ML17157A242

Application for Proposed Amend 132 to License NPF-14, Revising Tech Spec to Support Cycle 6 Reload

ML17157A242
Person / Time
Site:	Susquehanna
Issue date:	07/02/1990
From:	Keiser H PENNSYLVANIA POWER & LIGHT CO.
To:
Shared Package
ML17157A243	List: ML17157A242 ML17157A244 ML17157A245 ML18026A235
References
NUDOCS 9007120232
Download: ML17157A242 (11)

Text

BEFORE THE UNITED STATED NUCLEAR REGULATORY COMMISSION In the Matter of Docket No. 50-387 PENNSYLVANIA POWER &

LIGHT COMPANY PROPOSED AMENDMENT No ~ 132 FACILITY OPERATING LICENSE NO. NPF-14 SUSQUEHANNA STEAM ELECTRIC STATION UNIT NO ~ 1 Licensee, Pennsylvania Power & Light Company/ hereby files proposed Amendment No. 132 to its Facility Operating License No. NPF-14

. dated July 17, 1982.

This amendment contains a revision to the Susquehanna SES Unit 1 Technical Specifications.

PENNSYLVANIA POWER & LIGHT COMPANY BY:

H. W. Keiser Senior Vice President Nuclear Sworn this 'of to

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and subscribed

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before me 1990.

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Notary Public tVOTARIAL SEAL 4 j(gt)7 j ~OpttMs'K

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' )()702 Helen J. Vsotfer, rfotary Public t~m~~g7 gD<

City of Allentown, Lebi3b C~~nty, hly Commission Expires Apr. 4, 1993 Pa. I='

SUSQUEHANNA SES UNIT 1 CYCLE 6 PROPOSED STARTUP PHYSICS TESTS

SUMMARY

DESCRIPTION June 1990 II PENNSYLVANIA POWER 5 LIGHT COMPANY

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S U HANNA SES UNIT I CYC 6 PROPOSED STARTUP PHYSICS T STS

SUMMARY

0 SCRIPTION Susquehanna SES Unit I is planned to be shut down for its fifth refueling and inspection outage on September 8, 1990. During startup and initial Cycle 6 operation, PP8L plans to perform a series of startup activities and tests to assure that the reload core conforms to the design. A list of these proposed activities and tests along with a brief description for each is provided below.

I) Core Loadin Verification

Purpose:

To assure the core is correctly loaded per design.

Description:

The core will be visually checked to verify correct loading.

An underwater video camera or suitable device will be used to record fuel assembly serial numbers, orientations, core locations, and proper core plate seating. A review of the videotape will be performed and will serve as an independent verification of the core loading. Any discrepancies discovered will be promptly corrected and the affected areas reverified prior to Unit I Cycle 6 (UIC6) startup.

2) POW RP X In ut eck Validati n

Purpose:

To ensure the POWERPLEX Core Monitoring System input deck is updated correctly before the start of every new operating cycle.

Description:

This validation will ensure that POWERPLEX, the ANF software system designed to perform in-core monitoring of BWR cores, is correctly updated for monitoring UIC6 operation. Core monitoring calculations within POWERPLEX are performed by XTGBWR, a three-dimensional reactor simulation code. The POWERPLEX input deck consists of all constants needed for the execution of this code and subsequent calculation of the margin to thermal limits. These constants must be updated to reflect the new core loading prior to the star t of every new reload operating cycle in order to ensure satisfactory core monitoring. The deck is updated by ANF or PPKL and verified jointly by members of the PPEL Reactor Engineering group located at Susquehanna SES and the Nuclear Fuels Engineering group located at the corporate headquarters.

3) Co tr Ro F n i nal nser and With r w

Purpose:

To assure proper control rod function.

Description:

A control rod functional test, which includes mobility and overtravel checks, will be performed on each control cell loaded in its final configurati'on. Core subcriticality will be demonstrated and documented as each control rod is functionally tested.

4) h wn Mar in emonstrati n

Purpose:

To assure that at least the minimum required Shutdown Margin exists with the analytically determined strongest worth control rod fully withdrawn.

Description:

This test will verify that at least the required amount of Shutdown Margin is maintained without determining the actual amount of Shutdown Margin in the core. The analytically determined strongest worth control rod (or its symmetric counterpart) is fully withdrawn. Diagonally adjacent control rods are then slowly notched out (one at a time) and sub-criticality verified at each step, until the analytically determined reactivity worth of the diagonally adjacent control rods at their respective notch position just equals or slightly exceeds the required amount of Shutdown Hargin. Verification at this step that the core is still subcritical demonstrates that at least the required amount of Shutdown Hargin exists.

5) In-Se uence Critical and Shutdown Mar in Det rmination

Purpose:

a) To determine the actual amount of Shutdown Margin.

b) To compare predicted versus actual critical control rod positions.

Description:

This test will be performed as part of the normal startup.

Control rods are pulled in group order in their normal sequence until criticality is achieved. Taking into account the period and moderator temperature coefficient corrections, the Shutdown Hargin is determined by calculation. In addition, to assure that there is no reactivity anomaly, the actual critical control rod position is verified to be within 1% dk/k of the predicted critical control rod position.

6) Control Rod S ram Tim Testin

Purpose:

To demonstrate that the scram insertion time to notch position 5 of each control rod is less than 7.0 seconds following core alterations.

Description:

This test satisfies Susquehanna SES Technical Specification 3/4. 1.3.2 which requires that all control rods shall demonstrate scram insertion times of less than 7.0 seconds with reactor coolant pressure greater than 950 psig prior to exceeding 4K thermal power after core alterations.

7) TJJaa

Purpose:

a) To assure proper operation of the TIP system.

b) To check core symmetry.

Description:

A gross asymmetry check will be performed as well as a detailed statistical uncertainty evaluation of the TIP system. A complete set of TIP data will be obtained at a steady-state power level greater than 75K of rated power. A total average deviation or uncertainty will be determined for all symmetric TIP pairs as well as a maximum absolute deviation. The results will be analyzed to assure proper operation of the TIP system and symmetry of the core loading.

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HANNA S N T Y P A D T ONA STARTUP A T V T S HMARY SCRIPT ON The following is a short summary of additional activities performed during the Startup Testing Program.

Thermal imit Monitorin Hargins to the fuel thermal limits are checked throughout the startup period through review of the POWERPLEX core monitoring system output.

TIP S stem 1 Performance A full set of TIPs will be run at a low power level to update the core power distribution before the first POWERPLEX core performance calculation is initiated'ubsequent TIP sets will be performed in conjunction with LPRM calibrations. The LPRM currents will be updated and the LPRM GAFS verified to be within the acceptable range.

Power Oistribution Com arison with Offline Monitorin Actual online core power distribution data from the POWERPLEX Core Honitoring System will be compared to SIHULATE-E core simulation code calculations. The SIMULATE-E code, approved by the NRC (PL-NF-87-001-A) for use in the Susquehanna SES core design and licensing, was used by PP&L Nuclear Fuels Engineering personnel in the design of the UIC6 core and will be used for operations support applications throughout the cycle.

Core Flow Calibration A core flow calibration will be performed at -10K core flow. Get pump and recirculation loop flow instrumentation will be adjusted, if necessary, to, ensure correct core flow indication and correct calculation of the flow biased Rod Block Monitor, APRH Scram, and APRM Rod Block setpoints.

w Recirculation oo aselin ata Ac uisition Recirculation loop data will be collected throughout the startup program to provide baseline information for plant performance monitoring in two loop and single loop operation.

PP& Anal ti 1 H hod n hmarkin Core physics data obtained from startup testing is used for the continued benchmarking of PP&L's CPH-2/SIMULATE-E core analysis methodology, as recoamended by the NRC in its safety evaluation on PP&L's Topical Report PL-NF-87-001-A ("gualification of Steady State Core Physics Methods for BWR Design and Analysis" ).

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