Mechanical the MSIP welds every other outage. for Category C weldments. allows the welds to be returned to an ASME Code Stress Improvement 100% of treated welds shall examination program or approved alternative. This Process (MSIP) Areas be volumetrically inspected reduces required examination frequency for each of the before returning to service. welds from every other outage (approximately 3-1/3 50% of welds in each years) to once per lSI interval (10 years) per ASME mitigation type group shall Section XI Table IWB-2500-1 for Category B-J pressure be volumetrically inspected retaining welds in piping, or in accordance with an once during the next 6 approved risk-informed inservice inspection plan. This years. If no cracks are will reduce time, dose, and expense associated with POI found during these examinations without compromising plant safety.

inspections, weldments shall be inspected according to a schedule consistent with the existing ASME Code examination program or an approved alternative.

2001019 Response to Generic Ensure that Palisades has procedures Cancel Commitment Evaluation was completed and it was concluded that the Letter 91-11 , dated that include time limitation and recommended actions of Generic Letter 91-11 were met.

July 18, 1991 surveillance requirements for vital NRC closed GL 91-11 for Palisades in letter dated instrument buses. If plant procedures May 18, 1992. Vital bus surveillance is per Technical do not include time limitations and Specifications. The commitment is not needed to meet surveillance requirements for this the obligation.

item, ensure that you have adequately evaluated the basis for such a position. The evaluation should address existing regulations and plant design basis.

4

ATTACHMENT 2

SUMMARY

OF COMMITMENT CHANGES COMMITMENT SOURCE COMMITMENT DESCRIPTION REVISED JUSTIFICATION

• NUMBER DOCUMENTIDATE COMMITMENT

• 2010689 Response To NRC Proceduralize the conduct of one Cancel Commitment This commitment was established to align the 10CFR50.54(f) Letter: SSFI-type inspection per NRC assessment schedule with the frequency of NRC SSFls, Request for Information Inspection Manual, or perform a which were typically performed every refueling cycle.

Pursuant to SSDC per internal Admin Procedures.

10CFR50.54(f) Regarding This inspection will be performed at SSFI-type inspections are no longer performed. Plant Adequacy and Availability least once per fuel cycle. design inspections are now being performed as of Design Bases component design basis inspections (CDBls), which are Information, performed triennially, per NRC Inspection Procedure February 6, 1997 71111.21. Entergy fleet procedure EN-U-104, "Self Assessment and Benchmark Process," requires that a self-assessment be performed prior to NRC inspections, such as CDBls. Due to this procedure, this commitment is not required to ensure that self assessments are performed prior to a CDBI.

In addition, performing a SSFI-type assessment every refueling cycle does not meet the current regulatory commitment threshold established in the NRC-endorsed "Guidelines for Managing NRC Commitment Changes,"

which states that commitments involve restoring compliance with a violated obligation within a certain date. The sites' 10 CFR 50.54(f) response, which originally made this commitment, did not involve a violated obligation. This commitment is no longer applicable.

2010973 Technical Specification Prior to implementation of the Cancel Commitment Technical Specification change was implemented in Change Request, proposed Condition B for LCO 3.7.10, 2001. Required actions are per Technical Specifications.

December 7,2000 Control Room Ventilation Filtration, Procedure changes are performed in accordance with Consumers Energy will have station processes.

procedures available describing compensatory measures to be taken in the event of any entry into LCO 3.7.10, Condition (3.

5