ML20205P870

Proposed Tech Specs,Implementing Bwrog/Ge Enhanced Option I-A (EI-A) Reactor Stability long-term Solution

ML20205P870
Person / Time
Site:	River Bend
Issue date:	04/15/1999
From:	ENTERGY OPERATIONS, INC.
To:
Shared Package
ML20205P866	List: ML20205P870 RBG-44968, Submits Addl Info Re 981008 LAR 1998-02 Re Implementation of Bwrog/Ge Enhanced Option I-A (EI-A) Reactor Stability long- Term Solution.Clarifies Certain Aspects of Proposed Ts,Per 990406 Telcon with NRC
References
NUDOCS 9904210013
Download: ML20205P870 (6)

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Enclosure 1 Revised Proposed TS Pages License Amendment Request (LAR) 1998-02, " Stability" Additional Information RBF1-99-0122 RBG-44968 April 15,1999 i

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l 9904210013 990415 PDR ADOCK 05000458 p PDR

T FCBB

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B 3.2.4 1

l BASES (continued) i This Surveillance is modified by a Note which allows 15 minutes l to verify FCBB following entry into the Restricted Region if the

! entry was the result of an unexpected transient (i.e. an unintentional or unplanned change in core thermal power or core flow). The 15 minute allowance is based on both engineering l

judgment and the availability of the PBDS to prov1de the operator with information regarding the potential imminent onset of neutron 1c/ thermal hydraulic instability. The 15 minute allowance l Of the Note is not to be used to delay entry Into Condition B 1f l the entry into the Restricted Region was the result of an l unexpected reduction in feedwater heating, recirculation pump trip. recirculation pump down shift to slow speed. or significant flow control valve closure (small changes in flow control valve position are not considered significant).

! REFERENCES 1 NEDO 32339 A. Revision 1. ' Reactor Stability Long Term Solution: Enhanced Option'

!.A.' April 1998 l

RIVER BEND B3.2 18 Revision No. 0

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RPS Instrumentation

, B 3.3.1.1 BASES APPLICABLE 2.b. Average Power Range Monitor Flow Biased Simulated SAFETY ANALYSES, Thermal Power --High LCO, and APPLICABILITY The Average Power Range Monitor Flow Biased Simulated (continued) Thermal Power-High Function monitors neutron flux to approximate the THERMAL POWER being transferred to the reactor coolant. The {

APRM neutron flux is electronically filtered with a time constant representative of the fuel heat transfer dynamics to generate a signal proportional to the THERMAL POWER in the reactor. trip i level is varied as a function of recirculation drive flow and is l clamped at an upper limit that is always lower than the Average l

Power Range Monitor Fixed Neutron Flux -High Function Allowable Value. The Average Power Range Monitor Flow Blased Simulated Thermal Power-High Function provides a general definition of the licensed core power / core flow operating domain.

During continued operation with only one recirculation loop in service, the APRM flow biased setpoint is required to be conservatively set (refer to the Bases for LC0 3.4.1. " Recirculation Loops Operating" for more detailed discussion). The setpoint modification may be delayed in accordance with the allowances of LCO

! 3.4.1. After this time, the LCO 3.3.1.1 requirement for APRM l OPERABILITY will enforce the more conservative setpoint.

l The Average Power Range Monitor Flow Biased Simulated Thermal Power l - High Function is not associated with a limiting safety system setting. Operating limits established for the licensed operating I domain are used to develop the Average Power Range Monitor Flow l Biased Simulated Thermal Power - High Function Allowable Values to provide pre-emptive reactor scram and prevent gross violation of the licensed operating domain. Operation outside the licensed operating i domain may result in anticipated operational occurrences and l postulated accidents being initiated from conditions beyond those assumed in the safety analysis. Operation within the licensed

operating domain also ensures compliance with General Design Criterion 12.

General Design Criterion 12 requires protection of fuel thermal safety limits from conditions caused by neutronic/ thermal hydraulic instability. Neutronic/ thermal hydraulic instabilities result in power oscillations, which could result in exceeding the MCPR SL.

(continued)

RIVER BEND B 3.3-8 Revision No. 0

F PBDS B 3.3 1.3 l BA3ES f

SUR(EILLANCE SR 3.3.1.3.2 (continued)

RE0J1REMENTS l the other channel if it is available. It 1s based on the l assumption that the instrument Channel indication agrees with the  !

1mmediate Indication available to the operator and that Instrument channels monitoring the same parameter should read similarly. Deviations between the instrument channels could oe an Indication of instrument component failure. A CHANNEL CHECK will detect gross channel failure: thus. It is key to verify 1ng the instrumentation continues to operate properly between each CHANNEL  !

FUNCTIONAL TEST Agreement Criteria are determined by the plant staff based on a combination of the channel Instrument uncertainties. Including indication and readability.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is based on operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal. but more frequent, checks of channels during normal operational use of the displays associated with tre 4 channels required by the LCO.

SR 3 3.1 3 3 .'

A CHANNEL FUNCTIONAL TEST 1s performed for the PB05 to ensure trat the entire system will perform the intended function. The CHANNEL FUNCTIONAL TEST for the PSDS includes manual initiation of an l internal test sequence and verification of appropriate alar 11 anc inop conditions being reported.

Performance of a CHANNEL FUNCTIONAL TEST at a Frequency of 24 months ver1fles the performance of the PB05 and associated circuitry. The Frequency considers the plant conditions recuired to perform the test. the ease of performing the test. and tre 11kel1 hood of a change in the system or component status. The alarm circuit is designed to operate for over 24 months with l sufficient accuracy on signal amplitude and signal timing considering envirenment. Initial calibration and accuracy craft (Ref. 2).

i REFERENCES 1. NE00 32339. Revision 1. " Reactor Stability Long Term Solution: Enhanced Option I-A." April 1998

2. NE00 32339P A. Supplement 2. " Reactor Stability Long Term 3

' Solution: Enhanced Option I-A Solution Design." April 1998.

RIVER BLND B 3.3 39h Revision No.

l Rsporting Requirements l .

5.6 '

5.6 Reporting Requirements 5.6.I Annual Radiolog1 cal Environmental Operating Report (continued) results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.

5.6.3 Radioactive Effluent Release Report  ;

The Radioactive Effluent Release Report covering the operation of the unit during the previous calendar year shall be submitted by May 1 of each year. The report shall include a summary of the quantitles of radioactive liquid and gaseous effluents and solid {

! waste released from the unit. The material provided shall be  !

l consistent with the objectives outlined in the ODCM and process

' control program and in conformance with 10 CFR 50.36a and 10 CFR 50. Appendix 1.Section IV.B.1.

5.6.4 Monthly Operating Reports Routine reports of operating statistics and shutdown expertence.. 4 including documentation of all challenges to the main steam safety / relief valves. Shall be submitted on a monthly basis no ~

later than the 15th of each month following the calendar month covered by the report.

5.6.5 CORE OPERATING LIMITS REPORT (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle and shall be documented in the COLR for the following:

1) LCO 3.2.1. Average Planar Linear Heat Generation Rate (APLHGR).

2) LCO 3.2.2, Minimum Critical Power Ratio (MCPR)

(including power and flow dependent limits).

3) LC0 3.2.3. Linear Heat Generation Rate (LHGR)

(including power and flow dependent limits).

4) LCO 3.2.4. Fraction of Core Boiling Boundary (FCBB)

5) LCO 3.3.1.1. RPS Instrumentation (RPS)

6) LCO 3.3.1.3. Periodic Based Detection System (PBDS)

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC. specifically those described in the following document 5.

(continued) i

_ RIVER BEND- 5.0-18 Amendment No. M 100 e  ;

I Rsporting P.squirements s.e ,

5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

1) NEDE 24011-P A. " General Electric Standard Application for Reactor Fuel" (latest approved version):

2) NEDC-32489P (April 1996). "T-Factor Setdown Elimination Analysis for River Bend Station" (for power and flow dependent limits methodology only as evaluated and approved by Safety Evaluation and License Amenament 100 dated October 10, 1997).

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3) NE00-32339P-A. " Reactor Stability Long-Term Solut1cn:

Enhanced Option I-A." including Supplements 1 througn 4 (April 1998).

c. The core operating limits shall be determined such that all applicable limits (e.g. . fuel thermal mechanical limits.

core thermal hydraullc limits. Emergency Core Cooling Systems (ECCS) limits. nuclear limits such as SDM. transient I analysis limits, and accident analysis limits) of the safety. '

analysis are met. ,

d. The COLR. Including any midcycle revisions or supplements.

Shall be provided upon issuance for each reload cycle to the NRC.

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RIVER BEND 5.0-19 Amendment No. E % GG 100 l