ML20045E683

Safety Evaluation Supporting Amend 175 to License DPR-50

ML20045E683
Person / Time
Site:	Crane
Issue date:	06/23/1993
From:	Office of Nuclear Reactor Regulation
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ML20045E673	List: ML20045E672 ML20045E683
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NUDOCS 9307020314
Download: ML20045E683 (19)

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UNITED STATES

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WASHINGTON, D.C. 20553-0001

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.175 TO FACILITY OPERATING LICENSE NO. DPR-50 METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY GPU NUCLEAR CORPORATION THREE MILE ISLAND NUCLEAR STATION. UNIT NO. 1 DOCKET NO. 50-281

1.0 INTRODUCTION

By letter dated June 24, 1992, as supplemented on May 28, 1993, GPU Nuclear Corporation (GPUN/ licensee) proposed to revise the Three Mile Island, Unit 1 (TMI-1) Technical Specifications (TS) to extend the interval' for performance of selected surveillances to coincide with a 24-month operating cycle.

The proposed amendment would modify certain surveillances that are specified with an 18-month surveillance interval to accommodate a 24-month fuel cycle.

Guidance on the proposed TS changes were provided by Generic Letter 91-04, dated April 2,1991, to all holders of operating licenses or construction permits for nuclear power reactors.

2.0 EVALUATION 2.1 Gene.ral Chances To accommodate a 24-month fuel cycle, the licensee has proposed to modify some of the TS that presently imply a maximum 24-month surveillance interval to state that they are to be performed each REFUELING INTERVAL. The licensee proposed defining REFUELING INTERVAL in Section 1.2.8 as "once per 24 months."

With this change in surveillance intervals, the TS for some surveillances will no longer state that they are to be performed during shutdown.

In addition, the licensee proposed changes to Table 1.2 in the Definition Section of the TS.

to define a REFUELING INTERVAL with the existing "R" notation for certain surveillances that are generally performed during a refueling outage.

Under the definition of FREQUENCY NOTATION (Section 1.25), the provision for a 25%

maximum' allowable extension time is retained. - Therefore, with those surveillances having an "R" FREQUENCY-NOTATION, the maximum allowable time between surveillances would become 30 months. To maintain the present maximum time between sarveillances on those instruraents for which justification to extend the maximum time to 30 months has not yet been provided, the licensee proposed and defined a new FREQUENCY NOTATION of "F."

The maximum time between surveillances for those instruments with an "F" notation is 24 months (i_.e., the"25%. extension does not apply). Those changes are consistent with-the guidance'provided in Generic Letter 91-04 and-are, therefore, acceptable.

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i The licensee proposed changes to the Bases Section of TS Definition 1.25 on the 25% allowance for extending surveillance intervals. These changes clarify that refueling interval surveillances should not be performed during power operation unless such action is consistent with safe plant operation. The removal of the qualification to perform some surveillances during shutdown and.

the clarification of these Bases are consistent with the. guidance provided in Generic Letter 91-04.

t 2.2 Protective Instrumentation To support the proposed changes to accommodate a 24-month fuel cycle, the licensee examined calibration data from surveillance and' maintenance records-and confirmed that instrument drift has not, except on rare occasions, exceeded acceptable results. Therefore, historical data do not indicate any problems that would preclude an increase in the intervals for instrument calibration. The licensee has determined the values of drift for each type of instrument and application based on historical calibration data. The licensee.

provided a description of the methodology and assumptions used to determine the rate of instrument drift with time.

The licensee used this information'to determine the projected values of instrument drift that could occur with an increased calibration interval up to a bounding limit of 30 months.

The licensee compared the projected instrument drift. errors to the values of drift used in the analyses to determine acceptable,setpoints for safety systems. The licensee has confirmed that safety limits and safety analysis-assumptions are not exceeded with the consideration of the instrument ~ drift errors associated with the increase in instrument calibration intervals.

Therefore, no changes were required-in safety system setpoints'or safety analysis.

Calibration and~ drift histories were reviewed for the following instruments:

TS Table item Instrument 4.1-l '

8 High Reactor Coolant Pressure 9

Low Reactor Coolant Pressure 12 RCP/ Flux Comparator 51.a.2 HSPS-Loss of.all RC Pumps 15.a HPI Analog RCS Pressure 17.a LPI Analog RCS Pressure 19.f RB Emergency Cooling and Isolation - Line Break Isol.

22 Pressurizer _ Temperature 26 Pressurizer Level 23 Control Rod Absolute Position 24 Control Rod Relative Position 37 Reactor Building Sump Level 4.1-4 4

Containment Water Level 4.1-1 38 OTSG Full Range Level 39

' Turbine Overspeed Trip

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- TS Table Item Instrument (Continued) 42 Diesel Generator Protective Relaying (Includes Sync Check Relay, Reverse Power Relay, Loss of Field Relay,-

Negative Sequence Relay, Thermal Overload Relay, Field -

Ground Relay, Up-to Voltage Relays, Overvoltage Relay, Neutral Ground Relay, Field Overload Relay, Up-to-Frequency Relay, and 30% Differential. Relay) 4.1-1 43 4KV ES Bus Undervoltage Relays-(Includes Degraded Grid-Relays and Loss of Voltage Relays).

44 RCS Pressure Decay Heat Valve Interlock Bistable 45 Loss of Feedwater Reactor Trip

-47.b PORV Acoustic Flow Monitor 48 PORV Setpoints 49 Saturation Monitor 51.a.4 HSPS EFW Auto Initiation on OTSG Low Level 51.b HSPS MFW Isolation on Low OTSG Pressure-51.c.1 HSPS EFW Flow Control Valve Cont, System Level Loops 51.c.2 HSPS EFW Flow Control Valve Cont System Controllers 51.d HSPS Train Actuation Logic 52 Backup Incore Thermocouple Display 2.2.1 High/ Low Reactor Coolant (RC) Pressure-These instruments provide trip signal

• to the Reactor Protection 'ystem (RPS)l S

on high pressure to protect overpressurization of the system and on low pressure to prevent departure from nucleate boiling-in the core.

The licensee-reviewed pressure trip actuation setpoints from 1982 to 1990 to assess drift-

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history.

Random variations'in the setpoints were.. projected: upwards to a-maximum of 30 months between calibrations.and the projected-error figure of 12.0% was determined to be within the assumed 12.48% loop error band.

2.2.2 RCP/ Flux Comparator,. Heat Sink Protection System (HSPS) Loss of all RC Pump Circuitry This instrumentation provides a trip signal to the RPS if reactor power a's 1

indicated by neutron flux is too high relative to the number and location of-Reactor Coolant Pumps (RCPs) operating.

Evaluation of surveillance data. from 1982 to 1990 indicated that variations in the calibration data were random about the'setpoints and.that'no recalibrations were required..The same was true of the trip time delay variations.

Shiftly. channel checks and monthly' channel. tests. provide' additional' assurance of component reliability.

2.2.3 RC Pressure High' Pressure / Low Pressure, Injection Analog Channels These instruments provide low RC pressure signals to.the High Pressure ;

Injection (HPI) and Low Pressure-Injection (LPI) systems'to initiate system operation in the event of-decreasing pressure-(loss of inventory).

The-pressure transmitters, located in the Reactor Building'(RB), are-checked for-calibration each refueling interval' and.the electronic portions'- of-the i

circuitry are checked on a monthly basis.

Setpoints are 1640 psig and 540 psig, respectively. TS requirements are 1600 psig and 500 psig, respectively.

The Final Safety Analysis Repor.t (FSAR) accident analysis assumption for HPI initiation is 1480 psig, which calculates to a 6.4% margin between the actual setpoint and the FSAR requirement. Surveillance data were assessed for the period 1985 to 1990 to predict drift over a 30-month period. - The worst case combination drift error in the positive direction was calculated to be 4.74%.

Therefore, calibration drift over a 30-month period is within the 6.4% margin between the nominal setpoint and that assumed in the accident analysis.

Drift error in the negative direction is in the conservative direction and therefore is not of a safety concern.

Shiftly channel checks and monthly channel tests provide additional assurance of component reliability.

2.2.4 RB Emergency Cooling and Isolation - Line Break Isolation Signal This instrumentation provides RB isolation from the Intermediate Closed Cooling Water (ICCW) system or the Nuclear Services Closed Cooling Water (NSCCW) system in the event of a low surge tank level in either of those systems concurrent with an Engineered Safeguards (ES) actuation signal.

The licensee reviewed historical calibration data and calculated maximum errors assuming calibration at 30-month intervals.

These errors, including the estimated maximum drift and accounting for random uncertainty associated with the level transmitters, ranged from -9.44% to +0.31% for the ICCW instruments.

This translates to 7.245 inches for the maximum negative error and 8.0248 inches for the maximum positive error.

The setpoint is 8.0 inches with a maximum high error of 0.05 inches. The 8.0248 error falls within this tolerance.

Instrument actuation due to negative errors is in the conservative direction and, therefore, has no safety consequences.

2.2.5 Pressurizer Temperature and Level The pressurizer temperature instrument is used to generate input used to temperature-compensate the signals generated by the six independent pressurizer level instruments.

Pressurizer level signals are used for a number of control functions including RC inventory control by makeup and letdown, deenergizing pressurizer heaters on low level, and alarms. The licensee evaluated calibration data from 1985 to 1990 to predict drift and overall uncertainty assuming 30 months between calibrations. This evaluation showed that the drift and uncertainties at 30 months would be within the available margins. This instrumentation is not related to any FSAR safety analysis assumptions; therefore, the specific criteria in Generic Letter 91-04 were not addressed in the licensee's submittal.

2.2.6 Control Rod Position Two control rod position indication systems, absolute position indication (API) and relative position indication (RPI), provide rod position indication to the control room. The licensee evaluated data from the last five surveillances and determined that only one indication (out of 69) in each of the two systems was out of tolerance during one surveillance.

Prediction of T

drift assuming 30 months between surveillances indicated-that drift would remain within the specified tolerance range. _This. instrumentation is not related to any FSAR safety analysis assumptions; therefore, the specific.

criteria in Generic Letter 91-04 were not addressed in the licensee's submittal.

2.2.7 RB Sump Level / Containment Water Level This instrumentation provides alarm and control functions related to RB sump-level including high and low-low level alarms in the control room, a low level interlock to close the RB sump drain valve and a postaccident containment water level instrument required by Regulatory Guide (RG) 1.97.

An evaluation of historical calibration data was performed to predict drift errors assuming 30 months between calibrations.

The calibration tolerance allowed in procedure is 12.0%.. The predicted drift errors all fall within this tolerance.

The predicted maximum drift error in one of the alarm setpoints was 3.0%.

This would result in an alarm delay of 2.7 inches (69.0 to 71.7 inches). The licensee's submittal stated that this would still allow operators sufficient time to respond prior to sump overflow at 90 inches.

-This instrumentation is not'related to any FSAR safety analysis assumptions; therefore, the specific criteria in Generic letter 91-04 were not addressed in.-

the licensee's submittal.

2.2.8 Once Through Steam Generator (OTSG) Full Range _ Level.

This instrumentation is used to provide operators _with OTSG 1evel indication during normal heatups and cooldowns as well as postaccident as required by RG 1.97.

Review of historical calibration data' indicated 'that drift error at 30 months could be larger.than loop error calculation assume. LHowever, there;is no accuracy requirement specified'for this instrumentation and it is not used-by the operators to take action in a design basis event and it does not interface.with any control or protection system.

The licensee therefore concluded that extending surveillance.to a maximum of 30 months would. result-in no change to this instrumentation's functional capability. _This instrumen-tation is not related to any FSAR safety analysis ' assumptions; therefore, the specific criteria in Generic letter 91-04 were not addressed 11n the licensee's.

submittal.

2.2.9 Turbine Overspeed Trip This instrumentation provides overspeed protection for the main turbine.:;The mechanical and-electronic portions of the overspeed trips are nominally set' at 110% and 112%,.respectively.

Evaluation of test data since -1985 indicated -

that the' mechanical device has consistently actuated slightly above 108%.

Since this is conservative, the licensee has elected to not readjust the setpoint' closer to 110%. The electronic trip' has: consistently. operated'within.

the specified band of 2,016, +0,_-10 rpm. This instrumentationLis not-related'

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to any FSAR safety analysis assumptions; therefore, the specific criteria in Generic Letter 91-04 were not addressed in the licensee's submittal..

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i 2.2.10 Diesel Generator Protective Relaying These relays ensure that the two emergency diesel generators remain capable of supplying emergency power during degraded grid or loss of offsite power conditions.

They also ensure that the generators are synchronized and at proper voltage and frequency before being paralleled to their respective buses.

Highly reliable synchronization check relays were recently installed that have manufacturer's specified drift tolerances well within those specified in the licensee's procedures. Maintenance records for the reverse power relay surveillances since 1984 indicate that, with one exception, the relays have functioned within specified tolerances.

Loss of field relays, in testing since 1984, have generally functioned within specified tolerances.

However, some drift has been experienced with the " telephone" unit that provides a delay of 12-18 cycles before tripping the diesel to allow'for momentary transients.

The licensee has justified that these drifts were acceptable from the standpoint that the generator was still protected with significant margin.

The negative sequence relays have remained within the required tolerance band (without adjustment) as shown in surveillance testing since 1984. All other diesel protective relays have either demonstrated an acceptably low drift rate during surveillances since 1984 or have been replaced with new state-of-the-art relays with manufacturer's tolerance well within those specified in TMI-l procedures.

This instrumentation is not related to any FSAR safety analysis assumptions; therefore, the specific criteria in Generic Letter 91-04 were not addressed in the licensee's submittal.

2.2.11 4KV Engineered Safeguards (ES) Bus Undervoltage Relay There are two types of relays associated with this instrumentation, degraded grid relays and loss of voltage relays. Their purpose is isolate the 4KV ES buses from degraded voltage offsite power events and start the emergency diesel generators to supply these buses.

Historical surveillance data for these relays since 1984 have indicated that the "as found" pickup and dropout data settings during this period remained within the acceptance criteria, even for periods up to 36 months without recalibration. This instrumentation is not related to any FSAR safety analysis assumptions; therefore, the specific criteria in Generic Letter 91-04 were not addressed in the licensee's submittal.

2.2.12 RCS Pressure Decay Heat Removal (OHR) System Valve Interlock Bistable This instrument provides an interlock to preclude opening the DHR System isolation valves at pressures greater than 400 psig to prevent overpressur-ization of the system. TM1-1 procedures require calibration of this bistable once per month.

The licensee's submittal stated that since this instrument is already calibrated monthly, extending the calibration required by TS to 24 months has no impact on the functional capability.

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2.2.13 Loss of Feedwater Reactor Trip The purpose of this instrumentation is to provide an anticip& tory reactor trip response to a loss of feedwater that is faster than can be provided by the high RCS pressure trip. This instrumentation monitors control oil pressure on the main feedwater pumps.

Evaluation of historical surveillance data predicts that the maximum drift over a 30-month period would be -0.3 psig. The predicted uncertainty in the setpoint over a period of 30 months, accounting for drift and uncertainty of the model, is +1.89 to -4.18 psig.

The licensee's submittal states that the specified tolerance is 14.4 psig and that the expected uncertainty in a 30-month period is within the tolerance band.

This instrumentation is not related to any FSAR safety analysis assumptions; therefore, the specific criteria in Generic Letter 91-04 were not addressed in the licensee's submittal.

2.2.14 Power-0perated Relief Valve (PORV) Acoustic Flow Monitor This instrument provides an alarm in the control room indicating that there is flow through the PORV and senses mechanical vibration.

An evaluation of historical calibration data was performed to predict drift errors assuming 30 months.between calibrations. The calibration tolerances allowed in procedures enveloped the predicted maximum drift in all cases except one, the amplifier has a specified tolerance of 1% and the drift could be as high as_+1.047%.

The licensee considers this acceptable since it is in a conservative direction. This instrumentation is not related to any FSAR safety analysis assumptions; therefore, the specific criteria in Generic Letter 91-04 were not addressed in the licensee's submittal.

2.2.15 PORV Setpoints The purpose of the PORV is to minimize challenges to the Code safety relief valves on the RCS and to prevent low temperature overpressurization events.

The TMI-1 TS require a setpoint verification check on a monthly basis.

The setpoints are recalibrated, if necessary, and retested based on the monthly surveillance. The only additional step that occurs during the refueling interval surveillance is that the valve is actually stroked in addition to checking the setpoint.

2.2.16 Saturation Margin Monitor This monitor warns operators of an approach to saturation conditions in the RCS by comparing pressure to temperature. The temperature detectors are passive devices and do not require calibration. An evaluation of historical calibration data was performed to predict drift errors assuming 30 months between calibrations. The calibration tolerances allowed in procedures (10.25%) enveloped the predicted maximum drift in all cases.

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N 2.2.17 Heat Sink Protection System (HSPS)

The purpose' of this system is to isolate main feedwater (MFW) and initiate emergency feedwater (EFW) in the event of a low OTSG level condition such as from a main steam line break. OTSG level is the signal used for EFW~ and OTSG pressure is the predominant signal used for MFW isolation.

Portions of.the HSPS are given ' calibration checks on 'a quarterly basis including testing of-the setpoints, coincidence logic, and outputs up_ to and. including valve actuation and pump start. The licensee evaluated historical surveillance _ data since installation of the system in 1987. The OTSG-level loop instruments have an assumed accuracy of 11.3% for the startup range instrument and

+10/-20% for the operating range instrument. The calculated maximum error, based on the past five surveillances, is il.22% and +4.8 to -14.4%,

respectively.

Similar results were obtained for the EFW control valve control system controllers.

The licensee therefore concluded that extending the HSPS refueling interval calibrations would not affect the safety function of the system.

2.2.18 Backup Incore Thermocouple Display This system is a backup diverse means, independent of the plant computer, of determining incore temperature distribution and is used to determine margin to saturation in the natural circulation mode.

Surveillance data were assessed for the period 1986 to 1990 to predict calibration uncertainties, accounting for drift and model/ random uncertainties, over a 30-month period. The worst case uncertainties for the signal converters-were calculated to be.+0.03 VDC and -0.02 VDC. _The tolerance range specified for the converters is 10.05 VDC.

Likewise, the uncertainties associated with the RTD Comparison checks were-

+2.89'F and -0.99'F, compared to the specified tolerance 'of 4*F.

For control room indication, the calculated uncertainties range from -5.09'F to +4.05*F compared to the specified tolerance of 18'F.

Uncertainties associated with thermocouple simulated input range-from -3.71*F to' +4.65'F_ compared to the specified tolerance of i8'F.

Therefore, the calibration drift for this instrument over a 30-month period is expected-to be well within the specified tolerances.

This instrumentation is not.related -to any FSAR safety analysis assumptions; therefore, the specific criteria in Generic Letter 91-04 were not -

addressed in the licensee's submittal.

2.2.19 Summary The licensee has confirmed that projected instrument errors _ caused by drift are acceptable to control plant parameters in order to'effect a safe shutdown with the associated instrumentation. Also, the licensee confirmed that it has checked all conditions and assumptions of the setpoint and safetycanalyses and appropriately reflected these in the acceptance criteria of plant. surveil-lance procedures for channel checks, channel functional tests,'and channel

- calibrations.

Finally, the licensee provided a description of the program for monitoring and assessing the effects of increased calibration surveillance intervals on instrument drift and its effect on safety.

The staff reviewed this information and concludes that the licensee.has adequately analyzed the effect of-increased calibration intervals on instrument drift and its effect on safety.

2.3 Post-accident Monitorina Instrumentation-To support the proposed changes to accommodate a 24-month fuel _ cycle, the licensee examined calibration data from surveillance and maintenance records and confirmed that instrument drift has not, except'on rare occasions, exceeded acceptable results. Therefore, historical data do not; indicate any problems that would preclude an increase in the intervals for instrument-calibration. The licensee has determined the values of drift' for each type of instrument and application based on historical calibration data.- The licensee provided a description of the methodology.and assumptions used to' determine-the rate of instrument drift with time. The licensee used 'this information to determine the projected values of instrument drift that could occur with an increased calibration interval up to a bounding limit of 30 months. The-licensee compared the projected instrument drift errors to the values of drift-used in the analyses to determine acceptable setpoints for safety systems.

The licensee has confirmed that safety limits and safety analysis assumptions -

are not exceeded with the consideration of the instrument' drift errors assoc-iated with the increase in instrument calibration intervals. Therefore, no changes were required in safety system setpoints or safety analysis.

Calibration and drift histories were reviewed for the following' instruments:

TS Table Item Instrument 4.1-4 2

Containment High Range Radiation 7

RCS Cold Leg Water Temperature 8

RCS Hot leg Water Temperature 9

RCS Pressure 10 Steam Generator Pressure 2.3.1 Containment High Range Radiation These monitors provide readings of RB radiation of an extremely large-magnitude following an-accident. The licensee' evaluated previous surveillance -

data from 1985 to 1990 to predict detector response after 30 months'with 95%

confidence. The evaluation predicted uncertainty values ranging from -15.23%

to +10.76%.

The alarm setpoint determination for each of these detectors assumes a system inaccuracy of 36%.

2.3.2 RCS Hot and Cold leg Water Temperature

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This instrument provides a post-accident monitoring function as' required in accordance with RG 1.97.

The licensee evaluated previous surveillance data from 1984' to 1990 for the hot leg temperature to predict detector response after 30 months with 95% confidence. The evaluation predicted uncertainty

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values ranging from -0.47% to +0.295%.

The allowable tolerance for these instrument channels i0,5%. The surveillance for-the cold leg temperature was not initiated until 1990; thus, only one set of-surveillance data was available. The licensee's submittal states that,'since the cold leg instrumentation is equivalent to that used for the hot leg temperature, the above uncertainty values can be used to extend the surveillance interval for the cold leg instrumentation as well.

2.3.3 Reactor Coolant System Pressure This instrument provides a post-accident monitoring function as required in accordance with RG 1.97.

The licensee evaluated previous surveillance data predict detector response after 30 months. The evaluation predicted uncertainty values ranging from -0.06% to +0.17%.

The allowable tolerance for these instrument channels is 10.25%.

2.3.4 Steam Generator Pressure This instrument provides a post-accident monitoring function as required in accordance with RG 1.97.

The licensee evaluated previous surveillance data predict detector response-after 30 months. The evaluation predicted uncertainty values ranging from -11.9% to +8.7%.

The allowable tolerance 'for these instrument channels is 112.5%.

2.3.5 Summary The licensee provided a description of the program for monitoring and assessing the effects of increased calibration surveilla'nce intervals on -

instrument drift and its effect on safety. The staff reviewed this information and concludes that the licensee has adequately analyzed the effect of increased calibration intervals on instrument drift and its effect on safety.

2.4-Eouipment TestJi The licensee performed evaluations of the following component tests:

TS Table item /TS Section Test or Component 4.1-2 1/4.7.1.1 Control Rod / Control Rod Drive System Functional Test 3

Pressurizer Safety Relief Va: es 4

Main Steam Safety Valves 2.4.1 Control Rod / Control Rod Drive System (CRDS)

This testing insures that the CRDS functions properly and that, upon a reactor trip signal, the control rods fully insert within the time period assumed in the FSAR..The licensee analyzed all surveillance test data since.TMI-l restart in 1985 and determined'that-no significant trend of-system degradation i

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has occurred with respect to control rod drop time. 'The test data all exhibit large margins to the drop time criteria. Although drop time increased slightly with time (age) due to wear, the licensee has found no correlation between rod drop time-and fuel cycle length.

2.4.2 Pressurizer Safety'(Relief) Valves In addition to the PORV, there are two Code safet.,

9f valves,on the-pressurizer to prevent overpressurization of the RL he TS requires one of the two valves (50%) be tested to verify proper lift i-reset pressure each' refueling period. The licensee analyzed all surveillana test. data since TMI-l restart in 1985 and determined that the' valves-have fail i the "as found" test five out of six times. On three occasions, the valve under test lifted at a pressure higher than the required 2500 11% psig (2475 to 2525 psig).

The highest lift pressure during surveillance testing was 2579'psig.

The licensee evaluated the implication of these valve failures in terms of-extending the surveillance interval to 24 months. The ASME Code requirement is that the relief valve setpoint must be 110% of design pressure (2750 psig) or less. The FSAR analysis also uses the 2750 psig ASME setpoint:as the assumed maximum RCS pressure during a startup accident. The licensee's submittal stated that a lift pressure 3-4% higher than the design'setpoint would not invalidate' the ' accident analysis.

The licensee's submittal ' stated that, since the service time for these valves at the time of test ranged from 9 to 36 months, there is no evidence that setpoint drift is a function of time in service and that, therefore, the proposed change in surveillance interval will have no effect on the safety function of: the pressurizer safety valves.

2.4.3 Main Steam Safety (Relief). Valves (MSSVs)

There are a total of 18 MSSVs on the four main steam headers to protect the OTSGs and main steam piping from overpressure'. The TMI-1. TS require that -25%

of. these valves be-tested each ' refueling period.

The ASME Code requires testing each valve at least once every 5 years. Therefore, the licensee has proposed to test 50% of the valves each-refueling in addition to extending the surveillance interval to 24 months. Review of'these valves showed that they too have a history of setpoint drift but normally 'in the low ' direction.

No valve has failed to lift or to have been incapable of:providing overpressure protection in accordance with the ASME Code. There is no. indication that drift is a function of time in service. The licensee concluded that the proposed change in surveillance interval will have no effect on the safety function of the pressurizer safety valves.

2.5 Re' actor Buildina The licensee performed evaluations of 'the following component. tests:

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TS Section-Test or Component i

4.4.1.3 RB Isolation ~ Valve Functional Tests'

.4.4.1.7(2)-

Operability of RB' Purge Valves 4.4.4.1.b.2 &

Hydrogen Recombiner System

'4.4.4.1.b.3 4.4.4.1.b.4 Hydrogen Recombiner System 2.5.1 RB Isolation Valve Functional' Tests Surveillance data for the last 4-refueling outages (13 data points)' were_

reviewed.(1985 to 1990)..All valves were found to be acceptable (valve closure and stroke' time) and no repairs or corrective: actions ~were required-

'for the stroking mechanisms. The licensee concluded that extending the surveillance interval to 24 months would have no effect on the safety function.

of the valves.

2.5.2 Operability of RB. Purge Valves This surveillance requires a visual examination' andLdurometer test toL determine to_ detect degradation, and assure timely cleaning,- lubrication,: and ~

seat replacement. ;The valves are also leak tested on a quarterlysbasis..

The seats in all four purge valves were replaced in 1985. Surveillances.since si that time have' indicated no significant. degradation of f thervalve seat, which y

is consistent with industry and vendor experience... The licensee. believes thatL j

the proposed change will have no effect on the~ safety function.of the valves.

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l 2.5.3 Hydrogen'Recombiner System This system would be.used in a postaccident situation _to control combustib' l'e '

. gas (hydrogen) concentration following a lossLof coolant accident..Ones.

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required surveillance involves a visual. examination :for abnormalities' and a i

functional < test to ensure the reaction chamber can be' maintained 'at:1200*F.for at least 4. hours. - Surveillance data' from-1985 to.1990 indicate that1the '

system is capable of performing-its; function and-thattthere is no significant.

degradation._ ' A semiannual functional test requires l. temperature Lto be 9

maintained for 2' hours and gives added assurance of: operability. A'second

'i surveillance' requires checking continuity of the electrical heater. circuits in-the reaction chamber.. This surveillance also indicates satisfactory results?

since 1985. The licensee believes that the7 proposed change will have.' no

.effect on the safety; function of the valvesJ 1

2.6 Emeroency Core Coolina System (ECCS) and RB Coolina System (RBCS)'

2.6.1 Emergency Loading _ Sequence.and Power Transfer l

TS ;4.-5:1.1.a requires 1 esting of the-load l sequence and transfer l to-.' emergency?

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~ oserc(diesel generators) capability for all-- engineered; safeguards (ES).

p equipmentionce eachl refueling: interval.. The licensee reviewed deficiencies that have been-noted (and dispositioned);during past conduct.of this?

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- surveillance.

A recurring' deficiency has-been poor repeatability. of the timer that starts the EFW pumps. The licensee plans to replace these relays'with a more reliable model in the future.

Past surveillance testing has demonstrated the overall operability and reliability of this system..The licensee believes that the proposed change will have no effect on the safety function of the system.

2.6.2 HPI Pump Functional Test TS 4.5.2.1 requires a test each refueling interval to verify operability of the HPI pumps by demonstrating acceptable HPI flow and proper valve movement.

Tests conducted between.1985 and 1990 achieved _ satisfactory results.

Quarterly testing of the pumps and valves under the inservice test;ng. program provides additional assurance of system operability.- The licensee; believes that the proposed change will have no effect on the safety function of the-system.

2.6.3 LPI Pump Functional Test TS 4.5.2.2 requires a test each refueling interval to verify operability of the LPI pumps by demonstrating acceptable HPI flow and proper. valve movement.

Tests conducted between 1985 and 1990 achieved satisfactory results.

Quarterly testing of the pumps and valves under the inservice testing program provides additional assuranceL of system operability.

The licensee ~ believes that the proposed change will have no effect on the safety function of the system.

2.6.4 Core Floodine ?ystem (CFS)

The CFS is a passive system designed to inject borated-water into the core-following an RCS depressurization due to a loss of coolant' accident- (LOCA).

The purpose of the refueling interval surveillance is to verify operation of the system by demonstrating-proper operation of the system's check valves and' isolation valves.

Evaluation of the surveillance tests conducted between 1985 and 1990 indicates that the valves have always performed satisfactorily.

LThe licensee believes that the proposed change will have no effect on the safety function of the system.

2.6.5 Reactor Building Spray System (RBSS)

The RBSS is designed to provide RB cooling and fission product removal following a.LOCA. The purpose _of the refueling interval surveillance is to verify operation _of the system by demonstrating. proper operation of the system's pumps and valves.

Evaluation of the surveillance tests conducted between'1985 and 1990' indicates that the pumps and valves have always performed ~ satisfactorily.. Quarterly testing of.the pumps and valves 'under the -

inservice testing program provides additional assurance of system operability.

The licensee believes that the proposed change will have no effect on the.

safety function of the system.

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. 2.6.6 Reactor Building Cooling and Isolation System (RBCIS)

The RBCIS removes heat from the RB following a LOCA. The purpose of the refueling interval surveillance is to verify operation of the system by demonstrating proper operation of the system's pumps and valves.

Evaluation of the surveillance tests conducted between 1985 and-1990 indicates that the pumps and valves have always performed satisfactorily. Quarterly testing of the pumps and valves under the inservice testing program provides additional.

assurance of system operability. The licensee believes that the proposed change will have no effect on the safety function of'the system.

2.6.7 Decay Heat Removal (DHR) System This system is designed to remove decay heat from-the core during shutdown periods and to provide a source of low pressure injection following a LOCA.

The purpose of the refueling interval survelliance is to visually inspect the system piping and components for leakage and to leak test the system.

Results of past surveillance testing (1985 to 1990) indicates that there has been.only minor leaks found in the past. The leak rate criteria'is less than 6 gallons per hour. The largest amount of DHR System leakage found during past testing was 0.13 gph.

Leakage has been well below the acceptance criteria in the past There is no indication that leakage is a function of time in service. The licensee believes that the proposed change will have no effect on the safety function of the system.

2.7 Emeroency Power System 2.7.1 Emergency Diesel Generators (EDG)

The EDGs are automatically started and load tested once every refueling interval per TS 4.6.1.b and TS 4.5.1.1 (see section 2.6.1 tbove).

Past surveillance testing has demonstrated the overall ' operability and reliability of this system. The licensee believes that the proposed change will have no effect on the safety function of the system. ' Additionally, the TS require monthly manual start and load testing and annual inspections and overhaul, providing added assurance of EDG reliability and availability.

2.7.2 Station Batteries

.The station batteries provide a source of emergency de power for various controls, EDG auxiliaries, vital instruientation, and de-powered motors under station blackout conditions. The THI-1 Individual Plant Examination (IPE) predicts that the frequency of station blackout occurrence at THI-1 is 8X10'5.

There are two' redundant and identical batteries, each consisting of 116' cells.

Each battery has a nominal terminal voltage of 250 VDC and is divided into two sections of 125 VDC. The 1A battery has tu red and yellow ECCS sections; the IB battery has the green and blue sections. The.1A battery was replaced in 1986 and the IB battery was replaced in 1988. The typical lifetime of the batteries is between 15 and 20 years.

. The purpose of the refueling outage surveillance is to verify that battery capacity is sufficient to meet the calculated load requirements.

The load test conducted at THI-1 in the past to meet this surveillance requirement consisted of a constant current disrSarge at 475 amps for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (the rated capacity of the battery) to verify battery capacity, referred to as a performance test. The last test, conducted in October 1991, showed the capacities of the 1A and IB batteries to be 108.3% and 105.8%, respectively.

Monthly surveillance testing of the batteries includes measuring specific gravity, cell temperature, electrolyte levels, and individual cell voltages.

In addition, the inter-cell and inter-tier electrical connection resistance is measured each refueling interval to verify that it is within 20% of the baseline resistance. The licensee's request stated that, based on this surveillance program, a precipitous failure of the battery is extremely unlikely.

During its review of extending this surveillance, the staff questioned the use of a 2-hour performance test in lieu of the simulated load duty cycle service test recommended by IEEE 450-1980.

In response to the staff's question, the licensee provided information from the battery vendor endorsing use of a performance test (in lieu of a service test) because of the unique nature of the TMI-l load duty cycle. The duty cycle has only 2 very short periods of time (about I minute each) when current demand is above 475 amps and these are very early in the 2-hour duty cycle.

The peak discharge current required in the TMI-1 duty cycle for the red, yellow, green, and blue sections of the two batteries is 673.7 amps, 660.1 amps, 466.9 amps, and 564.8 amps, respectively.

These peaks all occur within the first 11 minutes after the station blackout occurs. Discharging the battery at a constant current of 475 amps actually discharges more amp-hours than a service test would. The battery vendor concluded that, if the battery is shown to be capable of meeting capacity requirements at its 2-hour constant current discharge rate, it will also be capable of meeting its load duty cycle discharge' rate. The staff notes, however, that the performance test, as conducted in the past, never challenges the battery to deliver full current that would be required in a station blackout (i.e., 673.7 amps). During discussions of this deficiency with the staff, the licensee agreed to perform a modified performance test as discussed in Section 5.4 of proposed IEEE Standard 450-1993, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications" for future surveillance load tests. Typically, this is a simulated duty cycle consisting of just two rates; the highest one minute rate of the duty cycle followed by the test rate employed for the perfor.mance test.

Since the ampere-hours r uoved by a rated one-minute discharge represents a very small portion 9f the battery's capacity, the test rate can be changed to that for the performance test without compromising the results of the performance test. The staff considers this to be an acceptable alternative.

The staff questioned the lack of a provision in the THI-1 TS for taking action when a battery shows signs of degradation. The Revised Standard B&W TS (RSTS), based on IEEE-450 recommended practice, specifies a service test each refueling outage and a performance test every 5 years. The performance test is to verify that battery capacity is at least 80%. The RSTS also requires a performance test to be performed every 12 months if the battery shows degradation or has reached 85% of its service life expected for the application (normally about 17 years). Degradation is indicated when the battery capacity drops more than 10% from its capacity on the previous performance test or is below 90% of the manufacturer's rated capacity.

The recommended practice is to replace the battery if its capacity reaches below 80% of the manufacturer's' rating if the battery was sized using a 1.25 aging factor. The batteries at THI-l were sized using an aging factor of 1.164 and a temperature correction factor of 1.04. The licensee informed the staff that plant procedures require replacement of the battery when it reaches

.90% of capacity for the 1A battery and 83% for the IB battery. To compensate for extending the surveillance interval from 18 to 24 months, the licensee agreed to add a provision to the TMI-l TS to replace the battery during the-subsequent refueling outage if the battery drops to 85% of its capacity. This agreement is documented in a supplement to the TS change request dated May 28, 1993. The 85% capacity criteria for replacing the battery is within the design limits of the battery and is acceptable.

The May 28, 1993, submittal provided supplemental information that did not change the initial proposed no significant hazards consideration determination.

2.7.3 Pressurizer Heaters A surveillance test is required by TS 4.6.3 to demonstrate that pressurizer heater groups 7 and.8 can' be transferred from their normal power supply to the-emergency bus and energized and that the ES actuation interlock functions properly.

An evaluation of the four test results from 1985 to 1990 indicates that all tests.were successfully performed with no deficiencies related to the heaters, power supply or controls. The licensee states that the proposed change to the surveillance interval has no effect on the safety function of the pressurizer heaters'.

2.8 liain Steam Isolation Valves TS Section 4.8.2 requires closure time testing each refueling outage to verify that the main steam isolation valves close within 120 seconds. The licensee reviewed the results of 14 tests from 1985 to 1990 and verified thst the valves closed properly within the specified time during each test.

In-addition to the refueling interval testing, the valves are exercised on a monthly basis by shutting them 10% of full closure. The licensee states that the proposed change to the surveillance interval has no effect on the safety function of the main steam isolation valves.

2.9

' Decay Heat Removal-Capability TS Sections 4.9.1.4 and 4.9.1.5 require refueling interval surveillance testing to verify operability of the'EFW system components. The testing requires demonstration that each EFW pump will automatically start upon an actuation signal, that the EFW flow control valve responds to both automatic and manual control signals and that the motor-driven.EFW pumps can pump' water from the condensate storage tanks to the OTSGs.

In addition to this testing, quarterly testing is performed on each pump _tc verify. automatic start as part

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of the_ inservice testing program. The licensee reviewed test records from 1983 to 1990 to verify that no deficiencies were encountered that would inhibit the ability of the EFW system to pump water into the OTSGs. The licensee states that the proposed change to the surveillance interval has no effect on the safety function of the EFW system.

2.10 Reactor Coolant System Vents TS Section 4.11.1 requires refueling interval testing to verify operability of each power operated valve associated with the RCS venting system.

Stroke time and remote position indication are verified during this testing.

The licensee reviewed test records from.1983 to 1990 ta verify that no deficiencies were encountered that would inhibit the ability of the RCS venting system to perform its safety function. The licensee states that the proposed change to the surveillance interval has no effect on the safety function of the system.

2.11 Air Treatment Systems 2.11.1 Emergency Control Room Air Treatment System This system provides a habitable environment in the control room in the event of release c. airborne radioactivity or toxic gases.

TS Sections-4.12.1.1 and.-

4.12.1.3 requi, a surveillance testing each refueling interval or every 18-months, whiche/cr comes first, to verify proper operation of the system dampers, response to an automatic initiation signal, and proper pressure drop across the HEPA filter and charcoal adsorber bank at rated flow.

In addition to_the refueling interval testing, this system is required to operate at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every month to verify proper operation. The licensee reviewed past test records to verify that the system has reliably demonstrated its abil_ity.

to perform its safety function. The licensee states that the proposed change to the surveillance interval has no effect on the safety function of the system.

2.11.2 Reactor Building Purge Air Treatment System The purpose of this system is to purge the RB atmosphere in arder to reduce airborne radioactivity levels prior to personnel. entry into the RB and during hydrogen purging of the RB. 'The purpose of the refueling interval testing is to verify proper. system air flow rate and pressure drop across the HEPA filter / charcoal adsorber bank. The purge isolation velves are not allowed to be opened to more than about 30% by TS and are required to be' closed except

. for certain reasons to purge-the RB. Therefore, the probability of the. filter.

banks becoming clogged to the point:of restricting flow is very small.

Past

-testing has indicated.that-the. pressure drop increases by-about 0.5 inches _of..

water every 24 months compared.to the maximum of 6 inches.

The licensee states-that the proposed change to the surveillance interval has no effect on the safety function of the system.

2.11.3 Auxiliary and Fuel Handling Building Air Treatment System The purpose of this sy:, tem is to purge the Auxiliary and Fuel Handling Building atmosphere in. order to maintain low airborne radioactivity levels and

. to maintain a slightly negative pressure in the building. The purpose of the refueling interval testing is to verify proper system air flow rate and pressure drop across the HEPA filter / charcoal adsorber bank and to verify iodine removal efficiency of at least 90%. TS also require-operation of the fans at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> each month, during which time a degrsdation in flow would be detected and corrected. Historically, the ' charcoal adsorber units maintain their efficiency for about 6-7 years. Surveillance testing in the past indicates that iodine removal efficiency after 2% years of operation is still about 96%. -Therefore, extending this surveillance testing from 18 months to 24 months is not expected to have a significant effect on the ability of the system to. perform its safety function.

2.12 Reactor Internals Vent Valves The eight reactor internals vent valves are designed to relieve pressure generated by steaming in the core following a LOCA. The purpose of the refueling interval surveillance is to conduct a visual inspection of_ the valve seating surfaces, verify valve position, and verify the amount of force required to operate the valves. An evaluation of surveillance results from 1985 to 1990 indicated no seat degradation and that all valves were in the closed position and operated freely. Therefore, extending this surveillance testing from 18 months to 24 months is not expected to have a significant effect on the ability of these valves to perform their safety function.

2.13 Shock Sucoressors (Snubbers)

The TS require snubber visual inspections on a schedule dictated by the number r

of snubbers determined to be inoperable during the previous inspection period with the longest interval presently specified being -18 months. if no snubbers are found to be inoperable. The TS also require functional testing of snubbers on a sampling basis. -The licensee reviewed surveillance test data from 1986 to 1990 and determined that, out of a population of 181. snubbers, only one snubber failed a visual test and no snubbers failed the functional test.

The licensee has proposed to extend the surveillance interval from 18 months to 24 months if no snubbers were found inoperable and from 12 months.to 16 months if one snubber is found inoperable.

Past. test results-indicate.that THI-l_ has implemented an effective snubber' maintenance program.

Therefore, extending these surveillance intervals would not be expected to have an effect on the snubbers fulfilling their safety function.

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2.14 Sumarv In summary, the licensee has evaluated the effect of the increase in the surveillance intervals from 18 months to 24 months and has concluded that the.

effect on safety is small. The licensee has confirmed that historical plant maintenance and surveillance data do not invalidate.this conclusion. The licensee also confirmed that the increase in surveillance intervals to accommodate a 24-month fuel cycle does not invalidate any assumption in the plant licensing basis. The staff reviewed this information and finds that the proposed TS changes do not have a significant effect on safety and are, therefore, acceptable.

S.0 STATE CONSULTATION In accordance with the Commission's regulations, the Pennsylvania State official was notified of 'he proposed issuance of the amendment. The state official had no comments.

4.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The staff has determined that the amendment involves no significant increase in the amounts and no significant changes in the types of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (57 FR 61112). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in conjunction with the issuance of.the amendment.

5.0 CONCLUSI0E The Commission has concluded, based on the considerations ~ discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation,in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and-(3) the issuance of the amendment will not be inimical to the common:

defense and security'or to the health and safety of-the public.

Principal Contributors:

Ronald W. Hernan Saba N' Saba Date: June 23,1993 k,

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