ML022600311

APS, Palo Verde Units 1, 2, and 3, Technical Specifications Bases Revision 18 Update

ML022600311
Person / Time
Site:	Palo Verde
Issue date:	09/11/2002
From:	Bauer S Arizona Public Service Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
102-04839-SAB/TNW/DWG
Download: ML022600311 (23)

Text

jL F Scott A. Bauer Department Leader Regulatory Affairs Tel 623/393-5978 Mail Station 7636 Palo Verde Nuclear Fax 6231393-5442 P 0 Box 52034 Generating Station e-mail sbauer@apsc com Phoenix, AZ 85072-2034 102-04839-SABITNW/DWG September 11, 2002 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station P1-37 Washington, DC 20555-0001

Subject:

Palo Verde Nuclear Generating Station (PVNGS)

Units 1, 2, and 3 Docket Nos. STN 50-52815291530 Technical Specifications Bases Revision 18 Update Pursuant to PVNGS Technical Specification (TS) 5.5.14, "Technical Specifications Bases Control Program," Arizona Public Service Company (APS) is submitting changes to the TS Bases incorporated into Revision 18, implemented on September 6, 2002.

The Revision 18 insertion instructions and replacement pages are provided in the Enclosure.

No commitments are being made to the NRC by this letter.

Should you have any questions, please contact Thomas N. Weber at (623) 393-5764.

Sincerely, SAB/TNW/DWG/kg

Enclosure:

PVNGS Technical Specification Bases Revision 18 Insertion Instructions and Replacement Pages cc: E. W. Merschoff J. N. Donohew N. L. Salgado

ENCLOSURE PVNGS Technical Specification Bases Revision 18 Insertion Instructions and Replacement Pages

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Cover page Cover page List of Effective Pages, List of Effective Pages, Pages 1/2 through Pages 1/2 through List of Effective Pages, List of Effective Pages, Page 7/blank Page 7/blank B 3.3.1-5/B 3.3.1-6 B 3.3.1-5/B 3.3.1-6 B 3.3.3-1/B 3.3.3-2 B 3.3.3-1/B 3.3.3-2 B 3.3.10-5/B 3.3.10-6 B 3.3.10-5/B 3.3.10-6 B 3.9.3-1/B 3.9.3-2 B 3.9.3-1/B 3.9.3-2 B 3.9.3-3/B 3.9.3-4 B 3.9.3-3/B 3.9.3-4 B 3.9.3-5/blank B 3.9.3-5/B 3.9.3-6

PVNGS Palo Verde Nuclear GeneratingStation Units 1, 2, and 3 Technical Specification Bases Revision 18 September 6, 2002

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B 3.6.1-4 0 B 3.7.2-1 0 B 3.7.2-2 0 B 3.6.1-5 0 B 3.6.2-1 0 B 3.7.2-3 6 B 3.7.2-4 0 B 3.6.2-2 0 B 3.6.2-3 0 B 3.7.2-5 PALO' VERDE UNITS 1, 2, AND 3 5 Revision 18 September 6, 2002

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RPS Instrumentation - Operating B 3.3.1 BASES BACKGROUND Measurement Channels (continued)

The LPD calculation considers APD, average power, radial peaking factors (based upon target CEA position), and CEAC penalty factors to calculate the current value of compensated peak power density. An LPD - High trip occurs when the calculated value reaches-the trip setpoint. The four CPC channels provide input to the four DNBR - Low and four LPD - High RPS trip-channels. They effectively act as the sensor (using many inputs) for these trips.

The CEACs perform the calculations-required to determine the position of CEAs within their subgroups for the CPCs. Two independent CEACs compare the position of each CEA to its subgroup position. If a deviation.is detected by either CEAC, an annunciator sounds and appropriate "penalty factors" are transmitted to all CPCs. These penalty factors conservatively adjust-the effective operating margins to the DNBR - Low and LPD - High trips. Each CEAC also drives a single visual display which isswitchable between CEACs.

The visual display indicates individual CEA positions from the selected CEAC.

Each CEA has two separate reedswitch assemblies mounted outside the RCPB. Each of the two CEACs receives CEA position input from one of the two-reed switch position transmitters on each CEA, so that'the position of all CEAs is independently monitored by both CEACs.

CEACs are addressed in LCO 3.3.3.

Bistable Trip Units Bistable trip units' mounted in the Plant Protection System (PPS) cabinet, receive an analog input from the measurement channels. They compare the analog input to trip setpoints and provide contact output to the Matrix Logic. They also provide local trip indication and remote annunciation.

There are four channel's of bistables. designated A, B. C, and D. for each RPS parameter. one for each measurement channel. Bistables de-energize When a trip occurs, in turn de-energizing bistable&relays mounted in the PPS relay card racks.

(continued)

'PALO VERDE UNITS 1,2.3 B 3.3.1-5 REVISION 18

I RPS Instrumentation - Operating B 3.3.1 BASES BACKGROUND Bistable Trip Units (continued)

The contacts from these bistable relays are arranged into six coincidence matrices, comprising the Matrix Logic. If bistables monitoring the same parameter in at least two.

channels trip, the Matrix Logic will generate a reactor trip (two-out-of-four logic).

Some measurement channels provide contact outputs to the PPS. In these cases, there is no bistable card, and opening the contact input directly de-energizes the associated bistable relays. These include the CPC generated DNBR - Low and LPD - High trips.

The trip setpoints used in the bistables are based on the analytical limits derived from the accident analysis (Ref. 5). The selection of these trip setpoints is such that adequate protection is provided when all sensor and processing time delays are taken into account. To allow for calibration tolerances, instrumentation uncertainties.

instrument drift, and severe environment errors for those RPS channels that must function in harsh environments as defined by 10 CFR 50.49 (Ref. 6), Allowable Values specified in Table 3.3.1-1, in the accompanying LCO, are conservatively adjusted with respect to the analytical limits. A detailed description of the methodology used to calculate the trip setpoints, including their explicit uncertainties, is provided in "Plant Protection System Selection of Trip Setpoint Values" (Ref. 7). The nominal trip setpoint entered into the bistable is normally still more conservative than that specified by the Allowable Value to account for changes in random measurement errors detectable by a CHANNEL FUNCTIONAL TEST. One example of such a change in measurement error is drift during the interval between surveillances. A channel is inoperable if its actual setpoint is not within its Allowable Value.

To maintain the margins of safety assumed in the safety analyses, the calculations of the trip variables for the DNBR - Low and Local Power Density - High trips include the measurement, calculational. and processor uncertainties and dynamic allowances as defined in the latest applicable revision of CEN-PSD-335-P, "Functional Design Requirements for a Core Protection Calculation" (Ref. 10) and CEN-PSD-336-P," Functional Design Requirements for a Control Element Assembly Calculator," (Ref. 11).

(continued)

PALO VERDE UNITS 1,2,3 B 3.3.1-6 REVISION 0

CEACs B 3.3.3 B 3.3 INSTRUMENTATION B 3.3.3 Control Element Assembly Calculators (CEACs)

BASES BACKGROUND The' Reactor Protective System (RPS) initiates a reactor trip to protect against violating the core Specified Acceptable Fuel Design Limits (SAFDLs) and breaching the Reactor Coolant Pressure Boundary (RCPB) during Anticipated Operational Occurrences (AOOs).- By tripping the reactor, the RPS also assists the Engineered Safety Features Systems in mitigating accidents.

The protection and monitoring systems have been designed to ensure safe operation of the reactor. This is achieved by

-specifying Limiting Safety System Settings (LSSS) in terms of parameters ;directly monitored by-the RPS. as well as LCOs on other reactor-system parameters and equipment performance.

The LSSS (defined in this Specification'as the Allowable Value)-.in conjunctionrwith-the LCOs,-establish the thresholds for protective system action to prevent exceeding acceptable limits during Design Basis Accidents.

During AOOs, which are those events expected to occur one or more times during the plant life, the acceptable limits are:

'The Departure from Nucleate Boiling Ratio (DNBR) shall

-be maintained 'above the Safety Limit (SL) value to prevent departure from nucleate'boiling:

° Fuel centerline melting shall not occur; and The Reactor.nCoolant System pressure'SL of 2750 psia shall'not be"exceeded.

Maintaining'the parameters within the above values ensures that the offsite dose will'be within the 10 CFR 50 (Ref. 1) and 10CFR 100 (Ref.-2) criteria-during ACOs.

Accidents are events that, are ,analyzed even though'they are not-expected to occur'during theplant life. The acceptable limit during accidents is that-the'offsite dose shall be maintained within an acceptable fraction of 10 CFR 100 (Ref. 2) limits. Different accident categories allow a (continued)

PALO VERDE UNITS 1,2.3 B 3.3.3-1 REVISION 0

I _____________

CEACs B 3.3.3 BASES BACKGROUND different fraction of these limits based on probability of (continued) occurrence. Meeting the acceptable dose limit for an accident category is considered having acceptable consequences for that event.

The RPS is segmented into four interconnected modules.

These modules are:

"* Measurement channels;

"* Bistable trip units:

"* RPS Logic: and

"* Reactor Trip Circuit Breakers (RTCBs).

This LCO addresses the CEACs. LCO 3.3.1, "Reactor Protective System (RPS) Instrumentation - Operating,"

provides a description of this equipment in the RPS.

The excore nuclear instrumentation, the Core Protection Calculators (CPCs), and the CEACs are considered components in the measurement channels of the Variable Over Power-High, Logarithmic Power Level - High, DNBR - Low, and Local Power Density (LPD) - High trips. The CEACs are addressed by this Specification.

All four CPCs receive Control Element Assembly (CEA) deviation penalty factors from each CEAC and use the larger of the penalty factors from the two CEACs in the calculation of DNBR and LPD. CPCs are further described in the Background section of LCO 3.3.1.

The CEACs perform the calculations required to determine the position of CEAs within their subgroups for the CPCs. Two independent CEACs compare the position of each CEA to its subgroup position. If a deviation is detected by either CEAC, an annunciator sounds and appropriate "penalty factors" are transmitted to all CPCs. These penalty factors conservatively adjust the effective operating margins to the DNBR - Low and LPD - High trips. Each CEAC also drives a single visual display, which is switchable between CEACs.

The visual display indicates individual CEA positions from the selected CEAC.

(continued)

PALO VERDE UNITS 1,2,3 B 3.3.3-2 REVISION 18

PAM Instrumentation B 3.3.10 BASES LCO 4. Reactor Coolant-System Pressure (wide range)

(continued)

Wide range RCS loop pressure is measured by pressure transmitters with a span of 0 psig to 4000 psig.

Redundant monitoring capability is provided by two trains of instrumentation. Control room indications are provided through the Qualified Safety Parameter Display System (QSPDS) visual display. The QSPDS visual display is the primary indication used by the operator during an accident. -Therefore, the PAM instrumentation Specification deals specifically with this portion of the instrument channel.

RCS pressure is also a Type A variable because the operator uses this indication to monitor the cooldown of the RCS following a steam generator tube rupture or small break-loss of coolant accident (LOCA). Operator actions to maintain a controlled cooldown, such as adjusting steam generator pressure or level, would use this indication. -Furthermore. RCS pressure is one factor that may be used indecisions to terminate reactor coolant pump'operation.

At PVNGS the RCS Pressure'(wide range) consists of:

RCA-PT-190A RCB-PT-190B

5. Reactor Vessel Water Level Reactor Vessel Water Level is provided for verification and long term surveillance of core cooling.

The Reactor.Vessel Water Level Monitoring System provides a direct measurement of the collapsed liquid levelabove the fuel-alignment plate. The collapsed level represents the'amount of liquid mass that is in the reactor vessel'above the core. Measurement of the collapsed water level is selected because it is a direct indication of.the'water inventory.

(continued)

PALO VERDE UNITS 1.2,3 B 3.3.10-5 REVISION 18

-- I PAM Instrumentation B 3.3.10 BASES LCO 5. Reactor Vessel Water Level (continued)

The collapsed level is obtained over the same temperature and pressure range as the saturation measurements, thereby encompassing all operating and accident conditions where it must function. Also, it functions during the recovery interval. Therefore, it is designed to survive the high steam temperature that may occur during the preceding core recovery interval.

The level range extends from the top of the vessel down to the top of the fuel alignment plate. The response time is short enough to track the level during small break LOCA events. The resolution is sufficient to show the initial level drop, the key locations near the hot leg elevation, and the lowest levels just above the alignment plate. This provides the operator with adequate indication to track the progression of the accident and to detect the consequences of its mitigating actions or the functionality of automatic equipment.

At PVNGS the Reactor Vessel Water Level is displayed on QSPDS A and QSPDS B.

6. Containment Sump Water Level (wide range)

Containment Sump Water Level is provided for verification and long term surveillance of RCS integrity.

At PVNGS, Containment Sump Water Level instrumentation consists of the following:

SIA-LT-706 SIB-LT-707

7. Containment Pressure (wide range)

Containment Pressure is provided for verification of RCS and containment OPERABILITY.

At PVNGS, Containment Pressure instrumentation consists of the following:

HCA-PT-353A HCB-PT-353B (continued)

PALO VERDE UNITS 1,2,3 B 3.3.10-6 REVISION 0

Containment Penetrations B 3.9.3 B 3.9 REFUELING OPERATIONS B 3.9.3 Containment Penetrations BASES BACKGROUND During CORE ALTERATIONS or movement of fuel assemblies within containment with irradiated fuel in containment, a release of fission product radioactivity within the containment will be restricted.from escaping to the environment when theLCO requirements-are met. In MODES 1.

2, 3, and 4. this is accomplished by maintaining containment OPERABLE as described ,in LCO 3.6.1. "Containment." In MODE 6, the potential-for containment pressurization as a

.result of an accident is not likely; therefore, requirements to isolate the containment from the outside atmosphere can be less stringent. The LCO requirements are referred to as "containment closure",rather than "containment OPERABILITY."

Containment closure means that-all potential escape paths are closed or capable-of being closed. Since there is no

• otential for containment pressurization, the Appendix J eakage criteria andtests are not required.

The containment serves to contain fission product radioactivity that may-be released from the reactor core following an accident,'such that offsite radiation exposures are maintained well within the requirements of 10 CFR 100.

Additionally, the containment structure provides radiation shielding from the fission products that may be present in the containment atmosphere'following accident conditions.

The containment equipment-hatch; whichis part of the containmentpressure boundary, provides 'ameans for moving large-'6quipment andcomponents into-and out of containment.

During CORE ALTERATIONS or movement of,irradiated fuel assemblies within containment.- the equipment hatch must be capable of being closed and-held in place by at least four I

,bolts. Good engineering practice dictates that the bolts required by'this LCO be approximately-equally spaced.

The containment-air locks, which are also part of the containment'pressure boundary, provide a means for personnel

" a.,ccess-during MODES 1,-2, 3..and 4 operation in accordance with'LCO,3.6.2. "Containment Air, Locks." Each air lock has doors atlboth-ends. ,The doors are'normally interlocked to prevent simultaneous opening when containment OPERABILITY is required. During periods of shutdown when containment (continued)

B'3.*9.3-1 REVISION 18 "PALO VERDE UNITS 1.2,3

I _________

Containment Penetrations B 3.9.3 BASES BACKGROUND closure is not required, the door interlock mechanism may be (continued) disabled, allowing both doors of an air lock to remain open for extended periods when frequent containment entry is necessary. During CORE ALTERATIONS or movement of irradiated fuel assemblies within containment, containment closure is required: therefore, the door interlock mechanism may remain disabled, but one air lock door must always remain closed.

The requirements on containment penetration closure ensure that a release of fission product radioactivity within containment will be restricted from escaping to the environment. The closure restrictions are sufficient to restrict fission product radioactivity release from containment due to a fuel handling accident during refueling.

The Containment Purge and Exhaust System includes two subsystems. The refueling purge subsystem includes a 42 inch supply penetration and a 42 inch exhaust penetration. The second subsystem, power access purge subsystem, includes an 8 inch supply penetration and an 8 inch exhaust penetration. During MODES 1, 2. 3, and 4.

the two valves in each of the refueling purge supply and exhaust penetrations are secured in the closed position.

The two valves in each of the two power access purge penetrations can be opened intermittently, but are closed automatically by the Engineered Safety Features Actuation System (ESFAS). Neither of the subsystems is subject to a Specification in MODE 5.

In MODE 6. large air exchanges are necessary to conduct refueling operations. The refueling purge system is used for this purpose and the valves are closed by the ESFAS in accordance with LCO 3.3.8, "Containment Purge Isolation Actuation Signal (CPIAS)."

The Power Access Purge System remains operational in MODE 6 and the valves are also closed by the ESFAS.

The other containment penetrations that provide direct access from containment atmosphere to outside atmosphere must be isolated on at least one side. Isolation may be achieved by an OPERABLE automatic isolation valve, or by a manual isolation valve, blind flange, or equivalent.

(continued)

PALO VERDE UNITS 1,2,3 B 3.9.3-2 REVISION 0

Containment Penetrations B 3.9.3 BASES BACKGROUND Equivalent isolation methods must be approved and may (continued) include use of devices designed to allow eddy current testing and sludge lancing of the steam generators. Devices which present a substantial restriction to the release of containment atmosphere may be considered equivalent.

APPLICABLE During CORE ALTERATIONS or movement of irradiated fuel SAFETY ANALYSES assemblies within containment' the most severe radiological consequences result from a fuel handling accident. The fuel handling accident is a postulated event~that involves damage to irradiated fuel (Ref. 2).- Fuel handling accidents, analyzed in Reference 2.'include dropping a single irradiated fuel' assemblyand handling-tool or a heavy object onto other irradiated fuel assemblies: 'The requirements of LCO 3.9.6. "Refueling Water Level-Fuel Assemblies," LCO 3.9.7, "Refueling Water Level-CEAs,"- and the minimum decay time of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to CORE.ALTERATIONS ensure that the release of fission product radioactivity, subsequent to a' fuel handling accident, results in doses that are well within the guideline values specified in 10 CFR 100. The acceptance limits for'offsite radiation-exposure are contained in Standard'-Review Plan Section 15.7.4. Rev. 1 (Ref. 3). which defines "well wiithin"_ 10 CFR 100 to be 25%

or less of the 10CFR 100,values.

Containment penetrations satisfy Criterion 3 of 10 CFR 50.36 (c)(2)(ii).

LCO This LCO limits the consequences of a fuel handling accident in containment by limiting the potential escape paths for fission product radioactivity released within containment.

The LCO requires any penetration providing direct access from the containment atmosphere to the outside atmosphere to be closed except for the OPERABLE containment purge supply exhaust penetrations and equipment hatch. For the OPERABLE I containment purge'supply and exhaust-penetrations, this LCO ensures that-these penetrations are isolable by a valve in the Containment Purge Isolation System. The OPERABILITY requirements for this-LCO ensure that the automatic purge valve closure-times specified in the-UFSAR can be achieved and therefore meet' the-assumptions used in the safety analysis to ensure releases-through-the valves are terminated, such that the radiological doses are within the acceptance limit. The equipment hatch is required to be kept free of obstructions that could impede its closure so I (continued)

REVISION 18

'PALO VERDE UNITS 1,2,3 B 3.9.3-3

Containment Penetrations B 3.9.3 BASES LCO it is capable of being closed with a minimum of four bolts (continued) should a fuel handling accident occur inside containment.

APPLICABILITY The containment penetration requirements are applicable during CORE ALTERATIONS or movement of irradiated fuel assemblies within containment because this is when there is a potential for a fuel handling accident. In MODES 1, 2, 3, and 4, containment penetration requirements are addressed by LCO 3.6.1, "Containment." In MODES 5 and 6, when CORE ALTERATIONS or movement of irradiated fuel assemblies within containment are not being conducted, the potential for a fuel handling accident does not exist. Therefore, under these conditions no requirements are placed on containment penetration status.

ACTIONS A.1 and A.2 With the containment equipment hatch, air locks, or any containment penetration that provides direct access from the containment atmosphere to the outside atmosphere not in the required status, including the Containment Purge Isolation System not capable of automatic actuation when the purge valves are open, the unit must be placed in a condition in which the isolation function is not needed. This is accomplished by immediately suspending CORE ALTERATIONS and movement of irradiated fuel assemblies within containment.

Performance of these actions shall not preclude completion of movement of a component to a safe position.

SURVEILLANCE SR 3.9.3.1 REQUIREMENTS This Surveillance demonstrates that each of the containment penetrations required to be in its closed position is in that position. The Surveillance on the open purge and exhaust valves will demonstrate that the valves are not blocked from closing. Also, the Surveillance will demonstrate that each valve operator has motive power, which will ensure each valve is capable of being closed by an OPERABLE automatic containment purge isolation signal.

(continued)

PALO VERDE UNITS 1.2,3 B 3.9.3-4 REVISION 18

Containment Penetrations B 3.9.3 BASES SURVEILLANCE SR "3.9.3.1 (continued)

REQUIREMENTS The Surveillance-is performed every 7 days during CORE ALTERATIONS or movement of irradiated fuel assemblies within the containment. The Surveillance interval is selected to be commensurate-with the normal duration of time to complete fuel handling operations. A surveillance before the start of refueling operations will provide two or three surveillance verifications during the applicable period for this LCO. As such. this Surveillance ensures that a postulated fuel handling accident that releases fission product radioactivity within the containment will not result in a release of fission product radioactivity to the environment.

SR 3.9.3.2 This Surveillance demonstrates that each containment purge valve actuates to its isolation position on manual initiation or on an actual or simulated high radiation signal. The 18 month Frequency maintains consistency with other similar ESFAS instrumentation and valve testing requirements. The CPIAS is tested in accordance with LCO 3.3.8, "Containment Purge Isolation Actuation Signal (CPIAS)." SR 3.6.3.5 demonstrates that the isolation time of each valve is in accordance with the Inservice Testing Program requirements. These surveillances performed during MODE 6 will ensure that the valves are capable of closing after a postulated fuel handling accident to limit a release of fission product radioactivity from the containment.

SR 3.9.3.3 This Surveillance demonstrates that the necessary hardware, tools, equipment and personnel are available to close the equipment hatch and that the equipment hatch is clear of obstructions that would impede its closure. The 7-day Frequency is commensurate with the normal duration of time to complete the fuel handling operations. The Surveillance is only required to be met for an open equipment hatch. If the hatch is closed, the capability to close the hatch is not required.

(continued)

PALO VERDE UNITS 1,2,3 B 3.9.3-5 .REVISION18

Containment Penetrations B 3.9.3 BASES REFERENCES 1. GPU Nuclear Safety Evaluation SE-0002000-O01, Rev. 0, May 20, 1988.

2. UFSAR, Section 15.7.4.

3. NUREG-0800, Section 15.7.4, Rev. 1, July 1981.

PALO VERDE UNITS 1,2.3 B 3.9.3-6 REVISION 18