ML20097J920

Proposed Tech Specs Allowing Up to Four Bundles to Be Loaded in Previous Positions Around Source Range Monitor to Establish Required Three Counts Per Second & Deleting Description of Control Rod Matl

ML20097J920
Person / Time
Site:	Hatch
Issue date:	09/17/1984
From:	GEORGIA POWER CO.
To:
Shared Package
ML20097J911	List: ML20097J909 ML20097J920
References
TAC-55813, NUDOCS 8409240113
Download: ML20097J920 (9)

Text

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ATTACHET 1 NRC DOCKET. 50-321 OPERATING LICENSE DPR-57 EDWIN I. HA'ICH NUCLEAR PLW1' UNIT 1 REQUEST FOR TECHNICAL SPECIFICATIONS CHANGE The proposed changes to Technical Specifications (Appendix A to Operating License DPR-57) would be incorporated as follows:

Applicable Significant Hazards Evaluation Its Rmove Page Insert Page Section 1 3.10-2 3.10-2 1 3.10-7 3.10-7 1 2 5.0-1 5.0-1 2 3 1.0-2 1.0-2 3 8409240113 840917 PDR ADOCK 05000321 p PDR

p.M C. Core Alteration - Core alteration shall be the addition, reoval, relocation, or movenent of fuel, sources, incore instrm ents, or reactivity controls within the reactor pressure vessel with the vessel head renoved and fuel in the vessel. Suspension of core alterations shall not preclude cmpletion of the movment of a cmponent to a safe conservative position.

D. Design Power - Design power refers to the power level at which the reactor is producing 105 percent of reactor vessel rated stem flow.

Design power does not necessarily correspond to 105 percent of rated reactor power. The stated design power in megawatts thermal (MWt) is the result of a heat balance for a particular plant design. For Hatch Nuclear Plant Unit 1 the design power is 2537 MWt. Design power is used as an initial condition in transient and accident analyses.

E. Ergineered Safety Features - Engineered safety features are those features provided for mitigating the consquences of postulated accidents, including for exmple contairraent, energency core cooling, and standby gas treat 2nent systen.

F. Hot Shutdown Condition - Hot shutdown condition means reactor operation with the Mode Switch in the SHITIDOWN position, coolant tenperature greater than 2120F, and no core alterations are permitted.

G. Hot Standby Condition - Hot standby condition means reactor operation with the Mode Switch in the START & HOT STANDBY position, coolant tenperature greater than 2120F, reactor pressure less than 1045 psig, critical.

H. Imediate - Imediate means that the rquired action shall be initiated as soon as practicable, considering the safe operation of the Unit and the importance of the required action.

I. Instrment Calibration - An instrment calibration means the adjustment of an instrunent output signal so that it corresponds, within acceptable range and accuracy, to a known value(s) of the paraneter which the instrunent monitors.

J. Instrument Channel - An instrunent channel means an arrangenent of a sensor and auxiliary quipnent rquired to generate and transnit to a trip systen a single trip signal related to the plant paraneter monitored by that instrunent channel.

Anenduent No. 1.0-2

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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS w.

73.10;C ? Core Monitoring During Core 4.10.C ' Core Monitoring During Core Alterations

. Alterations

~qr 1. During normal' core alterations, two , Prior .to making normal alterations 1Sm's shall be operable;'one in the to the core the Sm's'shall be

" core quadrant where fuel or control- functionally tested and checked for

' l rods.are being' moved and one in an-  : neutron response. Thereafter, 1 adjacent. quadrant, except as specified .while required to be operable, the S m's will be checked daily for

-in 2.and 3 below.

response..

For an'Sm to be considered operable,- .

it shall be inserted to the normal Use of special moveable, dunking 7 operating level and shall. have a - . type detectors during initial-fuel minim a of 3 cps with all rods capable- loading and major core alterations of normal insertion fully inserted. in place of normal detectors is permissible as long as the detector is connected to the normal SW

2.-Prior to spiral unloading ~the S M's

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1shall be proven; operable as stated circuit.

above, however, during spiral .

Prior to spiral unloading or unloading the count rate may drop.

below 3 cps. - reloading the S m's shall be functionally tested. . Prior to

'3. Prior to spiral relo6d,; up to four. (4) spiral unloading the S M's should fuel: assemblies will be loaded into also be checked for neutron

their previous core positions next to response.

Keach of the 4 SM's to obtain-the-l required 3' cps. .Until these assenblies have been loaded, the 3 cps requirenent is not necessary.

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Spent' Fuel Pool Water Imvel D. Spent Fuel Pool Water Level D.-

Whenever irradiated fuel is stored in Whenever irradiated fuel is stored F .the spent fuel pool, the pool water- in the spent fuel pool, the water

^ : level shall' be maintained at or above level shall be checked and recorded 8.5 feet above the' top of the active daily.-

, -fuel.

" E. Control Rod Drive Maintenance E. Cont-ol' Rod Drive Maint ance

l. Requirenents for Withdrawal' l. Raluirenents for Withdrawal of 1-or 2 Control Rods of 1 or 2 Control Rods p , L A maxistst of two control rods k' separated by at least two' control-

cells in~all directions may be

- . withdrawn or removed fran the core for

<  : the purpose of performing control' rod drive maintenance provided that:

as % e Mode Switch is locked in the- %is surveillance requirenent is REREL position. % e' refueling a.

interlock which prevents more than the sane as given in 4.10.A.

-one control rod from being withdrawn may be bypassed for:one of.the

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control rods on which maintenance is

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Amendment:No. 66 3,10 3.10.A.2 Fuel Grapple Hoist Ioad Setting Interlocks Fuel handling is normally conducted with the fuel grapple hoist. The total load on this hoist when the interlock is rquired consists of the weight of the fuel grapple and the fuel assenbly. h is total is approximately 1500 lbs. in conparison to the load setting of 485 1 30 lbs.

3. Auxiliary noists Ioad setting Interlock Provisions have also been made to allow fuel handling with either of the three auxiliary hoists and still maintain the refueling interlocks. % e 485 1 30 lb load setting of these hoists is adequate to trip the interlock when a fuel bundle is being handled.

B. Fuel Loading 7t minimize the possibility of loading fuel into a cell containing no control rod, it is re uired that all control rods are fully inserted when fuel is being loaded into the reactor core. %is rquirenent assures that during refueling the refueling interlocks, as designed, will prevent inadvertent criticality.

C. Core Monitoring During Core Alterations

%e SPM'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 SIE's in or adjacent to any core quadrant where fuel or control rods are being moved assures adquate monitoring of that quadrant during such alterations.

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

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

The loading of up to four fuel bundles around the SIN's before attaining the 3 cps is permissible because these bundles were in a subcritical configuration when they were renoved and therefore they will renain subcritical when placed back in their previous positions.

Spent Fuel Pool Water Level The design of the spent fuel storage pool provides a storage location for 3181 fuel assenblies in the reactor building which ensures adequate shielding, cooling, and the 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 10 CFR 20. W e 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 pousible 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 quipped 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 sane time.

An endment No. 66,74 3.10-7

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~ 5.0 MAXR ~ DESIGN FEA'IURES 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' Cepany, on the south side of the Altmaha River in fAppling County near Baxley, Georgia. The' Universal Transverse Mercator Coordinates of the center of the reactor building are: Zone 17R LF 372,935.3n E and 3,533,765.2n N.

B.. Reactor Core

-1. Fuel Assablies

%e core shall consist of not more than 560 fuel assablies of the licensed cambination of 7x7 bundles which contain 49 fuel rods and 8x8 fuel bundles which contain 62 or 63 fuel rods each.

2. ' Control Rods

%e reactor shall contain 137 cruciform-shaped control rods.

C. Reactor Vessel

%e 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. Contaiment '

l. Primary Contaiment

%e principal design parmeters and characteristics of the primary contalment shall be as given in Table 5.2-1 of the FSAR.

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2. Secondary Contaiment* (See Page 5.0-la) 7

%e secondary contaiment 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 Contaiment Penetrations Penetrations to the primary contaiment and piping passing through such penetrations shall be designed in ecordance with standards set forth in Section 5.2.3.4 of the ESAR.

E. Fuel Storage

1. Spent Fuel All arrangment of fuel in the spent fuel storage racks shall be maintained in a subcritical configuration having a keff not greater than 0.95.

2. Nw Fuel

%e- new fuel storage vault shall be such that the keff dry shall not be greater than 0.90 and the keff flooded shall not be greater than 0.95.

Amendment No.-74, 91 5.0-1

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• OPERATING LIONSE DPR-57 E0 WIN I. HA'IQI NUCIEAR PIANT UNIT 1 Nrzutsr FOR 'IECIMICAL SPECIFICATICE GANGES t-x cs 1 Pursuant; to 10_CER 50.92, the following statements provide a sunnary of and -

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the basis for ,the proposed ch_anges:

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. - 1; ,-Change the number:of fuel assenblies that can be' loaded around a Source

- Range Monitor -(SIN) in order to assure that' 3 counts per second (cps)-

can .ber achieved without the use of additional sources or dunking chambers.

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i BASISii

%e four Sm detectors are . located, one ' per quadrant, roughly half 'a

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. core radius frm ' the ~ center. . ~ Although these are incore detectors and

" ' thus very sensitive when the reactor is fully loaded,: they--lose see of

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- .their ~ effectiveness when the reactor is partially defueled and the

- detectors are located see distance fra the array of renaining fuel'. .

GE's . spent fuel pool studies, GESSAR - NEDO-10741, Chapters 4. and 9,

- hshow . that: 1) sixteen or more fuel assenblies ~(i.e., 'four or more l control ' cells) must . be loaded - together before criticality is' possible;

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j and .2) for an uncontrolled 2x2 array of maxim m reactivity bundles, K o

' will always be less than .95. . In spiral loading-sequences in the Hatch

core, .an' array containing four or more control cells'will be at most two control cells . (i.e., about two

feet) away fra an SIN detector. - %e ,

" sensitivity- loss. on such a case is at most one decade of sensitivity

< ' . (i;e., Jabout one fifth of' the SINS logarithnic scale) . n is means that criticality cannot be reached during a spiral reload without'an operable ,

- SM detecting it. A spiral ~ sequence is any sequence - in ~ which : the-l central control cell is -last ' unloaded and first reloaded, all' fueled

- locations tare contiguous,- and . no imbedded ~ cavities or major peripheral-

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!- concavities .are permitted.

%e _ IIatch 1 . Technical ' Specifications; would require that the -fuel' ,

.assenblies be loaded into their; previous core position next to each of '

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- the : four SINS. . %e : loading of the - bundles around the SINS before y attaining the -3 cps 'is permissible because these' ' bundles - were ' in b subcritical configuration 'when they were renoved ' and therefore will renain 'subcritical when placed back in the previous position. %is request is identical to a request by Plant Hatch Unit 2 which was

, granted as Amendnent No. 39.

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Page'2-6 1 n e=. possibility of~ occurrence of an accident different than any evaluated 'in the FSAR is:not. created because there is no design change to any plant systems. . %is change does not significantly increase the probability or consequences of a previously analyzed accident because 1the. referenced. studies demonstrate inadvertent criticality with 4

= bundles is ' not possible. Further, the see subcritical assablies and arrangement; that was discharged is ' returned to the see core location..

Finally, - the ' safety margin is not . reduced because' the bundles remain considerably subcritical. Consequently, this change does not represent a significant hazards consideration.

2. Change - the . ' description of the control rod assemblies in Section 5.0

'-(Major Design Features) .of the Hatch-1 Technical Specifications to

- delete references to the specific materials and details of construction of the'-control blades.

, BASIS:

I%is change is intended to support the use in Hatch-1 of an arbitrary neber (up to 137) Type I General Electric Hybrid I Control-Rod (HICR)

" assemblies containing sme.hafnim as absorber material in place of the

> - boron carbide control' rods- presently in use. HICRs are . intended to be standard replacement. control' rod assablies for the General. Electric

. BWR/4 D-lattice operating. reactors. The- HICRs fom, . fit - and function are identical to that of the blade it replaces. % e HICR is. designed to increase control rod assembly life and to eliminate cracking of absorber.

. tubes containing boron carbide (B4C) . 'The essential differences-between the LHICR and the IER .' Z-4 D-lattice ' control- rod assemblies

. currently ~in use'are:

'a)- improved .B4C absorber rod tube material to - eliminate cracking during the lifetime of the control rod assembly, and

.b) some B 4 C absorber rods are replaced with solid hafnim absorber rods to increase blade life.

Other ' minor material and. dimensional changes are described in detail in NEDE-22290-A, " Safety Evaluation of the General Electric Hybrid-I N . Control Rod Assembly,"=Septaber'1983.

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Adherence kto the guidelines ' established for replacement' of' the standard -

U G S;a boren carbide- control blades .any require that . blades; in certain . core -

ilocations.be replaced at ea d refueling outage.. It is expected that use-

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, ,  ;; of - the ' HICR! blades Tin' these ~ locations - will allow operation of _ at:1least u  : ' < V J - twoT18-month : fuel: cycles without t replacing those . blades, . thus reducing

- Voutagef time, ? equipment duty - and ' personnel exposure - otherwise ^ required lforjblade repl= = nartt.

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Se. details (of ' (design - and i materials for the new blades"will not be cincluded in the revised . text, because those details .are tenecessary and '

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) inconsistent with other portions of the Design Features Section which do

. *W ' not provide design or materials-details.. Safety design bases which must 1

1be : met. byJ control? rods are : enumerated in, the Hatch-1 FSAR, Chapter 3.

g.- Analyses documented. inithe : approved topica1 report, NEIE-22290-A, have

~~' H 'shown that those design bases are met by the HICR blades; therefore, use tx _

'of- these blades will cause no reduction in the margin of safety.

fY ' , )Pbr example,ithe HICR weight and rod worth are the same as those for the

. y'  ; currently 'used control rod - assesbly. _ %erefore,. the scram speed and-f ) scram reactivity are also the same.- It follows then that the IJUR, PCPR J Jand ISPIJIGR limits are not affected by the HICR.

.Because the control rod worth is the same,' the capability of the reactor

to

achievel the Design 1 Basis' cold Jshutdown reactivity margin is not -

m Jaffected.s In addition, existing methodology for analysis of the control rod' withdrawal error translent and'the. control rod drop accident remains-

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u- EvalidEwith HICR.' assemblies installed.. It follows then, that = the probability of Lor consequences 'of. all. accidents .and transients-l previously : evaluated in the PSAR will not be J affected by use' of - the-HICRs..

S ef' possibility? of occurrence of an 3 accident different than any i evaluated (in the FSAR is not ' created by use -~ of the ' HICR assemblies,

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because there isno functional' change lin the control rods.

. L As shown above,, use of the HICR ' assemblies in Hatch-1 does not: increase

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thel probability or consequences of a previously analyzed accident, nor

.does it significantly reduce- any safety margin. We result of this -

design change ' is clearly within all acceptance criteria given in ~ the

. -aa ' Hat &-1 FSAR as noted above.

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$ Consequently, this change _ will ? not result in a significant hazards Se same change .was approved by Amendment No. 39 for 4.I foonsideration.

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ATTAOMENT 2 -

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3. - Revise ' the definition of CORE ALTERATION to clarify that core

' alterations only occur when there is fuel in the reactor vessel.

- BASIS: -

Wis Technical Specifications change is proposed to correct an oversight in the current definition which implies that a core alteration can occur even when no fuel is in the reactor vessel. It is reasonable to only classify as ~ a core alteration movement of quipnent within the core shroud while the vessel is fueled. % is change will replace the current Unit: l' definition with the sme one found in Unit 2 Technical Specifications.

%e primary detrimental adninistrative effect of the current Unit 1

' definition results from Specification 6.2.2.e, which requires a Senior Reactor Operator (SRO) any time a core alteration is performed. When fuel is_ in the vessel, an SRO should be and must be present; however, even when the vessel is empletely defueled, as it will be during the upcming outage, current specifications require the SRO's presence.

Approval-of this~ rquest would free the SRO for other duties following reactor. vessel defueling.

Incorporation - of this change into Technical -Specifications would not affect the plant in any mode other than refueling. Because the change only clarifies the definition and makes it the see as Unit 2 such that a " core alteration" does not occur unless there is a core to alter (fuel

. in the . vessel)', Plant safety is in no way jeopardized. Consequently, this change is consistent with Item (i) of the "Exmples of Amendments that are Considered Not Likely to Involve Significant Hazards Consideration" listed on page 14870 of the April 6, 1983 issue of the Federal Register and will not result in a -signigicant hazards consideration.

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