ML19263D334

Forwards Licensing Application for Third Refueling.Also Forwards Rept NEDO-24175,Jan 1979, Suppl Reload Licensing Rept.

ML19263D334
Person / Time
Site:	Hatch
Issue date:	03/22/1979
From:	Whitmer C GEORGIA POWER CO.
To:	Office of Nuclear Reactor Regulation
Shared Package
ML19263D335	List: ML19263D334 ML19263D349
References
NUDOCS 7903270471
Download: ML19263D334 (10)

Text

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Gewp.i Powar Conmey

. 230 Pe3chtree ?rmt eze owe ec, aus Atla r ti G. A ni.30301 Telepr.or.o 0 4 522 6060 March 22, 1979 b Chas. F. Whitmer Geon>ia D Power V ce Pres' dent Eng neer.ng t!v mutwn ,w erre swem Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 NRC DOCKET 50-321 OPERATING LICENSE DPR-57 EDWIN I. HATCII NUCLEAR PLANT UNIT 1 .

RELOAD-3 LICENSING APPLICATION Gentlemen:

Georgia Power Company hereby submits a licensing application for the third refueling at Edwin I. Hatch Nuclear Plant Unit 1. The enclosed topical report, NED0-24175, January, 1979, " Supplemental Reload Licensing Report for Edwin I. Hatch Nuclear Plant Unit 1, Reload-3", presents the results of a plant unique analysis performed for the third reload. Also, enclosed are proposed changes to the Techilical Specifications required to implement the results of these analyses.

There are two editorial corrections to pige 10 of NED0-24175 which were not discovered until after printing. The coordinate system of "igure 6 is from one of the computer programs used to .tnalyze the Rod Withdrawal Error, and does not match the plant coordinat.s system. The coodinates will be revised to represent the plant coordinate system.

Additionally, the error rod is identifie.1 in Figure 6 as (18,35) when in actuality it is the fully inserted rod whi.:h will be seen to have coordinates (18,31). These corrections will le submitted formally when the change page becomes availabic.

The startup of cycle 4 is currently scheduled for the last half of June, 1979, therefore your timely review of this application will be appreciated.

Yours very truly, m

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%r= - /

Chas. F. Whitmer MRD/LIO!/mb Enclosure Sw rn to and subscribed before me this 22nd day of March,1979.

- & M Ndary Public, GeercisNb'thr9 Lpgbygg My Commission Expires sep'. 09.1981 xc: Rubic A. Thomas George F. T*owbridge, Esquire 79032764H

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ATTACIDiENT 1 1

! NRC DOCKET 50-371 i OPERATING LICENSE DPR-57 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS Pursuant to 10 C"R 170.12 (c), Georgia Power Company has evaluated the attached proposed amendment to Operating License DPR-57 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 v-auire an environmental impact statement; c) The proposed amende.ent 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 Hatch ' Technical Specification changes for Reload-3/ Cycle 4 operation; e) The proposed amendment is therefore a Class III amendment.

4 e

ATTACIIMENT 2 NRC DOCKET 50-321 OPERATIhG LICENSE DPR-57 EDWIN I. IIATCII NUCLEAR PLANT UNIT 1 PROPOSED CIIANCE TO TECllMICAL SPECIFICATIONS The proposed char.3e to Technical Specifications (Appendix A to Operating License DPR-57) would be incorporated as follows:

Remove Page Insert Page 1.2-3 1.2-3 1.2-5 1.2-5 3.6-20 3.6-20 3.11-1 to 3.11-2 3.11-1 to 3.11-2 3.11-4 3.11-4 Figure 3.11-2 5.0-1 5.0-1

BASES FOR SAFETY LIMITS 1.2 REACTOR COOLANT SYSTEM INTEGRITY The reactor coolant system integrity is an important barrier in the prevention of uncontrolled release of fission products. It is essential that the integrity of this system be protected by establishing a pressure limit to be observed for all operating conditions and whenever there is irradiated fuel in the reactor vessel.

A. Reactor Vessel Steam Dome Pressure

1. When Irradiated Fuel is in the Reactor The pressure Safety Limit of 1325 psig as measured by the reactor vessel steam dome pressure indicator is equivalent to 1375 psig at the lowest elevation of the reactor coolant system. The 1375 psig va'ue is derived from the design pressure of the reactor pressure vessel (1250 psig) and coolant system piping (suction piping: 1150 psig; discharge piping; 1350 psig). The pressure Safety Limit was chosen as the lower pressure

, resulting from the pressure transients permitted by the applicable design codes: ASME Boiler and Pressure Vessel Code,Section III for the pressure vessel and USASI B31.1 Code for the reactor coolant system piping. The ASME Boiler and Pressure Vessel Code permits pressure transients up to 10% over design pressure (110% x 1250 = 1375 psig), and the USASI Code permits pres-sure transients up to 20% over the design pressure (120% x 1150 = 1380 psig; 120% x 1350 = 1602 psig).

The pressure relief system (relief / safety valves) has been sized to meet the overpressure protection criteria of the ASME Boiler and Pressure Vessel Code,Section III, Nuclear Vessels.

The details of the overpressure protection analysis showing compliance with the ASME Boiler and Pressure Vessel code, Sc: tion III, Nuclear Vessels is provided in the FSAR, Appendix M, Summary Technical Report ,f Reactor Vessel Overpressure Protection. To determine the required steamfic a capacity, a parametric study was performed assuming the plant was operating at the turbine generator design condition of 105 percent rated steam flow (10.6 x 106 pounds per hour) with a vessel dome pressure of 1020 psig, at a reactor thermal power of 2537 Mw, and the reactor experiences the worst pressuriza-tion transient. The analysis of the worst overpressure transient, a 3 second closure of all main steam lire isolation valves neglecting the direct scram (valve position scram) results in a maximum vessel pressure (bottom) of less than 1375 psig if a neutron flux scram is assumed. In addition, the same event was. analyze' to determine the number of installed valves which would limit pressure to below the code limit. The results of this analysis show that the eleven installed relief / safety valves were adequate even if assuming the backup neutron flux scram.

Turbine trip from high power without bypass is the most severe transient resulting directly in a nuclear system pressure increase, assuming direct scram. This event is presented in Reference 5. The analysis shows that the peak pressure in the bottom of the vessel is limited to 1180 psig. Peak steam line pressure is 1149 psig, showing adequate protection for this ab-normal operational transient.

1.2-3

BASES FOR SAFETY I,IMITS 1.2.B. References

1. ASME Boiler and Pressure Vessel Code Section III.

2. USASI Piping Code, Section B31.1.

3. FSAR Section 4.2, Reactor Vessel and Appurtenances Mechanical Design.

4. FSAR Section 14.3, Analysis of Abnormal Operation Transients.

5. Cencral Electric Boiling Water Reactor Supplemental Reload Licensing Submittal for the Edwin I. liatch Nucicar Plant Unit 1 Reload 3, NEDO-24175, January, 1979.

1.2-5

BASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3.6.G. Reactor Coolant Leakage (Continued) would grow rapidly. However, the establishment of allowable unidentified leakage greater than that given in Specification 3.6.G on the basis of the data presently available would be premature because of uncertainties asso-ciated with the data. For leakage of the order of ~ gpm, as specified in Specification 3.6.G, the experimental and analytical data suggest a reason-abic margin of safety that such leakage magnitude would not result from a crack approaching the critical size for rapid propagation (Reference FSAR, Question 10.4.2). Leakage less than the magnitude specified can be detected reasonably in a manner of a few hours utilizing the available leakage detection scheme, and if the origin cannot be determined in a reasonably short time the plant shall be shutdown to allow further investigation and corrective action.

The total leakage rate consists of all leakage, identified and unidentified which flows to the drywell floor drain and equipment drain sump. The capacity of the drywell floor sump pumps is 100 gpm and the capacity of the drywell equipment sump pumps is also 100 gpm. Removal of 25 gpm from either of these sumps can be accomplished with considerable margin.

H. Relief / Safety Valves The pressure relief system (relief / safety valves) has been sized to meet the overpressure protection criteria of the ASME Boiler and Pressure Vessel Code,Section III, Nuclear Vessels.

The details of the overpressure protection analysis showing compliance with ASME,Section XII is provided in the FSAR, Appendix M, Summary Technical Report of Reactor Vessel Overpressure Protection. To determine the required steamflow capacity, a parametric study was performed assuming the plant was operating at the turbine-generator design condition of 105 percent rated steam flow (10.6 x 106pounds per hour) with a vessel dome pressure of 1020 psig, at a reactor thermal power of 2537 Mw, and the reactor experiences the worst pressurization transient. The reanalysis for Reload-3 (NEDO-24175) of the worst overpressure transient, a 3 second closure of all main ster , line isolation valves neglecting the direct scram (valve position scram) results in a maximum vessel pressure of 1232 psig if a reutron flux scram is assumed.

Turbine trip from high power without bypass is the most severe transient resulting directly in a nuclear system pressure increase, assuming direct scram.

This event is presented in NEDO-24175. The analysis shows that the peak pressure in the bottom of the vessel is limited to 1180 psig. Peak steam line pressure is 1149 psig, showing adequate protection for this worst ab-normal operational transient.

3.6-20

BASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILIANCE REQUIREMENTS 3.11 FUEL RODS ,

4.11 FUEL RODS Applicability Applicability, The Limiting Conditions for Operation The Surveillance Requirements apply associated with the fuel rods apply to to the paraneters which monitor the those parameters which monitor the fuel rod operating conditions.

fuci rod operating conditions.

Objective Objective The Objective of the Limiting Condi- The Objective of the Surveillance tions for Operation is to assure the Requirements is to specify the type performance of the fuel rods. and frequency of surveillance to be applied to the fuel rods.

Specifications Specifications A0. Average Planar Linear Heat Genera- A. Average Planar Linear Heat Genera-tion Rate (APLHGR) tion Rate (APLHGR)

During power operation, the APLHGR The APLHGR for each type of fuel as for each type of fuel as a function a function of average planar of average planar exposure shall exposure shall be determined daily not exceed the limiting value shown during reactor operation at 1 25%

in Figure 3.11-1, sheets 1 and 2. rated thermal power.

If at any time during operation it is determined by normal surveillance that the limiting value for APLHCR is being exceeded, action shall be initiated within 15 minutes to re-state operation to within the pre-scribed limits. If the APLHGR is not returned to within the pre-scribed limits within two (2) hours, then reduce reactor power to less than 25% of rated thermal power with-in the next four (4) hours. If the m limiting condition for operation is restored prior to expiration of the specified time interval, then further progression to less than 25% of rated thermal power is not requred.

B. Linear Heat Generation Rate (LHGR) B. Linear Heat Generation Rate (LHGR)

During power operation, the LHCR as The LHGR as function of core a function of core height shall not height shall be checked daily dur-exceed the limiting value shown in ing reactor operation at 1 25%

Figure 3.11-2 for 7 x 7 fuel or the rated thermal power.

Ifmiting value of 13.4 kw/ft for 8 x 8/

8 x 8R fuel. If at any time during operation it is detirmined by aormal surveillance that tt a limiting value for LHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the 3.11-1

BASES FOR LIMITING CONDITIONS FO" OPERATION AND SURVEILLANCE REQUIREMENTS 3.ll.B. Linear Heat Generation Rate (LHCR)

(Continued)

LHGR is not returned to within the prescribed limits within two (2) hours, then reduce reactor power to less than 25% of rated thermal poaer within the next four (4) hours. If the limiting condition for operatioa is restored prior to expiration of the specified time interval, then further progression to less than 25%

of rated thermal power is not re-quired.

C. Minimum Critical Power Ratio (MCPR) 4.11.C Minimum Critical Power Ratio (MCPR)

The MCPR limit is specified through- MCPR shall be determined daily out the cycle. From BOC4 to during reactor power operation at EOC4 the MCPR limit is 1.26 for > 25% rated thermal power and 7 x 7, 1.24 for 8 x 8, and 1.21 for following any change in power 8 x 8R fuels. During power level or distribution that would operation, MCPR shall be as cause operation with a limiting above at rated power and flow, control rod pattern as described if at any time during opera- in the bases for Specification tion it is determined by normal 3.3.F.

surveillance that the limiting value for MCPR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the steady state MCPR is not returned to within the prescribed limits within two (2) hours, then reduce reactor power to less tLan 25% of rated thermal power within the next four (4) hours. If the Limiting Condition for Operation is restored prior to expiration of the specified time interval, then further progression to less than 25% of rated thermal powcr is not required. For core flows other than rated the MCPR shall be Xg times the MCPR value applicabic above, where Kf is as shown in Figure 3.11-3.

D. Reporting Requirements If any of the limiting values iden-tified in Specifications 3.11.A.,

B., or C. are exceeded, a Reportable Occurrence report shall be submitted.

If the corrective action is taken, as described, a thirty-day written report will meet the requirements of this specification.

3.11-2

BASES FOR LIMITNG CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3.11.B. Linear Heat Generation Rate (LHGR)

This specification assures that the linear heat generation rate in any rod is less than the design linear heat generation if fuel pellet densification is postulated. The power spike penalty specified for 7 x 7 fuel is based on the analysis presented in Section 3.2.1 of Reference 4 and References 5 and 6, and assumes a linearly increasing variation in axial gaps between core bottom and top, and assures with a 95% confidence, that no more than one fuel rod exceeds the design linear heat generation rate due to power spiking. The LHGR as a function of core height shall be checked daily during reactor operation at > 25%

power to determine if fuel burnup, or control rod movement has caused changes in power distribution. For LHGR to De a limiting value below 25% rated thermal power, the MTPF would have to be greater than 10 which is precluded by a censi-derabic margin when employing any permissible control rod pattern.

C. Minimum Critical Power Ratio (MCPR)

The required operating limit MCPR . s specified in Specification 3.11.C is derived from the established fuel cladding integrity Safety Limit MCPR of 1.07 and an analysis of abnormal operational tranaients presented in Reference 7.

Various transient events will reduca the MCPR belo.w e operating MCPR.

To assure that the fuel cladding integrity safety 1..it (MCPR of 1.07) is not violated during anticipated abnormal operational transients, the most limiting transients have been analyzed to determine which one results in the largest reduction in critical power ratio (A MCPR) . Addition of the largest A MCPR to the safety limit MCPR gives the minimum operating limit MCPR to avoid violation of the safety limit should the most limiting transient occur.

The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.

The es aluation of a given transient begins with the system initial parameters shown .n Tabic 6-2 of Reference 9 that are input to a GE core dynamic behavior transient computer program described in Reference 8. Also, the void reactivity coefficients that were input to the transient calculational procedure are based on a new method of calculation termed NEV which provides a better agreement between the calculated and plant instrument power distributions. The outputs of this program along with the initial MCPR form the input for further analyses of the thermally limiting bundle with the single channel transient thermal hydraulic SCAT code described in Reference 1. The principal result of this evaluation is the reduction in MCPR caused by the transient.

From BOC4 to EOC4, the most limiting transient for the 8 x 8R fuel is the loss of 1000F f eedwater heating with a ACPR of 0.14. The most li n ' ting event through-out c;cle 4 for 8 x 8 and 7 x 7 fuel is the Rod Withdrawal Error (RWE) with a ACPr. of 0.17 for 8 x 8 and 0.19 for 7 x 7. Therefore, the MCPR's specified in 3.11.C are based on loss of 100 F feedwater heating and the Rod Withdrawal Erro".

3.11-4

5.0 ((AJOR DESIGN FEATURES A. Site Edwin I. Hatch Nuclear Plant Unit No. 1 is located on a site of about 2244 acres, which is owned by Georgia Power Company, on the south side of the Altamaha River in Appling County near Baxley, Georgia. The Universal Transverse Mercator Coordinates of the center of the reactor building are:

Zone 17R LF 372,935.2m E and 3,533,765.2m N.

B. Reactor Core

1. Fuel Assemblics The core shall consist of not more than 560 fuel assemblies of the licensed combination of 7 x / bundles which contain 49 fuel rods and 8 x 8 and 8 x 8R fuel bundles which contain 62 or 63 fuel rods each.

2. Control Rods The reactor shall contain 137 cruciform-shaped control rods. The control material shall be boron carbide power (B 44C) compacted to approximately 70% of its theoretical density.

C. Reactor Vessel The reactor vessel is described in Tabic 4.2-2 of the FSAR. The applicabic design specifications shall be as listed in Table 4.2-1 of the FSAR.

D. Containment

1. Primary Containment The princiapi design parameters and characteristics of the primary con-tainment shali be as given in Table 5.2-1 of the FSAR.

2. Cecondary Containment The secondary containment shell be as described in Section 5.3.3.1 of the FSAR and the applicable codes shall be as given in Section 12.4.4 of the FSAR.

3. Primary Containment Penetrations Penetrations to the primary containment and piping passing through such penetrations shall be designed in accordance with standards set forth in Section 5.2.3.4 of the FSAR.

E. Fuel Storage

1. Spent Fuel All arr.:ngements of fuel 4.n the cpent fuel storage racks shall be main-taine? in a subcritical configuration having a kefg not greater than 0.90 for normal conditions and a k egg not greater than _0.95 for abnormal conditions.

2. New Fuel ,

The new fuel storage vault shall be such that the kerr dry shall not be greater than 0.90 and the keff f1 ded shall nc t be greater than 0.95.

5.0-1