ML19294C264

Application to Amend App a to OL to Modify Average Power Range Monitor high-high Flux Scram Trip Logic to Minimize Inadvertent Scrams Caused by Neutron Flux Spikes,W/O Reducing Plant Safety.Forwards Mod Description & Tech Specs

ML19294C264
Person / Time
Site:	Hatch
Issue date:	02/28/1980
From:	Widner W GEORGIA POWER CO.
To:	Office of Nuclear Reactor Regulation
Shared Package
ML19294C265	List: ML19294C264 ML19294C266
References
NUDOCS 8003070392
Download: ML19294C264 (8)

Text

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• Georgia Power Compcny 230 Peachtree Street Post OMice Box 4545 Ananta, Georgia 30302 Te:cphone 404 522-6060 February 28, 1980 m

Power Generation Department Georgia Power Director of Nuclear Reactor Regulation '

U. S. Nuclear Regulatory Commission Washington, D. C. 20555 NRC DOCKET 50-366 OPERATING LICENSE NPF-5 EDWIN I. HATCH NUCLEAR PLANT UNIT 2 THERMAL POWER MONITOR Gcntlemen:

Pursuant to 10 CFR 50.90, as required by 10 CFR 50.59(c)(1), Georgia Power Company hereby proposes amendments to the Plant Hatch Unit 2 Technical Specifi-cations (Appendix A to the Operating License). The proposed change will modify the Average Power Range Monitor (APRM) high-high flux scram trip logic. A description of the modification and af' .t on the plant safety analysis is provided in Enclosure 3. As discusse, , the flow referenced logic design will be modified with a new logic scneme that will minimize inadvertent scrams caused by neutron flux spikes without reducing plant safety. This new design feature was previously approved for the Hatch Nuclear Plant Unit 1 by Amendment No. 69 to Facility Operating License DPR-57.

The Plant Review Board and Safety Review Board have reviewed and approved these proposed changes to the Plant Hatch Unit 2 Technical ,ecifications and have determined that they do not involve an unreviewed safety question. This modification to the APRM results in the maintenance of adequate thermal margins for fuel cladding integrity and the reduction of the cyclic duty of the reactor vessel and fuel by minimization of the number of spurious scrams. Thus, it can be concluded that the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased, nor is the possibility of a new accident or malfunction of equipment import e_ m safety created. The margin of safety, as defined in the Technical Specifications, is not reduced due to this change because no safety limits have been affected.

This modification will be completed prior to startup following our March 1980 maintenance outage. Therefore, your review of this submittal in a timely manner will be appreciated.

Yours very truly, h'Y h .AY W. A. Widner Vice President and General Manager Nuclear Generation MRD/mb Enclosures Sworn to and subscribed before.me this 28th day of February, 1980.

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. Notary Public xc: Mr. Ruble A. Thomas go ,ry Pubiic Geraia. StMe et Larco George F. Trowbridge, Esquire 14 Commis3icn Evpires Ecet. 20.1983 R. F. Rogers, III 80030703 %

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ATTACHMENT 1 NRC DOCKET 50-366 OPERATING LICENSE NPF-5 EDWIN 1. HATCH NUCLEAR PLANT UNIT 2 PROPOSED DETERMINATION OF AMENDMENT CLASS Pursuant to 10 CFR 170.12 (c), Georgia Power Company has evaluated the attached proposed amendment to Operating Licepse NPF-5 and have determined that:

a) The proposed amendment does not require the evaluation of a new Safety Analysis Report or rewrite of the facility license.

b) The proposed amendment does not contain several complex issues, does not involve ACRS review, or does not require an environmental impact statement; c) The proposed amendment does not involve a complex issue, an environ-mental issue or more than one safety issue; d) The proposed amendment does involve a single issue; namely, the modification of the Average Power Range Monitor (APRM) high-high flux scram trip logic by the addition of a thermal power monitor; e) The proposed amendment is therefore a Class III amendment.

ATTACHMENT 2 NRC DOCKET 50-366 OPERATING LICENSE NPF-5 EDWIN I. HATCH NUCLEAR PLANT UNIT 2 PROPOSED CHANCES TO TECHNICAL SPLCIFICATIONS The proposed changes to the Technical Specifications (Appendix n to Operating License NPF-5) would be incorporated as follows:

Remove Page Ins'ert Page 2-4 2-4 2-6 2-6 B 2-9 B 2-9 B 2-10 B 2-10 3/4 2-5 3/4 2-5 3/4 3-2 3/4 3-2 3/4 3-6 3/4 3-6 3/4 3-7 3/4 3-7 3/4 3-8 3/4 3-8

ATTACHMENT 3 NE',l APRM SCRAM TRIP LOGIC FCR THE E0',llN 1. HATCH NUCLEAR "LANT UNIT 2

1. ,PU.. POSE This report describes the new APRM scram trip logic and Thisdiscusses APRM logic the ir.: pact of this logic on plant saf ety analyses. -

is being installed in the Ed, tin I. Hatch Nuclear Plant Unit This 2 f or operation during the initial and all subsequent cycles.

report shows that the new APPM scram trip logic will reduce fuel and reactor cycle duty without compromising the safety of the ,

plant.

2. BACKGROUND Scrams have been reported at operating CWRs as a result of momen-tary; anomalous neutron flux spikes which exceeded the high-high APRM flow referenced trip setting. Frequent causes of these flux spikes are comentary ficw chtng2s in the recirculation system flow and small pressure disturbances during turbine stop valve and turnine control valve testing. Although many of these scrams occurred during operation with less than rated core flow, the neutron f. lux did not exceed the 100% ficw flux scram trip value (120%) assumed in the transient safety analysis. These small neutron flux spikes represent no danger to the fuel because their Therefore, duration is less than the feel thermal time constant.

th" fuel surface heat flux dcas not increase sufficiently to challenge The new APRM scram logic the f uel cicdding integrity safety limit.

will reduce the number of spurious scrams occurring along the power-flow line without reducing the fuel safety margins for any ~

accidents or abnurnal cperational transients for which the plant is licensed.

3. D E E C.^d PT I_ON The APRM flow referenced scra feature was designed and installed

" cn Hatch 2 as noted in subsection 7.6.2.2.4, of the Final Safety Tnalysis Peport. The trip setpoint is varied as a function of reactor recirculation driving loop flow relative to a value of 120%

of nuclear-boiler rated power at full flow.

There are six APFM charnels, three for each reactor protection trip system The trip unit for one of these three channels can supply the trip signal to the associated reactor protection trip system At. least one APF.M channel in each trip system must trip to cause a scrar Presently, each APRM channel cerives its trip signal frca The APRM scram trip setpoints LPM neutron flux neasurements.

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i ATTACHMENT 3 (Continued) given in the technical specifications

• are Tovaried as a f unction of obtain the proper reacter recirculation driving loop flcw.

(' reference signal, each APPM channel is supplied with two redundant and isolated flow signals associated with the trip systen.

Althounh the present APRM ficw reference scram system accurately predicts the thermal pcwer level for steady-state operation, it

. overpredicts the fuel heat powerlevel during power increase events.

During such events, the neutron flux leadsTherefore, the reactorneutron thermalflux power because of the fuel time constant.

trip levels are reached before the reactor thermal power has signifi-cantly increased. While thvs anticipato.ry response in the APCM -

scram provides additional protection to the core during abnormal operational transients or accidents, it results'in many unnecessary scrams for small neutron flux disturbances along the flow control .

line (figure 1).

Many of these unnecessary scrams will be avoided by replacing the present APRM trip logic with a Thermal Power Simulator and an APRM Simulated Thermal Power (STP) trip. unit. The APRM signal for the scram trip will be processed through a Thermal Power Simulator This circuit repre-consisting'of a time constant delay circuit.

sents the fuel dynamics which will approximate the reactor thermal pb'.eer during a transient or steady state condition. A faster '

response trip unit on APEM neutron flux utilizing a non-flow referenced 12 L ncutron flux scram trip setpoint has also been added. Figure 3 illustrates the new APRM scrca trip logic.

Figure 2 shows the response of the new APRM scram trip logic to the same flux spike as shown in Figure 1. In this case, a scram does not occur, since the transient peak of the simulated thermal power is below the flow referenced scram setpoint.

The total recirculation drive flow signal to the APRM STP trip unit remains the sama. Therefore, an APRM channel trip could be initiated fro:a either a non-flow referenced APEM neutron flux trip unit or the flow referenced APRM STP trip unit.

4. CONF 0MMNCE TO GUIDES AND STANDARDS The electrical corponents used in the new APRM scram trip logic circuitry are in ccnformance with all applicable IEEE Standards, with all applicable NEC Regulatory Guides, and with the Code of Federal Regulations, litle 10, Chapter 1, Part 50, Appendix A.

The new APRM scram trip logic was qualified to the following codes by which Hatch 2 was licensed:

a) IEEE 279-1971 - Criteria for Protection Systems for Nuclear Power Generating Stations.

b) IEEE 323-1971 - General Guide for Qualifying Class 1 Electrical

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Equipnent toi Nuclesr Power Gencrating Stations.

• Technical Specificatio'ns 2.2.1, 3.2.2, 3.3.1, and <l.3.1.

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ATTACHMENT 3 (Continued) c) IEEE 341-1971 - Recorcmen&d Practices for the Seismic Qualification of Class 1 Electric Equipment for Nuclear Porter Generat.ing Stations.

Both AC and DC APRM power supplies remain the same.

The Average Power Range Monitor (APRM) system is one subsystem of the neutron monitoring system. The APRM subsysten is augmented to include the Simulated Thermal Power Trip (STPT). The APRM sub-system has 6 APRM channels, each using input signals from a number of LPRM channels. Three APPM channels are associated with each of the trip systems of the Reactor Protection System. The APRM subsysttm is designed to meet the require:nents of JEEE-279 as documented in ,

Subsection 7.6.2.3.i of the FSAR.

The STPT augmento e7ch of the 6 APRM channels such that each APRM channel has a 120% neutron flux trip whose setpoint is not retircu- -

lation flow biased. The new thermal power upscale trip has a

. setpoint thit is' flow biased and is obtained by filtering the APRM .

signal to obtain a signal which represents the thermal flux of the fuel. This time delay is accomplished by conditioning the APRM 2 neutron flux through a first order low pass filter that has a 6 second.RC time consta,nt. Since each of the 6 APFM channels was identically modified to add the STPT, and the independence between the 6 APFM channels was not altered, the redundancy requirements of IEEE-279 are still maintained.

5. EFFECTS 09 SAFETY ANALYSIS

(~ Cumulative fatigue damage analyses are performed for'the fuel assembly, the reactor and reactor internals. The cyclic loads ccnsidered in these analyses include coolant pressure and thermal gradients. Details of the methodology used for the fuel analysis are given in Reference 1. Reactor and reactor in'.crnal analyses are addressed in Subsection 3.9 of the FSAR. Avoidance of spurious scrats will reduce the plant cyclic duty and will, therefore, -

provide additional margin to the fatigue damage limits. Because the present limiting abnormal operational transient analyses do not account for the present flow referer.ce APRM scram setpoints, the only transient which is affected by the ne. i.PRM scram trip logic is the loss of-feedwater heating transient. Because the flow referenced scram with the new APRM scram trip logic has a maximum setting at 113.5% neutron flux, a scran will occur earlier for slow transients, before the fixed APRM scram setpoint of 11S% neutron flux is reached. Therefore the i.ransient ACPR will be less for the most limiting slow transient, a loss-of-ieedwater heating. No credit for thic reduction is taken in cycle 1.

At less than rated power conditions, the new APRM scram trip logic providas greater thermal margins to the fuel cladding integrity safety limit than at rated pwer. As reactor poaer is reduced, total steam generation d creases. In the loss-of-feedwater heating 1.ransient, the reduced steam flow at low power results in a decrease g in both fcedwater flow and t.he maximum temperature rise across a LW:at/57J 3

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ATTACHMENT'3 (Continued)

. given heater. The core subcooling change, as well as the positive reactivity insertion, will then be less severe. lherefore, the change in critical power ratio (aCPR) '. sill decrt ast. with decreasing power, Thus, the dif fere::.ce between the safety limit and the transient MCPR increases with decreasing power, irrespective of the C screa systems logic.

In addition, at any given recirculation loop flow rate, tile STPT logic is designed to maintain a relatively constant margin between the reactor power and the STP trip setting. This margin, or power

, difference, between the reactor power and the thermal power t. rip setting is established by the STPT retpoint specification:

S < minimum of (10 6r,w + SLT g . .g, w . . ,

where "W" is the recirculation loop flew rate ~ sa a percent of rated. This specification requires that' the STPT setpoint be reduced as the recirculation locp flow rate (and hence reactor -

power) is reduced. The characteristic decrease in ACPR uith decreasing power, and the reduction in APRM STPT setpoint with decreased recirculation loop flow (and hence reactor.. power), both act to assure that the fuel cladding integrity Safet'y Liait is not violated during the loss of feedwater heating transient at less than rated power.

Analyses for Hatch Unit 2 initial core have demonstrated thst with only the 120; trip setting, none of the abnernal operational transients analyzed violates the. f uel cladding integrity safety linit, and that there is substantial margin frc fuel damage. Therefore, the

(' use of the flow referenced trip setpoint, with the fixed setpoint as backup, provides adequate thermal margins for fuel cladding -

integrity.

6. APPLICATICN TO CURRENT OPEilATING PLAMTS At present, Bronsuick Units 1 and 2 and the James A. FitzPctrick Nuclear Power Plant are operating wit.h the new APRM scrau trip .

logic. This logic was an integral part of the Brunswick Units 1 and 2 APRM scram trip system when these plants were initially licensed (see Section 7.5.7, Average Power Range Monitor Subsystem, in the Brunswick Units 1 and 2 FSAR). The new APRM scram trip lonic was licensed c3 a retrofit margin ic?rovc-ment cption cn the James A. FitzPatric.k plant.

Field experience from these plants has shown.that scrams from recirculation system excursions have been reduced by 50 to 75% due to this nodificatina. A similar reduction on spurious scrams is expected when the ne'.. APRM scram trip logic is installed in the Hatch Unit 2 plant.

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ATTACHMENT 3 (Continued)

7. SU.*"!.EY A? D CCh'CLUSIC,NS Thisdocumenthasdisc[ssedthedesignandsafetyaspectsofthe new APF.M scram trip logic. By reducing the cyclic duty on the plant, greater margins to cumulative fatigue damage limits exist.

In addition, the effects of slow loss of coolaat transients are less severe. Therefore, the safety of the plant will be . increased by operation of the plant with the new APRM scram trip logic.

8. REF EP.E!!CES

1. Generic Reload ' Fuel Application", flEDE-24011-P-A, August 1978.

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FIGURE 1 .

APRM SCPJ.JtS WITH STP fic n- Fl o.4 Ils referenced

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