ML20217G216

Errata Pages 2 & 3 for Safety Evaluation Supporting Amend 168 Issued to FOL DPR-63 Issued on 990921.New Pages Change Description of Flow Control Trip Ref Cards to Be Consistent with Application for Amend

ML20217G216
Person / Time
Site:	Nine Mile Point
Issue date:	10/15/1999
From:	NRC (Affiliation Not Assigned)
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ML20217G132	List: ML20217G129 ML20217G216
References
NUDOCS 9910210227
Download: ML20217G216 (2)

Text

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. 004 g A*4 UNITED STATES j

, g , NUCLEAR REGULATORY COMMISSION o e WASHINGTON, D.C. 20666 4001

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• .,*l October 15, 1999

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Mr. John H. Mueller Chief Nuclear Officer Niagara Mohawk Power Corporation N!ne Mile Point Nuclear Station 1 Operations Building, Second Floor P.O. Box 63 -

- Lycoming, NY 13093

SUBJECT:

CORRECTION TO SAFETY EVALUATION FOR AMENDMENT NO.168 FOR l NINE MILE POINT NUCLEAR STATION, UNIT NO.1 (TAC NO. MA4218)

Dear Mr. Mueller:

4 On September 21,1999, the Commission issued Amendment No.168 to. Facility Operating License No. DPR-63 for the Nine Mile Point Nuclear Station, Unit No.1 in response to your application dated November 16; .1998, as . supplemented June 21,1999. The amendment changed the Technical Specifications regarding the setpoints for the average power range monitor flow biased scrarn to limit the magnitude of reactor power oscillations during a reactor trip, and along with changes to the control rod block settings, allow for operation in the

- Extended Load Line Limit Analysis region of the power / flow operatitig curve.

The amendment also enclosed a copy of the related Safety Evaluation (SE). Please substitute the enclosed pages 2 and 3 for the corresponding pages of that SE. The new pages change the description of the Flow Control Trip Reference (FCTR) cards to be consistent with the application for amendment. The changes do not affect the NRC staff's conclusions or associated bases.

4 Sincerely, D- w M Darl S. Hood, Sr. Project Manager, Section 1 Project Directorate i Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-220 Enclosures. deplacement pages number 2 and 3 cc w/encls: See next page 9910210227 991015 PDR ADOCK 05000220 p PDR

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2.0 BACKGROUND

The principal purpose of the proposed amendment is to change the APRM flow biased scram settings to limit the reactor power oscillation magnitude during reactor trip, thereby limiting the  !

associated CPR change and, in conjunction with MCPR operating limits, ensure compliance with ,

the MCPR safety limit. '

The change to the APRM flow biased neutron flux scram will be implemented by APRM Flow Control Trip Reference (FCTR) cards and by making the required changes to the APRM flow biased neutron flux scram (and rod block) TS. The FCTR cards are similar to the GE FCTR cards that implement the Enhanced Option I-A solution, which prevents reactor instabilities by automatically excluding reror operations in regions of the power / flow operating domain that i are susceptible to reactor power instabilities. The changes to the TS implement more restrictive TS flow biased scram trip (and rod block) settings in the low flow regions of the power / flow operating map (i.e., the operating conditions most susceptible to reactor instabilities).

In addition, the amendment would increase the APRM flow biased neutron flux scram and control rod block settings to allow plant operation in the previously approved ELLLA region.

NMP1 is currently restricted from full use of the ELLLA region of the power / flow map because of the flow biased rod block setpoints. The NRC staff previously approved operation in the ELLLA region by NMP1 Amendment No. 92, dated March 24,1987.

Existing NMP1 TS 2.1.2, " Fuel Cladding Integrity," applies to trip settings on automatic protective devices related to variables on which fuel safety limits have been placed. TS 2.1.2a delineates the relationships that establish the flow biased APRM scram and rod block trip settings. The maximum values of the scram and rod block trip settings are currently 120% and 110%,

respectively. NMPC has concluded from analyses that none of the postulated accidents would violate the established criteria with a 120% scram trip setting.

Reactor power level may be varied by moving control rods or by varying the recirculation flow rate. The APRM system provides a control rod block to prevent rod withdrawal beyond a given point at a constant recirculation flow rate. This rod block trip setting, which is automatically varied with recirculation flow rate, prevents an increase in the reactor power level to excessive values due to control rod withdrawal. The margin to.the safety limit increases as the flow decreases for the specified trip setting versus flow relationship; therefore, the worst case MCPR which could occur during steady-state operation is at 110% of rated thermal power because of the APRM rod block trip setting.

The actual power distribution in the core is established by s'pecified control rod sequences and is monitored continuously by the in-core local power range monitor (LPRM) system. As with the APRM scram trip setting, the APRM rod block trip setting is adjusted downward if the core maximum fraction of limiting power density exceeds the fraction of rated thermal power, thus, preserving the APRM rod block safety margin.

The current NMP1 APRM system consists of eight identical channels. Each channelis provided

. with eight inputs from the LPRM system to enable it to compute a core average thermal neutron flux. Additionally, the APRM averaging and trip circuits receive total recirculation flow signals and calculate the Reactor Protection System (scram) and the Reactor Manual Control System

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- (control rod block) trip level setpoints in accordance with a specific relationship between core average power and total recirculation flow.

NMP1 has been identified as a stability long term solution Option 11 plant, in which the existing

. quadrant-based APRM trip system, typical of BWR/2 plants, would initiate a reactor scram for a postulated reactor instability and avoid violating the MCPR safety limit. However, at NMP1, the TS flow biased APRM flux trip setting must be modified for Option ll implementation. This change is required to limit the oscillation magnitude at reactor trip, thereby limiting the associated CPR change and, in conjunction with the MCPR operating limits, ensures operation within the MCPR Safety Limit. Specifically, changes to TS 2.1.2, " Fuel Cladding Integrity," are proposed. To implement this TS setting change, the eight APRM analog Flow Bias Trip Units were replaced with GE digital FCTR cards. With these FCTR cards in place, a secam setpoint l ,

can be established that will meet General Design Criterion 12 of Appendix A to 10 CFR Part 50  !

for fuel design limit protection. j l

GENE-A13-00360-02," Application of Stability Long-Term Solution Option ll to Nine Mile Point Nuclear Station Unit 1," demonstrated the application of Option ll methodology at NMP1 for operating Cycle 12. GENE-A13-00360-02 was submitted to the NRC by letter dated October 2, 1995. As Indicated in GENE-A13-00360-02, " detect and suppress" calculations are performed for two points along the rated rod line consistent with the Boiling Water Reactor Owners Group's (BWROG's) detect and suppress methodology (i.e., NEDO-32465, "BWR Owners' Group Reactor Stability Detect and Suppress Solutions Licensing Basis Methociology for Reload Applications," dated May 1995). The accepted version, NEDO-32465-A, was issued in August

- 1996. The calculations start at MCPR Operating Limits along the rated rod line for (1) a five