ML20217N906

Rept of Changes,Tests & Experiments for Period of 970422-990621

ML20217N906
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Site:	Pilgrim
Issue date:	06/21/1999
From:	ENTERGY OPERATIONS, INC.
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ML20217N895	List: ML20217N891 ML20217N906
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NUDOCS 9910290157
Download: ML20217N906 (22)

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I ATTACHMENT TO PILGRIM STATION REPORT OF CHANGES, TESTS, AND EXPERIMENTS l

Installation of an Oil Drain Line on the Main Transformer and Oil Diversion Curbing in the Turbine Building Safety Evaluation: 3083 PDC: 97-009.10.8.3.4 FSAR Section:

This safety evaluation supports a major FRN to PDC 97-009 which replaced the main transformer during RFO 11. It also supported an analysis of the fire protection aspects of curbs installed by FRNs to PDC 97-009. The former transformer experienced a major fault while the plant was shut down for refueling in RFO 11. The fault resulted in a significant oil spill in the Turbine Building. The flow path was through the low voltage bushing, the bushing box, the Isolated Phase Bus Duct, the Bus Duct Cooling unit, and onto the Turbine building floor at elevation 23' in the Generator Auxiliaries area. The oil continued to flow into the lower Switchgear room, in order to prevent oil from affecting safety related equipment in a similar event, the following modifications were i

made:

Provided a means to minimize the amount of oil that can actually drain from

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the transformer into the Turbine Building. A drainage system was installed at the low voltage bushing enclosures on the main transformer. The system provided an alternate flow path for transformer oil from damaged bushings.

Constructed curbs in the Turbine i3uilding to prevent transformer oil from

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reaching the lower Switchgear room.

This safety evaluation did not involve an unreviewed safety question. Fires are an analyzed event and conservative conclusions have been made as to the consequences of such events. A postulated fire in a fire area containing safa shutdown equipment is conservatively assumed to disable all of its associated aJe shutdown equipment whether any would be damaged by such a fire. These modifications reduce the potential for this event and as a result reduce the risk of a fire in the Switchgear room.

Installation of a Permanent Zinc Oxide injection System Safety Evaluation: 3103 PDC: 97-012-FSAR Section: 11.8.3 This change replaced the temporary " active" skid with an installed permanent " passive" skid to add depleted zine oxide into the reactor feedwater for source term (dose rate) reduction. The zine oxide is injected by means of a bypass loop around the feed pumps. This change was installed as part of PDC 97-012 which connected the General Electric (GE) provided skid to the plant piping.

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The change did not involve an unreviewed safety question. The Condensate and Feedwater systems are non-safety related. Installation of this permanent skid does not effect the safety functions of any other system. The zinc injection system cannot initiate any analyzed accidents and will not adversely effect the feedwater system design basis. Components being added by this modification cannot compromise the reactor coolant pressure boundary. Zinc injection does not degrade fuel performance and does not affect the heat transfer ability of the fuel. The additional deposits on piping, piping components, and fuel surface due to zine addition will be insignificant when compared to existing corrosion product deposits. Analysis and testing performed by GE found that ionic zinc is not detrimental to BWR piping, structural alloys, or fuel cladding at concentrations up to 65 ppb. Reactor water zinc concentrations at PNPS will be maintained at approximately 5-10 ppb. The loss of feedwater heating and the failure.of a feedwater controller have been analyzed in the FSAR. The zinc injection skid does not interact with the operation of either the feedwater controller or feedwater heaters.

Eliminate Active Safety Functions of MO2301-9 and MO130-48 Safety Evaluation: 3138 FSAR Section: 7.4.3.2.5 The safety functions of HPCI and RCIC pump discharge valves MO2301-9 and M01301-48 were changed from an active " function to open" (when closed for test) for supporting associated system operability, to a " pressure boundary only" passive safety function.

In conjunction with the other CSCSs, HPCI provides adequate core cooling under abnormal transient conditions and the range of small to medium pipe breaks. The original design safety function of MO2301-9 was to open (if closed for testing) on receipt of a HPCI initiation signal in time to support system operability requirements for injection flow. The RCIC system provides makeup water to the reactor following reactor vessel isolation in order to prevent the release of radioactive materials to the environs as a result of inadequate core cooling. The original design safety function of M01301-48 was to open (if closed for testing) on receipt of a RCIC initiation signal in time to support system operability requirements for injection flow. The valves still receive a signal to open on system initiation, but since the systems are unanalyzed with these valves initially closed (i.e. no analysis for adequate stroke time or system hydraulic),

HPCI and RCIC will be considered inoperable when these valves are closed. This is an analytical and operational change only and involves no changes to plant hardware.

This change did not involve an unreviewed safety question.

HPCI and RCIC are mitigative systems and are not associated with any accident initiators. Single failures of close torque switches in MO2301-9 and M01301-48 may lead to functional failure of either valve and the associated system, but this is bounded by other single failures in these systems. Because the injection check valve and other motor-operated isolation 2 of 22

valves (MO2301-8 and M01301-49) are available, leakage failure of MO2301-9 and MQ1301-48 is not a concern. This is not a new' failure mode of these components.

Since HFCI and RCIC are only operable when MO2301-9 and M01301-48 are fully open, the core cooling functions of HPCI and RCIC are unaffected. Inadvertent closure actuations of either valve represent a single failure that causes loss of the system, however, the systems are not single failure proof and this is not a new failure mode.

Revision to the Cooling Water Differential Pressure Setting l

Safety Evaluation: 3141 FSAR Section: 3.4.5.3.1 This change supports revisions to Pilgrim Procedures 2.1.35 and 2.2.87 to reflect changes to the CRD cooling water differential pressure (dp). The original CRD cooling l

water differential pressure design value was approximately 15 psid. Due to changes to the CRD return line under PDCR81-66, the cooling water pressure control valve lost its i

ability to control differential pressure.

Therefore, the differential pressure for the cooling water is no longer a variable, but is a resultant of the drive water pressure and the cooling water flow at approximately 10 psid.

This change did not involve an unreviewed safety question. The safety function of the CRO system is to provide a means to quickly terminate the nuclear fission process in the core so that damage to the fuel barrier is limited. CHD cooling water flow is directed to each drive through an orifice in the CRD flange and follows a passage between the outer tube and thermal sleeve to the outer screen. Cooling water is used to protect the graphitar seals from high reactor temperatures. Long exposures at high temperatures will result in brittle, fast wearing seals. As such, changes in cooling water flow and dp may have an effect on the maintenance of the CRDs, but have no effect on the scram function of the CRDs. No new or changed failure modes are postulated due to this minor change.

Addition of Correction Factor for Feedwater Flow Measurement Safety Evaluation: 3144 PDC: 98-008 FSAR Sections: 7.16.4.2.7,7.16.4.2.8 This change added a supplemental feedwater flow measurement system (flow transmitter and data collection system) independent of existing flow venturis FE 641 A/B which are used to establish the relationship of MWT versus MWE. The modification l

provides administrative controls to collect data, compare the data to the installed system, and compute a Correction Factor (CF) for EPIC. It also provides limitations of individual and total flow at rated power to assure conservatism is maintained within allowable uncertainty assumptions. The new circuitry includes a continuous EPIC comparison of feedwater determined power versus other parameters with an EPIC alarm, should the difference be too great.

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This change did not involve an unreviewed safety system. The installation of the flow transmitter and data collection system does not penetrate or physically interface with any control or safety related system. Its use provides additional capability for assuring feedwater flow input to the reactor power calculation is within assumed bounds, and provides additional benefit to the system as a routine monitor of the existing feedwater flow venturi's physical condition.

Corrections to the Design Description of the Electrolytic Hydrogen Water injection System Safety Evaluation: 3159 FRN: 85-057-051 FSAR Sections: 10.22.3.5,10.22.3.7,10.22.4,10.22.5 This safety evaluation supports minor corrections to the FSAR description of the Electrolytic Hydrogen Water Chemistry (EHWC) system design. The changes consist of:

Deletion of the reference to an emergency shower in the EHWC gas

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generator building.

Elimination of the specified rac7iation reduction levels when running the

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EHWC in the " Maintenance Flow" mode.

Deletion of the description of the oxygen injection built-in delay function

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which is a fixture of the Extended Test System (E TS) only.

Addition of an exception (a check valve) to the description of valves in the

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hydrogen flow path that have packing type stem seals.

The evaluation did not involve an unreviewed safety question. The changes were administrative, made to reflect the actual design details of the EHWC.

Installation of High Pressure Sodium Vapor Lights Safety Evaluation: 3204 FSAR Section Affected: 10.16.2 This evaluation revises the station lighting section of the FSAR to indicate the acceptability for use of modified high pressure sodium vapor lights in the spent fuel pool during all modes of operation and in the reactor cavity during refueling periods.

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The FSAR description of the station lighting system provides for the use of mercury vapor, fluorescent, or incandescent fixtures for both normal and emergency lighting throughout the plant. Mercury fixtures and switches are not normally used in primary containment or on the 117 ft elevation for environmental reasons. Mercury can cause embrittlement of steels, particularly zirconium alloys, at high temperatures leading to l

intergranular cracking.

L This change did not involve an unreviewed safety question. The analysis conducted in support of this change concluded use of high pressure sodium vapor lights is acceptable if the lights have been modified to:

contain mercury in the event that a lamp element breaks, and

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provide a thermal barrier so that the lamps can be used in wet / dry

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applications.

Leak Isolation Capabilities of Recirculation Pump Suction and Discharge Valves i

Safety Evaluation: 3205 FSAR Section: 4.3.4 This change was for the purpose of clarifying the FSAR. The previous verbiage of the FSAR implied that the recirculation pump suction and discharge valves must be capable of isolating loop leaks between them. However, research into the design i

bases for these valves with the assistance of General Electric (GE) identified this was not an original design function of these valves. The change clarifies this implication.

This change did not involve an unreviewed safety question.

It does not involve a physical change or alteration to the plant and does not effect the safety functions of any other system. Its purpose is only to clarify the FSAR wording relative to the design functions of the recirculation suction and discharge valves.

Revision to HPCI, RCIC, and RWCU Line Break High Flow '.uto isolation Time Delay l

Safety Evaluation: 3228 l

PDC: 99-013 l

FSAR Sections: 4.7.5, T5.2-4, 6.5.2.3 The change reduced the allowable analytical stroke time of the HPCI, RCIC, and RWCU line break outside containment DC isolation valves by 1.0 seconds to correspond with an attendant 1.0 second increase in analytical limit for the control loop sensing and actuation time delay, and 0.5 second correction to account for the process l

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change to reach the instrument sensing line. Accordingly, MO2301-5 and M01201-5 allowable stroke times were changed from 35.0 seconds to 34.0 seconds and M01301-17 allowable stroke time was changed from 30.0 seconds to 29.0 seconds.

This change did not involve an unreviewed safety question. Its purpose was to account for the total loop uncertainty in the overall system design as recently quantified by Calculation IN1-252. Previous estimates of loop uncertainty and relay actuation times were 0.5 seconds. The recently performed calculation indicated a worst case delay of 0.8 to 0.9 seconds. The change also included a 0.5 second correction to account for l

process flowstream response time. The time delay relays and containment isolation valves are mitigative. components in that they are part of the primary containment isolation of HPCI and RCIC system initiation safety functions.

The remaining i

operational functions of "le valves are applicable to normal plant operations such as l

placing the systems in service and are not associated with any accident or transient l

initiators. The time delay relays and isolation valves were not physically modified.

Valve structural and operational limits were not exceeded and conventional design l

approaches were used to set the relays.

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1 Revise the RBCCW and TBCCW FSAR Pressure Boundary Descriptions Safety Evaluation: 3247 FSAR Section: 10.6.3 This change redefined the classification of Class l Pressure Boundary Only (PBO) contained in Section IV of the Q list and changed the classification of the TBCCW I

I piping in the rooms containing the RBCCW pumps from PBO Class I to Management Quality Control items (MQCl)(Class 11/l).

I The PNPS Q-List identifies the structures, systems, and components that prevent or mitigate the consequences of postulated accidents that could cause undue risk to the i

health and safety of the public covered by the PNPS Quality Assurance Program in accordance with 10 CFR 50, Appendix B.

These items are regarded as " safety-related" In addition to safety related equipment, PNPS has a category of equipment called Management Quality Control items (MQCl), to which specified requirements of the j

Quality Assurance Program are applied. Section 4 of the Q-List contains the listing of these Management Quality Control items. This section provides a listing of the non-safety-related components or types of components which are to be considered "Q" These items are covered only by the work controls within the Quality Assurance Program specified in this section.

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Section 4.2.6 describes a category of items called " Pressure Boundary Only". Portions of certain systems are required to maintain their pressure boundary integrity, to maintain system fluid inventory, or due to proximity to safety related equipment which 6 of 22

could be degraded by a failure of the pressure boundary. Prior to this change, those portions of the RBCCW system that perform no active safety function, as well as those portions of the TBCCW which are located in the rooms containing the RBCCW pumps were designated as PBO Class 1. This means that they were seismically qualified by engineering analysis, analyzed as needed to meet Class I seismic standards, and serve only to maintain pressure boundary.

This FSAR change was developed to distinguish the classification of the RBCCW and TBCCW systems and thereby be capable of identifying the in-Service Inspection (ISI) boundary change from Class 3 to Class 0 at the normally open RBCCW non-essential loop isolation valves. These boundaries are established by PNPS to determine the applicable 181 requirements per ASME Code Section XI.

This change did not involve an unreviewed safety question. It serves to enhance the organizatica's effectiveness in dealing with the non-safety related portions of the RBCCW system by establishing an Augmented inspection Program for the RBCCW non-essential piping and components.

The revision to the MQCl classification in the Q-List more clearly defines the applicable Quality Assurance criteria established for piping and components classified as Class l-PBO.

Changing the classification of those portions of the TBCCW system which are located in the rooms containing the RBCCW pumps from Class l-PBO to Seismic Class ll/l does not affect the function of any safety related components. The original intent of classifying those portions of the TBCCW piping as Class 1-PEO was to prevent failures of the TBCCW piping from affecting the safety functions of systems in proximity of the

.TBCCW system. However, since the definition of-Class 1-PBO is changed to one -

appropriate for only the non-safety portions of RBCCW, designating those portions of the TBCCW which are located in the rooms containing the RBCCW pumps as Seismic Class ll/l provides an adequate level of quality control measures to assure the proper functioning of all safety related components which could be affected by the failure of TBCCW piping.

Station Blackout Diesel Generator Manual Start and Load Testing Safety Evaluation: 3251 FSAR Section: 8.10.4 This evaluation supported a temporary procedure (PNPS TP 99-073) which consisted of manually starting and loading the Station Blackout Diesel Generator (SBODG) using actual station loads from 'either 4.16kV safety bus A5 or A6 during refueling outage RFO 12. One 4.16kV safety bus was selected and taken out of service, while the other 4.16kV safety bus remained the protected loop.

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The temporary procedure was established as the means of conformance with FSAR Section 8.10.4 for testing during unit shutdown (outages). The purpose of the testing during outages is to prove the capability of the SBODG to accept large loads while subject to subsequent transients that would be representative of an actual SBODG.

This temporary procedure did not involve an unreviewed safety question. The SBODG is the alternate AC power source (in accordance with 10 CFR 50.63) capable of providing the required. loads necessary to achieve safe shutdown following a station

. blackout event. It is classified as management Q. Upon a total LOOP (luss of both the preferred and secondary (23kV) power sources) and failure of both EDGs, the SBODG can be manually started and supply adequate power for the operation of one low pressure CSCS pump from either AS or A6 and all associated loads on that train required for a LOOP without a LOCA.

The loads placed onto the 2000KW SBODG during the testing consisted of both RHR pumps and the CRD pump from the selected bus (A5 or A6). The total load was approximately 1440KW.

The SBODG has the capacity to accept large loads as demonstrated by the post work test when the SBODG was initially installed. This load test performed during turnover of the SBODG envelopes the tests to be performed by TP 99-073. Additionally, the full load capacity of the SBODG is proven quarterly in accordance with PNPS Procedure 8.9.16.1. Therefore, the temporary procedure was within the capability of the SBODG and demonstrated that the SBODG can accept SBODG loads.

FSAR Revision Regarding EDG Voltage Recovery Time Safety Evaluation: 3259 FSAR Section: 8.5.4 This change revised the FSAR. stipulation for an EDG to recover to 95% of rated voltage within 1 second during loading transients. The 1 second recovery requirement has been eliminated as it is an indication of long term performance of the system, rather than as the measure of a successful emergency load sequence. The measure of successful emergency load sequence is starting and bringing the emergency CSCS pumps up to speed in accordance with Technical Specification 4.9.A.1.b.

The design basis LOCA is the most demanding accident condition for the CSCS. In the event of a DBA LOCA coincident with LOOP, the EDGs will be started and will restore power to the emergency buses within approximately 10 seconds. Once power has been restored, time delay relays sequentially load the ECCS onto the buses. The purpose of the time delay is to allow the preceding load to come up to it's rated speed before the next load is applied. 95% voltage recovery in 1 second is not a necessary requirement for the EDG to successfully power it's CSCS loads within the required time periods.

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The purpose of voltage recovery within a specified time is to ensure that the required voltage is available for the large loads to accelerate and come to rated speed within their specified time limits. Thereby at the full load condition, the EDG loading will be reduced to a minimum before the start of the next large load thus eliminatir$9 the maximum coincident load.

i The current FSAR statement was added by FSAR Amendment 26 in response to a preoperational NRC question pertaining to the EDG system performance during load transients.

The statement was derived from the original EDG procurement specification. The one second time was an approximation and was bounded by initial EDG acceptance testing. Subsequent operating history has shown that recovery of voltage to 95% within one second is not a requirement for the EDG to perform it's safety function, namely the ability to power sequenced emergency loads within prescribed time limits.

This change did not involve an unreviewed safety question. It clarifies the impact of a measured process variable during the load sequence test performed during refueling outages. There is no physical change to any system nor any affect on any safety function. Components involved in the initiation of accidents previously evaluated are not affected. The EDGs ability to accept load and meet the requiremercs specified in the accident analysis is not affected.

HPCI and RCIC Pumps Test Acceptance Criteria Safety Evaluation: 3260 FSAR Sections: 4.7.5, 4.7.7, T4.7-1, T6.3-1, 6.4.1, F6.4-1, 6.6 The purpose of this change was to revise the FSAR to include design basis information on the operating range upper pressure for the HPCI and RCIC systems that is credited in the safety analysis.

Furthermore, pump test surveillance procedures were changed to include acceptance criteria from the safety analysis requirements.

The test procedure changes allow testing to demonstrate that the HPCI and RCIC pumps are capable of delivering the required flow rates to the reactor vessel under all system conditions considered within the design basis. The basic test method is unchanged, however, the test acceptance criteria for pump discharge pressure is increased and turbine speed limits are now imposed on all pump tests.

The revisions did not involve any unreviewed safety question. The proposed changes to the FSAR describe the requirements for system operation more fully than before, and the procedure changes are designed to provide assurance by test of system capability to meet the design requirements. Both HPCI and RCIC provide core cooling following transients and accidents. The added assurance of HPCI and RCIC capability that is provided by the proposed testing provides assurance that core cooling can be provided over the applicable range of reactor pressures and the proposed changes do not involve any mechanism whereby a greater amount of radioactive material can be released to the surroundings, 9 of 22

Feedwater Flow Control Valve Internals Replacement Safety Evaluation: 3261, supercedes SE 3240 PDC: 98-029 FSAR Section: 7.10.3.5 The internals of feedwater flow control valves FV-642A&B were replaced to allow for better sensitivity to valve sticking / friction and oscillation problems experienced with the valves. The valves were the cause of plant shutdowns.

The pneumatic controls associated with each valve were replaced. Additional components were added to the control racks to support the piston-type actuator. The function will remain the same as the existing valves. The replacement design allows approximately an 8" stroke and should eliminate the sensitivity to valve sticking / friction and oscillation problems experienced with the prior valves.

This change did not involve any unreviewed safety question.

There is no safety function of the valves, but they do effect plant operations since they can cause a reactor shutdown if they do not control feed flow correctly relative to steam flow and reactor water level. The valve internals are considered acceptable replacements since they have better sensitivity to flow and the pneumatic controls were equivalent to or better than those on the existing valves. The new internals and controls do not affect the consequences of previously analyzed accidents. The consideration of flow, seat leakage, and pressure class are all acceptable.

Replacement of Main Generator Step Up Transformer

- Safety Evaluation: 3078 PDC: 97-009 FSAR Section: T8.5-1 The existing PNPS Main Generator Step Up Transformer was replaced due to damage as a result of an internal fault. It was replaced in kind with a replacement transformer of similar design obtained from the Millstone site. The new transformer will perform the identical functions of the existing transformer. The new transformer has similar ratings, but its physical dimension differences required modifications to the main generator step up transformer auxiliaries and to the fire protection system.

The change did not involve an unreviewed safety question. The physical dimensional differences of the replacement transformer required modifications to the main generator step up transformer auxiliaries. These included modifications to the main generator bus duct terminals and their enclosure, rework for connections to the unit auxiliary transformer, modificaSon of the transformer's oil preservation system, and control panel wiring changes. None of these modifications affected any safety related system.

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The fire protection system protects the main generator step up transformer with a deluge of water spray system in the even of a fire. The dimensions of the replacement transformer's auxiliaries necessitated changes to the fire protection system.

The system piping was modified accordingly..The fire protection system was not adversely affected by this change. Piping was sized based upon the available water supply. The fire protection system does not perform any safety functions. Changes to the fire protection system were consistent with the fire protection system design basis.

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Installation of a Wireless Telephone System Safety Evaluation: 3031 PDC: 96-025 l

'FSAR Sections: 10.15.2,10.'15.3 l

This mddification provided a permanent installation of wireless telephone base stations in the Reactor Building, Turbine Building, and Radwaste Building. This also included a p

' communication interface system located in the O&M Building which connects to the I

existing telephone system.

The new system enhances the existing means of l

communication between personnel in the Process Buildings, supplements the existing l

Gai-Tronics hard wired communications system, and replaces the hand held two-way radios, formerly used by Station personnel.

This change to the communications system had no effect on any safety system arid did not involve an unreviewed safety question.

MSlV Live Load Packing Charige

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' Safety Evaluation:' 3124 i

FSAR Sections: '4.6.3, F4.6-1 This ' safety evaluation was prepared to support FSAR changes and document the effects.of changing the original MSIV stem chevron packing rings to new composite l

graphite packing with live loads. The evaluation also documents the effects of cutting and capping the gland leak-off which was no longer needed with the new packing configuration.

FRNs 95-003-076 and 95-003-085 to Standing Plant Design Change (SPDC)95-003 authorized the design change of installing new composite graphite packing with live load in the MSIV stem packing and cutting and capping the gland leak-off line which Lwas not needed as a result. These design improvements reduced the packing friction, reduced the packing leakage, and improved the valve operability; however, the FSAR description of the original chevron packing rings and leak-off drain connection was not revised as part of the original modification.

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These changes did not involve an unreviewed safety question. The composite packing

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with live load does not adversely affect the' safety function of the MSIV. The new L

packing with live load enhances its safety function by improving its reliability and i

operability.

The gland leak-off for' MSIVs with live load packing are no longer

-necessary since the live load packing does not incorporate a packing gland. Since the live load packing design does not provide the same number of packing rings below the leak-off port as did the original packing, it is possible that the leak-off line will see a l

higher operating pressure and thus it is better to cut and cap the non-safety related l

leak off connection to prevent overpressurization of the upstream non-safety related L

waste collection components.

Replacement of Flow Control Trip Reference Cards Safety Evaluation: 3167 FRN: 96-015-004 FSAR Sections: 3.7.3.3, 3.7.4.6, 3.7.7, F3.7-1, F3.7-2, 7.5.7.1, 7.5.7.4 L

PDC 96-015 and subsequent FRNs installed a modified version of one of the six flow-biased _APRM rod-block and scram logic cards currently installed at PNPS.

This modification (FRN 96-015-004) replaced the remaining five cards (of six) which designates them all now as Flow Control-Trip Reference Cards (FCTRCs).

The principal purpose of this change was to install this primary design feature at PNPS, and thereby provide automatic actions to protect an adequate margin of safety to core l

l thermal-hydraulic instability. The range of core power and core flow was previously restricted by interim corrective actions (ICAs) to address NRC Bulletin 88-07, Supplement 1.

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-The change is analogous to the ICAs in that operation in the power / flow region most susceptible to the onset of undamped core power oscillation is prohibited by scram.

The primary difference is reliance on automatic actions (reactor scram or rod block) to enforce compliance rather than manual _ actions based upon administrative procedures.

This change did not involve an unreviewed safety question. The modification, as part

of ~a NRC reviewed and approved solution to NRC Generic Letter 94-02, is a key requirement if_ the integrated solution that assures PNPS adequately maintains an acceptable. margin of safety to prevent core instability, i.e., the initiation of undamped

- oscillations of core power associated with reactor operation at low reactor core flow and relatively high reactor power. The margin of safety is expressed as core power decay j

L ratio less than.1.0, the maximum acceptable limit for decay ratio being 0.8.

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designed, the modification prevents (by generating control rod blocks) intentional entry into core operating regions where decay ratio is postulated under adverse conditions to exceed 0.6.

. In the event of ~ an abnormal operational transient th1t results in

. inadvertent entry of the reactor into destabilizing operating conditions, the reactor core e

power is terminated by control rod scram before undamped core oscillations can j

commence.

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l Installation of a Period Based Detection System Safety Evaluation: 3183, supercedes SE 3016 PDC: 96-006 l

' FSAR Sections: 7.5.3, 7.5.6.1, 7.5.10, 7.5.10.2.6, 7.5.13, T7.5-6, T7.5-7, T-7.5-8

.This modification installed the Period Based Detection System (PBDS). The PBDS is a defense-in-depth feature of the Enhanced Option 1-A Stability solution.

A Period l

Based Algorithra (PBA) is contained in the firmware of the PBDS card. Its purpose is to

- monitor LPRM signals and. compare the periodicity of selected LPRMs to criteria that identifies the: expected periodic behavior ~ of reactor powe,r that is_ a precursor to l

thermal-hydraulic instability.

The cards generate digital and analog outputs 'for personnel trending and action..The LPRM _ Group A and B instruments provide inputs because the PBDS function is defense-in-depth.

PBDS_ has no interface with the APRMs or the Reactor Protection System (RPS).

The Period-Based Algorithm correlation to thermal-hydraulic oscillations is a result of BWROG analysis.

It is documented in NEDO 31960-A "BWR Owner's Group Long-Term Stability Solutions L

Licensing Methodology Supplement 1".

This modification did not create an unreviewed safety question. The PBDS does not

. interface with systems performing a safety function. The required personnel action in cases of unacceptable loss of stability margin is to. initiate a manual scram. The initiation and plant response to a manual scram is included in FSAR discussions for l

other abnormal and accident events. The components are mounted in "Q" panels meeting seismic ll/l criteria and the design incorporates separate and independent design features to ensure reliability.

l Modification to Procedures Governing Operator Response to High Temperature i

in MCCs B17, B18, and B20.

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- Safety Evaluation: 3202 FSAR Section: 8.4.5.2 l

This change modified the procedures for consistency with the design basis for the L

environmental qualification and safety analysis of the equipment in Motor Control Centers (MCCs) B17, B18, and B20. This equipment was designed to operate at <

80 F under normal operating temperatures and then be capable of performing safety functions at elevated temperatures over a 30 day period after a DBA event. The j

existing procedures allowed the equipment to operate above 80 F for indefinite periods of time.

1 MCCs B-17, B-18, and B-20 are each separately enclosed by a physical environmental protective barrier designed to isolate'their sensitive electrical equipment from direct exposure to h,gh energy steam and water pipe breaks.

These three separate enclosures each include separate air conditioning and ventilation 13 of 22

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subsystems to maintain each enclosure's air temperature below 80 F normally and 120 F following a DBA event. Temperature alarm (s) warn the operator (s) of a possible malfunction (s) of the associated MCCs' enclosure air conditioning subsystems or an excessive heat load to the associated enclosure's airspace. The Alarm Response Procedure (ARP) provides instructions to the operator (s) to address the alarms. This procedure was revised to lower the threshold of allowable air temperature within the MCC enclosures from 104 F to 80 F which is consistent with the environmental qualification of electrical equipment within the MCCs.

This procedure revision did not involve an unreviewed safety question. The revised procedure more conservatively adjudges the operability of safety related equipment in j

the unlikely case of the failure of supporting HVAC subsystems to maintain temperature within environmental qualification limits. The operator response is corrective only and cannot cause an accident. By lowering the temperature threshold for operability of important safety related equipment, the change may prevent the equ ament's l

malfunction during normal operation or during its mission time post-DBA. The MCCs, their enclosures, HVAC subsystems, and the associated temperature components are j

not physically modified in any manner by this change.

l Cap Penetrations X-21 and X-228A Safety Evaluation: 3209 PDC: 98-041 1

l FSAR Sections: T5.2-3, T.5.2-4 The ' Service Air and Containment Reference Pressure Systems which attach to penetrations X-21 and X-228A respectively, are no longer active and have been j

abandoned. This change cuts and caps these lines at containment penetrations X-t 228A and X-21 outside the torus and outside the drywell respectively. The portion of safety class piping that remains at the penetration continues to be part of the IWE l

containment Section XI program, but will be removed from the Section XI piping program by removing the safety class markings on ISI P&lDs. This was done since the penetration does not have an active piping system function and the time and dose to l

keep it in the Section. XI piping program is not necessary.

Cutting and capping (welding) the piping at the penetrations X-21 and X-228A moves the containment l

- boundary from the check valve for service air and from the isolation valves for the l

containment pressure reference system, to the cut and capped locations.

The remaining piping and cap will be part of the Section XI IWE and Appendix J programs.

This modification did not involve an unreviewed safety question.

The capped penetrations became part of the containment boundary so there is no piping system through containment. The piping, including the caps at the penetrations, is a non-mechan lstic static change.

These are tested as part of the Section XI IWE and Appendix,1 programs. The modification does not change any containment function. It only moves the boundary location from the isolation valves to the capped penetrations.

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increased Short Circuit Ratings for DC Panels Safety Evaluation: 3224 PDC: 98-026 FSAR Sections: 8.4.5.2, 8.6.3, T8.6-1, F8.6-1 This modification replaced components in 125V DC and 250V DC panels D4, D10, D16, D17, D29, D30, and D31 with components having adequate short circuit ratings i

and meeting station requirements for breaker / fuse coordination. The breakers, fuses, and buswork in the 125V DC panels and the entire 250V DC Panel D10 were replaced with components having an adequate short circuit rating. In addition, the modification:

l Resolved breaker coordination concerns by replacing some of the

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breakers in panels D10, D16, and D17, with non-automatic breaker / fuse combinations and replaced the current limiting fuses in D29, D30, and D31 with in-line fuses.

improved battery maintenance activities by providing a separate sub-bus

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in D16 and D17 for the plant loads and a means to isolate the main bus for battery testing and to equalize its charge.

The change did not involve an unreviewed safety question. The change reduced the probability of a complete loss cf a DC power train and the possibility of personnel injury and extensive equipment damage by improving the ability of the system to handle large faults. This reduces the probability of occurrence of accidents involving damage to DC equipment and loss of DC power. The existing separation between the two safety trains is maintained and no new failure modes are introduced that could cause both trains to fail. By increasing the short circuit capability of the system, the probability that the system will fail or that large portions of the system will be failed or damaged is

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reduced. The new components are designed to operate under the same conditions as

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the existing components. Their failure probability is not larger than that for the original

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components. By improving coordination and capability to withstand failed currents, the consequences of some equipment malfunctions are reduced by limiting the number of components involved in the failure and by reducing the amount of damage associated I

with the failure.

l Reduction in EDG Breaker Closure Timing j

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Safety Evaluation: 3226 i

PDC: 99-005 FSAR Sections Affected: 6.5.4.3, 8.5.3 The change reduces the existing 21.0 second timing to a value in the proximity of 14.4 seconds as assumed in the PNPS Safety Analysis. The design was such that, in the event of a postulated DBA LOCA coincident with the detection of a degraded voltage condition, power to the emergency buses would not be available for 21.0 seconds.

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This timing exceeds the 14.4 second timing assumed in the GE LOCA analysis (time between the initiation of the postulated recirculation line break and the availability of emergency AC power to the emergency bus).

l The modification only reduces the time in the event of a DBA LOCA or Turhire Trip coincident with a degraded voltage condition. For other e,enarios, e.g., uJCA w/

LOOP, PBOCs w/ LOOP, the EDG closing time remains the sarne, i.e., the existing time of approximately 10.15 seconds.

The new timing does not alter the Shutdown Transformer Source (SDT) breaker closing time.

The modification did not involve an unreviewed safety question. The modification changes the EDG breaker closing logic. There is no adverse effect on the Standby AC Power System (EDG) and other auxiliary electrical syr. ems. There is na physical change to any plant aquipment involved except a modification to the EDG closing logic.

The modification eliminates the discrepa 'v between tl e assumed delay of delivery of emergency AC power to electrically po's ad ECCS >quipment as evaluated in the PNPS LOCA/ECCS analysis for the limiting LOCA ;cenario (LPCI injection valve failure). In the event of a LOCA coincident with degraied voitage, the new timing will be bounded by the LOCA analysis NEDC-31852P, Rev.1, as referenced in the PNPS FSAR.

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Revision to Surveillance Schedule of Non-Q MCC Breakers Safety Evaluation: 3234 FSAR Section: T8.4-3 This evaluation revises FSAR Table 8.4-3 " Periodic Tests of Auxiliary Power Systems" This table defines overhaul and maintenance intervals for safety and non-safety 4160V Switchgear,480V Load Centers, and 480V Motor Con'rol Centers (MCC).

Two changes are addressed by this evaluation. First, to al!N deferrals of non-safety related MCC breakers surveillances by one fuel cycle if certain conc'itions are met.

Second, to remove the EQ and non-EQ distinction of Q breakers.

These changes serve to:

lmprove the cost / benefit ratio of the PMs appMd to non-safety related

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MCC control units, such that frequently operated or severe duty units see mere maintenance and those with light duty are not disturbed by unnecessary PMs.

lmorove the operating economy of the station by reducing the frequency

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of PMs on MCC control units and breakers which operate infrequent!y.

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Reduce the possibility of maintenance induced errors which may result e

l from unnecessary PMs.

These changes did not involve an unreviewed safety question. The equipment affected by the first change is not safety related and is restricted to equipment, the failure of which, will not affect any safety related functions. Further, the allowance for a one fuel l

cycle deferral of specific surveillances is bounded by the recommended surveillance l

frequency in EPRI NMAC Report TR-106857-V4 " Prevents./e Maintenance Basics, i

Volume 4.

Motor Control Centers".

This will ensure the affected MCC breakers cperate reliably. Deleting the distinction of "O" and "EQ" breakers in the FSAR will not effectively lengthen any previously committed PM schedule. Therefore, all "Q" and l

"EQ" breakers will continue to perform with a high degree of reliability.

Reload 12 Core Design l

Safety Evaluation: 3243 PDC: 99-010 FSAR Sections: 3.2.4, 3.2.6, F3.2-1,14.4, Appendix Q f

. This safety evaluation addresses the changes to the PNPS reactor core during RFO 12 which involved the replacement of 160 irradiated fuel assemblies with 160 fresh fuel l

assemblies. The fresh fuel assemblies are based on the GE11 bundla type with a bundle average enrichment of 4.08 weight percent. A split batch was used similar to i

Cycle 12, with the high gadolinia batch fraction loaded ir the core interior and the low gadolinia batch loaded toward the periphery, for optiniizatior: of cycle energy and shutdown margin. For operation during Cycle 13, the PNPS core was configured i

according to the loading pattern specified in the Fuel Loading Plan.

The MCPR l

operating limits are based on an MCPR safety limit of 1.08.

This change 1d not inve!ve an unreviewed safety question. The Reload 12/ Cycle 13 l

core design does not impair the ability of either the fuel assembly or the pressure relief l

system to perform their respective design safety functions. MCPR is calculated to I

always exceed Operating Limit MCPR and therefore prevent MCPR from lowering l

below the Ss7ety Limit MCPR. Additionally, the other thermal limits, i.e. MAPLHGR and MFLPD are met by Cycle 13 design with adequate margin. The previously analyzed l

transients do not violate thermal limits and design criteria, and have acceptable l

consequences using NRC approved methods described in GESTAR 11, the generic fuel licensing document. The types of accidents and transients evaluated i. the FSAR are described in Chapter 14.

No changes are proposed which affect the reliability or performance of equipment serving a safety function.

Performance of the turbine bypass valves is conservatively modeled in the analysis of the feedwater controller failure at maximum demand to yield conser/ative MCPR operating limits and to demonstrate conservative margin to spring safety valve lift.

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Additional Information Added to FSAR Regarding RBCCW Design

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. Safety Evaluation: 3255 FSAR Sections: 5.2.3.5.2,10.5.G q

' Additional information was acMed to the FSAR.to provide a clear description of 'the safety design and licensing basis of the RBCCW and primary containment systems

-relative to the effects of Pipe Breaks inside Containment (PBICs), also referred to as

. High Energy Line Breaks (HELBs). Primary coolant system breaks inside containment

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Lare not assumed to result in pressure boundary failure of the RBCCW system, e.g. due to pipe whip, jet impingement, missiles, or water hammer. Failures of RBCCW system pressure boundary are passive failures and are not assumed coincident with other passive failures, i.e., a break in the primary coolant system. Thus, the RBCCW lines -

are considered closed inside primary containment.

These_ changes did not involve an unreviewed safety question. The effort consisted of l

researching and documenting the design and licens*ng bases of the RBCCW system and adding these findings in the FSAR as appropriate.

i Restoration of Turbine Vibration Trip Setpoint Safety Evaluation: 3258 FRN: 99-004-024 FSAR Section: 7.11.3.3.6 l

The settings for the six turbine vibration trip and alarm functions had previously been increased by FRNs 96-004-036 and 97-004-036. The trip point was raised to allow for break-in of new turbine bearing packing which would minimize the need to defeat a high vibration trip during startup and power ascension. The relaxed trip setting is no

' longer required as the likelihood of a packing rub has been reduced. Therefore, the setpoint at the applicable circuit cards was lowered from 14 mils to'12 mils (peak).

This change did not involve an unreviewed safety question. Turbine shaft vibration 's a l

variable monitored by turbine supervisory instrumentation.

The turbine shaft high vibration trip feature serves to automatically trip the turbine when measured vibration exceeds a preset limit, thus preventing possible rotor, shaft, or bearing damage. The

. setpoint value was'changee such that a turbine trip will occur well in advance of vibration levels which could result'in damage. ' The associated EPIC alarm and recorc'v high alarm continue to function at their more wnservative settings to provide advance warning of increasing vibration.

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Update FSAR Table 8.5-1 per Calculation PS 79 S5fety Evalua'. ion: 3262 FSAR Section: T8.5-1 The FSAR table was revised to reflect EDG loading changes based on an updated

. analysis incorporating a new time period (10 to 120 minutes), an evaluation of EDG operation at maximum generator voltage and frequency, Calculation Comment Sheets written against Calculation PS 79, revised pump loading criteria, and a slightly different

- method for evaluating motor operated valve loading.

lhe change did not involve an unreviewed safety question. The updated EDG loading analysis was performed using a new load flow computer program, ETAP, which models all electrical distribution system components and calculates total EDG loading including system losses. The ETAP computer program replaces the existing tabular listings of EDG loads. The analysis showed that maximum steady state load on the emergency diesel generators due to the worst case accident and operational transients is within the EDG's ratings and therefore the required safety related components will receive adequate electrical power. At least one EDG will be available to carry the emergency service bus loads required for the safe shutdown of the reactor, maintain the safe shutdown condition, and operate all auxiliaries necessary for station safety assuming the worst case single failure.

Installation of Class i Nitrogen Supply to the SRV Accumulators Safety Evaluation: 3185 PDC: 96-019 FSAR Sections: 4.4.5,10.11.3.2, 5.4.3, T5.2-3, T5.2-4, TL.2-1 4

As a result of field walkdowns for USI-A46, pneumatic piping inside the drywell was identified as seismically vulnerable. The accumulators are seismically qualified, but the supply is limited and subject to long term leakage. This modification provides a Class I manual makeup system to allow existing RN Accumulators to be recharged with j

nitrogen from outside the drywell, ar.d thus maintain pressure control capability beyond the 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or original design.

This change did not involve an unreviewed safety question. The modification relieves the time sensitivity of attaining SDC and precludes repressurization of the reactor coolant system (challenge RPV P/T limits). It provides a Class I passive piping route in parallel with a non-Class I piping system whose failures have been evaluated. The additional piping is separated from existing piping by a check valve which is the same isolation boundary as the existing routing from Non-Q source. The added equipment will not caure the RN to open or preclude it from opening when required by setpoint, ADS, or manual initiation. The existing piping system failures of loss of pressure and overpressure have been addressed and bounded by the conclusion that the existing accumulator's pressure rating and relief valve's capacity are greater than the makeup capability.

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Clarification of Design Function for EDG HVAC Dampers Safety Evaluation: 397 FSAR Section:.10.9.0 9 This evaluation was prepared to clarify FSAR Section 10.9.3.9 with respect to the positioning of "A" EDG outside air and exhaust damper control valves. The prior FSAR wording was not clear regarding the design relationship of these damper velves and the EDG start capability. This relationship is that the control logic fer the EDG dampers takes and EDG start signal and repositions the damper-solenoid valve.

The repositioning of the solenoid valve dumps air for the EDG HVAC control system, opening dampers to allow for proper EDG room cooling and radiator cooling flow.

This change to the FSAR to clarify the EDG damper control design did not involve an unreviewed. safety question.'.The change is. administrative in r,ature to ensure the FSAR properly reflects the design feature of the EDG HVAC dampers.

EDG Design Basis Loading Revisions

_ Safety Evaluation: 3166 FSAR Section: T8.5-3, T8.5-4 This evaluation was prepared to clarify FSAR Section 8.5 regarding design basis loading of the.EDGs. The continuous rated and overload capacities were revised to reflect the design documentation collected and reviewed as part of the design basis recovery effort. Additional information was also added to the FSAR for the EDG inspection and overhaul interval schedule which establishes the maximum EDG load as a function of time. Accordingly, allowable operating times were established between major inspections or overhauls for various loadings.

This evaluation _did not involve an unreviewed safety question. The FSAR changes are based on technical. data supplied by the EDG manufacturer. The continuous and overload rating capacities establish design maintenance intervals, not operational limitations.

Derign Description Change for SSW Sample Valves Safety Evaluation: 3173 PDC: 97-010C FSAR Section:.10.7.5 LThis evaluation was prepared in support of a change to the FSAR SSW description of the purpose for sample valves installed in each cooling water loop between the pumps and the heat exchangers.

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l This change did not involve an unreviewed safety question. The change removed the description of the SSW valves that formerly provided access to the headers for a pitot tube transverse. These valves instead are used for venting air out of the system to obtain air free fluid during SSW pump flow testing.

Containment Heatup Analyses Safety Evaluation: 3127, supercedes SE 3118 FSAR Sections:

4.8.5.2, 4.8.5.4, 4.8.5.5. T4.8-1, T5.2-1,10.5.5.2,10.5.5.3, T10.5-2, 14.5.3.1.2,14.5.3.1.3, T14.5-1, T14.5-6, F14.5-9, F14.5-10, F14.5-12, F14.5-13, F14.5-16, F14.54. F14.5-18, F14.5-19.

This evaluation was conducted to demonstrate the current licensing basis requirements are met for the containment heatup analyses which were updated and approved by the NRC in License Amendment 173. Specifically, the Containment Heat Removal systems were evaluated at conditions following the bounding Design Basis Loss of Coolant Accident (DBA-LOCA), Steam Line Breaks (SLBs), and other transient and shutdown cases using an updated decay heat method and an Ultimate Heat Sink (UHS) temperature of 75 F. RHR and Core Spray Pump Net Positive Suction Head (NPSH) was evaluated for the bounding conditions of the DBA-LOCA based on the current design basis analysis for LOCA-generated debris, new RHR and Core Spray Pump suction strainers, and requirements imposed by the NRC Safety Evaluation Report (SER) for License Amendment 173. FSAR sections affected by the updated analysis have been revised accordingly.

This evaluation did not involve an unreviewed safety question. It ensures that the current licensing basis requirements for the Containment Heat Removal systems, as defined in the FSAR and the NRC SER for License Amendment 173, are, net. The changes to the containment accident analyses satisfy an NRC stipulation cc.ained in Amendment 173 requiring the containment heat removal analyses using the ANSI /ANS 5.1-1979 decay heat methodology to include a 2a uncertainly factor for the fission product decay heat. The scope of this safety evaluation included a review of all potential effects from the updated decay heat on the containment heat removal analyses using a 75 F SSW heat sink.

l Installation of a New Liquid Radwaste Treatment Process Safety Evaluation: 3012 PDC: 92-073 FSAR Sections. 9.2A.1,9.3.4.4 This change installed a new liquid radwaste treatment process to replace the flatbed filter and radwaste demineralizer system as the preferred system. The new system, a 21 of 22

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• Thermex" filtration system meets or exceeds the flatbed filter's performance, but at a significantly reduced solid waste generation amount. The flatbed system is isolated and will be used only as a backup to the new system.

The Thermex system is a stand alone skid which receives the waste stream diverted from the flatbed filters and transfers it back to the plant's clean waste or treated water tanks located in the radwaste section. The system is a multi-staged, modular liquid waste concentrator consisting of a suspended solid separator, suspended solid polishar, reve se osmosis unit, and mixed bed demineralizer.

This change did not involve an unreviewed safety question. The Thermex system meets the existing water chemistry requirements and.is not susceptible to any failure that has not been already postulated an designed to be controlled. The liquid radwaste I

process system is not safety related and there is no safety related equipment in the area where the Thermex system is installed.

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