ML17317B023

Requests Revisions to Rod Position Indicator Tech Specs to Remove References to Part Length Control Rods & Change Reporting Requirements.W/Encl Addl Details & Revised Tech Spec Pages

ML17317B023
Person / Time
Site:	Cook
Issue date:	03/19/1979
From:	Disbrow R INDIANA MICHIGAN POWER CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
AEP:NRC:00145, AEP:NRC:145, NUDOCS 7903270461
Download: ML17317B023 (31)

Text

REGULARLY INFORMATION DISTRIBUTIO YSTEM (RIDS)

ACCESSION NBR 790327046i DOC ~ DATER 79/03/19 NOTARIZKDR NO DOCKET ¹ FACIL!50~315 DONALD C ~

COOK NUCLEAR POWER PLANTE UNIT ig INDIANA &

05000315" 50~316 DONALD C, COOK NUCLEAR POWER PLANTg UNIT 2g INDIANA &

05000316'UTH

~ NAME AUTHOR AFFILIATION DISBROW R ~ K ~

INDIANA & MICHIGAN POWER CO ~

REC IP ~ NAME REC IP IENT AFFILIATION

'DENTONt H ~ R ~ 'FFICE'F NUCLEAR REACTOR REGULATION

SUBJECT:

REQUESTS REVISIONS TO ROD POSITION INDICATOR TECH 'SPECS

'TO REMOVE REFERENCES TO PART LENGTH CONTROL~ RODS TO CHANGE REPORTING REQUIREMENTS'SW/ENCLT ADDL DETAILS

& REVISED TECH SPEC PAGES'ISTRIBUTION'ODE:

ASSIS COPIES RECEIVED:LTR Q" ENCL '

SIZEIJ TITLE! GEN RAL DISTRIBUTION FOR AFTER ISSUANCE OF.OPKRATING LZC

'OTES:~~C$

~ck

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RKC IP IENT'D CODE/NAME ACTIONS 05 BC QRB~f INTERNAL' REG FI 12 15 CORE PERF BR 17 ENGR BR 19 PLANT SYS BR 21 EFL'T TRT SYS EXTERNALS 03 LPDR 23 ACRS COPIES LTTR ENCl 7

7 1

1 2

2 1

1 1

1 1

1 1

1 1

1 16 16 RECIPIENT" ID CODE/NAMK'2 NRC PDR 14'A/EDO

'16 AD SYS/PROJ 18 REAC SFTY BR 20 EEB 22 BRINKMAN 04 NSIC COPIES LTTR ENCLt 1

1 1

'1 1

1 1

1 1

1 1

MAR 28 1978 wq ss $79.

TOTAL NUMBER OF COPIES REQUIRED'TTR 38 ENCL 38 LtL,,

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INDIANA 5 MICHIGAN POWER COMPANY P, o. Box 18 BOWLING GREEN STATION NEW YORK, N. Y. 10004 March 19, 1979 AEP:NRC:00145 Donal'd C.

Cook Nuclear Plant Unit Nos.

1 and 2

Docket Nos.

50-315 and 50-316 License Nos.

DPR-58 and DPR-74 Mr. Harold R.'Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.

20555

Dear Mr. Denton:

This letter requests revisions to the Rod Position Indicator (RPI) System Technical Specifications for the Donald C.

Cook Nuclear Plant Unit Nos.

1 and 2.

Appendix A Technical Specifications 3/4.1.3.1, and 3/4.1.3.2 have.been revised to indicate the following:

(1)

References to part-length control rods have been removed.

(2)

The Technical Specification Reporting Requirements for the RPI System have been changed to allow for the submittal of a Licensee Event Report (LER) only when the actual, as opposed to the indi-

cated, rod misalignment is greater than

+

12 steps from the bank demand position.

Please note, though, that the surveillance requirements for the RPI System have not been lessened.

Only the resulting reporting requirements are being changed.

More detailed discussions of the above Technical Specification change request and a copy of the corresponding revised pages are contained in the Attachments to this letter.

The revisions to the RPI Reporting re-quirements contained therein have been discussed with members of your staff and found acceptable.

AEPSC respectfully requests that this Technical Specification change request package be processed in an ex-peditious manner to preclude the future submittal of LERs due to false indic'ation of 'the RPI system.

V 9032704g(

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Mr. Harold R, Denton, Director

~\\2M AEP:NRC:00145 This Technical Speci'ficati'on change request has been reviewed by the PNSRC and the AEPSC NSDRC, in accordance with the appropriate provisions of our Techni'cal Specificati'ons.

The result of these re-vi'ews tndi'cates that in no i'nstance will the subject Technical Speci-fication change adversely affect the health and safety of the public.

Thi's Technical Specification change request is considered to be a Class I'I'icense Amendment as per the provisions o'f 1'0,'CFR 170.22.

As required by Part 170 Subsection 22 a c e k for $1,600.00 accom-panies thi's-submittal.

uly yours, RED:em R.

E. Disbrow Vice President Sworn and subscribed to before me this//~day of March, 1979 in New York County, New York Notary P

GREGORY ft GURtCAN NotarY Public, State of New York No. 31-4643431 Qualified in New York Countgj <>

Commission Expires March 30, 194, cc:

R.

C. Callen G. Charnoff R. Walsh P.

W. Steketee R. J. Vollen D. V. Shaller-Bridgman R.

W. Jurgensen

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m I ATTACH ZNT "i'c" TO iXHP: "~iC: 00~ 45 DOii7(LD C.

COOK VUCL;.:Zi PLA!iT U'lTT 'lO.

1

Techni ca 1 Speci fica tion Pa raqra oh 3. 1.3. 1 Basis for Technical Specification Change:

This proposed revision deletes all reference to "(indicated position)" which previously appeared in paragraph

3. 1.3. 1 and Action Statements (b) and (c).

The purpose for the revision is to allow alternate methods for physical rod position verification without altering the intent of the Specification.

As the intent of the Specification is primarily concerned with actual rather then indicated rod position, two alternate methods for oosition verification are available for use in addition to the position indication system.

The two alternate methods are (a) rod detector (L.V.D.T.) secondary coil voltage measurements and (b) the movable incore detectors.

The primary source of rod position determination will continue to be the rod position indication system although with the alterna.e method'vailable in the event of an indicator in excess of the

+

12 step requirement, the surveillance requirements and specification intent can still be satisfied.

It should also be noted that Specification

3. 1.3 '

imposes sufficient requirements to insure both the availability and accuracy of the position indicators.

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Technical Specification Paraqranh 3.1.3.1 Basis for Technical Specification Change:

This proposed revision deletes all reference to "(indicated position)" which previously appeared in paragraph

3. 1.3. l.and Action Statements (b) and (c).

4'he purpose for the revision is to allow alternate methods for physical rod posi tion verification without altering the intent of the Specification.

As the intent of the. Specification is primarily concerned with actual rather then indicated rod position, two alternate methods for oosition verification are available for use in addition to the position indication system.

The two alternate methods'are (a) rod detector (L.V.O.T.) secondary coil voltage measurements and (b) the movable incore detectors.

The primary source of rod position determination will continue to be the rod position indicati.on system although with the.alternate method available in the event of an indicator in excess of the

+ 12 step requirement, the surveillance requirements and specification intent can still be satisfied.

It should also be noted that Specification

3. 1.3.2 imposes sufficient requirements to insure both the availability and accuracy of the position indicators.

REACTIVITY CC'iTROL SYSTEMS 3/4.1.3 i"r VABLE CONTROL ASS""."BLIES GROUP HEIGHT LIMITING CONOITICN FOR OPERATION 3.1.3.1 All full length (shutdown and control} rods which are inserted

'in the core, shall be OPERA3LE and positioned within + 12 s.eps of their group step counter damand position.

APPLICABILITY:

MiOOES 1* arid 2+

ACiEGli:

a.

With one or more full length rods inoperable due to beina immovable as a result o. excessive riction or r echanical interferer.ce or known to be untripoable, de:er.,ine

.hat the SHUTOG'~N VIRGIN requirement of Soecification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in HOT STA,'iOBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b.

With-more than one "full length rod inoperable or misaligned from the group step counter demand position by more than

+ 12 steps, be in HOT STAIIOBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

I c.

With one full rod inoperable due to causes other than addr essed by ACTION a,

above, or misalign d

rom its group step coun.er demand heigh".

by more han

+ 1" s.eps, POWER OPERATION may continue proviaea that witnirr orZnour either:

The rod is restored to OPERABLE status within the above alignment requirements, or 2.

The rod is declared inoperable and the SHUTOO'~N "ARGIN requirement of Soecification

3. 1. 1. 1 is satisfied.

POWER OPERATION may then continue provided that:

a)

A reevaluation of each accident analysis of Table 3.1-1 is performed within 6 days; this reevaluation sh'all confirm that the previously analyzed results of these accidents remain valid for the duration of operatio'n under these conditions.

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C.

COOK - UNIT 2 3/4 1-18

b)

Tht" S i~JT':l!i VA.".0Ili reo Jlr~i":=nt 0 "ec> f'. cat1OTt 3.1.1.1 is deter;.tined at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

c)

A po::er distribu.ion

~",.ap is obt ined flo i the n;ovable i>>core dctec'"rs and F,{~) and F'".< are verified t" be.'ith';n their ',i.;:its.~ith'ir 72 hcurs.

d)

Eitner t.

E'8 "~L PGiER level is reduced to 75'T R, T 0

'....".." cL:

- EP, 't'i thin one h Jr

a. d

'Lli "flin the next

-'. ho;r-the high neutrcn flux trip setpo'.nt is r educed to

< 85.; of i <<TE0 li -".""..i';L Pt".'('".'c e)

The re,:ir"er of 'he rods in the group i"ith the inoperab'.e rod are ali-:~ed to.;:ithin 1

s-.eps of the inoperable rcd v!i "t n One hotJ? "htile l,".aintainino the l cd seQtJence antd inse t ion 1 i.".l1 ts of Fit'Ures 3.1-1 and

.1-."", the T.'-:=i'.".'<~ PO;:-2 ltevel shall be restricte" pursuant to Specification 3.1.3,6 during subsequen.

op ra'ion.

s SURVE! LLr"..'CE RE".; I.";"-

"=.'!T" 4.1.3.1.1'he posit, on of each full lengtn rod shall be d terained to be v>ithin the g Gup detialld liait by verifying the individual rod positions at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except durin'irt.e intort(als

~.hen the

.",od Position,0eviation t',onitor is inoperable, then veri ying tne group positions a". least once p r 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

4.1.3.1.2 Each full lerqth rod not fully inserted in th core shall be deterained to be GPEBA"LE by i::ove::;ent of at least 8 steps in any one directio'n at least o>>ce pe~

31 oays.

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NUCLEAR REGULA 0 I

WASHINGTON, D. C. 20555 triarch 15, 1979

,]rrh ~

ALL POWER REACTOR LICENSEES Gentlemen:

On September 14, 1978,. the Nuclear Regulatory Commission established a

new Pipe Crack Study Group which was to evaluate recent pipe and safe end cracking experience relative to previous staff conclusions and recommendations.

The bases for establishing the new Study Group were

( 1) the discovery of cracks in the inner surface of large-diameter austenitic stainless steel piping (recirculation lines) in a BWR and (2) questions concerning the capability of ultrasonic detection methods to detect small cracks.

The new PCSG reviewed existing information that either was contained in written records or had been collected through'eetings in this country and in foreign countries.

The review was in the context of changes occurring since the preparation by the original Pipe Cracking Study Group of NUREG-75/067 "Technical Re ort Investi ation and Evaluation of Crackin in Austenitic Stainless Steel Pi in of Boilin Water Reactor Plants The conclusions and recommendations of the new Pipe Crack Study Group are presented in the enclosed "Investi ation and Evaluation of Stress Corrosion Crackin in Pi in of Li ht Water Reactor Plants NUREG-0 31.

This report is for your information and comment.

Also enclosed is a copy of a related Federa1

~Re iater Notice.

The NRC staff will review the Study Group report and its conclusions/

recommendations and any comments received about the report.

Following this review, the staff will decide what further actions, if any, are required for the licensing and operation of reactors.

Sincerely,

Enclosures:

1.

NUREG-0531 2.

Notice Brian K. Grimes, Assistant Director for Engineering and Projects Division of Operating Reactors

c'h

[75%401-M)

DOMESTIC UCENSINO OF'PRODUCllON AND UTIUZATIOKFACIUTIES' Investlsotlon onrl Evolvotion of Stress Corrosion Crocking In Piping of Usht Woter Reactor Plonts AGENCY: U.S.

Nuclear Regulatory Commission.

ACTION: Request for public comment on NUREG-0531

-InvestfgaUon and Evaluation of Stress Corrosion Crack-ing ln Piping of Light Water Reactor Plants" February 1979.

SUMMARY

On September 14, 1978, the Nuclear Regulatory Commission established a new Pipe Crack Study Group. The Group was to evaluate recent pfpe and safe end cracking ex-perience relatfve to previous staff con-clusions and recommendations.

The NRC seeks public comment on the report which summarizes the Group's revfew and conclusions.

DATES: The public comment period expfres May 15, 1979.

. FOR FURTHER INFORMATION CONTACT:

Darrell G. Eisenhut,. Deputy Direc-tor for Operating Reactors, Division of Operating Reactors, Office of Nu.

clear Reactor Regulation.

U.S. Nu.

whar Regulatory Commission. Wash.

fngton. D.C. 20555. (Phone: 301-492-7221)

In 1975, a Pipe Cracking Study Group was established by the United States 9fuclear Regulatory Commission (USNRC) to review intorgranular stressworrosfon crackfng (IGSCC) fn Boiling Water Reactors (BWRs). The Group reported fts findings concern.

fng stress~rrosfon cracking in by-pass lines and core spray piping of austen-tic stt(fnless steel fn a report, Technf-cal Report-Investfgatfott and Evalua.

tfott of Crac)rfttg fn hustettftic Stain-less Steel Piping of Bofling Water Re-actor Plants (NUREG-75/067).

During 1978, IGSCC <<ss reported for the first time in larg~eter piping in a BWR. This discovery, to.

gether with questions concerning the capability of ultrasonic detection methods to detect small cracks, fed to the formation of a new Pipe Crack Study Group (PCSG) by USNRC on September 14, 1978.

The charter of the new PCSG was to specifically address the'five following questions:

"1. The significance of the cracks discovered in large4iameter pipes rel.

ative to the conclusions and recom.

mendatfons set forth in the referenced report (NUREG-75/067) and fts fmple.

mentatlon document, NUREG-0313;

2. Resolution of the concerns raised over the ability to use ultrasonic tech-rdques to detect cracks in. austenltlc stainless steel;

3. The significance of cracks found fn huge-diameter sensitized safe ends and any recommendations regarding the current NRC profpmn for dealing with.this matter,

4. The potential for stress corrosion cracking in PWRs:

S. Examine the significance of crack-ing fn the Inconel safe ends that has been experienced at the Duane Arnold Operating Facility. and develop any recommendatfons regarding NRC ac-tions taken or to be taken."

The PCSG limited the scope of the study to BWR and PWR piping and safe ends attached to the reactor pres.

sure vesseL The PCSG reviewed exist-ing information-either that contained in written records or that collected through meetings fn this country and in foreign countries. The specific areas considered are presented fn the chap-ters of this report:

~ BWR Cracking Experience and Correcth e Actions

~ PWR Cracking Experience and Corrective Actions

~ Metallurgy Associated 'with Pipe Cracking o Reactor Coolant Chemistry

4) Pipe Configuration and Stress Levels O Duane Arnold Safe.End Cracking O Meth()ds of Detecting Cracks O SlgnUlcance of Cracks O Recent Development Relevant to Control and Detection of IGSCC The review of'these topics ln the context of changes occurring since the preparation of NUREG-75/067 led to the preparation of specific conclusions and recommendatlons relevant to the current status of IGSCC, the signlfi.

cance of the problem. and the rellabil-lty of detection and measures available to correct or minimize IGSCC ln exist-fng and future plants. These conclu.

slons and recommendations are pre-sented fn the newly issued PCSG report.

The NRC staff willreview the Study Group report and its conclusions/rec.

ommendatlons and the public corn.

ments received during this comment period. Following this review. the staff will decide what further actions, ff any, are required for the fleet)sing and operation of reactors.

Requests for a single copy of the report should be made in writing to UN. Nuclear Regulatory Commission, Washington, D.C. 20555, Attention: Di.

rector, Division of Technical Informs-tfon and Document ControL Conunents on this,report should be sent to the Office of Nuclear Reactor Regulation, UN. Nuclear Regulatory Commission, Washington, D.C. 20555, Attention: Duputy Director. Division of Operating Reactors. The comment period expires.May 15. 1979. Copies of all oomments received wfffbe available for examination ln the Commission's Public Document

Room, 1717, Street, N.Wwashington. D.C.

'ated at Bethesda, Md this 6th day of March, 1979.

For the Nuclear Regulatory Com.

snfssfon.

Vtcrox SrEtzo, Jr.,

Dfrector, Dfvfsfon qf Operatfng

Reactors, Offfce ofNuclear Re.

actor Regulatfon.

EPR Doc. 79-7705 Filed 3-13-70: R:4S am)

FEDERAL RED(STEIN VOL44, NO. 50 TUESDAY, MARCH 13, 1919

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 OFF ICIAL BUSINESS PENALTY FOR PRIVATE USE, $300 POSTAGE ANO FEES PAIO U.S. NUCLEAR REGULATORY COMMISSION

Mr. John Tillinghast Indiana and Michigan Electric Company Indiana and Michigan Power Company cc:

Mr. Robert W. Jurgensen Chief Nuclear Engineer American Electric Power Service Corporation 2 Broadway New York, New York 10004 Gerald Charnoff, Esquire

Shaw, Pi ttman, Potts and Trowbridge 1800 M Street, N.W.

Washington, D.

C.

20036 David Dinsmore Comey Executive Director Citizens for a Better Environment 59 East Van Buren Street Chicago, Illinois 60605 Maude Reston Palenske Memorial Library 500 Market Street St. Joseph, Michigan 49085 Mr. D. Shaller, Plant Manager Donald C.

Cook Nuclear Plant P. 0.

Box 458 Bridgman, Michigan 49106 Kenneth R. Baker 2874 Robin Hood Drive Stevensville, Michigan 49127