Text

Annual 10 CFR 50.59 Evaluation Report
	ML020790619
Person / Time
Site:	Wolf Creek
Issue date:	03/11/2002
From:	Muench R; Wolf Creek
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	ET 02-0012
	Download: ML020790619 (52)
	v • d • e

W LF CREEK rNUCLEAR OPERATING CORPORATION Richard. A. Muench Vice President Technical Services MAR 1 1 2002 ET 02-0012 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Subject:

Docket No. 50-482: Wolf Creek Generating Station Annual 50.59 Evaluation Report Gentlemen:

Station This letter transmits the Annual 50.59 Evaluation Report for Wolf Creek Generating I provides a (WCGS), which is being submitted pursuant to 10 CFR 50.59(d)(1)(2). Attachment Evaluation 50.59 summary of the evaluation results. Attachment II provides the WCGS Annual Report.

2001, and contains a This report covers the period from January 1, 2001, to December 31, approved by the WCGS summary of 50.59 evaluations performed during this period that were onsite review committee.

There are no commitments contained in this correspondence.

364-4034, or Mr.

If you have any questions concerning this report, please contact me at (620)

Karl A. (Tony) Harris at (620) 364-4038.

Very truly yours, Richard . Muench RAM/DMH/pb Attachments cc: J. N. Donohew (NRC), w/a D. N. Graves (NRC), w/a E. W. Merschoff (NRC), w/a Senior Resident Inspector (NRC), w/a P0. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831 An Equal Opportunity Employer M/F/HC/VET

Attachment I to ET 02-0012 Page 1 of 2 WOLF CREEK NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Docket No.: 50-482 Facility Operating License No.: NPF-42 ANNUAL 50.59 EVALUATION REPORT Report No.: 17 Reporting Period: January 1, 2001 through December 31, 2001

Attachment I to ET 02-0012 Page 2 of 2

SUMMARY

This report provides a brief description of changes, tests, and experiments performed at Wolf Creek Generating Station and evaluated pursuant to 10 CFR 50.59(c)(1). This report includes summaries of the associated 50.59 evaluations that were reviewed and found to be acceptable and by the Plant Safety Review Committee (PSRC) for the period beginning January 1, 2001 in accordance with the requirements of 10 ending December 31, 2001. This report is submitted CFR 50.59(d)(1)(2).

10 On March 13, 2001, Wolf Creek Nuclear Operating Corporation implemented the revised CFR 50.59 regulation. One evaluation has been generated and included in this report that was performed in accordance with the revised rule.

On the basis of these evaluations of changes:

"* There is no increase in the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Report (USAR).

"* There is no possibility that an accident or malfunction of equipment important to safety of a different type than any evaluated previously in the USAR may be created.

"* The margin of safety as defined in the basis for any Technical Specification is not reduced.

an Therefore, all items contained within this report have been determined either not to involve unreviewed safety question or require a license amendment.

Attachment IIto ET 02-0012 Page 1 of 49 Evaluation Number: 59 1999-0057 Revision: 0

Title:

Updated Safety Analysis Report (USAR) Revision to Clarify Inservice Inspection Examinations Activity Description This activity clarifies the USAR to reflect the performance of in-service inspection examinations with the unit operating. Performance of ISI examinations is non-destructive to equipment.

Approved plant processes control the plant configuration. Therefore, there are no expected effects to the plant as a result of this activity.

50.59 Evaluation Performance of ISI examinations while the plant is on-line has no impact on the probability of occurrence of an accident described in the USAR. Performance of ISI examinations is not a credible initiator for any accident, whether the exam is performed with the plant on-line or shut down. ISI examinations do not manipulate plant systems, structures or facilities. ISI examinations do not affect the initial conditions or accident dynamics as described in the USAR.

Therefore, there are no design bases accidents discussed or referenced in the that are impacted by this activity.

Whether an ISI exam is performed with the unit shut down or the unit on-line is inconsequential to radiological consequences of accidents described in the USAR. ISI exams are non destructive and are not credible initiators of accidents nor do they impact the accident dynamics as described in the USAR. Therefore, ISI exams have no impact on accident consequences.

There are no new credible accidents that could be created by the performance of ISI exams on line. ISI exams are non-destructive, and do not manipulate plant equipment or facilities in any fashion. If equipment must be removed from service to perform the ISI exam, the plant configuration is controlled in accordance with plant procedures. ISI exams are non-destructive, and do not direct equipment manipulation or facility alteration. Therefore, performance of ISI exams will not create an accident of a different type, regardless of whether the exams are performed with the unit shut down or on-line.

ISI exams are non-destructive and do not direct equipment manipulation therefore; the possibility of equipment malfunction as a result of ISI exams is not credible. Performance of IS[

exams is intended to reduce the probability of equipment malfunctions by detecting service induced defects prior to failure. Performance of the exams with the plant on-line or shut down does not impact that probability.

Whether an ISI exam is performed with the unit shut down or the unit on-line is inconsequential to radiological consequences of malfunctions of equipment important to safety as described in the USAR. ISI exams are non-destructive and are not credible initiators of equipment malfunction, nor do ISI exams impact malfunction dynamics as described in the USAR.

Therefore, ISI exams have no impact on radiological consequences due to equipment malfunction.

There are no new credible malfunctions of equipment important to safety as a result of ISI exams. ISI exams are non-destructive, and do not require system or equipment manipulation.

Attachment II to ET 02-0012 Page 2 of 49 The ASME Section Xl acceptance criteria for the specific ISl exams are not being altered by this change. Technical Specification 4.0.5 (and Bases 4.0.5) requires that the ISI program be administered in accordance with ASME Section X1. The acceptance criteria for indications found through ISI examination are located within Section XI and are not altered by this USAR change. Since no acceptance criteria are being altered or changed, the margin of safety is not impacted.

Attachment IIto ET 02-0012 Page 3 of 49 Evaluation Number: 59 1999-0157 Revision: 0

Title:

System Changes To Boron Recycle System, Liquid Radwaste System, And Secondary Liquid Waste System Activity Description The activity reflects changes to the Boron Recycle System, Liquid Radwaste System, and Secondary Liquid Waste System.

The Recycle (HE) Evaporator will be considered permanently removed from service.

The Secondary Waste (HF) Evaporator will also be considered permanently removed from service. The Recycle Evaporator, Secondary Waste Evaporator, and related equipment including the boron recycle radioactivity monitor, HERE0016, are permanently out of service. Improved technology installed over the past several years by various change packages and temporary modifications has made it more effective to process reactor coolant effluent as liquid radwaste rather than to recycle it. HERE001 6 monitors the flow path from the recycle evaporator condensate pump to the reactor makeup water storage tank. A high radiation signal from HERE0016 redirects flow back to the recycle evaporator feed demineralizers. The flow path from the recycle evaporator to the reactor makeup water storage tank reactor is and will continue to be isolated by locked closed valves. Because the recycle and secondary liquid waste evaporators are not operated, they place no heat load on the component cooling water and auxiliary steam systems. Since evaporation is not being used to process liquid radioactive wastes, the Primary Bottoms Tank is no longer the limiting source for iodine.

Instead it appears that maximum potential radioactive iodine inventory would be in the drum dryer installed by Temporary Modification 98-018-HB. Rather than perform a specific analysis for this temporary modification, a bounding analysis was performed and documented in calculation change notice AN-99-014-00-CN002. This bounding analysis assumes that all of the iodine entering the liquid radwaste system is contained in a hypothetical hold up tank with no iodine removal other than radioactive decay.

Chapter 15.7.2 is being revised to replace the analysis of a Primary Bottoms Tank failure with a bounding hypothetical tank failure.

"* The USAR stated function of the Boron Recycle System is being changed to indicate that RCS effluent (letdown and equipment drains) is processed as liquid radwaste rather than recycled into makeup water and boric acid. (The USAR was previously updated to reflect this change, but some sections were missed.) The stated function of the Boron Recycle System is being changed from "receive and recycle RCS effluents by separationinto boric acid and makeup water"to "receive reactorcoolant effluent for the purpose of storage until it can either be reused or processedby the Liquid Radwaste System".

"* The current method for processing clean tritiated liquid radwaste is by filtration, reverse osmosis, and demineralization rather than by processing through the Liquid Radwaste (HB) Evaporator. The Liquid Radwaste Evaporator remains available for use if needed.

(Again, the USAR has been previously updated to reflect some of these changes, but additional obsolete descriptions are being removed.) Statements which indicate tritiated waste streams are recycled into reactor grade makeup water and supplied to the reactor makeup water storage tank (RMWST) are being changed to indicate the RMWST is supplied primarily from the demineralized water system. The Liquid Radwaste (HB)

Evaporator, while available, is generally not used. The Liquid Radwaste System using the Liquid Radwaste Processing Skid normally processes the bulk of the radioactive

Attachment IIto ET 02-0012 Page 4 of 49 liquid discharged from the reactor coolant system for discharge. The affected descriptions and data tables are being revised accordingly. The only radionuclide discharged to the environment in significantly increased quantities as a result of these changes is tritium. USAR Section 15.7.3 refers to section 2.4.13.3 which describes an analysis of a Liquid Radwaste Tank failure. Tritium levels in ground water above the 10 CFR 20 release limit at the point it enters the cooling lake are justified using the argument: "the tritium concentration would be highly diluted by the uncontaminated waters of the cooling lake". Because tritium (half life 12.3 years) is present in the lake due to operation of the plant, it is not accurate, strictly speaking, for the USAR to call the lake uncontaminated in this context. Calculation AN-99-029 was performed to document an analysis considering both the higher normal tritium release rates due to not recycling letdown and release of the Refueling Water Storage Tank to ground water.

The conclusions of AN-99-029 indicate that the lake tritium concentration still remain 3

well below the 10 CFR 20, Appendix B, Table II, Column 2 limit of 1 x 10 ýCilml even if the Refueling Water Storage Tank is released to ground water after the lake has attained equilibrium tritium levels due to continuous release of all the tritium released to the reactor coolant system during normal operation. Note that the USAR incorrectly states that the 10 CFR 20, Appendix B, Table II, Column 2 limit for tritium is 3 x 10 pCi/ml. The USAR is being revised accordingly. The term "Liquid Radwaste Demineralizer Skid" is being replaced with "Liquid Radwaste Processing Skid" and the description is being revised to indicate that it may consist of a series of components employing various processes which may include filtration, reverse osmosis, and/or demineralization.

" Heat tracing is made permanently out of service if it is associated with piping that is permanently out of service.

" Actual release data is contained in Annual Radioactive Effluent Release Reports filed with the NRC in accordance with Offsite Dose Calculation Manual (ODCM) requirements. USAR statements indicating that source terms are calculated using the GALE code are being removed. The GALE code is no longer used for updating source terms. Chapter 11 tables supporting chapter 15 and Chapter 2 accident analyses will continue to be maintained and are not considered historical.

The most significant radiological effects are:

"* Since evaporation is not being used to process liquid radioactive wastes, the Primary Bottoms Tank is no longer the limiting source of iodine in the radwaste tank rupture analysis. A bounding analysis was performed and documented. This analysis shows an increase in the calculated Exclusion Area Boundary thyroid dose from 0.124 rem to 1.49. This increase is much less than 10% of the available margin and therefore, is not considered an increase in consequences.

"* Since RCS effluent is processed as liquid radwaste, rather than recycled, more tritium is released to the cooling lake. A calculation was prepared to evaluate and justify this change. This calculation shows that cooling lake tritium concentration will remain below 3 x 105 pCi/ml which is well below the applicable 10 CFR 20 limit of 1 x 10-3 jCi/ml.

"* Minimal hardware changes are being made. These changes consist of lifting leads, pulling fuses, and removal of the Recycle Evaporator Concentrates Pump.

50.59 Evaluation USAR described accidents considered in this activity are:

Attachment IIto ET 02-0012 Page 5 of 49

1. Airborne release due to a waste gas decay tank failure (USAR section 15.7.1).

2. Airborne release due to a liquid radwaste tank failure (USAR section 15.7.2). Two liquid tanks were analyzed and discussed in section 15.7.2.

a) Boron recycle hold-up tank (RHUT) (highest total inventory of radioactivity) b) Primary evaporator bottoms tank (highest inventory of halogens)

3. Liquid release due to liquid tank failures (USAR section 15.7.3 which refers to USAR section 2.4.13.3). Three liquid tanks were analyzed and discussed in section 2.4.13.3.

a) Primary spent resin storage tank b) Boron recycle hold-up tank (RHUT) c) Refueling water storage tank (RWST)

The relevant accidents identified are tank failures. None of the changes being evaluated increase the pressure or temperature, or otherwise adversely challenge the integrity of any tank identified above. Therefore, there is no increase in the probability of an accident previously evaluated in the USAR.

The source of all of the radioactivity in all of the tanks identified in the analysis above is the reactor coolant system (RCS). The changes described above under headings 1-a and 1-b actually tend to decrease RCS radioactivity levels, since make-up water and boric acid do not have the low levels of residual activity that would exist in the recycled products. The affects of the process changes on each relevant event will be discussed individually.

Waste Gas Decay Tank Failure (USAR section 15.7.1).

Waste gas decay tanks are used to permit the decay of fission product gases as a means of minimizing the release of radioactive materials to the atmosphere. The principal radioactive components of the waste gas decay tanks are the noble gases krypton and xenon, the particulate daughters of some of the krypton and xenon isotopes, and trace quantities of halogens. The main source of these is the RCS via volume control tank (VCT) purging. RCS radioactivity levels are not being increased, therefore the main source of the radioactivity in the tanks is not increased. Smaller sources of gasses include vents from the recycle evaporator gas stripper, the reactor coolant drain tank, the pressurizer relief tank, and the recycle holdup tanks. The only one of these sources affected is the vent from the recycle evaporator gas stripper, which is no longer used. Therefore the total amount of radioactive material assumed to be in a failed waste gas decay tank is still bounding and radiological consequences are not increased.

Airborne Release Due To A Liquid Radwaste Tank Failure (USAR section 15.7.2).

1. Boron Recycle Hold-Up Tank The USAR states this tank was selected for evaluation because of all of the tanks listed in USAR Table 11.1-6 it contains the maximum total inventory of radioactivity. All of the noble gases and 10% of the iodine in the tank are assumed to escape and become airborne. The amount of radioactivity assumed for the accident analysis is based on an assumed 1% fuel defects and the process flow depicted in USAR Figures 11.1A-2 sheets 1 and 2, which show an RHUT receiving RCS effluent at the rate of 2140 gallons per day (1840 gpd letdown shim bleed and 300 gpd from the reactor coolant drain tank). Although

Attachment IIto ET 02-0012 Page 6 of 49 USAR Figure 11.1A-2 sheet 2 is affected and is being revised, all changes are downstream of the RHUT. Therefore the source terms used for evaluating an RHUT rupture are still bounding and radiological consequences are not increased.

2. Primary Evaporator Bottoms Tank The USAR states this tank was selected for evaluation because of all of the tanks listed in USAR Table 11.1-6 it contains the maximum total inventory of radioactive iodine. (Some demineralizers actually contain more, but it is considered fixed in the resin and not a credible airborne source.) Since evaporation is not being used to process liquid radioactive wastes, the Primary Bottoms Tank is no longer the limiting source. Instead it appears that maximum potential radioactive iodine inventory would be in the drum dryer installed by Temporary Modification 98-018-HB. Rather than perform a specific analysis for this temporary modification, a bounding analysis was performed and documented in calculation change notice AN-99-014-000-CN002. This bounding analysis assumes that all of the iodine entering the liquid radwaste system is contained in a hypothetical hold up tank with no iodine removal other than radioactive decay. The estimated Exclusion Area Boundary 0-2 hour thyroid dose due to releasing the equilibrium iodine curie content of this hypothetical tank is less than 1.5 rem. IE Circular 80-018 "10 CFR 50.59 Safety Evaluations For Changes To Radioactive Waste Treatment Systems" states that "radiological consequences of unexpected and uncontrolled gaseous releases of radioactivity stored in a waste system must be a small fraction of the 10 CFR 100 guidelines, i.e., less than 1.5 rem thyroid."

In addition, the USAR is revised to describe this scenario as having dose consequences that are a small fraction of the 10 CFR 100 guidelines (10 percent or 2.5 rem whole body and 30 rem thyroid). The 10 CFR 100 thyroid limit is 300 rem. An increase in dose of less than 10% of the available margin is generally not considered an increase in consequences.

Liquid Release Due To A Liquid Tank Failures (USAR section 15.7.3 / 2.4.13.3).

Radioactive liquids from the plant are postulated to enter the ground water as a result of the accidental rupture of tanks. Concentrations in ground water at the point it would enter the cooling lake (or the tributary to Wolf Creek in the event the cooling lake is drained) are calculated and compared to 10 CFR 20 limits.

The rupture of each of three tanks is postulated as a separate isolated event. The tanks were selected based on radioisotopes of relatively long half-lives of concern to human health, Sr-90, Cs-137, Co-60, and H-3.

Highest curie contents for Sr-90, Cs-137, and Co-60 are in the Primary Spent Resin Storage Tank. The highest concentration of H-3 is in the Boron Recycle Holdup Tank, while the greatest curie content of H-3 is in the Refueling Water Storage Tank.

USAR Table 2.4-35 gives each tanks curie content for the important radionuclides used in the analysis.

1. Primary Spent Resin Storaqe Tank (Sr-90, Cs-137, and Co-60)

Bechtel mechanical calculation HB-21 determined the curie content for this tank as given in USAR Table 11.1-6 sheet 17 and USAR Table 2.4-35. These values were based on the primary coolant activities determined by Bechtel mechanical

Attachment II to ET 02-0012 Page 7 of 49 calculation HB-19 given in USAR Tablel1.1-1 and the process flow paths, flow rates, and decontamination factors shown on USAR Figure 11 .1A-1 sheets 1 &

2. A careful review of these shows that virtually all of the calculated activity is from the CVCS Mixed Bed Demineralizers, the CVCS Cation Bed Demineralizer, and the Recycle Evaporator Demineralizer. Since the changes being evaluated are downstream of these demineralizers, the source terms used for evaluating a Primary Spent Resin Storage Tank failure are unaffected and radiological consequences are not increased.

2. Boron Recycle Hold-Up Tank; and

3. Refueling Water Storage Tank (H-3)

Since the Reactor Water Make Up Tank is being supplied exclusively from uncontaminated sources, and not from the Recycle Evaporator, the tritium levels in the RCS, and therefore the tritium levels in these tanks, would be decreased due to the changes described above under heading 1-a. This results in a reduction of the radiological consequences. However, the USAR analysis justifies tritium levels in ground water above the 10 CFR 20 release limit at the point it enters the cooling lake using the argument: "the tritium concentration would be highly diluted by the uncontaminatedwaters of the cooling lake".

Because tritium (half life 12.3 years) builds up in the lake over the life of the plant, it is not accurate, strictly speaking, for the USAR to call the lake uncontaminated in this context. Calculation AN-99-029 was performed to document an analysis considering both the higher normal tritium release rates due to not recycling letdown and release of the refueling water storage tank to ground water as described in USAR section 2.4.13.3. IE Circular 80-018 "10 CFR 50.59 Safety Evaluations For Changes To Radioactive Waste Treatment Systems" states that radiological consequences of unexpected and uncontrolled liquid releases of radioactivity stored in a waste system must be less than the radionuclide concentrations of 10 CFR 20, Appendix B, Table II, Column 2 for liquid releases at the nearest water supplies. The conclusions of AN-99-029 indicate that the lake tritium concentration still remain well below the 10 CFR 20, Appendix B, Table II, Column 2 limit of 1 x 10-3 gCi/ml even if the refueling water storage tank is released to ground water after the lake has attained equilibrium tritium levels due to continuous release of all the tritium released to the reactor coolant system during normal operation.

No new types of credible accidents have been identified.

The equipment and systems being evaluated are not considered important to safety. USAR Table 11.2-2 identifies tank uncontrolled release protection provisions. The Recycle Evaporator is listed on this table. Since it will now be permanently isolated the potential for an uncontrolled release from it will be reduced. No new credible malfunctions have been identified.

USAR Table 11.2-2, "Tank Uncontrolled Release Protection Provisions", does list and discuss the Recycle Evaporator. Since the Recycle Evaporator will be isolated, the probability of an evaporator malfunction resulting in an uncontrolled release will be reduced.

Attachment II to ET 02-0012 Page 8 of 49 USAR Table, 11.2-2, "Tank Uncontrolled Release Protection Provisions", does list and discuss the Recycle Evaporator. Since the Recycle Evaporator will no longer be used it will contain less radioactivity. Therefore, radiological consequences of an equipment malfunction will not be increased.

No new or unique procedures, processes, etc. are involved. No new types of credible accidents have been identified.

The changes all involve equipment and systems in the Radwaste building, which are physically distant and separated from equipment important to safety. Therefore, the possibility of a different type of a malfunction of equipment important to safety is not created.

No acceptance limits have been identified. Since no acceptance limits were identified, no related margins of safety have been defined.

Attachment II to ET 02-0012 Page 9 of 49 Evaluation Number: 59 1999-0161 Revision: 0

Title:

Piping Replacement Due to Flow Accelerated Corrosion Activity Description This activity provides guidelines for the modification/replacement of Main Turbine lines with a low alloy steel (2 1/4 Cr - 1 Moly) This modification will mitigate abnormal pipe-wall thinning due to Flow Accelerated Corrosion (FAC) and improve piping resistance to FAC.

The proposed pipe replacement did not change the cross sectional properties (section modulus, moment of inertia), or the geometric configuration. The mechanical properties such as tensile strength and code allowable stresses will remain unchanged. The lower yield strength and higher Young's Modulus is judged to have insignificant impact on the original analysis.

Therefore, the change does not adversely affect the existing safety margins or structural integrity of the affected piping system. The piping stresses will remain acceptable within code allowables.

The proposed activity will restore degraded sections of the affected piping system, to perform its original design intent. The proposed replacement does not involve or affect any safety related system or component. All system functions will continue to be performed as designed.

The proposed replacement does not adversely affect any system, component or procedures required to mitigate the consequences of an accident previously evaluated in the USAR. The proposed change will restore a degraded section of the affected piping system, to its original design configuration (piping geometry, cross section, support location, fittings).

50.59 Evaluation Since the proposed change will restore a degraded section of the affected piping system, to its original design configuration (piping geometry, cross section, support location, fittings), no accidents are identified as being associated with this change. Therefore, the probability of occurrence of an accident is not increased and radiological consequences are not increased.

Since the proposed change will restore a degraded section of the affected piping system, to its original design configuration (piping geometry, cross section, support location, fittings), no accidents could be created.

Ductile fracture, corrosion, erosion/corrosion, loss of mechanical properties, excess strain, mechanical creep etc., are credible failure modes for which the proposed piping replacement has been evaluated, through a critical characteristics comparison to the existent piping system design. Based on the evaluation, it was concluded that a new credible failure mode is not introduced. There are no malfunctions of equipment important to safety identified. Therefore, the probability of occurrence of an equipment malfunction is not increased and the radiological consequences are not increased.

Since, the proposed change will restore a degraded section of the affected piping system, to perform its original design intent, no acceptance limits which could affect the basis for any Technical specification are identified. Therefore, the margin of safety has not been decreased.

Attachment IIto ET 02-0012 Page 10 of 49 Evaluation Number: 59 1999-0162 Revision: 0

Title:

Piping Replacement Due to Flow Accelerated Corrosion Activity Description This activity provides guidelines for the modification/replacement of Main Turbine lines with a low alloy steel (2 1/4 Cr - 1 Moly). This modification will mitigate abnormal pipe-wall thinning due to Flow Accelerated Corrosion (FAC) and improve piping resistance to FAC.

The proposed pipe replacement did not change the cross sectional properties (section modulus, moment of inertia), or the geometric configuration. The mechanical properties such as tensile strength and code allowable stresses will remain unchanged. The lower yield strength and higher Young's Modulus is judged to have insignificant impact on the original analysis.

Therefore, the change does not adversely affect the existing safety margins or structural integrity of the affected piping system. The piping stresses will remain acceptable within code allowables.

The proposed activity will restore a degraded section of the affected piping system, to perform its original design intent. The proposed replacement does not involve or affect any safety related system or component. All system functions will continue to be performed as designed.

The proposed replacement does not adversely affect any system, component or procedures required to mitigate the consequences of an accident previously evaluated in the USAR. The proposed change will restore a degraded section of the affected piping system, to its original design configuration (piping geometry, cross section, support location, fittings).

50.59 Evaluation Since the proposed change will restore a degraded section of the affected piping system, to its original design configuration (piping geometry, cross section, support location, fittings), no accidents are identified as being associated with this change. Therefore, the probability of occurrence of an accident is not increased and radiological consequences are not increased.

Since the proposed change will restore a degraded section of the affected piping system, to its original design configuration (piping geometry, cross section, support location, fittings), no accidents could be created.

Ductile fracture, corrosion, erosion/corrosion, loss of mechanical properties, excess strain, mechanical creep etc., are credible failure modes for which the proposed piping replacement has been evaluated, through a critical characteristics comparison to the existent piping system design. Based on the evaluation, it was concluded that a new credible failure mode is not introduced. There are no malfunctions of equipment important to safety identified. Therefore, the probability of occurrence of an equipment malfunction is not increased and the radiological consequences are not increased.

Since, the proposed change will restore a degraded section of the affected piping system, to perform its original design intent, no acceptance limits which could affect the basis for any Technical specification are identified. Therefore, the margin of safety has not been decreased.

Attachment IIto ET 02-0012 Page 11 of 49 Evaluation Number: 59 2000-0011 Revision: 0

Title:

Updated Safety Analysis Report Change to Address Plant Safety Review Committee Responsibilities Activity Description This activity changes the wording of two sections in the Updated Safety Analysis Report (USAR) that discuss Plant Safety Review Committee (PSRC) responsibility. These sections did not contain specific wording do describe the types of information that require PSRC review.

Therefore, it was left to interpretation what these sections of the USAR referred to. Due to the generic in nature it could be understood that the review requirements included a greater set of information than was in the PSRC procedure. The sections are being revised to describe the actual intent of the review requirements in terms of the documentation associated to the specific regulations and or programs.

50.59 Evaluation There are no design bases accidents discussed or referenced in the USAR that take credit for PSRC activities. Therefore the probability of an accident or the consequences of an accident have not been increased This is an administrative change to PSRC review responsibilities. A change to these administrative sections will not create any credible accidents.

This change is administrative in nature. It will not affect equipment. Therefore, there will be no increase in equipment malfunctions or consequences of malfunctions. No new equipment malfunctions will be created.

There are no acceptance limits documented that this change would affect. Therefore, the margin of safety has not been reduced.

Attachment IIto ET 02-0012 Page 12 of 49 Evaluation Number: 59 2000-0049 Revision: 0

Title:

Changes to Fuel Elevator and Fuel Transfer System Activity Description The proposed activity will correct the following USAR inaccuracies:

" USAR Figure 9.1-11, New Fuel Elevator, will be revised to accurately reflect the physical arrangement of the new fuel elevator drive system. This figure incorrectly shows the hoist mounted to the north of the elevator, with cables and sheave assemblies supporting this arrangement. The actual configuration is with the hoist to the east of the elevator. The actual orientation of the drive system is per design (reference M-716-00167-W07, Instruction Manual for New Fuel Elevator.

" USAR Section 9.1.4.2.2, Fuel Transfer Tube and Associated Components, Item b, Lifting arm (transfer car position), will be revised to accurately reflect the physical location of the mechanical latch device for the mechanical interlock which allows fuel transfer system lifting arm operation only when the transfer car is at the end of its travel. The USAR incorrectly states that the mechanical latch device is located on the lifting arm. The actual location in on the transfer car per design (reference M-716-00158-W16, Instruction Manual for Fuel Transfer System).

For each of these items, this 50.59 evaluation credits the design process for ensuring that the new fuel elevator drive system and the fuel transfer system mechanical interlock meets all appropriate design requirements.

50.59 Evaluation Since the affected equipment handles irradiated fuel assemblies, the fuel handling accident is reviewed. No other design basis accident is applicable. The actual condition of the new fuel elevator and the fuel transfer system are per design requirements. Therefore, no credible accidents are identified as affected.

Neither the actual orientation of the new fuel elevator drive system nor the location of the fuel transfer system mechanical interlock increases the probability of dropping design loads.

Therefore, the proposed change does not increase the probability of dropping a fuel assembly.

The fuel handling accident in the fuel building assumes that all fuel rods in the dropped assembly are ruptured. Since the new fuel elevator and the fuel transfer system can handle a maximum of one fuel assembly, there is no potential for increasing the radiological consequences of a fuel handling accident.

For the same reasons the activity does not create accidents of a different type than any previously evaluated.

The only equipment important to safety that interfaces with the new fuel elevator and the fuel transfer system is nuclear fuel. The actual orientation of the new fuel elevator drive system and the location of the fuel transfer system mechanical interlock is per design requirements. No credible malfunctions of equipment important to safety are identified.

The only safety related equipment that the new fuel elevator and the fuel transfer system potentially affects is nuclear fuel. The actual orientation of the new fuel elevator drive system

Attachment II to ET 02-0012 Page 13 of 49 and the location of the fuel transfer system mechanical interlock has no impact on fuel assemblies, thus does not increase the probability of occurrence of a malfunction of equipment important to safety.

Either the new fuel elevator or the fuel transfer system may contain an irradiated fuel assembly.

The actual orientation of the new fuel elevator drive system and the location of the fuel transfer system mechanical interlock do not affect the assembly in any manner, therefore the activity does not increase the radiological consequences of a malfunction of equipment important to safety.

For the same reasons, the activity does not create the possibility of a different type of malfunction of equipment important to safety than any previously evaluated.

There are no acceptance limits in licensing basis documents are identified. Therefore, the change does not reduce the margin of safety.

1 to ET 02-0012 Page 14 of 49 Evaluation Number: 59 2000-0058 Revision: 0

Title:

Replacement of Obsolete Instrumentation and Controls Activity Description This activity replaces obsolete equipment in the Condensate Demineralizer (AK) system that is no longer supported by the manufacturers and resolves several technical problems in the system. The following changes are included in the scope of the activity:

"* Instruments and controls in panel AK182 will be replaced with a programmable logic controller (PLC) and a panel mounted industrial computer.

"* Conductivity and pH analyzers in panel AK182 will be replaced with new Rosemount analyzers.

"* The annunciators in panels HF133 and AK182 will be replaced with new Panalarm Series 90 annunciators.

"* Flow transmitters will be replaced with Rosemount differential pressure transmitters with LCD displays (0-100% flow indication).

"* The three valve manifolds used to isolate the flow transmitters will be replaced with five valve manifolds.

"* A pulsation dampener and isolation valves will be added to the Acid Regeneration Pump discharge header to reduce pressure spikes which cause the Acid Concentration Analyzer, AKCIS0312, to false alarm and the pumps to trip.

"* A temperature sensor and transmitter will be installed in the Sluice Water supply pipe upstream of valve AKPCV031 1.

"* A digital display will be installed in panel AK182 to display the Sluice Water temperature.

"* The circuitry for relays in panel AK 82 will be modified so that the relays can be reset without removing the interior panel cover to depress a manual reset button.

"* A latching relay will be installed in each circuit between hand-switches and the corresponding solenoid valves to prevent the solenoid valves from vibrating out of position.

The changes to be implemented within this activity will reduce system maintenance and improve operability. Overall system reliability is also expected to increase. However, the new PLC to be installed in panel AK182 may exhibit greater vulnerability to the nuclear power plant EMI/RFI environment and electrical transients than the existing Tenor drum controller.

Installation and wiring of the PLC in accordance with the modification instructions and the manufacturer's literature is expected to minimize the PLC vulnerability.

50.59 Evaluation The changes proposed by this activity do not affect the design basis accidents discussed or referenced in USAR. Therefore, the probability of occurrence or radiological consequences of an accident are not increased. The changes proposed by this activity do not create any new credible accidents.

The Condensate Demineralizer System (AK) serves no safety function. The equipment affected by this activity is not relied upon during or following design basis events to assure the integrity of the reactor coolant pressure boundary, the capability to shut down the reactor and maintain it in a safe shut down condition, or the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to 10 CFR 100.11 guidelines. The changes proposed to be implemented under this activity do not affect equipment important to safety. Therefore, the probability of malfunctions of equipment

Attachment II to ET 02-0012 Page 15 of 49 important to safety is not increased. Consequences of equipment malfunctions are not increased tan the possibility of a different type of malfunction is not created.

The changes to be implemented by this activity do not affect the Technical Specifications or plant licensing basis documents other than the USAR figure changes previously identified.

Therefore, no acceptance limits associated to those documents are affected, and the margin of safety has not been reduced.

Attachment I to ET 02-0012 Page 16 of 49 Evaluation Number: 59 2000-0079 Revision: 0

Title:

Modification of Spent Fuel Pool Bridge Crane Activity Description The objective of this activity is to enable modification of the Spent Fuel Pool Bridge Crane (HKE04). The seismically designed modification, temporarily adds an extension on the north side of the existing crane. The modification is necessary to enable fuel-handling access to the north two rows of the new spent fuel storage racks that were installed as part of another modification. The extension will be installed, using structural steel bolting, for a short duration to enable access to the otherwise inaccessible storage cells of the new racks for post installation drag testing as well as storage of spent fuel assemblies into those cells. The existing 2-ton motorized monorail trolley will be relocated onto the extension structure (monorail). Once the post-installation drag testing and the storage of spent fuel assemblies in these presently inaccessible storage cells have been completed, the 2-ton motorized monorail trolley will be replaced back to the south monorail beam, and the crane extension will be removed from the Spent Fuel Pool Bridge Crane. The only permanent change will be the additional bolt holes used for attaching the crane extension to the SFP bridge crane. The local change to the cross section of the existing members have been evaluated and found acceptable. The temporary configuration has been analyzed to insure the crane will maintain its structural integrity during a Safe Shutdown Earthquake (SSE).

The functional requirements of the Spent Fuel Pool (SFP) Bridge Crane are identified in the USAR. The existing 2-ton motorized monorail trolley is primarily used for moving fuel assemblies in the fuel storage pool area. The existing 2-ton motorized monorail trolley will be temporarily relocated onto the temporarily added monorail beam on the north side of the SFP Bridge Crane. As such, the SFP Bridge Crane will be operated within the same parameters as presently defined in the USAR. The USAR describes interlocks between the bridge crane and the new fuel elevator, fuel transfer canal gate and the cask loading gate, a limited maximum lift height (to maintain fuel shielding), a weight-operated hoist upper limit switch in addition to the geared-type upper and lower limit switches, and that the SFP Bridge Crane to be controlled from a pendant station supported from the trolley. Since the existing 2-ton motorized monorail trolley will be re-located to the north monorail, all of these functions will be retained with the hoist on the north monorail.

The Spent Fuel Pool Bridge Crane (HKE04) will be modified by temporarily adding a monorail beam on the north side of the bridge crane. The existing 2-ton electric hoist will be also temporarily relocated, onto the temporary monorail beam. This temporary change in the configuration of the Spent Fuel Pool Bridge Crane will be a configuration different than presently depicted on USAR Figures.

50.59 Evaluation The fuel handling accident was reviewed. It consists of dropping of a spent fuel assembly onto the fuel storage area floor, resulting in the rupture of the cladding of all the fuel rods in the assembly despite many administrative controls and physical limitations imposed on fuel handling operations.

There are no new or different types of credible accidents that the proposed modification to the Spent Fuel Pool Bridge Crane could introduce since the same administrative controls and physical limitations remain in effect as prior to the modification. Fuel assemblies will be 1 to ET 02-0012 Page 17 of 49 handled with the same motorized monorail trolley hoist utilizing the same controls and procedures as presently defined in the USAR. Therefore there is no increase in the probability of occurrence or no increase of radiological consequences of an accident previously evaluated in the USAR.

The proposed change has been analyzed to ensure structural integrity during an SSE.

Therefore, there are no credible malfunctions of equipment important to safety that will be changed or consequences that will be altered by the proposed modification to the Spent Fuel Pool Bridge Crane.

Fuel assemblies will be handled with the same motorized monorail trolley hoist utilizing the same controls and procedures as presently defined in the USAR. Therefore, the proposed modification to the Spent Fuel Pool Bridge crane will not create the possibility of an accident or equipment malfunction of a different type than any previously evaluated in the USAR.

There are no acceptance limits in the USAR that could be affected by the proposed modification to the Spent Fuel Pool Bridge Crane. Therefore, the margin of safety has not been reduced.

Attachment IIto ET 02-0012 Page 18 of 49 Evaluation Number: 59 2001 -0001 Revision: 0

Title:

Addition of Spool Piece to Replace Automatic Temperature Control Activity Description This activity deletes Service Water temperature control valve EATV0007 and associated instruments and adds a spool piece in its place. This activity will eliminate the possibility of EATV0007 binding closed. The spool piece will allow for full cooling water flow. However, the automatic temperature control performed by EATV0007 for maintaining generator hydrogen temperature will be lost. Therefore, Operations will throttle Service Water return isolation valve EAV0037 manually for the temperature control. Lake temperature does not change rapidly and the Generator load is fairly constant. Due to these conditions, Operations can maintain the Generator Hydrogen temperature. Replacing EATV0007 with approximately 18 inches of spool pipe and throttling EAV0037 does not change flow significantly. This modification should not cause any erosion on the hydrogen cooler tubes or challenge the hydrogen temperature limits.

The control room also has an alarm and generator hydrogen temperature indication. Operating procedures provide actions in response to abnormal conditions.

50.59 Evaluation The transient describing a decrease in heat removal by the secondary system on a turbine trip and the initiation signal of generator trip was reviewed for potential impact. Once the generator trips, the results of the turbine trip remain the same. Generator hydrogen temperature control is not relied on for this event. Therefore, this activity will not increase the probability or consequences of an accident described in the USAR.

Current design postulated that EATV0007 will fail open and EAV0037 is throttled to maintain generator gas temperature. Exchanging EATV0007 with a spool piece on the generator hydrogen cooling line mimics this condition. Therefore, this activity does not create any credible accidents.

This is a non-safety related system. Changing EATV0007 to a spool piece in the generator hydrogen cooler line reduces valve malfunctions and does not increase cooler degradation.

Therefore, it does not cause or create any credible malfunctions of equipment important to safety. Nor will it increase the consequences of a malfunction of equipment.

There were no acceptance limits associated to this activity. Therefore the margin of safety has not been reduced.

Attachment IIto ET 02-0012 Page 19 of 49 Evaluation Number: 59 2001-0002 Revision: 0

Title:

Compensatory Measures for Loss Sprinkler and Spray Systems of Activity Description This activity revises the compensatory measures for out of service containment fire suppression systems. The revision allows the verification of operable fire detection in the area and requires no additional compensatory measures if detection is operable.

This activity also adds compensatory measures for the loss of supervisory circuits for the valve for the underground fire main to be sealed or locked in its required position when the electrical supervisory circuit is demonstrated inoperable.

50.59 Evaluation Fire protection features provided inside containment consist of manually charged sprinkler systems in the North and South cable penetration areas, a manually charged standpipe and hose system, portable fire extinguishers, automatic linear heat detection above each reactor coolant pump, automatic linear heat detection above concentrations of cable trays, and automatic ionization smoke detection inside containment cooler ducts. The automatic fire detection system alarms in the control room.

Separation of fire safe shutdown components within containment meets the intent of Appendix R,Section III.G.2. The Fire Hazards Analysis (USAR Appendix 9.5B) summarizes the fire safe shutdown methodology inside containment and concludes that no postulated fire from fixed or transient combustibles in the reactor building will prevent safe shutdown of the plant.

Compensatory action options of an hourly fire watch, 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months tours, or hourly temperature monitoring for inoperable fire suppression and hose standpipe systems inside containment are impractical and considered inferior to operable detection. An hourly fire watch inside containment while at power is not recommended due to ALARA considerations. Also, an hourly fire watch and 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months tours of containment are not as reliable as operable detection, since detection is present at all times. Monitoring containment air temperature is also not as reliable an indicator of fire as fire detection since the fire detection system will operate long before a noticeable increase in containment air temperature. Operable fire detection inside containment provides a more reliable and faster indication of fire than an hourly fire watch, 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months tours, or hourly temperature monitoring.

During normal operation, total reliance is placed on the fire detection system to initiate action.

With an operable water supply and fire suppression systems, the only way to initiate fire suppression activities is by the automatic fire detection system. The currently analyzed condition specified that fire detection is required before a fire in containment is recognized.

Therefore, relying on fire detection when the suppression systems or standpipe hose system is inoperable is not different than the normal operation of the plant.

In accordance with Wolf Creek Generating Station procedures, when a fire is detected by the automatic detection system, an operator is sent to investigate the source of the alarm. Ifthere is a true fire emergency, then the Fire Brigade is called out. Operators only open the valve that supplies water to the hose stations and sprinkler systems in containment if directed by the Fire Brigade Leader. Not all situations require charging the fire protection piping inside containment.

Attachment II to ET 02-0012 Page 20 of 49 It has been established that fire extinguishers are a suitable backup to hose stations inside containment during postulated fires in Modes 1 through 4. During Mode 5, the air lock can be breached. This allows a backup water supply to be brought into containment. The Fire Hazards Analysis (USAR Appendix 9.5B) describes the fire safe shutdown methodology for a fire in containment. An RCP lube oil fire is not postulated during power operation due to the oil collection system. Adequate separation exists between redundant components within the containment such that a fire in any one area will not impact the redundant train. Aside from RCP lube oil, the only significant fire hazard in containment during operation is cable insulation.

The cable insulation is IEEE 383 rated. Tests have shown that IEEE 383 cable insulation does not propagate fire and self-extinguishes when flame is removed. Since the quantity of transient combustibles is low, and an RCP lube oil fire is not postulated, the probability of a significant fire of such magnitude that the cable insulation will continue to burn is very low. Under these postulated scenarios, it is reasonable to conclude that a sufficient alternate to hose stations for a fire inside containment is fire extinguishers. Therefore, a postulated fire inside containment can be extinguished with fire extinguishers in the absence of suppression and hose stations.

In addition to the above, one of the corrective actions for PIR 2000-2976 was to add compensatory measures for fire protection valve supervision out of service. This change adds the appropriate compensatory measures for loss of fire protection valve supervision. System 7, Fire Protection Valve Supervision will be added to USAR Table 9.5.1-3.

There are no design basis accidents affected by this change.

Since there are no physical changes and the design basis function of the system is not affected by this change, no new types of accidents not previously analyzed could be created. In addition, the probability of occurrence of an accident, or consequences of and accident has not been increased.

Since the proposed change would not affect the system's failure mode, the systems design function, the level of qualification, or equipment important to safety, no credible malfunctions of equipment important to safety are identified. Therefore, the probability and consequences of a malfunction of equipment important to safety has not been changed. No new malfunctions have been created.

There are no acceptance limits associated with these administrative changes. Therefore, no acceptance limits could be affected and the margin of safety has not been reduced.

Attachment IIto ET 02-0012 Page 21 of 49 Evaluation Number: 59 2001-0003 Revision: 0

Title:

Piping Replacement Due to Flow Accelerated Corrosion Activity Description This activity provides guidelines for the replacement of a six and eight inch line in the Feedwater Heater Extraction, Drains and Vents system, to mitigate abnormal pipe-wall thinning due to Flow Accelerated Corrosion (FAC), with a low alloy steel (2 1/4 Cr - 1 Moly) which improves piping resistance to FAC.

The proposed pipe replacement does not change the cross sectional properties (section modulus, moment of inertia), or the geometric configuration. The mechanical properties such as tensile strength and code allowable stresses remain unchanged. The lower yield strength and higher Young's Modulus is judged to have insignificant impact on the original analysis.

Therefore, the change does not adversely affect the existing safety margins or structural integrity of the affected piping system. The piping stresses will remain acceptable within code allowables.

The proposed replacement does not adversely affect any system, component or procedures required to mitigate the consequences of an accident previously evaluated in the USAR. The proposed change will restore a degraded section of the affected piping system, to its original design configuration (piping geometry, cross section, support location, fittings).

Ductile fracture, corrosion, erosion/corrosion, loss of mechanical properties, excess strain, mechanical creep etc., are credible failure modes for which the proposed piping replacement has been evaluated, through a critical characteristics comparison to the existent piping system design. Based on the evaluation, it was concluded that no new credible failure mode is introduced.

The proposed change will restore a degraded section of the affected piping system, to perform its original design intent. The proposed replacement does not involve or affect any safety related system or component. All system functions will continue to be performed as designed.

50.59 Evaluation No accidents are identified as being associated with or impacted by this change. The proposed change will restore a degraded section of the affected piping system to its original design configuration (piping geometry, cross section, support location, fittings).

Therefore, the probability of occurrence of an accident, or consequences of and accident has not been increased. No credible accidents could be created.

No credible malfunction of equipment important to safety are identified which could be associated with or affected by this change. The proposed pipe replacement does not change the cross sectional properties (section modulus, moment of inertia), or the mechanical properties (tensile strength, and/or code allowable stresses), or the geometric configuration of the piping. Therefore, the probability and consequences of a malfunction of equipment important to safety has not been changed. No new malfunctions have been created.

There are no acceptance limits contained in the bases of the technical specification or licensing basis documents that could be affected by this change. Therefore, the Margin of safety has not been reduced.

Attachment II to ET 02-0012 Page 22 of 49 Evaluation Number: 59 2001-0004 Revision: 0

Title:

Piping Replacement Due to Flow Accelerated Corrosion Activity Description This activity provides guidelines for the replacement of a twelve and twenty inch line in the Feedwater Heater Extraction, Drains and Vents system, to mitigate abnormal pipe-wall thinning due to Flow Accelerated Corrosion (FAC), with a low alloy steel (2 1/4 Cr - 1 Moly) which improves piping resistance to FAC.