ML20247F992

Application for Amends to Licenses DPR-44 & DPR-56, Consisting of Tech Spec Change Request 88-09,changing Tech Specs to Permit Removal of Rod Sequence Control Sys & Reduce Rod Worth Minimizer Low Power Setpoint

ML20247F992
Person / Time
Site:	Peach Bottom
Issue date:	07/19/1989
From:	Hunger G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20247F993	List: ML20247F992 ML20247G012
References
NUDOCS 8907270295
Download: ML20247F992 (27)

Text

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-10 CPR 50.90-4 y g..

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N PHIL.AD ELPHI A' ELECTRIC 'COM PANY NUCLEAR GROUP HEADQUARTERS -

$955-65 CHESTERBROOK BLVD.'

WAYNE, PA 19087 5691 (215) 640 6000 July 19, 1989 Docket Nos. 50-277 50-278 License Nos. DPR-44 DPR-56 U.'S.-Nuclear _ Regulatory Commission ATTN:: Document Control Desk Washington,'D. C.

20555

SUBJECT:

PeachiBottom Atomic Power Station, Units 2.and 3 Technical Specifications Change Request

Dear Sir:

Philadelphia Electric Company hereby submits-Technical Specification. Change. Request No. 88-09, in accordance with 10 CPR.

50'.90, requesting an amendment to the Technical Specifications

.(Appendix A) of Operating License Nos. DPR-44 and DPR-56.

Information supporting this Change Request is contained in-to this letter, and the proposed replacement pages.are contained in Attachment-2.

This submittal requests changes to the Technical Specifications to' permit removal of the Rod. Sequence' Control System and also to reduce the Rod Worth Minimizer low power setpoint.

If you have any questions regarding this matter, please contact us.

Very truly yours,

. h.

n G. A. Hunger, Jr.

Director Licensing Section i

Nuclear Support Division Attachments Qgoj 1

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8907270295 890719

'PDR ADOCK 05000277 p

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,cc:-

W. T. Russell,Admi.nistrator, Region I, USNRC T. P. Johnson,-USNRC Senior Resident Inspector R. E. Martin,. Project Manager, USNRC-T. E. Magette,-State of Maryland D,

J. Urban, Delmarva' Power R. A.

Burricelli, Public Service Electric & Gas H. C..Schwemm,. Atlantic Electric T. M. Gerusky, Director, PA Bureau of' Radiological Protection!

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-COMMONWEALTH OF PENNSYLVANIA :

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CHESTER COUNTY I'

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D. R. Helwig, being first duly sworn, deposes and says:

That he is Vice President of Philadelphia Electric Company; the Applicant herein;'that he has read the enclosed request-(Change Request Number 88-09) for amendment of the Peach Bottom Units 2 and 3 Facility Operating-Licenses (DPR-44 and DPR-56) and knows the-contents thereof; and that'the statements and matters set forth therein are true and correct to the best of his knowledge, information and belief.

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Vice Presi t

Subscribed and sworn to before me this/Yrec day of' July 1989.

h, Yllbtle Notary Public

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NOTARIAL SEAL CATHERINE A. MENDEZ Notary Public Media Boro. Delaware Co.

My Commission Expires Sept 4.1989 w2---_'___-

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ATTACHMENT 1 i

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1 PEACH BOTTOM ATOMIC POWER STATION UNITS 2 AND 3 Docket Nos. 50-277 50-278 License Nos. DPR DPR-56 TECHNICAL SPECIFICATION CHANGE REQUEST

" Removal of~the Rod Sequence Control System and Lowering of the Rod Worth Minimizer Low Power Setpoint" l

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Docket Nos. 50-277 e

50-278 License Nos. DPR-44 DPR-56 1

Philadelphia Electric Company (PECo), Licensee under

. Facility Operating Licenses DPR-44-and DPR-56 for Peach Bottom 9

Atomic Power Station, Units 2 and 3, hereby requests that the Technical Specifications contained in Appendix A to the License be amended as indicated by a vertical bar in the margin of the pages in and listed here:

19, 99, 100, 101, 102,.102a, 107, 108 l

and 109.

Licensee proposes to eliminate the requirement for use of the Rod Sequence Control System (RSCS) and to decrease the power level setpoint above which the Rod Worth Minimizer (RWM) would no longer be required to be used from the existing 25% power requirements at both units to a new 10% power level setpoint.

These proposed amendments to the Technical Specifications are based on and are consistent with the NRC Safety Evaluation Report issued to J.

S.

Charnley, December 27, 1987, which approved Amendment 17 of General Electric Topical Report NEDE-240ll-P-A, " General Electric Standard Application for Reactor Fuel".

We request NRC approval so that the physical modifications may be completed and the amended Technical Specifications be made effective prior to the startup of Peach Bottom Atomic Power Station Unit 3 which is currently scheduled for late September, 1989.

Discussion of Changes Amendment 17 to the General Electric Topical Report NEDE-240ll-P-A f

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Docket Nos. 50-277

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50-278-License-Nos. DPR-44':

DPR-56 submitted to-the NRC and approved for.use as a-reference requested Commission approval to:

(1)

Eliminate the required useLof the RSCS.on those reactors having such a system, while retaining the:RWM to provide backup to the operator for control rod pattern control.

(2)

Lower the setpoint'for turnoff!of RWM to 10% of rated thermal power from its current-25% level at both Units.

The~GE report justifies the proposed changes by concluding:

(1) the RSCS is redundant to the RWM and is'therefore not needed to mitigate the consequences of.a-control? rod drop accident (RDA);

(2). the existing calculations demonstrate that a RDA is not a significant concern above 10% power and, therefore, a mitigation system such as the RWM is not needed for higher power level operation (above 10%); and (3) existing NRC sponsored RDA analysis indicates a less severe peak fuel enthalpy for a dropped control rod reactivity worth than previous analyses.

The newer analysis methodology demonstrates an extremely low probability for any event wnich would exceed fuel damage criteria.

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Docket Nos. 50-277 1

50-278 i

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License Nos. DPR-44 DPR-56' The NRC accepted General Electric Topical Report NEDE-240ll-P-A (Amendment 17) for use as a reference in licensee applications and provided guidance for those licensees wanting to.

make the. changes suggested in the GE Topical Report.

Consistent

.with the NRC's Safety Evaluation Report guidance for removal of the RSCS, the proposed Technical Specification changes will minimize reactor operations without the RWM available by allowingLonly one reactor start-up per calendar year to commence with the RWM out of service prior to or during the withdrawal of the first twelve control rods.

A discussion of the procedures which would be used in the event the RWM is unavailable is also included as required.

Finally,-the Banked Position Withdrawal Sequence is currently in use at Peach Bottom Atomic Power Stations Units 2 and'3, to reduce potential maximum rod worths.

The proposed changes to the Technical Specifications are summarized as follows:

elimination of the requirements for the Rod Sequence o

Control System; j

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lowering of the required setpoint for reactor power turnoff 1

o of the RWM from the existing 25% reactor power to a new setpoint of 10% of reactor power, and the addition of a limitation on the number of reactor start-ups allowed when the RWM is unavailable.

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Docket Nos. 50-2771 a

50-278 License Nos. DPR-44:

,-DPR-56 h

.. List of Proposed-Page Changes:

Page (Unit)

Change

.Page.19 (Units.2 Land'3)

Delete "and Rod Sequence 2.1.A BASES,Jfourth Control System".

paragraph Page 19-(Unit 3)

Add " change of".

2.1.A BASES, fourth paragraph Page 99 (Unit.2)

Delete "30% (*) power... is Surveillance Requirement (SR) continuing with'one' fully".

4.3.A.2.a.

Replace with "the RWM low power setpoint.

Each partially or fully-withdrawn operable control' rod shall be exercised at least one notch at least every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when operating above the RWM low i

power setpoint if there are three

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or more inoperable control rods i

or when operating above the RMW low power setpoint if there is one fully".

Additionally, delete the corresponding,

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Docket Nos.'50-277

3 50-278 License Nos. DPR-44 DPR-56 Note at the. bottom of

-the LCO column.

Page 99-(Unit 3)

Delete "21%' power... ~1s

'SR 4.3.a.2.a.

continuing with one' fully".

Replace with "the RWM low power setpoint. Each partially-or fully withdrawn operable control rod shall be exercised at least one notch at least every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when' operating above the.RWM. low.

power setpoint if there are three or more inoperable: control rods or when operating above the RWM low power setpoint if there is one fully".

Page 100 (Units 2 and 3)

Delete LCO 3.3.A.2.d.

Limiting Conditions for and replace with " DELETED".

Operation (LCO) 3.3.A.2.d.

Page 100 (Unita 2 and 3)

Delete SR 4.3.A.2.d. and SR-4.3.A.2.d.

replace with " DELETED".

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i Docket'Nos. 50-277-li.

50-278' License'Nos. DPR-44 DPR-56 j

Page-101 (Units 2 and 3)~

Remove redundant "must".

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-LCO 3.3.A.2.f.

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Page 101 (Unit 2)

Remove redundant "after".

SR 4.3.B.1.a.

Remove "30%(*) power".and' replace with "the Rod Worth

. Minimizer low power'setpoint".

Delete corresponding Note at the bottom of the LCO-column.

Page 101 (Unit 3)

Remove redundant:"after".

SR 4.3.B.1.a.

Remove "21% power" and replace with "the Rod Worth Minimizer low power setpoint".

Page 102 (Units 2 and 3)

Delete LCO 3.3.B.3.a. and' LCO 3.3.B.3.a.

replace with " DELETED".

Page 102 (Units 2 and 3)

Delete SR 4.3.B.3.a.

SR 4.3.B.3.a.

and replace with " DELETED".

Page 102 (Unit 2)

Delete Note.

Note corresponding to.-

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Docket Nos. 50-277 50-278 1

License Nos. DPR-44

-l DPR-56 l

LCO 3.3.B.3.a.

Page 102a (Units 2 and 3)

Add the sentence _"The Rod LCO13.3.B.3.b.

Worth Minimizer (RWM) low power setpoint'is greater than or equal to 10% of

. rated power."

In the next sentence, delete "below 25% rated power" and replace with "with thermal power-less than or equal to the Rod Worth Minimizer (RWM) low power setpoint".

Additionally, delete "or a second licensed operator shall verify that the operator at the reactor console i

is following the control rod program" and replace with "except as follows:"

Add proposed sections 3.3.B.3.b.l.,

3.3.B.3.b.2., and 3.3.B.3.b.3. which provide procedural criteria when withdrawing control rods when l

the RWM is out of service.

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Docket Nos. 50-277 50-278 License'Nos. DPR-44 DPR-56 Page 102a (Units 2 and 3)

-Delete "25% of rated power"'

SR 4.3.B.3.b.l.

and replace'with "the Rod Worth Minimizer low power setpoint".

.i Page 102a (Units 2 and 3)

Add "or technically qualified SR.4.3.B.3.c.

member of.the station technical staff".

Page 102a (Unit 2)

Delete LCO 3.3.B.3.c.

LCO 3.3.B.3.c and corresponding note and replace with " DELETED".

Page 102a (Unit 3)

Delete LCO 3.3.B.3.c. and-l LCO 3.3.B.3.c.

replace with " DELETED".

Page 107 (Units 2 and 3)

Delete sentence "The 3.3 A and 4.3.A BASES, use of the individual rod Section 2 bypass switches in the...

control rod is known."

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Page 108 (Units 2 and 3)

Delete " rod sequence 3.3.B and 4.3.B BASES, control system Section 1 and the".

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.g Docket Nos. 50-277' 50-278 License Nos.'DPR-44 DPR !

- Page 108 (Units 2 and 3)'

Delete "and sequence 3.3.B and 4.3.B BASES, mode of the Rod Sequence Section 3 Control System'(RSCS) restrict" and replace with " restricts".

Additionally, delete the second sent'ence which discusses the group notch mode of the RSCS.

Page;109 (Unita 2 and 3)

In first paragraph,.first 3.3.B and 4.3.B BASES, sentence, delete

- Section 3 "and RSCS are" and replace with "is".

In first and fourthLsentences of the first paragraph, delete-

"20" and replace with "10".

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Delete second and third i

sentences of the first paragraph which relate to the justification for the

'l power limits.

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' Docket Nos. 50-277

' ' - ' 50-278 License Nos. DPR-44 L;

DPR-56 Delete all of the second paragraph of Page 109.

In third paragraph, first sentence, delete words "and the sequence mode of-the Rod Sequence Control-System provide" and replace with.

"provides".

In third paragraph, first sentence,-

delete "and the_ group notch mode of RSCS requires notch movement of rods".

Additionally, delete " systems limit" and replace with "RWM system limits" In third paragraph, delete the sentence "They serve l

as a backup to... worth of control rods."

Additionally, delete the following two j j I

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'50-278 License Nos. DPR-44 DPR-56'

' In the'eventEthat....

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conformance functions of-this system", and "In this cases the RSCS is backed up by independent' procedural controls".

Replace these sentences with'

a. discussion of the procedural controls for startup'with RWM inoperable..

In-the following' sentence.

of the third paragraph,' delete "funtions" and replace with

" function" and delete'"and RSCS make".and replace with 1

"makes".

In third paragraph, delete "20 percent of rated these devices force" and replace with "10 percent of rated, the RWM forces".

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• g Docket Nos. 50-277 50-278 License Nos. DPR-44 DPR-56 In third paragraph, last two sentences, delete "20" and replace with "10"..

Page 109 (Unit 2)-

Delete Note at bottom I'

.3.3.B and 4.2.B BASES of page.

Discussion of the Safety Significance:

Elimination of RSCS The NRC.SER from A. C. Thadani (NRC) to J. S. Charnley (GE) dated December 27, 1987, approving NEDE-240ll-P-A concludes that:

1) the RSCS is not needed for BWR 4 and 5 reactors; 2) operation without the RSCS is acceptable; 3) suitable provisions should be made for RWM use and operator backup; and 4) rod patterns should be in accord with Banked Position Withdrawal Sequence (BPWS) concepts.

The following items describe the methods for satisfying items 3 and 4.(above) as suggested by the Safety Evaluation Report:

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Docket Nos. 50-277 50-278 License Nos. DPR-44 DPR-56 1.

The proposed RWM Technical Specification changes allow only one reactor startup per calendar year with the RWM unavailable prior to or during the withdrawal of the first 12' control rods.

2.

Procedures are in place which ensure that a second reactor operator provides an effective and truly independent.

control rod movement monitoring function.

3.

A BPWS pattern is used at Peach Bottom (a BWR 4 reactor),

which reduces potential maximum rod worths.

The above discussion demonstrates consistency and applicability to the requirements for removal of the RSCS as described in the SER.

Based on the GE Topical Report and the NRC SER, removal of the RSCS with the above described controls in place, has been determined to not pose a significant safety concern.

Reduction of RWM Shutoff from 25% to 10% of Rated Thermal Power The NRC contracted Battelle National Laboratory (BNL) to perform studies of a Rod Drop Accident and to improve the RDA analysis methodologies.

These continuing studies indicate a substantial reduction (from earlier GE studies) in enthalpy for a given control rod worth as a result of better geometrical and :

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1 Docket Nos. 50-277 i

50-278 l

License Nos. DPR-44 l

l DPR-56 moderator reactivity feedback modeling.

The results from the BNL studies for "zero" power RDA events (the most conservative initial j

condition) indicate less than 130 cal /gm for maximum control rod l

worths (no control rod pattern errors).

The NRC concluded from i

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these studies and results that there is a large likelihood that l

control rod error patterns would not lead to a control rod worth l

which could exceed fuel damage limits in a 'zero' power RDA.

The BNL study also clearly showed that in the ten percent reactor power range (and above) peak enthalphies would always be well below limits.

Based on the NRC SER, reduction of the RWM turnoff setpoint does not pose a significant safety concern.

NO SIGNIFICANT HAZARDS CONSIDERATIONS:

The NRC has provided guidance concerning the application of the standards for determining whether a license amendment involves any significant hazards consideration by providing examples (51FR7744) of amendments that are considered not likely to involve significant hazards considerations.

Example (iv) is a relief granted upon demonstration of acceptable operation from an operating l

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restriction that was imposed because acceptable operation was not

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1 yet demonstrated.

This assumes that the operating restriction and

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l the criteria to be applied to a request for relief have been

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.a Docket Nos. 50-277 50-278 License Nos..DPR-44 DPR-56 a

. established in a prior review and that it is justified in a satisfactory way that the criteria have been met.

This proposed Technical Specification change request is similar to example (iv) in that the NRC has reviewed the proposed change and established criteria for removing the RSCS and lowering the RWM low power-setpoint..

Additionally, the proposed changes are in compliance with the NRC criteria.

The proposed Technical Specification changes are in conformance with the changes determined to be acceptable by the NRC as set forth in the NRC SER transmittal'from A.

C. Thadani, NRC, to J. S. Charnley, General Electric, " Accept 7nce for Referencing of Licensing Topical Report NEDE-240ll-P-A, "Ge.. Oral Electric Standard Application for Reactor Fuel," Rev. 8, Amendment 17," dated December 27, 1987.

Deletion of the RSCS and reduction of the RWM low power set point from 25% to 10% does not involve a significant hazards consideration because it does not:

1.

Significantly increase the probability of occurrence or the consequences of an accide'nt or malfunction of equipment related to safety as previously evaluated in the FSAR.

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'C DocketiNos. 50-277

'50-278-License Nos.JDPR-44 DPR-56 Deleting 3the-RSCS and. changing the low power set. point on

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the RWM has no effect on.the probability of equipment-malfunction in other systems or within the RWM.

1 The probability of occurrence of an. accident is not-affected by this change.

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

The consequences of an RDA as evaluated in the PBAPS'UFSAR will not be affected by this modification.

An extensive

.probabilistic-study was performed'by the NRC staff (letter and; enclosure from B. C. Rusche, NRR, to R. Fraley, ACRS, dated June 1, 1976, " Generic Item IIA-2 Control Rod Drop Accident (BWRs)").

This study indicated that there was not a need for the RSCS.

Furthermore, improved methodologies in.the RDA analysis methods (e.g. BNL-NUREG 28109,

" Thermal-Hydraulic Effects on Center Rod Drop Accidents in a Boiling Water Reactor," October 1980) indicated that the peak fuel enthalples resulting from an RDA are significantly lower than previously determined by less refined methodologies.

The RSCS duplicates the function of the RWM.

So long as the RWM is operable, the RSCS is not needed since the RWM L

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E Docket Nos. 50-277 50-278 License Nos. DPR DPR-56 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 withdrawal movement and compliance with the prescribed control rod pattern be verified by a second licensed operator or technically qualified member of the station technical staff.

The verification process is-controlled procedurally to ensure a high quality, independent review of contr61 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 start-up per calendar year with the RWM out of service prior.to or

.during the withdrawal of the first twelve control rods.

1 All.the above taken together demonstrate consistency and applicability to those conclusions reached in the ;eferenced NRC SER, and substantiate the conclusion that there will be no increase in the consequences of an RDA as evaluated in the PSAR as a result of eliminating the RSCS.

l There will also be no increase in the consequences of an RDA as evaluated in the UFSAR due to lowering the RWM set

~l point from 25% to 10%.

The effects of an RDA are more

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severe at low power levels and are less severe as power level increases.

Although the original calculations for i

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License Nos. DPR-44 DPR-56 the RDA were performed at 10% power, the NRC required that

.the generic BWR' Technical' Specifications be written.to require operation of the RWM below 25% power-in order to ensure conservatism.

However,-GE continued to. perform the g

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RDA. analyses at and below 10% power-because these produced f

more conservative analytical results.

More refined' I:

calculations by BNL (BNL-NUREG 28109, " Thermal-Hydraulic Effects on Center' Rod Drop' Accidents in a Bolling Water

. Reactor", October 1980) have shown that even with the maximum single control rod position error, and most multiple control rod error patterns, 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 set point from 25% to 10% will not-result in an increase in the consequences of an RDA as evaluated in the UFSAR.

The previously referenced NRC SER has concluded this RWM set point reduction to be acceptable.

2.

Create the possibility for an accident or malfunction of a different type than any evaluated in the FSAR.

Operation of the RSCS and RWM cannot cause t prevent an accident.

They function to minimize the consequences of an l

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Docket Nos. 50-277 50-278 License'Nos. DPR-44 DPR-56 RDA.

The RDA is already evaluated in the UFSAR, and-the effect of thisLproposed change on the analyses is discussed in Item 1 above.

Elimination of the RSCS and lowering the RWM set point'will have no impact on the operation of any other systems, and hence would not contribute to a malfunction in any other equipment nor create the possibility for an accident to occur which has not already been evaluated, 3.

Involve a significant reduction in the margin of safety.

Elimination of the RSCS will not lower the margin of safety.

for the. reasons discussed in Item 1 above and sammarized below:

a)

An extensive NRC study has determined that the possibility of an RDA resulting in unacceptable consequences is so low as to negate the requirement l

for the RSCS.

b)

Recent calculations have determined that the consequences of an RDA are acceptable above 10% power. - _ _ _ - _ _

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Docket Nos. 50-277-50-278 License Nos. DPR-44 DPR-56 c)-

The RSCS is redundant in function to the RWM.

Eliminating'the RSCS does-not eliminate the control-rod. pattern' monitoring function performed by the RWM.

-d)

HTo ensure that the RWM will be in service when-required, the proposed RWM Technical Specification-will be. revised to allow only'one startup per calendar year with the RWM out of service prior to or during-the withdrawal of the first twelve control rods.

If the RWM is out of service below 10% power, control rod movement and compliance with prescribed control rod patterns will be verified by a.second licensed operator or technically qualified member of the station technical staff.

This situation is controlled by a station procedure which specifically requires the following:

o Plant Management approval is required in order for the operator to bypass the inoperable RWM.

o A second operator or technically qualified staff member, with no other duties, is required to verify the first operator's actions while the first operator performs rod movements.

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DPR-56

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..The'startup and the-shutdown sequences with their-respective signoff sheets provide for verification by the second operator after each rod movement step is completed by the'first operator.

o The.startup and shutdown sequences follow the-same control-rod' patterns that the RWM enforces if it were not bypassed.

There is'no significant reduction in the margin of safety resulting'from lowering the RWM set point from 25% to 10%

because calculations by GE and BNL have shown-that even with the maximum single. control rod position error, and most multiple error patterns, the peak fuel rod enthalpy during an RDA from these patterns would not exceed the NRC limit (280 cal /gm) above 10% power.

In summary, GE has provided technical justification for the proposed changes in the Topical Report NEDE-240ll-P-A and associated references which justify the acceptability of l

the proposed changes.

The NRC has reviewed and accepted the GE analysis and l

L provided guidelines for licensees wanting to make the changes proposed in NEDE-240ll-P-A and approved in the NRC i 1

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Docket Nos. 50-277

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DPR-56 1

SER issued. December 27, 1987 to J. S. Charnley of General i

Electric.

3 The proposed changes are consistent with those approved in the NRC SER and the guidelines set'forth therein.

Therefore, there is no significant reduction in a margin of safety.

Environmental Considerations:

.)

This amendment would eliminate the requirement for use of the. Rod Sequence' Control System and would decrease the setpoint for Rod Worth Minimizer-shutoff from the. existing 25% of rated thermal power to a new setpoint of 10% of rated thermal power.

The proposed changes do not involve any increase in the amounts'and do not involve any changes in the types of effluents that may.be released offsite.

No increase will occur in the individual or cumulative occupational radiation exposure.

i Therefore, no environmental report is required.

Conclusion The Peach Bottom Plant Operations Review Committee and the Nuclear Review Board have reviewed the proposed changes to the

-Technical Specifications and have concluded that they do not involve l-

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License'Nos'. DPR-44.

.DPR-56' an..unreviewed. safe'ty questions or' involve Significant Hazards

. Considerations or an environmentaliconsideration,.and will not endanger;the health and' safety of the public.

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