ML20012C442

Application for Amends to Licenses DPR-71 & DPR-62,to Permit Removal of Rod Sequence Control Sys & Reduce Rod Worth Minimizer Cutoff Setpoint to 10% Rated Thermal Power

ML20012C442
Person / Time
Site:	Brunswick
Issue date:	03/14/1990
From:	Cutter A CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20012C443	List: ML20012C442 ML20012C444
References
NLS-90-044, NLS-90-44, NUDOCS 9003210295
Download: ML20012C442 (8)

Text

t t Carollne Power & Light Company SERIAL: NLS-90-044 e o no mi . n. c. N c risor 10CFR50.90 8BTSB25 A. B CUTTER Vice President Nuc5at Services Departrnent United States Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-325 & 50-324/ LICENSE NOS. DPR-71 & DPR-62 REQUEST FOR LICENSE AMENDMENT ROD SEQUENCE CONTROL SYSTEM / ROD WORTH MINIMIZER Gentlemen In accordance with the Code of Federal Regulations, Title 10, Parts 50.90 and 2.101, Carolina Power & Light Company (CP&L) hereby requests a revision to the Technical Specifications for the Brunswick Steam Electrio Plant (BSEP),

Unita 1 and 2.

The proposed amendment revises the Brunswick Technical Specirloations to (1) permit the removal of the Rod Sequence Control System (RSCS) and (2) reduce the ' lod Worth Minimizer (RWM) outoff setpoint from 20% rated thermal power to 10% rated thermal power. The proposed changes are consistent with the actions approved in the NRC Safety Evaluation Report issued to J. S.

Charnley on December 27, 1987, regarding Amendment 17 of General Electric Topical Report NEDE-24011-P-A, " General Eletrio Standard Application for Reactor Fue1."

Enclosure 1 provides a detailed description of the proposed changes and the basis for the changes.

Enclosure 2 details the basis for the Company's determination that the proposed changes do not involve a signifloant hazards consideration.

Enclosure 3 provides the proposed Technical Specification pages for Unit 1.

Enclosure 4 provides the proposed Technical Specification pages for Unit 2.

In order to allow time for procedure revision and orderly incorporation into copios of the Technical Specifications, CP&L requests that the proposed amendments, once approved by the NRC, be issued with an effective date to be no later than 60 days from the issuance of the amendment.

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Document Control Desk

< .(NLS-90-044) / Page 2 Please refer any questions regarding this submittal to Mr. M. R. Oates at (919) 546-6063.

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Enclosures:

1. Basis for Change Request

2. 10CFR50 92. Evaluation

3. Technical Specifloation Pages - Unit 1

4. Technical Specirloation Pages - Unit 2 00: Mr. Dayne 11. Brown Mr. S. D. Ebneter Mr. W. ll. Ruland Mr. E. G. Tourigny A. B Cutter, having been first duly sworn, did depose and say ' that the information contained herein is true and correct to' the' best of his information, knowledge and belief; and the. sources of his information are officers, employees, contractors, and agents of Carolina Power & Light Company.

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ENCLOSURE 1 BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 NRC DOCKETS 50-325 & 50-324 OPERATING LICENSES DPR-71 & DPR REQUEST FOR LICENSE AMENDMENT .

ROD SEQUENCE CONTROL SYSTEM BASIS FOR CHANGE REQUEST proposed Change  :

The proposed amendment revises the Brunswick Technical Specifications to - +

(1) permit the removal of the Rod Sequence Control System (RSCS) and -

(2) reduce the Rod Worth Minimizer (RWM) outoff low power setpoint from 20% ~  !

rated thermal power to 10% rated thermal power. The proposed changes are l consistent with the actions approved in the NRC Safety Evaluation: Report I issued to J. S. Charnley on December 27, 1987, regarding Amendment 17 of General Electrio Topical Report NEDE-24011-P-A, " General Electrio Standard i Application for Reactor Fuel." j Basis 4 The Rod Sequence Control System (RSCS) restricts' rod movement through the use of rod select, insert, and withdrawal blocks to minimize the individual worth of control rods to lessen the' consequences of a postulated Rod Drop Aooident (RDA) with control rod movement restricted. The RSCS is a hardwired ~(as opposed to the RWM which is computer controlled), redundant -backup to the Rod L Worth Minimizer. It is independent of the RWM in terms.of inputs and outputs '

but the two systems are compatible. The RSCS is designed to monitor and block necessary operator control rod selection, withdrawal and insertion actions, and thus assist in preventing signifloant control rod pattern errors which  !

could lead to a control rod with a large reactivity worth (if droppod).- A signifloant pattern error is one of several. abnormal events all of which must ,

occur to have a RDA which might exceed the fuel enthalpy limit criteria set 'i for the event. The RSCS was designed only for possible mit16ation.of the.RDA and is active only during low power operation (ourcently'less than 20% of rated thermal power) when a RDA might be signifloant.- It'provides rod blocks

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on detection of a pattern error. It does not prevent a RDA. A similar pattern control function is performed by the RWM, a computer controlled system.

In response to Amendment 17 of General Electrio Topical Report NEDE-24011-P-A,

" General Electric Standard Application for Reactor Fuel," the NRC issued- a > '

Safety Evaluation Report, dated December 27, 1987, approving (1) elimination -

of the RSCS while retaining the RWM to provide backup to the operator for control rod- pattern control and- (2) lowering the setpoint for outoff of the i RWM to' 10% of rated thermal power from its current 20%-level. The proposed '

amendment is consistent with the guidelines set forth in the NRC SER.  ;

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As discussed above, the RSCS is a redundant backup to the RWM. Also, the.RWM verifies compliance with Banked Position Withdrawal Sequence (BPWS) which prevents the RDA from exceeding the 280 cal /gm fuel enthalpy criterion. When the RWM is operable, the RSCS is not needed since the RWM prevents control rod pattern error. In the event the RWM is out of service, after the withdrawal of the first 12 control rods, the proposed Technical Specifications require that control rod movement and compliance with the prescribed control rod -

pattern (i.e., Banked Position Withdrawal Sequence per Technical Specification 4.1.4.1.2) be verified by a second licensed operator or technically qualified member of the technical staff. The verification process is controlled procedurally to ensure a high quality, independent review of control rod movement. In addition, to further minimize control rod movement at low power with the RWM out of service, the proposed Technical Specifications will permit only one plant startup per calendar year with the RWM,out of service prior to or during the withdrawal of the first twelve control rods. These measures, taken together, demonstrate consistency and applicability' to those conclusions reached in the referenced SER and prevent significant control rod pattern errors which could lead to a control rod with a large reactivity worth (if dropped).

Lowering of the RWM setpoint from 20% to 10% of rated thermal power is acceptable because the effects of an RDA are more severe at lower power levels and are less sovere as power level increases. Although the original calcula-tions for the RDA were performed at 10% power, the NRC required that the generlo BWR Technical Specification be written to require operation of the RWM below 205 power in order to ensure conservatism. However, GE continued to perform the RDA analysos at and below 10% power because these produced more conservativo analytical results. Recently, more refined calculations by Brookhaven National Laboratory (BNL) have shown that' even with the maximum single control rod position error, the peak fuel rod enthalpy reached during an RDA from these control rod patterns would not exceed the NRC . limit of 280 cal /gm for RDAs above 10% power. confirming the original GE analyses. Hence, lowering the RWM sotpoint from 20% to 10% will not result in an increaso.in the consequenoos of an RDA as evaluated in the FSAR. The previously referenced NRC SER has concluded this RWM setpoint reduction to be acceptable.

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L ENCLOSURE 2 BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 NRC DOCKETS 50-325 & 50-324 OPERATING LICENSES DPR-71 & DPR REQUEST FOR LICENSE AMENDMENT ROD SEQUENCE CONTROL SYSTEM 10CFR50 92 EVALUATION The Commission has provided standards in 10CFR50-92(o) for determining whether a significant hazards -consideration exists. A~ proposed amendment to anL operating license .for_ a facility involves. no significant hazards consideration :

if operation of the facility in accordance with' the proposed l amendment would not: (1) involve a signifloant-increase in_the probability or consequences of-an accident previously evaluated, (2). create _ the possibility of a new or-different kind _ of accident from any aooident1previously evaluated, or (3) involve a significant ' reduction in a margin _of safety. Carolina Power & LLight Company has reviewed this proposed license amendment request and determined that its adoption w0uld not involve a significant hazards consideration. The bases for this determination are as follows:

Proposed Change The proposed amendment revises the-Brunswick Technical Spooitioations to (1) permit the removal of the Rod Sequence Control. System (RSCS) and (2) . reduce the Rod Worth Minimizer (RWM) outoff setpoint from 205 ' rated thermal power to 105 rated -thermal power. The proposed changes are consistent with the actions

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approved in the NRC Safety Evaluation Report lasued to J. S. Charnley on December 27, 1989, regarding Amendment 17 of . General Electric Topical' Beport NEDE-24011-P-A, " General Electrio Standard Application' for Reactor Fuel."

pasia, The change does not involve a significant hazards consideration for the following reasons:

1. The proposed amendment does not involve a significant increase in' the probability or consequences of'an accident proviously evaluatedc The Rod Sequence Control System (RSCS) and Rod Worth Minimizer-(RWM) are-not required for nor do they support the proper. operation of any other sys te m. Hence, deleting the RSCS and changing the low power _setpoint on the' RWM has 'no effect on the probability of failure of, equipment in-other systems or. within the RWM. -

m The probability of occurrence of'an accident is not affected by thisl

, change. These changes impact onlyf the rod ~ drop accident _(RDA) analyses.

The probability of an RDA is dependent:only Don the control rod drive system and mechanisms themselves, and not'in any way on the 'RSCS or RWM.

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i The consequences of an RDA as evaluated in the Brunswick FSAR will not be -

affected by these changes. A probabilistic study was performed by the NRC staff (letter and enclosure from B. C. Rusche, NRR, to R. Farley, ACRS, . I dated June 1,1976, "Generio item II A-2 Control Rod Drop Accident '.

(BWRs)"). Furthermore, improvements in the RDA analysis methods (e.g.,

BNL-NUREG 28109, " Thermal Hydraulio Sffects on Control Rod Drop Accident in a BWR," _0ctober 1980) indicated that the peak fuel enthalples resulting from an RDA are significantly lower than previously determined by less _

refined methodologies. -

The RSCS in a redundant backup to the RWM. _ When the- RWM'is operable, the  ;

RSCS is not needed since the RWM prevents deviation from BPWS patterns.

In the ovent the RWM is out of servloe, after the. withdrawal of the first r 12 control rods, the' proposed Technical Spectrications require that: [

control rod movement and compliance with the prescribed control rod - i pattern (i.e., Banked Position Withdrawal-Sequence per Technical Spoolfication 4.1.4.1.2) be verified by a second licensed operator or  ::

technically qualified member of the technical staff.. The; verification proooss is controlled procedurally. to ensure _ a high quality, independent review of control rod movement. In addition, to further minimize control {

rod movement at low power with the RWM out of service, the proposed ' '

Technical Speoirications will permit only one plant start-up per eclendar year with the RWM out of service prior to or during the withdrawal of the- i first 12 control rods. Those measures, taken together, demonstrate -  !

consistency and applicability to those conotusions reached'in the  !

referonood SER, and substantiate the conclusion that there- will be no '

significant increase in the consequences of an RDA as evaluated in the  ;

FSAR as a result of eliminating the RSCS.

There will also be no significant increase in the consequenoea of an RDA  ;

an ovaluated in the PSAR due to lowering-of the RWM'setpoint from 205 to 10% of rated thermal power. The effects ~of an RDA are more sovere at icw .

power levels and are less sovero as-powcr level increases._ Although the original calculations for the RDA were performed at 10% power, the NRC  ;

required that the generio DWR Technical Specification be written to }

require operation of the RWM below 205 power in' order to. ensure . .

concorvatism. However, GE continued. to perform the RDA analyses at and ,

below 10% power because' thoso produced more conservativo analytioni _

results. Recently, more refined calculations by Brookhaven National .

Laboratory (BNL) have shown that-oven with the maximum'aingle control rod  !

position error, the peak fuel rod enthalpy reached during_an RDA from- -

those control rod patterns would not exceed the.NRC limit of 280 cal /gm j for RDAs abovo 105 power, confirming- the original GE analyses. Hence, 1 lowering the RWM sotpoint from 20% to 105 will not result in a significant t increaso-in the consequenoon of an RDA as evaluated in' the PSAR.

2. The proposed amendment does not create the possibility of a new or -

diiTerent kind of aooident from any acoldent previously evaluated.

Operation of the RSCS and RWM cannot cause er prevent' an aooident.- They i function to minimize the consequences of an RDA. The RDA is already evaluated in the PSAR, and the effect of this proposed change _ on the- '

analysos is discussed in -Item 1 above.

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l Elimination of the RSCS and lowering the RWM setpoint will- have no impact 5 on the operation of any other systems, and hence, would not contribute to i a malfunction in any other equipment nor create the possibility for any .,

accident to occur which has not already been' evaluated.

3. The proposed amendment does not involve a significant reduction in the  ;

margin of safety.  :

Elimination of the RSCS will not result in a significant reduction in the i margin of ' safety for the following reasons; (a) An NRC study discussed ~in Item 1 above has determined-that the probability of an RDA resulting in unacceptable consequences was so small that backfit of the RSCS was not needed.

(b) The RSCS is a redundant backup to the RWM. Eliminating the RSCS does not eliminate the control rod pattern monitoring function performed by the RWM. Furthermore, to ensure that the RWM will be in . service when' required, the proposed RWM Technical Specification, allows only one startup per calendar year with the RWM out of service prior to or during the withdrawal of the first 12 control rods. IT the RWM_is out of service below 105 power, control rod movement and compliance with prescribed control rod patterns- (i.e., Banked Position Withdrawal Sequence per Technical Specification 4.1.4.1.2) will be verified by a second licensed operator or technically qualified member of-the technical staff. This situation is controlled by a procedure which specifically requires the following: '

i) ?lant Hansgement approval is required in order for the operator

• to bypass the inoperable RWM.

11) A second operator or technically qualified staff member, with I no other duties, is required to V9rify the first operator's  !

actions while the first operator is performing rod movements.-

lii) The startup and shutdown sequences with their respective signoff sheets are provided to the second operator for verification of each step and rod movement made by the first operator. i iv) Additional plant procedures provide the operator with shutdown instructions that would result in a control rod pattern allowed by the RWM it _that system were not bypassed and was .,

controlling. These instructions identify,:for the operator,

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the RWM shutdown step to be initiated for further rod insertion ,

below the RWM low power setpoint (i.e.,105 of rated thermal .

power).

There is no significant reduction in the margin of fsafety resulting from lowering the RWM setpoint from 20% to 10% of rated thermal power because 1 calculations by GE and DNL have- shown that even with the. maximum single

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i control rod position error. and most multiple error patterns, the peak fuel rod enthalpy during an RDA from these patterns would not . exceed the. "

Nhc limit (280 cal /gm) when operating above 105 of rated thermal. power. .  ;

SUMMARY

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GE has provided technical justification for the proposed changes in the .,

Topical Report NEDE-24001-P-A and associated references which justify the _l acceptability of the proposed changes.  ;

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The NRC has reviewed and accepted the GE analysis and provided guidelines for licensees to follow when, requesting the changes proposed in NEDE- 1 24001-P-A and approved in the:NRC SER issued December' 27, 1987, to'J..S.. l Charnley of General Electrio. ,

The proposed changes are consistent with.those-approved'in the NRC SER and .-

the guidelines set forth therein. 'j reduction in a margin of safety. . Therefore, thereLis no significant' t i

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