ML19259C589

Proposed Tech Spec Changes Re Spent Fuel Pool Storage Expansion

ML19259C589
Person / Time
Site:	Hatch
Issue date:	07/09/1979
From:	GEORGIA POWER CO.
To:
Shared Package
ML19259C588	List: ML19259C587 ML19259C589 ML19259C590
References
NUDOCS 7907170069
Download: ML19259C589 (8)

Text

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NRC DOCKET 50-321 OPERATING LICENSE DPR-57 EDWIN I. HATCH NUCLEAR. PLANT UNIT 1 PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS The proposed changes to the Technical Specifications (Appendix A to Operating License DPR-57) would be incorporated as f'ollows:

Remove Page Insert Page 3.10-5 3.10-5 3.10-7 3.10-7 3.10-9 3.10-9 3.10-10 3.10-10 5.0-1 5.0-1 2i04 087 7 907170 Cd 7

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.10.F.3 Monorail Hoist (Continued) a.

The trolley and hoist shall be demonstrated to be operable by a trial lift of the spent fuel pool gate or an equivalent weight.

b.

A visual inspection shall be made to insure the structural integrity of the 5-ton mono-rail hoist.

3.10.G Spent Fuel Cask Lif ting 4.10.G Spent Fuel Cask Lif ting Trunnions and Yoke Trunnions and Yoke See note for Specification See note for Specification 3.10.F.1 above.

4.10.F.1 above.

3.10.H Time Limitation Irra'diated fuel shall not be handled in or above the reactor prior to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reactor shutdown.

3.10.I Crane Travel-Spent Fuel 4.10.I Crane Travel-Spent Fuel Storage Pool Storage Pool Loads in excess of 1600 Loads, other than fuel assemblies pounds shall be prohibited or control rods, shall be verified from travel over fuel to be 51600 pounds prior to move-assemblies in the spent ment over fuel assemblies in the fuel storage racks.

fuel storage pool racks.

2i34 083 3.10-5

3.10. A. 2.

Fuel Grapple Hoist Load Setting Interlocks Fuel handling is normally conducted with the fuel grapple hoist. The total load on this hoist when the interlock is required consists of the weight of the fuel grapple and the fuel assembly. This total is approximately 1500 lbs.

in comparison to the load setting of 485 30 lbs.

3 '.

Auxiliary Hoists Load Setting Interlock Provisions have also been made to allow fuel handling with either of the three auxiliary hoists and still maintain the refueling interlocks.

The 485130 lb.

load setting of these hoists is adequate to trip the interlock when a fuel bundle is being handled.

B.

_ Fuel Loading To minimize the possibility of loading fuel into a cell containing no control rod, it is required that all control rods are fully inserted when fuel is being loaded into the reactor core. This requirement assures that during refueling the refueling interlocks, as designed, will prevent inadvertent cri ticali ty.

C.

Core Monitoring During Core Alterations The SRM's are provided to monitor the core during periods of Unit shutdown and to guide the operator during refueling operations and Unit startup.

Requiring two operable SRM's in or adjacent to any core quadrant where fuel or control rods are being moved assures adequate monitoring of that quadrant during such al terations.

The requirements of 3 counts per second provides assurance that neutron flux is being monitored.

During sprial unloading, it is not necessary to maintain 3 cps because core alterations will involve only reactivity removal and will not result in criticality.

The loading of diagonally adjacent bundles around the SRM's before attaining the 3 cps is permissible lecause these bundles were in a subcritical configuration when they were removed and therefore they will remain subcritical when placed back in their previous positions.

D.

Spent Fuel Pool Water Level The design of the spent fuel storage pool provides a storage location for 3181 fuel assemblies in the reactor building which ensures adequate shielding, l

cooling, and reactivity control of irradiated fuel. An analysis has been performed which shows that a water level at or in excess of eight and one-half feet over the top of the active fuel will provide shielding such that the maximum calculated radiological doses do not exceed the limits of 10CFR20.

The normal water level provides 14-1/2 feet of additional water shielding.

All penetrations of the fuel pool have been installed at such a height that their presence does not provide a possible drainage route that could lower the water level to less than 10 feet above the top of the active fuel.

Lines extending below this level are equipped with two check valves in series to prevent inadvertent pool drainage.

E.

Control Rod Drive Maintenance During certain periods, it is desirable to perform maintenance on two control rod drives at the same time.

3.10-7 2i04 089

BASES FOR LIMITING CONDITIONS FOR OPERATION 3.10.H Time Limi+.ation The radiological consequences of a fuel handling accident are based upon the accident occurring at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter reactor shutdown.

3.10.I Crane Travel-Spent Fuel Storage Pool A maximum weight of 1600 pounds shall be permitted to be transported over stored spent fuel in order to minimize the consequences of a load handling accident.

3.10.J References 1.

FSAR Section 7.6, Refueling Interlocks 2.

FSAR Section 7.5, Neutron Monitoring System 3.

Morgan, W.

R., "In-Core Neutron Monitoring System for General Electric Boiling Wat:- Reactors," General Electric Company, Atomic Power Equipment Department, November 1968, revised April 1969 (APED-5706) 4.

FSAR Section 10.3, Spent Fuel Storage 5.

FSAR Section 3.6.5.2, Reactivity Control 2}34 09 3.10-9

BASES FOR SURVEILLANCE P.EQUIREMENTS 4.10.

REFUELING A.

Refueling Interlocks Complete functional testing of all refueling interlocks before any refueling outage will provide positive indication that the interlocks operate in the situations for which they were designed. By loading each hoist with a weight equal to the fuel assembly, positioning the refueling platform, and withdrawing control rods, the interlocks can be subjected to valid operational tc ~ ts.

Where redundancy is provided in the logic circuitry, tests can bc performed to assure that each re-dundant logic element can independently perform its functions.

C.

Core Monitoring During Core Alterations Requiring the SRM's to be functionally tested prior to any core alteration assures that the SRM,'s will be operable at the start of that alteration.

The daily response check of the SRM's ensures their continued operability.

D.

Spent Fuel Pool Water Level A daily record of the Spent Fuel Pool Water Level to determine that the minim 0m of 8.5 feet is not exceeded is considered sufficient to ensure that radiological shielding is maintained.

E.

Control Rod Drive Maintenance Refueling interlocks and core monitoring surveillance arc discussed in 4.10.A and 4.10.C above. The choice of the strongest (highest reactivity worth) rod which will be used for a determination of the relevant shutdown margins is based on prior core calculations supplemented by empirical data obtained from similar cores.

From similar data e-d calculations the reactivity worth of rods adjacent to a withdrawn rod aill also be known.

Thus the surveillance shutdown margins can be evaluated in terms of rod posi tion.

F.

Reactor Building Cranes Modifications to the main reactor building crane are being studied in order to increase its ability to withstand a single failure. A spent fuel cask will not be lif ted until these modifications have been accepted by the AEC and the AEC has approved the lifting of casks by the crane and the appropriate Technical Specifications.

G.

Spent Fuel Cask Lif ting Trunnions and Yoke See note for Bases 4.10.F above.

I.

Crane Travel-Spent Fuel Storage Pool

t. \\ ) 4 y j Refer to Bases 3.10.1.

q.

u 3.10-10

5.0 MAJOR 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 Ba~xley, 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 Assemblies The core shall consist of not more than 560 fuel assemblies of the licensed combination of 7x7 bundles wnich contain 49 fuel rods and 8x8 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 powder (B C) compacted to approximately 4

70% of its theoretical density.

C.

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

D.

Contai nment 1.

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

2.

Secondary Containment The secondary containment shall 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 arrangements of fuel in the spent fuel storage racks shall be main-not greater than 0.95.

tained in a subcritical configuration having a keff 2.

New Fuel dry shall not be The new fuel storage vault shall be such that the keff flooded shall not be greater than 0.95.

greater than 0.90 and the keff 9104 092-

/-

5.0-1

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

Remove Page Insert Page 5-4 5-4

?.'iO4 093

DESIGN FEATURES ORAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 185 feet.

CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 2845 fuel assemblies.

l 5.7 COMP 0NENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7.1-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7.1-1.

2 'i 0 4 09 'l HATCH - UNIT 2 5-4